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Compilation of FRN Notices (Prior to 1996). Title 10- Chapter 1, CFR Parts 50-199, Vol 4
	ML18215A067
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Issue date:	12/31/1996
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{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION Rules and Regulations Title 1 O -Chapter 1 CODE of FEDERAL REGULATIONS Volume IV STATEMENTS OF CONSIDERATION JANUARY 1, 1987 THROUGH DECEMBER 31, 1996 PARTS 50-199 Division of Administrative Services UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 J UNITED STATES NUCLEAR REGULATORY COMMISSION Rules and Regulations Title 1 O -Chapter 1 CODE of FEDERAL REGULATIONS Volume IV STATEMENTS OF CONSIDERATION JANUARY 1, 1987 THROUGH DECEMBER 31, 1996 PARTS 50-199 Division of Administrative Services UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555

I 1.olll PREFACE The Division of Administrative Services, Office of Administration, distributed a reissuance of the "basic book" of the NRC Rules and Regulations in June 1999. This edition of the NRC Rules and Regulations monthly supplement program contains all codified text of 10 CFR effective through March 1999, including the 40 supplements that comprised the publication as of February 26, 1999, plus the supplement for March 1999. As part of this reissuance of the basic book that comprises the "loose-leaf' NRC Rules and Regulations, the Statements of Consideration for final rules published from January 1, 1987 thru December 31, 1996, were removed. This decision was made to reduce the unwieldy size
off this monthly publication.

The Statements of Consideration for final rules in this bound, two-volume set (Volumes Ill and IV) are to be used as a permanent companion to the NRC Rules and Regulations. These volumes are not a complete historical set of NRC rulemaking preambles but merely replace those Statements of Consideration for final rules that were removed as part of the March 1999 compilation. The Statements of Consideration for final rules prior to January 1, 1987 are also available in a bound, two-volume set (Volumes I and II). Volume I consists of the Statements of Consideration prior to January 1, 1987, for Parts 0-49 of 10 CFR. Volume II consists of the Statements of Consideration prior to January 1, 1987, for Parts 50-199. Volume Ill consists of the Statements of Consideration from January 1, 1987 through December 31, 1996, for Parts 0-49. Volume IV consists of the Statements of Consideration from January 1, 1987 through December 31, 1996, for Parts 50-199. iii I PART I 50 -RULES.and REGULATIONS TITLE 10, CHAPTER 1, CODE OF FEDERAL REGULATIONS-ENERGY DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES STATEMENTS OF CONSIDERATION '52 FR 1292 Published 1/12/87 Effective 2/11 /87 Bakkruptcy Fi/ing;.Notification.. Requi(em~nt~ ..

Part 30 Statements of Consideration
52 FR 1415 Published 1/13/87 10 CFR Part 50 Domestic Licensing of Production and Utilization Facilities; Minor Corrective Amendment Correction
In rule document 86-29170 beginning . on page 47206 in the issue of Wednesday, December 31, 1986, make : th!! following corree:tion:

§ 50.54 (Corrected] On page 47206, in § 50.54, in the third column, the first paragraph designation should be "(p)(l)".

52 FR 9453 Published 3/25/87 Effective 5/26/87 Operators' Licenses and Conforming Amendments See Part 55 Statements of Consideration 52 FR 16823 Published 5/6/87 Effective 5/6/87 10 CF.R Part 50 Production and UtHization Facilities; Timing Requirements for Full Participation Emergency Preparedness Exercises 1or 'Power Reactors Prior *to Receipt of an Operating

'License 50-SC-1 AGENCY: .Nuclear Regul~tozy Commission. ACTION: Final rule.

SUMMARY

~ The Nuclear Regulatoiy Commission tNRC or CommissiotJJ u ame:1*=.:r\g itf regu.l.a!io~s to !c:~a:~gt::: the timing re(2u.ire:::ie:itE for,a foll paMi:::i;:;ation emer!'enCJ prepare.cineS5 exercise for power .reactors prior ,io issuance of a full-power opera ting license [one authorizing-operation aho\'t: 5% of .ratedpower.of the .reactor}. The amendment .r.equir.es a full participation exer.cise. including State and local governments, lo be held within two years before lhe issuance of .a foll-power oper.itin_g license. as opposed .to the current reauirement of within one vear. An exerci;e which tests Llie Jicens~'.li onsite emergencr plan, but which need not include State or local governm!!nt participation, is .still required !o he held within one year before issuance of a full-power operating 1icense. This rule change is unrelated to the Commission':s notice of .,proposed rulemaking thai would ~tab!ish criteria for the evaluation of .emergency plannin,g for nudear plan ts in .those situations in which a State or localitj' has -elected not to participate in the emergency planning process. EFFECTIVE DAT.£: This ru1e 'is effective on May 6, 1987. FOR FURTHER INFORMATION CONTACT: Michael T. Jamgocbian, Regu1atory Applications Branch, Office of Nuclear Regulatory "Research. U.S. Nuclear Regulatory Commission. \\'ashington. DC 20555, ielephcne (301} 4!1&-:7657. SUPPLEMENTARY INFORMATION: I. Bacl<.ground This notice of final rulemaking will change the timing reguirements for a foll participation emergency preparedness exercise for power reaciors prior to issuance of .a full-power operatillB license. It is unrelated to the Commission's notice of proposed rulemaking, published in the Federal l __ PART 50

STATEMENTS OF CONSIDERATION Regislflr on March 6, 1987 {52 FR 6980), that would establish criteria for the .evaluation of emergency planning for nuclear plants in those situations in wbich a State 1oca1ity bas elected not to participate in the eme!ilency planning process. The 'Commission published .the proposed timing-requirements rule for comment on December .2, 1986 {51 FR 43369). A notice extending the .30-day comment period was published in the Federal Register on January 7, 1987152 FR 543). During the 40-day comment a total -0f 18 public .comments were received.

Nine supported the proposed rule and nine opposed it. As indicated below, the Commission has reviewed the comments and has decided to promulgate a final rule which includes a number of rnodifica lions from the one b;'. had been proposed. The Fecieral E:-'.,e;gency Management Agency. by memorandum dated March 27, 1987, has advised the Commis'sion of its concur~ence in the final rule that is being issued here. \-\7 hen the Commission decided to require a full-participation emergency planning exercise within one year prior to the licensing of a power plant. it based this scheduling decision on a ba:ance between foe des;rabilitv for an exe,cise close to the date of licensing in o,der to assess the adequacy of the emergency plan being tes!ed and the countervailing need to avoid scheduling and resource burdens. Based on the Commission's experience since the 'original promulgation of the scheduling requirement the Commission now believes that it is appropriat-e to strike a new balance. The new rule strikes that balance by requiring a full-participation emergency planning exercise within two years prior to the licensing of a power plant. the same scheduling requirement mandated for full-participation emergency planning exercises after 1* .

1censmg. Since the promulgation of its emergency planning requirements in 1980. both the Commission and the Federal Emergency Management Agency (FEMA) have gained much experience in assessing the results of. a:id the requirement for. participation exercises.

Most of these exercises have been the post-licensing exercises that NRC and FEMA regulations now require to be held every two years. In setting the two-year requirement for operating plants in 1984, prior NRC and FEMA experience demonstrated that the reasonableness of emergency planning at a nuclear power pl ant can be fairly tested and* adequately assured by a participation exercises which are held every two years rather than on a more frequent basis. 49 FR 27733, 27734-27735 [July 6, 1984). Similarly. the Commission has concluded that no safety requirement mandates a participation exercise within one year prior to plant licensing. To the extent that an off site pre-licensing exercise is intended to reveal whether an emergency plan ha~ fundamental flaws, that purpose can be achieved at least as well by an exercise held within two years of licensing as within one year. To the extent that the exercise is designed to lest the preparedness of those individuals and organizations that must participate in offsite emergency planning. NRC and FE.\.iA experience with post-licensing exercises has convinced us that exercises every two years, including rem~dial exercises when necessary. perform this function satisfactorilv. Exercises on a more frequent ba;is are not necessarv to enable the Commission to dete~ine whether an emergency plan provides "reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency." lo CFR 50.47[a). Moreover, since the Commission's promulgation of its original requirement for a full-participation exercise within one year of the licensing of a power plant it has also become clear that the resource and scheduling burdens created by this t_iming requirement have proven far more onerous than originally

expected.

First, with the United States Court of Appeals for the District of Columbia Circuit's decision in Union of Concerned Scientists

v. NRC, 735 F.2d 1437 (D.C. Cir. 1984), cert. denied, 469 U.S. 1132 (1985), it has become necessary to permit litigation in contested proceedings over the results of pre-licensing exercises.

This litigation occasionally has not been completed ' within the year following the exercise. See e.g., Long Island Lighting Co. (Shoreham Nuclear Power Station, Unit 1). Such a delay makes it impossible to comply with the regulatory scheduling requirement and the dictates of UCS v. NRG. Even when the delavs do not make compliance with both requirements impossible, they unnecessarily complicate both the licensing proceeding and the scheduling of the required pre-licensing exercise. Second. utilities are finding it difficult to predict the actual date when their plants will be receiving an operating license. Thus, experience is proving that it is often difficult to know precisely when the pre-licensing exercise should be scheduled to comply with the one-year requirement. Certainly, an important indicator of this difficulty is the fact that in the last two years six plants that have been awarded operating licenses have sought and received exemptions from the scheduling requirements of the licensing exercise rule. See 52 FR 713 50-SC-2 (January 8, 1987) (Shearon Harris Nuclear Power Plant: eight-month extension); 51 FR 41035 (November 12, 1986) (Perry Nuclear Power Plant, Unit 1; eleven-month extension); 50 FR 32129 (August 8, 1985) (Limerick Generating Station, Unit 1; two-week extension); id. 28485 (July 12, 1986) (Enrico Fermi Atomic Power Plant, Unit 2; three-week extension); id. 9917 March 12, 1985) (Waterford Steam Electric Station. Unit 3; five-week extension); id. 5829 (February 12, 1985) (Byron Station, Unit t:-.. ;-ef::-::-::.::-1fr1 e>:tc~sjon1. Tt1E' freuue!i~V and circumsta!1ces surround 0 ing these exempiions support the Commission's conclusion that the one-year scheduling requirement has proven difficult to meet and that an emergency exercise conducted more than one year before plant full-power licensing is adequate to assess the propriety of an emergency plant. As a result of the Commission's experience with the one-year licensing exercise requirements and FEMA and NRC experience with the two-year post-licensing exercise requirements. which provides for remedial exercise when necessarv. the Commission has determined !hat *its previous pre-licensing requirement for a full-participation exercise wi'.hin one year of the licensing oi a power piant is not necessary. The benefits of a licensing exercise requirement can be fully achieved by allowing that requirement to be met within two years of the lic-ensing of a power plant. This approach also should reduce the unnecessary scheduling and resource burdens that have become e\*ident to the Commission based upon its experience with the one-year requirement. II. Summary of Public Comments and Commission Responses J_ Commonwealth Edison Summary of Comment. Commonwealth Edison supported that portion of the proposed rule extending from one to two years the period within which the pre-operational off site exercise must be held. However. Edison disagreed with the last two sentences of the proposed rule which require the applicant to conduct an exercise of its onsite plan if the offsite exercise is more than one year prior to issuance of the operating license. Edison argued that the additional test would be of marginal value and might tend to introduce additional issues into the operating license hearing. On this basis Edison recommended deletion of the last two sentences of the proposed rule. Commission Response. The Commission disagrees that a operation13l onsite exercise *within one year_before issuance of a full-power operating license is of marginal value. PART 50

STATEMENTS OF CONSIDERATION The importance of annual onsite emergency planning exercises by the licensee's operational staff has already been recognized in the Commission's regulations, which now require that after a facility is licensed to operate there must be an annual onsite exercise.

This annual emergency response function drill erisures that the licensee's new persanne1 are adeq~ai.el}* onC p;-o~!p1Jy trained and that exisitn~ hcensee pe:so:mei maintain thei; e:rnergenq response capability. The exisiing requi,er:ient of a pre-operational onsite exercise within one year prior to power license issuance -is consistent with this philosophy as well as foe Commission's general desire to haYe pre-operational emergency planning exercises as close *as practicable to the time of licensing. And since, unlike foe situation v.ith offsite exercises, no one has identified any existing response or timing difficulty with the cnsite requirement. we find no reason !o rev"ise the requirement at this time. Moreover, to mandate an onsite exercise within one year of operation while requiring -an offsite exerclse within two -years is a rncognition of the distinct nature of the participants involved in each ir1stance. The Stale .and local emergency planning organizations that are primarily involved in offsite emergenc:y planning are in .almost all instances .organized and trained to deai with eme~gency situations long before facility operation. Wniie the offsite emergency test is important to judge the abiiity of these existing organ;zations to respond to the particular of a radiological emergency, in iight of their ongoing responsibility for all types of emergencies a demonstration of offsite preparedness by such agencies withill two !*ears prior to licensing affords -reasonable assurance of their capabilities at the time-of1icensing.1n contrast, as an applicant mal<es a scale shift from a facility construction to a facility operation mode within tbelast twelve to eighteen months prior to operation, as a general .rule many new operational personne1 are retained wbo must .be ready to carry out the utility'.s onsite emergency response responsibiiities. It is also .in recognitwn -of this distinction that the Commission finds that an onsite exercise should be required within one year of licensing to provide assurance that the applicant*s onsite response capabilities are adequate. For the pur_poses ofdarjty, the Commission is revising the last two sentences, which pro,*ide that a operationa1 onsite exercise be held within one year before *operali-on above :i% of rated power. 2. Edison Electric .institute Summary ,of Comment. The EdiSGD Eleciric Institute supported the proposed rule hnd did not suggest any chanaes to Hs text or .rationale. "' Commission Response. None required.

3. 1-lunion and V.-'i1liams Summaryo_f Comment. This 1aw Ii.rrr: fiied comments on behalf oI Long ls1and Li~nting Compan_y fLILCO). LILCO stated that it suooorted the amendment anc:i agreed with the Commission*s basic premise that the two-_year inierrn] was adequate to ensure an acceptable lev.el cf emergency preparedness.

LILCO cited its experience with .the Shoreham facility as supporting the need for the amendment. and disagreed with Commissioner Asse1stine's view that the exemption process was the approprlBte means-to address the problem. LlLCO did not offer any suggestions Ior changes in tbe Jlroposed rule. Commission Response. None requir-ed.

4. l\,1arvin Lewis Summary of Comment. Mr. Lewis opposed the proposed rule, staling .that it ,would "weaken regulation and _pose a danger to the health and safety .of the public by allowin_g unlicensed operators more freedom lo act w1th nuclear bazards before baving proven thal they can act .responsibly." Commission Response.

Licensees are not being granted any additional ~*freedom~ by this rule. The fu11 participation exercise must still be he1d prior to fuli-power .operation of the facilitr and a pre-operational onsite exercise will continue to be reouirerl one year prior to full-power operation. The only change is the timing of the full participation exercise. .5. Atom1c Industrial Forum The Atomic Indusirial Forum {AlF) :supported the proposed rule but pointed out, with Tespect to its last two sentences, tha1 Section IV.F.Z of Appendix E already requires a 1icensee to conduct annua1 exercises o'f its rnergency p1an. A1F suggested thal this was .a redundant requirement and therefore the last two sentences -of the proposed rule should be deleted. Commission Response. The Commission disagrees that the 1ast lwo sentences of the proposed ru1e sbm.ild be deleted. bul has determined to revise those .sentences for ;pwposes of clarit_y. {Also. in the interests of clarity. the preceding sen_tence is modified to specify that an operating license "for full power~ is lo be taken, in this context, to be "one authoriz1ng operation iibove '5% of raled power." The prior reference fo 5% of Ia ted power was ambiguous.) 50-SC-3 li. Stone -& 11rebster Engineering Corporation Summwy of CommenL Stone & W ebst~r supported the proposed rah! *tmd -did not suggest any ~hanges. Commission Re-~ponse. '.',:cnt Tt'c;'..::~ed.

7. Seacoast Anti-Pollution Leat!ue Summon* of Comment. The Seacoast Anti-pollution League (SAPLj opposed the amendment and agreed with the views of Commissioner Asseistine.

SAPL argued that ernergencr response personnei experience fairly rapid turnover, and therefore "a full scale exercise is needed annuallv." SAPL did not accept the Commission's reliance on the fact that State -and local go\*emments are ofien called upon to respond :to a 1.*ariet!' on non-nuclear emergencies. Commission Response. The Commission does not agree that two years between full pa;ticipation exercises is unwarranted based on personne1 changes. The Commission*',s .and FEMA's rules have, since 1984, permitted the two-year cycle for fu11 participation exercises for operating plants. The Commission's view was .in 19B4. and is today, that there are more beneficial uses of State and local governments* resources, such -as providing for additional training and equipment, than using such resources 'to -support an annual full participation exercise. The Commission does not rely -so1e1}' .on the fact that State and loca1 governments routinely respond to a variety of public emergencies. However, the basic principles involved in handling non-nuclear .emergencies, such as evacuations due to an impending lmrrlcane or a leak of toxic chemicals. also apply in responding-to a nuc1eaT accident. This lends support to the rule because State and loca1 emergency response organizations are Ireguentl_y called upon and must maintain a hlgh degree of readiness independent of nuclear power plant exercises.

8. Liz Cu1lingtt:J11 SummaIJ' of Co!f!ment.

Ms. CuTiington opposed the proposed rule and stated as follows: In extending the time period Irom one yaar to two, the NRC would be essentially handing to the *utilities an across-the-board offer af total exemption from lhe reguirement -to prepare emergency response plane Tor reactors under licensing review, as long 11s un acceptable number of sheets of paper -ere submitted to the Commission with appropriate title pages. Under this propose&! rule change, -ll :utility could submit :a xeroxed ,eopy-of !fl'ebsler':i; Dictionary as iitll emergency response *plan. and have ruil deadline for completil\\: the plan itself. .15 * .reality, £or either exercisin&-it. Dr .demonstrating that it is Ieasible. PART 50

STATEMENTS OF CONSIDERATION Commission Respoose.

The proposed rule change is more limited in scope thar. the c;ommem sugt1ests. Ii does not affec~ elthe: tht re~:r..:.irPC content of emergency pi2!'ls n;r th:: need to exercise such p;an~ or. 2 ~egular basis. The amendmeni ur.]v exiends from one to two years the period within which the preoperational full-pa;ticipa tion exercise mus\ be held. All other Appendix E and 10 CF'R 50.47 requirements must continue to be met as a prerequisite for issuance of an operating license, including the requirement that a pre-operational onsite exercise be held within one vear before going above 5 percent ofrat~d power. 9. Georgia Power Company Summar}' of Comment. The Georgia Power Company supported the proposed rule and did not suggest changes in its text. Commission Response. None required.

10. Nuclear Information and Research Ser.'ice

Summon' of Comment. The ~uclear Information a'nd Research Service [NIRS) opposed the amendment, and stated three reasons for doing so: 1. Changes [in emergency procedures]

will be more likely to occur in a new plant where lest minute alterations in technical specifications. guidelines. newly trained operators. and actllal equipment ar~ common occurrences. It is precisely this kind of change which marks 2 new plant from an operating plant and which necessitates an exercise no more than one year prior io licensing.

2. An exercise no more.than one year prior to licensing

- would ensure that any new government officials or workers are familiar with the plans themselves.

and are capable of carrying them out."' 3. The one-year requirement has been easily satisfied in most cases, and a schedule exemption is an available option where needed. Commission Response. Changes of the type cited by NIRS do occur prior to issuance of an operating license and throughout the life of an emergency plan. However, these changes would be addressed in the utility's emergency plan. The proposed rule retains the requirement that a pre-operational onsite exercise be held within one year before going above 5 percent of rated power. When changes in offsite emergency procedures or offsite personnel occur, it is the responsibility of the State or local government to ensure that personnel are adequately trained to carry out their functions under the plan. The licensee is required by Commission regulations to assist in such training. See 10 CFR Part 50 Appendix E, Section F [introductory paragraph). The proposed rule would pennit the use oi a two-yea, cycle ior the holding of a pre-operationai off site exercise. This timing would be consistent with the two-year cvcie for the holding of a post-op~ratior{al off site exercise for operating plants which has been in effect since'1984. Sound principles of administrative law dictate that where agency policy is no longer correctly reflected in its rules, rulemaking should be undertaken and public corc,.rnent sought. The Commission now belie\'es that a year period between full participation exercises should be used in all cases. and therefore has proceeded with rulemaking to codify this policy. 11. Union of Concerned Scientists/New England Coalition on Nuclear Pollution Summary of Comment. The Union of Concerned Scientists [UCS) and the New England Coalition on Nuclear Pollution [NECNP) oppose the rule on the following grouJ1ds:

1. The Commission has not adequately explained its reasons for making a change in policy. 2. The proposed rule ignores a distinction previously drav..-n between pre-and operational exercises.

3. The Commission should have prepared a backfit analysis for the proposed rule. Commission Response.

The logic for the proposed rule was stated in the notice of proposed ruiemaking (51 FR 43369, December 2, 1986). as follows: The Commission in 1984 revised its emergency preparedness regulations to relax the frequency of fuU participation exercises by State and local governments for sites with an operating license. This was done in part because the Federal Emergency Management Agency [FEMA). based on its experience in observing and eva!ua ting exercises. adopted a biennial, rather than an annual. requirement for full participation exercises. Under the biennial requirement adopted by the Commission. State and local governments need only participate in one full participation exercise. at any site, every two years. The Commission revised this regulation because it found that annual exercises used a disproportionate amount of Federal. State, and local government resources, and that. State and local governments frequently exercised their emergency preparedness capabilities by responding to a variety of natural and man-made emergencies. such as chemical sprns. on a continuing basis. The Commission concluded that biennial full participation exercises were adequate to protect public health and safety. The Commission in revising its regulations for full participation exercises retained the requirement for annual exercises of each licensee's emergency plan {49 FR 27733, July 6. 1984). The Commission did not make a similar change regarding the required frequency of fuJI participation exercises at sites without an operating license. Because of the opportunity 50-SC-4 ir. ar, cpcra!~n~ license pro8eeding under Sec:ior: 18!'!~ of the Atomic Energy Act ior a hearing or, the results of a full participation exercise, this requirement created some difficulty in scheduling the exercise so that it would e!iow time for a hearing while still being conducted within one year of plant readiness to be licensed. In 1982 the Commission adopted a rule which, by finding that emergency preparedness exercises were not required for a Licensing Board, Appeal Board. or Commission decision, would have al!oweci the exercise to be conducted close enougl-. to a licensing decision lo avoid this difficulty and to avoid annual pre-iice:-ising exercises (47 FR 30232, July 13. 1982). However. the Court of Appeals for the District of Columbia Circuit vacaied that rulem2king. The court held that the Commission could not remove from the healing requirements of Section l B9a of the Atomic Energy Act a material issue relevant to its licensing decision, and that the prelicensing exercise was such a material issue. Union of Gonce med Scientists

v. NRC. 735 F.2d 1437 (D.C. Cir. 1984). cert. denied, 105 S.Ct. 815 [1985). The Commission has thus been left with a regulatory scheme for frequency of full participation emergency preparedness exercises that treats sites with an operating license difierentlv than sites without an operating license: The Commission does not believe this disparity in treatment is warranted.

The Commission is concerned . about the burden the present rule may place on State and local governments. The requirement that those governments participate in a full participation exercise every two years is in addition to the requirement for their participation at sites without an operating license. Requiring annual participation at sites without operating licenses could thus place a significant burden on State and local government resources.

The Commission in the prior rulemaking determined that emergency preparedness would be adequate if Slate and local governments participated in an exercise every two years. There seems to be little reason why State and local governments nonetheless should have to participate in full participation exercises on an annual basis in the pre-licensing stage solely because a license did not issue within 365 days of the exercise.

The only requirement should be that the participants be adequately in place and trained to make the exercise meaningful. This could well occur two years before issuance of an opera ting license. If the exercise demonstrates that preparedness was inadequate. then remedial steps. including another exercise, if appropriate, can be taken. r..foreover in accord with the Commission*s regulations for sites with operating licenses, applicants will still have to conduct annual exercises, i.e., if the full participation exercise is held more than one year before issuance of the operating license. then the applicant must conduct an exercise of its eme:-gency plan before license issuance. However, that latter exercise need not involve State or local governments. PART 50

STATEMENTS OF CONSIDERATION UCS no!nts ou! tha~ inc ~~e:: ruiem2king on emergency pia,ming.

the Commission remarked on the desirabiiity of ha\'ing the operational exercise close in ti:ne to commercial operation. The reason slated b\* the Commission was that the "exercis.es are best held at a later time. when the opera ling and irnrnagcment staff of the plant-who are central figures in an exercise-are in piace and trained in emergency functions." (47 FR 30233, July 13. 1982). As was explained earlier. the Commission continues to support this principle and has retained the reouirement that an onsite exercise of the ~mergency plan be held within one year prior to operation abo\'e 5 percent power. . . The backfit ru1e, 10 CFR 50.109, applies only where the Commission seeks to impose new or different reouirements on licensees. It does not apply where requirements are either relaxed or deleted. 12. M'ells Eddleman. et al. Summary of Comment. :l',.1r. Eddleman . and others joining him oppose the amendment for the following re&sons: 1. ** ... a one year time range before operation above 5% power is a practical maximum for giving an up to date "snap shot" assessment of the level and capability of emergenc~* preparedness existing when the planl begins to operate." 2. Om: year is adequate to litigate the results of the exercise, based on the Shearon Harris proceeding.

3. " .*. nuclear accidents have a tendency to occur early in the operation of a nuclear plant ... ", citing Three Mile Island and the Browns Ferryflre.

4. Ylle rule is iliegal because it is an attempt lo deny hearing rights to intervenors in the Shearon Harris case on the exemption gran led from the existing one-year requirement.

Commission Response. The Commission disagrees that a full participation exercise is needed within one year of operation to demonstrate adequate emergency preparedness. The Commission has determined that a year cycle for full participati,on exercises is sufficient for making a finding that adequate protective measures can and will be taken in the event of an accident. The Commission has not based its acceptance of the two year requirement for holding a full participation exercise on the time needed to litigate the results of such exercisE:.

,ather, as indicated abo\'e and in response to comment # 11, the Commission has determined that a two-year cycle is an appropriate period of time for holding full participation exercises.

With regard to litigation the results of the exercise. under UCS l' NRC. 735 F.2d 1437 rD.C. Cir. 1984). it is clear that the results of exercises are litigable in the operating license proceeding. irresuecti\'e of when those exercises are 'held: so long as the holding of an exercise is a pre-license requirement. However, while the two year time period pro\'ided in this rule was not premised on the time needed to litigate the results of an exercise. as was explained earlier, one of the faciors on which the Commission did base this amend:neni was the observed difficult\' in some cases [alihough not in the

Shearon Harris proceeding) in scheduling the exercise so that it would allow time for a hearing while still being conducted within one year of plant readiness to be licensed.

Another factor was the observed difficulty of utilities in predicting a plant's readiness for a full--power operating license. In this situation. as int.lie case of the Shearon Harris plant. while the holding of the full participation exercise and the licensing hearing would be completed within one year, due to unaniicipated construction delays the plant would not be ready for a full-power operating license within the one year time frame. With regard to the commenter's statement that nuclear accidents tend "to occur early," it is correct that the few major nuclear accidents that have occurred, i.e .. -the Three Mile isiand Accident and the Browns Ferry fire, did in fact occur early in the operational history of the plants. Howe\'er, the number of these occurrences is far too small to establish a "tendency." In any case. the commenter's suggestion that the need for emergency preparedness may be heightened during the initial period of plant operation, e\'en if well taken, does not present a valid objecfion to this rule change because, for the reasons given above, the rule change does not decrease the le\'el of emergency preparedness at a nuclear power plant. The license and exemption ha\'e already been issued in the Shearon Harris proceeding. This rulemaking was not the basis upon which a hearing on the exemption request was denied. Carolina Power & light Co. et al. (Shearon Harris Nuclear Power Plant) CLJ--86-24, 24 NRC __ (December

5. 1986). Certainly, if the exemption request were pending. it would now be mooted as a result of this rulemaking.

The scope of issues open for litigation may be changed by rulemaking. Engaging in such rulemaking has been held by the courts not to deny hearing rights of any person. See Siegel!'. AEC. 400 F.Zd 778 (1968). 50-SC-5 13. Laura Drey Summary of Com,7lent. Ms. Drey opposed the rule change bu! Slated no reasons. Commission Response. None required.

14. Kenneth Vic.kery Summary of Comment. Mr. Vickery opposed the amendment.

stating that "the !\'RC must know if the plants and the surrounding areas are ready for accidents when starting operation since many se:-ious accidents occur early in the opera ting li\*es of nuclear power plants." Commission Response. See response to comment of \'Velis Eddleman, .#12 abo\'e. 15. Rachel Allen Summary of Comment. This comment was a duplicate of Comment -#14. 16. Shaw, Pittman, Potts Er Trowbridge Summary of Comment. This law firm filed commen.ts on behalf of 9 entities holding nuclear power plant operating licenses or construction permits. These commenters supported the proposed rule. fundamentaliy for the reasons cited by the Commission in the notice of proposed rulemaking. The commenters also noted that the proposed rule uses the term "full-scale exercise" which is otherwise undefined in the regulations and recommended that the term "full participation exercise" be used. Commission Response. The term scale exercise" has been replaced with the tenn "full-participation exercise" and the last two sentences of the proposed rule have been re,*ised for purposes of clarity. 17. Carolina Power and Light Company Summary of Comment. Carolina Power and Light Company [CP&L) supported the amendment and cited reasons similar to those given by the Commission. CP&L noted that its recent experience in licensing the Shearon Harris facility bore out the need for the rule change .. , Commission Response. None required.

18. North Carolina Department of Crime Control and Public Safety Summar!/ of Comment. This commenter supported the proposed rule on three grounds: 1. It makes !he NRC rule consistent with FEMA's and increases internal consistency in NRC regulations.

2. It reduces undue burdens on State and locai go,*~rnments.

3. It allows more time for litigation of the results of a pre-operational exercise.

PART 50

STATEMENTS OF CONSIDERATION Cc.'T:r;, *:-t*:.Jr:

ht!:-J,on5t:. Tht: reasons g:ven by th!~ commenler support the Cor!"1missi::in*s position as slated in the notce of proposed miemaking.

m. Commission Decision The Commission has reviewed all commenis received and has decided to proceed with a final rule. The text of the. proposed rde has beer; altered as noted in the response to comments :;;: 5 and #16 above. Upon publieolion of the final rule. a full p;;.rticipation exercise must be held v.ithin two years prior to issuance of & nuc1ear pov.:er plant operating license for opera!ion above 5 percent rated power. If the full participation exercise is conducted more than one year pr;or to issuance of an operating license for full power. an onsite exercise which tests the licensee's emergency plans shall be conducted one vear before issuance of an operating li;ense for full power. Additional V'iews ofCommissioner Asseis ll~'le I continue to believe that fhe reauirement to conduct a full participation exercise, which inciudes State and local government participation, within one year prior to issuance of an operating license is needed to provide an accurate*

and

timely verification of the adequacy of emergency preparedness.

The purpose of this requirement is to provide an to-date assessment of the state of emergency preparedness for a new plant at the time the plant receives an operating license. This requirement has been easilv satisfied in most cases. In the few ca;es in which there has been some difficulty, the Commission's exemption process provides a suitable alternate method for addressing the situation. Given the satisfactory experience with the current rule and the benefit in having up-to-dale and accurate information on the state of emergency preparedness at new nuclear power plants, l would not relax the existing one-year requirement for a full participation exercise. Environmental Assessment and Finding of No Significant Environmental Impact The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule is not a major Federal action *significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. See 10 CFR 51.ZO(a)(l]. Moreover, the Commission has determined. pursuant lo 10 CFR 51.32, that the final ruie has no significant environmental impact. Tni~ determination has beer-; made beca:.Ise the Commission cannot identify any impact on lhe hum8l} environment assodated with changing the timing of full participation of State and local governments in pre-licensing emergency preparedness exercises from within one year of license issuance to within two years. The need for this rulema}...iiig is explained in the Supplementary Information accompanying this final rule. The alternative approaches that were considered in L'lis rulemaking proceeding were: 1. To retain the requirement for a full participaiion exercise within one year of issuance of an operating licel19e. 2.. To relax the requirement to within two years of issuance of an operating iicense. There were no environmental impacts identified from either of the altematfres considered. In addition, when promulgating the original emergency planning and preparedness regulations in 1980, the NRC prepared an "'Environmental Assessment for Final Changes to 10 CFR Part 50 and Appendix E of 10 CFR Part 50, Emergency Planning Requirements for Nuclear Power Plants" (NUREG-0685, June 1980}. and concluded that under the criteria of 10 CFR Part 51 an environmental impact sta!ement was not required for the Commission's emergency planning and preparedness regulations, which included 10 CFR Part 50, App. E a-s hereby revised. NUREG-0685 may be examined in the Commission's Public Document Room, 1717 H Street NW., Washington, DC. Copies are available for.purchase through the Superintendent of Documents, USGPO, Box 37082, Washington, DC, 20013-7082. Paperwork Reduction Aci The final rule contains no information collection requirements and therefore is* not subject to the requirements of the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Regulatory Analysis The Commission bas prepared a regulatory analysis for this regulation. The analysis examines the costs and benefits of the action and the alternatives considered bv the Commission. A copy of the regulatory analysis is available for isspection and copying, for a fee, at the NRC Public Document Room, 1717 H Street NW., Washington. DC. Single copies of the analysis may be obtained from Michael T. Jamgochian, Regulatory Applications* 50-SC-6 Branch, O:fice c,f ~:i;;:ear F,:>?c:2.1,:;ry Research. U.S. r\iucicar fc.t,f:;ia,0;y Commission, Washington. DC ZD.355. Telephone (301) 443-7657. Backfit Analysis This final rule does not moc!ify or add to systems. structures, components design of a facility; the design 1opp:-oval or manufactu,ing license for a facility-: the procedures or organization

eq;.:ired to design, construct or opera re a fad::y. Accordingly.

no backfi: analysis pursuant to 10 CFR 50.109 is rec;ui~ed for this final rule. Regulatory Flexibility Certification In accordance with the Regul2tory Flexibility Act of 1980. 5 U.S_C. 605(b). the Commission certifies that this rule will not have a significant economic impact upon a substantial nuwber of small entities. The mle concerns the timing of a full participation exercise of emergency plans for applicants for nuclear power plant licens~s. The electric utility companies owning and operating these nuclear power piants are dominant in their service areas anc do not fall within the definition of a small business found in the Small Business Act. 15 U.S.C. 632, or within the Small Business Size standards set forth in 13 CFR Part 121. Although part of the burden for the conduct of emergency preparedness exercises falls on Siate and local governments_ Ltie final rule. by changing the frequency of the requirement, if anything lessens the amount of the current burden. Thus. the final rule does not impose a significant economic impact on a substantial number of small entities, as defined in the Regulatory Flexibility Act of 1980. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire prevention. Incorporation by reference, Intergovernmental relations. Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For the ;easons set out in the preamble, and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended. and 5 U.S.C. 553, the !\i'RC is adopting the following amendment to 10 CFR Part 50: PART 50

STATEMENTS OF CONSIDERATION 52 FR 19303 Published 5/22/87 10 CFR Parts 50 and 51 Domestic Licensing of Production and Utilization Facilities; Communications Procedures Amendments; Correction AGcNCY: Nuclear Refmh,tor~*

Co:mnission. ACTION: Final rule; correction.

SUMMARY

This document corrects a final rule amending 10 CFR Part 50 and 51 with regard to procedures for submitting correspondence, reports. applications, and other written communications pertaining to the domestic licensing of production and utilization facilities.

The final rule was published on November 6, 1986 [51 FR 40303) and inadvertently omitted a recentiy published revision of one of the affected sections. This action is necessary to return the omitied information and to inform the public and licensees of this omission. FOR FURTHER INFORMATION CONTACT: Donnie H. Grimsley, Director, Division of Rules and Records, Office of Administration and Resources Management, U.S. *Nuclear Regulatory Commission. Washington, DC 20555, Telephone: 301-492-7211. In FR Doc. 86-25045, published in the Federal Register of Thursday, November 6, 1966. make the following correction:

1. On page 40309, in the first column (containing the revision to § 50.54[f)), in the fifth line ending with the word "revoked.", insert the following words after the period: Excep1 for infar;nation so"Jgh! tD Yerifr licemee compliance with the cc.rrent lice::.sing basis fo; tha~ facility.

the ]';RC r.rnst prepan, the n;a~or: or reasons for each info~mation

e~'c'est rrior to iEsunnce to er.sure that ,he. burden lo be in: posed on respondtnts is justified in Yiew of the poiential safeiy significance of the issue io be adcireFsed in the requested information.

Each such just}fication pr?vici!r.d for a;: eYc}?atior; perio~med by tnE: !,RC s!a:::1 m~s1 oe epproved by the Exermin: Director for Operations or his m he~ c:iesignee prior to issuance of the request. Dated at Be!hesda, Jv1a;-ylanci. fob 1 str. day For thr: Nuciea.r F..egu1al.o:y C.mT'.:::H:.:,~-::*:-;. James M. Taylor, Acting Deputy Execu/i\*e Directo,_for Operations. 52 FR 28963 Published 8/5/87 Effective 10/5/87 10 CFR Part 50 Changes in Property Insurance Requirements for NRC Licensed Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRCJ is amending its regulations requiring licensees to maintain substantial amounts of onsite property insurance to provide financial security for stabilizing and decontaminating their licensed reactors in the event of an accident.

These changes will increase the amount of insurance required to $1.06 billion, impose a modified decontamination priority on any proceeds from such

insurance, and require that proceeds subject to the decontamination priority shall be paid to an independent trustee. All commercial reactor licensees are subject to this rule. EFFECTIVE DATE: October 5, 1967. FOR FURTHER INFORMATION CONTACT: Robert S. Wood, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone 301/492-8413.

SUPPLEMENTARY INFORMATION: I. Background On November 8, 1964, a proposed rule was published in the Federal Register ! (49 FR 44645) which would increase the 'amount of onsite property damage insurance that commercial reactor licensees are required to carry pursuant to 10 CFR 50.54[w). Operating reactor licensees are currently required to carry both (1) the maximum amount of property insurance offered as primary coverage by either American Nuclear Insurers/Mutual Atomic Energy Reinsurance Pool (ANI/MAERP) or Nuclear Mutual Limited currently $500 million-and (2) any excess coverage in amount no less than that offered by either $120 million-or Nuclear Electric Insurance Limited [NEIL-11)-$610 million. Thus, the minimum currently required under the rule is $500 million primary coverage and $120 million excess coverage. By buying both excess layers, utilities are able to purchase a total of $1.23 billion in property insurance. The proposed rule also provided for a modified decontaminatio:i priority on any proceeds from such insurance and sought comment on several related issues. 50-SC-7 II. Summary and Analysis of Comments The NRC received 35 comments on the proposed rule to amend its property insurance regulations. The comments may be grouped as follows: ; 21 Utilities 5 Counsel to Utilities 3 Insurers/Insurance Trade Groups 2 Utility/Nuclear Trade Groups 2 Individuals 1 Bar Association Committee 1 Environmental Interest Group Because the issues considered in the rulemaking are complex and affect different utilities in different ways, the focus of the comments varied considerably. Nevertheless, the majority

of utility commenters tended to endorse the positions taken by two trade groups, : the Atomic Industrial Forum and the Edison Electric Institute, but chose to emphasize areas of their particular concern. The two individuals who commented tended to support the 'positions of those utilities with smaller plants worried about being required to carry more insurance than they believe is necessary.

The set of comments from the environmental interest group supported most aspects of the NRC's I proposal. The insurers provided a [ slightly different perspective from the *utilities but made many of the same ipoints. The Bar Association Committee 'comments provided the most distinct perspective on one aspect of the proposed rule, the decontamination priority. The following is a detailed discussion and analysis of the comments received arranged according to topic. The topics i are: (1) Amount of insurance; (2) method I of future adjustment of insurance; (3) !site-wide vs. unit coverage; (4) State jprohibitions against buying certain 1 types of insurance; (5) the !decontamination priority; and (6) other issues. Of greatest concern to the commenters was item five, NRC's proposal to require insurance proceeds to be used first to decontaminate the facility when so required to protect public health and safety and so ordered by the Director of the Office of Nuclear Reactor Regulation.

1. Amount of Insurance Summary of Comments:

Most of the commenters, including most utilities and their representatives, either supported the proposed coverage limit of $1.02 billion or at least found it acceptable or non-burdensome. However, most of those who accepted the proposed limit in principle suggested that the rule be modified to reflect the potential for reductions in capacity because of changes in the insurance markets. To avoid such a situation, several commenters suggested that the rule be PART 50

STATEMENTS OF CONSIDERATION modified to require either-$1.02 billion or, if that amount were not available, the total of the primary coverage offered by either ANI/MAERP or NML plus the total excess coverage offered by both ANI/MAERP and NEIL-II. Others suggested a modification to allow utilities which may incur property losses during a particular policy year to take reasonable steps to have their insurance reinstated so as not to be in technical violation of NRC's regulations.

To avoid this, commenters suggested that the NRC clarify its position that claims made by a utility which resulted in a reduction in policy limits for that policy year would not be considered a violation of NRC regulations. Alternatively, some proposed that *§ 50.54(w) be amended to require a licensee to take reasonable steps to obtain reinstatement of insurance, an action keeping that licensee in compliance with§ 50.54(w). Another commenter suggested that reductions in capacity could be handled by allowing a "cushion" of $50 million to $100 million between what is available and what the NRC requires. Special action should be required on reinstatement only if it . affected public health and safety. A few commenters took issue with the amount established. One suggested that the rufe is not sufficiently flexible because it would require insurance to be purchased from all insurers offering this coverage, which tends to limit a utility's options. It was alleged that this requirement might also lead to competitive behavior. Another slated that the NRC should not be concerned with the full cost of restoring a reactor lo service after an accident but rather with the costs associated with protecting public health and safety. This commenter suggested using decommissioning funds for cleaning up after an accident, and stated that, because of lessons learned from TMI. cleanup costs have been substantially reduced. Another commenter suggested rounding off required coverage of $1 billion because of the lack of precision in establishing a required amount. A non-utility commenter suggested that the "enormous" premiums generated for property insurance are directly removed from expenditures that would otherwise be made for nuclear safety. One commenter suggested that requiring a specific dollar amount, rather than all that is offered, would lead to the established figure being considered by utilities a maximum as well as a minimum. Several commented on the related issue of whether special provisions should be made for licensees of smaller reactors. Most who commented on this issue supported the NRC proposal of not specifically providing for smaller reactors in the rule but rather treating them on an ad hoc basis through the* exemption process. A few commenters indicated their support for special provisions for smaller reactors based on physical size, core inventory, etc. Even those against special treatment of smaller reactors suggested that existing exemptions should remain unaffected by the new rule and that the final rule should clarify that this interpretation is correct NRG Response: 1.a. Issue: Amend the rules to reflect reductions in available insurance through changes in insurance capacity. Response: The NRC believes there is some merit in commenters' proposals to address in the rule the problem of reductions in capacity. To require more than what is available would be meaningless because licensees' only realistic alternative to buying insurance would be to self-insure which the NRC determined would provide no additional assurance. This is discussed under issues "4". Futher, the Commission has traditionally never required more insurance than that generally available. Surety bonds, letters of credit, and other ~ethods of assurance may be available to certain licensees in relatively small amounts, but would probably not be generally available especially if capacity shrank substantially. Consequently, the NRC has modified the rule to allow the lesser of the specific amount or the maximum available from insurance sources. The Commission will continue to monitor the adequacy of the amount of property insurance that is available to reactor licensees. The NRC also believes that there is justification for amending the rule to provide reasonable lime for reinstatement of insurance when a claim filed by an insured causes coverage al a facility lo be reduced during the remainder of the policy year by the amount of the claim. Because a reduction would be for only the remainder of the policy year, after which the insurance would be automatically reinstated, the NRC does not believe a serious threat to public health and safety would exist. It is highly unlikely that any single utility would face two large accidents within one year at the same site. Therefore, the .NRC has modified the rule to require that licensees take reasonable steps to obtain reinstatement of insurance within 60 days. 1.b. Issue: Change the amount of insurance required. Response: Because most commenters supported the amount of coverage proposed in the rule, there was little discussion of whether that figure should be changed. A few commenters stated that coverage was higher than necessary to protect public health and safety or too 50-SC-8 imprecise to be specified to the degree done in the rule. However, in the proposed rule, the rationale for requiring $1.02 billion was not only that this was the most available al that time, but that it approximated the maximum estimated in a PNL study 1 to be necessary to clean up a reactor in its entirety after an accident. The actu~l maximum required was estimated to be $1.06 billion. The amount currently available is $1.23 billion. The Commission believes that $1.6 billion should be required, because at least that amount is now available and no other amount is as technically supportable. The NRC does not believe that most utilities would have difficulty obtaining the slightly higher amount since most choose lo buy the maximum available. The NRC agrees that the process of determining the actual amount needed to protect public health and safety is imprecise and that the threat to public health and safety usually decreases as cleanup progresses. Clearly, there is a sliding scale of effect on public health and safety so that after some undetermined point is passed, the contributions to public health and safety that additional amounts of insurance would make are less significant. although this problem is addressed more fully in the discussion on the decontamination priority. Thus, there currently appears to be no adequate basis for selecting any figure that would be better than the amount suggested in the PNL report. The NRC is not persuaded that there is any compelling reason to keep the $1.02 billion originally proposed, or to round off to $1 billion, or to choose some figure other than the $1.06 figure derived from the PNL report. The NRC is also not persuaded by the argument that requiring the purchase of insurance is unduly restrictive or may lead to anti-competitive behavior by forcing the purchase of insurance from all carriers. Other carriers could conceivably enter the insurance market. In fact, by requiring a specific dollar amount rather than insurance from specific carriers, NRC has avoided a ' Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents. (NUREG/CR-2601). Pacific Northwest Laboratory, November 1982. [See especially pages 2-26 to 2-28. The PNL study estimated that accident cleanup costs at a reference PWR following a scenario 3 accident would be approximately $404 million. However. additional costs lhat can appropriately be ascribed to such an accident include: Base operations and maintenance............................ $124M Design differences (when comparing to TMl-2)...... $84M Cost escalation during cleanup................................... $209M Additional decontamination of the containment building.................. $100M Net Stabilization cost.... . .......................... $139M When these costs are added to the $404 million basic cleanup cost estimate, a total of$1.06 billion is derived.) This report is available for purchase from the U.S. Government Printing Office, P.O. Box 37082, Washington. DC 20013-7082. PART 50

STATEMENTS OF CONSIDERATION restriction which would have prohibited others from offering the insurance.

Similarly, NRC does not accept the argument that the $2-3 million spent per reactor on insurance premiums each . year represent a drain on a licensee's expenditures on safety. These premiums . represent only small percent of a licensee's annual expenditures on operating the reactor safely. NRC also notes that the utilities did not state that they had been prevented from receiving ratepayer revenues to cover the insurance premium expenses. The argument that a specified dollar amount will be considered a maximum as well as a minimum is beside the point. The Commission has based its amount on what the PNL study determined would be needed for decontamination and cleanup following an accident. Any amount above that has not been shown to be significant in protecting health and safety, but represents the. economic replacement of the facility itself. If a utility decides not to insure for the replacement value of its plant, that is its decision to make. 1.c. Issue: Clarify that exemptions already granted will not be affected by new rulemaking. Response: The NRC agrees that the reasons for exemptions from the current rule have not been changed by the final rule. Exemptions have been granted to four licensees of small reactors from carrying the full amount of insurance currently required by 10 CFR 50.54(w). These exemptions were based on by-case analysis of accident costs at the specific plants. Increasing the required amount of insurance based on general technical studies in no way negates the continued validity of the specific studies upon which the existing exemptions were based. Thus, the exemptions for amounts of property insurance required should remain in effect. Licensees currently exempt from full requirements of 10 CFR 50.54(w) will not be required to reapply for an exemption.

2. Method of Future Adjustment of Insurance Summary of Comments:

Almost all who commented on this issue suggested that future changes should be effected by rulemaking based on a periodic assessment of need. Some suggested updating the PNL report. Commenters generally argued that relying on some formula tied to measures of inflation or increasing insurance based on availability would not be appropriate. They believe that an inflation-based formula probably would not reflect the actual needs for insurance necessary to protect public health and safety because decontamination cost changes could not be measured accurately by any existing cost index. Additionally, changes in an index might lead to property insurance requirements that would outstrip the availability of insurance. Similarly, requirements based simply on the capacity available would not necessarily equal the amount necessary for decontamination. Insurance capacity tends to increase based on the demand for it. This demand is not only comprised of funds necessary for decontamination and debris removal, but also reflects the needs of insureds and their owners for compensation to replace the facility, a concern beyond the province of the NRC. One commenter suggested that insurers determine the future amount to be required, which presumably is akin to requiring whatever capacity is available. NRG Response: The NRC basically accepts the reasoning propounded by most commenters on this issue. Property insurance capacity available now has apparently reached, if not exceeded, the maximum amount necessary to decontaminate and clean up a large commercial reactor as determined by the PNL study. Although there may be cost increases in the future, no satisfactory formula for measuring them exists. Unlike such measures as the Consumer Price Index or the Whitman construction index, measures of decontamination costs depend on a very limited universe of experience. Furthermore, present trends suggest that growth in insurance capacity will outstrip decontamination cost escalation. It can be expected that most licensees will purchase the maximum available notwithstanding changes in NRC requirements. Of course, NRC resources might be required to perform periodic analyses to determine changes in accident recovery costs and possibly to conduct rulemaking based on these analyses. Nevertheless, as one commenter has pointed out, any future rulemakings should be substantially less onerous and involved than the present one. Even PNL technical studies should be substantially less expensive if the same cost estimation methodology were to be used. Thus, the costs to the NRC of any future rulemaking should be lower. These costs would most likely be less than the excess premium costs incurred by those few licensees who would be required to buy more insurance than they otherwise would because of any automatic upward escalation of insurance requirements as discussed above. 3. Site-Wide vs. Unit Coverage Summary of Comments: Very few commenters addressed this issue directly, although some discussed it in relation to the reinstatement considerations covered under issue "1". Consequently, most recommendations were concerned with allowing a reasonable period for licensees to obtain 50-SC-9 reinstatement of coverage for their facilities. One commenter suggested that some latitude be allowed so that different licensees sharing the same site would not be required to maintain a single insurance limit for the site. Most others strongly endorsed the NRC's acceptance of site-wide as opposed to per-unit coverage. NRG Response: The NRC continues to believe that site-wide coverage is appropriate given that general per-unit coverage has not been and probably will not be available and that the chance of a second accident occu~rii1.g at one.site during a policy year prior to reinstatement of the full policy limits is extremely remote. As indicated under issue "1", the NRC also believes three is some merit in revising the rule to allow reasonable steps to be taken for reinstatement. With respect to the comments that provision be made for units of different licensees at the same site, the NRC agrees. Although per-site coverage is acceptable where a single -licensee owns more than one reactor, separate coverage is needed for different licensees operating different units at the same site. This is because differences in factors such as reactor design and utility management can result in different risks for different licensees. In addition, separate coverage conforms to the manner in which reactor property damage insurance is offered and held. For these reasons, the final rule requires each licensee to obtain property damage insurance for its nuclear facilities. Thus, different licensees operating different units at the same site would each have to provide coverage for their reactor units. 4. State Laws Prohibiting the Purchase of Certain Insurance Summary of Comments: Relatively few comments were received on the . issue of what should be done about those utilities subject to State laws or constitutional provisions prohibiting the purchase of mutual or retroactive insurance. 2 All those commenting on this issue insisted that the NRC must show greater flexibility to those utilities prohibited from buying full insurance coverage. Commenters suggested that 2 Prohibitions generally take two forms. One prohibits public entities in some States from owning stock in private corporations. Such prohibitions have been construed to include the purchase of insurance from mutual insurance companies whereby the insured becomes an owner of the company. A second prohibits public entities from extending credit to private corporations, an acti?n interpreted to include the payment of retrospechve premiums. The insurance offered by NML and NEIL-II is mutual insurance and may require the payment of retrospective premiums. PART 50

STATEMENTS OF CONSIDERATION

§ 50.54(w](3) 3 of the existing rule should be retained. This section provides some leeway for affected utilities either to obtain alternate coverage or to be exempted from unachievable NRC requirements. Several commenters indicated that they have taken a variety of steps to attempt to purchase additional insurance. One commenter described recent efforts to amend the sections of the Texas Constitution pertaining to the prohibitions against purchasing insurance. The only way the Texas Constitution can be amended is by referendum and the proposal submitted to Texas voters in November 1984 was defeated. Another commenter indicated that the Nebraska Constitution prohibits public entities in the State from becoming "a subscriber to the capital stock, or owner of such stock, or any portion of interest therein of any railroad, or private corporation, or association" (Article XI, Section I]. The New York Power Authority stated that NEIL-II excess insurance is unavailable to it at the present time but did not explain why. Some commenters indicated that they have made efforts to secure equivalent protection, including surety bonds or letters or lines of credit. However, instruments in the amount necessary to comply with the rule are not available. Even if available, however, these commenters concluded that "the existence of such protection on a liability (however contingent], backed by no assets and not covered by any current revenues, would effectively preclude them from future financing needed to maintain, improve, and expand their physical plant." Commenters concluded, therefore, that they would be unable to furnish evidence of equivalent protection in

accordance with the proposed rule. Commenters suggested that three alternatives exist. First, affected utilities could continue to explore with NEIL-II the possibility that through amendment to its by-laws and charter and to its current policies, NEIL-II could offer insurance structured to comply with State law. Nebraska publicly-owned utilities have done this and are awaiting a decision by the Nebraska Supreme Court. Second, the relevant portions of State laws or constitutions could be amended. However, as indicated above, this could be a protracted process with no assurance of success. Third, licensees could seek exemption under 10 CFR 50.12 from the proposed 3 This paragraph reads, "When a licensee is prohibited from purchasing on-si le property damage insurance because of state or local law, the licensee shall purchase the specific amount of such insurance found by the NRC to be reasonably available to that licensee, or to obtain an equivalent amount of protection

... " amendments. Several commenters mentioned that if these alternatives were unavailable, they would probably be forced to shut down their reactors at enormous cost. Finally, one commenter suggested that one result of the NR.C's proposal in this area was to inject itself into the regulation of terms and conditions of property insurance, a result the NRC previously indicated it wished to avoid. NRG Response: The proposed rule was designed to allow utilities either to purchase the requisite insurance or to demonstrate an equivalent amount of protection by alternate means. When the Commission published the proposed rule for public comment, it believed that such alternate means of financial assurance would be available. This assumption formed the basis for the Commission's conclusion that it probably could not promulgate a rule that would preempt the State prohibitions at issue. Based on the comments received, however, it now appears that alternate means of protection might not be available to some utilities. As noted above, these utilities indicated that surety bonds or letters or lines of credit are not available in the amount necessary to comply with the proposed rule. Self insurance, although discussed as a possibility in the legal analysis of the preemption question, would not be acceptable because the potential liability would greatly exceed the net assets of the affected utilities. Thus, a guarantee based on the financial strength of the licensee would be insufficient. A trust fund based on collections from customers would be prohibitively . expensive if funded immediately, because it would require collecting the entire amount in advance. If funded over a period of time, it would not meet the Commission's objective of providing financial assurance for decontamination and cleanup because an accident could occur before the full amount had been collected. In these circumstances, the Commission believes that a clear conflict exists between the requirements of Federal and State law. The Commission has determined that $1.06 billion of nuclear property damage insurance is needed to stabilize, decontaminate and clean up a reactor after an accident so as to mitigate potential threats to the health and safety of workers and the public and to the environment. Accordingly, the Commission is requiring reactor licensees to purchase this insurance. Some State laws or constitutions prohibit certain licensees from purchasing mutual or assessment insurance. No other insurance is available in an amount sufficient to comply with the Commission's 50-SC-10 requirement. There are no reasonable alternative means of financial assurance available to these licensees. Based on these changed circumstances, the Commission believes that its property insurance rule would preempt conflicting State laws. There can be no doubt that this rule is concerned with reactor safety. Indeed, in discussing the issue of whether to impose a decontamination priority in the proposed rule the Commission stated: In fact, the Commission has no reason to impose a property insurance requirement other than to protect the public health and safety. Proceeds from insurance would be used both to assure that contamination from a reactor immediately after the accident did not threaten public health and safety and the environment and to eliminate delays and degradations to the cleanup process that could cause threats to health, safety and the environment over time. Thus, the rule operates in an area of exclusive Federal control. Although the State prohibitions at issue do not attempt to regulate within this area, they prohibit what the Commission's rule requires. They would, therefore, be preempted because they are in direct conflict with Federal law. It should be noted that "[f]ederal regulations have no less preemptive effect than federal statutes." Capital Cities Cable, Inc. v. Crisp, 81 L.Ed.2d 580, 589 (1984), quoting Fidelity Federal Savings and Loan Assn. v. De La Cuesta, 458 U.S. 141 (1982). In addition, the State prohibitions concern insurance, a matter that the States have traditionally regulated. Accordingly, Congressional intent to preempt must be "clear and manifest." Rice v. Santa Fe Elevator Corp., 331 U.S. 218, 230 (1947). Such intent is present, however, when the requirement of Federal and State law are in conflict. See Florida Avocado and Lime Growers v. Paul, 373 U.S. 132, 142-43 (1963); Ferebee v. Chevron Chemical Co., 736 F.2d 1529 (D.C. Cir. 1984), cert. denied, 469 U.S. 1062 (1984). A conflict with Federal law arises when compliance with both Federal and State law is impossible or the State law frustrates the accomplishment of a Federal objective. Pacific Gas and Electric Co. v. State Energy Resources Conservation and Development Commission, 461 U.S. 190, 204 (1983). In Capital Cities, supra, the Supreme Court held that an FCC regulation requiring cable television operators to carry certain broadcast signals without alteration preempted an Oklahoma constitutional provision prohibiting the advertising of alcoholic beverages within the State. State law compelled deletion of wine commercials, an action that Federal law prohibited. Cable operators who complied with Federal law were subject to criminal prosecution under State law. PART 50

STATEMENTS OF CONSIDERATION The Court found a conflict based on the impossibility of complying with both Federal and State law, as well as obstruction of the Federal objective of increased program diversity.

81 L.Ed.2d at 593-94. Similarly, it is impossible for a reactor licensee to comply with both the Commission's rule requiring the purchase of nuclear property damage insurance and the State constitutional provisions prohibiting the purchase of that insurance. In addition, the State prohibitions interfere with the Federal objective of exclusive Federal regulation of reactor safety. Thus, the provisions of Federal and State law are clearly in conflict. This conclusion is consistent with recent Supreme Court decisions on Federal preemption under the Atomic Energy Act, notwithstanding the absence of a conflict in those cases. In Pacific Gas, supra, the Court upheld California's moratorium on the certification of new nuclear power plants pending Federal approval of a demonstrated technology or means for the permanent disposal of high level radioactive wastes. The Court reasoned that, under the Atomic Energy Act, Congress provided for a system of dual regulation of nuclear-generated electricity in which the Federal Government maintains complete control over reactor safety and the States exercise their traditional authority over economic questions, such as the need for power, type of facilities to be licensed, land use, and ratemaking. Because the moratorium was based on economic rather than safety concerns and did not conflict with Federal safety regulations, the Court held that the California statute was not preempted. The Court relied on the fact that the moratorium would only affect reactors at the planning stage, cautioning that "[s]tate regulations which affected the construction and opera lion of federally approved nuclear power plants would pose a different case." 461 U.S. at 223 n.34. The State prohibitions at issue here . affect a reactor licensee's ability to comply with the Commission's safety requirements. Thus, they would affect the operation of a Federally licensed reactor and would not escape Federal preemption under the rationale of Pacific Gas. In Silkwood v. Kerr-McGee Corp., 464 U.S. 238, (1984), the Supreme Court held that the Atomic Energy Act does not preempt an award of punitive damages under State tort law for personal injuries or property damage from radioactive materials. The Court examined the legislative history of the Price-Anderson Act and found ample evidence that Congress had assumed that remedies under State tort law would continue to be available to persons injured by nuclear incidents. There was no conflict with Federal law because licensees could pay both Federal fines and imposed punitive damages. Citing a failure of proof, the Court declined to decide whether an award of damages could be so large as to conflict with the NRC's policy of avDiding penalties which would put a licensee out of business or have an adverse effect on safety. Thus, the Court did not preclude the possibility of Federal preemption in an appropriate case based on a conflict with Federal law. Finally, the Commission rejects the notion that a conflict with Federal law could be avoided by the licensee's option of ceasing to operate the reactor. Rather, the Commission believes that in analyzing this issue, a court would conclude that a conflict exists where the licensee's only option is to withdraw from its Federally licensed activities. This is suggested not only by Pacific Gas, supra, but by other Federal 'preemption cases. See, e.g., Ferebee, supra, 736 F.2d at 1541; Douglas v. Seacoast Products, Inc., 431 U.S. 265 (1977); Sperry v. Florida, 373 U.S. 379 (1963). The Commission believes that, where a reactor licensee has no reasonable alternatives to purchasing nuclear property damage insurance, the State prohibitions at issue would conflict with Federal law. For all the foregoing reasons, the Commission concludes that its property insurance rule would preempt these State prohibitions.

5. Decontamination Priority Summary of Comments:

The great majority of comments received on the proposed rule focused on the NRC's proposal to require some form of decontamination priority, that is, proceeds from insurance should be used to decontaminate and clean up after an accident before any other purpose such as facility restoration or payment of investors. Although the comments in response to the Advance Notice of Proposed Rulemaking (ANPRMJ published on June 24, 1982 (47 FR 27371) evoked a similar degree of interest in the issue of decontamination priority, comments on the proposed rule provide more detail and elaboration of the reasons for the extensive opposition to the NRC proposal. The NRC proposal was drafted as a modified decontamination priority 4 4 As proposed, paragraph 50.54(w)[3) reads, "The proceeds of this insurance shall be used first to decontaminate the licensed reactors before any other purpose when and to the extent that such decontamination is required to protect public health and safety and is so ordered to be used by the Director of Nuclear Reactor Regulation." 50-SC-11 with hope that earlier objections by utilities could be ameliorated. In the ANPRM, utilities had objected to an absolute decontamination priority because such a priority: (1) Would conflict with trust indenture language and fuel leasing agreements that require licensees in general to maintain sufficient insurance to protect bondholders and fuel lessors in case of damage to property; (2) would be unnecessarily restrictive in allowing utilities to respond to an accident and thus would inhibit an expedited recovery and adversely affect public health and safety; (3) could, because of reduced investor protection, lead to reduce access to funds from investors or increased cost of funds; and (4) would be unnecessary because of the existing 1 priority offered in the excess layer through NEIL-II. (This priority covers ", .. all expenses necessarily incurred in discharging the obligation or liability of the Insured(s) to remove debris of and to decontaminate the property covered by this Policy following direct physical damage to such property caused by any peril covered under the Underlying Property Policy ... " (see NEIL's Decontamination Liability and Excess Property Insurance Policy, p. 12)). Many of these same arguments were raised again by commenters in response to the modified decontamination proposal. Some commenters indicated that a modified priority, while not as seriously affecting finances as an absolute priority, would still have substantial adverse impact. This point was made most comprehensively by the Edison Electric Institute (comment 13) and endorsed by several other commenters. It stated in part: While the EEi recognizes that the Commission sought to afford utilities some financial flexibility lo respond lo any future

accident by giving the Director of Nuclear Reactor Regulation the authority to require payments under the decontamination priority, the sole substantive standard guiding the use of this authority is that he may direct insurance proceeds be used 'to the extent that such decontamination is required to protect public health and safety.' ... While this approach provides more flexibility than an absolute decontamination priority, we continue to have serious concerns about this proposal.

In particular, there seems to be virtually no limit upon the discretion of the Director, other than the 'public health and safety' standard. If an accident were to occur, we fear that the Director would immediately impose the decontamination priority for all insurance proceeds, whether or not all funds are necessary to protect public health and safety since this step maximizes his future options. Once this happens, there would be no administrative mechanism to cause a reconsideration of this decision. Moreover, PART 50

STATEMENTS OF CONSIDERATION the Director is likely to be under pressure from multiple sources to (1) require decontamination prior to other actions that may be appropriate to restore the property to operating condition and (2) manage the course of repair and clean-up through his control of the insurance proceeds.

These factors cause EEi to fear that the broad discretion vested in the Director over the use of proceeds from the decontamination priority may effectively deprive utilities of any of the benefits of property insurance. In effect, the Director would become the ultimate manager of the decontamination and repair operation. EEi believes one way to balance the need for flexibility to respond to an accident with the need to limit the Director's authority over property insurance proceeds is to restrict the time period in which any order regarding the use of proceeds from property insurance for decontamination remains in effect. Limiting the dura lion of any such order will ensure that post-accident response decisions are periodically reevaluated as conditions change and new information becomes available. EEi proposes that the Director's initial determination respecting use of the property insurance proceeds for decontamination shall issue a new order regarding the use of such proceeds. This second order would remain in effect for no more than six months. It could be followed by new orders which terminate every six months. All such orders should be final orders which are subject to judicial review. Each should reveal the rationale, basis and record which are relied upon support the Director's determination. In addition, the Director should sequest [sic] only those portions of the insurance funds reasonably needed for decontamination so that funds also would be available to carry out other activities necessary to restore the utility's property. Finally, EEi interprets the Commission's proposal to allow the Director to act expeditiously without holding hearings since that could seriously delay the decisionmaking process. In order to avoid any confusion, we suggest the Commission confirm our understanding of this process. Other commenters suggested similar problems. For example, by not establishing standards or criteria by which the Director of NRR is to order a decontamination priority, the resulting uncertainty would be as unsettling to investors as a more rigid priority. Further, such an order would probably be subject to a protracted review process, open to intervention, and might. so tie up funds that the cleanup process could be hampered and public health and safety adversely affected. One commenter stated that, although the NRC has authority to impose decontamination requirements on a licensee, it does not have authority "to dictate either to insurers or to indenture trustees how they are to deal with the insurance proceeds" (comment 15). Similarly, another commenter suggested that requiring a decontamination priority would preempt the coverages filed by ANI/MAERP with State insurance departments and would thus be in effect a preemption of State law, a result the NRC sought to avoid (comment 7, p. 2). Another commenter analyzed several scenarios in which an accident caused varying amounts of damages and the resulting actions that probably would be taken by the trustee (Comment 15, p. 2): If the damaged plant can be reused, decontamination would be a necessary element of the repair, and the trustee would release insurance proceeds. After initial decontamination, repairs would proceed simultaneously with decontamination. The only effect of an order establishing a priority for decontamination would be to forbid use of the funds acquired from the trustee for anything but decontamination. This would presumably require the licensee to segregate and invest the portion of the insurance proceeds trhat would be spent on repairs rather than decontamination and to use other sources of funds for the repairs. When the priority order was lifted, the segregated funds could then be used to replace the funds spent on repairs. In short, the priority order would cause some added bookkeeping without serving any useful purpose. If the damaged plant could not be restored to use, the trustee might refuse to release funds for decontamination on the ground that they would not be used for repairs or replacement. However, the utility could still obtain the funds if it could provide additional unhanded property. While the utility would probably use the funds for decontamination without compulsion, NRC could require it to do so even in the absence of the proposal in § 50.54(w)(3). If the licensee did not have additional unbonded property to offer the trustee, it could obtain the proceeds only on the basis that it would use them to build or acquire new public utility property. Thus, an NRC priority order would require the use of the proceeds for decontamination while the indenture would forbid their use for that purpose, and the licensee might be unable to reach the proceeds. This could push the licensee into insolvency and possibly bankruptcy. We doubt that reorganization would be appealing to NRC since it might find itself with the responsibility for the cleanup and the possibility that in the post-bankruptcy period it could not recover the cleanup costs from the licensee. Absent a priority, the trustee, in its discretion, could still refuse to relinquish the proceeds if the licensee was unable to substitute additional property. But this is not a position it would take lightly. Such a step could also push the licensee into bankruptcy, and the trustee would run the risks that (i) its efforts to use the proceeds to benefit bondholders would be automatically stayed and [ii) it would be forced to turn the proceeds over to the licensee or reorganization trustee for use in bankruptcy operations. 50-SC-12 In the case of a possible bankruptcy, we believe that, as a practical matter, the licensee, its creditors and NRC would have a common interest in working out a plan that allowed the licensee to continue as an entity while decontaminating to the extent necessary to protect public health and safety. Our analysis is that the proposed priority could not facilitate that result and in some circumn*stances would impede it. We therefore urge that NRC reject 50.54(w)(3). A somewhat different conclusion was reached by another commenter. This commenter suggested two approaches. First, perhaps a licensee could "identify those decontamination costs which are not treatable as capital repairs and cause them to be insured separately or at least paid separately under existing insurance. It may then be possible to conclude that such insurance is not property insurance which must be payable to mortgage trustees." (Comment 20, p. 2-3). This approach would allow release of a significant amount of funds not subject to the control of the trustee. Second, funds would be released by the trustee if they were to be used for repairs that would be treated as additions "chargeable to plant accounts in accordance with sound accounting practice." (Ibid, p. 3) However, this commenter concludes, To the extent that decontamination expenses do not constitute repairs so chargeable, the trustee may refuse to release the funds unless the Company happens to have available property additions which have not previously been tendered to the trustee. There can certainly be no assurance that the Company will have independently available property additions to tender to the trustee to obtain such release. The Company simply may be unable to obtain release of funds to expend on decontamination, at least to the extent that such expenditures do not constitute repairs under the provisions of the mortgage. (Ibid, p. 3) Other commenters emphasized the impact on investor perceptions. One commenter indicated that because of the lack of clear standards for imposing a decontamination priority, bondholders probably would not have a "significantly greater sense of security for their investment than they could have had under the 1982 proposal of an absolute priority." (Comment 18, p. 4). Further, "existing bondholders purchased their bonds with an understanding that Federal regulatory policy would allow utilities to protect their investment with insurance." (Ibid, p. 5). A letter from Morgan Stanley & Co. was enclosed with EEi's comments to support this view. This letter states that particularly since TMI, investors expect that, "in the event of an accident, a portion of these insurance proceeds will PART 50

STATEMENTS OF CONSIDERATION be available to repair or replace their investment." Finally, the comment from the insurance pools suggested that they were able to pay the owner of TM! Unit 2 $68 million for loss to their nuclear fuel, even though decontamination was barely commenced.

This contributed to regaining financial stability and made it more likely that the owners could address the range plans to decontaminate the reactor. (Comment 31, p. 2). Despite their strong reservations about a decontamination priority, several commenters proposed ways of making such a priority less burdensome. Suggested revisions to the proposed rule would: (1) Provide specific criteria for determining when public health and safety is endangered, such as by following limits used in 10 CFR Part 20; (2) establish "sunset" provisions so that the Director's order would last only for a specified time, e.g., three to six months subject to renewal; and (3) have the Director of NRR issue a show cause order requiring the licensee to explain its prepared work plan and schedule and related expenditures are in the public interest. Following a hearing on the show cause order, the Commission could issue a modifica lion of the licensee's work plan or schedule. Until an order was issued, the licensee would be able to proceed with its work plan. Others suggested that the term "decontamination" be defined. These commenters argued that preparatory expenditures, such as stabilization costs and purchase of equipment and other materials necessary for cleanup might not be included in the definition of decontamination priority. Similarly, the degrees of decontamination should be tied to avoiding "exposure of the off-site public to radiation levels exceeding those allowed by NRC regulations" (comment 18, p. 6) or should otherwise indicate the level above which health and safety is endangered (comment 16, p. 3). One comment addressed the specific issue of timing the implementation of a decontamination priority to coincide with the renewal dates of the policies. Because policies issued by NML and NEIL-II are, in effect, bilateral contracts, changes would be difficult to make without mutual consent, or without awaiting policy expiration. In preparing the proposed rule, the NRC relied on comments and proposals submitted by the Association of the Bar of the City of New York (the Association] in response to the 1982 ANPRM. As indicated in the proposed rule (49 FR 44647, Co. 3), the Association determined that utility trust indentures do not in general give bondholders any vested rights to a given amount or type of coverage. It also recognized that to respond properly to a nuclear accident, a licensee may be required to take a range of actions apqrt from decontamination and debris removal. Consequently, this commenter favored priority for payment of decontamination and debris removal expenses only insofar as it is "necessary to remove any significant health or safety hazard." The Association suggested this goal could be accomplished if a regulation were properly drawn, although it proposed no wording for such a regulation in its comments. In responding to the proposed rule, however, the Association changed the focus of its comments. Although the Association indicated that it supported "in principle the form of decontamination priority which the Commission has proposed," it believes that, "as now proposed, the amendments to the Commission's property insurance regulations may not effectively provide for the decontamination priority which the Commission desires to achieve ... that in the event of a serious nuclear accident, the utility's indenture trustee may refuse to release property insurance proceeds for decontamination or debris removal purposes." (Comment 12, p. 5). The Association believes that "only by restricting payment of insurance proceeds to the trustee can there be any reasonable assurance that the proposed decontamination priority will prove effective." (p. 6). Such a restriction of payment would not be effective merely by payment directly to the utility because funds would still be vulnerable to creditor delays due to bankruptcy or insolvency. The Association perceives a fundamental conflict between the Atomic Energy Act, NRC regulations, and the terms of the license on the one hand and the provisions of the Bankruptcy Code on the other. The Association presents a legal analysis and synopsis of the arguments that the parties might use in obtaining property insurance proceeds and concludes from its analysis that a decontamination priority can only be effectively implemented if the decontamination proceeds are paid neither to the indenture trustee nor the utility itself but rather into a separate trust fund previously established for that purpose. The proceeds under the insurance policy would be paid to the trustee of the trust fund and disbursed to pay for costs incurred in decontaminating the reactor and removing radioactive debris. The remaining amounts, up to the limits of the policies. would then be available to pay for property damage and these funds would be paid to the indenture trustee." (p. 13}. 50-SC-13 Finally, the Association believes that such an approach should not conflict with utility indenture provisions because such indentures generally require a utility to maintain property insurance to the same extent as companies similarly situated and opera ting like properties, and not a particular level of coverage. Primary policies offered by ANI/MAERP and NML would have to be modified analogously to the NEIL-II excess policy, yielding a hybrid decontamination liability and property insurance policy. NRC Response: The NRC disagrees that a decontamination priority would conflict with bond indenture language. As the Association indicated, such language typically would require a utility "to insure its property against loss or damage to the same extent that property of a similar character is usually so insured by companies similarly situated and operating like properties." 5 The NRC agrees with the Association's conclusion that such language would allow the NRC to impose a decontamination priority because all utility licensees would face similar conditions, would be insured "to the same extent" and would thus comply with indenture language. More recently, the Association concluded that, after a large accident, conflict between the interests of bondholders as represented by their trustees on the one hand, and the NRC on the other, could seriously impede, in some circumstances, recovery from an accident. The Association's recommended solution, i.e., to require all insurance proceeds to be deposited in a trust apart from the utility or bondholder's trustee, would provide additional assurance that funds would be available for accident decontamination and cleanup but could exacerbate this conflict, if imposed. The extent to which this conflict would adversely affect investor perceptions and thus increase utility cost of capital and reduce access to capital is speculative. It is possible that utilities would incur some, perhaps significant, increased capital costs if a full decontamination priority were imposed. However, there are a number of factors which should temper the 'One commenter indicated that its mortgage provisions require it to maintain insurance in "a reasonable amount against loss or damage by fire and from other causes customarily insured against by similar companies." [Comment 20). This is not appreciably different from the language cited by the Association. The staff has attempted without success to obtain assistance from the Federal Energy Regulatory Commission (FERCJ and others to determine the extent to which indenture language varies. PART 50

STATEMENTS OF CONSIDERATION rational investor's possible disquiet.

First, as some commenters pointed out, the NRC has the authority under sections 161, 182(a), and other sections of the Atomic Energy Act to impose a decontamination priority after an accident whether or not such a priority is actually spelled out in 10 CFR 50.54(w). Given this, it can be argued that a conflict existed all along, although perhaps not known to certain investors. Second, investors now are covered for a relatively small portion of their investment, although it has increased from the maximum $300 million coverage at the time of the TMI-2 accident. Because NEIL-II coverage has its own decontamination priority, 6 the most coverage investors could currently expect is $620 million. When compared to the $5 billion or more to construct the la test nuclear power facilities, an investor would be covered for little over 10% of the value of the investment. The NRC doubts that an investor would derive security from this limited coverage but would tend not to invest if an NRC decontamination priority were imposed. Third, commenters essentially ignored in their comments the rapid growth in insurance capacity. The NRC cannot speculate on whether and to what extent capacity will grow further. Nevertheless, because of the shortfall in coverage as compared to facility value, it would be expected that demand for, if not supply of, insurance will remain high. If capacity continues to grow at a pace that substantially exceeds possible increases in estimated facility decontamination costs, it would be expected that investors would find a decontamination priority progressively less onerous as more funds would be made available exceeding those subject to a priority. Fourth, as some commenters have suggested, a decontamination priority is only necessary for a scenario in which a plant is completely written off and the interests of the NRC and indenture trustee would conflict to the extent that bankruptcy might occur. Without a priority, some commenters argue that both the NRC and the trustee would have incentive to work out a plan "that allowed the licensee to continue as an entity while decontaminating." (Comment 15) Again, the NRC disagrees that a decontamination priority would destroy that incentive. If the priority 6 The fact that NEIL-I! itself has a decontamination priority raises the question of whether NEIL-H's priority would not also conflict with bond indenture language. The comments directed to this area did not differentiate between the adverse effects that the NEIL-I! priority and the NRC priority might have. itself ever became an impediment to accident recovery and resulted indirectly in a threat to public health and safety, it could be rescinded or made part of a broader recovery framework as the previous commenter discussed. The NRC, no less than the investors, would not wish to precipitate bankruptcy and so impede accident recovery. The NRC rejects the argument that a decontamination priority would reduce flexibility in responding to an accident. Obviously, the NRC would not interpret a priority in so rigid a manner as to preclude prudent practices necessary to an orderly decontamination, such as equipment purchases, stabilization activities, etc. The decontamination priority was not meant to be applied sequentially in that all expenditures on cleanup would have to be made before any others. The priority has been worded to allow licensee flexibility, particularly after a reactor has been stabilized after an accident. Despite possible utility reluctance, the priority should be compatible with the broadest range of actions necessary to protect public health and safety. Further, the decontamination priority is meant to be invoked only when there would be serious concern over the availability of funds for decontamination. Although most commenters opposed imposition of a decontamination priority, many did recommend changes to the wording of the priority that would make it less onerous. One change, suggested with slight variation by several commenters, would require that a definite time limit be established for an order by the Director of NRR. Some suggested that a time limit, of three or six months duration, could be extended as necessary. The NRC agrees that periodic reevaluation of the need to continue the Director's order is desirable and is thus incorporating a variety of provisions in the final rule relating to duration of the priority. However, the NRC believes that, as a practical matter, orders for decontamination priority would be extended as necessary to protect public health and safety. Thus, the principal effect of sunset provisions would be to allow for additional consideration as each order was replaced or extended. Concerning the matter of hearings, the mechanism for imposing the decontamination priority would be an order to show cause by the Director of NRC. The Commission's rules in 10 CFR Part 2, Subpart B, afford the licensee the right to demand a hearing when the NRC staff seeks to impose requirements by order. In addition, any person whose 50-SC-14 interest may be affected by the proceeding could request a hearing or file a petition to intervene. If the Commission followed its usual practice of confining the scope of the proceeding to whether the order should be sustained, only pesons opposing the order could request a hearing or petition to intervene. Petitioners who did not object to the order but might seek further corrective measures would lack the requisite interest in the proceeding. See Bellotti v. NRC, 725 F.2d 1380 (1983). Thus as a practical matter, the Director's ability to impose the priority without a hearing would most likely depend on the licensee's response to the order. With regard to the commenter's concern about delay in the decisionmaking process, .the Commission notes that, even if a hearing were held, insurance proceeds would not necessarily be tied up. This is because the Commission's authority to impose the priority before holding a hearing would depend on the circumstances. Other suggested changes to the actual wording of the priority concern the definition of the degree to which public health and safety should be protected. Some commenters recommend defining what is meant by protection of public health and safety; others suggest referencing the radia lion protection guidelines used in 10 CFR Part 20. The NRC favors tying decontamination limits to 10 CFR Part 20 standards for radiation protection and the ALARA principle (as low as reasonably achievable). Notwithstanding reliance on these radiation protection standards, the NRC believes it is also necessary to work with a decontamination plan tailored both to the specific problems and characteristics of the particular site suffering the accident and to the general characteristics that differentiate decontamination after an accident from decommissioning. This approach would allow both the NRC and the licensee greater discretion in initiating and completing a safe recovery and would be particularly desirable because accidents are expected to be rare and, to a large degree, unique. The NRC also believes that some commenters have confused the purpose of this rule-to provide adequate funds for recovery after an accident-with the process of accident recovery itself which is more thoroughly covered in other parts of the NRC's regulations. This rule applies to decontamination after an accident; it does not encompass decommissioning. The NRC realizes that there may be an overlapping area of PART 50

STATEMENTS OF CONSIDERATION tasks which cannot be defined as uniquely decontamination or decommissioning.

For example, certain components would be removed and disposed of whether as a result of cleanup following an accident or as part of decommissioning operations at the end of a plant's useful life. The NRC has attempted to minimize this overlap, but acknowledges that it cannot be eliminated. The NRC understands, as confirmed by ANI officials, that insurers would likely pay for activities required or components damaged as a result of the accident. Insurers would not distinguish between accident decontamination and decommissioning when paying out insurance proceeds, as long as a licensees could offer proof of causation of damage. If a reactor suffered an accident so severe that restart would not be possible, the reactor would have to be decommissioned. In this situation, the distinction between decontamination and decommissioning would be difficult to maintain. For example, a licensee could use property insurance proceeds for decontamination activities, such as removal and disposal of certain components, that under normal circumstances would clearly be considered part of the decommissioning process. As a result, the licensee might be able to preserve some decommissioning funds for other purposes following completion of decommissioning. Conversely, a licensee might try to draw upon funds reserved for decommissioning to perform decontamination tasks if property insurance proceeds were expended prior to completion of decontamination. If an accident were severe enough to prevent restart of the reactor, however, essentially all of the property insurance proceeds would likely be needed for decontamination. Because property insurance also covers replacement of components, equipment, and structures, the insurance proceeds would be insufficient under current (and likely future) limits to cover all replacement as opposed to decontamination expenses. Thus, it is likely that all property insurance proceeds would be paid out regardless of whether used for accident decontamination or decommissioning activities. In sum, the Commission is implementing a decontamination priority further modified to reflect many commenters' concerns. The section of the rule containing the priority, 19 CFR 50.54[w)(3), begins by establishing a priority for stabilizing the reactor after an accident so as to prevent any significant risk to the public health and safety. After the reactor is safe and stable, the licensee is required under section 10 CFR 50.54(w)(3)(ii) to submit a cleanup plan that identifies all cleanup operations necessaNy to bring the reactor to the point of decommissioning or restart. Various cleanup operations are indentified and reference to 10 CFR Part 20 occupational exposure standards is made so as to differentiate between decontamination after an accident and decommissioning. Section 10 CFR 50.54(w)(3)(iii) addresses the scope of the decontamination priority. With respect to the Association's recommendation that all insurance proceeds be placed in trust so as not to be available to the bondholder's trustee, the NRC believes this requirement provides additional protection of public health and safety. As explained in the Association comments, utility bond indentures typically require available property insurance proceeds to be paid directly to the indenture trustee, not to the utility. This means that in the event of a serious accident, the insurance policy proceeds would not be under the control of the utility. If the NRC were to order the utility to spend all or part of the proceeds for protection of the public health and safety, the utility would be powerless to do so without asking for and receiving the proceeds from the trustee. However, the bondholder trustee might not be willing or even legally able to release the proceeds. The trustee's obligation is not to the public health and safety, but to bondholders, and the bondholders are interested in preserving their investment. Thus, utility indentures generally limit the trustee's ability to pay insurance policy proceeds to the utility to circumstances where the funds will be used for repairs to or replacement of the damaged property and the bondholder's interests will thereby be protected. Most importantly, if a utility were forced into default because of the financial consequences of a serious accident, the trustee might be legally prohibited from paying policy proceeds to the utility. As the Association points out and NRC's own experience confirms, the NRC's ability under the Bankruptcy Code to get priority for expenditures of funds for safety is very uncertain. For these reasons, the Commission is adopting the Association's proposed approach to the decontamination priority. Thus, 10 CFR 50.54(w)(4) requires that the policy proceeds be paid to a separate trust fund established for the sole purpose of protecting the public health and safety. The NRC also believes that an approach worth further evaluation is to 50-SC-15 seek legislation in Congress that would give preference in any bankruptcy proceeding to expenditures that mitigate threats to public health and safety. The NRC is currently studying the feasiblity of this approach.

6. Other Issues Summary of Comments:

Very few comments were received that were not related to the previously discussed issues. One commenter endorsed NRC's position stated in the proposed rule of not becoming involved in regulating insurance terms and conditions (comment 1). Another endorsed the NRC's position of not requiring licensees to carry coverage from both primary insurers [i.e., NML and ANI/MAERP) [comment 19). Finally, one commenter suggested that the term "financial protection" not be used in 10 CFR 50.54(w)(2). "Financial protection," as defined in the Atomic Energy Act and 10 CFR Part 104, is used in a specific sense not meant in Part 50. NRG Response: The NRC essentially agrees with these comments and is incorporating them in the rulemaking. The NRC agrees that the term "financial protection" might be misleading to some in the context used in Part 50. Thus, the rule will be revised to use the less ambiguous term, "financial security." Environmental Assessment and Finding of No Significant Environmental Impact These amendments to 10 CFR 50.54(w) will increase the amount of insurance that each commercial reactor licensee is required to maintain to clean up a licensed reactor site after an accident. The amount of required insurance will increase from a minimum of $620 million currently required to $1.06 billion. The rule also adds a requirement that proceeds from insurance must be used first to stabilize and then decontaminate the licensed reactors before any other purpose when and to the extent that decontamination is required to protect public health and safety and is consistent with the Commission's 10 CFR Part 20 radiation protection standards. These actions are required to provide greater assurance that commercial reactor licensees will have sufficient funds to clean up their reactors following an accident. Assurance of these funds is required so that public health and safety is not adversely affected during the cleanup process. Alterna lives to this action consist of maintaining the existing rule or establishing some other limit of insurance. Neither this action nor the alternatives to it have any significant impact on the environment. No other PART 50

STATEMENTS OF CONSIDERATION agencies or persons were contacted for this action. Consequently the Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required.

Although changes in insurance requirements affect the financial arrangements of licensees and have economic and social consequences, they do not alter the environmental impact of the licensed activities. As determined in the above environmental assessment, the alternatives to the proposed action likewise do not have any significant impact on the environment. No other documents related to this proposed action exist. The foregoing constitutes the environmental assessment and finding of no significant impact for this final rule. Paper Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of1980 (44 U.SC. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval number 3150-0011. Regulatory Analysis The Commission has prepared a regulatory analysis for this regulation. This analysis examines the costs and benefits of the alternatives considered by the Commission. Interested persons may examine and copy for a fee the regulatory analysis at the NRC Public Document Room, 1717 H St. NW., Washington, DC. Single copies of the analysis may be obtained from Robert S. Wood, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone (301) 492-8413. Backfit Analysis This final rule requiring an increase in property damage insurance does not require "the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct, or operate a facility." Accordingly, this action is not a backfit as defined in § 50.109. However, the staff has prepared an analysis of the rule's impact in light of the factors listed in§ 50.109(c). This analysis may be examined or obtained in the same manner as the regulatory analysis mentioned previously. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980 (5 U.S.C. 605[b)), the Commission cei:tifies that this rule will not have a significant economic impact on a substantial number of small entities. This rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, and the NRC is adopting the following amendment to 10 CFR Part 50. 52 FR 31601 Published 8/21/87 Effective 8/19/87 Statement of Organization and General Information See Port 1 Statements o! Consideration 52 FR 41288 Published 10/27/87 Effective 11 /27 /87 10 CFR Part 50 Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Commission is amending its regulations to broaden the scope of a recent modification to General Design Criterion 4 (GDC-4). The amendment would allow the removal of numerous pipe whip restraints and jet impingement barriers as well as other related changes in all reactor types. Implementation of the amendment will increase safety since inadvertent restriction of pipe growth due to thermal i 50-SC-16 expansion and associated stresses leading to pipe cracking is avoided. Also, the duration of inservice inspection will be reduced, yielding substantially less occupational exposures.

EFFECTIVE DATE: November 27, 1987. ADDRESSES: Copies of the written public comments are available for public inspection and copying for a fee at the NRC Public Document Room at 1717 H Street NW., Washington, DC. FOR FURTHER INFORMATION CONTACT: John A. O'Brien, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone (301)443-7854. SUPPLEMENTARY INFORMATION: Table of Contents !. Background II. Scope of Rulemaking Ill. Final Rule IV. Acceptance Criteria V. Invitation to Comment VI. Issues Analysis VII. Availability of Documents VIII. Finding of No Significant Environmental Impact: Availability IX. Paperwork Reduction Act Statement X. Regulatory and Backfit Analyses XI. Regulatory Flexibility Act Certification XII. List of Subjects in 10 CFR Part 50 On July 23, 1986, the Commission published a proposed rule on the board scope modification to General Design Criterion 4 of Appendix A, 10 CFR Part 50 (51 FR 26393). This proposed rule contained a summary of the acceptance criteria which the Commission had developed. A 60-day public comment period was allowed. Twenty-eight written comments were received from utilities, reactor vendors, engineering companies, industry groups, consulting firms and a citizen group. There was no overt opposition to the proposed rule; each commenter supported the proposed rule or its intent either in part or entirely. However, the citizens group expressed certain legal reservations which are addressed below in issues 20 and 21. A compilation of the twenty-one issues raised as a result of public comments and the accompanying Commission response is given under Issues Analysis. The text of the final rule is identical to the text of the proposed rule. The final rule should be applied consistently with the guidance in this Supplementary Information. I. Background Background to this rulemaking can be found in the limited scope modification to GDC-4 published as a proposed rule in the Federal Register on July 1, 1985 (50 FR 27006). Research performed by the NRC and industry, coupled with operating experience, has indicated that PART 50

STATEMENTS OF CONSIDERATION safety can be negatively impacted by the placement of protective devices such as pipe whip restraints near certain piping. The Commission adopted a step approach to the modification because safety improvements could be quickly realized without extensive and time consuming review and discussion if the scope were initially limited to the primary main loop piping of PWRs. The Commission decided not to defer the limited application of leak-before-break technology while the detailed provisions of the acceptance criteria were being reviewed and approved.

Many near term operating license (NTOL) nulcear power plant units and operating nuclear power plant units had requested exemptions from the requirements of GDC-4 and could benefit from the limited scope rule. A broader application of leak-before-break technology requires adoption of the general criteria published in NUREG-1061, Volume 3, Chapter 5, November 1984, entitled "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks". II. Scope of Rulemaking This rulemaking modifies General Design Criterion 4 to the extent that dynamic effects of pipe ruptures in nuclear power units may be excluded from the design basis provided it is demonstrated that the probability of pipe rupture is extremely low under conditions consistent with the design for the piping. Dynamic effects of pipe rupture covered by this rule are missile generation, pipe whipping, pipe break reaction forces, jet impingement forces, decompression waves within the ruptured pipe and dynamic or nonstatic pr*essurization in cavities, subcompartments and compartments. However, cavities, subcompartments and compartments necessary to the containment function are not affected by this modification. To retain high safety margins, the application of leak-before-break technology to various piping systems should not decrease the capability of _containments to perform their function of isolating the outside environment from potential leaks, breaks, or malfunctions within the containment. Containments will continue to be designed to accommodate loss of coolant accidents resulting from breaks in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Also, the functional design for emergency core cooling systems still retain.s nonmechanistic pipe rupture. Environmental qualification of electrical and mechanical equipment is discussed under issue 4 below. This amendment to GDC-4 allows exclusion from the design basis of dynamic effects associated with high energy pipe rupture by application of leak-before-break technology. Only high energy piping in nuclear power units that meets rigorous.acceptance criteria is covered. High energy piping is defined as those systems having pressures exceeding 275 psig or temperatures

exceeding 200 'F. Studies completed by Lawrence Livermore National Laboratory under contract to the NRC indicate that adverse safety implications can result from requiring protective devices to resist the dynamic effects associated with postulated pipe rupture. The placement of pipe whip restraints degrades plant safety when thermal growth is inadvertently restricted, reduces the accessibility for and effectiveness of inservice inspection, increases inservice inspection radiation dosages and adversely affects construction and maintenance economics.

III. Final Rule The final rule consists of a substitute sentence at the end of GDC-4 (replacing the sentence introduced by the limited scope rule) permitting the use of analyses to exclude dynamic effects of pipe ruptures in all high energy piping in all nuclear power units. A deterministic fracture mechanics evaluation is mandatory. Evaluations of the potential for water hammer, corrosion, creep damage, fatigue, erosion, environmental conditions, indirect failure mechanisms and other degradation sources which could lead to pipe rupture are also required. In order to demonstrate that the probability of fluid system piping rupture is extremely low, applicants and -licensees may follow procedures and acceptance criteria developed by the staff. The supporting safety analysis must demonstrate from the results of a fracture mechanics analysis that a substantial range of stable pipe crack sizes can exist for an extended period which provides detectable leaks, and that the fluid systems piping will not rupture under these conditions consistent with the design basis for the piping. The language of the rule specifies "conditions consistent with the design basis for the piping." The design basis for piping means those conditions specified in the FSAR. as amended, and may include 10 CFR Part 50 (especially the General Design Criteria in Appendix A to Part 50), applicable sections of the Standard Review Plan, Regulatory Guides and industry standards such as the ASME Boiler and Pressure Vessel Code. The term "extremely low" is used in this amendment to GDC-4 with 50-SC-17 reference to the probability of fluid system pipe rupture. For reactor coolant loop piping, a representative value which would qualify as "extremely low" would be of the order of lE-6 per reactor year when all rupture locations are considered in the fluid system piping or portions thereof. For other piping, representative values will be developed consistent with this definition as the need arises. Alternatively, a deterministic evaluation with verified design and fabrication, in addition to adequate inservice inspection, can meet the extremely low probability criterion. The deterministic evaluation is based on the requirement that structures and components are correctly engineered to meet the applicable regulations and NRC-endorsed industry codes. This rulemaking will introduce an inconsistency into the design basis by excluding the dynamic effects of postulated pipe ruptures while still retaining nonmechanistic pipe rupture for emergency core cooling systems, containments, and environmental qualification (see issue 4 below for additional information on potential relaxation with respect to environmental qualification). The Commission recognizes the need to address whether and to what extent leak-before-break analysis techniques may be used to modify present requirements relating to other features of facility design. However, this is a longer term evaluation. For the present, the rule allows the removal of plant hardware which it is believed negatively affects plant performance and safety, while not affecting emergency core cooling systems, containments, and environmental qualification. The Commission's primary justification for this rulemaking rests with its statutory responsibility to ensure an adequate level of protection of the public health and safety. This action also rests upon advances in the state of knowledge and technology that allow the Commission to better focus its regulatory requirements so as to improve safety of plant personnel. The Commission. decided to quantify the degree to which overall safety was improved by this action, and to set forth those economic impacts associated with these safety benefits. These are highlighted below. For existing PWRs, considering primary coolant loops only, cost savings of $186 million and reductions of 34,000 man-rem are estimated for a population of 85 PWRs. These figures do not include savings resulting from redesign of heavy component supports. One licensee taking advantage of the limited scope modification of GDC-4 has estimated a per plant cost savings of $20 million and reduced worker exposures of about 2,000 man-rem associated with a redesign of reactor coolant pump supports. PART 50

STATEMENTS OF CONSIDERATION The above-mentioned value-impacts were realized under the already published limited scope amendment to GDC-4. Additional benefits which can be achieved under this broader amendment are discussed below. For existing BWRs, considering only recirculation loop piping, cost savings of $30 million and reductions of 8,600 rem are estimated for a population of 38 plants. In existing PWRs and BWRs, offsite ris~ is estimated to be insignificantly impacted, or if credit is taken for improved inservice inspection and enhanced safety, to be reduced by an unquantified amount. The Commission has not quantified situations in existing plants other than those discussed above; however, it is believed that other high energy piping will also indicate favorable irnpacts.

Value-impacts resulting from this rule are greatest for future plants, where estimated costs can be reduced approximately $100 million per unit. Of this sum, about $30 million are direct *costs and the balance sterns from reduced financing costs and improved scheduling. Reduction in worker radiation exposures varies from plant to plant, but is in the range of 300 to 800 man-rem. Offsite risk is believed to decrease by an unquantified amount due to improved effectiveness of inservice inspection and enhanced safety. The above quoted figures are based primarily on the elirnina tion of pipe whip restraints and jet impingement barriers and do not treat other facility .changes that could result from this rule. IV. Acceptance Criteria The Commission developed a new Standard Review Plan Section 3.6.3 which gives more details on how applicant and licensee submittals will be evaluated. This document has been issued for public comment (52 FR 32626) prior to being adopted by the Commission. The Commission may also develop at some future time a Regulatory Guide after experience is gained with the use of SRP 3.6.3. V. Invitation to Comment Comment was invited on the following topics in the proposed broad scope amendment to GDC-4. 1. Value-impacts associated with this expanded modification to GDC-4, with particular reference to experience with the use of pipe whip restraints and jet impingement shields near nuclear reactor piping. (The value-impact analysis prepared by Lawrence Livermore National Laboratory is available for inspection and copying for a fee in the NRC Public Document Room, 1717 H Street NW., Washington, DC.) 2. The scope of piping which could or should be affected, supported by technical justification.

3. The decision to limit impacts of this modification of GDC-4 to only dynamic effects associated with pipe rupture. 4. The acceptance criteria which the Commission proposes to use to evaluate whether leak-before-break technology is applicable to specific situations.

5. Acceptable allowables for connected component supports which would provide adequate assurance that component support failure would not be a source of the pipe rupture loads being eliminated from the design basis. 6. The imposition of a temperature limitation as a way of avoiding concerns with creep damage. VI. Issues Analysis Issue 1: Margins for leak detection should not be rigidly fixed but should be based on uncertainties for each particular situation.

Commission Response: In the Commission's proposed acceptance criteria, the postulated through-wall crack used in the deterministic fracture mechanics evaluation is based on a detection margin of ten with respect to the leakage from the postulated crack. The Commission agrees that the selection of the margin should be derived from the uncertainties involved. As noted in issue 7 of the final limited scope-GDC-4 r-ule (51 FR iz°5i:iz, ApriT11.-- -1986), the Commission recognizes that the measurement or determination of leakage from a system under pressure involves uncertainties for which margins are needed. Cornrnenters suggesting relaxation in the detection margin cited only limited sources of uncertainty such as material properties and calculated flow rates through a crack. Other sources of uncertainty not mentioned include plugging of the crack with particulate material over time, stresses and number of cycles, and uncertainties associated with personnel and instruments used to detect leakage. For the present, the Commission will retain the leak detection margin of ten unless detailed evidence can be presented for other values. The Commission may require administrative controls to enforce adequate implementation of leakage detection and monitoring. Additionally, the Commission may undertake recurring inspections to verify that leakage detection and monitoring satisfy leak-before-break requirements. Issue 2: Margins on loads and leakage crack sizes used in the deterministic fracture mechanics evaluation should be relaxed. Commission Response: The Commission acknowledges that there are many situations where the margin is not required on loads resulting from the design basis piping analyses. However, there are situations where the 50-SC-18 uncertainty in the total procedure, including stress analyses and fracture mechanics evaluations, warrants some margin [see issue 7 below). Applicants or licensees must maintain the margin on loads at 1.4, except when the deadweight, thermal expansion, pressure, seismic inertial and seismic anchor motion loads are combined based on individual absolute values. In this case, the margin on loads may be reduced to 1.0. The evaluation of seismic anchor motion loads at SSE conditions may be omitted when these loads are shown to be small at DBE conditions. The Commission believes that, because of uncertainties associated with flaw geometry and the different analytical procedures, the margin between the leakage crack size and critical crack size stated in the proposed rule should not be reduced below the value of two. Issue 3: The acceptance criteria should refer to "crack detection" rather than "leakage detection". Commission Response: The fracture mechanics evaluation outlined in the Commission's acceptance criteria .examines a postulated throughwall flaw which may grow under service and .earthquake loads. The size of the postulated flaw for fracture mechanics evaluation purposes depends on the ability to detect the presence of the flaw during service with an adequate margin for detection. The standard methods to detect the throughwall flaw during service depend on the magnitude of flow or leakage through the flaw. Therefore, the methodology has to be based upon leakage detection rather than crack detection. Issue 4: Leak-before-break technology should be extended to relax pipe rupture requirements for containment design, emergency core cooling systems and environmental qualification of electrical and mechanical equipment. Commission Response: This was addressed as issue 3 in the final limited scope GDC-4 rule (51 FR 12502). The Commission plans to consider whether environmental qualifica lion requirements can be modified based upon leak-before-break technology. The Commission does not intend to consider near-term changes to emergency-core cooling system and containment design :bases as discussed in the Final Rule *section of this SUPPLEMENTARY INFORMATION. When leak-before-break technology is applied to dynamic effects design bases, these effects are reduced to zero; there are no replacement dynamic effects postulated. However, environmental qualification design bases cannot be reduced to zero when leak-before-break technology is applied to piping. The postulated pipe rupture has served as a convenient and conservative umbrella covering many sources of environmental PART 50

STATEMENTS OF CONSIDERATION qualification design bases, such as breaches in the fluid system pressure boundary from failed pump seals, leaking valve packings, flanged connections, bellow, manways, rupture disks and throughwall cracks. Thus, in applying leak-before-break technology to environmental qualification, the Commission faces the task of developing a replacement environmental qualification design basis. The Commission is not prepared at this time to propose new environmental design criteria for temperature, pressure, humidity and flooding.

If it can be shown that it is beneficial to apply before-break technology to environmental qualification, another modification to GDC-4 would be proposed. In the interim, the Commission recognizes that situations may arise where justification can be developed by the industry for alternative equipment qualification requirements. Such justifications, if accepted by the Commission pursuant to the existing exemption process, would aliow a limited number of case-by-case relaxations in environmental qualification requirements. The Commission encourages the development of generic alternative equipment qualification design bases by the industry. This could support future amendments to GDC-4 and other affected requirements addressing environmental qualification. Issue 5: Can minor modifications of piping systems not related to the exclusion of dynamic effects be made without examining impacts on the original leak-before-break evaluation? Commission Response: The original *leak-before-break evaluation must be applicable for the life of the plant. Changes in configuration or operating conditions must be examined to -determine impacts on the validity of the original leak-before-break evaluation, particularly as to how stresses are influenced. The Commission believes that many minor modifications, such as changing piping insulation, can be made without affecting the original analysis. However, modifications which are not minor in scope as for example when the number or type of pipe supports are changed extensively, require an evaluation of the applicability of the original leak-before-break analysis. Issue 6: Leak-before-break should be mandatory for plants which have not yet received their construction permit. Commission Response: The Commission believes that economic and operational considerations will motivate many utilities to apply leak-before-break technology. While it is estimated that an unquantified reduction in public risk results from this rule, the actual scope of piping contributing to the reduction will vary from plant to plant. The Commission encourages the use of high quality piping which does not require pipe whip restaints and jet impingement barriers. For any new application, the Commission would permit the applicant to decide whether or not to use before-break technology. Issue 7: Leak-before-break technology should be applicable to discrete locations. There should be no requirement that leak-before-break technology be applicable only to an entire piping system or analyzable portion thereof. Commission Response: Standard Review Plan [SRP) Section 3.6.2 of NUREG--0800 has been used for more than a decade to postulate the number and location of pipe ruptures in nuclear power plants. SRP 3.6.2 ignores or treats indirectly many factors, such as material properties, potential corrosion, and the potential for water hammer, which actually determine where and whether pipe rupture will occur: break procedures explicitly treat these factors. The Commission will not commingle SRP 3.6.2 with more advanced leak-before-break methodology. Leak-before-break is intended to be a substitute for SRP 3.6.2 only when all breaks in a fluid system piping are eliminated. This avoids consideration of synergistic effects, that is, the effects of a pipe break at one location on another potential break location. Additionally, the Commission, through long term and extensive piping research programs, has become aware that differences exist between analytically calculated stresses and actual stresses occurring at discrete locations in piping. The differences between calculated and

actual stresses usually stem from difficulties in modeling pipe supports under dynamic and static environments.

Leak-before-break will be applied only to an entire fluid system piping or analyzable portion thereof. Issue 8: Creep should not be an issue in applying leak-before-break technology. Commission Response: This rule gives guidance for reactors other than light water reactors. Creep can be an issue for gas and metal cooled reactors. Normally, creep damage is not an important concern in light water reactors. Issue 9: Extensive materials testing requirements should be relaxed. The use of generlc materials properties should be permitted. Commission Response: The ductile piping fracture mechanics analysis techniques that are applied in the before-break assessment are strongly dependent on the material tensile properties and material resistance to crack extension. The material testing requirements are necessary to provide reliable assessments of margins against unstable flaw extension when case-by-50-SC-19 case leak-before-break analyses are performed. However, if archival materials are not available or if actual plant material properties cannot be defined practically, generic plant specific or industry wide material data bases can be assembled and used to define the required material tensile and toughness properties. To provide an acceptable level of reliability, plant specific generic data bases must be reasonable lower bounds for sets of compatible material tensile and toughness properties associated with actual materials at the plant. Any industry generic data base must be a reasonable lower bound for the population of material tensile and toughness properties associated with anyTndividuai material specification (e.g., A106 Grade B), material type [e.g., austenitic steel), or welding procedure. Except as indicated in the Commission response to issue 13, industry generic data bases for the range of piping materials in light water reactors have not been assembled and proposed for leak-before-break analyses. Industry groups are encouraged to assemble and use reliable generic data bases so that analyses and evaluations can be performed efficiently and effectively. Issue 10: The temperature limitation of 750°F should not be adopted for evaluation of creep damage. Commission Response: The temperature limitation of 750°F is revised as follows: for ferritic steel piping, the temperature limitation will be 700°F; for austenitic steel piping the temperature limitation will be 800°F. These values more accurately reflect the creep performance of piping and are in accord with the ASME Code. Recent experience in fossil fuel plants operating at temperatures over 1000°F has indicated that creep-related ruptures in large diameter piping may not be low probability events, and suggests deficiencies in creep design standards after a service life of ten years or greater. Until creep behavior after long service intervals is better understood, the Commission will retain the temperature limitations cited above. Issue 11: Delete the words "reviewed and approved by the Commission" from the text of the rule. Commission Response: This comment is rejected. Leak-before-break technology is applicable only to high quality piping which is maintained in a high quality condition. Since much of the plant's piping is custom designed, the Commission would have to undertake detailed case-specific review to determine that acceptable standards of quality are achieved and maintained, and that the analyses meet the Commission's requirements. Detailed reviews are especially needed in piping other than PWR primary coolant loops PART 50

STATEMENTS OF CONSIDERATION to assure that failure mechanisms such as water hammer, corrosion, erosion, fatigue, and creep are not significant contributors to the potential for pipe rupture. Additionally, factors such as leakage detection, material properties and environmental conditions are more variable outside PWR primary coolant loops, and possible misuse of before-break technology can occur, unless careful review and evaluation of these aspects are performed by the Commission.

Consequently, the words "reviewed and approved by the Commission" were added specifically to ensure that a careful evaluation enforcing the Commission's rigorous acceptance criteria would be performed for each individual request from licensees and applicants. The adopted revision of GDC-4 requires NRC review and approval of the analyses on which the elimination of dynamic effects are based. As reflected in the limited scope rule (51 FR 12502), which is replaced by the adopted broad scope rule, the NRC has previously reviewed and approved the application of leak-before-break technology for eliminating design basis dynamic effects of postulated ruptures in PWR primary loop piping. No additional review and approval by the Commission in these cases is required under the adopted broad scope rule for elimination of design basis dynamic effects of postulated ruptures in PWR primary loop piping provided the conditions set forth in the Supplementary Information accompanying the rule (51 FR 12502) are satisfied. The proposed broad scope amendment (51 FR 26397, July 23, 1986) also stated that "Modifications of the licensed plant design of operating plants may involve an unreviewed safety question under 10 CFR 50.59 * * *. A simple removal of pipe whip restraints and jet impingement barriers would not involve an unreviewed safety question." The meaning of this last sentence is that after analysis reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low, then, without prior approval, pipe whip restraints and jet impingement barriers may be removed. Pipe whip restraints and jet impingement barriers cannot be removed, however, without conducting an appropriate leak-before-break evaluation, submitting the evaluation for Commission review and obtaining Commission approval. Moreover, removal of a pipe whip restraint which also serves as a seismic restraint would not be a "simple" removal of a pipe whip restraint and, therefore, would involve an unreviewed safety question. Issue 12. How is the demonstration of extremely low probability made for indirect sources of pipe rupture? Commission Response: Indirect sources of pipe rupture, as discussed in the plant FSAR, are investigated by applicants and utilities. These include seismic events and system overpressurizations due to accidents resulting from human error, fires or flooding which cause electrical and mechanical control,.5ystems to malfunction. The analysis of indirect sources should also confirm that snubber failure rates are maintained at *a low rate.-corniilfance with-the snubber surveillance requirements of the technical specifications can be used to demonstrate that snubber failure rates are low. Missiles from equipment, damage from moving equipment and failures of systems or components in close proximity to the piping are investigated as well. The results of prior analyses conducted to show compliance with Commission regulations can be applicable to potential sources of indirect pipe rupture. Issue 13. It is recommended that adequate material toughness be demonstrated when limit load analysis is applied and that the margin of three be on the applied force and moment combined rather than just on moment. The limit load analysis procedures in ASME Code, Section XI, Appendix C, Winter 1986 Addenda, should be allowed. Commission Response: The Commission is revising its requirements on limit load analysis procedures as stated in the proposed amendment to GDC-4 (51 FR 26393). The new requirements do not contain the arbitrary margin of three on applied moment, but instead are based on an experimentally verified ASME approved procedure. During preparation of NUREG-1061, Volume 3, there was significant uncertainty associated with reliable application of limit load analysis for austenitic steel, especially in the case of submerged arc welds (SAW) and shielded metal arc welds (SMAW). This uncertainty led to restrictions on the use of limit load analysis and application of methods originally used as the basis for IWB-3640 in Section XI of the ASME Code. Recently, the ASME Code approved revised evaluation procedures for austenitic steel piping (see Appendix C of Section XI); these procedures incorporate methods to account for reduced toughness associated with SAW and SMAW. The Commission has concluded that the evaluation method in Appendix C of Section XI (including the tensile and toughness properties defined for base metal and welds) is acceptable when performing leak-before-break analyses for austenitic steel piping, provided the margins presented in the Commission response to issues 1 and 2 are met. The value of flow stress used with this method will be evaluated by the Commission. Because generic 50-SC-20 evaluation procedures and materials properties have not yet been approved for ferritic piping by the ASME Code or the NRC, leak-before-break analysis for ferritic piping will continue case-by-case until approved Code procedures are available. Issue 14: Leak-before-break should not

be limited to high energy piping. Commission Response:

The Commission's rules require postulated pipe ruptures only in high energy piping. There is no reason to apply break technology in_moderate energy piping because there are no postulated pipe ruptures in such fluid system piping. Issue 15: Strict adherence to Regulatory Guide 1.45 should not be required outside the containment. Commission Response: The Commission does not require and did not intend to suggest the need for strict adherence to Regulatory Guide 1.45 outside the containment. The proposed rule stated only that"* *

leakage detection requirements equivalent to Regulatory Guide 1.45 must be satisfied for all piping within the scope of this rule." Scheduled operator walkdowns can be used as a means of leak detection outside the containment.

Issue 16: Older operating plants should not be held to the requirement that heavy component supports should meet ASME Code allowables as a condition for applying break. Commission Response: The use of ASME Code allowables for heavy component supports of older operating plants as a condition for applying before-break technology is not required. However, when heavy component supports are redesigned excluding the dynamic effects of pipe rupture, current *industry codes (such as the ASME or AISC code) may be required. Additionally, current NRC criteria for calculating seismic loads (coupled with the already existing SSE) may also be required. For example, a simple replacement of high strength fasteners with more ductile fasteners of lower yield strength would not require changes in the industry codes or seismic criteria from that used in the original design. On the other hand, modification of the heavy component supports that involves redesign and removal of snubbers in early vintage plants would require use of current industry codes and NRC seismic criteria. Dynamic effects from pipe ruptures in branch connections must be considered if the branch connections do not qualify for before-break. In heavy component support redesign, improved functional reliability must be demonstrated for any changes made. Structural capacity associated with the original steel and concrete, including struts, columns, pedestals, hangers, trusses and skirts PART 50

STATEMENTS OF CONSIDERATION cannot be diminished in the support system of operating plants or plants under construction.

Redesigns will be limited to replacing high strength fastener material and reducing the number and capacity of snubbers. Applicants and licensees undertaking heavy component support redesign, with dynamic effects of pipe rupture eliminated, should use independent design and fabrication verification -procedures to minimize design and construction errors. Displacements and rotations resulting from potential failure of redesigned lateral (horizontal] supports should not lead to the rupture of piping connected to the reactor coolant loop heavy components. Issue 17: Additional guidance is needed on the acceptability of remedial stress enhancement programs such as induction heating as it pertains to stress, . corrosion, cracking, residual stress states and sensitization. Commission Response: The rule precludes leak-before-break evaluations for systems that have materials that are susceptible to intergranular stress corrosion cracking (IGSCC]. The Commission recognizes that remedial residual stress improvement treatments are effective in reducing susceptibility to IGSCC. However, remedial stress improvement treatments of nonconforming materials alone do not provide a sufficient basis to support leak-before-break evaluations in the context of this rule. The Commission would, however, review such evaluations case-by-case if hydrogen water chemistry were used as an adjunctive measure with the remedial stress improvement treatments. Practices with regard to facility water chemistry would be an additional factor considered in the review. Nonconforming piping with any planar flaws in excess of the standards in IWB 3514.3 of Section XI of the ASME Code would not be permitted to use before-break analyses. However, nonconforming piping that has been treated by two mitigating methods may qualify for leak-before-break if the piping contains no flaws larger than those permitted by IWB 3514.3 of Section XI of the ASME Code. If piping has been repaired by weld overlays, leak-before-break technology cannot be applied. Issue 18: The fracture mechanics approach should not require that the location of highest stress utilize the poorest material properties. Commission Response: The proposed rule stated that, in conducting the deterministic fracture mechanics evaluation, investigators would "identify the location(s) at which the highest stresses coincident with poorest materials properties occur * * *". This sentence should have read "identify the location(s] which have the least favorable combinations of high stress and poor material properties

* *". The Commission did not intend to combine the highest stresses at one location with the poorest material properties of another location.

The critical rupture locations depend on stress and material properties, among other things, and investigators may need to examine several locations to decide which is the controlling case. Issue 19: The decision that before-break technology is not applicable to materials subject to cleavage type fracture should be reconsidered. Commission Response: This comment is rejected. The Commission will allow leak-before-break technology only to materials which are ductile under the full range of system operating temperatures in order to avoid sudden brittle piping failures. Issue 20: The clause in the rule requiring Commission review and approval of analyses demonstrating piping integrity will preclude litigation over the scope of piping affected and the adequacy of the analyses. This would amount to a de facto illegal removal of a material issue from an operating license amendment proceeding. Commission Response: The commenter cites in support of this proposition Cleveland Electric Illuminating Company (Perry Nuclear Power Plants, Units 1 and 2), ALAB-841, (July 25, 1986), reconsideration denied ALAB-844, (August 18, 1986). The Licensing Board case affirmed in ALAB-841 is Cleveland Electric Illuminating Company (Perry Nuclear Power Plant, Units 1 and 2), LBP-85-35, 22 NRC 514 (1985). A careful reading of these cases shows that they do not stand for the commenter's proposition. At issue was the scope and adequacy of the applicant's preliminary hydrogen control analysis required by 10 CFR 50.44(c](3). One of the criteria for this analysis is that it "use accident scenarios that are accepted by the NRC staff." (50.46(c)(vi)(B)(3)) The Licensing Board did not hold that the staffs approval of the applicant's analysis was binding and thus precluded a challenge to the scope and adequacy of the analysis. Rather, the Board permitted such a challenge on a number of issues. 22 NRC at 533-548. However, the Board did not allow the intervenor to raise other issues under this contention which went beyond the scope of the hydrogen control rule itself. SO-SC-21 22 NRC at 548-549. The Board's views on these matters were upheld on appeal. A direct application of this case to the GDC-4 context shows that the commenter's conclusion is incorrect. The staffs acceptance of a GDC-4 analysis will not preclude litigation of either the scope of piping included or the adequacy of the analysis itself. However, a challenge on either basis must be confined to the overall scope of GDC-4 and could not be used as a collateral challenge to other parts of the regulations or to argue that the rule itself is inadequate. Challenges of this type must be brought pursuant to 10 CFR 2.758. The staffs review and approval of the piping integrity analyses is an indispensable part of the implementation of the leak-before-break concept. Without such review (for piping other than the PWR primary loop]. the staff has no means to assure itself that the acceptance criteria have been properly applied. The comment is therefore rejected. Issue 21: The reallocation of resources within the NRC to review piping integrity analyses submitted under the amendment is barred by the Atomic Energy Act, which requires that public safety take precedence over cost savings to licensees. Commission Response: The Regulatory Analysis performed to support this rulemaking shows that there is a net safety benefit to be realized from proper application of before-break technology. The Commission has undertaken the rulemaking for that purpose. The positive results in terms of simplicity of the plant, ease of inspection, avoidance of improper removal and reinstallation of unneeded supports and restraints, and the reduction of personnel exposures have been shown to vastly outweigh any additional risk associated with removing supports and restraints. Therefore, reallocation of NRC resources to ensure that NRC acceptance criteria are rigorously adhered to is fully justified in terms of public safety. In addition to these issues, the Commission deleted the fatigue crack growth analysis specified in the proposed rule. This requirement was found to be unnecessary because it was bounded by the crack stability analysis. Having considered all of the above, the Commission has determined that a final rule be promulgated VII. Availability of Documents

1. Copies of NUREG-1061, Volume 3, may be purchased from the PART 50

STATEMENTS OF CONSIDERATION
Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC, 20013-7082.

Copies are also available from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. A copy is also available for public inspection and/or copying at the NRC Public Document Room, 1717 H Street NW., Washington, DC. 2. Copies of the ASME Boiler and Pressure Vessel Code may be obtained from the American Society of Mechanical Engineers, 345 East 47th Street, New York, NY, 10017. 3. Copies of Regulatory Guide 1.45 . entitled "Reactor Coolant Pressure Boundary Leakage Detection Systems" may be obtained by writing to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC, 20555. VIII. Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule if adopted, would not be a major Federal action significantly affecting the quality of the human environment and, therefore, an environmental impact statement is not required. Although certain plant hardware might be removed from the plant, consistent with this rule, the removal would not alter the environmental impact of the licensed activities as set out in the Final Environmental Impact Statement for each facility. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 1717 H Street, NW., Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from John A. O'Brien, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 443-7854. IX. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 350i et seq.). Existing information collection requirements under 10 CFR Part 50 were approved by the Office of Management and Budget approval number 3150-0011. . X. Regulatory and Backfit Analyses The regulatory analysis is available for inspection in the NRC Public Document Room, 1717 H Street NW, Washington, DC. Single copies of the analysis may be obtained from John A. O'Brien, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washlngton, DC 20555, telephone (301) 443-7854. A backfit analysis under 10 CFR 50.109 for the purpose of completeness was published in the proposed broad scope GDC-4 modification (51 FR 26393), although it was not required because the rule will not require licensees or applicants to make any changes. The Commission's primary justification for this rulemaking rests on its statutory responsibility to ensure an adequate level of protection of the public health and safety. Economic advantages or disadvantages resulting from this action did not affect such responsibilities. The Commission remains mindful of its statutory responsibilities pursuant to Union of Concerned Scientists et al. v. NRC, DDC No. 85-1757, August 4, 1987. The Commission has prepared, however, a regulatory analysis to set forth clearly the costs and benefits of the impacts of this rule and the examined alternatives. XI. Regulatory Flexibility Act Certification As required by the Regulatory Flexibility Act of 1980, (5 U.S.C. 605(b]J, the Commission certifies that this rule will not have a significant economic impact on a substantial number of small entities. This rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definitions of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. Final Rule For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974; as amended, and 5 U.S.C. 553, the NRC is adopting the following amendments to 10 CFR Part 50. so-sc-22 52 FR 42078 Published 11/3/87 Effective 12/3/87 10 CFR Part SO Evaluation of the Adequacy of Off-Site Emergency Planning for Nuclear Power Plants at the Operating Ucense Review Stage Where State and/or Local Governments Decline To Participate In Off-Site Emergency Plannlng AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Com.nission is amending its rules to provide criteria for the evaluation at the operating license review stage of prepared emergency plans in situations in which state and/or local governments decline to participa le furthe: in emergency planning.

The rule is consistent with the approach adopted by Cor,gress in sr.ction 109 of the NRC Authorization Act of 1980, Pub. L. 96-295, described in the Conference Report on that statute (H.96-1070, J:me 4, 1980), twice re-enacted by the Congress (in Pub. L. 97-415, Jan. 4. 1983. and Pub. L. 98-553, Oct. 30, 1984), and followed in a prior adjudicatory decision of the Commission, Long Island Lighting Co., (Shoreham Nuclear Power Station, Unit 1 ). CLl-86--13. 24 NRC 22 (1986). The rule recognizes that though state and local P?rticipati?n in emergency planning is highly desirable, and indeed is essential for maximum effectiveness of emergency planning and preparedness, Congress did not intend that the absence of such participation should preclude licensing of substantially completed nuclear power plants where

there is a u !iii ty-prepar~d emergency plan that provides reasonable assurnnce of adequate protection to the public. EFFECTIVE DATE: December 3, 1987. FOR FURTHER INFORl'I.ATION CONTACT: Peter G. Crane, Office of the General Counsel, USNRC, Washington, DC 20555,202-634-14~5 Michael T. Jamgochian, Office of Nuclear Regulatory Research, USNRC, Washington, DC 20555, 301-443-7657.

David B. Matthews, Office flf Nuclear Reactor Regulation, USNRC, Washinr"on, DC 20555, 301-492-9647. SUPPLEMENTARY INFOl'IMATION: Discussion . On ~arch 6, 1987, the NRC published its nohce of rr0posed rulemaking in *the Federal Register, at 52 FR 6980. The period for public comment (60 days, PART 50

STATEMENTS OF CONSIDERATION subgequently extended for an additional 30 days) expired on June 4, 1987. The proposed rule drew an unprecedentedly large number of comments.

Some 11,500, individual letters were sent to NRC, as well as 27,000 individually signed form letters sent to Congress or the White House and forwarded to NRC. Approximately 16,300 persons signed petitions to the NRC, Every comment was read, including form letters, which were examined one by one so that any indivdual messages added by the signatories could be taken into account. NRC attempted to send cards of acknowledgment to each commenter. The sheer volume of the comments received makes it clearly impracticable to discuss them individually. As a result. the following discussion will focus on the principal issues raised in the comments. Issue #1. ls the proposed rule legal? Specifically, is it in accord with the language and legislative history of the emergency planning provisions enacted by the Congress in 1980? Answer: Yes. The intent of the proposed rule, as clarified in Commission testimony and in other responses to the Congress, is to give effect to the Congress's 1980 compromise approach to emergency phmning, not go beyond it. To explain this requires a somewhat detailed discussion of the background of the actions taken in 1980 by Congress and by the Commi~sion with "egard to emergency planning. The backdrop for the actions taken by the Congress and the Commission in 1980 was, of course, the 1979 accident at Three Mile Island. The accideni changed the NRC's regulatory approach to radiological emergency planning. Before the accident, rimergency planning received relatively little attention from nuclear rlgulators. The prevailing assumption was that engineered safety features in nuclear power plants, coupled with sound operation and management, made ii unlikely that emergency planning would ever be needed. At that time, only a limited evaluation of offsite emergency planning issues took place in the pre-construction review of applications to build nuclear power plantu. The Three Mile Island accident led to the widespread recognition that, while there is no substitute for a wr.11 built, well run, and well regulated nuclear power plant, a substantial upgrading of the role of emergency planning was necessary if the public health and safety were to be adequately protr.cted. The Commission issued an advance notice of proposed rulemaki!!g in July 1979, 1md in September and December of ti, 0 sa.ne year it issued proposed emergency planning rules. 44 FR 54308 (Septembe, *g, 1979); 44 FR 75167 (December 19, 1979). Before the Commission took final action on the rnles, however, the Congress took action, writing emergency planning provisions into the NRC Authorization Act for fiscal year 1980, Pub. L. No. 96-295. It is extremely important to focus on what the Congress did in that Act, because CongreGs' actions were the starting point for all the NRC did subsequently in the emergency planning area, as the written record makes clear. Section 109 of the NRC Authorization Act directed the Commission to establish regulations making the existence of an adequate emergency plan a prerequisite for issuance of an operating license to a nuclear facility. The NRC was further directed to promulgate standards for state radiological response plans. In the same section of the 1980 Act, Congress specified the conditions under which the Commission could issue operating licenses, and in doing so, it made clear its preferences with regard to state and local participation. Its first preference, reflected in section 109(b)(l)(B)(i)(IJ, is for a "State or local radiological emergency response plan wliich provides for responding to any radiological emergency at the facility concerned and wr.,ch complies with the Commission's standards for such plans." In section 109(b)(l)(B)(i)(Il), however, the Congress set out a second option: "In the absence of a plan which satisfies the requirements of subclause (I), there exists a State, local, or utility plan which provides reasonable assurance that public health and safety is not endangered by operation of the facility concerned." [Emphasis added.) In addition, section 109 providecl that the Commission's determination under the first but not the second r.:* the two options could be made "only in consultation with the Director of the Federal Emergency Management Agency end other appropriate agencies." Section 109[b)(1)(B)(ii). The statute further directed the Commission to "establish by rule * *

a mechanism to encourage and assist States to comply as expeditiously as practicable" with the NRC's standards for Slate radiological emergency response plans. Section 109(b)(1 )(CI. The Conference Report on the legislation, H. 96-1070 (June 4, 1980) explained in clear terms,. at p. 27, the rationale for the two-tiered approach: "The conferees sought to avoid penalizing an applicant for an operating license if a State or locality does not submit an emergency response plan to the NRC for review or if the suLmitted plan does not satisfy all the guidelines or rules. In the absence of a State or local plan that complies with the guidelines or rules, the compromise pem1its NRC to issue an operating license if it determines that a State, local 50-SC-23 or utility plan, such as the emergency preparedness plan submitted by the applicant, provides reasonable assurance that the public health and safety is not endangered by operation of the facility." [Emphasis added.) The statute. which was enacted on June 30, 1980, and the Conference Report make abundantly clear that in Congress' view, the ideal situation was one in which there is a state or local plan that meets all NRC standards.

It is generally clear that in Congress* view, there could be emergency planning under a utility plan that to some degree fell short of the ideal but was nevertheless adequate to protect the health and safety of the public. That Congressional judgment was before the Commission when it considered final emergency planning rules only a few weeks later, and the Commission took pains to make clear on the ree,ord that it was following the Congress' approach. As the Commission stated in its notice of final rulemaking. published on August 19, 1980, at 45 FR 55402: Finally, on Juiy z3:-1*eiio:*iit1iie final Commission consideration of these rules, the Commission was briefed by the General Counsel on the substance of conversations with Congressional staff membr.rs who were involved with the passage of the NRC Authorization Act for fiscal year 1980, Pub. L. No. 96-295. The General Counsel advised the Commission that the NRC final rules were consistent with that Act. The Commission hes relied on ell of the above information in its consideration of these fine! rules. In addition, the Commission directs that the transcripts of these meetings shell be pert or the administrative record in this rulemeking. In addition, in a key portion of the rule, dealing with the question of whether NRC should automatically shut down nuclear plants in the absence of an NRG-approved slate or local emergency plan, or should instead evaluate ell the relevant circumstances before deciding on remedial action, the NRC again explicitly followed the Congress' lead. In determining what action to take, the Commission said, it would look at the significance of deficiencies in emergency planning, the availability of compensating measures, and any compelling reasons arguing in favor of continued operation. 10 CFR 50.47(c). The Commission explained: "This interpretation is consistent with the provisions of the NRC Authorization Act for fiscal year 1980, Pub. L 96-295." 45 FR 55403. Thus in deciding that the la!:k of an approved state or local plan should not be grounds for automatic shutdown of a nuclear power plant, the Commission expressly declared itself to be following the statutory approach. This background sheds considerable light on a passage from the Federal Register notice which some commenters saw as indication that the Commission consciously decided in 1980_ that states

PART 50
STATEMENTS OF CONSIDERATION end localities should have the power to exercise a veto over nuclear power plant operation.

The Commission said: The Commission recognizes that there is a possibility that the operation 0£ some reactors may be affected by this rule through inaction 0£ State and local governments or an inability to comply with these rules. The Commission believes that the potential restriction 0£ plant operation by State and local officials is not significantly di££erent in kind and effect from the means already available to prohibit reactor operation .... Relative to applying this rule in actual practice. however, the CommiFsion need not shut down a facility until all factors have been thoroughly examined. 45 FR 55404. (Emphasis added.) It has been argued that the language just quoted indicates that t~e .. Commission made a conscious dec1s1on in 1980 to allow states and localities lo exercise a veto power over completed nuclear power planl_s. Seen in context. however, It is apparent that the Commission did no such thing. Rather,

the Commission was acknowledging the fact that under the approach It was taking, the action (or inaction) of a state or locality had the potential to affect the operation of nuclear power plants, since state and local non-participation would clearly make It more difficult for an applicant to demonstrate the adequacy of emergency planning.

It is worth emphasizing the word "potential" in the quoted passage. It indicates that the Commission believed that In some cases, slate and local action or Inaction might have the effect of restricting plant operation, while in other cases ii would not. In other words, l~e Commission foresaw a case-by-case evaluation, with the result not foreordained either in the direction of plant operation or of shutdown. Clearly, neither the Cr,mmission nor the Congress envisioned that state or local participation should automatically bar plant operation without further inquiry. The mechanism adopted by the Commission for implementing the tiered approach was set forth in 10 CFR 50.47 0£ the Commission's regulations. For the first tier, sixteen planning standards for a state or local emergency plan were spelled out in 10 CFR 50.47(b)(1-16) of the Commission's regulations. The second tier, by contrast. was dealt with in a brief and unspecific provision, 10 CFR 50.47(c)(1): Failure to meet the 1161 applicable standards set forth in paragraph (b) of this

section may result in the Commission declining to issue an operating license: however, the applicant will have an opportunity to demonstrate lo the satisfaction of the Commission that deficiencies in the plans are not significant

£or the plant in question. that adequate interim con:pensating actions have been or will be taken promptly, or that there are other compelling reasons lo permit plant operation. In a 1986 decision. the Commission declared that in a situatic,1 in which state end local authorities decline to participate in emergency planning, the NRC has the authority and the legal obligation to consider a utility plan and render a judgment on the adequacy of emergency planning and preparedness. Long Island Lighting Co. (Shoreham Nuclear Power Station, Unit 1), CLl-86-13, 24 NRC 22. The Commission observed in LILCO that the emergency planning standards of 10 CFR the regulation which establishes the 16 planning standards by which a state and local plan is to be measured-"are premised on a high level of coordination between the utility and State and local governments," so that "(i]t should come es no surprise that without governmental cooperation [the utility) has encountered great difficulty complying with all of these detailed planning standards." 22 NRC 22. 29. The Commission noted, however, that Its emergency planning rules were intended to be "flexible," and that a utility plan will pass muster under 10 CFR 50.47(c) "notwithstanding noncompliance with the NRC's detailed planning standards

* * (1) if the defects are 'not significant':

(2) if there ere 'adequate interim compensating actions': or (3) if there are 'other compelling reasons.' " The Commission added: "The decisions below focus on (1) end (2) end we do likewise. The Commission then explained that the "measure of significance under (1) and adequacy under (2) is the fundamental emergency planning standard of§ 50.47(a) that 'no operating license * *

will be issued unless a finding is made by NRC that there is reasonable assurance that adequate protective measures can end will be taken in the event 0£ a radiological emer!jency.' "The "root question," the Commission said. was whether a utility plan "can provide for 'ad,!quate protective measures * *
in the event 0£ a radiological emergency.'" To answer that question, the Commission continued, requires recognition of the fact that emergency planning requirements do not have fixed criteria.

such as prescribed evacuation times or radiation dose savings. but rather aim at "reasonable and feasible dose reduction under the circumstances. 24 NRC 22, 30. Thus the Commission is already on record es believing itself legally obligated to consider the adequacy of a utility plan in a situation of slate and/or local non-participation in emergency planning. Likewise. it is on record as believing that the evaluation of a utility plan takes place in the context of the overriding obligation that no license can be issued unless the emergency plan is found to provide reasonable assurance of adequate protective measures in an emergency. The Commission believes that the planning standards of 10 CFR 50.47(b), which are used lo evaluate a slate or local plan, also provide an . 50-SC-24 appropriate framework to evaluate a utility plan. Therefore, the new rule provides for the first time that where a utility plan is submitted, in a situation of stale and/or local non-participation in emergency planning. it will be eve'lueted for adequacy against the same standards used to evaluate a slate or local plan. However, due allowance will be made both for the non-participation of the state and/or local governmental authorities and for the compensatory measures proposed by the utility in reaching a determination whether there is "reasonable assurance the: adequate protective measures can end will be taken. To sum up, therefore, the rule is in accord with legal requirements for emergency planning RI nuclear power plants because: -The rule is consistent with section 109 of the NRC Authorization Act of 1980, a measure which has twice reenacted by the Congress, though it has since expired. In addition, the House of Representatives recently rejected an amendment designed to bar implementation of the rule for two specific plants. -The rule is consistent with existing NRC regulations, end is well within NP.C's rulemaking authority. -Since the rule provides for no diminution of public protection from what was provided under existing regulations, it cannot be in contravention of any statutory requirements governing the level of NRC safely st,mdards. Issue #2: ls this a generic rule, or is this proposal really aimed et the Shoreham and Seabrook plants? The rule is generic in the sense that ii is of general applicability and future effect, covering future plants as well as existing plants. At present, however, there are on'.y two plants with pending operating license applications for which state and/or local non-participation is an issue. Those plants are Shoreham and Seabrook. The NRC's 1980 rules, perhaps because of optimism that states and localities would always choose lo be partners in emergency planning, included only a general provision, 10 CFR 50.47(c), dealing with cases in which utilities are unable lo satisfy the standards for stale end local emergency plans, and had no specific discussion of the evaluation of a utility plan in cases of state or local non-partic:ipalion. This does not mean that the NRC was compelled to adopt new regulations in order to act on the Shoreham and Seabrook license applications. On the contrary, the NRC hes always had the option of proceeding by case-by-case adjudication under its 1980 regulations. Issue #3: Will this rule assure licenses to the Shoreham and Seabrook plants? It will not assure a licer:se lo any particular plant or plants. It will establish a framework in which a utility PART 50

STATEMENTS OF CONSIDERATION seeking an operating license can, in a case of stale and/or local parlicipation, attempt lo demonstrate lo !he NRC Iha! emergency planning is adequate.

Whether a utility could

succeed in making !hat showing would depend on the record developed in a specific adjudication, the results of which would be subject to multiple levels of review within the Commission as well as lo review in the courts. Issue #4: Is state or local participation essential for the NRC to determine that there will be adequate protection of the public health and safety? We do not have a basis at this time for determining generically whether state and local participation In emergency planning la e11sential for NRC lo determine that there will be adequate protection of the public health and safety. There has yet to be a final adjudicatory determination in any proceeding on the adequacy of a utility plan where state and local governmental authorities decline to participate in emergency planning.

Clearly, it will be more difficult for a utility to satisfy the NRC of the adequacy of its plan in the absence of slate and local participation, but whether it would be impossible remains lo be seen. The fact that Congress provided for evaluation of a utility plan in section 109 of the NRC Authorizat'ion Act of 1980 (and in two subsequent Authorization.Acts) indicates that Congress believed that it was at least possible in some cases fore utility plan to be found to provide "reasonable assurance that public health end safety is not endangered by operation of the facility concerned, in the words of the "second tier" provided in section 109. Issue #5: Is emergency planning as impt1rtant to safety es proper plant design 11ncl operation? First of all, this issue does not have to be addressed in the ::ontext of the final rule announced in this notice, since the present rule involves no redrawing by NRC of the balance between emergency planning and other provisions for the protection of health end safety. Having said that, we tum to the question of the piece of emergency planning in the overall regulatory scheme for the protection of public health and safety. Though the Commission in its 1980 rulemaking explicitly described emergency planning as "essential," it is less clear what importance the Commission assigned to emergency planning, as compared to the importance accorded to other means of pro lee.ting public health end safety, notably sound siting, design, and operation. In the Supplementary Information explaining the 1980 rulemaking, the Commission stated that "adequate emergency preparedness is an essential aspect in the protection of the public health and safety," 55 FR 55404, and commented that "onsile and offsite emergency preparedness as well as proper siting and engineered design fe11tures ere needed to protect the health and safety of the public." (Emphasis added.) 45 FR 55403. The Commission also explained the: in light of the Three Mile Island accident it had become "clear that thP. protection provided, by siting and engineered design fee lures must be bolstered by the ability to lake protective measures during the course of an accicjent." Id. Though the word* "bolstered" suggests that the Commission of 1980 viewed emergency planning as a backstop for other means of public protection rather than as of equal importance to them, the issue cannot be resolved definitively by microscopic analysis of the particular words chosen in 1980. More relevant lo the task of ascertaining the intent of the 1980 rulemaking is the regulatory structure established under the 1980 rules. In 10 CFR 50.54(sl(2l(ii), the Commission provided that if it "finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can end will be taken in the event of e radiological emergency ' ' ' end if the deficiencies ' ' ' ere not corrected within four months of that finding. the Commission will determine whether the reactor shell be shut down until such deficiencies are remedied or whether other enforcement action is appropriate." In other words, e plant ordinarily may operate for at least four months with deficiencies in emergency planning before the NRC is required even lo decide whether remedial action should be taken. This approach. the Commission said in the Supplementary Information to the 1980 rule, was consistent with section 109 of the :-JRC Authorization Act of 1980. 45 FR 55407. At the time that the Commission creeled the so-celled "120-day clock" for deficiencies in emergency p_lenning, it was settled Commission Jew (end remains so today) that the NRC must issue en order direr;ing e licensee to show cause why its license should not be modified, revoked or suspended whenever it concludes that "substantial health or safety issues he[ve) been raised" about the activities authorized by the license. Consolidated Edison Company of New York (Indian Point. Units No. 1, 2 end 3), CLl-75-8, 2 NRC 173, 176. That standard was endorsed by the Court of Appeals for the District of Columbia Circuit in Porter County Chapter of the Izaak Walton League v. NRC, 606 F.2d 1363 (1978). In the context of that standard, the 120-dey clock provision for emergency planning deficiencies amounts to e Commission finding that, et least for the first 120 days, even a major deficiency in emergency planning does not . automatically raise e "substantial health or safety issur." with regard to plant operation. By contrast, e major safety deficiency relating to emergency 50-SC-25 conditions-for example. the availability of the emergency core cooling system-would warrant immediate shutdown. In sum, despite language indicating that emergency planning was "essential," the Commission in 1980 created a regulatory structure in which emergency planning was treated somewhat differently, in terms of the corrective actions to be taken when deficiencies are identified, from thu engineered safety features ("hardware") that would be relied on in an emergency. Issue :::ti: Assuming that NRC should consider e utility plan, what criteria should apply? In part.icular: (a) Should the utility plan provide just as much protection es a state or local plan, or may less protection be adequate? (b) If less protection may be adequate, must NRC still find reasonable assurance that under the utility plan, adequate protective measures can and will be taken? Or is it sufficient for NRC to find that the totality of the risk, including all relevant factors, including the likelihood of an accident, assures that there is adequate protection of public health end safety? Under the rule adopted in this notice, e utility plan, to pass muster, is required to provide reasonable assurance that adequate protective measures can and will be taken in an emergency. The rule recognizes-as did Congress when it enacted and re-enec.ted the provisions of Section 109 of the NRC Authorization Act of 1980-that no utility plan is likely to be able to provide the same degree of public protection that would obtain under ideal conditions, i.e. a state or local plan with full state and local participation, but that it may nevertheless be adequate. The rule starts from the premise that accidents can happen, and that et every plant, adequate emergency planning measures are needed to protect the public in the event an accident occurs. Whether in fact a particular utility plan will be found adequate would be a melter for adjudication in individual licensing proceedings. Issue #7: May NRC assume that a state or local government which refuses to cooperate in emergency planning will still respond to the best of its ability in an actual emergency? If so: . (a) May NRC assume that the state or local response will be In accord with the utility plan? (b) May NRC assume that the state or local response will be adequate? (c) If the NRC rule calls for reliance on FEMA, and FEMA says that it can't. judge emergency planning except when there is state and local participation in an exercise, how can the NRC ever make a judgment on emergency planning In a situation in which state and local authorities do not participate? In this rule, the Commission adheres to the "realism doctrine, enunciated in PART 50

STATEMENTS OF CONSIDERATION its 1986 decision in Long island Lighting Co. (Shoreham Nuclear Power Station. Unit 1), CLl-86-13, 24 NRC 22, which holds that in an actual emergency, stale end local governmental authorities will act to protect their citizenry, and that it is appropriate for the NRC to take account of that self-evident fact in evaluating the adequacy of a utility's emergency plan. The NRC's realism doctrine is grounded squarely in common sense. As the Commission statecl in L/LCO, even where state and local officials "deny they ever would or could cooperate with [a utility) either before or even during an accident," the NRC "simply cannot accept these statements at face value." 24 NRC 22, 29 fn. 9. It would be irrational for anyone to suppose that in a real radiological emergency, state and local public officials would refuse to do what they have always done in the event of r.mergencies of all kinds: do their best to help protect the affected public. The long Island lighting Co. decision included the observation that in an accident, the "best effort" of state and county officials would include utilizing the utility's plan as "the best source for emergency plr.nning information and options." 24 NRC 22. 31. This rule leaves it to the Licensing Board to judge whet form the "best efforts" of state and local officials would take. However, the rulemaking record strongly supports the proposition that state and loco! governments believe that a planned 1esponse is preferable to an ad hoc one. Therefore it is only reasonable to suppose that in the event of a radiological emergency, state and local officials.

in the absence of a state or local radiological emergency plan approved by state and local governments, will either look to the utility and its plan for guidance or will follow some other plan that exists. Thus the presiding Licensing Board may presume that state end local governmental authoritie I will look to the utility for guidance and i!enerally follow its plan in an actual emergency; however, this presumption may be rebutted by, for example, a good faith and a timely proffer of an adequate and feasible state or local radiological response plan which would in fact be relied upon in en emergency. The presiding Licensing Board should not hesitate to reject any claim that state and local officials will refuse to act to safeguard the health and safety of the public in the event of an actual emergency. In actual emergencies, state, local, and federal officials have invariably done their utmost to protect the citizenry, as two hundred years of American history amply demonstrateo. At the present time, the Commission does not have a basis in its adjudicatory experience to judge either that a utility plan would be adequate in every case or that ii would be inadequate in every case. Implementation of this rule may ultimately provide that informational basis.* The problem of how the NRC can decide the adequacy of emergency planning in the face of FEMA's declared reluctance to make judgments on emergency planning in cases of state and local non-participation does not appear insoluble. Though FEMA has expressed its reluctance to make judgments in such circumstances, because of the degree of conjecture that would in FEMA's view be called for. we do not interpret its position as one of refusal to apply its expertise to the evaluation of a utility plan. For FEMA to engage in the evaluation of a utility plan would necessitate no retreat from its stated view that ii is highly desirable to have. for each nuclear power plant, a state or local plan with full state and local participation in emergency pla~ning, including emergency exercises. (The Commission shares that view.) FEMA's advice would undoubtedly include identification of areas in which judgments are necessarily conjectural. and NRC's overall judgment on whether a utility's plan is adequate would in tum have to take account of the uncertainties included in FEMA's judgment. Beyond a certain point, uncertainty as to underlying facts would plainly make a positive finding on "reesonabli, e_ssurance" increasingly difficult. These are issues, however. which can bl' addressed in the case-by-case adjudications on individual fact-specific situations. It should be noted that while the rule makes clear*thet ultimate decisional authority resides with NRC, it does envision a role for FEMA in the evaluation of utility plans, although section 109 of the NRC Authorization Act of 1980 did not specify any role for FEMA in the eveluatio'J) of utility plans (as opposed to state and local plans). Issue #8: If this is a national policy question, why doesn't the Commission leave the issue to the Congress to resolve? Congress did address, in 1980, the issue of what should be done in the event there is no acceptable state or local emergency plan: it directed the NRC to evaluate a state, local, or utility plan to determine whether it provided "reasonable assurance that public health and safety is not endangered by operation of the facility concerned." Perhaps because it was overly optimistic that there would be an acceptable state or local plan in every case, the Commission did not, except in general terms (at 10 CFR 50.47(c)J, provide in its regulations for the evaluation of a utility plan. The present rule is an effort to make up for that omission by 50-SC-26 incorporaiing provisions implementing the Congress's 1980 policy decision into the NRC's rules. As noted elsewhere, the 1980 statute, twice re-enacted, has expired, but the NRC does not need the specific authority of that statute to adopt this rule, which is promulgated pursuant to the NRC's general authority, under section 161(bJ and other provisions of the Atomic Energy Act, to regulate the use of nuclear energy. The House of Representatives, as has been described above, voted 261-160 on August 5, 1987 to reject an amendment which would have barred the -application of this rule to two specific plants. The Congress is thus well aware of the Commission's emergency planning rulemaking. For the Commission to terminate its rulemaking and ask the Congress to address the policy issues involved thus seems unwarranted at this lime.The Commission is still well within the framework of the guidiance which the Congress gave it in 1980 (and in the two re-renactments of the statute) and also well within its rulemaking authority. It has yet to carry through that guidance to the point of making an adjudicatory decision on the adequacy of a utility plan. If and when the Commission determines, through adjudications in individual cases, that there is a continuing problem which only Congressional action can solve, it can so notify the Congress, but that point has not yet been reached. Issue #9: Doesn't the proposed rule still leave open the possibility that state or local action or inaction can have the effect of blocking operation of a plant? If so, how can the proposed rule be said to effectuate the Congressional intent that licensees not be penalized for the inaction or inadequate action of stale and local authorities? Yes, the proposed rule does leave open the possibility that state or local non-participation cari indirectly block the operation of a nuclear plant. This is so because under the particular facts of an individual case It may*be impossible for the NRC to conclude that a utility plan is adequate, as defined in this rule. That does not mean, however, that the Congress's intent, es expressed in the 1980 statute and its re-enactments, is thereby frustrated. The Congress was concerned that utilities not be "penalized," but not to the extent that it was willing to countenance operation of a nuclear power plant in a situation where the public was not adequately protected. Congress intended to give a utility the opportunity to demonstrate that its plan provided "reasonable assura1 *ce," but it also provided that the NRC cou;-1 not permit a plant to operate unless it fouud that the utility had met that burden. Issue #10: Will the proposed rule discourage cooperation between PART 50

STATEMENTS OF CONSIDERATION licensees end slate end iocel governments in emergency planning?

There is no reason to believe that the rule would discourage coopere tion between licensees end slate end local governments in P.mergency planning. Realistically, the only way in which the rule could discourage such cooperation would be if utilities were lo decide that because of the new rule, they had less of en incentive lo be accommodating to the needs end desires of slate end local euthorititls. That might be a possible result if it appeared that the new rule make it easy end fest for a utility to obtain approval for its plan in cases of slate end local non-participation. In reality, it is likely to be much more difficult and lime-consuming for a utility lo obtain approval of its plan in the face of slate end local opposition. The problems highlighted by this rulemeking ere likely, if anything, to impress utilities anew with the desirability of doing everything necessary to obtain end retain full stale and local participation in emergency planning. Issue #11: Is the proposed rule based on en NRC consideration of economic costs? The NRC rule is en effort to bring the NRC's regulations more clearly into line with a policy decision made by the Congress in 1980. *,*he NRC's rule is thus based on econonic considerations only lo the extent the1 the Congress's policy decision of 1980 was based on economic considerations. In the Conference Report on the NRC .l\uthorizetion Act of 1980 (H.96-1070, June 4, 1980?, che conferees slated that they did not wish utilities to be "penalized" in situations in which there was no acceptable elate or local plan. That could be taken es a reference to economic costs or simply lo coneiderati,me offairneee, in that the issue wee whether a utility wee to be barred from operating a plant by the action& of third parties over which it had no control. The NRC'e motivation in promulgating this rule is not economics. Ile motivation is to assure that the NRC is in a position to make the decisions that Congress intended that it make, end that the , .Commission has declared that it would m&ke. Issue #12: Is the;1>ropoeed rule intended lo reed elates and localities out of the emergency planning process? Emphatically not. The rule leaves the existing regulatory structure unchanged for cases in which elate end local authorities elect to participate in emergency planning. The NRC, in _ common with the Congress end FEMA. .regards full state end local participation in emergency planning to be necessary for optimal emergency planning. The rule change is directed lo the question of whet the NRC's regulatory approach should be in which elates end localities decide to take themselves out of the emergency planning process. Ideally, in the NRC's view. the new rule would never have lo be used, because states and localities would never refuse lo participate in emergency planning. Issue #13: Does the proposed rule 11lter the place of e111ergency planning in the overall safety finding that the Commission must make? It does not. As described above, the Commission must make both a finding of "adequate protective measures * *

in an emergency" end an overall safety fi~ding of "reasonable assurance that the health and safety of the public will not be endangered" (10 CFR 50.35(c), implementing section 182 of the Atomic Energy Act, 42 U.S.C. 2232). The rule does nothing to alter either the requirement that emergency planning must be found adequate or the place of emergency planning in the overall safely finding. Issue #14: What effect if any does the proposed rule have on nuclear plants that are already in operation?

The rule does not specifically apply to plants that already have operating licenses. As described above, 10 CFR 50.54(sl(2)(ii) of the Commission's regulations already provides a mechanism (the "120-day clock.) for addressing situations in which deficiencies are identified in emergency planning et operating plants. To the exten) that this rule provides criteria by which a utillly plan would be judged by state end local withdrawal from participation In emergency planning, those criteria would presumably be of assistance to decisionmekere In determining, under 10 CFR 50.54(e)(2J(ii), whether remedial action should be taken, end if so, what kind, where deficiencies In emergency planning remain uncorrected after 120 days. Issue #1!;: Does the Commiesion;e rule mean that the NRC does not have to find that a utility pl,.m would offer protection equivalent to whet a plan with full elate end local participation would provide? As elated previously, under the rule adopted in this notice, a utility plan, lo pees muster, is required to provide reasonable assurance that adequate protective measures can end will be taken In emergency. The rule recognizes-es did Congress when it enacted and re-enacted the provisions of Section 109 of the NRC Authorization Act of 1980--that no utility plan is likely to be able lo provide the same degree of public protection that would obtain under ideal conditions, i.e. a slate or* local plan with full stale and local participation, but' that it may nevertheless be adequate. The Commission's rule, as modified end clarified, would establish a process by which a utility plan can be evaluated against the same standards that ere used to evaluate a s!ale or local plan 50-SC-27 (with i:illowencee made both for those areas in which compliance is infeasible because of governmental participation end for the compensatorv measures proposed by the utility). It must be recognized that emergency planning rules ere necessarily flexible. Other then "adequacy," there is no uniform "passing g::'.!r.le" for emergency plans, whether they ere prepared by a state, a locality, or a utility. Rather, there is a case-by-case evaluation of whether the plan meets the standard of "adequate protective measures ... in the event of en emergency." Likewise, the acceptability of a plan for one plant is not measured against plans for other nuclear plants. The Commission, in its 1986 LILCO decision, stressed the need for flexibilty in the evaluation of emergency plans. In that decision, the Commission observed that it "might look favorably" on a utility plan "if there was reasonable assurance that it was capable of achieving dose reductions in the event of an accident that are generally comparable to what might be accomplished with government cooperation." 24 NRC 22, 30~ We do not read that decision es requiring a finding of the precise dose reductions that would be accomplished either by the utility's plan or by a hypothetical plan that had full state and local participation: such findings are never a requirement In the evaluation of emergency plane. The final rule makes clear that every emergency plan le to be evaluated for adequacy on Its own merits, without reference to the specific dose reductions which might be accomplished under the plan or to the capabilities of any other plan. It further makes clear that a finding of adequacy for any plan Is to be considered generally comparable to a finding of adequacy for any other plan. The rule change is designed to establish procedures end criteria governing the case-by-case adjudicatory evaluation, at the operating license review stage, of the adequacy of emergency planning in situations in which state and/or local authorities decline to participate further in emergency planning. It is not intended to assure the licensing of any particular plant or plants. The rule is intended lo remedy the omission of specific procedures for the evaluation of a utility plan from the NRC's existing rules. adopted In 1980. In providing for the evaluation of a utility plan, however. the rule represents no departure from the approach envisioned in 1980 by the Congress and by the Commission. In 1980, the supplementary information to NRC's final rule stated that the rule was consistent with the approach taken by Congress in Section 109 of the NRC Authorization Act of 1980 (which, in a compromise between House and Senate PART 50

STATEMENTS OF CONSIDERATION versions, provided for the NRC lo evaluate a utility's emergency plan in situations where a state or local plan was either nonexistent or inadequate), though the rule itself included no explicit provisions governing the NRC's evaluation of a utility plan in such circumstances.

It should be emphasized that the rule is not intended to diminish public protection from the levels previously established by the Congress or the Commission's rules, since the Commission's rules and the Congress have since 1980 provided for a two-tier approach to emergency planning. The rule takes as its starting point the Congressional policy decision reflected in section 109 of the NRC Authorization Act of 1980. That statute adopted a two-. tier approach to emergency planning. The preferred approach was for operating licensee to be issued upon a finding that there is a "State or local radiological emergency response plan * *

which complies with the Commission's standards for such plans," but failing that, it also permitted licensing on a showing that there is a "State, local, or utility plan which provides reasonable assurance that the public health and safety Is not endangered by operation of the facility concerned." Under the Commission's 1980 rules, tie regulatory provision that implemented the second of the two tiers of Section 109 was general and unspecific.

The relevant regulation, 10 CFR 50.47(c), allowed a nuclear power plant to be licensed to operate, *notwithstanding its failure to comply 'with the planning standard of 10 CFR

50.47(b), on a showing that "deficiencies in the plane are not significant for the plant in question, that adequate interim compensating measures have been or will be taken promptly, or that there are other compelling reasons to permit plant operation," without defining those terms further. The Commieeion currently believes that the planning standards of rn CFR 50.47(h), which are used to evaluate a state or local plan, also provide an appropriate framework to evaluate a utility plan. Therefore, the new rule provides for the first time that where a utility plan is submitted, in a situation of state and/or local participation in emergency planning, it will be evaluated for adequacy against the same standards used to evaluate a state or local plan. However. due allowance will be made both for the non-participation of the state and/or local governmental authorities and for the compensatory measures proposed by the utility in reaching a determination whether there is "reasonable assurance that adequate protective measures" can and will be taken. *
The approach reflected in this rule ampli~ies and clarifies the guidance provided in the Commission's decision in long ls/ond lighting Co .. (Shoreham Nuclear Power Station, Unit 1). CLl-86-13, 24 NRC 22 (1986). The rule incorporates the "realism doctrine,"'

set forth in that decision, which holds that in an actual emergeocy. state and local governmental authorities will act to protect the public, and that it is appropriate therefore for the NRC, in evaluating the adequacy of a utility's emergency plan, to take into account the probable response of state and local authorities, to be determined on a by-case basis. That decision also included language which could be interpreted as envisioning that the NRC must estimate the radiological dose reductions which a utility plan would achieve, compare them with the radiological dose reductions which would be achieved if there were a state or local plan with full state and local participation in emergency planning, and permit licensing only If the dose reductions are "generally comparable." Such an interpretation would be contrary to NRC practice, under which emergency plans are evaluated for adequacy without reference to numerical dose reductions .which might he accomplished, and without comparing them to other emergency plane, real or hypothetical. The final rule makes clear that every emergency plan is to be evaluated for adequacy on lie own merits, without reference to the specific dose reductions which might be accomplished under the plan or to the capabilities of any other plan. It further makes clear Iha t a finding of adequacy for any plan is to be considered generally comparable to a finding of adequacy for any other plan. The long Island lighting Co. decision included the observation that in an accident, the "beet effort" of state and county officials would include utilizing the utility's plan as "the best source for emergency plannin~ information and options." 24 NRC 22, 31. Thie rule leaves it to the Licensing Board to judge what form the "best efforts" of state and local officials would take, but that judgment would be made in accordance with certain guidelines set forth in the rule and explained further below. The rulemaking record strongly supports the proposition that state and local governments believe that a planned response is preferable lo an ad hoc one. Therefore it is only reasonable to suppose that in the event of a radiological emergency, elate and local officials, in the absence of a state or local radiological emergency plan approved by state and local governments, will either look to the utility and its plan for guidance or will follow some other plan that exists. Thus, the presiding Licensing Board may presume that state and local govermental 11uthorities will look to the 50-SC-28 utility for guidance and generally follow its plan in an actual emergency; however, this presumption may be rebutted by, for example, a good faith and timely proffer or an adequate and feasible state or local radiological response plan which would in fact be relied upon in an emergency. The presiding Licensing Board should not hesitate to reject any claim that state and local officials will refuse to act to safeguard the health and safety of the public in the event of an actual emergency. ln actual emergencies, state, local, and federal officials have invariably done their utmost to protect the citizenry, as two hundred years of American history amply demonstrates. The rule thus establishes the framework by which the adequacy of emergency planning, In cases of state and/or local non-participation, can be evaluated on a case-by-case basis In operating license proceedings. The rule does not presuppose, nor does It dictate, what the outcome or that case-by-case evaluation wlll be. As with other issues adjudicated In NRC proceedings, the outcome of case-by-case evaluations of the adequacy or emergency planning using a utility's plan will be subject to multiple layers of administrative review within the Commission and to judicial review in the courts. Backnt Analysis Thie amendment does not Impose any new requirements on production or utilization facilities; It only provides an alternative method to meet the Commission's emergency planning regulations. The amendment therefore is not a backfit under 10 CFR 50.109 and a hackfit analysis le not required. Regulatory Flexibility Certification ln accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(h), the Commission certifies that this rule will not have a significant economic impact upon a substantial number of small entities. The proposed rule applies only to nuclear power plant licensees which are electric utility companies dominant in their service areas. These licensees are not "small entities" as set forth in the Regulatory Flexibility Act and do not meet the small business size standards set forth in Small Business Administration regulations in 13 CFR Part 121. Paperwork Reduction Act This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were appro*,ed by the Office of Management and Budget, approval No. 3150-0011. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire protection, Incorporation by reference, PART 50

STATEMENTS OF CONSIDERATION Intergovernmental relations, Nuclear power plants end reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting end Recordkeeping requirements.

Environmental Assessment end Finding or No Significant Environmental Impact The Commission hes determined under the National Environmental Policy Act or 1969, es emended, end the Commission's regulations-in Subpart A of 10 CFR Pert 51, that this rule is not a major Federal action significantly effecting the quality of the human environment end therefore an environmental impact statement is not required. The Commission has prepared, in support of this finding, an environmental assessment which is available for inspection and copying, for a fee, at the NRC Public Document Room, 1717 H Street NW .. Washignton, DC. Regulatory Analysis The Commission has prepared a regulatory analysis for this regulation. This analysis further examines the costs end benefits of the proposed action and the alternatives considered by the Commission. The analysis is available for inspection end copying, fore fee, at the NRC Public Document Room, 1717 H Street, NW., Washington, DC. For the reasons set out in the preamble, and under the authority of the Atomic Energy Act or 1954, as amended, the Energy Reorganization Act of 1974, as emended, and 5 U.S.C. 553, the Commission is adopting the following amendments lo 10 CFR Part SO: (Editorial riote: Tbe.foDowlng regulatory." . enal)*ai, and en~ironmental assessment' w,11 not appear in the Code of Federal Regulation11]

Regulatory Anelysis-Eveluetion of the Adequacy of Off site Emergency Planning for Nuclear Power Plants al Iha Operating Ucense Review Siege Where Stele end/or Local Governments Decline lo Participate in Offsite Emergency Planning Statement of the Problem In 1980, Congress enacted provisions dealing with emergency planning for. nuclear power plants in the NRC Authorization Act for fiscal year 1980. Section 109 of that Act provided for the NRC to review a utility's emergency plan in situations in which estate or local emergency plan either did not exist or was inadequate.

The NRC published regulations later then year that were designed to be consistent with the Congressionally mandated approach, but they did not include specific mention of utility plans. The absence of such a provision hes led lo uncertainty about the NRC's authority to consider a utility plan end the , criteria by which such a plan would be judged. The present rulemeking is designed to clarify both the NRC's obligation to conside_r a utility plan et the operating license stage in cases of stale end/or local non-participation in emergency planning end the standards against which such a plan would be evaluated. Objective The objective of the proposed amendments ere to implement the policy underlying the 1980 Authorizeiton Act end to resolvr., for future licensing, whet off~ite emergency planning criteria should apply where state or local governments deolde not to participate in off site emergency planning or preperedneu. Alternatives Five alternalivea were considered, including leaving the existing rules unchanged. The pros and cons of these eltemellves ere discussed in the rule preamble published in the Federal Register. Consequences NRC The amendments will probably not impact on NRC resources currently being used in licensing cases because current NRC policy. developed in the adjudicatory case law, is to evaluate utility plans es possible interim compenseling actions under 10 CFR 50.47(c)(1). Thus, while there could be extensive litigation end review regarding whether the rule's criteria ere met. !hie would likely be similar lo the review end litigation under current preciice. Other Government Agencies No Impact on other agency resources should result with the possible exception that FEMA will need to devote resources to develop criteria for reviP.w of utility plans endjor to review the plans on a case-by-case uesis. lndusty Impacts on the industry ere speculative because there is no way to predict. in advance of their actual application, whether any pertculer utility plan will satisfy the rule. However, industry should generally benefit from knowing that rules ere in piece so that plans for compliance can be formulated. Public Under the rule being adopted e utility plan, to pass muster, is required to provide reasonable assurance that adequate protective measures con and will be. token in au emr.rgency. The rule recognizes-es did Congress when it enacted end re-enacted the provisions of Section 109 of the NRC Authorization Act of1980-thel while no utility plan Is likely to be able lo provide precisely the same degree of public protection that would obtain under ideal conditions. i.e. a stale or local plan with full slate end local participation. such a pion may nevertheless be edquete. The rule alerts from lhe premise that accidents can happen. end that el every plant, adequate emergency planning measures ere needed to protect the public in the evenl en eccidenl occurs. Whether in feet e particular utility plan will be found adequate would be a melter for adjudication in individual licensing proceedings. Impact on Other Requirements The proposed amendments would not affect other NRC requirements. Constraints No constraints have been identified that effect implementation of the proposed amendments. 50-SC-29 Decision Rationale The decision rationale is set forth In detail In the preamble lo the rule change published in the Federal Register. Implementation The rule should become effective 30 days after publication in the Federal Raglater. Implementation will involve coopere!ion with FEMA end the development of FEMA/NRC criteria for review of utility plans may be required before the rule Is applied lo specific cases. Environmental AaseHmant for Amendments lo Emergency Planning Regulations Dealing With Evaluation of Offslt11 Emergency Planning for Nuclear Power Pienta et the Operating license Review Siege Where Slate end/or Local Governments Decline lo Pertlclpela In Ofrslte Emergency Planning Identification of the Action The Commission is emending Its regule lions lo provide criteria for the evaluation et the operating license siege of off site emergency planning where, because of the non-participation of state end/or local governmental authorities, a utility hes proposed Its own emergency plan. The Need for the Action As described in the Federal Register notice accompanying the final rule, the Commission's emergency planning regulations, promulgated in 1980. did nol explicitly discuss the evaluation of a utility emergency plan. although Congress expressly pr1,vided that in the absence of a state or local emergency plan, or in cases where a state or local plan was inadequate, the NRC should*consider a utility plan. That omission

hes led to u::icerteinty es lo whether the NRC is empowered to consider a utility pion in cases of state end/or local non-participation, es well es about whet the standards for the evaluation of such a plan would be. Alternatives Considered The Commission published a proposed rule change on March 6, 1987, el 52 FR 8980. In deciding on a final rule. the Commission considered four options in addition to the one reflected in the final rule. These were: issuance of the rule es originally proposed end described:

issuance of a rule making clear that in cases of stale end/or local perticipetion, licenses could be issued on the basis of the utility's best efforts: issuance of a rule barring the issuance of licenses in cases of elate end/or local non-participation: end termination of the rulemaking without the issuance of any rule change. Environmental Impacts of the Action The rule does not alter in any way the requirement that for en operating license to be issued, emergency planning for the plant in question must be adequate. The rule is designed lo effectuate the second track of the two-track approach adopted by the Congress 15 the NRC Authorization Act of 1980 end two successive authorization eels, es described in detail in the Federal Register notice. The rule does not effect the piece of emergency planning in the overall safety finding which the Commission must make prlo~ lo the licensing of any plant. Accordingly, the rule change does not diminish public protection and hes no environmental Impact. PART 50

STATEMENTS OF CONSIDERATION Agencies and Persons Consulted A summary of the very numerous comments appears as pert or the Federal Register notice. Shortly before presenting en options paper to the Commission.

NRC representallves briefed representatives of the Federal Emergency Management Agency on the contents of the options paper. Finding of No Significant Impact Based on the above, the Commission hes decided not to prepare an envlronme"i1tel impact statement for the rule changes. 52 FR 49362 Published 12/31/87 Effective 2/1/88 Completeness and Accuracy of Information See Part 2 Statements of Consideration 53 FR 6137 Pubti*hed 3/1/88 Effective 3/1 /88 Relocation of Office of Nuclear Reactor Regulation See Part 19 Statements of Consideration 53 FR 8845 Published 3/17/88 10 CFR Part 50 Evaluation of the Adequacy of Off.Site Emergency Planning for Nuclear Power Plants at the Operating License Review Stage Whe.re,State and/or

Local Governments Decline To Participate in Off-Site Emergency Planning Correction In. rule document 87-25439 beginning
on p*age 42078 in the issue of Tuesday,
November 3, 1987, make the following .correction:

On :page 42079, in the third column. in the second complete paragraph, in the sixth line, "generally" should read "equally".

53 FR 16051. Published 5/5/88
Effective 5/5/88 10 CFR Part 50 Codes and Standards for Nuclear Power Plants AGENCY: Nuclear Regulatory Commission.

ACTION: final rule.

SUMMARY

The Commission is amending its regulations to incorporate by reference the Winter 1984 Addenda, Summer 1985 Addenda, Winter 1985 Addenda, and 1986 Bdition of Section Ill, Division 1, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), and the Winter 1983 Addenda, Summer 1984 Addenda, Winter 1984 Addenda, Summer 1985 Addenda, Winter 1985 50-SC-30 Addenda, and 1986 Edition of Section XI, Division 1, of the ASME Code. A limitation is placed on the use of paragraph IWIJ:-3640 as contained in the -_.Winter 1983 Addenda and Winter 1984 Addenda of Section XI, Division 1. This limitation requires that for certain types of welds,.IWB-3640 when implemented shall be used as modified by the Winter 1985 Addenda. In addition, the existing modification pertaining to the inservice .inspection of pressure retaining welds in ASME'Code Class 2 piping has been revised to limit its applicability up to the: 1963 Edition with addenda up through the Summer 1983 Addenda. The sections of the ASME Code being incorporated by reference provide rul~s for-the construction ofHght-water-cooled . nuclear power plant components and specify. requirements for inservice inspection of those components.

Adoption of these amendments would permit the use of improved methods* for construction and inservice inspection of nuclear power plants. . EFFECTIVE DATE: Ma:v 5. 1988. The incorporation by ref~rence of certain publications list~d in the regulations is approved by the Director of the Office of th~ Federal Register as of May 5, 1988. FOR FURTHER INFORMATION CONTACT: Mr. G.C. Millman. Division of .Engineering. Office of Nuclear Regulatory Research, U.S. Nuclear *Regulatory Commission, Washington, DC 20555, Telephone: (301) 492-3872. SUPPLEMENTARY INFORMATION: On June 26, 1987, the Nuclear Regulatory Commission published in the Federal Register (52 FR 24o15) a proposed amendment to its regulation, 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to update the reference to editions and addenda of the American Society of Mechanic;al Engineers Boiler and Pressure Vessel Code (ASME Code). This amendment revises § 50.55a to incorporate by reference all editions through the 1986 Edition and all addenda through the Winter 1985 Addenda that modify Division 1 rules of Section III, "Rules for the Construction of Nuclear Plant Components," and, subje.ct to certain.limitations and modifications, addenda through the Winter 1985 Addenda that mcidify Division 1 rules of Section XI, "Rules for the Inservice Inspection of Nuclear Power Plant Components," of the ASME Code. Specifically, this amendment to § 50.55a incorporates by r1;iference the Winter 1984 Addenda, Summer 1985 Addenda, Winter 1985 Addenda.and 1986 Edition for Division 1 rules of Section III, and the .Winter 1983 Addenda, Summer 1984 Addenda, Winter 1984 Addenda, Summer 1985 Addenda, Winter 1985 Addenda, and 1986 Edition for Division 1 rules of PART 50

STATEMENTS OF CONSIDERATION Section XI of the ASME Code. The 1986 Edition is equivalent to the 1983 Edition, as modified by the.Summer 1983
Addenda, Winter 1983 Addenda, Summer*1984 Addenda, Winter 1984 Addenda, Summer 1985 Addenda, and .. . Winter 1985 Addenda...The*Slim.riier 1984 . Addend1tand Summer, 1985 Addenda for ..Section XI do not include technical requirements, but are included-in the reference to prevent the confusion that _might occur with a lack of continuity in the addenda references.

Interested persons were invited to submit written comments for consideration in connection with the proposed amendmei;it by August 25, 1987. Comments were received frorr. three individuals in response to the notice oi proposed rulemaking. Two of the commenters were in favor of the proposed rule, and submitted suggestions for editorial clarifications. One of these commenters wa$ concerned that the manner proposed ior specifying the.endorsed editions and addenda in§§ 50.55a [b)(l)and [b)(2) was potentially confusing for thir. specific amendment because the latest addenda that is specified does not modifv the latest edition that is specified (i.e., the Winter 1985 Addenda modifie~ the 1983 Edition). The staff agrees with the commenter and has modified paragraphs (b](l) and (b](Z) to make it clear that the Winter 1985 Addenda applies to the 1983 Edition, and that the 1986 Edition is the latest ASME Code update being incorporated by reference into the regulation. The other commenter in favor of the proposed r_ule believes that the proposed additional sentence in § 50.55a(b)(Z)(i) which specifies a limitation on the use of IWB-3640 for certain addenda should be provided for clarity in a separate paragraph. The staff has considered and adopted this suggestion. In this final rule, the specified limitation is contained in a new paragraph (b)(ZJ(v). This commenter also recommended a revision to Footnote 6 to clarify details regarding implementation of the code cases specified.in the identified regulatory guides.'It is the opinion of the staff that the rule should not be cluttered with such information.. . Therefore, that proposed revision has

not been incorporated into the final rule. However, the staff is considering incorporating additional information directly into the regulatory guides to clarify their use. . Additionally, this commenter noted that the Winter 1983 Addenda to.Section XI included significant' imprcivements . .to .
the inservice insjlection*of Class 2 . . piping:-Tliis comment-is-correct:

In-. . : particular; the.rti.lesspecified in that addenda satisfy NRC staff-concerns associated with the rules specified in earlier addenda.for the examination or pressure retaining welds in ASME Code Class 2 piping, including residual heat removal systems, emergency core . . cooling systems, and containment heat removal systems. Thit staf~ previously addressed these concerns by specifying a modification in § 50.55a(b)(2](iv), whi::h reauireci that foi; exHml o: examimitions for pressure retainio~ weids in ASME Code Class 2 pipino be ~*.::iermined based upon specific* rni~s in tne 1974 Edition and Addenda \nrough foe Summer 1975 Addenda. Aifr,ou::'h the commenier did not make tr1e poin;

specifically.

the proposed ruie shouid have recognized the improvements in the Winter 1983 Addenda by incorporating a revision to limit the applicabilitv of the reaui.-red existino modificatio;l soeciii.ea" in para!ITaohc (b)(2)(iv) to ASW.i.E Corie editi;ns 0 and addenda up to the 1S63 Edition with addenda up through the Summer 1983 Addenda. This final rule incorporates this limitation to the *use of the modification specified in paragraph (b](Z){iv). The third commenter opposed the proposed amendment. That commenter believes that the NRC should not rely on industry standards, but rather should develop its own standards based upon NRC experience and data. NRC practice is to utilize national standards, such as the ASME Code, whenever possible to define acceptable ways of implementing the NRC's basic safety regulations. This is consistent with OJ\.m Circular No. A-119 (Revised), 1 which provides. policy and administrative.guidance to federal agencies regarding participation in the development and use of voluntary standards. Consistent with this policy, the NRC staff participates actively in the develop~ent of many national standards, including the ASME Code, to ensure that NRG experience and data is part of the information base used to support development of the standard. Although the NRC staff is heavily involved in the development of the ASME Code, endorsement of the ASME Code by the NRC without exception is not an automatic action BS evidenced by the existing linii.tations and modifications* specified in § 50.55a(b )(2) and the new limitation specified in paragraph-{b)(2)(v) by this final rule.

Paragraph IWB-3640 was
incorporated into the Winter 1983 Addenda of Section XI, Division 1, to provide procedures and acceptance criteria for determining the acceptability for continued service of austenitic
stainless*

steel piping with flaws in excess of the allowable indications 1 Single copies of 0MB Circular No. A-119 may be obtained from the 0MB Publications Office, 726 Jackson Place NW~ Washington, DC 20503, Telephone [2112) 39!>-7332, 50-SC-31 specified in IWB-3514.3. Concern was expressed by the NRC staff and others that IWB-3640, as presented initially in the Winter 1983 Addenda, did not provide an acceptable level of margin as:;ams! faiiure ior materiais with 10w t~ughness, such as might occur in fluxed welds [i.e., submerged arc welds (SAW) or shielded metal arc welds (SMA WJ). One concern with low toughness materials was that such materials might fail at load levels Lernw limit load. Additionaliv. there was c'oncern that secondary

tresses, which were not included in the stress analysis procedures required by JWB--3640, might contribute to the failure of low toughness materjals. .The ASME e5~;c;bli,;r;;;d .a special task group to address foe concerns associated wHb-paragrnph IWB-3640 as contained in the Vvinter 1933 Addenda. In the interim. the NRC staff required that licensees utilizing the procedures and acceptance criteria of IWB-3640.

as contained in the \,Vint.er 1983 Addenda, apply additional safety factors in their anatyses to be submitted to the staff to account for the above concerns. NRC staff acceptance criteria were provided in Generic Letterfl4-1.1. "Inspections of BWR Stainless Steel Piping." 2 In the opinion of the NRC staff, the concerns associated with material toughness have been adequately addressed by the ASME Code with the modification to paragraph IWB-3640 in the Winter 1985 Addenda. This addenda provides specific acceptance criteria for SAW amfSMAWtype welds, and these criteria address the concerns associated with fimit load and the need to

incorporate secondary stresses in the -evaluation.

This amendment to § 50.55a incorporates a limitation -in paragraph {b){2){v) that allows for the use of paragraph IWB-3640, as contained in the Winter 1983 Addenda and Winter

1984 Addenda, for all applications permitted in that paragraph except those . associated with SAW*and SMAW type welds. For these welds, this.amendment specifies that paragraph JWB...3640, as modified by 1he Winter 1985 Addenda, must be used. Footnote 8 of§ 50.55a provides reference to the NRC Regulatory Guides that denote which ASME Code* Cases *: have been determined to be acceptable to the NRC staff for implementation.

Previously, this footnote provided

reference to only Regulatory Guides 1.84 and 1:115, which denote acceptability of Section m, -Division 1, Code Cases on design and fabrication.

and on materials, respectively. This amendment 'A copy.of Generic Letter 64-11 is available for inspection or copying for a fee al the NRC Public Document Room, 1717 H Street NW., Washington. DC: PART 50

STATEMENTS OF CONSIDERATION revises Footnote 6 to incorporate a reference to Regulatory Guide 1.147, "lnservice ir1suecuon Code Ccsf Acceptability.:__i'.SME Section Xl Division l," which identifies the Code Cases acceptable to the NRC staff for impierpentation in the inservice inspection

[ISI} program of cooled nuclear power plants. At presen~. the hnplementation section of Regulatory Guide 1.147 specifies that applicants should make a specific request to the J\."RC to use Code Cases endm:sed in the regulatory guide. The next revision of Regulatory Guide 1.14i (i.e., Revision 6j wilt reflect thE proposed change in Footnote 6 of the regulation. It will permit the use of Code Cases endorsed in the regulatory guide without a specific request to the NRC for approval. In the interim, it is the intent of'the NRC that Code Cases listed in Regulatory Guide 1.147 be used without specific application to the NRC. This amendment further revises Footnote 6 to correct the referenced titles for Regulatory Guides 1.84 arid 1.85. Section 50.55a(g) provides . requirements for .selecting the edition and addenda of Section XI to be complied with during the preservice inspection (§ 50.55a(g)(3), for plants . whose construction pennit was issued on or after July 1, 1974); the initial 10-year inspection interval (§ 50.55a(g}{4)(i));*and successive 10-year inspection intervals (§ 50.55a{9)(4J(ii)). Paragraph IWA-2400 of Section XI. as revised by the Winter 1983 Addenda, incorporates rules for selecting the applicable edition and addenda of Section XI during :the preservice inspection (IWA-2411); the initial 10-year inspection interval (1W A-* 2412); and successive 10-year inspection intervals {1W A--i413). . The criteria provided in the regulations and Section XI are effectively 1he same for the preservice inspectioni-und the successive 10-year inspection intervals, but-differ for the initial 10-year inspection intervat For 'the !nitial 1.0-year inspection interval, the regulations specify that inservice examinations

of components and imiervice tests shall comply with the requirements in-the latest edition* and addenda of the Code incorporated by reference on the date 12 months prior to the date ofissuance
of the operating license while Section XI provides that the inspection-plan shall comply with the Edition and Addenda of Section XI that has been adopted by the regulatory authority 36 months after the date of issuance of the construction pennit. or subsequent Editions and Addenda that have been adopted by. the regulatory .authority.

In general..use of the Commission requirements will result in foe seiection ci a more receni eJnwr. and addenda than will use of L'-:-e Section Xl rules. Satisfying foe requirements of§ 50:5.5c\6)i4 J[i) for !:nf initial 10-vear insoec!ion tnfo:::,::il wi'.'.. ir: general. als0 5ati~fy foe ruies of Seciior, XI.

It is the Commission's intent that in all cases the existing requirements in § 50.55a(g) be the basis fcir selecting the edition and addenda of Section XI to be complied with during the preservice inspection.

the 10-vear inspection interval. and succ~ssive 10-veer inspection intervals.

Subsection IWE, Requirements for Clas*s MC Components of Light-Water Cooled Power Plants," was added to Section XI, Division 1, in the Winter 1981 Addenda. Howevei-, 10 CFR 50.55a presently incorporates only those portions of Section X1 that address the ISi requirements for Ciass 1, 2, and 3 components and t.lieir snpports.

The .regulation does not currently address the ISi of containments. Since this amendment is only intended to update current regulatory requirements to include the latest ASME Code edition and addenda,' the requirements ,of Subsection IWE would not be imposed upon Commission licensees by this amendment. The applicability of Subsection JWE is being considered separately. Environmental Impact: Ca.regork:al Exclusion The NRG has determined that this final role is the type of action described in categorical exclusion 1D CPR 51.22{cJ(3). Therefore, neither an environmental impact statement nor an enviromnental assessment has been prepared for this final rule.

Regulatory Analysis The Commission has ptepared a regulatory analysis for this amendment to the regulations.

The.analysis examines the costs and benefits of the alternatives considered by ilie Commisaion. Interested -persons may exami.De a copy of the regu).q_tory analysis at theNRCPnblic Docwnent Room, 1717 H SL NW., Washingten. DC. Single copies: of'the analysis may be obtained from Mr. G.C. Millman. Division of Engineering. Of&e of _ Nuclear Regulatory Resean:h, U.S. Nuclear *Regulatory.Commission, Washington, DC20555, Telephone (301) 492-3872. . Papeiwork Reduction "'-ct Statement This final.rule amenda information collection requirement& that are subject to the Paperwork. Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were approved by li:iC' Office of Management and BudgeL approval number 3150--0011. 50-SC-32 Regulalory Fiexibility Certification .A.s 1*eouired b,* the Reirn:alo,r Flexibili.ty Act of 1!'.180. 5-U.S.C.'605fb), tne Commission herehv certifies Lliat this rule does not have" a significant economic impact on a substantial number of small entities. This rule affects only the licensing and operation of nuclear power plan.ts. T1e con~panies that own these plants do not fall with.in the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121. List of Subjects in 10 CFR Part 50 Antitrust, Clas!rlfied information, Fire protection; Incorporation by reference, Intergovernmental relafimi.s, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements; Pursuant to t1;ie Atomic. F.ne.rgy* Act of 1954, as-amended, the Energy

Reorganization Act of 1974. as amended, and 5 U.S."C. 553, the NRC is adopting the following amendments to 10 CFR Part 50. 53 FR 19240 Published 5/27/88 Effective 7 /26/88 Retention Periods for Records See Part 4 Statements of Consideration 53 FR 20603 Published 6/6/88 Effective 7 /6/88 10 CFR Part 50 Revlslon of Backflttlng Process for Power Reactors AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.. .

SUMMARY

The Nuclear Regulatocy Commission is promulgating an amended rule which governs the backfitting of nuclear pow*er plants. This action is necessary in order to have a backfit rule which UIJambiguously conforms with the August 4, 1987 decision of the U.S. Court of Appeals for the District of Columbia in Union of Concerned Scientists, et al., v. U.S. Nuclear Regulatory Commission.

This action is intended to clarify when economic costs may be considered in backfitting nuclear power -plants. EFFECTIVE OATE;_luly 6, 1988. PART 50

STATEMENTS OF CONSIDERATION FOR FURTHER IMFORMATION CotHACT: Sleven F. Crockett, Office of the General Counsel, U*.S. Nuclear Regulatory Commission, Washinglon.

DC 20555. Phone: (202) 492-1600. SU?PLEFAENTARY lt~FORMATION: Background On September .20, 1985, after an extensive rulemaking proceeding which included sequential opportunities for public comment on an advanced notice of proposed rulemaking (48 FR 44217; September 28 1983) and a .notice .of proposed rulemaking (49 FR 47034; November 30, 1984); the Commission

adopted final amendments to its rule
which .governs the backfitting of nuclear power plants, 10 CFR 50.109 (50 FR 38097; September 20, 1985). Backfitting is defined in some detail in .the rule, but for purposes of discussion here it means measures which are directed by the Commission or by NRC staff in order to improve the safety of nuclear power reactors, and which reflect a change in a prior Comm_ission or staff position on. the safety matter in question.

decided not to appeal the decision. Instead, the Commission decided to amend both the rule and the related NRC Manual chapter (Chapier 05i4) so that they conform u,nambiguously to the Court's opinion. On September 10, 19S7. the Commission published pn;,posed amendments to the rule (52 FR 34223) and provided for a comment period ending on October 13, 1987.1 The final rule as set .out in this document is substantially the same as the proposed rule (52 FR 34223: September 10, 1987). In this rulemaking the Commission has adhered to the following safety principle for all of its backfitting decisions: The Atomic Energy Act commands the Commission to ensure that nuclear power plant operation provides adequate protection to the health and safety.of the public. In .defining, redefining or enforcing this statutory standard of adequate protection, the Commission will not consider economic costs. However, adequate protection is not absolute protection or zero risk. Hence safety improvements beyond the minimum *needed for adequate protection are . Judicial review of the amended backfit rule and a related int~rnal NRC Manuafchapter which partially implemented it was sought and, on 'August 4. 1987, the U.S. Court of Appeals for the DC Circuit rendered its decision vacating both the rule and the NRC Manual chapter which . possible. The Commission is empowered under section 161 of the Act to*irnpose additional safety requirements not . needed for adequate protection and to consider economic costs in doing so. implemented the rule in paPt. UCS v. NRC, 824 F.2d 103. The.Court concluded that the-rule, when considered along with certain statements in the rule . preamble published in the Federal . Register, 'ilid not speak unambiguously in terms that constrained the Commission from considering economic costs in establishing standards.to ensure adequate protection of the public health and safety as dictated by section 182 of the Atomic Energy Act. At the same time, the Court agreed with the Commission that once an adequate level of safety protection had been achieved .under-section 182, 1he Commission was -fully authorized under section 16li of the

Atomic Energy Act to consider and take economic costs into account in ordering further safety improvements.

The Court therefore rejected the position of

petitioners in the case;"Uriion of Concerned
scientists, that economic.

costs may never be a factor in safety decisions under the Atomic Energy Act.

Because the Conrt'.s opinion regarding the circumstances in which costs may
be considered in making safety decisions on nuclear power plants was completely in-accord with the Commission's own policy views on thii; important subject, the Commission The 1985 revision of the backfit rule, which was the subject of the Court's decision, required, with certain .exceptions., that backfits be _imposed only upon a finding that they provided a substantial increas~ in the overall protection of the public health and .safety or the common defense and security and that the direct and indirect costs of implementation were justified in view of this increased prptection.

The amended.rule, set out in this document, restates the exceptions to this requirement for a finding, so that the rule will clearly be in accord with the _safety principle stated above. I In jts Comments On the proposed amendm.entS, the Union of Concerned Scientists asserts that*the Federal-Register notice of !he proposed amendments was technically defective. UCS argues that si_nce the Court had vacated the entire rule, the Federal Register notice should have proposed * *enactment of an entire, amended, rule, rather than simply amendments to the vacated rule: In weighing the technical

merit of UCS' argument.

It should be note"d that as of the date of the Federal Register notice. the mandate of the Court bad not yet issued end the rule was thus still legally in effect. However, the more important consideration is that the notice clearly revealed the Commission's Intent to reissue the backfut rule once It had been conformed to the Court's decision. UCS understood this intent and took the*opportunity*to renubmit the *comments it had nubmitted during the rulemaking leading up to the 1985 revision of the rule. In any . event. the Commission is publishing the*entire rule In this document.

50-SC-33 Particularly 1n response to tht Court s decision, the rule now provides that if the contemplated backfit involves ddining or redefinirrg what level of protection to the public health and saletv or common defense and securitv shouid be regarded as adequate, neith~r the rule's "substantial increase

standard, nor its "costs justified" standard, see § 50.109(a)(3), is to be applied. (See § 50;109(a)(4)(iii).) Also in response to the Court's decision, see 824 F.2d at 119, the rule now also explicitly' says that the Commission shall always require the backfitting of a facility if it determines that such regulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the public and is in accord with the common defense and security . On instruction from the Commission, the NRC slaff has amended its Manual chapter on plant-specific backfitting to ensure consistency with the Court's opinion. Copies of the revised chapter are available for public inspection in the Commission's Public Document Room, 1717 H Street NW., Washington, DC .* 20555.2 . Response to Comments Comments were received from 12 utilities, one Federal agency (IJOE), one vendor, seven individuals, seven citizens' groups, and two industry .

groups. Lengthy and detailed comments were-submitted by the Union of Concerned Scientists (UCS) and the Nuclear Utility Backfitting and Reform Group (NUBARG).

Both organizations were active in the rulemaking which led to the 1985 revision of the rule. The comments submitted by these two groups encompassed most-of.the comments made by others. Below,.the Commission paraphrases the chief comments and responds to them. The .Commission has given careful consideration to every comment:The original comments,may be viewed.in the NRC's Public Document Room in Washington, bC. *

Several commenters argue that the revised Manual chapter should undergo what amounts to notice *and comment rulemaking.

However, the Manual chapter, If it is a rule at alJ, is a rule of agency organization. procedure. or practice, and ther~fore is not subject to the notice and comment requirements of the Administrative Procedure AcL See 5 U.S.C. S53{h){AJ; see also § 553(a){2). The Commission did publish for comment an earlier version of Manual Chapter {49 FR 16900: April 20, 1984), but that version was already in effect when It was published for comment, and it was published for comment only because the Commission was still in the process of making fundamental changes to *the backfilling process and wanted comment on the procedures then "in effect. See id. PART 50

STATEMENTS OF CONSIDERATION "Adequate Protoctiun
- The great majority of the commenter~

raised issues about the rule's use of thr phrase "adequate proter.tion".'This phrai,e is used in the rule's exceotioP. provisions. See § 50.109[a)l4). Generallv. the rule requires. among other things,

that it be shown for a given proposed backfit that implementation of the backfit would bring about a "substantial increa!;e" in overall protection to pubiic health and safetv. and that the direct and indirect costs of the backfit are justified by that substantial*increase.

See § 50.109(a)(3). However, § 50.109(a)(4) also requires that these two standards not be applied in three situations: First, where the backfit is required to bring a facility into compliance with NRC requirements or the licensee's own written commitments: Second; where the backfit is necessary to ensure that the facility provides adequate protection to the health and safety ofthe public and is in "accord with the .common defense and security; and Third, as noted above, where the backfit invoives defining or redefining what level bf protection to the public health and safety or common defense and security should be regarded as adequate. The comments on the rule's use of the

phrase "adequate protection" generally took two forms, each discussed more fully later on in this notice. The first form, most fully represented by ucs*
comments, wa,s that the rule itself should actµally include a definition of "adequate protection" (the final rule set out in this document does not), a phrase nowhere explicitly defined in general terms, either in the Atomic Energy Act, from which the phrase comes, or iii the Commission's regulations.

The second, more modest, form of the comments on **adequate protection", .. most fully represented by NUBARG's comments, was that one or another of .. the three exception provisions in the rule was redundant (none is). While not amounting to a call for a definition of "adeq{!ate protection", NUBARG's comments displayed some of UCS'

uncertaintv about what the Commission . meantby the phrase. Each group had difficulty applying the phrase to characterize.past Commission

iction in backfitting.

UCS claimed that the Commission had never backfitted in order to achieve something beyond *"adequate protection." NUBARG, however, claimed that the Commission had neverTeqµired e*backfit on the grormds that compliance with the regulations was not enough to provide adequate protection. These view5. imp!1ldent. and certainly notninp iliepil. differing in emphasis. reflect the twu about decisions which ultim.ately turn f!:-Oups' opposite concerns about the on the application-by duly constituted pcssibil:'.y that the Commission vvoulc authoritv and after full consideration of u~:e the ph~c!SC "adequate pro!eciior," all rf'iev',mt information-of ohrases arbiira~iiy. ucs is Concerned that the v,nid, are not fully defined. Consider. Commission might interpret the. phrase for instance. the "reasonable assurance" "adequete protection" to refer to a level determination the Commission must of safety such that every proposed make before issuing an operating improvement would be subjected to license.* Indeed. most of the cost-benefit analvsis. Converselv. the Commission's rules and regulations are industry appears-concerned that the ultimately based on unquantified and. as Commission might interpret the phrase we note below, presently unquantifiable "adequate protection" to refer to a level ideas of what constitutes "adequate of safety such that no proposed protection". improvement would be subjected to Were there something peculiarly cost-benefit analysis. critical about the role of "adequate The Commission certainly did not protection" in the backfit rule, the issue intend that this rulemaking should focus of the phrase's meaning could have been on the meaning of the phrase "adequate raised in the rulemaking for the 1985 protection". The main point of this rule. Two of the three exception rulemaking was simply to negate the provisions set out above were ,*n the misimpression.left by two statements in the preamble to the 1985 version-of the 1985 revision of the rule,-where*they backfit rule. UCS puts forward two used the equivalent phrase "undue risk" grounds for its emphasi's on the phrase instead of "adequate protection". Also. "adequate protection'\ First, UCS as the Court in UCS v. NRC noted, 824 asserts that w(t)he crucial decision as to F.2d at 119, the statement of whether cost benefit analysis will .be considerations which accompanied the . used in assessing the need for 1985 version of the rule quite explicitly backfitting is dependent on whether the 11-t least twice limited the sonsideration particular backfilling under of costs in backfilling decisions to

consideration is needed to ensure situations where "adequate protection" adequate safety * * * ." Second, UCS was already secured. 5 claims that the Court "ordered" the Nonetheless, an issue which is a Commission to "stop trying to obscure concern of almost every commenter in its intentions through ambiguous and this rulemaking should not be ignored. *vague language*
* ." *
Therefore, the Commission will answer . However, as *w:ill .be explained.

more as best it can the questions the fully below, 14e. Court's decision turned conimenters have raised concerning the not on the rule's lack of a definition of rule's use of the phrase* "adequate "adequate protection" but rather on two pi;otection". We begin with UCS'*call for statements which seemed to the Court to an objective.and generally applicable imply that the Commission*intended to -definition of "adequate protection~;We . take cosui"into consideration in argue that such a definition is not determining what "adequate protection" possible in the near future, buUhat the required; the .meaning of "adequate , public :and licensees are nonetheless

protection" was simply not an issue in protected against misuse of .the phrase. the litigation.

Moreover,-UCS In the course of.responding to UCS' . overestimateslhe*r.ole the phrase 'Comments; we*shall. of necessity. be** * "adequate protection" plays in the making at least preliminary. responses to backfit-rule. The threshold decision in most ofNUBARG's comments-also.*.

considering a proposed backfit, and very ucs argues that the rule permits the often the only decision that need be agency to escape its legal responsibility
. made, 3 is not whether adequate protection is at stake but rather whether the facility is in compliance with the Commission's requirements and the licensee's written commitments.

Even if UCS is right about the importance oflhe phrase "adequate protection;

th!lre is*nothing unusu.al or
For inlltance.

a majority of the plant-speclfi*c backfits carried out during the first year after the 1985 re"!ti=: !:'f the beckflt rale became effective were for the sake of ,eompliance. See SECY-416-46. Evaluation of Managing Plant-Specific Backfit

Requirements (November

21. 1986), Enclosure

'l. 50-SC-34 * "' * * {A)n operating.license may.be issued by

the Commission

' *

upon finding that: * * * * (t)here is reasonable assurance
*
that the activ/ties authorized by. the operating license an be conducted without endangering the health.and safetr df the public* * * .* 10 CFR 50.57(a)(3').
"The consideratioo and weighing of costs
contemplated by the ruJe.applieslo backfits 3hat are intended lo result in incremental aafety *
improvements for a plant that already. providef an acceptable degree of protection(,)" 50 FR 38103. col. 1: also, "(t)he 'COSls ausociuted with proposed new safety requirements may be considered by the * -Commission provided thal the Atomic F.nergy Act fihding 'no undue risk' can be made." Id. al 38101.
col. 3.

PART 50

STATEMENTS OF CONSIDERATION lo urticulaie

!he factors on which it bases its b<.1c.kfittmg decisions. UCS a~ser!s that lhe rule shot!ld "enunciate cri1e~u.1 a!ld iiuidelines about what consti:ules redefi!ling and defining adeau.ite protection le*,els. what con;titulcs an adequate as opposed to a beyond adequate protection level, and what factors place a particular circumstance within the rule or within the ex.ccptions." Another comment asseits that any definition of "adequate protection" should include the resolution , of all outstanding safety issues. Yet another calls for ."objective* criteria", -some real numbers" on releases, accident consequences, and the like. There does not exist, and cannot exist, at least not yet, a generally applicable definition of "adequate protection'! which would guard against every possible niisuse of the phrase. Congress.established "adequate protection*~ as.the standard the . Commission is to apply in licensing a

plant, see 42 U.S.C. 2232(a), and gave the .Commission authority to issue rules and regulations necessary for protection of . public health and safety, see 42 U.S.C. . 2201, but ,Congress did not define "adequate protection", nor did it
command the Commission*

to define it. -Such a definition would have to take one of two forms, one of them incapable of preventing the abuses the commenters are concerned about, and the other simply not possible*yet. The first of these would be a verbal

definition of the kind encountered in, for instance, the various "reasonabie man" .. standards in the common law. After -the pattern of fhese, the Commission could say, correctly, that "adequate emerge as a byproduct of tl,~ Commission's efforts, stili in their ear:, stages. to implement ib gene~al s;;f('t~:

goals. which take a nartl-,* quantitat1H form. (See 51 FR 30028; Aui'.Js! '.!1. 193G. Policy Statement o~ Saiety Goals.) However, given the state of ihe art in quantitative safety assessment, it is not reasonable to expect that the Commission could make licensing decisions-let alone decisions on whether to consider cost in wholly on a quantitative definil!on of "adequate protection". Surprisingly, some of the commenters who call for "objective_criteri(l", "spme real numbers; and the like; have in the past criticized quantitative risk assessments. Nonetheless, even in the absence cif a useful and generally applicable definition of "adequate protection", the Commission can still make sound judgments abou_t what "adequate protection" requires, by relying upon eY,pert-engineering and scientific . judgment, acting in the light of all relevant and material'information. As UCS itself s*aid in its comments on the proposed 1~85 revision of the rule, "(u)ltimately, the determination of what standards must be niet in order to provide-a reasonable assurance that the public health and safety will be protected comes down to the reasoned professional judgment of the responsible . official." **

The Commis.sion's exe.rcise of this judgment will take two familiar forms, of which the most important is rule and regulation; An essential point of the Commissio.n's Jiaving regulations is to flesh out the adequate protection" standard entrusted to the Commission by Congress.

See UCS v:NRC, 824 F.2d at 117-18.:Exercising engineering and scientific judgme~t in the light of all . relevant and material information, the NRC identifies potential h~zards and then requires that designs be able to cope with.such.hazards with sufficient safety*margiAS and reliable backup .. O'i1ysteIIis . .-Regulations and guidance arrived atin this way do not, sirictly .. speaking, define" adequate protection, since there will be times when the NRC

issues rules which-require something .
beyond adeqiiate*protection.

protection" is not zero risk. that it is* the .

same as "no undue risk", that it has long-term and s~ort-terin aspects, arid that it is that level of safety. which the Atomic Energy Act requires for initial ,and continued operation of a nuclear power plant.l-iowever, such a *definition clearly will not, ofitself, prevent *the .abuses UCSandNUBARG"iire:
concerned about, nor is such *a standard suffici.ently, heipful_to the NRC staff in actual practice ... Thus:ifthere is to be a useful arid generally applicable definition of *"adequate protection.;it must fake another, more precise form, namely,* quantitative.

Several of the commenters seem to have such 8 definition in mind when they call for "objective criteria", some "real numbers", and the like. In -fact, the Commission is actively . "pursuing reliable quantitative measures of safety, and some quantitative end *

generally applicable definition of * "adequate-protection" may eventually . Nonetheless,-compliance with such *regulations and guidance may be presumed to assure-adequate protection at a minimum, As the Commission has . said on many. occasions, compliance with the Commission's
regulations and guidance "should provide a leve1 of
safety sufficient for adequate protection
of the public.health and safety and
common defense and security under the Atomic Energy Act." (49 FR 47034, 47036, 50-SC-35 coi. 2. l'\ovember

30. 1984, proposed 1985 rule; see abo 50 FR 38097, 38101. col. 3, September

20. 1985, final 1985 rule: 51 FR 30023. col. 1. August 21, 1986. Policy S1a!~1:1cn1 on Safety Goals.) Because "adequate protection" is presumptively assured by compliance with the regulations and other license requirements, all the versions of the backfit rule-the 1970 rule, the 1985 rule, and the one set out in this document.

see § 50.109[a)[4)(i}-have a "compliance" exception: plants out of compliance may be backfitted without findings of "substantial increase" in protection or a **justification" of costs. However-and here is where the lack of a general definition for "adequate protection" poses a "adequate protection" is only presumptively assured by compliance. As the Commission said in promulgating the 1985 revision, the presumption may be overcome by, for instance, new information which indicates that

improvements are needed to ensure adequate protection.

(50 FR 38101, col. 3.) Such new information may reveal an unforeseen significant hazard or a substantially greater potential for a known one, or insufficient margins and backup capability. Engineering judgment may, in the light of such information, conclude that restoration of the level of protection presumed by.the regulations

requires more than co~pliance.

Thus both the 1985 revision and the revision below: contain exemptions for backfits necessary to assure "adequate protection", or, *as the 1985 rule equivalently said, "no undue risk". See § 50.109(a)(4)(ii) of the rule. set otit in this document.

If compliance does not assure adequate prote_c;_tion, the Commission

-. must be able to determine how much: more protection is required, and a precise and generally applicable . definition of "adequate _protection" would facilitate tl1at.determination. But such a definition would have only a limited role to play. The first and most crucial .question is *wh_ether the proposed backfit is required to bring a plant into compliance. Only if the* proposed backfit requires more.than compliance with NRC regulations and license conditions need there be a determination as to what .. adequate protection" requires. Given this relation between compliance and "adequate protection", the industry might be more concerned than UCS is about the lack of a general definition of "adequate protection", for UCS will at least have the comfort of knowing that compliance will be secured before cost is considered, but the industry cannot be PART 50

STATEMENTS OF CONSIDERATION sure !ioV\' m*;Jct' 1;1::~e thu!"i :umpliunce may uc asked c;J 1\ uesr.:le lhe cost. Where, as in the cases contemplated l,y the second exc.:eption provision of the rulc. mun: than cc,I:1pl1crnce is required and amrntilatin criteria do not define "udPquc1te r~o!ectioP.".

the agency mus\ fall back on the second familiar form in which engineering judgment is exercised by the Commission, namely. case. Administrative agencies are not required to proceed by rule alone. for . the method of .. ase-by-case judgment is quite capc:ble of meeting the requirement that the factors.on which administratke decisions are based be articulated. Rather than proceeding by an*almost ministerial application of "objective criteria", the Commission must fashion a series of case-by-case judgments into a well-reasoned and factually well-supported body of decisions which, acting as reasoned precedent;-can control and guide the Commission's exercise of the discretion

granted it by Congress in precisely the ,vay in which common-law precedents eontrol and guide the common law judge's exercise of his or her judgment.

See Nader v. Ray, 363 F.Supp.-946, 954-55 (D.D.C, 1973) (determining what constitutes adequate protection calls for exercise of discretion in a judgmental process very different from acting in accord with a clear, non-discretionary legal duty). The Commission foresaw the-need to proceed case-by-case on.occasion and therefore made *it a principal aim of the backfit rule to centralize the -responsibility and document the bases fQr case-by,:case decisions for such decisions. The Commission thereby hoped to better assure that such * -decisions as might of necessity be by-case. would form *a reasoned and. coherent body.6 * .: * *

UCS alleges that fn three instances the .Commission has abused its discretion by applying cost *considerations in specific cases where licensees are in compliance but adequate proteciion is at stake. Howev~r. UCS is misinformed about the firsi of the three cases, arid its allegations about the
other two. redu*ce simple to disagreement over what constitu!es adequate protection.

We briefly discuss the three cases below.* Citi~g ltade journal articles which quote unnamed NRC sources. UCS claims that the backfit rule

caused-the NRC staff to change its mind about requiring two licensees to conduct certain inspections and analyses in orrler to justify * . continued operations.

The two plants in question had reactor pump coolant shafts similar to ones which elsewhere had shown a high probability of . shearing off under certain condiiions. ucs asserts that "lw)e * * *. learn from this example the 'inherent lack of logic and circularity embedded in the rule: l'IRC is prevented, by.operation of the rule, frum asking questions needed to learn the degree of risk of a known equipment problem-beca,;se 11,ey do not know the an*wers in advance."_ Not~;:ig in the Court's rulin:; ir. UCS \. r,RC forbids the Commission's approach However. the fact,-of th*e situatiuns -...*..-ert no! what UCS alieges 11,em lo have been: indeed the Lac.kfit rule was not invoJv(:d. Letters were sen~ on April 23. 1986 requirm~ the licensees to submi! witl,m Z:J days information which would _"enable the Commission to determine whether or not (their) license(s) should be modified." Such information included information on design, operational history. schedules for inspection, 'plans for operator training. and "'any analysis performed subsequent to those done for the FSAR (Final Sufety Analysis Repo,t) which would address the consequences of a locked rotor or broken shaft event during plant operation." These letters were sent undP.r the first part of 10 CFR 50.54(1): This part authorizes such information requests without consideration of cost. As an earlier draft of the April 23 letter available in the NRC's Public Docwnent Room shows. the NRC had planned to ask for new analyses under a later part of § 50.54(1) which authorizes requests not required to assure adequate.protection if "the burden-to be imposed * * * -is justified in view of the potential safety significance of_the issue to be addressed in the rcquesteil information:'" 10 CFR 50.54[1). (This "safety significance'" standard, by its emphasis on "potential'", requires less than is required by the "[actural) substantial increase" stiindard in the backlit rule ,µid olso avoids the circularity UCS alleges.] However, the staff sensibly opted for first . asking whether iiuch*analyses had already been. done. In fact they bad, *or were underway when "the letters -were sent. The backfit rule played no part here.

ucs* second instance of alleged abuse involves the Mark I containment, about whose performance in beyond-design-basis accidents

[ ones which involve damage to the reactor cor~) there is

substantial uncertainty.

UCS asserts that cost considerations have blocked staff action which would have brdught about a significant reduction in some of the figures which estimate the prob&bility that the i,;fark I would foil in*certain kinds of beyond-design-basis accidents. UCS adds in passing that those-figures represent undue risk. The NRC staff has already made a formal reply to similar charges of undue risk. See. e.g . .-Boston Edison Co. [Pilgrim Nuclear Generating Station). Interim Director's Decision under 10 CFR Z.206, DD-87-14, . 26 NRC'87; 9&;-106 [1987). Suffice it-here to say the NRC staff ha!*by no means _completed its .. considerations of the Mark I.containment; but that, given present information; the-staff has concluded that overall severe-lfccidenf risks at plants with -Mark I coniainments are riot undue: Id. iii iD4--1Jl6. UCS is content_10* put fonrard only unsupported assertions to the contrary *. Thus the staff may legitimately consider cost when deciding whether to . backfit the_Mark_l con)ai_l).lllents. . , . ucs* third_ allegation of abuse rllhearses part of its February 10, 1987_ § 2.206 Petiiion to the Commission for immediate' action to relieve aliegedly undue risks posed by nude'~r pow~r * *plants designed by the.Babcock &-Wilcox The NRC's Director of Nuclear Reactor Regulation -responded fully to the Petition. den~ing it. on

October 19, 19B7 [UCS' COJ!lments on !he propo&etl backlit rule Were submitted ori Octobe*r 13). See Director's Decision Under 10 cm* 2.206. DD-87-1B,
26 NRC-{October 19, 1987). The Director concluded tha! "there are-no substantial*health*and safety issues that would warrant the suspension or revocation of any. license or permit.for such facilities:" Slip _Opinion at 63. Simply because UCS disagrees with such conclusions does not mean that the Commission is misusing° the "adequate
. protectjon~-

standard. 50-SC-36 tc: **ar!eq:;ate protec!!cn**. UCS L:~lc!:~ asserts that the prepused rult "completely fail[ed] lo co:nport with the orders and directions of the Court of Appeals in UCS v. NRC, that the Court "could no! have been more ciear about the defects of the backfit rule", that the proposed revised rule "suffers from the exact same defects" as the one vacated, that. inaeed, "the new proposal is even more devoid of objective guidance or criteria * *

than*was its predecessor." UCS' criticisms are based on part of a single paragraph in the Court's decision.

ln pertinent part, that.paragraph says, * .... *

In our view, the backfitting rule is an exemplar of ambiguity and vagueness; indeed, we suspect that-the Commission designed the rule lo achieve this very result. The rule does not explicate the scope or meaning of the three listed 'exceptions'.

The rule does not explain the action the Commission will (in italics) take when a backfit falls within one of-these exceptions. In short, the rule does not: speak in terms that constrain the Commission from operating

outside the bounds of the statutory scltetne;" 824 * . F.Zd al 119. . UCS says that this portion of a paragraph was an "order'.!

by: the Court to get the Commission to "stop !J"Ying to obscure its intentions through ambiguous and vague language * * * ."

Whether-the Court's laniuage amounts to an '-'order" or cinly strong advice, we have followed it. For one thing, the rule* explicitly says that backfits fallii;ig
within the exceptions will be imposed (inexplicably, UCS asserts that*the proposed rule did not have this provision).

See § 50.109(al{4). For another, bath in :what we haye already said, and in what_ we shall be saying in response to.N(JBARG's.commerits _on the exceptions provisions, we shall .have -explicated the scope imd meaning of the three listed exceptions.

* . 'However; we have-not taken the quoted language of the Court fo mean that, after years of making rules and adjudicating cases 'which ultimately

.. depend on the Commission's-judgment about what "adequate protection" requires; the-Commission should *be obliged to give a mechanically. applicable definition of "adequate -protection" in order to avoid using the time-honored method of case-by-case, . precedenhguided,. judgment to

implement only a part of the l:iackfit , rule. Ce~tainly, the Court never even -noted a lack.of a general definition*

of "adequate pmtection" *in the rule, let alone "ordered"- the ,Commission to provide such a definition.

PART 50
STATEMENTS OF CONSIDERATION ucs* position laci-;s all sense of proport,un.

We must emphasize the core of the Cour!"s decision, rafrier than get bogg~d 001\*n by !ransformini a suspi::ion a!lO a few criticisms of the rule inio an order to undertake an unpreceuen,ed task oi defini!ion. Reviewing the exceptions in the rule, and various statements in the Federal Register notice accompanying the rule. the Court said, "We conceivably could read the terms of this rule to comply with the statutory scheme we have described above {that is, a scheme in which economic costs can play no part in establishing what adequate protection requires)." Id. Moreover, the Court says this despite the lack of any summary, general. "objective" definition of "adequate protection" in the rule. But the Cow't'then went on to say, "Statements that the Commission has made in promulgating the rule and in

defending it before this court. however, disincline us from interpreting the rule in this fashion." Id, Again, it is not the lack of a definition*of adequate protection that disinclined-the Court from saving the rule, but rather certain statements the Commission had made which seemed to suggest that the Commission might consider economic cost when deciding what adequate protection required.

The Three Exceptions Echoing the Court's remark that the rule "does not explicate the scope or meaning of the three .listed 'exceptions' ", id., NUBARG "believes that there is e substantiatamountof overlap in these exceptions and that they have not been adequately defined or explained in the proposed rule." NUBARG and others representing the . industry are concerned that the two exception provisions which use the phrase "adequate protection., §§ 50.109(a)(4){ii).and (iii), may "swallow" the.rule. One industry commenter objects to the notion. implied by§ 50.109(a)(4)(ii). that adequate protection might require more than compliance. Another iii concerned that § "5o.109[a)(4){iHJ, the exception which has been added in response to the

Court's ruling, might lead to . redefinitions of "adequate protection" that .would threaten Joss of licenses.

To avoid these results, NUBARG and 'othersl"ecommend deleting one of the two exception provisions which use the -* phrase "adequate protection".* ... NUBARG's clioiceis § 50.l09{a)(4l{ii),

retained from the 1985 version of the rule, where.it.used the equivalent phrase, "no undue risk". This section provides that the "substantial increase"*

and '"costs justified"' standards will not apply to backfits necessary lo provide adequate protection to public health and safety. NUB.-\RG calls this provision redundant to the exception for backfits required for the sake of co;noliance. § 50.109lal(4J(i). As 'was noted abo\'e, l\'UBARG reports that its research has uncovered no case in which the Commission "has recognized that some additional measures not contained in existing requirements are necessary to ensure that a facility continues to meet the current level of adequacy." Two other commenters believe that U1e exception.pr{)vision added because of the litigation, § 50.109(a)(4)(iii), should be deleted, as being redundant to the provision NUBARG would like to see deleted. No matter which of the two provisions the commenter would like to see deleted, the commenter would like some restrictions placed on the use oi the ~remaining one. The restriction by far the most frequently proposed is that no action may be taken under the remaining exception provision in the absence of "significant new inform!ltion or the occurrence of an event which clearly shows" that the actiori is necessary. In sum, these commenters either reopen an issue settled in 1985 or they recommend deleting that part of the rule which directly responds to the Court's ruling. We take neither course, for, even putting the 1985 rule and the Court's ruling aside, if either of the two provisions were to be deleted, an . essential power of the Commission would be remain unimplemented. First, the exception for backfits necessary to secure adequate protection, § 50.109{a}(4){ii),must be retained, because :it must be made clear that Commission action is.not to be obstructed by cost considerations' in a situation where compliance has.indeed proved to be insufficient tosecure the level of protection presumed .in the rule, *order, or commitment in question.. ' Despit~ the results of NUBARG's research, such situations have arisen. See, e:g., SECY~346. "Evaluation of* Managing Plant-Specific Backfit ' Requirements", November 21, 1986. Accordingly, this exception provision is riot redundant to the exception for backfits necessary to restore

compliance.

Neither is it redundant to the ~xception for backfits involving the defining or redefining of "adequate protection", for,the latter exception ... -* assumes some changejn the NRC's .

judgment _of what level of protection . -should be regarded as *~dequate".

50-SC-37 Retaining§ 50.109(2!!~:iiii wi!! r;J1 give the Commission the power to proclaim at will that compliance is not enough. As we said in the statement of considerations accompanying the 1965 rule. and have in part reiterated in the resoonse to UCS' comments, the re~lations, though they do not define "adequate protection .. , *are presumed lo ensure it, .and, in the absence of a redefinition of '.'adequate protection", that presumption can be overcome only by significant new information or some showing that the regulations do not address some significant safety issue. * "(l)t may be presumed that the current body of NRC safety regulations provides adequate protection. Where new information indicates that improvements are needed to ensure there is 'no undue risk' on * *

a * *
basis which the Commission believes to be the minimum necessary, such requirements must be imposed." (50 FR at 38101-102.)
Second, the exception provision for backfits which *BI'e necessary under a defining or redefining of"adequate protection",§ 50.109{a)(4)(iii}, must be retained because it must be made clear _ that, as the Court held, cost may riot be a factor in setting the level of protection
judged as "adequate".

7 As NUBARG -acknowledges, citing Power Reactor Development Co. v. International, Union of Electrical, Radio and Machine Workers, AFL-CI0,..367 U.S. 396,408 {1961), the Commission has both the power to define ~adequate pmtectionN, arid the power to re-define-it. 8 Without this last exception provision, it might appear from the rule either that the Commission had no such power or that it was restricted by cost considerations, contrary to the Conrt_'_s_ruling. Nor

should this exception provision be limited to situations involving . "signific,ant new information,'.'

as proposrci in several comments. This la.st exception. may be thought by some to threaten to swallow the*backfit rule. We believe; however, that

instances ofbaclcfits based on.a. "redefinition" of "adequate protection" will be rare. Moreover, the case-by-case approach which is required in the
As the rule notes in I 50.109(e){7), cos! may . nonetheless be a conlriderstion in clloosing the means of achieving
adequate protection". , The words udefinlng or Tedcfuring" in this third exception 11hould not be constrned nece9Sarily to mean "providing Ii nseful and iienerally applicable
ilefmition~.

al leaat not until socb a definition becomes possible. Under present conditions. !he Commi1!sion wiH have ~defined or redefined what level of protection ill*to beyeganlcd as adequate* if * ~t makes a judgment thaL mthough compliance . -assures the le,rel of protection that. had been.

thought of 11s adequate.

that level of protectimi should no longer be considered adequate. PART 50

STATEMENTS OF CONSIDERATION absence of a general ddini1ion of adequate protection" provides licensees-and the public-a large measure of protecllon from arbitrarv action by thE: Commission.

Citing c~se law. NUBARG says thal. in applying H1is las! exception provision. the Commission "must act rationally and consistently in light of available evidence", and "must apply a reasoned analysis indicating the prior policies and standards are being changed, not casually ignored * * *."We wholly agree, and believe that the approach envisioned by the backfit.rule will facilitate the Commission's acting accordingly. Other Matters Two other comments bearing on the phrase .adequate protection" require an explicit response. First, several commenters from the industry would prefer that the rule state that the "documented evaluation" which the NRC must prepare in connection with any action under one of the exception

provisions, see § 50.109[a)(4), should include consideration of as many of the factors which§ 50.109(c) requires of a . "backfi.t analysis'\

as are appropriate.

The suggested modification of the rule

would have only limited utility. Few of the factors listed in§ 50.i09(c) of the rule are appropriate for consideration in a documented evaluation justifying action under the compliance e*xceptiori in the rule. It is true that several of the factors in § 50.109(c), indeed, all:of.them but those in paragraphs (c) (5) and (7) and some of those in paragraph (c)(8) thnt since rules "define" "c:dequatt* . protection", the Commission cannu! apply the rule's "substantial incn:,!sc*" nnd "cost justified" standards ir: rulemaking without applyin~ cos! considerations in setting the standa:c_;

c*i adequate protection, contrary to the Court's holding. The answer to this comment is, of course, .that the rules do not, strictly speaking, "defi~" "adequate protection", and they only presumptively assure it. Not only may there, as stated above, be individual cases that require actions that go _beyond what is necessary under the regulations to assure adequate protection, .there will also be times.when the l\o'RC issues a rule which requires something b~yona adequate protection. This follows directly from the Commission's power under section 161 of the Atomic Energy Act, affirmed by * .the Court, to issue rules or orders to "minimize danger to life or property." See 42 U.S.C. 2201; see also USC v. NRC, 824 F.2d at 118. If a proposed rule requires something more than adequate . protection, applying a cost standard to the proposed rule will not be introducing. cost considerations int.o the setting of

the adequate protection standard and is . therefore permitted.

Of course if the rule is directed at either establishing what

level of protection is "adequate" or assuring that such a level of protection is mel, then cost* will play no role.
The bat:kfit rule as set out below is substantially the same as the rule * . are approprill.te for considimition under . th~ "adequate protectiori" exception, to the extent that they require a showing of_ . proposed in.the Federal Register. (See 52 FR 34223; i,eptember 10, 1987.)
Provisions' which appeared at the end of § 50.109(a)(4) of the proposed rule, or in _ exactly what the licensees must do.and l} showing th~t the backfit-in question actually contributes to.safety; However, the Commission believes that the.rule's requirement that the documented . evaluation
~include a statement'of the objectives of and reasons .for the .. , .. . modification and the basis {or invoking the exception" adequately~.i.ssures that the fact_ors in § 50.109( c) 'r'{m be -considered to the extent relevant, witlrourtheir being listed and labeied as if they were a part of a § 50.109( c) analysis.

Thus, little; if anything, is to be gained.by an explicit requirement that * § 50.109(c) factors be considered in a *documented evaluation. Second, one citizens' group asserts * * *that the backfit _rule should not apply _to ruleniaking. This issue was thoroughly discussed in 1985. However, this group's . comment puts the issue in a slightly alteredlight,. and provides another

opportunity to clarify the meaning .of "adequate protection".

The-group argues the footnote to that paragraph, appear below in new paragraphs (a) (5) through (7). . En'rironmental Impact: Categm:ical

Exclitsion The NRC'has det~rmined that this final rule is the type of action described

'in*categorical exclusion 10 GfR 51.22(c)(3).Therefore, neither an environmental impact statement nor an environmental assessment has been prepared for this final rule. Paperwork Reduction Act.Statement This*final rule does not contain a new or amended information c*onection

requirement subject to the Paperwork -Reduction Act of 1980 (44 U.S.C. 3501 et -seq.). Existing requirements were approved by the Office of Management and Budget; Approval Number 3140-0011. . . 50-SC-38 The revision to 10 CFR 50.109 will bring it into conformance with the holding in Union of Concerned Scientists.

et al.. i.*. U.S. Nuclear Regulator~* Commission. D.C. Cir. Nos. 85-1757 and 86-1219 (August 4. 19B7). The revision clarifies the backfit rule to reflect NRC practice that in determining whether to adopt a backfit requirement. economic costs will be considered only when addressing those backfits involving safety requirements beyond those needed to ensure the adequate protection of public health and safety. Such costs are not considered when establishing the adequate protection of public health and safety. This revised rule does not have a significant impact on State and local governments and geographical regions, public health and safety, or the environment; nor does it represent substantial costs to licensees, the NRC, or other Federal -agencies. This constitutes the regulatory analysis for this rule. Regulatory.Flexibility Act Certification Jn accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the .Commission hereby certifies that this final rule, if promulgated, will not have a significant economic impact on a substantial number of small entities. The affected facilities are licensed under the provisions of 10 CFR e0.21(b) and 10 Cffl 50.22. The companies that own these facilities do not fall within the scope of "small entities"as set forth in the Regulatory Flexibility Act or the Small ijusiness Size Standards set forth in regulations issued by the Small* Business Administration in 13 CFR*Part 121. Backfit Analysis

The NRC has determined that a backfit analysis is* not required for this . rule because it does not impose requirements on 10 CFR Part 50 licensees.

List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and Recordkeeping requirements. . . . .

Fo~ ihe reasons set out in the
preamble and under the authority
of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following
amendments.

to 10 CFR Part 50. PART 50

STATEMENTS OF CONSIDERATION
53 FR 23203 Published 6/21/88 Effective 7 /21 /88 10 CFR Part SO Station Blackout AGENCY: Nuclear Regulatory Commission..

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission is amending its regulations to require that light-water-cooled nuclear power plants be capable of

withstanding a total loss of alternating current [ac) electric power [called "station blackout")

for a specified duration and maintaining reactor core cooling during that period. This requirement is based on information

developed under the Commission's study of Unresolved Safety Issue A-44, "Station Blackout." The amendment is intended to provide further assurance t~at a loss of both offsite power and onsite emergency ac power systems will not adversely affect the public health
and safety. EFFECTIVE DATE: July 21, 1988. FOR FURTHER INFORMATION CONTACT: Aleck Serkiz, Division of Reactor and Plant Systems, Office *of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 2.ll555, Telephone:

(301) 492-3555. SUPPLEMENTARY INFORMATION: Background The alternating current (ac) electric power for essential and nonessential service in a nuclear power plant is supplied primarily by offsite power. Redundant onsite emergency ac power systems are also provided in the event that all offsite power sources are lost. These systems provide power for various safety functions, including . reactor core decay heat removal and containment heat removal, which are essential for preserving the i~tegrity of the reactor core and the containment building, respectively. The reactor core decay heat can also be removed for a limited time period by safety systems that are independent of ac power. The term "station blackout" means the loss of offsite ac power to the essential and nonessential electrical buses concmrent with turbine trip and the unavailability of the redundant onsite emergency ac power systems [e.g., as a result of units out for service of maintenance or repair, failure to start. on demand, or failure to continue to run after start). Ha station blackout persists for a time beyond the capability of the ac-independent systems to remove decay heat. core melt and containment failure could result. The Commission's existing regulations* establish requirements for the design and testing of onsite and offsite electric power systems that are intended to reduce the probability of losing all ac power to an acceptable level. (See General Design Criteria 17 and 18, 10 CFR Part 50, Appendix A.) The existing regulations do not require explicitly that nuclear power plants be designed to assure that core cooling can be maintained for any specified period of loss of all ac power. As operating experience has accumulated, the concern has arisen that the reliability of both the onsite and offsite emergency ac power systems might be less than originally anticipated, even for designs that meet the requirements of General Design Criteria 17 and 18. Many operating plants have experienced a total loss of offsite power, and more occurrences can be expected in the future. Also, operating experience with onsite emergency power systems has included many instances when diesel generators failed to start. In a few cases, there has been a complete loss of both the offsite and the onsite ac power systems. During these events, ec power was restored in a short time without any serious consequences. 50-SC-39 In 1975, the results of the Reactor Safety Study (WASH-1400) 1 showed that station blackout could be an important contributor to the total risk from nuclear power plant accidents. Although this total risk was found to be smell and not undue, the relative importance of the station blackout accident was established. Subsequently, the Commission designated the issue of station blackout es an Unresolved Safety Issue (USI); a Task Action Plan (TAP A-44) was issued in July 1980, and studies were initiated to determine whether additional safety requirements were needed Factors considered in the analysis of risk from station blackout included: [1) The likelihood and duration of the loss of offsite power; (Z) the reliability of the onsite ac power system; and [3) the potential for severe accident sequences after a loss of all ac power, including consideration of the capability to remove core decay heat without ac power for a limited time period The technical findings of the staff's studies of the station blackout issue are presented in NUREG-1032, "Evaluation of Station Blackout Accidents at Nuclear Power Plants, Technical Findings Related to Unresolved Safety Issue A-44." Additional information is provided in supporting contractor reports: NUREG/CR-3226, "Station Blackout Accident Analyses," published in May 1983; NUREG/CR-2989, "Reliability of Emergency AC Power Systems et Nuclear Power Plants," published in July 1983; NUREG/CR-3992, "Collection and Evaluation of Complete and Partial Losses of Offsite Power at Nuclear Power Plants," published in February 1985; and NUREG/CR-4347, "Emergency Diesel Generator Operating Experience, 1981-1983," published in December 1985. The major resulta of these studies ere given below.

Losses of offsite power can be .characterized as those resulting from plant-centered faults, utility grid blackout, and severe-weather-induced failures of offsite power sources. Based on operating experience, the frequency of total losses of offsite power in operating nuclear power plants was found to be ebout one per 10 site-years.

The median restoration time was about one-half hour, and 90 percent of the offsite power losses were restored within approximately 3 hours [NUREG/ CR-3992).

The review of a number of representative designs of onsite emergency ec power systems has 1 Copies of ell NRC documents ere available for public inspection end copying fore Fee al the NRC Public Document Room at 1717 H Street. NW., Washington.

DC 20555. Copiea or published documenla may alao be purchaaed through the U.S. Government Printing Office by calling (202) 275-2060 or by writing to the Superintendent of Documenta, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082. PART 50

STATEMENTS OF CONSIDERATION indicated a variety of potentially important failure causes. However, no single improvement was identified that could result in a significant improvement in overall diesel generator reliability.

Data obtained from operating experience in the period from 1976 to 1980 showed that the typical individual emergency diesel generator failure rate was about 2.5 x 10*2 per demand (i.e., one chance of failure in 40 demands), and that the emergency ac power system unavailability for a plant which has two emergency diesel generators, . one of which was required for decay heat removal, was about 2 x 10*3 per demand (NUREG/CR-2989).

Compared to the data in NUREG/ CR-2989, updated estimates of emergency diesel generator failure rates Indicated that diesel generator reliability has improved somewhat from 1978 to 1983. For the period 1981 to 1983, the mean failure rate for all demands was about 2.0 x 10*1 per demand (i.e., one chance of failure in 50 demands).

However; the date also indicate that the probability of diesel generator failures during actual demands (I.e., during losses of offsite power) is greater than that during surveillance tests (NUREG/ CR-4347). .

Given the occurrence of a station blackout, the likelihood of resultant core damage or core melt is dependent on the reliability and capability of decay heat removal systems that are not dependent on ac power. If sufficient independent capability exists; additional time will be available to restore ac power needed for long-term
cpoling (NUREG/CR-3226).
It was determined by reviewing design. operational and site-dependent factors that the expected frequency of core damage resulting from station blackout events could be maintained near 10** per reactor-year with readily achievable diesel generator reliabilities, provided that plants are designed to cope with station blackout for a
specified duration.

The duration for a specific plant is based on a* comparison of the plant's characteristics to those factors that have been identified as the main.contributors to risk from station blackout (NUREG-1032). The stafrs technical findings show that station blackout does not pose an

undue risk to public health and safety.
The findings summarized above show that recovery from loss of offsite power occurs for the most part in less than 4
hours, emergency diesel generator reliability is high (i.e.,;..0.95), and that * *
given a station blackout the likelihood of core damage is more dependent

_on decay heat removal systems that are non-ac-dependenL However, plant design and operational characteristics, plus site-dependent factors (such as anticipated weather. conditions) introduce a level of variability which warrants a need foi; plant-specific coping analyses to provide greater assurance .that core cooling can be maintained until ac power is restored. Thus the Commission believes that § 50.63 of 10 CFR Part 50 will bring about a significant increase in protection to the public health and safety. As a result of station blackout coping analyses, improved guidance will be provided to licensees regarding maintaining minimum emergency diesel generator reliability to minimize the probability of losing all ac power. In addition, the Commission is amending its regualtions by adding a new§ 50.63 to require that all nuclear power plants be capable of coping with a station blackout for some specified period of time. The period of time for a specific plant will be determined based on a comparison of the individual plant"s design with factors that have been identified as the main contributions to risk of core damage resulting from station blackout These factors, which vary significantly from plant to plant because of considerable differences in design of plant electric power systems as well as site-specific considerations, include: (1) Redundancy of onsite emergency ac power sources (i.e., number of sources minus the number needed for decay heat removal), (2) reliability of onsite emergency ac power sources (usually diesel generators), (3) frequency of loss of offsite power, and (4) probable time to restore offsite power. The frequency of loss of, and time to restore, offsite power are related to grid and switchyard reliabilities, historical weather data for severe storms, and the availability of nearby alternate power sources (e.g .. gas turbines). Experience has shown that long duration offsite power outages are caused primarily by severe storms (hurricanes, ice, snow, etc.). The objective of the rule is to reduce the risk of severe accidents resulting from station blackout by maintaining

highly reliable ac electric power systems and, as additional defense-in-depth, assuring that plants can cope with a station blackout for some period of time. *
The rule requires all plants to be able to cope with a station blackout for a*
specified acceptable duration selected on a plant-specific basis. All licensees and applicants are required to assess the capability of their plants to cope with a station blackout (i.e., determine that the plant can maintain core cooling with ac power unavailable for an 50-SC-40 acceptable period of time), and to have procedures and training to cope with such an event. Licensees may use an alternate ac power source if that source meets specific criteria for independence*

and capacity and can be shown to be available within one hour to cope with *a station blackout. A coping analysis is not required for those plants that choose this alternate ac approach if the alternate ac can be demonstrated by test to be available to power the

shutdown buses* within 10 minutes of the onset of station blackout.

Use-of,an alternative ac source, one that

minimizes common mode failure, is a preferred option since this approach will also-benefit other safey concerns.

On the basis of station blackout studies conducted for USI A-44 and presented in tl1e reports referenced above, the NRC staff has developed Regulatory Guide 1.155 entitled "Station Blackout," which presents guidance on (1) maintaining a high level of reliability for emergency diesel generators, (2) developing procedures and training to restore offsite and onsite emergency ~c power should either one or both become unavailable, and (3) selecting a specific acceptable station blackout duration which the plant would be . capable of surviving without core damage. Application of the methods in this guide would result in selection of an acceptable station blackout duration (e.g., 2, 4, 8, or 16 hours) which depended on the specific plant design and related characteristics acceptable to the staff. However, applicants and licensees could propose alternative methods to those specified in the regulatory guide in order to justify other acceptable durations for station blackout capability. Additionally, the regulatory guide on station blackout presents guidance on quality assurance and specifications for alternate ac source(s) and non-safety-related equipment required for coping with station blackout. The equipment Installed to meet the station blackout rule must be implemented so that it does not degrade the existing safety-related systems. This is. to be accomplished by making the

non-safety-related equipment . independent to the extent practicable from existing safety-related systems. The guidance provided In the regulatory guide illustrates the specifications that the staff would find acceptable for safety systems and equipment.

The quality assurance guidance for the safety-related equipment for which there are no existing NRC quality assurance requirements (e.g., Appendix B. Appendix R) embody the following e'.ements: (1) Design control and PART 50

STATEMENTS OF CONSIDERATION procurement document control, (2) instructions, procedures and drawings, (3) control of purchased material, equipment and services, (4) inspection, (5) test end test control, (6) inspection, test and operating status, (7) conforming items, (8) corrective action, (9) records, (10) audits. NRC inspections will focus on the implementation and the effectiveness of these quality controls as described in the regulatory guide. Based on the rule end regulatory guide, those plants with an already low risk from station blackout would be required to withstand a station blackout for a relatively short period of time and probably would need few, it any, modifications as a result of the rule. Plants with currently higher risk from station blackout would be required to withstand somewhat longer duration blackouts.

Depending on their existing capability, these plants might need to make hardware modifications (such as increasing station battery capacity or condensate storage tank capacity) in order to cope with the longer station blackout duration. The rule requires that each light-water-cooled nuclear power plant licensed to operate must be able to withstand for a specified-duration and recover from a station blackout. The rule requires each plant to perform a coping analysis end identify the coping duration, along with the basis therefor and a description of procedures established for coping end recovery. If modifications to equipment or plant procedures are necessary, these are to be identified and a schedule provided for implementing such changes. It should be noted, based on all

evidence that staff has on hand, that no undue risk exists with, or without, the promulgation of the station blackout rule. However, station blackout may still remain an important contributor to residual risk. This station blackout rule will enhance safety by accident prevention and thereby reduce the likelihood of a core damage accident being caused by a station blackout occurrence.

This does not mean however, that further enhancements in reducing the overall residual risk ere not achievable by additional improvements in severe accident management, given the assumption that core damage occurs, whether from station blackout sequences or other causes (such as small or large loss-of-coolant accident sequences). Initiatives that provide such safety enhancements (through improvements of core damage management procedures) are currently being pursued apart from the station blackout rule. Therefore, this rule should be viewed as being in the same accident prevention context as the ATWS.rule ( § 50.62) and the fire protection rule (§ 50.48) in that it r.ecogniz.es, es the other two rules recognize, multiple failure possibilities resulting from common cause effects that should be addressed. This concern has been recognized in the Introduction to

Appendix A of 10 CFR Part 50. Proposed Rule On March 21, 1986, the Commission published a proposed rule In the Federal Register (51 FR 9829) that would require (1) light-water-cooled nuclear power plants to be capable of coping with a station blackout for a specified duration, end (2) licensees to determine the maximum duration for which their plants as currently designed are able to cope with a station blackout.

A 90-dey comment period expired on June 19, 1986. On April 3, 1986 (13 days after the proposed rule was published), the NRC published in the Federal Register (51 FR 11494) a notice of availability and request for comments on a draft regulatory guide entitled "Station Blackout" (Task SI 501-4). This draft guide provided guidance for licensees to comply with the proposed station blackout rule. Many letters commenting on the proposed rule also included comments on the draft regulatory guide. Responses to these comments provided below address the public comments on

the draft guide as well as on the proposed rule. Comments on* the Proposed Rule The Commission receives 53 letters commenting on the proposed rule.2 Forty-five of these were froitrthe nuclear industry, comprised of electric utilities, consortiums of electric utilities, vendors, a trade association, and ail architect/

engineering firm. Other lettera were submitted by the Union of Concerned Scientists, the Department of Nuclear Safety of the State of Illinois, a representative of the Professional Reactor Operator Society, a citizens group, .a consultant, and three individuals. Largely, the industry comments were opposed to generic rulemaking to resolve the station blackout issue. The Nuclear Management and Resources Council (NUMARC). formerly the Nuclear Utilities Management and Resources Committee, submitted, along with its comments on the proposed rule, a set of four industry initiatives that it believes 1 CopiH are available for public Inspection and copying for a fee at the NRC Public Doc:ument Room at 1717 H Street, NW .. Washington. DC. 50-SC-41 would resolve this issue without rulemaking. Thirty-nine of the industry letters supported NUMARC's submittal. NUMARC proposed a fifth initiative (see iteni 21) by letter dated October 6, 1987, On the other hand, the Union of Concerned Scientists, the Illinois Department of Nuclear Safety, and the citizens group supported the Commission's objective in the proposed rule, but did not believe the rule end guidance associated with the rule went for enough to reduce the possibility of a serious accident that could be initiated by a total loss of ec power. Every letter was reviewed and considered by the staff in formulating the final resolution of USI A-44. Because of the large number of comments, it was not practical to prepare formal responses to each one separately. However, since many comments were on similar subjects, the discussion and response to the comments have been grouped into the following subjects: 3 1. Quality classification of modifications.

z. Whether the backfit analysis adequately Implements the Backfit Rule. *, 3. Cost-benefit and whether § 50.63.meets "substantial increase in the overall protection of the public health and safety". 4. Whether NRC should require substantial improvements in safety that go beyond those proposed in this rulemaking.

5. The need for generic rulemaking.

6. Applicability of the proposed I 50.83 to specific plants. 7. Plant-specific features and capabilities.

8. The source term used to estimate consequences.

9. Specificity on the extent of required coping studies. . *

10. Acceptable duration for coping with a** station blackout.

11. Credit for alternate or diverse ec power sources. . 12. Trends on the reliability of ~c power sources. 13, Sharing 9f emergency diesel generators between unite at multi-unit sites. 14. Certification of the definitions of station blackout and diesel generator failure. 15. Specificity end clarification of requirements.

16. Technical comments on NUREG-1032.

I7. Relationship of USI A-44 to other NRC Generic lssu.es.

18. An alternative of plant-specific probabilistic assessments.

19. Procedure, and operator actions during ' station blackout.

20. Schedule provisions in the proposed I 50.63.

21. Industry lnltlaUv!!S, The comments end responses to each of these subjects are presented on the following pages.
The first four eubJecla are ones on which the Commlsolonera specifically requested public comments when the proposed rule wee published.

PART 50

STATEMENTS OF CONSIDERATION

1. Quality Classification of. Modifications The Commission requested comments on whether the staff should give further consideration to upgrading

_to safety grade the plant modifications needed (if any) to meet the proposed rule. Upgrading to safety grade would further ensure appropriate licensee attention is paid to maintaining equipment in a high state of operability and reliability. . Comments-The prevailing view by industry on this subject is representt:ld . by the following comments submitted by NUMARC: Quality classification is unnecessary...:.. Equipment used to prevent or respond to a station blackout should be sufficiently available and operable to meet its required function. To this extent, the Commission's desire that appropriate attention be paid to maintaining a sufficiently high state of operability and reliability is appropriate. The point of departure begins with the method for achieving this objective. Specifically, by Itself, a "safety grade" classification scheme does not solely equate with high states of equipment operability end reliability. Such classification systems too often can become a documentation exercise more than a process for providing the requisite level of system

functionality.

Duquesne Light agreed with this view and expressed the following comments: &_ty plant modifications or additional equipment required to meet the proposed rule should not be specified safety grade. For equipment which is to be manually started end placed in service-for testing or in the event of a loss of power condition there is no necessity for specifying safety grade since adequate reliability can be obtained through normal surveillance testing and the proper maintenance of commercial power plant equipment. The cost difference in safety grade vs. commercial grade modifications is significant and must he emphasized. The opposite point of view was taken by the Illinois Department of Nuclear Safety. No credit should be given for the capability of equipment to respond to a station blackout unless the equipment was originally designed, constructed, inspected, . performance tested. qualified, certified for !he intended safety-related purpose, and the eq1.1ipment is maintained to the highest industry safety standards. Gulf States Utilities commented, The proposed rule does not provide sufficient direction on the quality classification of plant modifications that may be required to meet the rule. *. *

the quality classification of plant modifications implemented to meet the proposed rule should be commensurate with classification of the system they support. Response-The proposed§ 50.63 does not specifically address the topic of safety classification of plant modifications; however, detailed guidance is provided in Regulatory Guide 1.155 dealing with quality assurance and equipment specifications for non-safety-related equipment.

Any safety,related equipment used either presently, or in modifications resulting from this rule, should meet the criteria currently applied to such equipment. The technical analyses performed for USI A--44 (NUREG-1032) show that plant-centered events (i.e., those events in which design and operational characteristics of "the plant itself play a role in the likelihood of loss of offsite power), and area-or weather-related events (e.g., grid reliability or external influences on the grid) are the dominant causes of loss of offsite power. Neither seismic events nor events related to single failure causes were found to be major contributions to loss of offsite power. Therefore, both the stafrs findings and public comments received do not support an explicit rieed for plant modifications for coping with station blackout to be seismically qualified. The substantial increase in protection sought by this rule can be achieved by modifications which meet criteria somewhat less stringent than generally required by safety grade criteria. related equipment modifications to meet all safety-grade-related criteria would be more burdensome and expensive and would likely achieve only a very small further reduction in risk. The major contributors to the residual risk of loss of offsite power are adequately dealt with by modifications which conform to the quality assurance and equipment specification guidance provided in Regulatory Guide 1,155.

2. Whether the iJackfit Analysis Adequately Implements the Backfit Rule In addition to comments on the merits of the proposed rule, the Commission specifically requested comments on whether the backfit analysis for this rule adequately implements the Backfit Rule, § 50.109 of 10 CFR Part 50. Comment.Of-The Commission received two differing views in response to this request. On one hand, NUMARC expressed the view that the proposed rule does not meet the backfit rule standard because the analysis of the factors set forth in§ 50.109(c) were not adequately considered by the staff. Specifically, NUMARC stated: 1, Installation and continuing costs associated with the backfit have been underestimated.

2. Potential impacts on radiological exposure of facility employees should be further addressed.

50-SC-42 3. The relationship to proposed and existing regulatory requirements should be considered further. 4. Potential impacts of differences in facility, type, design, or age should be considered further. 5. The reduction in risk from offsite releases to the public has been overestima tcd. On the other hand, the Ohio Citizens for Responsible Energy (OCRE) and the Union of Concerned Scientists commented that the backfit rule should not apply to the proposed rule. OCRE took the position that "application of the backfit rule to [NRC) rulemakings

*
is plainly illegal," and the Commission is not empowered to consider costs to licensees in deciding whether to impose new requirements.

The Union of Concerned Scientists commented that the cost-benefit analysis should not be . applied in this case because safety improvements are needed to secure compliance with existing NRC regulations, specifically General Design Criterion 17, Electric Power Systems (Appendix A to 10 CFR Part 50). Response-NUMARC's comments on the backfit analysis were taken i~to account by the staff in revising the draft version of NUREG-1109, "Regulatory Backlit Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout," and a separate appendix that addresses the factors in § 50.109(c) was added to that report. All but Item 2 above are on the same subjects as letters from other commenters and are discussed In more detail under subjects 3 [Item 1), 6 (Item 4), 8 (Item 5), and 17 (Item 3) in this section. NUMARC's Item 2, the potential impact on radiological

exposure of facility employees, would need to be assessed in detail only if it were a major factor in the value-impact analysis.

The effect of radiological exposure on facility employees, if any, would be extremely small in comparison to the reduction in radiological exposure to the public from accident avoidance. Therefore, this factor would have no impact on the overall value-impact analysis.

Contrary to OCRE's and the Union of Cemcemed Scientists' comments, the Commission may subject the rulemaking process to internal controls.

Moreover, the Commission is empowered to consider the costs of incremental safety improvements which go beyond the level of safety necessary to ensure no undue risk to the public health and safety. Sre UCS, et al., v. NRC, D:c. Cir. Nos. as-1-57 and 86-1219 (August 4, 1987). The improvements embodied In § 50.63 go beyond the level of safety necessary to ensure no undue risk. PART 50 ** STATEMENTS OF CONSIDERATION Finally, contrary to.the Union of Concerned Scientists' cominent on GDC 17, new station blackout measures cannot be imposed on licensees as a matter of compliance with GDC 17, under the compliance exception in the backfit rule, § 50.109(a)(4)(i). GDC 17 does not explicitly require that each plant be able to withstand station blackout for a specified time, or that each licensee perform a coping

assessment and make whatever modifications may be necessary In the light of that assessment.

Nor are any of these highly specific requirements logically compelled by any part of GDC 17. Moreover, GDC 17 has never been interpreted by the staff or the Commission to contain these specific requirements. Thus, to impose them under GDC 17 would amount to a backfit which resulted from a new staff and Commission interpretation of GDC 17. The issue in this rulemaking is whether some additional protection is warranted beyond that already provided. The Commission is entitled to inquire, and seek public comment on, whether additional safety measures should be imposed where there is a substantial increase in the overall protection of public health and safety and the cost of implementation is justified in view of this increased protection.

3. Cost-Benefit Analysis and Whether § 50.63 Meets the "Substantial Increase in the Overall Protection of the Public Health and Safety" Chairman Zech and Commissioner Roberts requested comments on the analysis of cost benefit, value impact, and safety improvements and the station blackout standing on the overall risk (e.g., is the reduction of risk only a small percentage of the overall risk, or is it a major component of an already small risk?). Chairman Zech and Gommissioner Roberts were particularly interested in specific comments assessing whether or not this proposal meets the "substantial increase in the overall protection of the public health and safety * * *" threshold now required by the backfit rule. Comments-{A)

One of the major comments by industry on the benefit analysis was that the costs of implementing the proposed requirements have been underestimated. NUMARC end the Atomic Industrial Forum (AIF) commented that the cost estimates for hardware modifications reported in NOREG/CR-3840, "Cost Analysis for Potential Modifications To Enhance the Ability of a Nuclear Plant To Endure Station Blackout," were too low. Commonwealth Edison and other* utilities felt that performance of en analysis to determine the maximum duration a nuclear plant could cope with a station blackout would be substantially costlier than what is estimated in NUREG-1109. Industry also expressed concern that the interpretations associated with the proposed rule could lead to substantial costs above those addressed by the NRC staff in its backfit analysis. AIF commented that "The estimate of 120 NRC man-hours per plant [for NRC review] * *

appears inadequate to account for technical review and evaluation of the determination of
maximum coping capability and of the description of station blackout procedures which the rule would require each licensee to submit." (B) Several commenters expressed the view that the NRC failed to consider all the risks associated with a station blackout in its value-impact assessment.

The Union of Concerned Scientists thought independent failures, in addition to failures that lead to a station blackout, should be included. One individual stated that "both NRC reports [NUREG-1109 and NUREG-1032] are completely deficient in that neither look at sabotage. OCRE commented that seismic events should also be considered: (CJ With respect to safety improvements and overall risk. different points of view were expressed. On one hand, NUMARC commented that, while the risk reduction might be large for a limited number of plants, the risk reduction associated with the majority of plants will be small. Thus, as a general matter, the reductions in risk offered by the proposed rule constitute a small percentage of the oveall risk, a risk which is already small (and acceptable). AIF stated that there is no standard by which to conclude that "substantial additional protection will be realized. A different view was expressed by the Union of Concerned Scientists who stated that "station blackout is clearly a major component of the total risk posed by operating nuclear plants. The magnitude of the total risk is largely unknowable due to the enormous uncertainty which surrounds probabilistic assessments." Response-CA) In order to adequately respond to industry's comments above, the staff and NRC contractors reviewed the cost estimates associated with implementing the station blackout rule. Based on this review, the estimated costs for hardware modifications were reviewed end are in the range of from 20 50-SC-43 percent to almost 140 percent greater than the* estimates in NUREG/CR-3840, depending on the specific modification considered. On average, the cost estimates for hardware backfit were found to be approximately 80 percent greater than estimated in NUREG/CR-3840. However, the cost estimates in NUREG/CR-3840 were not used by the staff in the value-impact analysis in the draft version of NUREG-1109 where estimates approximately 100 percent greater than the NUREG/CR-3840 estimates were used. Therefore, the revised cost estimates used in the final value-impact analysis are not significantly different from the estimates used In the draft version. Industry's comments on the costs to assess a plant's capability to cope with a station blackout were based on the proposed rule that required an assessment of the maximum coping capability and the potentially unbounded nature of such an assessment. Based on public comments, the Commission has revised the final rule to modify the requirement for licensees to determine the maximum coping capability. (See response to public comments in subject number 9.) Instead, a coping assessment is required only for a specific duration. The cost for such a study is estimated to be from 70 to 100 percent higher than the original estimates by the staff, and these revised costs are used in the final value-impact analysis. The staff revised iis estimate of the resource burden on NRC for review from 120 to 175 person-hours per reactor. This revision was based on technical review required for other comparable NRC activities.

(BJ The technical analyses performed for USI A-44 indicated that the contribution to core damage frequency from independent failures, in addition to failures that must occur to get to a station blackout, is low. Likewise, results of USI A-44 studies and other probabilistic risk assessments have shown that, for station blackout sequences, the contribution to core damage frequency from seismic events is low. Not all events can be analyzed on a probabilistic basis. Sabotage is an example. Even though sabotage was not explicitly considered in the stafrs impact analysis, it is discussed in NUREG-1109 under other considerations.

These considerations support the conclusion that a station blackout rule will provide a substantial safety benefit. (CJ The revised value-impact analysis performed for the resolution of USI A-44 PART 50

STATEMENTS OF CONSIDERATION indim&es that there me substantial heaefits in lemu of redm:ed COM da.mQge £,,eqaeocy and reduced risk to the pubac that result from the station blackoat rule. and the costs are wammted in light al. these beoefits.

The best e1tima lie for tne owrall impact ratio is 2AOO person-rem per million ~rs. Ewen if those p&ants wilh the risk (and tbelefore the greateet dlk ~n} were not conaidered. the "fllllle-impact .ratio for the re,naining pmilll ia still favorable (i.e., *oat t.500 pel'IIGlll-rem per million doll.an}. Analyses reported in NUREG-1150, "R.eactar Risk Reference Document ( draft issued for comment in February 1987). 4 indicate that .&tatioo Waclrout is a dwnmant risk Cllllltributor ta overall residual risk for m0.'1t of the aix plants analyzed. Thea l'eaulu support the commea.t by the Union al Qiooemed Scientists in resrpooae to the . Commissiooer'11 ,:equeat Cur Q)DlrneDts Oil tll.ia mihject. 4. Wltefher NRC Should Require Substanfial fmpn111en1enrs in Safety that Go Beyond Thm,e Proposed in tfris Rulemaki~ Co.lilDlhsloner .Asselstine requeated comments on wheiher Jhe NRC should r~uire substallilal improW!Dlellts in safety wilb respect to station blackout, like those beiug accomplished in some other coUDtrie1, which can be .achieved at reasonable cost and which go beyond those proposed in tlli& rulemaking.

Comments-NRC received eight letters that included comments ao this subject. Five of these were from the nucleal' iDduatry, none of which felt that the approach to station blackout taken in EurOj)ean countriea sbo11ld be used to justify safety improvements that go beyond the proposed J 50.63. 'Ihe main j11t1tification .mr indwitry's arsument is t'hat foreign cowitries may have reason, for requiring adivities tbat differ from, or exceed, those in the U.S. For example, Washington Public Power Supply Systems (WPPSS) commented. "It is not apparent that the details of U.S. grid stabilities end onsite power reliabilities are substantially similar enough to those found abroad to warrant a simple adoption of t'hese {European]

measures." In another comment from industry on this subject, NUMARC 1tated that tbere are several reasons why ma11y of the features for cupm,g wilh a statio11 blackout in new French nuclear power plants may already exist at most U.S.

Free single copies may be obtaaoed £ram the Division of Information Support Services.

UA NudeerReg,,1-,,Co!fflffleslon. Wmiington. 0C ZD551i. planls. 1n fact, t'hey said, "111e Fnmch. approadt to station blackout dOe11 not appear to depart significantty from current regulatory approaches in the U.S." Similarly, AIF stated, "The asserliom of extensive stalion blackout coping capability at foreign {notably European) nudelll' power planhs are not sufficiently substantiated to serve as even part of tile basis for the proposed req uire1m:11ts." '11!ree other letten {Union of Coocemed. Scientists, OCRE, and Illinon Department of Nuclear Saf-ety} supported the NRC rulemaking to require all plants to be ablt! to a,pe with a statkm bladcout, but urged the CA,,nmistrioo to 80 beyond the proposed rule. The Illinffl! Depertment of Nuclear Safety st1!.1ed that: The goal of holdins the expected frequency af core damage from station blackout to 10-5 per reactor-yell!' ill not snffitient\y striIJsenl With re111tnr~ fflOMfit momficlltiomr~ the pn,paed ni1e, a &eqa,mcy of 10-~ a,,peeni achienbe at reaammble mat. Specifically, tht! rule slmuld n,quin no lea dt1111 20 hours decay heat removal C11pacity imteed of only four or eigmt boars .in the pnlpOll8d nde. in the event of a blackout. lle.spoose-The staff agrees with industry'* romment1 that foreign countries may have valid reasons fur imposing requirements that differ from or exceed those in the U.S. For example, it appears that there is a highB fmqwmcy of losses of off site power in Frence than in the U.S. Thhl experience. along with Frend. safety objectives, led the French to deaip their new standard nuclear power plants to be able to COjle wi1h a very long duration. station blackoat {i.e., up to three days). "l1te Frem:h safety appmacli and their station blackuat deliga featme* are documented in NUREG--1208, "Analysis of Frendi {Paloel) Pressurized Wat-er Reador Desip Differences Compared to Current U.S. PWlt Dellisns," June 1988. The Commillllion beli1m!1I that the staff has adequately considered foreign approaches in preventing core melt from station blackout in developing the resolution of USI A-44. Although tht: rule req,aire9 plsntw kl be able to cope with station blackout for a epeci& duration, that duration is not si,ecified in the rule. Guidance to detennine an acceptable duration is included. in Regulatory Guide 1.155. This guidance should appiy to ID08t plants, but if there were adeqwtte justification, diffe!'ent requirements {either more or less stringent than the regulatmy gui&!) could be applied to-wpecific plantll. The use of alternate ac 90Ul'Cetl provides a nreans to acllieve further ;ncremental decrease. in core melt frequency. 50-SC-44 5. The Need for Generic Rulemalring CollUlHNltM-Five letters from the nucle4u-industry commented that generic ndemaltiog w not nece*ary to resolve the station blackout isBUe. Their reinona &:r tlm isne wel'e as follows: A genmc ndeimik;ng ill imlpprupriate sim:e the hislDric ffllfflber of 9iles experiencing a loss af aU of&iR ia AJalL (Texu Utilities) The station blamcmt iane slurnld be hamiJedOR a~ buia and does II01 ue.d lo be resolved by j8IM!ric rulemalwJ&. Each plant l!aa IIAiqae prabawlity for* a lDSB-of-pltWer event based on trerwmission sy11!mn, location of plam. *and Ot111ite power systems. [Duquesne Light) The Commission med nut pnmre generic rulemekm@ in order ro l"eSO!ve II non--g1meric issue. 1n the proposed atation btackom role. the numbs plants of concern ill. ackno11riec:li,:d 111 be limited. (NUMARCI Stalilm blackoat la& bes foand nat to be a

generic issue. Station blackout risk is plant specific and. accordiB8 lo die ataffa anaiyaa, the propoul ~a ollft! expected lo result i.n modi!u:atiam at DD more lhan a few facilities.

if al a11y. Requiring an 1icensees to undertake exte11Bive analyse& under Im! provisionB of lhe proposed rules when only II mnail gn,op of planbl 1llll1 have a need for n!mt!diat aetiffll ts not appropriate. (AIF) /tes~The Commission believes that a rule is appropriate to ensure that station blackout is addressed at all nuclear power plants. The plant-specific feature!I that oontnoute to ri9k for statron bladroat {e.!., di~ g-eireratm configmation. probabilitJ of loss of offsite power) are considered by lhe staff in the at.atim blackout regulatoiy guide to determine au acceptable coping duration fo,r each plant. Even though not all sites haw experienced a loss of offsite power, there ia not 91,ffici.ent aasurance that 1111ch eYent. would not oocur in the Iutnre.: Since historic experience has shown chat a total loss of offsite poser occun about once every 10 site-yeara. and many am:lear plants have operated fur less than 10 years, it is not surprismg that some plants have experiera:ied a loss of offaite power while others have not. Even though it is likely that many plants will not need hardwe.re modifications m comply with the role, the as11essnnmt of 5lation blackout copmg capam1ity fur II specim: duration and implementation m a"SSociated procedures will affect a Hfety benefit for all plants. 'I1nl "limited number of plants of com:em* in NUMARC's tetter refeni to those plant, havin,g the highest risk from statiml blaclrout {i.e~ those that would need ha,-dware modificatioml). Without a plant-specific assessment, these plant! cari not be PART 50

STATEMENTS OF CONSIDERATION identified.-

Even excluding these plants from consideration, the staffs analysis

has shown that the improvements in safety associated with the rule are consistent with backfit considerations set forth in §. 50.109. 6. Applicability of the Proposed§ 50.63 to Specific Plants Comments-Four letters included comments or questions regarding the applicability of the rule to specific plants. For example, does the rule apply to high-temperature gas-cooled reactors (HTGR) (i.e., Fort St. Vrain)? What about TMl-2 or plants that are near completion but will not have an operating license prior to the amendment's effective date? Houston Power and Lighting Company wrote: Proposed Section 50.63 provides schedular guidance for implementing station related modifications on plants that already hold operating licensees or will be licensed to oper,ite prior to the effective date of the amendment.

Plants who may be NTOL's [near-term operating license) but will not be licensed prior to the amendment*s effective date should be accorded the same compliance period under parts (c) and (d) of this section. Otherwise this proposed rule could be interpreted to imply that plants not licensed prior to the effective amendment date must comply with the rule and make all necessary modifications prior to receiving an O.L [operating license). The rule should be amended to address plants which are scheduled to receive an O.L. within a short time following implementation of this rule. Response-Rather than identifying specific plants for which the rule does not apply, § 50.63(a) specifies when it does apply (i.e., "each cooled nuclear power plant licensed to operate"). Since Fort St. Vrain is an HTGR, the generic rule would not apply. Station blackout will be considered individually for that plant based on its unique design. Since TMl-2 is not licensed to operate, likewise the rule would not apply to that plant. Any plant licensed to operate after the date the rule becomes effective will comply with the same 270-day schedule for information submittal applied to plants

previously licensed.

This affords NTOLs the same compliance features as plants already licensed to operate. 7. Plant-Specific Features and Capabilities Comments-A number of utilities described plant-specific features and capabilities that reduced the risk posed . by a station blackout event compared to the staffs analysis. Examples of such features are given below.

Availability of alternate, independent ac power sources such as diesel generafors; gas turbines, or nearby "black start" ac power sources.
Extremely reliable offsite power supplies because of multiple ways or underground feeders to back up above ground transmission lines.
Dedicated shutdown systems and associated diesel generators to meet the fire protection requirements of Appendix R to 10 CFR Part 50.
Common or shared systems between two units at multi-unit sites such as direct current (de) power, auxiliary feedwater, or diesel generators.

Response-The analyses performed for USI A-44 clearly show that specific features do affect the risk from station blackout, and the station blackout regulatory guide takes this into account in providing guidance on different acceptable coping durations depending on the most significant of these features. Those plants with extremely reliable offsite and onsite ac power supplies need only have a very short (e.g., 2-hour) coping duration to be acceptable. Plants that have a dedicated shutdown system with its own independent power supply* could take credit for this system to cope with a station blackout. The final rule and Regulatory Guide 1.155 have been clarified to give credit for alternate ac power supplies (see response to subject 11). Therefore, the Commission believes that for almost all sites, plant-specific differences have been adequately accounted for in the resolution of USI A-44, but the door is open to licensees who believe their plants have additional capability that should be considered by. the staff in demonstrating compliance with the rule. 8. The Source Term Used To Estimate Consequences Comments-NUMARC and others in the industry commented that the consequences of offsite releases that would result from a station blackout event are overestimated, and new source term information would lead to the prediction of much lower consequences for this event. Several commenterli felt that the approach taken by the staff to estimate consequences of a station blackout event was decreasing by a factor of three the estimated consequences of the siting source term (SSTt) from NUREG/CR-2723, "Estimates of the Financial Consequences of Nuclear Power Reactor Accidents" (September 1982). AIF felt that "implementation of any requirements resulting from the resolution of USI A-44 should be deferred until the results of the source 50-SC-45 term research can be taken into account." They based this statement on the premise that If the consequences used. in the staffs value-impact analysis were reduced by a factor of 10, none of the alternatives would be feasible. The Union of Concerned Scientists expressed a different point of view in their letter which said "* *

available evidence indicates that the consequences of an accident involving station blackout may be even worse than those estimated either in WASH-1400 or the NRC's more recent studies." Response-NRC has had an extensive research effort underway since about 1981 to evaluate severe accident source terms. The staff has reviewed the results of this research to take into account the public comments received on this subject. Since there is still a great deal of uncertainty regarding source terms and associated consequences, the staff revised .its value-impact analysis for USI A-44 considering a range of estimates for consequences of a station blackout.

The NRC research on severe accident source terms has resulted in the development of significant new. analytical tools by NRC contractors, as discussed in NUREG-0956, "Reassessment of the Technical Bases for Estimating Source Terms," July 1986. The analytical methods developed, generally referred to as the Source Term Code Package (STCP), have been used to analyze a number of severe accident sequences for five reference plants, namely: Peach Bottom, a BWR Mark I design; Sequoyah, a PWR ice condenser; Surry, a PWR with a sub-atmospheric containment; Grand Gulf, a BWR with a Mark III containment; and Zion, a PWR with a large dry containment (NUREG-1150, "Reactor Risk Reference Document," Draft for Comment, February 1987). The results of these analyses show that releases from station blackout sequences can be expected to vary significantly depending upon the plant and the specific sequence. Although generalizations are difficult, it appears that calculations using the STCP yield release fractions for most of the sequences range from about one third of an SSTt release (for the case of Surry, without condensation) to roughly one order of magnitude less than this. However, the uncertainties in our present understanding also do not preclude the possibility of a large release, approaching that of the SSTt estimate. To determine the consequences m terms of person-rem, given the above range of release fractions, data taken from NUREG/CR-2723 indicate that the PART 50

STATEMENTS OF CONSIDERATION variations HI peniOIH!etll associated
, wit!. reJeasea Qf mqrutacie SST1. SST2 and ssn are virtually identical to the variations in .latent t:ancer fatalities for the same three releases.

Hence, the estimated climJ,81! in iatent caocer fatalilies with releaae &actions provides a reliable mdicatwn of change in pec.aoo-rem a, we11. 'Table 10 in NUREX;/CR-D.23 presents variationa i11 eatimated latent canGel' fatalities .associated with clmnges i.n ssn release fractilJDS {fur all elements except aoble gasea). Thi9 table shows that a release fractioo of one-third of an SSTl release would yield a value of about 50 psrcent of the lat.ent cancer f.atallties {and persoo-rem) of an SSTl release. Similarly, a release fracliOll of one-third of an SST1 release would yield an estimated person-rem of about 15 percent of that associated with an SST1 release. Conseqmmtly, for value-impact calculations, the staff estimated the range of coni.equences of atation blackout, in tenm of person-rem, to be from 0.1.5 to 0.5 of the estimated rem of an SST1. release. As noted, the original value-impact analysis was based on 1J.3 times the estimated rem of an SST1 release. With regard to a possible delay in the re11Dlution of USI A-44 until "bettm"' source terms become available, key considerations appear to be when better soorce t-enns are l;kdy to become avmTuhh! and to what degree uncertain'tie in phenomenology as well as di~ces between invffligators will be resolTed. Althuugh Te9eal'ch on source tenmi m expected to continue weB intn the future. improvemenbl in our knowledge are expected m be tBI3ely e'ffllutionmy be,,md thi9 point, in that iliemejffl'phenomenaappear~h~e been accountied for, at lttst in a order fashien, botn in NRC 11.s well as industry models. Resolution and narrowing of lite remaining uncertainties would al.90 berie6t &om improved experiment& and analytieal model9 that are libly k> become UT.!ilable gradually. For thue rea&Dl\8, 'lisnfficantly better soan:e terms than thoae presentty availahle are likelJ to be forthooming only after a number of yean. Since the range of. aeYen! accident sout"Ce tel'ms and oomequences

usgeirted ah<Ml from estimating .tatiDll btaclcout sequeaoes is sufficiendy broad ti, arvel' likely improvements in source tenn knowledge.

the re!!Olutioo ofUSI A-44 should not be ~ayed. 9. Specificity on the Extent of Required Copil'/6 Studies Cam~ral letteni by ind\lStry expre55ed concern that the studies nece&sarJ to demonstrate that a plant CBll cope with a station blackout . are not welt defiDed and CIJllld potentially be unbounded. These commenta focused on twq mam.

points. First. the propoaed role NqDil'ed plants to determine the Dll!Xinymt dul'ation the plmt cmdd cope with a station blackout, yet the draft regulatmy guide indaded specific gu.idance on acceptable ooping durations

{e.g., 4 or II hoaraJ. Deterrnimng the maximum duration, rather than ~ng the plan1:"s capability fur a specific aa:epmt:M duration, could be an open-ended -* requirement Alon,g these linel'I, NUMARC stated: Un1esa the required coping demcnstration is ~ecifically bounded by clearly stated definitiOI111, BSsumptions, and criteria, there could conceivably be a~ of supporting special effect* analyseB whicl\ lu:eosees may have to comider aa a resi:il: of. the exercise of discretion by mdmdiaal staff re1fiewem. Under the rule .aa prapased. licenlll!ea cannot ascertam the ultimate l'equiremeDta they will be expected to meet (i11cluding lhe potential plant modificaUons they v.111 need to make) to demonstrate campaanoe. Second, industry al.ao commented on the potential open-endedness of analyses to determi.De the operability of equipment in.environmental conditions resulti!J8 from a station blackout (e.g., without heating. ventilation. and air conditioning}. Unless these analyset1 were well defined, industry felt ilie analyses coo1d be much more costly than estimated by the staff. Howevar, NUMARC made the following statement relating to tlie need for detailed prNCriptive requirements by NRC that appeani to conJradict their earlier statement. The 1JOHll * *

ftl IIOt ih11t reguhttiuM fflUllt be prescriptive by their very niltmle, Prescriptive reptiODS, which outliqe in detail exactl:y what steps are required by licensees to satisfy a proposed regulation.

are, iD many instaucea, -eaary and counterproductill'I!, Jlespo.rne-With resard Ca the proposed requiranent that each plant determine its maximum. dmation for coping with statian 1:iackoat, the staff agrees wiili 1he mdmtry oon:arneom. First of an. it would be difficidt ta adequately defim! '"maximam duration'" in tm. ae:me. Second. if m:ensees determine that their plan-ta en. cope with a station blackout fur a specif:ioo duration and restore ac pawer through an acceptable coping BDB.lysill, the additional safety benefit gained from simplJ the kDowledge that a Janger, ar "maximum duration," copmg dm-atian exists is small Third. the coahl far auesllUlg "maximum dtll'Btioa" will be higher since mme extensive analysea "MR be required to anBlyze a transient which 50-SC-46 woukl go beyond the coping analysis for a specified dtR'Btion and recowry from station btackout. Therefore, the rule and regulatory guicie have been revitted aooordingly to dell!ffl the requirement for licensees to determine a plant's maximum wping capability. With regard to the comments OD assessments to determine equipment . operaln1ity during a station blackout, the staff feeh strongly that .such assesmwrrt11 are 02cessary tu determine a plant's respome to station blaclwuL By delelini the requirement to determine a plant's "maxinmm" coping capability, the assessment of equipment operability would not be as costly as assumed by industry. Goidance on acceptable copin,g assessments is provided in the station blackout regulatory guide. Also, guidelines to evaluate the effects of loss of ventilation under station blackout cooditions are provided in Appendix E of NUMARC-a?'OO. "Guidelines and Technical Bases fur NUMARC Initiatirea Addressing Station Blackout at Light Water Reactoni. N The. efforts provide adrlitionai definitions, criteria, and utandarda for lk:ensees' asse9!1ment9 of eq\riplll1!nt operability without the med for .. prescriptive resn}ations" by NRC. In order to further ewlnat~ indwltry's comments on thia sohject. NRC requested Sandia National Laboratories to identify specifu: tasks neceesmy to det-ennkm c,perability of equipment duri.og a station blackout and to estimate the cost to perfann these tB11ks. Resol.ta oI thia 11tudy were ased in the revised value-impact analysis performed for this issue {"Equipment Operability Durillg Smti.oB Blackout E,vent," NUREG/ CR--41H2).

10. Ar:ceptable Durotioo.

for Coping with a Station Blackoul Cl1mlllBlts-Severai comments with differing YieWB were directed at guidautz in the dmft regulatury guide on acceptable station blackout coping durations in order for plants ta comply with the proposed mle.. W&11bmgton Public Power Supply commented that "it should be possible for certain uti1itie11 lo demonstrate {an acceptable) zero hour blackout." One individual recommended "that a 30 minute period be a margin, and that no duration under 4 hours be accepted by the staff.* NucleDyne Engineerins commented thut '"advanced reaetors should require the capability to sal'-ely withstand a 11tation blackout of at least 8 hours," and the IHinois Department of Nuclear Safety wrote that "the rult! should require no le1111 than 20 hours PART 50

STATEMENTS OF CONSIDERATION decay heat removal capability instead of only 4 or 8 houri." Response-Although diverse commenta were received on this subject, none provided supporting analysis or information to back up the opiniona expressed.

However, the staff did .reanalyze the estimated riak from station blackout evems fur different plant-and Bite-related characteriatk:s and .revised ita guidaooe on acceptable coping daratioe aa:ordingly based on a goal of .limitiDg the averaae cootrib111:ian to core damqe from &talion blackout to about 10*5 pl!!' reactor-year. Most planta would still Deed a 4-or 8-hoUI" COp!Il8 capability. Tlwee few plants with the most redmdant ontlite emergency ec power system, coincident with significantly lower than average expected frequeric:, of loss of offsite power, would need only a 2-hour capability to be acceptable. Any plant with minimum redundancy in the onsite emergency ac power system coincident with low reliability and a significantly higher than average expected frequency of loss of offsite power would need to substantially improYe its ac power reliability or be able tu cope with a station blackout for 16 hows. 11. Credit for Alternate or Diverse AC Power Smm:es Comments-Ten letters from the utility induatry commented that more credit should be allowed fOl" the availability of alternate power sources such as onsite gas turbines. The comment.a below represent the utilities' viewpoint. The station blm:kout rule should be clarified la allow aoedit for diverse and very reliable offsite power tlOllft:811 or d:ivene lllld very reliable ollllite electrical generation. (Public Service Company of Colorado) 'Mte option of providing an additional alternate source of ac power is eliminated by [the propoaed reeolution). The inconsistency in thi11 approach en best be underatood by considerins an example al a generic nuclear power 11tation. [Toledo Edison) If the li<;easee wera to provide an additional independence dieeel generalur capable of providing the necesaary ac power to prevent station blackout, the licensee * * . would still be required to withstand at least 4 hours without ac power. They wotdd receive no credit far the additional diesel generator in the copins analyllia. If the licensee -re to use that same dieael engine to power a charging pump. even thol.lJl!h it would bl! of lee, significance to mitigatioa al reactor core damage than the diesel senerator.. the licenaee could take credit for it in coping with the blackout. [Toledo Edison) Since a diesel-powered clmrsmB pump wilt not provide for eq8ip111ent loading flexibility. ligbtine, ventilation. in1trumentation, etc., it Is obvionly of lower value than an additional source of 11c power. The fixed category approach lalcen in {the proposed resolution). however, will not permit taking credit for the same diesel engine when used as a generator t1iough the actual reliability for the machine is the same. (Toledo Edison) Response-The proposed regulation did not in1end to isnore the alternative of addi~ additional power sources or takiQg credit for auch sources if they already exist. For example, as specified in the regulalo.ry guide. if a licensee 11dded an emergency diesel generator to one of its plant, that had minimum redundancy ill lhe onsite emergency ac power system. the acceptable station blackout coping dUl"ation could be reduced. For some plants, however, addins a diesel generator wou1d not result in a .reduction in the aa:eptable coping duration, and !he point made by Toledo Edison is a valid one. The rule and rll8ulatory guide have beeo revised to clarify that alternate ac power sources are gi.ven credit to cope with a station blackout provided that certain criteria Sl"e met le,g., independence, redundancy. high reliability, maintenance, and testing).

12. Trends 011 the Reliability of AC PmverSoarces Comments-rive letters included comments on the reliability of ac power sources. Four letters from industry felt that improved ac power reliability should be factored into the staff's technical analysis.

Examples of these comment.a include the foUowing:

*
the feqnem:y of Ion af off site power act.iritiea hu been decreasins
* * (Waabington .Public PDWl!II' Supply System); * *
offsile power availability in the abseRce of regulallon has significantly improved lffer lhe past decade. (Southern Califomla Bdillon CumpanJ);

{NUREG/CR-4341) ... *

1100W11 an
improvement in dieael generator reliability over that shown in the earlier document [NUREG/CR-29B9J (General Electric);

and Typically the reliability of ollllite power systems increases during the fir.t few yeara followins startup. (Gulf States Ulilitiell) '.I'he 1Uinoi1 Department of Nuclear Safety, on the other bend, felt that potential vulnerabilities etift exist in onsite emergency ac power systems, and licensees should demonstrate that they haw taken steps to reduce the probability of loa of ac power. Response-The staff and its contractors have extensively analyzed the industry experience and trends in ec power reliability as documented In NUREG-1032, NUREG/CR-2989, NUREG/CK-.3992, and ts41. Trend, have shown that two aspects of ac powei-reliability have improved aomewhat-the reduced frequency of losses of offsite power due 50-SC-47 to pbmt-centered tMmts, and 1l &tight improvement in aTerage diesel generator reliability from UJ78 throngh 1983. These factors have been taken into account in the stafrs analyses and the resolution of USI A-44.. However, data alao demooatrate lhat there are practical limits on ac power reliability. and die defense-in-aepth approach of being able to cope with a at.awm *Dlackuat is WIU'l'anted.

13. Sharing of Emetgency Diesel Generaton, UnittJ at Multi-Vnit Sita Cammeal.s-8everal letters from industry stated that same plants with two unite on a site have the capability to crosstie electrical buae1 between units and therefore haYe improved flexibility in providing ac power.. Smee the magnitude af the electrical loadi necessary to provide core cooling during a station blackout is significantly less than that required for a de&ip baais accident, it could be possible to provide ac power to both um.ts at the site using only a single diesel generator.

Response-The proposed rule and draft regulatory guide do not prohibit the approach discussed above.. U licensees can demonstrate that sucll cross tie capability exists. procedw-es are in place to accampliah the crosslie and shed nonessential loads (if necessary), and no NRC regulations are violated {such as separation, minimam redundancy,andindependence),then credit wouid be given fur this capability as shown in Regulatory Guide L155 (i.e., reduced acceptable station blackout coping duratitJrn1 for greater diesel generatorrednndancyi

14. CJarifi:oJJtian of the DefinitiDDB -Of Station Blackout and Diesel Generator Failures Comments-{A)

Three cammenters from the utility industry recommended that the definition of station blackout in § 50.2 should be clarified to exclude ac power from the station batteries through inverters.. This wun:e of ac power from the station batteries would be available in the event of a loss of both the offsite and onsite emef8enCY ac power 110urces (i.e., diesel generatorsJ. (B) Several from industry commented that the definition of diesel generator 'failnre 11h0111d be clarified, particularly with respect to the treatment of term failurets that can be recovered quickly. Sargeat and Lundy Engineers commented that A definition of failure an demand for emergency diesel generators needs to be provided. Umier the context of a station blackout. a die11el generator winch fails to PART 50

STATEMENTS OF CONSIDERATION start automatically upon de!eclion of an offsite power loss, but is successfully started manually from the main control room or from the local control panel, should riot be
considered a failure on demand. Response-{A)

The staff agrees with comment A and revised the definition of station blackout accordingly. (B) Based on actual experience, failures of diesel generators to start due to failures in the auto-start system make up less than 20 percent of all diesel generator failures. Therefore, discounting these failures would not have a significant impact on overall diesel generator reliability statistics. However, the staff agrees in principle with comment B and has clarified the station blackout regulatory guide so that auto-start failures of diesel generators need not be counted in determining the failure rate if the diesel generator is capable of being started manually immediately*after it does not start automatically.

15. Specificity and Clarification of Requirements

Comments-Public comments were received regarding the specificity and clarification of the proposed rule and draft regulatory guide. These ranged from general to specific comments as the following two excerpts indicate:

We are concerned that, it the proposed rule is adopted, the staff will promulgate regulatory guidance criteria which will be unrealistic and excessive, i.e., compounding the event with other accidents, imposing passive failure criteria, applying seismic, environmental qualification and.other qualifica lions to equipment that could otherwise be used in response to such an event, etc. {Maine.Yankee Atomic Power Company) Definitions of Pl and P2 (in Table 3 of the draft Regulatory Guide) use frequency of extremely severe weather and severe weather interchangeably, thus creating confusion In the definition. (Washington Public Supply System) Response-Some of the comments on this subject relate to other subjects discussed elsewhere in this section. Some comments were quite specific while others were general in nature or expressed views that were not substantiated with backup material. The staff has taken these comments into consideration and revised and clarified the rule and regulatory guide accordingly. Additional guidance is provided in NUMARC-8700 which has been reviewed by the staff and referenced in the regulatory guide as providing a method the staff finds acceptable for meeting the rule. 18. Technical Comments on NUREG-. 1032 Comments-In addition to comments on the proposed rule end draft regulatory guide, se..veral letters contained comments on the stafrs draft technical report, NUREG-1032, "Evaluation of Station Blackout Accidents at Nuclear Power Plants." Response-NUREG-1032 was issued in draft form for public comment in May 1985 (50 FR 24332). The comments received were reviewed and considered by the staff and resulted in a evaluation of the technical analysis. De.tails of the specific comments and responses are not presented here. Rather, NUREG-1032 was revised extensively over the past year to address the public comments. In general, the overall conclusions on the risk from station blackout events did not change significantly as a result of the reanalysis. One of the major changes resulting from the reanalysis was a revision to the definitions of plant characteristics, especially the clustering of plants into site and weather-related groups (Appendix A in NUREG-1032). These changes are reflected in revisions to the guidance in the station blackout regulatory guide to determine specific acceptable station blackout coping durations.

17. Relationship of US/ A-44 to Other NRC Generic Issues Comments-The major public comment regarding the relationship of USI A-44 to other NRC generic safety issues was that the proposed rule may not be necessary or should be postponed because of ongoing work to resolve related ,generic issues. Some comments were general in nature such as the following one from Southern California Edison Company: Promulgation of a rmal station blackout rulemaking at this time will unnecessarily complicate the final resolution of related generic technical issue * * *. The NRC must develop end implement a program to coordinate the resolution of all power-related generic issues prior to finalizing any individual proposed rule. AIF suggested that the implementation of any requirements for station blackout be deferred until the requirements from USI A-45, Shutdown Decay Heat Removal Requirements, are known and until the effect of source term changes can be evaluated.

NUMARC mentioned specific proposed and existing regulatory requirements that should be considered because they could reduce the need for a station blackout rule (e.g., B-56, Diesel Generator Reliability, and GI 23, Reactor 50-SC-48 Coolant Pump Seal Failures). Other related issues mentioned in the public comments were A-30, Adequacy of Safety-Related DC Power Supplies, and implementation of safe shutdown facilities to meet the fire protection requirements of Appendix R. Response-The question that needs to be addressed is "should a requirement be imposed now to reduce risk, or should it be postponed until related issues are resolved sometime in the future?" Potentially, this could result in substantial delays, thereby not resolving generic safety Issues in a timely manner. The staff has considered the resolution of USI A-44 in light of the related issues mentioned in the comments. Although these issues are identified es separate tasks within NRC, they are all managed in a well established program that coordinates all related issues. A brief discussion of the most relevant issues is presented below. (Additional information is provided in NUREG-1109, "Regulatory*Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout.") Resolution of USI A-45 will occur at some time following issuance of the station blackout rule(§ 50.63) and after plant-specific station blackout coping evaluations have been performed by licensees per NUMARC/NUGSBO Initiative 5, utilizing guidelines provided in NUMARC-8700. Further, the resolution of USI A-45 is expected to be highly plant-specific and focused on loss of decay heat removal considerations from other causes beyond station blackout. Utilization will be made of A-44 evaluations (as applicable) and any *plant equipment modification needs identified from A-45 will be carefully evaluated to maximize effective use of previously identified A-44 equipment needs. Maintaining emergency diesel generator reliability, the purpose of B-56, iii an integral part of the resolution of USI A-44. However, the Commission believes that additional depth will achieve a substantial Increase in protection to public health and safety. The resolution of GI 23 (reactor coolant pump seal leakage) deals with loss of reactor coolant system inventory and associated degraded core conditions. USI A-44 deals with station blackout induced effects, which result in loss of ac power, thereby impacting a broader spectrum of plant equipment end safety-related functions. Although the resolution of GI 23 will contribute to establishing a higher level of assurance that seal leakage will be minimized (thereby minimizing the need for power PART 50

STATEMENTS OF CONSIDERATION to repiace water inventory losses over the station blackout duraooo and recovery phase), resolution of GI 23 by itself will not 11ddreas the broader scope of USI A-44 safety cancem&. Some licenseea have implemented dedicated shutdown systems that are indepeodem of normal and emergency ac power to meet Appendix R requirements.

ff applicable, these features would be credited in the reeolution of USI A-44 by providing the capability to cope with a staticn blackout. Thus, the resolution of USI A-44 is coordinated with related generic issues, and implementation of a final resolution should not be delayed further. {Response to comments on the effect of . source term chaoses is included in subject uumber 8.) 18. An Alternative of PlanJ-Specific Probabilisuc Asseuments Comments-Several utilities suggested that, in lieu of the . requirements in the rule, licensees should be permitted to submit specific evaluationa to demonstrate that the frequeocy of COl'e damage from station blackout eveata ia 10-5 per reactor-year or len. In II similar vein, the suggestion was made that NRC should specify a target te'l'el of reliability for ac pcrwer systems in CJ1'der to satisfy NRC's criteria for core damase frequency. A few licensee. submitted limited probabilistic assessments to show that for some plants station blackout cowd have .a vecy small probability of seven consequences. Response-The Commission does not preclude licensees from submitting . pl.ant-specific probabimitic asS'l!Ssments to support a determination that'station blad:out would have a very small probability for causing core damage. However, the requlrements of the role must be met. The Commi.ssioa would olwerve that the use of probabilistic assessments was important a1 iDpUt to the regulatory decisirmmaking that culminated m tne station bladtout rule and related guidance.* As expressed in the Commission's Safety Goal Policy statement of Ausust 1986 (51 FR 2.8044), the Commission has acquired 11 reasonable degree of confidence about the usefulness and value of probabilistic assessments in assisting regulatory decisionmaking Oil complex safety *issues.In short, such assessments are of value in complementi!J8 aod focusing the more traditional and deterministic defense-in-depth appro11ches. On the other hand, any licensee mwit decide whether or not its plant-specific ac power configuration and other related equipment are sufficiently unique to merit the commct ami submittal of a probabilistic IUIBes&ment U part of. achievq m.mpoance of § 50.63. The Commiuklo's experience also indicat.es that probBbilistic asse11S111ents are 1"l:9UWQ!: intensm! .mt can be of marginal utility if their only end result is to delay nle compliance-

19. Procedures and Operator AcliOM During Stalion Blaakoul Commeata--{A)

Several letters from industry co.iunadl:d that. in response lo Generic Lettm 81,-0t, '"Emetl!dilCJ Procedures and 'li'ainiog for Station Blackout Ewllbl." dated February 21, 1981. utilities already hav4! procedures in place to prepare plant operations for station blackout events. Owners' groups have established generic guidance for station blackout opet"llting procedures for licensees to 11t1e in developing specific proceduree. A represent!! tiYe of the Profemonal Reactor Operator Society c,:,mmemed tht\t Generic procednres are med by most operating faci!itie&. These procedures are not carried into adequate deptb ofspecific power plant operations. The industry has relied too heavily on eeireric procedures and has not given a real look at what specific steps must be laken. Extn!polation of these procedures mU.11 be required. Speaf1C maimemmce procealll'e& must.be establiahed ud foilowed. (B) Otller comments an procedures related tu the timeliness of operator actiam, both inside and outside the control.room. Homton Lighting and Power sagsested that: In Section 3.1 (Part 6) tof the regulatoiy guide1, 1ffll first ~ente!ICt! -mould be revised lo read, 'Cmaaidemtiaa should be given to timely operator actions both inside and outside of lite CUlllml n,on, that * * * ,' eo that indil can be &8ke& b exiatilQI equi.jameat that may not hne acbultiaD BDd caotrol frWII tile c;oatrol .l'OOIII. Illinois Power Compa.i,:y remmmended that: * * *

Section C.:1.3, Item 3.a, of !he proposed regulatory guide llhould be modified lo t"e8d: a. The eydem should be <:apable of beiniJ actuated aDd Olllltrulled from the coalNI room, ar ii other meam of cmtlml are reqwred (e.g., 1D211Wal jumpinJ of conm,l logics or manual operation of valftll}.

it should be demonsttated that these lltep!I GaD be carried out in a timely fashion. Respoos~A} Licemees mar take credit for station blackout procedures alreadJ iD pace to comply with the station blackout rule. However, for the most part, these procedures were developed without having the benefit of a plant-specific asaeasment to detennine whether a plant comd withstand a station blackout for a specific duration. Therefore. these procedures may need to 50-SC-49 be modified after lieensees have determined a11 acceptable station blackout copins duration and evaluated their plant's response to a station blaclcout oI this duration. (B)The staff agrees with the comments related to operatar ac1iot111 outsid!! fire control room. and the regulatory guide was revised accordiugly_

20. Schedule 'Provisions in Proposed §50.63 Commem:s-Two lett.ers contained commentll on the proposed schedule in I 50.63. OCRE fult the achedulins provisions in the propo1ed nde were far too generous.

One iruijvkh1al ' recommended that the schedule be modified to require licensees lo mbmit, within 9 months of the date of the amendment, a liat of modifications along with a proposed schedule to implement those modificaUona. {According lo the propoaed. rule, licensees would not haw to submit a achedule for implementing equipment modifications until aftff the staff received and reviewed lit:enees* submittals on their planfs acceptable station blackout duration.) Resp01f8E'-The trtaff agreed in part with these comments, and the schedule was revised accordingly. Section 50.63(c)(1J(iii) now requires that Hcemrees !l'llbmit within 9 montrnl after the role is nisued a list af equipment modificati1ms and a proposed scbednle for implementing tht!m. A final schedule fflfflld be deTeiop1!d after NRC ha11 reviewed the licensee11' submittal of their plant's acceptable station blackout duration.

-21. Industry mitiativs Comments-In addition to comments 011 the proposed rule, NUMARC endorsed the following five initiallves 5 to address the more important . contributions to station blackout:

1. Each atilitJ wiH review their site(s) 118-1iaat the criteria specified in NURF.G-1108, aad ii the llite{a) fail int11 the ca tesory of an eight-hour site after utilizing aH power sources available, the utility will take actions to reduce tlie eite(e) omrtribation to tbe overall risk of sta lion blackout. hardware changes will be made within one y!lal'. Hardware changes will be made within a reasonable time thereafter.

NUMAllC initlall, p!UflDHd a set of fmr initiatives.

The fitli illi1lali*e reaar&Ds tlae performance of a copins assessment WBll provided in NUM.a.~'i'OII. which W1IS submitted by letter from J. Opelca (NUMARCl III T. Spei1 (RES! dated November 23. 1111'7, A OOH ia available far pablic iniiptct.ion und copyias for a fae at tbe NRC i';.blic Document Room at 1717 H Street NW., W.ashlng1on. DC. PART 50

STATEMENTS OF CONSIDERATION

2. Each utility will implement procedures et each of its site(s) for: a. Coping with a station blackout event, b. Restoration of ac power following a station blackout event, end c. Preparing the plant for severe weather conditions (e.g., hurricanes and tornadoes) to reduce the likelihood end consequences of a loss of offsite power end to reduce the overall risk of a station blackout event. 3. Each utility will, if applicable, reduce or eliminate cold fast-starts of emergency diesel genera tors 'for testing through changes to technical specifications or other appropriate means. 4. Each utility will monitor emergency ec power unavailability utilizing data utilities provide to INPO on a regular basis. 5. Each utility will assess the ability of its plant(s) to cope with a station blackout.

Plants utilizing alternate ec.power for station blackout response which can be shown by test to be available to power the shutdown busses within 10 minutes of the onset of station blackout do not need to perform any coping assessment. Remaining alternate ac plants will assess their ability to cope for 1 hour. Plants not utilizing an alternate ac source will assess their ability to cope for 4 hours. Factors identified which prevent demonstrating the capability to cope for the appropriate duration will be addressed through hardware and/or procedural changes so that successful demonstration is possible. NUMARC previously opposed generic rulemaking and felt that the first four initiatives would resolve the station blackout issue. Response-These five initiatives now include many of the elements that are included in the NRC resolution of USI A-44. The staff has followed up on the NUMARC initiatives through a series of meetings in 1986 through 1987. The result has been the development of NUMARC-8700 which provides guidelines and criteria acceptable to the staff. The procedures in NUMARC'riJ700 have been referenced in Regulatory Guide 1.155 as providing guidance acceptable to the staff for meeting the requirements of the rule. Table 1 in Regulatory Guide 1.155 provides a cross-reference to NUMARC-8700 and notes where the regulatory guide takes precedence. NUMARC's previous concerns have been addressed in the development of Regulatory Guide 1.155 and NUMARC'riJ700. Finding ~f No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's rules in Subpart A of 10 CFR Part 51, that this rule is not a major Federal action significantly affecting the quality of the human environment, and therefore, an environmental impact statement Is not required. There are not any adverse environmental impacts as a result of the rule because there is no additional radiological exposure to the general public or plant employees, and plant shutdown is not required so there are no additional environmental impacts as a result of the need for replacement power. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection and copying for a fee at the NRC Public Document Room, 1717 H Street NW., Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from Mr. Warren Minner&, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301) 492-782~. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Management and Budget approval number 315CHJ011. Regulatory Analysis The Commission has prepared a regulatory analysis on this final regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. A copy of the regulatory analysis, NUREG-1109, * "Regulatory/Backfit Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout," is available for inspection and copying for a fee at the NRC Public Document Room, 1717 H Street NW., Washington, DC 20555. Regulatory Flexibility Certification As required by the Re*gulatory Flexibility Act (5 U.S.C. 805(b)), the Commission certifies that this rule does not have a significant economic impact on a substantial number of small . entities. The rule requires that nuclear power plants be able to withstand a total loss of ac power for a specified time duration and maintain reactor core cooling during that period. These facilities are licensed under the provisions of§§ 50.21(b) and 50.22 ofto CFR Part 50. The companies that own these facilities do not fall within the scope of "small entities" as set forth in the Regulatory Flexibility Act or the small business size standards*set forth in regulations issued by the Small Business Administration in 13 CFR Part 121. List of Subjects in 10 CFR Part SO Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, 50-SC-50 Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. . For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is adopting the following amendments to 10 CFR Part 50. Backfit Analysis Analysis and Determination That the Rulemaking To Amend 10 CFR Port 50 Concerning Station Blackout Complies . With the Backfit Rule 10 CFR 50.109 The Commission's existing regulations establish requirements for the design and testing of onsite and offsite electrical power systems (10 CFR Part .50, Appendix A, General Design Criteria 17 and 18). However, as operating experience has accumulated, the concern has arisen regarding the reliability of both the offsite and onsite emergency ac power systems. These systems provide power for various safety systems, including reactor. core decay heat removal and containment heat removal. which are essential for preserving the integrity of the reactor core and the containment building, respectively. In numerous instances emergency*diesel generators have failed to start and run during tests conducted at operating plants. In addition, a number of operating plants have experienced a total lOSII of offsite electric power, and more such occurrences are expected. Existing regulations do not require explicitly that nuclear power plants be designed to withstand the loss of all ac power for any specified period.

This issue has been studied by the staff as part of Unresolved Safety Issue (USI) A-44, "Station Blackout." Both deterministic and probabilistic analysea were performed to determine the ti.mins and consequences of various accident sequences and to identify the dominant factors affecting the likelihood of core melt accidents from station blackout.

Although operational experience showa that the risk to public health and safety Is not undue, these studies, which have evaluated plant design features and 'dependent factors in detail, show blackout can be a significant contributor to the overall residual risk. Consequently, the Commission is amending Its regulations to require that plants be capable of withstanding a total loss of ac power for a specified duration and to maintain reactor core cooling during that period. An analysis of the benefits and costs of implementing the station blackout PART 50

STATEMENTS OF CONSIDERATION rule is presented in NUREG-1109, "Regulatory

/Baddit Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout. The estimated benefit from implementing the station blackout rule is a reduction in the frequency of core damage per year due to station blackout and the associated risk of offsite radioactive releases. The risk reduction for 100 operating reactors ii estimated to be 145,000 person-rem and supports the Commission's conclusion that I 50.63 provides a substantial improvement in the level of public health and safety protection. The cost for licensees to comply with the rule would vary depending on the existing capability of each plant to cope with a station blackout, as well as the specified station blackout duration for tl11tt plant. The costs would be primarily for licensees (1) to assess the plant'* capability to cope with a station blackout, (2) to develop procedures. (3) to improve diesel generator reliability if the reliability falls below certain levels, and (4) to retrofit plants with additional components or systems, as necessary, to

meet the requirements.

The estimated total cost for 100 operating reactors to comply with the resolution of USI A-44 is about $80 million. The average cost per reactor would be around $000,000, ranging from $350,000. if only a station blackout assessment and procedures and training are necessary, to a maximum of about $4 million if substantial modifications are needed, includins requalification of a diesel generator.

The overall value-impact ratio, not including accident avoidance costs, is about 2,400 person-rem averted per million dollars. If the net cost, which inchides the cost savings from accident avoidance (i.e., cleanup and repair of onsite damages and replacement power following an accident).

were used, the overall value-impact ratio*would improve significantly to about 8,100 person-rem averted per million dollars. These value-. which exceed the $1jm/ person-rem interim goidance provided by the Commi11Sion. support proceeding . with the implementation of§ 50.6.1.. The preceding quantitative impact analysis was one of the factors considered in evaluating the rule, but other factora also played a part in the decision-making proceu. Probabilistic risk assessment (PRA) studies performed for this USL a11 well as some plant-specific PRAa, have shown that station blackout can be a significant contributor to core melt frequency, and, with consideration of containment failure, station blackout events can represent an important contributor to reactor risk. In general, active systems required for containment heat removal are unavilable during station blackout. Therefore, the offsite risk is higher from a core melt resulting from a station blackout than it is from many other accident scenarios. . Although there are licensing* requirements and guidance directed at . providing reliable offsite end onsite ac

power, experience has shown that there are practical limitations in ensuring the reliability of offsite and onsite emergency ac power systems. Potential wlnerabilities to common cause failures associated with design, operational, and environmental factors can affect ac power system reliability.

For example, if potential common cause failures of emergency diesel generators exist (e.g., in service-water or de power support systems), then the estimated core . damage frequency from station blackout events can increase significantly. Also, even though recent data indicate that the average emergency diesel generator reliability has improved slightly since 1976, these data also show that diesel generator failure rates during unplanned demand (e.g., following a loss of offsite power) were higher than that during surveillance tests. The estimated frequency of core damage from station blackout events i11 directly proportional to the frequency of the initiating event. Estimates of station blackout frequencies for this USI were based on actual operational experience with credit given for trends showing a reduction in the frequency of losses of offsite power resulting from centered events. This is assumed to be a realistic indicator of future performance .. An argument can be made ~at the . * *

future performance will be better than the past. For example, when problems with the offsite power grid arise, they are fixed and, therefore, grid reliability should improve. On the other hand, grid power failures may become more frequent because fewer plants are being built, and more power is being transmitted among regions. thus placing greater stress on transmission lines. The factors discussed above support the determination that additional defense-in-depth provided by the ability of plant to cope with station blackout for a specific duration would provide substantial increase in the overall
protection of the public health and safety, and the direct and indire.ct costs .. of Implementation are justified in view of this increased protection.

Th~ Commission has considered how this backfit should be prioritized and scheduled in light of other regulatory activities ongoing at operating nuclear 50-SC-51 power plants. Station blackout*warrants a high proiority ranking based on both. its status as an "unresolved safety *

issue" and the results and conclusions reached in resolving this-issue. As noted in the implementation section of the rule (§ 50.63(c)(4)), the schedule for * . equipment modification (if needed to
meet the requirements of the rule) shall be established by the NRC s~aff in consulation and coordination with the licensee.

Modifications that CBIU\OI be *schedule for completion within two*

years after NRC accepts the licensee'.s.

specified station blackout durationmust* he justified by the licensee. The NRC : retains the authority to determine the schedules for modifications. In addition, some foreign countries, including France, Britian, Sweden, Germany and Belgium, have taken steps to reduce the risk from station blackout events. These steps include adding

design features to enhance the capability of the plant to cope with a station blackout for a substantial period of time and/or adding.redundant and . diverse emergency ac power sources. , Analysis of §50.109(c)

Factor!! 1. Statement of the specific objectives that the backfit is designed to achieve.

The NRC staff has completed a rev~ew. and evahiation of information develope~'

since 1980 on Unresolved Safety l!leue * (U~I) A-44, Station Blackput. As a* result . of these efforts, the NRC is amending 1P CFR Part 50 by adding a new I 50.63, , *. "Station Blackout." . . . . . . . _ . The objective of the s.tation blackout:

*
rule is to reduce the risk.of severe *
accidents associated with station blackout.

Specifically, the rule requires.

all light-water-cooled nuclear power. plants to be able to cope with a elation blackout for a specified duration.and to have procedures and training for such an event. A regulatory gulde,-to be-. *. issued aiong with the rule, provides an' acceptable method I<> determine the .' . station blackout duration for each plant. The duration is to be determined for.* each plant based .on* a ci>mpariso~.

of th~ individua!'plant design wit~ factors that have been identified as the mai~

contributors to risk of core melt resulting from elation blackout.

These.' factors are (1) the redundancy of on11ite *emergency ac power sources, (2) the reliability of onsite emergency ac power. sources, (3) the frequency of loBB of offsite power, and (4) the probable time needed to restore offsite power. 2. General description of the activity PART 50

STATEMENTS OF CONSIDERATION required by the licensee or applicant in ' significant increase in occupational order to complete the backfit. exposure is expected from operation In order to comply with the resolution and maintenance activities associated of USI A-44, licensees will be required with the rule. Equipment additions and to---:. modifications contemplated do not * * ~aintain the reliability of ~nsite . . require work in and around the reactor* emergency ac*power eo~rces at or above; coolant system and therefore are not sp*ecified acceptable reliability levels, *. expected to result in significant
Develop procedures and training to . radiation exposure.

restore ac power using nearby power* . 5. In~tallation and continuing costs sources if the emergency ac power

associated with the backfit, including sy!JtOm and the normal offsite power . . the cost of facility downtime or the cost sources are unavilable.
of construction delay.
Determine the duration that the plant should be able to withstand a For 100 operating reactors, the total
station blackout based on the factors estimated cost associated with the sp_ecified in§ so.63 , "Station Blackout," station blackout rule ranges from $42 to * $94 million with a beet estimate of $60 and Regulatory Guide 1.155, "Station million. This estimate breaks down as Blackout."
If available, an alternate ac power follows: source that meets specific criteria for independence and capacity can be used to cope with a station blackout.
Evaluate the plant's actual capability to withstand and recover from a station blackout.

This evaluation includes:

-Verifying the adequancy of station battery power, condensate
storage tank capacity, and plant/instrument air for the station blackout duration.* -Verifying the operability of equipment' needed to operate during a station blackout and the recovery from the .. blackout-for environmental conditions . associated with total loss of ac power (i.e., loss of heating, ventilation, and air conditioning).
- * *
Depending on the plant's existing capability to cope with a station blackout,.licensees may or may not need to backfit hardware modificatiomi'(e*.g.;

adding battery capacity) to comply with* the rule. (See item 8 of this analysis for-. addltionaldiscussion.) Licensees will be required to have procedures and training to cope with and recover from a station blackout.

3. Potential change in the risk to the ... public from the accidental.

offsite release of radioactive material.

Implementation of the station blackout rule will result in an estimated total risk redudion to the public ranging* from 65,000 to 215,000person-rem with a best estim8;te of about 145,000 person. rem. 4. Potential impact on radic~logical exposure of facility employees

.. For 100 operating reactors. the* estimated total reduction in occupational exposure resulting from* . reduced core damage frequencies and . associated postaccident cleanup and repair activities is 1,500 person-rem. No, Esti-Estimated total cost mated (million dollars) Activity num-berol reac-Best High Low tors Assess plant's capability to cope with station blackout ***********-****** 100 25 40 20 Develop procedures and training .............. 100 10 15 5 Improve diesel generator reliability ................... 10 2.5 4 1.5 Requalily diesel generator ....... *-****** 2 5.5 11 ?-5 Install hardware to increase plant capability to cope with station b!ackout ........... -...... 27 17 24 13 Totals ................ .............. 60 94 42 6. The J)(lte~tial safety impact of changes in plant or operational . . . complexity, including the relationship to propos~ and existing regulatory

requirements.

The rule requiring plants to,be able to cope with a station blackout should not .add to plant or operational complexity, The station blackout rule is closely related to several NRC generic programs and proposed and existing regulatory requirements as the following discussion . indicate9.

Gene.~ic /s~ue Ii-ss. Diesel Ge;,~rator . Reliability, .. * . *
The resolution of USI A-44 includes a regulatory guide on station blackout that specifies the following guidance on,
diesel generator reliability (Regulatory Guide 1,155, Sections*CU.
and C.1.2);,.
The minln'uii!J eri1ei'gency dre~1il generator. (EDGJ reliability should be ta'igilted al 0.95 per demand for each EDG for plants In* emergency ac Group* A, B. and C and at 50-SC-52 0.975 per demand for aach EDG for plant, in emergency ac Group D (see Table 2). These reliability levels will be considered minimum target reliabilities and each plant should have an EDG reliability program containing the principal elements, or their equivalent, outlined In Regulatory Position 1.2. Plants that select a target EOG reliability of 0.975 will use the higher level as the target in their EOG reliability programs. . The reliable operation of onsite emergency ac power sources should be ensared by a reliability program designed to maintain and monitor the reliability level of eacla power source over time for assurance that the selected reliability levels are being achieved.
An EOG reliability program would typically be composed of the following elements or activities (or their equivalent):
1. Individual EOG reliability target levels consistent with the plant category and coping duration selected from Table 2. 2. Surveillance testing and reliability monitoring programs designed to track EOG
performance and to support maintenance activities.
s. A maintenance program that ensures that the target EOG reliability is being achieYed and that provides a capability for failure analysis and root-cause lnwstigations.

f, An information and data collection system that services the elements of the reliability program end that monitora achieYed EOG reliability levels against tal!flt value,. S. Identified responsibilities for the major program elements and a management oversight program for reviewing reliability levels being achieved and ensuring that the

program is functioning properly.

The resolution of B-58 will provide specific guidance for use by the staff or industry to review the adequacy of diesel genera~ reliability programs consistent with the resolution of USI A-, 44. Generic Issue 23, Reactor Coolant Pump Seal Failul'e8 Reactor coolant pump (RCP) seal integrity is necessary for maintaining primary system inventory during station blackeut conditions. The estimates of core damage frequency for station blackout events for USI A-44 assumed that RCP seals would leak at a rate of 20 gallons per minute. Results of analyses performed for GI 23 will provide the information necessary to estimate RCP seal behavior during a station blackouL The industry coping analysis guidelines (NUMARC-8700) recognize the possibility of leakages exceeding an assumed 25 gpm per pump and *incorporate the need to reevaluate the plant-specific coping analysis if the resolution of GI 23 identifies higher levels. ' PART 60

STATEMENTS OF CONSIDERATION US/ A-45, Shutdown Decay Heat Removal Requirements The overall objective of USI A-45 ia to evaluate the adequacy of current *licensing design requirements to ensure that the nuclear power plants do not pose an unacceptable risk as a result of failure to remove shutdown decay heaL The study includes an assessment of alternative mean11. of ahutdown decay heat removal and of diverse "dedicated" systems for thia purpose. Results will include proposed recommeodationa regarding the desirability of. and possible design requirements for, improvements in existing systems or an alternative dedicated decay heat removal method. The USI A-44 concern for maintaining adequate core cooling under _station blackout conditions can be considered a subset of the overall A-45 issue. However, there are significant differences in scope between these two issues. USI A-44 deals with the . probability of loss of ac power, the capability to remove decay heat using systems that do not require ac power, and the ability to restore ac power in a timely manner. USI A-45 deals with the overall reliability of the decay heat removal function in terms of response to transients, small-break loss-of-coolant accidents, and special emergencies such WI fires, floods, seismic events, and sabotage.

Although the recommendatioruJ that might result from the resolution of USI A-45 are not yet final, some could affect the station blackout capability, while others would not. Recommendations that involve a new or improved decay heat removal system that is ac power dependent but that does not include its own dedicated ac power supply would have no effect on USI A-44. Recommendations that involve an additional ac-independent decay heat removal system would have a very modest effect on USI A-44. Recommendations that involve an additional decay heat removal system with its own ac power supply would have a significant effect on USI A-44. Such a new additional system would receive the appropriate credit within the USI A-44 resolution by either changing . the emergency ac power configuration group or providing the ability to cope with a station blackout for an extended period of time. Well before plant modifications, if any, will be implemented to comply with the station blackout rule, it i11 anticipated that the proposed techncial resolution of USIA-45 will be published for public commenL Those plants needing hardware -modifications for station blackout could be reevaluated before any actual modifications are made 110 that any contemplated design changes resulting from the resolution of USI A-45 can be considered at the Sf!me time. Generic Issue A-30, Adequacy of Safety-Related DC Power Supply The analysis performed for USI A-44 assumed that a high level of de power system reliability would be maintained so that (1) de power system failures would not be a significant contributor to lo&1es of all ac power and (2) should a station blackout occur, the probability of immediate de power system failure would be low. Whereas Generic Issue A-30 focuses on enhancing battery reliability, the resolution of USI A-44 ia aimed at ensuring adequate station battery capacity in the event of a station blackout of a specified duration. Therefore, these two issues are consistent and compatible. Fire Protection Program Section 50.48 of 10 CFR Part 50 states that each operating nuclear power plant must have a fire protection plan that satisfies GDC 3. The fire protection features required to satisfy GDC 3 are specified in Appendix R to 10 CFR Part 50. They include certain provisions regarding alternative and dedicated shutdown capability. To meet these provisions, some licensees have added, or plan to add, improved capability to restore power from offsite sources or onsite diesels for the shutdown system. A few plants have installed a safe shutdown facility for fire protection that includes a charging pump powered by its own independent ac power source. In the event of a station blackout, this system can provide makeup capability to the primary coolant system as well as reactor coolant pump seal cooling. This could be a significant benefit in terms of enhancing the ability of a plant to cope

with a station blackout.

Plants that have added equipment to achieve alternate safe shutdown in order to meet Appendix R requirements could take credit for that equipment, if available, for coping with a station blackout event. 7. The estimated resource burden on the NRC associated with the backfit and the availability of such resources. The estimated total cost for NRC review of industry submittals required by the station blackout rule is $1.5

million based on submittals for 100 reactors and an estimated average of 175 person-hours per reactor. 50-SC-53 8. The potential impact of differences in facility type, design, or age on the relevancy and practicality of the backfit. The station blackout rule applies to all pressurized water reactors and boiling water reactors.

However, in determining an acceptable station blackout coping capability for each plant, differences in plant characteristics relating to ac power reliability (e.g., number of emergency diesel generators, the reliability of the offsite and onsite emergency ac power systems) could result in different acceptable coping capabilities. For example, plants with an already low risk from station blackout because of multiple, highly reliable ac power sources are required to withstand a station blackout for a relatively short period of time; and few, if any, hardware backfits would be required as a result of the rule. Plants with currently higher risk from station blackout are required to withstand somewhat longer duration blackouts; and, depending on their existing capability, may need some modifications to achieve the longer station blackout capability.

9. Whether the backfit is interim or final and, if interim, the justification for imposing the backfit on an interim basis. The station blackout rule is the final resolution of USI A-44; it is not an interim measure. 53 FR 24018 Published 6/27/88 Effective 7/27/88 General Requirements for Decommissioning Nuclear Facilities See Part 30 Statements of Consideration 53 FR35996 Published 9/16/88 Effective 10/17 /88 10 CFR Part 50 Emergency Core Cooling Systems; Revisions to Acceptance Criteria AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations to allow the use of alternative methods to demonstrate that the emergency core cooling system [ECCS) would protect the nuclear reactor core during a postulated design basis loss-of-coolant accident (LOCA). The Commission is taking this action because research, performed*since the current rule was written, has shown that calculations performed using current methods and in accordance with the current requirements result in estimates PART 50 ** STATEMENTS OF CONSIDERATION of cooling system performance that are significantly more conservative than estimates based on the improved knowledge gained from this research.

While the existing methods are conservative, they do not result in accurate calculation of what would actually occur in a nuclear power plant during a LOCA and may result in less than optimal ECCS design and operating procedures. In addition, the operation of some nuclear reactors is being unnecessarily restricted by the rule, resulting in increased costs of electricity generation. This rule, while continuing to allow the use of current methods and requirements, also allows the use of more recent information and knowledge to demonstrate that the ECCS would protect the reactor during a LOCA. This amendment, which applies to all applicants for and holders of construction permits or operating licenses for light water reactors, also relaxes requirements for certain reporting and reanalyses which do not contribute to safety. EFFECTIVE DATE: October 17, 1988. FOR FURTHER INFORMATION CONTACT: L.M. Shotkin, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3530. SUPPLEMENTARY INFORMATION: Background On March 3, 1987, the Nuclear Regulatory Commission published in the Federal Register proposed amendments (52 FR 6334) to 10 CFR Part 50 and Appendix K. These proposed amendments were motivated by the fact that since the promulgation of § 50.46 of 10 CFR Part 50, "Acceptance Criteria for Emergency Core Cooling Systems (ECCS) in Light Water Power Reactors," and the acceptable and required features and models specified in Appendix K to 10 CFR Part 50, considerable research has been performed that has greatly increased the understanding of ECCS performance during a LOCA. It is now confirmed that the methods specified in Appendix K, combined with other analysis methods currently in use, are highly conservative and that the actual cladding temperatures which would occur during a LOCA would be much lower than those calculated using Appendix K methods. In soliciting the public's comments on the proposed rule, the NRC specifically requested its views on questions posed by Commissioner Asselstine and the Advisory Committee on Reactor Safeguards (ACRS). The ACRS requested that the Commission solicit the public's comments on whether the existing rule should be "grandfathered" indefinitely. That is: 1. Should the conservative ECCS evaluation method of Appendix K be permitted indefinitely or should this aspect of the ECCS rule be phased out after some period of time? Commissioner Asselstine requested the public's comments on the following:

2. Should this rule change include an explicit degree of conservatism that must be applied to the evaluation models? 3. This rule change would allow a 5 to 10 percent increase in the fission product inventory that could be released from any core meltdown scenario.

Should this rule change explicitly prohibit any increase in approved power levels until all severe accident issues and unresolved safety issues are resolved?

4. Should the technical basis for this proposed rule change be reviewed by an independent group such as the American Physical Society? Summary of Public Comments The comment period for the proposed rule revision and the draft regulatory guide (52 FR 11385) expired on July 1, 1987. Twenty-seven letters addressing the proposed rule were received by the expiration date, as well as nine responses to the request for comments on questions in the regulatory guide. A number of late comments were also received.

These were also considered to the extent that new and substantial comments were provided. The public comment on the proposed rule revisions have been divided into thirteen categories and are summarized in the following paragraphs. Categories one through four represent the responses to the specific questions posed by the ACRS and Commissioner Asselstine. In general, consideration of the public comments resulted in no substantive revision to the proposed rule. 1. Grandfathering of Conservative EGGS Methods of Appendix K (Question 1). Twenty-one of the commenters specifically addressed the ARCS question concerning the grandfathering of the current Appendix K approach. Seventeen of these commenters recommended indefinite grandfathering of the existing Appendix K evaluation models. Most cited the known conservatism as the basis of their recommendation. In addition, several commenters stated that in light of the known conservatisms not allowing continued use of existing Appendix K evaluation models would be unfairly burdensome to licensees who determine 50-SC-54 that they would not derive an economic benefit by performing realistic analysis of ECCS performance. The position of an additional commenter is unclear concerning grandfathering. The remaining commenter was not opposed to grandfathering but thought the question is premature. This commenter believes that indefinite use of existing ECCS evaluation methods should be considered when significant experience has been gained with the implementation of the new features of the rule but makes no recommendation as to what policy the Commission should pursue in the meantime. The Commission agrees with the majority of the commenters that existing Appendix K evaluation models should be permitted indefinitely. The Commission also believes that the decision to permit continued use of such models can and should be made at this time because it believes that both methods provide adequate protection of the public health and safety. As described in the regulatory analysis, the probability of a large break is so low, that the choice of best estimate versus Appendix K has little effect on public risk. The TMI action plan calls for industry to improve their small break LOCA evaluation models to be more realistic when evaluating the more probable small break accident scenario. This has been done within the context of § 50.46 and Appendix K compliance and .. was entirely appropriate since small breaks are not limiting in design basis performance and a better understanding of small break behavior is a desirable safety goal from a risk perspective. Therefore, the grandfathering provision has been retained in the final rule. 2. Specification of Explicit Degree of Conservatism (Question 2). The majority of the responses to this question indicated that the proposed rule already contains conservatism in the required uncertainty evaluation. The use of additional conservatism would be inconsistent with the objective of the rule which is to provide a realistic evaluation of plant response during a LOCA. The NRC has not included an additional explicit degree of conservatism in this rule. 3. Resolution.of all Safety Issues Prior to Allowing Power Level Increases (Question 3}. Some commenters pointed out that fission product inventory is not a direct function of total power, but rather it is the rate of fission product formation that is a direct function of power. Fission product inventory available for release during a core meltdown would be a function of burnup, not total power. PART 50

STATEMENTS*Qf'CONSIDERATION Actually, the inventory of fission products is a complex function of both time and power and not as simple as described by the commenters.

Short lived isotopes, such as xenon and iodine, quickly reach an equilibrium inventory and total steady state inventory of these fission products is a direct function of power. Inventories of long-lived isotopes, such as strontium and cesium, are functions of total fuel burnup, as described by the commenters. Intermediate-lived isotopic inventories are complex functions of time, power, and integrated power. In an independent study, documented in chapter XII of NUREG 1230, the staff determined that the change in risk due to a 5% power increase is negligible. The arguments above do not alter the Commission's position that the increase in fission products available for release during a core meltdown caused by a 5% power increase is negligible compared to the uncertainty in fission product release. The Commission has decided not to delay the proposed rule revision -pending resolution of all unresolved safety issues or severe accident issues and therefore will proceed with this final rulemaking, as planned. 4. Independent Review of Technical Basis {Question 4). Several commenters indicated that the technical basis for the proposed rule has had adequate review as the research was being performed. A number of commenters stated that it was the role of the ACRS to perform any review of the proposed rule revision because it is uniquely qualified due to its familiarity with the research. The Commission agrees that the technical basis has had adequate review, except for the uncertainty methodology which is new and untried except for the General Electric Company's use of an uncertainty evaluation of their SAFER code. As a proof of principle and demonstration of feasibility, the ACRS and a second independent peer group has reviewed

the uncertainty methodology developed by the NRC for use in quantifying the uncertainty of NRC developed thermal hydraulic transient codes. Both the ACRS and the peer group made generally favorable comments concerning the methodology; however, both groups recognized that a complete demonstration (i.e., application to small break LOCA and the reflood portion of large break LOCA) has not yet been accomplished and certain reviewers questioned whether such a demonstration could be performed successfully.

The only objectives of the NRC methodology demonstration are to demonstrate feasibility, to develop an audit tool, and to provide the necessary experience to audit licensee submittals. The staff does not believe that an NRC demonstration of the methodology is a prerequisite to this rulemaking. Licensees wishing to'adopt the best estimate approach permitted as a result of this rule are neither required to use this methodology nor to model their own methodologies after it. This methodology will play an important part in the estimate model review process. The NRC has determined through twenty years of experience that independent analysis with independent methodologies is the most effective way to intelligently review new vendor or licensee methodologies. It is therefore appropriate that this new methodology be subjected to stringent technical scrutiny, as directed by the Commission. The NRC staff is committed to completing this demonstration by the time that it will be needed to review licensee submittals and is confident that such a demonstration will be successful. Based on the paucity of negative response concerning the technical basis for the proposed rule revision and generally favorable review of the NRC uncertainty methodology, the Commission plans no further review of the technical basis. 5. General Comments on Proposed Rule. Twenty-one commenters made comments of this nature. The majority of the comments came from the nuclear industry of which 19 expressed support of the proposed rule. The industry also strongly supports the specific ECCS rule approach proposed by the NRC. One commenter neither supported nor opposed the proposed approach. One negative comment was received from an anonymous individual within the nuclear industry who implied, without specifics, that the ECCS rule is not sound and that public comment is not a fair hearing because expert insiders would be afraid to comment. Based on the absence of any supporting justification for the negative response and the unprecedented amount of research supporting the rule revision, the NRC does not consider this comment to be valid and has proceeded with this rulemaking with no major revisions. One commenter suggested that fuel reload suppliers should not be required to complete full LOCA/ECCS analyses because the hydraulics are not changed by a fuel change. Although this point is valid, the Commission believes that it is an unworkable situation to allow fuel suppliers to make use of previous analyses performed by others. It is believed that serious questions of accountability would arise in cases 50-SC-55 where errors are discovered in evaluation models, requests are made to revise plant technical specifications, or some other questions regarding the analyses are raised. The NRC believes that shared responsibility for evaluation models would not be in the best interest of the public health and safety and therefore has not implemented the suggestion of this commenter. The NRC recerved two requests for an extension of the comment period to allow time for review of NUREG-1230, which describes the research supporting the proposed rule revision. The NRC believes the comment period was sufficient since most of the research is not new and has been extensively reviewed in the past. Both commenters were contacted and told that comments received after the comment period would be considered if time permitted. Comments from both parties were received late and were indeed considered by the NRC. 6. Reporting Requirements. Some commenters viewed the proposed reporting procedures as new requirements needing consideration in the backfit analysis while others stated that they are a major relaxation and clarification of existing reporting requirements. The NRC position is that the reporting requirements are new in the sense that they will now appear in the Code of Federal Regulations. However, in practice, these reporting requirements are indeed a clarification and relaxation over the current interpretation for the existing requirements and therefore the net effect of these requirements will be to reduce the frequency for reporting and reanalysis. A number of commenters requested that only significant errors or changes in the non-conservative direction or only those that result in exceeding the 2200°F limit be required to be reported. In addition, a number of commenters suggested that the NRC require only annual reporting of significant errors or changes. The NRC considers a major error or change in any direction a cause for concern because it raises potential questions about the adequacy of the evaluation model as a whole. Therefore, the NRC requires the reporting of significant errors or changes, in either direction, on a timely basis so that the Commission may make a determination of the safety significance. Thus, the final rule contains no change in this requirement. One commenter recommended that the word "immediate" be deleted from the requirement to propose steps to be taken to demonstrate compliance in the PART 50

STATEMENTS OF CONSIDERATION event that the criteria in § 50.46(b) are exceeded.

The Commission considers this a very serious condition in which the plant is not in compliance with the regulations and may be operating in an unsafe manner. The word "immediate" reflects this seriousness and is further defined by reference in other sections of Part 50. Several commenters questioned the need to report minor or inconsequential errors or changes, even on an annual basis, as required in the proposed rule. While errors or changes which result in changes in calculated peak clad temperatures of less than 50°F are not considered to be of immediate concern, the NRC requires cognizance of such changes or corrections since they constitute a deviation from what previously has been reviewed and accepted. The proposed annual reporting is believed to be a fair compromise between the burden of reporting and the Commission's need to be aware of changes and error corrections being made to evaluation models. Therefore, the annual reporting of minor errors remains in the final rule. One commenter interpreted the use of the words "or in the application of such a model" as requiring reporting when facility changes (already reportable under § 50.59), resulting in model input changes, occur. The regulatory language referred to is intended to ensure that applications of models to areas not contemplated during initial review of the model do not result in errors by extending a model beyond the range that it was intended. The Commission does not believe that further clarification of this requirement is necessary and has not done so in the final rule. Several commenters requested a further relaxation of the reporting requirement by changing the definition of significant code errors from 50°F to 100°F. While justification for the 50°F criteria is largely judgmental, the NRC believes that it is sufficiently large to screen the code error corrections and changes which have little safety significance while providing a mechanism for timely reporting of more serious errors and changes. Since 50°F is a threshold for reporting and no further action is required pending NRC determination of safety significance, the Commission has retained this criteria in the final rule. One commenter requested consideration for allowing that the cumulative effect of several errors and corrections be applied towards the 50°F threshold. The requirement, which states that the 50°F criteria applies to the sum of the absolute magnitudes of temperature changes from numerous error corrections or model changes was formulated specifically because the Commission requires knowledge of serious deficiencies in evaluation models in use by licensees. Allowing errors or corrections which offset one another to relieve a licensee of the thirty-day reporting requirement, would be counter to this objective. If this recommendation were accepted, two errors or changes, having a large impact on the calculated peak cladding temperature but in the opposite direction, would not be reportable if the net magnitude of their difference was less than 50°F. For this reason, and the fact that no further action (beyond

reporting within thirty days] is required, the Commission retained this requirement in the final rule. 7. Continued Use of Rohsenow.

Five comments that addressed this aspect of the proposed rule were received. One commenter believed that this correlation should not be permitted without further verification and should be phased out. Other commenters supported continued use of the correlation subject to the provisions of the proposed rule. The NRC position is that no safety concern is created by continued use of the correlation, as long as the evaluation model is overall conservative. Therefore, the Commission can not justify the burden of requiring licensees to modify their evaluation models and to perform reanalysis. As discussed in SECY 83-472, current evaluation models contain more conservatisms than just those required by Appendix K. However, error corrections or changes could alter the conservatism of the model. Therefore, the Commission believes that it is necessary to ensure continued overall conservatism in the evaluation models as a basis for continued use of the correlation. Therefore, the final rule does not modify this requirement except for the correction of a typographical error identified by one commenter.

8. Uncertainty Evaluation.

The comments received on the uncertainty evaluation support the proposed rule, particularly the flexibility provided by a non-prescriptive requirement. Therefore, the Commission is publishing the final rule without modification of this requirement.

9. Acceptance Criteria.

The three comments received on this topic were all supportive of the existing criteria, as contained in § 50.46(b), and thus the 50-SC-56 Commission did not give consideration to altering them in the final rule. 10. Cladding Materials. Three commenters requested that the Commission consider broadening the language of the rule to allow the use of a range of zirconium based alloys for cladding material. The Commission believes that this modification is beyond the scope of the current rule revision and should be considered in a separate rulemaking action in which it would receive appropriate public review and comment prior to implementation. In addition, zircaloy cladding material is specified in other portions of the Code of Federal Regulations, such as § 50.44. Making a change of this type is more suitable in a broader regulatory context. Therefore, the Commission is not broadening the definition of cladding materials within this rulemaking.

11. Other Suggested Expansions to Rule Scope. One commenter believes that hydraulic loads occurring during a LOCA could cause steam generator tubes to rupture and that the NRC should resolve steam generator tube integrity safety issues prior to publishing this rule. Steam generator tubes are designed to withstand LOCA loads at allowed thinning, and there is no evidence to contradict this. If anything, the problem would be with inspection techniques to detect the actual tube thinning and whether there is an unacceptably high probability that a tube rupture during a LOCA due to tube thinning is in excess of the design basis. However, the risk from LOCA with concurrent tube rupture will not be greatly affected by the proposed rule change. As a result of the commenter's concerns, this issue has been assigned as a generic issue (GI-141) to be prioritized by the NRC staff. The results of the prioritization process will determine if further action is required.

A second commenter believes that the ECCS rule does not adequately address a plant's long term decay heat removal capability, and recommends a "short/ long term integrative analysis approach." Both the existing requirements and the proposed rule contain the requirement to provide for long term cooling subsequent to a LOCA. Small increases in power that may result from the proposed rule should not greatly change decay heat removal requirements following a LOCA or any other accident or transient. Thus, the issue of decay heat removal is not materially impacted by this rulemaking. Moreover, any proposed increase in power resulting from this rule PART 50

STATEMENTS OF CONSIDERATION promulgation would be approved only after the licensee demonstrates that decay heat removal capacities remain adequate.

The Commission is planning no further action with regard to this issue. 12. Acceptability of Models Approved Under SECY-83-472. One commenter requests that the rule language be modified to state explicitly that ECCS evaluation models that have been previously approved under SECY-83-472 continue to be acceptable under this rule. SECY 83-472 provides an alternative, acceptable method for developing ECCS evaluation models. Licensees were still required, however, to demonstrate that evaluation models developed using the SECY-83-472 approach complied with the requirements of Appendix K to Part 50. This final rule explicitly finds that ECCS evaluation models, which have been previously approved as satisfying the requirements of Appendix K, remain acceptable. Therefore, the Commission sees no need for further clarification of this issue. 13. Comments Received After Comment Period. Six letters commenting on the proposed rule were received subsequent to the end of the comment period. The Commission considered these comments to the extent that the comments provided substantive information not previously considered. One commenter believes that the proposed § 50.46(a)(2) expands the discretion of the Director of the Office of Nuclear Reactor Regulation (NRR) by allowing imposition of immediate effective restrictions on reactor operation without a prior determination that such action is required to protect the public health, safety, or interest. NRC's intent is not to alter the responsibilities of the Director of NRR but to simply retain the description of the scope of the authority that is currently found in§ 50.46(a)(l)(v).

Furthermore, the provisions of § 50.46(a)(2) do not specify the procedure to be followed by the Director of NRR. These procedures are set out in Part 2 and remain unchanged by this rulemaking.

One commenter believes that the rule is illegal because it is based solely on cost savings considerations and that there is nothing wrong with large conservatisms. The Commission disagrees with this assessment. Safety factors are required to protect the health and safety of the public when uncertainties in plant response exist. As these uncertainties are reduced, it is appropriate to modify these safety factors to provide more realistic evaluation of actual plant response. The large conservatisms of Appendix K served the public well in 1974 when there was great uncertainty in ECCS performance. However, these conservatisms are nQw known to be very large, and there is no need to "over regulate" by maintaining this unnecessary margin. This type of activity can often result in the expenditure of resources that would be better spent improving safety in other areas. The benefits to safety, while difficult to quantify, are believed to be substantial. While cost savings may have been one factor resulting in the rule change, the Commission believes that the conservatisms contained in the acceptance criteria themselves, as well as those required in the uncertainty evaluation required in this rule, are adequate to protect the health and safety of the public. This commenter also cites portions of the 1975 General Electric Company's Nuclear Reactor Study (Reed Report), which claims that there is a lack of understanding of phenomena and small safety margins. Many of the conclusions of the "Reed Report" were valid in 1975 when it was written and due to this fact it was difficult to show that sufficient safety margins existed. Most of the research discussed in NUREG-1230 has been conducted since the "Reed Report" was written and has resulted in significant improvement in understanding LOCA phenomena. We now know that significant margin to the ECCS acceptance criteria exists, particularly for the BWR/6 which was of concern in the "Reed Report." The contents of this report have been reviewed by the Commission on several occasions, most recently in NUREG-1285, and the finding has been made that no new significant safety issues are identified. For these reasons, the NRC is proceeding with this rulemaking, as proposed. The same commenter also recommends that credit for ECCS margins be taken in the Individual Plant Examinations (IPE) and not through generic rulemaking. The Commission agrees that plant specific differences may justify the application of different margins and that these may be addressed through Individual Plant Examinations. However, the requirement for licensees to evaluate ECCS performance and meet the acceptance criteria specified in 10 CFR 50.46(b) is generic. The Commission believes that margins that may be reduced due to a better understanding of a reactor's response to a LOCA should be applied through a generic rulemaking action because it allows a broad range 50-SC-57 of technical review of the issues, enhances public participation in the process, and provides a complete public record. Therefore, the Commission has decided to proceed with the rulemaking as planned. Finally, this commenter questions the experimental basis for this rule because full-scale ECCS bypass data is not yet available. The 2D/3D tests which will provide this important data represent a small portion of the total research upon which this rule relies. Significant research on ECCS bypass has already been completed in small scale vessels and the full-scale work is required only to confirm the smaller scale results and quantify any uncertainty due to scale effects. One full-scale ECCS bypass test has already been completed under the 2D/3D program which showed that more margin exists than expected from the small scale tests. Completion of the scale tests only affects the uncertainties in the calculations, and reduces them. Uncertainties must be addressed by licensees in any analysis under the revised rule whether 2D/3D results are available or not. The Commission concludes that there is no need to delay the final rule, while awaiting these data. Summary of Rule Changes Section 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors. Section 50.46(a)(l) is amended and redesignated § 50.46(a)(l)(i) to delete the requirement that the features of Section I of Appendix K to Part 50 be used to develop the evaluation model. This section now requires that an acceptable evaluation model have sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a LOCA. The NRC expects that the analytical technique will, to the extent practicable, utilize realistic methods and be based upon applicable experimental data. The amended rule also requires that the uncertainty of the calculation be estimated and accounted for when comparing the results of the calculation to the temperature limits and other criteria of§ 50.46(b) so that there is a high probability that the criteria would not be exceeded. The Commission expects the realistic evaluation model to retain a degree of conservatism consistent with the uncertainty of the calculation. The final rule does not specifically prescribe the analytical methods or uncertainty evaluation techniques to be used. However, guidance has been provided in the form PART 50

STATEMENTS OF CONSI.DERATION of a Regulatory Guide.1 In SECY-83-472, the NRC has found acceptable an approach for estimating the 95th percentile of the probability distribution.

This percential is considered adequate to meet the high level of probability required by the rule. It is also recognized that the probability cannot be determined using totally rigorous mathematical methods due to the complexity of the calculations. However, the NRC requires that any simplifying assumptions be stated so that the Commission may evaluate them to ensure that they are reasonable. The NRC has independently developed and exercised a methodology to estimate the uncertainty associated with its own thermal-hydraulic safety codes. This methodology is described in the "Compendium of ECCS Research." 2 This document also provides reference to the large body of relevant hydraulic research, documents NRC studies on the effects of reactor power increases on risk, and provides background information on the ECCS rule. While this method has not been reviewed for acceptability from the standpoint of safety licensing, it may provide additional guidance on how the uncertainty may be quantified. In addition to providing guidance to industry, this work was undertaken to provide a proof of principle and a tool to independently audit submittals. Appendix K, Section II, "Required Documentation," remains generally applicable, with only minor revisions made to be consistent with the amended rule. A new paragraph (ii] has been added to § 50.46(a][1) to allow the features of Section I of Appendix K to be used in evaluation models as an alternative to performing the uncertainty evaluation specified in the amended § 50.46(a)(l](i). This method would remain acceptable because Appendix K is conservative with respect to the realistic method proposed in the amended § 50.46[a][l)(i). This would allow both current and future applicants and licensees to use existing evaluation models if they did not need or desire relief from current operaih1g restrictions. In § 50.46, paragraphs (a) (2) and (3) have been revised to eliminate portions of those paragraphs concerned with historical implementation of the current file. These provisions have been 'Regulatory Guide, "Best Estimate Calculations of Emergency Core Cooling Systems Performance," RG 1.157. '"Compendium of ECCS Research for Realistic LOCA Analysis," NUREG-1230, TBP. replaced as described in the following paragraphs. Section 50.46(a][2) has been revised to indicte that restrictions on reactor operation may be imposed by the Director of Nuclear Reactor Regulation, if the ECC cooling performance evaluations are not consistent with the requirements of§ 50.46(a][1) [i] and [ii]. This section has been added to retain similar requirements that have been deleted from § 50.46(a)(l](i) by this rule revision. This section does not specify the procedures to be followed by the Director. These procedures are found in Part 2 and are unchanged by this rulemaking. The current rule contains no explicit requirements concerning reporting and reanalysis when errors in evaluation models are discovered or changes are made to evaluation models. However, current practice has required reporting of errors and changes and reanalyses with the revised evaluation models. This final rule explicitly sets forth requirements to be followed in the event of errors or changes. The definition of a significant change is currently taken from Appendix K, Section 11.1.b which defines a significant change as one which changes calculated cladding temperature by more than 20 °F. The revised§ 50.46(a)(3) states specific requirements for reporting and reanalyses when errors in evaluation models are discovered or changes are made to evaluation models. It requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported. Thereafter, although reanalysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected. The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the siguificance of the error or change and to maintain cognizance of modifications made subsequent to NRC review of the evaluation model. Past experience has shown that many errors or changes to evaluation models are very minor and the burden of immediate reporting cannot be justified for these minor errors because they do not affect the immediate safety or operation of the plant. The NRC therefore requires periodic reporting to satisfy NRC's need to be apprised of changes or errors without imposing an unnecessary burden on the applicant or licensee. This 50-SC-58 report is to be filed within one year of discovery of the error and must be reported each year thereafter until a revised evaluation model or a revised evaluation correcting minor errors is approved by the NRC staff. Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as to the adequacy of the overall evaluation model. This final rule defines a significant error or change as one which results in a calculated peak fuel cladding temperature different by more than 50 °F, or an accumulation of errors and changes such that the sum of the absolute magnitude of the temperature changes is greater than 50 °F. More timely reporting (30 days] is required for significant errors or changes. This definition of a significant change is based on NRC's judgment concerning the importance of errors and changes typically reported to the NRC in the past. This final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements. Errors or changes that result in the calculated plant performance exceeding any of the criteria of§ 50.46[b] mean that the plant is not operating within the requirements of the regulations and require immediate reporting as required by § 50.55[e], § 50.72 and§ 50.73 and immediate steps to bring the plant into compliance with § 50.46. Appendix K EGGS Evaluation Models Amendments have been made to Appendix K, Section I.C.5.b, to modify the post-CHF heat transfer correlations listed as acceptable. The "McDonough" reference has been replaced with a more recent paper by the same authors entitled "An Experimental Study of Partial Film Boiling Region With Water at Elevated Pressures in a Round Vertical Tube" which is more generally available and which includes additional data. The heat transfer correlation of Dougall and Rohsenow, listed as an acceptable heat transfer correlation in Appendix K, paragraph I.C.5.b, has been removed, because research performed since Appendix K was written has shown that this correlation overpredicts heat transfer coefficients under certain conditions and therefore can produce nonconservative results. A number of applicants and licensees currently use the Dougall-Rohsenow correlation in approved evaluation models. The NRC has concluded that the continued use of this correlation can be allowed. This is appropriate (even though parts of the approved evaluation model, Dougall-PART 50

STATEMENTS OF CONSIDERATION Rohsenow, are known to be nonconservative) because the existing evaluation models are known to contain a large degree of overall conservatism even while using the Dougall-Rohsenow correlation.

This large overall conservatism has been demonstrated through comparisons between evaluation model calculations and calculations using NRC's best-estimate computer codes. Thus, requiring that the applicants and licensees remove the Dougall-Rohsenow correlation from their current evaluation models cannot

be justified as necessary to maintain safety. The stipulation that the Rohsenow correlation will cease to be acceptable for previously approved evaluation models applies only when changes to the model are made which reduce the calculated peak clad temperature by 50 °For more. However, the requirement to report any changes or culmination of changes, such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 °F, still applies. A new Section I.C.5.c has been added to Appendix K to state the Commission's requirements regarding continued use of the Dougall-Rohsenow correlation in existing evaluation models. Evaluation models which make use of the Dougall-Rohsenow correlation and have been approved prior to the effective date of this rule may continue to use this correlation as long as no changes are made to the evaluation model which significantly reduce the current overall conservatism of the evaluation model. If the applicant or licensee submits proposed changes to an approved evaluation model, or submits corrections to errors in the evaluation model which significantly reduce the existing overall conservatism of the model, continued use of the Rohsenow correlation under conditions where nonconservative heat transfer
coefficients result would no longer be acceptable.

For this purpose, significant reduction in overall conservatism has been defined as a "net" reduction in calculated peak clad temperature of at least 50°F from that which would have been calculated using existing evaluation models. A reduction in calculated peak clad temperature could potentially result in an increase in the actual allowed peak power in the plant. An increase in allowed plant peak power with a known nonconservatism in the analysis would be unacceptable. This definition of a significant reduction in overall conservatism is based on a judgment regarding the size of the existing overall conservatism in evaluation model calculations relative to the conservatism required to account for overall uncertainties in the calculations. Appendix K, Section 11.1.b, has been removed since this requirement has been clarified in the amended § 50.46(a)(3). Likewise, Appendix K, Section 11.5, has been amended to account for the fact that not all evaluation models will be required to use the features of Appendix K, *section I. These minor changes to Appendix K do not affect any existing approved evaluation models since the changes are either "housekeeping" in nature or are changes to "acceptable features," not "required features." Availability of Documents

1. Copies of NUREGs 1230 and 1285 may be purchased from the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082.

Copies are also available from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. A copy is also available for public inspection and/or copying at the NRC Public Document Room, 2120 L Street NW., Washington, DC 20555. 2. Copies of SECY--83-472, an information report entitled "Emergency Core Cooling Systems Analysis Methods," dated November 17, 1983, is available for inspection and copying at the NRC Public Documents Room, 2120 L Street NW., Washington, DC 20555. Single copies of this report may be obtained by writing L. M. Shotkin, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555. 3. Regulatory Guide, "Best Estimate Calculations of Emergency Core Cooling Systems Performance, Task RS 701-4, may be obtained by writing to the Division of Information Support Services, U.S. Nuclear Regulatory Commission, Washington, DC 20555. 4. The Paraphrased Summary of Public Comments on the ECCS Rule is available for public inspection at the NRC Public Documents Room, 2120 L Street NW., Washington, DC 20555. Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule is not a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. The primary effect of the rule is to allow an increase in the peak local power in the reactor. This could be used 50-SC-59 either to tailor the power shape within the reactor or to increase the total power. Changing the power shape without changing the total power has a negligible effect on the environmental impact. The total power could also be increased, but is expected to be increased by no more than about 5% due to hardware limitations in existing plants. This 5% power increase is not expected to cause difficulty in meeting the existing environmental limits. The only change in non-radiological waste will be an increase in waste heat rejection commensurate with any increase in power. For stations operating with an open (once through) cooling system, this additional heat will be directed to a surface water body. Discharge of this heat is regulated under the Clean Water Act administered by the U.S. Environmental Protection Agency (EPA) or designated state agencies. It is not intended that NRC approval of increased power level affects in any way the responsibility of the licensee to comply with the requirements of the Clean Water Act. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW., Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from L. M. Shotkin, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington DC 20555, telephone (301) 492-3530. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These reporting requirements were approved by the Office of Management and Budget (Approval Number 3150-0011). Regulatory Analysis The Commission has prepared a regulatory analysis for this final regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The regulatory analysis is available for inspection and copying for a fee at the NRC Public Document Room, 2120 L Street NW., Washington, DC. Single copies of the analysis may be obtained from L. M. Shotkin, Office of Nuclear Regulatory Research, Washington, DC 20555, telephone (301) 492-3530. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b ), PART 50

STATEMENTS OF CONSI.DERATION the Commission certifies that this rule will not have a significant economic impact upon a substantial number of small entities.

This rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration in 13 CFR Part 121. Since these companies are dominant in their service areas, this rule does not fall within the purview of the Act. Backlit Analysis A backfit analysis is not required by 10 CFR 50.109 because the rule does not require applicants or licensees to make a change but only offers additional options and provides a clarification and relaxation of existing reporting requirements. Nonetheless, the factors in 10 CFR 50.109[c) have been analyzed for the entire rule. 1. Statement of the specific objectives that the backfit is designed to achieve. The objective of the rule is to modify 10 CFR 50.46 and Appendix K to permit the use of realistic ECCS evaluation models. More realistic estimates of ECCS performance, based on the improved knowledge gained from recent research on ECCS performance, may remove unnecessary operating restrictions. Also experience with the previous version of § 50.46 has demonstrated that a clearer definition of reporting requirements for changes and errors is very desirable.

2. General description of the activity that would be required by the licensee or applicant in order to complete the backfit. The amendment allows alternative methods to be used to demonstrate that the ECCS would protect the nuclear reactor core during a postulated design basis loss-of-coolant accident (LOCA). While continuing to allow the use of current Appendix K methods and requirements, the rule also allows the use of more recent information and knowledge currently available to demonstrate that the ECCS would perform its safety function during a LOCA. If an applicant or licensee elects to use a new realistic model they will be required to provide sufficient supporting justification to validate the model and include comparisons to experimental data and estimates of uncertainty.

In accounting for the uncertainty, the analysis would have to show, with a high level of probability, that the ECCS performance criteria are not exceeded. Whether or not a licensee or applicant chooses to use realistic analysis, complete with an uncertainty analysis, each licensee must comply with the requirement to report changes to their evaluation models (i.e'., less than 50°F change in calculated peak cladding temperature) annually to the NRC. In addition, significant changes (those which have a greater than 50°F change in calculated peak cladding temperature) have to be reported within 30 days. 3. Potential change in risk to the public from the accidental offsite release of radioactive materials. The rule could result in increased local power within the reactor core and possible increases in total power. Power increases on the order of 5 will have an insignificant effect on risk. One effect of increased power could be to increase the fission product inventory. A five percent power increase would result in a less than five percent increase in fission products. Thus, less than five percent more fission products might be released during core melt scenarios and potentially released to the environment during severe accidents. The rule still requires the fuel rod peak cladding temperature (PCT) remain below 2200°F. Reactors choosing to increase power by about five percent will be operating with less margin between the PCT and the 2200°F limit than previously. The increased risk represented by this decrease in margin and increase in fission product inventory is negligible and falls within the uncertainties of PRA risk estimates. In addition, other safety limits, such as departure from nucleate boiling [DNBJ, and operational limits, such as turbine design, will limit the amount of margin reduction permitted under the rule. The rule could also potentially reduce the risk from pressurized thermal shock by allowing the reactor to be operated in a manner which reduces the neutron fluence to the vessel. 4. Potential impact on radiological exposure to facility employees. Since the primary effect of the rule involves the calculational methods to be used in determining the ECCS cooling performance, it is expected that there will be an insignificant impact on the radiological exposure to facility employees. Because of the reduced LOCA restrictions resulting from the new calculations it is possible for the plant to achieve more efficient operation and improved fuel utilization with improved maneuvering capabilities. As a result, it is conceivable that there could be a reduction in radiological exposure if the fuel reloads can be 50-SC-60 reduced. This effect is not expected to be very significant.

5. Installation and continuing costs associated with the backfit, including the cost of facility down times or the cost of construction delay. LOCA considerations resulting from the present rule are restricting the optimum production of nuclear electric power in some plants. These restrictions can be placed into the following three categories:

(1) Maximum plant operating power, (2) Operational flexibility and operational efficiency of the plant, and (3) Availability of manpower to work on other activities. The effect of the rule will vary from plant to plant. Some plants may realize savings of several million dollars per year in fuel and operating costs. Significantly greater economic benefit would be realized by plants able to increase total power as a result of this final rule. The regulatory analysis cited above indicates that the total present value of the energy replacement cost savings for a five percent power upgrade would vary between 18 and 127 million dollars depending on the plant. Additional information concerning these potential cost savings are included in the regulatory analysis. The costs associated with the new reporting requirements are deemed to be minimal. Although the existing Appendix K has no official reporting requirements, paragraph 11.1.b was interpreted by the staff to require a reanalysis and report to NRC when significant changes are made which change the peak cladding temperature by more than 20 °F. Therefore, this rule change, by changing the definition of significant changes to 50 °F, is actually a relaxation of current practices. The annual reporting of changes that are not significant is not viewed by the NRC as a major burden since no other action is required.

6. The potential safety impact of changes in plant or operational complexity including the effect on other proposed and existing regulatory requirements.

There are safety benefits derivable from alternative fuel management schemes that could be utilized. The higher power peaking factors that would be allowed with the final rule provide greater flexibility for fuel designers when attempting to reduce neutron flux at the vessel wall. This can result in a corresponding reduction in risk from pressurized thermal shock. The reduced cladding temperatures that would be calculated under the revised rule offers the possibility of other design and operational changes PART 50

STATEMENTS OF CONSIDERATION that could result from the lower calculated temperatures.

ECCS equipment numbers, sizes or surveillance requirements might be reduced and still meet the ECCS design criteria (if not required to meet other licensing requirements). Another option may be to increase the diesel/generator start time duration. In summary, the effect of this rule on safety would have both potential positive and negative aspects. The potential for reduction of ECCS system capability in existing or new plants is present. However, several positive aspects may also be realized under the final rule. The net effect on safety would be plant specific. However, the probability of a large break LOCA is so low that the choice of best estimate versus Appendix K would have little effect on public risk. 7. The estimated resource burden on the NRG associated with the proposed back/it; and the availability of such resources. The major staff resources required under the final rule are to review the realistic models and uncertainty analysis required by the revised ECCS Rule. Based on previous experience with the General Electric Company's SAFER model and the learning that has resulted from these efforts, it is estimated that approximately one staff year would be required to review each generic model submitted. There are four major reactor vendors (GE already has a revised evaluation model approved under the existing Appendix K for both jet pump and non-jet pump plants and may update their methodology under this new rule) and several fuel suppliers and utilities which perform their own analyses and potentially might submit generic models for review. However, it is expected that only 3 or 4 generic models would be submitted since not all plants would benefit from this rule. Thus, about 3-4 staff years would be required to review the expected generic models. Once a generic model is approved, the plant specific review is very short. In addition, several vendors are currently planning to submit realistic models in conjunction with the use of

SECY-83-472.

Therefore, staff resources would be expended to review these models in any event. Since these models would not change as a result of the revised ECCS rule, there should be no net increase in resources required over that already planned to be expended. In summary, while it is difficult to estimate accurately, it is expected that the rule change will have a small overall impact on NRC resources.

. 8. T~f:; potential impact of differences m facility type, design or age on the relevancy and practicality of the back/it. . The degree to which the rule would affect a particular pl.ant depends on how limited the plant is by the LOCA restrictions.

General Electric Company (GE) plants do tend to be limited in operation by LOCA restrictions and would benefit from relief from LOCA restrictions. However, this relief is already available for most GE plants through the recently approved SAFER evaluation model. Any additional relief due to a rule change would be of little further benefit. Westinghouse (W) plants would appear to directly benefit from relaxation of LOCA limits. W plants represent the largest number of plants, with 47 plants operating and 10 additional plants being constructed. W indicates that most of these plants are limited by LOCA considerations. The potential benefit for plants of B&W and CE design is uncertain at this time. 9. Whether the proposed back/it is interim or final and if interim, the justification for imposing the proposed back/it on an interim basis. The rule, when made effective, will be in final form and not interim form. It will continue to permit the performance of ECCS cooling calculations using either realistic models or models in accord with Appendix K. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting crite 7 ia, Reporting and Recordkeeping reqmrements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendments to 10 CFR Part 50. 53 FR 36955 Published 9/23/88 Effective 10/24/88 10 CFR Part 50 Emergency Planning and Preparedness Requirements for Nuclear Power Plant Fuel Loading and Low-Power Testing AGENCY: Nuclear Regulatory Commission. ACTION: Final rule. 50-SC-61

SUMMARY

The Nuclear Regulatory Commission is amending its regulations to establish more clearly what emergency planning and preparedness requirements are needed for fuel loading and low power testing of nuclear power plants. The rule itself will now require NRC findings on the licensee's emergency plans for dealing with accidents that could affect persons on site. The Commission's prior practice of considering certain offsite elements of licensee's plans has been modified and codified in this regard to provide that NRC findings will be required before fuel loading or low power testing in coordination with offsite personnel and agencies so that necessary resources can be applied on site for mitigating and containing accidents, and so that offsite agencies may be kept informed of plant events. The rule will also change the prior practice, never included in the prior rule itself. of reviewing plans for prompt public notification in the event of an accident.

This practice of reviewing an offsite element of licensee emergency plans that has no onsite application is being discontinued as not necessary for public safety. The rule does not change the emergency planning requirements that must be satisfied before full power operation can be authorized. No new requirements are being imposed by the rule beyond those that have been previously required by rule and by prior NRC practice. The rule makes clear that no offsite elements of the applicant's emergency plan, other than those set forth in this revised rule, need be considered in connection with low power licensing. EFFECTIVE DATE: October 24, 1988. FOR FURTHER INFORMATION CONTACT: Carole F. Kagan, Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555; Telephone (301) 492-1632, or Michael T. Jamgochian, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555; Telephone (301) 492-3918. SUPPLEMENTARY INFORMATION: I. Background On May 9, 1988, the Commission published in the Federal Register (53 FR 16435) a notice of proposed rulemaking which would establish more clearly what emergency planning and preparedness requirements are needed for fuel loading and low power testing of nuclear power plants. As detailed in the notice of proposed rulemaking, 10 CFR 50.47(d) as promulgated on July 13, 1982 (47 FR 30232) provided that only a finding as to the adequacy of an PART 50

STATEMENTS Of CONSIDERATION applicant's onsite emergency planning and preparedness is required for low power. However, the provision in the Statement of Considerations that systems for prompt notice to the public in the event of an accident would also be reviewed before low power focused on protection of persons off site. The Statement of Considerations for the 1982 rule change gave no clear and consistent rationale for why the particular element dealing with public notification should be included.

The foundation for that rulemaking was the Commission's determination, described in more detail below, that the degree of emergency planning and preparedness necessary to provide adequate protection of the public health and safety is significantly less than that required for full power opera lion in light of the significantly lower risks associated with even low likelihood accidents at that stage (47 FR 30233 and note 1). Thus, the stated rationale for the 1982 rule would seem to undercut the need for any prompt public notification requirement. The Commission indicated in 1982 that, although at low power plant operators typically have less experience and there is a greater potential for undiscovered defects, the risk to public health and safety at low power is significantly lower than at full power as a result of several factors. Those reasons were stated by the Commission as follows: First, the fission product inventory during low power testing is much less than during higher power operation due to the low level of reactor power and short period of operation. Second, at low power there is a significant reduction in the required

capacity of systems designed to mitigate the consequences of accidents compared to the required capacities under full power operation.

Third, the time available for taking actions to identify accident causes and mitigate accident consequences is much longer than at full power. This means that operators should have sufficient time to prevent a radioactive release from occurring. In the worst case, the additionaftime available (at least 10 hours], even for a postulated low likelihood sequence which could eventually result in release of the fission products accumulated at low power into the containment, would allow adequate precautionary actions to be taken to protect the public near the site (47 FR 30233). The safety basis for the 1982 rule was reviewed as a necessary part of the instant proposed rulemaking. The Commission reexamined the need at low power to review those aspects of applicants' onsite plans that seem relevant only to offsite protective measures that might be needed if there were an accident with offsite dose effects (53 FR 16436-7). The proposed rule indicated that the Commission saw no need to review those aspects of applicants' plans that did not have a direct relationship to onsite dose effects in light of the significantly less risk to. offsite persons presented by fuel loadmg and low power testiJ;ig as contrasted with full power operation. On reexamination in light of public comment, the Commission has reaffirmed the safety conclusion that the safety risk to the public from low power testing is significantly less than the risk to the public from full power operation. Accordingly, the rule is being issued in final form substantially as proposed. However, a number of changes have been made in the rule in response to public comments. II. Analysis of Public Comments Nearly 1700 comments were received on the proposed rulemaking. The overwhelming majority were from private citizens, mostly in the New England area. Comments also came from utilities, industry groups, State and local government agencies and officials, members of Congress, one Federal agency and several local and national environmental groups. The comments ran approximately two to one in favor of promulgation of the proposed rule. Of those opposed, approximately 500 were form letters from residents of the area surrounding the Seabrook nuclear power plant. The remaining 60 to 70 comments in opposition were from private citizens, State and local government officials and environmental groups. The comments in favor came primarily from private citizens, with a sprinkling from utilities, nuclear industry organizations, one local government official and one Federal agency. Because of the large volume of comments received, it would be impractical to discuss each individually. The great majority of comments, both for and against the proposed rule, turned on the commenter's opinion on the impact of the rule on licensing the Seabrook facility. Most of the individuals who wrote in support of the rule expressed the opinion that the facility was ready to be licensed, that the power the facility would generate was needed, and that licensing should not be allowed to be held up by political forces. Most commenters in favor of the rule also expressed the opinion that the risks to the public from low power testing were considerably less than those from full power operation, and that prompt emergency notification to the general public should not be necessary at low power. The significant comments against the rule fall within the scope of fifteen separate major comments. These major comments and the Commission's response to them are set forth below. Comment 1. The risk assessments upon which the rule is _based are based 50-SC-62 on operation over a short time frame. However, there is no time limit for low power testing. Response. For many years, Commission policy has been to issue separate "low power" licenses which allow a plant to load fuel and perform testing and operator training at power levels up to 5 percent whenever to do so would expedite the licensing process without prejudicing the rights of any intervening parties. The purpose of the low power test program is to demonstrate that the overall plant performance conforms to the established design criteria and to confirm the operability of plant systems and design features that could not be completely tested during the preoperational test

phase. Tests that are performed during the program are specific to the type of light-water reactor (boiling water reactor versus pressurized water reactor), but typically include_ determination of in-core flux distribution, moderator temperature coefficients, control rod worths and adequacy of neutron instrumentation and associated protective functions.

Also, during this time operators obtain some valuable additional training manipulating the controls of the reactor at low power levels. In practice, many of these tests and manipulations are performed with the reactor at less than 1 percent of rated power, and those tests and manipulations which are performed with the reactor at "peak" low power (typically 3 percent to 4 percent of rated power) are completed within a day or two. Based on experience with U.S. commercial power plant startup test programs, the period over which a reactor would actually operate at or near 5 percent power during the low power test program is expected to be at most a few weeks; likewise, operation at 5 percent power beyond these few weeks would not be economically feasible. The safety evaluation supporting this rule assumes that operation under the rule would be consistent with this prior history and practice. To further clarify this point, low power licenses issued under this rule will be for purposes of fuel loading and low power testing and operator training only: steady-state operation at or near 5 percent for the full license term would not be authorized. Comment 2. The technical basis for both the current rule and the 1982 rule is flawed in that, at 5 percent power, substantial inventories of biologically significant fission products will be developed in from eight to forty days. Thus, while the inventory of all radionuclides developed during low power testing is reduced compared to full power operation, the inventory of radionuclides with public health significance still poses a substantial prompt public health hazard. In addition, the inexperience of the PART 50

STATEMENTS OF CONSIDERATION operators during low power testing and the newness of the system create a greater potential for undiscovered defects and incidents.

Response. Yes, there are some biologically significant fission products generated in the reactor core during the low power operation contemplated by this rule. But, although during low power testing plant operators typically have less experience and there is a greater potential for undiscovered defects, the risk at low power is still sufficiently low to provide reasonable assurance that public health and safety is protected even in the absence of the requirement for a prompt notification system and other purely offsite elements of emergency plans. This is a result of three factors, which were stated earlier by the Commission and which the Commission reaffirms in this rulemaking as follows: First, the fission product inventory during initial low power testing is much less than during higher power operation due to the low level of reactor power and short period of operation at this power level. The available inventory of fission products that are significant contributors to public health consequences would be reduced by about a factor of 20 for continuous operation at 5 percent power compared to continuous full power operation. However, as explained above, based on experience with commercial nuclear power plant startup test programs, operation at or near 5 percent power is only expected for a maximum of a few weeks. This would result in a further reduction in available fission product inventory. Second, at low power there is a significant reduction in the required capacity of systems designed to mitigate the consequences of accidents compared to the required capacities under full power operation. For example, the coolant flow required to dissipate decay heat at 10 hours following a loss of coolant accident in a typical pressurized water reactor would be less than 10 gallons per minute, which is well within the capacity of normal make-up systems. Most of the regulatory requirements for safety systems during reactor power operation, including containment integrity, emergency core cooling, and redundant power supplies, are the same for 5 percent power operation as they are for 100 percent power. Third, the time available for taking actions to identify accident causes and mitigate accident consequences is much longer than at full power. This means the operators should have sufficient time to prevent a radioactive release from occurring. The above safety evaluation makes no assumptions about the time that would be needed to notify the public off site and to implement an offsite emergency response if one would assume hypothetically that an off site release would occur: it is based solely on an analysis of the likelihood that an offsite release could occur and of the possible magnitude of that release. However, as an additional, separate consideration, the Commission al90 believes that, in the worst case, the additional time available (at least 10 hours), even for a postulated low likelihood sequence which could eventually result in release of the fission products accumulated at low power into the containment, would allow notification of both onsite and offsite emergency response organizations. These organizations would likely have adequate time to implement some offsite response should that be necessary. Without a prompt public notification system in place and an approved and tested offsite emergency plan, there obviously cannot be the same kind ofreasonable assurance of offsite protective measures that there would be with a fully reviewed and tested offsite emergency plan should there be an offsite release at low power. However, given the extremely low likelihood of any accident resulting in significant offsite releases, the requirements for procedures to notify emergency response organizations and the additional time that will likely be available would provide sufficient time for the emergency response organizations to implement some form of public notification and to carry out some reasonably effective offsite emergency response, even if such a release were to eventually occur. Comment 3. Testing at low power is riskier than full power operation because it involves deliberately defeating safety systems. Response. While some selected safety systems may be disabled during low power testing, the heat load and fission product inventory are significantly less than at full power. There are a number of methods available to remove this very low heat load generated at low power. In addition, special procedures are developed and followed for these tests, which are closely monitored by plant personnel. Therefore, because of the reduced heat load, small fission product inventory and special attention by plant operators, testing at low power does not place the plant at greater risk and presents a significantly lesser risk than does full power operation. Comment 4. The Chernobyl accident occurred while the reactor was at low power. Why does the NRC still say that the risk of low power testing is low? Response. The reactor physics characteristics of U.S. light-water reactors are very different from those of the.graphite-moderated RBMK type of reactor at Chernobyl. Positive void (and moderator temperature) coefficients, 50-SC-63 which played a central role in the accident at Chernobyl, are generally absent in U.S. reactors. Where they are present, they have a limited reactivity insertion potential, which precludes their causing any significant reactivity transient and power level increase. Substantial required shutdown reactivity margins in conjunction with fast automatic insertion of control rods on signals Indicative of unsafe conditions provide protection against the occurrence of reactivity excursions, such as that which took place at Chernobyl, in commercial U.S. reactors. U.S. light-water reactors do not have the inherent potential to rapidly elevate their reactor power to levels at which plant risk becomes significant. Additionally, the Chernobyl reactor operated at full power prior to its accident. Therefore, the buildup of fission product inventory was much higher than the buildup of fission product inventory at U.S. reactor operating under a fuel loading or low power testing license. Comment 5. Low power licensing fails the cost-benefit analysis required by NEPA. Response. This issue falls outside the scope of this rulemaking, which is only designed to address the requirements under the Atomic Energy Act for emergency planning at fuel loading and low power. The establishment of these safety requirements does not have a significant environmental impact under NEPA. The question of the correct NEPA .

analysis to be done in support of a low power license for any specific facility is made by case-by-case determination, and is not the subject of this rulemaking.

Comment 6. A low power license should not be issued where it is not certain that a full power license will ever be granted. The Shoreham reactor was irradiated unnecessarily. Response. This again is an issue that is not the subject of this generic rulemaking. In the past the Commission has addressed this issue in individual adjudicatory opinions, e.g., Long Island Lighting Company (Shoreham Nuclear Power Station), CLl-85-12, 21 NRC 1587 (1985), and does not believe that the issue warrants resolution generically by rulemaking. Comment 7. The proposed rule states t_hat the safety analysis performed in 1982 is still valid. After performing that analysis, the NRC decided to require that certain offsite aspects of emergency plans be in place prior to low power licensing. The NRC has given no rationale for changing the rule, while admitting that the previous analysis is valid. Response. One reason for this rule change is to clarify language in the rule itself that can easily be read to suggest that no offsite emergency planning elements need to be reviewed prior to PART 50

STATEMENTS 0-F CONSIDERATION fuel loading or low power testing. The 1982 safety analysis supported the proposition that those offsite aspects of emergency planning which are pertinent to protecting persons on site need be considered prior to low power. This rule change will incorporate this important safety consideration.

The provision in the 1982 rulemaking which is being reconsidered is the provision in the Supplemental Information that systems for prompt notification of the public in the event of an accident should be in place and reviewed at low power. However, this change is consistent with the 1982 safety analysis. Plans will still be required for notification of offsite planning and *response agencies so that these agencies and licensees may, as appropriate, keep the media and the public informed. But given the relatively low risk to the ,public from low power operation, a requirement for prompt notification of the public is far in excess of what is reasonably needed. Nothing in the 1982 rulemaking logically supports the contrary. Comment 8. The NRC has previously stated that review of the licensee's onsite response mechanism will necessarily include aspects of some offsite elements. Why is the NRC changing this position? Response. See the Response to Concern 7. The NRC is not changing its expert conclusion as to the lower level of risk from low power operation. However, this rulemaking is a more logical result of this expert conclusion than the positions stated in the 1982 Supplemental Information. Comment 9. The new rule does not address the risk of a terrorist attack or sabotage at low power. Response. Prior to receiving a low power license, a licensee must fully meet the requirements of.10 CFR 73.55. These requirements assure the implementation of an acceptable security plan around a nuclear power plant. These are the same security requirements that a licensee must meet prior to receiving a full power license. While the risk from terrorism or sabotage cannot be quantified, it is the Commission's judgment that compliance with § 73.55 will reasonably assure that the risk from terrorism or sabotage at low power is sufficiently low so as not to undercut the conclusion that low power safety risks to the offsite public are relatively low. Comment 10. The risks of an accident at low power are not confined to those onsite. If an accident were to occur at low power, public panic could ensue. Response. The Commission responded to a similar comment in promulgating the 1982 rule. See Issue 6, 47 FR at 30234. The Commission is not unmindful that, regardless of the objective lack of danger, members of the public may be made uneasy and could panic unnecessarily if an accident were to occur at low power. It was in response to this comment that the Commission agreed to review, and will continue to review, certain offsite notification elements of emergency plans prior to low power testing. In particular, prior to low power, means to keep state and local response organizations informed in the event of an onsite accident will be reviewed and approved. These organizations, through. normal communication mechanisms, have the capability to inform the public, if needed, in order to avert panic. However, the Commission has found that the immediate direct notification of the pubiic call~d fo~ by the language in the 1982 rule preamble is far in excess of what is necessary to keep the public informed. Comment 11. The change in proposed § 50.47(d)(5) to modify the requirement for provisions for monitoring offsite consequences from "in use" to "available" will create unacceptable delay in the identification of an actual or potential hazard to the public stemming from a radiological emergency. Response. The final rule will retain the phrase "in use". The wording change in the proposed rule was not intended to change current NRC staff practice of reviewing licensee onsite plans to assure they meet the intent of § 50.47(b)(9) and Planning Standard I of NUREG-o654 prior to issuance of an operating license limited to fuel loading and low power testing. While the safety evaluation which supports the elimination of the prompt public notification requirement for low power suggests that an offsite release is extremely unlikely, the Commission still considers it prudent to have release _monitoring equipment in use on site so that, at a minimum, the licensee is in a position to verify objectively that no release has occurred. Comment 12. The original rule justified retention of emergency planning for research reactors, but not for commercial reactors, since research reactors were perceived to be located in areas of high population density. This contradicts the Commission's current posture that the relatively lower risks of low power testing justify elimination of offsite safety measures, since it concedes that there is an accident risk at low power serious enough that a research reactor (much smaller than a power reactor) needs a full emergency plan. Response. The premise for the comment that research reactors with power levels approximating those of commercial nuclear power plants operating at 5 percent of full power are required to have approved offsite emergency plans is incorrect. Rather than requiring a "full emergency plan" for research reactors, the Commission's 50-SC-64 regulations (10 CFR Part 50, Appendix E, 10 CFR 50.47(c), 10 CFR 50.54(q)) provide that emergency plan requirements will be determined on a case-by-case basis. In making this determination the guidance of NRC Regulatory Guide 2.6 and American National Standards Institute/ American Nuclear Society 15.16 is used. In accordance with this guidance, and based on the relatively small risks posed by typical research reactors, (i.e., less than 50 megawatts) emergency planning involving offsite state and local plans and public notification has not been required. The guidance does, however, provide for consideration of more extensive planning, including all or a portion of the requirements listed in section IV of 10 CFR Part 50, Appendix E for research reactors with power levels greater than 50 megawatts. This graded approach to required emergency planning is consistent with the current rule. Comment 13. The Atomic Energy Act prohibits authorization of low power testing prior to completion pf public hearings on all issues material to full power licensing. Response. This comment is more properly addressed to § 50.57(c), which provides for low power licenses and which is not being.amended here. That section provides that a hearing is required prior to low power on those contentions "relevant to the activity to be authorized"-that is, low power testing, as opposed to full power operation. Comment 14. The proposed rule was designed to allow 'the Seabrook facility to receive its low power license. The Commission should promulgate a rule to promote the public health and safety and not one designed to license a

specific facility.

The issue should be addressed in the pending Seabrook adjudication, not in a rulemaking. Response. In the proposed rule, the Commission stated that its attention was focused on the emergency planning requirements for low power testing because of an Appeal Board decision in the Seabrook operating license proceeding,ALAB-883.And,forthe near term, the only reasonably foreseeable effect of the rule change will be on the Seabrook low power application. But this does not make the use of rulemaking inappropriate. As the Commission explained, the rule change was proposed to correct a possible discrepancy between the language of the 1982 rule and the language of the Statement of Considerations which potentially affects all license applicants, not just the applicants for Seabrook. Also, the questions involved in the proposed rule are generic safety questions and the Commission preferred to obtain (and, in fact, did obtain) a broad spectrum of public comment, rather than just the comments of the litigants in the Seabrook proceeding. PART 50

STATEMENTS Of CONSIDERATION The Commission is free to address a generic issue generically, even if the rule ch~~ge may currently apply only to one fac1hty. See, e.g., Siegel v. Atomic E1:ergy Commission, 400 F.2d 778 (D.C. Cir. 1968). Also see Securities and Exchange Commission v:*chenery, 332 U.S.194, 202 (1947) (choice of how to proceed lies within the informed discretion of the agency). The rule is not intended to overrule Public Service Company of New Hampshire, et al. (Seabrook Station, Units 1 and 2), CLl-87-2, 25 NRC 267 or CLl-87-3, 25 NRC 875 (1987). Comment 15. Members of the public may need immediate medical attention in the event of an accident at low power. The new rule does not provide that arrangements for medical services will be in place for those off site. Response.

The purpose for the requirement in 10 CFR 50.47(b)(12) that arrangements for medical services be made was described in the "

SUMMARY

" section of the Commission's policy statement on medical services (51 FR 32904) dated September 17, 1986, as follows: The Nuclear Regulatory Commission

["NRG" or "Commission")

believes that 10 CFR 50.47(b](1Z) ("planning standard (b)(1Z)") requires pre-accident arrangements for medical services (beyond the maintenance of a list of treatment facilities) for individuals who might be severely exposed to dangerous levels of offsite radiation following an accident at a nuclear power plant. ;However, it is highly unlikely that :members of the general public would be exposed to dangerous levels of radiation following an accident at 'low power. Therefore, the safety premise for the full power requirement that arrangements 'be made for medical services does not apply to fuel loading or low power testing. Conclusion As indicated in the responses to the comments, the Commission has decided to proceed with the proposed rule change with some clarifications and modifications. The rule reconciles a discrepancy between the language of the Commission's 1982 emergency planning rule change and the language of the Supplemental Information and provides an interpretation of that rule which appears to be fully consistent with the Commission's goals and safety conclusions in 1982. The majority of the public, as expressed in the comments, supports the rule. The comments opposing the rule have given no sound reasons for the Commission to alter its basic course. Finding of No Significant Environmental Impact: Availability The Commission has determined that under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A 9f 10 CFR Part 51, this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. The envir~nmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW., Washington, DC 20555. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 ( 44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget, approval number 3150-0011. Regulatory Analysis The Commission has prepared a regulatory analysis for this final regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The analysis is available for inspection in the NRC Public Document Room, 2120 L Street NW., Washington, DC. Single copies of the analysis may be obtained from Michael T. Jamgochian, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555; Telephone (301) 492-3918. Regulatory Flexibility Certification This final rule will not have a significant impact on a substantial number of small entities. The final rule will reduce or at least postpone the burden on NRC licensees by reducing the process required before a low power license may be granted. Nuclear power plant licensees do not fall within the definition of small businesses in section 3 of the Small Business Act, 15 U.S.C. 632, the Small Business Size Standards of the Small Business Administration in 13 CFR Part 121, or the Commission's Size Standards published at 50 FR 50241 (Dec. 9, 1985). Therefore, in accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission here by certifies that the final rule will not have a significant economic impact on a substantial number of small entities and that, therefore, a regulatory flexibility analysis need not be prepared. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this final rule, and therefore, that a backfit analysis is not required for this final rule because these amendme_!l~S do not involve any provisions whi-ch would impose backfits as defined in 10 CFR 50.109(a)(1). 50-SC-65 List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting record. keeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the Commission is adopting the following amendments to Part 50. 53 FR42939 . Published 10/25/88 Effective 10/25/88 10 CFR Part 50 Licensee Announcements of Inspectors AGENCY: Nuclear Regulatory 'Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission is amending its regulations to ensure that the presence of NRC inspectors on nuclear power reactor sites is not widely communicated or broadcast to licensee and contractor personnel without the expressed request to do so by the inspector.

This change will allow the NRC inspectors, badged at the facility, to observe ongoing a~tivities as they are being performed without advanced notification of the inspection to licensee and contractor personnel. There is a need for this change because of the possible altering of attention and performance levels of a licensee and/ or its contractors when the licensee is aware of NRC surveillance. Past occurrences where site and/ or contractor personnel have been notified of NRC's presence on site have heightened concern in this area. EFFECTIVE DATE: October 25, 1988. FOR FURTHER INFORMATION ,coNTACT: George Barber, Office of Nuclear Reactor Regulation, U.S. Nuclear ,Regulatory Commission, Washington, pc 20555, Telephone (301) 4,92-1234. SUPPLEMENTARY INFORMATION: I. Background By clarifying the meaning and intent of 10 CFR 50.70(b)[3), this final rule should ensure that NRC inspectors will be granted immediate and unannounced access to licensee facilities so as to provide the inspector with unfettered access equivalent to that provided a regular plant employee following proper identification and compliance with applicable access control procedures. This rule provides that no access control measures or other means may be employed by the licensee or its PART 50 STATEMENTS OF CONSIDERATION contractors to intentionally give notice to other persons of the arrival and presence of an NRC inspector at a facility, unless the licensee is specifically requested to do so by the . NRC inspector. There have been . instances in the past at several facilities that compromised the ability of properly badged NRC inspectors to inspect and access, on an unannounced basis, activities related to the license or construction permit when licensee employees or contractor employees informed others at the facility of the presence of the NRC inspectors. This change to 10 CFR 50.70 is to clarify that NRC inspectors, badged at the facility, have immediate, unescorted access to ongoing activities as these activities are being performed without advanced notification of the inspection. This is especially important during non-normal business hours when operating personnel might assume NRC inspectors would not be on site. II. Summary of Comments On March 18, 1988, the Commission published in the Federal Register (53 FR 8924) a notice of proposed' rulemaking on "Licensee Announcement of Inspectors." The Commission invited the public to comment on the proposed rule and received six letters of comment by April 18, 1988 (the specified closing date for public comments). After April 18, 1988, 26 additional letters of comments were-received. All 3fietters of comments were considered in NRC's review of this final rule. The comments are discussed below. Comment. A majority of the commenters believed the rule was unnecessary and characterized it as being too broad and vague. They asserted that it: was redundant with current regulations; would lead to unfair and impractical enforcement; be impossible to implement; inhibit inspector assistance by plant personnel; limit the ability of facility management to perform their safety functions; promote lying among the facility staff; require formal training and . recordkeeping; and indicates a distrust of licensees. NRG Response. NRC does not agree with the comments, but to ensure that the intent of the rule is clear and focused, adds the following clarification of the rule. The intent of this rule is to prevent site and contractor personnel from widespread dissemination or broadcasting the presence of an NRC inspector. Broadcasting, as used here, is defined as unsolicited one-way communications. Implementing or enforcing this rule should be no more difficult than implementing or enforcing any rule that involves personnel performance. Adopting this rule does not indicate a predisposition on the part of the NRC that licensees are not acting properly. It is human nature for an individual to be more conscious of his or her performance when the individual realizes he or she is being observed. The NRC inspection program evaluates licensee performance on the basis of a sampling of its activities. It is critical .that the sampling portion of the licensee's activities that are relied upon for this evaluation be representative of its overall activities. Therefore, the rule is more prophylactic than proscriptive, although it does carry enforcement sanctions should it be violated. Recognizing the possibility of inadvertent communication of an inspector's presence, the NRC expects to reserve enforcement action for significant intentional violations of the rule. An honest response by an employee to an innocent inquiry that he/she just saw an NRC inspector is not within the proscriptive perimeter of the rule. Therefore, an employee would not be required to lie, in response to a question, about the presence of an NRC inspector. Based on this discussion, formalized training will not be necessary, and NRC Form 3 need not be modified to reflect this requirement. The NRC does not agree that this rule will prevent management from performing its safety functions. It should be noted the rule does not affect software security systems which monitor the presence of persons in certain areas. Such systems should provide the licensee with needed jinformation on space occupancy in the ;case of an emergency or evacuation. For those licensees who havl) these systems 1in place, or will put them in place, the lrule does not affect such systems. If a 1licensee were, however, to design or modify these systems (or use them) for the purpose of monitoring the NRC inspector's movements in order to alert other plant personnel of the inspector's whereabouts, those actions would violate the rule.

In sum, the licensee is prohibited from taking affirmative action which would compromise the NRC inspector's mission of gaining unfettered access to the plant and its various areas of interest to the inspector.

Comment. Some commenters expressed a concern that the rule could raise Constitutional questions under the First and Fourth Amendments. NRG Response. As discussed above, the purpose of the rule is to enhance the credibility of the inspection process. Inspections are specifically authorized under section 1610 of the Atomic Energy Act of 1954, as amended, 42 U.S.C. 2201(0). The regulation is narrowly drawn to achieve a legitimate governmental interest (effective NRC inspections) without infringing on an individual's right to express ideas and opinions on any subject. Thus, the 50-SC-66 regulation does not impermissively intrude upon freedom of speech protected by the First Amendment to the Cons ti tu lion. The regulation does not raise any significant Fourth Amendment considerations. The Atomic Energy Act creates a pervasive regulatory scheme that puts licensees on clear notice that they will be subject to inspection, and the granting of a license is conditioned on consent to reasonable inspections. Thus, NRC inspections of licensees' premises, activities and records do not require a warrant under the Fourth Amendment. United States Nuclear Regulatory Commission vs Radiation Technology, Inc., 519 F. Supp. 1266, 1288-91 (D.N.J. 1981): Union Electric Co. (Callaway Plant, Units 1 & 2), ALAB-527, 9 NRC 126, 139-41 (1979). The new regulation is a reasonable exercise of the Commission's inspection authority. Inspectors will continue to identify themselves and comply with other .reasonable access control measures and, as always, inspections will be conducted for purposes authorized under the Atomic Energy Act and the Energy Reorganization Act. The regulation does not run afoul of the Fourth Amendment to the Constitution. Comment. A number of commenters suggested that the rule be implemented only by written request of the NRC inspector. NRG Response. NRC rejects the suggestion. With this suggested modification, the rule would only apply to those individuals who had been given notice of the NRC inspector's presence on site. If implemented, this suggestion would defeat the intent of the rule .. Environmental Impact: Categorical Exclusion The NRC has determined that this change is the type of action described in categorical exclusion 10 CFR 51.22(c)(2). Therefore neither an environmental impact statement nor an environmental assessment has been prepared for this final rule. Paperwork Reduction Act Statement The final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval number 3150--0011. Regulatory Analysis This final rule will have no significant impact on state and local governments and geographical regions. It may have a significant impact on health, safety, and the environment, but only in the sense of preventing adverse impacts on health, safety, and the environment through more effective inspections. The rule will PART 50 STATEMENTS OF CONSIDERATION make it clear that NRC inspectors are to have a realistic picture of the actual conditions at a site during the inspection process and, therefore. be better able to identify potentially dangerous conditions and/or practices for corrective action and to ensure that licensees comply with laws, regulations, and orders administered by the NRC. This constitutes the regulatory analysis for this final rule. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act, 5 U.S.C. 605!b), the Commission certifies thut this final rule does not have a significant economic impact on a substantial number of small entities. The final rule applies only to licensees authorized to construct or operate nuclear power reactors, who are not small business entities within the meaning of the act or implementing regulations. Therefore, a regulatory flexibility analysis has not heen prepared. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does apply to this final rule. The backfit analysis for announcement of inspectors rule in accordance with each of the factors specified in 10 CFR 50.109(a)(4)(ii)(c) is as follows: (1) This rule provides that no means may be employed by the licensee or its contractors to intentionally give notice to other persons of the arrival and presence of a NRC inspector at a facility, unless the licensee is specificaily requested to do so by the NRC inspector. (2) The licensee will have to communicate the requirements of this rule to its personnel and to contractor personnel working at its site. (3) The purpose of this rule is to enhance the credibility of the inspection process. By requiring that the presence of NRC inspectors (either resident or off site) is not announced, the NRC. public and licensees will have more confidence that the activities inspectors are witnessing are representative of licensee performance. Ensuring that NRC inspectors are witnessing representative licensee performance could substantially increase the likelihood that NRC inspectors will discover unsafe or potentially unsafe practices, bring about corrective actions and thereby lower the risk of accidents occurring which could lead to the accidental off-site release of radioactive material. It is not possible, without before and . after data, to quantitatively evaluate the benefits of implementing this rule. Still. a recent significant enforcement action concerning licensee employee's inattention to dutv demonstrates the premise advanced in the above paragraph. In this enforcement action. over 30 licensee personnel, both management and staff were cited for inattention to duty. The primary concern was sleeping on watch. It is not difficult to envision accidents that could occur because of this type of licensee performance. Coupling inattention to duty with equipment failure adds a new dimension to the risk of accidents occurring which could lead to the off-site release of radioactive material. In the enforcement action mentioned above, had the licensee announced the presence of the NRC inspector, the inattention to duty would have gone unnoticed. It should be noted that the licensee facility where this incident occurred did, on one past occasion, announce the presence of NRC inspectors.

(4} Not appropriate.

There is no radiological exposure of faci'ity employees resulting from the rule's implementation. (5) Very minor costs are associated with the rule's implementation. There are no training requirements or record keeping requirements associated with this rule. The only cost to the licensee would be communicating this ru!e to its employees and contractors. (6) Not appropriate. There is no potential safety impact of changes in plant or operational complexity associated with this rule. (7) Not appropriate. There is no resource burden on the NRC from the implementation of this rule. (8) Not appropriate. There is no potential impact of differences in facility type. design or age on the relevancy and practicality of the proposed backfit. (9) The proposed backfit is final. Conclusion Based on the above analysis, the Commission concludes that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from this backfit and that the direct and indirect costs of implementation for facilities are justified in view of this increased protection. List of Subjects in 10 CFR Part 50 Antitrust, Classified information. Fire protection. Incorpora lion by reference. Intergovernmental relations. Nuclear power plants and reactors. Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as 2menc!ed, the Energy Reorganization Act cf 1974, as amended, and 5 U.S.C. 552 and 553. the NRC is to adopt the following amendment to 10 CFR Part 50. 50-SC-67 53 FR 43419 Published 10/27 /88 Effective 10/27 /88 Relocation of NRC's Public Document Room; Other Minor Nomenclature Changes See Part 1 Statements of Consideration 53 FR 45890 Published 11/15/88 Effective 11/15/88 10 CFR Part 50 Alternative Method for Leaka!~e Rate Testing AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations to modify the requirements applicable to the leakage testing of containments of light-water-cooled

nuclear power plants. The rule explicitly permits the use of a new statistical data analysis technique that the NRC considers to be an acoeptable method of calculating containment leakage rates in addition to previously acceptable methods. EFFECTIVE DATE: November 15, 1988. FOR FURTHER INFORMATION CC:*NTACT:

Mr. E. Gunter Arndt, Office of Nuclear Regulatory Research, U.S. Nudear Regulatory Commission, Washington, DC 20555, telephone 301-492-3814. SUPPLEMENTARY INFORMATIO~I: Background The Nuclear Regulatory Commission is amending 10 CFR Part 50, Appendix J, "Primary Containment Leakage Testing for Water-Cooled Power Reac:tors," to explicitly permit the use of the Mass Point statistical data analysis method for calculating containment le!akage rates. The Mass Point method involves calculation of the air mass at a series of points in time and the plotting of mass against time. A linear regression line is. plotted through the mass-time points using a least squares fit. The slope of this line is divided by the int1ircept of this line, and the result is multiplied by. an appropriate constant to obtain the calculated leakage rate. This Mass Point method was incorporated in a newer ANB.I standard, ANSI/ ANS-56.8-1981, "Containment System Leakage Testing Requirements" (revised 1987) and in fact ha11 been accepted by the NRC staff as an improved alternative method of calculating containment leakage rates . However, it was recently rec.ognized by the NRC staff that a strict interpretation of the specific wording of Appendix' J, III.A.3, by referencing only tl1e older ANSI standard, would preclude use of PART 50 STATEMENTS OF CONSIDERATION the newer, improved method. To alleviate this restriction on the use of an improved alternative methodology, it is necessary to clarify the language in Section III.A.3 to explicitly permit the use of the newer Mass Point method in addition to the earlier methods covered by ANSI N45.4-1972.1 A proposed rule concerning the addition of the Mass Point method was

published for comment on February 29, 1988 (53 FR 5985). The complete history and background for the proposed action were discussed in detail in the * ;SUPPLEMENTARY INFORMATION section !which accompanied the proposed rule. \The effect of this amendment will be to ,permit licensees to use the Mass Point analysis as an altern~tive to the Total Time and Point-to-Point analyses incorporated by reference into Appendix J by ANSI N45.4-1972.

The final rule is identical to the proposed rule published for comment, and adds the following wofds to Section 111.A.3: . In addition to the Total Time and Point methods described in that standard, the Mass Point method, when used with a test duration of at least 24 hours, Is an acceptable method lo use to calculate leakage rates. A typical description of the Mass Point method 'can be found in the American National . standard, ANSI/ ANS 56.8-1987, "Containment System Leakage Testing Requirements", January 20, 1987. *. Also, as in the proposed rule, in order to allow a change. in the methods now permitted, the final rule deletes the following sentence from Section III.A.3 of Appendix J: The method chosen for the initial test shall normally be used for the periodic tests. The NRC believes the wording of the revision as published for comment, and as finally amended, clearly and accurately represents the NRC's position. All comments have been reviewed. In spite of the objections raised in the comments to the wording or content of the proposed rule, the wording in the final rule is identical to that published for public comment. A Public Comment Resolution Memo has been prepared and sent to all who commented. It is available for inspection and copying at the NRC's Public *Document Room at 2120 L Street, Washington, DC. The memo addresses

in more detail the NRC's decision to 1 ANSI N45.4-1972, "Leakage Rate Testing of Containment Structures*

for Nuclear Reactors" (dated March 16, 1972). Incorporation of ANSI N45.4-1872 by reference was approved by the Director of the Federel Register on October 30, 1972. Copies of this standard, ea well ea ANSI/ ANS-56.8-1987, "Containment System Leokoge Testing Requirements" (doted January 20, 1987] inay be obtained from the American Nuclear Society, 555 ;North Kensington Avenue, La Grange Perk, IL 60525. !A copy of each of these atenderds*is eveileble for * 'inspection et the Commission's Public Document Room et 2120 L Street NW., WBBhington, DC.* keep the wording the same as in the proposed revision. A brief summary of the comments received is set out in the following paragraphs.

Summary of Public Comments Twenty-one comtnent letters were received.

In general, three principal comments were presented. First, all conimentors supported the addition of the Mass Point analysis to the list of acceptable analysis methods. Second, all but two commentors objected to requiring a 24-hour test duration in combination with the Mass Point method. As noted in the proposed rule of February 29, 1988, the position stated in the text is consistent with the position that has been taken by the NRC staff when granting exemption requests on this matter. In particular, the description of the Mass Point method and its coupling with a test duration of at least ;24 hours reflect prior exemption

approvals and maintain necessary consistency.

i The intent of this limited amendment is not to endorse ANS 56.8, nor to propose any of the changes and updating represented by the October 29, 1986 proposed general revision to Appendix J (51 FR 39538). Instead, this action does no more than eliminate the need for exemptions to the existing rule by permitting the use of a statistical method that has been generally accepted for several years. This revision makes available to all reactor licensees 'the use of the Mase Point method for 24-hour tests. Inclusion of the 24-hour duration is considered necessary because a considerable difference of opinion exists as to what is a sufficient test duration. Until an acceptable set of alternative technical criteria is-developed to replace the 24-hour duration criterion, the NRC staff intends to continue the 24-hour criterion. Some alternative technical

criteria were presented for public ireview and comment in proposed ,regulatory guide MS 021-5, "Containment System Leakage Testing," 2 on October 28, 1986 .. These

criteria and others proposed are still being evaluated in order to determine what is an appropriate set of test termination criteria to include in the final regulatory guide. * : Third, one objection was raised to the !degree of flexibility permitted by the :proposed wording in defining the Mass 'Point method. 1 A free single copy of draft regulatory guide MS 021-5, to t!ie extent of supply. may be obtained by wriling to the Distribution Seclion. Document Control Branch, Division of Information Support Services, U.S. Nuclear Regulatory Commission.

Washington, DC 20555. A copy ie also uveilable for inspection, or copying for e fee, In the NRC Public Docwnonl Room, Z120 L Street NW .. Weahington. DC. * .. 50-SC-68

If this comment were to be followed, the effect would be to incorporate by reference into 10 CFR Part 50 the exact Mass Point analysis as defined in ANSI/ ANS 56.S-1987, along with the portions of that standard that are relevant to setting the conditions of use of this analysis.

The existence of proposed regulatory guide MS 021-5 demonstrates that this degree of compatibility between ANS 56.8 and a position acceptable to the NRC staff does not exist. Therefore, in order to define in detail a Mass Point analysis that would be acceptable to the NRC staff, such an Jncorporation by reference would also !!!_ave to contain the portions of proposed [~egulatory Guide MS 021-5 that modify l the ANS 56.8 definition and use of the Mass Point analysis. This approach would be undesirably cumbersome, !inflexible, and restrictive in the ability Ito keep the legally acceptable Mass 1 Point analysis current with any future 1 improvements, simplifications, or ichanges in the state-of-the-art of !statistical reduction of test data to a !leakage rate. , An alternative perhaps could be to :simply state that the Mass Point method ibe defined in a manner acceptable to the :NRC staff, and leave that definition to !the finalization of proposed regulatory iguide MS 021-5. However, this would :probably be a less acceptable /'alternative because it would be more flexible than the current wording and *would depend heavily on the as yet . uniseued regulatory guide. Finally, as noted in the proposed rule of February 29, 1988, the wording was intentionally made instructive but flexible in the event that the proposed general revision to Appendix J and its ,proposed associated regulatory guide [are not issued as final documents. 1Should that happen, then a clear need 'would exist for some flexibility in the ;ability of Appendix J to keep up with *changes to ANS 56,8 and potential

future modifications to the Mass Point ;analysis.

Effective Date Since the amendment set forth below is intended lo provide relief from, rather than to impose, restrictions currently in ,effect, the Commission.is, pursuant to 5 :U.S.C. 553(d)(1), making the final rule ;effective on November 15, 1988 without (the customary 30-day waiting period. [Environmental Impact: Categorical 1 Exclusion The Commission has determined that this rule is the type of action described in the categorical exclusion in 10 CFR 51.22(c)[2).

Therefore, neither an environmental impact stetemc*nt nor an environmental assessment have been

prepared for this rule.

PART 50 STATEMENTS OF CONSIDERATION Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 *et seq.). Existing requirements were approved under the Office of Management and Budget approval number 3150--0011. Regulatory Analysis The Commission hae prepared a regulatory analysie on this regulation. The analysis examines the costs and benefits of the.alternatives considered by the Commission. Interested persons may examine a copy of the regulatory analysis at the NRC Public Document Room, 2120 L Street NW., Washington, DC. Regulatory Flexibility Certification As req~ired by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b ), the Commission certifies that this rule does not have a significant economic impact on a substantial number of small entities. This rule eff~cts only the licensing and operation of nuclear power plants. The companies that own .these plants do not fall within the scope *of the definition of"small entities" set* *forth in the Regulatory Flexibility Act or , the Small Business Size Standards set . out in the regulations issued by the ,Small Business Administration at 13 , CFR Part 121. !Backfit Analysis

The NRC has determined that a 'backfit analysis is not required for this rule because, although the rule is applicable to all current or future :operating nuclear power plants, the ;provisions of the rule codify and permit the continuation of a previously accepted practice.

This action will not encumber those using this accepted practice with the a,dded burden of seeking exemptions to the existing rule. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear* power plants and reactors, Penalty, Radiation protection, Reactor siting criteria. Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, es amended, and 5 U.S.C. 552 and 553, the NRC is adopting 'the following

amendment to 10 CFR Part 50. 54 FR7178 Published 2/17 /89 , Effective 3/20/89 10 CFR Part 50 Licensee Action During National Security Emergencw AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission is amending its regulations to allow a licensee to take action that departs from approved technical specifications in a national security emergency.

The amendment is necessary to specify in the regulations

that for a national security emergency a licensee is permitted to take a needed action although it may deviate from technical specifications.

This amendment will allow the licensee to implement national security objectives as designated by the national command authority through the NRC. EFFECTIVE DATE: March 20, 1989. FOR FURTHER INFORMATION CONTACT: Joan Aron, Office for Analysis and *

Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone (301) 492-9001 . SUPPLEMENTARY INFORMATION:

Background On April 1, 1983, the Commission published in the Federal Register (48 FR 13966), a final rule that set out § 50.54 of 10 CFR entitled, "Conditions of Licenses," that contains a provision permitting a license to take reasonable action that departs from a license condition or a technical specification (contained in a license issued under this part) in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and technical specifications that-can provide adequate or equivalent protection is immediately apparent. However, thie provision does not apply to a national security emergency. The final rule in this notice allows a licensee to take action that departs from approved technical specifications in a national security emergency when this action is immediately needed to implement national security objectives as designated by the national command authority through the NRC and no action consistent with license conditions and. technical specifications that can meet national security objectives is immediately apparent. The rule was published for comment on July 19, 1988 (53 FR 27174). A thirty-day comment period expired on August 18, 1988. Comments were received from four respondents. 50-SC-69 Summary of Public Comments A summary of the public comments follows: (1) Flexibility. One commenter, writing on behalf of the nuclear power industry, supported the proposetl amendment, stating that it provides licensees with desirable regulatory authority and operational flexibility to accommodate exigencies that may be associated

with a declared national emergency.

(2) Need for the amendment. One commenter questioned the need for the proposed amendment. claiming that § § 2.204, 50.54(x), and 50.103 offer more than enough authority to permit a licensee to deviate from technical specifications during a national emergency when such action is needed to implement na~ional security objectives The fmal rule does not" duplicate existing requirements. Sectio11 2.204 deals with the Commission's ability to issue an order for modification of a licensee and § 50.103 deals with the Commission's ability to suspend a license, recapture special nuclear material or order the operation of a facility during a state of war or national emergency. Paragraph (x).of I 50.54 grants authority to nuclear power plant licensees to take reasonable action that departs from a licenee condition or a technical specification in an emergency when such action is necessary to protect public health and safety and no action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is . immediately apparent. The amended rule provides the same flexibility to licensees* but for the purpose of attaining national security objectivee during a declared national security emergency. (3) Implementation. One commenter questioned the lack of di~cussion relative to implementation requirements and suggested a delay in issuing the final rule until proper implementation guidance can be formulated. The final rule provides a basis for the licensee to take action in accordance with governmental directives in a national security emergency, when this action is immediately needed to implement national security objectives as designated by the national command authority through the NRC and no action consistent with license conditions and technical specifications that can meet

national security objectives is immediately apparent.

Guidance concerning implementation will be formulated by the appropriate federal agencies and will be issued some time in the future. (4) Definitiun of a "national security emergency." One commenter requested definition of a "national security emergency." PART 50 STATEMENTS OF CONSIDERATION NRG Manual Chapter 0601, Continuity of Government Program, approved June 30, 1988, defines a national security emergency as "any occurrence, including nuclear attack, a national disaster, or other emergency, which seriously degrades or seriously threatens the national security of the United States or has been declared by the Congress." A national security

emergency is established by a law enacted by the Congress or by an order or directive issued by the President pursuant to statutes or the Constitution of the United States. (5) Reporting requirements.

One commenter suggested that § 50.73(a)(2)(c) be revised to include the reporting requirements of the amended § 50.54(dd). At present, there is no reporting requirement include in§ 50.54(dd) and none is comternplated for the immediate future. Thus, there is no need to revise 10 CFR 50.73(a)(2)(c). .Environmental Impact: Categorical Exclusion . . . The NRC has determined that this final regulation is the type of action described in categorical exclusion 10 CFR 51.22(c)(2). Therefore, neither an environmental impact statement nor-an environmental assessment has been prepared for this proposed regulation. Papenvork Reduction Act Statement* This final rule does not contain a new or amended information collection requirement subject to The Paperwork Reduction Act of19BO (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval number 3150-0011 Regulatory Analysis The Commission previously has granted authority pursuant to 10 CFR .50.54(x) to nuclear power reactor

licensees to take reasonable action that departs from a license condition or a technical specification in an emergency when the action is immediately necessary to protect the public health and safety and no action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent.

This final rule will provide the same flexibility to licensees for the purpose of attaining national security objectives in accordance with governmental directives during a declared national security emergency. The final rule docs not significantly impact state and local governments and geographic locations: health, safety, and the environment: or costs to licensees, the NRG, or other Federal agencies. The final rule is in the interest of the common defense and security of the United States because it would facilitate operation of nuclear facilities in a national security emergency during which some deviation from facility technical specifications may be appropriate. This constitutes the regulatory analysis for this final rule. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980, 5 U .S.C. 605(b ), the Commission certifies that this rule will not have a significant economic impact upon a substantial number of small entities. The final rule affects only licensing and operation of nuclear

power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121. Because these companies are dominant in their service areas, this rule does not fall within purview of the Act.. Backlit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this rule and, therefore, that a backfit analysis is not required for this rule, because these amendments do not involve any provisions which would impose backfits as defined in 10 CFR 50.109(a)(l).
List of Subjects in 10 CFR Part 50 Antitrust, Classified Information, Fire Protection, Incorporation by Reference, Intergovernmental Relations, Nuclear power plants and reactor, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.
* , For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is adopting the following amendment to 10 CFR Part 50. . 54 FR 11161 Published 3/17 /89 Effective 3/17 /89 10 CFR Part 50 Extension of Time for the Implementation of the Decontamination Priority and Trusteeship Provisions of Property Insurance Requirements AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule. SUMP.,ARY: The Nuclear Regulutory Commission is amending the implementation schedule to change the effective date for the stabilization and decontamination priority and 50-SC-70 trusteeship provisions of its property insurance regulations. This delay in implementation is necessary because the insurers that offer property insurance for power reactors have informed the Commission that they will be unable to include the stabilization and decontamination priority and trusteeship provisions in their insurance policies within the date required by current regulations. Concurrently, the extension of the effective date of the rule allows the NRC to consider three petitions for rulemaking that propose changes to improve the efficacy of the NRC's stabilization and decontamination priority and trusteeship provisions. EFFECTIVE DATE: March 17, 1989. FOR FURTHER INFORMATION CONTACT: Robert S. Wood, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-1280. SUPPLEMENTARY INFORMATION: I. Background On September 19, 1988, the Commission published a proposed rule in the Federal Register (53 FR 36338) that proposed to amend the implementation schedule for the stabilization and decontamination priority and trusteeship provisions of its property insurance regulations contained in 10 CFR 50.54(w)(5)(i) to change the effective date from October 4, 1988, to April 4, 1990. As explained in the proposed rule, this implementation. schedule was part of a final rulemaking published on August 5, 1987 (52 FR 28963) which, for the first time, explicitly required power reactor licensees to purchase on-site property damage insurance policies in which $1.6 billion of the proceeds from these policies are to be used first for stabilization of a reactor after an accident end then for

decontamination of the facility before any other purpose. The 1987 final rule also required that these insurance proceeds be paid to an impartial trustee who would be required to disburse * :funds according to the stabilization and decontamination priority.

Subsequent to the publication of the :1987 final rule, the NRC was informed 'that the trusteeship provision and, to a '.lesser extent, the stabilization and :decontamination priority provisions of that rule were sufficiently complex and ,problematic that the insurers were :unable to incorporate such provisions in 'their policies by the required October 4, 1988, date. As explained in the.September 19, 1988, proposed rule, the insurers and their counsel gave two reasons why they were unable to comply with the date specified in the final rule for adding the stabilization and decontamination priority and trusteeship provisions. First, PART 50 STATEMENTS OF CONSIDERATION with respect to the trusteeship provision, counsel for insurers assured the NRC

staff that they had made a good-faith effort to obtain trustees, but were unsuccessful.

They believed the reason for their lack of success was the potential trustees' conflicts of interest end reluctance to assume, on the one hand, responsibility for disbursing potentially over $1 billion in insurance proceeds and the resulting exposure to possible litigation for wrongful disl:iursement, while, on the other hand, being eligible for only modest fees for this service. A second reason insurers gave for being unable to comply with the

effective date of the 1987 rule was essentially logistical.

As a contract, an insurance policy can only *be modified *

with the consent of all affected parties. Because the Conunission's mandated stabilization and decontamination priority and trusteeship provisions adversely affect the current rights under the policy of the bondholders' trustee, it is unlikely that policies could be legally changed before the end of the policy years. Because of insurers' policy renewal procedures end the policy anniversaries, these dates would have fallen after the effective date specified in the rule. II. Summary of Comments, NRC Response aod Conclusions By the end of the comment period on October 19, 1988, the NRC received five comments.

One of these was misdirected to this rulemaking. (Conunent 1 was directed to rescinding § 50.54 (x) aod (y) rather than I 50.54{w).) The remaining four either ,supported the proposed rulemeking

(comment

4) or sought clarification of the applicability of 10 CFR 50.54(w){5)(i) to specific licensees while the ;rulemaking was being considered (comments 2, 3, end 5). In addition, comment 4 suggested that, rather than provide a date certain in the rule, the stabilization and decontamination ,priority and trusteeship provisions of § 50.54{w) (3) and (4) be suspended indefinitely pending completion of consideration of three petitions for rulemaking (PRM-50-5~.

PRM-50-51A, and PRM-50-518; 53 FR 36335, September 19, 1988). The only issue of any controversy raised by commenters was whether the extension of time for implementing the stabilization and decontamination priority and trusteeship provisio.ns of § 50.54(w) should be for a date certain (i.e., April 4, 1990) or indefinite until consideration of the above-cited petitions for rulemaking has been completed. The Commission continues to believe that an 18 month extension is more appropriate than an open-ended extension. First as commenter 4 acknowledged, 18 months should be sufficient to complete consideration of the issues raised in the three petitions for rulemaking. Second if 18 months is insufficient, the Conunission can act to further extend the implementation date. Finally, the Commission imposed the

stabilization and decontamination priority and trusteeship provisions for valid health and safety reasons. Indefinitely deferring these provisions prior to a substantive reevaluation of their efficacy could conflict with the Commisiiion's mandate to protect health and safety. The proposed rule analyzed why an 18 month delay would have minimal health and safety impact. The NRC believes that analysis remains valid. For the foregoing reasons, the Commission concludes that a delay froni October 4, 1988, to April 4, 1990, in the implementation schedule of the stabilization and decontamination priority and trusteeship provisions is .justified and is amending 10 CFR 50.54(w)(5)(i) accordingly.

Because the amendment to § 50.54{w)(5)(i) relates solely to extending the time for implementing the stabilization and decontamination priority and trusteeship provisions of the property insurance rule and therefore provides relief from restrictions under regulations currently in effect, the Commission has found that good cause* exists for making the rule effective on the date of publication in the Federal

Register without the customary 30 day waiting period. III. Environmental hnpact: Categorical Exclusion The NRC has determined that this rule constitutes a minor corrective amendment that does not substantially modify existing regulations and, therefore, is the type of action eligible for categorical exclusion under 10 CFR 51.22{c)(2).

Accordingly, neither an environmental impact statement nor an environmental assessment is required. IV. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork . Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval nwnber 3160--0011, V. Regulatory Analysis On August 5, 1987, the NRC published in the Federal Register a final rule amending 10 CFR 50.54(w). The rule increased the amount of on-site property damage insurance required to be carried by NRC's power reactor licensees. The rule also required these licensees to obtain by October 4, 1988, insurance policies that prioritized insurance proceeds for stabilization and 50-SC-71 decontamination after an accident and provided for payment of proceeds to an independent trustee who would disburse funds for decontamination and cleanup before any other purpose. Subsequent to publication of the August 5, 1987 rule, the NRC was informed by insurers who offer nuclear property insurance that the decontamination priority and trusteeship provisions would not be able to be incorporated into the policies by the time required in the 1987 rule. In petitions for rulemaking, insurers' representatives further stated that the trusteeship provisions might actually have an effect counter to their intended purpose by delaying claims payment and thus possibly the. cleanup process. By deferring implementation of these provisions by 18 months, the Commission is allowing sufficient time either to secure the required coverage or to reconsider the mechanism by which accident cleanup funds may be assured to be used for their intE;nded purpose. Even without formal stabilization and decontamination priority and trusteeship provisions, NRC has authority to take appropriate enforcement action to order cleanup in the unlikely event of an accident. Thus, this rule will not have a significant impact on public health and safety. Furthermore, this rule will not have significant impacts on state and local governments and geographical regions; on the environment; or, create substantial costs to licensees, the NRC, or other Federal agencies. The foregoing discussion constitutes the regulatory analysis for this rule. VI. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605[b ), the Commission certifies that this final rule does not have a significant economic impact on a substantial nwnber of small entities. The final rule affects only those companies licensed lo operate nuclear powerplants. The companies that own these plants do not fall within the scope of the definition of. "small entities" set {orth in the Regulatory Flexibility Act or the Small Business Size Standa.rds set out in regulations issued by the Small Business Administration at 13 CFR Part 121. . VII. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this rule because this rule would not impose a backfit as defined in § 50.109[a)(1). Therefore, a backfit analysis is not required for this rule. List of Subjects In 10 CFR Part 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear powerplanto and reactors, Penalty, Radiation protection, Reactor siting PART 50 STATEMENTS OF CONSIDERATION criteria, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and Ii U.S.C. 552 and 553, the NRC is adopting the following amendment to 10 CFR Part 50. 54 FR 13361 Published 4/3/89 Effective 5/3/89 10 CFR Part 50 RIN: 3150-AC72 Flow Control Conditions for the Standby Liquid Control System In Boiling Water Reactors AGENCY: Nuclear Regulatory*

Commission. ACTION: Final rule ..

SUMMARY

The NRC is amending its regulations to set forth conditions and considerations for determining reactivity control capacity for boiling waler. reactor standby liquid control systems. The changes are necessary to clarify the* existing regula lion.

EFFECTIVE DATE: May 3, 1989. FOR FURTHER INFORMATION CONTACT: William R. Pearson, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3764.

SUPPLEMENTARY INFORMATION: On Monday, October 24, 1968, the Commission published in the Federal Register (53 FR 41607) a proposed rule, entiiled "Flow Control Conditions for the Standby Liquid Control System in Boiling Water Reactors," that proposed amendments to 10 CFR 50.62. Interested parties were invited to submit written .... comments within a 60-day comment period, which ended on December 23, 1988. One comment was received, which agreed with the proposed clarification. No change to the proposed rule was suggested in the public comment received by the NRC. The Commission believes that the proposed rule adequately clarifies reactivity control conditions for boiling water reactor standby liquid control systems (SLCS), thus, a final rule is being issued adopting the proposed rule without modification. Environmental Impact: Categorical Exclusion The NRC has determined that this rule is the type of action described as a categorical ei:cc;:lusion in 10 CFR 51.22[c)(2). Thus, neither an environmental impact statement nor an environmental assessment has been prepared. Paperwork Reduction Act Statement This rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the O!Jice of Management and Budget under control number 3150-0011. Regulatory Analysis Because this rule is of a clarifying nature and does not substantially change existing regulatory requirernen ts, the regulatory analysis prepared for the *final rule entitled "Reduction of Risk from Anticipated Transients Without Scram [ATWS) Events for Cooled Nuclear Power Plants," . published June 26, 1984 (49 FR 26036) is still valid for this rule. The analysis is available for inspection in the Public Document Room, 2120 L Street NW., -Washington, DC, Lower Level. Sii1gle copies of the analysis may be obtained from William R. Pearson, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3764. Regulatory Flexibility Act Certification In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 605[b)), the Commission certifies that this rule will not have a significant economic impact on a substantial number of small entities and that therefore a regulatory flexibility analysis is not needed. This rulemaking action affects only licensees that own and operate nuclear utilization facilities licensed under sections 103 and 104 of the Atomic Energy Act of 1954, as amended. These licensees do not fall within the definition of small businesses set forth in section 3 of the Small Business Act (15 U.S.C. 632) or within the Small Business Size Standards set forth in the regulations. issued for the Small Business Administration at 13 CFR P.art 121. Backlit Analysis The NRC has determined that the backfit rule,.10 CFR 50.109, does not apply to this rule, and therefore, that a backfit analysis is not required, because these amendments do not involve any provisions which impose backfits as defined in 10 CFR 50.109(a)(l). List of Su_bjects in 10 CFR Part 50 Antitrust, Classified information, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and record.keeping requirements. 50-SC-72 For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendment to 10 CFR Part 50. 54 FR 15372 Published 4/18/89 Effective 5/18/89 Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Reactors See Part 52 Statements of Consideration 54 FR 50735 Published 12/11/89. Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Reactors; Correction See Part 52 Statements of Consideration 55 FR 10397 Published 3/21/90. Effective 4/20/90 Preserving the Free Flow of Information to the Commission See Part 30 Statements of Consideration 55 FR 12163 Published 4/2/90. Effective 4/2/90 10 CFR Part 50 RIN 3150-AD19 Stabilization and Decontamination Priority and Trusteeship Provisions AGENCY: Nuclear Regulatory Commission. ACTION: Final rule. ,

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending the . provisions of its property/

accident recovery insurance regulations applicable to commercial power reactor licensees. The changes (1) clarify the scope and timing of the stabilization and decontamination processes after an accident at a covered reactor; (2) specify that the insurance is required to ensure that commercial power reactor licensees will have sufficient funds to carry out their obligation to clean up and decontaminate after an accident; and (3) eliminate the requirement that insurance proceeds after an accident are paid to an independent trustee. This rule responds to issues raised in three petitions for rulemaking.

EFFECTIVE DATE: April 2, 1990. FOR FURTHER INFORMATION CONTACT: Robert S. Wood, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone (301) 492-1280.

PART 50 STATEMENTS OF CONSIDERATION SUPPLEMENTARY INFORMATION: I. Background After the receipt of three petitions for rulemaking assigned Docket Nos. (PRM50-51) from Linda S. Stein, Steptoe & Johnson, counsel to American Nuclear Insurers and MAERP Reinsurance Association (ANI/MAERP); (PRM-50--51A) from J.B. Knotts, Jr., Bishop, Cook, Purcell & Reynolds, counsel to the Edison Electric Institute (EEI), the Nuclear Utility Management and Resources Council (NUMARC) and several power plant licensees; and (PRM-50--51B) from Peter D. Lederer, Baker & McKenzie, counsel to Nuclear Mutual Limited and Nuclear Electric Insurance Limited (NML and NEIL-II), the Commission published a notice of receipt requesting public comment on the petitions in the Federal Register of September 19, 1988 (53 FR 36335). The petitions were filed in response to a final rule on changes in property insurance requirements published by the Commission on August 5, 1987 (52 FR 28963). These petitions sought (1) clarification of the scope and timing of the stabilization process after an accident at a covered reactor; (2) clarification of the procedures by which the NRC determines and approves expenditures of funds necessary for decontamination and cleanup, and clarification of how such procedures affect both insurer's needs to secure appropriate proofs of loss and when payments may be made for non-cleanup purposes; (3) a change in the terminology of the required insurance from "property" insurance to "decontamination liability" insurance so as to better forestall claims on insurance proceeds by a licensee's bondholders; and (4) rescission of the provision that proceeds of the required insurance are to be paid to an independent trustee, who will disburse the proceeds for decontamination and cleanup of the facility before any other purpose. Four comments were received on the petitions for rulemaking, all of which supported the amendments recommended in the petitions. The Commission responded to the comments received on the petitions in a proposed rule published on November 6, 1989 (54 FR 46624). This final rule, in effect, grants these petitions and completes NRC action in response to PRMs 50--51, 50--51A, and 50--51B. II. Analysis of and Response to Comments On November 6, 1989, the Commission published in the Federal Register (54 FR 46624) a proposed rule to amend 10 CFR 50.54(w). The rule was developed in response to the three petitions for rulemaking discussed above. As of January 18, 1990, the NRC received seven comments on the proposed rule. Six comments came-from electric utilities or their representatives. One comment came from the Association of the Bar of the City of New York. All commenters essentially supported the Commission's rulemaking, although some took issue with specific provisions. Two aspects of the proposed rule, in particular, were opposed by several commenters. The first is the statement in the preamble of the proposed rule that the NRC retains the authority to require an independent trustee to hold and to disburse insurance proceeds in individual cases, if warranted. Further, the NRC expressed its intention that if the NRC obtains authority to receive and retain insurance proceeds itself, it will consider whether to exercise this authority and the best method of implementing the authority (54 FR 46624, at p. 46627). In support of their objections, the commenters refer to the case cited in the proposed rule-In re Smith-Douglass (Nos. 87-1683, -1684 (4th Circuit, September 6, 1988))-and take issue with the Commission's conclusion that the decision in this case justifies future reimposition of a trusteeship requirement. The Commission continues to believe that uncertainties remain with respect to interpretation of this and similar decisions. Consequently, if the Commission concludes that future conditions warrant reinstitution of the trusteeship requirement, it will reopen this issue for reconsideration. If the Commission does make such a decision, however, it will provide ample opportunity for public comment at that time. Because no provision of this final rule is affected by these concerns, the Commission proposes no further discussion or action at this time. The second issue raised by several commenters concerns how the Commission might address possible

increases in accident cleanup costs resulting from inflation or other factors. Commenters expressed the opinion that there is insufficient experience from which to develop an effective formula to estimate future accident cleanup costs. Furthermore, such a formula would not be able to account for advances in technology that might reduce future costs. Commenters suggest that rather than use a formula to estimate future cleanup costs and consequently establish future insurance requirements, the NRC reevaluate accident cleanup costs every 3 to 5 years by conducting 50-SC-73 specific studies using then-current technology.

One commenter recommended using a simple formula based on the Consumer Price Index to estimate future cleanup costs. Since publication of the proposed rule, the NRC's contractor has updated NUREG/CR-2601 1 (hereinafter cited as Addendum 1) which provided the basis for the $1.06 billion in insurance currently required. The report found that in 1989 dollars, approximately $1.03 billion would be needed for cleanup after a severe accident at a reference boiling water reactor. In addition, depending on whether a 4 percent or an 8 percent inflation rate is assumed, an additional $186.5 million to $409.9 million would be needed to cover incremental cost escalation during the cleanup process. In evaluating these costs, the contractor considered labor, energy, waste disposal, and nuclear insurance as those cost components with the greatest potential effect on cost escalation. Except for nuclear insurance, these factors are the same as those used in the Commission's decommissioning rule, although the relative weights of the factors vary (53 FR 24018, June 27, 1988) (See 10 CFR 50.75(c)(2)). The Commission notes, however, that commenters had ample opportunity to evaluate and comment upon the technical studies that the NRC used as the basis for its decommissioning requirements. No such opportunity has been available heretofore for Addendum 1. Consequently, the Commission concludes that the public interest would best be served if the issue of whether and to what extent the amount of accident cleanup insurance should increase is deferred pending public comment on Addendum 1. As part of its conclusion, the Commission further notes that most licensees already carry accident cleanup insurance in amounts that exceed the maximum amount predicted by the formula in Addendum 1. Thus, there is no compelling health or safety reason to increase the required amount of insurance in advance of public comment. Concurrently, the Commission believes that the public comments on Addendum 1 will enable the Commission to make more informed decisions in connection with any future 1 "Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents-Addendum 1," Pacific Northwest Laboratory, to be published. This report will be available by approximately May 1990 for purchase from the U.S. Government Printing Office, P.O. Box 37082. Washington. DC 20013-7082. A notice of availability will be published. PART 50 STATEMENTS OF CONSIDERATION rulemaking proceeding to increase the amount of required insurance. Individual commenters also have raised specific concerns with the proposed rule. These concerns include the stabilization priority threshold, the 60-day priority period, and the cleanup plan. One commenter indicates that, pursuant to proposed 10 CFR 50.54(w)(4)(i), insurance proceeds would only be required to be dedicated to stabilization and decontamination if the estimated costs exceeded $100 million. Further, this priority would initially apply to stabilization costs for 60 days and could be extended in 60-day increments. Within 30 days after the reactor is stabilized, the licensee is required to submit a cleanup plan which must be approved by the Director of the Office of Nuclear Reactor Regulation. This commenter also suggests that the rule should clarify (a) whether the NRC or the licensee provides the cost estimate, and (b) how the Director of the Office of Nuclear Reactor Regulation determines the length of the stabilization priority and the criteria for approving the cleanup plan. The NRC believes that these and similar issues have been discussed in previous rulemaking and that additional specificity may be cumbersome and counterproductive. The Commission clearly intends to rely on licensees to prepare initial cost estimates of accidents, although it is conceivable that the Commission could prepare its own confirmatory estimates if unusual circumstances warranted. Furthermore, a cut-off figure of $100 million represents a relatively minor accident where the availability of funds would not, as a practical matter, be at issue. Thus, it is very unlikely that the Commission would dispute estimates unless they significantly exceeded $100 million. Further, § 50.54(w)(4)(i) explicitly defines what constitutes stabilization. Therefore, it is unlikely that serious disagreements would arise concerning when a reactor is stabilized. However, if disputes over stabilization should arise, the Commission's Rules of Practice under 10 CFR part 2 provide adequate procedures to resolve them. Similarly, part 2 procedures are also available to resolve , disputes that may arise over the content of cleanup plans. The Commission notes i that the proposed rule was drafted in response to the suggestions of petitioners representing most power reactor licensees and their insurers. The petitioners did not raise these specific issues in their petitions or in comments on the proposed rule. Consequently, the Commission concludes that the suggested changes to the proposed rule are not needed. One commenter takes issue with the following statement in the Regulatory Analysis published in connection with the proposed rule: "Although the effect of these formulas, if developed and adopted, would be to increase the required amount of insurance for some licensees, there should be little impact on insurance costs to licensees because almost all licensees buy the maximum amount of insurance available" (54 FR 46624, at p. 46628, November 6, 1989). This commenter states that, "This may have been true in the past, however we do not agree with this assessment. In fact, we did not automatically purchase the maximum amount of insurance available this year following an increase in available coverage." Notwithstanding this commenter's decision not to buy additional insurance, the Commission notes that the maximum amount of insurance currently offered exceeds by a significant margin the amount that would be required if the maximum figure suggested in Addendum 1 were adopted. Most licensees currently purchase substantially more than this maximum. Thus, the Commission stands by the statement in question. These amendments provide relief from restrictions under regulations due to take effect on April 4, 1990. Therefore, pursuant to 5 U.S.C. § 553(d)(l), the Commission is making the rule effective on the date of publication in the Federal Register without the customary 30-day waiting period. III. Finding of No Significant Environmental Impact; Availability Noting that the text of the final rule is identical to that of the proposed rule, the Commission has reviewed the environmental assessment and finding of no significant environmental impact published in the Federal Register on November 6, 1989 (54 FR 46624, at 46627) in connection with the proposed rule. On the basis of that review, and after considering the public comments and determining that such comments do not affect the conclusion reached in the earlier finding of no significant impact, the Commission has concluded that this amendment to 10 CFR 50.54(w) is not a major Federal action significantly affecting the quality of the human environment, and therefore, an environmental impact statement is not required. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection and copying at the NRC Public Document Room, 2120 L 50-SC-74 Street, NW. (Lower Level), Washington, DC. IV. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). The final rule has been referred to the Office of Management and Budget for review and approval. Public reporting burden for this collection of information is estimated to average 2,000 hours per response, including time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB-7714), U.S. Nuclear Regulatory Commission, Washington, DC 20555; and to the Paperwork Reduction Project (3150-0011 ), Office of Management and Budget, Washington, DC 20503. V. Regulatory Analysis On November 6, 1989, the Commission published in the Federal Register (54 FR 46624) a proposed rule to amend 10 CFR 50.54(w). The rule was developed in response to three petitions for rulemaking. Notice of receipt of these petitions was published in the Federal Register on September 19, 1988 (53 FR 36335). These petitions sought clarification of the stabilization and decontamination priority provisions and rescission of the trusteeship provisions currently contained in 10 CFR 50.54(w). The petitions further stated that the trusteeship provisions may actually have an effect counter to their intended purpose by delaying the payment of claims and thus possibly the cleanup process. The rule developed in response to the petitions for rulemaking should help clarify the mechanism by which accident cleanup funds may be guaranteed to be used for their intended purpose. Even without formal stabilization and decontamination priority and trusteeship provisions, the NRC has authority to take appropriate enforcement action to order cleanup in the unlikely event of an accident. By rescinding the trusteeship requirement, the Commission would be eliminating licensees' costs to obtain trustee services. Thus, the rule will not create substantial costs for licensees. The rule will not have significant impacts on State and local governments PART 50 STATEMENTS OF CONSIDERATION and geographical regions, on the environment, or create substantial costs to the NRC or other Federal agencies. The foregoing discussion constitutes the regulatory analysis for this rule. VI. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the Commission certifies that this final rule does not have a significant economic impact upon a substantial number of small entities. The rule only . affects licensees of nuclear power plants. None of the holders of these licenses fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the 'Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR part 121. VII. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this rule because this rule will not impose a backfit as defined in § 50.109(a)(l). Therefore, a backfit analysis is not required for this rule. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, . as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendment to 10 CFR part 50. . 55 FR 29181 1Published 7 /18/90 .Ef(ective 8/17 /90 Storage of Spent Fuel in Approved Storage Casks at Power Reactor Sites See Part 72 Statements of Consideration . 56 FR944 Published 1/10/91 *Effective 1/10/91 Operations Center Area Code Telephone Number Change See Part 20 Statements of Consideration 50-SC-75 56FR22300 Published 5/15/91 Effective 6/14/91 1 o cm Part so RIM 3.150-AD01 Fracture Toughness-Requirements for Protection Against Pressurized Thermal Shock.Event&

AOENCY:-NucleM" Regu-l'atory -Commissiom

ACTION: Finatrule.

SUMMARY

The Nucl'ear Regulatory*

Commission, (NRCJ'* is amending rt8' regulations*for*light-water nuclear power plants*tochange the procedure fOl"calculatin1rthe*amount ofradiation embrittlement that a* reactor vesse~ reeeiws. Toe-pressunzed thermaFshock rule-(P'l'Srule)'establfshes a screening criterion. This criterion limits the amount of embrittlement ofa:reactor vessef beltline1materiaV beyond' which . the plant cannot contiilue:to.operate: without justifu:ation based:on specific: analysis. T-he final, amendment does not change-the screening criterion. The E"l'S rnle,aht0: prescribe&', procedure. that must' be used. for calculating: the,amount of. embrittlement. for, comparison to the-.screening criterion. The finalamendment*updates the pnocedure-.and*makes.it consistent wi:th. the,one-gi*ven in Regulatory Guide 1.99, Revision 2, published.in, Muy-1988. EP.FECTIVE:DAiTE:: June 14..1991~ FO~ll'JHl!R-INl'ORMATION' COllfT'AC'I': Allen t. Hiser; Jr;, Division-of Engineering; Office of Nuclear Regulatory Reseamh; U.S. Nuclear Regulatory Commission, Washington;. DC 20555; Telephone: f301J *192-3988. SUPPLEMENTARY INF.ORMATION~ . *~. Background Pressurized thermal shock events are system transients in. a pressurized water reactor (PWRl that can cause severe overcooling followed by immediate PART 50 STATEMENTS OF CONSIDERATION repressurization to a high level. The thermal stresses caused by rapid cooling of the reactor vessel inside surface combine with the pressure stresses to increase the potential for fracture if an initiating flaw is present in low toughness material. This material may exist in the reactor vessel beltline, adjacent to the core, where neutron radiation gradually embrittles the material during plant lifetime. The degree of embrittlement depends on the chemical composition of the steel, especially the copper and nickel contents. The toughness ofreactor vessel materials is characterized by a "reference temperature for nil ductility transition" lRTNDTl, which is determined by destructive tests of material specimens. For many reactors now in operation, toughness of the beltlinc materials at room temperature is low. As temperature is raised, toughness increases slowly at first; but at the temperature defined* as RT NDT* toughness begins to increase much more rapidly. The transition in toughness from low values to high th1,1.t takes place above RT NDT means that vessel materials are quite tough at normal operating temperatures. Radiation embrittlement moves RTNDT to higher temperatures. Correlations based on test results for unirradiated and irradiated specimens have been developed to calculate the shift in RT NDT as a function of neutron fluence for various material compositions. The value of RT NDT at a given time in a vessel's life is used in fracture mechanics calculations to determine whether assumed pre-existing flaws would propagate as cracks when the vessel is stressed. The Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61, adopted on July 23. 1985 (50 CFR 29937), establishes a screening criterion. This screening cl'iteriori establishes a limiting level of embrittlement beyond which operation . cannot continue without further specific evaluation. The screening criterion is given in terms of RT NDT* calculated as a*function of the copper and nickel contents of the material and the neutron fluence according to the procedure given in the PTS rule. and called RT Pl'S to distinguish it from other procedures for calculating RT NDT* The PTS rule requires each PWR licensee to report the results of the calculations of predlcted RT ns values for each beltline material (including the copper, nickel and fluence values that pmvided the basis for the calculations) from the time he submits his report to the expiration date of the operating license (EOI.). The PTS rule further provides that if RTPTs for the controlling material is predicted to exceed the * . screening criterion before EOL,,the licensee should *submit plans and a schedule for flux reduction programs that are reasonabll( practicable to avoid reaching the screening criterion. Finally, the PTS rule requires licensees of plants that would reach the screening criterion before EOL despite the flux reduction program to submit a plant-specific safety analysis justifying operation beyond the screening criterion. The licensee must submit the.analysis at

least 3 years before the plant is* predicted to reach that limit. Regulatory Guide 1.154, "Format and Content of Plant-Specific Pressurized Thermal .: . Shock Safety Analysis Reports for
Pressurized Water Reactors" provides guidance for the preparation of the report and describes acceptance criteria that the NRC staff would use. In response to the PTS rule, the licensees of operating reactors have submitted the fluence predictions.1md material composition data and these have now been accepted.

Of greater** importance are ~he flux reduction_ programs that have b.een undertaken by licensees for those plants having high . values of RT Prs* * * *

On December 26, 1989 (54 FR 52946), the Commission published the proposed rule to change the procedure for* calculating RTPTS to reflect recent findings that embrittlement is occurring faster than predicted by the PTS rule for some reactor vessel materials.

Although the PTS rule was adopted on July 23, 1985, the procedure for calculating RTPTS* was developed in 1981-1982 and not

updated because a number of licensees were using the 1982 formulations as the basis for flux reduction programs.
Meanwhile, plant surveillance data were being added to the data base !lnd there were extensive new and more accurate correlations made. These culminated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials;"'

published*in May 1988. Revision 2 provides the basis for pressure-temperature limit calculations. Peer review. of the new correiations was provided by the public comments on Revision 2.

In the regulatory analysis prepared for Revision 2, and repeated in the regulatory analysis for this amendment, the NRC evaluated the impact of amending the PTS rule to be consistent with the Guide. Copper and nickel contents and fluence values for each PWR reactor vessel were taken from the PTS submittals from licensees.

When the values of RTPl'S were recalculated using these.quantities and the procedure 50-SC-76 developed for Revision 2, the results were higher for approximately half the vessels, including three vessels where the value may be over 60°F higher than previously thought. This would increase the probability of PTS-induced vessel failure by a factor of at least 30 for those plants. The NRC believes these changes in the nonconservative direction are greater than can be absorbed by the uncertainties believed to exist and taken into account by the NRG.when the RT M's-based screening limit was set. (A margin of 48°F is added in the calculation of RT Prs to cov:er not only the uncertainty in the formula for embrittlement but also the uncertainties in the copper, nickel, and fluence values entered in the formula.) Based on this new information, the probability of reactor vessel failure by fracture.during a PTS event is presently higher in some vessels than the probability based.on the procedure for calculating RTPTs which is given in the present PTS rule. Moreover, a few of those reactor vessels. will reach the screening criterion in the 1990.'s, Thus, the current PTS rule needs to be amended. A 75 day comment period expired on March 12, 1990. Comments were re.ceive!f from 15 respondents;

Summ~ry of Public Comments The'proposed amendments have been modified in response to the comments received and will be published in final form, as modified, to.become effective 30 days after publication of this final rule. Changes were made in response to the public comments to introduce flexibility and technical improvement in the calculation of RTPl'S by requiring consideration of the plant-specific surveillance data and operating conditions when they would have a significant effect on the date the screening criterion would be reached. Another change was made to loosen the reporting schedule for licensees whose reactor vessels will not become highly embrittled.

A summary of the public comments and staff responses follows; 1. Validity of a Limited Revision Seve:al comments questioned broad issues in the PTS rule and urged that a limited revision not be undertaken. Some comments said that the screening criterion should be raised (made less cons~rvative) because they.believed that the calculated probability of fracture would be reduced if the new embrittlement formula was substituted for the old in those calculations. Other comments pointed out changes in the assumptions about flaw size and PART 50 STATEMENTS OF CONSIDERATION location, as weU as updated information a bout the expected severity and frequency of PrS transients as .. reasons to revisit the* screening criterien. Still other comments questioned tlie*use of a single parameter; RT PTS* in the screening criterion and asked for consideration of a multiparameter criterion. Staff Response. A general response to the comments is as follows. Fil'St, the-scope of-the proposed'amendment ts-nan-ow: to make technical-corrections in the embrittlement formula for calculation of RTPTS values to*eompare*to the screening* criterion*, A general revision* of tlte PTS-ruie mu&t wait until further research is done. Second, the screening criterion is not: a, safety limit. It is tripwire which triggers* a plant--specific safety* analysis; i.e., it: defines which licensees need* to*do that analysis and when it should be done,.Thi'rd, the screening criterion is not linked directly to a predicted frequency* of through-wall cra11:king. Only when the plant-specific analysis is dene (using plant-specific systems: and* fFacture parameters) is-the criterion for continued operation* based on a through-wall crack*frequency of sx10-* per reactor year. It is Regulatory Guide 1.154, "Format' and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors" (not 10 CFR 50.61) which s.tates that this frequency is the staff'a primary acceptance criterlon,for-continued operation. In specific response' to* the issue* of conservatism, the-Regulatory Analysis for the proposed rule* summarized the resuttllo of some studies of the effects-on through~waU crack frequency*when calculated-using the*proposed embrittlement formula, instead*of* the* one used:in* me, origina.t. Monte Carlo,analyses done*earlier; These* studies showed tha-Mhe PTS rule* is more conservative* than previously thought-for some accident scenarios; but. not for alt The results: did not justify raising the screening criterion.

2. Alternative Use of P1a~t 0 Specific Survemance Data. Eleven* out ofl fifteen comments urged the addition ofthis alternative to the proposed RTrn* calculation method based on copper and-nickel contents and fluence, noting that this alternative is allowed in caTculating temperature limit1J usihg*RG.

1.99, 1 The 'Radiution E'mbritUement or Reactor Vessel' Materiuls. Regulatory Guide 1'.99, Rcvi*ion 2. May. 1968. strongest need for this, altemative is for plants nearing the screening criterion. In* the plant-specific PRA [probabilistic risk analysis} required as the basis. for all owing a. plant t0. operate beyond the screening criterion, any embrittlement information may be'used if justification* is* given. Noting this, commenters said that the use of plant-specific surveillance data would in some cases make the PRA result& favorable: therefore, lt-should'be permissible*to use* such data in caleutating* RTm. thereby avoiding the*time-and,expense of:the PRA analysis.

Staff Response The proposed amendment to the PTS rule is prescriptive.on' the issue of calculating*RTPl'S*

becrause not many plants meet the criteria ,for-"credible" surveillance data. given in. lil..G. 1.99-in all respects and because-the criteria are somewhat. subjective, Lengthy disputes over credibility are anticipated, based on experience' in-applying R.G. 1.99 elsewhere .. Moreover. in many cases there is a, difficult choice to. be made between reliance on,a very small amount of pla-nt-&pecific surveillance data, or a cali:ule\ed value,based on a large data base of specimens,most of which were irmdiated in other reactors. Nevertheless, in response* to* the widespread comments, it:is agreed that there is need for some flexibility in the PTS*rule to permit consideration of all available information~A.new pw:agraph* (b)(~).has been. aElded and the existing. paragFaphs (b){;J} through. been,renwnbered. .The. intent of the new paragraph (b )(3,) is to provide ~xibility fo[ use in two kinda of.special situations. Commenters dwelt on the situation where surveillance data showed. the vessel to, be sismficantl.y less embrittled than indicated .by; the : proposed embrittlement formula. In the. other situation.. there.is iruormation from surveillance data or other. information, such. as the operating; tempeira ture reactor vessel that shows the vessel. may be significantzy IDOl!e:embrittled than calculated by the proposed rule. 2 Thus.. some*flexibility has been added,to the rule to ensure that. significant information is not ignored. Several of the commenter& on this issue recommended that Position C 2, * 'Surveillance Data. Available," as well as the criteria for credibility of the sur.veillance data, given. in. R.G. 1.99, be incorporated in the PTSrule in total. ~Tb& irradiation temp&fflture& represented in the data baae that. waa correlated ta obtain the,forraule in the PTS.rute range,Hrom 525 to 590*degrees Fahrenheit. Operatio1J below thet*lemperature ranee is considered to cause mare embrittl&ment. 50-SC-77 The staff has rejected this suggestion in an effort to keep* the implementation of the PTS-rule. as.-simple as po&Sible. It is anticipated that only*those* lli::ensees . who11e vesselil areapproaching*the* screening criterion: would make-use of paragraph (b)(3'). Its use requires. review and approvat by*tlJe.staff.at, which time the guidelines in: R.G. 1.99 may be appropriate. butnotnecessarily so. 3. U&e of'Measured Values af RT= Sevei:al* comment& said: that the changes iii wordill8 of the.requirement in paragraph (b)(2)(i) that "measure_d values mustbeusedifava+/-lable ..... represented Bl change i.D! the rule, which reduced ifs flexibility; Staff Response There is: !IO* change in intent. The words, were* changed in the proposed* rule ta remove any ambiguity: A further clarification wumade in the*fina-J.rule by adding the words*"if credible-values are available.'.* The intent ilr. to,allow a licensee. to offer justification for not. using, a. parlicular measured. value if he does-not have confide.nee: indL 4. Only a Few Plants are Affected' Significantly, burthe Proposed Rule Adds a Regulatory Burden on A1Iand a Public.Relations Burden Also-The proposed. PTS rule "reordel'S" the list of reactor vessela*in: terms-of their sensitivityfo:l?TS:events, and shoul&be revised to reduce these impact& by incmasing flexibiliqr in the requiJ:ements or by:*a:multiparameterapproack Staff Response To limi.t:tbe effortrequired:by*the* industry, the PTS rule prescribes 11 screening:criterion to,separate*out those plants that should.do,the-PRA analysis, based on,tludeveLof.embrittlement of the reacto11 vessels,.ii.e .. : the.rule* describes,who:should do, the. analysis, and when: they should do, it' .. Yet. the foregoihg:comments, request*. that either some kind,oHntermediate*screening procedure ~establishi!d that considers several: pal'ameters, instead of. only RT m, or-that.the, objective should be acg?mpli.shedt b:,t intltoducing flexibility int!, the-rule, The: staffhanejected the,suggestion of a "mini-PRA: as an.intennediate procedure,. because that opens; the door to very misleading*coaclusions. When the, P1'S rule. was-.in the, early formative stages, there,were proposala*.for:a deterministic-criterion.However. it.soon became,clear that there wus,no way to choose.the*design transient from,among the array of transients, of-increasing severity but lower frequency. Extending PART 50 STATEMENTS OF CONSIDERATION this reasoning to PRAs, the staff concludes that a partial PRA is inappropriate. These comments have been rejected, but paragraph (b)(l) has been modified to:reduce the reporting burden for all plants except those expected to reach the screening criterion before the end of their operating life. These modifications are in addition to the amendments to paragraph (b)(l) that were published in the proposed rule to simplify the reporting. requil'ements.

5. Use of "Adequate Protection" Exception to the BackfitRule One comment said that flexibility in granting exemptions to the rule or exceptions to the required submittal schedules would be i'educ*ed if exception was taken to the backfit rule (10 CFR 50.109) on the basis that the amendments to the PTS rule were needed to provide adequate protection to the health and safety of the public. Staff Response The staff has continued to cite "adequate protection," because it believes that the amendment to the PTS rule is necessary*to assure that there is no undue risk to public health and safety from pressurized thermal shock. Characterizing the amendment as "necessary lo assure adequate protection" does not preclude the NRC from granting exemptions to the rule, so long as licensees propose alternatives which assure adequate protection.

The staff also notes that the PTS rule, paragraphs (b)(5), (b)(6) and (b)(7), provides procedures for the kind of . case-by-case review that would normally be the basis for an exemption. There is even what amounts to an appeal procedure in paragraph (b)(7) whereby a licensee whose plant-specific analysis and proposed corrective actions are not approved can again request consideration of additional modifications to equipment, systems and operation or the facility in addition to those previously proposed. Finding of No Significant Environmental Impact The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 CFR part 51, that. this rule is not a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. The PTS rule is one of several regulatory requirements the function of which Is to ensure reactor vessel integrity. This amendment to the PTS rule updates the procedure for calculating the level of embrittlement of the reactor vessel beltline as a result of neutron radiation. Use of the updated procedure will not result in any adverse changes in power level, effluents, or other operational' characteristics of a nuclear power reactor. Therefore, this rule is not expected to have any significant effect on the environment. Moreover, since the use of-the updated procedure is likely to result in more *accurate and conservative predictions of transition to nil ductility, the i:isk of an accident and attendant environmental consequences is likEly to be reduced under the new amended rule. The environmental assessment and finding of 110 significant impact on . which this determination is based are available for inspection at the NRC Public Document Room, 2J.,20 L Street NW. (Lower Level), Washington, DC .. Single copies of the environmental assessment and the.finding of:no significant impact are available from Allen L Hiser, Jr.,.Division of Engineering. Office of Nuclear Regulatory Research,.U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone:,(301) 492.-3988. Paperwork Reduction Act Statement This rule amends information collection requirements that are subject to the Paperwork Re~uction Act of ~80 (44 U.S.C:3501 et seq.). These information collection requirements were approved by the Office of

Managementand'Budgl!I AP,proval No. 3150-0011. . Public reporting burden for this collection of information is estimated to average approximately 331 hours per response, including time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing.and reviewing the collection of information.

Send comments regarding thi's burden estimate or any other aspect of'this collection of information, including suggestions for reducing this burden. to the Information and Records Management Branch (MNBB-7714), Di vis ion of Infonna lion Support Services, Office of Information Resources Management. U.S. Nuclear Regulatory Commission, Washington. DC 20555; and lo the Desk Officer, Office of Information and Regulatory Affairs, NEOB-3019 (3150-0011), Office of Management and Budget, Washington. DC 20503. Regulatory Analysis The NRG staff prepared a regulatory, analysis for the final rule, which describes the factors and alternatives 50-SC-78 considered by the Commission in deciding to propose this rule.-A copy of the regulatory analysis is available for inspection and copying for a fee at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC 20555. Single copies of* the analysis may be obtained from Allen L. Hiser, Jr., Office of.Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission. Washington,* DC 20555,. Telephone*(301}492--3988. Regulatory Flexibility Act Certification As required by the Regulatory Flexibility Act, 5 U.S.C. 605(b), the Commission certifies that this rule does not have a significant economic impact on a substantial number of small entities. This*rule specifies minimum fracture* toughness properties of irradiated pressure vessel-materials to ameliorate the effects of PTS events on nuclear facilities licensed under the provision*oflO CFR 50.21(b) and 10 CFR 50.22. The companies that own these facilities do not fall within the scope of the definition of"small.entities" as set forth. in the Regulatory Flexibility Act or the Small Business* Size Standards in regulations issued' by the Small'Businesil Administration at 10 CFR part 121. Backfit Analysis The NRC has concluded, on the basis of. the documented evaluation required by w CFR 50.109(a)(4), that the backfit re(luiremente contained in this amendment are necessary to ensure that the facility provides adequate protection to the public health and* safety, and, therefore, that a backflt analysis is not required and the cost-benefit standards of 10 CFR 50.109(a){3) do not apply. The documented evaluation given in the regulatory analysis includes a* statement of the objectives of and reasons for the backfits that would be required by the rule and sets forth the basis.for the NRC's conclusion that these*backfits are not subject to the cost~benefit sta1Jdards oflO CFR 50.109(a)(3). List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminlfl penalties, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors. Radiation protection, Reactor siting criteria,. Reporting and recordkeeping requirements. For the reasons set *out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, and 5 U.S.C. 553, the NRC is adopting the following amendments to 10 CFR part 50. PART 50

STATEMENTS OF CONSIDERATION 56 FR 23360 Published 5/21 /91. Effective 6/20/91 Standards for Protection Against Radiation See Part 20 Statements of Consideration 56 FR 31306 Published 7/10/91 Effective 7 /10/96 10 CFR Part 50 RIN 3150-ADOO Monitoring the Effectiveness of Maintenance at Nuclear Power Plants AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The Commission is amending its regulations to require commercial nuclear power plant licensees to monitor the effectiveness of maintenance activities for safety significant plant equipment in order to minimize the likelihood of failures and events caused by the lack of effective maintenance.

The Commission believes that, to maintain safety, it is ne 0 cessary to monitor the effectiveness of maintenance, and take timely and appropriate corrective action, where necessary, to ensure the continuing effectiveness of maintenance for the lifetime of nuclear power plants, particularly as plants age. The final rule requires that licensees monitor the performance or condition of certain structures, systems and components (SSCs) against licensee-established goals in a manner sufficient to provide reasonable assurance that those SSCs will be capable of performing their intended functions. Such monitoring would take into account industry-wide operating experience. Where monitoring proves unnecessary, licensees would be permitted the option of relying upon an appropriate preventive maintenance program. Licensees will be required to evaluate the overall effectiveness of their maintenance programs on at least an annual basis, again taking into account industry-wide operating experience, and adjust their programs where necessary to ensure that the prevention of failures is appropriately balanced with the minimization of unavailability of SSCs. Finally, in _performing monitoring and maintenance activities which require taking equipment out of service, licensees should assess the total plant equipment" that is out of service and determine the overall effect on the performance of safety functions. EFFECTIVE DATE: The final rule will become effective July 10, 1996. However, the information collection requirements contained in 10 CFR 50.65 are not effective until the NRC publishes the Office of Management and Budget (0MB) clearance in the Federal Register. FOR FURTHER INFORMATION CONTACT: Robert Riggs, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, (301) 492-3732. SUPPLEMENTARY INFORMATION: Background On March 23, 1988 (53 FR 9430), the Commission published a final Policy Statement on Maintenance of Nuclear Power Plants. In the Policy Statement, the Commission stated that it expected to publish a notice of proposed rulemaking and provided the general framework for the proposed rule. On November 28, 1988 (53 FR 47822), the Commission published a notice of proposed rulemaking to require commercial nuclear power plant licensees to implement a maintenance program to reduce the likelihood of failures and events caused by the lack of effective maintenance. In support of this rule, the Commission published a draft regulatory guide on maintenance on August 17, 1989 (54 FR 33988) for public comment. On December 8, 1989, the Commission issued a revised policy statement on maintenance (54 FR 50611) that stated the Commission's intention to hold rulemaking in abeyance for 18 months while it monitored industry initiatives and improvements and to assess the need for rulemaking in the maintenance area at the end of the 18 month period. On April 13, 1990, in response to a Commission request, the staff forwarded the following four proposed criteria to be used in determining the need for maintenance rulemaking: Criterion 1-Licensees have effectively implemented an adequate maintenance program or are committed to and proceeding towards this goal. Criterion 2-Licensees exhibit a favorable trend in performance related to maintenance. Criterion 3-Licensees are committed to the implementation of a maintenance performance standard acceptable to the NRC. Criterion 4-Licensees have in place or are committed to an evaluation 50-SC-79 program for ensuring sustained performance in the maintenance area. On May 25, 1990, the Commission approved these criteria and advised the staff that additional factors which may influence the Commission in determining the need for maintenance rulemaking were: (1) The ability to enforce maintenance programs or standards; (2) the presence of a strengthened commitment by the industry to monitor equipment performance to identify problematic components, systems, and functions, to conduct root cause analysis, to track corrective actions, and to feedback information into the maintenance program; and (3) provision of a mechanism by which the NRC could verify the effectiveness of the program. On May 23, 1990, the Commission directed the staff to develop a second proposed rule that would be based. In addition, the Commission directed the staff to develop two procedural approaches for implementation of a rule. The first implementation approach, which allowed licensees to use an alternate NRC approved maintenance standard, was incorporated into both rules. The second approach was to include conceptual considerations for application of a maintenance rule only to licensees exhibiting poor performance in the maintenance area. In SECY-91-110 dated April 26, 1991, the staff reported the results of the stafrs evaluation of the need for maintenance rulemaking. The evaluation was based upon an assessment of licensee progress against the four Commission-approved criteria and the additional factors identified by the Commission. The staff also presented for Commission consideration options and recommendations pertaining to: (1) The issuance of a final policy statement; (2) the issuance of a final oriented" rule and accompanying regulatory guide, based upon the November 1988 proposed rule, the August 1989 draft regulatory guide, and public comments received on both the proposed rule and draft regulatory guide; (3) the issuance of a proposed "reliability-based" rule and accompanying draft regulatory_ guide; (4) the application of a maintenance rule only to poor performers. Need for a Rule The Commission's determination that a maintenance rule is needed rests first on the conclusion that proper maintenance is essential to plant safety. As discussed in the Regulatory Analysis and the Backfit Analysis for this rule, PART 50

STATEMENTS OF CONSIDERATION there is a clear link between effective maintenance and safety as it relates to such factors as number of transients and challenges to safety systems and the associated need for operability, availability and reliability of safety equipment.

In addition, good maintenance is also important in providing assurance that failures of other than safety-related SSCs that could initiate or adversely affect a transient or accident are minimized. Minimizing challenges to safety systems is consistent with the Commission's defense-in-depth philosophy. Maintenance is also important to ensure that design assumptions and margins in the original design basis are either maintained or are not unacceptably degraded. Therefore, nuclear power plant maintenance is clearly important in protecting the public health and safety. The results of the Commission's Maintenance Team Inspections (MTls] indicated that licensees have adequate maintenance programs and have exhibited an improving trend in program implementation (Criterion 1). However, some common maintenance-related weaknesses were identified, such as inadequate root cause analysis leading to repetitive failures, lack of equipment performance trending, and the consideration of plant risk in the prioritization, planning and scheduling of maintenance. In general, as evidenced by plant operational performance data and the results of NRC assessments, the industry has exhibited a favorable trend in maintenance performance (Criterion 2). With regard to licensee commitment to an NRG-approved maintenance performance standard (Criterion 3), the industry, through NUMARC, expressed to the Commission its commitment, in general, to the goal of improving performance in the area of maintenance. The industry asserted that all licensees are committed, by virtue of their membership in the industry-sponsored Institute for Nuclear Power Operations [INPOJ, to meeting, or striving to meet, the performance objectives contained in INPO 90-008, "Maintenance Programs in the Nuclear Power Industry." INPO 90-008 is primarily a compilation of preexisting objectives and criteria developed by INPO relating to maintenance. These objectives and criteria largely relate to maintenance program content and programmatic measures of performance. No written commitments were received from licensees and the industry-wide commitment which was received was at best indirect. The Commission believes that a sufficient commitment by licensees to a maintenance standard approved by the NRC has not been received. With regard to licensees having in place or being committed to an evaluation program for ensuring sustained performance in the area of maintenance (Criterion 4), the industry, through NUMARC, indicated that all licensees will perform a comprehensive assessment of their maintenance programs against the performance objectives of INPO 90-008. These time assessments were to be conducted over a four year period. Additionally, periodic INPO evaluations which include the maintenance area will continue to be performed. However, the Commission believes that the industry's largely programmatic assessments and evaluations of licensee maintenance programs will not alone suffice. Instead, the Commission believes that the effectiveness of maintenance must be assessed on an ongoing basis in a manner which ensures that the desired result, reasonable assurance that key structures, systems, and components are capable of performing their intended function, is consistently achieved. Further, there is a continuing need for feedback of the results of such assessments and to factor those results into programmatic requirements, where assessment results indicate ineffective maintenance. Considering the above points, the Commission is satisfied that the industry has been generally successful in bringing about substantial improvement in maintenance programs. Further, the improving trend established over the past several years has continued. However, the necessity for ongoing results-oriented assessments of maintenance effectiveness is indicated by the fact that, despite significant industry accomplishment in the areas of maintenance program content and implementation, plant events caused by the degradation or failure of plant equipment continue to occur as a result of instances of ineffective maintenance. Additionally, operational events have been exacerbated by or resulted from plant equipment being unavailable due to maintenance activities. Under existing requirements and industry maintenance initiatives, with relatively few exceptions, the availabilities of safety significant structures, systems, and components are _not routinely assessed. These events and circumstances further attest to the need for ongoing results-oriented assessment of maintenance effectiveness since, together with equipment reliability, 50-SC-80 equipment availability is an important measure of maintenance effectiveness. Regarding the additional factors considered by the Commission in determining the need for a maintenance rule, the Commission believes that there exists a need to broaden its capability to take timely enforcement action where maintenance activities fail to provide reasonable assurance that safety significant SSCs are capable of performing their intended function. With regard to the presence of a strengthened industry commitment to: Monitor equipment performance to identify problematic components, systems and functions; to conduct root cause analysis; to track corrective actions; and to feedback information into maintenance programs, the Commission has determined, based upon the weaknesses identified by the MTis and the lack of sufficient commitments by licensees to a maintenance standard, that additional regulatory attention to these matters is warranted. Concerning the provision of a mechanism by which the NRC could verify the effectiveness of maintenance programs, neither the Commission nor the industry have been able to develop overall performance indicators which would readily provide unambiguous indication of overall maintenance effectiveness at any given plant. Thus, the Commission's consideration of these additional factors also weighs in favor of promulgating a rule that requires the monitoring and assessment of maintenance effectiveness. Additionally, consideration of these factors leads the Commission to conclude that it is necessary for such a rule to include requirements for corrective action to address instances of ineffective maintenance, and feedback of the results of monitoring and assessment into licensee maintenance programs. In consideration of the above, the Commission has determined that a regulatory framework must be put in place which provides a mechanism for evaluating the overall continuing effectiveness of licensee maintenance programs, particularly as the plants continue to age. As noted previously, areas directly related to this issue were identified as common weaknesses during the NRC's Maintenance Team Inspections. These areas included inadequate root cause analysis, lack of equipment performance trending, and lack of consideration of risk in the prioritization, planning, and scheduling of maintenance. The Commission therefore concludes that a rule requiring that licensees monitor and assess the PART 50

STATEMENTS OF CONSIDERATION effectiveness of maintenance activities is necessary.

In addition to all of the above considerations, the Commission's conclusion that a rule requiring that the effectiveness of maintenance be monitored is also predicated on the fact that the Commission's current regulations, regulatory guidance, and licensing practice do not clearly define the Commission's expectations with regard to ensuring the continued effectiveness of maintenance programs at nuclear power plants. The Commission has many individualized requirements relative to maintenance, including SSCs in the balance of plant (BOP), throughout the regulations. These include 10 CFR 50.34(a)(3)(i); 50.34(a)(7); 50.34[b)[6) (i), (ii), (iii), and (iv); 50.34(b)(9); 50.34(f)(l) (i), (ii), and (iii); 50.34(g); 50.34a(c); 50.36(a); 50.36(c) (2), (3), (5), and (7); 50.36a(a)(l); 50.49(b); 50.55a(g); part 50, appendix A, criteria 1, 13,18,21,32, 36, 37,40, 43,45,46, 52,53; part 50, appendix B. More generally, 10 CFR 50.34(b)(6)(iv) requires licensees to address their plans for the conduct of "maintenance, surveillance, and periodic testing of structures, systems, and components." However, there is no guidance on exactly what these "plans for the conduct of maintenance" should include with regard to the monitoring of maintenance effectiveness. The Commission's rules, guidance, and practice also require clarification as to what structures, systems, and components should be subject to maintenance requirements. Although § 50.34(b)[6)[iv) references maintenance for "structures, systems, and components" without further qualification, the guidance in Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants-LWR Edition," (Revision 3, November 1978) is silent on the scope of SSCs that the maintenance program should cover (see Regulatory Guide 1.70, section 13.5.2). Regulatory Guide 1.70 also refers to Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)." Regulatory Guide 1.33, which implements portions of 10 CFR part 50, appendix B, indicates in appendix A that "maintenance that can affect the performance of related equipment should be properly preplanned and performed in accordance with written procedures

* *." The sample listing of maintenance operations requiring procedures also is limited to related equipment.

Regulatory Guide 1.70 also endorses industry standards for nuclear power plant operations that are limited to maintenance or modifications "which may affect the functioning of safety-related structures, systems, or components*

*." The Commission has previously interpreted its rules and guidance as requiring licensees to address the safety aspects of certain SSCs in the BOP. For example, 10 CFR 50.34[g) requires applicants for licenses after 1982 to evaluate their facility against the Standard Review Plan (SRP), NUREG-o800.

The SRP requires licensees to evaluate a number of SSCs in the BOP [this is further discussed in the Commission's response to Question 7 in the summary of public comments). Requirements and guidance for monitoring maintenance effectiveness and for taking corrective action when maintenance is ineffective should enhance the Commission's capability to take timely and effective action against licensees with inadequate or poorly conducted maintenance in order to ensure prompt resumption of effective maintenance activities. For these reasons, the Commission concludes that a regulation that requires all nuclear power plant licensees to monitor the effectiveness of maintenance activities is warranted. The rule provides for continued emphasis on the defense-in-depth principle by including selected BOP SSCs, integrates risk consideration into the maintenance process, provides an enhanced regulatory basis for inspection and enforcement of BOP related issues, and provides a strengthened regulatory basis for ensuring that the progress achieved to date is sustained in the future. Description of Rule The objective of the final rule is to require the monitoring of the overall continuing effectiveness of licensee maintenance programs to ensure that: (1) Safety related and certain non-safety related structures, systems, and components are capable of performing their intended functions; and (2) for safety related equipment, failures will not occur which prevent the fulfillment of safety-related functions, and failures resulting in scrams and unnecessary actuations of safety related systems are minimized. All references to the rule are to the new § 50.65. Two approaches, which are prescribed in paragraphs (a)(l) and (a)(2) of the rule, are provided for assuring maintenance effectiveness. The intention of paragraph (a)(l) of the rule is that the licensee establish a monitoring regime which is sufficient in scope to provide reasonable assurance that (1) intended safety, accident mitigation and transient mitigation 50-SC-81 functions of the structures, systems, and components [SSCs) described in paragraphs [b)[l) and (b)(2)(i) can be performed; and (2) for the SSCs described in paragraphs (b)[2)(ii) and (b)[2)[iii), failures will not occur which prevent the fulfillment of safety-related functions, and failures resulting in scrams and unnecessary actuations of safety related systems are minimized. Where failures are likely to cause loss of an intended function, monitoring should be predictive in nature, providing early warning of degradation. Monitoring activities for specific SSCs can be performance oriented (such as the monitoring of reliability and availability), condition-oriented [parameter trending), or both. The results of monitoring are required to be evaluated against the established goals. Goals should be established commensurate with an SSC's safety significance. Where available, the assumptions in and results of probabilistic risk assessments [PRAs) or individual plant examinations (IPEs) should be considered when establishing goals. The licensee is encouraged to consider analytical techniques, such as system unavailability modeling studies, which may be useful in developing goals; however, such analyses are not required. The purpose of paragraph (a)(2) of the rule is to provide an alternate approach for those SSCs where it is not necessary to establish the monitoring regime required by (a)(l). For example, this provision might be used where an SSC, without preventive maintenance, has inherently high reliability and availability (e.g., electrical cabling) or where the preventive maintenance necessary to achieve high reliability does not itself contribute significantly to unavailability (e.g., moisture drainage from an air system accumulator). The licensee is encouraged to consider the use of reliability-based methods for developing the preventive maintenance programs covered under this section of the rule; however, the use of such methods is not required. The purposes of paragraph (a)(3) of the rule are two-fold: (1) This provision requires that SSC performance or condition goals, performance or condition monitoring, and preventive maintenance activities implemented pursuant to paragraphs (a)(l) and (a)[2) be evaluated in light of SSC reliabilities and availabilities. In the case of SSCs treated under paragraph (a)(l), adjustments are to be made to goals, monitoring, or preventive maintenance requirements where equipment PART 50

STATEMENTS OF CONSIDERATION performance or condition have not met established goals. Conversely, at any time the licensee may eliminate monitoring activities initiated in response to problematic equipment performance or industry experience once the root cause of the problem has been corrected or the adequacy of equipment performance has been confirmed.

In the case of SSCs treated under paragraph (a)(2), adjustment of preventive maintenance requirements may be warranted where SSC availability is judged to be unacceptable. SSCs treated under

paragraph (a)(2) which experience one or more maintenance-preventable failures, should become subject to the requirements of (a)(1) (see discussion below) or, where this is not feasible, may require other remedial action, such as modification or replacement.

(2) This provision provides that the planning and scheduling of maintenance should consider the cumulative impact of all equipment simultaneously out of service on plant safety. A regulatory guide providing an acceptable methodology for implementing this rule will be developed by the NRC staff and issued for public comment. To permit ample opportunity for licensees to comply with the five year implementation schedule specified in the rule, the regulatory guide is expected to be available in final form two years from the date this rule is promulgated. Additional Guidance Scope of Monitoring It is not the intent of the Commission to require a monitoring program so extensive that it detracts from licensees' ability to otherwise maintain equipment. The extent of monitoring may vary from system to system depending upon system importance to plant risk. Some monitoring at the component level may be necessary; however, it is envisioned that much of the monitoring could be done at the system or train functional level. For example, for less significant systems, indicators of system reliability (where sufficient performance data exist) and availability may be all that is necessary. Some parameter trending, beyond that already required by NRC requirements to provide early warning of degradation, may also be necessary for critical components whose unavailability causes a system train to be unavailable or whose failure is otherwise unacceptable. Rather than monitoring the many SSCs which could cause plant scrams, the licensee may choose to establish a performance indicator for unplanned automatic scrams and, where scrams due to equipment failures have been problematic or where such scrams are anticipated, choose to monitor those initiators most likely to cause scrams. It is not intended that this monitoring requirement duplicate activities currently being conducted, such as technical specification surveillance testing, which could be integrated with, and provide the basis for, the requisite level of monitoring. Consistent with the underlying purposes of the rule, maximum flexibility should be offered to licensees in establishing and modifying their monitoring activities. Reliability and Availability of SSCs Subject to Either Paragraph (a)(1) or (a)(2) SSCs which are treated under paragraph (a)(1) may have formally established reliability and availability goals against which they are explicitly monitored, where goals of this nature are appropriate. In addition, and regardless of the nature of the monitoring and goals established to satisfy paragraph (a)(1), reliability and availability over the longer term must be assessed periodically pursuant to the requirements of paragraph (a)(3), as part of the evaluation of goals, monitoring requirements, and preventive maintenance requirements. The reliability and availability of SSCs which are treated under paragraph (a)(2) are required to be considered under the requirements of paragraph (a)(3), as part of the periodic assessment of preventive maintenance requirements. Paragraph (a)(2) Is Not Intended To Be Used To Justify Continuing the Status Quo, Where the Status Quo Is Not Effective in Ensuring Acceptable Levels of Availability and Reliability Under the terms of paragraph (a)(2), preventive maintenance must be demonstrated to be effective in controlling the performance or condition of an SSC such that the SSC remains capable of performing its intended function. Hence, it is expected that, where one or more preventable failures occur on SSCs treated under this paragraph, the effectiveness of preventive maintenance is no longer demonstrated. As a result, the SSC would be required to be treated under the requirements of paragraph (a)(1) until such time as a performance history is established to demonstrate that reliability and availability are once again effectively controlled by an established preventive maintenance regimen. Once such a demonstration has been made, it would be acceptable to 50-SC-82 return to treating the SSC under paragraph (a)(2). Paragraph (al(3)-Assessing the Cumulative Impact of Out-of-Service Equipment on Performance of Safety Functions-Use of PRA Assessing the cumulative impact of out-of service equipment on the performance of safety functions, as called for under paragraph (a)(3), is intended to ensure that the plant is not placed in risk-significant configurations. These assessments do not necessarily require that a quantitative assessment of probabilistic risk be performed. The level of sophistication with which such assessments are performed is expected to vary, based upon the circumstances involved. The assessments may range anywhere from simple deterministic judgments to the use of an on-line living PRA. It is to be expected that, over time, assessments of this type will be refined based upon technological improvement and experience. Derivation of the Final Rule The final rule is comprised of a subset of the aspects of the proposed maintenance rule and its associated draft regulatory guide, which were issued for public comment on November 10, 1988, and on August 17, 1989, respectively. The final rule includes only those aspects that are oriented", including those addressing establishment of goals, monitoring and assessment of maintenance effectiveness, feedback and corrective actions, and, in a more* limited manner, predictive and preventive maintenance. These aspects were detailed in Regulatory Positions C.3, C.5, and C.6 of the draft regulatory guide and were the subject of considerable public comment in response to Questions 3, 9, 10, and 11 posed by the Commission when it issued the proposed maintenance rule. These comments are addressed in the summary of public comments accompanying the final rule. Details of the derivation are discussed below. Establishment of Goals and Monitoring Section 50.65(a)(1) requires the monitoring of performance or condition of structures, systems, and components (SSCs) against licensee-established goals. These requirements were drawp from the requirements of the proposed -* rule, in § § 50.65(c) (1) and (2), and elements (b) (1)(iii), (5), (10), and (17). The statement of considerations (SOC) for the proposed rule also discussed the process of establishing goals, monitoring, and taking appropriate corrective action, see 53 FR 47825. PART 50

STATEMENTS OF CONSIDERATION Comments on appropriate methods of monitoring, the need for, form of, and possible kinds of effectiveness criteria, and the use of performance indicators for component reliability and maintenance performance were requested, see questions 9 and 10, 53 FR 47825. Comments on criteria and quantitative goals were also requested in the Federal Register notice accompanying the publication of the draft regulatory guide, see 54 FR 33983. The draft regulatory guide discussed goal setting and monitoring in sections C.1.1, C.1.3, C.3.2, C.4.6.4, C.5.2.2, C.5.2.3, C.5.2.4, and C.6. Consideration of industry-wide operating experience under § 50.65(a)(1) as well as § 50.65(a)(3) of the final rule were anticipated by: (1) The proposed rule's discussion of a draft NUREG report which surveyed maintenance practices, 53 FR 47824, (2) a recommendation in the SOC concerning use of the NPRDS, id., and (3) Questions 10 and 11 of the SOC, 53 FR 47825. It was also alluded to in section C.5.2.3 of the regulatory guide, and discussed in section C.3.2. Corrective Action The final rule's requirements that corrective action be taken in response to the results of monitoring, and that at least an annual evaluation of the monitoring, goal establishment and corrective action activities were presaged by the proposed rule's requirement in § 50.65(c)(2) for assessment the effectiveness of the maintenance program and making appropriate improvements, Element (l)(ii) of the proposed rule, and the regulatory guide's discussion on the functioning of the maintenance process, e.g., sections C.1, C.1.3 and C.1.4, C.3.2, C.4, C.5.1, and C.6. Preventive Maintenance Preventive maintenance, which is endorsed by § 50.65(a)(2) of the final rule, was one of the elements of the proposed rule, see 53 FR 47828, Element 1(ii). The regulatory guide addressed preventive (also referred to as "proactive")

maintenance in sections C.2 and C.4.6.1. Scope of SSCs Subject to Maintenance The scope of SSCs subject to the final maintenance rule includes safety-related SSCs, and certain "non-safety" SSCs in the BOP which meet one or more of four specific criteria. See final rule, § 50.65(b). The matter of scope was addressed in the proposed rule, which suggested that all SSCs in a nuclear power plant, including those in the balance of plant (BOP) were to be subject to the proposed rule's maintenance requirements. See proposed rule,§ 50.65(b). The regulatory guide indicated that the rule applies "to all parts of the plant'that could significantly impact safe operation and security, including the BOP". See Sections B., C.1. Comments on scope of SSCs were solicited in the SOC for the proposed rule at Question 7 (53 FR at 47825), and in the proposed regulatory guide at Question 2 (see 54 FR 33983). As shown by the above, all of the significant provisions of the final rule were presaged in the proposed rule and in the proposed regulatory guide. The final rule is not a significant departure from NRC proposals offered for public comment except that, as noted, the final rule is a subset of those proposals. Since all of the elements of the final rule were the subject of extensive public comment, there is no need to publish the final rule as a proposed rule for still more comment. As noted, there will be further comment on the rule's implementing guidance. Clearly, given the period allowed for implementation, there can be adjustments made to the rule before it becomes effective should further developments so require. Industry Programs The Commission encourages industry initiatives and responsibility for problem identification and resolution. Several guidelines exist in the industry (e.g., INPO 90---008, "Maintenance Programs in the Nuclear Power Industry," Institute of Nuclear Power Operations] that are directed toward providing performance objectives and criteria for effective maintenance programs. With regard to the programmatic aspects of maintenance, the Commission encourages the industry to continue the development and improvement of such guidelines and to standardize recommendations and guidance for plant maintenance programs. In acknowledgement of the generally satisfactory state of maintenance programs, the final rule provides great flexibility for the industry to continue developing, improving and implementing recommendations and guidance concerning maintenance programs. The Commission encourages such activities, especially as they support improvements in the evaluation of maintenance program effectiveness. Implementation and Compliance The focus of the rule is on the results achieved through maintenance and, in this regard, it is not the intent of the rule Iha t existing licensees necessarily develop new maintenance programs. However, because the Maintenance 50-SC-83 Team Inspections identified weaknesses in some licensees' maintenance programs, it is expected that each licensee will assess its program and take appropriate action to improve those areas where weaknesses were identified. The rule has a five year implementation schedule with supporting regulatory guide development and promulgation expected within the first two years. This schedule allows three years for licensee development beyond the time that final guidance is expected to be available. Implementation and compliance with the rule is achieved through SSC performance or condition monitoring against appropriate licensee-established goals or, as an alternative, through the conduct of preventive maintenance that has been demonstrated to be effective. Where the performance or condition of SSCs is determined to be unacceptable, corrective action is required. Additionally, compliance is achieved through the periodic assessment of monitoring, goals, and preventive maintenance activities to ensure that the objective of minimizing SSC failures is being met, consistent with the objective of minimizing SSC unavailability due to monitoring and preventive maintenance. Summary of Public Comments The comment period for the proposed rule closed February 27, 1989, and for the draft regulatory guide October 17, 1989. Thirty-five comments on the proposed rule were received during the official comment period and fifty-seven were filed after the comment period closed. Thirty-six comments were received on the regulatory guide. All comment letters were considered in formulation of the final rule. Comment letters were also considered in arriving at the Commission's decisions to revise the accompanying regulatory guide to reflect the final rule's narrowed focus on results, to provide an opportunity for public comment on the revised regulatory guide, and to issue final guidance well in advance of the date specified for rule implementation. Of the 92 comments on the proposed rule, 67 were filed by utilities, 11 by industry groups and trade ass,ociations, 4 by individuals, 3 by vendors, 3 by public interest groups, 2 by Federal Agencies, and 2 by state groups/ individuals. Of the 36 comments on the regulatory guide, 22 were filed by utilities, 5 by industry and professional groups, 1 by State, 5 by corporations, 2 by individuals, and 1 by a vendor. The Commission is appreciative of the time and effort expended by those who submitted comments. Maintenance is a PART 50

STATEMENTS OF CONSIDERATION matter of considerable priority and importance, and the views expressed in the comments have been very helpful to the Commission in its deliberation.

Many comments came from individual licensees, but most supported the comments prepared by the Nuclear Management Resource Council (NUMARCJ. In summary, most of the commenters on the proposed rule stated that there was no need for a separate rule on maintenance for nuclear power plants because (1) the NRC already has regulatory authority and methods in place to provide an overview of maintenance program capability to ensure adequate protection of the public health and safety, (2) there has been no demonstration that the rule will increase public safety and it may actually decrease safety by diverting industry efforts away from maintenance to support activities directed toward demonstrating compliance, (3) good maintenance assessment indicators already exist for both industry and the NRC, such as the Institute of Nuclear Power Operations (INPOJ performance indicators, Systematic Assessment of Licensee Performance (SALP) reviews, the NRC Maintenance Inspection Program, and Licensee Event Reports (LER's), and (4) the industry already has maintenance initiatives under way and, as a whole, the industry is improving in the maintenance area. Many commenters considered the proposed rule unbounded in scope because there are no limits established* for the BOP. They were concerned that, with such a broad and undefined scope, the industry cannot assess the impact of the proposed rule. Therefore, it was suggested that, at the very least, the final rule should be postponed until issuance of the regulatory guide. NUMARC and most utilities commented that, without measures of effectiveness stated in the proposed rule, they did not know what requirements or expectations would be needed to implement the proposed rule and determine regulatory compliance. There was concern that effectiveness, as specified in the proposed 10 CFR 50.65(c), is a qualitative matter and subject to different interpretation by both licensees and the NRC. There was also concern that the lack of criteria describing adequate programs places a burden on the industry and public to assess what is needed for the broad subject area defined in the proposed rule by the NRC and that the proposed rule establishes requirements for specific program elements (10 CFR 50.65[b)J that are not defined. Most commenters felt that a prescribed set of maintenance performance indicators (MPis) cannot be used as the sole basis for evaluating the effectiveness of a maintenance program. NUMARC believes that the existing regulations do not establish requirements similar to the proposed rule, especially with regard to BOP equipment. Therefore, licensees will be forced to modify their maintenance programs to satisfy new requirements, which means the standards of a backfit analysis (10 CFR 50.109) apply. NUMARC further stated that the "adequate protection" standard of 10 CFR 50.109(a)(4) does not apply with regard to implementing the proposed rule. They feel that this was not supported by data provided in the proposed rule or the accompanying regulatory analysis. They felt that the public risk reduction data used in the regulatory analysis was outdated, that recent data by both the industry and the NRC should be used to evaluate public risk reduction, and that the increased costs associated with implementation were grossly underestimated. NUMARC further believes that industry objectives and programs are consistent with the NRC expectations stated in the March 1988 Policy Statement on Maintenance of Nuclear Power Plants. NUMARC believes that increased emphasis has been placed on maintenance, improvements in performance and reliability have been achieved, and therefore the promulgation of a rule is now unnecessary and unjustified. They believe that the NRC should take action against the few poor maintenance performers, rather than promulgate a rule across the whole industry. Two individuals, three public interest groups, and two State representatives were supportive of a maintenance rule but were not necessarily in total agreement with the way the rule was formulated or how it should be implemented. They believed that nuclear power plant maintenance directly affects the health, safety, and economic well-being of the public and that nuclear facilities not properly maintained will be unsafe and uneconomical, even with the best design, construction, and operation. They believe that improper maintenance, even of components not previously associated with safety, can have adverse safety consequences. Furthermore, they believe that the superior performance of nuclear power plants in other countries is attributed to their maintenance program. One State representative believes that the 50-SC-84 maintenance standard should be published initially as a guide and not as a rule that utilities should have the prerogative to organize in the most resource-effective manner their approach to meeting the key components of the standard. The Commission could then evaluate experience under the regulatory guide to determine whether a rule is required. One individual was against a rule because the industry has a good safety record and the rule would be costly and an unnecessary burden on the industry. The comments on the regulatory guide raised many of the same issues as those comments associated with the proposed rule. In general the issues addressed were the level of detail in the regulatory guide; the scope of structures, systems, and components covered by the guide; the criteria to be used to determine if a maintenance program is effective; the use of quantitative goals for determining satisfactory level of performance for plant maintenance programs; the quantitative measures for such goals; the usefulness of NPRDS data for assessing effectiveness of plant maintenance programs; the usefulness of PRAs for plant maintenance programs; the timeliness of corrective actions; the definition of maintenance; the documentation of the technical basis of a maintenance program; and the extent of root cause analysis and feedback. These comments on the proposed rule were either repeated or expanded in the commenters' responses to the 12 questions posed by the Commission in the Statement of Considerations for the proposed maintenance rule. These questions are listed below; and each response contains a synopsis of the public comment and the Commission response for that particular question. Where appropriate, the responses reflect the revisions to the final version of the maintenance rule. The responses also include consideration of the public comments received on the draft regulatory guide. 1. Is it appropriate for the nuclear power industry to develop a Maintenance Standard and, if so, would the industry develop such a Maintenance Standard? Comments-Most commenters feel that another maintenance standard is not needed. They believe that the guidelines developed by INPO provide the basic framework of a standard and could be expanded to accommodate NRC requirements. The Policy Statement on Maintenance, existing industry standards, and the INPO Guidelines for the Conduct of Maintenance at Nuclear Power Plants PART 50

STATEMENTS OF CONSIDERATION contain the information needed to ensure effective maintenance programs.

If a standard is to be developed, all utilities prefer a standard developed by industry rather than by NRC with INPO or NUMARC taking the lead. One citizen's group stated that the NRC, not the industry, should develop the maintenance standard. No commitment was received during the comment period to develop a maintenance standard. Response-The Commission encouraged the industry to develop a maintenance standard because the Commission believed that the development of a standard would allow maximum utilization of current industry initiatives toward developing and implementing effective maintenance programs and that licensee participation in the development of the standard would provide additional incentive and responsibility for improving plant maintenance programs. In addition, the Commission believed that the effort would benefit from industry's expertise in this area and that it would be more likely that the maintenance practices from plants with good maintenance programs would become part of the industry-developed maintenance standard. On April 17, 1990, NUMARC submitted INPO 90-008, "Maintenance Programs in the Nuclear Power Industry," as the industry maintenance standard. The Commission reviewed this document and found that, with minor modification, it formed a comprehensive description of the necessary attributes of a maintenance program. In acknowledgement of this document, the generally favorable results of the NRC's Maintenance Team Inspections regarding the adequacy of licensees' maintenance programs, and the many other industry initiatives in this area, the Commission revised the rule to emphasize the effectiveness or results of maintenance programs and emphasize the programmatic aspects of maintenance. Also, in acknowledgement of the generally satisfactory state of maintenance programs the final rule provides great flexibility for the industry to continue developing, improving and implementing recommendations and guidance concerning maintenance programs. The Commission encourages such activities, especially as they support improvements in the evaluation of maintenance program effectiveness. However, because the rule has been modified to de-emphasize programmatic requirements of maintenance, the Commission does not currently intend to formally endorse an industry maintenance program standard.

2. What level of detail should be included in the Maintenance Standard?

Comments-NUMARC and the utilities believe that ,any maintenance guidelines or standard should provide a general description of the necessary elements of a good maintenance program, but the details for implementation should be left to the individual utility. The emphasis should be on meeting the intent so as not to force a utility to change a well-working individual program solely for the purpose of standardization across the industry. The standard should have a balance of flexibility and specificity to avoid vague criteria that will lead to areas of varying interpretation and dispute. The current industry performance objectives, criteria, and guidelines developed by INPO allow the flexibility for individual utilities to meet the intent of the guidelines by meeting the criteria directly or by other appropriate means. One utility feels that it would be counterproductive to develop a minimum standard that could potentially lower the level of performance for the entire industry when only a few plants are experiencing problems. Another utility stated that a new rule or regulatory guidance will result in increased documentation, decreased flexibility to change and adjust programs as conditions or technology change, and decreased incentive for the maintenance staff to improve or enhance their maintenance capability. This could lead to a diversion of utility resources from safety-related activities and increase costs with minimal benefits. The commenters generally feel that any maintenance standard requiring an analysis of all SSCs for function and objective was practically unattainable and would significantly divert technical resources necessary for safe and reliable operation of a nuclear plant, with questionable benefit. Any standards, guidelines, or criteria should be tailored appropriately to the safety significance of the equipment being maintained and the function being performed. Response-As noted in the Commission response to Item 1, the final rule has been modified to establish a framework for evaluating the effectiveness of maintenance programs. As such, the rule describes the basic elements for measuring the effectiveness of maintenance and taking appropriate corrective action where maintenance is found to be ineffective. These elements include establishing goals, monitoring and assessment against these goals, feedback, and appropriate corrective 50-SC-85 action. The regulatory guide will be revised to reflect the rule's narrower focus on results and maintenance program effectiveness, and will describe a means for meeting the requirements of 10 CFR 50.65 acceptable to the staff. The rule and regulatory guide combination will provide a framework for evaluating the continuing overall effectiveness of maintenance, focusing on the objective of an effective maintenance program, while at the same time permitting licensees broad discretion and flexibility in the formulation and implementation of their individual maintenance programs. The rule does not require a monitoring program so broad in scope that it detracts from a licensee's ability to otherwise maintain its equipment. The extent of monitoring may vary from system to system, depending upon system importance to risk. Some monitoring at the component level may be necessary; however, it is envisioned that the majority of monitoring could be done at the system or train functional level. This monitoring requirement is not intended to duplicate activities currently being conducted which could be integrated with, and provide the basis for, the requisite level of monitoring. The Commission response to Question 7 has further details on scope and level of detail. 3. Is two years a reasonable time to develop and implement a standard? Comments-NUMARC and the utilities feel that two years was enough time to develop a standard depending on the scope of the BOP SSCs and components that need to be addressed. They stated that the systematic evaluation of all SSCs as described in the proposed rule alone would require more than two years. Most of the industry agrees that it would take two years to develop the standard and three to five years to implement it. One citizen's group feels that two years is too long for developing and implementing a standard; one year would be more appropriate. Response-During the time the Commission held rulemaking in abeyance, the industry developed and submitted INPO 90--008 to the Commission. The Commission also developed a regulatory guide that incorporated appropriate public comments. Furthermore, the MTis found that licensee maintenance programs have improved, and there are programs for improving maintenance developed by the industry. Therefore, the Commission believes that two years was ample time to develop and implement a standard. PART 50

STATEMENTS OF CONSIDERATION The Commission acknowledges that a systematic evaluation of SSCs could require as much as two or more years. Consequently, the final rule has a five year implementation schedule which allows at least three years for these evaluations beyond the time when final guidance is expected to be available.

4. Is it appropriate for a designated third party to certify plant maintenance programs to comply with the Maintenance Standard; if so, would an organization be willing to perform such certification?

Comments-Of the comments that addressed this question, most stated that it would be inappropriate for the NRC to delegate certification responsibility to a third party. The degree of opposition ranged from "not necessary" to "vigorously opposed." Most comments stated that third party certification would be unnecessary because existing measures that accomplish this function such as maintenance inspections and INPO evaluations. Some comments indicated that INPO could perform certification but not if a rule existed since that would place INPO in the position of a regulator. One respondent clearly stated that INPO should not be allowed to perform maintenance certifications for the NRC. Response-It was the Commission's intent to build upon industry initiatives to encourage good maintenance practices and common standards. A certification process against a maintenance standard by a third party was raised as an option that would have provided some degree of consistency and independence without relieving NRC of its regulatory responsibility to oversee the process. Because a viable third party certification process was not offered by the industry, the Commission is no longer pursuing this as an option. Additionally, as noted in Question 1, because the rule has been modified to de-emphasize programmatic requirements of maintenance, the Commission does not currently intend to formally endorse an industry maintenance program standard.

5. The Commission plans to issue by November 1989, a regulatory guide establishing standards and criteria for determining what constitutes an effective maintenance program. This regulatory guide is being developed in parallel with the final rulemaking.

The Commission encourages the industry to develop standards and acceptance criteria. If an acceptable industry standard is available in this timeframe, the Commission will consider endorsing the industry standard in the regulatory guide. An industry commitment to develop a maintenance standard, consistent with the Commission's schedule to issue a final regulatory guide by November 1989, would be necessary during this public comment period. Comments-Most respondents believe that issuance of a rule without public comment on a regulatory guide was inappropriate. Many feel that the most important NRC document concerning maintenance will be the regulatory guide and not the maintenance rule. Industry feels that the current standards as embodied in publications such as INPO 85-038 are sufficient and that a rule and regulatory guide are unnecessary. Several industry respondents said that they would be willing to participate with the NRC in developing a standard but that the November 1989 time constraint was unrealistic. Several respondents appeared to feel that the proper way to upgrade maintenance would be by first developing a regulatory guide and then a rule if use of the guide indicated that such a rule was needed. If the current industry standards were not enough, most feel that the NRC has the responsibility to develop the regulatory guide, though the industry respondents feel that they should have input to such a guide. INPO's position is that use of INPO 8~38 as a basis for a regulatory guide would be inappropriate. Response-The Commission believes that, by clearly putting forth a standard for an effective maintenance program in one document, guidance and stability would be provided to help ensure that the maintenance programs of all licensed plants achieve and maintain a satisfactory level of effectiveness. The Commission believes that the development of a standard by industry would support industry's current initiatives toward developing and implementing effective maintenance programs, and that utility participation in preparing a maintenance standard would provide additional experience, incentive, and responsibility for improving plant maintenance programs. The Commission was encouraged by NUMARC's submittal of INPO 90-008 as an industry maintenance standard. In acknowledgement of this document, the generally favorable results of the NRC's Maintenance Team Inspections regarding the adequacy of licensees' maintenance programs, and the many other industry initiatives in this area, the Commission revise the rule to emphasize the effectiveness or results of maintenance programs and emphasize the programmatic aspects of maintenance. Also, in acknowledgement 50-SC-86 of the generally satisfactory state of maintenance programs, the final rule provides great flexibility for the industry to continue developing, improving and implementing recommendations and guidance concerning maintenance programs. The Commission encourages such activities, especially as they support improvements in the evaluation of maintenance program effectiveness. However, because the rule has been modified to de-emphasize programmatic requirements of maintenance, the Commission does not currently intend to formally endorse an industry maintenance program standard. The Commission does not agree with commenters who suggested the issuance of a regulatory guide without a rule. The Commission desires to put forth requirements for evaluating the effectiveness of maintenance programs, including the issuance of implementing guidance, to clarify NRC regulatory purview and to provide additional enforceability. The revised regulatory guide will reflect the narrower, oriented focus of the rule. The details for the conduct of activities supporting maintenance will not be specified and should be developed by the licensee to ensure the adequate performance of plant equipment. Several guidelines exist in the industry (e.g., INPO 90-008 "Maintenance Programs in the Nuclear Power Industry," Institute of Nuclear Power Operations, and others sponsored by ANS, ASME, and EPRI) directed toward providing detailed recommendations for the effective conduct of maintenance activities. The industry is encouraged to continue the development and improvement of such guidelines and to standardize recommendations and guidance for plant maintenance programs.

6. The Commission believes that the proposed maintenance rule should be considered under 10 CFR 50.109(a)(4) of the backfit rule which would exempt the maintenance rule from backfit requirements based on the precepts that effective maintenance is necessary to assure adequate public protection and that the proposed rule codifies and standardizes previously existing Commission requirements, both explicit and implicit, in plant technical specifications, licensee safety analysis reports, and 10 CFR part 50, appendix B. The Commission requests public comment concerning the need for a backfit analysis for this rulemaking.

Comments-The nuclear industry commenters uniformly believe that a backfit analysis must be prepared for the maintenance rule. The most comprehensive responses were PART 50

STATEMENTS OF CONSIDERATION submitted by two nuclear industry groups: The Nuclear Utility Backfitting and Reform Group (NUBARGJ, and NUMARC. Many utility commenters endorsed NUMARC's response or repeated arguments made by NUMARC. A law firm, Conner and Wetterhahn.

also provided substantial comments that were generally consistent with those from NUMARC and NUBARG. In addition, a number of utility commenters joined in NUBARG's comments. The U.S. Department of Energy also agrees with the industry on a need for a backfit analysis. Only one commenter, Nuclear Information and Resource Service (NIRS), supported the Commission's position. NUBARG contends that the Commission "misapplied" the adequate protection exemption in the backfit rule in four respects. First, NUBARG asserted that the Commission prevented the public from reasonably commenting on the backfit issue by failing to specify whether it was relying on 10 CFR 50.109(a)(4)(ii), which exempts from analysis those rules that are "necessary to ensure that [a] facility provides adequate protection to the health and safety of the public," or the provisions of § 50.109(a)(4)(iii), which exempts those rules that involve "defining or redefining what level of protection to the public health and safety or common defense and security should be regarded as adequate." Next, after quoting from two passages in the notice of proposed rulemaking for the maintenance rule that suggest that the Commission is relying on both § 50.109(a)(4) (ii) and (iii), NUBARG appeared to contend that such reliance is logically inconsistent. No reasoned argument was presented by NUBARG in support of its contention, nor did . NUBARG specifically criticize the Commission's reliance on § 50.109(a)(4)(ii). Rather, NUBARG focused on§ 50.109(a)(4)(iii), arguing that the Commission's position that effective maintenance is necessary for adequate protection must logically rest on the presumption that none of the currently operating nuclear power plants do provide adequate protection. In any event, NUBARG also argued that the Commission's decision not to prepare a backfit analysis for the maintenance rule represents an unwarranted departure from the policies underlying the backfit rule-an "alarming retreat." Lastly, NUBARG argued that the Commission's reliance on the "adequate protection" exemption of§ 50.109(a)(4) is in "logical conflict" with the Commission's alternative ground that the rule is justified on the basis of the criteria contained in the backfit rule. NUMARC followed and expanded on NUBARG's arguments. NUMARC asserted that a backfit analysis is necessary solely because the maintenance rule would impose substantial new requirements on licensees and require the expenditure of significant resources by virtue of the maintenance rule's expansion of maintenance to the BOP. This argument was echoed by several other utility commenters. Next NUMARC attacked the Commission's assertion that the maintenance rule codifies and standardizes previously existing requirements by pointing out that the rule would require maintenance for SSCs in the BOP. NUMARC also followed the NUBARG reasoning that any redefinition of the standard of adequate protection to include maintenance must necessarily presume and admit that "all U.S. nuclear power plants are currently operating at a level below the 'adequate protection' baseline until they improve their maintenance program." Although NIRS agreed with the Commission that a backfit analysis need not be prepared for the maintenance rule, their agreement was partially couched on their position that the 10 CFR 50.109 is an invalid rule. Response-The Commission has determined to prepare a backfit analysis for the final rule. 7. The Commission believes that the inclusion of balance of plant (BOP) equipment in the proposed maintenance rule is necessary and proper. However, the Commission also recognizes that some licensee maintenance programs, as presently configured, apply to structures, systems, and components that are without question, irrelevant to protection of public health and safety from radiological hazards associated with the operation of the nuclear power plant. The Commission requests public comment concerning what limitation, if any, should be placed on the final maintenance rule to provide some licensee flexibility in this regard. Comments opposing including BOP equipment are summarized as follows: BOP equipment is outside the NRC's jurisdiction; the statutory jurisdiction of the NRC to regulate BOP components is limited to those BOP structures, systems, and comments that are related or important to nuclear safety; the economic impact of including nonsafety BOP equipment would be staggering; and the resulting improvement to safe operation of the plant would be disproportionate to the cost involved or 50-SC-87 could divert resources that would be more profitably spent on critical safety systems and components. The proposed rule did not define BOP SSCs, thereby not providing a meaningful opportunity for public comment. NRC should withdraw the proposed rule and develop a definition and a list of typical BOP SSCs that .are related or important to nuclear safety. BOP systems were not built to the standards of safety-related equipment and will not be capable of being maintained at the same level of readiness. For example, the proposed rule would require the proper maintenance of a component that is not required to be properly installed. However, if NRC proceeds with rulemaking and if BOP SSCs must be considered, it should be on a graded approach depending on a given BOP system's potential impact on safety functions. The utility must retain the ability to determine the requirements applicable to specific SSCs based on safety, reliability, and economic considerations. Instead of including all BOP SSCs, the rule must focus on the maintenance of functions whose failure would threaten public health and safety. Comments in favor of including BOP SSCs are summarized as follows: The maintenance rule should cover the whole plant. Unplanned reactor trips often originate in BOP systems. Furthermore, seemingly irrelevant parts of the plant can affect plant operations in unforeseen ways-for example, at Surry in the aftermath of the pipe break. Response-The Commission does not agree that maintenance of SSCs in the BOP is beyond the statutory jurisdiction of the Commission. Pursuant to section 161 and 182 of the Atomic Energy Act (AEA), the Commission has broad authority to protect the public health and safety, and the common defense and security and to minimize losses to life and property. Maintenance of SSCs in the BOP falls within this regulatory authority because such SSCs can and do have a significant effect on safety. With regard to safety, SSCs in the BOP have initiated transients and caused scrams and safety injection. Probabilistic risk assessments (PRAs) confirm that, for many plants, dominant accident sequences are initiated by transients in the BOP such as loss of offsite power or loss of feedwater. Therefore, to ensure that licensees operate safely, NRC's regulatory program is intended to ensure both a low frequency of transients that challenge safety systems and a high reliability of safety systems to respond to these challenges. This approach to regulation is part of the fundamental \ PART 50

STATEMENTS OF CONSIDERATION principle of defense-in-depth that underlies all NRC regulation. in-depth provides for both accident prevention and accident mitigation with principal emphasis on prevention.

Therefore, the Commission is well within its statutory jurisdiction in requiring that all SSCs that can significantly affect safety, including those in the BOP, be properly maintained. Indeed, the Commission's regulations already reflect the importance of maintenance of SSCs in ensuring adequate protection to public health and safety. Section 50.34(b][6)(iv) requires an FSAR to include the "plans for conduct of normal operations, including maintenance, surveillance, and periodic testing of structures, systems, and components." The Standard Review Plan (SRPJ (NUREG-0800), against which applicants for licenses after 1982 are required to evaluate their facility (see 10 CFR 50.34(q]J, requires applicants to evaluate a number of SSCs in the BOP, including design and installation as they affect safety. For example, the pressurizer relief tank system, which is "nonsafety related," is addressed in section 5.4.11 of the SRP. Of note is the rational for reviewing the design of the pressurizer relief tank: "The review is primarily directed toward assuring that its operation is consistent with transient analyses of related systems and that failure or malfunction of the system could not adversely affect essential systems or components is accordance with applicable criteria." Thus, the Commission has previously recognized that certain SSCs in the BOP can have a significant effect on safety and has exercised its regulatory authority by requiring the evaluation of the potential effect of nonsafety-related SSCs on safety. This is the same rationale for requiring maintenance of SSCs, including those in the BOP, that can significantly affect safety. The Commission agrees with the comments that the scope of the rule should be narrowed; not all of the BOP has the same safety significance. Accordingly, the scope has been modified to include only those BOP SSCs whose failure could most directly threaten public health and safety. Therefore, the scope of the rule has been modified as follows: The scope of the monitoring program * *

shall include safety related and nonsafety related structures, systems, and components as follows: (1) Safety related structures, systems, or components that are relied upon to remain functional during and following design basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shutdown the reactor and maintain it in a safe shutdown condition, and the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the 10 CFR part 100 guidelines.

(2) Nonsafety related structures, systems, or components: (i) That are relied upon to mitigate accidents or transients or are used in plant emergency operating procedures (EOPs); or (ii) Whose failure could prevent related structures, systems, and components from fulfilling their safety-related function; or (iii) Whose failure could cause a reactor scram or actuation of a safety-related system. This scope does not go beyond the jurisdiction of the NRC. This clarification of the scope should bound the scope, focus licensee resources on SSCs with the most safety significance, and reduce the cost impact projected by the comments. The Commission recognizes that BOP SSCs may have been designed and built with normal industrial quality and may not meet the standards in appendix B to 10 CFR part 50. It is not the intent to require licensees to generate paperwork to document the basis for the design, fabrication, and construction of BOP equipment not covered by appendix B. Instead, it is the intent to ensure that each licensee's maintenance program minimizes failures in those BOP SSCs that affect safe operation of the plant. In response to comments, security has been deleted from 10 CFR 50.65 as it is adequately addressed in § 73.46(g] and § 73.55(g).

8. The Commission believes that individual worker accountability plays an important role in an effective maintenance program. The Commission is, therefore, soliciting comments on the means for incorporating this consideration into a licensee's maintenance program. Comments-Respondents consistently agreed that worker accountability was an important and necessary part of a good maintenance program. Several of them gave examples for how their utility holds its employees accountable for their work. These examples all fell within the broad context of the personnel management system, i.e., selection, training, performance appraisal, supervision, promotional policies, etc. Most feel that rulemaking on worker accountability is impossible, unnecessary, or inappropriate.

Several cited the fact that worker accountability was a subject of negotiation between utility management and labor bargaining units. Several cited existing regulations (10 CFR part 2, appendix C, and 10 CFR 50.110) as already requiring worker accountability. One respondent said that the licensee should be responsible, not 50-SC-88 the worker. One respondent expressed a concern that a rule that included worker accountability would be interpreted as punitive by workers. Response-The Commission and industry have both recognized the importance of developing an attitude of accountability on the part of each and every worker in a nuclear power plant. The Commission agrees with industry that regulation of this area would be difficult to enforce objectively. The Commission concludes that each licensee should include considerations for emphasizing worker accountability based on local conditions; and the Commission will not attempt to deal specifically with this issue in the rule or regulatory guide. 9. The Commission desires to establish criteria within the maintenance rule which would form the basis for determining when a maintenance program is fully effective and additional improvement is not warranted from a safety standpoint. Such criteria might be either quantitative or qualitative and could be based on specific measurable attributes, on overall plant performance, on program results, or on other attributes. The Commission requests public comment concerning the need for such criteria, the form of such criteria, and the criteria themselves. Comments-Of the commenters that addressed this issue, most believe that quantitative indicators could not be used solely to evaluate effectiveness and that the determination of effectiveness was subjective. Further, the commenters believe that sufficient tools already existed in the form of SALP, QA assessments, regulatory inspections, monthly operating report data, and management reviews. One commenter noted that effectiveness needs to be defined in terms of a particular objective. Another stated that performance goals such as the number of maintenance-related reactor trips, LERs, etc., should be established. One individual commented that effectiveness needs to focus on functional failures affecting public health and safety; another suggested goals associated with general plant safety performance measures. Several commenters expressed concern that the lack of defined performance criteria could generate either complacency or a continuous ratchet since there would be no criteria for a "fully effective program." Response-The Commission agrees that determination of effectiveness depends on many factors and that, with regard to programmatic features, it is PART 50

STATEMENTS OF CONSIDERATION subjective.

The rule provides flexibility for each licensee to decide how to structure a maintenance program and conduct maintenance to achieve established performance goals. Specifically, the rule addresses (1) the development of licensee-established goals for performance, (2) the use of goals and other quantitative and qualitative means as a measure of the effectiveness of maintenance programs, and (3) the use of monitoring and assessment of equipment performance or condition against goals, or, alternatively, the demonstration of preventive maintenance effectiveness. In general, the Commission does not intend to define specific parameters or numerical criteria in either the rule or regulatory guide; each licensee is to establish appropriate goals to assist in monitoring the effectiveness of maintenance.

10. Are performance indicators that are being used by industry, may be used in the future, or have been used in the past, appropriate candidates as quantitative measures of maintenance effectiveness?

The Commission is particularly interested in experience or analysis concerning indicators or the use of indicators of component reliability as maintenance performance indicators. Comments-In addressing this item, NUMARC and most utilities stated that general plant performance indicators that have been developed and used by the industry were not appropriate for use as the sole effectiveness indicators because of the number of nonmaintenance-related factors included in them. Many of the proposed maintenance indicators are process indicators, which may or may not accurately reflect the state of the overall maintenance program. Such indicators are useful, .but only as one tool for management evaluation of the maintenance program. Although stating that there are presently no performance indicators in use by the industry that directly measure performance, NUMARC and the utilities recognized that some of the current industry indicators, taken in the proper context, can provide an indication of maintenance performance. Indicators can be used effectively by a specific utility as a management tool to assess the trend of performance within a given indicator or set of indicators. However, NUMARC admonished that there are individual plant variations that make absolute comparisons misleading, even for plants with the same licensee. NUMARC also stated that the comparison of plant-specific indicators to industry averages can be misleading. Two utilities stated that there was no need to develop new performance indicators. One added that the Commission should continue to evaluate a given licensee using its current technology. The other suggested that the existing INPO Performance Indicators be revised to meet the need for a maintenance standard. NUMARC expressed the opinion that a good maintenance program would use a combination of indicators based upon the condition, type, age, etc., of the plant and specific equipment in question. NUMARC believes that prescribing a rigid set of indicators would not achieve necessary plant flexibility and may preclude focusing on areas of more appropriate concern. Flexibility is needed to revise, delete, or add performance indicators as appropriate to provide information to management to fit circumstances, methods, and . conditions that may pertain to a given plant in a specific situation. In this vein, efforts to obtain consistent data would have questionable benefit for regulatory purposes and may have deleterious effects on plant programs. Another utility does not believe that any prescribed set of indicators can be used to judge the effectiveness of a plant's maintenance program. It also stated that no indicator or combination of indicators can give an overall measure of maintenance effectiveness. In its view, such a task must be left to the judgment of the individual licensee, INPO, and the NRC. One individual stated that maintenance effectiveness is a measure focused on economics. He went on to say that this view clouds the focus on public health and safety. According to this commenter, the proper focus of maintenance effectiveness is on functional failures that threaten public health and safety. NUMARC warned that component reliability by itself is not a good indicator of maintenance performance. The reason given for this position was that component reliability may be an indicator of an application, design, component, operating, or maintenance problem. NUMARC added that assessments by the plant staff or by the corporate staff, including observation of work in the field, are necessary ingredients in the measurement of maintenance performance. NUMARC pointed out that a given component failure or degradation could be allowable based on engineering judgment without indicating an ineffective maintenance program, especially for cases involving redundant or nonsignificant equipment. 50-SC-89 Response-The Commission agrees that plant performance indicators that have been developed and used by the NRC and industry are not appropriate as the sole indicators of maintenance effectiveness. The Commission also agrees that, because of individual plant variations, performance indicators are not appropriate for making absolute plant-to-plant comparisons. However, as recognized by commenters, indicators taken in context can be used as an indication of maintenance performance. More importantly, indicators can be used by licensees as an effective management tool to assess the need for corrective actions within a maintenance program. Operating characteristics such as consistently high availability or low equipment-caused forced outage rates over a number of operating cycles are indicators of good maintenance effectiveness. However, the plant material condition can degrade significantly before these indicators provide identification of degraded maintenance effectiveness; thus these indicators are not very timely. Based on the results of extensive work on indicator development, the Commission concludes that indicators that are based upon actual in-service component reliability and failure history provide a useful measure of maintenance effectiveness. Also, these indicators can be defined and implemented independent of the definitions and procedures that the licensee deems necessary to manage the flow of maintenance work. Knowledge of data showing component failure in excess of the industry average has the desirable property of alerting licensees to determine whether improved maintenance performance is needed. In general, the Commission agrees with NUMARC that a good maintenance program would use a combination of indicators based upon the condition, type and age of the plant and the specific equipment in question. Accordingly, the Commission has modified the final rule to allow licensees flexibility to determine the details of their individual maintenance programs.

11. Should an industry-wide component failure reporting system, e.g., NPRDS, be used by all plants in order to support the sharing of generic maintenance experience and facilitate monitoring of maintenance effectiveness?

Comments-Of the commenters, including NUMARC, who addressed this item, most recognized the usefulness of the NPRDS as a source of generic failure data. However, most of the commenters, \..4 PART 50

STATEMENTS OF CONSIDERATION including NUMARC, oppose the unqualified use of the NPRDS for monitoring maintenance effectiveness for a number of reasons. Some commenters, including NUMARC, perceive such use of the NPRDS as an inappropriate regulatory intrusion into a program designed to improve communications regarding

!!quipment performance within the industry that would tend to stifle the free exchange of information. NUMARC cited the necessary expansion of the reportable scope of the NPRDS to cover the entire BOP as a tremendous undertaking that could be prohibitively expensive. NUMARC, two utilities, and one individual believe that, although the NPRDS can be used to obtain gross indications of a problem, its usefulness is restricted because of plant-to-plant differences in maintenance practices, component application, design, environment, and the detail with which failures are reported. Response-The Commission generally agrees with the above comments. However, the NPRDS may provide useful information for comparing specific experience on equipment with a broader range of industry operating experience on similar equipment. The data does provide useful insights into maintenance trends at an individual plant. 12. Commissioner Roberts had the following views: I cannot join the majority in supporting the proposed rulemaking on maintenance. In order to have the benefit of the public's comments, it has been my custom to agree to publication of proposed rulemakings. I cannot do so in this instance. I have asked one fundamental question. What are we trying to accomplish with this rule that cannot more effectively and innovatively be accomplished without a regulation? I have not received a satisfactory answer. I do not believe the case has been made that licensees do not have

established maintenance programs.

Most importantly to me, there has been no demonstration that this rule would improve implementation of existing programs. Neither have I been provided with compelling documentation on what the problem is and how, specifically, this rule will fix it. On the contrary, the trends staff has provided show continued improvement in the maintenance area. The proposed rule the Commission is now publishing fails to provide a basis for determining when a maintenance program is effective or when improvements are "appropriate." We are even delaying publication of the accompanying regulatory guide until the final rule. Without being afforded the opportunity to review this implementation document, the Commission is left in the position of approving a specious rule. It is no wonder that this rulemaking would elicit such widespread opposition. The public is being asked to comment on a rule of form but no substance. I believe it would be more productive to delay issuance of this proposed rule until the draft regulatory guide is available for comment. Only then can we receive meaningful comments on the rulemaking package. , I am concerned that this rule goes beyond our authority. I cannot agree with a rule that would have the NRG regulating maintenance on all systems, structures, and components regardless of whether they have a nexus to radiological safety or not. I am troubled by the attitude demonstrated when we request public comments on what limitations, if any, should be placed on the final rule to address structures, systems and components that are "without question irrelevant (my emphasis) to the protection of public health and safety." This clearly abdicates our responsibility to show that a regulation is needed. We must ask ourselves: Are we proceeding with this rulemaking for the sake of the rule itself? As a !tested to by the cases where the Commission cited licensees, the NRG already has the authority to enforce compliance in the maintenance area. The arguments advanced by both the staff and the Commission in trying to comply with the requirements of the backfit rule have played a significant role in my decision not to support this proposed rulemaking. The staff argument for the rule's compliance with 50.109 has been made on the basis of cost. The staff states that the backfit analysis shows that"* *

the rule will provide a substantial increase in the protection of the public health and safety without any additional cost." I am skeptical of the assumptions made in the backfit and regulatory analysis and request comments on both these documents.

I also request comments on the views of the ACRS. They state that"* *

there are characteristics of regulations, and especially the way in which they are typically enforced, that lead us to believe that, under a rule, a move toward uniformity would occur, and this is likely to decrease the effectiveness of some of the better existing programs." I share their concern that the existence of this rule could make things worse and diminish rather than enhance the protection of the public. Regarding "adequate protection," the Commission appears to be saying that since effective maintenance is necessary to maintain adequate protection, this rule should be excepted under 50.109(a)(4).

This exemption would prohibit staff from taking implementation costs into consideration. However, it would require that a documented evaluation be prepared for public comment. Therefore, my opposition to the exception is not to the exception itself but to the precedential nature of the use of the adequate protection argument. Let me state that I, too, strongly believe that effective maintenance is necessary to assure that nuclear power plants are safe and to provide adequate protection to the public. I also believe, just as strongly, that this rule is not necessary to provide that protection, and that as the ACRS noted, it may well have the opposite effect. I believe that we cannot afford to be careless about the use of the "adequate protection" argument for exception to the backfit rule. The Commission is in litigation about this 50-SC-90 very issue. The Commission addressed this point in detail under the heading "Adequate Protection" in the Response to Comments on the final 10 CFR part 50 Revision of Backfit Process for Power Reactors. Let us remember that there had been concerns that in dealing with the backfit rule, the Commission would use the phrase "adequate protection" arbitrarily. The Commission could unwittingly be giving credence to that view. Additionally, it seems to me that the Commission position on adequate protection is internally inconsistent. The Commission needs to recognize that when it states that this rule is needed to maintain adequate* protection, it is saying that the current operating plants now pose undue risk to the public which we are presently tolerating. If I believed that, I would suggest (as I'm sure would the rest of the Commission) that this rule become immediately effective. This is clearly not the case. As the Commission in the very same comment shows, "* *

the proposed rule codifies and standardizes previously existing (my emphasis)

Commission requirements, both explicit and implicit, in plant technical specifications, licensee safety analysis reports, and 10 CFR part 50, appendix B." It seems to me that the Commission can't have it both ways. I request comments on my views. Comments-Of the commenters who responded to this question, most agreed with the views of Commissioner

Roberts, while only three commenters disagreed with the Commissioner.

Some commenters did not provide any basis for their agreement or disagreement. However, a number of commenters expressed concerns beyond the views expressed in Question 12. These are summarized below. A majority of the utility commenters implicity agreed with Commissioner Roberts that the proposed rule went beyond the current authority of the Commission by requiring maintenance of all SSCs in the BOP. According to these commenters, since many SSCs.in the BOP have no nexus to pubic health and safety, the maintenance rule would require licensees to spend their resources on unimportant areas, potentially decreasing the level of safety. One individual stated that regulators have a bias in favor of overboard regulations, pointing to the FAA's regulations on air transportation. This commenter noted that, unlike the scope of FAA's statutory charter which encompasses the development of the air transportation industry, the NRC's authority is limited to the regulation of the nuclear industry to protect public health and safety. Two utilities argued that the maintenance rule fails to provide meaningful definitions and standards of the activities required. In their view, this can lead to misinterpretation, arbitrary enforcement, and endless PART 50

STATEMENTS OF CONSIDERATION reinterpretations of the rule. One utility suggested that any industry standard on maintenance would be tailored to the lowest common denominator, and therefore there would be no net improvement in the level of safety. It also argued that, once codified, a
regulatory standard of acceptance maintenance would be difficult to improve. Finally, NUMARC and the utilities also repeated their general arguments why a maintenance rule is not necessary, in particular, on the gradual improvement in the industry maintenance performance, and the INPO Self-Assessment Program. NUMARC also asserted that the Commission has sufficient authority to ensure adequate protection.

A Commissioner on the Public Service Commission of the State of Vermont stated that there is safety significance in the BOP, pointing out that recent NRC staff and industry evaluations show that improper maintenance of components not previously associated with safety has resulted in adverse safety consequences. In addition, the Commissioner indicated that superior performance of nuclear plants internationally has been associated with maintenance programs that are stricter than those in the U.S., citing the experience of Japan and France. Response-Two of the issues raised by Commissioner Roberts and by the majority of commenters are similar to those issues raised in response to Questions 6 and 7. As discussed in the response to comments on Question 6, the Commission agrees that a backfit analysis is required for the maintenance rule. Because the current regulations provide an assurance of adequate protection of the public health and safety, the Commission is no longer proposing to exempt the maintenance rule from the requirements of a backfit analysis. The Commission does not agree that the maintenance rule will result in decreased safety by requiring licensees to divert their resources away from SSCs and activities with greater importance to safety. The maintenance rule is being issued to ensure that the effectiveness of maintenance programs is maintained for the life of the facility and is not expected to require significant modifications to current licensee programs. The regulatory guide will provide flexibility for a licensee to structure its maintenance program in accordance with the safety significance of those SSCs. However, the Commission does agree with the comments that not all SSCs in the BOP are rela led to the protection of public health and safety. Accordingly, as discussed in the response to the comments on Question 7, the scope of the rule has been modified to focus on those SSCs whose failure could most directly threaten public health and safety. Finally, during the time the Commission held rulemaking in abeyance, the public had the opportunity to comment on the draft regulatory guide. Considering the narrowing of the focus of the final rule to a results/performance-oriented approach, the supporting regulatory guide will require revision. During the revision process, previous public comments will be considered and appropriately reflected in the regulatory guide. The regulatory guide will be revised to reflect the rule's narrower focus on results and maintenance program effectiveness, and will describe a means for meeting the requirements of 10 CPR 50.65 acceptable to the staff. Revision of the regulatory guide will again include the opportunity for public comment. Implementation of the rule is to be delayed for five years after the issuance date, with the regulatory guide expected to be available within the first two years. This schedule will allow at least three years for licensee development beyond the time when final guidance is expected to be available. Additional Comments of Commissioner Curtiss I believe that the approach adopted by the Commission in this final rule is sound and appropriate. The entire Commission agrees that it is important for this agency to have a regulatory framework in place that will provide a mechanism for evaluating the overall continuing effectiveness of licensees' maintenance programs. This final rule will provide that regulatory framework. I strongly disagree with those who contend that the Commission rushed out with this maintenance rule without the benefit of public comment and with the attendant implication that the final rule was not well-considered. In point of fact, the reliability-based aspects of maintenance reflected in this final rule have been at the very heart of what the Commission has been considering in the maintenance area since as early as 1988. Indeed, it is abundantly clear from even a cursory review of the history of this issue that considerable time and attention have been devoted to the basic concepts reflected.in this final rule. That history is briefly summarized below: In the Final Commission Policy Statement on Maintenance of Nuclear Power Plants [53 FR 9430; March 23, 50-SC-91 1988), the Commission made it clear that-[i]t is the objective of the Commission that all components, systems and structures of nuclear power plants be maintained so that plant equipment will perform its intended function when required. To accomplish this objective, each licensee should develop and implement a maintenance program which provides for the periodic evaluation, and prompt repair of plant components, systems and structures to ensure their availability

* *. [T]he program should include the feedback of specific results to ensure corrective actions, provisions for overall program evaluation, and the identification of possible component and system problems ... An adequate program should consider
Technology in the area Predictive Maintenance
Equipment history and trending [and]
Measures of overall program effectiveness The Commission went on to indicate in that same 1988 Policy Statement that-The Commission expects to publish a Notice of Proposed Rulemaking in the near future that will establish basic requirements for plant maintenance programs.

We believe that the contents and bounds of the proposed rule will fall within the general framework described in this Policy Statement

* *. We encourage interested parties to provide their views on this important subject to the Commission, even at this early stage of the rulemaking process. 53 FR 9430-31. Thus, early on, the Commission began to consider the principal elements of the final rule adopted here by the Commission, called on licensees to incorporate those elements into their maintenance programs, and solicited public comment on such proposals.

In conjunction with the issuance of the Final Commission Policy Statement on Maintenance of Nuclear Power Plants, the Commission directed the NRC staff to develop a preferred maintenance rulemaking option requiring licensees to track certain maintenance performance indicators [See Staff Requirements Memorandum on COMKC-88-03, June 17, 1988). In response, the staff advised that the proposed rules should contain "provisions for performance assessment which licensees would implement to track the effectiveness of their maintenance programs" [See SECY-81::1-277, Amendment to 10 CPR part 50 Related to Maintenance of Nuclear Power Plants, p. 2, September 30, 1988). Although the staff was not in a position to suggest the use of specific PART 50

STATEMENTS OF CONSIDERATION maintenance performance indicators, it formulated a proposed rule that-emphasizes that an integral part of a good maintenance program is the monitoring and feedback of results. In this regard, the maintenance programs should utilize quantitative indicators that are based upon actual component reliability and failure history to provide the best measure of maintenance effectiveness.

SECY-88-289, Preliminary Results of the Trial Program on Maintenance Performance Indicators, p. 5, October 7, 1988. Indeed, the staff specifically noted that the goal of the recommendations contained in the proposed maintenance rule was to provide the NRC staff and licensees "with a practical near-term method to track maintenance effectiveness

* *" (SECY-88-289, p. 5)-the very core of the proposal that the Commission endorses in this final rule. The resulting Notice of Proposed Rulemaking on Maintenance and the proposed rule published for comment on November 28, 1988 (53 FR 47822) contain the same equipment history and trending, effectiveness monitoring, and feedback elements as the Final Commission Policy Statement on Maintenance.

They also contain clear indications that the Commission intended to include requirements for monitoring, trending, and feedback with regard to the effectiveness of maintenance in any maintenance rules that might ultimately be adopted. The need for, and details of, such provisions were emphasized in the draft Regulatory Guide that was subsequently published for comment as part of this maintenance rulemaking effort. 54 FR 33983. In turn, a number of commenters acknowledged the maintenance effectiveness measurement, trending, and feedback aspects of the proposed rule and provided their views on these matters. In sum, it is abundantly clear from all of this that the Commission has long been considering maintenance effectiveness monitoring of the sort that a majority of the Commission now adopts in this final maintenance rule, and that the industry and the public were given clear notice and the opportunity to comment on such considerations throughout this maintenance rulemaking process. The final rule that has resulted from this careful deliberation will provide the regulatory framework that all Commissioners agree this agency must have in order to ensure the continuing effectiveness of maintenance efforts at nuclear power plants, while at the same time providing licensees broad latitude in how they fashion their individual maintenance programs. Commissioner Remick's Separate Comments I respectfully differ with my colleagues inasmuch as I do not believe that there is a demonstrated need for a rule in light of significant improvements in maintenance programs resulting from Agency attention and licensee initiatives. The Commission indicates in its decision to promulgate this rule that"* *

the Commission is satisfied that the industry has been generally successful in bringing about substantial improvement in maintenance programs." Substantial improvements and favorable results are the goals that the Commission should strive for in its regulatory activities by utilizing the most effective regulatory tools for accomplishing those goals. As I argue below, I am not convinced that in this case a rule is the most effective regulatory tool for accomplishing those goals. Further, I differ inasmuch as I strongly believe that this rule should not be issued as a final rule. Although the rule is a concept worthy of discussion, it should not have been rushed out but should have been issued for the benefit of public comment. The Commission approved criteria to be used in determining when industry progress in the area of maintenance would be sufficient to obviate a need for rulemaking

[SECY memorandum from S. Chilk to J. Taylor, dated May 25, 1990). The staff performed a detailed evaluation of industry progress and concluded that the criteria had been satisfied (SECY-91-110, Staff Evaluation and Recommendation on Maintenance Rulemaking). Based upon its conclusions, the staff recommended that the Commission not proceed with a maintenance rulemaking. The ACRS agreed with the staffs recommendations. In general, I agree with the bases for the staffs conclusions. Therefore, I approved the staffs recommendation in SECY-91-110 not to proceed with maintenance rulemaking, but instead to issue a final policy statement on maintenance of nuclear power plants. I also approved the staffs recommendation to remove the maintenance escalation factor and revise the enforcement policy supplement of 10 CFR part 2, appendix C to include a specific maintenance-related example. Further, I agree with the staffs conclusion that the industry document, INPO 90--008, "Maintenance Programs in the Nuclear Power Industry," delineates the necessary elements of effective maintenance programs. The industry's commitment to monitor the progress of maintenance implementation using the performance objectives of INPO 91}-008, and the staffs intention to assess industry performance and report to the Commission after four years with an interim report after two years, are sufficient in my view to assure that there will be no backsliding of the level of industry performance of maintenance. In general, I support a regulatory approach which stimulates licensees' and industry's initiatives, encourages innovation, permits self-management and produces positive results, under agency monitoring, in contrast to prescriptive, process-oriented regulations 50-SC-92 which require rote adherence, stifle initiatives and depend on punitive enforcement actions for compliance. There appears to be a near-unanimous consensus that the agency and the industry have stimulated initiatives which have produced positive results, an outcome not necessarily assured even by result-oriented rulemaking. I agree with the view that routine use of the staffs maintenance inspection approach, utilizing the Maintenance Team Inspection (MTI) Criteria proposed in conjunction with the revised policy statement, could ultimately lead to essentially the same prescriptive result as a process-oriented rule. In the interest of ensuring that the responsibility for improving, sustaining and verifying adequate maintenance performance (using industry's standard document INPO 90--008) remained with the industry, I believe that the Commission should have directed the staff to develop an approach to its routine inspections which would have concentrated on inspecting for the effective results of maintenance programs rather than inspecting the details of the process. The MTI approach would then have been reserved for use as diagnostic inspection tool in those special cases where there was a perceived maintenance problem. In my approach, the staffs proposed final policy statement on maintenance would have been revised to include these future activities. I agree with the view that it is important for this agency *10 have a regulatory framework in place that will provide a mechanism for evaluating the overall continuing effectiveness of the maintenance programs, particularly as the plants continue to age. I believe that a revised final policy statement, together with the development of oriented inspection programs, would have provided an effective regulatory framework for such evaluation. I believe that the performance-based rule that the majority of the Commission has approved has some innovative features, and may be particularly appropriate for monitoring the effectiveness of maintenance programs for the advanced reactors. However, I do not agree with the view that the proposed rule in no way interferes with the process-related activities which the licensee community, to its considerable credit, has undertaken voluntarily. It may be argued that licensees will not have to change their maintenance programs to meet the provisions of the rule as it is written. Nevertheless the focus of the NRC's attention on implementation of a new rule almost always carries with it the strong potential for impact on the licensees' initiatives and programs and thus an inherent disincentive to not innovate or participate in new initiatives. One way of determining the potential impact of this rule would have been to issue it for public comment. I think that issuing the proposal for public comment would be good policy, and consistent with the Commission's Principles of Good Regulation, which state that all available facts and opinions be sought openly from licensees and other interested members of the public. To rush a final rulemaking package that contains some fundamental changes from the direction the PART 50

STATEMENTS OF CONSIDERATION Commission has taken over the past several years, without seeking all available facts and opinions, is likely to lead to implementation problems that the Commission may not be aware of now. The final rule represents a significant departure from the proposed rule. The proposed rule issued in 1988 focussed on what the Federal Register notice for the proposed rule called "maintenance practices" and "the adoption of common maintenance standards"-in a word, "processes", or "systems" of maintenance (53 FR 47824). The notice stated that "regulation

[of maintenance] by outcomes rather than processes" would be the subject of on rulemaking" (id.). The final rulerhowever, is focussed on outcomes and thereby seems to have concluded the "follow-on rulemaking" before it was begun. Although the proposed rule contained monitoring and trending components, they were only a few among seventeen maintenance activities covered by the proposed rule (see the proposed 50.65[b)), and so clearly were in no way intended as a surrogate for a oriented rule. However, monitoring is the focus of the final rule. The significant shifts in the focus of the rule and in the role of monitoring in the rule deserved public comment. The notice of the proposed rule invite responses to questions on monitoring, but the questions were confined largely to the issue of what specific measures might be used to assess the effectiveness of a maintenance program [see 53 FR 47825). Not addressed in the notice were certain matters which are crucial to the final rule. These include, for example, the final rule's requirement to monitor "against licensee-established goals" which are "commensurate with safety". Also, § 50.65[b) of the final rule defines the structures, systems, and components (SSCs) to be included in the scope of maintenance monitoring programs. This definition is both similar to and different from the definition of SSCs important to license renewal in part 54, a final rule which the Commission affirmed along with the final rule on maintenance. Public comment might have addressed whether the differences between the definitions of SSCs in these two maintenance-related rules are justified or will present interpretation and implementation problems. If I were convinced that a rule was needed to produce positive results, I could support the majority's rule as a proposed rule, provided that I could see how the staff would implement the rule through the development of regulatory guides and inspection modules, and provided that the public was given an opportunity to comment before promulgation of a final rule. But I am not convinced that a rule is needed to produce positive results. The staff has shown that we're seeing substantial positive results of the industry's maintenance program initiatives, and the stafrs findings have been verified in my discussions with Regional staff and Resident Inspectors. Therefore, I have concluded that the Commission should not change its direction now and that there is no need to promulgate a maintenance regulation which could be counterproductive to further maintenance program development and innovation. I fear that licensees will halt further development of their maintenance initiatives to await the development of the regulatory guidance to implement the rule, and that licensees will refrain from participating in future safety initiatives because they will interpret this Commission action as a significant retreat from its goals of achieving a stable regulatory environment. The development of an industry maintenance program standard, the industry's commitment to self-assessment against that standard, INPO's evaluation of maintenance progress against the objectives of the standard, NRC inspection programs which would concentrate on effective results, and the NRC's existing enforcement authority are adequate to ensure proper maintenance without a new rule. I would stress, however, the importance of the Commission's continuing to monitor the industry's progress in this area. A policy statement would be a suitable approach for continuing the Commission's necessary emphasis on maintenance, and at the same time allowing for continuing improvement in maintenance through flexibility, diversity and innovation in the industry's programs. Finding of No Significant Environmental Impact: Availability The Commission has determined that, under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 CFR part 51, this rule is not a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. Since this action is directed toward maintaining the level of maintenance effectiveness of existing plant SSCs to minimize the likelihood of failures and events caused by the lack of effective maintenance and does not require any modification of the plant, it will not adversely affect the quality of the human environment. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW., Lower Level, Washington, DC. Single copies of the environmentai assessment and finding of no significant impact are available from Robert Riggs, Office of Nuclear Regulatory Research, Telephone: (301) 492-3732, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). The information requirements will be submitted by the NRC to the Office of Management and Budget (0MB) for review and approval 50-SC-93 of the information requirements before they will become effective. Notice of NRC submission of the information collection requirements to 0MB, and issuance of the required 0MB approval, will be published by the NRC in the Federal Register. Regulatory Analysis The Commission has prepared a regulatory analysis on this final regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The analysis is available for inspection in the NRC Public Document Room, 2120 L St., NW., Washington, DC. Single copies of the analysis may be obtained from Robert Riggs, U.S. Nuclear Regulatory Commission, Washington, DC 20555, (301) 492-3732. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission hereby certifies that this regulation does not have a significant economic impact on a substantial number of small entities. This regulation affects licensees that own and operate nuclear utilization facilities licensed under sections 103 and 104 of the Atomic Energy Act of 1954, as amended. These licensees do not fall within the definition of small business set forth in section 3 of the Small Business Act, 15 U.S.C. 632, or within the Small Business Size Standards set forth in 13 CFR part 121. Backfit Analysis Pursuant to 10 CFR 50.109(a)(2), the Commission has prepared the following backfit analysis for the maintenance rule. The Commission has determined, on the basis of this analysis, that backfitting of the requirements in the maintenance rule will provide a substantial increase in the level of protection of public health and safety beyond that currently provided by the Commission's regulations, and that the costs of implementing the rule are justified in view of this increased protection. The maintenance rule requires licensees to monitor the effectiveness of maintenance activities for certain structures, systems and components based upon licensee-established goals for performance or condition, and take corrective action where necessary (the requirements of the maintenance rule are set forth in greater detail in the discussion below which addresses the nine factors of 10 CFR 50.109(c)). It is the Commission's judgement that maintenance, and in particular th~ goal-PART 50

STATEMENTS OF CONSIDERATION setting, monitoring and corrective action activities required by the maintenance rule, provide a substantial increase in the safety of nuclear power plant operation.

This judgement is based on the direct impact of maintenance on the reliability and operability of nuclear power plant safety systems, and its effect on the other plant structures, systems and components that are important to the protection of the public health and safety and common defense and security. The Commission's judgement that effective maintenance is an important contributor to safety is confirmed by studies of maintenance practices for domestic nuclear power plants, LERs, composite data from the Commission's Systematic Assessment of Licensee Performance (SALPJ, and the Commission's inspections at domestic nuclear power plants, as well as studies of maintenance practices at foreign nuclear power plants, the military, and the aerospace industry. The Commission first began focusing on maintenance as a result of its observation that plant performance, as reflected in such indicators as the number of unanticipated scrams, was not improving in the early 1980s. The Commission had expected that as newly licensed power plants gained operating experience and took advantage of lessons learned and other information distributed throughout the industry, problems in plant operation would gradually decrease to a relatively low level. To understand why industry performance was not improving as expected, the Commission performed an assessment of maintenance at domestic nuclear power plants in NUREG-1212, "Status of Maintenance in the Nuclear Power Industry." The study found that in 1985, maintenance safety problems were evident to varying degrees across the U.S. nuclear industry. Wide variations were found in maintenance practices and effectiveness, and a significant proportion of operational problems was found to be attributable to improper or inadequate maintenance. This finding was confirmed by an industry study of maintenance conducted about the same time. This industry study, which was performed by NUMARC Working Group 4, was discussed by the Working Group Chairman during the July 1988 Public Workshop on the Maintenance Rulemaking [NUREG/CP-0099, pp. 1.21-1.31). The industry study found that 38% of the root causes of 650 significant events examined were maintenance related. To obtain a broader perspective on maintenance, the Commission performed a survey and assessment of maintenance practices in other countries and industries to identify varying approaches to maintenance and to determine if there was any linkage between safety and effective maintenance. Specifically, the aim of the study (NUREG-1333) was to:

Review various regulatory approaches and determine their applicability to the maintenance rulemaking, and
Determine foreign and domestic maintenance practices that contribute significantly to effective maintenance.

The study covered Japanese, French, and German (FRGJ nuclear maintenance regulations and practices; the Federal Aviation Administration's regulatory approach to the maintenance of U.S. commercial aircraft; and the maintenance programs of the U.S. Navy and Air Force. The results of the study were used in formulating the proposed rule. These studies confirm the Commission's view that good maintenance is correlated with high reliability and minimization of plant transients, and therefore with nuclear power reactor safety. An additional concern of the Commission is the need to assure effective maintenance at nuclear power reactors throughout the terms of their . operating licenses (and any renewed operating licenses). While the current performance of the nuclear power in_dustry in the area of maintenance is acceptable and improving in the aggregate, the NRC Staffs Maintenance Team Inspections indicate that there are still common weaknesses in discrete areas of maintenance at nuclear power plants. Thus, while the Commission acknowledges the increased emphasis by licensees on maintenance and significant improvement in performance of maintenance programs in the aggregate, additional attention is warranted. Moreover, in the absence of a rule, there is no assurance that licensees would not relax their commitment to effective maintenance practices in the future. In this regard, the Commission notes that no licensee has made a formal docketed commitment to implement the Institute for Nuclear Power Operations (INPOJ performance objectives and criteria on maintenance (INPO 90-008). By adopting a maintenance rule now, the Commission will have a regulatory basis for preventing licensee "backsliding" in the area of maintenance. The absence of Commission maintenance requirements covering a 50-SC-94 broad scope of .structures, systems and components also represents a safety concern because of the potential adverse effect on the ability of the Commission to take timely and effective regulatory action against licensees with poor maintenance practices. It is true that there are a number of existing Commission requirements that are directly or indirectly relevant to maintenance, including 10 CFR 50.34(a)[3J[i); 50.34(a)(7); 50.34(b)(6) [i], (ii), (iii) and (iv); 50.34(b)[9); 50.34[f)[l) (i), (ii), and [iii); 50.34(g); 50.34a[c]; 50.36[a); 50.36[c) (2), (3), (5) and (7); 50.36a[a)[l); 50.49(b); 50.55a(g); part 50, appendix A, Criteria 1, 13, 18, 21, 32, 36, 37,40,43,45,46, 52,53;part50, appendix B. However, these requirements do not apply uniformly to all "safety-related" structures, systems and components, and only occasionally apply to structures, systems and components which could adversely affect the functioning of safety-related structures, systems and components. Any attempt on the part of the NRC to take regulatory action against a licensee with inadequate or poorly-implemented maintenance must be pursued on an individualized, case-by-case consideration of the adequacy of that licensee's maintenance practices and their effect on safety. This regulatory approach is costly in terms of agency resources. It also risks the possibility that the NRC will be unable to take timely enforcement action in the event of a finding of inadequate licensee performance in maintenance. By contrast, timely regulatory action could easily be taken if a licensee were found not to be implementing specific actions required by a rule which addresses maintenance. In sum, the Commission concludes that substantial safety benefits are to be achieved from adopting the final maintenance rule. The Commission also concludes that the costs of implementing the maintenance rule at all nuclear power plants are justified in view of the safety benefits identified above. A regulatory analysis has been prepared to assist the Commission in determining the benefits and costs of implementing the maintenance rule through a quantitative approach. However, the quantitative estimates in the regulatory analysis have proved to contain varying degrees of uncertainty. Depending upon the specific assumptions used in the analysis, a broad range of values is possible for the estimated risk reduction attributable to the maintenance rule [the uncertainties and their effect on the overall risk reduction and value/impact ratios are discussed in greater detail in PART 50

STATEMENTS OF CONSIDERATION the regulatory analysis).

Because of these uncertainties, the Commission has considered qualitative safety considerations and benefits. Thus, the regulatory analysis' quantitative estimates comprise a component of, but are not the primary factor with respect to the Commission's conclusions on the safety benefits and costs attributable to the final maintenance rule. The regulatory analysis estimates that implementation of the final maintenance rule could result in a point estimate of 52,000 person-rems avoided, with an upper bound of 72,000 and a lower bound of 7,300 person-rems. The net costs associated with implementation of the maintenance rule are estimated to entail a point estimate of 44 million dollars, with an upper bound of 2100 million dollars in cost savings and a lower bound of 1500 million dollars. The resulting value/impact ratio is a point estimate of 1200 person-rems/million dollars. Furthermore, the regulatory analysis for the maintenance rule also contains some conservatisms which the Commission believes underestimates the cost-effectiveness of the final maintenance rule. In the regulatory analysis, it was assumed that the damage frequency and forced outage downtime reductions associated with the results-oriented rule would be the same as those for a process-oriented rule. However, the Commission believes that the results-oriented approach, by focusing to a greater extent on equipment performance, would be more likely to achieve additional reductions in core damage frequency and forced outage downtime. The regulatory analysis also assumed that licensees under the final results-oriented rule would incur most of the costs of implementing programmatic elements similar in scope to those contained in the 1988 proposed maintenance rule in addition to the costs of implementing the results-oriented elements which were drawn from the proposed maintenance rule and incorporated into the final rule. The Commission projects that because the results-oriented rule is not a prescriptive programmatic rule, licensees will achieve some cost savings because they will have flexibility in determining the manner in which to improve the programmatic elements of their maintenance programs. Accordingly, the Commission projects that the costs for the performance-based final maintenance rule will be somewhat smaller than that assumed in the regulatory analysis. In view of the safety benefits discussed above, the Commission judges that the costs of implementing the maintenance rule are justified. The Commission recognizes that regulatory action in the area of maintenance should not be overly prescriptive, but rather be carefully directed to ensuring that unnecessary activities are not required, in view of the large degree of uncertainty in quantifying the costs and benefits of the maintenance rule. Accordingly, the final maintenance rule is carefully tailored to eliminate prescriptive programmatic, procedural arid organizational requirements. Rather, the final maintenance rule represents a oriented approach to assuring that maintenance is effectively conducted at nuclear power reactors.The licensee is responsible for establishing goals for structure, system and component performance or conditions, and the licensee is free to determine the monitoring method, the need for corrective action, and the nature of that action. Furthermore, the maintenance rule contains a provision(§ 50.65[a)(2)) whereby licensees may forego monitoring. The Commission believes that the final maintenance rule provides the necessary flexibility for licensees to tailor their maintenance programs to their specific plant design and configuration, organizational structure, and personnel, thereby permitting compliance with the maintenance rule in the most cost-effective manner. The Commission is confident that the regulatory goal of maintaining safety has been achieved in the most reasonable and cost-efficient manner and is consistent with the public interest. For the reasons set forth above, the Commission concludes that, the maintenance rule will result in a level of safety beyond that currently provided by the Commission's regulations and that is a substantial increase in the overall protection of the public health and safety, and that the net costs of the rule are justified in view of this increased level of safety. The nine factors listed in 10 CFR 50.109[c) are discussed below. 1. Statement of the specific objectives that the backfit is designed to achieve. The purpose of the maintenance rule is to maintain the effectiveness of maintenance at operating nuclear power reactors, thereby maintaining the level of safety at operating nuclear power reactors.

2. General description of the activity required by the licensee or applicant in order to complete the backfit. Under § 50.65(a)(1) of the maintenance rule, licensees will be required to: (i) 50-SC-95 Establish goals for the performance or condition of certain structures, systems and components to assure that they will meet their intended function, (ii) monitor these structures, systems and components to determine whether the licensee-established goals have been met, and (iii) take appropriate corrective action if the goals are not met. These goals are to be established by taking into account industry-wide operating experience.

Monitoring is not required, however, where the licensee demonstrates that preventive maintenance is sufficient to assure that the structures, systems and components will remain capable of performing their intended functions. See § 50.65(a)(2). Licensees will be required to evaluate the effectiveness of their goal-setting, monitoring and corrective action activities on at least an annual basis, taking into account industry-wide operating experience, and adjust their programs where necessary to ensure that failure prevention is balanced against unavailability of structures, systems and components. See § 50.65(a)[3). In addition, when performing monitoring and preventive maintenance activities, an assessment of the total plant equipment service should be taken into account to determine the overall effect on performance of safety functions. See § 50.65(a)(3). The structures, systems and components which are subject to the goal-setting, monitoring, and corrective action requirements of the rule are those which are safety-related, and certain non-safety related systems, structures and components as defined in § 50.65(b).

3. Potential change in the risk to the public from the accidental offsite release of radioactive material.

According to the Regulatory Analysis for the maintenance rule, a point estimate of the potential risk reduction to the public is approximately 52,000 person-rem, with an upper bound of 72,000 person-rem and a lower bound of 7,300 person-rem. The bases of these projections are provided in the discussion in the Regulatory Analysis. However, as suggested by the range between the upper and lower bounds of risk reduction to the public, the estimates possess a certain relatively high degree of uncertainty. One factor contributing to this uncertainty, and which tends to suggest that the values for the results-oriented final rule are conservative, is that the core damage reduction frequency [CDFJ and forced outage downtime reductions associated with the results-oriented rule are assumed to be the same as the process-PART 50

STATEMENTS OF CONSIDERATION
oriented rule. However, it is believed that the results-oriented rule, by focusing on equipment performance, would be more likely to achieve additional reductions in CDF and forced outage downtime.

4. Potential impact on radiological exposure of facility employees.

The goal-setting, monitoring, and availability evaluation requirements of the maintenance rule are not likely to result in any significant change, either positive or negative, in occupational exposures. Implementation of corrective actions, as required by § 50.65(a](l) of the maintenance rule can affect collective occupational exposures both positively and negatively. Increases in maintenance activity due to expanded preventive maintenance or more aggressive corrective maintenance (to reduce backlogs, for example] will tend to increase exposure, while productivity increases and reductions in the amount of rework will tend to reduce exposures. The net effect of these positive and negative trends is believed to be beneficial but small compared to the other costs and benefits of improved maintenance. Because of the uncertainty in this projection and the relatively small magnitude of the reduced exposures, the cost-benefit analysis of the Regulatory Analysis does not account for any changes in occupational exposures.

5. Installation and continuing costs associated with backfit, including the cost of facility downtime or the cost of construction delay. The Regulatory Analysis for the maintenance rule discusses the costs to the industry and the NRC associated with the maintenance rule. The maintenance rule does not require any change in the design or construction of any nuclear power plant. Nor does the rule apply to activities associated with the planning, design, and installation of plant modifications.

Therefore, there will be no installation, downtime, or construction costs associated with the rule. Rather, the maintenance rule will require licensees to establish goals for the performance or condition of certain structures, systems and components, monitor the performance or condition of those structures, systems and components, and implement corrective action if the licensee-established goals are not met. It also requires an annual evaluation of monitoring, establishment and corrective action activities to take into account wide operating experience and to make adjustments where necessary to balance failure reduction against structure, system, and component unavailability. For 110 operating reactors, the estimated net cost associated with implementation of this rule is $44 million. This estimate breaks down as follows: Industry cost element Implementation and operating ................... . Power replacement due to increased Millions of 1990 dollars 1050 availability................................................... (998) Onsite cleanup and power replacement.. .. f-------'(9--'-) Total industry cost............................. 44 The above cost figures are point estimates with a relatively large degree of uncertainty. The cost estimates in parentheses represent cost savings. 6. The potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing regulatory requirements. As discussed above, the maintenance rule does not require any design modifications. Therefore, safety impacts attributable to changes in plant design are not assumed to result from the maintenance rule. With regard to changes in operational complexity, maintenance is often considered a part of operations. The maintenance rule requires licensees to establish goals for the performance or condition of certain structures, systems and components, monitor the performance or condition of those structures, systems and components, and implement corrective action if the licensee-established goals are not met. It also requires an annual evaluation of monitoring, establishment and corrective action activities. In addition, in performing monitoring and maintenance activities, the overall effect of equipment service on the performance of safety functions must be assessed. These maintenance activities should provide a significant enhancement in safety by contributing to reduced operational complexity as a result of fewer maintenance reworks, fewer unplanned transients, and higher reliability of safety-significant SSCs, thus reducing the need for operator actions in response to events. Thus, operational complexity is not likely to be adversely affected. There are a number of existing Commission requirements directly or indirectly relevant to maintenance, including § § 50.34(a)(3)(i]; 50.34(a)(7); 50.34(b)(6) (i], (ii], (iii] and (iv]; 50.34(b)(9); 50.34(f](l) (i), (ii), and (iii); 50.34(g]; 50.34a(c]; 50.36(a]; 50.36(c)(2), (3), (5) and (7); 50.36a(a)(l]; 50.49(b]; 50-SC-96 50.55a(g); part 50, appendix A, criteria 1, 13, 18,21, 32, 36, 37,40,43,45,46, 52, 53; part 50, appendix B. Licensees must continue to comply with these requirements. However, 10 CFR 50.65 should provide added assurance that these requirements will be complied with. No duplication of requirements is intended.

7. The estimated resource burden on the NRC associated with the backfit and the availability of such resources.

The estimated resource burden to the NRC associated with the maintenance rule can be divided into two elements: (a) Development of a regulatory guide on maintenance effectiveness monitoring ($800,000); and (b] inspection and enforcement to ensure compliance with the rule (assumed to be negligible over and above existing inspection efforts.) With regard to enforcement, the maintenance rule does not require licensees to submit their maintenance program to the NRC for review and approval, and no agency resources have been included in the cost estimates for this activity. NRC does not expect to allocate any additional resources for inspections as a result of this rule. 8. The potential impact of difference in facility type, design, or age on the relevancy and practicality of the backfit. The maintenance rule establishes generic requirements that are applicable to all types of facilities and designs regardless of their age. These requirements (and therefore the cost of complying with these requirements] are essentially the same regardless of the type or design of the facility.

9. Whether the backfit is interim or final and, if interim, the justification for imposing the backfit on an interim basis. The maintenance rule is a final requirement.

Licensees will have up to five years following publication of the final rule in the Federal Register to be in compliance with the requirements of the rule. List of Subjects in 10 CFR Part 50 Administrative practice and procedure, Antitrust, Classified information, Fire prevention, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reporting and recordkeeping requirements. In consideration of the foregoing, the Nuclear Regulatory Commission amends part 50 of title 10 of the Code of Federal Regulations as set forth. PART 50

STATEMENTS OF CONSIDERATION 56 FR 36081 Published 7 /31 /91 Effective 10/29/91 Criteria and Procedures for the Reporting of Defects and Conditions of Construction Permits See Part 21 Statements of Consideration 56 FR 40178 Published 8/13/91 Effective 9/12/91 10 CFR Part 50 RIN 3150 -AD32 Emergency Response Data System AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations to require licensees of all operating nuclear power facilities except Big Rock Point to participate in the Emergency Response Data System (EROS) program. This action requires licensees to submit to the NRC timely and accurate data on a limited set of parameters whose values indicate the condition of the plant during a declaration of an alert or higher emergency classification.

This action will ensure that all licensees establish a definite schedule for implementation of the EROS program. EFFECTIVE DATE: September 12, 1991. ADDRESSES: Copies of all NRC documents are available for public inspection and copying for a fee at the NRC Public Document Room at 2120 L Street NW., Lower Level of the Gelman Building, Washington, DC. Copies of NUREG documents may be purchased from the Superintendent of Documents, U.S. Government Printing Office by calling (202) 275-2060 or by writing to the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082. Copies are also available from the National Technical Information Service, 5285 Port Royal Road. Springfield, VA 22161. FOR FURTHER INFORMATION CONTACT: M.L. Au, P.E., Office of Nuclear Regulatory Research. Nuclear Regulatory Commission, Washington, DC 20555, telephone: (301) 492-3749. SUPPLEMENTARY INFORMATION: Background On October 9, 1990 (55 FR 41095) the Commission published a proposed rule in the Federal Register that would require licensees to participate in the Emergency Response Data System (ERDSJ Program and to set a schedule for its implementation. EROS is a direct

electronic data link between computer data systems used by licensees of operating reactors and the NRC Operations Center (NRCOCJ during the declaration of an alert or higher emergency classification.

The EROS supplements the voice transmission of information over the currently installed Emergency Notification System (ENS), and is activated by a licensee when an alert or higher emergency occurs at a licensed nuclear power facility. This rule applies to all licensed nuclear power reactor facilities, except Big Rock Point and those that are permanently or indefinitely shut down. However, units shut down for maintenance, or authorized for fuel loading only, or low power opera lions, are required to report under EROS. Big Rock Point is exempt because configuration of the facility does not make available as transmittable data a

sufficient number of parameters for effective participation in the EROS program. The objective of the final rule is to ensure timely and effective implementation of ERDS to provide NRC increased assurance that a reliable and effective communication system that will allow the NRC to monitor critical parameters during an emergency is in place at operating power reactors.

Many of the elements of the rule are currently implemented under the EROS voluntary program in which over half of the licensed units have volunteered to participate. The EROS program is not expected to require any advancements in the state of the art, and the configuration of most power reactors is such that the relevant parameter values are available as transmittable data. Therefore, there should be no cause for delay in timely implementation of this rule. Public Comments Interested parties were invited to submit comments on the proposed rule. There were 113 comments made by 31 commenters on the proposed rule: Two from interested individuals. one from a citizens' group, one from a former Senior,. Reactor Operator and Emergency Director at a utility, one from the Nuclear Management and Resources Council (NUMARC), one from the Nuclear Utility Backfilling and Reform Group (NUBARG), 20 from power reactor licensees, one from a non-power reactor licensee, and four from State authorities. Many of the letters contained comments that were similar in nature. These comments were grouped and addressed as one issue. The NRC identified 21 separate issues that cover the significant points raised by commenters. Public comments received on the proposed rule were docketed and may be examined at the Commission's 50-SC-97 Public Document Room located at 2120 L Street NW. (Lower Level), Washington, DC. Upon consideration of the comments received, the Nuclear Regulatory Commission has adopted the proposed regulations, with certain modifications as set forth below. Analysis of Public Comments 1. Comment. The EROS data would be subject to distortion by terrorists or computer hackers which could cause the NRC to respond improperly in their. recommendations to the licensee, Federal agencies, and State and local governments. If the EROS were hardened, or essential data elements were verified by voice communication, this potential problem would be eliminated. Response. It is highly unlikely that a computer hacker would be able to locate EROS transmissions in the NRC's

communications network because of the limited access to this system. Also, the communication protocol incorporated for EROS transmission would make the data unintelligible without knowledge of the specific site link configuration.

Error detection/ correction has been incorporated into the transmission protocol which would, in all probability, detect any alteration in the data. And finally, as stated in NUREG-1394, "Emergency Response Data System (EROS) Implementation," and in this final rule, the NRC will continue the requirement for the licensee to maintain voice communication with the NRC during emergencies. Any data indicating rapid unrealistic changes or unexpected conditions would be immediately suspect and subject to verbal corroboration. Therefore, the NRC does not believe the probability for intentional data distortion is sufficiently large to justify resources for further countermeasures.

2. Comment. There is inadequate justification that implementing the EROS would substantially increase the overall protection of the public health and safety. This contention was made by nine commenters, in addition to the seven commenters who endorsed the consolidated comments from NUMARC and NUBARG without further elaboration.

The commenters stated that if there was a substantial increase this should be quantitatively demonstrable. They also stated that the utility is solely responsible for the protection of the public health. They argued that because this rule does not improve the manner in which the emergency director makes decisions, the claim of "unquantifiable but significant increase" in the protection of the public is invalid. One commenter stated the ERDS is an improvement to a system that has been deemed "adequate," and therefore is not necessary. PART 50

STATEMENTS OF CONSIDERATION Response.

Many have argued, as the commenters have, that the NRC Dackfit Rule (10 CFR 50.109) require9 quantitative evid,mce that r:ew NRC requirements will result in a substantial increase in the overall protection of the public health and safety or the common defense and security. The NRC does not agree with this interpretation and believes that the safety enhancement justification for a backfit can also be met on the basis of qualitative considerations. In such cases, the NRC believes that the evidence that a substantial increase in overall protection would occur must be clearly defensible and meaningful: The NRC has used this test in its assessment of the ERDS requirements. The Commission has previously determined that there exists both a regulatory and statutory basis for having emergency planning as a cdtical element in the protection of public health and safety. In il.s July 17, 1979, Advance Notice of Rulemaking, the following statement is made: "The Nuclear Regulatory Commission, in discharging its statutory responsibilities to protect the public health and safety, has given its primary attention to aspects of the reactor site and the facility design. In this regard, emergency planning, including evacuation planning, has been conceived of as a measure that adds to the level of public protection

* *" The NRC, in its mandated role to protect public health and safety, has a respcnsibility in the event of a reactor accident to monitor the actiona of the licensee, who has the primary continuing responsibility for limiting the consequences of the accident.

The NRG also has an important role in assuring the flow of accurate information to affected offsite officials and the public regarding the status of the emergency and, as requested or needed, giving advisory support or assistance in diagnosing the situation, isolating critical problems and determining what remedial actions are appropriate. The NRC must be capable of providing to State and local authorities, and to other Federal agencies, an independent assessment of protective actions recommended by the licensee. Given the regulatory and statutory basis, and given the importance of emergency planning and response in the defense-in-depth context, when an accident has occurred, the NRC believes that a significant increase in its ability to perform its role would constitute a substantial increase to the overe.ll protection of the public health and safety. Since thP. principal effect of EROS will be a marked improvement in the availability, iimeliness, and reliability of key information about what is taking place at the reactor during an accident, particularly during the critical early hours before the NR,C Site Team arrives, it is the judgment of the NRC that the implementation of ERDS will provide a gignificant improverner.t in the NRC's .ability to accurately and promptly assess the situation at the site. In emergency drills conducted by the NRC and power reactor licen.;ees, the information on the status of the reactor is typically 15-30 minutes old by the time it is received at the NRC Operations Center when transmitted via the existing Emergency Notification System [ENS). Moreover, inaccuracies and errors have been found in that information which in some cases has led to confusion and misunderstanding of the situation. In drills which have employed a prototype of the ERDS, there has been profound improvement in the availability, timeliness, and reliability of the information transmitted. The actual experience of the NRC using the existing ENS alone contrasted with drills where beth ENS and a prototype ERDS were used is the basis for its conclusion, that ERDS will provide significant improvements in the NRC's ability to underntand what is taking place during an emergency, and thereby more effectively perform its role of monitoring and advising th.e licensee. More impcrtantly, the improvement in assessment performance significantly improved the NRC's ability to provide appropriate recommendations and advice to the State and local officials who are required to make the decisions regarding the offsite protective actions which are necessary to protect the public. Because the decision made by the State and local authorities with regard to offsite protective actions could significantly affect the public health con3equences of a reactor accident, it is the judgment of the NRC that a significant improvement in the NRC's ability lo provide the right recommendation at the right time provides a substantial improvement in the overall protection to the public. An effective emergency response capability in the event of a reacior accident is an essential element of the depth approach to protection of the public health and safety. The NRC's role during an emergency is part of that capability. Because the ERDS wiil provide a significant improvement in the NRC's ability to perform that role in an emergency, the proposed ERDS improvements are therefore justified, 50-SC-98 and the ccsts of implementing those improvements are modest. 3. Comment. One commenter believed that the limited group ofreactor parameters monitored through EROS would be inadequate to provide a sound basis for NRC recommendations and therefore requested modifications to ERDS. One commenter urged the NRC to consider a continuous monitoring system, e.g., the Nuclear Data Link considered by the Commission following the Three Mile Island accident. Other commenters stated that the EROS design uses cumbersome hardware and software, that NRC's communication hardware should be able to accept dnta from a multiple unit plant through one r.:odem, and allow state-of-the art hardware. Response. Although the ERDS data does not portray every detail of a nuclear power reactor in an emergency situation, the Commission believes it does provide the data required by the NRC to perform its role during an emergency. The ERDS para!Ileter list was selected based on the information the NRC Technical Teams need to perform their emergency response functions. Moreover. the set of ERDS data will not be the only input to the NRC. The Emergency Notification System (ENS], a voice communication* system, will still be available to transmit data and any other relevant information that is not available through ERDS. In combination, the NRC will receive the necessary information to develop timely and appropriate evaluations of the event and to develop the necessary support actions to ensure protection to public health and safety. The ERDS is designed to transfer needed reactor data from a nuclear power plant only during emergencies. It is not a system to constantly monitor any licensee. The concept of constant monitoring, such as the Nuclear Data Link, was considered after the Three Mile Island accident in 1979. But after much evaluation and deliberation, Congress did not approve the concept for funding. The current protocol is already in use at several reactors under the volunteer program and is in the process of being implemented at other facilities. The NRC is not requiring additional redesign and retest costs on voluntary licensees who already have an acceptable system inplace or have submitted an acceptable implementation plan. The EROS was designed to use commercially available (off-the-shelf] computer3 which could effectively handle the data requirements, establishing a single link with each unit. PART 50

STATEMENTS OF CONSIDERATION To group several units into a single link would result in a data base size incompatible with the ERDS configuration.

The ERDS design has been frozen in order to maintain configuration control and standardization in implementing the ERDS volunteer program. 4. Comment. Submittal of an ERDS implementation plan should not be required of licensees that have implemented ERDS under the voluntary program. Similarly, licem1ees that have submitted the information required by the voluntary program altJng with a proposed implementation schedule should also be exempt foJm the schedule and system requirements contained in 10 CFR part 50, paragrap 1 1ls Vl.1, Vl.2 and Vl.4 of appendix E of the proposed rule. Response. The NRC aErees that it is unnecessary for licensees that have implemented the ERDS in an acceptable manner to submit an implementation plan. The final rule (appnndix E to part 50, section VI, paragraphs 4.c and d) has been modified so that licensees who have submitted all infonnation consistent with the timetable set in paragraph 4.b of appendix E to part 50, section VI, are not requked to submit an implementation plan. 5. Comment. (a) Ninet1~en of the commenters, including three that endorsed the NUMARC comments, were concerned that implementing the ERDS would increase the opentors' labor burden because the NRC, as well as State or local government agencies receiving the ERDS data, would not be staffed by personnel with sufficient system specific knowledge to understand the data. This would result in extensive inquiries to. the licensees to explain the data, thereby distracting the operating staff from theiir primary functions of accident re,~ponse and emergency managemen:t. [b) Some of these commenters urged the NRC to limit the data provided to States and local government and direct them regarding the use ,of the ERDS information to preclude the improper use or release of the da'ta. [c) Other commenteru stated that with the availability of ERDB parametric reactor data, the NRC would modify its oversight role into one 1:>f more active participation in event management, a function, the commenters claimed, is solely the responsibilitir of the licensee. Response. (a) The NHC does not believe that ERDS will impose an additional burden on Hcensees during an emergency. Rather, the reduction in the potential for miscommunication and misunderstanding afforded by ERDS should enhance the lic1msee's efficient use of its resources in dealing with an emergency. The NRC acknowledges that ERDS will impose small additional burdens on licensee resources during periods of non-emergency and typically involving non-operator personnel. These impacts are discussed in the regulatory analysis that accompanies this rule and include incremental licensee hours for development of the ERDS program and necessary software, periodic testing, and the preparation of configuration control reports. These incremental costs are judged commensurate with the enhanced protection of the public attributable to ERDS. Concern over the capability of NRC staff to understand the ERDS data are unfounded. The NRC Operations Center staff are experienced professionals with extensive knowledge of reactors, sufficient to allow them to use the data provided by the ERDS to follow the course of the emergency, chart and analyze trends, and support appropriate recommendations relating to the health and safety of the public. Further, the NRC is aware that while not all States have the technical knowledge required to interpret raw ERDS data, some have developed significant expertise in responding to emergencies at nuclear power plants. The NRC believes that since the States are responsible for protective actions to ensure the health and safety of their citizens, they should have available sufficient data upon which to base decisions. (b) The ERDS link will be established with a State government through a Memorandum of Understanding [MOU) with the NRC. The proper use, control, and dissemination of the ERDS data is one of the subjects addressed by the MOU. Under the MOU, the NRC will provide a liaison to the State at the NRCOC for EROS data interpretation if such help is requested. [c) The implementation of ERDS will not alter the respective responsibilities of the utilities and the NRC with respect to emergency management. The utility will retain primary responsibility for emergency management activities at the site locations. The NRC will continue to monitor, inform, and upon request, advise licensees and other local, State and Federal authorities who are responsible for the safety of their citizens, as well as to provide timely advice to the licensees as needed. 6. Comment. States may require the licensee to pay for equipment required to receive and process the EROS data. Furthermore, providing ERDS data to the States and local governments would increase NRC costs beyond that estimated in the Backfit Analysis. 50-SC-99 Response. The NRC has no control or authority over the State governments regarding their funding of ERDS receiving equipment. Each individual State government should determine its equipment and data requirements. However, through a Memorandum of Understanding [MOU) between the State and the NRC regarding the ERDS link, the ERDS data can be made available to a State. One of the functions of the NRC is to provide appropriate support to the States during a nuclear power plant emergency. This responsibility exists independent of the ERDS, and in the stafrs view, the EROS interface between the NRC and the States should not result in additional costs to the NRC. 7. Comment. Implementing the EROS seems to imply some general concern that the NRC neither trusts its abilities nor those of the licensees to respond correctly to emergencies using current practices. Response. ERDS is an enhancement of existing procedures that provides a superior method of assembling and transmitting to the NRC near real time data from a licensee during an alert or higher emergency classification. . Accurate and timely data assists the NRC in conducting informed analyses of the plant condition, and facilitates NRC consultation with State or local governments regarding action to ensure protection of public health and safety. 8. Comment. Will the time in the header of the ERDS data packet be some standard time such as GMT, EST, etc.? Response. The time from the licensee's plant computer will be used with ERDS data. Included in each licensee's EROS implementation plan will be the time standards used in their computers. This practice will ensure that the particular licensee and all monitors of ERDS data relating to a particular emergency or test are using the same time. There is no requirement for all licensees to adhere to a common standard time. 9. Comment. Non-power reactors should be explicitly exempt from the ERDS requirements. Response. Since 10 CFR 50.72 *of the regulations applies only to nuclear power reactors, it is not necessary to explicitly exempt non-power reactors in the rule. 10. Comment. Licensees are requested by Generic Letter 8~9 to ~ransmit a significant number of data sheets to the NRC during emergencies. With the implementation of ERDS, this should be relieved to allow better use of licensee resources to support ERDS. PART 50

STATEMENTS OF CONSIDERATION Response.

The information cited is an Information Notice (IN), and as such, it requires no action on the part of the licensee. The form contained in IN 89-89 is a copy of the work sheet used by NRC Headquarters Operations Center officers in recording routine Event Reports over the ENS. IN 89-89 was provided as information to licensees to aid in structuring their normal event report. 11. Comment. The NRC should provide the software required for ERDS communications to the utilities. Response. The NRC will develop software which may be used in a utility provided personal computer (PC) interface for ERDS. The NRC will provide software and source code for a program that will perform ERDS communications protocol and data transmission functions.

12. Comment. There were several concerns regarding the configuration control of ERDS hardware and software.

Five commenters stated the requirement to notify the NRC within 30 days following changes in individual parameters is overly prescriptive, and they proposed extending the maximum allowable notification period to 90 days, annually, or during Final Safety Analysis Report (FSAR) updates. Two commenters believed the time estimated to perform the configuration control functions was low by a factor of two or three, and therefore the ERDS would be more costly to the utilities than estimated. One commenter stated there should be specific guidance provided for the configuration control requirements of the utility/EROS interface; and two were concerned that if the NRC changes its format the licensees are automatically required to change their transmission of data. They recommended that the data should be limited to an initial format with no later changes. Response. In establishing the current reporting requirement for changes in the ERDS Data Point Library, the staff balanced the time needed by the licensees for its design change-control and review processes against the staffs need to know based on safety considerations. The staff views the 30 days as reasonable for the licensees to prepare such a report, and given that such changes can influence the NRC's interpretation of ERDS data does not view any further delay as warranted. For some licensees, plant to plant variation could result in a greater labor burden associated with configuration control tasks than the 5-person days per reactor year used in the regulatory analysis. However, that value represents an average that, considering the entire nuclear power industry, appears substantially correct. There is an economy of scale for those utilities that can combine submissions from multiple reactor units that retluce the industry average. The basic guidance information for configuration control of the EROS is contained in NUREG--1394. Based on the experience of the utilities that have implemented ERDS voluntarily, the configuration control requirements appear to be appropriate. The proposed rule would require the licensee to change its data transmission if the NRC changes its format, and the staff agrees that this is an unreasonable requirement on the licensees. Therefore the final rule has been revised to require all data transmission to conform to the initial format. As the ERDS matures, or as technical advances increase capabilities, there may be some modifications. However, any such changes will be coordinated with the licensees.

13. Comment. The ERDS rulemaking should clearly state that the ERDS is available to the States; and that all future State and local government requests for on-line data should be made through the NRC. Furthermore, the licensees should have access to the same screens as those available to the NRC. Response.

It is not within the authority of the Commission to specify to the States what data they may or may not receive. However, the NRC does recommend that States desiring an emergency data link to nuclear power plants within their jurisdiction use an ERDS connection from the NRC Operations Center. A Memorandum of Understanding with the NRC will . p'rovide the State with ERDS data. A provision allowing States to receive ERDS data should not be part of the rule since there is no NRC requirement imposed upon licensees to establish a data link with a State. The concept of providing each licensee with the same work stations as the NRC was considered. However, it was not deemed cost beneficial to expend in excess of $900,000 for the sole purpose of sending back to the licensees that data which they originally sent to the NRC. Any licensee desiring to do so may establish their own work station based on NRC design. 14. Comment. The requirement for the reactor parametric data to be transmitted to the NRC Operations Center at time intervals of not less-than 15 seconds or more than 60 seconds is too prescriptive and may eliminate the use of some existing computer systems currently supporting the licensee's 50-SC-100 Technical Support Center (TSC)/ Emergency Operating Facility (EOF), etc. One commenter suggested that data update frequency should be plant specific. Others argued that the wording in the proposed rule puts the licensee in jeopardy of non-compliance in the event of system or telecommunications line failure, and that considering the conditions, the proper descriptor for the data is "near real time" instead of "real time." Response. Originally the desired update frequency for ERDS data was 15 seconds, but to minimize the use and impact on the central processing unit (CPU), the minimum frequency was reduced by a factor of four, i.e., to at least every 60 seconds. Based upon the experience of those manning the NRCOC, the staff believes that less frequent data collection would diminish the NRC monitors' ability to adequately follow the course of the emergency. Furthermore, allowing update frequencies to range between 15 seconds and 60 seconds should provide sufficient latitude to allow most licensees to use th'eir existing computer systems. Exceptions to this requirement will be considered on a case by case basis by the NRC. Consistent with the NRC's . enforcement policy, licensees are not cited for matters beyond their control, such as equipment failures that are not avoidable by reasonable licensee quality assurance measures or management controls. Nonetheless, in the wording of the final rule, the term "near real time" has been used to describe the ERDS data. 15. Comment. The requirement to activate the EROS at the time the NRC is notified of the declaration of an alert or higher emergency classification should be relaxed because.it places a heavy labor burden on the plant operators at this critical time. Several commenters suggested a delay of one hour in order to allow actuation from the. Technical Support Center, thus rei:noving the burden from control room personnel. Four commenters stated the ERDS should not be operated from an on-site computer, and two suggested the rule should allow the ERDS to be activated by computer operations personnel or a software switch. One commenter stated the licensee should be the only entity to activate or deactivate the ERDS for a given plant. _ Response. There is no requirement for the ERDS to be activated from the control room or by control room personnel. The use of computer operations personnel or a software switch is acceptable to activate the PART 50

STATEMENTS OF CONSIDERATION*

ERDS. The only requirem,~nt is to initiate ERDS data transn'iission as soon as possible but not later than one hour after declaring an emergency class of alert, site area emergency, or general emergency. This change i:~ reflected in the final rule. The specific: methods selected to achieve this ri\quirement should be fully described' 1 in each licensee's ERDS impleme*ntation plan. The notification requirement is valid in order for NRC to fulfill ilf1 mandated role to monitor the licensee dining an emergency. A delay of 011le hour or more could deprive the NRC of vital information necessary to 1 perform its advisory and monitoring role. The licensee is currently required in 10 CFR 50.72 to have a shift communicator maintain continuous contact with the NRC Operations Center. :This request is not being changed. and this person could be responsible for !initiating the ERDS link. Similarly, the requirern'ent to use an on-site computer does no;t mean this equipment must be locat,~d in the control room. Any on-site~ location, such as the Technical Support Center or a computer facility, which is capable of meeting the requirement for notification is an acceptable location. However, site computers. e.g*., at sc,me central location used to service more than one plant site could be prone to additional commercial link. vulnerability. This could potentially decrea1;e the ERDS availability and reliabi!ilty beyond acceptable limits. The ERDS link will be activated or deactivated by the licen11ee to transmit the ERDS data to the NRC Operations Center via the NRC-provided telephone lines. In the event that NRC perceives the need to disconnect a plant from the NRC Operations Center to allow another plant onto the system. for example, terminating th1i transmission of exercise data to allow a unit with a real emergency to acces:~ the system, this capability must be s;vailable to the NRC. ' 16. Comment. The 18 month EROS implementation scheduhi does not provide adequate flexibiility for all utilities to install the syi1tem. Adhering to that schedule will cause serious operational and cost impacts to some utilities because the system requires extensive hardware modifications. Response. The volunt:ary program demonstrated that an iniplementation period of 18 months is g,enerally adequate. However. the NRC realizes there are plant to plant variations which, in certain cases, may require more extensive and tim,e conrnming modifications. Utilities 1;hat experience exceptional difficulties iin meeting the 18 month implementation schedule should request an extension from the !\'RC. Extension requests will be reviewed on a case-by-case basis. Extensions wiil not be granted in th!! absence of reasonable and good faith efforts to meet the schedule. , 17. Comment. The requirement in the proposed rule contained in appendix E to part 50, section VI .2, should be clarified to indicate that the licensee will provide data from each unit via an output port on.the appropriate data system and necessary software to assemble the data to be transmitted. Response. The staff agrees with this clarification. This section of the final rule will be modified appropriately.

18. Comment. Quarterly testing of the EROS is too frequent.

Testing on a annual or peifodic but unspecified schedule should be sufficient. One commenter noted that the rule does not address reporting requirements for system failures during testing. Also for consistency between the discussion section and the rule, the following statement regarding the use of ERDS during emergency training exercises should be added to 10 CFR 50.72(a)(4) of the rule. Although there is no requirement, the ERDS may also be activated lr.f the licensee during emergency drills or exercises if the licensee's computer system has the capability to transmit the data. Response. Quarterly testing during the initial year or 18 months of the ERDS program is necessary for both the licensees and the NRC monitors to gain experience and confidence with the system, as well as prove the availability and reliability of the system. An established schedule allows both the NRC and licensees to plan and allocate time and resources for testing rather than trying to accommodate testing on an unregimented basis. After a period of approximately one year of demonstrated system performance, i.e .. proper functioning during quarterly testing, the test frequency may be relaxed to annually. There are rio explicit repm:ting requirements for failures during testing because the quarterly testing will be conducted with NRC. If there are failures during these tests, the NRC, because of its participation in the tests, will be aware of them. It is unlikely there will be any system testing of which the NRC is unaware, e.g., with State or local governments, since the State links will most probably be through the NRC Operations Center. The recommended additional statement regarding use of ERDS during emergency training exercises has been included in the final rule. 50-SC-101

19. Comment. Three comrr:enters stated that this rule should impose no new isolation requirements, and suggested that references should be deleted to H potential requirement for additional isolation requirements.

Response. The reference to the potential need for isolation devices is not a new requirement. It is intended merely to serve to reinforce requirements as a design control mechanism in 10 CFR 50.55a and adds emphasis for adequate protection against spurious electrical sig..'1als. More recently constructed nuclear power reactors have adequate isolation of their computer interfaces, but in some older reactors it is conc*eivable the computer assembling the ERDS data may not be fully buffered, and as such. could require appropriate isolation devices. The statement alerts the licensees to the potential need for additional isolation devices. 20. Comment. There should be more flexibility in acceptable quality indicators [tags) for the EROS data, thus allowing greater use of existing plant methodologies. Requiring the utilities to use the quality tags prescribed by the NRC would force major software changes and added costs for some licensees. Response. Using the data quality indicators prescribed by the NRC should necessitate, at the most, only very minor licensee software changes. A simple translation matrix that converts the quality tags used by the licensee to the form to be used by the NRC Operations Center is sufficient. This can be applied to the ERDS data prior to transmission. There is no requirement for the utilities to change the quality tags used at their facility. However, if each utility transmits ERDS data to the NRC Operations Center using their own quality tags, variation from licensee to licensee could cause confusion to the NRC monitors, thereby necessitating additional telephonic consultation with the licensee.

21. Comment: Four commenters stated that when ERDS is implemented the requirement for full time manning of the Emergency Notification System (ENS) should be relaxed. Without this relaxation the affected utility will not be able to redirect its efforts as claimed. Response:

It is not the intent to replace the ENS with ERDS; rather, ERDS is a supplemental system specialized in automatic coilection and transmission in near real time of a selected set of parametric reactor data required by the NRC in its emergency monitoring role. Although implementing ERDS will diminish the current ENS PART 50

STATEMENTS OF CONSIDERATION burden, not all functions of the ENS will be subsumed Into the EROS. Therefore, telephone contact will still be required via the ENS. Ne\'ertheless, the effort required by the licensee's personnel to gather the data for periodic relay to*the NRC will be reduced, thus permitting their use of personnel in other
emergency functions.

Environmental Impact: Categorical Exclusion The NRC has determined that this final regulation is the type of action described in categorical exclusion 10 CFR 51.22(c)(3)(iii). Therefore, neither an environmental impact statement nor an environmental assessment has been prepared for this final regulation. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Management and Budget 1u1.der approval number 3150-0011. Public reporting burden for this collection of information is estimated to average 115 hours per response the first year and 38 hours per response* thereafter, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments regarding this burden estimate or any other aspects of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB-7714), U.S. Nuclear Regulatory Commission, Washington, DC 20555; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOD-::!019 (3150--0011), Office of Management and Budget, Washington, DC 20503. Regulatory Analysis The NRC has prepared a regulatory analysis for the final rulemaking on this subject. The analysis examined the costs and benefits of the alternatives considered by the NRC. The NRC requested public comments on the preliminary reguiatory analysis. Comments received were considered, but no changes to the regulatory analysis are considered necessary. Therefore, the preliminary regulatory analysis is adopted as the final regulatory analysis without change. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the Commission certifies that this rule does not have a significant economic impact .on a substantial number .of small entities. This.fin~ l'llle affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall with.in the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR part 121. Backlit Analysis As required by 10 CFR 50.109, the Commi11sion has completed a backfit analysis for this rule. The Commission concluded that the rule will provide a substantial increase in the overall protection of the public health and safety by ensuring far more accurate and timely flow of data for the NRC to fulfill its role during an alert or higher emergency. The direct and indirect costs estimated for the implementation of this rule are justified in view of this increased protection. Further, the implementation and maintenance requirements of the rule will have no effect on occupational radiological exposure. The backfit analysis on which this determination is based is as follows: Item 1: Statement of the specific objective that the back/it is designed to achieve. Response: The objective of this rulemaking is the timely and effective implementation of EROS so as to provide increased assurance that a reliable and effective communications systein, that will allow the NRC to monitor available critical parameters during an emergency, is in place. The availability of accurate, near re.al-time data depicting what is taking place at a rel:lctor facility during an alert or higher emergency will improve the NRC's understanding of the event as it is happening, and thereby better enable the NRC to perform its role of (i)

providing State and local authorities recommmendations and advice on offsite action that they may need to take to protect their citizenry; (ii) supporting the licensees efforts to manage the accident by providing technical analysis and logistic support; (iii) keeping other Federal agencies ar.d entities informed of the status of the incident; and (iv) keeping the media informed of the NRC's knowledge of the status of the incident.

Item 2: General description of the activity that would be required of the licensee or applicant in order to complete the back/it. Response: All licensees or applicants would be required to install an NRC-50-SC-102 supplied ,communication link, provide the necessary hardwar.e from the plant .computer to interface with the NRC-supplied communication link. provide support for periodic testing af the EROS, and report any configuration changes to the licensee's EROS-related hardware and software. Initially, the ERDS will be tested quarterly, unless otherwise determined by NRC based on demonstrated system performance. Item 3: Potential change in the risk to the public from the accidental offsite release of radioactive material. Response: The principal effect of EROS will be a marked improvement in the availability, timeliness, and reliability of key information about what Is taking place at the reactor during an accident, particularly during the critical early hours before the NRC*Site Team arrives. Hence, ERDS will provide significant improvements in the NRC's ability to understand whatis taking place during an emergency, and thereby more effectively perform its role of monitoring and advising the licensee. More importantly, the improvement in assessment performance will improve the 1'1'RC's ability to provide appropriate recommendations and advice t.o the State and local officials who are required to make the decisions regarding offsite protective actions which are

necessary to protect the public. Because the decisions made by the State and local authorities with regard to offsite protective actions could so significantly affect L'J.e public health consequences of a reactor accident, it is the judgment of the NRC that a significant improvement in the NRC's ability to provide the right recommendation at the righttime provides a substantial improvement in t.li.e overall protection to the public. Because the EROS will provide that significant improvement in the NRC's ability to provide the right recommendation at the right time, the proposed ERDS requirements are justified.

Item 4: Potential impact on radiological exposure of facility employees. Response: The implementatiCln of the proposed EROS rule would have no effect on routine occupational radiological exposure and would not result in increased radiological exposure of facility employees. Item 5: Installation and continuing costs associated with the backfit, including the cost of facility downtime or the cost of construction delay. Response: The cost impact of the rule was estimated to be approximately $153,000 for one nuclear power reactor PART 50

STATEMENTS OF CONSIDERATION (one unit). This figure, expressed in 1990 dollars, represents the inc1~emental worth of installing and op13rating ERDS for 30 years using a 5 percent discount rate. The overall industry cost of implementing the rule for '.l18 nuclear power reactor units was estimated at approximately

$18 million. No downtime costs were considered in the cost impact estimates because the installation and operation of the ERDS sho,uld have no impact on the operation ol: a nuclear power plant. . Item 6: The potential safety impact of changes in plant or operaaonal complexity, including the relationship to proposed and existing regulatory requirements. Response: The ERDS rule should have little or no impact on the c1perational complexity of the nuclear power reactor units since the required modifications to the hardware and softwaie are minor. The redirection in the labor burden provided by the automatlu collection and transmission of selected reactor data would increase the efficiency and effectiveness of nuclear p,Jwer plant operating personnel during an emergency. This rule is cl1Jsely associated with Generic Letter 89-15 and complements the ENS that exists at every nuclear power reaci:or. Item 7: The estimated n~source burden on the NRC associated wlth the backfit and availability of such msources. Response: The impact on the NRC resulting from the implementation of the EROS rule is anticipated to be a time cost of about $200,000 for the current population of operational/ licensed nuclear reactor units. This figure provides for initial 1reviews of licensees' implementation plan submittals. After implementation. the NRC cost is estimated to be approximately $4.4 millio11 for 118 nuclear power reactor units. This figure represents the costs for pnriodic testing and configuration control expressed as the present worth in 1990 dollars and uses a 5 percent discount rate over 30 years. . Item 8: The potential impact of the differences in facility typ,~. design, or

age on the relevancy and practicality of the backfit. Response:

The rule is independent of the facility's type, design, or age. There are considerable variations in the instrumentation systems of the nuclear power plants, and the estiimated cost impacts were based on an average value for current nuclear power plants to implement the ERDS. There will be no differences. however, in potential impacts between the vari,~us facilities on a yearly basis. The rule does not require that licensees mo11itor more parameters than are presently monitored at each facility.* Item 9: Whether the proposed back/it is. interim or final and, if interim, the justification for imposing the proposed back/it on tin interim basis. Response: Implementation of the ERDS in accordance with the final rule will require that all licensees develop and submit an ERDS implementation plan to the NRC within 75 days of the publication of the final rule in the Federal Register. The implementation plan should provide a schedule which identifies the earliest possible time frame for ERDS implementation by the licensee as well as proposed alternate implementation dates. The NRC will establish an industry-wide ERDS implementation schedule which will take into account such factors as planned computer modifications and scheduled outages. The ERDS must be implemented within 18 months of the publication of the final rule in the Federal Register. Licensees that have submitted the required information under the voluntary implementation program will not be required to resubmit this information. However, they will be required to meet the implementation schedule of 18 months after the effective date of the final rule or before initial escalation to full power, whichever comes later. Licensees with currently operational ERDS interfaces approved under the voluntary ERDS implementation program will not be required to submit another implementation plan and will be considered to have met the requirements under this rule. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminal penalty, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors. Radiation protection, *Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as ameI).ded, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendment to 10 CFR part 50. 56 FR 40664 Published 8/15/91 Effective 9/16/91 Revisions to Procedures to Issue Orders; Deliberate Misconduct by Unlicensed Persons See Part 2 Statements of Consideration 50-SC-103 56 FR 61352 Published 12/3/91 Effective 6/20/91 Standards for Protection Against Radiation; Correction See Part 20 Statements of Consideration 56 FR 64943 Published 12/13/91 Effective 1 /13/92 Nuclear Power Plant License Renewal See Part 54 Statements of Consideration 57 FR 18388 Published 4/30/92 Effective 6/1 /92 Uranium Enrichment Regulations See Part 40 Statements of Consideration 57 FR 30383 Published 7/9/92 Effective 8/10/92 10 CFR Part 50 RIN 3150-AOS9 Decommissioning Funding for Prematurely Shut Down Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations on the timing of the collection of funds for decommissioning for those nuclear power reactors that have shut down before the expected ends of their operating lives. These amendments require that the NRC evaluate decommissioning funding ,;ilans for power reactors that shut down PART 50

STATEMENTS OF CONSIDERATION i prematurely on a case-by-case basis. The NRC's evaluation would take into account the specific safety and financial
situations at each nuclear power plant. EFFECTIVE DATE: August 10, 1992. FOR FURTHER INFORMATION CONTACT: Robert Wood, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, telephone (301) 504-1255.

SUPPLEMENTARY INFORMATION: Background On June 27, 1988 (53 FR 24018), the NRC published a final rule that amended 10 CFR parts 30. 40, 50, 51, 70, and 72. This final rule established . several acceptable methods by which power reactor licensees may provide assurance that they will have sufficient . funds to decommission their plants by

the time the plants are permanently shut doWll (53 FR 24043). In considering this final rule, the Commission
acknowledged that, in certain instances, reactors mtght be permanently shut
down before completing the full term of , their operating lives._However, because the Commission determined that such * 'instances would be infrequent, it did not ' explicitly include remedies for this . I situation in the final rule. In establishing the June 27,l988, final rule, the Coriuni!lsion recognized that power reactor licensees generally have access to significant amounts of financial capital and are closely regulated.

Therefore, the Commission allowed these licensees the option of accumulating decommissioning funds over the projected operating life of the faci}jty rather than requiring that these funds be available or guaranteed prior to opera~on, or at some time before the end of the.projected operating life of the facility. The Commission recognized the risk that, if some reactors did not operate for their entire operating lives, those licensees might have insufficient decommissioning funds at the time of permanent shutdown. After the NRC published the June 27, 1988, final rule, .four power reactor facilities shu_t down prematurely: The Fort St. Vrain Nuclear Generating Station, the Yankee Rowe Nuclear Power Station, the Rancho Seco Nuclear Generating Station, and the Shoreham Nuclear Power Station. The NRC staff sought the Commission's guidance on the appropriate period for collecting funds to compensate for any shortfall of decommlssl.oning funds for plants such as these that shut down prematurely: The Commission elected to determine the appropriate collection period for any decommissioning funding shortfall for prematurely shut down power reactors on a case-by-case basis. As part of its decision, the Commission directed the NRC staff to prepare a rulemaking that would codify this case-by-catle approach. A proposed rule implementing this approach was published in the Federal Register on August 21, 1991 (56 FR41493J. Analysis of and Response to Comments The NRC received 17 comments in response to the proposed rule. Eleven comments were from NRG-licensed electric utilities; two from utility trade groups; one from a utility counsel; two . from bond rating/investment advisory companies; and one from a public interest group . Except for the comment from the public interest group, all comments supported that part of the proposed rule that would allow the period of funds accumulation for a prematurely shut down reactor to be determined on a case-by-case basis. However, most utilities and their representatives opposed the guidance in the preamble to the proposed rule that would use a bond rating of "A" as a criterion for determining the future solvency of and

thus the extent of the funding period for a licensee with a prematurely shut down power reactor. 1. Comment: The use of bond ratings. The conunenters offered a variety of reasons why they considered bond ratings, particularly at the "A" level, to be inappropriate for judging a licensee's ability to pay for deconunissioning for a prematurely shut down reactor. These reasons included the following:

(1) Bond ratings are too restrictive and do not allow for variations in licensees . situations as contemplated by the by-case approach. (2) Bond rating may not be.an accurate indicator of a licensee's future ability to pay for decommissioning. (3) Not all licensees issue debt that is rated. In the i;:ase of poser plants with several owners, owners will likely have different ratings. (4) Bond ratings would likely decline by virtue of a premature reactor shutdown, thus precipitating further financial problems and further downratings.

(5) Differences in ratings by different services or for different classes of debt issues were not addressed.

(6) Reliance on bond ratings may . result in unsound business decisions to avoid accelerated fund accumulation or may discourage use of the SAFSTOR decommissioning opticµ. (7) A "BBB. rating, or equivalent, is still considered investment grade and is used as a criterion in Regulatory Guide 50-SC-104 1.159 1 and Appendix A to 10 CFR part 30. (8) A one-year trigger period is too short and may be disruptive. (9) Possible adverse tax consequences may accrue if accelerated payments are required. Response.: The NRC continues to believe that bond ratings can serve as one of several criteria to estimate the ability of a licensee to pay future decommissioning costs. The NRC did not intend that this rule set a mandatory requirement that a minimum "A" rating must be met before the NRC would approve funding into a shut down reactor's safe storage period. Rather, one reason the "A" rating criterion was proposed was to serve as a screening test of whether additional financial data was required to determine whether the licensee should be allowed to fund decommissioning into s storage period. If a licensee met *this criterion, the licensee would not have to prepare and submit additional documentation of its financial situation to be allowed to fund decommissioning into a storage period. A benefit of the proposed screening test was a potential saving of licensee and NRC resources to develop and review the additional financial data. With respect to the level of rating (i.e .. "A" vs "BBB" or equivalent), the preamble to the proposed rule presented a rating of "A" as a threshold below which a licensee would be required, if other criteria were not met, to accelerate payment of any decommissioning funding shortfall. The staff chose an "A" rating because a downrating from "A" to "BBB" would* allow a licensee to secure funds-or meet other criteria while Its rating was still Investment grade. To that extent, an "A" rating is not inconsistent with the use of "BBB" ratings in Regulatory Guide 1.159 and appendix* A to 10 CFR part 30. In Regulatory Guide 1.159, a "BBB" bond rating was used as a minimum suggested standard for a mixed portfolio of investments in a decommissioning trust. Because of investment diversification, a "BBB" investment-standard represents a 1 Regulatory Guide 1.159 is available for inspection and copying for a fee at the Commission's Public Document Room 2120 L Street. N.W .. (Lower Level). Washington. D.C. Copies or issued guides may be purchased from the Government Printing Office at the current GPO price. Information on current GPO prices mny be obtained by contacting the Superintendent or Documents, U.S. Government Printing Office, P.O. Box 37082. Waahington. D.C. 20013-2171. Issued guidea may also be purchaaed from the National Technical Information Service on a standing order basis. Details on this service may be obtained by writing Nl1S. 5825 Port Royal Road. Springfield. VA 22161. PART 50

STATEMENTS OF CONSIDERATION relatively low level of financial risk. Similarly, appendix A of part 30 used a "BBB" rating as a minimum for a parent company of a licensee to uuarantee decommissioning costs. B11cause a parent company is a separate legal entity from its subsidiary, the NRC would potentially have access to two sources of funds (the licen'see and its parent) thus red1,1cing the Iisk of decommissioning funding shortfalls.

For this reason, the NRC disaurees that an "A" bond rating standard .ls too stringent as a screening test. For these reasons, the NIRC will continue to use the "A" bond rating as a screening test for determining the decommissioning funding period for prematurely shut down po,wer reactors. If a power reactor licenset! cannot pass the initial screening test, cir if it has passed the screening test lout subsequently loses its "A. bond rating, this licensee could still be ,allowed to fund into the storage peric1d by meeting other criteria as described: below. These criteria include: . ' (1) A licensee's financial history including its past funding ,of reactor safety expenditures: (2) The local rate regulatory . environment and other relevant State laws including public utility commission (PUC) commitments; (3) The number of other genera ting plants, both nuclear and non-nucTear, in its system. This is anothe1: way of measuring the relative impact of decommissioning costs on a particular licensee's finances: and ' (4) Other factors that a !licensee can demonstrate as being relevani. The NRC wishes to cla~ify that it assumes that most license.es with "BBB" bond ratings would still hie able to obtain NRC approval of decommissioning funding into the safe storage period for a .prematurely shut down reactor. This is because most licensees will be able to successfully meet the other criteria deucribed above even if they are unable to pass the "A" bond rating screening test. Recent exemptions issued to two prematurely shut down plants (Ranchc1 Seco and Shoreham) intlicate that bond ratings are only one of several fa1;tors that the NRC will use to determimi whether a licensee has demonstrated reasonable assurance of Its ability to pay decommissioning costs. Fiinally, this discuss.ion on the use of bond ratings is intended as non-binding EUidance only: this final rule includes no such detailed criteria.

2. Comment: It is not necessary to require that all funds should be available in external funds or guaranteed by the tiine final dismantlement activities commence.

A few commenters disagreed with the NRC's statement that all funds are required to be available or guaranteed in external funds by the time final dismantlement activities commence. Some commenters hypothesized , scenarios In which relatively small funding shortfalls may be covered in rates already approved by a licensee's PUC or the FederalEnergy Regulatory Commission (FERC) prior to actual *collection. In thls,situation, funds, although not strictly available at the time final dismantlement activities. commence, would have a high degree of assurance of being obtained by the time the licensee needed to complete the dismantlement activities. Another commenter suggested that the NRC's requirement of full funding prior to the start of final dismantlement operations is inconsistent with a case-by-case approach. This commenter recommends that licensees be required to provide assurance that funds are available to complete specific dismantlement activities, rather than the entire dismantlement process. Response: The NRC disagrees with recommendations that the NRC should abandon its general policy of requiring all funds needed for decommissioning be available prior to the start of final dismantlement. As described in the proposed rule (56 FR 41493), the June 27, 1988, final rule clearly requires funds at the time of permanent end of operations. Section50.75(e)(1) defines the three methods of financial assurance acceptable for power reactor licensees. Two of the methods, prepayment and

surety or insurance, provide all funds by the time of.permanent shutdown.

The* third acceptable method, an external sinking fund, is defined as "a fund established and maintained by setting funds aside periodically in an account segregated from licensee assets and outside the licensee's administrative control in which the total amount of funds would be sufficient to pay decommissioning costs at the time termination of operation is expected." This requirement was imposed to avoid a situation where lack of funds could delay and degrade the decommissioning process to the detriment of public health and safety. Although the dismantlement process can be completed in discrete stages, the potential unavailability of funds at a later stage may conceivably affect the dismantlement process at an earlier 50-SC-105 stage by creating incentives to "cut comers." Thus, this requirement was not altered in the proposed rule on funding for plants that shut down prematurely and will remain in the final rule. 3. ComIJ7ent: Accelerated funding when there is a risk of premature shutdown. One commenter asked the NRC to allow accelerated funding in cases where there is a risk of premature

shutdown.

This commenter specifically referred to its reque.st for FERC to accelerate funding over a shorter period than the fill remaining operating life so that adequate funds would be available at the time of permaneilt*shutdown. The commenter also Indicated that its

request was denied by FERC. Response:

The NRC.strongly supports any effort by its licensees to accelerate funding, especially when a serious . possibility of premature shutdown is . anticipated. The NRC further believes that existing regulations (i.e., 10 GFR 50.75(e)) would allow accelerated funding and that, In appropriate circumstances, accelerated funding could be ordered If necessary of desirable for safety. In any case, the NRC would defer to FERC or the appropriate PUC for the appropriate rate treatment of accelerated* funding. 4. Comment: Amending 10 CFR 50.82 to clarify issuance of possession-only licenses apd other procedural aspects of decommissioning. One commenter recommended that § 50.82 be amended to indicate the timing and procedures for decommissioning. The commenter requests th/lt the NRC specify when it will issue a possession-only license or other license amendments to permits scaling back site operations. Response: The NRC is evaluating its regulations concerning decommissioning and is considering issuing proposed

amendments to clarify Us procedures in the areas suggested by the commenter.

To expedite this rule, however, the NRC will consider the timing of only licenses and other license amendment procedures as part of a separate rulemaking action. 5. Comment: The case-by-case approach '1ails to provide sufficient protection to* the public's health and safety." A commenter argues that many p_lants will shut down prematurely in the future and safe storage is neither risk free nor cheap. Thus, adequate funds for safe storage must be assured, In addition to funds for actual decommissioning. Therefore, plants must have adequate . funding available for the time of PART 50

STATEMENTS OF CONSIDERATION shutdown and not be allowed to fund into the safe storage period. Further, this commenter asserts that "A" bond ratings are inadequate because "in many instances, utility (and other) bonds have oone from investment grade 0
status to junk or default status m one step." In the event of a rrecip~~ating.

incident such as an accident, there 1s no likelihood at all that the derating process will be gradual * * * ." This commenter concludes by stating that the NRC "should determine how to insure, in each and every case, that adequate funds are available." . Response: This conunenter makes several assertions to support the . commenter's opposition to funding during a safe storage period. These assertions, however, are not supported by facts. It is not true that ~oat ~?~d ratings, especially for electric ut1hties: move quickly through several categones of ratings. The process is almost always gradual and, therefore,. would almost always give the NRC time to t~e steps. to assure the adequacy of fundings during a storage period. In addition. this commenter also ignores NRC's requirement that its power reactor licensees carry accident recovery insurance of at least $1.06 billion (10 CFR 50.54(w)) to provide a source of funds for accident cleanup and decontamination. This requirement reduces the likelihood that premature decommissioning resulting from an accident would be particularly financially stressful. More importantly, the NRC would, as stated, evaluate each instance of premature decommis~io~ing _on a by-case Qasis. The cntena ~1scusse~ above provides the NRC with a vanety of measures to assure the adequacy of funding. The case-by-case approach that is being adopted in this 1:11~ all_ows the NRC to consider the parllc1pation financial situation for each licensee that shuts down its power reactor before the expected end of operation life. In spite of the commenter's assertions, the Commission does not expect this to be a frequent occurrence. When it do_es. occur, in most situations ~e ma1onty of decommissioning funds will have been collected during the operating life of the shut down reactor. Most licensees currently have substantial amounts collected and would, at the least. be able to fund activities necessary to place a shut down reactor into safe storage. Whatever funding shortfall remains can be collected or guaranteed in a time frame and through funding mechanisms commeI1cSurate with a licensee's financial situation. As that financial situation changes, the licensee. under NRC monitoring, would alter funding methods accordingly. For the reasons presented in the discussion of issues raised, the NRC is issuing this final rule as proposed. Finding of No Significant Environmental Impact: Availability agencies or persons were contacted for this action, and no other documents related to the environmental impact of this action exist. The foregoing constitutes the environmental assessment and finding of no significant impact for this final rule. Paperwork Reduction Act Statement This final rule clarifies This final rule does not contain a new decommissioning funding arrangements or amended information collection for those licensees whose power requirement subject to the Paperwork reactors are shut down prematurely. Reduction Act of 1980 (44 U.S.C. 3501 et This action is required so that the seq.) Existing requirements were Commission may evaluate on a case-by-approved by the Office of Management case basis the unique financial situation and Budget, approval number 3150-0011. that could confront those licensees. The Commission would continue its Regulatory Analysis requirements for assurance of On June 27, 1988 (5a'FR 24018), t~e decommissioning costs but could alter . NRC published in the Federal Register a the timing of ftmds collection ac_cording final rule* amending 10 CFR parts 30, 40, to a licensee's individual financial 50, 51, 70 and 72 regardin~ g~ne~al situation. The Commission believes that requirements for decomnuss1omng if utility licensees were required to have nuclear facilities. In that rule, the all funds for decommissioning by the ' Commission provided the option that time of permanent shutdown as required power reactor licensees may collect by the existing rule, some utilities could funds for decommissioning over the be unnecessarily financially stressed projected operating life of the facility without significantly increasing the but required that all funds needed for protection of the public health and decommissioning be accumulated by the safety and of the environment. . time of permanent shutdown: Under the Neither this action nor the alternative existing rule, power reactor licensees of maintaining the existing rule would that shut down prematurely would no! significantly affect the environment. have the remaining term of the operating Although changes in the timing.of_ . license to accumulate decommissioning collection of funds for decomm1ss1omng funds and could be unduly b~dened prematurely shut down power reactors financially if required to raise all may affect the financial arrangements of remaining decommissioning funds licensees and may have economic and shortly after shutdown. Consequently, social consequences, they would not the NRC will evaluate the schedule for alter the effect on the environment of collecting decommissioning funds for the licensed activities considered in the prematurely shut down facilities on a final rule published on June 27, 1988 (53 case-by-case basis. A case-by-case FR 24018) as analyzed in the Final approach allows the NRC to evaluate Generic Environmental Impact the particular financial cir.cumsta~ce_s of Statement on Decommissioning of each affected licensee while conhnmng Nuclear Facilities (NUREG-0586, August to ensure that the public health anil 1988).

The alternative to this action safety and the environment are would not significantly affect the . . adequately protected.

This final.rule environment. Therefore, the Comm1ss10n would generally reduce financial costs

has determined, under the National for those licensees allowed to extend Environmental Policy Act of 1969, as the collection period of amended, and the Commission's decommissioning funds. regulations in subpart A oflO CFR pa1: This final rule would not create 51 that this final rule will not be a ma1or substantial costs for other licensees.

The Fe 0 deral action significantly affecting the rule will not signficantly affect state and quality of the human environment, and local governments and geographical therefore an environmental impact regions, or the environment, or. create statement is not required. No other substantial costs to the NRC or other

Copieo or NUREG--0586 may be pun:hased from the Superintendent or Doc:umenta.

U.S. Government Prinlill!! Offica. P.O. Box 3711112, ~a,hi113ton. D.C. 20013-7082. Copies are also .available &om the National Technical Information Senrlce, 5285 Port Royal Road. Springfield, YA 22161. A copy la all!IO available for inspection and copying ror II fee In the NRC Public Document Room. 2120 L Street. NW~ (Lower Level). Wa11hington. DC. 50-SC-106 Federal agencies. The foregoing discussion constitutes the regulatory analysis for this final rule. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. ~b), the Commission certifies that this final PART 50

STATEMENTS OF CONSIDERATION rule will not have a significant impact upon a substantial numbe:r of small entities.

The rule will pote,ntially affect licensees of approximately 118 nuclear power reactors. Nuclear p~wer plant licensees do not fall wiH1i11 the definition of small businenses as defined in Section 3 of the Small Business Act, 15 U.S.C. 632, the Small Business Size Standards of the Small Business Administrator (13 CFR pa:rt 121), or the Commission's Size Stand;;1rds (50 FR 50241; December 9, 1985). Backfit Analysis The NRC has determined that this final rule does not impose a backfit as defined in 10 CFR 50.109(a1)(l). Therefore, a backfit analysis is not required for this rule. List of Subjects in 10 CFR Part 50. Antitrust. Classified information. Criminal penalty, Fire protection. Incorporation by referenl:4~, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reacttir siting 1c:riteria, Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended. the Energy Reorganizatio111 Act of 19'74, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendment to 10 CFR part 50. 57FR34666 Published 8/6/92 Effective 9/8/92 10 CFR Part 50 RIN 3150-ADOS Codes and Standards for Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Final rule. 50-SC-107

SUMMARY

The Commission is amending its regulations to incorporate by reference the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section III, Division 1, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), and the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section XI, Division 1, of the ASME Code. The final rule imposes an augmented examination of reactor vessel shell welds and separates the requirements for inservice testing from those for inservice inspection by placing the requirements for inservice testing in a separate paragraph.

The ASME Code addenda and edition incorporated by reference provide updated rules for the construction of components of light-water-cooled nuclear power plants, and for the inservice inspection and inservice testing of those components. This final rule permits the use of improved methods for construction, inservice inspection, and inservice testing of nuclear power plant components; requires expedited implementation of the expanded reactor vessel shell weld examinations specified in the 1989 Edition of Section XI; and more clearly distinguishes in the regulations the requirements for inservice testing from those for inservice inspection. EFFECTIVE DATE: September 8, 1992. The incorporation by reference of certain publications listed in the regulations is approved by the Office of the Director of the Office of the Federal Register as cif September 8, 1992. FOR FURTHER INFORMATION CONTACT: Mr. G.C. Millman, Division of Engineering, Office of Nuclear . Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: {301) 492-3848. SUPPLEMENTARY INFORMATION: Background On January 31, 1991 (56 FR 3796), the Nuclear Regulatory Commission published in the Federal Register a proposed amendment to its regulation, 10 CFR part 50, "Domestic Licensing of Production and Utilization Facilities," to update the reference to editions and addenda of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). This proposed amendment would revise § 50.55a to incorporate by reference the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section III, Division 1, of the ASME Code, and the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section PART 50

STATEMENTS OF CONSIDERATION XI, Division l, of the ASME Code, with a specified modification.

The modification would require implementation of certain requirements for containment isolation valve (CIV) testing that appear in Section XI Subsection IWV prior to the 1988 Addenda, but which do not appear in the later addenda. The amendment would impose an augmented examination of reactor vessel shell welds, and separate in the regulations the requirements for inservice testing from those for inservice inspection by placing the requirements for inservice testing in a separate paragraph. Summary of Comments Interested parties were invited to submit written comments for consideration in connection with the proposed amendment by April 16, 1991. Comments were received from 29 separate sources. These sources consisted of 23 utilities, one service organization representing four nuclear power plants, the Nuclear Management and Resources Council (NUMARCJ, one owners group (BWR Owners Group (BWROG)), one state entity (Illinois Department of Nuclear Safety (IDNS)), one public citizens group (Ohio Citizens for Responsible Energy (OCRE)), and one independent consultant. The submitted comments generally addressed one of the following subject areas: (1) The incorporation by reference of the specified later addenda and edition of Section III, Division l, and Section XI, Division l, of the ASME Code into § 50.55a; (Z) the endorsement of comments submitted by NUMARC; (3) the proposed modification to Section XI Subsection IWV rules for CIV testing; (4) the proposed augmented reactor vessel examination; (5) the separation of the rules for inservice inspection and inservice testing; (6) the existing scope of § 50.55a for pump and valve testing; and (7) the potential endorsement in § 50.55a of ASME/ ANSI OM part 4 on snubbers. Those who commented on the updating of existing references to Section III and Section XI of the ASME Code in § 50.55a generally noted their approval. One commentor, however, expressed significant concern with the new provision initially specified in the Section XI 1988 Addenda which expands the existing requirement to examine one circumferential and one longitudinal reactor vessel shell weld during the Znd and subsequent inspection intervals to essentially 100 percent of all reactor vessel shell weld during those intervals. Volumetric examination of all reactor vessel shell welds during the first inspection interval has been a requirement in Section XI since the 1975 Addenda. The commentor believes that the expanded examination is unnecessary and that examination efforts should focus on the beltline welds or welds that exceed a specified fluence level. The NRC agrees with the ASME action to expand the reactor vessel examination on the basis that the importance of the reactor vessel. and previous unexpected cracking of primary coolant pressure boundary components, requires that the expanded examinations be performed to ensure the integrity of the reactor vessel. The importance of reactor vessel integrity in protecting the public health and safety demands that periodic, comprehensive inservice examinations of the reactor vessel be made to ensure that structural degradation, if it occurs, does not go undetected. Although the beltline welds do receive the highest radiation, there is simply no assurance that service induced cracking would be limited to those welds. An examination once every ten years of essentially 100 percent of all reactor vessel shell welds is both reasonable and necessary. The comments submitted by NUMARC relate to: (1) The proposed endorsement of a later edition and addenda of the ASME Code, which NUMARC considers to be a positive step; (Z) the proposed modification to Section XI Subsection IWV (i.e., the reference to part 10 of ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987 (OM Part 10)), which NUMARC considers to be inappropriate and unnecessary on the basis that 10 CFR part 50, Appendix J testing is adequate; (3) the proposed augmented reactor vessel examination, which NUMARC recognizes to be important, but suggests that more flexibility be incorporated into the implementation provisions; and (4) the scope of§ 50.55a which NUMARC believes should not be influenced by Generic Letter 89-04. Approximately one-half of the utility cornmentors specifically endorsed the comments by NUMARC. In general, comments from the other utilities were consistent with one or more of the comments from NUMARC. The comments from NUMARC are discussed below, along with comments from others on the same subject. Most of the comments addressed, in part, the proposed modification to Section XI Subsection IWV rules for containment isolation valve testing. Utility comments supported the NUMARC comment, which expressed the belief that the current Appendix J containment leakage testing program already provides an adequate basis for assessing and controlling containment 50-SC-108 leakage and that the modification could result in a valve having to be declared inoperable immediately, in spite of the fact that the total containment leakage may be substantially less than allowable. NUMARC suggested that, in lieu of reinstating requirements for specific valves, NRC recommend to the ASME Operations & Maintenance (O&MJ Committee that it perform a comprehensive review of the testing requirements for containment isolation valves and acceptance standards for those tests. IDNS agreed with the NRC position that the requirements for leakage rate analysis and provisions for corrective action should be maintained, but believed that it would be less confusing for licensees if those requirements were incorporated into the existing requirements for Type C testing in Appendix J. OCRE strongly supported the action by NRC to modify the Section XI rules for CIV testing. The NRC concern that resulted in the proposed modification to Section XI Subsection !WV sterns from the findings of two reviews and a follow-on study of

Appendix J leak test results. The overall findings show that valve leakage is the primary contributor to occurrences of containment unavailability and that such occurrences generally involve small, rather than large, leaks. Risk to the public from small leakage events is very low, but the NRC is concerned that eliminating the existing Section XI requirement to analyze leakage rates and to take corrective action in the event of abnormally high leakage rates for those CIVs that do not provide a reactor coolant system pressure isolation function could reduce the ability to detect degrading valves and, thereby, could permit an unacceptable reduction in the present safety margin associated with the leak tight integrity of those CIVs and, thereby, the containment.

It was specifically noted in the proposed rule that the NRC was interested in receiving comments on the discussed basis for and content of the proposed modification, and was particularly interested in receiving comments that would provide insight and justification, based upon plant experiences, relative to the need for revising or possibly eliminating the proposed modification. Many comments were received that express concern with the proposed modification. However, these comments, which generally state the opinion that Appendix J requirements are adequate and sufficient with regard to ensuring containment integrity, are of a qualitative nature and no specific plant PART 50

STATEMENTS OF CONSIDERATION data or operational experiences were provided or referenced that updated the results of the earlier studies. No additional substantive information was provided for the NRC to consider relative to the need for revising or possibly eliminating the proposed modification.

It has not been demonstrated, by analyt1is of more recent and comprehensive containment leakage test data, that containment leakage integrity can be maintained at an acceptable level without continued implementation of the existing Appendix J valve leak rate test program in conjunction with the Section XI requirement for analysis of leak rates. Consistent with the comment by NUMARC, the NRC staff discussed the basis for OM part 10 CIV testing requirements with repret1entatives from the ASME O&M Committee. Based on these discussions and in concert with the O&M Committee organization, the O&M Committee has ini1iated action to (1) perform a comprehensive review of OM part 10 CIV testing requirements and acceptance standards and (2) develop a basis document that would provide, as a minimum, ~! documented basis for not including the requirements for analysis of leakage rates and corrective actions in OM part 10 for those CIVs that do not pmvide a reactor coolant system pressure lsolation function. The NRC will mevaluate the need for the modification to Section XI Subsection IWV, following review of this basis document. It is anticipated that this will occur as pa1:t of a future rulemaking proceeding that will address the incorporation by refe:rence of the ASME O&M Code into § 50.55a. In the meantime, this final rule incorporates by reference the 1988 Addenda and 1989 Edition of Section XI, Division 1, with a specifiHd modification for CIV testing that is provided in a new § 50.55a(b)(2)(vii). The modification substantially preserves the existing requirements for analysis of leakage rates and corrective actions that exist in Subsection IWV prior to the 1988 Addenda. Specifically, th,e modification requires that licensees implement the requirements of ParagrapJ1 4.2.2.3(e), "Analysis of Leakage Rates," of part 10 and Paragraph 4.2.2.3(f), "Corrective Action," of part 10, in addition to the requirements of Paragraph 4.2.2.2 of part 10, for all Category A valves that are CIV s, regardless of whether or not they provide a reactor coolant system pressure isolation function. Because paragraph 4.2.2.3( e) of part 10 is specified in the modification rather than the existing IWV-3426, th,~ existing Section XI requirement is somewhat relaxed by permitting valve combinations rather than specific valves to be analyzed. This recognizes that, in the past, requests for relief have been granted where design constraints necessitate testing-combinations of valves with permissible leak rate limits applied to valve groups. The specified modification does not require the present practice of trending NPS 6 and larger valves because that requirement has not been carried from IWV-3427[b) to OM part 10. Section XI Subsection IWV (1988 Addenda and 1989 Edition), Subsection IWP (1988 Addenda and 1989 Edition), and Subsection IWF (1987 Addenda, 1988 Addenda, and 1989 Edition) reference ASME/ ANSI OM part 10, ASME/ ANSI OM part 6, and ASME/ ANSI OM part 4, respectively. During preparation of this final rule, it was recognized that Table IWA-1600-1 in the applicable Section XI addenda and edition specifies a nonexistent revision for OM part 10 and part 6, and does not specifically identify the applicable revision for OM part 4. The Section XI Subcommittee on Inservice Inspection has taken action to correct the revision reference, which, for these standards, should be the ANSI/ASME OMa-1988 Addenda to the ASME/ANSI OM-1987 Edition. To ensure that licensees are aware of the correct revision reference to the OM standards, an additional modification, § 50.55a[b)(2)(viii), has been added to specify that the OMa-1988 Addenda is the applicable revision to the OM-1987 Edition for OM part 4, part 6, and part 10 when using the noted Section XI addenda and edition. The NUMARC comment relative to the proposed augmented examination of the reactor vessel indicates an understanding of the NRC position on the need for this examination, but notes concern with the specifics of the proposed implementation. Specifically, NUMARC expresses concern that: (1) Better utilization of available inspection resources could be accomplished by limiting application of the augmented inspection program to the reactor vessel beltline shell welds, or by limiting implementation of the augmented examination to reactor vessel shell welds that exceed a specific neutron flux exposure (this comment differs from the one utility comment noted above relative to updating later edition and addenda of Section XI in that it only refers to the augmented examination); (2) tooling for the older Boiling Water Reactors (BWRs) may generally not be available in the time-frame needed; (3) only those reliefs which address the scope and extent of shell weld 50-SC-109 examinations should be revoked, and they should be revoked on a plant specific basis; and (4) the NRC should state its willingness to accept requests for specific new exemptions, based on the availability of suitable equipment and technology at the time of the scheduled inspection and the appropriate technical justification. Other comments on the augmented examination include those from: BWROG, which noted concern for those plants close to the end of the current interval that could not practically incorporate the augmented examination into the current interval and would have to perform that examination during the first period of the next interval (Note: The deferred augmented examination may be used as a substitute for the reactor vessel shell weld examination normally scheduled for the interval in which the deferred examination was performed(§ 50.55a(g)(6)(ii)(A){3}, therefore, the impact of deferring the augmented examination will be reduced); IDNS, which strongly supports the NRC position regarding the augmented examination of the reactor vessel; and OCRE, which also strongly supports the augmented examination and notes that the examination will not only provide an additional assurance of safety, but will aid in understanding aging degradation phenomena which will assist licensees that wish to pursue license renewal. The NRC position with regard to the augmented examination of the reactor vessel, as previously stated in the Supplementary Information to the proposed rule, is that degradation of reactor vessel materials has become more of a concern recently, because: (1) Results from irradiation surveillance material tests show that certain reactor vessel materials undergo greater radiation damage than previously expected, (2) indications from operational data show that stress corrosion cracking of BWR reactor vessels is more probable than was thought several years ago, and (3) significant service induced cracking has occurred in large vessels (i.e., pressurizer, steam generators) designed and fabricated to the ASME Code. It is the judgment of the NRC that, because of new information and previous limited examinations of reactor vessels, there may exist a substantially greater potential for reactor vessel degradation, in all areas of the reactor vessel, than previously considered and that maintenance of the level of protection presumed by the regulations requires more than compliance to existing regulatory requirements. The NRC has PART 50

STATEMENTS OF CONSIDERATION determined that the augmented examination of reactor vessels will result in a substantial increase in the overall protection of the public health and safety, and that the costs of implementation are justified in view of the increased protection.

The backfit analysis required by § 50.109, "Backfitting," is provided as part of the regulatory analysis that supports this final rule. However, the NRC agrees with comments that additional flexibility and specificity will improve implementation of the augmented examination of reactor vessel examination. To this end, the augmented examination of reactor vessel shell welds specified in this final rule includes the following new provisions and clarifications: (1) The revocation of previously granted reliefs is limited to those reliefs that deal with the extent of volumetric examination of reactor vessel shell welds; (2) the augmented examination will be performed in accordance with the section XI edition and addenda applicable to the inspection interval in which the examination is actually performed; (3) "essentially 100%" as used in § 50.55a(g)(6)(ii)(A) means "more than 90 percent of the examination volume of each weld, where the reduction in examination volume is due to interference from another component, or part geometry;" (4) licensees that defer the augmented examination to the next interval are permitted to retain all existing approved reliefs for the current interval; (5) licensees with fewer than 40 months remaining in the inspection interval in effect when the rule becomes effective are permitted to extend the interval in accordance with the provisions of section XI (1989 Edition] IW A-2430(d); (6) licensees that are unable to satisfy completely the requirements for the augmented examination may request to perform alternate examinations in accordance with § 50.55a(g)(6)(ii)(A)(5]. These items are addressed individually in the discussion below regarding provisions of the augmented reactor vessel shell weld examination. Section 50.55a(g)(6)(ii) addresses augmented inservice inspection programs for those systems and components for which the Commission determines that added assurance of structural reliability is necessary. For that purpose, and consisent with the discussion in this final rule, § 50.55a(g)(6)(ii)(AJ has been added to require expedited implementation of the reactor vessel shell weld examinations specified in the 1989 Edition of section XI, Division 1, in item Bl.10, "Shell Welds," of Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel," in Table 2500-1 of subsection IWB, "Requirements for Class 1 Components of Light-Water Cooled Power Plants." In order to ensure the applicability of the new augmented examination to all licensees, § 50.55a(g)(6)(ii)(A](l) revokes all previously granted reliefs relating to the extent of volumetric examination of the reactor vessel shell welds that apply to examinations for the inservice inspection interval that is in effect when the rule becomes effective subject to a specified modification. Limiting the revocation of previously granted reliefs to those that deal with the extent of the volumetric examination permits the retention of those approved reliefs that deal with issues such as specification of calibration blocks. Licensees that choose to defer the augmented examination to the next interval in accordance with § 50.55a(g)(6)(ii)(A)(3] should note that paragraph (iv) of that section modifies the revocation of approved reliefs to permit retention of previously approved reliefs for the current interval when the augmented examination in deferred. This provision recognizes that plants that previously received relief from the section XI reactor vessel shell weld examination and satisfy the condition to defer the augmented examination may find it impractical to implement the section XI examination during the current inspection interval. Section 50.55a(g)(6)(ii)(A)(2) requires all licensees to implement the specified augmented examination of reactor vessels during the inspection interval in effect when this rule becomes effective, subject to conditions specified in § 50.55a(g)(6)(ii)(A)(3) and (4). Section 50.55a(g)(6)(ii)(A)(2) specifically permits the use of the augmented examination, when not deferred, as a substitute for the reactor vessel shell weld examinations scheduled for the inspection interval in effect when the rule becomes effective, ancl. specifies that, for the purpose of this rule, "essentially 100 percent" as used in Table IWB-2500-1 means "more than 90 percent of the examination volume of each weld, where the reduced examination volume is due to interference from another component, or part geometry." This is consistent with section XI Code Case N--460, which previously has been approved for use in Regulatory Guide 1.147. It is recognized that it may be necessary to implement a combination of internal and external diameter examinations to achieve "essentially 100%" examination volume 50-SC-110 coverage for each weld. A clarification has been included in this section to note that the augmented examination may be used as a substitute for the reactor vessel shell weld examination in the interval in effect when the rule becomes effective when the augmented examination is not deferred. This is a reinforcement of § 50.55a(g)(6)(ii)(A)(3], as it appears in both the proposed and final rule, which specifies that the deferred examination may not be used as a substitute for the reactor vessel shell weld examination scheduled for implementation during the inservice inspection interval in effect when the rule becomes effective. The NRC recognizes that plants with fewer than 40 months remaining in the inspection interval when this rule becomes effective may find it impractical to implement the augmented examination of the reactor vessel during that inspection interval. Therefore, § 50.55a(g)(6)(ii)(AJ(3) permits plants with fewer than 40 months remaining in the inspection interval when this rule becomes effective to defer the augmented examination until the first period of the next inspection interval. However, this same paragraph specifically prohibits the use of the deferred augmented examination as a substitute for reactor vessel shell weld examinations scheduled for the inspection interval in effect when the rule becomes effective. The intent is to ensure that the examinations are deferred only when necessary and not to have the rule encourage a 40-month delay in reactor vessel shell weld examinations. Further, § 50.55a(g)(6)(ii)(A)(3) permits using the deferred examination, with a condition, as a substitute for reactor vessel shell weld examinations scheduled for the inspection interval in which the deferred examinations are performed. The condition is that subsequent reactor vessel shell weld examinations for successive inspection intervals be performed in the first period of the inspection interval. This condition is necessary to prevent a potential 160-month gap between reactor vessel shell weld examinations. This gap would occur if a plant used the deferred examination performed in the first period as a substitute for the scheduled examination and then deferred the examination for the next inspection interval to the end of that interval as permitted by section XI. In addition, this section specifies that licensees with fewer than 40 months remaining in the inservice inspection interval in effect when the rule becomes effective may extend that interval in accordance with PART 50

STATEMENTS OF CONSIDERATION the provisions of section XI (1989 Edition] IWA-2430(d]

to permit implementation of the augmented examination during the current interval. It is not the intent of the NRC to permit licensees in the second period of an inspection interval to red 0 t1ce the interval length for the purpose of "being within 40 months of the end of the interval" and, thereby, deferring the augmented examination to the first p,~riod of the subsequent interval. Section 50.55a(g](6](ii)(A)(4) specifies that a licensee that has either completed or has scheduled an inspe'Ction of essentially 100 percent of the length of all Examination Category B-A shell welds during the inservice inspection interval in effect when the rule becomes effective does not have to implement the required augmented examination of the reactor vessel shell welds. Primarily, this paragraph is intended. to permit licensees who are in the first inspection interval to use the essentially 100 percent reactor vessel shell weld examination required for that interval by section XI to satisfy the requirement for the augmented examination of the reactor vessel. The technical objective of the augmented examination will be accomplished under these conditions. These licensees will continue to apply the current requirements of § 50.55a(g](4) until the next inspection interval when future examinations will be performed based on ASME section XI, 1989 Edition, or later Code edition and addenda specified in !i 50.55a(b ]. The augmented examination specified in § 50.55a(g)(6J(ii](A) is not an ASME Code requirement. It is a mquirement specifically developed and additionally imposed by the Commission. Therefore, except for the specific provisions in § 50.55a(g)(6)(ii](A)(2) and (3) that permit using the augmented examination as a substitute for section XI required reactor vessel shell weld examinations, the closing out of an inservice inspection interval is not dependent on completion of the augmented examination. In the specific instance where the augmented examination is deferred to the first period of the next inspection interval. the current inspection interval could be closed out relative to reactor vessel shell weld examinations by implementing the regularly scheduled reactor vessel shell weld examinations as modified by previously a.pproved applicable relief requests for that interval. The NRC recognizes that, as noted by commentors, there may exiBt conditions that prevent licensees from completely satisfying the requirements for the augmented reactor vessel shell weld examination as specified in § 50.55a(g)(6)(ii](AJ. For this reason, § 50.55a(g)(6](ii](A)(5) has been added to permit licensees that make a determination that they are unable to completely satisfy tbe specified augmented examination to propose and use alternatives that have been authorized by the NRC's Director of the Office of Nuclear Reactor Regulation. This final rule amends § 50.55a to separate the requirements for inservice testing from those for inservice inspection by moving the requirements for inservice testing to a separate paragraph. Previously, § 50.55a(g], "lnservice inspection requirements," specified the requirements for (1) preservice and inservice examinations for Class 1, Class 2, and Class 3 components and their supports, (2) system pressure tests for Class 1, Class 2, and Class 3 components, and (3) inservice testing of Class 1, Class 2, and Class 3 pumps and valves. In order to emphasize the importance of inservice testing and to distinguish more clearly its requirements from those of inservice inspection, this final rule moves the requirement for inservice testing from § 50.55a(g], "lnservice inspection requirements," to a separate (previously reserved] § 50.55a(f), which is titled "lnservice testing requirements." All existing requirements for inservice examination and system pressure testing are retained in § 50.55a(g]. There is overall favorable acceptance of the separation of the requirements in the regulation for inservice testing and for inservice inspection. It is generally believed by the commentors, as it is believed by the NRC, that the separation serves to clarify and emphasize the requirements for inservice testing. Two administrative changes were made in the development of § 50.55a(f) relative to existing § 50.55a(g]. First, § 50.55a(f)(6)(ii) has been added to indicate the Commission's intent to impose an augmented inservice testing program if added assurance of operational readiness is deemed necessary. This paragraph only indicates intent and does not impose a specific requirement. It does parallel the existing § 50.55a(g)(6)(ii] which specifies that the Commission may require an augmented inservice inspection program for systems and components for which it deems that added assurance of structural reliability is necessary. One utility commentor expressed concern that the addition of § 50.55a(f](6)(ii] would permit the Commission to impose an augmented inservice testing program without further justification. This is not the case. Any program for augmented 50-SC-111 inservice testing will be fully justified with a documented regulatory analysis that includes the appropriate backfit analysis. The intent of the NRC to perform the necessary backfit analysis is clearly demonstrated by the backfit analysis that was performed to require the augmented examination of the reactor vessel that is specified in § 50.55a(g)(6)(ii)(A) of this final rule. Second, this final rule includes the addition of introductory text to § 50.55a(g) which states that the requirements for inservice testing of Class 1, Class 2, and Class 3 pumps and valves are located in § 50.55a(f). This change is necessary because the placement of inservice testing requirements into a separate § 50.55a(f), as included in the proposed rule, would have caused administrative inconsistencies with regard to existing references to § 50.55a(g] for inservice testing in documents such as technical specifications, safety analysis reports, procedures, and records. With this change, existing references to § 50.55a(g] for inservice testing will refer the user to § 50.55a(f), where the specific requirements for inservice testing are located. The NRC recommends that as the governing documents are updated, the direct reference to § 50.55a(f) be incorporated, as appropriate. Two editorial revisions, relative to the previous § 50.55a(g], are included in the new § 50.55a(f). These editorial revisions: (1) Reserve § 50.55a(f)(3] (i] and (ii) so that the structure of § 50.55a(f) will parallel that of § 50.55a(g) for the purpose of promoting easier cross-referencing between the two paragraphs; and (2) modify the reference to 120-month inspection interval in § 50.55a(g] to 120-month interval in § 50.55a(f), because the term "inspection interval," as used in Section XI, is used only in the context of inservice inspection. (The term "test interval" was not used because, unlike inspection interval, the 120-month time frame does not designate a period of required actions for the testing program. The 120-month interval used in § 50.55a(f) and the 120-month inspection interval used in § 50.55a(g] are considered by the staff to be coincident for the purpose of 120-month updating requirements.] A number of comments were received regarding the scope of § 50.55a as applied to pump and valve testing. These comments ranged from recommending that the scope of§ 50.55a be expanded to be consistent with the scopes of OM part 6 and part 10, which go beyond Class 1, Class 2, and Class 3 components, to recommending that the PART 50

STATEMENTS OF CONSIDERATION scope of § 50.55a be limited to ASME Code classified components.

One commentor expressed concern that the Supplementary Information in the proposed rulemaking addressed Generic Letter 89-04 in a way that seemed to include the letter in the rulemaking. That was not intended. To the contrary, the intent of this rulemaking is to maintain the existing scope of § 50.55a for pump and valve testing. For plants whose construction permits were issued on or after January 1, 1971, that scope constitutes Code classified components as specified in existing § 50.55a(g] (2) and (3) (i.e., § 50.55a(f] (2) and (3) by this rulemaking). For those plants whose construction permits were issued prior to January 1, 1971, that scope constitutes components of the reactor coolant pressure boundary which must meet the requirements applicable to components that are classified as ASME Code Class 1, and other safety-related pumps and valves which must meet the requirements applicable to components that are classified as ASME Code Class 2 or Class 3, as specified in e~isting § 50.55a[g)(1) (i.e., § 50.55a(f][1) by this rulemaking]. The reference to the generic letter has not been included in the final rule. A number of comments were received with regard to snubber testing which is outside the scope of this rulemaking. Commentors generally suggested that ASME OM part 4, "Examination and Performance Testing of Nuclear Power. Plant Dynamic Restraints (Snubbers]," which is referenced in Subsection IWF in the 1987 Addenda, 1988 Addenda and 1989 Edition of Section XI, be incorporated by reference into § 50.55a. Subsection IWF, "Component Supports," provides rules for the examination of component supports, and the testing of snubbers. Prior to the 1987 Addenda, Subsection IWF provided self-contained rules for the testing of snubbers. Section 50.55a does not specify requirements for the testing of snubbers. This was clarified by the separation of requirements for inservice testing and inservice inspection. lnservice testing requirements specified in § 50.55a(f] apply only to pumps and valves. The testing requirements specified in OM part 4 and referenced in Section XI Subsection IWF article IWF-5000 are not incorporated by reference into § 50.55a. Requirements for the testing of snubbers are generally governed by plant technical specifications. NRC is in the process of initiating a proposed rulemaking that would, among other things, address the incorporation by reference of the ASME OM Code, which contains rules for pump, valve, and snubber testing, into § 50.55a(f]. The NRC will as a part of this future rulemaking determine the need for and acceptability of endorsing the ASME OM Code rules for snubber testing. However, in accordance with requirements for examination of component supports specified in § 50.55a(g), licensees are required to implement the rules for examination of snubbers that are provided in OM part 4 as referenced in Subsection IWF Article IWF-5000 in the applicable Section XI addenda and edition of this final rule. Section 50.55a[g) provides requirements for selecting the ASME Code edition and addenda of Section XI to be complied with during the preservice inspection(§ 50.55a(g](3], for plants whose construction permit was issued on or after July 1, 1974); the initial 10-year inspection interval (§ 50.55a(g](4)[i]J; and successive 10-year inspection intervals (§ 50.55a[g)(4)(ii)J. As noted in the final rule codifying the most recent amendment to § 50.55a (May 5, 1988; 53 FR 16051), paragraph IWA-2400 of Section XI (as revised by the Winter 1983 Addenda] incorporated rules for selecting the applicable edition and addenda of Section XI during the preservice inspection (IWA-2411), the initial 10-year inspection interval (IWA-2412), and successive 10-year inspection intervals [IWA-2413). The criteria provided in the regulations and Section XI are effectively the same for the preservice inspection and the successive 10-year inspection intervals, but differ for the initial 10-year inspection interval. In general, use of the Commission requirements will result in the selection of a more recent edition and addenda than will use of the Section XI rules. Satisfying the requirements of § 50.55a(g)(4J[i] for the initial 10-year inspection interval will, in general, also satisfy the rules of Section XI. Although the Section XI requirements for selecting editions and addenda remain unchanged in the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition, the Commission is reaffirming its intent that in all cases the existing requirements in § 50.55a[g] be the basis for selecting the edition and addenda of Section XI to be complied with during the preservice inspection, the initial 10-year inspection interval, and the successive 10-year inspection intervals. This final rule makes a number of editorial changes to § 50.55a for the purpose of adopting a standard convention for imposing an obligation or expressing a prohibition. In this convention "shall" is used to impose an obligation on an individual or legal 50-SC-112 entity capable of performing the required action, "must" is used as the mandatory form when the subject of the sentence is an inanimate object, and "may not" is used to impose a prohibition. The following paragraphs are amended solely to be consistent with this convention: The introductory paragraph to the section; paragraphs (a](l), (a)(3), (b](2)(iii], (b][2)(iv], [g](1), (g)(3](ii], (g)(3)(iii), (g)(3](iv], introductory paragraph to (g)(4), [g)(4)(i), [g)[4)(ii], (g)(5)(i], (g](5)(iv], (g)(6](i], (h], and footnote 8. Other paragraphs are amended for the same editorial reason, but they also contain technical revisions relevant to other parts of this final rule. Section 50.55a(f] has been developed consistent with the noted convention. Subsection IWE, "Requirements for Class MC Components of Cooled Power Plants," was added to Section XI, Division 1, in the Winter 1981 Addenda. Since § 50.55a does not currently address the inservice inspection of containments and the scope of § 50.55a is not affected by this final rule, the requirements of Subsection IWE are not imposed upon Commission licensees by this amendment. The incorporation by reference of Subsection IWE into § 50.55a is presently the subject of a separate rulemaking action. Section 50.55a(b](2)(vi) is reserved for that action. The NRC previously alerted all holders of operating licenses or construction permits for nuclear power reactors, through NRC Information Notice No. 88--95 (IN 88-95), "Inadequate Procurement Requirements Imposed by Licensees on Vendors," to the potential that inadequate licensee procurement requirements or implementation by vendors in supplying components under the ASME Code could result in failure by these vendors to fully implement 10 CFR part 50, Appendix B (Quality Assurance Criteria]. The problem, which was revealed during routine NRC inspections of vendors, resulted from the belief by some vendors that if an item was exempted by the ASME Code from Code requirements, the item was exempt from all other regulatory requirements. The apparent belief of some vendors was that since NRC endorses the ASME Code in its regulations and has accepted the various exemptions, there are, therefore, no other applicable regulatory requirements. This belief is not consistent with the NRC position. The NRC reaffirms its position which, as previously put forth in IN 88-95, states that all safety-related items, even those exempted from ASME Code requirements, are required to be PART 50

STATEMENTS OF CONSIDERATION manufactured under a quality assurance program that meets the requirements of 10 CFR part 50, appendix B. Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 CFR part 51, that this rule is not a major Federal action that significantly affects the quality of the human environment and therefore an environmental impact statement is not required.

This final rule is one part of a regulatory framework dirncted to ensuring pressure vessel i:11tegrity, and the operational readiness of pumps and valves. Therefore, in the general sense, this rule will have a positive impact on the environment. This rule incorporates by reference into the NRC regulations improved rules contained in the ASME Code for the construction, inservice inspection, and inservice testing of components used in nuclear power plants. In addition, this rule requires an augmented examination of reactor vessel shell welds to further ensure the structural integrity of the r,eactor vessel. The occupational exposureis attributable to the expanded reactor vessel examinations contained in the ASME Code and the augmented examination are not expected to be significant because exposures will be limited by the use of remote examination equipment. Occupational exposures associated with the augmented reactor ves!1el examination will be further limited by provisions in the final rule that permit, under certain conditions, the licensee to satisfy the requirement for the augmented examination by previously scheduled or implemented reactor vessel examinations, or by deferring the examination to the next int,~rval and using the deferred examination as a replacement for the previously scheduled examination for that interval. The actions required by applicants and licensees to implement the final rule are of an established nature tha.t should not increase the potential for a negative environmental impact. The environmental asses!1ment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from Gilbert C. Millman, Division of Engineering, Office of Nucle.ar Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301) 492-3848. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Management and Budget approval number 3150--0011. The public reporting burden for this collection of information is estimated to average 42 hours per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB-7714), U.S. Nuclear Regulatory Commission, Washington, DC 20555, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-3019, (3150-0011), Office of Management and Budget, Washington, DC 20503. Regulatory Analysis The Commission has prepared a regulatory analysis for this amendment to the regulations. The analysis examines the costs and benefits of the alternatives considered by the Commission. Interested persons may examine a copy of the regulatory analysis at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the analysis may be obtained from Mr. G.C. Millman, Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone (301) 492-3848. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission hereby certifies that this rule will not have a significant economic impact on a substantial number of small entities. This rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR part 121. Since these companies are dominant in their service areas, this rule does not fall within the purview of the Act. 50-SC-113 Backfit Analysis The final rule incorporates by reference a later edition and addenda to Section III, Division 1, and, with both a technical and nontechnical modification, Section XI, Division 1, of the ASME Code; imposes an augmented __ examination on reactor vessels; and. separates the requirements for inservice inspection from those for inservice testing. The incorporation by reference into the regulations of later editions and addenda of Section III and Section XI of the ASME Code is not a backfit because Section III requirements apply only to new construction, except as voluntarily implemented* by licensees, and because updated Section XI requirements are an integral part of the longstanding § 50.55a(g)(4)(ii) requirement to update inservice inspection and inservice testing programs to reflect the requirements of the latest edition and addenda of Section XI incorporated by reference in § 50.55a(b) 12 months prior to the start of the 120-month inspection interval, subject to specified limitations and modifications. The technical modification to part 10 of ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987 specified in§ 50.55a(b)(2)(vii) is not a backfit because it simply retains an existing Section XI requirement for containment isolation valve testing that licensees now are required to implement in accordance with § 50.55a(g). The nontechnical modification specified in § 50.55a(b)(2)(viii) is not a backfit because it only serves to properly identify an incorrectly referenced standard in Section XI. The NRC has concluded, based on the analysis required by § 50.109(a)(3) which is provided in the regulatory analysis, that the backfit that will be imposed by the augmented reactor vessel examination specified in § 50.55a(g)(6)(ii)(A) will result in a substantial increase in the overall protection of the public health and safety, and that the direct and indirect costs of implementation are justified in view of the increased protection. The separation in the regulation of the inservice inspection and inservice testing requirements is an administrative reorganization of § 50.55a that has no impact on existing technical requirements and, therefore, has no effect on backfitting. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear PART 50

STATEMENTS OF CONSIDERATION power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C .. 552 and 553, the NRC is adopting the following amendments to 10 CFR part 50. 57 FR 39353 Published 8/31/92 Effective 10/1 /92 10 CFR Parts 20 and 50 RIN 3150-AE30 Reducing the Regulatory Burden on Nuclear Licensees AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations to reduce the regulatory burden on nuclear licensees.

This action reflects an initiative undertaken by the Commission in response to a Presidential memorandum requesting that selected Federal*agencies review and modify regulations that would eliminate any unnecessary burden of governmental regulation and ensure that the regulated community is not subject to duplicative or inconsistent regulation. In that spirit, the NRC's Committee to Review Generic Requirements (CRGR) identified eight areas where regulations could be revised to reduce the regulatory burden on licensees without in any way reducing the protection for the public health and safety or the common defense and security. The final amendments address unnecessary* regulatory requirements related to the frequency of reporting information, analysis of emergency core cooling systems for operating power reactors, 50-SC-114 and clarification and update of regulations affecting certain material licensees. EFFECTIVE DATE: October 1, 1992. FOR FURTHER INFORMATION CONTACT: Mr. C.W. Nilsen, telephone (301) 492...: 3834 or Mr. Joseph J. Mate, telephone (301) 492-3795, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555. SUPPLEMENTARY INFORMATION: Background On January 28, 1992, the President of the United States signed e memorandum addressed to selected Federal Agency Heads who are concerned with energy production end protection of the environment. The memorandum requested the addressees work together to streamline the regulatory process end ensure that the regulatory community is not subject to duplicative or inconsistent regulatic;>n. On the same day, the President signed a second memorandum entitled "Reducing the Burden of Government Regulation." This memorandum, which was sent to all Federal agencies, set aside a 90-dey period to review end evaluate existing regulations end programs and to identify and accelerate action on initiatives that will eliminate any unnecessary regulatory burden. At the end of the review period, agencies were to submit e written report indicating the regulatory changes recommended or made during the review period end the potential savings es e result of the changes.

In response to the Presidential memoranda, the Commission-decided that it would be consistent with its policy to monitor the impact of complying with NRC regulations by its licensees to instruct its Committee to Review Generic Requirements (CRGR)
to review existing NRC regulations to determine whether regulatory burdens can be reduced without in any way reducing the protection for the public health and safety and the common defense end security.

In accomplishing their review, the CRGR drew upon previous studies end solicited comments from the public, other Federal agencies, end the Commission's staff. A Federal Register Notice was published on February 24, 1992 (57 FR 6299) seeking public comment in connection with the review, end e second Federal Register Notice on March 23, 1992 (57 FR 9985) discussed likely or possible candidates for action, based on CRGR's preliminary evaluation of comments. An as8ocleted PART 50

STATEMENTS OF CONSIDERATION public meeting was held on March 27, 1992, in Bethesda, Maryland.

After completing their s;pecial review, the CRGR recommended 1:evising the regulations in eight areas. The proposed revisions met the criteria for reducing the burden without in any way reducing the protection for public health and safety and common defense and security. The Chairman of the NRC sent a report to the President of the United States on April 27, _1992, which summarized NRC's activities concerning the President's directive and advised the President that NRC would pursue the CRGR's recommendations expeditiously within the framework of the procedures and practices for rulemaking. On June 1, 1992, in respCtnse to a memorandum from the President of the United States, dated April 29, 1992, the Commission directed the staff to strive to publish the proposed rule changes in the eight areas identified by the CRGR in the Federal Register for comment as soon as possible, but not later than June 15, 1992, with a view to issuing the final rules in the Federal Registur no later than August 27, 1992. On June 16, 1992 (57 FR 27167), the NRC published the proposed rulemaking in thn Federal Register-for comment. The comment period expired on July 20, 1992. Summary and Analysis of li>ublic Comments Thirty comment letters were received on the proposed rule and a,:e available for public inspection, and copying for a fee, at the Commission's Pu1blic Document Room located at 2120 L street, NW. (Lower Level), Washington, DC. The comments on the proposed rule came from a variety of sources. These included private citizens, publicly-held corporations, citizens' groups, the armed forces, industry representatives, electric power companies or their representatives, and legal firms. Eleven significant points were rais11d by the commenters. Of the 30 comment letters received, 26 letters were favorable and 2 letters were partially opposed to the regulation changes. The comments end their resolutions are discuss,ed below. 1. Comment. One commenter suggested that the Commission not only amend§ 20.190B(b) concem:ing contamination monitoring, but also issue a statement that those licemiees still operating under the old part 20 not be required to monitor packages for contamination that meet the conditions of§ 20.1906(b). Response. The NRC does not believe that the suggested change by the commenter is necessary bec,ause the amendment of§ 20.1906(b) will make the subject contamination monitoring requirements of the new part 20 essentially the same as those contained in the existing part 20 (§ 20.205(b)(1)(iii) and (c)(l)). 2. Comment. One commenter opposed the rule on the basis that sealed sources routinely leak and, therefore, should not be excluded from monitoring. The commenter cited an example where a driver and a truck were contaminated because of a failure to conduct a proper radiation sweep. Response. The final rule does not exempt licensees from monitoring or surveying any packages with evidence of degradation of package integrity, including evidence of potential contamination. Likewise, this revision does not relax the preshipment requirements for monitoring of packages contained in 10 CFR part 71. The NRC does not have any evidence that supports the commenter's assertion that sealed sources routinely leak and, thus, the NRC believes that the requirements in place are sufficient to detect potential abnormal situations. No amount of regulation can, a priori, preclude all incidents involving leaking sources. However, these incidents can be dealt with through followup inspection and enforcement under the present regulatory scheme. 3. Comment. Several commenters addressed in general terms the need for the NRC to continue its efforts to reduce any unnecessary regulatory burden on licensees through amendments to 10 CFR chapter I. Response. The NRC will continue its efforts to identify additional amendments that will provide for a reduction in regulatory burden while still assuring adequate protection of the public health and safety. 4. Comment. One commenter questioned the basis for exempting from external monitoring for radiation levels only nuclear material that was either in the form of a gas or in a special form since the external radiation levels are dependent upon radionuclides, quantity, shielding, and distance between radioactive material and the point of interest rather than material form. Response. The NRC agrees with the commenter that the requirement to survey, upon receipt, the radiation levels on the package exterior should be based on the potential radiation hazard. Therefore. the requirement specified in 10 CFR 20.1906[b)(2) that monitoring of radiation levels be.performed on labeled packages is being revised to delete the exemption that the radioactive material be in the form of a gas or in special form as defined in 10 CFR 71.4. 50-SC-115

5. Comment. One commenter questioned whether the monitoring requirements were applicable for packages that show evidence of damage. Response.

The wording of 10 CFR 20.1906(b)(3) has been revised to indicate more clearly that packages with evidence of damage are to be monitored for both radioactive contamination and for radiation levels. 6. Comment. Several commenters requested that the proposed wording to 10 CFR 50.71(e)(4) concerning FSAR updates be revised to decouple the FSAR updates from the refueling cycle and that the 24-month requirement for updates is an unnecessary restriction. Response. The proposed changes were not accepted. The majority of facility design changes reflected in an updated FSAR are effected during the refueling outage. The use of the refueling cycle interval provides for a current plant status document that is coordinated with plant changes. The wording of§ 50.71(e)(4) is not restrictive to plants that will eventually increase their refueling cycle to 24 months. 7. Comment. Three electric utilities requested that the proposed wording in 10 CFR 50.36(a)(2) concerning radiological effluent reporting be revised to specify a particular date. One commenter suggested: "The report must be submitted as specified in § 50.4 prior to March 31 of each year." Response. The wording of 10 CFR 50.36(a)(2) gives the licensee maximum flexibility for scheduling submission_ of radiological effluent reports with the only restriction being that the interval between reports must not exceed 12 months. The reporting requirements remain as proposed.

8. Comment. Two commenters suggested that the amendments indicate that the changes in reporting requirements of the new regulations take precedence over the existing license technical specifications or license conditions where there may be a conflict.

Response. The proposed amendments are generic and licensees may request administrative amendments to any conflicting license condition or technical specification as needed. 9. Comment. Two commenters suggested that NRC reconsider the need for licensees to submit 10 CFR 50.36a(2) effluent release reports and 10 CFR 50.59 reports concerning annual design changes. The commenters noted the requirement for these reports was issued before the Final Safety Analysis Reports were required to be updated periodically and before resident inspectors were assigned to all reactor sites. The .---------------------------------------------

PART 50

STATEMENTS OF CONSIDERATION commenters also observed that these reports are now available on site for review by inspectors at any time and that most design changes are reflected In the FSARs. Further, the commenters did not believe that these reports are routinely reviewed by the NRC staff. The commenters believed that if the requirements to submit such a report were eliminated, there would be no impact on safety, the required evaluations could continue to be performed, and the reports would continue to be available for review. The commenters believed that the deletion of these requirements would contribute to significant increased savings by licensees.

Response. The consequence of eliminating the requirements for these reports requires significant additional assessment. Thus, the proposed revisions have not been modified in order not to delay the benefit of burden reduction. Although this proposal will not be addressed in the current rulemaking, these suggested revisions will be evaluated as part of an ongoing NRC effort. 10. Comment. One commenter questioned whether the changes in reporting frequency of facility changes under 10 CFR 50.59, FSAR updates, and radiological effluent reports would impair the ability of the NRC to review the information in a timely manner. Response. The resident inspector program along with regional regulatory programs provide timely and in some cases day-by-day review of facility operations. The changes being made will not impair NRC's ability to review the information.

11. Comment. One commenter (Yankee Atomic Electric Co.) stated that the FSAR update changes discussed in Action Item 1 in the proposed rule and In Action Item 7 of this document emanated from a petition for rulemaking that they submitted to the NRC on February 9, 1990 (PRM 50-55). The notice of receipt for this petition was published in the *Federal Register on May 3, 1990 (55 FR 18608). The petitioner originally requested that nuclear power plant licensees be allowed to file FSAR reports at periods greater than annually.

They suggested that § 50.71(e)(4) be revised to read as follows: "Subsequent revisions shall be filed no later than 6 months after completion of each planned refueling outage for a licensee's facility. If two or more facilities share a common FSAR, the licensees shall designate the refueling outage schedule on one of the multiple facilities to establish the schedule for revisions of the common FSAR. The FSAR revisions shall reflect all changes up to a maximum of 6 months prior to the date of filing." . During the comment period on this proposed rule, Yankee Atomic Electric Co. stated that the period between successive FSAR upaates should not be limited to 24 months as proposed. Their. rationale was that the restriction of 24 months was unnecessary. Response. Upon receipt of the Yankee Atomic Electric Co. comment letter of July 20, 1992, the NRC again reviewed the petition (PRM 50-55) submitted by Yankee Atomic Electric Co. and the comments submitted in response to the Notice of Receipt. Based on this review, the NRC believes that the current action being taken to reduce the burden on nuclear licensees is substantially similar to the relief requested in the petition. The 24-month interval for successive FSAR updates is addressed in comment number 6 above. It should be noted that the petition did not contain. a specific reference to a number of months regarding successive FSAR updates. With respect to the petitioner's concern about multiple facilities sharing a common FSAR, licensees will have maximum flexibility for scheduling updates on a case-by-case basis. This final rule does not address multiple facilities. This final rule is considered by the NRC to grant the petition submitted by the Yankee Atomic Electric Co. This final rule constitutes final NRC action on the petition. Discussion The Nuclear Regulatory Commission . is amending 10 CFR parts 20 and 50 to implement the eight proposed actions identified below end also identified in the report on "Special Review of Existing NRC Regulatio.ns" that was completed by the CRGR and that was attached to Chairman Selin'e letter to the White House dated April 27, 1992. These actions will not reduce the protection of the public health and safety or the common defense and security. Each of the eight actions is discussed below. 1. Posting of Rooms Occupied by Diagnostic Nuclear Medicine Patients (10 CFR 20.190.1/h,1) The revision reduces the posting requirements for rooms in hospitals occupied by patients administered radioactive materials who might otherwise be released from confinement under the provisions of 10 CFR 35.75. The estimated savings to licensees is $300,000 for elimination of the need for posting. 50-SC-116

2. Contamination Monitoring of Packages {10 CFR 20.1906{b))

This action clarifies the regulations and reduces the monitoring burden for packages containing radioactive material in the form of a gas or in a special form as defined in 10 CFR 71.4. The estimated savings to licensees is $10.1 million. 3. Frequency of Radiological Effluent Reports (10 CFR 50.36a} This action reduces the requirements for the submission of reports concerning the quantity of principal nuclides released to unrestricted areas in liquid and gaseous effluents from semiannually to annually. The estimated savings for this action, assuming an average remaining plant life of 26 years, is $16,600,000 for licensees and $360,000 for the NRC. 4. Use of Fuel with Zirconium-Based (Other than Zircaloy} Cladding (10 CFR 50.44, 50.46, and Appendix K to Part 50} This action revises the acceptance criteria In 10 CFR 50.44 and 50.46, relating to evah,iations of emergency core cooling systems and combustible gas control applicable to zircaloy clad fuel to include ZIRLO clad fuel. This revision to include ZIRW as an acceptable zirconium based cladding material along with zircaloy will reduce the licensee burden but will not reduce the*protection of the public health or safety. The NRC will address, through an appropriate separate rulemaking, the use of other similar zirconium based cladding materials when all of the necessary safety evaluations for those materials have been completed. The estimated savings for eliminating the need to process recurring exemptions to the regulations to licensees is $2 million end the savings lo the NRC is $50,000. This estimate is based on six plants per year requesting the use of ZIRLO clad fuel over the next 8 years. 5. Receipt Back of Processed Low Level Waste (10 CFR 50.54) This action is addressed in*a separate rulen;iaking. For additional information, see the proposed rule entitled "Receipt of Byproduct and Special Nuclear Material" published in the Federal Register on April 24, 1992 (57 FR 15034). 6. Annual Design Change Reports {10 CFR50.59} This action revises the requirements for the annual submission of reports for facility changes under§ 50.59 (Changes, tests, and experiments) to confm:m with the proposed change for updating the PART 50

STATEMENTS OF CONSIDERATION FSAR (see Item 7). This action does not affect the substance of the evaluation or the documentation requimd for§ 50.59 type changes. It only affec:ts the interval for submission of the info:rmation to the NRC. Instead of submitting the information annually, the information can be submitted on a refueling cycle basis, provided the interval between successive reports does not exceed 24 months. The estimated savings for this action, assuming an average remaining plant life of 26 years, is $1,500,0l)O for licensees and $400,000 for the NRC. 7, Frequency of Final Safety Analysis Report (FSAR) Updates (W CF_R 50.71} This action provides licensees with an option from the current requirements for the annual updating of the Final Safety Analysis Report (FSAR). In lieu of an annual submission, licensues may choose to provide the required information once per each refueling outage. Updates to the FSAR can*be submitted 6 months after each refueling outage, provided the inter\l'al between successive updates to the FSAR does not exceed 24 months. Thi11 action does not affect the substance of FSAR updates. The estimated savings for this action. assuming an average remaining plant life of 26 years. is $11,100,000 for licensees and $910,000 for the NRC. 8. Elimination of Unnecess,Ji"y Event Reports (10 CFR 50.72 and 60.73} This action is addressed in a separate rulemaking.

For additional information, see the proposed rule entitled "Minor Modifications to Nuclear Power Reactor Event Reporting Requirements" published in the Federal Register on June 26, 1992 (57 FR 28642). Environmental Impact: Cattigorical Exclusion The NRC determined that the final regulation is the type of action described in categorical exclusions 10 CFR 51.22(c) (2) and (3). Therefore, neither an environmental impact statement nor an environmental assessment has been prepared for this final regulation.

Paperwork Reduction Act Statement This final rule amends ink,rmation collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Management and Budget approval numbers. 3150-QOH and 3150-0011. The reduction of the publi,c reporting burden for this collection of .information is estimated to average 208 hours per response for operating power reactors and 1 hour per response for certain materials licensees, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.

Send comments regarding this burden reduction or any other aspect of this decrease in the collection of information including suggestions on this reduced burden to the Information and Records Management Branch (MNBB-7714), U.S. Nuclear Regulatory Commission, Washington, DC 20555; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-3019 (3150-0011, 3150-0014), Office of Management and Budget, Washington, DC 20503. Regulatory Analysis The NRC is amending its regulations to reduce the regulatory burden on nuclear licensees. This action reflects an initiative on the part of the NRC and responds to the spirit of President Bush's memoranda of January 28, 1992, which requested that selected Federal agencies review and modify regulations that will reduce unnecessary burden of governmental regulation and ensure that the regulated community is not subject to duplicative or inconsistent regulation. The NRC has identified eight rulemaking actions that would eliminate duplicative or inconsistent regulatory requirements. Six of the actions arc included in this package. Two of the eight actions are being processed as separate rulemakings and are not discussed here. The eight actions are as follows: 1. Posting of Rooms Occupied by Diagnostic Nuclear Medicine to include exceptions for posting requirements for rooms in hospitals for patients administered radiopharmaceuticals for diagnostic tests (10 CFR 20.1903(b)l.

2. Contamination Monitoring of Packages-to eliminate certain provisions for contamination monitoring of packages containing certain types of radioactive material (10 CFR 20.1906(b)).

3. Frequency of Radiological Effluent Reports-to change the frequency of reports on power reactor radiological effluents from twice per year to once per year (10 CFR 50.36a). 4. Use of Fuel with Zirconium-Based Cladding-to eliminate the need to obtain exemptions in order to use certain fuel cladding material not presently addressed in the regulations (10 CFR 50.44, 10 CFR 50.46 and 10 CFR part 50, appendix K). 5. Receipt Back of Processed Low Level Waste-separate rulemaking (10 CFR 50.54). 50-SC-117

6. Annual Design Change Reports-to change the frequency of reporting changes at power reactors from once per year to once per refueling cycle (10 CFR 50.59(b)).

7. Frequency of Final Safety Analysis Report Updates-to change the frequency of safety analysis report updatesJrom once per year to once per refueling cycle (10 CFR 50.71 ). 8. Elimination of unnecessary event reports-separate rulemaking (10 CFR 50.72 and 50.73). Each of these actions considers the elimination or relaxation of regulatory requirements currently imposed on NRC licensees.

Action Items 1 and 2 would affect material licensees while Action Items 3 through 8 would affect power reactor licensees. For each regulatory action, the NRC has evaluated the health and safety implications and the cost impacts relative to a status quo alternative. The NRC finds that each would result in a reduction in burden without reducing protection of the public health and safety. The public health and safety determination appears in a document entitled "Report on Special Review of Existing NRC Regulations by the Committee to Review Generic Requirements" issued on April 13, 1992. Additionally, an analysis of the safety implications of Action Item 4 is available in a U.S. NRC letter to Westinghouse Corporation dated July 1. 1991, entitled "Acceptance For Referencing of Topical Report WCAP-12610 "Vantage+Fuel Assembly Reference Core Report" (TAC NO. 77258)." The cost savings to both the licensee population and the NRC appear below. Dollar impacts are expressed on a 1992 present worth basis in 1992 dollars. The basis for these cost estimates is available in a report entitled "Analyses of Potential Cost Savings for Selected NRC Reforms" dated June 10, 1992. TOTAL DISCOUNTED I COST SAVINGS AS* SOCIATED WITH PROPOSED TORY REVISIONS [In millions of 1992 dollars] Regulatory revision Licensees Item 1................................... 0.3 llem 2................................... 10.1 Item 3 ................................... 16.8 Item 4 ................................... 2.0 Item 5 ................................... 0 N/ A llem 6................................... 1.5 llem 7................................... 11.1 llem B ...................................

NI A NRC 2-0.100 2-0.100 0.360 0.050
NIA 0.400 0.910 *NIA.
Assumes en annual real discount rate of 5% ' Negative cost savings represent a cost ture.
Not applicable-separate rulemeking.

PART 50

STATEMENTS OF CONSIDERATION The NRC concludes that each of these proposed regulatory revisions is justified due to the net cost savings that will accrue without reducing public health and safety. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act, 5 U.S.C. 605(b). the Commission certifies that, this rule will not have a significant adverse economic impact on a substantial number of small entities.

The NRC has adopted size standards that classify a small entity as a small business or organization, one whose gross annual receipts do not exceed $3.5 million, or as a small governmental jurisdiction whose supporting population is 50,000 or less. The first two issues involve the relaxation of requirements which will affect approximately 5,000 material licensees. Although many of these licensees may be small entities, there should be no adverse impact on these small licensees because the regulations are being relaxed. The remaining six issues affect 112 power reactor licensees. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the NRC Size Standards. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this final rule and, therefore, that a backfit anaiysis is not required because these amendments do not involve any provisions that would impose backfits as defined in 10 CFR 50.109(a)[1). List of Subjects 10 CFR Part 20 Byproduct material, Criminal penalty, Licensed material. Nuclear materials, Nuclear power plants and reactors, Occupational safety and health, Packaging and containers, Radiation protection, Reporting and rccordkeeping requirements, Source material, Special nuclear material, Waste treatment and disposal. 10 CFR Part 50 Antitrust, Classified information, Criminal penalty, Fire protection, _Incorporation by reference, Intergovernmental rel a lions, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the . NRC is adopting the following amendments to 10 CFR parts 20 and 50. 57 FR41378 Published 9/10/92 Effective 10/13/92 10 CFR Part 50 RIN 3150-AE12 Minor Modifications to Nuclear Power Reactor Event Reporting Requirements AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) has amended its regulations to make minor modifications to the current nuclear power reactor event reporting requirements.

The final rule applies to all nuclear power reactor licensees and deletes reporting requirements for some events that have been determined to be of little or no safety significance. The final rule reduces the industry's reporting burden and the NRC's response burden in event review and assessment. EFFECTIVE DATE: October 13, 1992. FOR FURTHER INFORMATION CONTACT: Raji Tripathi, Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone (301) 492-4435. SUPPLEMENTARY INFORMATION: Background The Commission is issuing a final rule that amends the nuclear power reactor event reporting requirements contained in 10 CFR 50.72, "Immediate Notification Requirements for Operating Nuclear Power Reactors," and 10 CFR 50.73, "Licensee Event Report System." The final rule is issued as part of the Commission's ongoing activities to improve its regulations. Specifically, this final rule amends 10 CFR 50.72 (b)(2)(ii) and 10 CFR 50.73 (a)(2)(iv). On June 26, 1992 (57 FR 28642), the Commission issued a proposed rule requesting public comments on these amendments. Over the past several years, the NRC has increased its attention to event reporting issues to ensure uniformity, consistency, and completeness in reporting. In September 1991, th~ NRC's Office for Analysis and Evaluation of Operational Data (AEOD) issued for comment a draft NUREG-1022, Revision 1, 1 "Event Reporting Systems 10 CFR

Free single copy may be requested by writing to the Distribution and Mail Services Section, U.S. 50-SC-118 50.72 and 10 CFR 50.73-Clarification of NRC Systems and Guidelines For Reporting." Following resolution of public comments, the NUREG will be issued in the final form. The NUREG will contain improved guidance for event reporting.

NRC's reviews of operating experience and the patterns of licensees' reporting of operating events since 1984 have indicated that reports on some of these events are not necessary for the NRC to perform its safety mission and that continued reporting of these events would not contribute useful information to the operating reactor events database. Additionally, these unnecessary reports would have continued to consume both the licensees' and the NRC's resources that could be better applied elsewhere. The NRC has determined that certain types of events, primarily those involving invalid engineered safety feature (ESF) actuations, are of little or no safety significance. Valid ESF actuations are those actuations that result from "valid signals" or from intentional manual initiation, unless it is part of a preplanned test. Valid signals are those signals that are initiated in response to actual plant conditions or parameters satisfying the requirements for ESF_ initiation. Invalid actuations are by definition those that do not meet the criteria for being valid. Thus, invalid actuations include actuations that are not the result of valid signals and are not intentional manual actuations. Invalid actuations include instances where instrument drift, spurious signals, human error, or other invalid signals caused actuation of the ESF (e.g., jarring a cabinet, an error in use of jumpers of lifted leads, an error in actuation of switches or controls, equipment failure, or radio frequency interference). NRC's evaluation of both the reported events since January 1984, when the existing rules first became effective, and the comments received during the Event Reporting Workshops conducted in Fall of 1990 identified needed improvements in the rules. The NRC determined that invalid actuation, isolation, or realignment of a limited set of ESFs including the systems, subsystems, or components [i.e., an invalid actuation, isolation, or realignment of only the reactor water clean-up [RWCU) system, Nuclear Regulatory Commission, Washington, DC 20555. A copy is also available for inspection or copying for a fee at the NRC Public Document Room, 2120 L Street, NW., (Lower Level), Washington, DC 20555. ~--r PART 50

STATEMENTS OF CONSIDERATION the control room emergency ventilation (CREV) system, the reactor building ventilation system, the fuel building ventilation system, or thE: auxiliary building ventilation system, or their equivalent ventilation sy:,tems]

are of little or no safety signific.mce. However, these events are currently reportable

under 10 CFR 50.72 (b)(2)[ii) and 10 CFR 50.73 (a](2](iv).

The final rules for the current event reporting regulations, 10 CFR 50.72 and 10 CFR 50.73 (48 FR 3903E1; August 29, 1983, and 48 FR 33850; July 26, 1983, respectively), stated that ESF systems, including the reactor protection system (RPS), are provided to mitigate the consequences of a significant event. Therefore, ESFs should (1) work properly when called upon and (2) should not be challenged frequently or unnecessarily. The Statements of Consideration for these final rules also stated that operation of an ESF as part of a pre-planned operational procedure or test need not be reported. The Commission noted that EBF actuations, including reactor trips, are frequently associated with significant plant transients and are indicative of events, that are of safety significance. At that time, the Commission alsc, required all ESF actuations, including the RPS actuations, whether manual or automatic, valid or invalid-except as noted, to be reported to the NRC by telephone within 4 hours c,f occurrence followed by a written Licensee Event Report (LER) within 30 days of the incident. This requirement on timeliness of reporting remains unchanged. The reported information is used by the NRC in confirmation of the licensing bases, identification of precursors to severe core damage, identi:fication of plant specific deficiencies, generic lessons, review of management control systems, and licensee performance assessment. Discussion The NRC has determined that some events that involve only invalid ESF actuations are of little or no safety significance. However, not all invalid ESF actuations are being exempted from reporting through this rule. The relaxations in event reportiing requirements contained in the final rule apply only to a narrow, limited set of specifically defined invalid ESF actuations. These events include invalid actuation, isolation, or realignment of a limited set of ESFs including systems, subsystems, or component!: (i.e., an invalid actuation, isolation, or realignment of only the R\IVCU system, or the CREV system, reactor building ventilation system, fuel building ventilation system, auxiliary building ventilation system, or their equivalent ventilation systems). The actuation of the standby gas treatment system following an invalid actuation of the reactor building v~tilation system is also exempted from reporting. In addition, the final rule excludes invalid actuations of these ESFs (or their equivalent systems) from signals that originated from non-ESF circuitry. However, invalid actuations of other ESFs would continue to be reportable. For example, emergency core cooling system isolations/ actuations; containment isolation valve closures that affect cooling systems, main steam flow, essential support systems, etc.; containment spray actuation; and residual heat removal system isolations (or systems designated by any other names but designed to fulfill the function similar to these systems and their equivalents), are still reportable. If an invalid ESF actuation reveals a defect in the system so that the system failed or would fail to perform its intended function, the event continues to be reportable under other requirements of 10 CFR 50.72 and 10 CFR 50.73. If a condition or deficiency has (1) an adverse impact on safety-related equipment and consequently on the ability to shut down the reactor and maintain it in a safe shutdown condition, (2) has a potential for significant radiological release or potential exposure to plant personnel or the general public, or (3) would compromise control room habitability, the event/ discovery continues to be reportable. Invalid ESF actuations that are excluded by this final rule, but occur as a part of a reportable event, continue to be described as part of the reportable event. These amendments are not intended to preclude submittal of a complete, accurate, and thorough description of an event that is otherwise reportable under 10 CFR 50.72 or 10 CFR 50.73. The Commission relaxed only the selected event reporting requirements specified in this final rule. Licensees are still required under 10 CFR part 50, appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to address corrective actions for events or conditions that are adverse to quality whether the event is reportable or not. In addition, minimizing ESF actuations (such as RWCU isolations) to reduce operational radiation exposures associated with the investigation and recovery from the actuations, are consistent with ALARA requirements. 50-SC-119 This rule excludes three categories of events from reporting: (1) The first category excludes events in which ah invalid ESF or RPS actuation occurs when the system is already properly removed from service if all requirements of plant procedures for removing equipment from service have been met. This includes required clearance documentation, equipment and control board tagging, and properly positioned valves and power supply breakers. (2) The second category excludes events in which an invalid ESF or RPS actuation occurs after the safety function has already been completed (e.g., an invalid containment isolation signal while the containment isolation valves are already closed, or an invalid actuation of the RPS when all rods are full inserted). (3) The third category excludes events in which an invalid ESF actuation occurs that involes only a limited set of ESFs [i.e., when an invalid actuation, isolation, or realignment of only the RWCU system, or any of the following ventilation systems: CREV system, reactor building ventilation system, fuel building ventilation system, auxiliary building ventilation system, or their equivalent ventilation systems, occurs]. Invalid actuations that involve other ESFs not specifically excluded, (e.g., emergency core cooling system isolations or actuations; containment isolation valve closures that affect cooling systems, main steam flow, essential support systems, etc.; containment spray actuation; residual heat removal system isolations, or their equivalent systems), continue to be reportable. Licensees continue to be required to submit LERs if a deficiency or condition associated with any of the invalid ESF actuations of the RWCU or the CREV systems (or other equivalent ventilation systems) satisfies any reportability criteria under § 50.72 and § 50.73. Impact of the Amendments on the Industry and Government Resources Relaxing the requirement for reporting of certain types of ESF actuations reduces the industry's reporting burden and the NRC's response burden. This reduction is consistent with the objectives and the requirements of the Paperwork Reduction Act. These amendments have no impact on the NRC's ability to fulfill its mission to ensure public health and safety because the deleted reportability requirements have little or no safety significance. It is estimated that the changes to the existing rules will result in about 150 (or PART 50

STATEMENTS OF CONSIDERATION 5-10 percent) fewer Licensee Event Reports each year. Similar reductions are expected in the number of prompt event notifications reportable under 10 CFR 50.72. Some respondents, in their comments on the proposed rule, dated June 26, 1992, submitted an estimate of approximately 15 percent reduction in their reporting burden. Summary of Comments The NRC received 19 comments-2 from individuals, 3 from supported organizations, and 14 from utilities.

Except for two respondents, all commenters welcomed the Commission's efforts to reduce the licensee burden and to save the agency's resources in event review and processing. The utilities and the industry-supported organizations expressed their desire for a broader relaxation to include all invalid ESF actuations from reporting. Other comments from the respondents concerned the following: clarification of the definition of "invalid" actuations; examples of events being exempted from reporting; consideration of specific situations; exemption from reporting of the actuation of the standby gas treatment system following an invalid actuation of the reactor building ventilation system; and possibly extending relaxation of invalid actuations/isolations of RWCU from reporting to include those of the chemical and volume control system in a pressurized water reactor. The Statement of Considerations for this final rule addresses most of these concerns. Other issues and clarifications concerning event reportability will be addressed in NUREG-1022, Revision 1. However, it is not practical to address a plant-specific situation unless it relates to a generic concern. The Commission stresses that only certain specific invalid ESF actuations are being exempted from reporting through the present amendments. NUREG-1022, Revision 1 will contain specific examples and additional guidance on events which are presently reportable as well as those which are being exempted from reporting through these amendments. In the future, the Commission will give due consideration to other proposed relaxations from event reporting after the NRC staff has had an opportunity to reassess the data needs of the agency and performed safety assessments to justify initiating a separate general rulemaking. Until such time, all events not specifically exempted in these amendments continue to be reportable. The two respondents who opposed the proposed amendments expressed their concerns about eliminating the selected event reporting requirements. These commenters believe that the elimination of these event reporting requirements may adversely affect the NRC's information database and ultimately affect the agency's ability to carry out its mission to protect public health and safety. For many years, the NRC staff has been systematically reviewing information obtained from Licensee Event Reports. These assessments of reactor operational experience have included data on the types of events included in the three categories that the NRC is deleting from reporting. The staffs reviews and assessments of nearly 1000 years of operational experience have identified essentially no safety significance associated with the type of events included in the aforementioned three categories. The Commission has reviewed the scope of these amendments, and on the basis of the staffs assessment of the past reactor operational experience, has subsequently concluded with a reasonable confidence that relaxation from reporting of events in the three categories does not affect the agency's ability to protect public health and safety. Based on the input from the utilities, these amendments will reduce the industry's reporting burden by about 15 percent. The estimated savings of the NRC's response burden in event review and assessment is about 5-10 percent. Environmental Impact: Categorical Exclusion The NRC has determined that this final rule is the type of action described in categorical exclusions 10 CFR 51.22 (c)(3)(ii) and (iii). Therefore, neither an environmental impact statement nor an environmental assessment has been prepared for this final rule. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These amendments were approved by the Office of Management and Budget approval numbers 3150-0011 and 3150-0104. Because the rule will relax existing reporting requirements, public reporting burden of information is expected to be reduced. It is estimated that about 150 fewer Licensee Event Reports (NRC Form 366) and a similarly reduced number of prompt event notifications, made pursuant to 10 CFR 50.72, will be required each year. The resulting reduction in burden is estimated to 50-SC-120 average 50 hours per licensee response, including the time required reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and reviewing the collection of information. Send comments regarding the estimated burden reduction or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB--7714), U.S. Nuclear Regulatory Commission, Washington, DC 20555; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB--3019, (3150-0011 and 3150-0104), Office of Management and Budget, Washington, DC 20503. Regulatory Analysis The Commission has prepared a regulatory analysis on this final rule. The analysis examines the costs and benefits of the alternatives considered by the Commission. The analysis is available for inspection in the NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555. Single copies of the analysis may be obtained from: Raji Tripathi, Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone (301) 492-4435. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 605 (Bl), the Commission certifies that this rule does not have a significant economic impact on a substantial number of small entities. The final rule affects only the event reporting requirements for operational nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration Act in 13 CFR part 121. Backfit Analysis As required by 10 CFR 50.109, the Commission has completed an assessment of the need for Backfit Analysis for this final rule. The proposed amendments include relaxations of certain existing requirements on reporting of information to the NRC. These changes neither impose additional reporting requirements nor require modifications to the facilities or their licenses. Accordingly, the NRC has concluded that this final rule does not constitute a PART 50

STATEMENTS OF CONSIDERATION backfit and, thus, a backfit analysis is not required.

List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminal penalty, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation* protection, Reactor siting 1;riteria, Reporting and recordkeep;ing. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1964, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 1552 and 553, the Commission is adopting the following amendments to 10 CFR part 50. 57 FR 47978 Published 10/21/92 Effective 11/20/92 10 CFR Part 50 RIN 3150-AED4 Receipt ;Of Byproduct and Special Nuclear Material AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission is amending its regulations governing the condition of licenses for production and utilization facilities to allow a reactor licensee to receive back byproduct and special nuclear material that is produced by operating the reactor after that material has been sent off-site for processing, such as compaction or incineration.

EFFECTIVE DATE: November 20, 1992. FOR FURTHER INFORMATION CONTACT: LeMoine J. Cunningham, telephone (301) 504-1086, or Paul H. Lohaus on (301) 504-2553. U.S. Nuclear Regulatory Commission, Washington, DC 20555. SUPPLEMENTARY INFORMATION: Contents I. Background. II. Response to Public Comments on the Proposed Rule. I. Background On April 24, 1992 (57 FR 15034), the Commission published a proposed rule that would amend its regulations in 10 CFR 50.54, "Conditions of licenses." This addition to the regulations is needed primarily because of changing circumstances surrounding the treatment, storage, and disposal of level radioactive waste (LLW) generated by operating nuclear power reactors. At the time when most operating licens\!s were issued, the Commission expected that LLW would be promptly treated and sent off-site for disposal in a licensed LLW disposal facility. Therefore, licensees were not authorized to receive byproduct or special nuclear material except in the form of sealed sources for analysis, calibration, or other special purposes, in the form of fuel for use in the reactor, or associated with radioactive apparatus or components. Except for LLW generated in Michigan, where generators have been denied access to LLW disposal capacity since November 10, 1990, companies providing nuclear power reactors with off-site LLW processing and reduction services currently transfer treated waste directly to one of three operating commercial LLW disposal 50-SC-121 facilities. These companies providing off-site treatment and volume-reduction services may have several reasons for needing to return treated LLW to the generator, rather than shipping it to a disposal site. First, access to LLW disposal facilities may be restricted for the generator whose waste has been treated. Under the Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA), States with operating disposal facilities may deny access to generators in other States. Second, the licensee offering off-site services may not have adequate capacity for storing the waste until disposal. Finally, the legal relationships among States and regions, established under the various compacts ratified by Congress, in conjunction with the LLRWPAA, may force return of treated LLW to the generator in order to ensure that the waste is disposed of at the appropriate disposal facility. Accordingly, although a reactor licensee may send its LLW off-site to another licensee for treatment (e.g, compaction, or incineration), the licensee treating the waste will have a need to return the waste to the generator, but may not do so because the generator lacks authority to receive it. The Commission identified two principal options for addressing this issue. The first option was to use a by-case licensing-action approach, either by amending each facility license, or by issuing a separate license under 10 CFR part 30. This action would require each facility licensee either to request an amendment to its operating license, authorizing the receipt of processed LLW, or to request a separate license. However, addressing the issue for each licensee individually would be inefficient, requiring both the licensee and NRC to expend significant resources. The second option, adopted by the-* Commission as the final rule, amends 10 CFR 50.54, "Conditions of licenses," to allow reactor licensees to receive back LLW generated at the plant and shipped off-site for processing. This approach not only resolves the authorization issue, but also eliminates the need for significant NRC and licensee efforts to complete and approve the amendment requests to part 50 licenses, and ensures that a uniform approach is applied in all cases. The Commission considers the final rule a minor amendment that does not represent any change in Commission regulatory policy regarding radioactive waste. On October 16, 1981, the Commission published its policy statement on LLW volume-reduction (46 PART 50

STATEMENTS OF CONSIDERATION FR 51100), in which it called upon all generators of radioactive waste to reduce the volume of LLW for disposal, to extend the life of disposal sites and alleviate storage concerns.

The final rule will further enhance licensees' options to reduce the volume of waste, by using services performed off-site and permitting the return of the treated waste to the generator. The Commission anticipates that many reactor licensees will take steps to process or reduce the volume of generated LLW, typically by off-site compaction and incineration, before storing the waste at their facilities, on an interim basis. However, the Commission does not look favorably on long-term on-site storage. The final rule is intended to ensure that licensees will have adequate short-term on-site storage capacity for self-generated LLW, until permanent disposal capacity is available. The commission does not believe this minor amendment represents a change in the stated Commission position that it "* *

does not look favorably on term on-site storage." The Commission expects licensees to ship generated wastes for disposal to the extent possible.

Storage of LLW should be used only for the short-term management of LLW, when disposal is interrupted or unavailable. The final rule applies to both power and non-power reactor licensees. The final rule does not authorize the receipt of any material recovered from the reprocessing of irradiated fuel. In addition to publishing the proposed rule in the April 24, 1992, Federal Register NRC sent a copy of the proposed rule to all Agreement and Agreement State radiation control program directors, all State liaison officers, and those organizations on the low-level waste compact distribution list, on May 26, 1992. The comment period ended on July 8, 1992. Various comments were received, all of which were evaluated in developing the final rule. II. Response to Public Comments on the Proposed Rule The Commission received a total of 31 comment letters on the proposed rule. Responses were received from utilities or their counsel (19); nuclear power and nuclear material user-groups (2); State departments of health and radiation protection agencies (2); public interest groups (2); disposal facility operators and developers (2); and private citizens (4). Copies of these letters are available for public inspection and copying for a fee at the NRC Public Document Room at 2120 L Street NW. (Lower Level), Washington, DC. Of the 31 letters received, 26 endorsed adoption of the proposed rule. Many of these commenters cited the benefits from off-site waste processing technology that alloi,vs operating reactor licensees to maximize current storage capacity and minimize radioactive waste volumes. They emphasized that implementation of the rule is an efficient and cost-effective solution to a practical problem, and several favorably cited the rule's timeliness and potential benefits in light of future uncertainties concerning LLW disposal capacity. One commenter, after endorsing the proposed rule, encouraged NRC to become more involved in the licensing process for new disposal sites. Both Illinois and Arkansas, the only states to provide comments concerning the proposed rule, endorsed the rule. Approximately 20 specific questions or suggestions were received that address the proposed rule. The majority of comments received may be grouped into one of four basic categories that include: (1) Clarification and enhancement of the proposed rule, (2) Waste accountability at the processor licensee, (3) Radioactive waste transportation concerns, and (4) Worker health and safety. Public comments and NRC responses follow: Comment: Six commenters requested that the rule be clarified to authorize reactor licensees to receive back processed waste originally generated by any reactor or reactors located at the same site. Commenters pointed out that some plants do not distinguish one unit's waste from an adjacent unit's waste when shipping off-site for treatment, and that oftentimes, wastes generated at a particular site with multiple operating reactors will be commingled. They argued that the proposed rule, if strictly interpreted, would not allow a reactor licensee to receive back processed waste containing waste from another reactor located at the same site. Three commenters explicitly requested that NRC change the proposed rule to permit wastes generated at a single site to be consolidated. Two of the three, Edison Electric Institute and the law firm of Winston & Strawn, provided NRC with specific language for such clarification. ' Three other commenters concurred with this recommendation by citing Edison Electric Institute's submittal to NRC concerning the proposed rule. Response: The Commission agrees that the rule should permit radioactive waste from multiple units of one licensee at a particular site to be 50-SC-122 received back under the license of any of the units at that site. The Commission has revised the final rule to reflect this change. The following language is added to modify § 50.54(ee)(1) of the final rule: "Each license issued under this part authorizing the possession of byproduct and special nuclear material produced in the operation of the licensed reactor includes, whether stated in the license or not, the authorization to receive back that same material, in the same or altered form or combined with byproduct or special nuclear material produced iri the operation of another reactor of the same licensee located at that site * * *" Comment: One commenter suggested that the proposed rule be modified to authorize the receipt back of byproduct or special nuclear material from a licensed entity that is authorized to possess the radioactive material, but is not "a licensee of the Commission or an Agreement State," as the proposed rule had originally stipulated. The commenter suggested that a common or contract carrier transporting source or byproduct radioactive material may not be able to return such material to the reactor licensee generating the material, although this may be necessary in several situations, such as the return of waste because of ineffectual waste packaging. Similarly, the commenter contended, a non-licensed government agency, such as the Department of Energy, may be unable to return treated waste to a reactor licensee if a strict interpretation of the proposed rule were adopted. Response: The Commission agrees that the rule should permit receipt back, by a reactor licensee, of byproduct and special nuclear material produced by the reactor licensee from a non-licensed entity that is authorized to possess the material. The Commission has revised the final rule to reflect this enhancement. In § 50.54(ee)(1) of the final rule, authority has been granted to power and non-power reactor licensees to receive back byproduct and special nuclear material produced in the operation of the reactor "from a licensed entity authorized to possess the material," as well as from Commission or Agreement State licensees. Comment: One commenter suggested that NRC modify the proposed rule to authorize the transfer of byproduct or special nuclear material for reduction or decontamination purposes . among reactor sites with a common licensee. The commenter stated that "* *

under the Commission's restrictive interpret\ltion of the scope of reactor operating licensees, a reactor PART 50
STATEMENTS OF CONSIDERATION licensee may not receive LLW for processing that was produced in the operation of a reactor for which it also has licensee responsibilities." Response:

The Commis11ion believes authorization permitting a reactor licensee to receive byproduct or special nuclear material, at one reactor site, that is produced at another reactor site for which it also has licensee responsibilities for the purpose of performing decontamination or reduction services, is beyond the scope of this rulemaking. The Commission does not agree that the proposed rule should be modified to allow transfer of byproduct and special nudear material among reactor sites for thfo purpose. As a matter of policy, the Commission opposes practices at reactor facilities that may divert the attention of licensee management from the primary task of safe operation of the powe1: reactor. Accordingly, the Commission believes that such situations should continue to be reviewed and authorized on a specific basis. Comment: One commenter suggested that NRC clarify that, if the proposed rule is adopted, the final rul.e will also authorize the transfer of decontamination equipment that is slightly contaminated with '.byproduct or special nuclear material among reactor sites with a common license. Response: The Commission believes this suggestion is beyond the scope and intent of the rulemaking. Accordingly, the Commission does not agree that the proposed rule should addrens this issue. However, part 50 licenses typically contain conditions that permit transfer of decontamination equipmeint among reactor sites with a common license. Licenses which permit the licensee to "* *

receive, possess * *
any byproduct, source or special nuclear material * *
associated with radioactive apparatus or components," may authorize the receipt of transferred decontamination equipment.

Comment: Three commenters expressed concern that the mle will result in an increase in the m1mber of miles traveled by radioactive waste on our nation's highways, and that this is not in the public interest. One commenter suggested that po,tential hazards are greater from the transportation of radioactive waste on the return leg from a processor to a generator, because the radioactive material within the waste han been concentrated. Response: The proposed rule may result in an increase in the number of miles radioactive waste is transported. The additional dose associatl!d with this transportation, however, represents only a small increase in doses [that are already small) from transportation activities. Further, all waste shipments must meet the applicable regulatory requirements of the U.S. Department of Transportation ancf the U.S. Nuclear Regulatory Commission. The regulations require the packaging to be commensurate with the potential hazard of the contents. Comment: Two commenters suggest that off-site processing creates health risks to "additional attending personnel." One commenter, therefore, advocates on-site processing of waste. This commenter contends that on-site processing minimizes the number of people exposed to the hazards of radiation; additional handling, storage, and transportation will result in higher exposure to personnel and greater risks of harmful effects to the public. Response: All processing, storage, and transportation of LLW must meet regulatory requirements. This minor rule change, which authorizes receipt of waste, does not change any of the requirements concerning waste processing, storage, and transportation. The collective occupational exposure [dose) would be essentially the same for the waste processing whether the processing were done on-site or off-site, assuming that the same process were used in either location. Additional handling for shipping and receiving of wastes sent off-site for processing is needed; however, this incremental dose from this activity would be a small fraction of the dose for waste processing and other shipping and receiving activities. The corresponding doses to members of the public would also represent a very small incremental increase in doses that already are very small. Comment: Several comments were received about accountability of waste at the waste processor, and subsequent changes to the waste as a result of waste processing activities. Specifically, two commenters suggested that the possible intermingling of wastes at the processor facility would make difficult the task of ensuring that waste received from a particular generator is returned to that generator alone, as required by the rule. Another commenter expressed concern that accompanying waste manifests may become inaccurate as a result of changes to the waste by the processor licensee. Response: An individual reactor licensee's decision to ship waste off-site for processing, with the intention of receiving back such LLW for temporary storage, will require the reactor licensee to coordinate with the waste processor, to ensure that waste shipped back to the 50-SC-123 facility fulfills the criteria of the new rule and any other applicable regulations. Processors who choose to accept reactor waste intended for return back to the reactor site licensee currently satisfy a host of substantive requirements governing transfer and recordkeeping of radioactive waste cited in 10 CFR 20.311, "Transfer for disposal and manifests," or appendix F to new § § 20.1001 through 20.2401, "Requirements For Low-Level-Waste Transfer for Disposal at Land Disposal Facilities and Manifests." These rules require that manifests accompanying radioactive waste shipments to licensed waste processors and land disposal facilities. The manifest must indicate the identification of the waste generator, the physical description of the waste, waste volume, waste radionuclide identity and quantity, total radioactivity, and the principal chemical form of the waste. In addition to .the manifesting requirements, licensed waste processors who treat or repackage waste must also fulfill waste classification, identification and labeling requirements found in 10 CFR 61.55, 61.56, and 61.57. The final rule exclusively authorizes reactor licensees to receive back LLW sent off-site for treatment. The final rule does not allow a reactor site to accept any waste that is not originally generated at the*site, and the processor must fashion its operations to comply with this condition. The individual processor licensee, when receiving waste intended for return to the reactor site, may have to perform individual "batch" processing, for the reactor licensee to accept processed waste in compliance with the rule. Comment: One commenter pointed out that waste processing may result in changes to waste classification, and in fact, may concentrate radioactivity enough to approach or exceed than-class-C waste concentrations. This commenter was concerned that these potential shifts in waste classification might go unreviewed. This commenter also expressed concern that waste processing operations may produce waste products containing mixed waste and may result in the release of gases or particulates into the atmosphere. Waste processing operations may increase the toxicity or concentration of the waste, this commenter argued, suggesting that this is an undesirable outcome from the perspective of the Host State responsible for ultimate disposal of the treated waste. Response: If a reactor licensee intends to receive back LLW shipped off-site from its facilities and is to comply with the rule, the processor licensee will have PART 50

STATEMENTS OF CONSIDERATION to segregate wastes by individual reactor licensee.

Excluding this segregation operation, the rule will not affect waste processor licensee generated waste product or operations. The final rule will not result in changes to waste product currently generated by waste processors. The final rule may lead processors to treat some generated wastes in individual "batches," to allow return of the waste back to the reactor licensee, but otherwise, processor licensees will simply continue performing LLW volume-reduction activities, as they have before promulgation of the final rule. NRC and Agreement State regulations are in effect which authorize treatment and handling operations at waste processor licensee facilities and ensure that these operations are conducted safely .and without adverse effects on the environment. These conditions are not affected by the final rule, and will continue in force after promulgation of the final rule. Comment: One commenter expressed concern that it was unclear what authority had responsibility to oversee waste processing and enforcement at the processor. This same commenter questioned whether waste ownership or title to the waste may, in fact, become obscured as a result of waste being shipped and managed by several licensees. Response: NRC will continue to license and inspect processor licensees in non-Agreement States, and the Agreement State authorities will continue to license and inspect their licensees. Tracking and manifest requirements will continue to apply to reactor-generated wastes, and title to the waste and waste ownership can be adequately communicated and documented between reactor licensees and processor licensees. Before the final rule, waste processors have received waste from reactor licensees, have processed the waste to reduce its volume, and have repackaged and shipped for disposal the final, reduced waste product. The transfer of waste from generator to processor and/ or broker, and from processor and/ or broker to a disposal facility licensee, has not resulted in a significant number of disputes concerning transfer of title or possession of radioactive wastes, among licensees. The contractual arrangements between licensees and the laws of the various States pertaining to transfer of ownership continue to provide licensees with the means by which they can negotiate the transfer of title to the waste. The Commission does not envision any difficulties concerning radioactive waste title transfer originating from promulgation of the final rule. Comment: One commenter asked whether the languag_e in the proposed rule prohibiting receipt back of material "recovered from the reprocessing of irradiated fuel" could be applied to LLW containing fuel fines from damaged fuel rods. The commenter questioned whether NRC prohibits off-site treatment of such wastes, and whether the generator of such wastes may refuse to accept back such wastes once processed. The commenter asked whether the term "reprocessing of irradiated fuel" is "narrowly" interpreted by the NRC to refer only to the reprocessing of spent fuel rods. Response: The proposed rule contains no new authorizations for, or prohibitions against, LLW processing. The term "material recovered from the reprocessing of irradiated fuel" does refer to the reprocessing of "spent fuel rods" and does not apply to LLW containing fuel fines from damaged fuel rods. NRC does not prohibit off-site treatment of LLW that may contain very small quantities of fuel fines from damaged fuel rods. The rule authorizes, but does not require, the receipt of waste. Reactor licensees are not authorized to reprocess irradiated fuel or to possess the wastes from such reprocessing. (Wastes from fuel reprocessing are, by regulatory definition, high-level wastes, not LLW.) The intent of the proposed rule is to allow reactor licensees to receive the radioactive materials that they produce and that they already are allowed to possess. The sentence in question, concerning fuel reprocessing, was added to the proposed rule to make it clear that reactor licensees are not authorized to receive materials that they are not already authorized to possess (reprocessing wastes). Comment: One commenter expressed concern that by allowing all reactor licensees to receive back waste after processing, many States will simply require these licensees to store all other LLW generated within the State. Response: The Commission does not believe this issue is germane to the final rule. The final rule affects license conditions allowing receipt of radioactive material, but does not alter conditions concerning storage of radioactive waste. The rule addresses the receipt back of LLW generated only at reactor sites, and shipped off-site for processing. The rule does not authorize the storage of LLW, generated throughout the State, at reactor sites. 50-SC-124 Comment: One commenter inquired as to the actions to be taken if a generator refuses to accept waste back after processing. Response: NRC does not consider this a likely scenario. Both the generator and the processor involved in the transfer of waste for treatment will likely have agreed, by contract, on the terms of treatment and transfer of the LLW. If the waste, on return to the generator, is not accepted by the generator, the processor licensee would have grounds to seek legal recourse to force the generator to take possession of the treated LLW. However, if a threat to the public health and safety were to present itself at any time as a result of a reactor licensee's refusal to accept waste from a processor shipped off-site for processing, NRC would use its authority to compel the appropriate party to take possession of the waste, and store it safely. Comment: One commenter suggested that a loophole in the Low-Level Radioactive Waste Policy Act of 1980 allows a licensee to forward a shipment of radioactive material to another State to be stored or treated, and then avoid all responsibility for disposal of the material by declaring the material a waste. The 1980 Act places the receiving State in an untenable position by requiring it to provide disposal capacity for wastes its licensees become burdened with in this manner. Response: The 1980 Act encourages States to form regional compacts to collectively provide for disposal capacity ofLLW. These compacts, authorized by Congress, were allowed to exclude waste generated outside their borders, beginning January l, 1986. This date was later extended to January 1, 1993, when Congress approved the Level Radioactive Waste Policy Amendments Act of 1985. The 1985 Act further authorizes that each State shall provide disposal capacity for LLW "* *

generated within * * *"the State. Accordingly, generators who ship radioactive material out-of-state for processing, and then declare that material to be LLW, will likely be unable to shift responsibility for the disposal of their waste. Therefore, the Commission does not consider this to be a significant issue, nor one which is affected, in any way, by the final rule. Comment: One commenter suggested that the rule clarify considerations for non-reactor licensees concerning receipt of waste back at their facilities after processing.

Response: The Commission has drafted the final rule to apply to reactor licensees, only. Reactor licensees have reported that processor licensees are PART 50

STATEMENTS OF CONSIDERATION unwilling to accept reactor-generated waste without some assurance that reactor licensees will be authorized to receive back processed LLW initially generated at the reactor facility.

Currently, no problems have been identified concerning parts 30, 40, and 70 licensees and the return of processed LLW to their facilities. Tbe Commission can address this issue in the future if parts 30, 40, and 70 licensees encounter problems in this area. Comment: One commenter noted that currently available on-sit,~ reduction technology is more effective than the "double*-handling" mandated by off-site processing. Response: This rule does not mandate off-site processing. The licensee is free to evaluate the cost effectiveness of a given technology or proce.ss and choose either on-site or off-site prncessing. Comment: Two commeuters suggested that the rule be decided in conjunction with the proposed rule on export and import of radioactive waste. Commenters were concerned that wastes may be imported for disposal from NRC licensees operal!ing abroad. Response: The Commission does not agree that these concerns 1ire applicable to the final rule. The Commission does not believe the final rule v;il] have any impact on the proposed rule on export and import of radioactive waste, nor will the proposed rule on export and import of radioactive wast,~ impact this rulemaking. Comment: Two commenters argued that the implications of the rule are uncertain in the aftermath of the Supreme Courfs decision rejecting the "Take Title" provision of the LLRWPAA. One commente-r asked whether States, under the proposed rule and in light of the Supreme Court's decision in New York v. United States, could refuse to allow interim on-site LLW storage. Response: On June 19, 19!32, the U.S. Supreme Court issued its d1!cision on the New York challenge to the constitutionality of the LLRWPAA. Although the Supreme Court decision, in this case, is currently being evaluated for its possible general impact on the management ofLLW in this country, the Supreme Court decision dous not impact the final rule. Comment: One commenter expressed concerns that volume reduction of a "source" containing some intrinsic value would lose its remaining value after being compacted to reduce the volume of the waste. Response: The Commission does not believe this issue is pertinent to the final rule. Comment: One commenter in Florida expressed concern that the proposed amendment would allow nuclear power plants, including Turkey Point in Florida, to store LLW on-site for an indefinite period. The commenter believes that South Florida's unique hydrology and geology raise serious questions about its suitability for storage of LLW. The commenter states that any decisions by NRC to allow LLW to be stored at reactor sites should be made on a specific basis and that an environmental impact statement [EIS) should be prepared under the National Environmental Policy Act [NEPA) for Turkey Point, because the original EIS [in 1972) did not address storage of LLW. One commenter, from Michigan, expressed concern that, given the

importance of the Great Lakes, nuclear power plants in the Great Lakes area be phased out so that no further waste accumulates.

Response: Current reactor license conditions allow licensees to store wastes generated in the operation of the reactor. While the rule authorizes reactor licensees to receive back waste shipped off-site for processing, the final rule makes no changes to existing requirements concerning storage of LLW, nor does it modify waste processing or transportation requirements. specific concerns associated with storage of wastes authorized under terms of existing licenses should be addressed on a case-specific basis. Finding of No Significant Impact: Availability The Commission previously determined that the selected action was of the type described in the categorical exclusion of 10 CFR 51.22(c)(2). After having received several comments addressing the transport and storage of processed LLW, however, the Commission has chosen to conduct an environmental assessment pertaining to these environmental concerns and the consequences of the proposed rule. The Commission has determined, under the National Environmental Policy Act of 1969, and the Commission regulations in subpart A of 10 CFR part 51, that this rule would not be a major Federal action significantly affecting the quality of the human environment and therefore, an environmental impact statement is not required. The handling and temporary storage of concentrated waste will not significantly increase risks to workers or the public. Similarly, this rule will not pose significant risks to the public or the environment resulting from additional miles traveled by radioactive wastes on our Nation's 50-SC-125 highways. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection and/ or copying for a fee at the NRC Public Document Room, 2120 L Street, NW. [Lower Level), Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from Richard H. Turtil, U.S. Nuclear Regulatory Commission, Washington, DC 20555, (301) 504-3447. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval number 3150---0011. Regulatory Analysis The Commission has considered alternatives to, as well as the costs and benefits of, the final regulation. There is no alternative to amending the regulations that would satisfy questions concerning the legality of transfer on a generic basis. The final regulation will not impose any additional cost nor burden on a licensee or other individual. The final rule is intended to facilitate actions necessary to ensure that licensees will have adequate short-term on-site storage capacity for LLW, until permanent disposal is available. The Commission does not look favorably on long-term on-site storage. The foregoing constitutes the regulatory analysis for the final rule. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act (5 U.S.C. 605[b)), the Commission certifies that this rule does not have a significant economic impact on a substantial number of small entities. The rule does not affect small entities. The final regulation is entirely permissive in nature and will predominately affect large entities, nuclear power reactor licensees, and persons who provide volume-reduction services to these licensees. Backfit Analysis The Commission has determined that the backfit rule, 10 CFR 50.109, does not apply to this final rule, and therefore, that a backfit analysis is not required for this final rule, because these amendments do not involve any provisions that would impose backfits, as defined in 10 CFR 50.109[a)[1). PART 50

STATEMENTS OF CONSIDERATION List of Subjects in 10 CFR Part 50 Antitrust, Classified information,
Criminal penalty, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, NRC is adopting the following amendment to 10 CFR part 50. 57 FR 53191 Published 11/6/92 10 CFR Part 50 [3150-AE04] Receipt of Syproduct and Special Nuclear Material Correction In rule document 92-25504 beginning on page 47978 in the issue of Wednesday, October 21, 1992, make the following corrections:

1. On page 47979, in the 3d column: a. In the 14th line, "of' should read "or". b. In the second complete paragraph, in the 15th line, "authority" should read "authorized".

2. On page 47980, in the first column, in the 4th line, "license" should reaq "licensee".

3. On page 47982, in the second column, in the second paragraph, in the second and third lines, delete "to store wastes" the first time it appears. 57FR55062 Published 11/24/92 Effective 12/24/92 Clarification of Statutory Authority for Purposes of Criminal Enforcement See Part 11 Statements of Consideration 57 FR 61785 Published 12/29/92 Effective 12/29/92 Material Approved for Incorporation by Reference; Maintenance and Availability See Part 35 Statements of Consideration 58FR21904 Published 4/26/93 Effective 5/26/93 10 CFR Parts 50 and 52 RIN 3150-AD80 Training and Qualification of Nuclear Power Plant Personnel AGENCY: Nuclear Regulatory . Commission.

ACTION: Final rule. plant in a safe manner in all modes of operation. This action is being taken to meet the directives of section 306 of the Nuclear Waste Policy Act of 1982. EFFECTIVE DATE: May 26, 1993. ADDRESSES: Copies of all referenced NRC documents are available for public inspection and copying for a fee at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC 20555. Copies ofNUREG documents may be purchased from the Superintendent of Documents, U.S._ Government Printing Office by calling (202) 275-2060, or by writing to the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082: Copies are also available from* the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. FOR FURTHER INFORMATION CONTACT:*Dr. Rajender Auluck, P.E., Office of Nuclear Regulatory Research, telephone: (3Q1) 492-3794 or Mary Ann Biamonte, Office of Nuclear Reactor Regulation, telephone: (301) 504-1073, U.S. Nuclear* Regulatory Commission, Washington, DC20555. SUPPLEMENTARY INFORMATION: Background. Nuclear Waste Policy Act of 1*982 In section 308 of the Nuclear West~ Polley Act of 1982 (NWP A), Public Law 97-425, the NRC was "directed to promulgate regulations, or other appropriate Commission regulatory guidance for the training and qualifications of civilian nuclear power plant operators, supervisors, technicians and other operating personnel. Such regulations or guidance shall establish

*
instructional requirements for civilian nuclear power plant licensee personnel training programs." In order SUIIIIARY:

The Nuclear Regulatory to meet this directive, on March 20, Commission (NRC) is amending its 1985, the Commission published a regulations to require each applicant for -Policy Statement on Training and and each holder of a license to aperate Qualification of Nuclear Power Plant a nuclear power plant to establish, ;[ Personnel (50 FR 11147). The policy implement, and maintain a training statement endorsed a training_' program for. nuclear power plant , accreditation* program managed by the personnel based on a systems approach ' Institute of Nuclear Power Operations .to training (SAT). The training program (INPO). It encompaesed the elements of will provid~ qualified personnel to effective performance-based training operate and maintain the nuclear .power and provided the basis to ensure that 50-SC-126 PART 50

STATEMENTS OF CONSIDERATION personnel have qualifieli:tions commensurate with the .Performance requirements of their jobs. In addition to endorsing the managed training accreditation program, the 1985 Policy Statement also recognized the INPO-managed accreditation of utility tr!rlnihg programs for the following categories of nuclear power plant pen:onnel:

(1) Non-licensed operator; . (2) Control room operator.

* (3) _Senior control roon1 operator/shift supervisor. " (4) Shift technical advisor. (5) Instrument and control technician.

(6) Electrical niaintenaJJce personnel. (7) Mechanical maintenance J)_i!r&onnel.

(8) Radiological protection technician.

(9). Chemistry techniciah. (10) On-site technical staff and managers. While issuing the policy statement, the Commission decided to defer rulemaldng in this area for a minimum of 2 years in order to allo,v the industry to continue its initiatives to upgrade training programs .through the managed training accreditation program. Following issuance of the policy statement, the NRC evalm1ted the managed training accreditation program over a 2-year period and concluded that it was an effective program. On November 18, 1988 (53 FR 466073), the NRC published an amendod policy statement in order to: (1) Provide additional ir:iformation regarding the NRC's exper:ience with industry accreditation, (2) Change the policy regarding enforcement to eliminate discretion in inspection and enforcement in the areas covered by the 1985 Polle,* Statement, and -(3) Reflect current Commission and industry guidance. The NRC continues to perform inspections at differe~t utilities to ensure that these training p,rograms remain effective.

' U.S. Court of Appeals Decigion On April 17, 1990, the U.S. Court of Appeals for the District of Columbia Circuit concluded that the Commission's Policy Statei:111ent did riot meet the-intent of the Congi,issional directive to create mandatot1' requirements for personnel training programs at civilian nuclear power plants. The Court remanded'.

the issue back to the NRC for action consistent with*the. Court's findings. Sne, Public Citizen v. NRG, 901 F.2d 14'.7 (DC Cir. 1990). The Commission requested a rehearing of the decision by 'the full court, which was denied on June 19, , 1990. On November 26, 1990, the Supreme Court denied certiorari on petition by the Nuclear Utility . Management and Resource Council. See, Nuclear Management arid Resources Council, Inc. v. Public Citizen 111 S. Ct. 536 (1990). Actions Taken in R.esponse to the Court. Decision In response to the court decision; the ' NRC developed the proposed rule that . would amend 10 CFR parts 50 and 52, entitled "Training and Qualification of Nuclear Plant Personnel." The proposed* rule was published in the Federal * .

Register on January 7; 1992 (57 FR 537). The amendments would *require that each applicant for and each.holder oh* license to operate a nuclear power pll$t establish, implement, and maintairi

_a . training program for nuclear power

plant personnel that provides qualified personnel to operate and maintain the facility in a safe manner in all modes of operation.

The proposed rule met the

directives contained in section ~06 of the Nuclear Waste Policy Act of 1982 (NWPA), Public Law 97-425, as interpreted-by the U.S. Court of Appeals for the District of Columbia Circuit, that mandatory requirements be established for the training and qualification of personnel at civilian nuclear power plants. The proposed rule would require training programs that are derived from a systematic analysis of job perfonnance requirements that can include botli specific and industry-wide experiences.

Current industry training programs have been developed consistent with this . . approach. Based on monitoring industry training programs since the 1985 Policy Statement went into effect. the NRC has concluded that these programs have been generally effective in ensuring that personnel have qualifications

commensurate with the performance requirements of ~eir jobs. Summary and Analysis of Public Comments The comment period for the proposed rule expired March 9, 1992. Public comment letters received on the proposed rule are available for public inspection and copying for a fee at the Commission's Public Document Room. Comments were received from 30 individuals and corporate entities, virtually all of whom are directly involved in the nuclear power industry.*

Many of the letters contained_ simJlar . comments and have been grouped . together and addressed as a single issue. All comments have been grouped into* eight broad issues. For each broad issue, 50-SC-127 the NRC baa included a summary of the comments received and an analysis and response to those comments.

i. Responsibility for Training and Acceptability of Third-Party Accreditr,rtion Programs Comment. Several commenters indicated that the ~C should clarify who under the proposed requirements wiUhave responsibility for training amtnctor personnel.

Given the proposed rule'a requirement that training programs be b8.86d on a systems approach to training, they indicated that the-NRC should clarify its intention reg~ding the acceptability of licensees re.lying on third-:,party training programs other than INPO-managed training ~aeditation-certification programs in evaluating the training needs and qualifications of personnel. The Radiation Protection Association's program of registration-certification of Health Physics Technicians was cited as en example of an other than maneged training acczeditation-. certification program that the Commission should explicitly endorse *. Commenters also indicated that the NRC should clarify that if the evaluation of ~cmnel does not indicate that additional training is needed (i.e .* they are already qualified), then additional training ia not required. Finally, commenters questioned whether the NRC has developed acceptance criteria for licensees to use in determinin~ the acceptability of vendOMieveloped and . other third-party training programs. and if the NRC anticipated cwriving such

criteria &om NUREG-1220. .
Response.

The intent of the rule is to that nuclear power plant ~l have the necessary . ,' mowledge, skills, and abilities to perform their assigned jobs competenUy; i.e., they 8l'8 qualified to independenUy perform specific activities. Therefore, it is the responsibility of each licensee and applicant to ensure that personnel specified by the rule, regardless of whether they are employees or contractors. are qualified. The mquirement that each licensee or applicant develop, implement, and maintain a SAT-based training program iu*pplicable only to licensee personnel, nol contractors, and establishes a process that provides a high degree of assunmce that personnel will be qualified to perform their assigned duties. This assurance arises from the five.major elementA of the SAT process: (1) Aoalyaia of job perlormance iequlnmlenl$ and training needs; (2} derivaUon of learning objectives; (3} design and implementation of the PART 50

STATEMENTS OF CONSIDERATION training programs:

(4) trainee evaluation: and (5) program evaluation

and revision.

Training is onlyrequired when a comparison of job performance requirements for tasks being assigned and the skills and knowledge of a specific Pl?ll'son indicate a training need. Third-party (including developed) training programs, although not specifically endorsed by the NRC. are acceptable provided that the licensee has evaluated the programs to ensure that they will ISSult in pro~ qualification. Because the acceptallility of vendor-developed programs will vary based on individual f~cilit¥. needs, the NRC is not providing spec1fic , acceptance criteria. Licensees should evaluate vendor-4eveloped training programs against the facility's job and task analysis results to ensure that the vendor programs will m.eet the licensee's specific qualification requirements.

2. Appropriateness of SAT-Based Training Comment. Numerous commenters questioned the appropriateness of requiring SAT-based training.

At the most basic level was a concem* that the NRC has not placed sufficient emphasis on the fact that the required SAT-based training is not intended to be a simple "cookbook" approach and that personnel should be encouraged to acquire additional knowledge, training, and academic instruction to gin them e deeper understanding of the technical principles underlymg their trainine, A more widely held coocem was that the proposed rule could be construed as requiring additional rigorous Job and taslt analysis, particularly smcB NRC Inspection Procedure 41500 is more restrictive in this area then the current requirements for INPO-managed training accreditation.

These commenter& noted that the job and task analysis is not necessarily appropriate or sufficient for all of the types of personnel covered .by the rule. They requested that the NRC explicitly acknowledge that varying degrees of rigor in the-performance of fob* and task analysis are appropriate for differing types of positions, as are analyses completed through cooperative generic industry efforts. Specifically, it was recommended that training programs affecting the Shift Supervisor, Shift Technical Advisor, end Technical Staff and Managers be allowed BDd encouraged to rely on additiooal bases for deteimining training needs and that Inspection Procedunt "1500 and NUREG-1220, Training R&view Criteria and ProceduJea~" be wvised to make them fully consistent with cummt INPO guidance. Finally, one commenter noted that SAT-based training is inconsistent with the requirements in 10 CFR part 55 for licensed operators and requested that the NRC explain why it hes determined that only SAT-based treinin~ fo 11cceptable. Response. The Commission shares the concern that SAT-based training not be treated in a "cook.book" manner. It is not the intent of the NRC that the iDdustry simply approach the based program in a "cookbook" manner, since Ute NRC does not intend to discourage licensees from imposing additional requirements above thos!3 developed from the SAT-based training program. The decision to require based training programs reflects both . the industry's success with this approach and the fact that the process has the advantage ofincorporating ongoing review and revision of the program to reflect changing needs. Additional rigorous job and task analysis will not be required for any of the positions listed in this rule. The NRC has monitored and evaluated the development and implementation of the current industry programs. The NRC believes that the*job, task, and needs analyses underlying the currently accredited programs 81'8 adequate, as are the criteria that are used in determining the acceptability of programs for future accreditation. In order to clarify its position that additional jab and task analyses are not being required, thir NRC has revised Inspection Procedure 41500 and NUREG-1220 to make them consistent with this regulation. . The NRC recognizes that 10 CFR part 55, which only applies to licanaed operators, allows non-SAT based approaches to training. This provision was necessary to accommodate existing industry programs for training licensed operators at the time part 55 was promulgated, because industry-wide implementation of SAT-based trainiDg was not complete. At this time, based training bas been broadly implemented by the industry for both licensed operators and other plant personnel. Virtually all of the initial and requalification programs for licensed operators ere SAT-based. The NRC believes that, based on SAT's success and its wide use by industry, that it is appropriate to incorporate SAT-based training es _a requirement in this rule. 3. Defmition of Personnel To Be Trained Comment. A number of both general and specific commenta were n,ceived that addNlssecl the issue of the definition of the penonnel that would be awered by the pJ1>POll8CI rule. Seme held the view that the specification of 50-SC-128 the paraomiel CC>Vered by*tha p12posed i,de was too narrow while* otliers beiiavad it was too broad; Some *

commentera recommended that the
proposed i,de cover all penonnal who perform or oversea design, operaUon, or maintenance activiUes regardless of whether they ere physically located on~ site or off-site.

At the same time, the commenters*iDdicated that the proposed rule should explicitly exclude all personnel who, regardless of location. only perform support services. In addiUon, the commenter& iDdicated that personnel worlting under direct

supervision, such as short-tenn contractor personnel, should be excluded from these trahiing requirement$.

More concretely, numerous commenter& requested that the proposed rule be clarified in terms of personnel to be trained. For example:

rs a ."Radwaste Operator" to be considered a "Non-Licensed Operator"?

Does "Shift Supervisors" iDclude * "Assistant Shift Supervisors"*? While the job Utle "l&C Technicians" is precise, "electrical personnel" and "mechanical personnel" ar,pear to be, broader than "electricians.' and "mechanics." Which functions or personnel are included in on-site technical staff?

Finally, a number of commenters noted that to be consistent with INPO's current eccreditaUon program, managers should not be included in the rule. This could be accomplished by amending § 50.120(b) so that the'::C!Fonnel designated as (9) T cal Staff and Managers" be designated "Engineering Support Personnel." Almost a11 the commenters seeking clarification of tlie ** personnel to be trained recommended that the regulatiqn e~licitly state that it applies only to the training programs currently included in the.INPO accreditation program.
Response.

The scope of the list of personnel is consistent with and -incorporates those posiUons that are currently covered by*accreditetion and exisUng-industry practice. The NRC believes that the exisUng training programs for the personnel listed are saUsfactory. For these.reasons, the NRC does not believe it is appropriate to revise the scope of the regulation. The shift supervisor posiUori is the only supervisory function included within the scope of the rule end does not include the assistant shift supervisor. However, the regulation has been revised to reflect a change from the title of the "on-site technical staff and managers" category to "engineering support personnel," which is now consistent with the name for the corresponding lNPO accredited training . ,"-, PART 50

STATEMENTS OF CONSIDERATION program. The scope of the personnel covered by that progr&Jll to be trained in accordance witn*this_re,gulaUon,
however, is 1,1Dchanged, . *. . The NRC does not beilieve that a ch. anse to the rule is nettded in. order to clarify the applicability of this rule to short-term contractor pursonnel. . Contractor personnel Brit not cove)'.8d_

by this rule unless they occ:upy regular. positions working inde11endenily within the licensee's orgenization. However, if short-term contractor pe,rsonnel are

assigned to work indepe,ndently, they must be qualified to per:rorm the
assigned tasks; Finally, lhe issue of including off-site persoD1Del in the final *rule has been considered.

The NRC h8' concluded that the requirements apply to Job functions in the id'.entified

categories of personnel relating to site activities regardless of the lcicatio~

of the personnel.

4. Relationship BetM--een Troining and Qualification Comment. One commEnter expressed .concern that the relation:,hlp between training and qualification has been .

blurred. The commenter indicated that while the proposed rulit" l.s entitled "Training and Qualificatlion of Nucleill' Power Plent Personnel," the requirements appear to n,late to training only. It was recommended that the tenn qualification be eliminattid, or, alternatively, that necessury
. qualifications be explicitly listed. The commenter also indicated that the NRC should clarify that successful completion.of a training 11rogram is-not in end of itself sufficient, in lieu of eny
specific qualifications Imposed by other regulations, for a.particulnr position.

Response. The NRC disigrees that the distinction between training and

qualification hes been blumd by the rule. As stated in the preamble for the proposed rule,qualificetietn in the context of this rule meens Job task qualification.

The propoSEd rule contained the requirement-that licensees end applicants develop, iE11plement, end maintain a SAT-based training program to ensure that nuclear power personnel are qualified to perform the tasks of their Jobs. Because licensees and

applicants must comply with ell applicable regulations, there should be no ambiguity concerning the fact that . successful completion of a training program does not obviate the need to
comply with any other training or qualification requirements imposed by other regulations or licenSE,, conditions.

This means that nuclear power plant personnel must also meet the-licensees' initial Job qualification req11irements imposed as part of initial employment. -Therefore, no changes were made to the rule-in.response to this comment. s. Applicability of the Rule Comment. Several commenters expressed the opinion that the applicability of the rule was too broad with respect to licensees who are undergoing decommissioning or are part *. 52 applicants. Specifically, they

recoi:nmended that the rule apply only to applicants for or licensees with en operating license .. The commenters
suggested that facilities engaged in decommissioning where all fuel has been permanently removed from the reactor vessel or those with a possession only license (POL) should not be subject to this rule._Additionally, they questioned why part 52 needed to be . amended to include the requirements of § 50.120(b), since the.provisions of part 52 already automatically incorporate ell*
of the standards in part 50 that _ere . technically relevant.

Response. The NRC believes that making the provisions of the rule applicable to all pert 50 licensees and app_licants is appropriate. The process ensures that as plant conditions change, training programs will be revised to reflect these changes. These revisions could include*the development of new programs or the elimination of obsolete programs. However, the process also ensures that the modification of the program to reflect the changed environment is performed in. an orderly fashion. If permanent changes in the condition of the plant (i.e., decommissioning or POL) make some_ or all existing .training programs unnecessary, the licensee would obtain relief from these requirements by applying for an .

exemption eliminating or modifying the affected programs.

Also, the reason that 10 CFR part 52 needs to be amended is to ensure that .part 52 applicants have considered the requirements of-10 CFR 50.120(b) in their applications.

6. Implementation of the Rule Both general and specific concerns were raised regarding implementation of the rule, the time periods allowed for implementation, and the means to be used by licensees to demonstrate compliance of a training program that is not accredited by the INPO-managed . training accreditation program. (a) General Concerns Comment~ Numerous commenters expressed concerns regarding the manner in which the NRC will monitor implementation of the rule to ensure that it is consistent with the Commission's intentions and that the

50-SC-129 guidance provided by the NRC and INPO is consistent.

Specifically, it was recommended that the Commission carefully monitor the implementation of the final rule to ensure a consistent understanding of the regulatory goals es was identified in SECY-91-172, "Regulatory Impact Survey-Final" In addition it was suggested that the principles in the Staff Requirements Memorandum dated December 20, 1991, regarding the Systematic Assessment of Licensee Performence (SALP) program should be applied to this traini,ng rule. Commission monitoring of the training rule would ensure that there is and inter-regional consistency in the requirements, evaluation

criteria, and results, and would preclude the
imppsition of additional requirements based on rising expectations.

Commenters indicated that the NRC should clarify the process that INPO and the NRC will use to avoid giving licensees conflicting guidance. Commenters indicated that the NRC should explicitly state that maintenance of en accredited training program will be construed es complete compliance with these training requirements. Other commenters indicated that NRC should consider.delaying the effective date of the rule until it hes completely reviewed implementing guidance (e.g., Reg. Guide 1.8,-Rev.

2) and made it consistent with the final rule. Response.

The NRC believes that the requirements and implementation of this rule will be consistent with the accredited programs already developed and implemented by the industry. Therefore, the p9licy the Commission* expects to follow in implementing the rule is that continued accreditation along with effective implementation of the accredited program is considered to be an acceptable means of demonstrating compliance. This . conclusion is based on staff inspections which have found the accredited programs to be generally acceptable, and the NRC review of documents that provide the industry program objectives

and criteria.

An applicant or licensee . could also comr ly with the . requirements o this rule withoµt being accredited. Inspection Procedure 41500 end NUREG-1220 have been revised to make them consistent with this regulation. This guidance will be used by the NRC staff when monitoring implementation of this rule or inspecting training programs and is intended to ensure consistent interpretation of training criteria by an NRC regions. The NRC, therefore, does not intend to revise Reg. Guide 1.8. PART 50

STATEMENTS OF CONSIDERATION (b) Implementation Period Comment. With regard to the specific time frames allowed for implementation, several commenters expressed the opinion that if the rule is truly consistent with established programs, that an implementation period of 180 days was reasonable.

However, other commenters stated that additional time should be granted to accommodate the industry's implementation date of December 31, 1993, for the new "Engineering Support

Personnel" accredited program and for the review and documentation activities that are believed by the commenters to be necessary to demonstrate compliance with the rule. Similarly, many believed that the requirement in § 50.120(b) that applicants must have established and
implemented the required training program 18 months prior to fuel load is not reasonable, given that the accredit~tion process for training programs provides for verification and revision of training programs based on experience gained from operations.

It was recommended by several commenter& that applicants simply be required to have training programs established and reedy for accreditation prior to initial fuel load. Finally, several commenters poted that linking the. required program review and revision cycle to the industry's current 4-year schedule is unnecessarily prescriptive. Response. The NRC bas considered the issues raised by the commenters regarding the appropriate implementation time periods for both. licensees and applicants. For licensees, the Commission believes that the 180-day implementation period is sufficient, because all licensees have developed, implemented, and are maintaining accredited programs. Implementation of the new "Engineering Support . Per:sonnel" program, which replaces the current "Technical Staff and Managers" program or other future accredited

program changes*, does not negate the fact that SAT-based training is continuing for the personnel covered by the rule, therefore, compliance with the regulation would ba maintained.

The requirement that applicants establish and implement the training program 18 months prior to fuel load is

also considered appropriate.

The NRC realizes that an applicant would not . have a training program accredited 18 months prior to fuel load, and this rule does not require accreditation. The rule only requires that a training program be established for those portions of the plant programs nece~sary to support

ongoing activities covered under the rule. In addition, the NRC believes that having the S.A T-based training program in place prior to fuel load allows significant benefits in terms of program review and revisions based upon experience gained prior to fuel loading. The NRC concurs that linking the program review-revision cycle to existing practice (i.e., a 4-year accreditation-renewal cycle) is unnecessarily prescriptive, therefore reference to specific 4-year review cycle has been deleted from the supplementary information section of the final rulemaking notice. (c) Review and Recordkeeping Requirements
Comment. Several commenters requested that the NRC clarify the requirements for recorclkeeping and for program reviews and revisions.

Specifically, the NRC was requested to clarify (1) what records need to be maintained in order to meet the requirements of § 50.120, (2) whether any special retention periods apply to these records, and (3) what "associated programs" must be readily auditable, or that this language be dropped from the discussion. They also requested that NRC clarify the rule so that it is clear that the ,periodic reviews of training programs are to be conducted by appropriate functional managers, not just training managers. Response. The records the licensees will need to maintain to meet the requirements of§ 50.120 are the same records currently being maintained by licensees for their existing training programs. The prorosed rule does not impose any specie retention periods for these records; The words "associated programs" will be deleted from the discussion related to being readily auditable. The final rule does not require clarification since the proposed rule notes "licensee management," which NRC talces to include functional line managers:. . (d) bemonstratioµ.ofCompliance Comment. The NRC should clarify how compliance with the rule is to be . demonstrated by facilities without an accredited program. Response. An accredited program is considered to be an acc!3ptable means of .demonstrating compli!lJlce with the

rule. Facilities that do not have an accredited program would demonstrate complimce with the final rule through the development of training programs
using the systems approach to training as defined in 10 CFR 55:4. The NRC will conduct' inspections of non-accredited facility programs to ensure tha~ the
requirements of the final rule are meL 50-SC-130

7. Recommend That the Commission T,y One More Time To Reverse the Court Decision Comment. Most commenters expressed their strong opinion that the rule is unnecessary given the industry's initiatives in developing and implementing effective training programs, but accept the rule as necessary given the Court's decisions.

However, one commenter requested that, given the President's January 28, 1992, directive that agencies are to "identify and accelerate action on initiatives which will eliminate any unnecessary regulatory burden," the Commission seek, through the Executive Branch, if necessary, a judicial review of the Court's ruling. . Response. The Commission believes that the President's directive does not supersede the Court's ruling and the NRC has exhausted all reasonsble avenues of judicial review .. 8. Reconsideration of Other Training Requirements in Ught of This Rule Comment. One commenter requested that the NRC review part 55 in its entirety to ensure that it is consistent with this rule, stating that it is possible that many of the prescriptive requirements in existing part 55 could be eliminated if it were amended to reflect existlng industry practice for identifying the need for and developing* training programs. . Response. Part 55 currently states that a SAT-based training program and a certified simulator is an acceptable alternative to the prescriptive requirements of part 55 and would meet the existing requirements for licensed operator training. In fact, 'most of the initial and requalification programs for licensed operators are based on SAT. Thus, part 55 is consistent with this rule. Furthennore, some programs retain elements of the prescriptive portion of 10 CFR part 55 and to eliminate these elements would create an unnecessary perturbation to these programs. Discu111,ion The safety of nuclear power plant operations and the assurance of general public health and safety depend on personnel performing at adequate perfonnance levels. The systematic determination of qualifications and the provision of effective initial training and periodic retraining .will enhance confidence that workers can perform at adequate performance levels. Qualification in the context of this rule means that nuclear power plant personnel have completed the training program, or parts thereof, as evidenced PART 50

STATEMENTS OF CONSIDERATION by meeting the job.perfonnance requirements, end are pem1itted to
independently perform spocific . activities.

The Commission has taken en approach in this rule that would specify the process to be impleniented by applicants an<J. licensees through*which job performance criteria 8IJ1d associated personnel training would. he derived. This approach provides foir' flexibility and site-specific edaptatio111s in the training programs. No additional c*ost is anticipated with this approach for licensees with accredited programs because th11 rule is believed to be consistent with existing industry . practice for personnel training. Summary of Final R.ule Each applicant for and each holder of an operating license for a imclear power plant shall: . (1) Establish a training p,rogram for . certain nuclear power plant personnel who perform operating, m,aintenance, and technical support activities;

* (2) Use a systems approach to training:

(3) Incorporate instructkmal requirements to provide trained and qualified personnel who cm safely operate the facility in all modes of operation: . (4) Periodically review, evaluate, and revise the training progra111i: and (5) Maintain sufficient records, available for NRC inspection, to verify the adequacy of the trafning program.

Although no written response is required, licensees are exi1ected to review their license conditions and other commitments for co:nsistency with this rule. The Commission has alno developed conforming amendments to 10 CFR parts 50 and 52 to accompany this rule. Pert of these amendments to perts 50 and 52 are considered minor. The other -change to pert 52 is more substantive and has been developed t1l ensure that applicants for a combined.

license (construction and operation) will establish, implement, and maintain a training program In accordance with the requirements in 10 CFR 50.120. This rule is not intended to pn1elude vendor training programs developed in conjunction with standardization of design. Discussion of Final R.ule A new 5 50.120, hes beon added to 10 CFR part 50, entitled "Tr11ining and qualification of nuclear p11JWer plant

personnel."
This section establishe11 the requirements for and the ;8ssential . . elements of the process tu be used by applicants and Ucensees to: * (1) Deterrtilrie training and qualification requirements for all appropriate personnel:
(2) Develop training programs to ensure that each licensee has trained and qualified personnel to operate and maintain the facility in a safe manner: and (3) Implement and maintain these , pro~ams effectively on a continuing
basis. * *
Paragraph (a), "Applicability," indicates that the rule applies to each applicant for and each holder of an operating license for a nuclear power plant.
Paragraph (b), "Requirements," requires that each applicant or licensee establish, implement, and maintain a program for training nuclear power plant personnel which addresses all modes of operation and is derived from a systems approach to training (SAT). The SAT process was selected because : it has the followin~

characteristics:

(1) Training design ~d content are . derived from job performance

' requirements:

(2) Training is evaluated and revised , in terms of job performance . '. requirements end observed results on the job: . . (3) Success in training can predict ' satisfactory on-the-job performance; and (4) A training program can be audited . because it involves clearly delineated

' process steps and documentation. .

The SAT process contains five ma1or elements and is intended to require a training system that will ensure
successful performance on the job by trained individuals.

The elements ere: (1) Analysis of job performance requirements and training needs; (2) Derivation of learning objectives; (3) Design and implementation of the training l?rograms: (4) Tramee *evaluation: (5') Program evaluation end revision. The SAT process also provides a . sequential method of generating the type of documentation needed for treinlng review. Use of SAT will obviate the need for additional documentation for NRC review. The SAT process is a generic process, and its application is not limited to a certain subject matter or to specific licensee personnel. Training programs based on job perfo~ance requirements have been successfully used by the military for over 20 years and by the nuclear industry for much of the past decade. Furthermore, the Commission has recognized the appropriateness of using this approach to training in its requirements for operator licensing prescribed in § 55.31(a)(4), and for operator requalification prescribed in § 55.59(c}.

50-SC-131 This rule would provide for the training and qualification of the following nuclear power plant personnel:

(1) Non-licensed operator. (2) Shift supervisor. (3) Shift technical advisor. (4) Instrument and control technician. (5) Electrical maintenance personnel. (6) Mechanica.l maintenance personnel. . (7) Radiological protection technician. (8) Chemistry technician. (9) Engineenng support personnel. Licensed operators, sucli as C!'.>ntrol room operators and senior control room operators, are not covered by this rule. They will continue to be covered by 1 O CFR part 55 for both initial and

requallfication training.

Because some senior control room operators may also be shift supervisors, only those aspects of training r~leted to their shift supervisor function would be covered by this rule. This rule would require that training programs be periodically evaluated and revised as appropriate, and also be periodically reviewed by management for effectiveness. Current industry

objectives in this regard involve the evaluation by management of individual training programs on a continuing or periodic basis to identify program strengths, weaknesses, and effectiveness.

These evaluations are norm11lly completed within a 3-to 6-month period following completion of training within the programs. The sum of these evaluations results in a comprehensive review. Periodic evaluations of the overall training programs are being performed by the industry as part of accreditation renewa_l. The Commission expects this . practice to continue. Determination of job performance requirements and training needs is part of the analysis in the SAT process and is reflected in qualification requirements. The facility applicant or licensee will be responsible for ensuring that all personnel within the scope of this rule.have the training end resulting qualffications commensurate with job performance requirements for their assigned tasks. Initial and continuing. training, as appropriate, is expected to be provided t~ job incumbents in positions covered by this rule. Each applicant and licensee.is Jequired to maintain and keep available for NRC inspection the materials used to establish and implement required training progrenis for the affected personnet Current industry practice in this regard involves retention of tltose records necessary to support management 'information needs .end to PART 50

STATEMENTS OF CONSIDERATION provide required historical da~. In general, these include records of program development, evaluation, and revision related to the existing training program. The NRC has found through inspections of training programs that sufficient records are being retained for periods that are adequate for regulatory purposes.

The Commission believes that no additional guidance for recordkeeping is necessary. No written response is required by . this rule. However, applicants and licensees would be expected to compare their current training commitments and licensing bases with the requirements of this rule. Licensees should use the results of this comparison to evaluate and revise, as appropriate, existing technical specifications or previous commitments. This approach will ensure a common understanding between applicants, licensees, and the NRC staff of training commitments when future inspections are conducted. Impact of This R.ule on Existing Industry Training Programs. This rule would supersede the Policy Statement on Training and Qualification of Nuclear Power Plant Personnel. The Commission believes that this rule would not result in any change t.o accredited training programs. The NRC has found through inspections that the programs are generally acceptable. The Commission expects that training programs accredited and implemented consistent with the industry program objectives would be in compliance with the requirements of this regulation. An existing Memorandum of Agreement between INPO and the Commission assures that the NRC will be made aware of any modifications or updates to the industry's program objectives and criteria. Having seen such modifications, the NRC will review to determine if they warrant any modification in the Commission's position expressed above. The NRC will continue to monitor.the industry accreditation 1;>rocess by.: (a) Nominating individuals who are not on the NRC staff to serve as members of the National Nuclear Accre~iting Board with full voting privileges; . (b) Having an NRC staff member attend and observe selected National Nuclear Accrediting Board meetings with the .INPO staff or the utility representatives; . (c) Having NRC staff observe ~lected INPO accreditation team site visits; * (d) Reyiewing any subsequent

revisions to the program objectives and criteria es currently described in the . National Academy for Nuclear Training document "The Objectives and Criteria for Accreditation of Training in the Nuclear Power Industry" (ACAD 91-015);1 and . (e) Verifying licensee programs through the NRC inspection process. As noted above, the NRC lies the ability to verify compliance with this regulation through the inspection program and will do so as appropriate.

In its inspections, tho NRC staff will use Inspection.Procedure 41500, "Training and Qualification Effectiveness," which references the guidance in NUREG-1220, Revision 1, 2 "Training Review Criteria and Procedures." Based on NRC inspections conducted to date, the Commission believes that the objectives developed by the industry provides sufficiently clear guidance to allow applicants and licensees to implement effective training programs in compliance with this rule. Therefore, the Commission does not believe it is neoossary to issue e regulatory guide to provide additional guidance for complying with this rule. Vendor-Developed Programs for Standardized Plants In 10 CFR part 52, the Commission articulated the goal of safety through standardization of design. The Commission believes that the benefits of standardization could involve the standardization of some types of training associated with the 10 CFR part 52 design certification. Therefore; nothing in this rule is intended to preclude standard training programs being developad or implemented by a vendor. For example, the initial training for instrument and control technicians related to a particular standard design may be conducted by a vendor. As a result, there could be a pool of technicians trained by the vendor on the certified design available for hire at a nuclear power plant site. These personnel, however, would need to complete site-specific training related to the administrative end operating philosophy of the site as well.as any other specific requirements of. the licensee. Thus, the requirements for personnel training programs prescribed by 1 A copy of ACAD 91--015 Is available for public lnspec:Uon or copying et the NRC Public Document Room, 2120 L Street, NW. (Lower Levell, Washington, DC. 2 Copies of NURF.G-1220, Rev. 1 may be purchased from the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC :zopt:1-7082; Cople, ere also available from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. A copy Is also available for public lnspecUon or copying at the NRC Public Document Room, :ZlZO L Slr9et, NW. (Lower Level), Washlnston, DC. 50-SC-132 S 50.120 do not prevent a vendor from training personnel or from developing a training process. However, it is important to note that vendor training programs are not governed by this rule and that the licensee is ultimately responsible for ensuring that personnel are qualified. Applicants for a Combined License Part 52 is being emended to require that applicants for combined licenses establish, implement, and maintain training programs in accordance with the requirements in 10_.CFR 50.120. Criminal Penalties As a result of the addition of§ 52.78 by this rulemaking, the criminal penalty provision, § 52.113, is being modified to add§ 52.78 to the list of sections in subsection (b), since the new section is not identified es substantive, 88 that criterion is expressed in the Federal Register Notice: Clarification of Statutory Authority for Purposes of Criminal Enforcement, 57 FR 55062 (November 24, 1992). In addition, S 52.101 ls deleted and S 52.103 is added to the list, to reflect changes made to part 52 in a previous rulemeking, 57 FR 60978 (December 23, 1992). Finding of No Significant Environmental Impact: Availability The Commission hes determined under the National Environmental . Policy Act of 1969, as emended, en~' Commission's regulations in subpart A ** of 10 CFR part 51, that this rule is not \ a major Federal action significantly l affecting the quality of the human ; environment and, therefore, an environmental impact statement is not required. Numerous studies have shown that In complex man-machine systems, human error has often been the overriding contributor to actual or potential system failures that may be precursors to accidents. With this rulemaking, the NRC is emphasizing the need to ensure that industry personnel training programs are based upon job performance requirements. Personnel who are subjected to training based on job performance requirements should be able to perform their jobs more effectively, and with fewer errors. Therefore, the environmental effect of implementing this rule would, if anything, be positive because of the reduction in human error. The environmental assessment and finding of no significan*t impact on which this determination is based are available for inspection et the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC 20555; Single PART 50

STATEMENTS OF CONSIDERATION copies of the environmental assessment end finding of no sign~ficant impact are available from Rajender Auluclc, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone:

(301) 49i-3794. Paperwork R.eductio~ Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Management end Budget approval numbers 315~011 and 3150-0151. Public burden for update end maintenance of information is estimated to average 780 hours per utility per year, including the time for reviewing the present program, searching existing data sources, gathering and maintaining the data needed, end co~pleting and reviewing the collection ofinformation. Send comments regarding this burden estimate or any other aspect of this collection-of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB-7714), U.S. Nuclear Regulatory Commission, Washington, DC 20555; and to the Desk Officer, Office of Information end Regulatory Affairs, NEOB-3019, (3150-0011), Office of Management and Budget, Washington, DC 20503. R.egulatory Analysis : A regulatory analysis has been prepared for this final regulation. The analysis examines the values (benefits) and impacts (costs) of implementing the regulation for personnel training and qualification. This analysis is available for inspection in the ,-:Re Public Document Room, 2120 L Street, NW., (Lower Level), Washington, DC 20555. Single copies of the analysis may be obtained *rrom Rajender Auluck (see ADDRESSES heading). R.egulatory Flexibility Certification As required by the: Regulatory Flexibility Act of 1989, 5 U.S.C. 605(b), the Commission certifies that this rule will not have a significant economic _impact on a substantial number of small entities. This final rule primarily affects the companies that own and operate light-water nuclear power reactors and the vendors of those 'reactors. The companies that own 'and operate these reactors do not fall within the scope of the definition of "small entity" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulatiqns issued by the Small ' ' Business Administration in 13 CFR part 121. Backlit Analysis The Commission has determined that: the backfit rule, 10 CFR 50.109, does not apply to this final rule because these amendments are mandated by section 306 of the Nuclear Waste Polley Act of 1982, 42 U.S.C. 10226. Therefore, a . backfit analysis is not required for this , rule. : List of Subjects . 10 CFR Part 50 Antitrust, Classified information, Criminal penalty, Fire protection,. Incorporation-by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation . protection, Reactor siting criteria,

Reporting end recordkeeping

requirements.

i 10 CFR Part 52 . Administrative practice end : procedure, Antitrust, Backfitting, Combined license, Early site permit,

Emergency planning, Fees, Inspection, Limited work authorization, Nuclear power plants and reactors, Probabilistic risk assessment, Prototype, Reactor
siting criteria, Redress of site, Reporting and recordkeeping requirements, Standard design, Standard design certification.

For the reasons set out in the preamble and under the authority of the , Atomic Energy Act of 1954, as amended, : the Energy Reorganization Act of 1974, : as amended, the Nuclear Waste Policy , Act of 1982, and 5 U.S.C. 552 and 553,

the NRC is adopting the following

' amendments to 10 CFR parts 50 and 52

as follows: 50-SC-133 58 FR 33993 Published 6/23/93 Effective 7 /10/96 10 CFR Part 50 RIN 3150-AE55 Monitoring the Effectiveness of Maintenance at Nuclear Power Plants AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is emending its regulations for monitoring the effectiveness of maintenance programs at commercial nuclear power plants. The current regulations require that nuclear power plant licensees evaluate performance and condition monitoring activities and associated goals and preventive maintenance activities at least annually.

This amendment changes the time interval for conducting evaluations from a mandatory once every year to at least once every refueling cycle, but not to exceed 24 months. EFFECTIVE DATE: July 10, 1996. ADDRESSES: Copies of comments received on the proposed rule may be inspected and copied for a fee at the Public Document Room localed at 2120 L Street, NW. (Lower Level), Washington, DC. Singh~ copies of the environmental assessment are available from Joseph J. Mate, Office of Nuclear Regulatory

Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone:

(301) 292-3795. FOR FURTHER INFORMATION CONTACT: Joseph J. Mate, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3795. PART 50

STATEMENTS OF CONSIDERATION SUPPLEMENTARY INFORMATION:

Background On July 10, 1991 (56 FR 31324) the NRC published the final rule "Requirements for ?vfonitoring the Effectiveness of Maintenance at Nuclear Power Plants" (§ 50.65). The final rule, which will become effective July 10,* 1996, requires commercial nuclear power plant licensees to monitor the effectiveness of maintenance activities . for safety-significant plant equipment in order to minimize the likelihood of failures an*d events caused by the lack of effective maintenance. Section 50;65 (a)(3) requires nuclear power plant licensees to evaluate the overall effectiveness of their maintenance activities on an annual basis. An industry consensus guidance document and a regulatory guide to provide an acceptable methodology for implementing the final rule are expected to be published by June 30, 1993. Discussion Since the Maintenance Rule was published in July 1991, two events have occurred that led the Commission to reconsider the annual evaluation requirements in § S0;65(e)(3). I:irst, in the Summer of 1991, the Nuclear Management Resources Council (NUMARC) Steering Group was formed to develop an industry guide for implementing the Maintenance Rule. While developing the guide, the Steering Group suggested to the. NRC in a public meeting held on February 26, 1992, that instead of annual assessment requirements, the NRC should consider assessments based on e refueling cycle interval. The NUMARC Steering Group stated that: (1) Significantly more date would be available during refueling cycles than is available on an annual basis; (2) Key date from some surveillance tests can only be obtained during refueling outages and is not available on an annual basis; and (3) Adjustments to maintenance activities that may be made after such an evaluation would be typically performed after e refueling outage. The NUMARC Steering Group further added that the evaluation process is a time consuming* activity and that with limited data available, the annual evaluation would not provide for meaningful results. With only limited data, changes to maintenance programs will likely not be made because there would not be sufficient information available for spotting trends or doing trend analysis.

. Second, the NRC conducted e regulatory review to eliminate or revise unnecessarily burdensome regulations and published a final rule on August 31, 1992 (57 FR 39353) that amended several regulations identified by its Committee to Review Generic Requirements (CRGR). One of those amended regulations was 10 CFR 50.71 (e) (Final Safety Analysis
Report Updates) where the frequency of licensee reporting to the NRC was changed from annually to once per refueling cycle. The change was made because the use of e refueling cycle interval provided a more coordinated and cohesive update since a majority of design changes end major modifications were performed during refueling outages. In addition, it had no adverse impact on the public health and safety and reduced the regulatory burden on the licensees.

The Commission is now changing the required frequency of maintenance activity evaluations from annually to once per refueling outage.*Evaluation of do.ta collected over the period of e refueling cycle wiU provide a substantially better basis for detecting problems in degraded performance of structures; systems, and components (SSC's) and weakness in maintenance practices. Evaluations conducted on e refueling cycle basis would also consider and integrate data available only during refueling outages with the data available during operations; under the existing requirements this may not occur depending on whether .the annual assessment coincides with the refueling outage. Furthermore, evaluations of date accumulated over the period of e refueling cycle, es opposed to the shorter annual period required by the rule, will provide e more meaningful ho.sis for the recognition and interpretation of trends. The Commission understands that a normal frequency of refueling outage ranges from 15 to 18 months; however, the conditions may vary from plant to plant. In order to ensure that an indefinite period of time does not occur between maintenance evaluations, the Commission is establishing en upper limit of 24 months between the 50-SC-134 maintenance evaluations. This would address those licensees that have extended their refueling cycle beyond 24 months for any reason including numerous short outages or extended shutdown periods. Although the Commission believes that it is generally the case that maintenance evaluations will be more effective if conducted in conjunction with refueling outages, licensees would still have the option of conducting them more frequently. In light of the above discussion, the NRC is changing the requirement for evaluation of the Qverall effectiveness of maintenance activities to be performed . once per refueling cycle provided the interval between evaluations does not exceed 24 months. Summary and Analysis of Public Comments On March 22, 1993 (SB FR915303), the NRC published II notice of the proposed rulemaking for public commenL The comment period expired on May 6, 1993. The NRC received 17 comments on the proposed rule. All of the comments except for one favored the change identified 'in the proposed rule. The comments on the proposed rule came primarily from public utilities with comments also received from a public utilities representative end a private citizen. The NRC has identified and grouped all comments into six broad issues. For each broad issue, the NRC has included a summary of the. comments received and their resolution as follows: 1. Comment. One commenter stated that the proposed change in the rule would unfairly require nuclear plants on an annual refueling cycle to perform twice es many evaluations es plants on a 24-month cycle. *The commenter believes that the NRC should consider a fixed maximum period of 2 years and give the utilities the latitude to menage the timing of the evaluation within that framework. Response. The intent of the propo.sed modification of the maintenance rule is to allow sufficient flexibility in the scheduling of Maintenance Programs evaluations so that the additional information available from the refueling activities could be factored into the evaluation. The refueling cycle hes also been adopted as the basis for FSAR updates. It is recognized that those licensees who refuel more frequently will have to conduct these activities more frequently than others. The Commission believes that this is neither

an undue burden nor one that is outside the control of the licensee to impact by reducing the frequency of refueling.

PART 50

STATEMENTS OF CONSIDERATION

2. Comment. Some commenters stated that, as a result of the verificetlon and yalidation I!rog~am to test the proposed industry gu1delmes, it was determined that several systems are neither risk* significant nor able to be monitored for performance by currently known plant level performance criteria.

Some commenters believe that these systems have no public healt~ or safety significance end that they should be excluded from the scope of the rule end the rule modified ecc~rdingly. Response. The suggestion to change the scope of the rµle to exclude those* systems that hove no 'public health or safety significance or* that hove no current plant level performance criteria is clearly beyond the scope of the rule, and cannot be considered et this time. However, if, es a result of any further verification and validation programs. changes to the rule or regulatory guidance are warranted, the NRC will consider such changes at that time. 3. Commen(. One commenter stated, "one of the clear lessons learned from the recently comple(ed verification and validation program is that the major expense of the rule's implementation will be the detailed documentation.(for

NRC audit purposes) of performance monitoring

* *" ' Response.

The documentation developed by a licensee in response to 10 CFR 50.65 is that level which the licensee determines necessary to support the progra~ developed by the licensee to monitor performance of a structure, system or component. The purpose of this rule modification is not to address the level of documentation required for NRC audit purposes. It is merely to provide more flexibility iµ the timing of Maintenance Program evaluations.

4. Comment. One commenter stated that "The NRC is mesmerized by a suggestion by NUMARC (Nuclear Management and Resources Council), to extend the annual assessment of plant maintenance from 'an annual schedule to a refueling outage schedule." The commenter further stated that the extension does not provide an improvement in Sl!fety end mey help hide maintenance .that was improperly deferred.

Response. As stated earlier, the NRC decided to make the proposed change in the assessment requirement for the following reasons: (1) Evaluation of data collected over the'period of a refueling cycle will provide o substantially bettor basis for detecting problems in degraded performance of SSC's and weakness in maintenance practices; (2) Evaluations conducted on a refueling cycle basis would also consider and integrate data available only during refueling outages with the data available during operations; under the existing requirements this may not occur depending on whether the annual assessment coincides with the refueling outage; and (3) Evaluation of data accumulated over the p!;!riod of a refueling cycle, as opposed to the shorter annual period required by the rule, will provide a more meaningful basis for tho recognition and interpretation of trends. In addition, adjustments to maintenance activities that may be made after such a review and evaluation would be typically performed after o refueling outage. Periodic evaluation of maintenance activities is a time consuming process end with limited data available, the annual evaluations not conducted in conjunction with a refueling wc;iuld not provide for as meaningful a result. These conclusions havo been reached bnsed on tho NRC's independent assessment. Therefore, the commontor incorrectly implies that the NRC simply accepts NUMARC's suggestions without independent review and consideration. Another reason for changing the annual assessment of plant maintenance

concerned a change made by the NRC in August of 1992. As part of the regulatory review to eliminate or revise unnecessary burdensome regulations, the NRC revised the frequency of licensee reporting of the Final Safoty Analysis Reports from annually to once por refueling cycle. This change was made because the NRC believes that the use of a refueling cycle interval provided a more coordinated and cohesive update since the majority of the design changes and modifications were made during refueling outages. This was not a rationale relied upon by NUMARC and further contradicts the *commenter's view that the NRC accepts the suggestions of NUMARC without indepundent consideration.

In summary the Commission disagrees with the commenter's view thot the extension does not improve safety. The change In requirements will improve the quality of assessments by ensuring that each assessment will include a review of all maintenance activities conducted during the refueling cycle including the refueling outoge. 5. Comment. One commenter stated that effective maintenance is an ongoing duty and need and that allowing licensees to put off monitoring the effectiveness of maintenance from annually to 18 to 24 months sends the wrong message that the NRC does not care about safety. SO*SC-135 Response. The NRC agrees that effective maintenance is an ongoing duty and need. The NRC does not &g!ee, however, that tho rule change allows licensees to_ put off monitoring the effectiveness of maintenance. Section 50.65 (a)(1) which is not being changed, requires licensees to monitor the performance or conditions of SSC's against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSC's are capable of fulfilling their intende4 functions. It also requires appropriate corrective action to be token when the performance of the SSC does not meet established goals. The only thing that is being changed is the frequency of the periodic evaluation of the maintenance program. Tho NRC does care about safety and it does not awoe with the commenter that changing the evaluation cycle sends the wrong message to the industry. The NRC believes that this additional flexibility will not reS\Jlt ln any increase in risk to public health ond safety, and In fact, should result in a more effective maintenance and improved plant s..1.fety.

6. Comment. One of the commonters stated that the amendments' maximum time period of24 months would ho restrictive for tho/IC plants planning to Increase their refueling cycle to 24 months. The commenter explained that the Standard Technical Specification, Revision 0, retains the option for performance of surveillance requirements within 1.25 times the interval specified and thus, could extend the refueling outage inter\'al of ploilts with a 24-month refueling cycle by upwards of 6 months. Accordingly, the refueling cycle for these plants would not meet the maximum time period of 24 months allowed by the amendment.

Another commenter stated that this rule could be further improved by the elimination of the requirement for a specific time Interval. Response. The NRC believes that it is necessary to assure that maintenance effectiveness is periodically assessed and that this period is not unacceptably long nor.indefinite. Thus, e balance was necessary between obtaining the Improved reviews associated with assessments conducted during refueling outages and the extended or indefinite periods associated with plants with extended plant cycles or experiencing extended plant shutdown or outages. In weighing this balance, the Commission estaolished an upper limit of 24 months between maintenance evaluations in order to obtain improved evaluations for the maiority of the plants having e frequency of refueling cycle from 15 to 18 months, and yet not allow PART 50

STATEMENTS OF CONSIDERATION maintenance effectiveness to continue without being assessed for periods in excess of 2 years. The NRG does not agree that the rule could be improved further by elimination of the requirement of a specific time interval.

Finding of No Significant Environmental Impact: Availability The Commission has datermined that, under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 Cf'R part 51, that this rule, is not a major Federal action that significantly affects the quality of the human environment end therefore en environme~tal impact statement is not required. The final amendment does not require any change to nuclear power plant design or require any modifications to a plant. Nor does the rule change the scope of the maintenance rule or affect the nature of the activities to be performed, e.g., monitoring. corrective action, end assessments of compJiance. The final rule change only extends the time period for performing evaluations of the effectiveness of licensees' maintenance program from at least once a year to at least once every refueling cycle, not to exceed 24 months. The extension should not result in any significant or discernible reduction In the effecUveness of a licensee's maintenance program: rather the change will increase the meaningfulness and quality of the maintenance evaluations. For these reasons, the Commission finds .that the final amendment will not result in any significant increase in either the probability of occummce of an accident or the consequences of an accident and therefore concludes that there will be no significant effect on the environment es a result of the amendment. The environmental assessment is available for inspection at the NRG Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environmental assessment are available from Josep;h J. Mate, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone: (301) 492-3795. Paperwork Reduction Act Statement This final rule amends the Information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Management and Budget, approval number 3150-0011.

Because the rule relaxes existing requirements related to the assessment of maintenance activities,the public burden for this collection of information is expected to be reduced by 150 hours per licensee.

This reduction Includes the time required for reviewing Instructions. searching existing doto sources, gathering and maintaining the data needed.and completing end reviewing the collection of information. Send comments regarding the estimated burden reduction or any other aspect of this oollection of information, Including suggestions for reducing this burden. to the Information and Records Management Branch (MNBB-7714), U.S. Nuclear Regulatory Commission. Washington, DC 20555; and to the Desk Officer, Office of Information end Regulatory Affairs, NEOB-3019, (3150-0011), Office of Management and Budget, Washington, DC 20503. Regulatory Analysis The Nuclear Regulatory Commission hes considered the costs and benefits of the final rule. With rasped to benefits. the amendment will allow those licensees who choose to exercise the option to perform evaluations of their maintenance program in conjunction with refueling outages but no less frequently than evmy 24 months. The Commission beUevM that thio additional 0exJ1Jillty will not result In any Increase In risk to the public health and safety, and may result in a more effective maintenance and improved plant safety. . Under tlie rule, the frequency of periodic assessments would change from annually to et least once per refueling.cycle but not to exceed 24 months. Because most refueling outages normally occur In the 15* to 18-month range, the time between periodic assessments assuming a *16-month average would be Increased by about 33 percent. Therefore, the licensee staff hours to accomplish a periodic assessment under the proposed rule would be reduced from approximately . 460 staff hours to about 310 staff hours per plant. This would save the licensee . approximately 150 staff hours per plant. There are-no additional changes in costs lo be incurred by the NRC. The foregoing constitutes the regulatory analysis for this final rule. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, (5 U.S.C. 605(b)), the Nuclear Regulatory Commission

certifies that, this rule will not have a significant economic impact on a substanUal number of small entitles.

This rule affects only the operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of 50-SC-136 "small entities" as set forth In the Regulatory Flexibility Act or the Small Business Size Standards set out In the regulations issued by the Small Business Administration at 13 CFR part 121. Backlit Analysis The NRG has determined that the backfit rule, 10 CFR 50.109, does not apply to this rule and, therefore, that a backfit analysis is not required for this final rule because this amendment does Involve any provisions which would Impose backfits as determined in 10 CFR 50.109. ** List of Subjects in 10 CFR. Part 50 Antitrust, Classified Information, Criminal penalties, Fire protection, Incorporation by reference,. Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeplng requirements. For reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as , amended, and 5 U.S.C. 552, 553, the NRG ls adopting the following amendment.to 10 CFR part 50. 58 FR39092 Published 7/21/93 10 CFR Part 50 RIN 3150-ADBO

Training and Qualification of Nuclear Power Plant Personnel Correction In rule document 93-9651 beginning on page 21904 in the issue of Monday, April 26, 1993, make the followin{
corrections:

§ 50.120 [Corrected] On page 21912, in the first column, in § 50.120(b)(1). beginning in the second and sixth lines, "(October 25, 1993, publication]" should read "November 22, 1993" each time it appears.* PART 50

STATEMENTS OF CONSIDERATION 58 FR 45243 Published 8/27/93 Effective 9/27 /93 10 CFR Parts 50 and 54
RIN 3150 -AE63 FSAR Update Submittals AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule ..

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations on power reactor safety in order to consistently apply the requirement that n~clear power pl~t licensees submit final safety analysis report (FSAR) updates annually or six months after each refueling outage. These amendments eliminate confusion regarding two references to an existing reporting requirement.

The final rule. does not require additional reporting requirements. EFFECTIVE DATE: September 27, 1993. FOR FURTHER INFORMATION CONTACT: Claudia M. Craig, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 504-1281. SUPPLEMENTARY INFORMATION: Background In February 1993, the Commission approved the establishment of a regulatory review*group (RRG) to conduct a comprehensive and disciplined revie~ of p*ower reactor regulations end related NRC pr?cesses, programs, and practices for their implementation. J'he RRG found two areas in the regulations that may cause confusion regarding a recent amendment to another section of the regulations. On August 31, 1992, the Commission amended 10 CFR 50.71(e) to allow nuclear power reactor licensees to submit FSAR updates either annually or 6 months after each refueling outage. The RRG discovered that 10 CFR 50.54(a)(3) and 10 CFR 54.37(b) still referenced the previous requirement for annual FSAR submittals. This conflict may confuse licensees in determining how often quality assurance pro~am changes and FSAR updates for license renewal should ~e submitted. Description The amendments delete the references to the annual submittal of updates in 10 CFR 50.54(a)(3) .and 10 CFR 54.37(b). The amended sections reference the regulation, 10 CFR 50.71(e), not the. specific requirements of the regulation. Licensees with a QA program description that is common to multiple units or several sites may submit Reduction Act of 1980 (44 U.S.C. 3051 et seq.). Existing requirements were approved by the Office of Management and Budget approval numbers 3150-0011 and 315CHl155. changes to the common quality , Regulatory Analysis assurance (QA) pro~am description that The Commission has prepared a do not reduce commitmen~ annually or regulatory analysis on this final 6 months after each.refueling outage at regulation. The analysis examines the only one of the sites if the interval costs and benefits of the alternatives . between submittals does not exceed 24 considered by the Commission. The months end all applicable dockets ere analysis is available for inspection in referenced. This would allow licensees the NRC Public Document Room, 2120 with multiple plants to tie the submittal L Street NW. (Lower Level), . of changes to the common QA program Washington, DC. Single copies of the to the refueling outage schedule of only analysis may be obtained from Claudia one plant end would eliminate the need M. Craig, U.S. Nuclear Regulatory for a separate submittal for each plant. Commission, Washington, DC 20555, The amendment will eliminat?~e (301) 504-1281. confusion cau~ed ~y the confl.1cting R ul t Fl 'bility Certification requirements m different sections of the eg a o~ exi regulations. As reqmred by the Regulatory . Flexibility Act of 1980, 5 U.S.C. 605(b), Summary of Pubbc Comments the Commission certifies that this On May 14, 1993 (58 FR 28523), the regulation will not have a significant NRC published a proposed rule that economic impact on a substantial would delete the references to the number of small entities. This annual submittal of updates in 10 CFR regulation affects only the licensing and 50.54(a)(3) and 10 CFR 54.37(b). The operation of nuclear*power plants. The comment period ended on June 14, companies that own these plants do not 1993, and the NRC received five letters fall within the scope of the definition of of public comment on the proposed "small entities" as given* in the . rules. Four commenters fully supported Regulatory Flexibility Act, or the Small the proposed changes; one commenter Business Size Standards promulgated in submitted s.tatements f~r § 50.54(a)(3) to the regulations issued.by the Small further clanfy the reqwrem~nts and Business Administration at 13 CFR part recommended that NRC revise 10 CFR 121 54.37(c) to duplicate the reporting

frequency of§ 50.59(b)(2);

one Backfit Analysis commenter also recommended that NRC The NRC has determined that the consider extending the reporting backfit rule, 10 CFR 50.109, does not frequency associated with 10 CFR apply to this final rule. The rule affects 50.59(b)(2) to be consistent with the recordkeeping end reporting FSAR update submittal. The requirements which have been deemed Commission agrees with the proposed not subject to the batjcfit rule and the statements for 10 CFR 50.54(a)(3) end changes are voluntary relaxations of has incorporated the statements into the requirements which are not being final rule. All other sections of the final imposed upon licensl!es. Therefore, a rulemaking remain unchanged. Copies backfit analysis is not required for this of those letters and the NRC staff final rule because these amendments do response to the*public comments ere not involve any-provisions that would available for public inspection an.d impose backfits as defined in 10 CFR copying for a fee at the NRC Public 50,109(a)(l).

Document Room at 2120 L Street NW.* (Lower Level), Washington, DC. Environmental Impact: Categorical Exclusion The NRC has determined that this final rule is the type of action described as a categorical exclusion in 10 CFR 51.22(c)(3) (i) and (iii). Therefore, neither en environmental impact statement nor an environmental assessment has been prepared for this final rule. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork 50-SC-137 List of Subjects 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants end reactors, Radiation protection, Reactor siting criteria, Reporting end recordkeeping requirements.

1 O CFR Part 54 Administrative practice and procedure, Age-related degradation, Backfitting, Classified information, PART 50

STATEMENTS OF CONSIDERATION Criminal penalties, Environmental protection, Incorporation by reference, Nuclear power plants and reactors, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is adopting the following amendments to 10 CFR parts 50 and 54. 58 FR52406 Published 10/8/93 Effective 11/8/93 Whistleblower Protection for Employees of NRG-Licensed Activities See Part 19 Statements of Consideration 58 FR54646 Published 10/22/93 Whistleblower Protection for Employees of NRG-Licensed Activities: Correction See Part 19 Statements of Consideration 58 FR 67657 Published 12/22/93 Effective 1 /1 /94 Standards for Protection Against Radiation; Removal of Expired Material See Part 20 Statements of Consideration 58 FR 68726 Published 12/29/93 Effective 1 /28/94 Self-Guarantee as an Additional Financial Assurance Mechanism See Part 30 Statements of Consideration 59 FR 1618 Published 1112/94 Effective 1 /28/94 Self-Guarantee as an Additional Financial Assurance Mechanism; Correction See Part 30 Statements of Consideration 59 FR 5519 Published 2/7 /94 Effective 2/7 /94 Minor Clarifying Amendments See Part 21 Statements of Consideration 59 FR 10267 Published 3/4/94 Effective 4/4/94 10 CFR Part 50 RIN 3150-AE46 Notification of Spent Fuel Management

and Funding Plans by Licensees of Prematurely Shut Down Power . Reactors AGENCY: Nuclear Regulatory Commission. . ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC).is amending its regulations to clarify the timing of notification to the NRC of spent fuel management and funding plans by licensees of those nuclear power reactors that have been shut down before the expected end of their operating lives. The final rule requires that a licensee submit such notification either within 2 years after permanently ceasing operation of its licensed power reactor or no later than 5 years before the reactor operating license expires, whichever event occurs first. Licensees of nuclear power reactors that have already permanently ceased operation by the effective date of this rule are required to submit such notification within 2 years after the effective date of this rule. EFFECTIVE DATE: April 4, 1994. FOR FURTHER INFORMATION CONTACT: Robert Wood, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, telephone (301) 504-1255.

50-SC-138 SUPPLEMENTARY INFORMATION: Background On June 30, 1993, the NRC published in the Federal Register a notice of proposed ruleniaking to clarify the timing of notification to the NRC of spent fuel management and funding plans by licensees of those nuclear power. reactors that have been shut down prematurely (58 FR 34947). 1. Comments Received The NRC received four comments on the proposed rule. Three of the four comments crone from licensees or their representatives and supported the rule as proposed. These commenters agreed with the NRC assessment that the proposed rule is administrative in nature and would produce consistency with the decommissioning rule. However, each of the three recommended that the rule amendments

should apply only prospectively; that is, the rule. should not apply to licensees whose power reactors have already permanently ceased operating.

The commenters requested that the NRC allow licensees of these plants to submit spent fliel management funding plans on a case-by-case schedule. One commenter recommended that the NRC add a statement to this effect to the final rule . A fourth commenter supported the concept of requiring the submittal of spent fuel management and funding plans soon after permanent shutdown, but recommended that licensees be required, to submit these plans within 60 days after permanent shutdown. The three commenters representing licensees also supported the NRC intent . to initiate rulemaking on including spent fuel costs as part of decommissioning costs only after careful consideration of the database that the NRC is developing in this area. In a related area, one of these commenters noted that the NRC currently has regulations in place in 10 CFR part 72 to ensure a licensee's financial qualifications for the safe construction, operation, and decommissioning of an independent spent fuel storage installation (ISFSI). The fourth commenter supported rulemaking. on funding assurance for spent fuel storage costs that would be similar to, but separate from, decommissioning costs. 2. NRC Response to Comments The NRC responds as follows to the issues raised by the commenters: (1) The role should only apply prospectively. PART 50

STATEMENTS OF CONSIDERATION NRG response:

The NRC disagrees that this rule should not apply to licensees of plants that have already permanently ceased ~perating. This rule should be consistent:with the provisions of 10 CFR 50.82(a), which requires all power plant licensees to submit decommi_ssioning plans no later than 2 years after permanently*ceasing operations regardless of how long the plant operated. The ~C recently amended 10 CFR 50.82(a) to allow the collection period of any shortfall of decommissioning funds to be determined on a case-by-case basis for plants that had been, shut down prematurely (57 FR 30383, July 9, 1992). However, even licensees of these plants must subm~t their decommissioning plans' within the 2-year time* frame, notwithstanding the collection period ultimately adopted. : To maintain consistency, the NRC believes that the 2-year limit should be applied to plants already shut down. However; to assure that the NRC does not impose unnecessary burdens on these licensees, the final rule has been modified to allow these licensees 2 years from the effective date of the rule to submit their spent fuel management and. funding plans.*. (2) Submittal of spent fuel management and funding plans should be required within 60 days of permanent shutdown of the fac.ility, rather than with}n 2 years. NRG Response: The NRC disagrees with this comment. Sixty days is too short a period in which to develop a meaningful spent fuel management and funding plan. Because licensees will normally develop these plans in conjunction with their decommissioning plans, the NRC should maintain consistency by req.uiring the same 2-year limit for both spent fuel management and funding plans and the overall decommissioning plan, which includes decommissioning funding. (3) Costs associated with the construction, operation, and decommissioning of JSFS1s are already assured by provisions in 10 CFRPart 72. *

NRG Response:

The NRC agrees that part-72 contains provisions to ensure * .In practice, licensees, of most of the nuclear power plants that have already permanently shut down have developed plans for the management and funding of the disposition of spent fuel at their sites .. For example, Fort St. Vrain has either shipped . spent fuel offsite to DOlj or moved It to an ISFSI onsite. Shoreham *1s shipping Its fuel to Limerick. Yankee-Rowe and Rancho Seco have developed plans for ODsite storage facilities. Humboldt Bay end LeCrosse are maintainiDg fuel hi their spent fuel. pools. Dresden 1, San Onofre 1, and Indian Point 1 are maintaining fuel ln their spent fuel

pools or.In pools of other unita still operating et the site. Peach Bottom 1 has no fuel _onsite. that licensees ha,ve adequate funds to construct, operate, and decommission ISFSis. Spent fuel management and funding plans submitted in compliance with the amended§ 50.54(bb) need not cover spent fuel wh1le it is being stored in an ISFSI in compliance with part 72. The NRC will consider whether these provisions are adequate when it evaluates whether it is necessary to include spent fuel management and funding as part of decommissioning costs. Finding of No Significant Environmental Impact: Availability This final rule clarifies the timing of the submittal of plans for managing and providing funding for managing all irradiated fuel for those licensees whose power reactors are shut down prematurely.

This action is required to coordinate the submittal of spent fuel management and funding plans with the submittal of decommissioning plans for prematurely shut down reactors. Because management and funding of spent fuel can have a significant impact on the method and timing of decommissioning, licensees should submit their plans for spent fuel management and funding to be consistent with the timing provisions for decommissioning plans in § 50.82(a) (i.e., no later than 2 years after permanent shutdown). Neither this action nor the alternative of maintaining the existing rule would significantly affect the environment. Changes in the timing of the submittal of spent fuel management and funding for prematurely shut down power reactors would not alter the effect on the environment of the licensed activities considered in either the final spent fuel disposition rule (49 FR 34689; August 31, 1984) or the final decommissioning rule (53 FR 24018; June 27, 1988) as analyzed in the Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities (NUREG--0586; August 1988). The alternative to this action would not significantly affect the environment. Therefore, the Commission has determined, under the National Environ.mental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 CFR part 51, that this rule will not be a major Federal action significantly affecting the quality of the human environment and, therefore, an environmental impact statement is not required. No other agencies or persons were contacted for this action, and no other documents related to the environmental impact of this action exist. The foregoing constitutes the environmental 50-SC-139 assessment and finding of no significant impact for this final rule. Paperwork Reduction Act Statement This* final rule does-not contain a new* or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget, approval number (3150-0011). Regulatory Analysis On August 31, 1984, the NRC . published a final rule, "Requirements fc;>r Licensee Actions Regarding the

Disposition of Spent Fuel Upon Expiration of Reactor Operating Licensees." (49 FR 34689). As part of this rule, the NRC required power reactor licensees to submit for NRC review and approval, no later than 5 years before expiration of the reactor operating license, their plans for managing spent fuelat their site until title to the spent fuel is transferred to the Department of Energy (DOE). These plans are to include plans for funding of spent fuel management before transfer to DOE. On June 27, 1988, the Commission promulgated its final decommissioning rule (53 FR 24019). Section 50,82 of this rule provides that licensees of all power reactors that permanently cease operation after July 27, 1988, including those that shut down prematurely, must apply to *the NRC to decommission their facilities within 2 years following permanent cessation of operations.

Section 50.82(b)(l)(iii) further provides that the proposed decommissioning plan*submitted by the licensee should consider such factors.as the "unavailability of waste disposal capacity and other site-specific factors affecting the licensee's capability to carry out decommissioning safely * * *." The Commission requires licensees to submit decommissioning plans iii a timely manner after they permanently cease operations at their facilities. The NRC's regulations recognize that a licensee's ability to plan prop!;!rly and safely for decommissioning depends on a licensee's ability to manage and dispose of its spent fuel. Thus, the timing of requirements for submittal of plans for spent fuel management and storage should be consistent with the timing for submittal of decommissioning plans, including those for power reactors that have been shut down prematurely. Therefore, the NRC is amending 10 CFR 50.54(bb) to require each power reactor licensee to notify the NRC of its program to manage and provide _funding for PART 50

STATEMENTS OF CONSIDERATION management of the irradiated fuel at its reactor either within 2 years after the licensee permanently ceases operation
or its reactor or no later than 5 years before its reactor operating license expires, whichever occurs first. Licensees of nuclear power reactors that have already permanently ceased operations by the effective date of this rule are required to submit such notification within 2 years after the effective date of this rule. Although the timing of preparation and submittal of.plans for management and funding of spent fuel would be formally advanced for licensees that shut down their power reactors prematurely, these licensees typically would have already evaluated spent fuel management and funding "issues before submitting decommissioning plans . required under 10 CFR 50.82. This rule merely makes 10 CFR 50.54(bb) submittal schedular requirements consistent with 10 CFR 50.82. Thus, there should be no.substantive impact on power reactor licensees.
This final rule would not create substantial costs for other licensees.

This final rule also will not significantly affect State and local governments and geographical regions, or the environment, or create substantial costs to the NRC or other Federal agencies. The foregoing discussion constitutes the . regulatory analysis for this final rule.

Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission certifies that this final rule will not have a significant impact upon a substantial number of small entities.

The rule will potentially affect approximately 115 nuclear power reactor operating licenses. Nuclear

power plant licensees do not fall within the definition of small businesses as defined in section 3 of the Small Business Act, 15 U.S.C. 632, the Small Business Size Standards of the Small Business Administrator (13 CFR part 121), or the Commission's Size Standards (56 FR 56671, November 6, 1991). Backfit Analysis The NRC has determined that this final rule does not impose a backfit as defined in 10 CFR 50.109(a)(l).

Therefore, a backfit analysis is not required for this final rule. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminal penalty, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons given in the preamble and l!llder the authority of the Atomic Energy A.ct of 1954, as amended, the Energy Reorganization Act of 1974; as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendmentto 10 CFR part 50. 59 FR 14085 Published 3/25/94 Effective 5/31/94 NRG Operations Center Commercial Telephone Number Change See Part 20 Statements of Consideration 59 FR 14087 Published 3/25/94 Effective 6/23/94 10 CFR Part 50 RIN: 3150-AD40 Emergency Planning and Preparedness Exercise Requirements for Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its emergency planning regulations in order to update the Commission's emergency planning exercise requirements for nuclear power plants and clarify ambiguities that have surfaced in the implementation of the regulations.

These amendments also make the NRC regulations consistent with FEMA regulations. EFFECTIVE DATE: June 23, 1994. FOR FURTHER INFORMATION CONTACT: Michael T. Jamgochian, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301-492-3918). 50-SC-140 SUPPLEMENTARY INliORMATION: Background On August 19, 1980 (45 FR 55402), the NRC published a final rule that revised its emergency planning regulations. The final rule became effective on November 3, 1980. On July 6, 1984 {49 FR 27733), the NRC amended its emergency planning regulations to relax the frequency of participation by State and local governmental authorities in emergency preparedness exercises at nuclear power reactor sites. The amendments were based on the NRC's experience gained in observing and evaluating emergency preparedness exercises since 1980. Further experience has shown that the language setting forth the requirements in 10 CFR piu-t 50, Appendix E, Section IV.F.3 concerning full or partial participation by State or local governments in the bieQDial (offsite) exercise is wmecessarily complicated. The NRC published a notice of proposed rulemaking in the Federal Register on June 28, 1993 (58 FR 34539). Public comments were requosted by September 13, 1993. The proposed rule did not seek to change the requirements set forth in Appendix E, Section IV.F.3 (a), (b), and (d) but to clarify and simplify the text of the regulation. Offsite authority responsibilities remain unchaDged. Under the proposed rule the offsite plans for each site were to be exercised biennially with full participation by each offsite authority having a role under the plan. Further, where the offsite authority has a role under the plan for more than one site, it would be required to participate in one exercise fully every two years and partially participate in other offsite plan exercises in this period. The only amended requirements were those set forth in Appendix E, Section IV.F.3(e) where the interval for an ingestion exposure pathway exercise was changed from 5 to 6 years, and Appendix E, Section IV.F.3(c) where the requirement that all States within the plume exposure pathway emergency planning zone (EPZ) for a given site fully participate in an offsile exercise for that site at least once every 7 years was deleted. Public Comments A total of 12 comment letters were received, of which 5 were from utilities, 6 were from State emergency management agencies and one from NUMARC. All commenters generally agreed with the proposed rulemaking except for one Stale agency. PART 50

STATEMENTS OF CONSIDERATION Comment: The one commentor that opposed the rule change noted that, We do not believe, however, the. NRC has substantiated its claim that the seven-yell!'

return requirement Is ~nnecessary. SimilBr arguments have surfaced in previous emergency planning is~ues, and our response is the same: The high level of industry sensitivity to emergency preparedness is a direct result of comprehensive requirements for emergency preparedness programs end exercises. Elimination of those requirements runs the risk of returning the industry to pre* TMI levels of preparedness. Response: The Coin.mission does not agree that deleting the 7 year return frequency "* *

runs the risk of returning the industry to pre-TMI levels of preparedness." The Commission is confident that this will not occur because the Commi~ion has foWld that multi-sites states, when not fully participating in an exercise at a specific site will usually piu:tially participate at a significant level of activity every 2 years et that specific site in order to support the participation of the appropriate local governments.

The Commission has found that this level of exercise particlpati(?n provides adequate ' emergency response training for State and local govemme,nts. The Commission believes that this rulemakiug does nqt have an adverse impact on public h~lth and safety because State emergency response personnel continuously respond to actual emergencies

and experience hes shown that states through a combination of full and partial participation exercises maintain an adequate level of . response capability.

A formal requirement for a State to return to a specific site every 7 years to participate in an exercise has proven to be unnecessary. Nonetheless, nothing prevents a State from returning to a specific site to participate in an exercise whenever it deems warranted. Comment: Several comments suggested additional clarification to the emergency planning regulations. Response: Although the Commission always appreciates: suggestions on clarifying its reguhi-tions, the Commission at this time believes that all of the suggested changes would be inappropriate to include in this rulemaking proceeding because the suggested revisions are beyond the scope of this ruleniaking. Comment: Several commenters noted that the proposed wording for the ingestions pathway exercise was not consistent with the FEMA requirement and could be interpreted differently than intended. They suggested the following requirement, "A State should fully participate in the ingestion pathway portion of exercises at least once every six years. In States with more than one site, the State should rotate this participation from site to site." Response: The Commission agrees with the suggested wording and has incorporated this comment in the final rule.

Discussion The Commission finds that the current regulation has resulted in a relatively complicated description of the requirements for exercise participation by State and local governments who have offsite planning responsibility for more than one nuclear power plant. This final rule simplifies and clarifies this requirement.

In addition, Appendix E is revised to reflect that the interval for an Ingestion exposure pathway exercise be changed from at least once every 5 years to at least once every 6 years (FEMA's ingestion pathway exercise requirement is at least once every 6 years). The change in the interval would match the biennial frequency required for exercises of offsite plans. Further, Appendix E is also revised to eliminate the 7 year return frequency requirement because it bas proven to be unnecessary to achieve the underlying purpose of the rule as well as being burdensome to states which are within the plume exposure pathway for multiple sites (FEMA does not have a return frequency requirement). Both changes assure compatibility with FEMA requirements and thus avoid confusion among licensees and State governments. Notwithstanding elimination of the 7 year return frequency requirement, the Commission.believes that offsite authorities should rotate their full participation in exercises among sites if they are within the plume exposure

pathway for more than one site. The Commission codified the 7 year return frequency in the July 6, 1984 (49 FR 27733), amendment to the emergency planning regulations.

This amendment provides that at least once every 7 years, all States within the plume exposure pathway EPZ of a given site must fully participate in an offsite exercise for that site. In doing so, the Commission noted that "the final rule is not totally consistent with FEMA*s final regulation (44 CFR part 350). This inconsistency lies in the area of return frequency of multiple-site states as previously discussed. The FEMA position on return frequency is a significant departure from the NRC's proposed regulation ofJuly 21, 1983 (48 FR 33307). The Commission oolieves that more study is needed before 50-SC-141 deletion of the return frequency requirement can be justified." The Commission now believes that sufficient experience has been gained in the observation and evaluation of emergency preparedness exercises at nuclear power reactor sites to conclude that the 7 year return frequency should be deleted.

The Commission has found that multi-site States, when not fully * . participafing in an exercise at a specific site will usually partially participate at a significant level of activity every 2 years at that specific site in order to support the participation of the appropriate local governments.

The Commission has found that this level of exercise participation provides adequate emergency response training for State and local governments. Additionally, a provision still exists in the regulation which permits State or local government participation in any licensee's drills or exercises. A State or local government may consider its response capability to be less than optimal because of an unusually large personnel turnover or because there have been limited responses to real emergencies in the community. The regulation still requires the licensees to provide for State or

local government participation if they indicate such a desire. This final revision does not have any adverse impact on public health and safety because State emergency response personnel continuously respond to actual emergencies and experience has shown that states through a combination of full and partial participation exercises maintain an adequate level of response capability.

A formal requirement for a State to return to a specific site every 7 years to participate in an exercise has proven to be *unnecessary.This rulemaking deletes that unnecessary, unwarranted and burdensome requirement. Nonetheless, nothing prevents a State from returning to a specific site to participate in an exercise whenever lt deems warranted.

Lastly, this revision deletes past due dates (see section F(2) (a)) because they are now meaningless. . FEMA concurs with .the amendments in this rulemalcing.

Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 CFR part 51, that this rule is not a major Federal action significantly

affectirig the quality of the human environment; and therefore, an et: vironmental impact statement is not PART 50
STATEMENTS OF CONSIDERATION required.

This regulation updates and clarifies the emergency planning regulations relating to exercises. It does not involve any modification to any plant or revise the need for or the standards for emergency plans, and there is no adverse effect on the quality of the environment. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW (Lower Level), Washington, DC 20036. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval number 3150-0011. Regulatory Analysis The Commission has prepared a regulatory analysis on this regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The analysis is available for inspection in the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC 20036. Single copies of the analysis may be obtained from Michael Jamgochian, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone: (301) 492-3918. Regulatory Flexibility Act Certification The regulation does not have a significant impact on a substantial number of small entities. The final rule updates and clarifies ambiguities in the emergency planning regulations relating to exercises. Nuclear power plant licensees do not fall within the definition of small business in Section 3 of the Small Business Act, 15 U.S.C. 632, the Small Business Size Standards of the Small Business Administration in 13 CFR part 121, or the Commission's. Size Standards published at 56 FR 56671 (November 6, 1991). Therefore, in accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission hereby certifies that this final rule, will not have a significant economic impact on a substantial number of small entities and that, therefore, a regulatory flexibility analysis need not be prepared. Backfit Analysis This regulation does not impose any new requirements on production or utilization facilities. The regulation deletes the requirement that all states within the plume exposure pathway EPZ for a given site fully participate in an offsite exercise for that specific site at least every 7 years. It also relaxes the requirement to perform an ingestion exposure pathway exercise from every 5 years to every 6 years. These changes would permit, but do not require, licensees to change their emergency plans and procedures. Therefore, these changes are not considered backfits as defined in 10 CFR 50.109 (a)(l). List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection; Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons set out in the preamble, and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendments to 10 CFR part 50. 59 FR 50688 Published 10/5/94 Effective 10/5/94 NRG Library; Address Change See Part 35 Statements of Consideration 50-SC-142 60 FR 13615 Published 3/14/95 Effective 4/13/95 10 CFR Parts 50, 55, and 73 RIN 3150-AF18 Reduction of Reporting Requirements Imposed on NRC Licensees AGENCY: Nuclear Regulatory Commission. ACTION~ Finai rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations to reduce reporting requirements cUITently imposed on water-cooled*

nuclear power reactor, research and test reactor, and nuclear material licensees. This rule reduces the regulatory burden on NRC licensees; and partially implements a recent NRC initiative to revise or eliminate duplicative or unnecessary reporting requiremerits. The amendments will: . Eliminate the current requirement for licensees to submit summary reports of *containment leakage rate tests to the NRC (10 CFR Part SO-Appendix J), but preserve the*reqtiirements:in §§ 50.72 and 50.73 under which licensees currently report any instances ,of leakage exceeding authorized limits in the technical specifications of the*license; revise 19 CFR 55.25 to refer licensees to a similar reporting requirement in 10 CFR 50.74(c) and require notification of operator incapacity only in case of permanent disability or illness; and eliminate the requirement for quarterly submittal of safeguards event logs presently contained in 10 CFR 73.71(c)(2) and Appendix G to Part 73. EFFECTIVE DATE: April 13.1995. FOR FURTHER INFORMATION CONTACT: Naicm S. Tanious, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone (301) 415-6103. SUPPLEMENTARY INFORMATION: Background On January 7, 1994, the Executive Director for Operations (EDD) sent to the Commission SECY-94-003, "Plan for Implementing Regulatory Review Group Recommendations.'.' The Commission .approved these recommendations for reducing regulatory burden on its Hcensees. This PART 50

STATEMENTS OF CONSIDERATION -final rule is one of several rulemakings and other regulatory actions currently being developed by the NRC staff to implement the Regulatory Review Group recommendations to eliminntc duplicative or unnecessary reporting requirements.

The NRC believes that this action will reduce the regulatory burden on NRC licensees without causing adverse effects on the protection of public health and safety.

On November 2: 1994 (59 FR 54843), the NRC published the notice of proposed rulemaking that reporting requirements on licensees tinder Parts 50, 55, and 73. Specifically, the proposed amendments were intended to: (1) Eliminate the current requirement for licensees to submit summary reports of containment leakage rate tests to the NRC (10 CFR part 50-appendix J), but preserve the requirements in§§ 50.72 and 50.73 under which licensees currentty report any instances of leakage exceeding authorized limits in the technical specifications of the license; (2) revise 10 CFR 55.25 to refer licensees to a similar reporting requirement in 10 CFR 50.74(c) and require notification of operator incapacity only in case of permanent disability or illness; and (3) eliminate the requirement for quarterly submittal of safeguards event logs presently contained in 10 CFR 73.71(c)(2) and Appendix G to Part 73. The public comment period expired December 19, 1994._ Analysis of Public Comments on the Proposed Rule The NRC received seven comments:

one from Nuclear Energy Institute (NE!), an organization that represents the nuclear power industry, five from the nuclear power industry, and one from Ohio Citizens for Responsible Energy, Inc. (OCRE). The comments from NEI and the nuclear power industry are supportive of the proposed rule to reduce the reporting requirements. OCRE opposes the proposed rule.* However, all coinmenters believe that elimination 9f these reports will not adversely impact public health and safety. The following section addresses

the public comments received and provides NRC's response to them. Of the six comments received which favor the proposed rule, several of those endorsing the nile pointed out that the proposed -changes eliminate
unnecessary or redundant requirements and conserve both NRG and licensee resources.

Two of the commenters felt that the NRC sliould assess additional reporting requirements to determine whether they can be eliminated or reduced.in frequency. As discussed in the background section of this rulemaking, the NRC has underway several regulatory activities to implement the Regulatory Review Group's recommendations to eliminat,i duplicative or unnecessary reporting requirements. This rulemaking is limited to tlm requirements set out in the proposed rulemaking. Licensees do not Need to Assemble ihe Summary Report One commenter from the nuclear power industry states that the requirement to generate but not submit a summary report for the containment leakage tests provides no additional , benefit and is an unnecessary burden since the summary report contains data readily available from other sources. The commenter suggests that the requirement to generate the summary report be eliminated. The NRC disagrees. The NRC believes that the results of containment leakage tests, the licensee analysis verifying the acceptability of the results, as well as any necessary interpretations of the results, is necessary information which might not be documented absent this documentation requirement. Furthermore, the assembly of a summary report will provide access by NRC inspectors and auditors to this information in a more timely fashion. Public.Participation in the NRG Regulatory Process Will Diminish _OCRE opposes the proposed rule because it believes that adoption of the rule will diminish the public"s access to information. OCRE states that the public's health and safoty is not the only factor to consider when NRC proposes to eliminate some licensee reports. Access to these reports, OCRE states, is vital for effective public participation in the regulatory process. To that end, OCRE has filed a petition for rulemaking with the NRC (59 FR 30308, June 13, 1994). The purpose of the petition is to establish public to-know provisions which would ensure public access to licensee-held information. ln each case where the NRC considers eliminating a reporting requirement, the NRC first considers the public health and safety impact of the proposed elimination. If there is no direct impact on public health and safety, the NRC also considers the reduced administrative burden on the licensee and the extent to which the proposed elimination will deprive the public nf important health and safety infomrntion. OCRE's comments have raised the . generic issue of the incremental and cumulative P.ffect of this and similar 50-SC-143 rulemakings in depriving the public of access to licensee information that.was previously avaiiable from the NRC. In that regard, OCRE has directly presented

this issue to the Commission through its petition for rulemaking referenced above and the NRC finds that this generic issue is better addressed in the context of that petition, rather .than in individual rulemakings such as this one. The NRC also finds that the effect of this rulemaking will be to reduce the administrative burden on licensees and that the loss of the information in this particular case will not adversely affect the public interest in access to information regarding adequate protection of the public health and safety. Having considered all comments received and other input, the NRC has determined that the following final rule should be promulgated.

Written Reports This final rule would not require additional written reports. On. the contrary, under this final rule, reporting will be reduced for all licensees under 10 CFR Parts 50. 55, and 73. Criminal Penalties For purposes of Section 223 of the Atomic Energy Act of 1954, as amended, relating to willful violations of requirements notice is hereby given that these amendments are being adopted and promulgated pursuant to Sections 161 b, 161i, or 1610 of the Act. Environmental Impact: Categorical Exclusion The NRC has determined that this final rule is the type of action described in the categorical exclusion 10 CFR 51.22(c)(3)(iii). Therefore, neither an environmental impact statement nor an envirnnmental assessment has been prepared for this regulation. Paperwork Reduction Act Statement This final rule amends information col-lection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Management and Budget, approval numbers 3150-0011, -0018, and-0002. Because the rule will relax existing information collection requirements, the annual public burden for this collection of information is expected to be reduced by approximately 20 hours per licensee. This reduction includes the time required for reviewing instructions. searching existing data sources, gathering and maintaining the data needed and completing and reviewing PART 50

STATEMENTS OF CONSIDERATION the collection of information.

Send comments regarding the estimated burden reduction or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011.

-0018, -0002), Office of Management and Budget, Washington, DC 20503. Regulatory Analysis The Commission has prepared a regulatory analysis on this final rule. The analysis examines the costs and benefits of the alternatives considered by the Commission. The Commission requested public comment on the draft regulatory analysis, but no comments were received. Therefore, no changes to the draft regulatory analysis have been -made. The draft regulatory analysis is. adopted as the final regulatory analysis without change. The analysis is .--* available for inspection in the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington.DC._ Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, (5 U.S.C. 605(b)), the Commission certifies that this rule will not have a significant economic impact on a substantial number of small entities. This final rule . affects the nuclear power reactors, research and test reactors, and some material licensees. The companies and organizations that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act of the size standards established by the NRC (56 FR 56671; November 6, 1991). Backfit Analysis The NRC has determined that the backfit rule,. 10 CFR 50.109, does not apply to this final rule because these amendments do not involve any -provisions which would impose backfits . on licensees as defined in § 50_.109(a)(l). In addition, information collection and reporting requirements are not subject to the backfit rule. List of Subjects 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire *protection, Intergovernmental relations, Nuclear power plants and*reactors, Radiation protection, Reactor sitin'g'criteria, Reporting and recordkeepirig requirements.

10 CFR Part 55 Criminal penalties, Manpower training programs, Nuclear power plants and reactors, Reporting and. recordkeeping req~irements.

10 CFR Part 73 Criminal penalties, Hazardous materials transportation, Export, Import, Nuclear materials, Nuclear power plants and reactors, Reporting and recordkeeping requirements, Security measures. 60 FR 24549 Published 5/9/95 Effective 5/9/95 Changes to NRG Apdresses and , Telephone Numbers See Part 2 Statements of Consideration 50-SC-144 60 FR 36953 Published 7/19/95 Effective 8/18/95 10 CFR Part 50 RIN 3150-AF06 Technical Specifications _ AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory

Commission (NRC) is amending its: regulations pertaining to technical specifications for nuclear power reactors.

The rule codifies criteria for determining the content of technical specifications. Each licensee covered by these regulations may voluntarily use* the criteria as a basis to propose the relocation of existing technical* . specifications that do not meet any of the criteria from the facility license to licensee-controlled-documents. The vol~tary conversion of current technical specifications in this manner is expected to produce an improvement in the safety of nuclear power plants through a reduction in unnecessary plant transients and more efficient use of NRC and industry resources. EFFECTIVE DATE:,August 18, 1995. FOR FURTHER INFORMATION CONTACT: . Christopher I. Grimes, Chief, Technical Specifications Branch, Division of Project Support, Office of Nuclear Reactor Regulation, U.S. Nuclear

Regulatory Commission*, Washington, DC 20555;..ooo1, Telephone:

(301) 415-. 1161. SUPPLEMENTARY INFORMATION: Background Section 182a. of the Atomic Energy Act of.1954 (Act), as amended (42 U.S.C. 2232), mandates the inclusion of technical specifications in lli:enses for*

the operation of production and utilization facilities.

The Act requires

that technical specifications include information concerning the amount, kind, and source of special nuclear material; the place of use; and the specific characteristics of the facility.

That section also states that technical specifications shall contain information the Commission requires through regulation to enable it to find that the utilization of special nuclear material will be in accord with the common defense and security and will provide adequate protection of public health and safety. Finally, that section requires

technical specifications to.be made a part of any license issued; The Commission promulgated

§ 50.36, "Technical Specifications," which PART 50

STATEMENTS OF CONSIDERATION implements section 182a. of the Atomic Energy Act on December 17, 1968 (33 FR 18610). This rule;delineates requirements for determining the contents of technical:

specifications. Technical specifications, at a minimum, must set forth the specific characteristics of the facility and the conditions for its operation that are required to provide adequate protection of the health and safety of the public. Specifically, § 50.36 ,requires the following:

Each license authorizing operation of a production or utilization facility of a type described in§ 50.21 or§ 50.22 will include technical specifications.

The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto submitted pursuant to § 50.34. The Commi~sion may include such additional technical specifications as the Commission finds appropriate. Technical specific;:ations cannot be changed by licensees without prior NRC approval. However,:since 1969, there has been a trend to~ard including in technical specifications not only those requirements derived from the ana~yses and evaluation in the safety analysis report but also essentially all othe~ Commission requirements governing the operation of nuclear power reactors. This extensive use of technical specifications was due in part to a lack of well-defined criteria (in either the body of the rule or in some other regulatory document! for wh~t sho~ld be included in technical specifications.

Since 1969, this use has contributed to the volume of technical specifications and to the several-fold increase in the number of license amendment applications to effect changes to _the technical specifications.

It has diverted both NRC staff and licensee attention from the more important requirements in these documents to the extent that it

has resulted in an adverse but unquantifiable impact on safety. On March 30, 1982 (47 FR 13369), the NRC published in the Federal Register a proposed amendment to Part 50. The proposed rule would have revised § 50.36, "Technical Specifications," to establish a new system of specifications divided into'two general categories.

Only those specifications contained in the first general category as technical specifications would have become part of the operating license and would have required prior NRC approval for any changes. Those specifications contained in the second general category would have become supplemental specifications and;would not have required prior NRC approval for most changes. The NRC, review of the first

general category of specifications would have been the same as that currently performed for technical specification changes; which are amendments to the operating license. For the second category, "supplemental specifications," the licensee would have been allowed to . . make changes within specified conditions without prior NRC approval.

The-NRC :would have reviewed these changes when they*~ere made an~ : . would have done so m a manner similar to that currently used for reviewing

design changes, tests, and experiments

' performed under the provisions of : § 50.59. Because of difficulties with ' defining the criteria for dividing the

technical specifications into the two categories of the p~oposed.

ru~e and because of other higher pnonty ' licensing work, the proposed

amendment was deferred. . 1n the early 1980s, the nuclear industry and the NRC staff began studying whether the existing system of establishing technical.

specification requirements for nuclear power plants needed improvement. During this. period, an*NRC task group known as the* Technical Specifications Improvement Project (TSIP) and a Subcommittee of ; the Atomic Industrial Forum's (AIF's) : Committee on Rea'Ctor Licensing_and Safety performed two studies of this issue,t The overall conclusion of these studies was. that many improvements. in the scope and content of technical

specifications were.needed and*that a joint NRC and industry program should be initiated to implement these improvements.

Both groups made specific*recommendations; these are summarized as follows: . (1) The NRC should adopt the criteri'a

for defining the scope of technical . ; specifications proposed in the AIF and.

TSIP reports. Those criteria should then be used by the NRC and-each of the
nuclear steam supply system vendor* owners groups to completely rewrite and streamline the existing standard technical specifications (STS). This process would result in the transfer of many requirements from control by
technical specification requirements to ' control by other mechanisms (e.g., the final safety analysis report (FSAR), . operating procedures, quality assurance (QA) plan) that would not require a license amendment or prior NRC approval when changes were needed. 1 SECY-86-10, "Recommendations for Improving Technical Specifications," January 13, 1986, contains both "Recommendations for Improving Technical Specifications," NRC Technical

Specifications Improvement Project, September 30, * *19e5, end "Technical Specifications improvements," AIF Subcommittee on Technical Speclficetionll Improvements, October 1, 1985. 50-SC-145 The new STS should place greater emphasis on human factors principles in order to make the text of the STS clearer and easier to understand.

The new STS should also improve the bases section of technical specifications, which gives the purpos~ for ~ach requirement in-the specification . (2) A parallel program of short-term improvements in both the scope and substance of the existing technical specifications should be initiated in addition to developing new STS as *stated in recommendation

1. On February 6, 1987* (52 FR 378~). the NRC published in the F.~dera~ Reg1S!er for public comment an Intenm*Pohcy Statement.

on Technical Specification Improvements for Nuclear Power Reactors" (interim policy statement) containing proposed criteria in,response to recommendation

1. These cntena were generally derived from the criteria proposed in the AIF and TSIP reports. and were modified slightly on the basis of discussions between the NRC staff

and the industry.

The public comm(lnt period for the interim policy statement expired on March 23, 1987. . The criteria were developed with the intention that they would apply to limiting conditions for operation (LCOs). The NRC staff belie".ed that the safety limits needed to ~emam . unchanged in the technical . . specifications because of their more direct link to protection of the physical barriers that guard against the . . uncontrolled release of radioactivity. At the time the criteria*were developed, the industry did not wish to addres~ administrative controls and design features in the effort to improve the . STS. Later, however, both the industry and the NRC staff realized that it would be beneficial to include upgraded administrative controls and design features in the improved STS, and these were handled separately from the

application of the criteria to the LCOs. The NRC bas developed a program for short-term improvements as. described in recommendation 2 (above). These are known as ".line-item" improvements and are generic improvements developed and promulgated by the NRC staff for voluntary adoption by licensees.

Subsequently, improved specific STS were developed and issued by the NRC in September. 1992. The improved STS were pubhshed as the following NRC reports:

NUREG-1430, "Standard Technical Specifications, Babcock and Wilcox Plants"
NUREG-1431, "Standard Technical Specifications, Westinghouse Plants" PART 50
STATEMENTS OF CONSIDERATION
NUREG-1432, "Standard Technical Specifications, Coinbustion Engineering Plants"
NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4"
NUREG-1434, "Standard Technical Specifications, General Electric Plants, BWR/6" Copies of these NUREGs, as revised,*

may be purchased from the Superintendent of Documents, U.S. Government Printing Office, by calling (202) 275-2060 or by writing to the Superintendent of Documents, U.S. Government Printing Office, PO Box 37082, Washington, DC 20013-7082. Copies are also available from the National Technical Information Service, 5825 Port Royal Road, Springfield, VA. 22161. These improved STS were the result of extensive technical meetings and discussions among the NRC staff, industry owners groups, yendol'l!, and the Nuclear Management and Resources Council (NUMARC).

On July 22, 1993 (58 FR 39132), the Commission published a "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (final policy statement), which incorporated experience.and lessons learned since publication of the interim policy statement.

The Commission has decided not to withdraw the final policy statement* because it contains detailed discussions of the four criteria and guidance. on how the NRC staff and licensees should apply the criteria. The interim policy statement identified three criteria to be used to define which of the current technical specification requirements should be retained or included in technical specifications and which LCOs could be relocated to licensee-controlled documents, as follows: Criterion 1: Installed instrumentation thatis used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product.barrier. Criterion 3: A structure, system, or component that is part of the primary success path and which fuI).ctions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presen.ts a chalienge to the integrity of a fission product barrier. The interim policy statement.also stated that, in addition to structures, systems, and components captured by the three criteria, it was the Commission's policy that licensees retain in the technical specifications LCOs for a specified list of systems that operating experience and probabilistic risk assessment (PRA) had generally shown to be important to public health and safety. In the final policy statement, the Commission retained this thought as a fourth criterion as follows: Criterion 4: A structure, system,.or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. As stated in the final policy statement, if a requirement meets any one of the four criteria, it should be retained or included in technical specifications; The final policy statement also . addressed comments received on the, interim poli~y statement and described th~ Commission's intent with regard to* use of the criteria and their codification through rulemaking.

This .final rule codifies the four criteria contained.in the final policy statement for defining the scope of LCOs in technical specifications.

These criteria are intended to be consistent with the scope of technical specifications as st!l,ted in the Statement of Consideration for the final rule issuing § 50.36 (33 FR 18610, o*ecember 17, 1968). The Statement.of Consideration discussed the scope of technical specifications as including the following:

In the revised system, emphasis is placed on two general classes of tech.nice!

matters: (l) Those related to prevention of acciden\J;, and (2) those related to mitigation of the consequences of accidents. By systematic analysis and evaluation of a particular facility, each applicant is required to identify at the com1truction pennit stage those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity. Such items are expected to be the subjects of Technical Specifications in the operating license. The first of these two general classes of technical matters to be included in technical specifications is captured by Criteria 1, 4, and, to some extent, Criterion 2, in that they.address systems and process variables that alert the operator to a situation when accident initiation is more likely. The second general class of technical matters is explicitly addressed and captured by Criteria 2, 3, and 4. By applying the four criteria contained in thi~ rule, a licensee should capture the conditions for operation of its facility that are required 50-SC-146 to meet the principal operative standard in Section 182a. of the Atomic Energy Act, that is, that adequate protection is provided to the health and safety of the public. .

The Commission recognizes that the four criteria carry a theme of focusing on the technical requiJ:ements for features of controlling importance to safety. Since many of the requirements are of significance to the health and safety of the public, this rule reflects the subjective statem~nt of the purpose of technical specifications expressed by the Atomic Safety and Licensing Appeal Board in Portland General Electric Company (Trojan Nuclear Plant), ALAB-531, 9 NRC 263 (1979). There, the Appeal Board interpreted technical specifications as being reserved for tht1se conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety . The Commission wishes to emphasize that this rule is intended to be consistent with the language of section 182a. of the Atomic Energy Act, the current § 50.36 rule, and previous interpretations of the regulations.

This rule merely clarifies the scope and purpose of technical specifications by identifying criteria which can be used to establish, more clearly, the framework for LCOs in technical specifications. The Commission believes that amending§ 50.36 to.include the four criteria contained in the final policy statement will codify a viable, potentially safety-enhancing and saving method for technical specification improvement. The Commission

continues to encourage licensees to use the improved STS as the basis for plant-specific technical specifications.

As stated in the final policy statement, the Commission will place the highest priority on requests based on the criteria for individual license amendments that are used to evaluate all of the LCOs for an individual*plant to determine which LCOs should be included in the technical specifications. Related surveillance requirements and actions would be retained for each LCO that remains in the technical specifications. Each LCO, action, and surveillance requirement should have supporting bases. Such requests would constitute complete conversions to the improved STS. In addition, the Commission will also entertain requests to adopt portions of the improved STS, _!3ven if the licensee does not adopt all STS improvements. These portions will include all related requirements and will be developed as PART 50

STATEMENTS OF CONSIDERATION line-item improvements by the NRC staff when they are clearly generic in nature, when*there is evidence that a significant number of licensees could benefit from the improvement, and when the industry expresses interest in the improvement.

The Commission encourages all licensees who submit

technical specification related submittals based on these criteria to emphasize human factors principles to the extent practical consistent with the fol'IIltlt and content of their current technical specifications.

LCOs that do not meet any of the criteria, and their associated actions arid surveillan~e requirements, may be. proposed for relocatiQn from the technical specifications to controlled documents, such as the FSAR. The criteria may be applied to either standard or custom technical specifications. The' Commission will also consider the criteria in evaluating future generic requirements for inclusion in technical specifications. The Commission expects that licensees, in-preparing their technical specificati1;,n subntjttals, will utilize-any plant-specific PRA or risk survey and any available literature on risk insights and PRAs. This material should be employed to strengthen the technical bases for-those provisions that remain in technical specifications, when applicable, and to indicate whether the provisions to be relocated contain

constraints of importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk. Similarly,.the NRC staff has and will contipue to employ risk insights in evaluating technical specifications subµiittals.

~11 actdition to the use of PRA in Criterion 4 to determine the scope of . technical specifications, PRA has been .used as a*basis for a number of ~h~vements to the content of t cal specifications over the last several years. The NRC staff bas appro:ved several relaxations in technical specification allowed outage times and surveillance test intervals which. were based on PRA. In addition, the NRC staff used PRA to develop screening criteria to evaluate all of the* changes in allowed outage times and surveillance test intervals that were made during the development of the improved STS. The industry and the NRC staff have used PRA to an even greater.extent in the development and review of the tecJ:mical specifications for advanced reactor designs. The industry and the NRC staff are currently exploring several new approaches to utilizing PRA for technical specification improvements including the use of on-line risk assessment tools. In addition, the industry and the NRC staff are using PRA to explore further improvements in technical specifications by examining the risks during shutdown and during the transition between modes of operation. As a part of this ongoing program of improving technical spi;icifications, the Commission will continue to consider methods to make better use of risk and reliability information for defining future generic technical specification requirements. During technical specification conversions, the staff will apply the backfit rule (§ 50.109) when adding new requirements from the improved STS to individual plant technical specifications, provided the licensee does not voluntarily accept the new requirements. If, however, the staff suggested additional changes are needed to make the licensee requested changes acceptable from the standpoint of adequate protection or compliance with NRC regulations, § 50.109(a)(2) and § 50.109(a)(3) do not apply and the .request may be denied without the additional items. Summary of Public Comments The Commission received three letters commenting on the proposed rule. Each letter contained several comments. One commenter representing the commercial nuclear industry expressed concern that there is insufficient regulatory guidance *on how the NRC staff intends to implement this rule with respect to the fourth criterion

(§ 50.36(c)(2)(ii)(D)).

The commenter believes that this rule should not be modified until the NRC and the industry have reached a common miderstanding of the application, tlueshold, and intent of Criterion

4. The commenter stated, It is our view, and the Commission apparently recognizes, that this criterion gc,>es beyond the adequate protection standard for publi~ health and safety and license compliance purposes embodied in the first three criteria." Similar to this comment on the proposed rule, the Advisory Committee on Reactor Safeguards (ACRS) commented in a June 18, 1993, letter to the Chairman that the NRC staff needs to provide more detailed guidance on the definition of "significant to public health and safety," as it is used in Criterion

4. Criterion 4 is intended to capture those constraints that probabilistic risk

assessment or operating experience show to be significant to public health . and safety, consistent with the Commission's PRA Policies.

The level of significance either would need to be 50-SC-147 such that it justified including the constraints in the technical specifications to ensure adequate protection of the public health and safety or that the addition of such constraints provides substantial additional protection to the public health and safety.

The Commission identified four systems that meet Criterion 4 in the final policy statement based on previous qualitative reviews of operating experience and risk. They are reactor core isolation cooling/isolation condenser, residual heat removal, standby liquid-control, and recirculation pump trip. The Commission recognizes, however, that other structures, systems, or components may meet this criterion.

Plant-and design-specific PRAs have yielded valuable insight to unique plant *vulnerabilities not fully recognized in the safety, design basis accident, or

transient analyses.
The NRC's current regulatory requirements are largely based on deterministic engineering criteria involving the use of multiple barriers and defense in depth. Recently, the NRC staff has formulated a comprehensive plan for the application of PRA technology and insights throughout the agency. It is expected that the PRA Implementation Plan will serve as the framework for continued and future applications of PRA at the NRC. Implementation ofthis plan will increase the systematic use of risk nssossll)ent techniques.

To ensure consistent and appropriate making that incorporates PRA methods and results,it is important that coherent and clear application guidelines are applied. As part of the PRA

Implementation Plan, such guidelines*

will be established (incorporating safety. goals and backfit rule consi_derations) that address the interdependence of probabilistic risk and deterministic engineering principles. The process of developing these guidelines will involve

communications among the NRC staff, the nuclear industry, and the public to ensure that all parties understand the role of PRA methods and results in NRC's risk management efforts. The NRC staff anticipates that, as it gains experience with the development and use of such PRA application guidelines, it will be better able to refine phrases as "significant to public health and safety," and other phrases that are used in many of the Commission's regulations.

The Commission could delay publication of this final rule until the. PRA application guidelines are in place . However, the Commission believes that the experience gained while using the PART 50

STATEMENTS OF CONSIDERATION criteria under the interim and final policy statements combined with the limitations imposed on the NRC staff by the backfH rule provide assurance that, in the interim, the stafrs use of Criterion 4 to apply PRA to technical specification content will be properly controlled.

The Commission has

concluded that it is appropriate to publish this final rt,1le, which provides the framework for technical specifications, at this time.
One commenter stated thai the PRA portion of the fourth criterion should be clarified to include only those equipment items important to significant sequences as defined in Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities," Appendix 2, and reported in licensees' individual plant examination (IPE) reports. The IPE program nas resulted in commercial reactor licensees using assessment methods.to identify specific severe accident vulnerabilities.

Since submittal of their IPE reports, .. many licensees have enhanced their plant-specific PRAs and have gained additional insights into unique plant vulnerabilities. These additional insights from PRAs are being used by licensees in such areas as implementation of the maintenance rule. As stated in the Commission's "Proposed Policy Statement.on the Use of Probabilistic Risk Assessment

Methods in Nuclear Regulatory Activities," the'use of PRA technology should be increased in all regulatory matters to the extent supported by the state of the art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional in-depth,philosophy.

The Commission will continue to apply PRA to technic;al specifications in accordance with its proposed policy statement on the use of PRA. In addition, guidance for specific applications cir classes of applications will be developed under the PRA * . Implementation Plan. The Commission believes this is a more appropriate means to define how Criterion 4 will be used in practice, rather than to limjt the structures, system.s, and compo:nents captured by Criterion 4 to those items important to risk-significant sequences as defined in Generic Letter 88-20, Appendix 2, and reported in licensees' IPE reports; The Commission belie.ves that this process will provide the NRC staff and the industry with additional risk insights, beyond those identified through the IPE program. The same commenter said that the operating experience portion of the fourth criterion should be deleted because it is subjective and because no equipment would satisfy only that portion of the fourth criterion arid none of the other criteria. While operating experience is an important part of PRA, not all PRA models are sophisticated enough to capture all operating experience. The Commission believes* that operating experience can play an important role in determining the safety significance of structures, systems, and components and that there will be no adverse impact by including operating experience as . part of Criterion

4. One commenter emphasized that the development of implementation guidance, especially with respect to Criterion 4, should be consistent with the implementation guidance of the maintenance rule.

As stated previously, the Commission believes that the improved STS, the final policy statement, the backfit rule (§ 50.109), and the statement of consideration for this rule contain sufficient guiqance on implementation of the criteria to proceed with rulemaking.

Supplementary guidance will continue to be provided to the NRC staff that will support the process for -implementing the four criteria on both a generic and plant-specific basis, and will be publicly available. The NRC staff will ensure that any guidance documents that relate to the implementation of the four criteria will be consistent with the implementation guidance of the maintenance rule along with the guidance for other rules promulgated by the Commission. One commenter expressed a concern With respect to the level of PRA information necessary to support the relocation of existing technical specifications which do not meet the first three criteria. If a technical specification provision does not meet any of the first three criteria, and if the current PRA . knowledge or operating experience does not identify the structure, system, or component as risk significant, the NRC staff will not preclude relocating such technical specifications. The level of PRA information necessary to support relocation would be considered as part of the overall review of the supporting basis for the proposed change. The Commission expects that licensees will utilize PRA insights to indicate whether the provisions to be relocated contain constraints of importance fn litniting the likelihood or severity of the nccidont sequences that are commonly found to dominate risk. One commenter stated that the implementing guidance needs to be 50-SC-148 , clear on how the proposed criteria would be used to determine if new requirements are to be incorporated into technical specifications. The Commission believes that the improved STS, the final policy. statement, the backfit rule(§ 50.109), and the statement of consideration for this rule contain sufficient guidance on implementation of the criteria. The staff' will also ensure that application of the criteria to new requirements is consistent with the guidance in the draft "Regulatory Analysis Guidelines," Revision 2, published in August 1993 (NUREG/BR-0058), and the final version of Revision 2 when it is approved by the Commission. In addition.the NRC has recently published NUREG/CR-6141, "Handbook of Methods for Risk-Based Analyses of Technical Specifications," December 1994, which summarizes systematic risk-based methods to improve various aspects of technical specification requirements. The handbook was developed through research sponsored by the NRC and will be used as a reference document to assist the NRC staff in reviewing licensees' risk-based analyses submitted* as part of the bases for proposed *

changes in facility technical specifications.

This guidance will be . updated periodically to incorporate lessons learned and changes in the state

of the art, will help ensure the criteria are applied in a consistent and controlled manner, and will be publicly available.

As stated above, as part of the PRA Implementation Plan, PRA application guidelines ~ill be. established (incorporating safety goals and backfit rule considerations) that address the interdependence of probabilistic risk and deterministic engineering principles. As these application guidelines develop, they will progressively be used to provide guidance to the NRC staff on the use of the criteria contained in this rule and the application of the backfit rule to new regulatory requirements. One commenter stated that the same or similar criteria to those in the rule should also be applied to 10 CFR 50.36(c)(3), (4), and (5), so that surveillance requirements, design features, and administrative controls which do not provide the necessary "adequate protedtion of the health and safety of the public" can be relocated to other licensee-controlled documents. With respect to§ 50.36 (c)(3), "Surveillance Requirements," the Commission stated in the final policy statement that appropriate surveillance reqµirements and actions should be retained for each LCO which remains or PART 50

STATEMENTS OF CONSIDERATION is included in the technical specifications.

The criteria in § 50.36(c)(2) apply to safety functions. Therefore, the Commission does not believe that these

criteria can be appropriately applied to the 'types of requirements found in the "design features" and administrative controls" sections of the technical specifications.

The NRC staff has, however, been pursuing separate improvements to these requirements, in cooperation with industry, using the intent of the criteria to identify the optimum set of requirements in each of these areas and to eliminate redundancy to other regulations consistent with the minimum requirements of§ 50.36 and the Atomic Energy Act, as amended. One commenter stated that the removal of items from plant technical specifications may decrease enforceability and licensee attention to safety. The Commission does not agree that the removal of items from plant technical specifications will decrease licensee attention to safety. On the contrary, the Commission believes that implementation of the criteria contained in this rule will produce an improvement in the safety of nuclear power plants through.the Use of more operator-oriented technical specifications, i~proved technical specification bases, reduced action statement induced plant transients, and more efficient use of NRC and industry resources. Clarification of the scope and purpose of technical specifications has provided useful guidance to both the NRC and industry and has resulted in improved technical specifications that are intended to focus licensee and plant operator attention on those plant conditions most important to safety. The Commission also does not agree that the removal of items from plant technical specifications will have any adverse impact on the NRC's ability to take enforcement action on significant issues. The improved STS are intended specifically to focus on the operating plant parameters and associated surveillance criteria of safety significance. The Commission requires compliance with technical specifications, and expects adherence to commitments contained in controlled documents. Violations and deviations will, as in the past, be handled in accordance with tho NRC enforcement policy in 10 CFR Part 2, Appendix C. Any changes to a licensee's technical specifications to apply these criteria will be made by the license amendment process prior to implementation. When a licensee elects to apply these criteria, some requirements are relocated from technical specifications to tlre FSAR or to other controlled documents. Licensees are to operate their faoilities in conformance with the descriptions of their facilities and procedures in their FSAR. Changes to the facility or to procedures described in the FSAR are to be made in accordance with 10 CFR 50.59. The Commission will talce appropriate enforcement action to ensure that licensees comply with 10 CFR 50.59. Changes made in accordance with the provisions of other licensee-controlled documents (e.g., QA plan, security plan) are subject to the specific requirements for those documents. Nothing in this rule limits the authority of the NRC to conduct necessary inspections and to take appropriate enforcement action when regulatory requireme1~ts or commitments are not met. The same commenter stated that the removal of items from plant technical specifications will diminish public participation rights in the regulation of operating nuclear power plants by

diminishing the universe of potential oporatin_g license amendment cases. Any changes to a licensee's technical specifications to apply these criteria will be made by the license amendment process before implementation.

The review of each license amendment will involve an opportunity for public participation. One of the goals of the technical specifications improvement program was to make more efficient use of NRC and industry resources by focusing attention on those plant conditions most important to safety and, in turn, reducing the number of license amendment requests. Since 1969, there has been a trend toward including in technical specifications not only those requirements derived from the analyses and evaluations included in the safety analysis report but also essentially all other Commission requirements governing the operation of nuclear power reactors. This extensive use of technical specifications is due in part to a lack of well-defined criteria (in either .

the body of the rule or in some other regulatory document) for what should be included in technical specifiGations.

This has contributed to the volume of technical specifications and to the several-fold increase, since 1969, in the number of license amendment applications to effect changes to the technical specifications. It has diverted both NRC staff and licensee attention from the more important requirements in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety. 50-SC-149 The commenter found it curious that an industry and an agency that claim to be able to quantify the risks of nuclear power are unable to quantify this impact on safety, and stated, "Perhaps if it is unquantifiable, the alleged adverse impact does not really exist." The Corrmission agrees that there are limitations and uncertainties in the ability to quantify the impact on safety described above. Uncertainties exist in any regulatory approach and these uncertainties are derived from knowledge limitations. A probabilistic approach has exposed some of these limitations and yielded an improved framework to better focus and assess their significance and assist in developing a strategy to accommodate them in the regulatory process. The Commission does not intend, however, to let these limitations prevent it from talcing steps to improve the regulations in a manner that will have substantial safety benefits. The Commission believes the public will be better served by focusing both NRC and industry attention on the most safety-significant items. The NRC staff has made three changes to this rule since it was published in its proposed form. The first change was made in order to maintain consistency with other NRC staff end Commission documents that have been issued since this rule was publishe~ in its proposed form. In § 50.36(c)(2)(ii)(D), the term "probabilistic safety assessment" hes been changed to "probabilistic risk assessment. The second and third changes are in § 50.36(c)(2)(iii). The beginning of the first sentence was changed to read, "A licensee is not required to propose to modify technical specifications

* *" rather than "A licensee is not required to modify technical specifications
* *" This change was made to clarify that a licensee would be required to modify their technical specifications if the Commission determined that a new requirement was necessary in accordance with the. backfit rule and the new requirement met one of the four criteria contained in § 50.36(c)(2)(ii).

The third change is the deletion of the last sentence in § 50.36(c)(2)(iii). The sentence read, "However, for technical specification amendments a licensee proposes after August 18, 1995, the criteria in paragraph (c)(2)(ii) of this section provide an acceptable scope for limiting conditions for operation." This sentence was deleted because it did not add or modify any requirements and the thought is adequately expressed in this statement of consideration. PART 50

STATEMENTS OF CONSIDERATION Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969: as amended, and the Commission regulations in Subpart A of Part 51, that this final rule is not a major Federal action significantly affecting the quality of the human environment and will not degrade the environment in any way. Therefore, the Commission concludes that there will be no significant impact on the environment from this rule. This discussion constitutes the environmental assessment and finding of no significant impact for this rule; a *separate assessment has not been prepared.

Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement.subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget, approval number 3150-0011. . Regulatory Analysis The Commission has determined that a regulatory analysis is not required for this rule. The Commission believes that the intent of the regulatory analysis has been met through the extensive consideration given to the development of the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" and the improved STS, both of which gave the public an opportunity for comment. In addition, the determination that no regulatory analysis is necessary was noted in the FederalRegister Notice for the proposed rule, and the NRC received no comments on this issue. The criteria being added to § 50.36 are the same as those contained in the final policy statement and have been used by the NRC and the nuclear power industry to define the content of technical specifications since September 1992. The rule does not impose any requirements but, rather, allows nuclear power reactor licensees to voluntarily use the criteria to relocate existing technical specifications that do not meet any of the criteria to licensee-controlled documents, The NRC staff also uses these criteria to determine whether technical specifications arc. appropriate to provide regulatory control over new requirements or positions that have been justified consistent with the backfit rule. The Commission considered the need for and consequences of this action when it made the decision not only to publish the criteria in the final policy statement but also to codify the criteria through rulemaking. Appropriate alternative approaches to this action have been identified and analyzed over the life of the Technical Specifications Improvement Program, beginning with an earlier attempt to define the content of technical specifications through rulemaking. As described in the background discussion, the Commission published a proposed amendment to § 50.36 (47 FR 13369) on March 30, 1982. However, because of difficulties with defining criteria for technical specifications and because of other higher priority licensing work, the rule change was deferred. In February 1987, the Commission published an "Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors," and in July 1993, published the final policy statement. During its review of the final policy statement, the Commission concluded that the four criteria should pe codified in a rule. Thus, alternative approaches to regulatory objectives have l:ieen identified and analyzed, and the Commission has decided that there is no preferable alternative to codifying the four criteria in a rule. With regard to evaluation of values and impacts of alternatives, the Commission believes there is no difference in the values or impacts of applying the criteria under the final policy. statement or through a rule, *except that the criteria are more readily available to future users in a rule rather than in a policy statement. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the Commission certifies that this final rule does not have a significant economic impact on a substantial number of small entities. This rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of"small entities" as given in the Regulatory Flexibility Act or the Small Business Size Standards in regulations issued by the Small Business Administration at 13 CFR part 121. Backfit Analysis The NRC has determined that the backfit rule, § 50.109, does not apply to this final rule and, therefore, a backfit analysis is not required for this final rule because these amendments do not involve any provisions that would impose backfits as defined in § 50.109(a)(l). 50-SC-150 List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons given in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendment to Part 50. 60 FR 49495 Published 9/26/95 Effective 10/26/95 10 CFR Part 50 . RIN 3150-AFOO Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors AGENCY: Nuclear Regulatory . Commission. . ACTION: Final rule; .

SUMMARY

The Nuclear Regulatory

Commission is* amending its regulations to provide a performance-based option for leakage-rate testing*of containments of light-water-cooled nuclear power
plants. This option is available for_ voluntary,'adoption by licensees in lieu of compliance with the prescriptive
requirements contained in the current
regulation.

This action improves the. focus of the regulations by eliminating prescriptive requirements that are ..

marginal to safety. The final rul~ allows PART 50
STATEMENTS OF CONSIDERATION tes(intervals to be ~d-on system and.
list also selected Appendix J a~ a component performanCl!*&nd provides . potential candidate for modification.

licensees greater flexibility for cost-The NRC published in the Federal effectiye implementation methods of Register, for cOJnment, a proposed regulatory safety obje~ves.

* * . .revision to Appen4ix J on October 29, EFFECTIVE DATE: Oct_.ober 26, 1995. .
1986 (5 l FR 3Q538),to update . . . FOR FURTHER INFORMATION CONTACT: Dr. Moni Dey, Office of Nuclear_Regulatory Research/U.S.

Nuclear Regulatory CommiB8ion;Washington, DC 20555, telephqne (301) 415-6443,*e-mail nikd~.gov .*

1 **
SUPPLEMENTARYlNFORMATION: . B41ckground.,-Developnient of Proposed Rule *
acceptance criteria and test methods based o~ experience in applying the existing requirements and advances in co_ntainment leak testing methods, to ~solve interpretive questions, and to reduce the number of exemption requests.

This proposed *rule was withdrawn from further consideration

and superseded with a more comprehensive revision of Appendix J. Tlie NRC published a notice in the Federal Register on February 4, 1992 NRC's Marginal-to-~afety Program (57 FR 4166), presenting its conclusion that Appendix J was a candidate whose In 1984, the NRC staff initiated a requirements may be relaxed or program to make regulatory eliminated based on cost-benefit requirements more efficient by considerations.

On the basis of NRC eliminating those with marginal impact staff analyses of public comments on the on safety. The NRC'~ initiative to proposal, the Commission approved and eliminate requirements marginal to announced on November 24, 1992 (57 safety recognims both the dynamic FR 55156) its plans to initiate nature of the regulatory process and that rulemak.Qig for developing a the importance and safety contribution performance-oriented and risk-based of some existing regulatory regulation for containment leakage-requirements may not have been testing requirements. On January 27, accurately predicted when adopted or. 1993, (58 FR 6196) the NRC staff may have diminished With time. The . pu~lished a general framework for availability of new technical

developing pepormance-oriented and information and methods justify a . risk-based regulations and; at a public review and modification of existing workshop on April 27 and 28, 1993, requirements.

' invited discussions of specific.proposals The NRC solicited comments from for modifying*containment leakage-. industry on specific regulatory testing ~ants. Industry and. . requirements iµid associated regulatory*

public commenl!>

on the proposals, and.: positions th~ needed l'etlValuation. The,. oth~rmcommendatimis and*innovative

Atomic Industrial Forum conducted a .
ideas raised at the. public workshop;
. survey providing:most of-industry's.

were documented in the.proceedings of .input, publishe~ fo~ the NRC as *the workshop (NUREG/CP-0129, NUREG/CR-4330 1, "Review of Light September 1993). ~pecifically, the NRC

Water Reactor Regulatory
concluded that the allowable Requirements,".Vol.

1, April 1986. A containment leakage rate utilized in list of 45 candidates for potential

containment testing may be increased regulatory modification were identified.

and other Appendix J requirements The NRC's.review of the list selected

need not be as prescriptive a.s the* Appendix J a~ one of seven areas current requirements.

To increase requiring further analysis (NUREGiCR-flexibility, the detailed and prescriptive 4330, Vols .. 2 and 3; dated June 1986 technical requirements contained in and 1$.y 1987). The NRC also,

Appendix J regulations could be . conducted a survey ofits staff on.the improved .and replaced with . same issue. The NRC staff survey
perform~ce-based requirements.

and identified 54 candi\lates for regulatory supporting regulatory guides; The . modification, a number of which were

regulatory guides would allow
previously identified in the industry
alternative approaches, although survey. The NRC's assessment of this *
compliance with existing regulatory

'Copies of NUREGs may be purchased from the* Superintendent of Doc11111ents, 'U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082. Copies are also available from the National Technical Information Service. 5285 Port Royal Road, Springfield, VA 22161. A copy is . available for inspection and/or copying in the NRC Public Document R09m,'2120 L Street, NW. (Lower Level), Washington, rx;:. * . . . . .

requirements would continue to be acceptable; The performance-based requirements would reward superior operating practices.
The. present rulemaking is part of this overall effort and initiative for . eliminating requirements that are marginal to safety and is guided by.the policies,.framework and criteria for the 50-SC-151 program. A more comprehensive proposed rule than that proposed in 1986 that accounts for the latest technical information and regulatory framework, using performance-oriented and risk-based approaches, was published by the NRC in the Federal. Register on February 21, 1995. The public comment period for the proposed rule closed May 8, 1995. NRC's Regulatory Improvement Program The NRC's marginal-to-safety initiative is part of a broader NRC initiative for regulatory improvement.

Through its Program for Regulatory Improvement, the NRC has institutionalized an ongoing effort to eliminate requirements marginal to safety and to reduce the regulatory burden on its licensees. The NRC stafrs plan, summarized in SECY-9~90, dated March 31, 1994, satisfies the

requirement for a periodic review of existing regulations given in Executive Order 12866 of September 30, 1993. This plan was approved by: the Commission on May 18, 1994. The Regulatory Improvement Program is aimed at the fundamental principle adopted by the Commission that all regulatory burdens must be justified and that its regulatory process must be efficient.

In practice, this means the elimination or modification of requirements for which burdens are not commensurate with their safety

significance.

The activities of the Regulatory Improvement Program should result in enhanced regulatory focus in areas that are more safety significant. As a result, an overall net increase in safety is expected from the program.

The Regulatory Improvement Program will include, whenever feasible and appropriate, the consideration of performance-oriented and risk-based approaches.

The program will review requirements or lice~e conditions that are identified as a significant burden on licensees. If review and arialysis find that the requirements are marginal to safety, they will be eliminated or

relaxed. By performance-oriented, the NRC means establishing regulatory objectives without prescribing the methods or hardware necessary to accomplish the objective, and allowing licensees the flexibility to propose effective methods for implementation.

By risk-based, the NRC me~s regulatory approaches that use probabilistic risk analysis (PRAr as the systematic framework for developing or modifying requirements. In institutionalizing the Regulatory Improvement Program and adopting a performance-based regulatory approach, PART 50

STATEMENTS OF CONSIDERATION the NRC has formulated the following framework for revisions to its regulations:

(1) The new performance-based regulation will be less prescriptive and will allow licensees the flexibility to adopt cost-effective methods for implementing the safety objectives of the original rule. (2) The regulatory safety objectives will be derived, to the extent feasible and practical, from risk considerations with appropriate consideration of uncertainties, and will be consistent with the NRC's Safety Goals. (3) Detailed technical methods for measuring or judging the acceptability of a licensee's performance relative to tho regulatory safety objectives will be, to the extent practical, provided in industry standards and guidance . documents which are endorsed in NRC regulatory guides. ( 4) The new regulation will be optional for current licensees so that licensees can decide to remain in compliance with current regulations. (5) The regulation will be supported by necessary modifications to, or development of, the full body of regulatory practice including, for example, standard review plans, inspection procedures, guides, and other regulatory documents. (6) The new regulation will be formulated to provide incentives for innovations leading to improvements in safety through better design, construction, operating, or maintenance practices. Current Appendix J Requirements Appendix J to 10 CFR Part 50, * "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," became effective on March 16, 1973. The regulatory safety objective of reactor containment design is stated in 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion No. 16, "Containment Design." GDC Criterion 16 mandates "an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment

* *" for postulated accidents.

Appendix J to 10 CFR Part 50 implements, in part, General Design Criterion No. 16 and specifies containment leakage-testing requirements, including the types of tests required. For each type of test required, Appendix J specifies how the tests should be conducted, the frequency of testing, and reporting requirements. Appendix J requires the following types of containment leak tests: (1) Measurement of the containment integrated leakage rate (Type A tests, often referred to as ILRTs). (2) Measurement of the leakage rate across each pressure-containing or

leakage-limiting boundary for various primary reactor containment penetrations (Type B tests). (3) Measurement of the containment isolation valves leakage rates (Type C tests).
Type B and C tests are referred to as local leakage-rate tests (LLRTs). Leak-Tightness Requirements Compliance with 10 CFR Part 50, Appendix J, requirements is determined by comparing the measured containment leakage rate with the maximum allowable leakage rate. Maximum allowable leakage rates are calculated in accordance with 10 CFR Part 100, "Reactor Site Criteria," and are incorporated into the technical specifications.

Typical allowable leakage rates are 0.1 percent of containment volume per day for pressurized water reactors (PWRs) and one volume percent per day for boiling water reactors (BWRs). Tes.t Freqw;mcy Requirements Schedules for conducting containment leakage-rate tests are specified in Appendix J for both preoperational and periodic tests. Periodic leakage-rate test schedules are as follows: Type A Tests (1) After the preoperational rate test, a set of three Type A tests must be performed at approximately equal intervals during each 10-year service period. The third test of each set must be conducted when the .plant is shut down for the 10-year plant in-service inspection.

. (2) The performance of Type A tests must be limited to periods when the plant facility is nonoperational and secured in the shutdown condition under administrative control and in accordance with the safety procedures defined in the license. (3) If any periodic Type A test fails to meet the applicable acceptance criteria, the test schedule applicable to subsequent Type A tests will be reviewed and approved by the Commission.

If two consecutive periodic Type A tests fail to meet the app1icable acceptance criteria, a Type A test must be performed at each plant shutdown for refueling or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet the 50-SC-152 acceptance criteria, after which time the regular retest schedule may be resumed. Type B Tests (1) Except for airlocks, Type B tests must be performed during reactor shutdown for refueling, or other convenient intervals, but.in no case at intervals greater than 2 years. If opened following a Type A or B test, containment penetrations subject to Type B testing must be tested prior to returning the reactor to an operating mode requiring containment integrity. For primary* reactor containment penetrations employing a continuous leakage monitoring system, Type B tests, except for tests of airlocks, may be performed at every other reactor* shutdown for refueling but in no case at intervals greater than 3 years. (2) Airlocks must be tested prior to initial fuel loading and at six-month intervals thereafter. Airlocks opened during periods when containment integrity is not required by the plant's . technical specifications must be tested at the end of such periods. Airlocks opened during periods when containment integrity is required by*the plant's technical specifications must be tested within 3 days after being opened. For airlock doors opened more frequently than once every 3 days, the airlock must be tested at least once every 3 days during the period of frequent openings. For airlock doors having testable s.eals, testing the seals fulfills the 3-day test requirement. Airlock door-seal testing must not be substituted for the 6-month test of the entire airlock at not less than P., the calculated peak containment pressure related to the design basis accident. Type CTests Type C tests must be performed during each reactor shutdown for refueling, but in no case at intervals greater than 2 years. There have been two amendments to this Appendix since 1973; The first amendment, published September 22, 1980 (45 FR 62789), modified the Type B penetration test requirements to conform to what had become accepted practice through the granting of

exemptions.

The second am!Jndment, published November 15, 1988 (53 FR 45890), incorporated the Mass Point Statistical Analysis Technique as a permissible alternative to the Total Time and Point-to-Point techniques specified in Appendix J. In~ernational Experience . A combination of Type A tests and an on-line monitoring (OLM) capability is being actively pursued in Canada and PART 50

STATEMENTS OF CONSIDERATION Europe, notably in Frimce and Belgium, and is currently being considered in Sweden. OLM is used to identify a "normal" containmei;it pressurization pattern and t.o detect deviations from that pattern. With on 7 line, low-pressure testing, Hydro-Quebec's Gentilly-2 station is able to mo~itor the change in containment leaktightness between Type A tests. The Belgians conduct a, leakage test using OLM during reactor operation after each cold shutdown longer than 15 days with the objective of detecting gross lea).<s. The objective of the Belgian apprQach to Type A testing is to reduce the*frequency and duration of the tests. The Typi:i A test is conducted at a containment pressure (P,) not less than half of the peak pressure (0.5 P.). It i~ performed once every 10 years. In France, containment leaktightness is continuously monitored during reactor operation in all of the French PWR plants using the SEXTEN system. It is also bei~g evaluated by the Swedes for their PWR units. Leaks may bo detected during t~o positive or negative pressure periods in the containment by evaluating the air mass balance in the contai,nment.

Type A tests are conducted at containment peak pressure (loss-of-coo.lant accident pressure) before init~al plant startup, during the first reful!ling, and thereafter every 10 years unless a degradation in containment leaktightness is detected. In that case, tests are conducted more frequently. .

i Further details of international approaches to containment testing are provided in NUREGl-1493.

Advance Notices fo~*Rulemaking Over time, it has become apparent that variations in plant design and operation frequently make it difficult to meet so~e of the requirements cQntained in Appendix J because of its prescriptive nature. Economic and occupational expostire costs are directly related to the frequency of containment

testing: Containment integrated rate tests (Type A) preclude any other reactor maintenance' activities and thus are on the critical path for return to service from reactor:outages.

In addition to the costs of the tests, integrated leak tests impose the ad~ed burden of the cost of replacement power .. Containment-penetration leak tests (Type B and C) can be conducted during reactor shutdowns in parallel with other activities and thus tend to be less costly; however, the large number of penetrations impose a significant burden on the utilities. Additionally, risk assessments pe~formed to date indicate that the allowable leakage rate from containments can be increased, and that control of containment leakage at the current low rates is not as risk significant as previously assumed.23 In August of 1992, the NRC initiated a rulemaking to modify Appendix J to make it less prescriptive and more performance-oriented. The Commission also initiated. a plan to relax the allowable containment leakage rate used to define performance standards for containment tests. In the Federal Register ofJanuary 27, 1993 (58 FR 6196), the NRC indicated the following potential modifications to Appendix J of 10 CFR Part 50 would be considered: (1) Increase allowable containment leakage rates based on Safety Goals and PRA technology (i.e., define a new performa~ce standard); and (2) Modify Appendix J to be a performance,based regulation: A. Limit the revised rule to a new regulatory objective. In order to ensure the availability of the containment during postulated accidents, licensees should either: (i) Test overall containment leakage at intervals not longer than every 10 years, and test pressure-containing or limiting boundaries and containment. isolation valves on an interval based on the performance hist_ory of the equipment; or (ii) Provide on-line (i.e., continuous) monitoring of containment isolation status. B. Remove prescriptive requirements from Appendix J and preserve useful pcirtiims as guidance in an NRC regulatory guide. C. Endorse industry standards on: (i) Guidance for calculating specific allowable leakage rates based on new NRC performance standards; (ii) Guidance on the conduct of containment tests; and (iii) Guidance for on-line monitoring of containment isolation status. D. Continue to accept compliance with the current detailed requirements in Appendix J (i.e., licensees presently in compliance with Appendix J will not need to do anything if they do not wish to change their practice). 2 "Severe Accident Risks: An assessment for five U. S. Nuclear Power Plants, Final Summarv Report." NUREG-1150, December 1990. C~pies of NUREGs may be purchased from the Superintendent of Documents, U.S. Government Printing Office, P. 0. Box 37082, Washington, DC 20013/7082. Copies are also available from the National Technical Information Service. 5285 Port Royal Road, Springfield, VA 22161. A copy is available for inspection and/or copying in thr. NRC Public Document Room, 2120 L Street. NW. (Lower Level), Washington, DC. > "Performance-Based Containment Leak Test Program," NUREG-1493, July 1995. 50-SC-153 A public workshop on the subject was held by the NRC on {\pril 27 and 28, 1993.4 Febru.ai-y 1995 Proposed Revision Based on several advance notices for rulemak.ing and significant public comment and discussion, evaluation of risks and costs, and consideration of which modifications have become feasible and practical, in the February 21, 1995, Federal Register the NRC proposed two phases for modifications of requirements to containment leakage testing. The first phase allowed rate testing intervals to be based on the performance of the containment system structures and components. The second phase will further examine the needed requirements of the containment function (i.e. structural and leak-tight integrity of containment system structures and components, and prevention of inadvertent bypass), and include consideration of the potential for on-line monitoring ofcontainment integrity to verify certain functions. . Public comments were solicited to guide this future work. The February 21, 1995, proposed rule applies to all NRC licensees who operate light-water-cooled power reactors. The proposed rule allows licensees the option of continuing to comply with the current Appendix J or to adopt the new performance*based standards. The NRC's analyses are based upon the insight gained through the use of probabilistic risk assessment techniques and the significant data base of practical, hands-on operating .' experience gained since Appendix J was promulgated in 1973. This operating experience provides solid evidence of the activities necessary to conduct Appendix J testing, and the costs of those activities both in monetary terms and occupational radiation exposure. The proposed rule is based on analytical efforts documented in NUREG-1493 which, like NUREG-1150, confirms previous observations of insensitivity of population risks from severe reactor accidents to containment leakage rates. The current Appendix J requirements continue to achieve the regulatory criterion of assuring an essentially tight boundary between the power reactor system and the external environment (General Design Criterion 16). Costs associated with complying with current Appendix J requirements are estimated to be $165,000 for a * "Workshop on Program for Elimination of Requirements Marginal lo Safety," NURE~/CP-0129. September 1994. PART 50

STATEMENTS OF CONSIDERATION complete battery of Type B/C tests and $1,890,000 for Type A tests. Over the average reactor's remaining lifetime of 20 years, the present value of all remaining containment leakage testing at a 5 percent discount rate is estimated to be about $7 million per reactor. Estimates of the remaining wide costs of implementing current Appendix J requirements ranged from $720 to $1,080 million, approximately 7 s* percent of which could be averted with a performance-based rule. The Regulatory Analysis.for the proposed rule finds that by allowing requirements to remain in effect.with marginal impact on safety, but which impose a significant cost on li_censees, is to have missed an opportunity to
improve regulatory coherence and to focus NRC's regulations to areas where the return in terms of added public safety is higher. Specific alternatives for modifying the current Appendix J_were identified by the public in response to the NRC's Federal Register notice published*

on January 27, 1993 (58 FR 6196). Those whose characteristics matched the NRC's established criteria-for the marginal to safety program were selected for further review. Modifications of Advance NRG Proposal* Allowable Leakage Rate The NRC had initially planned to establish, by rulemaking, a risk-based allowable leakage rate commensurate with its significance to total public risk. Specific findings from-NUREG-1493 on* the allowable leakage rate include: 1. Allowable leakage could be increased approximately two orders of magnitude (100-200 fold) with marginal impact on population dose estimates from reactor accidents.

2. Calculated risks to.individuals are several orders of magnitude below the NRC's Safety Goals for all reactors considered.

3 .-Increases in the allowable leakage rate are estimated to ha_ve a negligible impact on occupational exposure. Relaxing the allowable leakage rate is estimated to reduce future_industry testing costs by$50 to $110 million, a 10 percent decrease in rate testing costs. . A risk-based allowable leakage rate would be based on an evaluation, using PRA, of the sensitivity and significance ofcontainment leakage to-risk, and the determination of an appropriate containment leakage limit commensurate with*its significance to the risk to the public and plant room *operators. However, this would have entailed a major change in policy and restructuring of the current licensing basis and a more complete understanding of the uncertainties asso_ciated with the threat of severe accidents to the containment, and therefore, the NRC planned to develop a modification of the performance standard (allowable leakage level)-in the second phase separate from modifications of testing requirements. This modification would be part of a broader effort to further examine the risk significance of various attributes of containment performance, i.e., structural and leak-tight integrity of containment-system structures and components, *and-inadvertent bypass. On-Line Monitoring (OLM) Systems Currently, there is no NRC requirement for systems which continuously monitor the containment to detect.unintentional breaches of containment integrity. Studies discussed in NUREG-1493, "Performance-Based Containment Leak Test Program," found that, based on operating.experience, OLM would not significantly reduce the risk to the . public from .nuclear-plant operation and, thus, could not.be justified solely on the basis ofrisk-based considerations; Specific findings include: 1. Existing continuous monitoring methods appear technically capable _of detecting leaks in reactor containments within 1 day to several weeks. OLM systems are in use or planned in several European countries and Canada. 2. OLM systems are, capable of detecting leaks only in systems that are .. open*to the ccintainment atmosphere during normal operation (approximately 10 percent of the mechanical penetrations).

3. The.technical and administrative objectives of OLM systems and Type A tests are different.

4. OLM could not be considered as a complete replacement for Type A tests because it cannot challenge the structural and leak-tight integrity of the . containment:system at elevated pressures.

5. Analysis of the.history of operating experience indicated a limited need for, and benefit of, OLM in the*u.S. Although OLM can not be justified solely based on risk considerations, a plant already possessing such a system has a greater assurance of achieving certain attributes of containment integrity.

Therefore, OLM systems could contribute towards an overall monitoring scheme. Some capability for on-line monitoring already exists as a byproduct of specific containment designs. For example, licensees with 50-SC-154 inerted BWR containments, or subatmospheric PWR containments, could possibly detect gross leakages that develop during normal operation. Given that the application of on-line monitoring is specific to containment design, and generic application can not be justified solely on i;isk considerations, the NRc;: did not propose a requirement for OLMs. However, licensees with such a capability (e;g. inerted BWR containments, and subatmospheric PWR containments) were encouraged to propose specific application of such a capability, and to take credit for any added assurance of containment integrity provided by such a system compared to other testing methods. The NRC proposed _to reconsider the role of OLM in the second phase of modifications in this area along with the allowable leakage rate. . Proposed Modification of Type A, B, and C Test Intervals In the February 1995 proposed rule, the NRC proposed a new risk-based regulation based on the-performance history of components (containment, penetrations, valves) as the means to justify an increase in the inferval for Type A, 8, and C tests. The revised regulation requires tests to be conducted on an interval based on the performance of the containment structure, penetrations and-valves without specifying the interval in the regulation. Currently, three Type A tests are conducted in every 10 year period. Type 8 (except airlocks, which are tested more frequently) and C tests are conducted on a frequency not to exceed 2 years. The NRC proposed to base the frequency of Type A tests (ILRTs) _on the historical performance of the overall containment system. Specific findings documented in NUREG-1493 that justify the*proposal include:

1. The fraction of leakages detected only by ILRTs is small, un the order of a few percent. 2.-Reducing the frequency of ILRT testing from 3 every 10 years to 1 every 1 O years leads to a marginal increase in risk. 3. ILRTs also test ths strength of the containment structure.

No alternative to ILRTs has been identified to provide assurance that the containment structure would-meet allowable leakage. rates during design-basis accidents.

4. At a frequency of 1 test every 10 years, industry-wide occupational exposure would be reduced by 0.087 person-sievert (8.7 person-rem) per year. Based on specific, detailed analyses of <lata from the North Anna and Grand I PART 50

STATEMENTS OF CONSIDERATION Gulf nuclear power plants, and data from twenty-two nuclear plants (see
NUREG-1493), performance-based alternatives to current LLRT methods are feasible with marginal impact on risk. Specific findings include: 1. Type B and C tests are capable of detecting over 97 percent of containment leakages.

2. Of the 97 percent, virtually all leakages are identified by LLRTs of containment isolation valves'.(Type C tests). 3. Based on the detailed evaluation of the experience of a single two-unit station, no'Correlation of failures with type of valve or plant service could be found. 4. For the 20 years of remaining operations, changing the Type B/C test frequency to once every 5 years for good-performing components is estimated to reduce industry-wide

occupational radiation exposure by o.n person-sievert (72 person-rem) per year. . If 20-year license extension is assumed, the estimate is 0.75 person-sievert (75 person-rem) per year. .Future industry testing costs are reduced by approximately

$330 to $660 million ifILRT tests*are conducted once every 10 years rather than the current 3 per 10 years. ILRT savings represent about 65 percent of the remaining costs of current Appendix J. requirements. . Performance-based LLRT alternatives are estimated to reduce future industry testing costs by $40 million to $55 millio_n. LLRT savings represent about 5 percent of the total remaining costs of Appendix J testing. Therefore, based on the risks and costs evaluated, and-other considerations discussed above, a *performance-based Appendix J was proposed which encompnssed tho following principles, which differ .moderately from those first described in the Federal Register (January 27, 1993 58 FR 6197).

General (1) Make Appendix J less prescriptive and more oriented; (2) Move details of Appendix J tests to a regulatory guide as guidance; (3) Endorse in a regulatory guide the industry guideline (NEI 94--01) on the conduct of containment tests (The methods for testing are contained in an industry .standard (ANSI/ ANS 56.8-1994) which is referenced in the NE! guideline);

and (4) Allow voluntary adoption of the new regulation, i.e., current detailed requirements in Appendix J will con.tinue to be acceptable for compliance with the modified rule. Leakage Limits Acknowledge the Jess risk-significant nature of allowable containment leakage but pursue its modification as a separate action. Type A Test Interval (1) Based on the limited value of integrated leakage-rate tests (lLRTs) in detecting significant leakages from penl'ltrations and isolation valves, establish the test interval based on the performance of the containment system structure; (2) The performance criterion of the test will continue to be the allowable leakage rate (La); (3) The industry guideline allows extension of the Type A test interval to once every 10 years based on satisfactory performance of two previous tests, inclusive of the pre-operational ILRT; (4) In the regulatory guide, the NRC takes exception to industry guidance for the extension of the interval of the general visual inspection of the containment system, and limits the interval to 3 times every 10 years, in accordance with current practice. Type B Fr C Test Interval (1) Allow local leakage-rate test (LLRTs) intervals to be established based on the performance history of each component; (2) The performance criterion for the tests will continue to be the allowable leakage rate (La); (3) Specific performance factors for establishing extended test intervals (up to 10 years for Type B components, and 5 years for Type C components) are contained in the regulatory guide and industry guideline. In the regulatory guide, the NRC has taken exception to the NE! guideline allowing the extension of Type C test intervals up to 10 years, and ..limits such extensions to 5 years. Summary of Public Comments Twenty-six letters were received that addressed the policy, technical, and cost aspects of the proposed rulemaking, including the nine questions posed by the NRC in the February 21, 1995 .proposed rule. All comments, including the ones received by the NRC after the deadline were considered. The commenters included 4 private citizens, 1 public interest group, 18 utilities, 1 nuclear utility industry group, 1 State regulatory agency. and 1 foreign regulator.

Although the proposed rule did not generate a significant number of public comments, the commenters did align themselves into twa distinct groups: those who supported publishing the rule and those against. Those who supported publishing the rule comprise the vast majority of the commenters (22) and included the Nuclear Energy Institute (NE!). which represents the nuclear utility licensees, eighteen individual nuclear power plant licensee respondents, a Spanish regulatory authority and two private citizens (Mr. 50-SC-155 Hill and Mr. Barkley).

This group is very supportive of the Commission's based regulatory program. and supports proceeding with the rule in an expeditious manner, despite having reservations about three specific provisions. The issues ofmost concern to this group are: (1) Licensee commitments to certain requirements of the regulatory guide implementing Appendix J testing via use of the technical specifications (industry would prefer using a plant's final safety analysis report); (2) requirements to conduct visual internal and extmnal inspections of the containment on a frequency of 3 times per 10 years (industry would prefer once per 10 years to coincide with Type A tests); (3) making Option B of the proposed rule mandatory (industry would prefer to . retain the optional feature); and (4) Type C test frequency (industry would prefer a 10-year test interval for certain Type C valves). Industry supports a future rulemaking to increase the allowable leakage rate. Two private citizens (Mr. Arndt and Dr. Reytblatt) are opposed to the proposed rule. The issues of most concern to these citizens are: (1) Type A test frequency (Mr. Arndt would prefer that frequencies be held at current levels); (2) Type A test methodology (Dr. Reytblatt wants to halt Type A testing until the test accuracy is improved); (3) Type C test frequencies (Mr. Arndt believes the existing database does not support 10-year test intervals, and suggests 5-years as an upper limit at the present time); and (4) Leakage rate (a future rulemaking to increase the allowable leakage rate should not be undertaken). Two organizations are opposed to the proposed rule. The Bureau of Nuclear Engineering of the state of New Jersey and the Ohio Citizens for Responsible Energy (OCRE, represented by Ms. Hiatt). a public interest group, expressed skepticism in the risk-based approach to regulation as embodied in the philosophy of the Marginal-to-Safoty l'rngram. The issuns of most concorn to this group are that: (1) Increases in public risk are not acceptable, no matter how marginal; and (2) A future rulemaking to increase the allowable leakage rate should not be undertaken. NRC Position. With respect to the areas of disagreement between the NRC and those who generally support the proposed rule, no new information has been provided in the public comments that was not already addressed in ongoing dialogue. Accordingly. the NRC has not made any substanti\'e changes to its proposed regulation. Specifically, the NRC has retained: (1) Its position of PART 50

STATEMENTS OF CONSIDERATION requiring the use of technical specifications; (2) The intervals established for visual examinations of containment; and (3) The 5-year Type C test interval.

With respect to the optional feature of the rule, the NRC agrees with the industry and has retained this feature .. With respect to Mr. Arndt and Dr.

Reytblatt, the NRC agrees in part with Mr. Arndt and has decided not to.alter the LLRT test interval as noted in item (3). The other issues raised by Mr. Arndt and Dr. Reytblatt contain no information that has not been considered previously in a public forum. Therefore, the NRC has decided to make no substantive changes to its proposed rule as a result of the issues raised. With respect tci the two organizations opposed to the proposed rule (OCRE and the NJ Bureau of Nuclear Engineering), neither has provided new information or a compelling reason to abandon the based approach to regulation.

In its preliminary criteria for developing performance-based regulations, the NRC identified several issues to be addressed by the rulemak.ing process as a measure of the viability of the revised rule. These issues were addressed in the proposed rule and the NRC sought further public input on them. Comments were received on these topics in addition to other areas of interest to the public. The

following is a summary of comments received on these issues and areas, and NRC's response.

A complete discussion of all comments is included in the Public Comment Resolution Document.5 1. Can the new rule and its implementation yield an equivalent level of, or would it only have a marginal impact on safety? Twenty-four commenters addressed this issue, offering a wide variety of opinions. Twenty comrnenters believe that implementation of the proposed rule will provide an equivalent level of safety to that provided by the current rule. A majority of commenters, representing for the most part nuclear utilities, believe that the proposed regulation will reduce the testing burden currently imposed on the nuclear industry, and will result in more efficient use of utility resources, while ensuring the health and safety of the public. They believe that the practical experience gained from more than 1,500 reactor-years of commercial nuclear power-plant operation provides 'Copies are available for inspection or copying for a fee from the NRC Public Document Room at 2120 L Street NW., Washington, DC; the PDR's mailing address is Mail Stop LL-6, Washington, DC 20555; telephone (202) 634-3273; fax (202) 634-3343. an appropriate basis to adjust the Appendix J testing in~ervals which were established over 20 years ago on the basis of engineeriFlg.judgment. Further, these comrnenters believe that a significant reductiun in occupational exposures can be achieved with reduced testing frequency.

Mr. E. Gunte;i-Arndt, a private citizen, believes that the NRC has neither sufficient objective data nor perspective to jusUfy.inq-easing containment leakage rates, decreasing test frequencies, relaxing testing criteria, and reducing containment-system maintenance standards.

Dr. Reytblatt, a private citizen, believes that Type A testing must be immediately suspended bl:lcause the.current testing methodology is flawed. Mr .. Kent W. Tosch, Manager of New Jersey,'s Bureau of Nuclear Engineering, points out that the containment is an extremely important barrier to a release ofradioactivity, but the.philosophy reflected in this rulemak.ing is that this barrier can be allowed to become less reliable, even . when some nuclear plants are showing signs of aging. Ms. Susan L. Hiatt, Director of Ohio Citizens for Responsible Energy, notes that relaxing the frequency of Appendix J tests leads to an increase in overall reactor risk of approximately 2 percent and, while the NRC may deem.this to be marginal, it nonetheless is an increase in risk: The NRC believes it has.collected sufficient subjective and independent data to conduct its risk analysis. Detailed data from two independent power plants, representing four units, data supplied by the NEI representing approximately 30 additional units, and approximately 180 ILRT and licensee

event reports .were analyzed.

These data produced consistent resu_Jts. Dr. Reytblatt's views, while technically corrl)ct, have been opposed by several technically competent organizations including the American National Standards Institute, and Oak Ridge National Laboratory because the improvements he suggests will have an insignificant effect on measured containment leakage rates in practice and thus have no safety significance. The NRC believes there has b'een ample opportunity for public discussion of the basis for the Appendix J revisions. Based on the foregoing, Jhe NRC reaffirms its prior conclusion (stated in* the February 21, 1995, Federal Register notice) that its safety objective for containment integrity can be maintained while at the same time reducing the burden on licensees. Additionally, the final rule provides a greater level of worker safety than that provided by the previous rule. 50-SC-156

2. Can the regulatory/safety objective (qualitative or quantitative) be established in an objective manner to allow a common understanding between licensees and the NRC on how the performance or results will be measured or judged? To avoid repetition, the NRC incorporated responses to this question with those of Question 3. :I. Can t hu regulation and implementation documents be developed in such a manner that they can be objectively and consistently inspected and enforced against? Approximately 20 commenters expressed opinions on Questions

2 and #3. The majority of the commenters

believe that regulatory/safety objectives can be established objectively, and can be consistently enforced, although opinions differ on the optimum enforcement mechanism.

Mr. Fernando Robledo of the Spanish nuclear regulatory agency states that the use of probabilistic risk assessment in the regulatory process provides a more realistic and objective assessment of nuclear safety, and thus supports its increased use in the regulatory process. The NEI believes the use of technical specifications for inspection and enforcement is neither necessary nor warranted and that, rather than a licensee commitment in the plant technical specification, future licensee commitments to implement Option B should be provided by documentation in the updated Final Safety Analysis Report. To assist in the common understanding of new methods of establishing Type A, 8, and C test frequencies between the NRC and power reactor licensees, the NRC has had ongoing discussions with licensees. These discussions included participation in workshops designed to elicit a common understanding. Also, the NRC wishes to retain the current practice which requires its review and approval of changes to Appendix J performance limits and surveillance requirements. Therefore, the NRC has required that the regulatory guide should be specified in the technical specifications, an approach not inconsistent with the Commission's policy on technical specifications. Based on the foregoing, the NRC reaffirms its pricir conclusion (stated in the February 21, 1995, proposed rule) that it expects that its activities to date, the review and endorsement of a industry guideline in a regulatory guide, and the general reference of the regulatory guide in plant technical specifications, will provide a common PART 50

STATEMENTS OF CONSIDERATION understanding on the measures of compliance.

4. Should the proposed revision be made even less prescriptive?

Except for Mr. Hill and Mr. Barkley, commenters did not t?xplicitly address this question, which was directed at the possibility of reducing. even further, the testing frequency of ILRTs based on the fact that there does seem to be a strong statistical link hntwi,nn passing or failing successivn lLRTs. Mr. I !ill believes that there is no need to make the rule less prescriptive, and it may be inferred that is no desire on the part of industry to further increase the testing interval between ILRTs or to eliminate them completely. Rir.hard Barkley, although strongly supporting an adjustment to tlrn frequency of Type A testing to once every 1 ll years, also discourages the NRC: from adopting a Type A surveillance inti:rval any longer than 10 years because of aging considerations. The NRC has decided. in general, to maintain the present level of prescriptiveness in the proposed rule and, in particular, to not decrease further the test frequency for ILRTs. The NRC's position is guided by the desire to maintain some conservatism to address uncertainties and adopt an evolutionary approach wherein incentives remain for good performance.

5. Should the proposed revisions be made mandatory?

To avoid repetition, the NRC incorporated responses to this question with those of Question 7. 6. Was the definition of "backfit" in § 50.109(a)(1) intended to encompass rulemakings of the type represented by this proposed rule? To avoid repetition, the NRC incorporated responses to this question with those of Question 7. 7. Is it appropriate for the Commission to waive the applicability of the Backfit Rule? The majority of the 20 comnwnters believe that compliance with the performance-based Appendix J program should not be made mandatory. The NE! believes that rulemakings that provide relief from a current regulation but would also contain one or more new requirements (as is the case here) would be subject to the backfit rule. These commenters believe that application of the backfit rule would be necessary before the NRC could promulgate the performance-based Appendix J program as a requirement, believing some licensees might select, for reasons of cost, to continue to comply with the existing Appendix J. The majority of commenters believe that the backfit rule would apply and should not be waived. Several utilities have no objection to waiving a back.fit analysis when clear relief is available, but are concerned with the generic implications of waiving the applicability of th!! backfit rule. The NE! believes that while the proposed Appendix J revisions would provide much needed performance-based improvements to the existing Appendix J, it would also impose new requirements; thus, the proposed rule constitutes a backfit. Further, this

commenter believes that, as a matter of administrative law, an agency lacks authority to depart from its own rules, thus, it cannot waive its own regulations.

The NRC believes that if the rule were made mandatory, all licensees would incur costs setting up the procedures for implementing the rule's requirements following the guidance provided in the regulatory guide and the NE! guidance document. For those utilities whose circumstances (e.g., remaining plant life) would lead them to follow the current Appendix J, costs would be incurred with no additional benefit. Thus, the NRC agrees with the opinions expressed by the NE! and has decided to retain the proposed rule in its present form, which provides a non-mandatory alternative to the current Appendix J requirements. Because the NRC has decided to retain the optional feature of the proposed rule, the question of backfit is not addressed.

8. Should NRC pursue a fundamental modification of its regulations in this area by establishing an allowable leakage rate based on risk analysis (as presented in draft NUREG-1493, Chapter 5), as compared to the current practice of using deterministic design basis accidents and dose guidelines*

contained in 10 CFR Part 100; or should the NRC modify the allowable leakage rate within the current licensing basis by revising source terms and updating regulatory guides (R.G.s 1.3 and 1.4) 6 for calculating doses to the public? What are the advantages and disadvantages of the two approaches? What are some other considerations than risk to public, e.g., plant control room habitability, that might limit the allowable leakage rate? The 20 commenters who responded to this question consist predominantly of the utilities endorsing the NE! position. These respondents encourages the NRC 6 Copics may be purchased at current rates from the Superintendent of Documents, U.S. Go\'ernmcnl Printing Office. P.O. !lox 37082. Washington. DC 20402-9328 [telephone 202-512-2249 or 202-512-2171); or from the National Technical Information Service by writing NTtS al Port Royal Road. Springfield. VA 22161. 50-SC-157 to pursue a rulemaking to alter allowable leakage rates using risk-based analysis, believing that a firm technical basis exists for relaxing leakage rates up to two orders of magnitude with only a marginal impact on population risk estimates. It was also suggested that a review of the present source terms, dose projection models, and associated assumptions against the revised source terms and dose methodologies should also be performed to determine if relief can be achieved while assuring public health and safety. Three commenters discouraged the NRC from relaxing containment leakage rates ranging from the opinion that little benefit would result (Mr. E. Gunter Arndt) to an unequivocal belief that such a move would violate a plant's licensing basis by eliminating the protection provided for the nearest public individual by the 10 CFR Part 100 siting criteria (Ms. S. Hiatt). Ms. Susan Hiatt, representing the Ohio Citizens for Responsible Energy, believes that containment leak rates should be periodically reexamined, not for the purpose of relaxing them, but to determine whether they should be made more stringent given increasing population density around operating nuclear power plants. The NRC has decided to continue to pursue further reductions in regulatory burden with marginal impacts on safety and will address the complexities noted in the public comments in its future efforts to relax the allowable leakage rate. 9. If the allowable leakage rate is increased, could on-line monitoring of containment integrity replace other current containment tests? Could the results of the on-line monitoring be used to establish a new performance basis for containment integrity involving less stringent reporting requirements if there is high assurance there are no large leakage paths in containment (> 1 in. diameter). The 18 commenters who responded to this question consist of the NE! and the utilities endorsing the NE! position, and Mr. Richard Barkley. The commenters do no~-believe that on-line monitoring (OLM) of containment integrity can replace many of the current containment tests, and state that OLM systems have very limited abilities to identify breaches in containment integrity. In the experience of Mr. Barkley, such systems add unnecessary plant complexity and cost. The NRC acknowledges the public comments rendered and will be guided by them in decisions yet to be made regarding the Phase 2 effort. 10. Are there any other regulatory approaches and tech_nical methods by PART 50

STATEMENTS OF CONSIDERATION which the NRC can adopt a complete performance and, risk basis to its regulations for containment leak-tight integrity?

What are some of the attributes for performance, and what risk-based methods can be used to analyze these attributes? The NEI. speaking for all other utilities, addressed this question by stating that it had*not conducted any analyses to determine whether any other regulatory approaches and technical methods by which the NRC can adopt a complete performance and.risk basis to its regulations for containment tight integrity.

11. Rulemaking Documents.

Seventeen commenters expressed opinions about NRC's regulatory policy decisions and/or specific language in the rule or its supporting documents. Mr. Hill believes that the NRC's and the NEl's guidance documents are not developed to the point of establishing a . common understanding of how to rieet NRC's regulatory and safety objectives (e.g., while NEI 94-01 contains a lot of information and solid guidance, it also contains inconsistencies, contradictions and unclear passages). The. NEI, whose comments were endorsed by most responding licensees, proposed modifications to several of the rulemaking documents, including the Federal Register notice and its own guidance document. The NRC has amended its rule and accepts most of the revisions to the implementing documents to clarify

language and achieve consistency between the rulemaking documents;

12. Technical Issues.

Testing Frequency Twenty-four commenters expressed opinions on test frequency, the majority were supportive of 10-year intervals for both Types A, B and C tests. Regarding ILRTs, the Nuclear Energy Institute, several individual utilities, and Mr. Howard Hill expressed views that the proposed rule provides an acceptable testing frequency for ILRTs. Mr. Fernando Robledo, of the Spanish nuclear regulatory agency, believes that 10 years is too long a time interval between Type A containment tests. Mr. E. Gunter Arndt's view is that a preoperational test should not count as one of the two successful ILRT tests required to go to a 10-year test interval because preoperational conditions are not at all representative of operating conditions.

The citizens' group, Ohio Citizens for Responsible Energy, believes the frequency of containment leak-rate testing should remain unchanged from the current practice. Several commenters also expressed opinions on the NRC's position on LLRT testing frequency. Mr. Fernando Robledo, while agreeing in general with the test frequency for"type B and C tests proposed in the draft regulatory guide, believes that certain.mechanical penetrations particularly important for plant safety should be leak tested every 24 months. Mr. E. Gunter Arndt's view is that the testing history of penetrations, and especially of valves, does not support leaving them untested for 10 years and suggested that an upper limit should be once every 5 years. One utility in particular, and the Nuclear

Energy Institute in.general believe that the NRC does not go far enough in citing that several sets of data justify 10-year LLRT intervals.

In contrast, Mr. Richard Barkley, who also endorses Type B & C testing frequency based on performance, strongly supports the NRC's proposal to prohibit the adoption of Type C surveillance intervals longer than 60 m_onths. In establishing the 5-year test interval for LLRTs, the NRC has designed a cautious, evolutionary approach as data are compiled to 111inimize the uncertainty .now believed to exist with respect to LLRT data. The NRC's judgment, based on risk assessment and deterministic analysis, continues to be that the limited database on unquantified leakages and common mode and repetitive failures introduces significant uncertainties into the probabilistic risk analysis; The NRC will be open to submittals from licensees as more performance-based data are developed. The extension of LLRT test interval to 5 years is a prudent first step. By allowing a 25 percent margin in testing frequency requirements, the NRC has provided the flexibility to accommodate longer fuel cycles. With respect to the 10-year interval for ILRTs, the NRC believes its technical support document (NUREG-1493) is persuasive by demonstrating that testing intervals could be increased up to once every 20 years with an imperceptible increase in risk, using actual ILRT data which

accounted for random and plant-specific failures and plant aging effects. Based on the foregoing discussion, the NRC has decided to retain the 60-month Type C test interval and the 120-month
interval for Type A and B tests. In response to public comments, the NRC -has revised the regulatory guide to limit the extension of test intervals for main steam and feedwater isolation valves in BWRs, arid containment purge and vent valves in PWRs and BWRs beyond 30
months given their operating experience and/or safety significance.

50-SC-158 Test Pressures Two commenters expressed opinions on the magnitude of the pressures used in conducting Type A leakage tests. Northern States Power Company_: believes that Type A testing at full pressure is unnecessary and believes that visual inspection coupled with a reduced pressure test will adequately assure that the containment structural members are leak-tight, especially since.* reduced pressure Type A tests are legally acceptable tests as prescribed in the current 10 CFR Part 50, Appendix J. Mr. E. Gunter Arndt states that while Type* A tests performed at reduced pressure rather than peak accident pressure are economically advantageous to the industry, the results of these tests are not necessarily indicative of leakage rates during accidents. The NRC believes that extrapolating low pressure leakage-test results ,to full pressure leakage-test results has turned out to be unsuccessful. The NRC

believes that the peak*calculated accident pressure:

(1) Is consistent with the typical practice for NRC staff . evaluations of accident pressure for the first 24 hours in accordance with Regulatory Guides 1.3 and 1.4; (2) Provides at least a nominal check for gross leak paths which might exist at high test pressures, but not at low test pressures; and (3) Directly represents technical specification leakage-rate limits, and provides greater confidence in conta,inment system leak-tight, integrity. . Based on the foregoing, the. NRC has decided to retain the calculated design basis loss-of-coolarit accident peak pressure as the ILRT test pressure.

containment Inservice Visual Inspection Eighteen commenters expressed opinions *on this issue._The NEI and most utilities oppose the NRC'.s. proposal to require visual examination of containment be performed 3 times every 10 years. These commenters suggest that this issue be taken up in a parallel rulemaking.

The NRC finds the industry's , arguments for relaxing the frequency of containment visual inspections to be unpersuasive .. Because the visual . examination is not integral to the ILRT (i.e., may be performed independently) and because the NRC sees benefits to.the early detection of unknown aging mechanisms which may be active, the NRC considers it prudent to conduct visual inspections on a frequency greater than the ILRT. Further, the NRC believes it is inappropriate to defer a requirement pertaining to containment PART 50

STATEMENTS OF CONSIDERATION strudural integrity to an ongoing rulemaking to incorporate ASME Section XI, IWE and IWL until its form and substance is finalized.

Based on the foregoing, the NRC has decided to retain its frequency for the inservice visual inspection. Reporting Requirements . Only one comment was received on this issue. Dr. Z. Reytblatt noted that the proposed rule's reporting requirements consist only of a c"over letter to the NRC and suggested this is intended to conceal information from the public. Dr. Reytblatt suggests that utilities should be required to submit ull computer files related to testing to the NRC immediately after the tests have been completed to prevent their alteration or destruction. It is not the intent of the NRC's reporting requirements to conceal information from the public; if tests fail, the information is required to be reported to the NRC, and the NRC will make such data avaiJable to the public. The NRC has decided to retain its reporting requirements as stated in the proposeq rule. Modifications to the Proposed Rule in Response to Public Comments The NRC has decided to amend its proposed rule and its implementing docu~ents tci clarify language. The NRC

has concluded that its regulatory analysis and its technical support document, NUREG-1493, do not require corrections to its technical or cost analyses or its findings.

Modifications to all documents will be restricted to clarifications and enhancements to assist in communications with the reader, specifically in areas discussed in the public comments. The proposed rule has been modified by changing "Acceptance criteria" to "Performance criteria" in Section II, Definitions, and various conforming text changes to reflect consistent use of that term. Other similar redundant terms in the proposed rule, e.g. goals, have been deleted to establish clear and concise language in the rule. Specific changes to the draft regulatory guide, Section C, Regulatory Position, include (1) in paragraph number 2, the inclusion of the rationale for denying the "3 refueling cycle" change requestec:l in the public comments; (2) the inclusion of a new paragraph number 4, taking exception to the NE! Industry Guideline, Section 10.2.3.3, which provides guidance that an as-found Type C test or an alternative test or analysis (emphasis added) shall be performed prior to any maintenance. repair, modification, or adjustment activity if it could affect a valve's tightness. "Alternate test or analysis" are not endorsed as appropriate substitutes for an as-found test, since the latter provides clear and objective evidence of performance of isolation components; and (3) limitation of the extension of test intervals for main steam and feedwater isolation valves in BWRs, and containment purge and vent valves in PWRs and BWRs beyond 30 months given their operating experience and/or safety significance. Regulatory Guide; Issuance, Availability The Nuclear Regulatory Commission has issued a new guide in its Regulatory Guide Series. This series has been developed to describe and make available to the public such information as methods acceptable to the NRC staff for implementing specific parts of the Commission's regulations, techniques used by the staff in evaluating specific problems or postulated accidents, and data needed bv the staff in its review of applications for permits and licenses. Regulatory Guide 1.163, "Performance-Based Containment Leakage-Test Program," endorses an industry standard which contains guidance on an acceptable based leakage-test program, leakage rate test methods, procedures, and analyses that may be used to implement the final regulation published in this notice. Comments and suggestions in connection with items for inclusion in guides currently being developed or improvements in all published guides are encouraged at any time. Written comments may be submitted to the Rules Review and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission. Washington, DC 20555. The NRC staff's response to public comments received on the draft version of this guide (DC'~1037. issued in Februar\' 1995) are available for inspect(on or copying for a fee in the NRC Public Document Room, 2120 L Street NW .. Washington, DC. Regulatory guides are available for inspection at the Commission's Public: Document Room. 2120 L Street N\V .. Washington. DC. Single copies of regulatory guides may be obtained frP.e of charge by writing the Office of Administration. Attention: Distribution and Services Section. U.S. Nuclear Regulatory Commission. Washington. DC 20555-0001; or bv fax at (301) 415-2260. Issued guides rnay also be purchased from the National Technical Information Service on a standing order basis. Details on this service may be 50-SC-159 obtained by writing NTIS, 5285 Port Royal Road, Springfield, VA 22161. Regulatory guides are not copyrighted, and Commission approval is not required to reproduce them. Implementation The proposed Option B to Appendix J will become effective 30 days after publication. At any time thereafter, a licensee or applicant may notify the NRC of its desire to perform containment leakage-rate testing according to Option B. Accompanying

this notification, a licensee must submit proposed technical specifications changes which would eliminate those technical specifications which implement the current rule and propose a new technical specification referencing the NRC regulatory guide or, if the licensee desires, an alternative implementation guidance.

Implementation must await NR<;: review and approval of the licensee's proposal. The NRC anticipates that a generic communication will be issued shortly which will provide the implementation procedure to all power reactor licensees. Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969. ns amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment, and therefore an environmental impact statement is not required. There will be a marginal radiological environmental impact offsite, and the occupational exposure onsite is expected to decrease by about 0.8 person-rem per year of plant operation for plant personnel if licensees adopt the performance-based testing scheme provided in the revised rngulation. Alternatives to issuing this revision of the regulation were considered. One alternative would also entail complex revisions to other NRC regulations and therefore the NRC has decided to pursue it separately in the future. A third alternative would add regulatory burden without a commensurate safety benefit and therefore was found not to be acceptable. The environmental assessment is available for inspection or copying for a fee in the NRC Public Document Room, 2120 L Street NW, (Lower Level). Washington, DC; the PDR's mailing address is Mail Stop LL-6, Washington, DC 20555; phone (202) 634-3273;fax (202) 634-3343. PART 50

STATEMENTS OF CONSIDERATION Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 lJ.S.C. 3501 et seq.). These requirements were approved by the Office of Management and Budget, approval number 3150-0011. . Becau~e the rule will relax existing mformat10n collection requirements by providing an option to the existing requirements, the public burden for this collection of information is expected to be reduced by approximately 400 hours per licensee per year. This reduction

!nclude~ the time required for reviewing mstruct1ons, searching existing data sourctis, gathering and maintaining the data needed and completing and reviewing the collection of information. Send comments regarding the estimated burden reduction or any aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011). Office of Management and Iludgnt, Washington. DC 20503. Regulatory Analysis The Commission has prepared a final regulatory analysis on this regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The analysis is available for inspection or copying for a fee in the NRC Public Document Room, 2120 L Street NW, (Lower Level), Washington, DC; the.PDR's mailing address is Mail Stop LL-6, *washington, DC 20555; phone (202) 634-3273; fax (202) 634-3343. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, (5 U.S.C. 605(b)), the Commission certifies that this rule will not, if promulgated, have a significant economic impact on a substantial number of small entities. This rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Size standard adopted by the NRC (10 CFR 2.810). Backfit Analysis This final rule amends a current regulation by establishing alternative requirements which may be voluntarily adopted by licensees. Therefore, the final rule does not constitute a backfit as defined in lOCFR 50.109(a)(l). Therefore, a backfit analysis is not necessary. List of Subjects in 10 CFR Part 50 Antitrust, Classrned information, Criminal penalties, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendments to 10 CFR Part 50. 60 FR 53505 Published 10/16/95 Effective 11 /15/95 10 CFR Parts 50, 70, and 72 RIN 3150-AF27 Physical Security Plan Format Changes AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations to eliminate the requirement for applicants for power reactor, Category I fuel cycle, and spent fuel storage licenses to submit physical security plans in two parts. This action is necessary to allow for a quicker and more efficient review of the physical security plans. EFFECTIVE DATE: November 15, 1995. FOR FURTHER INFORMATION CONTACT: Carrie Brown, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-8092. SUPPLEMENTARY INFORMATION:

Under current NRC regulations, applicants for power reactor, Category I fuel cycle, and spent fuel storage licenses must submit physical security plans in two parts. Applicants for power reactor, Category I fuel cycle, and spent fuel storage licenses are required to address, in Part 1 of their plans, how they will comply with the applicable regulations of 10 CFR Parts 11 and 73. They are required to list, in Part 2 of their plans, any test, 50-SC-160 inspections, audits and any other means to be used to demonstrate compliance with the regulations. The two-part format is restrictive and has no regulatory advantage. Existing licensees with physical security plans approved before the effective date of the final rule will not be required to adopt the new format. These licensees, however, may revise their plans on a voluntary basis, pursuant to the rules that permit licensees to make changes in security plans that do not decrease the effectiveness of the plans. This final rule will not change any of the substantive content currently required in the physical security plans. The benefit of this rulemaking is the elimination of an unnecessary

requirement and there are no expected*

adverse impacts. For those licensees who desire to revise their physical security plans, the staff has revised Regulatory Guide, 5.52, "Standard Format and Content of a Licensee Physical Protection Plan for Strategic Special Nuclear Material at Fixed Sites (Other than Nuclear Power Plants)," for use as guidance. NRC encourages applicants or licensees to follow such guidance in order to allow for a quicker and more efficient review of the plans. Summary of Public Comments The comment period for the proposed rule published April 17, 1995 (60 FR 19170), closed on May 17, 1995. Two comments were received. The following comment summary and resolution address these comments. Comment. This commenter complimented NRC for eliminating . unnecessary requirements and commented on one statement, in the "Supplementary Information" section, that says licensees may"* *

revise their plans on a voluntary basis pursuant to the rules that permit licensees to make changes in security plans that do not decrease the effectiveness of the plan." The commenter discussed a Generic Letter that is being developed by the Office of Nuclear Reactor Regulation entitled, "Standardization of Security Program Reviews," and encouraged the issuance of the draft Generic Letter for comment as soon as possible.

Response. The Generic Letter was published in the Federal Register on June 14, 1995 (60 FR 31326), with a 30-day comment period. Comment. This commenter noted that a similar requirement to submit physical security plans in two parts in 10 CFR 72.180 was not addressed and indicated that it should be included. Response. NRC agrees with the PART 50

STATEMENTS OF CONSIDERATION comment and 10 CFR 72.180 has been amended. Environmental Impact: Categorical Exclusion NRC has determined that this final rule is the type of action described as a categorical exclusion in 10 CFR 51.22(c)(2).

Therefore, neither an environmental impact statement nor an environmental assessment has been prepared for this final rule. Paperwork Reduction Act Statement This rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501, et seq.). Existing requirements were approved by the Office of Management and Budget, approval numbers 3150-0009, 0011, and 0132. Regulatory Analysis The Commission has not prepared a regulatory analysis on this regulation because the amendment does not involve a question of policy, will have no impact on public health and safety, and will require no additional burden on current licensees. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission certifies that this final rule does not have a significant economic impact on a substantial number of small entities. This final rule affects applicants for power reactor, Category I fuel cycle, and spent fuel storage licenses. Because these licensees are not classified as small entities, as defined by NRC's size standards (10 CFR 2.810), the Commission finds that this final rule does not have a significant economic impact on a substantial number of small entities. Backfit Analysis NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this final rule, and thereTcire: tluif a -. backfit analysis is not required, because this amendment does not involve any provisions that would impose backfits, as defined in 10 CFR 50.109(a)(l). List of Subjects 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. 10 CFR Part 70 Criminal penalties, Hazardous materials transportation, Material control and accounting, Nuclear materials, Packaging and containers, Radiation protection, Reporting and recordkeeping requirements, Scientific equipment, Security measures, Special nuclear material. 10 CFR Part 72 Manpower training programs, Nuclear materials, Occupational safety and health, Reporting and recordkeeping requirements, Security measures, Spent fuel. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, NRC is adopting the following amendments to 10 CFR Parts 50, 70, and 72. 50-SC-161 60 FR 65456 Published 12/19/95 Effective 1 /18/96 10 CFR Part 50 RIN 3150-AD57 Fracture Toughness Requirements for Light Water Reactor Pressure Vessels AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations for light-water-cooled nuclear power plants to clarify several items related to the fracture toughness.

requirements for reactor pressure vessels (RPV). The amendments will clarify the pressurized thermal shock (PTS) requirements, make changes to the Fracture Toughness

Requirements and the Reactor Vessel Material Surveillance Program Requirements, and provide new requirements for thermal annealing of a reactor pressure vessel. EFFECTIVE DATE: January 18, 1996. FOR FURTHER INFORMATION CONTACT: Alfred Taboada, Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-00001, telephone:

(301) 415-6014. SUP.PLEMENTARY INFORMATION: On October 4, 1994 (59 FR 50513), the NRC published in the Federal Register a proposed amendment to clarify several items related to fracture toughness requirements.for reactor pressure vessels (RPV) and to add a new section on thermal annealing of a reactor vessJl to 10 CFR Part 50. Background Maintaining the structural integrity of tho reactor prossuro vessel of water-cooled reactors is a critical concern related to the safe operation of nuclear power plants. To assure the structural integrity of RPVs, NRC regulations and regulatory guides have been developed to provide analysis and measurements methods and procedures to establish that each RPV has adequate safety margin for continued operation. Structural integrity of a RPV is generally assured through a fracture mechanics evaluation, including measurement or estimation of the fracture toughness of the materials which compose the RPV. However, the fracture toughness of the RPV materials varies with time. As the plant operates, neutrons escaping from the reactor core impact.the vessel heltlino materials (e.g. the materials thut PART 50

STATEMENTS OF CONSIDERATION surround the reactor core), causing embrittlement of those mntorials.

The NRC's regulations and regulatory guides related to RPV integrity provide.the

- * . criteria and methods needed to estimate.

the extent of the embrittlement, to evaluate the consequences of the erilbrittlement in temis of the structural integrity.of the RPV, and ~o provid11. methods to mitigate the deleterious

effects of the embrittlement.

The NRC has several regulations and regulatory guides that establish criteria and procedures for assuring the . structural integrity of RPVs. With the addition of the thermal annealing requirements in this rule and several . regulatory guides, the regulatory

- documents contribute to a comprehensive set of regulations and . regulatory guidance pertaining to RPV . integrity.

This final rule adds requirements for thermal annealing of the RPV as a , method for mitigating the. effects of neutron irradiation (10 CFR 50.66) and amends the following:

1. The Pressurized Thermal Shock (PTS) rule (10 CFR 50.61). 2. Appendix G of 10 CFR Part 50, "Fracture Toughness Requirements." 3. Appendix H of 10 CFR Part 5Q, "Reactor Vessel Material Surveillance Program Requirements." . . Overview of the Final Rule PTS Rule (1o'CFR 50.61) This arilend~ent to the PTS ruie

makes three changes: . 1. The rule incorporates in total, and therefore makes binding by rule, the method for determining the reference temperature, RT NDT, *including
treatment of the unirradiated RT NDT .
value, the margin term, and the explicit *. definition of "credible" surveillance

' data, which is currently described.in . *

Regulatory Guide i.99, Revision 2. . . *.* 2. The section is restructured to * . improve clarity, with the requirements section giving only the requirements for the value for the refe~ence temperature
for end of life fluence, RTPTs* The method for calculating RT PTS is m*oved to a new paragraph of the rule .. 3. Thermal annealing is i.dentified as a method for mitigating the effects of neutron irradiation, thereby reducing RTPTS*. . . . Thermal Annealing Rule (10 CFR 50.66) The thermal annealing rule, 10 CFR 50.66, provides a consistent set of . requirements for the use of thermal. annealing to mitigate the effects of neutron irradiation and replaces the requirements for annealing in the . current Appendix G of 10 CFR Part 50. Tho final rule requires, prior to initiation of thermal annealing, submittal of a Thermal Annealing Report containing:

(1) A.Thermal Annealing Operating Plan, (2) a Requalification IQspectiori and Test Program, (3) a Fracture Toughness Recovery and Reembrittlement Trend Assurance Program, arid (4) Identification of Unreviewed Safety . Questions and Technical Specifications . Changes. The report must be subltiitted at least 3 years before the date at which the limiting fracture toughness criteria in 50.61 and Appendix G*to Part 50 would be exceeded. This 3*year period is specified to provide the NRC stuff with sufficient time to review the thermal annealing program. Under § 50.66(a); the NRC will, within three years of submission of a licensee's .Thermal Annealing Report, document its views*on the plan; including whether thermal annealing constitutes.an unreviewed safety question. . In order to provide for public participation in the regulatory process, Section 50.66(f)(1) requires that the NRC hold a public meeting a minimum of 30 days before the licensee starts t~ thermal anneal the reactor vessel. The Commission will notify and solicit comments from cognizant local and state governments, and will publish a notice in the Federal Register and in a . forum, such as local newspapers, which is readily accessible to ip.dividuals in the vicinity of the-site, in order tQ solicit comments from the public. The thermal annealing operating plan must include an evaluation of the effects of temperature, and of mechanical and thermal stresses on the reactor and associated equipment such as containment 1 the biological shield, and attached piping, to.demonstrate that the

operability of the reactor will not be detrimentally affected.

The bounding conditions of the temperatures and times used in this. analysis define the proposed annealing conditions. If these conditions are exceeded during the vessel annealing, then the evaluation would no longer be valid, and the acceptability of the actual vessel annealing would have to be demonstrated as discussed below in the

next paragraph.
Upon completion of the thermal . annealing, the li.censee must confirm in
writing to the Director, Offil:;e of Nuclear Reactor Regulation (NRR), that the thermal annealing was performed in accordance with the Thermal Annealing Operating Plan and the Requalification . Inspection and Test Program. Within 15 days of the licensee's written confirmation that the thermal annealing was completed*in accordance with the 50-SC-162 Thermal Annealing Plan, and prior to restart, )he NRG shall: ( 1) Briefly document whether the thermal *
annealing was perfornt'ed in corilplilince*

with the licensee's Thermal Annealing Operating Plan and the Requalification {nspection and Test Program, with ihe

documentation to be placed in the NRC public document room, and (2) hold a public meeting to: (1) permit the licensee to explain the results of the
reactor vessel annealing to the NRC and the public, (2) allow the NRC to discuss its inspection of the reactor vessel annealing, and (3) provide an
opportunity for the public to comment to the NRC on the. ilierina\

annealing. . . The licensee may. restart_ its teactor.af\er the meeting has been completed, unless the NRC orders otherwise. Within 45 days of the licensee's written confirmation that the themial annealing was completed in accordance with the Thermal Annealing Operating plan and the Requalificatfon Inspection and Test Program, the NRC staff shall complete full documentation of the NRC's *

inspection of the" licensee's .annealing process and place the documentation in the Public Document.

Room *. If the thermal annealing was completed but not performed in accordance with the*Thermal Annealing Operating Plan and the Requalification Inspection.and Test Program, including the bounding conditions of the temperature and times as discussed above, the licensee must submit a

summary of lack of corlipliance and a . j\lstification for subsequent operations.
The licensee must also identify any changes to the facility which ere attributable to the-noncompliances . which constitute unreviewed safety questions and any changes to the technical specifications which are required for operation as a result of the noncompliances:

This identification does not relieve the licensee from complying with applicable requirements of the Commission regulations and the operating license, and if,.as a result of . the annealing operation, *these requirements cannot be met, the licensee must obtain the appropriate exemption per 1'0 CFR 50.12. If unreviewed safety questions or changes . to technical specifications are not

identified as necessary for resumed operation, the licensee may restart after the NRC staff places e summary of its inspection of the thermal annealing in the Public Document Room, and the NRC holds a public meeting on the thermal annealing.

On the other hand, if urireviewed safety questions.or changes to technical specific;ations are identified as necessary for resumed operation, the PART 50

STATEMENTS OF CONSIDERATION licensee may restary only after the Director of NRR authorizes restart, the summary of the NR:C staff inspection is placed in the public document room, and a public meeting on the thermal annealing is held. ;
The fimd Thermal Annealing Rule also sets forth the requirements that a licensee must follow if the thermal annealing was terminated prior to completion.

In general, the process and requirements for partial annealing are analogous to the situations where the thermal annealing was completed; viz., where the partial aµnealing was otherwise performed in compliance with the Thermal ~nnealing Operating Plan and relevant portions of the Requalification Inspection and Test Program, the licenl!ee submits written confirmation' of such compliance and may restart following, inter alia, holding of a public meeting on the annealing. By contrast, where the partial annealing was not performed, in accordance with the Thermal Annealing Operating Plan. and relevant portions of the Requalification Inspection and Test Program, the licensee is required to submit a summary of lack of compliance and a justification for subsequent operations, and identify any changes to the facility which ii.re attributable to the noncompliances which constitute unreviewed safety questions and changes to the technical specifications which are required for operation as a result of the nonci>mpliances with the Thermal Annealing Operating Plan and relevant portions of the Requalification Inspection and Test Program. If Unreviewed Safety Questions and/or changes to technical specifications are identified as necessary for resumed operation, the licensee may restart only after the Director of NRR authorizes restart and the public meeting on lhe thermal annealing, is held. Every licensee that either completes a thermal annealing or terminates an annealing but elec,ts to .take full or partial credit for the annealing shall provide a Thermal Annealing Results Report detailing: (1) The time and temperature profile of the actual thermal anneal, (2) the post-anneal RTNDT and Charpy upper shelf energy values of the reactor material to be used in subsequent operations, (3) the projected post-anneal reembrittlement trends for both RTNDT and Cherpy upper-shelf energy, end (4) the projected values of RTP"Ts and Charpy upper-shelf energy at the end of the proposed period of operation addressed in tho application. The report must be submitted within three months of C(?mpleting the thermal anneal, unless an extension is authorized by the:Oirector, NRR. Two items of particular importance to the overall annealing are .the recovery of fracture toughness and the degree of reembrittlement of the RPV beltline materials. This final rule provides alternative metlwds for detennining these values, ranging from assessments using plant-specific materials to an assessment using u generic computation. Two methods provided for evaluating annealing recovery are experimental methods to determine plant-specific annealing recovery.and a third method is a generic computational method. Experimental methods and the computational method are-also provided

for estimating recovery of RT NDT and Charpy upper-shelf energy of the beltline materials.

The experimental methods for estimating recovery of RTNDT*and the Charpy upper-shelf energy utilize either surveillance program specimens or material removed from the vessel beltline. The experimental methods provide a specific estimate of recovery, rather than

the generic value evaluated from the computational method. This final rule re'luires that surveillance specimens from "credible" surveillance programs must be used to develop plant-specific recovery data, if such specimens are available.

This final rule does not require the removal of material from the RPV beltline to permit plant-specific evaluation of recovery. As described previously, the computational method requires appropriate justification. Post anneal reembrittlement trends of both the RTNDTand the Charpy upper shelf energy must be estimated and monitored using a surveillance program described in the Thermal Annealing Report. The reactor pressure vessel is perhaps the most important single component in the reactor coolant system. As such, ensuring its integrity is a fundamental element of plant safety. Thermal annealing is a positive action that could be taken to reduce the level of embrittlement in the pressure vessel beltline and, thereby, improve the ability of a pressure vessel to withstand accident loadings. While thermal annealing is a positive action, there are* numerous complex technical questions regarding its application in the U.S. that are unanswered. Thermal annealing of a commercial reactor pressure vessel has never been accomplished in the United States. Thermal annealing has been successfully employed in Eastern Europe and Russia on Russian-designed pressure vessels. However, there are significant differences between the U.S. and Russian designs in terms of the 50-SC-163 geometry of the pressure vessels, the attached piping, and the surrounding structures. The staff has observed one of these annealing operations. While informative, the East European and Russian experience does not provide answers to all of the potential questions related to annealing of U.S. designed pressure vessels. Research analyses performed previously indicated the potential for plastic deformation of the main coolant piping for a typical U.S. plant design .. and anticipated annealing conditions. There are also questions regarding how thermal growth of the pressure vessel is treated, and the adequacy of the thermal and stress analyses used to predict response of the overall system under thermal annealing conditions. Additionally, there may be questions in other areas such as temperature limits for the concrete structures, and potential radiological hazards associated with removing and storing the reactor internals during the annealing process, and fire hazards associated with heating the vessel. Recognition of Lhe numerous complex technical questions related to thermal annealing, and of the potential benefits for operating nuclear power plants, has resulted in a cooperative effort, funded by the U.S. Department of Energy and the industry, to perform Annealing Demonstration Projects. Projects are planned t_o demonstrate two different annealing processes, evaluating heater designs and vessel designs. It is anticipated that the annealing demonstration projects will answer many of the generic questions regarding thermal annealing of U.S. pressure vessel and piping designs. The thermal annealing report, required by the thermal annealing rule; is designed to facilitate a detailed review by the licensee of plant-specific questions and considerations in . performing a thermal annealing. The proposed rule specifically discusses the potential for unreviewed safety questions and technical specification changes that may result from or be related to thermal annealing of the reactor pressure vessel. With completion of the demonstration projects and as the staff and industry gain experience with thermal annealing, many of the issues related to annealing will be better understood and related questions will be answered. However, until this experience is realized, the staff will critically review licensee determinations regarding unreviewed safety questions and the need for technical specification changes associated with each proposed thermal annealing. PART 50

STATEMENTS OF CONSIDERATION The thermal annealing rule has been
structured to provide time. for. the staff to thoroughly review the licensee's annealing plan and determination.

regarding unreviewed safety questions and the need for technical specification changes. If the staff identifies an unreviewed safety question or the need for a technical specification change, the licensee would be so notified and the existing NRC regulatory prnctice*s would be invoked to address the issues. . Appendix G of 10 CFR Part 50 . Appendix G of 10 CFR Part 50 specifies fracture toughness . requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light-water-cooled nuclear power reactors. These requirements provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests. The amendments to Appendix G are principally of a clarifying or a restructuring nature. Requirements for "volumetric inspection" and "additional evidence of fracture toughness" have been removed because they were unnecessary, given the inspection and performance demonstration programs currently required under 10 CFR 50.55a. The "additional evidence of fracture toughness" requirement in Section V.C.2 is incorporated in the "equivalent margins" analysis in Section IV.A.1 as a provisional method for developing fracture toughness data needed for thiat analysis. The pressure-temperatu~ and minimum permissible temperature requirements in Section IV have been restructured. The principal feature is the addition of a table which summarizes the pressure-temperature limit requirements and minimum temperature requirements as a function of the plant operating condition, the vessel pressure, whether fuel is in the vessel, and whether the core is critical. . In addition, Section IV has been reworded to clarify the minimum permissible temperature requirement by indicating the criteria for use in determining the location in the component or material which must satisfy the minimum temperature requirement. This minimum temperature is defined in Section IV as the metal temperature of the controlling material in the region which has. the least favorable combination of stress and temperature for the appropriate plant condition. An explicit statement has been added to require that pressure and leak tests of the reactor pressure v~ssel required by Section XI of the American Society of Mechanical Engineers Boiler & Pressure Vessel (B&P.V) Code (ASME Code) must be *completed before the core is critical. The requirement that all pressure and leak tests of the RPV required by Section XI of the ASME Code must be completed before the core is critical is intended to prohibit the use of nuclear heat, i.e., core criticality, in tho conduct of ASME, Section XI pressure and leak tests. The use of nuclear heat before the completion of such tests is not consistent with basic defense-in-depth nuclear safety principle for several reasons, including the hindrance of finding leaks with the vessel at such a high temperature an_d the potential for exacerbating the consequences of a vessel rupture (in the extremely unlikely event that it should occur) by having the core critical. The *explicit prohibition of nuclear heat in these cases was discussed in a letter to Messi:s, Reynolds and Stenger of the Nuclear Utility Backfilling and Reform Group from James M. Taylor, Executive Director of Operations, dated February 2, 1990. The current requirements in 10 CFR Part 50, Appendix G, Section V. D. with respect to reactor vessel thermal annealing are being replaced by a sentence which references the new Thermal Annealing rule, 10 CFR 50.66 .. Appendix Hof 10 CFR Part 50 Appendix H of lo CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements" provides the rules for monitoring the changes in the fracture toughness properties of the RPV beltline materials due to irradiation embrittlement using a surveillance progr_am. Appendix H references American Society for Testing and Materials (ASTM) standard E 185 ("Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels") for many of the detailed requirements of surveillance programs, and permits the use of integrated surveillance programs, wherein surveillance program capsules for one reactor are irradiated in another reactor. Integrated surveillance programs are permitted under Section II.C of Appendix H of 10 CFR Part 50. One provision of this section is that "the amount of testing may be reduced if the initial results agree with predictions." This provision was deleted, although previous authorizations granted by the Director, Office of Nuclear Reactor Regulation, continue in effect. A second change to Appendix H restructures Section 11.C to clarify the 50-SC-164 requirements for integrated surveillance programs, The other principal change to Appendix H clarifies the version of ASTM Standard E 185 that applies to the various portions of the surveillance programs. Appendix H recognizes the need to separate surveillance programs into two essential parts, specifically the design of the program and the *

subsequent testing and reporting of
results from the surveillance capsules.

Because the design of the surveillance program cannot be chariged once the pi:ogramjs in place, the requirements for design of the surveillance program are static for each plant. However, the testing and reporting requirements are updated along with technical improvements made to,ASTM standard E 185. Request for Public Comments At the request of the Commission, the proposed rule contained a request for public comments on the following specific issues related to the proposed regulation on thermal annealing:

1. The technical adequacy of the staffs guidance;

2. The sufficiency of the guidance and criteria to support a: certification that if satisfied, a plant with an annealed
vessel can safely resume operation;

3. Whether health and safety concerns are best served by approval of the thermal annealing plan or of readiness for restart; 4. The preferred regulatory process (including opportunities for public participation) and the commenter's basis for recommending a particular process; and 5. Whether there are health and safety issues concerning thermal annealing that cannot be addressed generically and would warrant plant-specific consideration.

The supplementary information section of the proposed rule also discussed the issue of opportunity for public participation in regulating thermal annealing of pressure vessels; The response to the request for public comments on the.se -issues, along with other items, are summarized below. Summary of Comments The following includes a summary of the comments receiv.ed on the proposed rule, on the five issues identified by the Commission, and ~n the options for public participation in thermal ann.ealing. Comments were received from nine separate sources. These sources consist of five utilities, the Nuclear Energy

Institute (NEI), the Nuclear Utility Backfitting and Reform Group PART 50
STATEMENTS OF CONSIDERATION (NUBARG) represented by the firm Winston & Strawn, one public citizens group (Ohio Citizens for Responsible Energy (OCREJ), and one nuclear steam system supplier (NSSS). NEI provided detailed comments on 10 CFR 50.61, 10 CFR 50.66, Appendix G to 10 CFR Part 50, and Appendix H to 10 CFR Part 50; responded to the request for comments on the five issues related to thermal annealing and included detailed comments on the opportunities for public participation.

The five utilities and the NSSS endorsed the NEI comments. Three of the five utilities provided additional comments on 10 CFR 50.61; one of the five utilities provided additional comments on 10 CFR Part 50, Appendix G; two of the utilities provided additional comments on 10 CFR Part 50, Appendix H; and one of the five utilities disagreed with the NEI position on the opportunity for public participation and submitted a separate comment. OCRE provided comments on the opportunity for public participation. NUBARG provided comments on the backfitting aspects of the proposed rule and the stafrs backfit justification. NEI and one of the utilities included comments on the Draft Regulatory. Guide DG-1027, "Format and Content of Application for Approval for Thermal Annealing of Reactor Pressure Vessels," that was discussed in the proposed rule. These comm.ants on Draft Regulatory Guide DG-1027 are being reviewed by the NRC staff and will be addressed separately in the resolution of comments on the regulatory guide. The NRC reviewed the comments received on the proposed rule, the comments on the five questions related to thermal annealing and the issue of opportunities for public participation. The resolution of these comments is presented below. PTS Rule (10 CFR 50.61) Sixteen specific comments in the submittals from NEI and three utilities addressed 10 CFR 50.61. A general comment argued that both the existing 10 CFR 50.61 and the proposed modifications contained an excessive amount of prescriptive technical detail that limits licensee compliance flexibility. The commenters prqposed that these prescriptive technical details be removed from the rule and placed in a regulatory guide. These commenters suggested that the rule not be issued until it has been written to contain only those requirements essential to regulate reactor pressure vessel embrittlement. A number of comments suggested changes that were clarifications to the proposed rule, including proposals to clarify the procedure for calculating the reference temperatures in the preservice condition, RT NOT, and, at end of reactor life, RT PTS* One comment noted that the proposed rule omitted part of the procedure in Regulatory Guide 1.99, presently being applied by the NRG, that permits adjustments for differences in chemistry between surveillance material and the vessel material when using credible surveillance data to calculate a best fit chemistry factor for transition temperature shifts due to irradiation. Several comments proposed changes in the criteria for establishing whether surveillance material data is credible that would result in a less restrictive basis for using surveillance data in determining the transition temperature shift. The comments argued that the proposed rule is ambiguous with respect to the use of information from other sources that contain limiting material for a specific plant and that the NRC must have the flexibility to approve use of such information on a case-by-case basis. Several comments proposed limiting the basis for making changes of RT PTS subject to the approval of the Director,* NRR. The NRC recognizes that 10 CFR 50.61 contains an unusual amount of .prescriptive material and that the comments proposing simplification have merit. Some changes to the rule have been made to provide flexibility, where.appropriate. The NRC staff is evaluating subsequent changes that would be more performance based. However, the NRC staff believes that this rule, as written, is needed to ensure that plants apply the appropriate method for determining RT PTS and that the appropriate reference to the thermal annealing rule be applied for the pressurized thermal shock situation. A number of clarifications were made to the rule. The paragraphs dealing with the determination of RT PTS were modified to make clear that RT PTS is a unique, end of life, case of RT NOT and to clarify the procedure for determining these values. As suggested, the adjustment procedure was added to the rule to permit accounting for differences in chemistry between surveillance materials and reactor vessel materials when calculating chemistry factors. With respect to the plant specific material surveillance data that is , permitted to be used in a surveillance program, the rule was modified to make clear that such data includes results from other plant's surveillance programs and test reactors. Several clarifications

were made to the criteria for determining credible material.

The NRC determined that the requirements for approval by the Director, NRR, for 50-SC-165 changes in RT PTS are appropriate and should not be modified. Thermal Annealing Rule (10 CFR 50.66) Twelve individual comments were received on the proposed Thermal Annealing Rule, 10 CFR 50.66. These comments included a number of suggestions for clarification of details of the proposed rule. Three of the comments addressed the requirements that, after the annealing operation, the reembrittlement rate of the reactor vessel due to neutron irradiation must be estimated and must be monitored using a surveillance program which conforms to Appendix H of 10 CFR 50, "Reactor Vessel Materials Surveillance Program." The comments are summarized as follows: (1) The supplementary information section for the proposed rule is silent on what is acceptable if limiting material is not available.The rule should provide appropriate requirements on the method for monitoring reembrittlement after annealing for those plants that do not have limiting material for their surveillance program and the monitoring plans should be consistent with the preannealing surveillance program approved by the NRC staff; (2) Appendix H does not define an acceptable post-anneal surveillance program, the reference to Appendix H should be deleted, and the post-anneal surveillance program should be defined in the annealing plan that is approved by the staff; and (3) The term reembrittlement rate is unclear as to the period of time to be used for its determination, and a wording change is proposed for the requirement that would relate change in toughness to fluence accumulated after the anneal. . Three of the comments addressed the requirements in the proposed rule that the Thermal Annealing Operation Plan include time-temperature profiles which represent the annealing conditions that. may not be exceeded during the annealing operation and are to be used for determining the.amount of recovery of the fracture toughness of the material due to annealing. The comments suggested that, instead of a single temperature profile, bounding time and temperature conditions be established for the maximum values that would be . used for thermal and stress analysis and to verify the re-qualification inspection and test program, and the minimum values that would be used to establish the amount of recovery of fracture toughness and for reembrittlement rate estimates. The bounding values would be based upon the estimated uncertainties in the times and PART 50

STATEMENTS OF CONSIDERATION temperatures and the actual annealing conditions should fall within these bounds. Two comments addressed the section on Certification of Annealing Effectiveness.

One comment suggested deleting the requirement in the proposed rule for certification of the annealing effectiveness and instead adding a provision in the Thermal Annealing Operating Plan that approval prior to subsequent power operation be required i:mly if the anneal was not performed in accordance with the approved plan. The comment also suggested that, if the licensee terminates the annealing before achieving the specified time but otherwise maintains

the annealing envelop such that no concern exists for stress or thermal damage, no additional constraints be imposed on subsequent operations and no credit be given for annealing.

The second comment suggested that (1) the

staffs review of the annealing report (certification report) need not be completed prior to reinitiating power operation if the anneal was performed in accordance with the approved _ Thermal Annealing Operating Plan, (2) . reporting and quantification of the actual recovery results need not be reported unless the vessel was at or_ above the PTS screening criteria when annealing was started, and (3) the Thermal Annealing Operating Plan should specify the_minimum content and *a schedule for reporting the annealing results. The commenter provided a proposed list of criteria, content, and schedule for reporting the annealing results. One comment stated that no guidance was provided in the .proposed rule on what constitutes components "affected" by tho annealing operation that ore required to be reported in the Thermal Annealing Operating Plan. The comment suggested

11ternative wording that components to be reported should be structures and components that are expected to experience significant temperature gradient or stress variations during the thermal annealing operation.

One comment suggested qualifying the provision in the proposed rule that the effects of localized high temperatures must be evaluated for changes in thermal and mechanical properties of the reactor vessel insulation for those coses where such changes may be negligible at annealing conditions. One comment suggested that the use of

applicable material data, such as data from integrated surveillance programs, be an optional part of the computational methods for determining fracture tough~ess recovery.

The NRC reviewed the comments received on the proposed rule _in detail. After consideration, the NRC reached the conclusion that most of the comments are notinconsistent with the intent of the proposed rule and in some cases reflect a need for clarification of the rule. In these cases, alterriative wording that clarified the intent of the rule was substituted in the text. With respect to the comments on the requirement that_reembrittlement rate after annealing must be monitored using a surveillance program, the NRC is aware that some plants do not have limiting materials.for their existing preannealing surveillance programs. For these situations the staff has approved alternative surveillance plans on a by-case basis. Clearly, these plants will not have limiting material for surveillance programs for use in determining reembrittlement rates afte,r annealing. ' . The NRC recognizes that Appendix H of 10 CFR Part 50, which is referenced in this rule, does not specifically _ address the surveillance of an annealed reactor vessel. However, the.

requirements of Appendix H to 10 CFR Part 50 apply to all reactors including*

the specific case of an annealed reactor vessel. To clarify the surveillance requirements of an annealed plant, the final rule has been µiodified to include, as suggested, that the post-anneal reembrittlement is to be monitored using a surveillance program defined in the Thermal Annealing Report and that the surveillance program must conform to the intent of Appendix H to 10 CFR Part 50. The term reembrittlement "ra:te" in the proposed rule was intended to mean the projected amount of reembrittlement over a specific fluence period. It is recognized that reembrittlomont is not n straight line function of fluence. Deterinination of reembrittlement rate is . discussed in more detail in Draft Regulatory Guide 1_.162, "Format and Content of Report for Thermal

Annealing of Reactor Pressure Vessels." In Regula~ory Guide 1.162, the approved method for estimating the reembrittlement rate, the lateral shift method; results in the same embrittlement trend as that used for the pre-anneal operating period. To avoid confusion the term "rote" hos been changed to "trend" in the final rule and the regulatory guide. The NRC agrees with the comments that the time and temperature profile required in the annealing operating plan should be bounding values. In this regard, Regulatory Guide DG-1027 calls
for the thermal annealing operating plan to include identification of the 50-SC-166 limitations and permitted variations in temperature, time, hentup and cooldown rate. For clarification, the final rule has been modified to use the terms "bounding conditions for till!e*s and temperatures and heatup and cooldown schedu.les" to describe _ conditions that may.not be exceeded during the annealing operation, and the -lower limit time and temperature of the actual anneal is used for determining the projected recovery of fracture . toughness by annealing.

.-The NRC considers that the intent of paragraphs (c), Completion or" Termination of Thermal Annealing, and (d), Thermal Annealing Results Report, of the final rule to be consistent with the two comments on that subject. The final rule .does not require that the NRC approve restart following the annealing operation if the Thermal Annealing Operating Plan and the Requalification Inspection and Test Program was

complied with. The NRC accepts the suggestion that the rule should be more specific on the items the licensee should include in the report and has included the list "in the final *rule; Finally, the NRC agrees with the suggestion to *make clear that a report is not required if: * (1) The licensee terminates the anneal prior to completion;

_ (2) The partial anneal was otherwise in accordance with the Thermal Annealing Plan; * (3) The licensee does not elect to take credit for any recovery. A statement was added to the Final Rule to cover the early termination situation.

The NRC has accepted the suggested clarifications of what constitutes an "affected" component and the qualification on the requirement to evaluate changes in properties on reactor vessel insulation if these are negligible.

The NRC considers it unnecessary to ini::lude a reference in the rule to data from integrated surveillance programs as an optional part of the computational methods to determine fracture toughness recovery. Generic computational methods for this. purpose are provided in the Regulatory Guide 1.162. However, the final rule does not prohibit use of alternative methods if adequate justification is provided. Appendix G to 10 CFR Part 50 Two comments were received on the Appendix G to 10 CFR Part 50 of the proposed rule. The NEI comment, which was endorsed by five utilities and one NSSS organization, included a table with six.items on Appendix G. The other comment on Appendix G was received from one of the five utilities. PART 50

STATEMENTS OF CONSIDERATION Two of the comments identified typographical errors and suggested a change in organization to improve clarity. One of the comments suggested revising the rule to change the definition of reference temperature, RTNoT, for cases where plants do not have data to comply with code procedures for determining RT NOT* One comment suggested a change in the title of Table 1., "Pressure and Temperature Requirements," by adding to the title "For the Reactor Pressure Vessel" to make clear that this table does not apply to other components in the reactor coolant pressure system and proposed adding a footnote to the table for the same purpose. One comment identified an error in the minimum temperature requirements for the hydrostatic and leak testing of the pressure vessel without fuel when the vessel pressure is equal or below 20 percent of the vessel design pressure.

One of the comments suggested that two of the entries in the table were new requirements when the table was intended to provide clarification. The utility's comment disagreed with the proposed rule change to prohibit the use of nuclear heat for the performance of vessel leak and hydrostatic testing. The utility contended that using nuclear heat, by providing a significant temperature margin above the pressure and temperature limit curves, greatly reduces the probability of brittle fracture and should be allowed. The NRC corrected the typographical errors and corrected the minimum temperature requirement for the hydrostatic and leak testing of the pressure vessel at low vessel pressures and without fuel. The title to Table 1 was changed, es suggested, for clarification. The NRC does not agree with the proposal to change the definition of RT NOT* The situation described in the comment, when data is not available to comply with code procedures, is presently handled on a case-by-case basis in accordance with MEB Branch position, MEB 5-2. The NRC staff does not agree with the comment that the two requirements cited are new requirements. Item 2.2.c. and Item 2.2.d of Table 1 are in the existing ASME code requirement and in Paragraph IV.A.3. in the rule. The NRC also does not agree with the utility's comment that using nuclear heat greatly reduces the probability of brittle fracture. The reasons for this are set forth in the February 2, 1990, letter to Messrs. Reynolds and Stenger of NUBARG from James M. Taylor, Executive Director for Operations. Appendix H to 10 CFR Part 50 Three comments were received on Appendix H to 10 CFR 50. The comment from NEI was endorsed by the five .utilities and the NSSS. Two of the five utilities submitted additional comments. NEI and one utility commented that the proposed change to Paragraph III.B.1, which establishes the applicable edition of ASTM standard E 185 for a reactor surveillance program, constituted a backfit that would require a substantial design change in the *survei1lance program for those plants fabricated to a code edition prior to 1973. The other two commenters suggested new changes to Appendix H to 10 CFR Part 50. One of the commenters noted that an existing provision -in Appendix H to 10 CFR Part 50, not part of the proposed rule change, dealing with requirements for attaching capsule holders to the vessel wall is a reiteration of a requirement in the ASME Code and should be removed. The other commenter suggested a new change to Appendix H to 10 CFR Part 50-to add a statement to the criteria for approval of an integrated surveillance program that would permit the use of surveillance specimens for extension of license purposes. The commenter also suggested that there is an apparent conflict between Paragraph III.C.2. and Paragraph 111.C.3. that address requirements for an integrated surveillance. The provision in the proposed rule was changed and reference to ASTM E 185 73 was deleted to make clear that the surveillance programs must be designed to the edition of ASTM 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased or to a later epition through 1982. The Commission agrees with the industry comments that imposing the ASTM E 185 1973 edition is impractical because vessels purchased prior to 1973 could not necessarily comply with all of the surveillance requirements in the 1973 edition of the ASTM standard. The NRG staff believes that the provision in the present rule on requirements for attaching capsule holders to the reactor vessel wall is required for clarity and should not be deleted. The comments related to the requirements for an integrated surveillance program were not persuasive to the NRC staff. The existing provisions of the rule do not preclude the application of the integrated surveillance program for extension of license purposes. The two paragraphs purported to be in conflict' address separate items; one addresses the number of materials to be irradiated, 50-SC-167 specimen types, and number of specimens per reactor; the other addresses amount of testing. Request for Comments on Issues Related to Thermal Annealing Comments were received from NEI on the five issues on thermal annealing that were included in the proposed rule at the Commission's direction. In addition, OGRE and one utility, Pacific Gas and Electric, submitted comments on Issue 4, concerning the preferred regulatory process (including opportunity for public participation). Public Comments on the five issues are summarized below:

Issue 1: The technical adequacy of the NRC staffs guidance.

Comment: The detailed comments submitted on 10 CFR 50.66 are summarized in the Summary of Comments section on the Thermal Annealing Rule. In addition, NEI suggested that draft Regulatory Guide, DG-1027, be revised to include acceptance criteria where an action is required, but the acceptance criteria was not defined. NEI further commented that the re-embrittlement rate equation (DG-1027, Equation 1) appeared to be very conservative and would result in a post-anneal operating life that is less than industry believes justified. Re.~ponse: The NRC is concurrently revising the noted draft regulatory guide and will address this comment in the resolution of comments for the guide. Issue 2: The sufficiency of the guidance and criteria to support a certification that if satisfied, a plant with an annealed vessel can safely resume operation. Comment: NEI noted that "The reactor pressure vessel thermal annealing rule and guide address appropriate issues to assure public health and safety and that the annealed reactor pressure vessel may be safely operated. The prior NRC staff approval of the reactor vessel annealing plan assures a clear process and criteria to restart following the vessel anneal. The licensee needs only to attest to compliance with the approved plan prior to resuming operations. The resumption of operations should not be needlessly delayed while a report documenting performance of the vessel anneal and recovery of the embrittled material properties is confirmed, because the vessel anneal will only improve the material properties. The

final report should be submitted on a schedule that considers when the vessel would have exceeded the RT PTs or uppershelf energy (USE) screening criteria without an anneal. The material property recovery will document prior PART 50
STATEMENTS OF CONSIDERATION to the time when *the vessel would have exceeded the screening criteria, thereby : assuring that the vessel is safe to Operate . at restart and for the duration justified by the material embrittlement
recovery." * . *. . * . Response:

NRC agrees with the NEI comment, except NRC believes it is. necessary for the licensee to submit the final report within three months of completing pr terminating the anneal, unless an extension is authorized by the Director, Office of Nuclear Reactor Regulation.

Issue 3: Whether health and safety concerns are best served by approval of the thermal annealing plan or of readiness for restart. Comment: NEI noted that "The performance of a reactor pressure *vessel anneal in accordance with an approved annealing plan improves the public health and safety by reducing the probability of core melt frequency.

This improvement occurs because of the increase in reactor vessel material ductility. The amount of recovery achieved by a thermal anneal "'.ill be documented prior to the original date when the reactor vessel would have exceeded the PTS or USE screening limit. Therefore, a demonstration for "restart readiness" is an extra burden that will not provide any further improvement of the public health and safety." * . Response: T)le NRC's determination as to the procedures for NRC review of the Thermal Annealing Operation Plan, Requalification Inspection and Test Program and justification for restart discussed below in further detail in the Opportunities for Public Participation section. Issue 4: The preferred regulatory process (including opportunities for public participation) and the commenter's basis for recommending a particular process*. Comment: NEI noted that "The industry recommends that a hearing opportunity be provided, but that it be a non-adjudicatory, 10 CFR Part 2, Subpart L type hearing on the docketed record. The essential features of the hearing process proposed are as follows. The NRC would at time of receiving the licensee proposed annealing plan issue a Federal Register announcement that staff is performing the review per 10 CFR 50.66. A Subpart L hearing could be held, if requested by an intervener, after the NRC staff has issued a safety evaluation report on the licensee annealing plan, but prior to commencement of the reactor vessel thermal an~ealing unless the NRC staff makes a "no significant hazards determination." Enclosure 4 provides additional details that support this . industry position." Additional detailed comments by NEI and the co_mments on this subject. by QClIB are discussed

: under the Opportunities for Public Participation heading. . Response:

The rule provides for public participation in the regulatory process by "incorporating a public meeting on the Licensee's Thermal Annealing Report a minimum of 30 days before the start of thei:mal annealing, and a public meeting after the licensee completes the anneal but before the reactor is restarted. The opportunity for public hearings in thermal annealing should be limited to those cases whore there is an unreviewed safety question or a change to the Technical Specifications or where the licensee did . not comply with the Thermal Annealing Operating Plan and Requalification Inspection and Test Program .. Expanded discussion on this issue is provided below under the Opportunities for .Public Participation heading. Issue 5: Whether there are health and safety issues concerning thermal annealing that cannot be addressed generically and would warrant specific consideration. Comment: NEI noted that "Thermal annealing to reduce material irradiation embrittlement is a well understood inetallurgical phenomenon. The supporting thermal and stress analysis used to demonstrate that the vessel is not damaged during the anneal are standard technologies used at nuclear plants. Because thermal annealing uses well understood technology, public health and safety is reasonably assured." Response: The NRC agrees with this comment. Opportunities for Public Participation The Supplementary Information section of the proposed rule discussed the four options the Commission considered for structuring the regulatory process related to public participation in the NRC's review and approval of a licensee's proposal for thermal annealing of a reactor vessel. The proposed rule, at the Commission's direction, requested comments on the preferred regulatory process (including opportunities for public participation). The four options included: (1) No hearings under the rule as proposed; (2) Discretionary opportunity for hearing under rule as proposed in which *situation the Commission would decide on a case-by-case basis to determine whether a hearing should be held; 50-SC-168 . (3) Required opportunity for hearing under rule as proposed, but. ~ork could commence if the .NRC were to make a * "no significant ha~d determination!'. on the*pro~osi:id thennal annealing; and _(4) Modify th~ proposed rule to * .* require suspension of license prior and

during the thermal annealing at which time no hearing wo.uld be afforded and the license would only be reinstated if the licensee demonstrates thatit has addressed the reactor embrittlement such that. it is acceptable to operate the plant. *.. ' .. Three comments were submitted ori the subject. OCRE and NEI addressed all of the alternatives in detail and they; \is well as one utility, identified and discussed individual preferred
alternatives.

NEI commented thaf each of the four alternatives has a sufficiently serious flaw to prevent adoption. With respect to the no hearing alternative, NEI agrees that annealing is presently subject to approval by the Director of NRR in

accordance with Part 50 Appendix G rather than being the* subjec~.of a license amendment as an u*ni:eviewed safety
question under§ 50:59 . .However, NEI
believes that annealing is an important process from* a regulatory standpoint and that public participation, in the form of.informal hearings, is appropriate.

NEI objected to a discretionary opportunity for a hearing because it provides significant

uncertainty in the process for licensees and members of.the j>ublic. NErs objection to requiri}!g*

a *hearing, as discussed in.staff Option 3, is thatit would aHow* those who object to the resumption of operat\on, on other-than technical ground!!, to 'use hearings to delay restart. Option 4 is objectionable, to NEI because it does not provide the licensee with any stability or predictability since the licensee would be required to demonstrate compliance after the annealing was performed, and does not provide the public with any opportunity to express its views .. NEI further commented that a license amendment is not necessary to approve a thermal annealing plan because annealing will not change the reactor vessel or*other cm;nponents in a manner inconsistent with the facility technical specifications nor will it require .

changes in the FSAR, and further, that a licensee is not required to modify its procedures to address or accommodate the annealing process. NEI noted that, while there is an incentive for the licensee to obtain credit for its improved P/T <:urves, and could seek a licensee amendment to do so, the licensee's existing P/T curves could remain in force.

PART 50

STATEMENTS OF CONSIDERATION Despite the conclusion that a license amendment .is not necessary for. thermal annealing, NEI recommended that a hearing opportunity be provided, but that it be a non-adjudicatory, Sul:>part L type hearing oil the record. NEI gave the following advantages for this approach:

(1) The NRC would be provided with a clear understanding of the licensee's annealing process, and the NRC's hearing process; (2) a Subpart L hearing is held on the .written record and typically does not include the discovery or live testimony _associated with adjudicatory hearings, but allows the public to participate.in a meaningful way without consuming the vast NRC, licensee, and public resources required for an adjudicatory hearing; and (3) it would provide predictability and stability by ensuring that all issues which could be subject to a hearing are addressed prior to restart. Any inspection or test performed in order to restart would be for the purpose of -confirming compliance with the rule. OCRE supported the proposed rule

provided that the public hearing rights were preserved with regard to reactor pressure.vessel annealing.

It is OCRE's position on the request for public comment that, based on the Sholly decision, the NRC must offer the opportunity for a formai adjudicatory hearing on the application for annealing and on the licensee's justification for subsequent operation where the licensee cannot certify that the thermal annealing was performed in accordance with the approved application. OCRE commented that approval by the

Director of NRR of the application for annealing and restart of the reactor, if the licensee cannot*certify that annealing was performed in accordance with the approved application, will.give the licensee the authority t~ operate in ways in which-they otherwise could not, and is thus, a de facto licens.e amendment.

OCRE fully supported Option 3 which requires opportunity for hearing under the rule as proposed. OCRE suggested that the adequacy of the thermal annealing plan, as well as the vessel's ability to perform its safety function after annealing, could be raised in the hearing on the thermal annealing plan and that the licensee's implementation of the thermal annealing plan could not commence until any hearing is concluded. or unless the NRC makes a "no significant hazards determination" with respect to thermal annealing. With respect to Option 1, OGRE concluded that the informal hearings or public meetings*proposed by the . Commission for the initial thermal annealing are not a substitute for. adjudicatory hearings required by the . Atoinic Energy Act (AEA) and do not give the interveners the same rights as they would have in a Section 189a hearing. OGRE found Option 2 pi:eferable to ha~ing no hearing. However, OCRE contended that this option is flawed by the assumption that "Section 189a of the AEA does not

afford an interested me.mber of the public a right to request a hearing." They contend that approval by the Director, NRR to anneal the reactor pressure vessel or to restart after' annealing does constitute a de facto . operating licensing amendment for which the opportunity for a hearing is required.

OCRE found Options 1 and 4

unacceptable in that they do not provide the opportunity for a formal adjudicatory hearing. The comment from tho utility suggested that Option 1 is the appropriate approach as long as the annealing process to be implemented is approved in advance by the NRC staff and the utility certifies that they have com,plied with the approved annealing process during the annealing operation, as provided for in the proposed rule. The utility further commented that if Technical Specifications ,changes or amendments to the operating license are required in order to perform tho annealing then the opportunity for hearings would be required due to the normal license amendment process and if the final safety analysis report (FSAR) were required to be updated to reflect the thermal annealing process, the provisions of 10 GFR 50.59 would apply. The utility suggested that if those changes did not constitute an "unreviewed safety question," no amendment would be needed and the license amendment process shou Id not be invoked and that if a member of the public is c*oncerned about a licensee's compliance with the NRC approved thermal annealing plan, those concerns could be addressed pursuant to the 10 CFR 2.206 petition process. The utility commented that, under its proposal, existing regulatory provisions for. public participation would apply as appropriate and no new prescriptive requirements would be necessary.

The Commission has considered the public comments and has modified the proposed rule as follows. A licensee that seeks to utilize thermal annenling to mitigate the effects of neutron irradiation of the nuclear reactor vessel must, at least three years prior to the date at which the limiting fracture toughness criteria in § 50.61 or Appendix G to Part 50 would be exceeded, submit a Thermal Annealing Report to the NRC staff for review. The 50-SC-169 report shall contain four sections: (i) Thermal Annealing Operating Plan, (ii) Requulificution Inspection und Test Progrnm, (iii) Program for determining Fracture Toughness Recovery and Reembrittlement Trend, and (iv) a section identifying any changes to the description of the facility as described in the updated final safety analysis report (FSAR) which constitute . unreviewed safety questions (USQs} under §'50.59, and changes to the facility's technical specifications, which are necessary either to perform the thermal annealing, or to operate following completion of the .annealing. Section 50.66(a) provides that the NRC will. within three years of submission of a licensee's annealing report, document . its views on whether the plan for conducting thermal annealing constitutes an unreviewed safety* question or otherwise requires a change to the plant's technical specifications. Such a determination is the threshold

determination for whether NRC approval is required before undertaking the activity.

In the event the NRG were to conclude, contrary to the licensee, that an unreviewed safety question is present or a change to the technical specifications is necessary, the NRG would, as a discretionary enforcement matter, issue an appropriate order to the licensee prohibiting annealing prior to issuance of a license amendment. An opportunity for formal adjudicatory hearing would be provided in connection with the license amendment; however, if the NRG makes a finding that the proposed change to the FSAR description or technical specification constitutes a "no significant hazards consideration" pursuant to Section 189.(a)(2)(A), the licensee may conduct the thermal annealing prior to completion of any hearing. In any event, at least 30 days before the licensee starts to thermal anneal and before the NRG completes its review, the NRG will hold a public meeting on.the licensee's proposed Thermal Annealing Plan and Requalificatfon Inspection and Test Program. Following the completion of the annealing operation, the licensee must confirm in writing to the Director, Office of Nuclear Reactor Regulation, that the thermal annealing was performed in accordance with the Thermal Annealing Operating Plan and the Requalification and Inspection Test Program. In support of this confirmation, the licensee must submit a report, within three months of completion or termination of the anneal, that presents the results of the annealing operation. Within two weeks of the PART 50

STATEMENTS OF CONSIDERATION licensee's.

written confirmatiori that the thermal *annealing was completed ~n

accordance w,ith the Thermal Annealing Plan, and prior to restart, the NRC shall: (1) Place in its public document room a summary of the NRC stafrs inspection of the licensee's thermal annealing
process to confirm that the thermal annealing was completed in accordance with the Thermal Annealing Operating Plan and the Requalification Inspection and Test Program, and (2) hold a public . meeting'with the licensee to permit the licensee to explain the results of the react~r vessel annealing .to the NRC and the pu~lic, for the NRC to discuss its inspection of the reactor vessel
annealing process,*and*to provide an opportunity for the public to.comment to the NRC on the annealing operation and the results of the Starrs inspection.

Witltin 45 days of the licensee's written confirmation that the thermal annealing was completed, the NRC shell complete full documentation. of the NRC's inspection of the .licensee's annealing process to confirm that the*. annealing was*completed.in accordance

with the Thermal Annealing Operating Plan and the Requalification Inspection and Test Program.*

.. The licensee may resume operation if: (1) The licensee concludes that the thermal annealing operation was performed in compliance with the Thermal Annealing Operating Plan, the. Requali~cation Inspection and Test P.rogram, and the provisions of Section 50.a&{b), (2) a summ:ary of the NRC's inspecthm of the thermal anne1ding is placed in the NRC public document room as required by Section 50.66(c) (2) and (3) the NRC holds the public meeting required by Section 50.66(£)(2), unless the staff takes action against the licensee. Since NRC approval to resume operation is not necessary, an opportunity for hearing would not be . provided in this situation. If, however, the licensee cannot conclude that the thermal annealing was performed in. compliance with the Thermal Annealing Operating Plan or the RequaUfication

Insp_ection and Test Program, the licensee.must submit a justification for
continued operation to .the Director.

If the noncompliance presents an unreviewed safety question, es determined by the licensee or directed by the NRC following its review of the report, then the plant may not restart unti~ the Director has approved restart. Those failures to comply with the Thermal Annealing Operating Plan and the Requalification Inspection and Test Program; which either (1) Are* considered to be unreviewed safety questions or (2) require changes to the technical specifications as a result of the. noncompliances, would also be subject to an opportunity for a formal adjudicatory hearing in accordance with the Commission's* regulations governing license amendments. However, the licensee may rest11rt prior to completion of the hearing if the Director makes a finding that such restart constitutes a "no significant hazards consideration," as provided uncl.er Section 189.(a)(2)(A) of the Atomic Energy Act of 1954, as amended.

The regulatory process for thermal annealing and the associated hearing opporturiitles are com;istent with standiµg NJ{C regu~tory practices defining thos.e matters which present . s1c1fficient potential effect on public health and safety (e.g., are unreviewed safety questions) to justify both prior NRC review of the change, and an opportunity for hearings (with the associated time and resource impacts on both the licensl;le and the NRC). With respect to lhe thermal anneaHng review pro~ss, the Commission reassessed the reguJatory requirementi;.

and processes for*assuring safety. The-Commission determined that the most important safety matters are normally addressed ln license conditions, technical* specifications, and the FSAR. The regulatory process for NRC consideration of licensee-initiated changes concerning these matters, and the associated opportunities for hearings . is in tO CFR 50.59. In view of this well~ established regulatory process iinportant safety information, the Commission determined that a regulatory. process re.quiring NRC -approval of a thermal annealing plan is not necessary, because the licensee is already required to comply with its license conditions, technical specifications, and FSAR. Important changes to license conditions, technical specifications, and FSARfrom a safety standpoint are subject to both prior NRC review and approval and an opportunity .for hearing. With respect to restart . following comph1tion of the annealing, the 15-day delay period should be sufficient time for review of the licensee's input given the NRC stafrs understanding of the annealing operation plan prior to implementation, ongoing resident inspections and headquarters inspections of the implementation of thermal annealing

operating plan. The Commission did not adopt NEI's suggestion for informal hearings where the Director must approve-restart if the Thermal Annealing Operating Plan and Requalificatfon Inspection and Test Program .were not complied with, _ because the Commission does not see 50-SC-170 nny distinction (in terms of safety implications) betweon the subject matter of hearings under this rule, as compared with other actions under Part-50 which, would require formal hearings.
As discussed earlier in the . supplementary information.

pre'(iously . performed research analyses indicated the potential for plastic.'defomiation of the main coolant.piping (or a typical U.S .. plant design and anticipated annealing conditions. There are also questions regarding how' thermal growth of the pressure vessel is treat~d. a_nd the adequacy of the thermal and stress analyses used to predict response of the .overall system under thermal annealing* conditions.-Additionally, there may be questions in other areas such as

temperature limits'for the concrete structures, and potential radiological

.. hazards associated with removing and storing the reactor internals.during the annealing process, and fire hazards associated with.heating the vessel. Recognition of the numerous complex technical questions related to 4 thermal annealing and of the potential benefits for operating nuc,ear power plants has.

resulted in a cooperative effort, fonded by the U.S. Department of Energy and the industry, to p81'form.Annealing . Demonstration.

Projects. Projects are -. planned to demonsb'&te two different . annealing processes, _evaluating heater designs and ves.sel designs. It is anticipated that the*anhealing demonstration projects will answer many of the generic questions regarding thermal annealing ofU.S_. pressure vessel and piping designs. The Thermal Annealing Report, required by the thermal ahnealing rule, . is designed to facilitate a detailed

review by the licensee of plent~specific questions and considerations in *
performing a: thermal annealing.

The . proposed rule specifically discusses the potential for unreviewe,hafety questions and technical specification

changes that may result from or be related to thermal annealing of the reactor pressure vessel. With completion of.the demonstration**

projects and as the staff and industry gain experience With therinal annealing, many of the issues related to annealing will be better understood and related questions will be answered. However, until this experience is realized, the staff will critically review licensee determinations regarding unreviewed safety questions and the need for

technical specification-changes associated with each proposed thermal annealing.

The level of staff effort is expected to be significantly greater during its review of the initial proposed PART 50

STATEMENTS OF CONSIDERATION vessel annealings than that which will* be required after experience is gained. The thermal annealing rule has been structured to provide time for the staff to thoroughly review the licensee's annealing plan and determination regarding unreviewed safety questions and the need for technical specification changes. If the staff identifies an unreviewed safety question or the need for a technical specification change, the licensee would be so notified and the existing NRC regulatory practices would be invoked to address the issues. Backfitting Issues Comments were received on
backfitting issues from the Nuclear Utility Backfitting and Reform Group (NUBARG).

NUBARG commented that they do not object to the new NRC position in Appendix G to 10 CFR Part 50 which prohibits core criticality before completion of hydrostatic pressure and leak tests as a conservative measure to *enhance safety. However, they are concerned that amending Appendix G on the basis of a compliance exception may set a bad precedent for avoiding backfilling analyses. NUBARG stated that "The logic of the proposed rule would seem to allow the NRC to avoid a backfitting analysis by (1) invoking the intent of one requirement.to override the explicit provisions of another, (2) using the compliance exception when the practice being eliminated seems specifically contemplated by and specified in the pertinent regulation; and (3) overlooking the fact that the NRC has apparently accepted this position in practice by some licensees*

*" In NUBARG's view, this proposed amendment should be supported by a backfit analysis.

The* Commission has reviewed this comment and has concluded that use of the qimpliance exception under§ 50.109 for the changes in Appendix G to 10 CFR Part 50 is appropriate. The Backfit Analysis section contains further discussion on this subject. The issue of explicitly prohibiting core criticality before completing pressure and leak tests has been addressed previously (letter from J.M. Taylor, EDO, to N. S. Reynolds and D. F. Stenger, NUBARG, dated February 2, 1990) and the NUBARG comment did not provide new information. The Commission hes concluded that any backfit requirements in this amendment are necessary to bring the facilities-into compliance with licenses, or the rules and orders of the Commission, or into conformance with written commitments by the licensees. Therefore, a backfit analysis is not required pursuant to 10 CFR 50.109(a)(4)(i). NUBARG also commented on the amendment to Appendix H lo 10 CFR Part 50 regarding surveillance that would preclude reducing the amount of testing if the initial test results agreed with predicted results. Although NUBARG recognizes the change would be prospective, it believes that NRC should provide flexibility to allow continue.cl relief for any licensee who lacks such on authorization but has relied on the provision. The Commission believes that sufficient flexibility already exists in that licensees who do not have an

authorization may seek an exemption under 10 CFR Part 50.12. Another aspect of the backfitting concern raised by NUBARG addresses the proposed amendment to § 50.61 which, based on the adequate protection exception, would impose a uniform methodology for calculating the reference temperature.

NUBARG contends that to rely on the adequate protection exception is arguably erroneous because the change in methodology is not likely an adequate protection issue (i.e., for most plants, the screening criteria will not be approached for many years). As discussed further under Backfit Analysis, the Commission believes that a new backfit analysis is not required for this conforming change, which corrects an inadvertent omission from the previous rulemaking. Therefore, the Commission concludes that the adequate protection basis for tho backfit continues to apply from the previous rulemaking (56 FR 22300; May 15, 1991) to §50.61. Criminal Penalties For purposes of Section 223 of the Atomic Energy Act (AEA), the Commission is issuing the final rule under one or more of Sections 161b, 161i or 1610 of the AEA. Willful violations of the rule will be subject to criminal enforcement. Finding of No Significant Environmental Impact The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule is not a major Federal action significantly affecting the quality of human environment and, therefore, an environmental impact statement is not required. The individual actions covered in this final rule would either serve to enhance safety of the reactor pressure vessel, thereby decreasing the environmental impact of plant operation, or have no 50-SC-171 impact on the environment. Therefore, in all cases these individual actions will not have an adverse impact on the* environment. PTS !,ule (10 CFR 50.61) The inclusion of thermal annealing as an option for mitigating the effects of neutron irradiation serves to decrease the environmental impact of plant operation by enhancing the safety of the reactor pressure vessel. The incorporation of the Regulatory Guide 1.99, Revision 2, method for determining RT NOT into the PTS rule has no impact on the environment because this change will result in values of RT PTS which are consistent with those currently used in plant operation. The restructuring of the PTS rule. is the type of action described iri categorical exclusion 10 CFR 51.22(c)(2). Therefore, an environmental asses*sment is not necessary for this change. Thermal Annealing Rule (10 CFR 50.66) The thermal annealing rule (10 CFR 50.66) permits and provides requirements for the thermal annealing of a reactor vessel to restore fracture properties of the reactor vessel material which have been degraded by neutron irradiation. This final rule only applies when a licensee elects to use it. The final rule provides an alternative for assuring compliance with the requirements in 10 CFR 50.61 and Appendix G of 10 CFR Part 50. The application of thermal annealing to a reactor vessel improves the condition of the reactor vessel material. In addition, this rule establishes requirements to avoid damaging the reactor system and to p*rotect against accidents during the annealing operation. This rule is one of several regulatory requirements that will function to ensure reactor vessel integrity. In that sense, this rule has a positive impact on the environment by reducing the potential for vessel failure. For th.ese reasons, the Commission has determined that there is no significant impact and, therefore, an environmental statement is not required. Appendix G to 10 CFR Part 50 The prohibition of core criticality before completion of the required pressure and leak tests will serve to reduce the potential for vessel failure, and thereby decrease the potential environmental impact of plant operation. . The restructuring of Sections IV and V of Appendix G is clarifying or corrective in nature, and is the type of I I PART 50

STATEMENTS OF CONSIDERATION action described in categorical exclusion 10 CFR 51.22(0)(2).

Therefore, an environmental assessment is not necl;lssary for this change. .

The changing of the reference from . Appendix G of Section III of the ASME
Code to Appendix G of Section Xl of the ASME Code has no impact on the environment because the requirements in the Appendices ere identical.

Therefore, there is no adverse impact on the environment from this change. The referenc~ng of the thermal annealing rule results in no adverse impact .on the environment because . Appendix G currently permits the use of thermal annealing to reduce fracture toughness loss of the RPV materials due to irradiation embrittleIIient. Appendix H to 10 CFR Part 50 Concerning the amendments to Appendix H to 10 CFR Part 50 in the final rule, the requirement that all irradiation surveillance tests be made (i.e., no reduction in testing is permitted) will have e*positive impact on the environment in helping to assure the integrity ofthere11ctor pressure vessel.**

. :
The restructuring of Section II.C is the type of action described in categorical exclusion 10 CFR 51.22(c)(2).

Therefore, an environmental assessment is not necessary for this change. . The clarification oftlie applicable version of ASTM Standard E 185 will result in no adverse impact to the environment since there will be no change _to current *surveillance programs. Changes to future surveillance programs will make the programs more effective in assessing irradiation embrittlement effects to the RPV materials, thereby helping to assure the integrity of the* reactor pressure vessel

Paperwork Reduction Act Statement
This*final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U:S.C. 3501 et seq.): These requirements were approved by the Office of Management and Budget, approval number 315CH>011.
The public reporting burden for this collection of information is estimated to . averag(;l 6,000 hours per response, including.

the time for reviewing instructions, searching existing data sources, .gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments regarding the burden estimate or any other aspect of this collection of inforp10tion, including suggestions for reducing the burden, to the Information and Records * *. *

Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011), OfficeofMan!lgement and Budget, Washington, DC ~0503. Public Protection Notification
The NRC may not conduct or sponsor,.

and a *person is not required to respond to, e collection of information unless it displays a currently° valid 0MB control . number. . Regulatory .Analysis The NRC staff has prepared a . regulatory analysis for*the amendments ,-to 10 CFR 50.61; Appep.dix G of 10 CFR . , Part so. and Appendix H of 10 CFR Part * : *50 that describes the factors and : alternatives considered by the ' Commission in.deciding to issue these amendments. A copy of the regulatory

analysis is available for inspection_

and copying for a fee at the NRC Public Document Room, 2120 L Street NW. * *(Lower Level), Washington, DC 20555-0001 .. Singl13 copies of the analysis may . be obtained from Alfred Taboada, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington* .. DC. 20555-0001, telephone (301)°415-6014. . j Regulatory Flexibility ~ct Certification As required by the Regulatory Flexibility Act, 5 U.S.C. 605(b), the ; Commission certifies that ,this final rule : ,will not have a.significant economi~ . imp11ct on a substa.ntial number of sipell

entities .. The rules which ai:e affected by the amendments will: (1) Preclude brittle fracture of enibrittled vessels . during PTS events, (2) provide the general fracture toughp.ess requirements for RPVs, including ductile fracture toughness requirements and temperature limits, (3) provide the requirements for. s*urveillance programs to monitor irradiation embrittlement of
RPV beltline materi11ls, and (4) provide for a method for*.re~toring the *fracture . toughness of RPV beltline materials used in nuclear facilities licensed under the provision of 10 CFR 50.21(b) and 10 CFR 50.22. The companies that own these facilities do not fall within the scope of the definition of "small .. entities" as set forth in the Regulatory Flexibility Act, the Small Business Size Standards in regulations issued by the Small Business Administration at 13 CFR Part 121, or the size standards established by the NRC at 10 CFR 2.810 (60 FR 18344; April 11, 1995). 50-SC-172 Backfit Analysis . PTS Rule (10 CFR 50.61)
The revision to *§ 50.6i requires:

licensees to calculate RT PTS using the same methodology. specified in Regulatory Guide 1.99, Revision 2, for.

determining RTNDT* This change*was logically e requisite part of the previous rulemaking (56 FR 22300; May 15, 1991) to § 50.61 that set forth a unified method for calculating radiation embrittlement of the reactor beltline materials in Pert 50. However, the Commission, at that t.ime, inadv!3rtently failed to make the *conforming change'to*.§ 50.61. The Commission believes that .the backfit statement for the previous amendment, which determined that the backfit was necessary to ensure that the facility continues to provide adequate.*

protection to the public health and safety, is applicable to this conforming change to § 50.61 .. The restructuring of the P.TS rule does not impose any backfits as defined in 10 CFR 50.109(e)(l) because there is no change in requirements due to this restructuring. . * * : * . The inclusion of thei:mel annealing in § 50.61 does not constitute a backfires defined in 10 CFR 50.109(a)(1) because the .decision to perform annealing is voluntary, no annealing has been**

conducted in this country, and there are no staff positions or* Commission requirements r.elied upon by licensees that are being.changed.

Thermal Annealing Rule (10 CFR 50.66) The final thermal annealing rule establishes requirements with respect to applications for thermal annealing. However, the Commission hes determined that. the rule does not impose a "backfit" as defined in 10 CFR 50.109(a)(1). The thermal annealing rule does not require any licensee to perfonp. thermal annealing. Under existing requirements, all licensees are required to evaluate whether they *exceed the PTS screening limits in 10 CFR 50.61 and the Cherpy tipper shelf screening limits in Appendix G of CFR Part 50. However, .these rules provide an alternative means for meeting these screening limits (e.g., performing thermal**annealing). No* licensee currently has pending before the NRC an application for thermal *annealing, nor has any current licensee . been granted permission*to conduct thermal annealing. The rule does not reflect any new or different NRC staff position which conflicts with a prior . . NRC staff position .or Commission rule. Thus, the final rule will have a purely prospective effect on future applications for thermal annealing. The Commission has stated in other rulemakings PART 50

STATEMENTS OF CONSIDERATION establishing prospective requirements (10 CFR Part 52 and the License Renewal Rufe, 10 CFR Part 54) that the Backfit Rule was not intended to protect the future applicant from current changes in Commission requirements.

Accordingly, the Commission concludes that the rule does not impose backfits and a backfit analysis need not be prepared for the final thermal annealing r.ule. Appendix G to 10 CfR Part 50 The restructuring 'of Sections IV and V of this appendix, referencing of the thermal annealing rule, changing the reference from App~ndix G of Section III of the ASME Code to Appendix G of

Section XI of the. ASME Code, and deleting the "design to perinit
annealing" requirement do not impose any backfits as defined in 10 CFR 50.109(a)(l), because they are either prospective in nature or are of a clarifying nature. 1 10 CFR Part 50, Appendix G,. Paragraph IV.2.d. of the final rule explicitly prohibits core criticality before completion ofASME Code hydrostatic pressure and leak tests. This is intended to make clear that licensees may not use nuclear heat in order to perform ASME Coqe hydrostatic tests. This amendment can be construed as a backfit, inasmuch as the prior version of 10 CFR Part 50, Appendix G, Paragraph IV.A.5 could be read to permit core criticality during ASME hydrostatic tests and Section XI of the ASME Code does not explicitly ,prohibit core . criticality prior to completion of these tests. However, the* Commission never intended the disputed language-in . Paragraph IV.A.5 of Appendix G to
permit core criticality before successful completion of the required ASME hydrost_atic tests. The scope of Appendix G is "fracture toughness requirements" only; that scope is stated clearly in the title of Appendix G, and Appendix G was n~t intended to specify system operational requirements; It is not correct, therefore, to interpret paragraph IV.A.5. as permitting nuclear hydrotesting.

The final phrase in IV.A.5, "depending on whether the core is . critical during the test," was included in the rule for the sake of completeness, to* specify appropriate fracture toughness requirements in th;e event that a licensee for some reason wanted to have the core critical during hydrotest, and was given approval to do so (e.g., as in the case of the Hatch units, where nuclear hydrotesting was allowed one last time as an approved exception.) The ASME Code's hydrostatic testing provisions for the reactor.coolant pressure boundary (RCPB) provides the necessary assurance that GDG-14 is met. GDC-14 inter alia requires RCPB testing in order to provide an extremely low probability of RCPB failure, in terms of abnormal leakage, rapidly propagating failure, and gross rupture. Using heat produced by a critical reactor core to perform such testing essentially undercuts the basic safety principle embodied in GDC-14 that testing should be completed prior to nuclear reactor operation. It makes little sense to allow core thereby allowing the reactor to be in an operational condition where a loss of coolant could have significant consequences-prior to successful completion_.of tests that are intended to ensure that the probability of such coolant losses during such an operational condition are extremely low.1 The ASME Code, Section XI, req1.1ires that the System Leakage Test be performed prior to plant startup following each refueling outage (Table-2500-1, Examination Category B-P, Note.2). The only way to interpret the ASME Code as permitting core criticality prior to completion of the hydrostatic tests is to read the term, "plant startup" as referring to something other than reactor criticality. This is neither the normal industry practice, nor has it been the NRC staffs longstanding interpretation of this provision of the ASME code. Indeed, it does not appear that the NRC staff has construed either Appendix G, Paragraph IV.A.5 nor Section XI of the ASME Code as permitting core criticality prior to successful completion of ASME Code hydrostatic tests. Moreover, the vast majority of nuclear utility licensees do not use nuclear heat to perform ASME code hydrostatic tests. This suggests that most licensees hold the same interpretation of Appendix G and Section XI of the ASME Code as the Commission. In sum, the Commission believes Section XI of the ASME Code, which is endorsed by 10 CFR 50.55a, implicitly prohibits core criticality prior to successful completion of hydrostatic testing. Therefore, the Commission concludes that the change in the language of Appendix G, Paragraph IV.2.d. is necessary to assure compliance with 10 CFR 50.55a and the ASMECode. 'The Commission is aware that NUBARG has presented ail argument to the NRC that per£ormance 0£ ASME Code hydrostatic tests are more e££ective at the higher temperatures. achieved when using nuclear heat, as compared with the heal sources normally employed by utilities in per£orming the .hydrostatic tests. However, for the reasons set £orth In the 1990 letter froin James M. Taylor, EDO to N. . S. Reynolds and D.F. Stenger, NUBARG, the

Commission reje_cts this argument.

50-SC-173 The Commission has concluded that

any backfit requirements in this *
amendment are necessary-to bring.the*

facilities into compliance with licenses;. or the rules and orders of the

Commission, or into conformance with:, written commitments.by the licensees.

Therefore, a backfit analysis is not* required pursuant to 10 CFR 50.109(a)(4)(i). Appendix H to 10 CFR Part 50 The amendments to Appendix H to 10 CFR Part 50 are either prospective in nature or of a clarifying nature, and hencu do not involve any provisions whkh would impose backfits as defined in 10 CFR *50.109(a)(1). . List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria,* Reporting and record keeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 552 and 553; the NRC is adopting the following amendments to 10 CFR Part 50. 61 FR 232 Published 1/3/96 10 CFR Part 50 RIN 3150-AD57 Fracture Toughness Requirements for Light Water Reactor Pressure Vessels Correction In rule document 95-30665 beginning on page 65456 in the issue of Tuesday, December 19, 1995, make the following correction: PART 50-[CORRECTED] On page 65468, in the first column,.in the authority citation for Part 50, in the *first paragraph, in the fourth line, "83 Stat. 1444" should read "83 Stat. 444". PART 50

STATEMENTS OF CONSIDERATION 61 FR 6762 Published 2/22/96 Effective 4/22/96 Employee Protection Policies; Minor Amendments See Part 19 Statements of Consideration 61 FR 30129 Published 6/14/96 Effective 7/15/96 10 CFR Part 50 RIN 3150-AF20 Production and Utilization Facilities; Emergency Planning and Preparedness Exercise Requirements AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRG) is revising its emergency planning regulations.

This amendment allows greater flexibility in the licensee's emergency preparedness training activities in the 2-year period between biennial full-.participation exercises. The amendment preserves the requirement that each licensee, at each site, conduct an emergency preparedness exercise biennially, with full participation by State and local governments that are within the plume exposure pathway emergency planning zone (EPZ); reduces the required frequency of exercising the licensee's onsite emergency plan from annual to biennial; requires licensees ,to ensure that adequate emergency response capabilities are maintained between biennial exercises by conducting drills, at least one of which must involve some of the principal functi anal areas of the licensee's onsite emergency response capabilities; and requires licensees to continue enabling State and local governments that are in the plume exposure pathway emergency planning zones (EPZs) to participate in drills. With this amendment, the Commission is granting, in part, a petition for rulemaking submitted by the Virginia Electric Power Company on December 9, 1992 (PRM-50-58). EFFECTIVE DATE: July 15, 1996. FOR FURTHER INFORMATION: Contact Michael T. Jamgochian, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555 (301--415-6534); E-mail MTJ1@nrc.gov. SUPPLEMENTARY INFORMATION: Backgrotmd The NRC received a petition for rulemaking submitted on December 9, 1992, by the Virginia Electric Power Company that was assigned Docket No. PRM-50-58. The petitioner requested that the NRC amend, Section IV.F.2., of 10 CFR part 50, appendix E, "Emergency Planning and.Preparedness for Production and Utilization Facilities," to change .. the requirement that each site exercise .its emergency plan biennially rather than annually. The petitioner's proposed amendment would have required each licensee to conduct a biennial full participation exercise of the emergency plan at each site and to take actions necessary to ensure that its emergency response capability is maintained during the 2-year interval. The petitioner believes. that the annual gradecj. exercise is but one of many indicators designed to provide reasonable assurance that . actions can and will be taken during an emergency situation that will provide for the health and safety of the public. The NRC published.a notice of receipt for the petition on March 4, 1993 (58 FR 12341). A total of 32 comment letters were received and considered when developing a proposed rule concerning the issues raised by the petitions. A notice of proposed rulemaking was published in the Federal Register on April 14, 1995 (60 FR 19002). Public comments were requested by July 13, 1995. A total of 18 comment letters were received, of which 12 utilities, 2 State emergency management agencies, and the Nuclear Energy Institute (NEI) supported the proposed rule change. One State emergency management agency and an environmental group opposed the proposed rule change. One letter received from a State emergency management agency had no comment on the proposed rule change. NRC Response to Public Comments The comment letters that were received provided many thought-50-SC-174 provoking and constructive comments. The Commission's evaluation of and response to these comments is presented in the following section.

Issue 1. While the biennial exercise provides the opportunity for broad based State and local participation in exercising offsite plans and procedures, the annual graded utility exercises enhance the biennial exercisP.

process by providing State liaison personnel and their utility counterparts the opportunity to remain proficient. A 2-year gap will lessen proficiency. Response. It is clearly not the Commission's intent to lessen the proficiency .at any level of the emergency planning organization (onsite or offsite) with .the rule change. The Commission believes that interaction and training problems*tbat might arise as a result of deleting the annual onsite exercise*would be resolved by requiring licensees to enable any State or local Govermnent to participate in the licensee's di:ills when requested by the State or local Government;The Commission is confident that, if a.State governmental* emergency response agency feels the need to participate in*a drill that would* require specific offsite interaction and decisionmaking capability, the licensee would accommodate the State agency's request within the framework of the drills that the licensee conducts throughout the 2-year period between. the biennial full participation exercise. In fact, a State.who was originally against granting the petition for rulemaking because of similar concerns stated the following in their comment on the proposed rule. "We were among those initially opposed to the Virginia Electric Power Company petition that prompted this rule change,. primarily because of a perceived potential for a diminution of emergency preparedness capability on the part of licensees. However, we acknowledge that the compromise embodied in the Commission's proposed rule . change offers adequate assurance that ongoing licensee emergency preparedness activities will continue at a reasonable level. Because of the number of licensees and the capacity of the State's emergency response organizations, when appropriate (this State) will invoke the language of the proposed rule change that requires licensees to " '*

enable any State or local government located within the plume exposure pathway EPZ to participate in the licensee's drills when requested by such State or local government.'" Issue 2. County, State, and utility emergency preparedness will degrade under a biennial schedule.

Mini-drills will not take the place of annual PART 50

STATEMENTS OF CONSIDERATION exercises as now constituted.

Further, States have been encouraging more Federal exercise participation by the Federal Emergency Management Agency (FEMA) and NRC. The proposed change would cut back on the opportunities to test current personnel and train new personnel. . *

Response.

The Commission.disagrees. The rule change does not require drills" to replace anb.ual exercises. The rule change does require that "the licensee shall take actions necessary to ensure that adequate emergency

response capabilities are maintained
*
by conducting drills, including at least one drill involving a combination of some of the principal functional areas of the licensee's onsite emergency response capabilities." (10 CFR part 50, appendix E, IV.F.2.b.)
Additionally, the.opportunity to test and train new perso~el is provided by requiring that "Licensees shall enable any State or local Government
*
to participate in the licensee's drills." (Id at IV.F.2.e.)

Issue 3. There is a need for clarity regarding State and local participation in the exercises and drills that are proposed to replace .the annual NRC graded exercise. At 60 FR 19002; dated April 14, 1995, licensees are charged to "enable" States* and,local governments

to participate in these exercises and drills, but at 60 FR 19006, activating all response facilities (Technical Support Center, (TSC); Operations Support Center (OSC), and tl;te Emergency Operations Facili,tylEOF))

is not , necessary. Because State.and local governments* coordinate interaction*

. through the EOF ann Media Centers, clarification is required.

For example, perhaps the utility would be charged with exercising the EOF and Media , Centers as a part of at least one exercise and/or drill each year. Response. Based on the extensive coordination and cooperation between licensees and State end local governments over the last 15 years, the

Commission is confident that, if a State or locai governmental emergency response agency felt the .need to participate in a drill that included interaction at the EOF end Media Centers, the licensee would accommodate the request within the framework of the drills that the licensee conducts throughout the 2-year period between the bienniiµ full participation.

exercises.

, Issue 4. Rather than eliminating any requirements, it is suggested that each site initially be granted a waiver for "off-year" integrated exercises.

The waiver would. be effective only as long as an acceptable level of emergency response capability is maintained .. Response. The Commission disagrees. . The Commission believes that the proposed rule would accamplish the commenter's objective without the extensive.NRC.resources that implementing the commenter's suggestion would require. Issue 5. The Commission does not appear to have addressed the quantitative question about expected turnover rates that would be important in determining whether biennial exercises could substantially reduce local team skills. Response. Please see the response to Issue 1. Additionally, the Commission has always been .and continues to be. committed to the principle that there exists "reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency." If, this finding is jeopardized either at the State or local governmental level, additional training* would be warranted and would be provided by participating in the drills the licensee conducts between biennial exercises. Issue:B. The Commission has not adequately addressed local Government comments on the importance of regular .exercises for improving coordination and communication.

Response.

The Commission*did not

receive any cpmmimts from local governments relating to this petition for rulemaking.

Nonetheless, the. Commission is confident that if a local Government wished to improve its coordination and communication capabilities, licensees would welcome its participation in one o,r more of the onsite drills thatwill be conducted between the biennial exercises.

Issue 1. The Commission has not addressed the FEMA concern that regular cooperation with offsite teams may.play a critical role in their prep!!fedness, which niay be especially important in view of the potential role such teams may play as first responders in actual *emergencies.

Response. Prior to publishing the proposed rule, the Commission received FEMA's assurance that their concerns with the petition for rulemaking had been resolved. Nonetheless, regular cooperation between offsite and licencee emergency response* teams will be ensured by the requirement that licensees enable any state or local . Government within the plume exposure pathway emergency planning zone to participate in the licensee's drills upon request.

50-SC-175 Issues Raised by Petitioner . The petitioner characterizes the present requirement as one that is resource intensive but of marginal importance to safety. The petitioner has identified grounds for change for a number of issues associated with the current requirement to conduct an . emergency plan exercise annually.

The issues presented by the petitioner follow: * (1) The requirement to conduct an integrated annual exercise is* not clearly defined. Therefore, the regulation should be clarified. (2) The existing regulation; 10 CFR

part 50, appendix E, is inconsistent.with other regulations that govern the frequency of offsite response _* organization integrated exercises (i.e., 44 CFR part 3110). (3) The performance ofoffsite
,, response organizations during biennial exercises has confirmed.

that a biennial frequency is sufficient to provide*the

reasonable assurance finding. * * * . (4) The existing regulation:, 10 CFR *
50.54(t), provides for an incj.ependent review of the adequacy of the program. (5) The existing requirement to conduct an annual exercise is not . necessary to achieve the underlying

.. purpose of the rule. A biennial exercise is sufficient to provide an acceptable formal confirmation of capability. (6) Reconsideration of the .* requirement is warranted in light of the

completion and implementation of *
enhanced emergency preparedness
facilities, the current level of industry proficiency and performance, and the increased industry sensitivity to emergency preparedness. , * . * (7) Personnel could be utilized more effectively in their normal professional function rather than by participating in a resource-intensive integrated test that only serves to confirm the already existing level of the response capability.

(8) Emergency planning resources

could be utilized more effectively to further the development and maintenance of emergency preparedness activities. . Commission Response The Commission believes that it is important;inlight of public comment, .* as well as the discussion provided in the petition, to clarify* NRC's intent (µnder the existing rule) that licensees need not conduct annual exercises with scenarios that progress to*severe core damage or result' in offsite releases.

Historically, these scenarios were used in both the biennial full-participation* exercise ofoffsite emergency plans and the annual exercise of the licensee's June 28, 1996 PART 50

STATEMENTS OF CONSIDERATION onsite emergency plan; this is no longer necessary for ~e _currently required annual exercises of the licensee's onsite emergency plan. Information Notice (IN) 87-54, "Emergency Response Exercises," was issued to clarify NRC intent in this regard arid to provide detailed guidance, specifically on the types of "off-year" training activities that licensees can perform during the interval between the biennial full participation exercises to maintain adequate EP response capabilities and to satisfy the rule. Some licensees have availed themselves of the flexibility afforded by the IN 87-54 guidance to conduct realistic, interactive "off-year" training activities that simulate less severe events, such as a minor fire, loss of electric power, or equipment failure, and focus on the capability of the onsite emergency response organization to diagnose problems and develop actions to successfully mitigate the scenario event. However, as noted in the petition, many licensees continue to employ severe accident scenarios in annual exercises of their onsite. emergency plans. Accordingly, the Commission is revising Section IV.F.2.b.

of 10 CFR part 50, appendix E, to (1) reduce from annual to biennial the frequency of exercising the licensee's onsite emergency plan (which may be included in the biennial full participation exercise specified in IV.F.2.c.) and (2) require licensees to conduct training drills, including at least one drill involving a combination of some of.the principal functional areas of the licensee's onsite emergency response capabilities. This drill would be conducted between biennial full participation exercises to ensure that adequate emergency response capabilities are maintained. The principal functional areas of emergency response include activities such as management and coordination of emergency response, accident assessment, protective action decisionmaking, and plant system repair and corrective actions. This approach is consistent with a comment from one State that favored the petition for rulemaking but preferred that some guidelines be included in appendix E requiring plant specific internal exercises during the "off-year" to ensure plant personnel familiarity with their response plans rather than the vague expectancy that this activity will be done. Furthermore, licensees would continue to enable State and local governments in the plume exposure pathway EPZs to participate in drills in the interval between exercises, thus, preserving their training opportunities. The Commission believes .that the final rule may result in the reallocation and more effective utilization of resources in some Ycensees' emergency preparedness (EP) programs as they further the development and maintenance of emergency preparedness capabilities during the "off-year" periods. However, it is not clear that these changes will result in significant overall cost savings. The Commission cautions specifically against expectations that the final rule will necessarily result in significant reductions in NRC inspection activity concerning licensees' "off-year" EP maintenance activities. Also, licensees will, upon request, submit scenaEios for NRC review as may be deemed necessary by NRC in support of future inspections. Conclusion Having considered the arguments presented by the petitioner as well as* evaluating all public comments received, and based on a further understanding-of the issues involved gained from 14 years of experience evaluating licensee emergency preparedness exercises, the Commission concludes that (1) the required frequency for exercising the licensee's onsite emergency plan should be reduced from annual to biennial, (2) the means by which licensees are expected to train and maintain their emergency response capabilities and readiness in the 2-year interval between evaluated exercises should be changed to require licensees to conduct drills, including at least one drill involving a combination of some*of the principal functional areas of the licensee's onsite emergency response capabilities, and (3) opportunities for training of State and local Government personnel must be preserved. The principal functional areas of emergency response include management arid coordination of emergency response, accident assessment, protective action. decisionmalcing, and plant system repair and corrective.actions: During the specified drills, activation of all of the licensee's emergency response facilities (Technical Support Center (TSC), Operations Support Center (OSC); and the Emergency Operations Facility (EOF)) would not be necessary. Licensees would have the opportunity to consider accident management strategies, supervised instruction would be permitted, operating staff would have the opportunity to resolve problems 50-SC-176 (success paths) rather than have controllers intervene, and the drills could focus on onsite training objectives. The final rule relieves licensees from the current requirement to conduct a full formal exercise of the licensee's onsite emergency plan annually, and gives licensees the flexibility to choose the activities to be conducted in the 2-year period between biennial participation exercises in order to maintain their emergency response capabilities. Greater flexibility in the training of the onsite emergency response organization can provide significant benefits to some licensees. For example, licensees can eliminate the practice of developing scenarios that proceed to severe core damage, offsite releases, or to higher emergency classification levels. Licensees will have greater opportunity to conduct realistic emergency response training with supervised instruction that allows the operating staff to consider accident management strategies, diagnose problems, and be given credit for actions thaLwould mitigate scenario events. This approach is also responsive to public commenters who expressed concern about a possible decrease in licensee training and readiness in the period between biennial exercises; Under this approach, licensees

wm still be required to conduct emergency response training and drills of the onsite emergency response organization, as well as provide training opportunities to State and local Government personnel during the interval between biennial exercises.

The final rule completes NRC action in response to PRM-50-58. The final rule grants the petitioner's request that the frequency of required onsite emergency response*plan exercises be reduced from annual to biennial. Additionally, 10 CFR 50.47(a)(l) is being revised in order to correct a typographical error that appeared in the 1993 edition of Title 10, Parts Oto 50 of the Code of Federal Regulations. In

the 1993 edition, the word "protection" was substituted for "protective measures" in 10 CFR 50.47(a)(l).

This action corrects this paragraph to read as follows:"*

reasonable assurance that adequate protective measures can and will be taken * * *" Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in 10 CFR part 51, subpart A, that this rule is not a major Federal action significantly PART 50
STATEMENTS OF CONSIDERATION affecting the quality of the human environment and ther~fore, an *environmental impact statement is not required.

The rule will update and clarify the emergency!planning regulations relating tci exercises. It does not involve any modification to any plant or revise the need for or the standards for emergency plans. There is no adverse effect on the quality of the environment. The environmental assessment and findi~g of no significant impact on which thisidetermination is based are available for inspection at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC. I Paperwork Reductiof Act Statement This final rule does not contain a new or amended information collection requirement subject tp the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 . et. seq.). Existing reqtiirements were approved by the Offii:e of Management and Budget approval 1 Number 3150-0011. R~gulatory Analysis ; The* Commission ~as prepared a regulatory analysis on this final regulation. The analjsis examines the costs and benefits of the alternatives considered by the Commission. The analysis is available for inspection in the NRC Public Document Room, 2120 L Street, NW. (Lowei Level), _Washington, DC. Single copies of the analysis may be obta.ined from Michael T. Jamgochian, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; Telephone: (301) 415-6534. i Regulatory Flexibility AcfCertification The final rule does not have a significant impact on a substantial number of small entities. The final rule updates and clarifies the emergency . planning regulations relating to exercises at nuclear power plants. Nuclear power plant licensees do not fall within the definition of small business in Section 3 of the Small Business Act (15 U.S.C. 632), the Small Business Size Standards of the Small Business Administr4tion in 13 CFR part 121, or the Commiss'ion's Size Standards published at 56 FR 56671 (November 6, 1991).: l).s required by the Regulatory Flexibility Act of 1980, 5 . U.S.C. 605(b), the Commission certifies that the final rule will not have a significant economic impact on a substantial number bf small entities. Therefore, a regulatory flexibility analysis is not required. I Backfit Analysis . The final rule clarifies the intent of the existing regulation and facilitates greater flexibility in licensees' conduct of "off-year" emergency response training activities. This action does not seek to impose any new or increased requirements in this area. The changes permit, but do not require, licensees fo change their existing emergency plans and procedures to employ scenarios in "off-year" training or drills that do not go to severe core damage or result in offsite exposures. No backfitting is intended or approved in connection with this. final rule change. List of Subjects in 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, reporting and record keeping requirements. For the reasons set out in the pr,eamble, and urider the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 553; the NRC is adopting the following amendments to 10 CFR part 50. 61 FR 39278 Published 7/29/96 Effective 8/28/96 10 CFR Parts 2, so, and 51 RIN 3150-AE96 Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. . -ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission is amending its regulations 50-SC-177 on the decommissioning procedures that lead to the termination of an operating license for nuclear power reactors.

The final amendments clarify ambiguities in the current rule and codify procedures that reduce the regulatory burden, provide greater flexibility,:.and allow for greater public partici patfon in the decommissioning process. Some minor amendments pertain to non-power reactors and are . for purposes of clarification and procedural simplification. The Commission believes that the final amendments will enhance efficiency and uniformity in the regulatory process of decommissioning nuclear power plants. EFFECTIVE DATE: August 28, 1996. FOR FURTHER INFORMATION CONTACT: Dr. Carl Feldman, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-6194; or S. Singh Bajwa, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-1013. SUPPLEMENTARY INFORMATION: Background On June 27, 1988 (53 FR 24018), the Commission promulgated decommissioning regulations. On July 20, 1995 (60 FR 37374), the Commission issued proposed amendments to these regulations. A discussion of the current requirements and proposed amendments follows. Current Requirements Within 2 years after a licensee permanently ceases operation of a nuclear reactor facility, it must submit a detailed decommissioning plan to the NRC for approval, along with a supplemental environmental report that addresses environmental issues that have not already been considered. Based on these submittals, the NRC reviews the licensee's planned activities, prepares a Safety Evaluation Report (SER) and an environmental assessment (EA), and either makes a negative declaration of impact (the usual case) or prepares an environmental impact statement (EIS). Upon NRC approval of the decommissioning plan, the Commission issues an order permitting the licensee to decommission its facility in accordance with the approved plan. As part of the approval process, the opportunity for a hearing under subpart G of 10 CFR part 2, is made available to the public. Once the decommissioning process is completed and the NRC is satisfied that the facility has been PART 50

STATEMENTS OF CONSIDERATION radioactively decontaminated.

t-0 an unrestricted release level, the NRC terminates the license, If the licensee chooses to place the reactor in storage and dismantle it at a later time, the initial decommissioning plan submittal need not be as detailed as a plan for prompt dismantlemef).t; ', However, before the licensee can begin dismantlement, a detailed plan and environmental report must be submitted and approved by the Commission. Before the decommissioning plan is approved, the licensee cannot perform major decommissioning activities. If a licensee desires a reduction in requirements because of the permanent cessation of operation, it must obtain a license amendment for possession-only status. This is usually granted after the licensee indicates that the reactor has permanently ceased operations and fuel has been permanently removed from the reactor vessel. A licensee is required to provide assurance that at any time during the life of the facility, through termination of the license, adequate funds will be available to complete decommissioning. For operating reactors, the amount of decommissioning funding required is generically prescribed in 10 CFR 50.75. Five years before license expiration or cessation of operations, a preliminary decommissioning plan containing a specific decommissioning cost estimate must be submitted and the financial assurance mechanism must be appropriately adjusted. Finally, the decommissioning plan, submitted within 2 years after permanent cessation of operations, must provide a specific cost estimate for decommissioning and a correspondingly adjusted financial assurance mechanism. For delayed dismantlement of a power reactor facility, an updated decommissioning plan must be submitted with the estimated cost of decommissioning and the licensee must appropriately adjust the financial assurance mechanism. Before approval of the decommissioning plan, licensee use of these funds would be determined on a case-specific basis for premature closure, when accrual ofrequired decommissioning funds may be incomplete. Proposed Amendments The degree of regulatory oversight required for a nuclear power reactor during its decommissioning stage is considerably less than that required for the facility during its operating stage. During the operating stage of the reactor, fuel in the reactor core undergoes a controlled nuclear fission reaction that generates a high neutron flux and large amounts of heat. Safe control of the nuclear reaction involves the use and operation of many complex systems. First, the nuclear reaction must be carefully controlled through neutron absorbing mechanisms. Second, the heat generated must be removed so that the fuel and its supporting structure do not overheat. Third, the confining structure and ancillary systems must be maintained and degradation caused by radiation and mechanical and thermal stress ameliorated. Fourth, the radioactivity resulting from the nuclear reaction in the form of direct radiation (especially near the high neutron flux areas around the reactor vessel), contaminated materials and effluents (air and water) must be minimized and controlled. Finally, proper operating procedures must be established and maintained with appropriately trained staff to ensure that the reactor system is properly operated and maintained, and that operating personnel minimize their exposure to radiation when performing their duties. Moreover, emergency response procedures must be established and maintained to protect the public in the event of an accident. During the decommissioning stage of a nuclear power reactor, the nuclear fission reaction is stopped and the fuel (spent fuel assemblies) is permanently removed and placed in the spent fuel pool until transferred offsite for storage or disposal. While the spent fuel is still highly radioactive and generates heat caused by radioactive decay, no neutron flux is generated and the fuel slowly cools as its energetic decay products diminish. The spent fuel pool, which contains circulating water, removes the decay heat and filters out any small radioactive contaminants escaping the spent fuel assemblies. The spent fuel pool system is relatively simple to operate and maintain compared to an operating power reactor. The remainder of the facility contains radioactive contamination and is highly contaminated in the area of the reactor vessel. However, because the spent fuel is stored in a configuration that precludes the nuclear fission reaction, no generation of new radioactivity can occur. Safety concerns for a spent fuel pool are greatly reduced regarding both control of the nuclear fission process and the resultant generation of large amounts of heat, high neutron flux and related materials degradation, and the stresses imposed on the reactor system. Contaminated areas of the facility must still be controlled to minimize radiation exposure to personnel and control the spread of radioactive material. This situation is now similar to a 50-SC-178 contaminated materials facility and does not require the oversight that an operating reactor would require. Based on the preceding discussion, it should be noted that during the operating stage of the reactor a nuclear reaction must be sustained that has the potential during an accident to generate significant amounts of energy and radiation whose consequences can be severe. Moreover, the nature of maintaining and controlling a nuclear reaction and the complexity of systems and operations requirements necessary to prevent and mitigate adverse consequences requires considerable oversight by the NRC. During the decommissioning stage of the reactor, the potential for consequences that could result from an inadvertent nuclear reaction are highly unlikely. The systems required for maintaining the spent fuel in the spent fuel pool as well as the operations required to contain the remaining residual contamination in the facility and spent fuel pool are relatively simple. Consequently, the activities performed by the licensee during decommissioning do not have a significant potential to impact public health and safety and these require considerably less oversight by the NRC than during power operations. The amendments proposed in July 20, 1995 (60 FR 37374), were intended to provide licensees with simplicity and flexibility in implementing the decommissioning process, especially with regard to premature closure. The proposed amendments were intended to clarify ambiguities in the current regulations, codify procedures and terminology that have been used in a number of specific cases, and increase opportunities for the public to become informed about the licensee's decommissioning activities. The amendments were designed to establish a level of NRC oversight commensurate with the level of safety concerns expected during decommissioning activities. A. Initial activities. The decommissioning process outlined in the proposed amendments was similar in approach to that in the current decommissioning rule, but included flexibility in the type of actions that can be undertaken without NRC approval. Once a licensee permanently ceases operation of the power reactor, no major decommissioning activities (as defined in the proposed rule) could be undertaken until the public and the NRC were provided information by the licensee. Information required from the licensee in a Post-Shutdown Decommissioning Activities Report (PSDAR) consisted of the licensee's PART 50

STATEMENTS OF CONSIDERATION proposed decommissioning activities and schedule through license termination, an assessment of whether such proposed activities are bounded by existing analyses of environmental impacts, and a general decommissioning cost estimate for the proposed activities.

The PSDAR would be made available to the public for comment. Ninety days after the PSDAR submittal to the NRC and approximately 30 days after a public information meeting is held in the vicinity of the reactor site, the licensee could perform major decommissioning activities if NRC does not offer an objection. Before undertaking these activities, the licensee must provide certifications to the NRC that operations have permanently ceased and fuel has been permanently removed from the reactor vessel (elements not formally addressed in the current rule). Once these certifications have been provided to the NRC, the licensee could no longer operate the reactor. Part 50 technical requirements would also be amended to properly cover the transition of the facility from operating to permanent shutdown status (which also is not explicitly covered in the current rule). Thus, a licensee who has permanently ceased operations and removed fuel from the reactor vessel would no longer need to obtain a license amendment to proceed with certain decommissioning activities within established regulatory constraints. B. Major decommis~ioning activities. A major change from the current rule is that power reactor licensees would no longer be required to have an approved decommissioning plan before being permitted to perform major decommissioning activities. Under the proposed rule, licensees would be allowed to perform activities that meet the criteria proposed in § 50.59. Section 50.59 would be amended to include additional criteria to ensure that concerns specific to decommissioning are considered by the licensee. Based on NRC experience with licensee decommissioning activities, the Commission recognized that the § 50.59 process used by the licensee during reactor operations encompassed routine activities that are similar to those undertaken during the decommissioning process. The Commission concluded that the § 50.59 process could be used by the licensee to perform major decommissioning activities if licensing conditions and the level of NRC oversight required during reactor operations are continued, commensurate with the status of the facility being decommissioned. These objectives were considered in the proposed rule as follows. (1) The proposed rule would clarify, modify, and extend certain licensing conditions to decommissioning activities.

(2) Aside from changes to part 50, the final safety analysis report (FSAR), which is a licensing basis document for performing activities under § 50.59, would need to be updated to cover decommissioning activities.

(3) A PSDAR would be submitted to the NRC that would contain a schedule of planned decommissioning activities and provide a mechanism for timely NRC oversight. The licensee would prqvide written notification to the NRC before performing any decommissioning activity that is inconsistent with or makes significant schedule changes from the PSDAR. C. License termination. A licensee wishing to terminate its license would . submit a license termination plan for approval similar to the approach that is currently required for a decommissioning plan. However, the plan would be less detailed than the decommissioning plan required by the current rule, because it would not need to provide a dismantlement plan, and could be as simple as a final site survey plan. The approval process*for the termination plan, as in the current rule, would provide for a hearing opportunity under 10 CFR part 2. The proposed rule recognized that, if the spent fuel is either offsite or in an independent spent fuel storage facility (ISFSI), that is covered under a part 72 license, the remaining facility licensed under part 50 is similar to a materials facility and a less formal hearing, under subpart L rather than subpart G of part 2, is more appropriate. As in the current rule, a supplemental !c)nvironmental report would be required from the licensee that considers environmental impacts that are not already covered in existing EISs. An additional requirement, proposed for the purpose of keeping the public informed, is that a public meeting be held, after the licensee submits the license termination plan to the NRC, similar to the one held after the PSDAR submittal. D. Financial assurance. The proposed rule would continue the same degree of financial assurance as the current rule, but provide more flexibility by allowing licensee's limited early use of decommissioning funds. This provision was presented in a draft policy statement entitled "Use of Decommissioning Trust Funds Before Decommissioning Plan Approval" (59 FR 5216; February 3, 1994) that was published by the Commission for 50-SC-179 comment and incorporated into the proposed-rule. Currently, licensee use of these funds is determined on a specific basis for prematurely shutdown plants. However, the proposed rule eliminated the requirement for a decommissioning plan and instead required a PSDAR submittal, which requires a decommissioning cost estimate. The proposed rule permitted some small percentage (3%) of the generically prescribed decommissioning funds to be available to the licensee for planning purposes ("paper studies") before permanent cessation of power reactor operations. Moreover, to permit the licensee to accomplish major decommissioning activities promptly, an additional generic funding amount would be made available (20%) before a site-specific cost estimate, which must be submitted to the NRC within 2 years after permanent cessation of operations (as in the current rule). The remainder of the funds would be made available after submittal of the site-specific cost estimate, as in the current rule. When the licensee submits the license termination plan, the same financial considerations as those in§ 50.82(c) of the current rule would be required to provide assurance that the licensee has adequate funds to complete decommissioning and terminate the license. E. License extension. The proposed rule clarified that a license that has expired is not terminated until the Commission terminates it and further clarifies what conditions prevail under such circumstances. F. Grandfathering. The proposed rule applied to power reactor licensees who do not have an approved decommissioning plan on the effective date of the final rule. Licensees that already have an approved plan could, at their option, follow the provisions of the proposed rule. G. Non-power reactors. There were some minor clarifications and procedural simplifications in the proposed rule for the non-power reactor decommissioning process. Otherwise, the current rule remained essentially unchanged. Response to Comments Thirty-four comment letters were received on the proposed rule from power reactor licensees, contractors, Government agencies, Agreement States, citizens groups, and individuals. The comment letters have been categorized into two groups representing commenters generally in favor of the proposed rule and those generally not in favor of the proposed rule. The commenters in favor of the rule (24) PART 50

STATEMENTS OF CONSIDERATION consisted of power reactor licensees, contractors, Government agencies, and an Agreement State. The commenters not in favor of the rule (10) consisted of citizens groups, individuals, and an Agreement State. The comments have been summarized and addressed through issue categories based on the proposed rule. Issue 1-Proposed Rule Approach.

Comments. Commenters in support of the proposed rule were, to varying degrees, supportive of the proposed rule. There were a few commenters in this group who fully supported the proposed rule because it would facilitate efficient decommissioning of power plants by reducing regulatory burden, clarifying the applicability of regulations originally intended for operating reactors, allowing a phased approach to decommissioning, and allowing early partial use of the decommissioning trust fund. A few commenters supported the use of lessons learned from ongoing decommissioning projects, expanding public participation, and providing the rationale behind less formal NRC policies and practices in a way that satisfies the requirements of the Atomic Energy Act (AEA), Administrative Procedure Act (APA), and National Environmental Policy Act (NEPA). While many commenters were generally supportive of the general concept of the proposed rule, they indicated that the proposed rule did not go far enough in reducing unnecessary regulatory burden. They noted that the existing NRC requirements regarding operating reactors were more than adequate to encompass decommissioning activities and, if anything, should be relaxed rather than expanded. These recommended relaxations pertained to such items as a more liberal attitude toward collection and use of decommissioning trust funds, elimination of unnecessary criteria concerning the use of the proposed § 50.59, elimination of proposed mandatory public meetings, elimination of the proposed Post-Shutdown Decommissioning Activities Report (PSDAR) submittal, and elimination of

the proposed license termination plan or eliminating its inclusion into the license by amendment, including elimination of the accompanying proposed Subpart L or G hearing opportunity.

Comm enters not in favor of the proposed rule were not supportive of the proposed rule to varying degrees. Many of these commenters were strongly opposed to the proposed rule and indicated that it allowed nuclear power generators to have discretionary powers to regulate themselves; that NRC was abdicating its responsibility for protecting the health and safety of workers and the public; that, in allowing the decommissioning plan to be included in the 'Final Safety Analysis Report (FSAR) it could be revised without license amendment, thereby excluding the public from the process; and that major component removal should not be allowed before the decommissioning plan is approved by the NRC. These commenters expressed a variety of views indicating that the existing rule should be left alone or that the current rule should be left basically in place but made more efficient through better implementation and should include greater opportunities for public participation. Finally, a few commenters indicated that significantly greater public participation and oversight are necessary than that prescribed in the proposed rule. Response. The proposed rule was developed to allow more flexibility in dealing with premature closures, the decommissioning process in general, and the experience gained from recent decommissioning activities such as those at Fort St. Vrain, Shoreham, and Rancho Seco, as well as early component removal at Yankee Rowe and Trojan. The justification and intent of the final rule is unchanged. The NRC's primary concern, as the licensee transitions to decommissioning, is that the licensee will have sufficient funds to complete decommissioning and that the activities undertaken by the licensee will protect the public and the environment. The intent of this final rule is to streamline some of the decommissioning requirements for power reactor licensees, especially in approval of the decommissioning plan before major decommissioning activities can be undertaken and in early use of decommissioning trust funds. Specific issues addressed in the final rule are discussed in greater detail below. Issue 2-PSDAR, FSAR, and update requirements. Comments. Commenters in favor of the rule had various comments concerning the PSDAR, its required update, and the proposed update to the FSAR. Several commenters indicated that the PSDAR requirement should be eliminated because it is more stringent than requirements imposed on operating reactors, that the PSDAR should only require information (detailed schedule) pertaining to the current phase of decommissioning because dismantlement and site restoration may not occur for many years, that the word "synopsis" should be used to make it 50-SC-180 clear that the PSDAR is a high-level summary, and that there should be consistency in the criteria for assessing environmental impacts between the PSDAR and the proposed § 50.59 requirements. A few comments suggested making the reporting requirements more efficient by combining them and updating the PSDAR and FSAR together, requiring updates no more than once every 36 months, or using a single PSDAR for multi-reactor sites. Several comments suggested that the updating requirement for the PSDAR be eliminated because § 50.59 already requires annual reporting requirements, that the term "significant" used in the proposed § 50.82(a)(6) should be tied to the § 50.59 safety evaluation, and that the extent of deviation in the PSDAR schedule that is permissible without notice to the NRC should be clarified. Finally, there was a comment that the final rule should make it clear that, if prompt decommissioning (dismantlement) is being pursued by the licensee, the PSDAR and license termination plan should be permitted to be the same document. Commenters not in favor of the rule did not specifically address Issue 2. However, those commenters believed that the current rule requirements should be followed and that an approved decommissioning plan should be required before a licensee is permitted to perform major decommissioning activities. Response. The purpose of the PSDAR is to provide a general overview for the public and the NRC of the licensee's proposed decommissioning activities until 2 years before termination of the license. The PSDAR is part of the mechanism for informing and being responsive to the public prior to any significant decommissioning activities taking place. It also serves to inform and alert the NRC staff to the schedule of licensee activities for inspection planning purposes and for decisions regarding NRC oversight activities. Because the final rule eliminates the need for an approved decommissioning plan before major decommissioning activities can be performed, the requirement to submit a PSDAR is less stringent than existing requirements for power reactor licensees. The information required to be in the PSDAR is less detailed than the information required in the FSAR. Therefore, the PSDAR should not be combined with the FSAR because the two documents have different purposes. The final rule requires a written notification if activities are anticipated that would be inconsistent with the PART 50

STATEMENTS OF CONSIDERATION PSDAR activities previously described.

The licensee's consideration of such inconsistency would include any milestone scheduling changes of dismantlement tasks and significant increases in decommissioning costs from those described in the PSDAR. The final rule will explicitly include the requirement that activities that would result in significant increases to decommissioning costs from those presented in the PSDAR must be a consideration in the notification requirements of§ 50.82(a)(7). It is intended that regulatory guidance addressing the PSDAR Standard Format and Content will be issued soon after the final rule is published. Currently, FSAR updates are required annually or 6 months after a refueling outage provided the interval between updates does not exceed 24 months. Because the FSAR is the basis for the use of§ 50.59, the updates will need to be timely, so the final rule specifies a 24-month FSAR update for decommissioning activities for those nuclear power reactor licensees that have submitted the certifications of permanent cessation of operation and permanent removal of the fuel from the reactor vessel. If prompt decommissioning is desired by the licensee, the licensee could elect early submittal of the PSDAR, before cessation of operation, and then use of § 50.59 would be permitted at cessation of operation, provided the certification of permanent fuel removal from the reactor vessel has been received and the public meeting had been held in advance. Although the PSDAR and license termination plan serve different purposes, and a formal approval process is required of the latter, the PSDAR and license termination plan can be combined. If a licensee chooses to combine the PSDAR and the license termination plan, the requirements for both would apply to the combined document, including the requisite waiting period, public meeting, and approval by amendment of the license termination plan. The procedure for *approval of a license termination plan is similar to that currently required for approval of a decommissioning plan. For a multi-reactor site, the PSDAR could address the activities for all the reactors at the site if decommissioning of each will be undertaken at the same time. Issue 3-Ninety-Day Time Period Prior to Undertaking Major Decommissioning Activities. Comment. Several commenters noted that the proposed 90-day waiting period before major decommissioning activities could be undertaken did not address a health and safety concern and that there are potentially high costs associated with such a delay because licensees could do a lot of dismantlement during this time that would be more efficient and cost advantageous. These commenters emphasized that all activities could be carried out under § 50.59 and the current licensing basis. They further stated that, if the 90-day hold is retained, clarification is needed regarding the NRC's opportunity to interpose an objection to proceeding with major decommissioning and that the NRC review should be based on areas of significant safety. Finally, one commenter expressed a concern that the 90-day waiting period would not allow enough time for public participation, including consideration of comments received from the public after NRC notices the licensee's PSDAR submittal and during a public meeting. Commenters not in favor of the rule did not specifically address Issue 3. However, those commenters believed that the current rule requirements should be followed and that an approved decommissioning plan should be required before a licensee is permitted to perform major decommissioning activities. Response. The commenters have correctly noted that the 90-day waiting period does not just address a health and safety issue. The NRC has chosen a 90-day waiting period prior to allowing major decommissioning activities to occur as the minimal time necessary for the NRC to evaluate the licensee's proposed activities and to conduct a public meeting. The public meeting is informational and may be chaired by a local official, with a presentation of the regulatory process for decommissioning by the NRC, presentation of planned decommissioning activities by the licensee, and participation by State representatives. A question and answer period would follow the presentations. By submitting the PSDAR before cessation of operation, a licensee could reduce the need for a waiting period (see the response to Issue 2 for an additional discussion on ways that the waiting period may be reduced). Issue 4-Proposed Rule Modifications to§ 50.59. Comment. Many commenters approved of some form of the proposed modifications to§ 50.59. Many of these commenters noted that § 50.59(e) in the proposed rule is more stringent than the existing requirements for operating reactors. These commenters believed that the existing § 50.59 criteria are adequate. Several commenters stated that the four proposed constraints contained in § 50.59(e) are somewhat 50-SC-181 redundant to the proposed requirements in § 50.82; the PSDAR content plus update and the 90-day waiting period envelopes issues addressed by these criteria. These commenters believed that if§ 50.59(e) criteria were kept they should be in a regulatory guide and not in a rule. Comments specific to the four criteria and why they should be eliminated follow. Section 50.59(e)(1)(i) concerning foreclosure of the site for unrestricted release. It was noted that any event that detracts from this effort would be accidental in nature, and that the proposed rule provided no explanation of the types of activities that could result in foreclosing the site for unrestricted use. Section 50.59(e)(1)(ii) concerning significantly increasing decommissioning costs. It was noted that cost estimate information is required prior to and through the decommissioning process, making this requirement unnecessary. Moreover, it was asserted that there is no logical correlation between the cost of a decommissioning activity and whether a license amendment should be required for that activity and that costs have never been a consideration in determining whether a proposed activity is consistent with the licensing basis for a plant. It was also noted that other regulatory bodies such as Public Utility Commissions and the Federal Energy Regulatory Commission, as well as economic pressure, will force a licensee to perform decommissioning cost effectively. It was recognized that actions taken by a licensee may diminish the decommissioning fund and it was suggested that the wording be changed to deal with actions that would "significantly inhibit the ability to fund decommissioning costs which would prevent successful decommissioning." Section 50.59(e)(1)(iii) concerned environmental impacts not previously reviewed. It was noted that compliance with the operating license, technical specifications, and § 50.59 regarding unreviewed safety questions adequately preclude having significant adverse environmental impact that have not been reviewed. Moreover, the requirement is redundant to the requirement concerning unreviewed environmental impacts required in the content of the PSDAR specified in § 50.82. Section 50.59(e)(1)(iv) concerned violating the terms of the existing license. It was noted that this requirement is redundant with language in § 50.59(a) that allows licensees to proceed with an activity so long as it does not violate technical specifications PART 50

STATEMENTS OF CONSIDERATION or constitute an unreviewed safety question as defined by § 50.59(a)(2).

Also, it was noted that a license amendment is required for changes in technical specifications under the current § 50.59(c). Most commenters who opposed the use of proposed § 50.59 were not in favor of the rule. One commenter stated that the analysis of the dismantlement activities proposed under§ 50.59 to determine whether or not the activity generates any unreviewed safety issue should be provided to the NRC, rather than rely on an NRC audit as existing regulations provide. This analysis would also provide this information to the public for examination. Several of the commenters indicated that an the-fact review of§ 50.59 activities would provide insufficient regulatory protection. Finally, a commenter stated that the presence of an NRC inspector is essential during decommissioning activities. Response. The Commission concluded that the proposed § 50.59(e)(1)(iv) is redundant and should be eliminated from the final rule. The Commission reconsidered the need for the remaining§ 50.59(e)(1) requirements and determined that placing them in § 50.82 would be more appropriate. The Commission also concluded that the requirement ensuring that no major decommissioning activities occur that would significantly increase decommissioning cost could be overly burdensome. Instead, an appropriate constraint would be to prohibit any decommissioning activities that result in there no longer being reasonable assurance that adequate funds will be available for decommissioning. However, the NRC needs to be aware of changes in decommissioning activities that would result in significantly increasing decommissioning costs and would require written notification of such intended actions. The other paragraphs in § 50.59(e) were placed in § 50.82(a) to ensure that they will be considered as overall constraints on the licensee's decommissioning activities, rather than separately for each contemplated activity as proposed in §50.59(e). The purpose of retaining these

requirements is to ensure that no decommissioning activities can occur that result in: (1) Eliminating the potential for unrestricted release, (2) significant environmental impacts not previously considered in EISs, and (3) there no longer being reasonable assurance that adequate funds will be available for decommissioning.

The basis for this final rule permitting the use of§ 50.59 activities to perform decommissioning activities is that environmental impacts have already been considered and that such consideration was for an unrestricted release condition where the licensee has sufficient funds to complete decommissioning (see final generic environmental impact statement (FGEIS). NUREG-o586).1 The major considerations of licensee decommissioning activities that could significantly affect the environment are at the license termination stage when the licensee submits a license termination plan for approval. If a licensee contemplates decommissioning activiti~s that would violate these requirements, the licensee may not use the § 50.59 process delineated in this rule to perform the activities. The licensee would then be required to obtain a license amendment to perform the activities. The final rule prohibits licensees from performing any decommissioning activities that foreclose release of the site for possible unrestricted use, result in significant environmental impacts not previously reviewed, or result in there no longer being reasonable assurance that adequate funds will be available for decommissioning (§ 50.82(a)(6)). Prior to the licensee's use of the § 50.59 process to perform major decommissioning activities, the PSDAR submittal and public information process must be completed. The licensee is required to include a discussion that provides the reasons for concluding that the environmental impacts that might occur during decommissioning activities have already been considered in site-specific or generic environmental impact statements, and to estimate the amount of funds necessary to complete decommissioning (see§ 50.82(a)(4)). The licensee is also required to submit a site-specific cost estimate within 2 years after permanent cessation of operations. Use of decommissioning trust funds are subject to the requirements (in§ 50.82(a)(8)) that adequate funds will be available to ultimately release the site and terminate the license. Moreover, the final rule requires the licensee to notify the NRC in writing before performing any decommissioning activity inconsistent 1 NUREG-0586. "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities,"" USNRC. August 1988. Copies are available for inspection or copying for a fee from the NRC Public Document Room 2120 L Street NW. (Lower Level). Washington. DC; the PDR's mailing address is Mail Stop LL-6. Washington. DC 20555-0001; telephone (202) 634-3273; fax (202) 634-3343. 50-SC-182 with, or making any significant schedule change from, those actions and schedules described in the PSDAR and states that this notification include consideration of significant increases in decommissioning costs (§ 50.82(a)(7)). The NRC intends to maintain an active inspection program to provide the requisite level of oversight of licensee activities during decommissioning. The PSDAR and any written notification of changes required of a licensee will be used to schedule NRC inspection resources for significant decommissioning activities. In addition to continuing requirements that the licensee must comply with, such as 10 CFR part 20, regarding protection of workers and the public from radiation, and appendix B to 10 CFR part 50 regarding quality assurance, the final rule explicitly extends certain technical requirements to cover decommissioning activities (e.g., §§ 50.36, 50.36a, 50.36b, and Appendix I regarding technical specifications for surveillance requirements, administrative controls, control of effluents, and conditions to protect the environment). Thus, there will be a licensing basis appropriate to the activities undertaken using the § 50.59 process during decommissioning. By maintaining certain requirements throughout the decommissioning process, licensees will be able to use the existing § 50.59 process to perform decommissioning activities and thus provide comparable assurance that protection of the public health, safety, and the environment will not be compromised. Issue 5-Environmental Impact Considerations During the Initial Phase of Decommissioning. Comments. Many commenters in favor of the rule fully supported the environmental impact considerations delineated in the proposed rule for the PSDAR submittal, with no mandatory ER or subsequent EA requirement. A few commenters suggested that no environmental impacts for decommissioning need be addressed further because the FGEIS for the 1988 decommissioning rule (NUREG-0586, August 1988) 1 and subsequent environmental assessments (for various actual power reactor decommissioning situations) demonstrate that decontamination and dismantlement do not significantly affect the human environment and have beneficial effects in restoring the site to an environmentally acceptable condition. A few commenters suggested that decommissioning should be considered a categorical exclusion as defined in 10 CFR 51.22. PART 50

STATEMENTS OF CONSIDERATION Most of the commenters who were not in favor of the rule believed that the NRC should define decommissioning as a major Federal action requiring an EA or EIS. They further indicated that a generic environmental impact statement cannot substitute for a site-specific EA because the FGEIS does not consider all possibilities.

A few of these commenters further stated that the proposed environmental impact consideration process is NRC's attempt to streamline the process for utilities and deregulate NRC current requirements. A few commenters stated that the process outlined in the proposed rule abdicates NRC's responsibility to protect the health and safety of the workers, the public, the environment, and it also undermines citizen's due process. Response. While the FGEIS (NUREG--0586) 1 for the 1988 decommissioning rule concluded that only minor negative environmental impacts would result from decommissioning in addition to substantial positive environmental impacts, it did not address site-specific situations that could differ from the assumptions used in the FGEIS analysis. However, it is expected that any site impacts will be minor. Any site impact should be bounded by the impacts evaluated by previous applicable GEISs as well as any site-specific EIS. To account for site-specific situations that may occur outside these environmental impact considerations, the final rule prohibits major decommissioning activities that could result in significant environmental impacts not previously reviewed. The review process for the PSDAR and the approval process for the license termination plan requires licensees to review the existing documents and address any discrepancies in their submittals. The environmental assessment conducted for this rulemaking relied on the FGEIS for the decommissioning rule (NUREG---0586, August 1988) 1 and determined that, insofar as the rule would allow major decommissioning activities (dismantlement) to proceed without an environmental assessment, application of the rule will not have a significant impact on the environment. Although not required by NEPA, NRC has required in this final rule that licensees indicate in the PSDAR the reasons for concluding that the planned activities are bounded by the FGEIS and previous site-specific environmental impact statements. This requirement is consistent with one of the primary goals of the PSDAR process, which is to promote public knowledge and provide an opportunity to hear public views on decommissioning activities before licensees commence decommissioning. At the license termination stage, the Commission must make decisions on the licensee-proposed actions described in the license termination plan. The Commission must consider: (1) The licensee's plan for assuring that adequate funds will be available for final site release, (2) Radiation release criteria for license termination, and (3) The adequacy of the final survey required to verify that these release criteria have been met. Therefore, the NRC has determined that submittal of the license termination plan should be treated as a license amendment. In addition, under 10 CFR part 51, an environmental assessment or impact statement would be required at the time the license is amended. Following resolution of another ongoing NRC rulemaking activity that is considering adoption of radiological release criteria, a categorical exclusion may be adopted that would eliminate the requirement for an environmental assessment or impact analysis, except in the case of a restricted release of a site. Issue 6---Public Participation. Comment. Most commenters supporting the rule commented on the public participation aspects of the proposed rule. They believed that the participatory role given to the public was appropriate, excessive, or in need of further clarification. Several questioned the need for expanded public participation on matters of public health and safety because the NRC regulatory framework already provides for such participation (e.g., license amendment process). These commenters also noted that the purpose of the public meeting following the PSDAR submittal was not properly explained and that the final rule should clearly state that the meeting is intended for exchange of information only. Many commenters indicated that the NRC should limit the scope of these meetings and hearings to issues that are related to health and safety during the decommissioning process. These commenters also indicated that the supplementary information should include a clear statement of the purpose and participation guidelines for these meetings and clearly identify NRC's role at these meetings (which should be significant). A comment stated that it is essential that adequate mechanisms be developed for addressing issues of concern raised by members of the public and that, absent such closure, the meeting would only compound frustrations felt by the interested public. Finally, there was a comment that the 90-day waiting period (after the submittal of the PSDAR to the NRC) 50-SC-183 before allowing licensees to undertake major decommissioning activities may not allow enough time for adequate public participation. Most commenters who did not favor the rule believed that the public participatory role proposed was inadequate. These commenters stated that NRC should retain the only license amendment (POLA) and decommissioning plan approval required in the current rule to truly enhance public participation. Public meetings were considered helpful, but no substitute for an adjudicatory hearing that includes the rights to discovery, to present evidence, and to cross examine. Along these lines, a commenter stated that a meeting does not afford citizens the level of institutional accountability necessary, given the dangers of environmental-toxic contamination inherent in reactor decommissioning activities and that citizens must have a substantive role in the decommissioning process in order to clarify, negotiate, and protect their community's interest. A few commenters suggested that specific advisory boards (SSABs) should be established early in the decommissioning process and that meaningful public involvement should be required at every stage of the decommissioning process, not only at the final termination stage. Response. As discussed previously, initial decommissioning activities (dismantlement) are not significantly different from routine operational activities such as replacement or refurbishment. Because of the framework of regulatory provisions embodied in the licensing basis for the facility, these activities do not present significant safety issues for which an NRC decision would be warranted. Therefore, it is appropriate that the licensee be permitted to conduct these activities without the need for a license amendment. However, the information meetings will be beneficial in keeping the public informed of the licensee's decommissioning activities. Although the primary purpose of these meetings is to inform the public of the licensee's planned activities, the NRC will consider public health and safety comments raised by the public during the 90-day period before the licensee undertakes decommissioning activities. A more formal public participation process is appropriate at the termination stage of decommissioning because the final disposition of the site is determined at that time. Under the current rule, the Commission issues an order permitting the reactor to be decommissioned, based on the approved decommissioning plan, which PART 50

STATEMENTS OF CONSIDERATION amends the license. NRC administrative procedures, in subpart G of 10 CFR part 2, now provide an opportunity for persons to request a hearing regarding the NRC's decision.

A similar procedure will be followed in the final rule for the license termination plan once the licensee has permanently removed fuel from the site. However, the hearing will be less formal because it will follow the procedures in Subpart L of 10 CFR part 2. The role of the SSABs will be evaluated when the rulemaking regarding radiological release criteria for license termination is finalized. Issue 7-Establishment and Use of the Decommissioning Trust Fund. Most of the comm enters on this issue were in favor of the rule. These commenters requested greater flexibility in what costs can be included in the fund, such as disposal costs of radioactive waste from plant operations, and greater flexibility in the use of the trust funds prior to and during decommissioning. Specific comments that reflect the full range of comments on financial issues are: Comment a. The proposed § 50.82(a)(7) proposes to regulate a licensee's use of, and rate of withdrawal from, the decommissioning trust fund. While NRC oversight is warranted to ensure that decommissioning activities can be funded, regulating the rate of withdrawal from the trust fund may unnecessarily impede the efficiency of a licensee's decommissioning activities. Because the NRC's generic estimates of decommissioning costs are substantially lower than most recent site-specific cost estimates, licensees would be constrained to withdraw small fractions of an unrealistically low estimate. Response. Limiting initial withdrawals to 23 percent of the generic cost estimate (using the§ 50.75 requirements), until the licensee has submitted a site-specific decommissioning cost estimate, preserves the integrity of the decommissioning trust accounts. The final rule permits licensees to withdraw up to 3 percent of the generic formula amount for planning at any time during the decommissioning planning process, including planning that occurs while a plant is still operating. This amount should be ample based on current planning costs for licensees recently undergoing decommissioning. Likewise, allowing withdrawals of 20 percent of the generic amount for decommissioning activities would allow funding of certain activities before receipt of a site-specific cost estimate. This amount is consistent with costs of large component removal activities undertaken or contemplated by licensees of shutdown plants (e.g., Yankee-Rowe and Trojan). Once the NRC has received the site-specific decommissioning cost estimate, a licensee would have access to the balance of trust fu'nd monies for the remaining decommissioning activities. Because the timing of the submittal of a site-specific cost estimate is within the control of the licensee, the Commission believes that unwarranted restraints on access to funds are not imposed by the final rule. Comment b. The scope of decommissioning-related activities that licensees may collect funds for should include disposal of low-level waste generated during operations, maintenance and storage of spent fuel after cessation of operations, costs to maintain an independent spent fuel storage installation, and non-radioactive demolition or "greenfield." State Public Service Commissions and the Federal Energy Regulatory Commission have authorized funding for these activities in some cases because it is in the best interests of the utilities' customers. The NRC regulation should not require segregation of these funds in separate accounts; restrictions on the withdrawal of trust funds in the proposed rule could lead utilities to create separate trust accounts for each nuclear facility funding component (e.g., decommissioning, spent fuel management, and greenfield). Finally, the rule should allow for the prudent and economic use, at the utility's discretion, of decommissioning trust funds during the years of normal plant operation even before end of life. Response. The NRC's authority is limited to assuring that licensees adequately decommission their facilities with respect to cleanup and removal of radioactive material prior to license termination. Radiological activities that go beyond the scope of decommissioning, as defined in § 50.2, such as waste generated during operations or demolition costs for "greenfield" restoration, are not appropriate costs for inclusion in the decommissioning cost estimate. Funds for interim spent fuel storage and maintenance are addressed in § 50.54(bb). The final rule does not prohibit licensees from having separate accounts for other activities in the decommissioning trust fund if minimum amounts specified in the rule are maintained for radiological decommissioning. Comment c. Section 50.82(a)(7)(ii) of the proposed rule specifies that a specific decommissioning cost estimate must be submitted to the NRC prior to 50-SC-184 the licensee being permitted to use any funding in excess of previously stipulated amounts. This could be interpreted to mean that the NRC must approve the additional expenditures. If this paragraph is retained, the intent of this "permitting" should be made clear. Expenditures made in accordance with the PSDAR and the decommissioning cost estimate should not require any additional NRC authorization. Response. The NRC's intent in the proposed rule was not to use a formal approval mechanism for decommissioning expenditures once the licensee submits its site-specific decommissioning cost estimate. The final rule has been modified as suggested by the commenter. Comment d. More guidance should be provided regarding what constitutes a decommissioning "planning" expenditure. Changes in the proposed rule regarding expenditure of funds from the NRC Draft Policy Statement on use of decommissioning funds before decommissioning plan approval (59 FR 5216; February 3, 1994), should be more fully explained. Response. The term "planning" used in § 50.82(a)(8l(ii) specifically means "paper" studies, not equipment removal. Percentages are used in the final rule rather than specific dollar amounts, as used in the Draft Policy Statement, to better allow for inflation of costs in the future. Other changes to the Draft Policy Statement are based on the response to comments, developed prior to this rulemaking activity, and presented in the section on the "Resolution of Comments on the Draft Policy Statement." Comment e. If a plant shuts down early, not only will there be insufficient funds to pay for planned decommissioning (because not all payments will have been made), but the actual cost of decommissioning can be 2 to 3 times higher than planned. The NRC should require external funds in the amount necessary to complete decommissioning upfront. Moreover, the NRC does not have a procedure in place for "replacing" a reactor licensee that goes bankrupt. Finally, the NRC should specifically allow the total financial approach to be made along the lines of industry self-insurance. Response. The revised regulations preserve the integrity of the decommissioning funds by tying the rate of expenditure to specific parts of the decommissioning process. At the same time they allow broad flexibility once a licensee submits its site-specific decommissioning cost estimate. The issue of bankruptcy, as well as the requirement for power reactor PART 50

STATEMENTS OF CONSIDERATION licensees to have the total amount of decommissioning funds upfront, was considered during the development of the current rule and found to be adequately addressed in current requirements.

Bankruptcy does not necessarily mean that a power reactor licensee will liquidate. To date, the NRC's experience with bankrupt power reactor licensees has IJeen that they file under Chapter 11 of the Bankruptcy Code for reorganization, not liquidation (e.g., Public Service Company of New Hampshire, El Paso Electric Company, and Cajun Electric Cooperative). In these cases, bankrupt licensees have continued to provide adequate funds for safe operation and decommissioning, even as bondholders and stockholders suffered losses that were often severe. Because electric utilities typically provide an essential service in an exclusive franchise area, the NRC staff believes that, even in the unlikely case of a power reactor licensee liquidating, its service territory and obligations, including those for decommissioning, would revert to another entity without direct NRC intervention. However, the NRC believes that with electric utility deregulation becomi:i;ig more likely, it may need to require additional decommissioning funding assurance for those licensees that are no longer able to collect full decommissioning costs in rates or set their own rates. Thus, the NRC proposed a rulemaking plan to, in part, evaluate these developments in SECY-95-223 (September 1, 1995). Issue 8--Court decision. Comment. Most commenters who were in favor of the rule indicated that the proposed rule did not conflict with the recent court decision regarding the Yankee Rowe decommissioning (Citizens Awareness Network, Inc. v. NRG, 59 F.3d 284 (1st Cir. 1995)). Most of the commenters who were not in favor of the rule believed that the proposed rule violated the court's decision, or the spirit of the decision, regarding Yankee Rowe. Response. A significant basis for the court's decision was that it perceived that the Commission had not adequately provided the reasoning for the NRC decision to allow decommissioning activities before NRC approval of a licensee-submitted decommissioning plan (59 F.3d at 291...:292), a decision that the court considered to be a modification of the Commission's decommissioning regulations. The court noted that the Commission had failed to provide either a rulemaking proceeding or a hearing to address what the court perceived to be NRC approvals of licensee decommissioning activities (59 F.3d at 291-92, 294-95). By initiation of a notice of proposed rulemaking and solicitation of comment (July 20, 1995; 60 FR 37374), the Commission addressed the reasoning underlying the proposed decommissioning process and allowed public review and comment on that reasoning. The final rule includes a public notice and meeting process, prompted by the licensee's submission of a report describing planned decommissioning activities, to hear public views before the licensee undertakes major decommissioning activi.ties. This process specifically provides that licensees may not begin major decommissioning activities until after they have submitted a PSDAR. The PSDAR will be made available to the public for written comment and a public meeting will be held to hear public views. Finally, the licensee is required to submit a license termination plan before release of the site. The final rule specifies that the license termination plan be approved by the NRC through the license amendment process. This process provides the public with hearing opportunities and ensures that any hearing on that plan must be completed prior to release of the site. This procedural framework assures that those citizens living near the site, potentially for years or decades after the facility is shut down, will be provided with information regarding the licensee's planned decommissioning activities, have an opportunity to ask questions regarding those activities at a public meeting early in the process, and have timely input into the decision to release the site. In its decision, the court also specifically addressed a concern about decommissioning activities taking place prior to any NEPA analysis (59 F.3d at 292-93). The final rule addresses this issue in several respects. First, the final rule explicitly prohibits the licensee from performing any major decommissioning activity that results in significant environmental impacts not previously reviewed or forecloses possible unrestricted release of the site. Also, when the licensee submits the PSDAR, the licensee must specifically include a section discussing how the planned activities fit within the envelope of environmental effects included in either the FGEIS (NUREG--0586, August 1988) 1 or the facility's site-specific environmental impact statement. Moreover, the licensee must provide written notification if the intended decommissioning activities are inconsistent with the PSDAR. This requirement helps ensure that, after submittal and public comment on the PSDAR, any changes to the planned 50-SC-185 decommissioning activities continue to be enveloped by the assessment of environmental impacts in prior environmental reviews. Any activities not meeting the environmental criteria would require the licensee to file an application for amendment to the license and a supplement to its environmental report under 10 CFR part 51. Finally, the rule requires a formal license termination plan by the licensee. The activities in the licensee's plan which do not meet the environmental criteria must be approved by the NRC by a license amendment that follows NRC procedures for amendments, including applicable hearing rights (under either subpart L or subpart G of 10 CFR part 2, as specified in the rule) and the preparation of environmental assessments. The court perceived that the agency "approval" of the expenditure of funds from the decommissioning funds may be a basis for triggering both NEPA reviews and hearing rights (59 F3d at 292-95). The final rule addresses this issue by providing generic guidance as to what expenditures can be made out of the decommissioning fund for decommissioning activities before submittal of a site-specific cost estimate. The revised regulations use generic criteria for expenditures from the decommissioning funds and do not require prior NRC approval of specific expenditures meeting the generic criteria (see § 50.82(a)(7)). These new provisions specifically require licensees to maintain sufficient funds for release of the site and termination of the license. The licensee will have to also include an updated, site-specific analysis of remaining costs in the license termination plan submittal. In publishing this final rule, the Commission has explained the rationale for the new decommissioning process, and has concluded that nothing in the court decision dictates that the Commission take a specific approach to this issue or otherwise raises questions concerning the validity of the approach adopted in this rulemaking. Issue 9-Definitions. Comment. Regarding the definitions in § 50.2, a few commenters indicated that the definition of decommissioning should include the concept of restricted release to accommodate the proposed rulemaking on acceptable residual radioactive criteria for decommissioning. Several commenters noted that the definitions of "major radioactive components" and "major decommissioning activities" were unnecessary because the use of the existing § 50.59 process does not require these considerations and is adequate to PART 50

STATEMENTS OF CONSIDERATION deal with decommissioning activities.

However, if a definition of "major radioactive components" must be kept, the definition should only be relevant to any components, that when dismantled for shipment, contain greater than class C waste. During decommissioning activities, these waste disposals have the greatest significance regarding environmental impacts and adequate funding and are unrelated to the physical size of components. Response. When the residual radiation criteria rule is final, the definition of decommissioning in § 50.2 will address use of the restricted release. It is necessary to have definitions of "major radioactive components" and "major decommissioning activities" to clarify what decommissioning activities may not occur before the end of the 90-day waiting period. However, the definition of "major radioactive components" in the final rule has been clarified so that large components, other than those named, are not prohibited § 50.59 activities if they contain small amounts of radioactivity. Dismantlement of these components is considered part of routine operating nuclear power reactor activities. Issue 10--Modifications to Specific Technical Requirements. Comment. Most of the commenters addressing this issue were in favor of the rule and indicated that there should be additional elimination or modification of requirements beyond those presented in the proposed rule. There was a spectrum of views on this issue: if a risk analysis were performed, it would demonstrate that the proposed rule would impose unnecessary burden on NRC licensees and NRC resources without commensurate benefit to health and safety; appropriate technical specifications for decommissioning would be for those activities for which there is a significant hazard; the final rule should include a discussion of the logic (i.e., philosophy) in making conforming revisions to part 50, especially with respect to provisions that did not change (e.g., §§ 50.55a, 50.63, 50.72, and 50.73 applicability); the study and survey by the NRC concerning additional amendments for non-applicability should be completed before this rule is finalized (one commenter); and that the proposed rule appears geared to permanently shut down reactors with fuel onsite and does not differentiate among th(:) aspects that apply once fuel is removed from the site, and the rule should consider such situations. Finally, one commenter requested that environmental qualifications remain in place for equipment important to safety pertaining to spent fuel management and storage. Response. This rulemaking is primarily directed toward the procedural process for decommissioning', with particular emphasis on premature closure situations. The modifications to technical requirements in the final rule are based on a consequence analysis that either leads to elimination of the requirement or extends its applicability to decommissioning. The modifications to the technical requirements in the final rule are incomplete, as noted in the proposed rule, and as the information base continues to develop, additional rulemaking actions to modify other requirements will be conducted. In the interim, licensees that no longer have fuel onsite may continue to request exemption for specific requirements on a case-by-case basis. The information base will address the storage of density packaging of hot spent fuel in the spent fuel pool with special consideration given to potential radiological consequences that could occur from loss of coolant in the pool. Consideration for amending rule requirements is also being given to situations in which the fuel is in dry storage at an Independent Spent Fuel Storage Installation (ISFSI). Comments on specific amendments were: Comment: Part 26. The final rule should explicitly state that the fitness for duty program does not apply to a permanently shut down and defueled facility. If it must apply, then it should apply to persons with unescorted access to the fuel storage building or buildings containing equipment necessary for the safe storage and handling of spent fuel. Response. Consideration of this issue is ongoing and may result in future rulemaking. However, until a decision is made, part 26 continues to be applicable. Comment: Section 50.36. Criteria are needed to ensure that technical specifications are appropriate for the conditions of a plant in a defueled state. The four criteria specified in § 50.59(e) would be appropriate additional guidance. Response. Consideration will be given at a later time to the development of additional guidance in the form of standardized technical specifications for decommissioning. However, licensees may apply for modification of their technical specifications on a case basis. Comment: Section 50.36 (c){6) and (e}. These requirements, which appear to imply that a new set of technical 50-SC-186 specifications will be developed for the plant decommissioning phase, are redundant and should be eliminated because§ 50.51(b)(2), the requirement to conduct activities in accordance with the specific part 50 license for the facility, is sufficient to ensure effectiveness of the technical specifications. Response. As a reactor facility transitions from operational to decommissioning status, numerous changes to technical specifications are expected. The regulatory experience with revisions to the technical specifications during this transition period has entailed case-specific evaluations of individual licensee requests. This has resulted in some inconsistency and variability of expectations among shutdown reactor facility license requirements. This revision provides the basis for developing a consistent framework for the development of "standardized technical specifications for decommissioning," as well as addresses the uncertainty regarding the applicability of the existing regulation to permanently shutdown reactors. Section 50.51 specifically addresses the continued effectiveness of expired licenses and limitation of licensee actions during any continued effectiveness period. As such, § 50.51 does not, nor is it intended to, provide specific license conditions and requirements. Section 50.36 addresses this issue. Comment: Section 50.36a(a){1}. This requirement should be clarified and revised because radioactive waste systems will have to be removed prior to license termination, and the present wording appears to require that these systems be used and maintained. Moreover, temporary systems are typically used for effluent treatment and the rule should be modified to describe only those systems that are appropriate. Response. Section 50.36a(a)(l) is intended to ensure that operating procedures for any waste treatment systems used to control effluents be maintained and used to existing release criteria, and not that the systems be used and maintained when no longer necessary. However, in response to the comment,§ 50.36a(a)(l) has been modified from the proposed rule so that systems that are no longer necessary can be eliminated from compliance requirements. Comment: Section 50.47. A defueled plant that has ceased operation warrants a material reduction in the scope of its offsite emergency planning requirements because the credibility of any offsite consequences are reduced. PART 50

STATEMENTS OF CONSIDERATION Beyond the spent fuel pool, there is not sufficient source term to justify emergency plans. This also pertains to appendix E to part 50 and the requirements in § 50.54(t) concerning periodic review (frequency and scope) of the licensee's emergency preparedness program. Response.

Consideration of the potential radiological consequences of hot, high-density packaged fuel in the spent fuel pool is still ongoing. Modifications to this requirement, if made, will be developed at a later time. Comment: Section 50.48. While some commenters agreed with the concept of a fire protection plan through the end of decommissioning, one found the proposed language overly restrictive, vague, and ambiguous. This commenter stated that once the permanently removed spent fuel is certified to no longer be a fire protection concern, an industrial fire protection program could be adequate in most cases. Several other commenters noted that there are other ongoing NRC activities to improve current fire protection regulations, and if actions are taken now, they should only be based on "significant hazards" considerations. Response. These modified requirements have been coordinated with ongoing NRC activities regarding the improvement of fire protection regulations. Also, see the response to § 50.47 regarding spent fuel considerations. As presently configured, fire protection regulations apply only to operating reactor facilities. The need for an ongoing fire protection program, albeit a modified one, remains after the facility has ceased reactor operations. The final rule provides a based program that can readily be modified during the decommissioning process to address residual hazards. Comment: Section 50.49. Electric equipment required for protection of spent fuel outside the reactor does not meet the definition of equipment defined by§ 50.49(b). The discussion in the final rule should be corrected to note that the environmental qualifications regulations apply to selected safety and non-safety related equipment as described in § 50.49(b). Response. No modifications to the proposed rule are necessary. However, the environmental qualifications regulations apply to selected safety and non-safety related equipment as described in § 50.49(b). Comment: Section 50.51. Section 50.51(b) should be deleted because it is redundant. If it is kept, the requirements on the continuation of a license should be clarified to affirm that other operating reactors would be unaffected when the operating license of one reactor has been terminated at a reactor site. Section 50.51(b)(1) should be clarified to indicate that, at sites that have an intervening reuse but do not require decontamination to unrestricted release, decontamination would not need to occur until the end of the reuse period. Response. Section 50.51(b) is not redundant and will not be deleted. This section in the final rule has been modified to clarify that an expired license for a nuclear reactor facility that has permanently ceased operations is not terminated until the Commission terminates it. This provision further clarifies what conditions prevail under such circumstances. At a multi-reactor site, each reactor is individually licensed and actions are applied accordingly. The final rule addressing the radiological criteria for decommissioning will address the issue of restricted release options. Under the proposed rule, such restrictions would have to ensure that members of the public, in the event the restrictions fail, would not receive a dose in excess of 100 mrem per year. Unless the facility remained under license, individuals having access to the facility would be considered members of the public. Comment: Section 50.54(g). The antitrust law requirements for a reactor that has permanently ceased operations and permanently defueled should be reevaluated for applicability. Response. Section 50.54(g) simply provides that the issuance of an NRC license does not relieve the licensee from compliance with the antitrust laws specified in Section 105 of the Atomic Energy Act, and that the NRC may take appropriate action, including suspension or revocation of the license, if a court finds the licensee to have violated any provisions of such antitrust laws. This subsection of the regulation is sufficiently flexible that there is no reason to modify or delete it with respect to a facility that has ceased operations or is permanently defueled. Comment: Paragraphs (k), (]}, and (m) of§ 50.54. The requirement for licensed operators should be eliminated or reduced because reactivity changes can only occur during the initial stages of decommissioning in connection with repositioning fuel assemblies in the spent fuel pool. With reference to § 50.54(i), the scope of the operator requal_ification program and limitations on a licensee's freedom to modify it should he reduced at facilities undergoing decommissioning. Response. Consideration of these issues is ongoing and may result in future rulemaking. 50-SC-187 Comment: Section 50.54(w). Onsite property damage insurance for a facility undergoing decommissioning should be eliminated or substantially modified. Response. Consideration of the potential radiological consequences of hot, high-density packaged fuel in the spent fuel pool is still ongoing. Modifications to this requirement, if made, will be developed at a later time. Comment: Section 50.55a. Pertaining to codes and standards requirements, it should be noted that §§ 50.55a (a), (f), and (g), inservice testing requirements, do not apply to permanently defueled reactors because the plant is not operating and there is no need to apply the regulation. Response. No change is necessary because these requirements provide assurance that relevant portions of the facility are maintained functional or operational to adequate standards so they are operationally capable. Comment: Section 50.63. The requirements on the loss of all ac power should not apply to decommissioning because the potential for significant radiological consequences is very low (there is a low probability of incident and long recovery time). Response. Consideration of the potential radiological consequences of hot, high-density packaged fuel in the spent fuel pool is still ongoing. Modifications to this requirement, if made, will be developed at a later time. Comment: Section 50.65. Monitoring maintenance for a permanently shutdown and defueled facility on any of its structures, systems, or components (SSC) to levels required by the current maintenance rule is unnecessary. Permanently shutdown and defueled facilities can no longer experience the levels of mechanical stresses associated with an operating plant. Therefore, the industry interprets the proposed rule to mean that the maintenance program only applies to the safe storage of fuel. The relative risks from a shutdown plant allow requirements in existing technical specifications and other administrative programs to provide adequate assurance for safe fuel storage. Response. The maintenance rule, § 50.65, requires that the performance or condition of all structures, systems, and components (SSCs) described in § 50.65(b) be included in the scope of the rule. Under the current rule, licensees are permitted flexibility in the goals that are established and the monitoring that is performed for these SSCs. The NRC agrees that the stresses on most SSCs in an operating plant are greater than those associated with a shutdown and defueled plant. The final rule allows the scope to he limited to PART 50

STATEMENTS OF CONSIDERATION those SSCs associated with the storage, control, and maintenance of spent fuel in a safe condition in a manner that *provides reasonable assurance that the SSCs are capable of performing their intended function.

Comment: Section 50.72. The immediate notification requirements for operating nuclear power reactors should not apply to permanently defueled reactors or, if applicable, should be significantly modified. Regarding § 50.72(a)(i), there should be no requirement to use the Emergency Notification System or Emergency Response Data Systems. Response. The NRC did not adopt this comment. Notification requirements for events such as abnormal releases and overexposures are examples of required reports that are necessary. Comment: Section 50.111. Criminal penalties should not be imposed for decommissioning activities because they are not so important to public health and safety that licensees need be subject to them. Decommissioning activities for reactor licensees should not be treated any differently than for other radioactive material licensees. Response. The Commission believes that certain actions are essential in initiating the decommissioning process (e.g., certifying to permanent cessation of operation and permanent removal of fuel from the reactor vessel, and submitting a PSDAR) and should, therefore, be treated as substantive with respect to the criminal penalty provisions of the Atomic Energy Act. Decommissioning actions, when initiated improperly, have a potential for significant consequences regarding health, safety, and the environment. Willful violations of, attempted violations of, or conspiracy to violate, § 50.82 would, therefore, be a matter of significant concern to the NRC. Thus, the NRC is retaining the addition of § 50.82 to the list of regulations to which criminal sanctions apply. Comment: Section 140.11. Concerning Price Anderson financial protection, permanently shutdown and defueled facility licensees should be permitted to withdraw from the secondary financial protection layer, and single units should be given a reduction in the primary level of coverage (e.g., $100,000,000). Response. Consideration of the potential radiological consequences of hot, high-density packaged fuel in the spent fuel pool is still ongoing. Modifications to this requirement, if made, will be developed at a later time, as will considerations of fuel stored in an ISFSI. Issue 10-Termination of License Requirements. Most of the commenters in favor of the rule supported the decommissioning requirements for termination of the license in the proposed rule. However, several of these commenters stated that approval of the license termination plan should not require an amendment or opportunity for a hearing. They believe that if the plan is made available for public comment, existing regulations provide ample opportunity for public participation and the AEA does not require a hearing. Another commenter noted that once the spent fuel is off the site, the hazard is reduced so there is no safety, technical, or legal basis for NRC approval of a detailed decommissioning plan or PSDAR. A commenter pointed out that the use of the proposed § 50.59, which includes the four criteria (§ 50.59(e)), addresses the unique circumstances associated with the decommissioning activities. If some activities do not satisfy the requirements of§ 50.59 and a license amendment is required, interested parties would have an opportunity to request a hearing. The approval of the plan by amendment and the opportunity for a hearing are not for reasons of health and safety; moreover, any interested party could always petition for a hearing under§ 2.206. Another commenter made similar comments and went even further in stating that if standards for radioactive release are clear, meeting the objective of terminating the license should be easily demonstrated without the need for approval of a plan or license amendment; and that the plan should be available to the NRC for information only. Response. The requirement for submittal of a termination plan is retained in the final rule because the NRC must make decisions, required in the current rule on the decommissioning plan, regarding (1) the licensee's plan for assuring that adequate funds will be _available for final site release; (2) radiation release criteria for license termination, and (3) adequacy of the final survey required to verify that these release criteria have been met. A public meeting is considered necessary at the license termination stage to inform the public about the licensee's proposed termination activities and to provide an opportunity for public comment on those proposed activities. The NRC has also made the determination that license termination is an action of sufficient significance as to warrant an opportunity for a public hearing on NRC's decision regarding the licensee's proposed termination activities. 50-SC-188 Specific comments concerning the license termination plan were provided by several commenters. Comment a. The timing of the license termination plan is not explicit in the proposed rule, § 50.82(a)(8), and it is not clear whether the rule permits dismantlement activities before submittal or approval of the license termination plan. Response. The final rule permits dismantlement activities 90 days after PSDAR submittal unless the NRC interposes an objection. The license termination plan must be submitted within 2 years of the licensee's expected date of license termination (the date specified in the PSDAR or supplement). Comment b. The NRC does not explain or support the need for the elements of the plan, discussed in proposed § 50.82(a)(8)(ii) (A)-(G). The current rule, under§ 50.82(d), simply requires updated, detailed plans before the start of decommissioning. Response. The final rule permits major decommissioning activities (dismantlement) to be performed using the § 50.59 process. Because a decommissioning plan is no longer required, the requirements for the license termination plan are less complex than those that are currently required for a decommissioning plan. The license termination plan provides documentation on the remaining activities necessary to terminate the license and includes consideration of remediation aspects that could involve license termination under either unrestricted or restricted release conditions (once the rulemaking on acceptable residual release criteria is final). The site characterization, description of the remaining dismantlement activities and plans for site remediation are necessary for the NRC to be sure that the licensee will have adequate funds to complete decommissioning and that the appropriate actions will be completed by the licensee to ensure that the public health and safety will be protected. The language of§ 50.82(8)(a)(ii) (B) and (F) in the proposed rule, now § 50.82(a)(9)(ii) (B) and (F) in the final rule, has been changed to more clearly reflect the intent of these requirements. Thus, element (A) now requires identification of remaining dismantlement activities, and element (F) now requires an updated specific estimate of remaining decommissioning costs. Comment c. One commenter questioned how multiple sites will be addressed. Another commenter stated that a single license termination plan PART 50

STATEMENTS OF CONSIDERATION should be encouraged for multi-reactor sites. Response.

Reactors at a multi-reactor site are individually licensed and licensing actions are applied to the individual licenses. A licensee would not be prohibited from submitting a single license termination plan for the multi-reactor site, but the NRC would address terminating each license separately. Issue 11-License Termination: Additional comments. Comment. A commenter stated that the need for a hearing when the licensee submits the license termination plan for approval should be reconsidered. If the licensee meets the requirements of the termination plan and applicable regulations, there would be no issues to adjudicate. Another commenter stated that, concerning the subpart L proceedings, the NRC should issue a clear statement of policy to eliminate the potential for significant litigation. Several commenters stated that if subpart L is to be used for hearings, it appears necessary to change the title of subpart L to include Part 50 licensees. Finally, a commenter stated that the applicability of Subpart L hearings should be incorporated into§ 2.700 as well as § 2.1201. Response. With respect to the termination plan, the Commission recognizes that ongoing rulemaking proceedings may result in establishing criteria for the restricted release of sites. Even if a hearing is not legally mandated at the termination stage as argued by some commenters, the Commission views it as appropriate to use the amendment process for approval of termination plans, including the associated opportunity for a hearing, to allow public participation on the specific actions required for license termination. In particular, the Commission has determined that, if a hearing is requested on the termination plan, the hearing must be completed before release of the site. This action will help ensure meaningful public input on any proposal for restricted release of the site. Given that a lengthy period (up to 60 years) may pass between the PSDAR stage and the termination stage, and given that final release criteria are still being developed that may include restricted release of a site, the Commission views a license amendment process as appropriate, along with the associated opportunity for a hearing, whether or not such hearings are mandated by legislation. Finally, the changes proposed by the commenters concerning the change of title of subpart L to include part 50 licensees and the incorporation of subpart L applicability into§§ 2.700 and 2.1201 are unnecessary because the rule already addresses these considerations. Comment. Many commenters expressed confusion on when a subpart L or subpart G hedring would be appropriate. One commenter noted that once fuel is out of the reactor vessel and in dry storage, there is no difference between storage on or off site and that reference to the subpart G hearing should be deleted. Another commenter wanted a clarification of what is meant by removing fuel from the site (i.e., under a part 72 license). Another commenter suggested that the wording to § 2.1201(a)(3) be clarified concerning permanent removal of fuel from the site to an authorized facility. One commenter inquired as to whether a license could be terminated if the licensee removed the fuel to an onsite ISFSI. Response. The final rule clearly indicates that once the fuel is removed from the licensed part 50 facility the power reactor facility can be treated as a materials facility where a subpart L hearing is appropriate. If fuel remains at the facility, a subpart G hearing is appropriate. If the fuel is in an ISFSI, that part of the affected site is regulated under a part 72 license and would no longer be regulated under the part 50 license. The wording in§ 2.1201(a)(3) has been changed to "removal of fuel from the part 50 facility," rather than "from the site," and means either removal offsite to an authorized facility or to an onsite facility (ISFSI) not under the part 50 license. Comment. Many commenters did not see the need for an environmental review at the license termination stage, and one suggested that it be considered a categorical exclusion. Another commenter stated that if there were to be an environmental review, its scope should be restricted to whether the licensee's controls and methods for mitigation of radiation will meet the standards adopted in § 20.1405 of the proposed residual radiation criteria rule. Response. At the license termination stage, an environmental assessment or impact statement will be required when the license is amended. Following resolution of another ongoing NRC rulemaking activity that is considering adoption of radiological release criteria, a categorical exclusion may be adopted that would eliminate the requirement for an environmental assessment or impact analysis, except in the case of a restricted release of a site. Comment. A few comments addressed proposed changes to§ 51.53 concerning requirements for environmental impact considerations. One commenter stated 50-SC-189 that the first sentence of the first paragraph of§ 51.53(b) should be deleted to be consistent with the concept that "a license amendment authorizing decommissioning activities" is no longer required. Revised wording should begin with "each applicant for a license amendment approving a license termination plan or decommissioning plan." Another commenter stated that § 51.53 should be revised to reflect the fact that the proposed rule, if adopted, would not require an amendment that authorizes the conduct of decommissioning activities, because neither the existing nor the proposed decommissioning process requires a license amendment to approve a decommissioning plan. Therefore the first paragraph of this section should be reworded as "[E)ach applicant for license termination upon submittal of the license termination plan under § 50.82 of this chapter either for unrestricted use or based on continuing use restrictions applicable to the site, * *

shall submit* * *"Asimilar change was stated to be needed in § 51.95 for the same reasons. Finally, a commenter noted that § 51.53(b) as well as § 51.95(b) refer to "applicants
*
for a utilization facility," which does not seem to be an element of the proposed rule and should be deleted; also, § 51.95(b) does not mention approval of a license amendment for license termination or a decommissioning plan, which is an omission and should be consistent with § 51.53(b).

Response. No change was made to this section because the non-power reactor facilitie's are still required to submit a decommissioning plan. For non-power reactors, the current rule remains essentially unchanged and requires submittal of a decommissioning plan that is approved through license amendment. The non-power reactor licensee must also submit an appropriate supplemental environmental report and the NRC must do an EA as part of the decommissioning plan approval process. Comment. Most of the commenters who were not in favor of the rule supported the license termination phase requirements but believe that these requirements were not timely and should be implemented in some manner at the initiation phase of decommissioning. Response. During the initial phase of decommissioning, the requirements in the final rule are designed to provide oversight commensurate with the level of safety concerns experienced in decommissioning, while providing PART 50

STATEMENTS OF CONSIDERATION additional opportunity for public comment on the licensee's proposed activities.

The final rule requirements are based on NRC's experience with licensees' use of the § 50.59 process during operations and consideration of the types of activities that licensees would undertake during the decommissioning process. Where appropriate, licensing requirements are continued through decommissioning and the NRC is informed of each licensee's planned decommissioning activities. (Additional discussion can be found in the response to Comment 5). Issue 12-Regulatory Guides. Comment. Several commenters requested regulatory guidance in the form of regulatory guides. These requests pertained to a standard format and content for the PSDAR and license termination plan as well as to transition guidance for licensees who are shut down and choose to adopt the new process. Additional guidance was also requested for a regulatory guide that dealt with the decommissioning process, such as a revision to Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors," that would include such topics as the objective and implementation aspects of public meeting and hearings, guidance on issues the NRC would consider in not giving negative consent approval to the PSDAR after the 90-day waiting period, guidance on interpretation and development of technical rule requirements, and guidance, on the particulars of "grandfathering." Additionally, several commenters requested additional financial guidance, through a regulatory guide, on the development and use of the decommissioning trust fund. Response. The NRC intends to issue regulatory guidance on the initial phase of decommissioning. Guidance on the standard format and content of the PSDAR will be issued after the final rule is published. Other guidance on the license termination phase is also being developed. Issue 13-Elimination of the Possession-only License Amendment (POLA). Comment. Generally, commenters in favor of the rule agreed with eliminating the POLA. Objections to POLA elimination from other commenters were that distinct categories between reactor operation and cessation of operation should be maintained and that eliminating the POLA process would eliminate a hearing opportunity prior to reactor decommissioning. Reflecting the views of many commenters against POLA elimination, a State commenter said that by deleting the POLA the NRC would eliminate the amendment process that expressly provides for State consultation (§ 50.91(b)) and that no subpart G hearing process would occur that would allow for discovery by parties to the proceeding and provide a mechanism for intervention. The State commenter held that the proposed rule delays the need for amendment to the license termination stage when it is too late; it is needed before major decommissioning activities are undertaken. Moreover, at the license termination stage, only a subpart L hearing is proposed (no discovery). Finally, a few commenters asked why non-power reactors, which are less hazardous facilities (smaller and less contaminated), can still request a POLA and still require decommissioning plan approval while power reactors no longer have this option or requirement. Response. If fuel is removed from the licensed part 50 facility, the activities undertaken during decommissioning are more like the kinds of activities undertaken at a typical materials facility where the subpart L process applies. The final rule requires that certain procedures be satisfied before a licensee can perform major decommissioning activities. These procedures include requiring a PSDAR submittal, conducting a public meeting, and allowing a specified time period for NRC review of the licensee's intended actions. Other final rule requirements prohibit the licensee from performing any major decommissioning activity that could result in significant environmental impacts not previously reviewed or foreclose the release of the site for unrestricted use. Written notification to the NRC is required for licensee decommissioning activities that are inconsistent with those described in the PSDAR, including significant changes in decommissioning costs. Finally, the final rule extends certain regulatory requirements to decommissioning. Thus, licensee activities that would require approval under a POLA are no longer necessary. The affected State(s) will be notified about the public information meeting as well as consulted on the licensee's planned decommissioning activities by the NRC prior to the public meeting. The final rule requires that a copy of the PSDAR and any written notification of inconsistent PSDAR activities be sent to the affected State(s). In response to the comment concerning why non-power reactors are still given the option of submitting a POLA and still require a decommissioning plan, it is noted that such reactors are required to 50-SC-190 immediately dismantle, except for extenuating circumstances, and are not permitted a storage period (because there is no significant health, safety or environmental reason for delay-see FGEIS, NUREG 0586).1 Issue 14-"Grandfathering" Considerations. Comment. There were several commenters who were concerned that the proposed rule did not significantly address nor provide necessary guidance for "grandfathering" issues. Specific comments in this area were that recognition should be given to those plants whose decommissioning plans have been approved on a case-by-case basis; that if existing facilities are grandfathered from any part of the proposed rule, it should clearly identify this; that the proposed rule does not adequately implement the grandfathering option because the current§ 50.82 would disappear from the rule and no explicit provisions would exist to rely on. It is suggested that the NRC keep the old provision as well as an applicable alternative and; that for grandfathering, an implementation provision should be added to the rule in a fashion similar to § 20.1008. Several commenters also noted that guidance needs to be given to those licensees who are in various aspects of decommissioning based on the current rule requirements and wish to switch to the proposed rule requirements. Response. The Commission has reconsidered the issue of "grandfathering" and modified the language in the final rule to provide more specific guidance for nuclear power reactor licensees whose facilities are currently at certain stages of decommissioning. The Commission has decided to eliminate the provision in the proposed rule that would give those licensees that have an NRC approved decommissioning plan, before the date when a final rule became effective, the option of either complying with the final rule requirements or continuing with the requirements of the currently existing rule. All licensees will be required to comply with the decommissioning procedures specified in the provisions of the final rule, when it becomes effective. The final rule addresses the process for converting from the existing rule requirements to those in the final rule for those nuclear power reactor licensees whose facilities are already at certain stages of decommissioning. For power reactor licensees who, before the effective date of this final rule, either submitted a decommissioning plan for approval or PART 50

STATEMENTS OF CONSIDERATION possess an approved plan, the plan will be considered as the PSDAR submittal and the licensee will be required to perform decommissioning in conformance with these final rule requirements.

However, for power reactor licensees who are involved in subpart G hearings of 10 CFR part 2, conversion to the new rule will not be permitted until the hearing process is completed. The public meeting and 90-day hold on decommissioning activities required in § 50.82(a) (4)(ii) and (5) will not apply. Those licensees will be subject to any orders arising from these subpart G hearings, absent any orders from the Commission. For nuclear power reactor facility licensees whose licenses have been modified, before the effective date of this rule, to allow possession but not operation of the facility, the certifications required in § 50.82(a)(l) will be considered to have been submitted. With regard to extending current rule requirements for "grandfathering" considerations, no current rule requirements need be .retained because the "grandfathering" provision in the

proposed rule has been eliminated in the final rule. The final rule covers conversion from the existing requirements for approval of a submitted or approved decommissioning plan, as described above, and is specific to existing licensee decommissioning plan situations.

Issue 15-Miscellaneous Comments. Comment. Several commenters stated that the backfit rule, § 50.109, should apply to decommissioning because a proper reading of the intent of that rule should cover rulemaking dealing with decommissioning. Otherwise, additional requirements could be imposed without a benefit cost analysis. Response. The Commission has concluded that the provisions addressed in this rulemaking do not involve a backfit because they address only reactors that have permanently ceased operations and § 50.109 only applies to design, construction and operation of a facility. These regulations are primarily procedural in nature and, to the extent they address nonprocedural matters, they are a codification of existing process. . Comment. A few commenters noted that the regulatory analysis for the proposed rule did not evaluate the alternatives to the proposed new regulatory requirements and existing requirements do not require a license termination plan or a license amendment to approve a license termination plan. The regulatory analysis does not accomplish the objective of ensuring that all regulatory burdens are needed, justified, and minimal. Response. The regulatory analysis did evaluate the alterdatives to the proposed new regulatory requirements. The license termination plan is not a new requirement because, under the existing rule, licensees are required to submit a proposed decommissioning plan for approval within 2 years of permanent shutdown. Currently, licensees who plan to delay decommissioning by including a period of storage must submit a final decommissioning plan for approval before starting decommissioning. Current NRC policy is to approve the decommissioning plan by license amendment. Because the proposed rule would permit the licensee use of the § 50.59 process to perform major dismantlement activities, the license termination plan is less complex than a decommissioning plan and covers the remainder of activities requiring completion to terminate the license, other than dismantlement activities. The changes adopted in the rulemaking primarily provide additional flexibility to licensees that reduces burden without reducing safety by allowing licensees to undertake the majority of decommissioning activities without first obtaining NRC approval. Comment. Several commenters wanted the option of entombment to be allowed because restricted release will be allowed when the residual radiation criteria rule is final. Aside from the difficulty of disposal, the money not spent on LLW burial is substantial. The interest on this money would be more than adequate to provide for the maintenance and surveillance required for the entombment option. The public, including local communities, may be interested in not transporting waste across state boundaries and in keeping funds that would otherwise be spent on disposal within the community. Response. The issue of entombment was not addressed in this rule. The NRC position on entombment is the same as in the current rule. Entombment would only be permitted for very special circumstances but would involve a continued license on a case-by-case basis. The concept of restricted release included in the proposed rule on residual radiation criteria would involve termination of the license with restrictions in place to limit the use of the facility by the public, but certain radiological criteria for restricted release would have to be met. Comment. Several individual commenters wanted to know whether NRC rules allow the optional period of 50-SC-191 storage of the reactor facility to be longer than 60 years and does the 60-year completion date for decommissioning specified in the current rule consider storage of fuel in an ISFSI. One commenter stressed that spent fuel should not be separated from any of the phases of decommissioning because this is a piecemeal approach and inappropriate. Another commenter stated that the licensee should be required to maintain capability to handle the fuel for dry cask storage. Response. The primary considerations of the proposed rule were procedural, with emphasis on the issue of premature closure. Other aspects of the existing rule were unchanged. A 60-year period for completion of decommissioning is still imposed, subject to other considerations delineated in the current rule requirements. The existing rule, as well as the proposed rule, consider the storage and maintenance of spent fuel as an operational consideration and provide separate part 50 requirements for this purpose. Regarding maintaining the capability to handle the fuel for dry cask storage, these requirements are maintained in 10 CFR part 72. Comment. Several commenters noted that the requirements of this proposed rule and the proposed residual radiological criteria rule should be coordinated to avoid redundancy. Response. The two rules will be coordinated. Comment. A few commenters noted that a complete site characterization should be included at the initiation of decommissioning activities and that mandatory site radiological surveys should be required before issuing a new license to establish background conditions. Response. These considerations are being addressed during finalization of the residual radiological criteria rule. Comment. Finally, several commenters requested that the NRC consider the impacts of the proposed "safeguards for nuclear fuel or high level radioactive waste" rule (60 FR 42079; August 15, 1995) (which affects parts 60, 72, 73, and 75) on this rule when that proposed rule is issued in final form. Response. This rule is primarily directed toward the procedural requirements necessary for power reactor decommissionings. Therefore, the requirements imposed by this rule can be treated independently from the other "safeguards" rule under development. That rule, when final, may modify some of the technical requirements imposed by this final rule. PART 50

STATEMENTS OF CONSIDERATION Resolution of Comments on the Draft Policy Statement On February 3, 1994 (59 FR 5216), the NRC published in the Federal Register a draft policy statement and accompanying criteria relating ~o J?O"'.er reactor licensee use of decomm1ss10mng trust funds before NRC approval of licensees' decommissioning plans. The proposed rulemaking to amen~ t~e . procedural aspects of decomm1ss10nmg (60 FR 2210; July 20, 1995) codifie? the position embodied in the draft pohcy statement.

Based on the NRC's resolution of comments on the proposed rule and incorporated into this final rule, the criteria in the draft policy statement have been modified. No final policy statement will be issue?. _Other* changes in the final rule pertammg to licensee use of decommissioning trust funds were discussed earlier in the section on Response to Comments. The NRC received comments on the draft policy statement from th~ . following individuals or orgamzat10ns:

1. Michigan Department of Commerce 2. Citizens Awareness Network 3. Mary P. Sinclair 4. Detroit Edison Company 5. Committee for a Safe Energy Future 6. Jon Block 7. Nuclear Energy Institute

8. Yankee Atomic Electric Company 9. Virginia Power Company 10. New England Coalition on Nuclear Pollution

11. Winston & Strawn 12. Consolidated Edison Company 13. Maryland Department of the Environment

14. TU Electric Company The public interest group, individual commenters, and one State oppose allowing any withdrawals from decommissioning trust funds before the NRC approves a licensee's decommissioning plan, a procedure that this final rule has discontinued.

The other commenters generally supported the draft policy statement, although they disagreed with certain provi~ions or .. took issue with the need for 1t. Specific comments and observations, and the NRC analysis of arid response to them, are discussed below. Specific Comments Comment 1. The trust agreements may need to be modified to include low-level radioactive waste storage and disposal (LLW) and interim spent fuel storage as allowable decommissioning costs when these costs are incurred as part of additional, temporary facilities at particular sites. LLW disposal cos~s, in particular, should be able to be paid from the decommissioning waste fund without waiting 60 days for NRC approval. Provisions should be in~luded for decommissioning nonradioactive structures associated with the reactor (Commenters 1 an,d 4). Response. The policy statement and this rule were not intended to address this issue. This issue is being addressed separately (see SECY 95-223; September 1, 1995). As provided in 10 CFR 50.75, financial assurance for decommissioning includes the cost of disposal of LLW associated with rea~t?r decommissioning. If a temporary fac1hty is built to store LLW under the Part 50 reactor license, the trust agreement should have been structured to include these costs. Although the NRC definition of decommissioning excludes interim storage of spent reactor fuel, a licensee is required to provide for the cost of interim spent fuel storage under 10 CFR 50.54(bb). With respect to the issue of waiving the 60-day NRC approval period for withdrawals to pay for LLW shipments, this final rule eliminates the procedure to which this comment referred. Comment 2. The NRC should not allow decommissioning trust fund withdrawals before an environmental assessment is performed while the reactor licensee has a possession-only license because: (1) It will allow scale decommissioning activities without a resident NRC inspector site during the removal of irradiated components; (2) it is inconsistent with the mandate of the NRC, which is to implement a submitted, reviewed, publicly evaluated, and approved decommissioning plan before large-scale decommissioning activities begin; (3) health and safety of the workers and the public can not be adequately served by the experimental process of the component removal process, and (4) existing NRC regulations sta!e ~hat a licensee may only conduct hm1ted activities prior to approval of the decommissioning plan (e.g., decontamination, minor component disassembly, shipment and storage of spent fuel). Reasonable interpretation of the rules does not require expansion of 10 CFR 50.59 and/or activities permitted under a license (Commenters 2, 3, 5, 6, and 10). There could be insufficient financial resources remaining to decommission Nuclear Power Plants thus, creating a potential burden on the State and, serious impairment of radioactive material licensee's ability to complete decommissioning. Most existing decommissioning 'certifications and funding plans' are generally acknowledged by the NRC to already be severely UNDERFUNDED. This rule would exacerbate that situation (Commenter 13). 50-SC-192 Response. This final rule addresses the process that licensees are to use for post-shutdown decommi~si?ning activities, as well as the hm1ts on the amounts to be withdrawn from decommissioning trust funds. By permitting a lJce1:1see to_p~~form certain decommiss10nmg achv1hes and to withdraw funds for those activities through use of the PSDAR submittal process required in the final !ule will allow the licensee to reduce its overall decommissioning costs by taking advantage of lower low-~eve~ radioac~ive waste disposal costs. This will b~nefit the licensee and its ratepayers without adversely affecting public health and safety. Comment 3. The NRC should develop a similar policy for operating plants and should allow licensees to withdraw decommissioning trust funds to dispose of structures and equipment no longer being used for operating plants (Commenters 7, 8 (by reference), and 14). Footnote 2 of the policy statement should be revised to clarify that the policy statement does not apply "to licensee withdrawals from decommissioning funds for operating plants" rather than stating that t~,e policy statement does not apply to licensees with operating nuclear reactors" (Commenter 11). Response. The NRC has concluded that allowing decommissioning trust fund withdrawals for disposals by nuclear power plants that continue to operate is not warranted. T~ese activities are more appropriately considered operating activities and should be financed in that way. Footnote 2 is not included in this final rule. Comment 4. The policy statement may become obsolete if the l;lR~ a?opts a new definition of decomm1ss10mng as proposed on February 2, 1994 (59 FR 4868). This definition states, "Decommissioning means to remove a facility or site safely from service and reduce residual radioactivity to a level that permits use of the pr~per~y for unrestricted use and termmahon of the license, or (2) release of the property under restricted conditions and termination of the license." To avoid obsolescence of the policy statement as a result of changes in the definition of decommissioning, the commenters recommend replacing all references to release of the site for unrestricted use with "decommissioning of the site consistent with the definition in§ 50.2" (Commenters 7, 8 (by reference), and 11). ; PART 50

STATEMENTS OF CONSIDERATION Response.

The NRG agrees with this recommendation and has changed this final rule accordingly. Comment 5. Two commenters disagree with a statement in the draft policy statement, "If a licensee of a permanently shut down facility spends decommissioning trust funds on legitimate decommissioning activities, the timing of these expenditures, either before or after NRC approves a licensee's decommissioning plan, should not adversely affect public health and safety, provided adequate funds are maintained to restore the facility to a safe storage configuration in case decommissioning activities are interrupted unexpectedly" (Commenter 7's emphasis). The commenters state that maintaining a viable SAFSTOR option beyond plan approval should not be required for cases where another option has been approved by NRG (Commenters 7 and 8). The draft policy statement misuses the term "SAFSTOR" to mean maintenance of a site in a safe storage condition prior to receipt of Decommissioning Plan approval and commencement of decommissioning rather than a specific decommissioning alternative defined in NRG regulations (Commenters 11 and 14). Response. Commenter 7 has misinterpreted the intent of this statement. First, this part of the policy statement was drafted to make the point that any expenditures for decommissioning activities normally viewed as necessary would not be detrimental to public health and safety, notwithstanding the timing of these expenditures, unless they were large enough to prevent the licensee from returning its facility to a safe storage configuration if the decommissioning process were to go awry. This is not the same as requiring a licensee to switch from DEGON (immediate dismantlement) to SAFSTOR after the NRG has approved the licensee's decommissioning plan. This final rule modifies use of the above-referenced criterion for decommissioning trust fund withdrawals. However, the rule corrects any references to SAFSTOR when it means to address the general ability of a licensee to return its reactor to safe storage while awaiting further decommissioning. Comment 6. Criterion 4 is redundant of the other criteria (Commenters 7 and 8). At a minimum, the statement should indicate that items (c) and (d) of criterion 4 do not require NRC approval before a licensee undertakes the proposed activities (Commenter 8). Redundancies can be eliminated by factoring the first three criteria into criterion

4. However, issuance of the policy statement based on criterion 4 (or the other criteria) is premature in that the NRG is curren,tly considering more definitive guidance on acceptable pre-p Ian-approval decommissioning activities (Commenter 11). Response.

The NRC agrees that some confusion may have arisen by including criterion 4 in the policy statement. The NRG included this criterion to provide guidance on the allowed decommissioning activities as opposed to the use of decommissioning trust funds for those activities. Criterion 4 is a quote from Commission guidance in the SRM of January 14, 1993, and, to some degree, overlaps the other criteria of the policy statement. The NRG has removed criterion 4 as a separate criterion in this final rule. Comment 7. The "ancillary issue" in the draft policy statement should be expanded to include a number of expenses that are paid out of decommissioning trusts by operating plants well in advance of licensee preparation and submission of the decommissioning plan. These expenses include, but are not limited to, trust fees, investment manager fees, income taxes, and periodic site-specific studies (Commenters 7, 8 (by reference), 11, and 14). The policy statement should be revised to state specifically that if a licensee determines that it meets the criteria for de minimis withdrawals, it need not request permission from the NRG to use these funds (Commenter 8). * *

The section dealing with 'de minimis' withdrawals for developing the decommissioning plan also seems to be outside the original intent for use of these funds. These withdrawals may seem to be a minor portion of funds allocated for decommissioning, but it starts a process that would allow utilities to tap these funds, if they can fit activities into the definition of decommissioning or simply request to use these funds for other purposes * *
Other uses are unacceptable, even if they are subject to prior regulator approval (Commenter 13). Response.

The intent of the ancillary issue was to allow de minimis withdrawals from decommissioning trust funds of up to $5 million for decommissioning-related administrative and other expenses without prior NRC consent notwithstanding the operating status of the plant. The final rule has changed this withdrawal amount to up to 3 percent of the generic amount specified in§ 50.75(c). This withdrawal amount is for purposes of planning for decommissioning (paper studies) and pertains to licensees of operating as well 50-SC-193 as permanently shut down plants. Permission from the NRG to use these funds in de minimis amounts is unnecessary as long as the amount and purpose of the withdrawal is documented. With respect to Commenter 13's concerns, the NRG has specified a maximum limit for de minimis withdrawals. If a licensee were to exceed this limit or use funds for decommissioning purposes, it would be subject to NRG enforcement action. Comment 8. "* *

The NRG has neither articulated the reasons why this detailed level of oversight (discussed in the policy statement) is needed, nor has the NRG provided specific examples of potential waste and misuse of funds that would warrant their proposed oversight
*
Absent an appropriate justification for the implementation of this policy statement, * *
this policy statement represents regulation without benefit (and that NRG concerns expressed in the policy statement) are not tangible for decommissioning." Thus, the policy statement should not be issued (Commenter 9). Also, "the draft policy statement provides no basis for the NRC's conclusion that prior NRG review of pre-plan-approval decommissioning fund expenditures should be required." The draft policy statement may satisfy the Commission's directive to the NRC staff to develop a policy without including an approval mechanism (Commenter 11). The draft policy statement is not clear as to the purpose of the NRC review of decommissioning expenditures before decommissioning plan approval.

The only reason for the review, given in the statement of policy, is to ensure the health and safety of the general public. There are other regulatory mechanisms for evaluating the activity for which the funds are withdrawn without reviewing the actual withdrawal from the fund. The expenditure of decommissioning trust funds for legitimate decommissioning activities is an economic and not a safety concern (Commenter 14). Response. Although the NRC did not include specific examples of waste and misuse of funds in the policy statement, as with any industrial process, costly mistakes can conceivably occur in decommissioning. The NRC also disagrees that codifying decommissioning trust fund withdrawals represents regulation without benefit. The NRG has specifically promulgated decommissioning requirements in 10 CFR 50.82 that include licensee PSDAR submittal process that is intended for PART 50

STATEMENTS OF CONSIDERATION keeping the NRC and public informed of the licensee's planned decommissioning activities.

The intent of the regulations is to require licensees to maintain the entire amount of funds needed for decommissioning in a specified assurance mechanism until the funds are used for their intended decommissioning activities. The PSDAR is closely tied to a licensee's provision of assurance to fund the decommissioning activities adequately. Without any NRC criteria for expenditures before the PSDAR submittal process is completed, the decommissioning trust fund could become a shell and thus defeat the purpose of NRC decommissioning funding assurance regulations. Because of the safety implications of inadequate decommissioning funds, the NRC believes it has responsibility for specifying withdrawal rates, notwithstanding the reviews that rate regulators may perform. Comment 9. Trust fund withdrawals should also be permitted for early decommissioning-related activities that, although not themselves directly reducing radioactivity at the site, will significantly facilitate such activities when they subsequently occur (Commenters 11 and 12). Response. In this final rule, withdrawals for planning activities are allowed before completion of the PSDAR process. Comment 10. The NRC should clarify footnote 2 to indicate that it applies to licensees of multi-unit sites. "So long as usage of trust withdrawals is identifiable with the shut down reactor and does not diminish decontamination funding subsequently available for reactors which are continuing to operate, there is no reason why reactor licensees should be treatecr* differently than single-reactor licensees for purposes of this policy statement" (Commenter 12). Response. The NRC agrees with this statement. However, footnote 2 is not included in this final rule. Comment 11. "If the NRC believes that NRC review and approval of plan-approval decommissioning expenditures is necessary, it should act through rulemaking rather than policy * *

Since prior NRC review of decommissioning fund withdrawals is not currently required, if the NRC wishes to impose such a requirement, it should initiate rulemaking to revise its decommissioning regulations accordingly" (Commenter 11). Response.

This final rule codifies criteria for decommissioning trust fund withdrawals. Thus, this commenter's concerns have been addressed. Comment 12. "The 'tacit consent' approach for reviewing licensee expenditure plans is inappropriate" and unsupported by the reasons the NRC stated for its policy. By expressly preserving the possibility that it would take action to prevent a fund withdrawal, the NRC blurs its asserted distinction between review and approval. Also, it is not clear that "tacit consent" and "approval" are legally distinguishable for purposes of determining whether the NRC is engaged in a "licensing action" that could involve public participation and environmental review (Commenter 11). Response. The NRC does not use "tacit consent" in this final rule. Thus, the concerns expressed in this comment should be assuaged. Comment 13. "Criterion 1 * *

should be revised to eliminate the provision that withdrawals must be for activities

'that would necessarily occur under most reasonable decommissioning scenarios."' This phrase adds nothing to the preceding provision that the withdrawal must be for "legitimate decommissioning activities." Because licensees may face decommissioning expenditures for activities that are within the NRC's definition of decommissioning but nonetheless unique to their plant(s), the proposed provision is inappropriately restrictive (Commenter 11). Criterion 1 is overly restrictive and burdensome*

If the NRC wants to prevent activities that preclude release of the site for (un)restricted use or are not in support of decommissioning efforts it should require review of the activity itself through any of the other available mechanisms such as 10 CFR 50.59 or special rulemaking
*
The basic premise is that in the event that there are circumstances or conditions which delay or preclude proceeding with the decommissioning effort there will be funds available to place the plant in a storage condition until the event or circumstance is resolved.

Thus, as long as the value of the fund does not fall below the regulatory required amount in effect at the time of the request the withdrawal should be allowed. Thus, the only requirement should be that the utility document that [the] activity was a legitimate decommissioning activity and the expenditure was reasonable (Commenter 14). Response. The NRC did not mean to imply that decommissioning activities unique to one site would not be eligible for early trust fund withdrawals. However, because we agree that the phrase, "legitimate decommissioning activities," is sufficient, the NRC has eliminated the phrase from this final rule. Comment 14. "* *

The explicit characterization as a decommissioning

'contingency' of the funding 'necessary 50-SC-194 to maintain the status quo' could be construed inappropriately to require that licensees include funding for that purpose in their decommissioning funds * *

If this criterion is retained, the language regarding provisions for this contingency should be deleted from the policy statement" (Commenter 11). Response.

This terminology has been eliminated in this final rule. Comment 15. "It does not seem necessary that NRC approve requests for the 'withdrawal of decommissioning funds for early equipment removal, prior to approval of the utilities[') decommissioning plans. This does not seem in concert with the intent of the sample statement under Background '* *

the fund trustee should only Telease funds upon certification that decommissioning is proceeding pursuant to an NRC-approved plan'" (Commenter 13). Response.

This final rule does not continue the language in question. Comment 16. "* *

This ruling may be judged as an item of Compatibility (for Agreement States). Because Maryland regulations, policies, etc., are expected to closely follow Federal rules and procedures, we would be forced to adopt and allow our licensees to use the same principle" (Commenter 13). Response.

The NRC does not believe that this is an issue of State compatibility because this final rule only applies to power reactor licensees, which are exclusively NRC licensees. Summary of Changes in the Final Rule Based on the response to comments, a few changes were made in the final rule. Otherwise, the final rule provisions are the same as those presented in the "background" section under the section titled proposed amendments. Specific changes made to the proposed rule in the final rule are summarized as follows: (1) Section 50.2. The definition of "major radioactive components" has been clarified. (2) Section 50.36a(a)(l). The amendment has been changed to exclude systems that are no longer necessary for compliance. (3) Section 50.59. Proposed§ 50.59(e) was eliminated. However, three of the proposed rule requirements contained in § 50.59(e) were moved to § 50.82(a) (6) and (7). Placing these requirements in § 50.82 as overall constraints, rather than specific requirements for each § 50.59 activity, required modification of the constraint that the decommissioning activities not result in significantly increasing decommissioning costs. Thus, the final rule (§ 50.82(a)(6)(iii)) prohibits PART 50

STATEMENTS OF CONSIDERATION decommissioning activities that would result in there no longer being reasonable assurance that adequate funds will be available to complete decommissioning.

In addition, the final rule requires in § 50.82(a)(7) that changes from those specified in the PSDAR that would result in significantly increasing decommissioning costs require written notification to the NRC. The fourth requirement that the terms of the existing license not be violated was eliminated. The requirement to consider environmental impact in the PSDAR, § 50.82(a)(4) was modified to explicitly require the reasons for concluding that any environmental impacts will be bounded by existing analysis. (4) Section 50.71. Section 50.71(e)(4) was revised to permit nuclear power reactor licensees that have submitted the certifications required under § 50.82(a)(1) to update the FSAR every 24-months. (5) Sections 50.82(a)(4)(i) and (6). The licensee is required to send a copy of the PSDAR and written notification of departure from the PSDAR to the NRC and affected State(s). (6) Section 50.82(a)(8)(ii). The phrase "being permitted to use" was removed from this section to avoid any incorrect interpretation that the NRC must explicitly approve decommissioning funding expenditures. (7) Section 50.82. Specifies that once the rule is effective, all power reactor licensees must comply with it. Power reactor licensees that possess an approved plan as well as licensees that applied for plan approval before the rule took effect would have the plan considered a PSDAR submittal, and licensees would be permitted to perform decommissioning activities in accordance with § 50.82. However, for power reactor licensees who are involved in subpart G hearings of 10 CFR part 2, conversion to the new rule will not be permitted until the hearing process is completed and those licensees will be subject to any orders arising from these hearings absent any orders from the Commission. (8) Section 50.82(a)(1)(iii). Specifies that once the rule is effective, power reactor licensees whose licenses have been modified, before the effective date of this rule, to possess but not operate the facility, will be considered to have submitted the certifications required in § 50.82(a)(l). (9) To improve clarity, the first sentence in§ 2.1205(d)(l) has been rewritten from that proposed to that found in the existing regulation. (10) To improve clarity and maintain parallelism of requirements, the last sentence of§ 51.53(b) has been rewritten from that found in the proposed rule to correspond with the language found in § 51.95(b) of the proposed (and existing) rule. * (11) To improve clarity, § 50.82(a)(9)(ii) (B) and (F) have been rewritten. Finding of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore, an environmental impact statement is not required. The final rule clarifies current decommissioning requirements for nuclear power reactors in 10 CFR Part 50 and presents a more efficient, uniform, and understandable process. The Commission has analyzed the major environmental impacts associated with decommissioning in the Generic Environmental Impact Statement (GEIS), NUREG-0586, August 1988, 1 published in conjunction with the Commission's final decommissioning rule (53 FR 24018; June 27, 1988). Insofar as this rule would allow major decommissioning (dismantlement) to proceed without an environmental assessment, the environmental impacts of this rule are within the scope of the prior GEIS. The environmental assessment for the final rule and finding of no significant impact on which this determination is based are available for inspection and photocopying for a fee at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from Carl Feldman, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, (301) 415-6194. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Management and Budget, approval number 3150-0011. Because the rule will relax existing information collection requirements, the public burden for this collection of information is expected to be decreased hy 12,202 hours per licensee. This reduction includes the time required for reviewing instructions, searching 50-SC-195 existing data sources, gathering and maintaining the data needed and completing and reviewing the collection of information. Send comments on any aspect of this collection of information, including suggestions for further reducing this burden, to the Information and Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001, or by Internet electronic mail to BJSl@NRC.GOV; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management and Budget, Washington, DC 20503. Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid 0MB control number. Regulatory Analysis The NRC has prepared a regulatory analysis for this final rule. The analysis qualitatively examines the costs and benefits of the alternatives considered by the NRC. In the response to comments, the NRC concluded that only some minor changes to the draft regulatory analysis were necessary, corresponding to some minor procedural changes in the final rule. The regulatory analysis is available for inspection in the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC 20555-0001. Single copies of the analysis may be obtained from Dr. Carl Feldman, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-6194. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the Commission certifies that this rule will not have a significant economic impact on a substantial number of small entities. The final rule modifies requirements for timely decommissioning of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of small entities as given in the Regulatory Flexibility Act or the Small Business Size Standards promulgated in regulations issued by the Small Business Administration (13 CFR Part 121). This discussion constitutes the analysis for the regulatory flexibility certification requirement. PART 50

STATEMENTS OF CONSIDERATION Small Business Regulatory Enforcement Fairness Act In accordance with the Small Business Regulatory Enforcement Fairness Act of 1996, the NRC has
determined that this action is not a major rule and has verified this determination with the Office of Information and Regulatory Affairs, 0MB Backfit Analysis The Commission has determined that the backfit rule, 10 CFR 50.109, does not apply to these final amendments, and therefore, a backfit analysis has not been prepared for this rule. The scope of the backfit provision in 10 CFR 50.109 is limited to construction and operation of reactors.

These final amendments would only apply to reactors that have permanently ceased operations and, as such, would not constitute backfits under 10 CFR 50.109. List of Subjects 10CFRPart2 Administrative practice and procedure, Antitrust, Byproduct material, Classified information, Environmental protection, Nuclear materials, Nuclear power plants and reactors, Penalties, Sex discrimination, Source material, Special nuclear material, Waste treatment and disposal. 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements. 10 CFR Part 51 Administrative practice and procedure, Environmental impact statement, Nuclear materials, Nuclear power plants and reactors, Reporting and recordkeeping requirements. For reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendments to 10 CFR parts 2, 50, and 51. 61 FR 41303 Published 8/8/96 Effective 9/9/96 10 CFR Part 50 RIN 3150-AC93 Codes and Standards for Nuclear Power Plants; Subsection IWE and Subsection IWL AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations to incorporate by reference

the 1992 Edition with the 1992 Addenda of Subsection IWE, "Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water.

Cooled Power Plants," and SubsectionlWL, "Requirements for Class CC Concr_ete Components of Water Cooled Power Plants," of Section* XI, Division 1, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) with specified*modifications and a limitation.

-Subsection IWE of the ASME Code provides rules for inservice inspection, repair, and replacement of Class MC pressure retaining components and their integral attachments and of metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments in light-water cooled power plants. Subsecti.on IWL of the ASME Code provides rules for inservice inspecUon and repair of the reinforced concrete and the post-50-SC-196 tensioning systems of Class CC components.

Licensees will be required _ to incorporate Subsection IWE and Subsection IWL into their inservice inspection (ISI) program. Licensees will also be required to expedite implementation of the containment examinations and to complete the expedited examination in accordance with Subsection lWE and Subsection IWL within 5 years of the effective date of this rule. Provisions have been --included that will prevent unnecessary duplication of examinations between the expedited examination and.the routine 120-month ISI examinations. Subsection IWE and Subsection IWL have not been previously incorporated by reference into the NRC regulations. The final rule specifies requirements to assure that the critical areas of

containments are routinely inspected to detect and take corrective action for defects that could compromise a containment's structural integrity.

EFFECTIVE DATE: September 9, 1996; The incorporation by reference of certain . publications listed in the regulations is * *. approved by the Office of the Director of the Office of the Federal Register as of September 9, 1996. FOR FURTHER INFORMATION CONTACT: Mr. W. E. Norris, Division of Engineering

Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6796.

SUPPLEMENTARY INFORMATION: The NRC is amending its regulations to incorporate by reference the 1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL to assure that the critical areas of PART 50

STATEMENTS OF CONSIDERATION
cqntainments are routinely inspected to detect and take corrective action for .. defects that could compromise a containment's structural integrity.

The rate of occurrence of degradation in containments is increasing. Appendix J to 10 CFR part 50 requires a general visual inspection of the containment but does .not provide specific guidance on* how to perform the necessary containment examinations. This has resulted in a large variation with regard to the performance and the.effectiveness of containment examinations. The rate of occurrence of corrosion and degradation..of containment structures -has been increasing at operating nuclear power plants. There have beeri 32 reported occurrences of corrosion in metal containments and the liners of concrete containments. This is fourth of all operating nuclear power plants. Only four of the 32 occurrences \Vere detected by clirrent containment inspection programs. Nine of these occurrences were first identified,by. the

NRC through its inspections or
structural audits. Eleven occurrences . were detected by licensees after they were alerted to a degraded condition at another site or 'through activity other than,containment inspection.

There have,been 34 reported occurrences of . degradation of the concrete or of the post-tensioning systems* of concrete containments. This is nearly one-half of these*types of containments. It is clear that current licensee containment

inspection programs have not proved to be adequate to detect the types of* degradation which have .. been reported.

Examples of degradation not found by . licensees; but initially detected at*plants throµghNRC inspections include: (1) Corrosion of steel containment shells in the drywelLsand cushion region,

nisulting in.wall'thickness reduction to below the minimum design thickness; (2) corrosion of the torus of the steel
containment shell (wall thickness below minimum design. thickness);

(3) corrosion:of the liner ofa concrete containment to-approximately half-* depth; (4) grease leakage from the tendons-of prestressed concrete containments; and (5) leaching as well . as excf;!ssive cracking in concrete containments. There are several General Design Criteria (GDC) and ASME Code sections which establish minimum requirements . for the design, fabrication, construction, testing, and performance of structures, systems, and components important to safety in water-cooled nuclear power plants. The GDC serve as fundamental underpinnings for many of the most . safety important commitments in licensee design and licensing bases. GDC 16, "Containment design," requires the provision of reactor containment , and associated systems to establish an essentially leak-tightbarrier against the uncontrolled releAse of radioactivity into the environment and to ensure that . the-containment design conditions . important to safety are not exceeded for as long as required for postulated accident conditions. Criterion 53, "Provisions for containment testing and inspection,"* requires that the reactor containment design permit: (1) Appropriate periodic inspection of allimportant areas, such as penetrations; (2) an appropriate surveillance program; and (3) periodic testing at *containment design pressure of the leak-tightness ofpenetrations which have resilient seals and expansion bellows. Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," of 10 CFR part 50 contains specific rules for leakage testing of containments. Paragraph

m. A. of* Appendix J requires that a general. inspection of the accessible interior and exterior surfaces of the containment structures and components be performed prior to any Type A test to uncover any evidence of structural deterioration that may affect either the containment structural integrity or tightness (Type A test means tests intended to measure the primary reactor containment overall integrated leakage rate: (1) after the containment has been completed.and is ready for operation, . and (2) at periodic.intervals thereafter).

The metal containment structure of operating nuclear power plants were* designed in accordance with either . Section m, Subsection NE, ~*class MC Components," or Section vm, of the ASME Code. These subsections contain provisions for the design and construction of metal containment structures, including methods for -determining the minimum required wall thicknesses .. The minimum wall thickness is that thickness that would ensure that the metal containment structure would.continue to maintain its structural integrity under the various stressors and degradation mechanisms which could act on it. The prestressed concrete containments of most operating nuclear reactors were designed in accordance with ACl-318 provisions taking into. consideration their unique features in the design of the post-tensioning system and in determining the prestressing forces. The post-tensioning system is. designed so that the concrete containment structure will continue to maintain its structural-integrity under 50-SC-197 the various stressors and degradation mechanisms which act on it. The liners of concrete containments provide a tight barrier. These requirements for minimum design wall thicknesses and prestressing forces as provided in these industry standards used to design co.ntainment structures are reflected in license co~ditions, technical specifications, and licensee commitments (e.g., the Final Safety Analysis Report). None of the .existing requirements, however, provide specific guidance on how to perform the necessary . containment examinations. This lack of guidance has resulted in a large . variation with*regard to the performance and .the effectiveness*of licensee containment. examination programs. Based on the results of inspections and audits, as well as plant.operational experiences, it is clear that many licensee containment examination programs have not detected degradation that could ultimately result in a compromise to the pressure-retaining capability. Some containment structures have been found to have.undergone a significant level of degradation that was not detected by these .programs. The Nuclear Management and Resources Council (NUMARC) (which has since become the Nuclear Energy Institute (NEI)) developed irnumber*of industr.y reports.to address,license renewal issues. Two of those, one for Pressurized Water Reactor. (PWR). containments and the other for Boiling . Water Reactor (BWR) containments; were developed for the purpose of managing age-related degradation of containments on a generic basis. The NUMARC.plan for co~tainments relies. on the examinations contained in Subsection IWE and Subsection IWL .to manage age-related degradation, and ... this plan assumes that these

examinations are '.'in current and effective use." In thaBWR Containment.

Industry Report, NUMARC concluded

that "On account of these available and established methods andtechniques to adequately manage potential degradation due to general corrosion of freestanding metal containments, no additional measures need to be developed and, as such, general corrosion is not a license renewal concern if the Containment minimum wall thickness is maintained and verified." Similarly;,in the PWR. Containment

'Industry Report, NUMARC concluded that potentially significant degradation of concrete surfaces, the posMensioning system, and the liners of concrete containments could be managed effectively if periodically* examined in accordance with the PART 50

STATEMENTS OF CONSIDERATION .requirements contained in Subsection IWE and Subsection IWL. The NRC agrees with NEI that these ASME standards, which the industry has participated in developing, would be an effective means for managing age-related containment degradation.

Thus, the NRC believes that adoption of these standards is the best approach. Background On January 7, 1994 (59 FR 979), the NRC published in the Federal Register a proposed amendment to its regulation, 10 CFR part 50, "Domestic Licensing of Production and Utilization Facilities," to incorporate by reference the 1992 Edition with.the 1992 Addenda of Subsection IWE, and Subsection IWL, of Section*Xl, Division 1, of the ASME Code with specified modifications and a limitation. Five modifications were specified in the proposed rule to address two concerns of the* NRC. The first concern is that four recommendations for tendon

examinations that are included in Regulatory Guide 1.35, "Inservice Inspection of Ungrouted Tendons in Prestressed Concrete .Containments," Rev. 3, are notaddressed*in Subsection . IWL (this involves four of the modifications, (§ (D)). Regulatory Guide 1.35, Rev. 3, describes a basis acceptable to the NRC staff-for developing an.appropriate inservice inspection and surveillance program for ungrouted tendons in prestressed concrete containment.

structures. The four recommendations contained in Regulatory Guide 1.35,

Rev. 3, whic;h are not.addressed by Subsection IWL, provide positions on issues such as failed wires and tendon sheathing filler grease conditions. (The ASME Code has considered the four issues involved and is in.the process of adopting them into addenda of Subsection IWL). The second NRC concern ,is that if there is visible evidence of i:iegredation of the concrete (e.g., leaching, surface cracking) there may also be degradation of inaccessible areas. The fifth modification
(§ 50.55a(b)(2)(ix)(E))

requires that inaccessible areas be evaluated when visible conditions exist that suggest the possibility of degradation of these areas. The limitation which was included in the proposed rule specified the 1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL as the earliest version of the ASME Code the NRC finds acceptable. This is because this is the first edition including addenda combination acceptable to the NRC staff that incorporates the concept of base metal examinations and also provides a comprehensive set of rules for the examination of post-tensioning systems. As originally published in 1981, Subsection IWE preservice examination and inservice examination rules focused on the examination of welds. This based examination philosop,hy* was established in the 1970s as plants were being constructed. It was based on the premise that the welds in pressure vessels and piping w:ere the areas of greatest concern .. As containments have aged, degradation of base metal, rather than welds, has been found to be the issue of concern. The 1991 Addenda to the 1989 Edition, the 1992 Edition and the 1992 Addenda to Section XI, Subsection IWE, have promoted the . incorporation of base metal examinations. The proposed rulemaking incorporated a provision for an expedited examination schedule. This

expedited e?(amination schedule .is necessary to prevent the delay in implemen~ation of Subsection IWE and Subsection IWL (the Summary of Documen!ed E;valuation lists each *plant and the delay in implementation which w.ould be encountered if the subsections
were implemented through routine
updates of.the' ISi programs).

Provisions were incorporated in the proposed rule to ensure that the expedited examination which would be completed within 5 years from the effective date of the rule and the routine 120-month examinations did not duplicate examinations. On March 4, 1994, the NRC received a request from the Nuclear Management and Resources Council (which has since become part of the Nuclear Energy Institute (NEil) to extend the public comment period from March 23, 1994 until April 25, 1994, to enable NEI to "provide necessary and constructive comments on the proposed rule change." This was granted, and on March 28, 1994 (59 FR 14373), the NRC published in the Federal Register a notice of extension of the public comment period. Summary of Comments Comments were received* from 25 separate sources. These sourc;es . consisted of 15 utilities, one service organization (Entergy Operations, Inc.) representing five nuclear plants, the Nuclear Energy Institute (NEI), the Nuclear Utility Backfitting and Reform Group (NUBARG) represented by the finn of Winston & Strawn, one owner's group (BWR Owner's Group (BWROG)), one architect and engineering firm (Stone & Webster Engineering Corporation), one public citizens group (Ohio Citizens for Responsible Energy 50-SC-198 (OCRE)), three individuals, and one consulting finn (VSL Corporation). Comments received could be divided into three groups. The first group contains those comments which.address the administrative aspects of the rule (e.g., backfit considerations,

effectiveness of current containm*ent
examinations), and the modifications specified by the NRC in the proposed rule. The second and third groups contain those comments which address the.technical provisions of Subsection IWE, and Subsection IWL, respectively.

The summary and resolution of public comments and all of.the verbatim comments which were received (grouped by subject area) are contained. in the Summary of Documented Evaluation. The majority of comments generally addressed one of the following subject areas: (1) The incorporation by reference of Subsection IWE and Subsection 1WL into § 50.55a; (2) the development of guidance documents -instead of . regulatory requirements; (3) the rationale for the proposed backfit; (4) . endorsement of the BWROG comments; and (5) the 5-year expedited implementation. These subject areas * .encompass the comments submitted by NEI and NUBARG, and their comments, if any, are discussed separately in each subject area. The comments on subject area number one from those that approve of the incorporation by reference of Subsection IWE and Subsection IWL into § 50.55a, can be -summarized as follows: (1) There is a need for the periodic examination of containment structures to assure the containment's pressure-retaining and leak-tight

capability; (2) Section XI requirements define concise, technically sound programs to assure continuing containment integrity; and (3) input in the development of .these rules was provided by all interested parties involved in containment inservice inspection-users, regulators, manufacturers, engineering organizations, and enforcement organizations. . The comments on the other four subject areas are summarized below. The resolution of public comments contains.

all of the comments which were received. Some of the comments resulted in modifications to the rule, and some of the comments have been transmitted to the ASME for their consideration. A discussion of the . comments which led to modifications follows the summary of comments on subject area number *five. The resolution of public comments pack.age contains

those comments transmitted to the PART 50
STATEMENTS OF CONSIDERATION

-ASME. Those comments asked for interpretations of the ASME Code rules. Regarding subject area number twci, eleven commenters believe that .additional specific guidance in the form of a guidance document" would be more appropriate than a regulation. They concur with NEI that current regulatory requirements for containment integrity and examinations are already provided *by existing regulations (GDC 16 and 53 and Appendix J) ari.d licensee commitments. If more detail on how to perform containment examinations is needed, the commenters (including NEI) state that the details could be provided. in a regulatory guide, Information Notice, Generic Letter, or In an industry developed guidance document. The NRC does not believe that existing . regulations and licensee commitments are adequate. Existing regulations and licensee commitments have not proved to be adequate to detect the types cif problems which have been experienced in operating reactors. This-is evidenced by the large number of instances of

degradation that were found by the NRC through its inspections or audits of . plant structures, or by licensees because th~y were alerted to a degraded condition at another site .. Licensee containment inspection programs have generally not detected the types of degradation being reported (only four of th_e 32 reported instances of corrosion in Class MC containments were discovered as a result of the Appendix J general . inspection).

Further, the NRC does not believe that providing guidance through a regulatory guide or industry report would generally improve containment examination practices. Licensees were made aware of containment degradation through several industry notices, and yet the staff is still detecting many of occurrences of degradation. The increasing.rate of occurrence of containment degradation, the number of occurrences, the extent to which some containments were degraded.-the high number of instances discovered through NRC inspections or by licensees because they were alerted to a degradation condition at another site, the time-. dependent mechanisms, and-the results of the survey performed by the NRC Regional Offices regarding current containment inspections all point to the necessity of imposing additional requirements to ensure containments comply with design wall thicknesses and prestressing forces. This is a compliance backfit.

With regard to subject area number three, six general comments were received-from the Nuclear Utflfty Backfitting and Reform Group (NUB/\RG) and from the Nuclear Energy Institute (NEI) (which were endorsed by other commenters) regardi_ng the incorporation by reference of Subsection IWE and Subsection IWL which are similar in nature. The first comment is that the application of the compliance exception to this rulemaking is inappropriate, and that the proposed rule constitutes a backfit for which a cost-benefit analysis should be performed.

The NRC agrees that the rulemaking is a backfit. However, as discussed under the Backfit Statement,. the NRC believes that the compliance exception to the backfit rule is appropriate. The second comment was a citation of a paragraph from the Statement of Considerations to the 1985 final backfit rule which addressed the compliance exception. That paragraph addressed "Section 50.109(a)(4) which creates exceptic;ms for modifications necessary to bring a facility into compliance or to ensure through immediately effective regulatory action that a licensee meets a standard of no undue risk to public health and safety."*Both NEI and NUBARG assert that the proposed rule is a new interpretation of how to demonstrate compliance with*existing standards and therefore constitutes a backfit under 10 CFR 50.109(a)(l). The NRC does not believe that the use of the compliance exception must be confined only to the situation addressed in the Statement of Consideration to the 1985 final backfit rule-"omission or mistake of fad." In any event, the current unsatisfactory status of containment inservice inspections can be characterized fairly as, in retrospect, a mistake about and omission from the necessary elements of a satisfactory inspection program. The third comment is that containments must experience corrosion or degradation that is so unanticipated and excessive so as to constitute a genuine compliance concern. Another commenter expressed the idea somewhat differently believing that a broad-based concern with the operability of containment-structures through the industry must be demonstrated to be a compliance issue. The NRC agrees with those criteria and concludes, in fact, that there is a based concern regarding the structural integrity of containment structures. The NRC's approach focuses on two questions: (1) Is the corrosion such that there is a basis for reasonably concluding that additional instances of noncompliance with the relevant GDCs, Appendix J, and/or licensee commitments at numerous plants; and (2) whether there is a basis for reasonably believing that the corrosion 50-SC-199 would have been identified and properly addressed by the licensees in the absence of additional regulatory requirements._Based on the: (1) Number of occurrences of containment degradation; (2) increasing rate of containment degradation; (3) locations of the degradation; ( 4) two instances where containment wall thicknesses were below minimum design wall thickness; (5) number of corrosion paths which have been reported; and (6) higher than anticipated corrosion rates in many of the occurrences, the NRC believes that containments are experiencing corrosion or degradation that_is unanticipated and excessive. Further, based upon factors (1) to (6) above, the NRC concludes that additional criteria are necessary to ensure that compliance with existing requiremepts for minimum accepted design wall thicknesses and prestressing .forces are maintained (and thereby the ability of the containment to continue to . perform its intended safety function). The fourth comment by NUBARG and NEI suggested that it is part of the anticipated process for the industry to rely upon NRC inspections and audits to identify problems and then alert the industry through NRC documents such as information notices and generic letters. During the presentation to the ACRS*on February 10, 1995, NEI asserted that "[i)t really doesn't matter how the utilities identify these instances of degradation." The NRC believes that inspections conducted by licensees should be adequate to ensure that containment degradation is identified without reliance upon NRC inspections. The fifth-NEI and NUBARG comment is that to ensure compliance the NRC could take individual enforcement action rather than endorse ASME standards. The NRC believes that the best *approach is to adopt the industry consensus standards (i.e., endorse ASME Section XI Subsection !WE and Subsection IWL). Containment corrosion and degradation have been reported since 1986. The patterns of d'egradation and the corrective actions were not immediately obvious. Given the number and the extent of the occurrences, and the variability among plants with regard to the performance and the effectiveness of containment inspections, the NRC believes that the best course of action is to endorse IS! requirements to ensure that containments comply with desi_gn wall thicknesses and prestressing forces. The sixth comment is that GOC 16 required containments to be designed and constructed with an allowance for corrosion or degradation of the containment wall over the projected PART 50

STATEMENTS OF CONSIDERATION design life of the plant. NEI and NUBARG assert that "[i]t is therefore hardly surprising that, as noted in the Statement of Considerations, '[o]ver one-third of the containments have experienced corrosion or other degradation.' " Therefore, they believe there is not a broad-based concern with operability of containment structures.

The NRC rejects the argument that because containments have corrosion allowances and corrosion was expected to occur that, ipso facto, further inspections are not necessary and the compliance exception is inappropriate. As previously pointed out, in many cases, the corrosion rate has been found to be greater than.that for which the containment was designed (in some cases the rate was twice that predicted). . Some of the more extreme cases of wall thinning occurred in plants with corrosion allowances. The existence of a corrosion allowance at any given plant is, of course relevant, but only in the context of determining whether a . relevant requirement or commitment is likely to be violated during the OL term. A corrosion allowance simply increases the tolerance (time period) for corrosion. However, once the allowance is ercided, then concern with compliance.becomes relevant. Based upon the staffs finding of the number and extent of corrosion to date, and the lack of activities to manage the degradation by many licensees, the NRC concludes that it is likely that those licensees will be in violation of applicable requirements for containment structural integrity and leak-tightp.ess during the OL term, absent the imposition of Subsections IWE and IWL. Because licensees have been unable to ensure compliance with current regulatory requirements, the NRC believes that more specific ISi requirements, which expand upon existing requirements for the examination of containment structures in accordance with GDC 16, 53, Appendix A to 10 CFR part 50; and Appendix J to 10 CFR part 50, are . needed and are justified for the purpose of ensuring that containments continue to maintain or exceed minirnwn accepted design wall thicknesses and prestressing forces as. provided for in industry standards used to design containments (e.g., Section ill and Section VIII of the AS.ME Code, and the American Concrete Institute Standard ACI-318), as reflected in license conditions, technical specifications, and written licensee commitments (e.g., the Final Safety Analysis Report). The NRC believes that the occurrences of corrosion and other degradation would have been detected by licensees when conducting the periodic examinations set forth in Subsection IWE and

Subsection IWL. With regard to subject area number four, six commenters believe that the . Boiling Water Reattors Owner's Group (BWROG) containment fospection plan (CIP) will adequately address examinations for the primary containment when used in conjunction with other existing examination requirements such as Appendix J. The staff does not believe that the CIP is a comprehensive containment examination program. In the CIP, there is a comparison between the CIP and Subsection IWE. The CIP dismisses seven of the eighteen identified . Subsection IWE examinations as not being justifiable even_ though some of these areas are likely to experience
accelerated corrosion.

The*CIP enumerates the conservatisms and margins against failure in the design of Mark I and II containments and concludes that in a typical plant probabilistic*r!sk 11,ssess.ment of failure, the contribution to failure of the containment steel structure is negligible. The NRC believes that the conservatisms and margins referred to are not additional tolerances which . allow areas of containments to go unexamined. These conservatisms and margins were required allowances in the design because of the uncertainties in loadings, in material properties, in analysis*, and in the variation of steel thicknesses. Examination of large areas of the containment cannot be dismissed as being non-critical based on* conservatisms and margins when corrosion has clearly eroded the margin of safety in some cases. In addition,* given that o.1ly four of the 32 occurrences of corrosion in metal containments and the liners of concrete containments were detected during the pre-integrated leakage rate test examination, the NRC does not believe that the CIP used in conjunction with other existing examination requirements such as Appendix J will adequately address examinations for the primary containment as asserted. The industry initiative that allows a decrease in the frequency of Appendix J leakage rate

testing further erodes confidence in the acceptability of the BWROG approach.

Comments were received from.ten sources on proposed § 50.55a(g)(6)(ii)(B) which would require a 5-year expedited examination. schedule (subject area number five). Most of these comments asked for clarifications of the NRC staff intent of this provision. Some commenters interpreted this provision as a requirement to perform all of the examinations specified for a 10-year * . 50-SC-200 interval in 5 years, which was not the intent. § 50.55a(g)(6)(H)(B) has been changed to clarify that for Subsection IWE, the baseline inspection will be the inservice examinations which are. to be performed during the first period of the. . first interval. For Subsection IWL, the*. baseline inspection will be the required inservice*examinations which *

correspond to the year of operation for
each unit. The result of the clarification is that § 50.55a(g)(6)(ii)(B)(1) addresses Subsection IWE and . . § 50.55a(g)(6)(ii)(B)(2) addresses Subsection IWL. § 50.55a(g)(6)(ii)(B)(2) in the proposed rule has become § 50.55a(g)(6)(ii)(B)(3) and § 50.55a(g)(6)(ii)(B)(3).has become. § 50.55a(g)(6)(ii)(B)( 4) 'in the final rule. There was one additional comment submitted by NEI. The proposed rule discussed NEl's (then NUMARC)
position on the role of Subsection lWE and Subsection IWL in license renewal. Subsections IWE and IWL were referenced many times as one acceptable*

approach for managing age~ related degradation. The plan for managirig age-related degradation assumes that these examinations are '.'in current and effective use.NEI commented on the above statements in , the proposed rule;_.Although the BWR and PWR containment IRs [Industry .

Reports) do reference Subsections IWE and IWL, their identification in the IRs should not be misrepresented to imply that Subsections IWE and .IWL. are being implemented or that they are requi~d for operating plants.during thefrinitial licensing term." The NRC agrees that the IRs were nofto be represented as a requirement for operating licensees to . implement Subsection IWE and Subsection IWL or their equivalent, and that these subsections were referenced as one acceptable approach of managing
age-related degradation for the license renewal period. However, present licensee containment examination programs have nor proved to be effective in detecting the. types of degradation
which have been reported.

The number of occurrences and the extent of

degradation (which includes cases of noncompliance) leads to the conclusion . that additional requirements are needed for managing containment degradation during the operating term. Because Subsections IWE and IWL were developed by the ASME with industry input and found to be acceptable by NEI. for managing age-relate.d degradation for the license renewal period, the NRC believes that adoption of those programs at this tim,e is the best approach.

The NRC also believes that with implementation of Subsections !WE.and PART 50

STATEMENTS OF CONSIDERATION IWL, the detrimental effects of containment aging will be managed during the current operating term, as well as during the license renewal term. As a result of the comments received, there is one editorial change,. two clarifications, and four modifications in the final rule. With respect to the editorial change, a commenter suggested that the wording of § 50.55a(b)(2)(ix)(D)(2) in the proposed rule be revised to be consistent with . § 50.55a(b)(2)(ix)(D)(1) and § 50.55a(b)(2)(ix)(D)(3) of the same paragraph.

§ 50.55a(b)(2)(ix)(D) addresses the sampling of the grease contained in post-tensioning systems, and conditions, which if found, are reportable. The suggested wording has .

been adopted in the final rule.
One of the clarifications was to . proposed § 50,55(g)(6)(ii)(B).

This . change was discussed previously in subject area number five. § 50.55a(g)(6)(ii)(B)(1) and § 50.55a(g)(6)(ii)(B)(2) require that licensees conduct the first containment examinations in accordance with Subsection IWE and Subsection IWL (1992 Edition with the 1992 Addenda),

modified by § 50.55a(b)(2)(ix) and § 50.55a(b)(2)(x) within 5 years of the effective date of the final rule. This expedited.examination schedule is necessary to prevent possible delays in the implementation of Subsection IWE by as much as 20 years and Subsection IWL by as much as 15 years. Subsection IWE, Table IWE-2500-1, permits the . deferral of many of the required examinations until the end* of the 10-year inspection interval.

Adding the 10 years that could pass before some utilities are required to update their ISI plans, a period of20 years could pass before the first examinations would take place. Subsection IWL is based on a 5-year inspection interval. Adding the possible 10 years before update of existing ISI plans, a period of 15 years could pass before the examinations were performed by plants that have not voluntarily adopted the provisions of Regulatory Guide 1.35, Rev. 3. Expediting implementation of the containment examinations is considered

necessary because of the problems that have been identified at various plants, the need.to establish expeditiously a baseline for each facility, and the need to identify any existing degradation.

Paragraphs (g)(6)(ii)(B)(3J and (g)(6)(ii)(B)(4) each provide a mechanism for licensees to satisfy the

requirell!ents of the routine containment examinations and the expedited examination without duplication.

Paragraph (g)(6)(ii)(B)(3) permits licensees to _avoid duplicating examinations required by both the periodic routine and expedited examination programs. This provision is intended to be useful to those licensees that would be required to implement the expedited examination during the first periodic interval that routine containment examinations are required. Paragraph (g)(6)(ii)(B)(4) allows licensees to use a recently performed examination of the post-tensioning system to satisfy the requirements for the expedited examination of the containment post-tensioning system. This situation would occur for licensees who perform an examination of the post-tensioning system using Regulatory Guide 1.35 between the effective date of this rule and the beginning of the expedited examination. The four modifications are: (1) § 50.55a(b)(2)(x)(A) expands the evaluation of inaccessible areas of concrete containments (Class CC) to metal containments and the liners of concrete containments (Class MC); (2) § 50.55a(b)(2)(x)(B) permits alternative lighting and resolution requirements for

remote visual examination of the containment; (3) § 50.55a(b)(2)(x)(C) makes the examination of pressure retaining welds and pressure retaining dissimilar metal welds optional; and (4) § 50.55a(b)(2)(x)(D) has been added to provide an alternative sampling plan.* Section 50.55a(b)(2)(x)(E), a clarific.ation, more clea,rly defines the frequency of the Subsection IWE general visual examination.

The first modification, § 50.551i(b)(2l(x)(A), which expands the

evaluation of inaccessible areas* of concrete containments (Class CC) to . metal-containments and the liners of concrete containments (Class MC), was the result of-a comment received on § 50.55a(b)(2)(ix)(E) of the proposed rule. The commenter believed that given the number of occurrences of corrosion*

in Class MC containments, the proposed provision (which only addressed concrete containments) should be expanded in the final rule to include metal containments and the liners of concrete containments. The second modification, § 50.55a(b)(2)(x)(B), was added to the final rule to permit alternative lighting and resolution requirements for remote visual examination of the containment. Subsection IWE references the lighting and resolution requirements contained in 1W A-2200. The lighting and resolution requirements contained in 1W A-2200 would on a practical basis preclude remote containment examination. The third modification, § 50.55a(b)(2)(x)(C), makes the 50-SC-201 examinations of Subsection IWE, Examination Category E-B (pre~sure retaining welds) and Subsection IWE, Examination Category E-F (pressure retaining dissimilar metal welds) optional. The NRC siaff concludes that requiring these examinations is not appropriate. There is no evidence of problems associated with welds of this

type under the given operating conditions.

In addition, the occupational radiation exposure that would be incurred while performing these examinations cannot be justified. It is estimated that the total occupational exposure that would be incurred yearly.in the performance of the containment weld examinations in accordance with Examination Categories E-B and E-F would be 440 person-rems. The fourth modification, § 50.55a(b)(2)(x)(D), provides an alternative to the ASME Section XI requirements for "additional examinations" (note: additional examinations*~ are required during the same outage when acceptance criteria are exceeded). The alternative would allow. licensees to determine the number of additional components to be examined based on an evaluation to determine the extent and nature of the degradation. Five commenters believe that the requirements for additional examinations used in other subsections of Section XI is inappropriate for containment components. Additional examinations are incorporated into Section XI to determine the extent to which degradation found in one component exists in other similar. components. In some instances, a large number of additional examinations could be required. The commenters believe that a review of the operational history of containment components shows that the degradation is limited to the area in question and is not widespread. This makes the Section XI requirements for additional examinations burdensome and inappropriate for application to containments. The NRC agrees and revised the rule to permit the alternative to the Section XI requirements for . additional examinations. The NRC believes that these modifications improve the final rule and will improve the containment inspection program as set forth by Subsection IWE and Subsection IWL. Some of the public comments cited failure data which have been accumulated in recent years in support of various NRC staff activities and industry initiatives. Most of this data has been accumulated since the ASME committees developed these. subsections. Without the benefit of this PART 50

STATEMENTS OF CONSIDERATION recently accumulated operational data, the ASME committees respor..sible for developing Subsection IWE and Subsection IWL modelled those subsections on other subsections of Section XI and the experience gained from application of those other subsections.

With the additional insights drawn from analysis of this new data, it is apparent that many aspects of containments are unique compared to components of other systems. Some of the containment components which were expected to experience degradation, based on experience with other systems, have proved not to be susceptible to the same type of degradation. The ASME working groups are considering these issues. However, based on initial committee discussion, it is anticipated that similar changes will be made to Subsection IWE and Subsection IWL, but the length of the ASME consensus process precludes the possibility of the changes being adopted into the ASME Code in the near term. Hence, the NRC has determined to adopt the 1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL with the modifications which were previously discussed. Other Provision~ Contained in the Final Rule The following paragraph was contained in the proposed rule and has not been discussed previously. This paragraph received comments which resulted in the provision being dropped in the final rule. Section 50.55a(b)(2)(x) was a provision in the proposed rule intended to pr.ovide licensees with a mechanism to merge the Subsection IWE and Subsection IWL ISI program

with their routine 120-month ISi program. Those licensees who were near the end of their present 10-year ISi interval when the final rule becomes effective would have been given an additional 2 years to submit their containment ISi program. Several commenters responded that due to the time constraints of having to develop the containment ISi program and then perform the required examinations within 5 years, the additional 2 years could not be utilized.

Therefore, § 50.55a(b)(2)(x) as it appeared in the proposed rule has been deleted, and § 50.55a(b)(2)(x) in the final rule contains the modifications which were added as a result of public comment on the proposed rule. The provisions in this paragraph and the following four paragraphs were contained in the proposed rula and have not changed due to comments. Section 50.55a(b)(2)(vi) incorporates a limitation specifying the 1992 Edition with 1992

Addenda of Subsection IWE and Subsection IWL as *the earliest ASME Code version the NRC finds acceptable.

This edition and addenda incorporate the concept ofbas.e*metal examinations and also provide a comprehensive set of niles for the examination of post-* tensioning systems. It should be noted that the wording of t,his provision has been changed in the final rule In order to make it consistent with other

provisions in § 50.55a(b).

Section 50.55a(b)(2)(ix) specifies five

modifications that must be implemented when using Subsection IWL. Four of these issues are identified in Regulatory Guide 1.35, Revision 3, but are not currently addressed in Subsection IWL. Section 50.55a(b)(2)(ix)(A) .requires that . grease caps which are accessible must be visually examined to detectgrease leakage or grease cap deformation.

Section 50.55a(b)(2)(ix)(B) requires the preparation of an Engineering Evaluation Report when consecutive surveillances indicate a trend of prestress loss to below the minimum prestress requirements. Section 50.55a(b)(2)(ix)(C) requires an evaluation to be performed for instances of wire .failure and slip of wires in anchorages. Section 50.55a(b)(2)(ix)(D) addresses sampled sheathing filler

grease and reportable conditions.

A comment was received on this provision which resulted in an editorial change (this was discussed on page 12). Section 50.55a(b)(2)(ix)(E) requires that licensees evaluate the acceptability of inaccessible areas of concrete containments when conditions exist in accessible areas that suggest the possibility of degradation in inaccessible areas. Existing§ 50.55a(g), "Inservice inspection requirements,'.' specifies the requirements for preservice and inservice examinations for Class 1 (Class 1 refers,to components of the reactor coolant pressure boundary), Class 2 (Class 2 quality standards are applied to water-and steam-containing pressure vessels, heat exchangers (other than turbines and condensers), storage tanks, piping, pumps, and valves that are part* of the reactor coolant pressure boundary* (e.g., systems designed for residual heat removal and eme~ency core cooling)), and Class 3 (Class 3 quality standards are applied to containing pressure. vessels, heat exchangers (other than turbines and condensers), storage tanks, piping, pumps, and valves (not part of the reactor coolant pressure boundary)) components and their supports. Subsection IWE (Class MC-metal containments) and Subsection IWL (Class CC-concrete containments) are 50-SC-202

incorporated by reference into the NRC regulations for the first time. Section 50.55a(g)(4) specifies the containment coinponents to which the ASME Code Class MC and Class CC inservice inspection classifications incorporated by reference in this rule will apply. Section 50.55a (g)(4)(v)(A), (v)(B), and. (v)(C) specify the Subsection IWE and Subsection IWL rules for inservice inspection, repair, and replacement of
metal and concrete containments.

This is consistent with the long-standing intent and ongoing application by NRC and licensees to utilize the rules of Section XI when performing inservice inspection, repairs; and replacements of applicable components and their

supports.

Small Business Regulatory Enforcement Fairness Act *

In accordance with the Small .
Business Regulatory Enforcement Fairness Act of 1996,the NRC has
determined that this action is not a major rule and has verified this determination with the Office of Information and Regulatory Affairs of 0MB. . .. Finding of No Significant Environmental Impact The Gommission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of.10 CFR part 51, that this rule is not a major Federal action that significantly affects the quality of the human environment and therefore an environmental impact statement is not required.

This final rule is one part of a regulatory framework directed to* . ensuring containment integrity. Therefore, in tho general sense, this rule will have a positive .impact on the environment. This rule incorporates by reference into the NRG regulations requirements contained in *the ASME Code for the inservice inspection of the containments of nuclear power plants. The performance of containment

examinations, as*set forth by the provisions of this final rule, for PWRs, Ice.Condensers, and BWR Mark Ils .and Ills is not expected to result in significant occupational l'adiation . exposure (1.0 person-rems per year or 0.04 person-rems per unit averaged over 27 examinations each year). The above categories of plants, for which the occupational radiation exposure is . . insignificant, represent the vast majority of units (89). For BWR Mark I containments, the estimated occupational radiation exposure which

. ; ,, ', ",' PART 50

STATEMENTS OF CONSIDERATION would be incurred, per year while performing BWR Mark I containment examination is 29.4 person-rems per .year or 4.2 person-rems per unit averaged over 7 examinations per year. However, the estimated occupational radiation exposure per unit does-not
provide an accurate representation of the actual radiological exposure that would be incurred by any one individual.

10 CFR 20.101, "Radiation dose standards for individuals in restricted areas" only permits a whole body dose of 1.25 rem per calendar quarter. As a practical matter, licensees carefully manage the exposure incurred by any one individual by practicing and applying "as low as reasonably achievable" (ALARA) principles to protect the health and safety of personnel. In the performance of the examination of BWR Mark I containments, this is accomplished'liy having several individuals perform* the examinations to "spread out" the exposure. In this manner, no one

individual will suffer any significant health effects. It also must be kept in mind that-these containment examinations are scheduled to occur at the interval of once every 31/J years. This provides licensees ample time for* . planning.

the examinations, and scheduling personnel in accord with ALARA considerations. Therefore, the

occupational radiation exposure is insigntficant given the relatively low . exposure on a unit basis and the licensees' programs for controlling the impact of exposure for any one individual.
. Actions required of applicants and licensees to implement containment examinations are of the same nature that applicants and licensees have been performing for many years in other Section XI ISi programs.

Extension of .these actions to additional components, therefore, should not increase the potential for a negative environm11ntal impact. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW .. (Lower Level), Washington, DC. Single copies of the environmental assessment ap.d the finding of no significant impact are available from Mr. W. E. Norris, Division of Engineering Technology, Office of Nuclear

Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6796.

Pape"".ork Reduction ActStatement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Manageinent and Budget, approval number 3150--0011. The public reporting burden for this collection of information is estimated to average 4,000 hours per response for development of an initial inservice inspection plan;-and 8,000 hours per response for the update of the plan and periodic examinations, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. The estimate of 8,000.hours for plan update and performing periodic examinations is a 2,000 hour reduction from the estimate given in the proposed rulemaking. This reduction results from changes made in response to public comment; A number of examinations have been modified or made optional greatly reducing the effort required to comply with the requirements contained in the final rule. Send comments on any aspect of this collection of information, including suggestions for reducing the burden, to the Information*and Records

Management Branch (T-6 F33), IJ;S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail at
BJSl@NRC.GOV; and to the Desk Officer, Office of Information and . Regulatory Affairs, NEOB-10202, . (3150-0011), Office of Management and Budget, Washington, DC 20503. Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid 0MB control number. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission hereby certifies that . this rule will not have a significant economic impact on a substantial number of small entities.

This rule affects only the operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entities" set* forth in the* Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Adminii;tration at 13 CFR part 121. Since these companies are dominant in their service areas, this rule does not fall within the purview of the Act. 50-SC-203 Backfit Statement The NRC is amending its regulations . to incorporate by reference the 1992 *Edition with the 1992 Addenda of . Subsection IWE and Subsection IWL to assure that the critical areas of containments are routinely inspected to detect defects that could compromise a containment's structural integrity. Based on a preponderance of reliable information, the NRC concludes that this rule is a compliance. backfit, and therefore a backfit.analysis is not required pursuant to 10 CFR 50.109(a)(4)(i). A summary of. noncompliance is set forth below. The documented evaluation required by § 50.109(a)(4) to support this conclusion is available for inspection in the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the analysis may be

obtained from Mr. W.E. Norris, Division of Engineering Technology.

Office of Nuclear Regulatory Research*, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6796,. . . The rate of occurrence cif corrosion and degradation of containment structures has been increasing at operating nuclear power plants. There have been 32 reported occurrences of corrosion in metal containments

and the liners of concrete containments.

This is approximately one-fourth of all operating nuclear power plants. Only four of the 32 occurrences were detected by current licensee containment inspection programs. Nine of these occurrences were first identified by the NRC through its inspections or , structural audits. Eleven occurrences were detected by licensees after they were alerted to a degraded condition at . another site* or through activity other than containment inspection. There

have been 34 reported occurrences of degradation of the concrete or of the* post-tensioning systems of concrete containments.

This is nearly one-half of these types of containments. It is clear that current licensee containment inspection programs have not.proved to be adequate to detect the types of . degradation which have been reported. Examples of degradation not found by licensees, but initially detected at plants . through NRC inspections include: (1) Corrosion of steel containment shells in the drywall sand cushion region,

resulting in wall thickness reduction to below the minimum design thickness; (2) corrosion of the torus of the steel containment shell (wall thickness below minimum design thickness);

(3) extensive

corrosion of the liner of a concrete containment with local PART 50
STATEMENTS OF CONSIDERATION degradation at many locations to approximately half-depth; (4) grease leakage from the tendons of prestressed concrete containments; and (5) leaching as well as excessive.cracking in concrete containments.

None of the existing requirements for containment inspection provide specific

guidance on how to perform the necessary containment examinations.

This.lack of guidance has resulted in a large variation with regard to the performance and the effectiveness of licensee containment examination programs. Based on the results of inspections and audits, and plant operational experiences, it is clear that many licensee containment examination programs have not detected degradation that could result in a compromise of . pressure-retaining capability. Most of those occurrences were first identified by ~e NRC through its 'inspections or audits of plant structures, or by licensees while performing an unrelated activity or, after they were fllerted to a degraded condition at another site. In analyzing the reported containment degradation, it is apparent that all containments are subject to certain type(s) of degradation depending on the design. Information gathered by the staff indicates that many licensees still have not reacted to this serious safety concern and have not initiated comprehensive containment inservice inspection. As a result of the rate of occurrence of containment degradation, and the extent of containment degradation, the NRC believes that there is a basis for reasonably concluding that such degradation is widespread and affects virtually all plants. Because of the serious degradation which has occurred, the belief that additional occurrences of noncompliance with required minimum wall thicknesses and prestressing forces will be reported, and the high likelihood that some of those occurrences could result in loss of structural integrity and leak-tightness, the NRC has determined that imposition of these containment inservice inspection requirements under the compliance exception to 10 CFR 50.109(a)( 4)(i) is appropriate. The NRC believes that the final action would also result in a substantial safety increase and that the direct and indirect costs of implementation are justified in view of the significant safety benefit to be gained. The NRC believes that the inspections contained in Subsections IWE and IWL will improve significantly the ability to detect degradation and take timely action to correct degradation of containment structures. A review of early implementation of the maintenance rule (10 CFR 50.65) at nine nuclear power plants, which is documented in NUREG-1526, indicates that most licensees assigned a low priority to the monitoring of structures. Several licensees incorrectly assumed that many of their 'structures are inherently reliable. This is true so long as there is no degradation. However, the degradation of structures can reduce high margins of safety to a low or negligible margin of safety. As discussed earlier, such substantial containment degradations have been detected at a large number of nuclear power plants, and their detection to date can best be characterized as happenstance. The final rule will provide for improved periodic examination of containment structures assuring that the critical areas of containment are periodically inspected to detect and take corrective action for. defects that could compromise the containment's pressure-. retaining and leak-tight capability. The NRC believes, therefore, that tho final action can be justified es a cost-justified safety enhancement backfit, as well as a compliance backfit. List of Subjects in 10 CFR Part 50. Antitrust, Classified information, Criminal Penalties, Fire protection, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation . protection, Reactor siting criteria, .Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 533, the NRC is..adopting the following amendments to 10 CFR part 50. 61 FR 65157 Published 12/11/96 Effective 1/10/97 Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants See Part 100 Statements of Consideration 50-SC-204 UNITED STATES NUCLEAR REGULATORY COMMISSION RULES and REGULATIONS TITLE 10, CHAPTER 1, CODE OF FEDERAL REGULATIONS-ENERGY fPAiffi L_!!__J ENVIRONMENTAL PRQTECTION REGULATIONS FOR DOMESTIC LICENSING AND RELATED REGULATORY FUNCTIONS 52 FR 8225 Published 3/17/87 Effective 7/14/87 Licenses and Radiation Safety Requirements for Well Logging See Part 39 Statements of Consideration 52 FR 19303 Published 5/22/87 Domestic Licensing of Production and Utilization Facilities; Communications Procedures Amendments; Correction See Part 50 Statements of Consideration

i52 FR 31601 'published 8/21/87 Effective 8/19/87 . Statement of Organization and General Information
See Part 1 Statements of Consideration 53 FR 13399 Published 4/25/88 : Effective 4/25/88 10 CFR Part 51 Revision of Telephone Numbers for : Environmental Inquiries AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule .. su~MARY: The Nuclear Regulatory Commission [NRC) is emending its regulations pertaining*to environmental matters to indicate the revision of five telephone numbers that enable prospective applicants or petitioners to consult with members of the NRC's STATEMENTS OF CONSIDERATION staff. These amendments are required because of the assignment of new telephone numbers fa conjunction with the recent consolida lion of approximately one-half of the NRC's headquarters staff to its new location in . Rockville, Maryland. These amendments are being made to inform NRC licensees and members of the public of the new telephone numbers. -EFFECTIVE DATE: April 25, 1~88. FOR FURTHER INFORMATION CONTACT: Donnie H. Grimsley, Director, Division

of Rules and Records, Office of .Administration and Resources Management, U.S. Nuclear Regulatory
Commission, Washington*.

DC 20555, * . Telephone: 301-492-7211. SUPPLEMENTARY INFORMATION: On March 28, 1988, the Office of the

Executive Director for Operations and portions of the Office of Governmental anp Public Affairs (GPA}-the Director of GPA and the Public Affairs staff-; relocated at the agency's new office
building in Rockville, Maryland.

A notice to that effect was published in the Federal Register on March 31, 1988 (53 ** FR 10449). These amendments reflect the a111;ignment of new telephone numbr.rs

to certain relocated agency personnel

.. Because these amendments deal solely with the organization and relocation of agency personnel, the notice and comment provisions of the Administrative Procedure Act d6 not apply. under 5 U.S.C. 553(b){A). These amendments.are effective

upon ' publication in the the Federal Register.

'. Good cause exists to dispense with the usual 30-day delay in the effective date, because these amendments are of a minor and a"dministrative nature. Environmental Impact: Categorical Exclusion The NRC has determined that this final rule is the type of action described 'in categorical exclusion 10 CFR . *51.2Z(c)(2). Therefore, neither an environmental impact statement nor an *. environmental assessment has been prepared for this final rule. Paperwork Reduction Act Statement This final rule contains no information collection requirements and therefore is 51-SC-1 not subject to the requirements.of the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). List of Subjects in.10 CFR Part 51 Administrative practice-and -_procedure, Environmental impact statement, Nuclear materials, Nuclear power-plants and reactors, Reporting and recordkeeping requirements. For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is adopting the following amendments to .10 CFR Part 51. 53 FR 24018 Published 6/27/88 Effective 7/27/88 General Requirements for Decommis_sioning Nuclear Facilities See Part 30. Statements of Consideration 53 FR 31651 Published 8/19/88 EffecUve 9/19/88 Ucensing Requirements for the Independent Storage of SpentNuclear Fuel and High-Level Radioactive Waste See Part 72 Statements of Consideration 53 FR 43419 Published 10/27 /88 EffecUve 10/27 /88 Relocation of NRC's Public Document Room; Other Minor Nomenclature Changes See Part 1 Statements of Consideration 54 FR 15372 .Published 4/18/89 Effective 5/18/89 Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Reactors See Part 52 Statements of Consideration 54 FR27864 Published 7 /3/89

Effective 8/2/89 NEPA Review Procedures for Geologic *Repositories for High-Level Waste See Part 60 Statements of Consideration PART 51 STATEMENTS OF CONSIDERATION 51-SC-2 54 FR39767 Published 9/28/89 Comment period expires 12/27 /89. 10 CFR Part 51 Waste Confidence Decision Review AGENCY: Nuclear Regulatory Commission.

ACTION: Review and proposed revision of waste confidence decision.

SUMMARY

On August 31, 1984, the Nuclear Regulatory Commission (NRC) issued a final decision on what has come to be known as its "Waste Confidence Proceeding." The purpose of that proceeding was "* *

to assess generically the degree of assurance now available that radioactive waste can be safely disposed of, to determine when such disposal or offsite storage will be available and to determine whether radioactive waste can be safely stored onsite past the expiration of existing facility licenses until offsite disposal or storage is available." (49 FR 34658). The purpose of this notice is to present for public comment the proposed findings of a Commission review of that Decision.

The Commission noted in 1984 that its Waste Confidence Decision was unavoidably in the nature of a prediction, and committed to review its conclusions "* *

should significant and pertinent unexpected events occur or at least every five years until a repository is available." The Commission has reviewed its five findings and the rationale for them in light of developments since 1984. This proposed revised waste Confidence Decision supplements those 1984 PART 51 STATEMENTS OF CONSIDERATION findings and the environmental analysis supporting them. The Commission proposes that the second and fourth findings in the Waste Confidence Decision be revised as follows: Finding 2: The Commission finds reasonable assurance that at least one mined geologic repository will be available within the first quarter of the twenty-first century, and that sufficient repository capacity will be available within 30 years beyond the !incensed life for operation of any reactor to dispose of the commercial high-level radioactive waste and spent fuel originating in such reactor and generated up to that time. Finding 4: The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the licensed life for operation (which may include the term of a revised license] of that reactor at its spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations.

The Commission proposes to reaffirm the remaining findings. Each finding, any proposed revisions, and the reasons for revising or reaffirming them are set forth in the body of the review below. The Commission also issued two companion rulemaking amendments at the time it issued the 1984 Waste Confidence Decision. The Commission's reactor licensing rule, 10 CFR part 50, was amended to require each licensed reactor operator to submit, no later than five years before expiration of the operating license, plans for managing spent fuel at the reactor site until the spent fuel is transferred to the Department of Energy (DOE) for disposal under the Nu.clear Waste Policy Act of 1982 (NWPA]. 10 CFR part 51, the rule defining NRC's responsibilities under the National Environmental Policy Act (NEPA), was amended to provide that, in connection with the issuance or amendment of a reactor operating license or initial license for an independent spent fuel storage installation, no discussion of any environmental impact of spent fuel storage is required for the period following expiration of the license or amendment applied for. In keeping with the proposed revised Findings 2 and 4, the Commission is providing elsewhere in this issue of the Federal Register proposed conforming amendments to its 10 CFR part 51 rule providing procedures of considering in licensing proceedings the environmental effects of extended onsite storage of spent fuel. Finally, the Commission proposes to extend the cycle of its Waste Confidence reviews from every five years to every ten until a repository becomes available. In its 1984 Decision, the Commisson said that because its conclusions were "*' *

unavoidably in the nature of a prediction," it would review them "* *
should significant and unexpected events occur, or at least every five years until a repository
*
is available." As noted below, the Commission now believes that predictions of repository availability are best expressed in terms of decades rather than years. To specify a year for the expected availability of a repository decades hence would misleadingly imply a degree of precision now unattainable.

Accordingly, the Commission proposes to change its original commitment in order to review its Waste Confidence Decision at least every ten years. This would not, however, disturb the Commission's original commitment to review its Decision whenever significant and pertinent unexpected events occur. DATES: The comment period expires December 27, 1989. Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except to comments received on or before this date. ADDRESSES: Mail written comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Service Branch. Deliver comments to One White Flint North, 11555 Rockville Pike, Rockville, MD betweeen 7:30 a.m. and 4:15 p.m. weekdays. FOR FURTHER INFORMATION CONTACT: Rob MacDougall, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (202) 492-3401; or John Roberts, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (202) 492--0608. SUPPLEMENTARY INFORMATION: Background In November 1976, the Natural Resources Defense Council (NRDCJ petitioned NRC for a rulemaking to determine whether radioactive wastes generated in nuclear power reactor can be subsequently disposed of without undue risk to the public health and safety. The NRDC also requested that NRC not grant pending or furture requests for operating licenses until the petitioned finding of safety was made. 51-SC-3 On June 27, 1977, NRC denied the NRDC petition. The Commission said that in issuing operating licenses, NRC must have assurance that wastes can be safely handled and stored as they are generated. It also said that it is not necessary for permanent disposal to be available if NRC could be confident that permanent disposal could be accomplished when necessary. NRC added that Congress was aware of the relationship between nuclear reactor operations and the radioactive waste disposal problem, and that NRC would not refrain from issuing reactor operating licenses until the disposal problem was resolved. The Commission also stated that it"* *

would not continue to license reactors if it did not have reasonable confidence that the wastes can and will in due course be disposed of safely." Also in November 1976, two utility companies requested amendments to their operating licenses to permit expansion in the capacity of this spent nuclear fuel storage pools: Vermont Yankee Nuclear Power Corporation for the Vermont Yankee plant; and Northern States Power Company for its Prairie Island facility.

In both cases, the utilities planned to increase storage capacity through closer spacing of spent fuel assemblies in existing spent fuel pools. The New England Coalition on Nuclear Power and the Minnesota Pollution Control Agency intervened. The NRC staff evaluated the requests and found that the modifications would not endanger public health and safety. The staff did not consider any potential environmental effects of storage of spent fuel at the reactors beyond the dates of expiration of their operating licenses. NRC's Atomic Safety and Licensing Board Panel (ASLBP) adopted the staffs safety and environmental findings and approved the license amendments for the two plants. It too did not consider the effects of at-reactor storage beyond the expiration of the facility opera ting license. The Board's decision was appealed to the Atomic Safety and Licensing Appeal Board (ASLAB). The ASLAB affirmed the Licensing Board's decision, citing the Commission's "* *

reasonable confidence that wastes can and will in due course be disposed of safely * * *" in the Commission's denial of the NRDC petition.

The decision of the ASLAB was appealed to the U.S. Circuit Court of Appeals. On May 23, 1979 the Court declined to stay or vacate the license amendments, but remanded to NRC the question of "* *

whether there is reasonable assurance that an offsite storage PART 51 STATEMENTS OF CONSIDERATION solution will be available by the years 2007-2009, the expiration of the plants' operating licenses, and if not, whether there is reasonable assurance that the fuel can be safely stored at the reactor sites beyond those dates." In its decision to remand to NRC, for consideration in either a generic rulemaking or an adjudicatory proceeding, the Court observed that the issues of storage and disposal of nuclear waste were being considered by the Commission in an ongoing generic proceeding known as the "S-3 Proceeding" on the environmental impacts of uranium fuel cycle activities to support the operation of a light water reactor, and that it was appropriate to remand in light of a pending decision on that proceeding and analysis.

On October 18, 1979, NRC announced that it was initiating a rulemaking' proceeding in response to the Appeals Court remand and as a continuation of the NRDC proceeding. Specifically, the purpose of the proceeding was for the Commission "* *

to reassess its degree of confidence that radioactive wastes produced by nuclear facilities will be safely disposed of, to determine when any such disposal will be available, and whether such wastes can be safely stored until they are disposed of." The Commission recognized that the scope of this proceeding would be broader than the Court's instruction, which required the Commission to address only storage-related questions.

The Commission believed, however, that the primary public concern was the safety of waste disposal rather than the availability of an off-site solution to the storage problem. The Commission also committed itself to reassess its basis for confidence that methods of safe permanent disposal for high-level waste would be available when needed. Thus, the Commission chose as a matter of policy not to confine itself exclusively to the narrower issues in the court remand. In the Notice of Proposed Rulemaking, the Commission also stated that if the proceeding led to a finding that safe site storage or disposal would be available before expiration of facility operating licenses, NRC would promulgate a rule providing that the impact of onsite storage of spent fuel after expiration of facility operating licenses need not be considered in individual licensing proceedings. The Waste Confidence Decision was issued on August 31, 1984 (49 FR 34658). In the Decision, the Commission made five findings. It found reasonable assurance that: (1) Safe disposal of high-level radioactive waste and spent fuel in a mined geologic repository is technically feasible. (2) One or more mined geologic repositories for commercial high-level radioactive waste and spent fuel will be available by the years 2007-2009, and sufficient repository capacity will be available within 30 years beyond expiration of any reactor operating license to dispose of existing commercial high-level radioactive waste and spent fuel originating in such reactor and generated up to that time. (3) High-level radioactive waste and spent fuel will be managed in a safe manner until sufficient repository capacity is available to assure the safe disposal of all high-level radioactive waste and spent fuel. (4)' If necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the expiration of that reactor's operating license at that reactor's spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations. (5) Safe independent onsite or offsite spent fuel storage will be made available if such storage capacity is needed. On the day the Decision was issued, the Commission also promulgated two rulemaking amendments: (1) An amendment to 10 CFR part 50, which required that no later than five years before expiration of reactor operating licenses, the licensee must provide NRC with a written plan for management of spent fuel onsite,. until title for the spent fuel is transferred to the DOE; and (2) an amendment to 10 CFR part 51 which provided that environmental consequences of spent fuel storage after expiration of facility licenses need not be addressed in connection with issuance of or amendment to a reactor opera ting license. In issuing the part 51 amendment, the Commission stated that although it had reasonable assurance that one or more repositories would be available by 2007-2009, it was possible that some spent fuel would have to be stored beyond those dates. The part 51 amendment was based on the Commission's finding in the Waste Confidence Proceeding that it had reasonable assurance that no significant environmental impacts will result from storage of spent fuel for at least 30 years beyond expiration of reactor opera ting licenses. Enactment of the NWPA contributed significantly to the basis for the Commission's 1984 Decision and companion rulemakings. The Act established a funding source and process with milestones and schedules 51-SC-4 for, among other things, the development of a monitored retrievable storage [MRS) facility and two repositories, one by early 1998 and a second, if authorized by Congress, at a later date, initially planned by DOE for 2006. For each repository, the Act required DOE to conduct in-situ investigations of three sites and recommend one from among them to the President and Congress for repository development. The NWPA also required DOE to recommend, from among alternative sites and designs, a site and design for an MRS for spent fuel and high-level waste management before disposal. The Commission's licensing and regulatory authority over both storage and disposal facilities was preserved by the Act. In the four years after enactment of the NWPA, DOE met a number of the Act's early program requirements, but also encountered significant difficulties. It published a final Mission Plan for the overall NWPA program, and followed with a Project Decision Schedule for DOE and other Federal agency actions. It promulgated, with Commission concurrence, a set of guidelines for repository siting and development. It published draft and final environmental assessments for nine candidate repository sites, and recommended three for characterization. It completed and submitted to Congress an environmental assessment, a program plan, and a proposal with a site and design for an MRS. All these actions followed extensive interactions with interested Federal agencies, State, Indian tribal, and local governments, and other organiza lions. In the course of these activities, however, DOE also slipped its schedule for operation of the first repository by five years, indefinitely postponed efforts toward a second repository, and had to halt further MRS siting and development activities pending Congressional authorization. In December, 1987, Congress enacted the Nuclear Waste Policy Amendments Act (NWPAA). The NWPAA redirected the high-level waste program by suspending site characterization activities for the first repository at sites other than the Yucca Mountain site, and by suspending all site-specific activities with respect to a second repository. The Amendments Act also authorized and set schedule and capacity limits on the MRS. The purpose of these limitations, according to sponsors of the legislation, was to assure that an MRS would not become a substitute for a geologic repository. Consistent with its commitment to revisit its Waste Confidence conclusions at least every five years, the PART 51 STATEMENTS OF CONSIDERATION Commission has undertaken the current review to assess the effect of these and other developments since 1984 on the basis for each of its five findings. In this document, the Commission supplements the basis for its earlier findings and the environmental analysis of the 1984 Decision. The Commission proposes to amend its second finding, concerning the timing of initial availability and sufficient capacity of a repository, and its fourth finding, concerning the duration of safe spent fuel storage. These proposed revisions are based on the following considerations: (1) The five-year slippage, from 1998 to 2003, in the DOE schedule for repository availability; (2) The additional slip of at least 18 months since January 1987 in the DOE schedule for the next step in the repository program, the excavation of the exploratory shaft; (3) The need to continue accounting for the possibility that the Yucca Mountain site might be found unsuitable and that DOE would have to initiate efforts to identify and characterize another site for the first repository; (4) The statutory suspension of specific activities for the second repository; (5) DOE's estimate that site screening for a second repository should start about 25 years before the start of waste acceptance; and (6) Increased confidence in the safety of extended spent fuel storage, either at the reactor or at independent spent fuel storage installa lions. The Commission is also proposing elsewhere in this issue of the Federal Register that 10 CFR § 51.23(a) be amended to confirm with the proposed revisions to Findings 2 and 4. Organization and Table of Contents In conducting this review, the Commission has addressed, for each of its 1984 Findings, two categories of issues. The first category consists of the issues the Commission considered in making each Finding at the time of the initial Waste Confidence Decision. For these issues, the Commission is interested in whether its conclusions, or the Finding these conclusions support, should be changed to address new or foreseeable developments that have arisen since the first Waste Confidence Decision. The second category of issues consists of those the Commission believes should be added to the 1984 issues in light of subsequent developments. [To enable the reader to follow more easily, the lengthy discussions of Findings 1 and 2 have been organized to address each original and new issue under subheadings.) The Commission seeks comment on whether it has identified all the issues relevant to its proposed findings, and on whether its analyses of these issues supports the conclusions and findings proposed. Table of Contents I. First Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 1: 1. Identification of acceptable sites. 2. Development of effective waste packages: [al Considerations in developing waste package. [bl Effect of reprocessing on waste form and waste package. 3. Development of effective engineered barriers for isolating wastes from the biosphere: [a) Backfill materials. [bl Borehole and shaft sealants. B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 1: 1. Termination of Multiple Site Characterization.

2. Relevance to NRC's "S-3 Table" proceeding.

3. International develoments in spent fuel disposal technology.

C. Conclusion on Finding 1. II. Second Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 2: 1. Technical uncertainties: [al Finding technically acceptable sites in a timely fashion. [bl Timely development of waste packages and engineered barriers.

2. Institutional uncertainties: (al Measures for dealing with State-local concerns. (bl Continuity of the management of the waste program. [cl DOE's schedule for repository development.

B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 2: 1. Potential delay under the program of single site characterization.

2. Potential limitations on timing of availability of disposal capacity: (al Impact of possible limited disposal capacity at Yucca Mountain, indefinite suspension of second repository program. [bl Impact of uncertainty in spent fuel projections on need to consider second repository program. 3. Impact of slippages in DOE program on availability of a repository when needed for health and safety reasons. 4. Effect of NRC emphasis on completeness and quality. C. Conclusion on Finding 2. III. Third Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 3: Licensee compliance with NRC regulations and license conditions; Safe management of spent fuel past expira lion of opera ting licenses; Availability of DOE interim storage. 51-SC-5 B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 3: Responsibility for spent fuel storage beyond 1998; Delay in second repository; Potential for license renewals.

IV. Fourth Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 4: Long-term integrity of spent fuel under water pool storage conditions; Structure and component safety for extended facility operation for storage; Safety of dry storage of spent fuel; Potential risks of accidents and acts of sabotage of spent fuel storage facilities. B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 4: Radiological and non-radiological consequences of extended spent fuel storage; Potential delay in first repository, license renewals, delay in second repository; Environmental assessment and finding of no significant impact of at-reactor storage beyond 30 years after reactor's licensed life for operation. V. Fifth Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 5: Adequacy of NWPA for determining responsibility for timely spent fuel storage; Spent fuel discharge projections; Industry commitment to implement from-reactor storage. B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 5: Responsiblity for spent fuel storage beyond 1998; Advances in technology for dry storage; Benefits of monitored retrievable storage facility under NWPAA; License renewals; Options for offsite storage under NWPAA. Original Finding 1 The Commission finds reasonable assurance that safe disposal of level radioactive waste and spent fuel in a mined geologic repository is technically feasible. Proposed Finding I Same as above. 1.A. Issues Considered in Commission's 1984 Decision on Finding 1 1.A.1. The Identification of Acceptable Sites Under the Nuclear Waste Policy Act of 1982 (NWPAJ, the Department of Energy (DOE) had responsibility for identifying candidate sites for a geologic repository and for repository development. The first requirement leading to recommendation of candidate sites was formal notification of States with one or more potentially acceptable sites for a repository within 90 days of PART 51 STATEMENTS OF CONSIDERATION enactment of the NWPA. In February 1983, the DOE identified nine potentially acceptable sites for the first repository. Four of the sites were in bedded-salt formations, three were in salt domes, one in volcanic tuff, and one in basalt. The NWPA required that each site nomination be accompanied by an environmental assessment (EA). In December 1984, DOE published Draft EAs (DEAs) for each of the nine sites identified as potentially acceptable and proposed the following sites for nomination: The reference repository location at Hanford, WA; Yucca Mountain, NV; Deaf Smith County, TX; Davis, Canyon, UT; and Richton Dome, MS. In May 1986, DOE released Final EAs (FEAs) for the five sites nominated. At that time, DOE recommended that the Yucca Mountain, Hanford, and Deaf Smith County sites undergo site characterization. The President approved the recommendation. The NRC staff provided extensive comments on both the DEAs and the FEAs. NRC concerns on the FEAs related primarily to DOE's failure to recognize uncertainty inherent in the existing limited data bases for the recommended sites, and the tendency of DOE to present overly favorable or optimistic conclusions. The primary intent of the comments was to assist DOE in preparing high-quality Site Characterization Plans (SCPs) for each site, as required under the NWPA, before excavation of exploratory shafts. NRC concerns can only be addressed adequately through the site characterization process, because one of the purposes of this process is to develop the data to evaluate the significance of concerns relative to site suitability. NRC did not identify any fundamental technical flaw or disqualifying factor which it believed would render any of the sites unsuitable for characterization. Further, NRC did not take a position on the ranking of the sites in order of preference, because this could be viewed as a prejudgment of licensing issues. NRC was not aware of any reason that would indicate that any of the candidate sites was unlicenseable. Nor has NRC made any such finding to date with respect to any site identified as paten ti ally acceptable. In March 1987, Congress began drafting legislation to amend the repository program. NRC provided comments on a number of these draft amendments. In December 1987, the NWPAA was enacted. In a major departure from the initial intent of the NWPA, the new law required that DOE suspend site characterization activities at sites other than the Yucca Mountain site. This decision was not based on a technical evaluation of the three recommended sites or a conclusion that the Hanford and Deaf Smith sites were not technically acceptable. According to sponsors of the legi~lation, the principal purpose of the requirement to suspend characterization at these sites was to

reduce costs. In effect, the NWPAA directed DOE to characterize candidate sites sequentially, if necessary, rather than simultaneously.

If DOE determines at any time that the Yucca Mountain site is unsuitable, DOE is to terminate all site characterization activities and report to Congress its recommendations for further actions. The NRC staff has identified numerous issues regarding the Yucca Mountain site that may have a bearing on the licenseability of that site. These issues will have to be resolved during site characterization. An example of a site issue that may bear on the question of suitability is tectonic activity, the folding or faulting of the earth's crust. In the 1984 Waste Confidence Decision, NRC noted that"* *

the potential sites being investigated by DOE are in regions of relative tectonic stability." The authority for this statement came from the Position Statement of the U.S. Geological Survey (USGS). NRC has raised concerns regarding tectonic activity at the Yucca Mountain site in the comments on the draft and final EAs, and in the draft and final Point Papers on the Consultation Draft Site Characterization Plan. If it appears during site characterization that the Yucca Mountain site will be unable to meet NRC requirements regarding isolation of waste, DOE will have to suspend characterization at that site and report to Congress.

DOE's program of site screening in different geologic media was consistent with section 112(a) of the NWPA, which required that DOE recommend sites in different geologic media to the extent practicable. This strategy was to ensure that if any one site were found unsuitable for reasons that would render other sites in the same geologic medium unacceptable, alternate sites in different host rock types would be available. NRC referred to this policy in its 1984 Waste Confidence Decision, when it said, in support of its argument on technical feasibility, that"* *

DOE's program is providing information on site characteristics at a sufficiently large number and variety of sites and geologic media to support the expectation that one or more technically acceptable sites will be identified.

NRC recognizes that simultaneous site characterization is not necessary to identify a repository site that would 51-SC-6 meet NRC's technical criteria for isolating wastes. Sequential site characteriza lion does not necessarily preclude or hinder identification of an acceptable site for a repository. NRC did express concern to Congress, on several occasions during deliberations over the proposed legislation, that sequential site characterization could delay considerably the schedule for opening a repository if the site undergoing characterization were found to be unlicenseable. NRC also indicated that this potential for delay would have to be considered by NRC in reevaluating the findings in its Waste Confidence Decision. The impact of this redirection of the high-level waste program on the Commission's Waste Confidence findings is not on the ability to identify technically acceptable sites, but on the timing of availability of technically acceptable sites. Because characterization of multiple sites appears to be more directly related to the timing of repository availability than to the feasibility of geologic disposal, consideration of the above statement in light of the NWPAA program redirection will be discussed under Finding 2. Another question bearing on whether technically acceptable sites can be found is whether compliance with Environmental Protection Agency (EPA] environmental standards for disposal of spent fuel and high-level waste can be demonstrated. These standards, originally promulgated in final form in September 1985, were vacated in July 1987, by the U.S. Court of Appeals, and remanded to EPA for further consideration (see NRDCv. EPA, 824 F. 2d 1258). As originally promulgated, the standards set limits on releases of radioactive materials from the site into the accessible environment over a 10,000-year period following disposal. They also required that there be less than one chance in ten that the release limits will be exceeded in 10,000 years, and less than one chance in 1,000 that releases will exceed ten times the limits over 10,000 years. In past comments on draft and proposed EPA standards, and in related NRC rulemaking efforts, NRC has expressed concern that probabilistic analyses should not be exclusively relied on to demonstrate compliance with EPA release limits. NRC's comments said in part that"* * * [t]he numerical probabilities in [the standards] would require a degree of precision which is unlikely to be achievable in evaluating a real waste disposal system." The comments went on to explain that"* *

identification of the relevant processes and events PART 51 STATEMENTS OF CONSIDERATION affecting a particular site will require considerable judgment and will not be amenable to accurate quantification, by statistical analysis, of their probability of occurrence." NRC believed then, and continues to believe, that it must make qualitative judgments about the data and methodologies on which the numerical probabilities were based. In response to NRC concerns, EPA incorporated language into its 1985 standards that appeared to allow flexibility to combine qualitative judgments with numerical probability estimates in a way that might have made implementation of the EPA standards practicable.

The text of those standards recognized that "proof of the future performance of a disposal system is not to be had in the ordinary sense of the word" with the substantial uncertainties and very long performance period involved. The 1985 standards emphasized that a "reasonable expectation"-rather than absolute proof-is to be the test of compliance. "What is required," the text of the standards said, "is a reasonable expectation, on the basis of the record * * *, that compliance

*
will be achieved." In an additional attempt to provide flexibility for implementation of the standards, EPA also provided that numerical analyses of releases from a repository were to be incorporated into an overall probability distribution only "to the extent practicable." This phrase appeared to allow some discretion for NRC to incorporate qualitative considerations into its license making, rather than having to rely solely on numerical projections of repository performance.

On the strength of these and other EPA assurances, the Commission did not object when the final standards were published in 1985. Pursuant to the remand by the Federal court in 1987, EPA is currently revising its standards for disposal of spent fuel and high-level waste. The court's decision directed that the remand focus on the ground water and individual protection requirements of the standards. Although the EPA standards are still undergoing development at this time, the Commission does not currently see a sufficient basis to withdraw its confidence in the feasibility of evaluating compliance with such standards. NRC staff will closely monitor the development of repromulgated standards to assure that EPA methodologies for demonstrating compliance with them can be applied by NRC to evaluate DOE's demonstration of compliance. In sum, considering both past and current programs for characterizing sites, the Commission concludes that technically acceptable sites for a repository can be found. The Commission is confident that, given adequate time and resources, such sites can be identified, evaluated, and accepted or rejected on their merits, even if no more than one site is undergoing site characterization. This judgment does not rest on the acceptability of the Yucca Mountain site or any one future candidate site. 1.A.2. The Development of Effective Waste Packages 1.A.2.a. Considerations in developing waste packages. The NWPA required NRC to promulgate technical requirements and criteria to be applied in licensing a repository for high-level radioactive waste. Under section 121 of the Act, these technical criteria must provide for use of a system of multiple barriers in the design of the repository and such restrictions on the retrievability of waste as NRC deems appropriate. The system of multiple barriers includes both engineered and natural barriers. The waste package is the first engineered barrier in the system of multiple barriers to radionuclide escape. The waste package is defined as the "waste form and any containers, shielding, packing and other absorbent materials immediately surrounding an individual waste container." Before sinking an exploratory shaft for site characterization, DOE is required to prepare an SCP including a description of the waste form or packaging proposed for use at the repository, and an explanation of the relationship between such waste form or packaging and the geologic medium of the site. The multiple barrier approach to radioactive waste isolation in a geologic repository is implemented in NRC requirements by a number of performance objectives and by detailed siting and design criteria. The NRC performance objective for the waste package requires substantially complete containment for a period of not less than 300 years nor more than 1,000 years after permanent closure of the repository. The technical design criteria for the waste package require that interaction of the waste package with the environment not compromise performance of the package, the underground facility, or the geologic setting. Therefore, the waste package design must take into account the complex site-specific interactions between host rock, waste package, and ground water that will affect waste package and overall repository performance. 51-SC-7 Under the NWPAA, DOE was required to suspend site characterization activities at sites other than the Yucca Mountain, NV site. Consequently, DOE has narrowed the range of waste package designs to a design tailored for unsaturated luff at the Yucca Mountain site. This aspect of the high-level waste program redirection may facilitate and expedite the waste package design process insofar as it enables DOE to concentrate its efforts on developing a single design for a single site instead of three designs for sites in bedded salt, basalt, and unsaturated tuff. Currently, DOE is evaluating uncertainties in waste package design related to waste form, container type, and environment. The current conceptual design for the waste package is based on several assumptions. the waste form is presumed to be old spent fuel or high-level waste in the form of borosilicate glass in steel canisters. (In addition to spent fuel and high-level waste, the waste form may include greater-than-Class C (GTCC) low-level waste. This waste is not routinely acceptable for near-surface disposal under NRC regulations for disposal of low-level wastes, but is acceptable for disposal in a repository licensed for disposal of spent fuel and high-level wastes. This waste might include such materials as sealed sources and activated metals from the decommissioning of reactors and production facilities.) Six materials are being considered for fabrication of containers, including austenitic steel (3161), nickel-based alloys (Alloy 825), pure copper (CDA 102), copper-based alloys bronze, CDA-613, and 70--30 Cu-Ni, CDA-715), and a container with a metal outer shell and ceramic liner. The reference container for the spent fuel and high-level waste is a 1.0-cm thick cylinder to be made of American Iron and Steel Institute (AISI) 3041 stainless steel. This will be DOE's benchmark material, against which other materials are to be compared. DOE currently intends for spent fuel containers to be filled with an inert gas, such as argon, before being welded closed. The reference repository location is in the unsaturated luff of the Topopah Spring Formation underlyign Yucca Mountain. According to DOE, little flowing water is thought to be present there to contribute to corrosion of the waste containers, although the degree of saturation in this luff is estimated to be 65+/-19 percent of the available void space in the rock. DOE has acknowledged, however, that the PART 51 STATEMENTS OF CONSIDERATION greatest uncertainties in assessing waste package performance at Yucca Mountain stem from difficulty in characterizing and modeling the coupled geochemical-hydrologic processes that represent the interactions between the host rock, waste package, and ground water. The final waste package design will depend on the results of site characterization and laboratory testing to reduce uncertainty in predicting these interactions in the reference repository horizon. The final design will also be shaped by research in understanding the degradation of candidate container materials, and the characteristics of the likely reference waste forms. Regarding the state of technology for developing long-lived waste package containers, the Swedish Nuclear Fuel and Waste Management Company (SKB), the organization responsible for radioactive waste disposal in Sweden, has described a container for spent fuel rods that consists of a 0.1-m thick copper canister surrounded by a bentonite overpack. The design calls for pouring copper powder into the void spaces in the canisters, compacting the powder using hot-isostatic pressing with an inert gas, and sealing the canisters. SKB estimates that the copper canister waste package has a million-year lifetime. (See also 1.B.3. below.) As noted in NRC's Final Point Papers on the Consultation Draft Site Characterization Plan, the Commission does not expect absolute proof that 100 percent of the waste packages will have 100 percent containment for 300 to 1,000 years. Since that time, the NRC staff has completed its review of the December 1988 Site Characterization Plan for Yucca Mountain. Although the Commission continues to have concerns about DOE's waste package program, nothing has occurred to diminish the Commission's confidence that as long as DOE establishes conservative objectives to guide a testing and design program, in tuff or in other geologic media if necessary, it is technically feasible to develop a waste package that meets the performance objective for substantially complete containment. 1.A.2.b. Effect of reprocessing on waste form and waste package. The Draft 1988 Mission Plan Amendment estimates that a total of about 77,800 metric tons of heavy metal (MTHM) of spent nuclear fuel and high-level radioactive waste will be available for disposal by the year 2020. (This estimate is based on a "no new orders" assumption for commercial nuclear reactors and a 40-year reactor lifetime.) Of this 77,800 MTHM, about 9,400 MTHM will consist of reprocessed defense waste and a small amount of commercial reprocessed waste from the West Valley Demonstration Project. The decision to locate the defense high-level waste in the repository for wastes from commercial power reactors resulted from the requirement in Section 8 of the NWPA that the President evaluate the possibility of developing a waste-only repository. In February 1985, DOE submitted a report to the President recommending a combined commercial and defense repository. In April 1985, the President agreed that no basis appeared to exist for a defense-only repository and directed DOE to dispose of defense waste in the commercial repository. About 8,750 MTHM of reprocessed high-level waste from defense facilities at Savannah River, SC, Hanford, WA, and Idaho Falls, ID will be available by 2020 for disposal in the repository, according to the Draft 1988 Mission Plan Amendment. This waste will likely be solidified into a borosilicate glass matrix. About 640 MTHM of reprocessed high-level waste will come from the West Valley Demonstration Project, a facility for wastes from discontinued commercial reprocessing of spent fuel at that site. This reprocessed waste also will be solidified, probably in a borosilicate glass waste form. Waste-form testing for the Yucca Mountain site is focusing on both spent fuel and reprocessed high-level waste. The performance of the waste form in providing the first barrier to radionuclide migration is being evaluated on the basis of the physical and chemical environment of the waste form after disposal, the performance of the waste container, and the emplacement configuration. A major limitation on glass form testing is that the actual waste glasses to be disposed of are not available, and their exact composition will not be established until after further testing. Reference waste-glass compositions are being used for studies on the effect of variation in glass composition on performance. (These glass compositions are designed by Savannah River Laboratory (SRL) for defense high-level waste, and by Pacific Northwest Laboratory (PNL) for the commercial high-level wastes to be vitrified under the West Valley Demonstration Project Act.) The reference compositions will be revised when better analyses of the composition of the wastes at SRL and West Valley are available. The test program will seek to establish upper bounds on leaching of important radionuclides, and the extent 51-SC-8 to which glass fracturing increases leach rate. Other factors influencing leach rate are temperature, pH of the leaching solution, formation of solid layers on the surface of the waste glass, irradiation, water volume, and chemistry. It is possible that renewed reprocessing of spent fuel from nuclear power reactors may result in a greater proportion of reprocessed waste to spent fuel than is currently anticipated. Although such a departure from the current plan to dispose of mostly unreprocessed spent fuel in the repository does not appear likely at this time, the Commission believes it is important to recognize the possibility that this situation could change. The possibility of disposal of reprocessed waste as an alternative waste form to spent fuel assemblies was recognized by the Commission in the 1984 Waste Confidence Decision. The Commission noted that the disposal of waste from reprocessing had been studied for a longer time than the disposal of spent fuel. and that the possibility of reprocessing does not alter the technical feasibility of developing a suitable waste package. The Commission went on to say that there is evidence that the disposal of reprocessed high-level waste may pose fewer technical challenges than the disposal of spent fuel. As long as DOE uses conservative assumptions and test conditions for evaluating the performance of different waste forms against NRC licensing requirements, the Commission has no basis to change its finding that there is reasonable assurance that reprocessing does not reduce confidence in the technical feasibility of designing and building a waste package that will meet NRC licensing requirements in a variety of geologic media. 1.A.3. The Development of Effective Engineered Barriers for Isolating Wastes From the Biosphere 1.A.3.a. Backfill materials. At the time of the 1984 Waste Confidence Decision, DOE was developing conceptual designs for backfill in several geologic media. Most candidate sites at that time were in saturated rock, and the conceptual designs included backfilling or packing around waste containers to prevent or delay ground water flow which could enhance corrosion and radionuclide transport near the waste containers. The conceptual design for the engineered barrier system at the Yucca Mountain site has different parameters because the site is unsaturated; instead of backfill or packing around the waste container, there is to be an air gap PART 51 STATEMENTS OF CONSIDERATION between sides of the waste canister and the host rock. Backfill material around the container is not required under NRC regulations for the waste package. NRC regulations require that"* *

containment of level waste within the waste packages [which includes the container]

will be substantially complete for a period to be determined by the Commission

*
provided, that such period shall not be less than 300 years nor more than 1000 years after permanent closure of the repository" (10 CFR 60.113(a)(l)(ii)(BJ), and that the entire engineered barrier system meet the release rate performance objective of 1 part in 100,000 per year. Backfill is also a component of the borehole, shaft, and ramp seals, which are not part of the engineered barrier system or the underground facility.

Boreholes, shafts, and ramps must be sealed when the repository is permanently closed. This aspect of backfilling is discussed below under "Development of sealants." Backfill may also include crushed rock used to fill openings such as drifts in the underground facility. At the Yucca Mountain candidate site, DOE currently plans to fill openings in the underground facility at closure of the repository. Backfilling is not planned before repository closure because it is not needed for structural support for the openings, and it would make waste retrieval more difficult. At closure of the facility, however, openings will be backfilled with coarse tuff excavated for the facility. In the conceptual design provided in the SCP, the selection of coarse tuff as backfill material is based on numerical simulations performed by DOE which suggest that coarse luff would be a more effective barrier to capillary flow in the backfill matrix than fine materials. DOE's design for the engineered barrier system submitted with the license application will have to contain information sufficient for NRC to reach a favorable conclusion regarding the overall system performance objective. Backfill or packing around waste containers is not required by NRC regulations if DOE can demonstrate that applicable performance objectives can be met without it. If, on the basis of testing and experiments during site characterization, DOE decided that backfill would enhance engineered barrier system performance, the design would have to reflect this conclusion. DOE has already conducted research on a wide variety of candidate materials for backfill around waste packages in a variety of geologic media. The Commission contim;es to have confidence that backfill or packing materials can be developed as needed for the underground facility and waste package to meet applicable NRC licensing criteria and performance objectives. ' 1.A.3.b. Borehole and shaft seals. The engineered barrier system described above is limited to the waste package and the underground facility as defined in 10 CFR part 60. The undergound facility refers to the underground structure, including openings and backfill materials, but excluding shafts, boreholes. and their seals. Containment and release-rate requirements are specified for the engineered barrier system, but not for the borehole and shaft seals. Seals are covered under 10 CFR 60.112, the overall post-closure system performance objective for the repository. Among other things, this provision requires that shafts, boreholes and their seals be designed to assure that release of radioactive materials lo the accessible environment following permanent closure conform to EPA's generally applicable standards for radioactivity. Although the criteria for seals given in 10 CFR part 60 do not specifically mention seals in ramps and the underground facility, it is reasonable to consider them together with borehole and shaft sealants, because the seals and drainage design in ramps and the underground facility could also affect the overall system performance of the geologic repository. Construction of the exploratory shaft facility (ESF) will be the first major site characterization activity. The ESF will consist of two vertical shafts, one for testing and the other for support, and underground excavations for at-depth testing. The repository surface facilities will be connected to the underground facility by two additional shafts (a and-materials shaft and the emplacement area exhaust shaft) and two ramps, a waste ramp for bringing radioactive waste and spent fuel into the repository, and a luff ramp for removing rock from the underground facility to a luff pile. In addition to these shafts and ramps, there will be exploratory boreholes for obtaining samples of rock, water, and gases in strata at varying depths. Exploratory boreholes have the potential to provide information on hydrologic properties of the Yucca Mountain site, with emphasis on movement of water in unsaturated luff. Other properties which will be studied using exploratory boreholes are lithologic, structural, mechanical, and thermal properties of the host rock. 51-SC-9 When the repository is decommissioned, NRC expects that most, if not all, shafts, ramps, and boreholes will probably have to be sealed to reduce the possibility that they could provide preferential pathways for radionuclide migration from the underground facility to the acccessible environment. DOE estimates that as many as 350 shallow and 70 deep exploratory boreholes may be emplaced by the time site characterization has been completed at the Yucca Mountain site. Decommissioning may not occur for up to 100 years after commencement of repository operations. Because the final design for seals will likely have been modified from the initial license application design (LAD), DOE is viewing the seal LAD as serving two primary functions. As set forth in DOE's SCP for the Yucca Mountain candidate site, the seal LAD is to establish that: (1) "* *

technology for constructing seals is reasonably available;" and (2) "* *
there is reasonable assurance that seals have been designed so that, following permanent closure, they do not become pathways that compromise the geologic repository's ability to meet the closure performance objectives." To establish the availability of technology for seal construction, DOE has identified at least 31 site properties that need to be characterized in determining necessary seal characteristics.

These properties include saturated hydraulic conductivity of alluvium near shafts, the quantity of water reaching the seals due to flooding events, and erosion potential in the shaft vicinity. The SCP also discusses material properties that need to be identified to determine sealing components such as initial and altered hydrologic properties of materials. The SCP indicates that DOE is planning to use crushed tuff and cements in the sealing program at the Yucca Mountain candidate site. The stated advantages of using luff include minimizing degradation of seal material and avoiding disruption of ambient ground-water chemistry. DOE's current design concept for meeting the overall performance objectives includes a combination of sealing and drainage. Seal requirements may be reduced in part by: (1) Limiting the amount of surface water that may enter boreholes, shafts, and ramps; (2) selecting borehole, shaft, and ramp locations and orientiation that provide long flow paths from the emplaced waste to the accessible environment above the repository; and (3) maintaining a sufficient rate of drainage below the repository horizon level so PART 51 STATEMENTS OF CONSIDERATION that water can be shunted past the waste packages without contacting them. Although DOE's program is focusing on seals for the Yucca Mountain candidate site, the Commission finds no basis for diminished confidence that an acceptable seal can be developed for candidated sites in different geologic media. The Commission finds no evidence to suggest that it can not continue to have reasonable assurance that borehold, shaft, ramp, and repository seals can be developed to meet 10 CFR part 60 performance objectives. J.B. Relevant issues That Have Arisen Since the Commission's Original Decision 1.B.1 In support of its argument on technical feasibility, the Commission stated in its 1984 Waste Confidence Decision that "* *

DOE's program is providing information on site characteristics at a sufficiently large number and variety of sites and geologic media to support the expectation that one or more technically acceptable sites will be identified." The NWPAA required, however, that DOE suspend site-specific site characterization activities under the Nuclear Waste Policy Act of 1982 at all sites other than the Yucca Mountain, NV site. Under the NWPAA, the DOE program has been redirected to characterize candidate repository sites in sequence rather than simultaneously.

If the Yucca Mountain site is found to be unsuitable, DOE must terminate site characteri;.mtion activities there and provide Congress with a recommendation for furture action, such as the charaterizaiton of another site. Because characterization of multiple sites now appears to be more directly related to the timing of repository availability than to the technical feasibility of geologic disposal as a concept, consideration of the , Commission's aforementioned 1984 statement in light of the NWPAA will be discussed under Finding 2. 1.B.2. What is the relationship, if any of the "S-3 Proceeding" to the current review of the Commission's 1984 Waste Confidence Findings? Would the planned revision of the S--3 rulemaking be affected if the Commission had to qualify its current confidence in the technical feasibility of safe disposal? In its decision to remand to NRC the questions of whether safe off site storage would be available to 2007-2009, or, if not, whether spent fuel could be safely stored onsite past those dates, the U.S. Circuit Court of Appeals observed that the issues of storage and disposal of nuclear waste were being considered by the Commission in an ongoing generic proceeding known as the "S-3" Proceeding. The S--3 Proceeding was the outgrowth of efforts to address generically the NEPA requirement for an evaluation of the environmental impact of operation of a light water reactor (LWR). Table S-3 assigned numerical values for environmental costs resulting from uranium fuel cycle activities to support one year of LWR operation. NRC promulgated the S-3 rule in April 1974. In July 1976, the U.S. Circuit Court of Appeals found that Table S--3 was inadequately supported by the record regarding reprocessing of spent fuel and radioactive waste management, in part because the Commission, in reaching its assessment, had relied heavily on testimony of NRC staff that the problem of waste disposal would be resolved. When the U.S. Circuit Court of Appeals issued the remand on what were to become the "Waste Confidence" issues in May 1979, NRC had pending before it the final amended S-3 rule. The Court regarded the resolution of the issue of waste disposal in the S-3 proceeding as being related to the issue raised by the petitioners in the appeals of the NRC decisions on the expansion of spent fuel storage capacity. The Court said that the " * *

disposition of the S-3 proceeding, although it has a somewhat different focus, may have a bearing on the pending cases." The Commission approved the final S-3 rule in July 1979. In October 1979, the Commission issued a Notice of Proposed Rulemaking (NPRM) on the Waste Confidence issues in response to the remand by the Court of Appeals. In the NPRM, the Commission stated that the proceeding would " * *
draw upon the record compiled in the Commission's recently concluded rulemaking on the environmental impacts of the nuclear fuel cycle, and that the record complied herein will be available for use in the general fuel cycle rule update discussed in that rulemaking." In the final Table S-3 rule issued in 1979, the Commission had said that " * *
bedded salt sites can be found which will provide effective isolation of radioactive waste from the biosphere." When the Commission issued the 1984 Waste Confidence Decision, part of the basis for the discussion of waste management and disposal in the August 1979 final S-3 rule had changed. For example, in 1984 the repository program was proceeding under the NWPA, which required that DOE recommend three sites for site characterization.

Although NRC is preparing to amend the S-3 Table, and add a new appendix 51-SC-10 to explain the basis for and significance of the data in the table, it is unlikely that the revisions will have any impact on the Commission's generic findings in the Waste Confidence proceeding. Nor is it likely that this reexamination of the Waste Confidence findings will affect the S-3 rule; the Waste Confidence Proceeding is not intended to make quantitative judgments about the environmental costs of waste disposal. Unless the Commission, in a future review of the Waste Confidence decision, finds that it no longer has confidence in the technical feasibility of disposal in a mined geologic repository, the Commission will not consider it necessary to review the S-3 rule when it reexamines its Waste Confidence findings in the future. 1.B.3. To what extent do developments in spent fuel disposal technology outside of the United States (e.g., Swedish waste package designs) enhance NRC's confidence in the technical feasibility of disposal of level waste and spent fuel? Spent fuel disposal technology is the subject of extensive research investigation in both Europe and North America. Advances in this technology are being communicated to the NRC staff both through bilateral agreements and the presentation of research results at international meetings. Outside the United States, studies of spent fuel as a waste form are now being conducted primarily in Canada and Sweden, although both France and West Germany have small programs in this area. The Swedish studies have been mainly concerned with boiling water reactor (BWR) spent fuel, whereas the Canadian studies focus on spent fuel from that country's CANDU reactors, which use unenriched uranium in a core immersed in "heavy" water made from deuterium. BWR and CANDU fuel, like pressurized water reactor (PWR) fuel, are uranium dioxide fuels clad in zircaloy. However, the burnup rates for these three fuel types vary considerably. Ongoing research studies on spent fuel include: work on the characterization of spent fuel as a waste form; the corrosion of spent fuel and its dissolution under oxidizing and reducing conditions; the radiolysis of ground water in the near vicinity of the spent fuel, and its effects on the dissolution of the fuel; and the development of models to predict the leaching of spent fuel over long time periods. The results of this work are steadily increasing our understanding of spent fuel as a waste form. High-level radioactive waste, whether it is spent reactor fuel or waste from reprocessing, must be enclosed in an PART 51 STATEMENTS OF CONSIDERATION outer canister as part of the waste package. The canister surrounding the waste is expected to prevent the release of radioactivity during its handling at the repository site before emplacement. After emplacement in the repository, it is expected to prevent the release of radioactivity for a specified period of time after the repository is closed, by providing a barrier to protect the waste from coming into contact with the ground water. For practical reasons, canister materials may be divided into the following classes: (1) Completely or partially thermodynamically stable materials such as copper; (2) passive materials such as stainless steel, titanium, Hastelloy, Inconel, and aluminum; (3) corroding or sacrificial materials such as lead and steel; and (4) non-metallic materials such as alumina and titanium dioxide ceramics and cement. Sweden has been conducting an extensive canister research program over the past several years. The main canister of interest is copper, but titanium, carbon steel, and alumina and titanium dioxide are also being studied as reasonable alternatives, should unexpected problems be discovered with using pure copper. The present Swedish canister design is a 100-mm thick copper container (as described previously in section A.2.a.), which is claimed to provide containment, in conjunction with an appropriate backfill material, for a period on the order of one million years. The critical factors for the isolation period for copper canisters are: (1) The presence of corrosive substances such as sulfphide ions in the ground water; (2) the possibility of these substances reaching the canister surface; and (3) the degree of inhomogeneity, or pitting, of the resulting corrosion. Studies are continuing to obtain more information on pitting corrosion of copper and on techniques for welding thick-walled copper containers. Several conceptual designs for canisters for the safe disposal of unreprocessed spent fuel have also been developed in Canada. One canister design option is the supported-shell, metal-matrix concept, which involves packing the spent fuel bundles into a thin corrosive-resistant shell and casting the remaining space with a low melting point metal or alloy. Structural support for the shell would be provided by the resulting metal matrix. Lead is a possible matrix material because of its favorable casting properties, cost, and low melting point. Other supported shell canister concepts include the packed-particulate and structurally-supported designs. In these designs, a thin outer shell is supported by a particulate material packed around a steel internal structure that contains the spent fuel bundles. Several materials h~ve been identified for the fabrication of the corrosion resistant outer shell, including commercially pure and low-alloy titanium, high nickel-based alloys such as Inconel 625, and pure copper. Detailed designs have been produced for all three types of supported shell canisters incorporating either a titanium or nickel alloy shell less than 6-mm thick. A conceptual design has also been produced for a copper-shell supported canister and a metal-matrix container with a relatively thick (25-mm) copper shell and a lead matrix material. This last canister is intended to contain 72 used CANDU fuel bundles in four layers of 18 bundles each. Both the Canadian and Swedish conceptual designs for the disposal of spent fuel in canisters provide for surrounding the canister with backfill material as part of the waste package when it is emplaced in the repository. This backfill material would be packed around the canister to retard the movement of ground water and radionuclides. Investigations of backfill material at the Stripa mine in Sweden have shown that bentonite and silica sand can be employed successfully as backfill, both around the canister and in repository tunnels. A bentonite-silica mixture is the recommended backfill material on the basis of its thermal and mechanical properties. Bentonite backfills have been shown to produce hydraulic conductivities that are very similar to the surrounding granite at Stripa. Problems concerning the variability of bentonite samples from different geographic locations can be eliminated if material from a single source is used. The presence of sulfur and some organic material, including bacteria, in many bentonites poses some problems related to accelerated corrosion. Treatment with hydrogen peroxide may be used to oxidize these organics. Heating the bentonite to 400 degrees C can also be effective, although this may alter the crystal structure of the bentonite. Many countries intend to dispose of their high-level radioactive waste by first converting the wastes into a solid, vitrified form after reprocessing. Since the leaching of the waste form by circulating ground water after disposal is the most likely mechanism by which the radionuclides might be returned to the biosphere, the waste form must be composed of a highly stable material with an extremely low solubility in 51-SC-11 ground water. Thus, the waste form itself should function as an immobilization agent to prevent any significant release of radionuclides to the biosphere over very long time periods. The two primary materials currently being considered for use as solidified waste forms are borosilicate glass and SYNROC, a man-made titanate ceramic material. SYNROC was initially developed in Australia as an alternative material to borosilicate glass. It is composed primarily of three minerals (hollandite, zirconolite, and perovskite) which collectively have the capacity to accept the great majority of radioactive level waste constituents into their crystal lattice strucuture. These three minerals, or closely related forms, occur naturally, and have been shown to have survived for many millions of years in a wide range of natural environments. SYNROC has the property of being extremely resistant to leaching by ground water, particularly at temperatures above 100 degrees C. In addition, the capacity of SYNROC to immobilize high-level wastes is not markedly impaired by high levels of radiation damage. The high leach-resistance of SYNROC at elevated temperatures increases the range of geologic environments in which it may be used, such as deep geologic repositories in both continental and marine environments. Research and development work on improving SYNROC production technology is currently being done jointly in Australia and Japan. New methods of using metal alkoxides in the fabrication of SYNROC to obtain high homogeneity and lower leachability have recently been developed in Australia. The Japanese have recently developed a new method that uses titanium hydroxide, as a reducing agent to produce SYNROC with a high density and low leach rate. A pilot facility for the production of non-radioactive SYNROC is not in operation in Australia, and a small pilot facility for producing SYNROC with radioactive constituents is being completed in Japan. On the basis of current information from the foreign studies just described on canisters, spent fuel as a waste form, backfill materials, and alternatives to borosilicate glass waste forms, the Commission concludes that there is no basis for diminished confidence that an acceptable waste package can be developed for safe disposal of high-level waste and spent fuel. PART 51 STATEMENTS OF CONSIDERATION J.C. Conclusion on Finding 1 The Commission has reexamined the basis for its First Finding in the 1984 Waste Confidence Decision in light of subsequent program developments, and concludes that Finding 1 should be reaffirmed. The technical feasibility of a repository rests initially on identification of acceptable sites. At this time, the Commission is not aware of any evidence indicating that Yucca Mountain is not acceptable for site characterization. There are many outstanding questions regarding the licenseability of the site, however, and they must be answered satisfactorily in order for NRC to issue a construction authorization for that site. If data obtained during site characterization indicate that the Yucca Mountain site is not suitable for a repository, DOE is required by the NWPAA to terminate site characterization activities and report to Congress. Within six months of that determination, DOE must make a recommendation to Congress for further action to assure the safe, permanent disposal of spent fuel and high-level waste. DOE could recommend, for example, that Congress authorize site characterization at other sites. Considering DOE's investigations of other potentially acceptable sites before its exclusive focus on Yucca Mountain, the Commission has no reason to believe that, given adequate time and program resources, a technically acceptable site cannot be found. The technical feasibility of geologic disposal also depends on the ability to dev,elop effective engineered barriers, such as waste packages. DOE is currently evaluating six candidate materials for waste containers, including austenitic steel and copper-and based alloys, and is planning form testing based on both spent fuel and high-level waste in borosilicate glass. On the basis of DOE's program, and results from Swedish investigations of a copper waste container, the Commission is confident that, given a range of waste forms and conserva live test conditions, the technology is available to design acceptable waste packages. In addition to the materials testing for the waste container and waste form, there may be additional measures that can be taken to improve the effectiveness of the engineered barriers. It is known, for example, that the loading characteristics of the wastes diminish with time. Also, the longer wastes are stored before disposal, the smaller will be the quantities of radionuclides available for transport to the accessible environment. It is also technically feasible to separate from radioactive wastes the radionuclides that constitute the principal source of heat from the nuclides of greatest 1ong-term concern. The former radionuclides, mainly fission products such as cesium-137 and strontium-90, could then be stored for a period of years while the fission products decay to the point where they could be disposed of either in a manner that does not require the degree of confinement provided by a geologic repository, or in a repository with less concern for thermal disturbance of the host rock's expected waste isolation properties. Meantime, the longer-lived remaining radionuclides, such as transuranic wastes with elements heavier than uranium, could be disposed of in a repository away from the fission products and without the high thermal loadings that would otherwise have to be considered in predicting the term waste isolation performance of the geologic setting. France, Great Britain, and Japan are currently pursuing this waste management strategy or a variant of it. The Commission emphasizes here that it does not believe that recycling technolgies are required for the safety or feasibility of deep geologic disposal in the United States. Other countries, such as Canada, the Federal Republic of Germany, and Sweden are pursuing disposal strategies based on a similar view. Reprocessing, if employed in its current stage of development, would result in additional exposures to radiation and volumes of radioactive wastes to be disposed of. For the purpose of finding reasonable assurance in the technical feasibility of geologic disposal, however, it is worth noting that technology is currently available to permit additional engineering control of waste forms if, for reasons not now foreseen, such control were deemed desirable at some future time. Meanwhile, the Commission continues to have confidence that safe geologic disposal is technically feasible for both spent fuel and high-level waste. DOE's current reference design for the waste package does not include backfill or packing around waste containers in the emplacement boreholes. Neither is required under NRC rules so long as DOE can show that applicable regulatory criteria and objectives will be met. An air gap between the container and the host rock is currently one of the barriers in DOE's design for meeting the performance objective. DOE has conducted investigations on a variety of 51-SC-12 candidate materials for backfill in a variety of geologic media, and the Commission finds no basis to qualify its past confidence that backfill materials can be developed, if needed, to meet applicable NRC requirements. The current reference design for sealing boreholes, shafts, ramps and the underground facility at the Yucca Mountain candidate site employs crushed luff and cement. Regardless of the geologic medium of the candidate site, DOE will have to show that the license application design meets NRC post-closure performance objectives. The Commission continues to have reasonable assurance that DOE's program will lead to identification of acceptable sealant materials for meeting these objectives. Overall, from its reexamination of issues related to the technical feasibility of geologic disposal, the Commission concludes that there is reasonable assurance that safe disposal of level waste and spent fuel in a mined geologic repository is technically feasible. Original Finding 2 The Commission finds reasonable assurance that one or more mined geologic repositories for commercial high-level waste and spent fuel will be available by the years 2007-2009, and that sufficient repository capacity will be available within 30 years beyond expiration of any reactor operating license to dispose of existing commercial high-level radioactive waste and spent fueld originating in that reactor and generated up to that time. Proposed Finding 2 The Commission finds reasonable assurance that at least one mined geologic repository will be available within the first quarter of the first century, and that sufficient repository capacity will be available within 30 years beyond the licensed life for operation of any reactor to dipose of the commercial high-level radioactive waste and spent fuel originating in such reactor and generated up to that time. 2.A. Issues Considered in Commission's 1984 Decision on Finding 2 2.A.1. Finding Technically Acceptable Sites in a Timely Fashion In order for the Commission to find that any candidate site for a repository is technically acceptable (that is, in compliance with NRC licensing requirements], the site must undergo comprehensive site characterization to assess its hydrologic, geologic, geochemical, and rock mechanics PART 51 STATEMENTS OF CONSIDERATION properties. It is possible that a site may be found unacceptable on the basis of early in-situ testing or other site characterization activities. It will not be possible, however, for the NRC staff to take a position before a licensing board that a site will meet NRC requirements for construction authorization until the results of all site characterization activities are available. Even then, the staff may conclude that the evidence from site characterization does not constitute reasonable assurance that NRC performance objectives will be met. Also, the results of the licensing hearings on construction authorization cannot be predicted. If construction is authorized and when it is substantially complete, DOE is required to obtain, in addition to the construction authorization permit, a*license to receive and possess waste at the geologic repository operations area in order to commence repository qperations. These considerations argue for maintaining the ready availability of alternatives sites if, after several years, site characterization or licensing activities bring to light difficulties at the leading candidate site. In support of its argument on technical feasibility, the Commission stated in its 1984 Waste Confidence Decision that "* *

DOE's program is providing information on site characteristics at a sufficiently large number and variety of sites and geologic media to support the expectation that one or more technically acceptable sites will be identified." At the time, DOE was required under the NWPA to characterize three candidate repository sites. The NWPAA had a major impact on DOE's repository program, however. Under the NWPAA, DOE was required to suspend site-specific activities at the Hanford, WA and Deaf Smith County, TX sites, which had been approved by the President for site characterization for the first repository.

Redirection of the repository program to single-site characterization (or, if necessary, sequential site characterization if the Yucca Mountain site is found to be unsuitable] will permit DOE to concentrate its efforts and resources on information gathering at a single site, as opposed to spreading out its efforts over a range of sites. The possible schedular benefits to single-site characterization, however, must be weighed for the purposes of this Finding against the potential for additional delays in repository availability if the Yucca Mountain site is found to be unsuitable. By focusing DOE site characterization activities on Yucca Mountain, the NWPAA has essentially made it necessary for that site to be found suitable if the 2007-2009 timeframe for repository availability in the Commission's 1984 Decision is to be met. Clearly, the Commission cannot be certain at this time that the Yucca Mountain site will be acceptable. Although Commission has no reason to believe that another technically acceptable site can not be found if the Yucca Mountain site proves unsuitable, several factors raise reasonable doubts as to the availability of even one repository by 2007-2009. These include: (1) The current reliance on a single site with no concurrently available alternatives; (2) the probability that site characterization activities will not

proceed entirely without problems; and (3) the history of schedular slippages since passage of the NWPA. For example, DOE's schedule for the first repository slipped five years (from 1998 to 2003) between January 1983, when the NWPA was enacted, and January 1987, when the first Draft Mission Plan Amendment was issued. The schedule for excavation of the exploratory shaft for the Yucca Mountain site slipped by more than three years since the issuance of the PDS in March 1986. DOE has cited numerous reasons for past program slippages, including the need for a consultation process with States and Tribes, Congressional actions (e.g., the barring of funds in the 1987 budget appropriation for drilling exploratory shafts], and DOE's recognition that the EIS and license application would require more technical information than previously planned. ' Given this history of delays, and given its understanding of current developments, the Commission can not be sure that current milestones for the repository program will be met, at least in the foreseeable future. For example, DOE has taken the position, with which NRC agrees, that sinking of exploratory shafts should not occur before it has a qualified quality assurance (QA] program in place. The Commission believes that the aggressive, oriented schedule for this milestone has not allowed for unexpected developments.

Indeed, the effort to develop an approvable QA program has in itself identified problems in design control and other processes that must be resolved in order to establish a qualified program that addresses all applicable NRC licensing requirements. Thus, although the NWPAA is a clear and strong reaffirmation of Congressional support for the timely development of a repository, the Commission in this Waste Confidence review cannot ignore the potential for delay in repository availability if the 51-SC-13 Yucca Mountain site, or any other single si le designated for site characterization, is found to be unsuitable. Without alternative sites undergoing simultaneous characterization or even surface-based testing, DOE will have to begin characterizing another site if the site currently selected for characterization proves unsuitable. The earlier a determination of unsuitability can be made, the smaller the impact of such a finding would be on the overall timing of repository availability. DOE has estimated conservatively that it would required approximately 25 years to begin site screening for a second repository, perform site characterization, submit an EIS and license applications, and await authorizations before the repository could be ready to receive waste. In its June 1987 Mission Plan amendment, DOE stated "It * *

seems prudent to plan that site-specific screening leading to the identification of potentially acceptable sites should start about 25 years before the start of waste acceptance for disposal." DOE went on to say that it considered this estimate to be conservative because it does not account for expected schedular benefits from the first repository program, including improvements in such areas as site screening, site characterization, and performance assessment techniques.

Although DOE's estimate was permitted on the successful completion of a program for the first of two repositories, schedular benefits from improvements in the understanding of waste isolation processes would still be available. The glass waste form from the Defense Waste Processing Facility now under construction at Savannah River, SC, for example, will be available for testing under simulated repository conditions well before the turn of the century under current DOE schedules, and improvements in the modelling of spent fuel behavior within waste canisters can be applied in performance assessments largely irrespective of the geology of a site. It may also be pertinent that when DOE made its 25-year estimate for the second repository program in mid-1987, the law at the time required the simultaneous characterization of three sites, so that DOE could not proceed to develop one site for a repository until the completion of characterization at the site that required the most time. Although it is still possible for a repository to be available by 2007-2009 if the current schedule does not incur major additional delays, the Commission does not believe it would be prudent to reaffirm the Agency's 1984 PART 51 STATEMENTS OF CONSIDERATION finding of reasonable assurance that the 2007-2009 timetable will be met. As the Court of Appeals noted in remanding this isue to NRC, the ultimate determination of whether a disposal facility will be available when needed "* *

can never rise above a prediction." The Commission is in the position of having to reach a definitive finding on events which are almost two decades away. We believe that the institutional timescale for this question can more realistically be framed in decades than in years. As the program proceeds into the next century, it will become easier for NRC to make more definitive assessments, if necessary, of the time a repository will be available.

It should be noted here that the basis for the 2007-2009 timeframe in the Court remand on the "Waste Confidence" issues has changed in the past five years. These dates no longer represent the expected dates of expiration of the Vermont Yankee and Prairie Island facilities. When the operating licenses were originally issued for nuclear power reactors, license durations were computed on the basis of a 40-year operating lifetime starting from the date of the construction permit (CP) for the facility. For many facilities, five years or more elapsed from the date of issuance of the CP until issuance of the operating license (OL). In response to requests from utilities, the NRC staff has agreed to extend the dates of expiration of the OLs by computing the 40-year period of the license from the date of issuance of the OL instead of from the date of the CP. The NRC staff has already changed the expiration date for Prairie Island Units 1 and 2 from the year 2008 to the years 2013 and 2014. The staff currently expects Vermont Yankee to request a change in its current expiration date of December 11, 2007. On the basis of the date of issuance of the OL for Vermont Yankee, it is eligible for extension of its operating license expiration to March 2012. Therefore, if the remand were to occur today, NRC would likely be evaluating the availability of a repository by 2012-2014, as these years are expected to represent the timeframe in which the OLs of the Vermont Yankee and Prairie Island facilities are due to expire. In light of all these considerations, the Commission believes it can have reasonable assurance that at least one repository will be available within the first quarter of the twenty-first century. This estimate is based on the time it would take for DOE to proceed from site screening to repository operation at a site other than Yucca Mountain, if this should prove necessary. Assuming for the sake of conservatism that Yucca Mountain would not be found suitable for repository development, it is reasonable to expect that DOE would be able to reach this conclusion by the year 2000. This would leave 25 years for the attainment of repository operations at another site. 2.A.2. Timely Development of Waste Packages and Engineered Barriers DOE's current conceptual design for the waste package is discussed in the SCP for the Yucca Mountain site. As information is obtained from site characterization activities and laboratory studies, the conceptual design will evolve in successive stages into the Advanced Conceptual Design (ACDJ, the LAD, and the final procurement and construction design. DOE has identified four areas of investigation related to the waste package LAD: (1) Waste package environment; (2) waste form and materials testing; (3) design, analysis, fabrication, and prototype testing; and (4) performance assessment. Numerous uncertainties exist in each of these areas. DOE's testing program will attempt to reduce uncertainties in these areas where possible. For example, situ testing is expected to decrease significantly uncertainties regarding the repository host rock mass in which the waste packages will be emplaced. In the area of performance assessment, however, where results of relatively short-term testing of complex waste-ground water interactions must be extrapolated over as many as 10,000 years, it may be necessary to rely more heavily on the use of simplifying assumptions and bounding conditions than in other areas of investigation. As discussed under Finding 1, the Commission continues to have reasonable assurance that waste packages and engineered barriers can be developed which will contribute to meeting NRC performanfe objectives for the repository. The timing of availability of a complete and high quality waste package and engineered barrier LAD, specifically their availability on a schedule which would permit repository operation by 2007-2009, is more difficult to assess at this time. In contrast with the technical feasibility issues discussed under Finding 1, development of acceptable waste packages and engineered barriers for a repository in the 2007-2009 timeframe does depend on the overall acceptability of the Yucca site. If the site is found to be unsuitable, waste package and _engineered barrier development will have to begin for a different site, because, under the NWPAA, DOE may not carry out site 51-SC-14 characterization and waste package development work at sites other than the Yucca Mountain site. Although much of the work related to waste form, materials, and performance assessment for the waste package can proceed independently of in-situ testing, the investigations related to waste package environment depend on the schedule for this testing. DOE's current schedule calls for completing the ACD for the waste package in 1992, and the waste package LAD in 1994. The ability to meet these dates will depend on whether DOE is able to resolve outstanding QA issues which have impeded shaft sinking and in-situ testing. In sum, the Commission is not aware of any scientific or technical problems so difficult as to preclude development of a waste package and engineered barrier for a repository at Yucca Mountain to be available within the first quarter of the twenty-first century. Moreover, even given the uncertainty regarding the ultimate finding of site acceptability, and the uncertainty concerning the range of site-related parameters for which the engineered facility and waste package will have to be designed, the Commission finds reasonable assurance that waste package and engineered barrier development can be completed on a schedule that would permit repository operation within the first quarter of the twenty-first century. If necessary (that is, if Yucca Mountain were found unsuitable late in the program), DOE could initiate site characterization and develop waste packages and engineered barriers at another site or sites and still commence operation before the end of the first quarter of that century. 2.A.3. Institutional Uncertainties 2.A.3.a. Measures for dealing with Federal-State-local concerns. In its 1984 Waste Confidence Decision, the Commission found that the NWPA should help to minimize the potential that differences between the Federal Government and States and Indian tribes will substantially disrupt or delay the respository program. The Commission noted that the NWPA reduced uncertainties regarding the role of affected States and tribes in repository site selection and evaluation. The Commission also said that the decision-making process set up by the NWPA provides a detailed, step-by-step approach that builds in regulatory involvement, which should also provide confidence to States and tribes that the program will proceed on a technically sound and acceptable basis. Despite the PART 51 STATEMENTS OF CONSIDERATION expected and continuing State opposition to DOE siting activities, the Commission has found no institutional developments since that time that would fundamentally disturb its 1984 conclusions on this point. NRC regulatory involvement, for example, has indeed been built into the process. DOE has continued its interactions with NRC regarding repository program activities since the Commission's 1984 Waste Confidence decision was issued. NRC provided comments to DOE on major program documents such as the Siting Guidelines and the PDS as required by the NWPA, and NRC concurred on those documents. NRC also reviewed and provided comments to DOE on the DEAs and FEAs. In the December 22, 1986 letter to DOE on the FEAs, the NRC staff noted that"* *

significant efforts were made by DOE to respond to each of the NRC staff major comments on the DEAs, and in fact, many of these comments have been resolved." NRC provided comments to DOE on the 1987 Draft Mission Plan Amendment, and DOE responded to most of these comments in the Final Mission Plan Amendment provided to Congress on June 9, 1987. Since enactment of the NWPAA in December 1987, DOE-NRC interactions have focused on the Yucca Mountain site. In January 1988, DOE issued the Consultation Draft Site Characterization Plan (CDSCPJ for the Yucca Mountain site. The NRC staff provided comments in the form of draft and final "point papers" on the CDSCP. The NRC comments included several objections related to: (1) The failure to recognize the range of alternative conceptual models of the Yucca Mountain site; (2) the status of the quality assurance (QA) plans for site characterization activities; and (3) concerns related to the exploratory shaft facility.

Although the December 1988 SCP shows improvement over the CDSCP, NRC continues to have an objection involving the need for implementing a baselined QA program before beginning site characterization and an objection involving the need for DOE to demonstrate the adequacy of both the ESF design and the design control process. DOE is committed to having a qualified QA program in place before sinking the exploratory shaft at the Yucca Mountain site. DOE has also taken measures to clarify and institutionalize the roles of other Federal agencies in addition to NRC. In the Draft 1988 Mission Plan Amendment, DOE described interactions with these agencies. DOE has a Memorandum of Understanding (MOU) with the Mine Safety and Health Administration of the Department of Labor for technical support and oversight for shaft construction and other site characterization activities, and with the Department of Transportation to define the respective responsibilities of the two agencies in the waste disposal program. DOE also has interagency agreements with the Bureau of Mines and the U.S. Geological Survey of the Department of the Interior. DOE's efforts to address the concerns of States, local governments, and Indian tribes have met with mixed results. For example, DOE has not succeeded in finalizing any consultation and cooperation (C&CJ agreements as required under Section 117(c) of the NWPA, as amended. These agreements were to help resolve State and Tribal concerns about public health and safety, environmental, and economic impacts of a repository. Publication of the Siting Guidelines under section 112(a) of the NWPA resulted in numerous lawsuits challenging the validity of the Guidelines. Similarly, the FEAs were challenged in the Ninth Circuit by affected States and tribes. The NWPAA did not curtail financial assistance to affected States and tribes, except to redefine and redistribute it if DOE and a State or tribe enter into a benefits agreement. The State of Nevada and affected local governments are currently receiving financial assistance. DOE has attempted to negotiate an agreement with the State of Nevada for monetary benefits under section 170 of the NWPAA. This section would provide for payments of $10 million per year before receipt of spent fuel. and $20 million per year after receipt of spent fuel until closure of the repository. These payments would be in addition to certain monetary benefits for which the State is eligible under the NWPA, as amended. Also under a benefits agreement, a Review PaRel would be constituted for the purpose of advising DOE on matters related to the repository, and for assisting in the presentation of State, tribal, and local perspectives to DOE. The beneficiary to a benefits agreement must waive its right to disapprove the recommendation of the site for a repository and its rights to certain impact assistance under

sections 116 and 118 of the NWPA, as amended. To date, the State of Nevada has declined DOE's offer to negotiate a benefits agreement.

The NWPAA introduced several new organizational entities to the repository program with responsibilities that may contribute to resolving concerns of Federal, State, and local governments involved in the program. Under section 51-SC-15 503 of the NWPAA, the Nuclear Waste Technical Review Board (NWTRBJ is to evaluate the technical and scientific validity of DOE activities under the NWPAA, including site characterization and activities related to packaging or transportation of spent fuel. The NWPAA also established the Office of Nuclear Waste Negotiator, who is to seek to negotiate terms under which a State or Indian tribe would be willing to host a repository or MRS facility at a technically qualified site. Among the duties of the Negotiator is consultation with Federal agencies such as NRC on the suitability of any potential site for site characteriza lion. At the time of this writing, the President has not appointed the Negotiator. On February 24, 1989 Congressman Morris K. Udall and Senator J. Bennett Johnston requested that the President take action to appoint an individual to this office. A Negotiator could contribute to the timely success of the respository program by providing an alternative site to the Yucca Mountain site that would still have to be techically acceptable, but that would enjoy the advantage of reduced institutional uncertainties resulting from opposition to Sta le or affected Indian tribes. An additional measure which may facilitate documentation and communication of concerns related to a repository is the Licensing Support System (LSSJ. The LSS is to provide full text search capability of and easy access to documents related to the licensing of the repository. Although the primary purpose of the LSS is to expedite NRC's review of the construction authorization application for a repository, it will be an effective mechanism by which all LSS participants, including the State and local governments, can acquire early access to documents relevant to a repository licensing decision. DOE has the responsibility for designing the LSS and bearing the costs associated with it, and NRC will be responsible for implementing it. Procedures for the use of the LSS are part of revisions to 10 CFR Part 2, NRC's Rules of Practice for the adjudicatory proceeding on the application to receive and possess waste at a repository. These revisions were the result of a "negotiated rulemaking" process in which affected parties meet to reach concensus on the proposed rule. The members of the negotiating committee included: DOE; NRC; State of Nevada; coalition of Nevada local governments; coalition of industry groups; and a coalition of national environmental groups. The coalition of industry groups PART 51 STATEMENTS OF CONSIDERATION dissented on the final text of the proposed rule, but the negotiating process enabled NRC to produce a proposed rule reflecting the consensus of most of the interested parties on an important repository licensing issue. NRC is committed to safe disposal of radioactive waste and the protection of public health and safety and the environment. Any State with a candidate site for a repository should be assured that a repository will not be licensed if it does not meet NRC criteria. NRC has its own program for interaction with the State of Nevada and affected units of local government, and will continue to provide information to Nevada and consider State concerns as requested. Given the difficult nature of sit_ing a repository, the Commission believes that the NWPA, as amended, has achieved the proper balance between providing for participation by affected parties and providing for the exercise of Congressional authority to carry out the national program for waste disposal. The NWPAA provides adequate opportunity for interaction between DOE and other Federal agencies, States, tribes, and local governments such that concerns can be presented to DOE for appropriate action. Both the NRC and the State or tribe can exercise considerable prerogative regarding repository development. The State or tribe may disapprove the recommendation that the site undergo repository development. This disapproval can be overridden only by vote of both houses of Congress within 90 days of continuous session. If the State disapproval is overridden, DOE may submit an application for authorization to construct the repository, and, if approved, a subsequent application to receive and possess waste for emplacement. NRC will make decisions on the license applications according to the requirements of its statutory mission. Despite the complexity of the overall process and the strong views of the participants in it, the Commission sees no compelling reason to conclude that current institutional arrangements are inadequate to the task of resolving State, Federal, and local concerns in time to permit a repository to be available within the first quarter of the twenty-first century. 2.A.3.b. Continuity of the management of the waste program. At the time the Commission issued its 1984 Waste Confidence Decision, the possibility that DOE functions would be transferred to another Federal agency was cited as the basis for concerns that the resolution of the radioactive waste disposal problem would likely undergo further delays. The Commission responded that in the years since the Administration had proposed to dismantle DOE in September 1981, Congress had not aQted on the proposal. The Commission further stated that even if DOE were abolished, the nuclear waste program would simply be transferred to another agency. The Commission did not view the potential transfer in program management as resulting in a significant loss of momentum in the waste program. The commission also concluded that the enactment of the NWPA, which gave DOE lead responsibility for repository development, further reduced uncertainties as to the continuity of management of the waste program. Section 303 of the NWPA did, however, require the Secretary of Energy to "* *

undertake a study with respect to alternative approaches to managing the construction and operation of all civilian radioactive waste facilities, including the feasibility of establishing a private corporation for such purpose." To carry out this requirement, DOE established the Advisory Panel on Alternative Means of Financing and Managing Radioactive Waste Facilities, which came to be known as the "AMFM" Panel. The Panel's final report, issued in December 1984, concluded that several organizational forms are more suited than DOE for managing the waste program, including an independent Federal agency or commission, a public corporation, and a private corporation.

The report identified a public corporation as the preferred alternative on the basis of criteria developed by the Panel for an acceptable waste management organization. In particular, the report indicated that a public corporation would be stable, highly mission-oriented, able to maintain credibility with stakeholders, and more responsive to regulatory control than a Federal executive agency. Commenting on the AMFM Panel's report in April 1985, DOE recommended retaining the present management structure of the waste program at least through the siting and licensing phase of the program. Congress did not take action to implement the Panel's recommendations, and DOE's management of the waste program has remained uninterrupted. By enacting the NWPAA, Congress effectively reaffirmed DOE's continued management of the waste program. Congress did not revise DOE's role as the lead agency responsible for development of a repository and an 51-SC-16 MRS. Congress did establish several new entities for the purpose of advising DOE on matters related to the waste program, such as the NWTRB and the Review Panel, to be established if DOE and a State or tribe enter into a benefits agreement under section 170 of the NWPAA. Congress provided further indication of its intent that DOE maintain management control of the waste program for the foreseeable future in requiring, under section 161, that the Secretary* of DOE "* *

report to the President and to Congress on or after January 1, 2007, but not later than January 1, 2010, on the need for a second repository." This is not to say, however, that there have been no management problems in the DOE program. Since the enactment of the NWPA in 1983, only one of the five Directors of DOE's Office of Civilian Radioactive Waste Management (OCRWMJ has held the position on a permanent basis. Inadequate progress toward an operating repository has concerned several Congressional observers, including Senator J. Bennett Johnston, Chairman of the Senate Energy and Natural Resources Committee.

In February 1989 confirmation hearings for then-Secretary-of-Energy-designate James Watkins, Senator Johnston strongly criticized mounting cost projections and lack of progress in the program, and called for new and stronger management. Whether the management structure of the respository development program should in fact be changed is a decision best left to others. The Commission believes that a finding on the likely availability of a repository should take management problems into account, but finds no basis to diminish the degree of assurance in its 1984 conclusion on this issue. Events since the submission of the AMFM Panel report do not indicate that there will be a fundamental change in the continuity of the management structure of the program any time soon. In addition, it cannot be assumed that the program would encounter. signficantly less difficulty with a new management structure than it would continuing under the present one. Under either scenario, however, the Commission believes it would be more prudeIJt to expect respository operations after the 2007-2009 timeframe than before it. Neither the problems of a new management structure nor those of the existing one are likely to prevent the achievement of repository operations within the first quarter of the next century, however. PART 51 STATEMENTS OF CONSIDERATION 2.A.3.c. Continued funding of the nuclear waste management program. Section 302 of the NWPA authorized DOE to enter into contracts with generators of electricity from nuclear reactors for payment of 1.0 mill (0.1 cent) per kilowatt-hour of net electricity generated in exchange for a Federal Government commitment to take title to the spent fuel from those reactors. In the 1984 Waste Confidence Decision, the Commission noted that all such contracts with utilities had been executed. After the 1984 Decision, President Reagan decided that defense high-level wastes are to be collocated with civilian wastes from commercial nuclear power reactors. DOE's Office of Defense Programs is to pay the full cost of disposal of defense waste in the repository. DOE is required under section 302(a)(4) of the NWPA, as amended, "* *

annually [to] review the amount of the fees * *
to evaluate whether collection of the fees will provide sufficient revenues to offset the costs * * *." In the June 1987 Nuclear Waste Fund Fee Adequacy Report, DOE recommended that the 1.0 mill per kilowatt-hour fee remain unchanged.

This assessment was based on the assumption that an MRS facility would open in 1998, the first repository would open in 2003, and the second repository in 2023. These assumptions do not reflect changes in the waste program brought about by the NWPAA enacted in December 1987. Two such changes with significant potential impacts were the suspension of site-specific activities related to the second repository until at least 2007, and the linkage between MRS construction and operation and the granting of a repository construction authorization, which will probably occur no earlier than 1998. According to the Draft 1988 Mission Plan Amendment, DOE should currently be preparing the 1988 fee-adequacy analysis on the basis of the changes to the waste program brought about by the NWPAA. The new fee adequacy report will reflect overall program cost savings to the utilities resulting from: (1) Limiting site characterization activities to a single site at Yucca Mountain, NV; and (2) the DOE Office of Defense Programs' sharing other program costs with generators of electricity"*

on the basis of numbers of waste canisters handled, the portion of the repository used for civilian or defense wastes, and the use of various facilities at the repository," in addition to paying for activities solely for disposing of defense wastes. An additional factor which may eventually also contribute to the overall adequacy of Nuclear Waste Fund fees is the likelihood that a significant number of utilities will request renewals of reactor operating lifetimes beyond their current OL expiration dates. OL renewal would provide additional time during which Nuclear Waste Fund fees could be adjusted, if necessary, to cover any future increase in per-unit costs of waste management and disposal.

The Commission recognizes the potential for program cost increases over estimates in the 1987 Nuclear Waste Fund Fee Adequacy Report. If there is a significant delay in repository construction, for example, it is reasonable to assume that construction costs will escalate. There may also be additional costs associated with reactor dry cask storage of spent fuel, if DOE does not have a facility available to begin accepting spent fuel by the 1998 date specified in the NWPA. These costs would be further increased if one or more licensees were to become insolvent and DOE were required to assume responsibility for storage at affected reactors before 1998. The full impact of the program redirection resulting from the NWPAA and the outlook for the timing of repository availability will continue to be assessed annually. If it does appear that costs will exceed available funds, there is provision in the NWPA for DOE to request that Congress adjust the fee to ensure full-cost recovery. Thus, the Commission finds no reason for changing its basic conclusion that the long-term funding provisions of the Act should provide adequate financial support for the DOE program. 2.A.3.d. DOE's schedule for repository development. At the time that the 1984 Waste Confidence Decision was issued, the Nuclear Waste Policy Act of 1982, enacted in January 1983, had been in effect for less than 20 moriths. The NWPA had established nun\erous deadlines for various repositdry program milestones. Under section 112(b)(l)(BJ, the NWPA set the schedule for recommendation of sites for characterization no later than January 1, 1985. Section 114(a)(2) specified that no later than March 31, 1987, with provision for a 12-month extension of this deadline, the President was to recommend to Congress one of the three characterized sites qualified for an application for respository construction authorization. Under section 114(d), NRC was to issue its decision approving or disapproving the issuance of a construction authorization not later than January 1, 1989, or the expiration of three years after the date of submission of the application, whichever occurs 51-SC-17 later. Section 302(a)(5)(B) required that contracts between DOE and utilities for payments to the Waste Fund provide that DOE will begin disposing of spent fuel or high-level waste by January 31, 1998. In little more than a year after enactment, the schedule established by the NWPA began proving to be optimistic. In the reference schedule for the repository presented in the April 1984 Draft Mission Plan, for example, DOE showed a slip from January 1989 to August 1993 for the decision on construction authorization. In the 1984 Waste Confidence Decision, the Commission recognized the possibility of delay in respository availability beyond 1998, and did not define its task as finding confidence that a repository would be available by the 1998 milestone in the NWPA. The Commission focused instead on the question of whether a repository would be available by the years 2007-2009, the date cited in the court remand as the expiration of the OLs for the Vermont Yankee and Prairie Island reactors. The NRC believed that the NWPA increased the chances for repository availability within the first few years of the first century, by specifying the means for resolving the institutional and technical issues most likely to delay repository completion, by establishing the process for compliance with NEPA, and by setting requirements for Federal agencies to cooperate with DOE in meeting program milestones. Finding that no fundamental technical breakthroughs were necessary for the repository program, the Commission predicted that "* *

selection and characterization of suitable sites and construction of repositories will be accomplished within the general time frame established by the Act [1998] or within a few years thereafter." In January 1987, DOE issued a Draft Mission Plan Amendment to apprise Congress of significant developments and proposed changes in the repository program. In the Draft Amendment, DOE announced a five-year delay in its schedule for repository availability from the first quarter of 1998 to the first quarter of 2003. DOE's reasons for the delay included the need for more time for consultation and interaction with States and Tribes, the requirement in DOE's 1987 budget the funds not be used for drilling exploratory shafts in 1987, and the need for more information than previously planned for site selection and the license application.

The 1987 Draft Mission Plan Amendment set the second quarter of 1988 as the new date for exploratory shaft construction at the PART 51 STATEMENTS OF CONSIDERATION Yucca Mountain Site. When the final 1987 Mission Plan Amendment was submitted to Congress in June 1987, the schedule for shaft sinking at the Yucca Mountain site had slipped six months to the fourth quarter of 1988. Congress did not take action to approve the June 1987 Mission Plan Amendment as DOE had requested. On December 22, 1987, the NWPAA was enacted. The NWPAA has its major impact on the respository program in suspending site characterization activities at the Hanford and Deaf Smith County sites and authorizing DOE to characterize the Yucca Mountain site for development of the first repository. DOE subsequently issued the Draft 1988 Mission Plan Amendment in June 1988, to appraise Congress of its plans for implementing the provisions of the NWPAA. In the Draft 1988 Mission Plan Amendment, DOE's schedule for shaft sinking at Yucca Mountain had slipped another six months to the second quarter of 1989. At this writing, the schedule for shaft sinking is November 1989, but NRC and DOE have agreed that DOE Must first have a qualified QA program in place. DOE efforts to date to qualify its QA program have revealed issues requiring DOE attention before shaft excavation can begin, and it is possible that additional issues affecting DOE's readiness will come to light. Realistically, as the date for shaft sinking slips, the date for repository operation must be adjusted to reflect this slip. This might not be the case if the original schedule had provided for periods of tirQe between critical milestones that could absorb delays without affecting the schedule for repository operation. This is not the case with the schedule for the repository. The repository schedule has always been aggressive and highly success-oriented. In comments on the Draft 1988 Mission Plan Amendment, the Commission noted that the schedule has not allowed adequately for contingencies, and that, given the compression in the schedule for near-term program milestones, DOE has not shown how it will be able to meet the 2003 milestone for repository operation. Another potential source of delay in repository availability may arise from NRC regulations. The Commission believes that current NRC rules are fully adequate to permit DOE to proceed to develop and submit a repository license application, but further clarification of these rules is desirable to reduce the time needed to conduct the licensing proceeding itself. In order to meet the three-year schedule provided in the NWPA for a Commission decision on repository construction authorization, the NRC staff has undertaken to refine its regulatory framework on a schedule that would still permit DOE to prepare and submit an application for repository construction authorization under its current schedule. The Commission fully expects to avoid delaying DOE's program, while working to reduce the uncertainties in NRC regulatory requirements that could become contentions in the licensing proceeding. Even if there are any delays resulting from a need for DOE to accommodate more specific regulatory requirements in its site characterization or waste package development programs, however, the Commission is confident that the time savings in the licensing proceeding will more than compensate for them. In view of the delays in exploratory shaft excavation since the 2003 date for repository availability was set, it may be optimistic to expect that Phase 1 of repository operations will be able to begin by 2203. As DOE's schedule for repository availability as slipped a year and half since the date was changed from 1998 to 2003, the earliest date for repository availability would probably be closer to 2005. An institutional issue that may further affect DOE's schedule is the status of EPA standards for disposal of spent fuel and high-level waste. These standards are required under section 121(a) of the NWPA. Under 10 CFR 60.112, NRC's overall postclosure system performance objective, the geologic setting shall be selected and the engineered barrier system, which includes the waste package, must be designed to assure that releases of radioactive materials to the accessible environment, following permanent closure, conform to EPA's standards. 40 CFR part 191, the EPA standards, first became effective in November 1985. In July 1987, the U.S. Court of Appeals for the First Circuit vacated and remanded to EPA for further proceedings subpart B of the high-level radioactive waste disposal standards. As noted under the aforementioned 1.A.1., !lie standards have not been reissued.

A significant modification in the reissued EPA standard may affect the schedule for completing the design of the waste package and engineered barrier to the extent that design testing is planned to demonstrate compliance with the standards.

DOE's current site characterization plans for demonstrating compliance with 40 CFR part 191 are based on the standards as promulgated in 1985. DOE is proceeding to carry out its testing program developed for the original EPA standards. DOE has stated that if the EPA standards are changed 51-SC-18 significantly when they are reissued, DOE will reevaluate the adequacy of its testing program. The Commission believes that DOE's approach is reasonable. Much of the information required to demonstrate compliance with the EPA standards is expected to remain the same regardless of the numerical level at which each standard is set. Considering the importance of developing the repository for waste disposal as early as safely practicable, it would be inappropriate for DOE to suspend work on development of engineered barriers pending reissuance of the standards, unless EPA had given clear indications of major changes in them. Another possiblity is that, regardless of any changes in the repromulgated EPA standards, they will be litigated in Federal court. Even if this proves to be the case, however, the Commission believes that any such litigation will still permit EPA to promulgate final standards well within the time needed to enable DOE to begin repository operations at any site within the first quarter of the twenty-first century. Given the current pace of the DOE program, and assuming that the QA program can be qualified and shaft excavation begun within the next year, the Commission finds it is still possible, though less likely, that a repository at Yucca Mountain will be available by 2007-2009. To the extent that the expiration of the OLs for Prairie Island and Vermont Yankee continue to be relevant in this proceeding, the Commission believes it is more likely that a repository will be available by the anticipated dates of extension of the OLs for those plants in 2012-2014. If DOE determines that the Yucca Mountain site is unsuitable, the Commission considers it reasonble to expect that DOE could make this determination by the year 2000 and have a repository at another site available within the first quarter of the next century. 2.B. Relevant Issues That Have Arisen Since the Commission's Original Decision 2.B.1. NRC stated in 9-14-87 correspondence to Sen. Breaux on pending nuclear waste legislation that under a program of single site characterization, "* *

there may be a greater potential for delay of ultimate operation of a repository than there is under the current regime where three sites will undergo at-depth characterization before a site is selected." To what extent does the NWPAA raise uncertainty about the identification of a technically acceptable PART 51 STATEMENTS OF CONSIDERATION site and potential delay in repository availability by limiting site characterization to a single candidate site [Yucca Mt.) and by raising the possibility that a negotiated agreement might influence repository site selection?

Does this uncertainty affect confidence in the availibility of a respository by 2007-2009? In providing comments to Congress on proposed amendments to the NWPA, NRC took the position that simultaneous site characterization of three sites, as required by the NWPA, was not necessary to protect public health and safety. NRC further stated that the adequacy of a site for construction authorization would ultimately be determined in a licensing proceeding, and that NRC would only license a site that satisfied NRC licensing requirements. As described next, the Commission believes that the NWPAA contains numerous provisions to ensure that a technically acceptable site will be identified. The NWPAA does not reduce the scope of site characterization activities that DOE is authorized to undertake. The Amendments Act establishes a Nuclear Waste Technical Review Board composed of individuals recommended by the National Academy of Sciences and appointed by the President to evaluate the scientific validity of DOE activities, including site characterization activities, and to report its findings at least semiannually to Congress and DOE. The Amendments Act also provides funding for technical assistance to States, tribes, and affected units of local government. Finally, section 160(1) of the NWPAA provides that "Nothing in this Act shall be construed to amend or otherwise detract from the licensing requirements of the NRC established in Title II of the Energy Reorganization Act of 1974 (42 U.S.C. 5841 et seq.)." In providing for these reviews and in reaffirming NRC's licensing authority, the NWPAA ensures that a candidate site for a repository must satisfy all NRC requirements and criteria for disposal of high-level radioactive wastes in licensed geologic repositories. Section 402 of the NWPAA establishes the Office of the Nuclear Waste Negotiator. The duty of the Negotiator is to attempt to find a State or tribe willing to host a repository or MRS at a technically qualified site. The Negotiator may solicit comments from NRC, or any other Federal agency, on the suitability of any potential site for site characterization. Section 403[d)(4) strengthens the Commission's confidence that a technically acceptable site will be identified by providing that DOE may construct a repository at a negotiated site only if authorized by NRC. Given these safeguards on selection of a technically acceptable site, the Commissioi;i does not consider that the possibility of a negotiated agreement reduces the likelihood of finding a technically qualified site. The Commission raised the concern as early as April 1987 that under a program of single-site characterization, there could be considerable delay while characterization was completed at another site or slate of sites if the initially chosen site were found inadequate. By terminating site characterization activities at alternative sites to the Yucca Mountain site, the NWPAA has had the effect of increasing the potential for delay in repository availability if the Yucca Mountain site proves unsuitable. The provision of the NWPAA for a Negotiator could reduce the uncertainty and associated delay in restarting the repository program by offering an alternate to the Yucca Mountain site; but at the time of this writing, a Negotiator has not been appointed. It should be noted here that the repository program redirection under the NWPAA does not, per se, have a significant impact on the Commission's assurance of repository availability by 2007-2009. The Commission's reservations about reaffirming this timeframe derive from other considerations, including delays in sinking shafts and the potential for other delays in meeting program milestones, that would have arisen without the NWPAA. The Amendments Act does, however, effectively make it necessary that Yucca Mountain be found suitable if the 2007-2009 timeframe is to be met; this target period would almost certainly be unachievable if DOE had to begin screening to characterize and license another site. Thus, confidence is repository availability by 2007-2009 implies confidence in the suitability of Yucca Mountain. The Commission does not want its findings here to constrain in any way its regulatory discretion in a licensing proceeding. The Commission has therefore concluded that even if the program were on schedule, it would be inappropriate to reaffirm the 2007-2009 timeframe in the 1984 Decision. 2.B.2. In the Draft 1988 Mission Plan Amendment, DOE stated that "* *

the date indicate that the Yucca Mountain site has the potential capacity to accept at least 70,000 MTHM [metric tons heavy metal equivalent]

of waste, but only after site characterization will 51-SC-19 it be possible to determine the total quantity of waste that could be accommodated at the site." a. Do the issues of limited spent fuel capacity at Yucca Mountain, indefinite suspension of the second repository program, and the likelihood that no more than one repository will be available by 2007-2009 undermine the NRC's 1984 assurance that "sufficient repository capacity will be available within 30 years beyond expiration of any reactor operating license to dispose of existing commercial high level radioactive waste and spent fuel originating in such reactor and generated up to that time?" b. Is there sufficient uncertainty in total spent fuel projections (e.g., from extension-of-life license amendments, renewal of operating licenses for an additional 20 to 30 years, or a new generation of reactor designs) that this Waste Confidence review should consider the institutional uncertainties arising from having to restart a second repository program? 2.B.2.a. Although it will not be possible to determine whether Yucca Mountain can accommodate 70,000 MTHM or more of spent fuel until after site characterization, the Commission does not believe that the question of repository capacity at the Yucca Mountain site should be a major factor in the analysis of Finding 2. This is because it cannot be assumed that Yucca Mountain will ultimately undergo development as a repository. The generic issue of repository capacity does add to the potential need for more than one repository, however. As noted earlier, the NWPA established deadlines for major milestones in the development of the first and the second repository programs. The Act also required NRC to issue a final decision on the construction authorization application by January 1, 1989 for the first repository, and January 1, 1992 for the second [or within three years of the date of submission of the applications, whichever occurred later). The July 1984 Draft DOE Mission Plan set January 1998 and October 2004 as the dates for commencement of waste emplacement in the first and second repositories, assuming that Congressional authorization was obtained to construct the second repository. Thus, at the time the 1984 Waste Confidence Decision was issued, DOE was authorized and directed to carry out two repository programs under a schedule to make both facilities operational by 2007-2009. DOE and NRC were also working under the constraint, . PART 51 STATEMENTS OF CONSIDERATION still in force under the NWPA as amended, that no more than 70,000 MTHM may be emplaced in the first repository before the second is in operation. Because DOE estimated at the time that commercial U.S. nuclear power plants with operating licenses or construction permits would discharge a total 160,000 MTHM of spent fuel, it appeared that at least two repositories would be needed. In the 1984 Waste Confidence Decision, reactors were assumed to have a 40-year operating lifetime, and because the earliest licenses were issued in 1959 and the early 1960's, the oldest plants' licenses were due to expire as early as 1999 and 2000, as discussed in more detail below. Although it was expected that at least one repository would be availablt, by this time, there was also a limit as to how quickly spent fuel could be accepted by th.e repository. DOE had estimated that waste acceptance rates of 3400 MTHM per year could be achieved after the completion of Phase 2 of the first repository. This rate could essentially double if two repositories were in operation. At 6000 MTHM/year, it was estimated that all the anticipated spent fuel could be emplaced in the two repositories by about the year 2026. This was the basis for the Commission's position that sufficient repository capacity would be available within 30 years beyond expiration of any reactor 01 to dispose of existing commercial high level waste and spent fuel originating in such reactor and generated up to that time. In May 1986, however, DOE announced an indefinite postponement of the second repository program. The reasons for the postponement included decreasing forecasts of spent fuel discharges, as well as estimates that a second repository would not be needed as soon as originally supposed. With enactment of the NWPAA in December 1987, DOE was required to terminate all site-specific activities with respect to a second repository unless such activities were specifically authorized and funded by Congress. The NWPAA required DOE to report to Congress on the need for a second repository on or after January 1, 2007, but not later than January 1, 2010. Current DOE spent fuel projections, based on the assumption of no new reactor orders, call for 87,000 MTHM to have been generated by the year 2036, including approximately 9000 MTHM of defense high-level waste. With the likelihood that there will be reactor lifetime extensions and renewals, however, the no-new-orders case probably underestimates total spent fuel discharges. Also, the NWPAA did not change the requirement that no more than 70,000 MTHM could be emplaced in the first repository before operation of the second. It therefore appears likely that two repositories will be needed to dispose of all the spent fuel and level waste from the current generation of reactors, unless Congress provides statutory relief from the 70,000 MTHM limit, and the first site has a de qua te capacity to hold all of the spent fuel and high-level waste generated. The Commission believes that if the need for an additional repository is established, Congress will provide the needed institutional support and funding, as it has for the first repository. For all but a few licensed nuclear power reactors, 01s will not expire until some time in the first three decades of the twenty-first century. Several utilities are currently planning to have their 01s renewed for ten to 30 years beyond the original license expiration. At these reactors, currently available spent fuel storage alternatives effectively remove storage capacity as a potential restriction for safe operations. For these reasons, a repository is not needed by 2007-2009 to provide disposal capacity within 30 years beyond expiration of most 01s. If work is begun on the second repository program in 2010, the repository could be available by 2035, ' according to DOE's estimate of 25 years for the time it will take to carry out a program for the second repository. Two repositories available in approximately 2025 and 2035, each with acceptance rates of 3400 MTHM/year within several years after commencement of operations, would provide assurance that sufficient repository capacity will be available within 30 years of 01 expiration for reactors to dispose of the spent fuel generated at their sites up to that time. There are several reactors, however, whose 01s have already expired or are due to expire within the next few years, and which are now licensed or will be licensed only to possess their spent fuel. If a repository is not available until about 2025, these reactors may be exceptions to the second part of the Commission's 1984 Finding 2, which was that sufficient repository capacity will be available within 30 years beyond the expiration of any reactor 01 to dispose of the commercial high-level waste and spent fuel originating in such reactor and generated up to that time. The basis for this second part of Finding 2 has two components: (1) A technical or hardware component: and (2) an institutional component. The 51-SC-20 technical component relates to the reliability of storage hardware and engineered structures to provide for the safe storage of spent fuel. An example would be the ability of spent fuel assemblies to withstand corrosion within spent fuel storage pools, or the ability of concrete structures to maintain their integrity over long periods. In the 1984 Decision, the Commission found confidence that available technology could in effect provide for safe storage of spent fuel for at least 70 years. The Commission's use of the expression "30 years beyond expiration of any reactor operating license" in the 1984 Finding was based on the understanding that the license expiration date referred to the scheduled expiration date at the time the license was issued. It was also based on the understanding that, in order to refuel the reactor, some spent fuel would be discharged from the reactor within twelve to eighteen months after the start of full power operation. Thus, the Commission understood that, depending on the date of the first reactor outage for refueling, some spent fuel would be stored at the reactor site for most of the 40-year term of the typical 01. In finding that spent fuel could be safely stored at any site for at least 30 years after expiration of the 01 for that reactor, the Commission indicated its expectation that the total duration of spent fuel storage at any reactor would be about 70 years. Taking the earliest licensed power reactor, the Dresden 1 facility licensed in 1959, and adding the full 40-year operating license duration for a scheduled license expiration in the year 1999, the Commission's finding would therefore entail removal of all spent fuel from that reactor to a repository within the succeeding 30 years, or by 2029. Even if a repository were not available until the end of the first quarter of the twenty-first century, DOE would have at least four years to ship the reactor's 683 spent fuel assemblies, totalling 70 metric tons initial heavy metal [MTIHMJ, from Dresden 1 without exceeding the Commission's 30-year estimate of the maximum time it would take to dispose of the spent fuel generated in that reactor up to the time its 01 expired. (MTIMH is a measure of the mass of the uranium in the fuel [or uranium and plutonium if it is a mixed oxide fuel] at the time the fuel is placed in the reactor for irradiation. J Considering the experience from the 1984 and 1985 campaigns to return spent fuel from the defunct West Valley reprocessing facility to the reactors of PART 51 STATEMENTS OF CONSIDERATION origin, 70 metric tons of BWR spent fuel can easily be shipped within four years. The first campaign, involving truck shipments of 20 metric tons from West Valley, NY, to Dresden 1 in Morris, IL, took eleven months. The second, involving truck shipments of 43 tons from West Valley to the Oyster Creek reactor in Toms River, NJ, took six months. (See Case Histories of West Valley Spent Fuel Shipments, Final Report, NUREG/CR-4847 WPR-86(6811)-1, p. 2-2.) This estimate assumes, moreover, that no new transportation casks, designed to ship larger quantities of older, cooler spent fuel, for example, would be available by 2025. The institutional part of the question concerning the availability of sufficient repository capacity required the Commission to make a finding as to whether spent fuel in at-reactor storage would be safely maintained after the expiration of the facility OL. This question related to the financial and managerial capability for continued safe storage and monitoring of spent fuel, rather than to the capability of the hardware involved. The Commission determined, in Finding 3 of its 1984 Decision, that spent fuel will be managed in a safe manner until sufficient repository capacity is available to assure safe disposal, which was expected under Finding 2 to be about 30 years after the expiration of any reactor OL. (See discussion of Finding 3 below for additional discussion of the institutional aspects of spent fuel storage pending the availability of sufficient disposal capacity.] The availability of a repository within the first quarter of the twenty-first century holds no significant adverse implications for the Commission's institutional concern that there be an organization with adequate will and wherewithal to provide continued term storage after reactor operation. This could be a concern if a significant number of reactors with significant quantities of spent fuel onsite were to discontinue operations indefinitely between now and 1995, and the owners of these reactors did not appear to have the resources to manage them safely for up to 30 years pending the assumed availability of a repository in 2025. No such development is likely. No licenses for currently operating commercial nuclear reactors are scheduled to expire until the year 2000, and most such licenses will expire during the first two decades after 2006. (See Nuclear Regulatory Commission 1989 Information Digest, NUREG-1350, Vol. 1, p. 33.) The availability of the first repository by 2025, and of a second repository within one or two decades thereafter, would provide adequate disposal capacity fqr timely removal of the spent fuel generated at these reactors. There are several licensees, however, whose authority to operate their commercial reactors has already been terminated. These are Indian Point 1, Dresden 1, Humboldt Bay, and Lacrosse. They are also the only licensed power reactors that are retired with spent fuel being stored onsite. Assuming conservatively that a repository does not become operational until 2025, it appears likely that spent fuel will remain at these sites for more than 30 years beyond the time their reactors were indefinitely shut down, at which point their operating licenses could be considered to have effectively expired, although they will continue to hold a possession license for the storage of the spent fuel. In considering the means and motivation of the owner of an indefinitely retired reactor to provide safe long-term storage, the Commission believes it is useful to distinguish between the owner with only one reactor, and the owner of a reactor at a multi-unit site or an owner with operating reactors at other sites. In the case of a retired reactor at a multi-unit site, the owner would have a clear need to maintain the safety of storage at the retired reactor sufficiently to permit continued generation at the site. If the owner of the retired reactor also owned other reactors at other sites, the spent fuel at the retired reactor could be transferred, if necessary, to the storage facilities of other units still under active management. Of the four reactors just cited, Indian Point 1 and Dresden 1 fit this description, and the sibling reactors at their sites are operating under licenses that do not expire until well beyond the year 2000-that is, well within the post-OL period during which the Commission has found that spent fuel could be safely stored pending the availability of a repository. For the Lacrosse and Humboldt Bay reactors, the Commission is confident that, even if a repository is not available within 30 years following their retirement, the overall safety and environmental acceptability of extended spent fuel storage will also be maintained for these exceptional cases. Because there will still be an NRG possession license for the spent fuel at these facilities, the Commission will retain ample regulatory authority to 51-SC-21 *require any measures, such as removal of the spent fuel remaining in storage pools to passive dry storage casks, that might become necessary until the time that DOE assumes title to the spent fuel under contracts pursuant to the NWPA. It should also be borne in mind that Humboldt Bay and Lacrosse are both small early reactors, and their combined spent fuel inventory totals 67 metric tons of initial heavy metal. (See Spent Fuel Storage Requirements (DOE/RL 88-34) October 1988, Table A.3b., pp. A.15-A.17.) If for any reason not now foreseen, this spent fuel can no longer be managed by the owners of these reactors, and DOE must assume responsibility for its management earlier than currently planned, this quantity of spent fuel is well within the capability of DOE to manage onsite or offsite with available technology financed by the utility either directly or through the Nuclear Waste Fund. Nor does the Commission see a significant. safety or environmental problem with premature retirements of additional reactors. In the Commission's original Waste Confidence Decision, it found reasonable assurance that spent fuel would have to spend no more than 30 years in post-operational storage pending the availability of a repository. For a repository conservatively assumed to be available in 2025, this expected 30-year maximum storage duration remains valid for most reactors, and would be true for all reactors that were prematurely retired after 1995. Based on the past history of premature shutdowns, the Commission has reason to believe that their likely incidence during the next six years will be small as a proportion of total reactor-years of operation. Historically, 14 of the 125 power reactors that have operated in the U.S. over the past 30 years have been retired before the expiration of their operating licenses. These early retirements included many low-power developmental reactors, which may make the ratio of 14 to 125 disproportionately high as a basis for projecting future premature shutdowns. The Commission is aware of currently operating reactors that may be retired before the expiration of their OLs. including: the recently-licensed Shoreham reactor, which has generated very little spent fuel; the Fort St. Vrain high-temperature gas-cooled reactor. which its owner plans to decommission; and the Rancho Seco reactor, which has operated for the past 12 years and may or may not be retired. Assuming that all these and perhaps a few more reactors do retire in the next several years, their PART 51 STATEMENTS OF CONSIDERATION total spent fuel storage requirements would not impose an unacceptable safety or environmental problem, even in the unlikely event that all these reactors' owners were rendered financially or otherwise unable to provide adequate care, and DOE were required to assume custody earlier than currently envisioned under the NWPA. Licensed non-power research reactors provide an even more manageable case. DOE owns the fuel for almost all of these reactors, many of which have been designed with lifetime cores that do not require periodic refueling. For those reactors that do discharge spent fuel, DOE accepts it for storage or reprocessing, and not more than an estimated 50 kilograms of such spent fuel are generated annually. Thus, given these worst-case projections, which are not expectations but bounding estimates, the Commission finds that a delay in repository availability to 2025 will not result in significant safety or environmental impacts due to extended operational spent fuel storage. To put it another way, the Commission is confident that, even if a repository were not available within 30 years after the effective expiration of the OLs for both currently retired reactors and potential future reactor retirements through 1995, the overall safety and environmental impacts of extended spent fuel storage would be insignificant. 2.B.2.b. Although it is clear that there is uncertainty in projections of total future spent fuel discharges, it is not clear that the institutional uncertainties arising from having to restart a second repository program should be considered in detail in the current Waste Confidence Decision review. License renewals would have the effect of increasing requirements for spent fuel storage. The Commission understands that some utilities are currently planning to seek renewals for 30 years. Assuming for the sake of establishing a conservative upper bound that the Commission does grant 30-year license renewals, the total opera ting life of some reactors would be 70 years, so that the spent fuel initially generated in them would have to be stored for about 100 years if a repository were not available until 30 years after the expiration of their last OLs. Even under the conservative bounding assumption of 30-year license renewals for all reactors, however, if a repository were available within the first quarter of the twenty-first century, the oldest spent fuel could be shipped off the sites of all currently operating reactors well before the spent fuel initially generated in them reached the age of 100 years. Thus, a second repository, or additional capacity at the first, would be needed only to accommodate the additional quantity of spent fuel generated during the later years of these reactors' operating lives. The availability of a second repository would permit spent fuel to be shipped offsite well within 30 years after expiration of these reactors' OLs. The same would be true of the spent fuel discharged from any new generation of reactor designs. In sum, although some uncertainty in total spent fuel projections does arise from such developments as utilities' planning renewal of OLs for an additional 20 to 30 years, the Commission believes that this Waste Confidence review need not at this time consider the institutional uncertainties arising from having to restart a second repository program. Even if work on the second repository program is not begun until 2010 as contemplated under current law, there is sufficient assurance that a second repository will be available in a timeframe that would not constrain the removal of spent fuel from any reactor within 30 years of its licensed life for operation. 2.B.3. Are early slippages in the DOE repository program milestones significant enough to affect the Commission's confidence that a repository will be available when needed for health and safety reasons? The 2007-2009 timeframe imposed on the Commission by the May 23, 1979 remand by the Court of Appeals was based on the scheduled expiration of the OLs for the Vermont Yankee and Prairie Island nuclear reactors. The specific issues remanded to the Commission were: (1) Whether there is reasonable assurance that an off site storage solution will be available by the years 2007-2009 (the expiration of the plants' operating licenses); and, if not, (2) whether there is reasonable assurance that the fuel can be stored safely at the sites beyond those dates. There was no finding by the Court that public health and safety required offsite storage or disposal by 2007-2009. In directing the Commission to address the safety of at-reactor storage beyond 2007-2009, the Court recognized the possibility that an offsite storage or disposal facility might not be available by then. In any case, the years 2007-2009 no longer have the same meaning for this proceeding as they had in 1984; the OLs for Prairie Island and Vermont Yankee have been or will soon be extended to 2012-2014, on the basis of NRC's past willingness to approve a 40-year operating lifetime from the date of issuance of the OL. *s1-sc-22 The Commission has not identified a date by which a repository must be available for health and safety reasons. Taking into account institutional requirements for spent fuel storage, the Commission found, under Finding 3 in the 1984 Waste Confidence Decision, that spent fuel would be safely managed until sufficient repository capacity is available. The Commission also found, however, that in effect, under the second part of Finding 2, safe management would not need to continue for more than 30 years beyond expiration of any reactor's OL, because sufficient repository capacity was expected to become available within those 30 years. Considering that spent fuel would not have to be stored more than 30 years after any reactor's 40-year OL expiration, and taking into account the technical requirements for such storage. the Commission went on to determine under Finding 4-that, in effect, spent fuel could be safely stored for at least 70 years after discharge from a reactor. Thus, the Commission's 1984 Decision did not establish a time when sufficient repository capacity would be required; it established a minimum period during which storage would continue to be safe and environmentally acceptable pending the expected availability of sufficient repository capacity. Bearing in mind that reactor facilities were originally designed and OLs issued for a licensed life for opera lion of 40 years, the Commission is proposing elsewhere in this Federal Register notice a clarifying revision of Finding 4 to say that spent fuel can be safely stored at a reactor for at least 30 years after the "licensed life for opera lion" of that reactor. Implicitly, the proposed use of the phrase "licensed life for operation" clarifies that the Commission found in 1984 that NRC licensing requirements for reactor facility design, construction, and operation provide reasonable assurance that spent fuel can be stored safely and without significant environmental impacts for at least the first 40 years of the reactor's life. The Commission's proposed finding also implies that, barring any significant and pertinent unexpected developments, neither technical nor institutional constraints would adversely affect this assurance for at least another 30 years after that first 40 years. Another implication of this revised finding is that, where a utility is able to meet NRC requirements to extend that reactor's operating lifetime by license renewal, spent fuel storage for at least 30 years beyond the end of the period of extended life will also be safe and PART 51 STATEMENTS OF CONSIDERATION without significant environmental impacts. In assessing the effect of early slippages in DOE repository program milestones, therefore, the most important consideration is not the earliest date that an operating license actually expired, but the earliest date that an OL was issued. The earliest OL to be issued was for Dresden 1 in 1959, followed by a number of reactors licensed for operation in 1962. The OLs for all of the 111 power reactors now licensed to operate are currently scheduled to expire sometime within the first three decades of the twenty-first century, which is also the period in which their currently licensed life for operation would end. (See Nuclear Regulatory Commission 1989 Information Digest, NUREG-1350, Vol. 1, p. 33.) Thus, conservatively assuming here that there will be no license renewals, the earliest timeframe when a repository might be needed to dispose of spent fuel from the majority of reactors is 2029-2050. As proposed in the first part of Finding 2, the Commission has reasonable assurance that a repository will be available within the first quarter of the twenty-first century. Even if a repository were not available until 2025, this would be several years before the beginning of the earliest timeframe within which, based on an assumed 30-year storage after an assumed 40-year licensed life of reactor operation, a repository might be needed for spent fuel disposal. Thus, early slippages in DOE's program milestones do not affect the Commission's confidence that a repository will be available within that timeframe. 2.B.4. NRC has stated that the 3-to 4-year license application review schedule is optimistic, and that for NRC to meet this schedule, DOE must submit a complete and high-quality license application. In the September 16, 1988 NRC comments to DOE on the Draft 1988 Mission Plan Amendment, the Commission requested that DOE acknowledge its commitment to develop this complete and high-quality application, "even if this would result in longer times to collect the necessary information and subsequent delays in submitting the license application." Will NRC's emphasis on the completeness and quality of the license application have a significant effect on the timing of the submittal of the license application and subsequent licensing proceeding to grant construction authorization in time for repository availability by 2007-2009? As the NRC indicated to DOE in NRC's October 25, 1985 comments on the draft PDS, the t!uee-year statutory schedule for the NRC licensing proceeding on the application for construction authorization is optimistic. The Commission has sought ways to improve the prospe 4 ts for meeting this schedule, for example by developing the LSS for expedited document discovery during the licensing proceeding. In the same correspondence on the PDS, NRC also stated that the adequacy of the three-year review period depends on DOE's submittal of a complete and high-quality application. A license application supported by inadequate data may lead to findings during the licensing proceeding that the results of certain tests cannot be admitted as part of the license application. If it is not possible to repeat the tests in question, NRC may have no alternative but to deny the application-with a consequent loss of program momentum and considerable financial cost. NRC recognizes that emphasis on a complete and high-quality license application may cause some near-term delays that could make it difficult to achieve the current schedule calling for submittal of the construction authorization application in 1995. Notwithstanding any such delays, the Commission has reasonable assurance that if the Yucca Mountain site is not found unsuitable, a repository at that site could be available by the 2012-2014 timeframe, consistent with the rescheduled OL expiration dates for Prairie Island and Vermont Yankee. For reasons discussed previously, this timeframe now appears more relevant to the Waste Confidence proceeding than the 2007-2009 timeframe. In any case, the Commission remains convinced that the benefits to the repository program of submitting a quality license application would outweigh the cost of delay in preparing the application. NRC has always placed great emphasis on early resolution of potential licensing issues in the interest of expeditious review of the license application and timely repository availability. It is in the same spriit of timely repository operation that the Commission is urging greater attention to quality than to meeting the schedule for submittal of the license application. NRC believes that a complete and quality license application offers the best available assurance that timely repository licensing and operation can be achieved. In addition to expediting the review of the application, a high-quality license application and site characterization program should enhance overall confidence that any site granted a construction authorization will prove to . 51-SC-23 be reliable during the period of performance confirmation. It will also increase public confidence that the program is being carried out in a thorough and technically sound manner. 2.C. Conclusion on Finding 2 In reexamining the technical and institutional uncertainties surrounding the timely development of a geologic repository since the 1984 Waste Confidence Decision, the Commission has been led to question the conservatism of its expectation that a repository would be available to 2007-2009. At the time of the 1984 Decision, the Commission said that timely attainment of a repository did not require DOE to adhere strictly to the milestones set out in the NWPA, and there would be delays in some milestones. It did not appear to the Commission at the time that delays of a year or so in meeting any of the milestones would delay the date of repository availability by more than a few years beyond the 1998 deadline specified in the act. Since then, however, several developments have made it apparent that delays of more than a few years are to be the norm rather than the exception in the early years of this program. There has been a five-year slip in DOE's estimate of repository availability from 1998 to 2003, and DOE has been unable to meet such near-term repository program milestones as excavation of the exploratory shaft and the start of in-situ testing. There remains the possibility that potential repository availability at the Yucca Mountain site wil be further delayed due to unforeseen problems during site characterization. These developments do not in themselves rule out the possibility that DOE will still be able to achieve repository operation by 2007-2009, but they do suggest that to expect repository operation by then may be optimistic. In the Commission's view, 2012-2014 is now a more relevant timeframe than 2007-2009. When the Court issued its 1979 remand, 2007-2009 was when the OLs for Vermont Yankee and Prairier Island were scheduled to expire. The opera ting licenses for the two Prairie Island units have since been extended to 2013 and 2014, and the operating license for Vermont Yankee is eligible for extension to 2012. These extensions have been made available under the Commission's policy that the allowable opera ting life of a licensed reactor should not be foreshortened because of construction delays. It therefore seems reasonable for NRC to make its finding on the timing of repository availability PART 51 STATEMENTS OF CONSIDERATION by 2012-2014, rather than by 2007-2009. The Commission has a greater degree of assurance that if the Yucca Mountain site is suitable, a repository would be available there by 2012-2014. For the sake of conservatism, however, the suitability of Yucca Mountain should not be assumed. Yucca Mountain is now the only candidate site available; the NWPAA required that DOE terminate site characterization activities at all sites other than the Yucca Mountain site. In effect, the 2007-09 schedule for repository availability could be met only if Yucca Mountain survived the repository development process as a licensed site. If this site were found to be unlicenseable or otherwise unsuitable, characterization would have to begin at another site or suite of sites, with consequent further delay in repository availability. The final decision on the suitability of the site to proceed to licensing and repository development will rest with DOE, but the position of the NRC staff will figure in that decision. The staff will not be able to make a recommendation to a licensing board to authorize repository construction at Yucca Mountain until all site characterization activities have been completed. DOE might thus be able for several more years to determine whether there will in fact have to be a delay to find and characterize another site. Another reason the Commission is unwilling to assume the suitability of Yucca Mountain is that NRC must be mindful of preserving all its regulatory options-including a recommendation of license application denial-to assure adequate protection of public health and safety from radiological risk. In our view, it is essential to dispel the notion that for schedular reasons there is no alternative to the currently preferred site. This view is consistent with past Commission statements that the quality of DOE's preparations for a license application should take precedence over timeliness where the two conflict. It is also consistent with the view that because we are making predictions about completion dates for a unique and complex enterprise at least some 20 years hence, it is more reasonable to express the timescale for completion in decades rather than years. In order to obtain a conservative upper bound for the timing of repository availability, the Commission has made the assumption that the Yucca Mountain site will be found to be unsuitable. If DOE were authorized to initiate site screening for a repository at a different site in the year 2000, the Commission believes it is reasonable to expect that a repository would be available by the year 2025. This estimate is based on the DOE position that site screening for a second repository should begin 25 years before the start of waste acceptance. The consideration of technical and institutional issues presented here has found none that would preclude the availability of a repository within this time frame. For the second part of its 1984 finding on repository availability, the Commission found reasonable assurance that sufficient repository capacity will be available within 30 years beond expiration of any reactor OL to dispose of existing commercial high level waste and spent fuel originating in that reactor and generated up to that time. The Commission believes that this finding should also be modified in light of developments since 1984. When the Commission made this finding, it took into consideration both technical and institutional concerns. The technical concern centered on the ability of the spent fuel and the engineered reactor storage facilities to meet the requirements for extended operational storage before shipment for disposal. The institutional question concerned whether the utility currently responsible for post-operational reactor storage, or some substitute organization, would be able to assure the continued safety of this storage. The principal new developments since 1984 that bear on these questions are: (1) That dry spent fuel storage technologies have become operational on a commercial scale; and (2) that several utilities are proceeding with plans to seek renewals of their OLs, with appropriate plant upgrading, for an additional period up to 30 years beyond the 40-year term of their current licenses. The accumulation of operating experience with dry-cask storage, a technology requiring little active term maintenance, provides additional assurance that both the technical and institutional requirements for extended post-operational spent fuel storage will be met. License renewals, however, would have the effect of increasing requirements for both the quantity and possibly the duration of storage. If the commission were to grant 30-year license renewals, the total operating life of some reactors could be 70 years, so that the spent fuel initially generated in such reactors would have to be stored for about 100 years. if a respository were not available until 30 years after the expiration of their last OLs. This raises the question as to whether that spent fuel, and the hardware and civil 51-SC-24 engineering structures for storing it, can continue to meet NRC requirements for an additional 30 years beyond the period the Commission supported in 1984. For all the reasons cited in the discussion of Finding 4, the Commission believes there is ample technical basis for confidence that spent fuel can be stored safely and without significant environmental impact at these reactors for at least 100 years. If a repository were available within the first quarter of the twenty-first century, the oldest spent fuel could be shipped off the sites of all currently opera ting reactors well before the spent fuel initially generated in them reached the age of 100 years. The need to consider the institutional aspects of storage beyond 30 years after OL expiration was not in evidence in 1984 because the Commission was confident that at least one repository would be available by 2007-2009. On that schedule, waste acceptance of spent fuel from the first reactor whose operating license had expired (Indian Point 1, terminated in 1980) could have begun within 30 years of expiration of that license. If a repository does not prove to be available until 2025, however, it would not be available within 30 years of the time that OLs could be considered effectively to have expired for Indian Point 1 and the three other plants with spent fuel onsite that were retired before the end of their licensed life for reactor operation. The same would be true of any additional reactors prematurely retired between now and 1995, when the 30-year clock starts for the availability of a repository by 2025. Premature shutdowns notwithstanding, the Commission has reasons to be assured that the spent fuel at all of these reactors will be stored safely and without significant environmental impact until sufficient repository capacity becomes available. Considering first the technical reasons for this assurance, it is important to recognize that each of these reactors and its spent fuel storage installation were originally licensed in part on the strength of the applicant's showing that the systems and components of concern were designed and built to assure safe operation for 40 years under expected normal and transient severe conditions. All of the currently retired reactors have a significant portion of that 40-year expected life remaining, and all have only small quantities of spent fuel onsite in storage installations that were licensed to withstand considerably larger thermal and radiation loadings from much greater quantities of spent fuel. Of the four reactors currently PART 51 STATEMENTS OF CONSIDERATION retired with spent fuel onsite, the two with far the longest terms of operation, Lacrosse and Dresden, were operated for 19 and 18 years, respectively. For the continued safe management of the spent fuel and storage installations at any existing or potential prematurely retired plant, the Commission believes it can reasonably rely on the continued structural and functional integrity of the plant's engineered storage installations for at least the balance of its originally licensed life as if the OL were still in effect. This is to say that for the purposes of Finding 2, no foreseeable technical constraints have arisen to disturb the Commission's assurance that spent fuel storage at any reactor will remain safe and environmentally accpetable for at least 30 years after its licensed life for operation, regardless of whether its OL has been terminated at an earlier date. The Commission also sees no insurmountable institutional obstacles to the continued safe management of spent fuel during the remainder of any shutdown reactor's initially licensed life for operation, or for at least 30 years thereafter. Because there will still be an NRC possession license for the spent fuel at any reactor that has indefinitely suspended operations, the Commission will retain ample regulatory authority to require any measures, such as removal of the spent fuel remaining in storage pools to passive dry storage casks, that might appear necessary after an OL expires. Even if a licensed utility were to become insolvent, and responsibility for spent fuel management were transferred to DOE earlier than is currently planned, the Commission has no reason to believe that DOE would have insufficient Nuclear Waste Fund resources or otherwise be unable to carry out any safety-related measures NRC considers necessary. Thus, in the case of a premature reactor retirement, the Commission has an adequate basis, on both technical and institutional grounds, for reasonable assurance that spent fuel can be stored safely and without significant environmental impacts for at least 30 years beyond not only the actual end of that reactor's OL, but the end of its originally licensed life for opera lion. In sum, considering developments since 1984 in the repository development program, in the operating performance of U.S. power reactors, and in spent fuel storage technology, the Commission finds that: (1) The overall public health, safety, and environmental impacts of the possible unavailability of a repository by 2007-2009 would be insignificant; and (2) neither 30-year renewals of reactor licenses nor a delay in repository availablility to 2025 will result in significant safety or environmental impacts from extended post-operational spent fuel storage. The Commission finds ample grounds for its proposed revised findings on the expected availability of a repository. The institutional support for the repository program is well-established. A mechanism for funding repository program activities is in place, and there is a provision in the NWPA for adjusting, if necessary, the fee paid by utilities into this fund. Congress has continued to provide support for the repository program in setting milestones, delineating responsibilities, establishing advisory bodies, and providing a mechanism for dealing with the concerns of States and affected Indian tribes. Technical support for extended spent fuel storage has improved since 1984. Considering the growing availability, reasonable cost, and accumulated operating experience with new dry cask spent fuel storage technology since then, the Commission now has even greater assurance that spent fuel can be stored safely and without significant environmental impact for at least 30 years after the expected expiration of any reactor's OL. Where a reactor's OL has been terminated before the expected expiration date, the Commission has an adequate basis to reaffirm what was implicit in its initial concept, namely: that regardless of the actual date when the reactor's operating authority effectively ended, spent fuel can be stored safely and without significant environmental impacts for at least 30 years beyond that reactor's licensed life for operation. There is thus no foreseeable health and safety or environmental requirement that a repository be made available within the 2007-2009 timeframe at issue in the Commission's original proceeding. Nor does the Commission see a radiological safety or environmental requirement for repository availability at the end of the expected revised timeframe of 2012-2014 for the expiration of the Prairie Island and Vermont Yankee OLs. Indeed, the Commission sees important NRC mission-related grounds for avoiding any statement that repository operation by 2007-2009 is required. Geologic disposal of high-level radioactive wastes is an unprecedented endeavor. It requires reliable projections of the waste isolation performance of natural and engineered barriers over millennia. After the repository is sealed, retrieval of the emplaced wastes will no 51-SC-25 longer be practicable, and the commitment of wastes to that site will, by design, be irreversible. In DOE's testing, both in the laboratory and at the candidate repository site, in its development of facility and package designs, and in all other work to demonstrate that NRC requirements will be met for a repository at Yucca Mountain, the Commission believes that the confidence of both NRC and the public depends less on meeting the schedule for repository operation than on meeting safety requirements and doing the job right the first time. Thus, given the Commission's assurance that spent fuel can safely be stored for at least 100 years if necessary, it appears prudent for all concerned to prepare for the better-understood and more manageable problems of storage for a few more years in order to provide additional time to assure the success of permanent geologic disposal. This is not to say that the Commission is unsympathetic to the need for timely progress toward an operational repository. It is precisely because NRC is so confident of the national commitment to achieve early repository operation that the Commission believes it no longer need add its weight to the considerable pressures already bearing on the DOE program. There is ample institutional impetus on the part of others, including Congress, the nuclear power industry, State utility rate regulatory bodies, and consumers of nuclear-generated power, toward DOE achievement of scheduled program milestones. With continuing confidence in the technical feasibility of geologic disposal, the Commission has no reason to doubt the institutional commitment to achieve it in a timeframe well before it might become necessary for safety or environmental reasons. Indeed, the Commission believes it advisable not to attempt in this review a more precise NRC estimate of the point at which a repository will be needed for radiological safety or environmental reasons, lest this estimate itself undermine the commitment to earlier achievement of repository operations. The Commission continues to hope that a repository will in fact be available by 2007-2009, and has found nothing to date that would conclusively prevent this achievement. To find reasonable assurance that a repository will be available by 2007-2009, however, is a different and more consequential proposition in the context of this review. In light of the delays the program has encountered since its inception, and the regulatory need to avoid a premature commitment to the PART 51 STATEMENTS OF CONSIDERATION Yucca Mountain site, the Commission cannot prudently describe a basis for assurance that the current DOE schedule for repository operation in 2003 will not slip another four to six years under any reasonably foreseeable circumstances. The Commission could more easily substantiate a finding that a repository will be available within the revised 2012-2014 timeframe that would be created by extending the OLs of the reactors in question when the Waste Confidence proceeding began. Even this revised estimate, however, could too easily be misinterpreted as an NRC estimate of the time at which continued spent fuel storage at these sites would be unsafe or environmentally significant. The Commission's enhanced confidence in the safety of extended spent fuel storage provides adequate grounds for the view that NRC need not at this time define more precisely the period when, for reasons related to NRC's mission, a permanent alternative to post-operational spent fuel storages will be needed. The Commission therefore proposes the following revision of its original Finding on when sufficient repository capacity will be available. The Commission finds reasonable assurance that at least one mined geologic repository will be available within the first quarter of the twenty-first century, and sufficient repository capacity will be available within 30 years beyond the licensed life for operation of any reactor to dispose of the commercial high-level radioactive waste and spent fuel originating in such reactor and generated up to that time. Original Finding 3 The Commission finds reasonable assurance that high-level radioactive waste and spent fuel will be managed, in a safe manner until sufficient repository capacity is available to assure the safe disposal of all high-level waste and spent fuel. Proposed Finding 3 Same as above. 3.A. Issues Considered in Commission's 1984 Decision on Finding 3 In the Commission's discussion of Finding 3 in its Waste Confidence Decision (49 FR 34658, August 31, 1984), in section 2.3 'Third Commission Finding,' the Commission stated. Nuclear power plants whose operating licenses expire after the years 2007-09 will be subject to NRC regulation during the entire period between their initial operation and the availability of a waste repository. The Commission has reasonable assurance that the spent fuel generated by these licensed plants will be managed by the licensees in a safe manner. Compliance with the NRC regulations and any specific license conditions that may be imposed on the licensees will assure adequate protection of the public health and safety. Regulations primarily addressing spent fuel storage include 10 CFR part 50 for storage at the reactor facility and 10 CFR part 72 for storage in independent spent fuel storage installations (ISFS!s). Safety and environmental issues involving such storage are addressed in licensing reviews under both parts 50 and 72, and continued storage operations are audited and inspected by NRC. NRC's experience in more than 80 individual evaluations of the safety of spent fuel storage shows Iha t significant releases of radioactivity from spent fuel under licensed storage conditions are extremely remote. Some nuclear power plant operating licenses expire before the years 2007-09. For technical, economic or other reasons, other plants may choose, or be forced to terminate opera lion prior to 2007-09 even though their operating licenses have not expired. For example, the existence of a safety problem for a particular plant could prevent further operation of the plant or could require plant modifications that make continued plant operation uneconomic. The licensee. upon expiration or termination of its license, may be granted (under 10 CFR part 50 or part 72) a license to retain custody of the spent fuel for a specified term (until repository capacity is available and the spent fuel can be transferred to DOE under sec. 123 of the Nuclear Waste Policy Act of 1982) subject to NRC regulations and license conditions needed to assure adequate protection of the public. Alternatively, the owner of the spent fuel, as a last resort, may apply for an interim storage contract with DOE, under sec. 135(b) of the Act, until not later than 3 years after a repository or monitored retrievable storage facility is available for spent fuel. For the reasons discussed above, the Commission is confident that in every case the spent fuel generated by those plants will be managed safely during the period between license expiration or termination and the availability of a mined waste repository for disposal. Even if a repository does not become available until 2025, nothing has occurred during the five years since its original Decision to diminish the Commission's confidence that high-level waste and spent fuel will be managed in a safe manner until a repository is available. The same logic just stated continues to apply through the first quarter of the twenty-first century. NRC regulations remain adequate to assure safe storage of spent fuel and radioactive high-level waste at reactors, at independent spent fuel storage installations (ISFSls), and in an MRS until sufficient repository capacity is available. 10 CFR 72.42(a) provides for renewal of licensed storage at ISFSls for additional 20-year periods for interim storage, or for additional 40-year periods for monitored retrievable storage of spent fuel and solidified radioactive 51-SC-26 high-level waste if an MRS facility is constructed, licensed, and operated. This would ensure that spent fuel and solidified high-level waste, if any were to be delivered to an MRS facility, would remain in safe storage under NRC regulation throughout its storage. The Commission has also published for public comment a proposed amendment to part 72, to issue a general license to reactor opera ting licensees to use approved spent fuel storage casks at reactor sites. If this proposed amendment is promulgated, no specific part 72 license would be required. Operating license holders would register with NRC to use approved casks on their sites. Spent fuel may continue to be stored in the reactor spent fuel pool under a part 50 "possession only" license after the reactor has ceased operating. In addition, DOE's policy of disposing of the oldest fuel first, as set forth in its Annual Capacity Report, makes it unlikely that any significant fraction of total spent fuel generated will be stored for longer than the 30 years beyond the expiration of any operating reactor license. This expectation, established in the Commission's original proceeding, continues to be reasonable, even in the event that a repository is not available until some time during the first quarter of the twenty-first century. Even in the case of premature shutdowns, where spent fuel is most likely to remain at a site for 30 years or longer beyond OL expiration (see Finding 2, previously discussed), the Commission has confidence that spent fuel will be safely managed until safe disposal is available. Until the reactor site has been fully decommissioned, and spent fuel has been transferred from the utility to DOE as required by NRC regulations, the licensee remains responsible to NRC. Furthermore, under 10 CFR 50.54bb, originally issued in final form by the Commission with its 1984 Waste Confidence Decision, a reactor licensee must provide to NRC, five years before expiration of an OL, notice of plans for spent fuel disposition. Accordingly, the Commission concludes that nothing has changed since the enactment of the Nuclear Waste Policy Act of 1982 and the Waste Confidence Decision in August 1984 to diminish the Commission's "* *

reasonable assurance that high-level radioactive waste and spent fuel will be managed in a safe manner until sufficient repository capacity is available
* * ." Pursuant to the NWPA, the Commission issued in final form 10 CFR part 53, "Criteria and Procedures for Determining Adequacy of Available PART 51 STATEMENTS OF CONSIDERATION Spent Nuclear Fuel Storage Capacity," addressing the determination of need, if any, for DOE interim storage. No applications were received by the June 30, 1989.NWPA deadline incorporated into the Commission's rule, and it seems unlikely that any applications will be made to NRC for interim storage by DOE. Even if NRC were to make an exception for a late application, a determination must be made before January 1, 1990 to comply with the NWPA. 3.B Relevant Issues That Have Arisen Since the Commission's Original Decision on Finding 3 Although a DOE facility will not be available to enable the Department to begin accepting spent fuel in 1998, as provided in the contracts under the NWPA, the Commission's confidence in safe storage is unaffected by any potential contractual dispute between DOE and spent fuel generators and owners as to responsibility for spent fuel storage. In the event that DOE does not take title to spent fuel by this date, a licensee under either 10 CFR part 50 or part 72 cannot abandon spent fuel in its possession.

Further, the Commission notes that only two reactors are currently scheduled for shutdown before 2003, DOE's anticipated repository startup date. (See Nuclear Regulatory Commission 1989 Information Digest, NUREG-1350, Vol.1, p.33). To resolve any continuing uncertainties, however, it would be helpful if DOE and utilities and other spent fuel generators and owners could reach an early and amicable resolution to the question of how and when DOE will accept responsibility for spent fuel. This would facilitate cooperative action to provide for a smoothly operating system for the ultimate disposition of spent fuel. The Commission recognizes that the NWPA limitation of 70,000 NTHM for the first repository will not provide adequate capacity for the total amount of spent fuel projected to be generated by all currently operating licensed reactors. The NWPAA effectively places a moratorium on a second repository program until 2007-2010. Either the first repository must be authorized and able to provide expanded capacity sufficient to accommodate the spent fuel generated, or there must be more than one repository. Since Congress specifically provided in the NWPAA for a first repository, and required DOE to return for legislative authorization for a second repository, the Commission believes that Congress will continue to provide institutional support for adequate repository capacity. The Commission's confidence about the availability of repository capacity is not affected by the possibility that some existing reactor licenses might be renewed to permit continued generation of spent fuel at these.sites. Because only two reactor licenses are scheduled to expire before 2003, the impact of license renewals (a matter not considered in the Commission's 1984 Decision] will have no significant effect within the first quarter of the twenty-first century on scheduling requirements for a second repository. Renewals may slightly alleviate the need for a second repository in the short term, because spent fuel storage capacity will be expanded for extended storage at these reactor sites. Over the longer term, renewals might increase spent fuel generation well into the latter half of the twenty-first century. Nonetheless, nothing in this situation diminishes the Commission's assurance that safe storage will be made available as needed. In summary, the Commission finds no basis for changing the Third Finding in its Waste Confidence Decision. The Commission continues to find "* *

reasonable assurance that high-level radioactive waste and spent fuel will be managed in a safe manner until sufficient repository capacity is available to assure the safe disposal of all high-level waste and spent fuel." Original Finding 4 The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the expiration of that reactor's operating license at that reactor's spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations.

Proposed Finding 4 The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impact for at lest 30 years beyond the licensed lite for operation (which may include the term of a revised license) of that reactor at its spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations. 4.A. Issues Considered in Commission's 1984 Decision on Finding 4 In the Commission's discussion of Finding 4 in its Waste Confidence Decision (49 FR 34658, August 31, 1984) section 2.4 "Fourth Commission Finding," the Commission said that: 51-SC-27 Although the Commission has reasonable assurance that at least one mined geologic repository will be available by the years 2007--09, the Commission also realizes that for various reasons, including insufficient capacity to immediately dispose of all existing spent fuel, spent fuel may be stored in existing or new storage facilities for some periods beyond 2007--09. The Commission believes that this extended storage will not be necessary for any period longer than 30 years beyond the term of an operating license. For this reason, the Commission has addressed on a generic basis in this decision the safety and environmental impacts of extended spent fuel storage at reactor spent fuel basins or at either onsite or offsite spent fuel storage installations. The Commission finds that spent fuel can be stored safely and without significant environmental impacts for at least 30 years beyond the expiration of reactor operating licenses. To ensure that spent fuel which remains in storage will be managed properly until transferred to DOE for disposal, the Commission is proposing an amendment to its regulations (10 CFR part 50). The amendment will require the licensee to notify the Commission, five years prior to expiration of its reactor operating license, how the spent fuel will be managed until disposal. The Commission's finding is based on the record of this proceeding which indicates that significant releases of radioactivity from spent fuel under licensed storage conditions are highly unlikely. It is also supported by the Commission's experience in conducting more than BO individual safety evaluations of storage facilities. The safety of prolonged spent fuel storage can be considered in terms of four major issues: (a) The long-term integrity of spent fuel under water pool storage conditions, [b) structure and component safety for extended facility operation, (c) the safety of dry storage, and [d) potential risks of accidents and acts of sabotage at spent fuel storage facilities. For reasons discussed above, the Commission arrived at a provisional figure of 10 years or more for storage (i.e., a 40-year reactor OL span, plus 30 years or more]. The 70-year-plus estimate is supported by oral testimony from the nuclear industry to the Commission in the Waste Confidence Proceeding. (See Transcript of Commission Meeting, "In the Matter of: Meeting on Waste Confidence Proceeding," January 11, 1982, Washington, DC, pp. 148-160). This testimony specifically addressed safety issues related to water pool storage of spent fuel and supported the position that spent fuel could be stored for an indefinite period, citing the industry's written submittal to the Commission in the proceeding. (See "The Capability for the Safe Interim Storage of Spent Fuel" [Document 4 of 4), Utility Nuclear Waste Management Group and Edison Electric Institute, July 1980). Some of this PART 51 STATEMENTS OF CONSIDERATION material alluded to in the oral testimony was subsequently referenced by the Commission in its discussion of water pool storage issues and its Fourth Finding of reasonable assurance that spent fuel and high level waste " * *

will be managed in a safe manner." (See 49 FR 346758 at pp. 34681-2, August 31, 1984). If a reactor with a 40-year initial license were to have that license renewed for another 30 years, the Commission believes that the spent fuel generated at that reactor can be safely stored for at least several decades past the end of the 70-year operating period. Adding to these 70 years the expected 30-year post-OL period during which the Commission believes, udner Finding 2, that sufficient repository capacity will be made available for any reactor's spent fuel, the total storage time would be about 100 years. In making the original Fourth Finding, the Commission did not determine that for technical or regulatory reasons, storage would have to be limited to 70 years. This is apparent from the Commission's use of the words " * *
for at least 30 years beyond the expiration of that reactor's operating license * * * [emphasis added]." Similarly, in using the words "at least" in its proposed revised Finding Four, the Commission is not suggesting 30 years beyond the licensed life for operation (which may include the term of a revised license) represents any technical limitation for safe and environmentally benign storage. Degradation rates of spent fuel in storage, for example, are slow enough that it is hard to distinguish by degradation alone between spent fuel in storage for less than a decade and spent fuel stored for several decades. The Commission's proposed revised Finding here is meant to apply both to wet storage in reactor pools and dry storage in engineered facilities outside the reactor containment building.

Both dry and wet storage will be discussed in detail next. Since the original Waste Confidence Decision, which found that material degradation processes in dry storage were well-understood, and that storage systems were simple, passive, and easily maintained, NRC and ISFSI operators have gained experience with dry storage which confirms the Commission's 1984 conclusions. NRC staff safety reviews of topical reports on storage-system designs, the licensing and inspection of storage at two reactor sites, and NRC promulgation of the part 72 amendment for MRS, have significantly increased the agency's understanding of the confidence in dry storage. Under NWPA section 218(a), DOE has carried out spent fuel storage research and development as well as demonstration of dry cask storage at its Idaho National Engineering Laboratory. Demonstration has been carried out for metal casks under review or previously reviewed by NRC staff. DOE has also provided support to utilities in dry storage licensing actions (see Godlewski, N.Z., "Spent Fuel An Update," Nuclear News, Vol. 30, No. 3, March 1987, pp. 47-52). Dry storage of spent fuel has become an available option for utilities, with reactor dry storage licensed and underway at two sites: The H.B. Robinson Steam Electric Plant, Unit 2, in South Carolina, and the Surry Nuclear Station in Virginia. NRC has received an application for dry storage at Duke Power Company's Oconee Power Station site as well. This application is still under review, bu the environmental review is completed and an environmental assessment and finding of no significant impact have been issued (see 53 FR 44133, November 1, 1988). Based on utility statements of intent, and projections of need for additional storage capacity at reactor sites, the NRC staff expects numerous applications from utilities over the next decade (see "Final Version Dry Cask Storage Study," DOE/RW-0220, February 1989). Since the original Waste Confidence finding, the Commission has reexamined long-term spent fuel storage in issuing an amendment to 10 CFR part 72 to address the storage of spent fuel and high-level radioactive waste in an MRS, as envisioned by Congress in section 141 of the NWPA. Under the rule, storage in an MRS is to be licensed for a period of 40 years, with the possibility for renewal. The Commission determined not to prepare an environmental impact statement for the proposed amendments to 10 CFR part 72, however. (See 53 FR 31651, p. 31657, August 19, 1988.) An environmental assessment and finding of no significant impact were issued because the Commission found that the consequences of long-term storage are not significant. The environmental assessment for 10 CFR part 72, "Licensing Requirements for the Indpendent Storage of Spent Fuel and High-Level Radioactive Waste," NUREG-1092, assessed dry storage of spent fuel for a period of 70 years after receipt of spent fuel from a reactor: The basis chosen for evaluating license requirements for the long-term storage of spent nuclear fuel and high-level radioactive waste in an MRS is an installation having a 70-year design lifetime and a 70,000 MTU storage capability. This assessment focuses 51-SC-28 on the potential environmental consequences for a long-term storage period, a period for which the Commission needs to assure itself of the continued safe storage of spent fuel and high-level radioactive waste and the performance of materials of construction. This means the reliability of systems important to safety needs to be established to ensure that long-term storage of spent fuel and HLW does not adversely impact the environment. For example, the staff needs to establish that systems, such as concrete shielding, have been evaluated to determine how their physical properties withstand the consequences of irradiation and heal flux for about a 70-year period. The Commission addressed structure and component safety for extended operation for storage of spent fuel in reactor water pools in the matter of waste confidence rulemaking proceeding. The Commission's preliminary conclusion is that experience with spent fuel storage provides an adequate basis for confidence in the continued safe storage of spent fuel for at least 30 years after expiration of a plant's license. The Commssion is therefore confident of the safe storage of spent fuel for at least 70 years in water pools at facilities designed for a 40-year lifetime. The Commission also stated that its authority to require continued safe management of spent fuel generated by licensed plants protects the public and assures them the risks remain acceptable. In consideration of the safety of dry storage of spent fuel, the Commission's preliminary conclusions were that [its] confidence in the extended dry storage of spent fuel is based on a reasonable understanding of the material degradation processes, together with the recognition that dry storage systems are simpler and more readily maintained. In response to Nuclear Waste Policy Act of 1982 authorizations, the Commission noted; ** *

the Commission believes the information above [on dry spent fuel storage research and demonstration]

is sufficient to reach a conclusion on the safety and environmental effects of extended dry storage. All areas of safety and environmental concern [e.g., maintenance of systems and components, prevention of material degradation, protection against accidents and sabotage) have been addressed and shown to present no more potential for adverse impact on the environmental and the public health and safety than storage of spent fuel in water pools.' At this lime, the Commission is confident it can evaluate the long-term integrity of material for constructing an installation and provide the needed assurance for safe storage of spent fuel and HLW to establish the licensibility of an MRS over extended periods of time. The MRS fuel storage concepts discussed here for revision of 10 CFR part 72 covers only dry storage concepts. [References omitted] The Commission believe that its 1984 Fourth Finding should be changed to reflect the environmental assessment in the 10 CFR part 72 MRS rulemaking and other evidence that spent fuel can be stored, safely and without significant PART 51 STATEMENTS OF CONSIDERATION environmental impact, for extended periods. Although the Commission does not beleive storage in excess of a century to be likely, with or without an MRS, there is the potential for storage of spent fuel for times longer than 30 years beyond the expiration of an initial, extended, or renewed reactor 01, if a reactor operating under such a license were prematurely shut down. The Commission does not, however, see any significant safety or environmental problems associated with storage for at least 30 years after the licensed life for operation of any reactor, even if this effectively means storage for at least 100 years, in the case of a reactor with a 70-year licensed life for operation. Under the environmental assessment for the MRS rule, the Commission has found confidence in the safety and environmental insignificance of dry* storage of spent fuel for 70 years following a period of 70 years of storage in spent fuel storage pools. Thus, this environmental assessment supports the proposition that spent fuel may be stored safely and without significant environmental impact for a period of up to 140 years if storage in spent fuel pools occurs first and the period of dry storage does not exceed 70 years. The Commission has also found that experience with water-pool storage of spent fuel continues to confirm that pool storage is a benign environment for spent fuel that does not lead to significant degradation of spent fuel integrity. Since 1984, utilities have continued to provide safe additional reactor pool storage capacity through reracking, with over 110 such actions now completed. The safety of storage in pools is widely recognized among cognizant professionsls. Specifically, the Commission notes one expert's view that: During the last 40 years there has been very positive experience with the handling and storing of irradiated fuel in water; thus wet storage is now considered a proved technology. There is a substantial technical basis for allowing spent fuel to remain in wet storage for several decades. For the past two decades, irradiated Zircaloy-clad fuel has been handled and stored in water. There continues to be no evidence that clad fuel degrades significantly during wet storage-this includes: fuel with burnups as high as 41,000 MWd/MTU: continuous storage of low-burnup fuel for as long as 25 years; and irradiation of fuel in reactors for periods up to 22 years. Cladding defects have had little impact during wet storage, even if the fuel is uncanned. [References omitted.] [See Bailey, W.J. and Johnston, Jr. A.B., et al., "Surveillance of LWR Spent Fuel in Wet Storage," NP-3765, Electric Power Research Institute [EPRI), October 1984, pp. 2-10.] This last conclusion has been reaffirmed by the same authors, who recently wrote: "There continues to be no evidence that LWR spent fuel with Zircaloy or stainless steel cladding degrades significantly during wet storage [EPRI 1986; international Atomic Energy Agency (IAEA) 1982]." (See "Results of Studies on the Behavior of Spent Fuel in Storage," Journal of the Institute of Nuclear Materials Management," Vol. XVI, No. 3, April 1988, p. 27.IV A). In addition to the confidence that the spent fuel assemblies themselves will not degrade significantly in wet storage, there is confidence that the water pools in which the assemblies are stored will remain safe for extended periods: As noted in the recent IAEA world survey, the 40 years of positive experience with wet storage illustrates that it is a fully-developed technology with no associated major technological problems. Spent fuel storage pools are operated without substantial risk to the public or the plant personnel. There is substantial technical basis for allowing spent fuel to remain in wet storage for several decades. Minor, but repairable, problems have occured with spent fuel storage pool components such as liners, racks, and piping. [See Bailey, W.J., and Johnston, Jr., A.B., et al., "Surveillance of LWR Spent Fuel in Wet Storage," EPRI NP-3765, prepared by Battelle Pacific Northwest Laboratories, Final Report, October 1984, p. 6-1.] The studies just cited support the view that rates of uniform corrosion of spent fuel cladding in storage pools are low over time. Localized corrosion on cladding surfaces has also been gradual and can be expected to remain so. Cladding that has undergone damage while in the reactor core has not resulted in significant releases of radioactivity when stored in pools. Furthermore, the operational experience accumulated since the 1984 Waste Confidence Decision and NRC experience in licensing and inspection reinforce the conclusions in that Decision that wet storage involves a relatively benign environment. There are no driving mechanisms, such as temperature and pressure, to degrade storage structures or components or the fuel itself, or to spread contamination. Degradation mechanisms are gradual and well understood; they allow ample time for remedial action, including repair or replacement of any failing systems. This extensive experience adequately supports predictions of term integrity of storage basins. The Commission also notes the endorsement of this basic confidence by cognizant professional organizations: The American Nuclear Society issued a policy statement [ANS 1986] in 1986 51-SC-29 regarding storage of spent nuclear fuel. The statement indicates that continued wet storage of spent fuel at nuclear power plant sites until the federal government accepts it under existing contracts with the utilities is safe, economical and environmentally acceptable. [See Gilbert, E.R., Bailey, W.J .. and Johnston, A.B., "Results of Studies on the Behavior of Spent Fuel in Storage," Journal of the Institute of Nuclear Materials Managemnt, Vol. XVI. No. 3, April 1988, p. 27.IVA).J Thus, supported by the consistency of NRC experience with that of others, the Commission has concluded that spent fuel can be stored safely and without significant environmental impact, in either wet storage or in wet storage followed by dry storage, for at least 100 years. The Commission considers it unlikely, however, that any fuel will actually remain in wet storage for 100 years or even for 70 years. We anticipate that, consistent with the currently developing trend, utilities will move fuel rods out of spent fuel pools and into dry storage to make room in pools for freshly-discharged spent fuel. Although the Commission has concluded that reactor spent fuel pools can safely be used to store spent fuel for 100 years, there is no technically compelling reason to use them that long. If reactor licenses are renewed for as long as 30 years, making a total of 70 years of operation, it will be necessary to store the spent fuel discharged at the end of the reactor's operation in a spent fuel pool for several years to allow for radioactive decay and thermal cooling. After this period, the fuel could be placed in dry storage and the spent fuel pool decommissioned. Thus, for most reactors, the most likely maximum period of storage will be well within the extended 30-year post-operational period under the Commission's proposed revision to Finding 4. Moreover, considering that under certain conditions spent fuel can be stored safely and without significant environmental impacts for up to 140 years, the Commission believes there is ample basis for confidence in storage for at least 100 years. In its 1984 Waste Confidence Decision, the Commission also concluded that "there are no significant additional non-radiological impacts which could adversely affect the environment if spent fuel is stored beyond the expiration of operating licenses for reactors" (see 49 FR 34658 at p. 34686, August 31, 1984). The Commission did not find anything to contradict this conclusion in its 1988 rulemaking amending 10 CFR part 72 for long-term spent fuel and high-level waste storage at an MRS: PART 51 STATEMENTS OF CONSIDERATION In August 1984, the NRC published an environmental assessment for this proposed revision of part 72 NUREG-1092, 'Environmental Assessment for 10 CFR part 72, Licensing Requirements for the Independent Storage of Spent Fuel and Level Radioactive Waste.' NUREG-1092 discusses the major issues of the rule and the potential impact on the environment. The findings of the environmental assessment are '[1) past experience with water pool storage of spent fuel establishes the technology for long-term storage of spent fuel without affecting the health and safety of the public, (2) the proposed rulemaking to include the criteria of 10 CFR part 72 for storing spent nuclear fuel and high-level radioactive waste does not significantly affect the environment, (3) solid high-level waste is comparable to spent fuel in its heat generation and in its radioactive material content on a per metric ton basis, and (4) knowledge of material degradation mechanisms under dry storage conditions and the ability to institute repairs in a reasonable manner without endangering the health [and safety] of the public shows dry storage technology options do not significantly impact the environment.' The assessment concludes that, among other things, there are no significant environmental impacts as a result of promulgation of these revisions of 10 CFR part 72. Based on the above assessment, the Commission concludes that the rulemaking action will not have a significant incremental environmental impact on the quality of the human environment. [53 FR 31651 at pp. 31657-31658, August 19, 1988.] Thus, the 1988 amendments to 10 CFR part 72 provide the basis for the Commission to conclude that the environmental consequences of term spent fuel storage, including radiological impacts, are not significant. Finally, no considerations have arisen to affect the Commission's confidence since 1984 that the possibility of a major accident or sabotage with offsite radiological impacts at a spent-fuel storage facility is extremely remote. NRC has recently reexamined reactor pool storage safety in two studies, "Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two Representative Nuclear Power Plants" [NUREG/CR-5176) and "Beyond Design Basis Accidents in Spent Fuel Pools" (NUREG-1353). These studies reaffirmed that there are no safety considerations that justify changes in regulatory requirements for pool storage. Both and dry-storage activities have continued to be licensed by the Commission. In its recent rulemaking amending 10 CFR part 72 to establish licensing requirements for an MRS, the Commission did choose to eliminate an exemption regarding tornado missile impact"* *

to assure designs continue to address maintaining confinement of particulate material." (53 FR 31651, p. 31655, August 19, 1988). However, NRC staff had previously considered tornado missile impacts in safety reviews of design topical reports and in licensing reviews under 10 CFR part 72. 4.B. Relevant Jssues*That Have Arisen Since the Commission's Original Decision on Finding 4 In its original Finding 4, the Commission found reasonable assurance of safe storage without significant environmental impacts for at least 30 years beyond reactor 01 expiration.

Delays and uncertainties in the schedule for repository availability since the 1984 Decision have convinced the Commission to allow some margin beyond the scheduled date for repository opening currently cited by DOE. As noted in Finding 2, the Commission has reasonable assurance that at least one repository will be available within the first quarter of the twenty-first century. For all currently operating reactors, this would still be within the period of 30 years from expiration of their OLs, which the Commission previously found to be the minimum period for which spent fuel storage could be considered safe and without significant environmental impact. Under the NWPA as amended, DOE is authorized to dispose of up to 70,000 MTI--IM in the first repository before granting a construction authorization for a second. Under existing licenses, projected spent fuel generation could exceed 70,000 MTI--IM as early as the year 2010. Possible extensions or renewals of OLs also need to be considered in assessing the need for and scheduling the second repository. It now appears that unless Congress lifts the capacity limit on the first and unless this repository has the physical capacity to dispose of all spent fuel generated under both the original and extended or renewed licenses-it will be necessary to have at least one additional repository. Assuming here that the first repository is available by 2025 and has a capacity on the order of 70,000 MTHM, additional disposal capacity would probably not be needed before about the year 2040 to avoid storing spent fuel at a reactor for more than 30 years after expiration of reactor OLs. Although action on a second repository before the year 2007 would require Congressional approval, the Commission believes that Congress will take the necessary action if it becomes clear that the first repository site will not have the capacity likely to be needed. If DOE were able to address the need for a sec"ond repository earlier, for 51-SC-30 example by initiating a survey for a second repository site by the year 2000, DOE might be able to reduce the potential requirement for extended spent fuel storage in the twenty-first century. The Commission does not, however, find such action necessary to conclude that spent fuel can be stored safely and without significant environmental impact for extended periods. The potential for generation and onsite storage of a greater amount of spent fuel as a result of the renewal of existing OLs does not affect the Commission's findings on environmental impacts. In Finding 4, the Commission did not base its determination on a specific number of reactors and amount of spent fuel generated. Rather, the Commission took note of the safety of spent fuel storage and lack of environmental impacts overall, noting that individual actions involving such storage would be reviewed. In the event there were applications for renewal of existing reactor OLs, each of these actions would be subject to safety and environmental reviews, with subsequent issuance of an environmental assessment or environmental impact statement, which would cover storage of spent fuel at each reactor site during the period of the renewed license. The Commission also notes that the amount of spent fuel expected to be discharged by reactors has continued to decline significantly, a trend already noted in the Commission's discussion of its Finding 5 (49 FR 34658 at p. 34687, August 31, 1984). At the time of the Commission's decision, "* *

the cumulative amount of spent fuel to be disposed of in the year 2000 [was] expected to be 58,000 metric tons of uranium" [see "Spent Fuel Storage Requirements" (Update of DOE/RL 17) DOE/RL-83-1, January, 1983). Today, that figure has declined to 40,384 11!-etric tons [see "Spent Fuel Storage Requirements" (DOE/RL-88-34), October 1988, p. A. 17). Thus, the amount of spent fuel considered likely to be discharged by the year 2000 in the Commission's 1984 decision will not be attained until well into the second decade of the twenty-first century, if then. The Commission believes that its 1984 Finding 4 should be revised to acknowledge the possibility and assess the safety and environmental impacts of extended storage for periods longer than 70 years. The principal reasons for this proposed revision are that: (1) The term material and system degradation effects are well understood and known to be minor; (2) the ability to maintain PART 51 STATEMENTS OF CONSIDERATION the system is assured; and (3) the Commission maintains regulatory authority over any spent fuel storage installation.

On the basis of experience with wet and dry spent fuel storage and related rulernaking and licensing actions, the Commission concludes that spent fuel can be safely stored without significant environmental impact for at least 100 years, if necessary. Therefore, the Commission proposes to revise its original Fourth Finding thus: "The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the licensed life for operation (which may include the term of a revised license] of that reactor at its spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations." Original Finding 5 The Commission finds reasonable assurance that safe independent onsite spent fuel storage or offsite spent fuel storage will be made available if such storage capacity is needed. Proposed Finding 5 Sarne as above. 5.A. Issues Considered in Commission's 1984 Decision on Finding 5 In its discussion of Finding 5 of its Waste Confidence Decision (49 FRN 34658, August 31, 1984), the Commission said that: The technology for independent spent fuel storage installations, as discussed under the fourth Commission Finding, is available and demonstrated. The regulations and licensing procedures are in place. Such installations can be constructed and licensed within a five-year time interval. Before passage of the Nuclear Waste Policy Act of 1982 the Commission was concerned about who, if anyone, would take responsibility for providing such installations on a timely basis. While the industry was hoping for a government commitment, the Administration had discontinued efforts to provide those storage facilities.

*
The Nuclear Waste Policy Act of 1982 establishes a national policy for providing storage facilities and thus helps to resolve this issue and assure that storage capacity will be available.

Prior to March 1981, the DOE was pursuing a program to provide temporary storage in off-site, or away-from-reactor [AFR]. storage installations. The intent of the program was to provide flexibility in the national waste disposal program and an alternative for those utilities unable to expand their own storage capacities. Consequently, the participants in this proceeding assumed that, prior to the availability of a repository, the Federal government would provide for storage of spent fuel in excess of that which could be stored at reactor sites. Thus, it is not surprising that the record of this proceeding prior to the DOE policy change did not indicate any direct commitment by the utilities to provide AFR storage. On March 27, 1981, DOE placed in the record a letter to the Commission stating its decision 'to discontinue its efforts to provide Federal government-owned or controlled reactor storage facilities.' The primary reasons for the change in policy were cited as new and lower projections of storage requirements and lack of Congressional authority to fully implement the original policy. The record of this proceeding indicates a general commitment on the part of industry to do whatever is necessary to avoid shutting down reactors or derating them because of filled spent fuel storage pools. While industry's incentive for keeping a reactor in operation no longer applies after expiration of its operating license, utilities possessing spent fuel are required to be licensed and lo maintain the fuel in safe storage until removed from the site. Industry's response to the change in DOE's policy on sponsored away-from-reactor (AFR) storage was basically a commitment to do what is required of it, with a plea for a clear unequivocal Federal policy. * *

The Nuclear Waste Policy Act of 1982 has now provided that policy. The Nuclear Waste Policy Act defines public and private responsibilities for spent fuel storage and provides for a limited amount of federally-supported interim storage capacity.

The Act also includes provisions for monitored retrievable storage facilities and for a research development and demonstration program for dry storage. The Commission believes that these provisions provide added assurance that safe independent onsite or offsite spent fuel storage will be available if needed. [References omitted] The policy set forth in the NWPA regarding interim storage remains in place. Therefore, the Commission's confidence remains unchanged. The only policy change affecting storage involves long-term storage in an MRS. The NWPAA sets schedule restrictions on an MRS by trying it to the repository siting and licensing schedule. These restrictions effectively delay implementation of an MRS. Consequently, its usefulness in providing storage capacity relief to utilities is likely to be lost. Although the Commission's confidence in its 1984 Decision did not depend on the availability of an MRS facility, the possibility of such a facility, as provided for in the NWPA, was one way in which needed storage could be made available. The NWPAA makes an MRS facility less likely by linking it to repository development. The potential impact of the decreased likelihood of an MRS on the Commission's confidence is, however, more than compensated for by 51-SC-31 operational and planned spent fuel pool expansions and dry-storage investments by utilities themselves-developments that had not been made operational at the time of the original Waste Confidence Decision. Consequently, the statutory restrictions that may make an MRS ineffective for timely storage capacity relief are of no consequence for the Commission's finding of confidence that adequate storage capacity will be made available if needed. Although the NWPAA limits the usefulness of an MRS by linking its availability to repository development, the Act does provide authorization for an MRS facility. The Commission has remained neutral since its 1984 Waste Confidence Decision with respect to the need for authorization of an MRS facility. The Commission does not consider the MRS essential to protect public health and safety. If any offsite storage capacity is required, utilities may make application for a license to store spent fuel at a new site. Consequently, while the NWPAA provision does affect MRS development and therefore can be said to be limiting, the Commission believes this should not affect its confidence in the availability of safe storage capacity. 5.B. Relevant Issues That Have Arisen Since the Commission's Original Decision on Finding 5 DOE will not be able to begin operation of a repository before 2003 under current plans, and operation might begin somewhat later. Given progress to date on an MRS, the link between MRS facility construction and repository construction authorization established by the NWPAA, and the absence of other concrete DOE plans to store the spent fuel, it seems unlikely that DOE will meet the 1998 deadline for taking title to spent fuel. (Under section 302(a)(5)(B] of the NWPA, "* *

the Secretary, beginning not later than January 31, 1998, will dispose of the high-level radioactive waste or spent nuclear fuel [ subject to disposal contracts].")

This potential problem does not, however, affect the Commission's confidence that storage capacity will be made available as needed. The possibility of a dispute between DOE and utilities over the responsibility for providing spent fuel storage will not affect the public health and safety or the environment. Uncertainty as to contractual responsibilities raises questions concerning: (1) Who will be responsible; (2) at what point in time responsibility for the spent fuel will be transferred; (3) how the fuel will be PART 51 STATEMENTS OF CONSIDERATION managed; (4) how the transfer of management responsibility from the utilities to DOE will take place; and (5) how the cost of DOE storage might differ, if at all, from utility storage. Utilities possessing spent fuel in storage under NRC licenses cannot abrogate their safety responsibilities, however. Until DOE can safely accept spent fuel, utilities or some other licensed entity will remain responsible for it. If DOE and the utilities can amicably resolve their respective responsibilities for spent fuel storage in the interest of efficient and effective administration of the overall waste management system, including the Nuclear Waste Fund, NRC would gain added confidence in the institutional arrangements for spent fuel management (see also Finding 3 on this issue). Estimates of the amount of spent fuel generated have continued to decline. At the time of the Commission's Decision, the Commission cited in Finding 5 the cumulative figure of 58,000 metric tons uranium of spent fuel generated in the year 2000 (See 49 FR 34658, p. 34697, August 31, 1984.) More recently, DOE estimated 40,384 metric tons (See "Spent Fuel Storage Requirements," DOE/RL-88-34, October 1988, p. A. 17). Although estimates may show an increase at some date well into the twenty-first century if licenses of some reactors are renewed or extended, this possibility does not affect the Commission's confidence in the availability of safe storage capacity until a repository is operational. The industry has made a general commitment to provide storage capacity, which could include from-reactor (AFR) storage capacity. To date, however, utilities have sought to meet storage capacity needs at their respective reactor sites. Thus, a new industry application for AFR storage remains only a potential option, which currently seems unnecessary and unlikely. Utilities have continued to add storage capacity by reracking spent fuel pools, and NRC expects continued reracking where it is physically possible and represents the least costly alternative. Advances in dry-storage technologies and utility plans both have a positive effect on NRC's confidence. At the time the Commission reached its original findings, dry storage of LWR spent fuel was, as yet, unlicensed under 10 CFR Part 72, and DOE's dry-storage demonstrations in support of dry-cask storage were in progress at the Idaho National Engineering Laboratory (INEL). Today, DOE's demonstration efforts have been successful (See Godlewski, N. Z., "Spent Fuel Storage-An Update," Nuclear News, Vol. 30, No. 3 March 1987, pp. 47-52, at p. 47.) Dry storage has been licensed at two reactor sites, and a third application is under review. Dry cask storage is licensed at Virginia Electric Power Company's Surry Power Station site (see License, SNM 2501 under Docket No. 72-2), and concrete module and stainless-steel canister storage is licensed at Carolina Power and Light Company's (CP&L's) H. B. Robinson, Unit 2, site (see License SNM 2502, under Docket No. 72-3). An application is under review for a similar modular system at Duke Power Company's Oconee Nuclear Station site (See Letter to Director, Division of Fuel Cycle and Material Safety, NRC, from Hal B. Tucker, Duke Power Company, dated March 31, 1988, under Docket No. 72-4). A new application has been received in 1989 for CP&L's Brunswick site, and another is expected in 1989 for the Baltimore Gas and Electric Company's Calvert Cliffs site. Applications are also expected for CP&L's Robinson 2 site (at another onsite location to allow for greater storage capacity), Wisconsin Electric Power Company's Point Beach site, and Consumer Power's Palisades site. The Tennessee Valley Authority has indicated that it will apply for its Sequoyah plant site. Thus, the successful demonstration by DOE of dry cask technology for various cask types at INEL, utilities' actions to forestall spent fuel storage capacity shortfalls, and the continuing sufficiency of the licensing record for the Commission to authorize increases in reactor storage capacity all strengthen the Commission's confidence in the availability of safe and environmentally sound spent fuel storage capacity. Renewal of reactor OLs will involve consideration of how additional spent fuel generated during the extended term of the license will be stored onsite or offsite. There will be sufficient time for construction and licensing of any additional storage capacity needed. In summary, the Commission finds no basis to change fhe Fifth Finding in its Waste Confidence Decision. Changes by the NWPAA, which lessen the likelihood of an MRS facility, and the potential for some slippage in repository availability to the first quarter of the twenty-first century (see our discussion of Finding 2) are more than offset by the continued success of utilities in providing safe at-reactor-site storage capacity in reactor pools and their progress in providing independent onsite storage. Therefore, the Commission continues to find"* *

reasonable assurance that safe independent onsite 51-SC-32 spent fuel storage or offsite spent fuel storage will be made available if such storage is needed." Dated at Rockville, Maryland.

this 25th day of September, 1989. For the Nuclear Regulatory Commission. John C. Hoyle, Assistant Secretary of the Commission. 54 FR 43576 Published 10/26/89 Effective 10/26/89 10 CFR Part 51 RIN 3150-AD41 Clarifying Amendment Relating to Enforcement Activities AGENCY: Nuclear Regulatory Commission. ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its environmental regulations to make clear that the provision excluding NRC enforcement activities from the requirements of the National Environmental Policy Act of 1969, as amended, not only encompasses formal enforcement actions but also encompasses informal administrative mechanisms relating to enforcement such as bulletins, information notices, generic letters, notices of deviation, notices of nonconformance and confirmatory action letters. This minor amendment will help to clarify the manner in which this provision will be applied. EFFECTIVE DATE: October 26, 1989. FOR FURTHER INFORMATION CONTACT: Stuart A. Treby, Assistant General Counsel for Rulemaking and Fuel Cycle, U.S. Nuclear Regulatory Commission, Washington, DC 20555; Telephone (301) 492-1636.

SUPPLEMENTARY INFORMATION: On March 12, 1984 (49 FR 9352), the Commission promulgated final regulations implementing section 102(2) of the National Environmental Policy Act of 1969, as amended, (NEPA) in a manner which is consistent with the NRC's domestic licensing and related regulatory authority. These regulations, 10 CFR part 51, subpart A, reflect the Commission's policy of developing regulations voluntarily subject to certain conditions to take account of the regulations of the Council on Environmental Quality (CEQJ implementing the procedural provisions of NEPA. Section 51.14[b) of the NRC regulations adopts certain definitions used in the CEQ regulations. These definitions include the definition of PART 51 STATEMENTS OF CONSIDERATION "Major Federal action" in 40 CFR 1508.18.1 As this definition makes clear, for NEPA purposes, the term "Major Federal action" does not include "* *

bringing judicial or administrative civil or criminal enforcement actions * * * ." This portion of the CEQ definition of "Major Federal action" is highlighted in § 51.10 of the Commission's regulations which addresses the purpose and scope of NRC's regulations implementing section 102(2) of NEPA and states in paragraph

[d): (d) Commission actions initiating or rel a ting to administra live or judicial civil or criminal enforcement actions or proceedings are not subject to section 102(2) of NEPA. These actions include issuance of notices, orders, and denials of requests for action pursuant to subpart B of part 2 of this chapter, and matters covered by part 15 and 160 of this chapter. Although the Commission's regulations make clear that enforcement matters are not subject to the NEPA

process, there has been some uncertainty as to whether certain types of informal administrative actions used by the NRC staff as an adjunct to the Commission's formal enforcement 1 40 CFR 1506.18 states: " 'Major Federal action' includes actions with effects that may be major and which are potentially subject to Federal control and responsibility.

Major reinforces but does not have a meaning independent of significantly(§ 1508.27). Actions include the circumstance where the responsible officials fail to act and that failure to act is reviewable by courts or administrative tribunals under the Administrative Procedure Act or other applicable law as agency action. "[a) Actions include new and continuing activities, including projects and programs entirely or partly financed, assisted, conducted, regulated, or approved by federal agencies; new or revised agency rules, regulations, plans, policies, or procedures; and legislative proposals [ § § 1506.8, 1508.17). Actions do not include funding assistance solely in the form of general revenue sharing funds, distributed under the State and Local Fiscal Assistance Act of 1972, 31 U.S.C. 1221 et seq., with no Federal agency control over the subsequent use of such funds. Actions do not include bringing judicial or administrative civil or criminal enforcement actions. "[b) Federal actions tend to fall within one of the following categories: "[1) Adoption of official policy, such as rules, regulations, and interpretations adopted pursuant to the Administrative Procedure Act, 5 U.S.C. 551 et seq.; treaties and international conventions or agreements; formal documents establishing an agency's policies which will result in or substantially alter agency programs. "[2) Adoption of formal plans, such as official documents prepared or approved by federal agencies which guide or prescribe alternative uses of federal resources, upon which future agency actions will be based. "[3) Adoption of programs, such as a group of concerted actions to implement a specific policy or plan; systematic and connected agency decisions allocating agency resources to implement a specific statutory program or executive directive. "[4) Approval of specific projects, such as construction or management activities located in a defined geographic area. Projects include actions approved by permit or other regulatory decision as well as federal and federally assisted activities." mechanisms, which include the issuance of orders pursuant to 10 CFR part 2, subpart B, are intended to be included within the scope of 10 CFR 51.10[d). See 10 CFR part 2, Appendix C, General Statement of Policy and Procedure for NRC Enforcement ~ctions, section V.H. These informal administrative actions include, among others, various written notices such as bulletins, information notices, generic letters, notices of deviation or non-conformance and confirmatory action letters. As use of the word "include" in the second sentence of§ 51.10[d) makes clear, § 51.10(d) does not purport to provide a comprehensive list of Commission activities relating to enforcement. The types of enforcement actions mentioned were intended to be only illustrative. As described in the Commission's General Statement of Policy on Enforcement, "Confirmatory Action Letters are letters confirming a Jicensee's or a vendor's agreement to take certain actions to remove significant concerns about health and safety, safeguards, or the environment." A confirmatory action letter is an informal enforcement tool issued by the NRC staff pursuant to 10 CFR part 2, Appendix C, V, H, 3. The letter memorializes commitments made by the licensee to the NRC staff that the licensee will take certain specific actions with regard to a facility. The NEPA status of such informal tools, including resumption of plant operation, was not explicitly addressed in § 51.10 since the Commission believed, as it argued in Commonwealth of Massachusetts

v. United States Nuclear Regulatory Commission, No. 88--2211, that such informal enforcement tools did not involve agency action. However, in that case the First Circuit Court of Appeals concluded that resumption of operation of the Pilgrim facility, after an extended shutdown for corrective actions reflected in a Confirmatory Action Letter, involved NRC action to reinstate the license. The Court went on to uphold the NRC actions against related challenges.

The case did not raise and the Court did not address any NEPA related issues. Licensee actions undertaken voluntarily, as documented in a confirmatory action letter, are generally directed to restoring compliance with NRC regulations, thereby enabling the licensee to resume licensed activities. Consequently, the only environmental effects of the licensee's voluntary actions to reestablish that licensed activities will be undertaken in accordance with the license are those evaluated at the time the facility or activity was licensed and assessed in the NRC Environmental Impact Statement prepared in connection with the initial issuance of the license and in 51-SC-33 subsequent environmental evaluations in connection with license amendments. The environmental effects of NRC activities associated with the supervision of such licensee actions, including NRC approval and supervision of the licensee's subsequent resumption of licensed activities are the same and do not require additional environmental review. Although the Commission did not intend§ 51.10(d) of its NEPA regulations to be read as if it applied exclusively to the types of enforcement activities specifically enumerated therein, it recognizes that clarification would be helpful. Accordingly, the Commission is promulgating this final rule. It should be clearly understood that it has always been contemplated, under § 51.10(d), that when licensee actions to remediate the matters underlying the enforcement action have been completed to the satisfaction of the Commission, the conditions of operation previously reviewed in an environmental context will be restored. Accordingly, when the NRC authorizes licensed activities to resume, no additional environmental review pursuant to NEPA or the Commission's regulations is needed. If it should be necessary for the licensee to obtain a license amendment to restore compliance with the Commission's safety requirements in order to satisfy the concerns underlying the enforcement action, any environmental effects associated with issuance of the license amendment would either be addressed in an Environmental Assessment or iencompassed by a categorical exclusion

under 10 CFR 51.22[c).

In this way, )appropriate consideration of any lenvironmental impact would be assured. ; Because this amendment is merely ,clarifying and interpretative in nature, .relates solely to matters of agency *practice and does not involve a significant question of policy, good cause exists for omitting notice of proposed rulemaking and public :procedures thereon as unnecessary and *for making the amendment effective upon publication in the Federal Register ;without the customary thirty day notice. Environmental Impact: Categorical

Exclusion The NRC has determined that this :final rule is the type of action described

in categorical exclusion 10 CFR ,51.22[c)[2).

Therefore, neither an ,environmental impact statement nor an environmental assessment has been ;prepared for this final regulation. Paperwork Reduction Act Statement This final rule contains no new or ,amended information collection requirements and therefore is not subject to the requirements of the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). PART 51 STATEMENTS OF CONSIDERATION Regulatory Analysis Questions have arisen as to whether 10 CFR 51.lO(d), which excludes Commission actions relating to enforcement from the NEPA process, encompasses informal administrative actions such as those described in section V.H. of the Commission's General Statement of Policy and Procedure for NRC Enforcement Actions, 10 CFR part 2, Appendix C, i.e., bulletins, information notices, generic letters, notices of deviation or nonconformance and confirmatory action letters. Section 51.10(d) of the Commission's regulations is not limited to a portion of the Commission's enforcement activities but is inclusive. The NRC staff has a need to assure a uniform understanding of the scope of actions encompassed by this regulation. This rule change revising the text of§ 51.lO(d) to make clear that it applies to the entire spectrum of the Commission's enforcement activities is the appropriate means to achieve this end. Backfit Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this final rule because this amendment to 10 CFR 51.10(d) does not contain any provisions which impose backfits as defined in 10 CFR 50.109[a)(l) and therefore a backfit analysis is not required. List of Subjects in 10 CFR Part 51 Administrative practice and procedure, Environmental impact statement, Nuclear materials, Nuclear power plants and reactors, Reporting and recordkeeping requirements. For the reasons set forth in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, the National Environmental Policy Act of 1969, as amended, and 5 U.S.C. 552 and 533, the NRC is adopting the following amendment to 10 CFR part 51: 54 FR 50735 Published 12/11 /89. Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Reactors; Correction See Part 52 Statements of Consideration 54 FR 53312 Published 12/28/89. Effective 12/28/89 Statement of Organization and General Information; Minor Amendments See Part 1 Statements of Consideration 55 FR 38472 Published 9/18/90 Effective 1 0/18/90 10 CFR Part 51 RIN 3150*AD26 Consideration of Ehvlronmental Impacts of Temporary Storage of Spent Fuel Alter Cessation of Re:1ctor Operation AGENCY: Nuclear Regulatory Commission. ACTION: Final rule. *-----

SUMMARY

The Nuclear Reguh1tory Commission is revising its generic determinations on the timing of availability of a geologic repository for commercial high-level radioactive waste and spent fuel and the environmental impacts of storage of spent fuel at reactor sites after the expiration of reactor operating licenses.

These revisions reflect findings of the Commission reached in a five-year update and supplement to its 1984 "Waste Confidence" rulemaking proceeding, which are published elsewhere in this issue of the Federal Register. The Commission now finds that spenl fuel generated in any reactor can be stored safely and without significant environmental impacts in reactor facility storage pools or independent spent fuel storage installations located at reactor or from-reactor sites for at least 30 years beyond the licensed life for operation [which may include the term of a revised or renewed license). Further, the Commission believes there is reasonable assurance that at least one mined geologic repository will be available within the first quarter of the first century, and sufficient repository capacity will be available within 30 years beyond the licensed life for operation of any reactor to dispose of the commercial high-level waste and spent fuel originating in such reactor and generated up to that time. EFFECTIVE DATE: October 16, 1990. FOR FURTHER INrORMATION CONTACT: John P. Roberts, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone: (301) 492-0608. SUPPLEMENTARY INFORMATION: Background In 1984, the Commission conclm!ed a generic rulemaking proceeding, the "Waste Confidence" proceeding, to reassess its degree of confidence that radioactive wastes produced by nuclear facilities will be safely disposed of, to determine when any such disposal would be available, and whether such wastes can be safely stored until they 51-SC-34 are safely disposed of. The Comm*Gsion found that there was reasonable assurance that one or more min*'ri geologic repositories for com*ni.~t::id h:gh-level radioactive waste an..: r.p,,r..t fuel will be available by 2007-200S. However, some reactor operating licenses might expire without being renewed or some reactors might be 11crmanently shut down prior to this pi1*,od. Since independent spent fuel storuge installntions had not yet been extensively developed, there was a probability that some onsite spent fuel storage after license expiration might be necessary or appropriate. In addition, the possibility existed that spent fuel might be stored in existing or new storage facilities for some period beyond 2007-2009. The Commission also found

that the licensed storage of spent.fuel for at least JO years beyond the reactor operating license expiration either at or away from the reactor site was feasible, safe. and would not result in a significant impact on the environment.

Consequently, the Commission adopted e rule, codified in 10 CFR 51.23, providing that the environmental impacts of at-reactor storage after the tennination of reactor operating licenses need not be considered in Commission proceedings related to issuance or amendment of a reactor operating license. The same safety and environmental considerations applied to fuel storage installations licensed wider part 72 as for storage in reactor basins. Accordingly, the rule also provided that the environmental impacts of spent fuel storage at independent spent fuel storage installations for the period following expiration of the installation storage license or amendment need not be considered in proceedings related to issuance or amendment of a storage installation license. Amendment to Part 51 At the time of issuance of its Waste Confidence decision and the adoption of 10 CFR 51.23, the Commission also announced that while it believed that could, with reasonable assurance, reach favorable conclusions of confidence, it also recognized that significant unexpected events might affect its decision. Consequently, the Commission stated* that it would "review its conclusions on* waste confidence should significant and pertinent unexpected events occur, or at least every 5 years until a repository for high-level radioactive waste and spent fuel is available." The Commission has now completed a five-year review of its . earlier findings. A description of this review end the supplement and update to the earlier findings is announced elsewhere in this issue. As a result of this review, the Commission is modifying two of its earlier findings as follows: PART 51 STATEMENTS OF CONSIDERATION

n,e Commission finds reasonable assurance th:,t-a,i lt~ast one mined geologic repositorJ will be.a~u1!.1ble wilhin the first quarter of 11:e twenty-first century, and sufficient ri;pos,l~ry copaci!y will be available wilhin 30,years beyond the licensed life for operation (which may include the term of a revised or renewed license) of any reactor to dispose of the commercial high-level waste and spent'fael originating in such reactor and generated up to that time; and The Commission finds reasonable assurance that. if necessary, spent fuel generatr.d in any reactor can be stored safely and without significant environmental impacts for at least 30 years bey1md the licensed life for operation (which may include the term of a revised or renewed license) of that reactor at its spent fuel storage basin, or at either onsite or offsi te independent spent fuel storage installations.

In this proceeding, the r.ommission is revising 10 CFR 51.23[a) to !*f'! consistent with these revisions to the Wn.:irn CW1fidence decision. Summary of Comments . The Commission received 11 *~omments on iti.i proposed revision to 10 CFR 51.23[a) from the following entities lusted in the order of receipt of comments: Duke Power Company Public Citizen F..dison Electric Institute Malachy Murphy (State of Nevada) Yankee Atomic Electric Company Department of Energy (DOE) Philadelphia Electric Company Commonwealth Edison Virginia Electric end Power Company Mervin I. Lewis, Registered Professional Engineer Florida Power & Light The revision to this rule was supported by Duke Power Company, Edison Electric Institute, Y&nkee Atomic Electric Company, Department of Energy, Philadelphia Electric Company. and Virginia Electric and Power Company and generally supported by Commonwealth Edison. Malachy Murphy, for the State of Nevada, suggests that 10 CFR 51.23(a) be amended to reflect reasonable assurance that spent fuel can be stored safely and without significant environmental risk in dry casks at 1eactor sites for up to one hundred years. The Commission, in the notice of proposed rulemaking, discussed its conclusion that even if storage of spent fuel were necessary for at least thirty years beyond the licensed life for operation of reactor.s, which for a reactor whose license is renewed for thirty years would mean a period of at least 100 years, such storage is feasible, safe and would not result in a significant impact on the environment. The Commission's conclusion on this issue considers both wet and dry storage. Al though the Commission does not dispute the statement that dry spent fuel storage is safe and environmentally acceptable for a period of 100 years, the Commission does not find it necessary to make that specific finding in this *proceeding. Marvin I. Lewis avers that 100 years is an excessive amount of time to predict that at-reactor storage will be available and safe. The commenter suggests that our institutions may not survive in a form that will provide safe onsite storage 100 years in the future. The commenter requests that the Commission reverse its finding that storage will be available and safe for 100 years. The Commission does not agree with the commenter that this finding should be reversed. The Commission believes that adequate regulatory authority exists and will remain available to require any measures necessary to assure safe storage of spent fuel. Conclusions The Commission is adopting the proposed revision with one small clarifying change. The proposed revision to 10 CFR :'il.23[a) [and the proposed revision to the Waste Confidence decision) stated that spent fuel can be stored safely for at least 30 years beyond the licensed life for operation of any reactor which may include the term of a "revised license." As the discussion in the notice made explicit, the term *"revised" license was intended to embrace a "renewed" license. To reflect more accurat.ely the inclusion of the term of a re~ewed license, the parenthetical phrase which refers to this subject is being revised to read: "which may include the term nf a revised or renewed license." The necessity for foe proposed revisions to the Waste Confidence decision and to 10 CFR 51.23(a) is based on the timing of repository availability, and premised on the following factors: The potential for delays in DOE's program; the mandate of the Nuclear Waste Policy Act Amendments of 1987 to characterize only the Yucca Mountain site which m*eans that if that site is found unsuitable, characterization will have to begin at another site or,suite of .. sites with consequent delay in repository availability; the regulator~ . need to avoid premature commitment to the Yucca Mountain site; and the questionable value of making predictions about completion of a project as complex and unique as the repository in terms of years when decades would be more realistic. But even with this chal}ge the Commission has concluded that it has reasonable assurance that on such a schedule for repository availability, sufficient repository capacity will be available within 30 years beyond the licensed !ifo 51-SC-35 for operation of reactors. Adequate regulatory authority is available to require any measures necessary to assure safe storage of the spent fuel until a repository is available. In addition, the Commission has concluded that even if storage of spent fuel ware necessary for at least 30 years beyond the licensed life of reactors, which in the case of a reactor whose operating license is renewed for 30 years would mean for a period of at least 100 years, such *storage is feasible, safe and would not result in a significant impact on the environment. The Commission's conclusions with* respect to safety and environmental impacts of extended storage are supported by NRC's Environmental Assessment (EA) for the 10 CFR part 72 rulemaking "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste" (53 FR 31651, August 19, 1988). Ongoing licensing and operational experience as well as studies of extended pool storage continue to demonstrate that such storage is a benign environment for spent fuel which does not lead to significant degradation of spent fuel integrity. Significant advances in the processes of dry storage of spent fuel continue to demonstrate that dry storage systems are simple, passive and easily maintained. NRC staff safety reviews of topical reports on dry storage system designs and dry storage installations nt two reactor sites, ns well as the EA for part 72, support the finding that storage of spent fuel in such installations for a period of 70 years does not significantly impact the environment. No significant additional non-radiological consequences which could adversely effect the environment for extended storage at reactors and independent spent fuel storage installations have been identified. In sum, the long-term material and system degradation effects are well understood and known to be .minor, the ability to maintain a spent . fuel storage system is assured, and the Commission maintains regulatory authority over any spent fuel storage installation. Environmental Impact This final rule amends 10 CFR part 51 of the Commission's regulations to modify the generic determina lion currently codified in part 51 which was made by the Commission in the Waste Confidence rulemaking proceeding. That generic determination was that for at least 30 years beyond the expiration of a reactor's operating license no significant environmental impacts will result from the storage of spent fuel in reactor facility storage pool or independent spent fuel storage installations located PART 51 STATEMENTS OF CONSIDERATION -at reactor or away-from-reactor sites. The modification provides that, if necessary, spent fuel generated in a reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the licensed life for operation of any reactor. The licensed life for operation of a reactor may include the-term of a revised or renewed license. The environmental analysis on which the revised generic determination is based can be found in the revision and supplement to the Waste Confidence findings published elsewhere in this Issue; This final rulemaking action formally incorporatirig the revised generic determination in the Commission's regulations does not have separate independent environmental impact. The supplemental assessment and revisions to the Waste Confidence findings are available for inspection at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 ( 44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget approval number 3150-0021. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(bJ, the Commission certifies that this rule will not have a significant economic impact on a substantial number of small entities. The rule describes a revised basis for continuing in effect the curre_nt provisions of 10 CFR 51.23(b) which ' provides that no discussion of any environmental impact of spent fuel -storage in reactor facility storage pools . or independ,mt spent fuel storage installations [ISFSI] for the period following the term of the reactor operating license or amendment or initial ISFSl licen:;e or amendment 'for whil:h application is made is required in any environmental report, environmental impact statement, environmental assessment or other analys1s prepared in connection with cedain actions. This rule affects only the licensing and opc~ration of nucle!1r power pl,mtg. Entities seeking or holding Commission licenses for such facilities do not fall within the scope of the definition of small businesses found in section 34 of the Small Business Act, 15 U.S.C. 632, in the Small Business Size Standards set out in regulations issued bv the Small Business Administration at 13 CFR pr,rl 121. or in the NRC's size standards published December 9. 1985 (50 FR 50241]. Backfit Amilysis This final rule does not modify or add to systems. stmctures, components or design of B facility: the design approval or manufacturing license for a facility: or the procedures or organization required to design, construe~ or operate a facility. Accordingly, no backfil analysis pursuant to 10 CFR 50.'lo<J(c) is required for this final rule. List of Subjects in 10 CFR Part 51 : Administration practice and iprocedure. Environmental impact /statement, Nuclear materhils, Nuclear µower plants und reuctors. Reporting ,md rccordkeeping requirements. For the reasons set out in the preamble and under the autbority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendment to 10 CFR part 51. 55 FR 38474 Published 9/18/90 10 CFR Part 51 Waste Confidence Decision Review AGENCY: Nuclear Regulatory Commission. ACTION: Review and Final Revision of Waste Confidence Decision.

SUMMARY

On August 31, 1984, the Nuclear Regulatory Commission (NRC) issued a final decision on what has come to be known as its "Waste Confidence Proceeding." The purpose of that proceeding was " ... to assess generically the degree of assurance now available that radioactive waste can be safely disposed of, _to determine when such disposal or offsite storage will be available and to determine whether radioactive waste can be safely stored onsite past the expiration of existing facility licenses until offsite disposal or storage is available." (49 FR 34658). The Commission noted in 1984 that its Waste 51-SC-36 Confidence Decision was unavoidably in the nature of a prediction, and committed to review its conclusions " ... should significant and pertinent unexpected events occur or at least every five years until a repository is available." The purpose of this notice is to present the findings of the Commission's first review of that Decision.

The Commission has reviewed its five findings and the rationale for them in light of developments since 1984. This *revised Waste Confidence Decision supplements those 1984 findings and the environmental analysis supporting them. The Commission is revising the second and fourth findings in the Waste Confidence Decision as follows: Finding 2: The Commission finds reasonable assurance that at least one mined geologic repository will be available within the first quarter of the twenty-first century, and that sufficient repository capacity will be available within 30 years beyond the licensed life for operation [which may include the term of a revised or renewed license) of any reactor to dispose of the commercial high-level radioactive waste and spent fuel originating in such reactor and generated up to that time. Finding 4: The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the licensed life for operation (which may include the term of a revised or renewed license) of that reactor at its spent fuel storage basin, or at either onsite or off site independent spent fuel storage installations. The Commission is reaffirl)Jing the 'remaining findings. Each finding, any revisions, and the reasons for revising or reaffirming them are set forth in the body of the review below. The Commission also issued two companion rulemaking amendments at the time it issued the 1984 Waste Confidence Decision. The Commission's reactor licensing rule, 10 CFR part 50, was amended to require each licensed reactor operator to submit, no later than five years before expiration of the operating license, plans for managing spent fuel at the reactor site until the spent fuel is transferred to the Department of Energy (DOE) for disposal under the Nuclear Waste Policy Act of 1982 [NWPA). 10 CFR part 51, the rule defining NRC's responsibilities under the National Environmental Policy Act (NEPA), was amended to provide that, in connection with the issuance or amendment of a reactor operating license or initial license for an independent spent fuel storage installation, no discussion of any PART 51 STATEMENTS OF CONSIDERATION environmental impact of spent fuel storage is required for the period following expiration of the license or amendment applied for. In keeping with the revised Findings 2 and 4, the Commission is providing elsewhere in this issue of the Federal Register conforming amendments to its 10 CFR part 51 rule providing procedures for considering in licensing proceedings the environmental effects of extended onsite storage of spent fuel. Finally, the Commission is extending the cycle of its Waste Confidence reviews from every five years to every ten until a repository becomes available. In its 1984 Decision, the Commission said that because its conclusions were " ... unavoidably in the nature of a prediction," it would review them " ... should significant and pertinent unexpected events occur, or at least every five years until a repository .. .is available." As noted below, the Commission now believes that predictions of repository availability are best expressed in terms of decades rather than years. To specify a year for the expected availability of a repository decades hence would misleadingly imply a degree of precision now unattainable. Accordingly, the Commission is changing its original commitment in order to.review its Waste Confidence Decision at least every ten years. This would not, however, disturb the Commission's original commitment to review its Decision whenever significant and pertinent unexpected events occur. The Commission anticipates that such events as a major shift in national policy, a major unexpected institutional development, and/ or new technical information might cause the Commission to consider reevaluating its Waste Confidence Findings sooner than the scheduled ten-year review. FOR FURTHER INFORMATION CONTACT: John Roberts, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (202) 492-0608. SUPPLEMENTARY INFORMATION: Analysis of Public Comments on the Proposed Waste Confidence Decision Review. 1.0 Introduction Comments were received from a Federal agency, the public interest sector, the nuclear industry, and one State as listed below in order of their receipt: Duke Power Company Public Citizen Edison Electric Institute Malachy Murphy (State of Nevada) Yankee Atomic Electric Company Department of Energy Philadelphia Electric Company Commonwealth Edison Virginia Electric and Power Company Marvin I. Lewis, Registered Professional Engineer' Florida Power & Light Company The majority of the commenters were supportive of the Commission's proposed decision and rule. The comments were consolidated into a total of 19 issues to be addressed. Each of these issues is discussed under the Commission finding to which it relates. Two additional issues, not raised by commenters, are treated under the heading "Other Relevant Issues." The "Other Relevant Issues" section includes consideration of the petition by the State of Vermont to intervene in the consideration of the extension of the operating license for Vermont Yankee and the potential for non-payment of the one-time fee for spent nuclear fuel generated prior to April 1983 into the Nuclear Waste Fund. 2.0 Analysis of Issues Related to Commission Findings 2.1 The Commission's First Finding The Commission finds reasonable assurance that safe disposal of high-level radioactive waste and spent fuel in a mined geologic repository is technically feasible. Issue No. 1: Technical Feasibility of Safe Disposal in a Mined Geologic Repository Comment The commenter representing Public Citizen (PC) stated that there is still not adequate assurance that permanent, safe disposal of high-level radioactive waste in a mined geologic repository is technically feasible. In support of this, the commenter indicated that a number of major scientific panels have pointed out that there is no technical or scientific basis for knowing for sure that geologic disposal is possible. As an example, PC stated that President Carter's Office of Science and Technology Policy (OSTP) found in 1979 a rather general consensus among scientists that a technology base "sufficient to permit complete confidence in the safety of any particular repository design or the suitability of any particu\ar site" was still lacking. PC further stated that more recently, a Waste Isolation Systems Panel of the National Academy of Sciences pointed out many areas of the geologic disposal problem where technical uncertainties exist, and where "more information is needed." PC also stated that the technical difficulties presented by a million-year disposal 51-SC-37 problem are unprecedented and enormous, and that there have been no major findings since [the above studies) that have resolved the uncertainties to the point where it is possible to be assured that geologic disposal is technically feasible. NRG Response The issue of the technical feasibility of the safe disposal of spent nuclear fuel and radioactive waste has been addressed at length in the Commission's 1989 Proposed Waste Confidence Decision Review (54 FR 39767; September 28, 1989) as well as in the original 1984 Waste Confidence Decision (49 FR 34658; August 31, 1984). While those discussions addressed the concerns raised by the comment, it is useful to provide additional specific responses to them. The comment that major scientific panels have pointed out that there is no technical or scientific basis for knowing for sure that geologic disposal is possible makes reference to President Carter's OSTP statement in 1979. Contrary to the comment, the OSTP statement does not support the contention that there is no technical or scientific basis for knowing for sure that geologic disposal is possible. Rather, it remarks on the lack of a technology base sufficient to permit complete confidence in the safety of any particular repository design or the suitability of any particular site. The information base necessary to license a repository is still being developed. This includes information on site characterization, repository design, waste package design, and the performance assessment of the entire disposal system. The complete body of such necessary information is expected to be in hand only at the completion of the developmental studies and characterization work being undertaken by the DOE. It is at this point that the DOE will be in a position to apply for a license from the NRC and seek NRC's approval of the safety of its proposed site and repository design. The Commission also notes that the OSTP statement was made over a decade ago, prior to the completion of a substantial amount of work which has addressed many of the issues related to disposal technology. While the Commission recognizes that more information is needed and that the technical difficulties are challenging, there is no basis to believe that safe disposal in a repository is impossible, or even that it is not likely. No major breakthrough in technology is required to develop a mined geologic repository. Rather, there is a need to add to the current extensive body of technical PART 51 STATEMENTS OF CONSIDERATION information already available and apply it to an evaluation of specific sites and engineering designs. Regarding the commenter's emphasis on the need for resolution of uncertainties to assure the technical feasibility of geologic disposal, we would respond that the Commission did not state that the feasibility of a mined geologic repository was assured, in the absolute sense, but that it had found reasonable assurance in the feasibility of mined geologic disposal on the basis of a thorough review of the technologies needed to achieve this disposal. Issue No. 2: Difficulty in Evaluating Compliance with Repository Safety Standards Over Long Time Periods Comment The PC commenter also raised the issue of what he termed the "inability to predict with a reasonable degree of certainty that, once buried, the waste will remain contained [in the geologic repository] for the required time period." The commenter noted uncertainties related to geologic stability, engineered barriers, rock-waste interactions, and groundwater hydrology which contribute to the difficulty of evaluating compliance with safety standards over the long time periods involved in radioactive waste isolation. The commenter concluded that although these problems may be able to be resolved, there is not a basis for assurance that this will be the case. NRG Response The NRC believes that existing safety assessment techniques have the potential to provide a basis for deciding whether proposed radioactive waste disposal systems are acceptable. We recognize the difficulty of predicting with a high degree of accuracy the maximum impacts a repository would have on human health and the environment, especially in the very far future. It will likely not be possible to test empirically the ability of models to predict long-term repository performance to the same extent as models for short-term performance. However, we believe existing technology can provide a sufficient level of safety for present and future generations under certain conditions. These conditions include addressing the uncertainties inherent in projecting far into the future and in modelling complex heterogeneous natural systems, and acquiring and evaluating data on specific sites. We also note that the language of the original Environmental Protection Agency's (EPA) Environmental Radiation Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Wastes (40 CFR part 191) does not require absolute assurance that containment requirements will be met. Rather, it recognizes the uncertainties involved in projecting repository performance far into the future, and states "Instead, what is required is a reasonable expectation, on the basis of the record before the implementing agency, that compliance with Sec.191.13(a) will be achieved." Issue No. 3: Unanticipated Difficulties in Developing the WIPP Facility Comment PC also indicated that the Waste Isolation Pilot Plant (WIPP) has not opened because of numerous unanticipated difficulties, including leakage of salt water into the site. PC states that this leakage, which was not anticipated prior to the beginning of construction in the early 1980s, shows that even on a scale of a few years, geologic events in a repository are unpredictable--to say nothing of events on a time scale of hundreds of thousands of years. NRG Response Although the NRC does not have oversight responsibility for the WIPP project, NRC does monitor DOE progress on WIPP insofar as it may offer valuable insight into efforts to license a repository for commercial high-level waste and spent fuel. For example, DOE must demonstrate compliance with the EPA standard in order to operate the WIPP facility. NRC cognizance of DOE efforts to implement the EPA Standard at WIPP could help provide information and consensus-building in the implementation of the EPA Standard for the commercial high-level waste repository. The NRC does not consider the occurrence of brine pockets at the WIPP site as a factor that might diminish its confidence in the technical feasibility of a mined geologic repository. The Commission does not expect that site characterization of a candidate site will proceed free from all difficulty. We have urged DOE to establish a planning mechanism for timely development and implementation of contingency plans at Yucca Mountain to address problems during site characterization as they arise. DOE has announced a new focus on surface-based testing for the Yucca Mountain site in its Reassessment Report to Congress. Under this program, the primary goal of testing is to identify features of the site which would render it unsuitable for a repository. If such features are identified, DOE would notify Congress and the State of Nevada, and terminate site specific 51-SC-38 activities. A finding that the Yucca Mountain site is unsuitable would likely lead to delays in repository availability while another candidate site is identified and characterized, however it would not diminish confidence in the technical feasibility of geologic disposal. Issue No. 4: Impact of the BEIR V Report on the Commission's Decision Comment Marvin Lewis drew attention to the recent findings of the Committee on the Biological Effects of Ionizing Radiation (BEIR VJ in their report on the Health Effects of Exposure to Low Levels of Ionizing Radiation. The commenter stated that the BEIR V study indicated that the danger from radioactivity is four or more times higher than previously known. The commenter further stated that the BEIR V findings will require that the NRC change many of its radiation protection guidelines and rules. He also requested that the NRC stop all action on the Waste Confidence Decision Review until the Commission can determine the effect of the BEIR V report on the Decision. NRG Response The Commission has been aware for some time of the scientific data underpinning the estimate of risk from radiation exposure contained in the BEIR V report. Much of this information has been incorporated in the Commission's forthcoming revisions to its radiation protection requirements (10 CFR part 20). For reasons stated below, however, the Commission does not foresee any impact of the BEIR V report on the Waste Confidence Decision. The BEIR V report is the latest in a series of reports dealing principally with the effects of low-LET radiation in humans, e.g., radiation such as beta particles and gamma photons. The report covers radiation carcinogenesis, genetic effects, and effects on the developing embryo/fetus. The report also includes new information related to the dosimetry of the Japanese atomic bomb survivors, and new epidemiological information. The NRC staff, other Federal agencies, and national and international organizations are currently reviewing both the BEIR V report and the report issued in 1988 by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEARJ. The estimates of risk due to low-LET radiation in the BEIR V report are based principally upon effects observed in populations exposed to high doses and at high dose rates. These effects are then extrapolated using statistical modeling to predict effects at low doses PART 51 STATEMENTS OF CONSIDERATION and dose rates. The extrapolations to low dose and dose rate lead to significant uncertainties in the estimates of risk in the BEIR V report. The estimates of risk for fatal cancer induction in the BEIR V report are from three to four times larger than the estimate from the preferred model of the BEIR III report in 1980. However, the new BEIR V estimate is within the overall range of risk estimates and uncertainties from the different models presented in BEIR III. It is important to note that the BEIR V report only addresses the issue of risk estimates for radiation effects. The BEIR committee did not make any recommendations on acceptable risk or on the potential impacts of the risk estimates to dose limits or standards for radiation protection. Efforts are underway by the International Commission on Radiological Protection [ICRP), National Council on Radiation Protection and Measurements [NCRP), and the Committee on Interagency Radiation Research and Policy Coordination (CIRRPC) of the Executive Office of the President to reach some measure of consensus on the impacts of the revised risk estimates to radiation protection standards. Under section 121[a) 'of the Nuclear Waste Policy Act (NWPA), NRC is required to issue technical requirements and criteria that it will apply in approving or disapproving a repository. These requirements and criteria must be consistent with the high-level waste disposal standards promulgated by the Environmental Protection Agency. Demonstration of compliance with the EPA standard was discussed under the rationale for Finding 1 in the Commission's Proposed Waste Confidence Decision Review. The NRC does not believe that numerical criteria for individual protection requirement~ are at issue in its Waste Confidence Proceeding. The broader issue of demonstrating compliance with EPA release limits using probabilistic analyses was a concern of the NRC staff and the NRC's Advisory Committee on Nuclear Waste in preparing the Proposed Waste Confidence Decision Review. As stated in the Proposed Waste Confidence Decision Review, the NRC staff is closely monitoring EPA's progress on issuing its revised standards to assure that EPA methodologies for demonstrating compliance with them can be applied by NRC to evaluate DOE's,demonstration of compliance. NRC will also monitor DOE efforts to demonstrate compliance with the EPA standard at the Waste Isolation Pilot Plant facility for transuranic wastes. 2.2 The Commission's Second Finding The Commission fi~ds reasonable assurance that at least one mined geologic repository will be available within the first quarter of the twenty-first century, and that sufficient repository capacity will be available within 30 years beyond the licensed life for operation [which may include the term of a revised or renewed license) of any reactor to dispose of the commercial level radioactive waste and spent fuel originating in such reactor and generated up to that time. Issue No. 5: Expected Date for Repository Availability Comment Malachy Murphy (State of Nevada) and Public Citizen expressed a lack of support for the Commission's proposed second finding. These commenters argue that the finding should be revised to reflect the 2010 date for repository availability announced in DOE's November 1989 Reassessment Report to Congress. They believe that the NRC's confidence" date of 2025 for repository availability may be exceeded if the Yucca Mountain site is found to be unsuitable sometime after the year 2000 because there might not be enough time to locate, characterize, license and construct a repository at another site by 2025. The commenter from Public Citizen also finds that even if the Yucca Mountain site were found to be suitable, a repository there might not be available until after 2025. This commenter concluded that it would be more conservative to assume that four candidate sites would be found to be unsuitable during the course of site characterization and that there is no basis for assurance that a repository would be available before 2055. NRG Response The NRC does not believe it is necessary to change the proposed second finding to reflect DOE's revised date for repository availability of 2010. NRC anticipated an extension of several years in DOE's schedule when it issued its proposed revised second finding. NRC took the position that if the Yucca Mountain site were found to be unsuitable on or before the year 2000, it was reasonable to expect that an alternative site could be identified and developed in time for repository availability by 2025. NRC continues to believe that if DOE determines that the Yucca Mountain site is unsuitable, it will make this determination by about the year 2000. DOE's program is now focused on surface-based testing designed to identify features of the site which would 51-SC-39 render it unsuitable for a repository. The only significant barriers to DOE proceeding with site characterization at Yucca Mountain are the development of a quality assurance [QA) program acceptable to NRC, completion of study plans for site characterization activities they wish to begin, and resolution of the impasse between DOE and the State of Nevada regarding permits for drilling. DOE has made significant progress in the development of a QA program for its site characterization activities. It is possible that this work will be completed and accepted by late 1990 or early 1991. Regarding the impasse with the State of Nevada, both DOE and the State of Nevada have filed lawsuits in Federal Court in an effort to resolve the question of site access. While any litigation of this matter has the possibility of an unfavorable outcome for DOE, the Commission believes that Congress has aggressively demonstrated in both the Nuclear Waste Policy Act of 1982 and the Nuclear Waste Policy Amendments Act of 1987 that it is committed to an orderly progression of the repository program and a resolution of the radioactive waste disposal problem. Accordingly, NRC believes that it is reasonable to assume that Congress will not allow the uncertainties related to the start of site characterization to continue for many more years. For these reasons, NRC believes that the coming decade will be ample time for the DOE to determine whether or not Yucca Mountain is unsuitable and to begin work on an alternate site, if necessary. We believe that Congress is committed to a resolution of the waste problem and will take measures to bring this issue to a close. We would also point out here that the Court decision that led to the Waste Confidence Proceeding did not require NRC to determine when a repository would be available. The Court remanded to NRC the question of " ... whether there is reasonable assurance that an offsite storage solution will be available by the years 2007-2009, the expiration of [Prairie Island and Vermont Yankee's] operating licenses, and if not, whether there is reasonable assurance that the fuel can be safely stored at the reactor sites beyond those dates." NRC chose as a matter of policy not to confine itself to the storage-related questions in the Court's remand, but to address the broader issues of whether radioactive wastes could be safely disposed of, when such disposal would be available, and whether such wastes can be safely stored until they are disposed of. NRC PART 51 STATEMENTS OF CONSIDERATION was not requested to determine nor has it made a determination that a repository must be available by 2025 in order to protect public health and safety. NRC does not find a reasonable basis for the argument that even if the Yucca Mountain site were found to be suitable, it might not be available by the year 2025. Surface-based and in-situ testing are expected to take approximately ten years. The NWPA provides that NRC's review of DOE's license application is to be completed in three years (with the possibility of an additional year]. Construction is scheduled to take another six years. Even if each of these activities were to take several years longer than planned, a repository at Yucca Mountain could be available well before the year 2025. The limiting condition appears to be the timing of DOE's access to the site to begin testing. Finally, we do not believe it is realistic to assume for conservatism that four candidate sites will be found unsuitable before an acceptable site is characterized, licensed and built. To date, no candidate site for a repository has been found to be unsuitable for technical reasons. However, if the Yucca Mountain site is found to be unsuitable, an alternative site would have to undergo a similar process of site-screening and characterization to determine its suitability. We believe it is reasonable to expect that experience gained in the Yucca Mountain site characterization effort would provide a better basis for choosing an alternative site. Furthermore, it may be possible to complete site suitability testing at another site at a faster pace than at Yucca Mountain given the benefits of lessons-learned at that site. Issue No. 6: Clarification of the NRC's Role in the Licensing Support System (LSS} Comment The DOE commented that it was not clear what NRC meant by the words "implementing it" in the statement "DOE has the responsibility for designing the LSS and bearing the costs associated with it and NRC will be responsible for implementing it." NRG Response In its Proposed Waste Confidence Decision Review, NRC included a description of the Licensing Support System (LSS) under its discussion of "Measures for dealing with State-Local concerns." The LSS is intended to provide participants in the repository licensing proceeding early access to documents relevant to the licensing decision. To eliminate any confusion regarding NRC's responsibilities for the LSS, the above sentence in the Proposed Decision Review will be eliminated and the following description will be inserted in its place: "DOE is responsible for the design, development, procurement and te~ting of the LSS. LSS design and development must be consistent with objectives and requirements of the Commission's LSS rulemaking and must be carried out in consultation with the LSS Administrator and with the advice of the Licensing Support System Advisory Review Panel. NRC (LSS Administrator] is responsible for the management and operation of the LSS after completion of the DOE design and development process." Issue No. 7: Suggestion for Reducing Licensing Uncertainties Related to Spent Fuel Transshipments Comment Commonwealth Edison commented that in order to enhance the viability of the option of transferring spent fuel from retired reactors to others under active management, the NRC should reduce, to the maximum extent possible, licensing uncertainties related to such fuel transfers. The commenter also stated that by predetermining that spent fuel pool densification and alternative site spent fuel storage methods do not raise any significant hazards considerations, the NRC's final decision would be strengthened. NRG Response The Commission evaluates applications for modification of spent fuel storage at licensee's facilities or for transshipment from one site to another on an individual basis. Such a case consideration of the merits of each application ensures that all significant safety issues are addressed in a thorough manner and provides a conservative approach for arriving at a decision on the merits of the license application. Issue No. 8: Appropriate Use of Nuclear Waste Fund Monies Comment Commonwealth Edison Company (CECo) refers to the NRC's statement that DOE could accept responsibility for management of spenf"fuel until a repository is available in the event that a licensee becomes insolvent prior to the time a geologic repository is ready to accept spent fuel. Funds from either the Nuclear Waste Fund (NWF) or from the utility itself could be used (54 FR 39767, at 39786 and 39790). CECo comments that the use of the NWF monies for this purpose would involve the solvent utilities funding the storage of spent fuel generated by the bankrupt licensees. CECo believes that it is not clear 51-SC-40 whether the Nuclear Waste Policy Act would allow NWF monies to be used for this purpose and suggests that NRC should seek and analyze comments on this issue. Until further evaluation and analysis has taken place, CECo believes NRC should delete this as a basis for confidence. NRG Response The Commission believes that there are two related issues presented in the above comment. The first is whether DOE can accept responsibility for spent fuel if a utility is insolvent or otherwise no longer capable of managing it. A second related issue is, given DOE's acceptance of responsibility for the spent fuel, where would DOE obtain the funds needed to pay the costs of this* responsibility? The NRC continues to believe that DOE would accept responsibility for spent fuel management in the event that a licensee is unable to exercise its own responsibility. Further, the NRC believes that DOE would have sufficient resources to carry out any safety-related measures. As indicated in the discussion under Issue 21, because DOE is not precluded from accepting responsibility for the waste in those situations, default is an issue of equity rather than public health and safety. As such, the Commission does not believe that a licensee's potential default has a direct bearing on the Commission's Waste Confidence Decision. Nevertheless, because the source of funds, but not DOE's ultimate responsibility is ambiguous, the NRC has decided to change the references that CECo cites with the bracketed words to be deleted in the Final Waste. Confidence Decision Review: If for any reason not now foreseen. this spent fuel can no longer be managed by the owners of these reactors, and DOE must assume responsibility for its management earlier than currently planned, this quantity of spent fuel is well within the capability of DOE to manage onsite or offsite with available technology [financed by the utility either directly or through the Nuclear Waste Fund]. (p.39786, col.1) Even if a licensed utility were to become insolvent, and responsibility for spent fuel management were transferred to DOE earlier than is currently planned, the Commission has no reason to believe that DOE would [have insufficient Nuclear Waste Fund resources or otherwise] be unable to carry out any safety-related measures NRC considers necessary. (p.39390, col.1] Issue No. 9: Costs Incurred Due to Delayed Acceptance of Spent Fuel at Repository Comment PART 51 STATEMENTS OF CONSIDERATION Commonwealth Edison Company (CECo) observed that additional costs will be incurred by licensees as a result of delayed acceptance of spent fuel at the repository. CECo believes that consideration should be given as to whether these costs will be covered by the Nuclear Waste Fund or whether the costs will be incurred directly by the licensee. NRG Response The Commission believes that this is a matter which will have to be resolved in another forum in the context of the contracts between DOE and the utilities/ owners of spent fuel. The individual contracts currently specify the dates by which DOE has agreed to accept responsibility for the disposal of spent fuel. If DOE must delay its acceptance of spent fuel, the responsibility for the financial consequences of that default would have to be determined at that time by reference to and interpretation of the pertinent contracts. The ultimate answer to this question will not affect the findings of the Waste Confidence Decision. Issue No. 10: Clarification of Discussion of Period of Safe Spent Fuel Storage at Dresden 1 Comment Commonwealth Edison Company (CECo) comments that the discussion in the Proposed Decision Review of the possible extended storage of spent fuel from Dresden 1 is not clear and should be clarified. On the basis of assumptions discussed in the Proposed Decision Review, CECo concludes that three different dates could be derived to indicate the maximum time for onsite spent fuel storage. For Dresden 1, which was licensed to operate in 1959 and permanently shut down in 1978, 30 years after shutdown would yield a maximum date of 2008; 30 years after a full 40-year license term yields a maximum date of 2029; and 30 years after a full 40-year license term plus a 30-year extension of the operating license would yield a date of 2059. NRG Response The NRC believes that CECo has misinterpreted the discussion pertaining to the maximum term of onsite spent fuel storage in the Waste Confidence Decision and the bases and assumptions underlying that discussion as they pertain to the specific circumstances of Dresden 1. The generic discussion of the derivation of the maximum safe storage term for the purposes of the Waste Confidence Decision is contained in pp.39785-90 and pp.39783-96. The Commission concluded on a generic basis that "spent fuel generated in any reactor can be stored safely and without significant environmental impacts in reactor facility storage pools or independent spent fuel storage installations located at-reactor or from-reactor sites for at least 30 years beyond the licensed life for operation (which may include the term of a revised license) of that reactor at its spent fuel storage basin or at either onsite or offsite independent spent fuel storage installations" (proposed 10 CFR 51.23(a) at p. 39968 (Finding 4) (emphasis added)). The discussion and findings were based on technical and institutional considerations that, for the sake of completeness, considered situations like those at Dresden 1 that differ from those with most reactors that are expected to operate to full term plus a possible extended license term. For Dresden 1, based on proposed § 51.23(a), the applicable storage period would be 30 years beyond the licensed life of operation, or until 2029. 2.3 The Commission's Third Finding The Commission finds reasonable assurance that high-level radioactive waste and spent fuel will be managed in a safe manner until sufficient repository capacity is available to assure the safe disposal of all high-level waste and spent fuel. Issue No. 11: Resolution of Contractual Conflicts Between DOE and Licensees Comment Commonwealth Edison Company (CECo) comments that the NRC has unnecessarily interjected itself into issues involved in the contracts between the DOE and licensees by NRC's statement that it would have more confidence if the DOE and licensees could resolve any uncertainties by reaching an early and amicable resolution as to how and when the DOE will accept responsibility for spent fuel. CECo believes that the implication in this statement is that licensees should amend their contracts with DOE to allow DOE additional time to perform under the contracts or that licensees should refrain taking action against DOE if it defaults under the contracts. CECo notes that NRC has stated that its confidence in safe storage is unaffected by potential contractual disputes between DOE and the spent fuel owners (54 FR 39792), therefore CECo believes that it would be appropriate for NRC to strike the statement and express no opinion regarding possible future disputes between DOE and licensees. NRG Response The Commission did not intend the implication that CECo perceives regarding any particular preferred outcome or suggested resolution of 51-SC-41 future potential contract disputes between DOE and contract holders. The Commission has stated that its confidence in safe storage is unaffected by any potential contractual dispute between DOE and spent fuel generators and owners as to responsibility for spent fuel storage. The Commission's further statement that it would be helpful if any future potential contract disputes could be resolved amicably merely expressed a concern that the waste management system operates smoothly and efficiently. The statement did not imply any additional impact on or repercussion from the Waste Confidence Decision upon the resolution of future potential contract disputes between DOE and contract holders. The Commission believes that it has made its position clear that its confidence is not diminished by any potential contractual disputes between DOE and spent fuel owners. However, in order to avoid any further misunderstanding in this regard, the Commission has decided to delete the following statements in its Proposed Waste Confidence Decision Review from its Final Waste Confidence Decision Review: To resolve any continuing uncertainties, however, it would be helpful if DOE and utilities and other spent fuel generators and owners could reach an early and amicable resolution to the question of how and when DOE will accept responsibility for spent fuel. This would facilitate cooperative action to provide for a smoothly operating system for the ultimate disposition of spent fuel. (54 FR 39792) and IfDOE and the utilities can amicably resolve their respective responsibilities for spent fuel storage in the interest of efficient and effective administration of the overall waste management system, including the Nuclear Waste Fund, NRG would gain added confidence in the institutional arrangements for spent fuel management. (54 FR 39797) Issue No. 12: NRG Responsibility to Identify Need for Utilities to Provide Interim Storage and to Notify Congress of This Requirement Comment Malachy Murphy (State of Nevada) comments that, in light of DOE's Reassessment Report to Congress, the NRC should explicitly state that utilities will need to have interim spent fuel storage available well into the next century. The commenter also states that NRC should explicity request that Congress take note of this requirement. The commenter believes that such action would be in keeping with NRC's responsibilities to the public and to nuclear utilities. NRG Response PART 51 STATEMENTS OF CONSIDERATION The standard contracts between DOE and generators of spent nuclear fuel or persons holding title to spent fuel currently provide.that in return for payment to the Nuclear Waste Fund, DOE will dispose of high-level waste and spent fuel beginning no later than January 31, 1998. The Commission believes it would be inappropriate for NRC to take any position on the need for generators and those holding title to such material to provide interim storage for it beyond 1998. This is a matter that will have to be resolved between the parties to the standard contracts. NRC, in its original Waste Confidence Decision and in the Proposed Waste Confidence Decision Review, addressed the issue of storage of spent fuel until a repository becomes available and has expressed its confidence that spent fuel will be safely managed until a repository is available. Furthermore, in its original Waste Confidence Proceeding, NRC amended its reactor licensing rule, 10 CFR part 50 to require each licensed reactor opera tor to submit, no later than five years before expiration of the operating license, plans for managing spent fuel at the reactor site until the spent fuel is transferred to DOE for disposal. In the Nuclear Waste Policy Act [NWPA), Congress placed primary responsibility for interim storage of spent fuel on the nuclear utilities until disposal becomes available. Section 132 of the NWPA requires that DOE, NRC, and other authorized Federal officials take such actions as they believe are necessary to encourage and expedite the effective use of available storage, and necessary additional storage, at the site of each civilian nuclear power reactor. Sections 218(a) and 133 of the NWPA also provide that NRC by rule establish procedures for the licensing of any technology approved by NRC for use at the site of any civilian nuclear power reactor. NRC may by rule approve one or more dry spent fuel storage technologies for use at the sites of civilian power reactors without, to the maximum extent practicable, the need for additional site-specific approvals. Congress is eminently aware of the likely need for at-reactor storage of spent fuel and has taken legislative action with resped to this matter. Therefore, the NRC believes it is not necessary to inform Congress of this need. However, the NRC will continue to exercise its responsibility to assure that spent fuel is managed safely until a repository is available and will notify Congress of any actions it believes are necessary to provide this assurance. 2.4 The Commission's Fourth Finding The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30.years beyond the licensed life for operation (which may include the term of a revised or renewed license) of that reactor at its spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations. Issue No. 13: Consideration of the Cumulative Impacts on Waste Management in the NRC's NEPA Documentation Comment DOE commented that the cumulative impacts on waste management of potential reactor operating license extensions should be considered in the NRC's National Environmental Policy Act (NEPA) documentation for license renewals. NRG Response DOE has observed that renewal of operating licenses would increase the total amount of spent fuel requiring disposal or interim storage which would be taken into account in DOE program planning and should be considered in NRC's NEPA documentation for license renewals. This is generally consistent with the discussion in the Commission's proposed decision, especially 54 FR 39795 [third column). The greater amount of spent fuel which must be stored as a result of license renewal does not affect the Commission's overall finding of no significant environmental impacts. Issue No. 14: Need for NRG to Facilitate ISFSI License Extensions to Reflect the Commission's Revised Fourth Finding Comment The Virginia Electric & Power Company [VEPCo) states that the current license on the Independent Spent Fuel Storage Installation [ISFSI) for its Surry nuclear power plant expires on July 31, 2006. VEPCo states that the NRC should initiate actions to facilitate ISFSI license extensions to reflect the proposed revised Fourth Finding that spent fuel generated in any reactor can be safely stored for at least 30 years beyond the licensed life for operation of that reactor either onsite or offsite. NRG Response The Commission's Waste Confidence finding on the duration of safe storage of spent fuel is generic in nature. specific licensing procedures remain effective. Pursuant to § 72.42, an ISFSI license is issued for a period of 20 years but may be renewed upon application by the licensee. Part 72 in no way precludes licensees from requesting 51-SC-42 additional extensions of license terms for ISFSls. The licensee thus has the option of requesting an ISFSI license renewal to coincide with whatever operating term and post-operation spent fuel storage period is in effect for a particular reactor. For example, a single renewal could extend the Surry ISFSI license expiration date to the year 2026. The NRC does not believe that further revisions to § 72.42 to facilitate these license extensions are warranted at this time. Issue No. 15: Insufficient Assurance on

Duration of Safe Storage and Risk of Fire at a Spent Fuel Pool Comment Public Citizen stated that there is not adequate assurance that spent fuel will be stored safely at reactor sites for up to 30 years beyond the expiration of reactor operating licenses.

This is even more the case if license extensions of up to 30 years are included. Public Citizen further stated that "the (Waste Confidence) policy statement fails to recognize that spent fuel buildup at reactor sites poses a growing safety hazard. The pools are not well protected from the environment (in many cases they are outside the reactor's containment structure) and have leaked in the past. For example, in December 1986 at the Hatch nuclear power plant in Baxley, Georgia, 141,000 gallons of radioactive water leaked out of the plant's fuel pool. More than 80,000 gallons of the water drained into a swamp and from there into the Altamaha River near the plant." Public Citizen added that "More recently, on August 16, 1988, a seal on a fuel pool pump failed at the Turkey Point nuclear plant near Miami, FL, causing some 3,000 gallons of radioactive water to leak into a nearby storm sewer. The shoes and clothing of approximately 15 workers were contaminated." Public Citizen also stated that the danger posed by an accident in which enough pool water escaped to uncover the irradiated fuel assemblies would be greater than the operational incidents described above. According to the commenter, if a leak or pump failure caused the water level in a spent fuel pool to drop to a level which exposed the fuel assemblies, the _remaining water might be insufficient to provide adequate cooling. The pool water could then heat to the boiling point, producing steam and causing more water to boil away. The danger then is that heat could continue to build up even further until the cladding which encloses the irradiated fuel pellets catches fire. The commenter continued saying that the PART 51 STATEMENTS OF CONSIDERATION NRC itself, in the time since the original Waste Confidence Decision, has studied the issue of storage in reracked spent fuel pools and concluded in a 1987 report that the consequence of such a cladding fire could be a "significant" radiation release. The NRC report found: (1) the natural air flow permitted by high-density storage racks is so restricted that potential for sustaining cladding fire exists; and (2) with high-density racks providing "severely restricted air flow" the oxidation (burning) would be "very vigorous" and "failure of both the fuel rods and the fuel rod racks is expected." Public Citizen states that nowhere in the Proposed Waste Confidence Decision Review does the NRC take into account the findings of this report, which should have been included. NRG Response The Commission has addressed the safety of extended post-operational spent fuel storage at considerable length in the discussion of its proposed revised Fourth Finding. Operational occurrences cited in Public Citizen's comment have been addressed by the NRC staff at the plants listed. The NRC has taken inspection and enforcement actions to reduce the potential for such operational occurrences in the future. We would like to note, however, that the event at the Hatch plant occurred in a transfer canal between spent fuel pools during an operation that would not normally be performed following expiration of a reactor operating license. In the case of the event at Turkey Point, the water that flowed outside the building went back into the intake of the plant cooling canal. The canal is a large, closed loop onsite flow path. There was no radiation release offsite, and the safety significance of the event appears to have been very low. Regarding the risk of fire at spent fuel pools, the NRC staff has spent several years studying in detail catastrophic loss of reactor spent fuel pool water possibly resulting in a fuel fire in a dry pool. The 1987 report, "Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue 82" (NUREG/CR-4982), referred to in Public Citizen's comment represents an early part of the NRC's study. Its findings were based on generic data on seismic hazards and response of spent fuel pools, which resulted in calculated risk numbers with wide ranges of uncertainty. (See p. xiii.) Subsequent study of the consequences and risks due to a loss of coolant water from spent fuel pools was conducted by the NRC, and the results were published in NUREG/CR-5176, "Seismic Failure and Cask Drop Analysis of the Spent Fuel Pools at Two Representative Nuclear Power Plants," January 1989, and NUREG-1353, "Regulatory Analysis for the Resolution of Generic Issue 82, ;;,Beyond Design Basis Accidents in Spent Fuel Pools'," April 1989. These reports were cited in the Commission's Proposed Waste Confidence Decision Review (54 FR 39767-39797, at p.39795, September 28, 1989). Also issued in 1989, as part of the NRC staffs study, was "Value/Impact Analyses of Accident Preventive and Mitigative Options for Spent Fuel Pools" (NUREG/CR-5281). The analyses reported in these studies indicate that the dominant accident sequence which contributes to risk in a spent fuel pool is gross structural failure of the pool due to seismic events. Risks due to other accident scenarios (such as pneumatic seal failures, inadvertent drainage, loss of cooling or make-up water, and structural failures due to missiles, aircraft crashes and heavy load drops) are at least an order of magnitude smaller. For this study, older nuclear power plants were selected, since the older plants are more vulnerable to seismic-induced failures. It should be noted that for a zircaloy cladding fire in a spent fuel storage pool, an earthquake or other event causing a major loss of cooling water would have to occur within two years after operation of a PWR or six months after operation of a BWR. (See NUREG-1353, p. 4-11.) Thus, during the decades of post-operational storage, even a major loss of cooling water would not be sufficient to cause a cladding fire. During the time the pool would be most vulnerable to a fire, the most-recently discharged fuel assemblies would have to be adjacent to other recently discharged assemblies for a fire to propagate to the older fuel. Considering that a third of the reactor core is typically unloaded as spent fuel each year, the probability of a fire involving even the equivalent of a reactor core--a small portion of a pool's capacity--is quite remote. It should also be noted that even if the timing of a spent fuel pool failure were conducive to fire, a fire could occur only with a relatively sudden and substantial loss of coolant--a loss great enough to uncover all or most of the fuel, damaging enough to admit enough air from outside the pool to keep a large fire going, and sudden enough to deny the operators time to restore the pool to a safe condition. Such a severe loss of cooling water is likely to result only from an earthquake well beyond the conservatively estimated earthquake for which reactors are designed. Earthquakes of that magnitude are extremely rare. 51-SC-43 The plant-specific studies following the 1987 generic study found that, because of the large safety margins inherent in the design and construction of their spent fuel pools, even the more vulnerable older reactors could safely withstand earthquakes several times more severe than their design basis earthquake. Factoring in the annual probability of such beyond-design-basis earthquakes, the plant-specific and generic followup studies calculated that the average annual probability of a major spent fuel pool failure at tin opera ting reactor was ten to thirty times lower than the average probabilities in the 1987 study. (See NUREG/CR-5176, p. xiii, and NUREG-1353, pp. ES-2-3.) For either BWR or PWR designs, this probability was calculated at two chances in a million per year of reactor operation. (See NUREG-1353, pp. ES-3-4.) After evaluating several regulatory options for reducing the risk of spent fuel pool fires, the NRC regulatory analysis concluded that "[t]he risk[s] due to beyond design basis accidents in spent fuel pools, while not negligible, are sufficiently low that the added costs involved with further risk reductions are not warranted." (See NUREG-1353, pp. ES-6-8.) Issue No. 16: Need for NRG Requirement for Dry Cask Storage Instead of Storage in Spent Fuel Pools Comment Public Citizen states that the use of dry cask storage for spent fuel would help address some of the concerns described above, but that NRC has no plans to require dry cask storage instead of storage in spent fuel pools. The commenter notes that NRC has explicitly stated in its Proposed Decision Review that storage in a reactor's "spent fuel storage basin" is considered safe, and (the commenter) apparently disagrees with this conclusion. NRG Response The record of operational experience with reactor spent fuel storage pools, as discussed in the Commission's Proposed Decision Review and in response to the preceding comments, strongly supports the conclusion that reactor spent fuel pool storage, which has continued for decades, is safe. Accordingly, the NRC has reached the conclusion that past experience and available information amply support the safety of spent fuel storage, both in pools and dry storage casks, for at least 30 years past the expiration of reactor operating licenses (including the term of a revised license). PART 51 STATEMENTS OF CONSIDERATION Issue No. 17: Suggestion to Revise Proposed Fourth Finding to Reflect Reasonable Assurance That Spent Fuel Can Be Safely Stored in Dry Casks at Reactor Sites for Up to One Hundred Years Comment Malachy Murphy [State of Nevada] commented that NRC's Proposed Revised Fourth Finding did not go far enough with respect to the duration of safe storage in dry storage casks. The commenter suggested that both the proposed finding and the Proposed Amendment to 10 CFR 51.23 be amended to reflect reasonable assurance that spent fuel can be stored safely and without significant environmental risk in dry casks at reactor sites for up to one hundred (100) years. NRG Response The Commission does not dispute a conclusion that dry spent fuel storage is safe and environmentally acceptable for a period of 100 years. Evidence supports safe storage for this period. A European study published in 1988 states, "In conclusion, present-day technology allows wet or dry storage over very long periods, and up to 100 years without undue danger to workers and population." [See Fette!, W., Kaspar, G., and Gunther, H., "Long-Term Storage of Spent Fuel from Light-Water Reactors" [EUR 11866 EN], Executive Summary, p.v, 1988.J Although spent fuel can probably be safely stored without significant environmental impact for longer periods, the Commission does not find it necessary to make a specific conclusion regarding dry cask storage in this proceeding, as suggested by the commenter, in part because the Commission's Proposed Fourth Finding states that the period of safe storage is "at least" 30 years after expiration of a reactor's operating license. The Commission supports timely disposal of spent fuel and high-level waste in a geologic repository, and by this Decision does not intend to support storage of spent fuel for an indefinitely long period. Issue No. 18: Maintenance of Institutional Controls for One Hundred Years Comment Marvin Lewis commented that the Commission's Proposed Revised Decision and Amendment to 10 CFR part 51 both require that at-reactor storage be available and safe for at least 100 years, which is an excessive amount of time to depend on institutional memory. The commenter states that to look into the future and have confidence that our institutions will survive in a form which will provide that safe onsite storage is available for at least 100 years into the future lacks any merit. The commenter asked that the Commission arrive at the opposi.l:e conclusion, namely that "Due to the Department of Energy's lack of quality control of data and analysis, inability to qualify acceptable sites, accusation against subcontractors when data contradicts DOE's preconceived assumptions, and general adherence to the political solution instead of scientific veracity, the NRC cannot find that temporary storage at reactors will ensure that geological storage for spent fuel will be available and safe when needed." NRG Response The Commission believes there is an adequate basis from the record of Federal regulations, historical experience and current practice to support the Commission's finding regarding institutional controls over spent fuel storage activities. The Environmental Protection Agency's standards for high-level waste disposal provide that "active institutional controls over disposal sites should be maintained for as long a period of time as is practicable after disposal; however, performance assessments that assess isolation of the wastes from the accessible environment shall not consider any contributions from active institutional controls for more than 100 years after disposal" (40 CFR 191.14[a)). The finding that repository licensing performance assessments can take credit for active institutional controls for 100 years is not one of the issues involved in the judicial action which vacated the EPA standard, and it is not expected that this section will be disturbed when the standard is reissued. It should also be noted that this language does not suggest that active institutional controls are unlikely for a period greater than 100 years. In the summary of the Final Rule (50 FR 38066; September 19, 1985), EPA noted that many commenters on the Proposed Rule felt that "a few hundred years" which was the proposed period for reliance on active institutional controls was too long. EPA agreed to limit the period to 100 years, noting that "this was the time period [EPA] considered in criteria for radioactive waste disposal that were proposed for public comment in 1978 (43 FR 53262), a period that was generally supported by the commenters on that proposal" (50 FR 38066, at p. 38080). NRC would add that there are abundant examples of institutions in human society which have maintained a continuity in institutional controls far 51-SC-44 exceeding 100 years. The government of the United States, which is relatively young, is over 200 years old. The governments of some European countries have been in existence for time periods between 700 to 1000 years. While invading armies and civil wars have been disruptive, archival information of interest to the safety of the population can be expected to be preserved. In the United States today, real estate contracts are commonly executed to cover a period of 100 years, or a significant fraction thereof. One hundred-year land-lease agreements are common. Major civil construction projects such as harbors, bridges, flood control systems, and dams are often planned and executed--and investments made in them--with the view of recovering the benefits over a period of 100 years or more. 2.5 The Commission's Fifth Finding The Commission finds reasonable assurance that safe independent onsite or offsite spent fuel storage will be made available if such storage capacity is needed. Issue No. 19: Impact of Extension of Time for Repository Availability on the Increased Generation of Low-Level Radioactive Waste Comment Commonwealth Edison (CECo) commented that the Proposed Waste Confidence Review does not address low-level waste concerns resulting from delayed acceptance of spent fuel by the repository under DOE's extended schedule for repository availability. CECo commented that if they store spent fuel in pools and implement rod consolidation to conserve space during the extension, additional low-level waste may be generated. CECo believes that NRC should determine if this additional low-level waste should go to a Federal Repository or to a sited compact for disposal. NRG Response The disposition of high-level and level radioactive wastes has already been determined by Congress in the Nuclear Waste Policy Act of 1982 (NWPA) and in the Low-Level Radioactive Waste Policy Act (LLWPAJ. Congressional designation of the method of disposal of each type of waste was not dependent on the DOE's schedule for development of the repository; rather, Congress designated the method of disposal according to characteristics of the waste which are associated with its hazard (i.e., radioactive source strength, radioactive species of the emanating radiation, and half-life]. It is not within the NRC's regulatory PART 51 STATEMENTS OF CONSIDERATION jurisdiction to change the directives provided by Congress in the NWPA and the LLWPA. 3.0 Consideration of Other Events Relevant to the Commission's Decision Issue No. 20: Petition by the State of Vermont to Intervene in the Consideration of the Extension of the Operating License for Vermont Yankee In the Commission's Proposed Waste Confidence Decision Review, it was stated that the basis for the 2007-2009 timeframe in the Court remand leading to the Waste Confidence Proceeding had changed since the original Decision. This discussion was based on the fact that it appeared likely that these dates no longer represented the expected expiration dates for the operating licenses of the Vermont Yankee and Prairie Island nuclear plants. The NRC staff has been granting extensions of the dates of expiration of nuclear plant operating licenses to reflect a 40-year period from the date of issuance of the operating license rather than from the date of the construction permit. The dates of expiration of the Prairie Island Units 1 and 2 had already been extended from the year 2006 to the years 2013 and 2014. The NRC staff anticipated that on the basis of the date of issuance of its operating license, Vermont Yankee would be eligible for an extension of its opera ting license to March 2012. In the time since the drafting of the Proposed Decision Review, several pertinent events have occurred. NRC published a notice of consideration of amendment to the Vermont Yankee Operating License, a proposed "no significant hazards" consideration determination, and opportunity for a hearing (54 FR 31120; July 26, 1989). On August 22, 1989, the State of Vermont filed a petition for leave to intervene. On October 30, 1989, Vermont filed a supplement to its petition to intervene proposing nine contentions for litigation on Vermont Yankee Nuclear Power Corporation's application to extend its operating license. On November 15, 1989, the NRC's Atomic Safety and Licensing Board (ASLB) heard oral argument by counsel for the licensee, the NRC staff, and the State of Vermont concerning the State's petition for leave to intervene and supplemental petition for leave to intervene. The ASLB granted the State of Vermont's petition for leave to intervene, admitted one contention (which did not concern waste disposal) as an issue in controversy for litigation, and granted the request for hearing. The ASLB's ruling was issued in a Prehearing Conference Memorandum and Order dated January 26, 1990 (Docket Nci.50-271-0LA-4). It is now apparent that the extension of Vermont Yankee's operating license expiration date will be dependent on the outcome of this contested hearing. There is the possibility tha't a shorter extension or that no extension will be granted. In view of the uncertain outcome, the Commission will delete all discussion of a possible revised date for the Vermont Yankee operating license expiration and the revised date for expiration of the Prairie Island operating license. This deletion, however, does not affect the Commission's Proposed Revised Second Finding in its Waste Confidence Decision Review. Assuming that no extension or a lesser extension is granted and Vermont Yankee's operating license expires in 2007, the basis for the Commission's finding that a repository will be available within the first quarter of the twenty-first century and that sufficient repository capacity will be available within 30 years beyond the licensed life for operation of any reactor, would be unaffected. Issue No. 21: PotentialNeed for Additional Financial Security for the Nuclear Waste Fund The NRC staff has been informed by DOE's Office of Civilian Radioactive Waste Management that a pending final report from DOE's Inspector General has indicated a potential problem for certain nuclear utility licensees to pay the one-time fee into the Nuclear Waste Fund (NWF) for spent fuel generated prior to April 1983. This issue arises because several utilities elected to defer payment into the fund and, instead, themselves hold the money that was collected from ratepayers for the time fee. DOE's Inspector General believes that some of those utilities may not be able to make their payments when due. The NRC staff met with DOE's Office of Civilian Radioactive Waste Management (OCRWM) on December 13, 1989 to discuss this issue and determine the potential impact on both NRC's Decommissioning Rulemaking and on the Waste Confidence Decision, and, more generally, on protection of public health and safety. In addition, NRC discussed at that meeting and in follow-up telephone conversations potential actions that DOE might take. These actions could include modifying DOE's spent fuel contracts with electric utilities, seeking legislative amendments, and working with the National Association of Regulatory Utility Commissioners to increase assurance of one-time contributions into the NWF. 51-SC-45 The NRC understands from OCRWM staff that, if a nuclear utility licensee were to default on its one-time contribution to the NWF, DOE is not precluded from accepting for disposal all spent fuel from that utility. Thus, the NRC does not view this issue as affecting its confidence that the spent fuel will be disposed of. Rather, the issue is one of equity--that is, will a utility and its customers and investors or U.S. taxpayers and/ or other utilities ultimately pay for disposal of spent fuel genera led prior to April 1983. Background In November 1976, the Natural Resources Defense Council (NRDC) petitioned NRC for a rulemaking to determine whether radioactive wastes generated in nuclear power reactors can be subsequently disposed of without undue risk to the public health and safety. The NRDC also requested that NRC not grant pending or future requests for operating licenses until the petitioned finding of safety was made. On June 27, 1977, NRC denied the NRDC petition. The Commission said that in issuing operating licenses, NRC must have assurance that wastes can be safely handled and stored as they are generated. It also said that it is not necessary for permanent disposal to be available if NRC could be confident that permanent disposal could be accomplished when necessary. NRC added that Congress was aware of the relationship between nuclear reactor operations and the radioactive waste disposal problem, and that NRC would not refrain from issuing reactor operating licenses until the disposal problem was resolved. The Commission also stated that it " ... would not continue to license reactors if it did not have reasonable confidence that the wastes can and will in due course be disposed of safely." Also in November 1976, two utility companies requested amendments to their opera ting licenses to permit expansion in the capacity of their spent nuclear fuel storage pools: Vermont Yankee Nuclear Power Corporation for the Vermont Yankee plant; and Northern States Power Company for its Prairie Island facility. In both cases, the utilities planned to increase storage capacity through closer spacing of spent fuel assemblies in existing spent fuel pools. The New England Coalition on Nuclear Power and the Minnesota Pollution Control Agency intervened. The NRC staff evaluated the requests and found that the modifications would not endanger public health and safety. The staff did not consider any potential PART 51 STATEMENTS OF CONSIDERATION environmental effects of storage of spent fuel at the reactors beyond the dates of expiration of their operating licenses. NRC's Atomic Safety and Licensing Board Panel (ASLBP) adopted the staffs safety and environmental findings and approved the license amendments for the two plants. It too did not consider the effects of at-reactor storage beyond the expiration of the facility operating license. The Board's decision was appealed to the Atomic Safety and Licensing Appeal Board (ASLAB). The ASLAB affirmed the Licensing Board's decision, citing the Commission's " ... reasonable confidence that wastes can and will in due course be disposed of safely .... " in the Commission's denial of the NRDC petition. The decision of the ASLAB was appealed to the U.S. Circuit Court of Appeals. On May 23, 1979 the Court declined to stay or vacate the license amendments, but remanded to NRC the question of " ... whether there is reasonable assurance that an offsite storage solution will be available by the years 2007-2009, the expiration of the plants' operating licenses, and if not, whether there is reasonable assurance that the fuel can be safely stored at the reactor sites beyond those dates." In its decision to remand to NRC, for consideration in either a generic rulemaking or an adjudicatory proceeding, the Court observed that the issues of storage and disposal of nuclear waste were being considered by the Commission in an ongoing generic proceeding known as the "S-3 Proceeding" on the environmental impacts of uranium fuel cycle activities to support the operation of a light water reactor, and that it was appropriate to remand in light of a pending decision on that proceeding and analysis. On October 18, 1979, NRC announced that it was initiating a rulemaking proceeding in response to the Appeals Court remand and as a continuation of the NRDC proceeding. Specifically, the purpose of the proceeding was for the Commission " ... to reassess its degree of confidence that radioactive wastes produced by nuclear facilities will be safely disposed of, to determine when any such disposal will be available, and whether such wastes can be safely stored until they are disposed of." The Commission recognized that the scope of this proceeding would be broader than the Court's instruction, which required the Commission to address only storage-related questions. The Commission believed, however, that the primary public concern was the safety of waste disposal rather than the availability of an off-site solution to the storage problem. The Commission also committed itself to reassess its basis for confidence that methods of safe permanent disposal for high-level waste would be available when needed. Thus, the Commission chose as a matter of policy not to confine itself exclusively to the narrower issues in the court remand. In the Notice of Proposed Rulemaking, the Commission also stated that if the proceeding led to a finding that safe site storage or disposal would be available before expiration of facility operating licenses, NRC would promulgate a rule providing that the impact of onsite storage of spent fuel after expiration of facility operating licenses need not be considered in individual licensing proceedings. The Waste Confidence Decision was issued on August 31, 1984 (49 FR 34658). In the Decision, the Commission made five findings. It found reasonable assurance that: (1) Safe disposal of high-level radioactive waste and spent fuel in a mined geologic repository is technically feasible. (2) One or more mined geologic repositories for commercial high-level radioactive waste and spent fuel will be available by the years 2007-2009, and sufficient repository capacity will be available within 30 years beyond expiration of any reactor operating license to dispose of existing commercial high-level radioactive waste and spent fuel originating in such reactor and generated up to that time. (3) High-level radioactive waste and spent fuel will be managed in a safe manner until sufficient repository capacity is available to assure the safe disposal of all high-level radioactive waste and spent fuel. (4) If necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the expiration of that reactor's operating license at that reactor's spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations. (5) Safe independent onsite or offsite spent fuel storage will be made available if such storage capacity is needed. On the day the Decision was issued, the Commission also promulgated two rulemaking amendments: (1) an amendment to 10 CFR part 50, which required that no later than five years before expiration of reactor operating licenses, the licensee must provide NRC with a written plan for management of spent fuel onsite, until title for the spent fuel is transferred to the DOE; and (2) an 51-SC-46 amendment to 10 CFR part 51 which provided that environmental consequences of spent fuel storage after expiration of facility licenses need not be addressed in connection with issuance of or amendment to a reactor operating license. In issuing the part 51 amendment, the Commission stated that although it had reasonable assurance that one or more repositories would be available by 2007-2009, it was possible that some spent fuel would have to be stored beyond those dates. The part 51 amendment was based on the Commission's finding in the Waste Confidence Proceeding that it had reasonable assurance that no significant environmental impacts will result from storage of spent fuel for at least 30 years beyond expiration of reactor operating licenses. Enactment of the NWPA contributed significantly to the basis for the Commission's 1984 Decision and companion rulemakings. The Act established a funding source and process with milestones and schedules for, among other things, the development of a monitored retrievable storage (MRS) facility and two repositories, one by early 1998 and a second, if authorized by Congress, at a later date, initially planned by DOE for 2006. For each repository, the Act required DOE to conduct in-situ investigations of three sites and recommend one from among them to the President and Congress for repository development. The NWPA also required DOE to recommend, from among alternative sites and designs, a site and design for an MRS for spent fuel and high-level waste management before disposal. The Commission's licensing and regulatory authority over both storage and disposal facilities was preserved by the Act. In the four years after enactment of the NWPA, DOE met a number of the Act's early program requirements, but also encountered significant difficulties. It published a final Mission Plan for the overall NWP A program, and followed with a Project Decision Schedule for DOE and other Federal agency actions. It promulgated, with Commission concurrence, a set of guidelines for repository siting and development. It published draft and final environmental assessments for nine candidate repository sites, and recommended three for characterization. It completed and submitted to Congress an environmental assessment, a program plan, and a

proposal with a site and design for an MRS. All these actions followed extensive interactions with interested Federal agencies, State, Indian tribal, and local governments, and other PART 51 STATEMENTS OF CONSIDERATION organizations.

In the course of these activities, however, DOE also slipped its schedule for operation of the first repository by five years, indefinitely postponed efforts toward a second repository, and had to halt further MRS siting and development activities pending Congressional authorization. In December, 1987, Congress enacted the Nuclear Waste Policy Amendments Act [NWPAA). The NWPAA redirected the high-level waste program by suspending site characterization activities for the first repository at sites other than the Yucca Mountain site, and by suspending all site-specific activities with respect to a second repository. The Amendments Act also authorized and set schedule and capacity limits on the MRS. The purpose of these limitations, according to sponsors of the legislation, was to assure that an MRS would not become a substitute for a geologic repository. Consistent with its commitment to revisit its Waste Confidence conclusions at least every five years, the Commission has undertaken the current review to assess the effect of these and other developments since 1984 on the basis for each of its five findings. The Commission issued its proposed Waste Confidence Decision Review and proposed revised findings for public comment on September 28, 1989. The comment period expired December 27, 1989. A total of eleven comments were received. In this document, the Commission supplements the basis for its earlier findings and the environmental analysis of the 1984 Decision. The Commission is amending fts second finding, concerning the timing of initial availability and sufficient capacity of a repository, and its fourth finding, concerning the duration of safe spent fuel storage. These revisions are based on the following considerations: (1) the five-year slippage, from 1998 to 2003, in the DOE schedule for repository availability prior to issuance of its November 1989 "Report to Congress on Reassessment of the Civilian Radioactive Waste Management Program" and its new target date of 2010 for repository availability announced in that report; (2) the additional slip of four and half years since the January 1987 Draft Mission Plan Amendment in the DOE schedule for the excavation of the exploratory shaft; (3) the need to continue accounting for the possibility that the Yucca Mountain site might be found unsuitable and that DOE would have to initiate efforts to identify and characterize another site for the first repository; (4) the statutory suspension of specific activities for the second repository; (5) DOE's estimate that site screening for a second repository should start about 25 years before the start of waste acceptance; and (6) increased confidence in the safety of extended spent fuel storage, either at the reactor or at independent spent fuel storage installations. The Commission is also issuing an amendment to 10 CFR 51.23(a] to conform with the revisions to Findings 2 and 4 elsewhere in this issue of the Federal Register. Organization and Table of Contents In conducting this review, the Commission has addressed, for each of its 1984 Findings, two categories of issues. The first category consists of the issues the Commission considered in making each Finding at the time of the initial Waste Confidence Decision. For these issues, the Commission is interested in whether its conclusions, or the Finding these conclusions support, should be changed to address new or foreseeable developments that have arisen since the first Waste Confidence Decision. The second category of issues consists of those the Commission believes should be added to the 1984 issues in light of subsequent developments. [To enable the reader to follow more easily, the lengthy discussions of Findings 1 and 2 have been organized to address each original and new issue under subheadings.] Table of Contents I. First Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 1. 1. Identification of acceptable sites 2. Development of effective waste packages (a) considerations in developing waste package [b] effect of reprocessing on waste form and waste package 3. Development of effective engineered barriers for isolating wastes from the biosphere (a] backfill materials (b] borehole and shaft sealants 51-SC-47 B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 1 1. Termination of Multiple Site Characterization

2. Relevance to NRC's "S-3 Table" proceeding

3. International developments in spent fuel disposal technology C. Conclusion on Finding 1 II. Second Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 2 1. Technical uncertainties (a] finding technically acceptable sites in a timely fashion [b] timely development of waste packages and engineered barriers 2. Institutional uncertainties (a] measures for dealing with State-local concerns (b] continuity of the management of the waste program (c) continued funding of the nuclear waste management program [d] DOE's schedule for repository development B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 2 1. Potential delay under the program of single site characterization

2. Potential limitations on timing of availability of disposal capacity (a] impact of possible limited disposal capacity at Yucca Mountain, indefinite suspension of second repository program [b] impact of uncertainty in spent fuel projections on need to consider second repository program 3. Impact of slippages in DOE program on availability of a repository when needed for health and safety reasons 4. Effect of NRG emphasis an completeness and quality C. Conclusion on Finding 2 Ill. Third Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 3: Licensee compliance with NRC regulations and license conditions; Safe management of spent fuel past expiration of operating licenses; Availability of DOE interim storage PART 51 STATEMENTS OF CONSIDERATION B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 3: Responsibility for spent fuel storage beyond 1998; Delay in second repository; Potential for license renewals IV. Fourth Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 4: Long-term integrity of spent fuel under water pool storage conditions; Structure and component safety for extended facility operation for storage; Safety of dry storage of spent fuel; Potential risks of accidents and acts of sabotage of spent fuel storage facilities B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 4: Radiological and non-radiological consequences of extended spent fuel storage; Potential delay in first repository, license renewals, delay in second repository; Environmental assessment and finding of no significant impact of at-reactor storage beyond 30 years after reactor's licensed life for operation V. Fifth Commission Finding A. Issues Considered in Commission's 1984 Decision on Finding 5: Adequacy of NWPA for determining responsibility for timely spent fuel storage; Spent fuel discharge projections; Industry commitment to implement away-from-reactor storage B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 5: Responsibility for spent fuel storage beyond 1998; Advances in technology for dry storage; Benefits of monitored retrievable storage facility under NWPAA; License renewals; Options for offsite storage under NWPAA Reaffirmed Finding 1: The Commission finds reasonable assurance that safe disposal of high-level radioactive waste and spent fuel in a mined geologic repository is technically feasible.

I.A. Issues Considered in Commission's 1984 Decision on Finding 1 J.A.1. *The identification of acceptable sites Under the Nuclear Waste Policy Act of 1982 (NWPA), the Department of Energy (DOE) had responsibility for identifying candidate sites for a geologic repository and for repository development. The first requirement leading to recommendation of candidate sites was formal notification of States with one or more potentially acceptable sites for a repository within 90 days of enactment of the NWPA. In February 1983, the DOE identified nine potentially acceptable sites for.the first repository. Four of the sites were in bedded-salt formations, three were in salt domes, one in volcanic luff, and one in basalt. The NWPA required that each site nomination be accompanied by an environmental assessment (EA). In December 1984, DOE published Draft EAs (DEAs) for each of the nine sites identified as potentially acceptable and proposed the following sites for nomination: the reference repository location at Hanford, WA; Yucca Mountain, NV; Deaf Smith County, TX; Davis Canyon, UT; and Richton Dome, MS. In May 1986, DOE released Final EAs (FEAs) for the five sites nominated. At that time, DOE recommended that the Yucca Mountain, Hanford, and Deaf Smith County sites undergo site characterization. The President approved the recommendation. The NRC staff provided extensive comments on both the DEAs and the FEAs. NRC concerns on the FEAs related primarily to DOE's failure to recognize uncertainty inherent in the existing limited data bases for the recommended sites, and the tendency of DOE to present overly favorable or optimistic conclusions. The primary intent of the comments was to assist DOE in preparing high-quality Site Characterization Plans (SCPs) for each site, as required under the NWPA, before excavation of exploratory shafts. NRC concerns can only be addressed adequately through the site characterization process, because one of the purposes of this process is to develop the data to evaluate the significance of concerns relative to site suitability. NRC did not identify any fundamental technical flaw or disqualifying factor which it believed would render any of the sites unsuitable for characterization. Further, NRC did not take a position on the ranking of the sites in order of preference, because this could be viewed as a prejudgment of licensing issues. NRC was not aware of any reason that would indicate that any of the candidate sites was unlicenseable. Nor has NRC made any such finding to date with respect to any site identified as potentially acceptable. In March 1987, Congress began drafting legislation to amend the repository program. NRC provided comments on a number of these draft amendments. In December 1987, the NWPAA was enacted. In a major departure from the initial intent of the 51-SC-48 NWPA, the new law required that DOE suspend site characterization activities at sites other than the Yucca Mountain site. This decision was not based on a technical evaluation of the three recommended sites or a conclusion that the Hanford and Deaf Smith sites were not technically acceptable. According to sponsors of the legislation, the principal purpose of the requirement to suspend characterization at these sites was to reduce costs. In effect, the NWPAA directed DOE to characterize candidate sites sequentially, if necessary, rather than simultaneously. If DOE determines at any time that the Yucca Mountain site is unsuitable, DOE is to terminate all site characterization activities and report to Congress its recommendations for further actions. The NRC staff has identified numerous issues regarding the Yucca Mountain site that may have a bearing on the licenseability of that site. These issues will have to be resolved during site characterization. An example of a site issue that may bear on the question of suitability is tectonic activity, the folding or faulting of the earth's crust. In the 1984 Waste Confidence Decision, NRC noted that " ... the potential sites being investigated by DOE are in regions of relative tectonic stability." The authority for this statement came from the Position Statement of the US Geological Survey (USGS). NRC has raised concerns regarding tectonic activity at the Yucca Mountain site in the comments on the draft and final EAs, in the draft and final Point Papers on the Consultation Draft Site Characterization Plan, and in the Site Characterization Analysis for the Yucca Mountain site. If it appears during site characterization that the Yucca Mountain site will be unable to meet NRC requirements regarding isolation of waste, DOE will have to suspend characterization at that site and report to Congress. DOE's program of site screening in different geologic media was consistent with section 112(a) of the NWPA, which required that DOE recommend sites in different geologic media to the extent practicable. This strategy was to ensure that if any one site were found unsuitable for reasons that would render other sites in the same geologic medium unacceptable, alternate sites in different host rock types would be available. NRC referred to this policy in its 1984 Waste Confidence Decision, when it said, in support of its argument on technical feasibility, that " ... DOE's program is providing information on site characteristics at a sufficiently large number and variety of sites and geologic PART 51 STATEMENTS OF CONSIDERATION media to support the expectation that one or more technically acceptable sites will be identified." NRC recognizes that simultaneous site characterization is not necessary to identify a repository site that would meet NRC's technical criteria for isolating wastes. Sequential site characterization does not necessarily preclude or hinder identification of an acceptable site for a repository. NRC did express concern to Congress, on several occasions during deliberations over the proposed legislation, that sequential site characterization could delay considerably the schedule for opening a repository if the site undergoing characterization were found to be unlicenseable. NRC also indicated that this potential for delay would have to be considered by NRC in reevaluating the findings in its Waste Confidence Decision. The impact of this redirection of the high-level waste program on the Commission's Waste Confidence findings is not on the ability to identify technically acceptable sites, but on the timing of availability of technically acceptable sites. Because characterization of multiple sites appears to be more directly related to the timing of repository availability than to the feasibility of geologic disposal, consideration of the above statement in light of the NWPAA program redirection will be discussed under Finding 2. Another question bearing on whether technically acceptable sites can be found is whether compliance with Environmental Protection Agency [EPA) environmental standards for disposal of spent fuel and high-level waste can be demonstrated. These standards, originally promulgated in final form in September 1985, were vacated in July, 1987, by the U.S. Court of Appeals, and remanded to EPA for further consideration [see NRDC v. EPA, 824 F. 2d 1258). As originally promulgated, the standards set limits on releases of radioactive materials from the site into the accessible environment over a 10,000-year period following disposal. They also required that there be less than one chance in ten that the release limits will be exceeded in 10,000 years, and less than one chance in 1,000 that releases will exceed ten times the limits over 10,000 years. 'In past comments on draft and proposed EPA standards, and in related NRC rulemaking efforts, NRC has expressed concern that probabilistic analyses should not be exclusively relied on to demonstrate compliance with EPA release limits. NRC's comments said in part that " ... [t]he numerical probabilities in [the standards] would require a degree of precision which is unlikely to be achievable in evaluating a real waste disposal system." The comments went on to explain that " ... identification of the relevant processes and events affecting a particular site will require considerable judgment and will not be amenable to accurate quantification, by statistical analysis, of their probability of occurrence." NRC believed then, and continues to believe, that it must make qualitative judgments about the data and methodologies on which the numerical probabilities were based. In response to NRC concerns, EPA incorporated language into its 1985 standards that appeared to allow flexibility to combine qualitative judgments with numerical probability estimates in a way that might have made implementation of the EPA standards practicable. The text of those standards recognized that "proof of the future performance of a disposal system is not to be had in the ordinary sense of the word" with the substantial uncertainties and very long performance period involved. The 1985 standards emphasized that a "reasonable expectation"--rather than absolute proof--is to be the test of compliance. "What is required," the text of the standards said, "is a reasonable expectation, on the basis of the record ... , that compliance ... will be achieved." In an additional attempt to provide flexibility for implementation of the standards, EPA also provided that numerical analyses of releases from a repository were to be incorporated into an overall probability distribution only "to the extent practicable." This phrase appeared to allow some discretion for NRC to incorporate qualitative considerations into its license making, rather than having to rely solely on numerical projections of repository performance. On the strength of these and other EPA assurances, the Commission did not object when the final standards were published in 1985. The Commission also notes that the EPA standards, as promulgated in 1985, contained a provision for development of alternative standards by EPA. The Federal Register text (50 FR 38074, September 19, 1985) describing this alternative standards provision stated: There are several areas of uncertainty the Agency [EPA) is aware of that might cause suggested modifications of the standards in the future. One of these concerns is implementation of the containment requirements for mined geologic repositories. This will require collection of a great deal of data during site characterization, resolution of the inevitable uncertainties in such information, and adaptation of this information into probabilistic risk 51-SC-49 assessments. Although the Agency is currently confident that this will be successfully accomplished, such projections over thousands of years to determine compliance with an environmental regulation are unprecedented. If--after substantial experience with these analyses is disposal systems that clearly provide good isolation cannot reasonably be shown to comply with the containment requirements, the Agency would consider whether modifications to [the standards) were appropriate. This statement suggests to the Commission that EPA would be willing to consider modifications to the standard's containment requirements in the event that their probabilistic formulation is found to hamper or preclude an adequate evaluation of a proposed repository's capability to isolate radioactive waste. Pursuant to the remand by the Federal court in 1987, EPA is currently revising its standards for disposal of spent fuel and high-level waste. The court's decision directed that the remand focus on the ground water and individual protection requirements of the standards. Although the EPA standards are still undergoing development at this time, the Commission does not currently see a sufficient basis to withdraw its confidence in the feasibility of evaluating compliance with such standards. NRC staff will closely monitor the development of the repromulgated standards. In sum, considering both past and current programs for characterizing sites, the Commission concludes that technically acceptable sites for a repository can be found. The Commission is confident that, given adequate time and resources, such sites can be identified, evaluated, and accepted or rejected on their merits, even if no more than one site is undergoing site characterization. This judgment does not rest on the acceptability of the Yucca Mountain site or any one future candidate site. I.A.2. The development of effective waste packages. I.A.2.a. Considerations in developing waste packages. The NWPA required NRC to promulgate technical requirements and criteria to be applied in licensing a repository for high-level radioactive waste. Under Section 121 of the Act, these technical criteria must provide for use of a system of multiple barriers in the design of the repository and such restrictions on the retrievability of waste as NRC deems appropriate. The system of multiple barriers includes both engineered and natural barriers. PART 51 STATEMENTS OF CONSIDERATION The waste package is the first engineered barrier in the system of multiple barriers to radionuclide escape. The waste package is defined as the "waste form and any containers, shielding, packing and other absorbent materials immediately surrounding an individual waste container." Before sinking an exploratory shaft for site characterization, DOE is required to prepare an SCP including a description of the waste form or packaging proposed for use at the repository, and an explanation of the relationship between such waste form or packaging and the geologic medium of the site. The multiple barrier approach to radioactive waste isolation in a geologic repository is implemented in NRC requirements by a number of performance objectives and by detailed siting and design criteria. The NRC performance objective for the waste package requires substantially complete containment for a period of not less than 300 years nor more than 1000 years after permanent closure of the repository. The technical design criteria for the waste package require that interaction of the waste package with the environment not compromise performance of the package, the underground facility, or the geologic setting. Therefore, the waste package design must take into account the complex site-specific interactions between host rock, waste package, and ground water that will affect waste package and overall repository performance. Under the NWPAA, DOE was required to suspend site characterization activities at sites other than the Yucca Mountain, NV site. Consequently, DOE has narrowed the range of waste package designs to a design tailored for unsaturated tuff at the Yucca Mountain site. This aspect of the high-level waste program redirection may facilitate and expedite the waste package design process insofar as it enables DOE to concentrate its efforts on developing a single design for a single site instead of three designs for sites in bedded salt, basalt, and unsaturated luff. Currently, DOE is evaluating uncertainties in waste package design related to waste form, container type, and environment. The current conceptual design for the waste package is based on several assumptions. The waste form is presumed to be old spent fuel or high-level waste in the form of borosilicate glass in steel canisters. (In addition to spent fuel and high-level waste, the waste form may include greater-than-Class C (GTCC) low-level waste. This waste is not routinely acceptable for near-surface disposal under NRC regulations for disposal of low-level wastes, but is acceptable for disposal in a repository licensed for disposal of spent fuel and high-level wastes. Tpis waste might include such materials as sealed sources and activated metals from the decommissioning of reactors and production facilities.) Six materials are being considered for fabrication of containers, including austenitic steel (3161), nickel-based alloys (Alloy 825), pure copper (CDA 102), copper-based alloys bronze, CDA-613, and 70-30 Cu-Ni, CDA-715), and a container with a metal outer shell and ceramic liner. The reference container for the spent fuel and level waste is a 1.0-crn thick cylinder to be made of American Iron and Steel Institute (AISI) 3041 stainless steel. This will be DOE's benchmark material, against which other materials are to be compared. DOE currently intends for spent fuel containers to be filled with an inert gas, such as argon, before being welded closed. In addition to these six materials, DOE also plans to assess the merits of alternative waste package materials and designs. The reference repository location is in the unsaturated tuff of the Topopah Spring Formation underlying Yucca Mountain. According to DOE, little flowing water is thought to be present there to contribute to corrosion of the waste containers, although the degree of saturation in this tuff is estimated to be 65 (plus or minus) 19 percent of the available void space in the rock. DOE has acknowledged, however, that the greatest uncertainties in assessing waste package performance at Yucca Mountain stern from difficulty in characterizing and modeling the coupled geochemical-hydrologic processes that represent the interactions between the host rock, waste package, and ground water. The final waste package design will depend on the results of site characterization and laboratory testing to reduce uncertainty in predicting these interactions in the reference repository horizon. The final design will also be shaped by research in understanding the degradation of candidate container materials, and the characteristics of the likely reference waste forms. Regarding the state of technology for developing long-lived waste package containers, the Swedish Nuclear Fuel and Waste Management Company (SKB), the organization responsible for radioactive waste disposal in Sweden, has described a container for spent fuel rods that consists of a 0.1-rn thick copper canister surrounded by a 51-SC-50 bentonite overpack. The design calls for pouring copper powder into the void spaces in the canisters, compacting the powder using hot-isostatic pressing with an inert gas, and sealing the canisters. SKB estimates that the copper canister waste package has a million-year lifetime. (See also I.B.3. below.) As noted in NRC's Final Point Papers on the Consultation Draft Site Characterization Plan, the Commission does not expect absolute proof that 100 percent of the waste packages will have 100 percent containment for 300 to 1000 years. Since that time, the NRC staff has completed its review of the December 1988 Site Characterization Plan for Yucca Mountain. Although the Commission continues to have concerns about DOE's waste package program, nothing has occurred to diminish the Commission's confidence that as long as DOE establishes conservative objectives to guide a testing and design program, in tuff or in other geologic media if necessary, it is technically feasible to develop a waste package that meets the performance objective for substantially complete con tainrnen t. I.A.2.b. Effect of reprocessing on waste form and waste package. The Draft 1988 Mission Plan Amendment estimates that about 77,800 metric tons of heavy metal (MTHM) of spent nuclear fuel will be available for disposal by the year 2020. (This estimate is based on a "no new orders" assumption for commercial nuclear reactors and a 40-year reactor lifetime.) Also, approximately 9400 MTHM of reprocessed defense waste and a small amount of commercial reprocessed waste from the West Valley Dernonstra tion Project is estirna ted to be available for disposal by 2020. The decision to locate the defense high-level waste in the repository for wastes from commercial power reactors resulted from the requirement in Section 8 of the NWPA that the President evaluate the possibility of developing a waste-only repository. In February 1985, DOE submitted a report to the President recommending a combined commercial and defense repository. In April 1985, the President agreed that no basis appeared to exist for a defense-only repository and directed DOE to dispose of defense waste in the commercial repository. About 8750 MTHM of reprocessed high-level waste from defense facilities at Savannah River, SC, Hanford, WA, and Idaho Falls, ID will be available by 2020 for disposal in the repository, according to the Draft 1988 Mission Plan Amendment. This waste will likely be solidified into a borosilicate glass PART 51 STATEMENTS OF CONSIDERATION matrix. About 640 MTHM of reprocessed high-level waste will come from the West Valley Demonstration Project, a facility for wastes from discontinued commercial reprocessing of spent fuel at that site. This reprocessed waste also will be solidified, probably in a borosilicate glass waste form. Waste-form testing for the Yucca Mountain site is focusing on both spent fuel and reprocessed high-level waste. The performance of the waste form in providing the first barrier to radionuclide migration is being evaluated on the basis* of the physical and chemical environment of the waste form after disposal, the performance of the waste container, and the emplacement configuration. A major limitation on glass form testing is that the actual waste glasses to be disposed of are not available, and their exact composition will not be established until after further testing. Reference waste-glass compositions are being used for studies on the effect of variation in glass composition on performance. (These glass compositions are designed by Savannah River Laboratory (SRL) for defense high-level waste, and by Pacific Northwest Laboratory (PNL) for the commercial high-level wastes to be vitrified under the West Valley Demonstration Project Act.) The reference compositions will be revised when better analyses of the composition of the wastes at SRL and West Valley are available. The test program will seek to establish upper bounds on leaching of important radionuclides, and the extent to which glass fracturing increases leach rate. Other factors influencing leach rate are temperature, pH of the leaching solution, formation of solid layers on the surface of the waste glass, irradiation, water volume, and chemistry. It is possible that renewed reprocessing of spent fuel from nuclear power reactors may result in a greater proportion of reprocessed waste to spent fuel than is currently anticipated. Although such a departure from the current plan to dispose of mostly unreprocessed spent fuel in the repository does not appear likely at this time, the Commission believes it is important to recognize the possibility Ip.at this situation could change. The possibility of disposal of reprocessed waste as an alternative waste form to spent fuel assemblies was recognized by the Commission in the 1984 Waste Confidence Decision. The Commission noted that the disposal of waste from reprocessing had been studied for a longer time than the disposal of spent fuel, and that the possibility of reprocessing does not alter the technical feasibility of developing a suitable waste package. The Commission went on to say that there is evidence that the disposal of reprocessed high-le'lel waste may pose fewer technical challenges than the disposal of spent fuel. As long as DOE uses conservative assumptions and test conditions for evaluating the performance of different waste forms against NRC licensing requirements, the Commission has no basis to change its finding that there is reasonable assurance that reprocessing does not reduce confidence in the technical feasibility of designing and building a waste package that will meet NRC licensing requirements in a variety of geologic media. l.A.3. The development of effective engineered barriers for isolating wastes from the biosphere I.A.3.a. backfill materials. At the time of the 1984 Waste Confidence Decision, DOE was developing conceptual designs for backfill in several geologic media. Most candidate sites at that time' were in saturated rock, and the conceptual designs included backfilling or packing around waste containers to prevent or delay ground water flow which could enhance corrosion and radionuclide transport near the waste containers. The conceptual design for the engineered barrier system at the Yucca Mountain site has different parameters because the site is unsaturated; instead of backfill or packing around the waste container, there is to be an air gap between sides of the waste canister and the host rock. Backfill material around the container is not required under NRC regulations for the waste package. NRC regulations require that " ... containment of high-level waste within the waste packages [which includes the container] will be substantially complete for a period to be determined by the Commission ... provided, that such period shall not be less than 300 years nor more than 1000 years after permanent closure of the repository" [10 CFR subsection 60.113(a)(1J(ii)(B)], and that the entire engineered barrier system meet the release rate performance objective of 1 part in 100,000 per year. Backfill is also a component of the borehole, shaft, and ramp seals, which are not part of the engineered barrier system or the underground facility. Boreholes, shafts, and ramps must be sealed when the repository is permanently closed. This aspect of backfilling is discussed below under "Development of Sealants." Backfill . 51-SC-51 may also include crushed rock used to fill openings such as drifts in the underground facility. At the Yucca Mountain candidate site, DOE currently plans to fill openings in the underground facility at closure of the repository. Backfilling is not planned before repository closure because it is not needed for structural support for the openings, and it would make waste retrieval more difficult. At closure of the facility, however, openings will be backfilled with coarse luff excavated for the facility. In the conceptual design provided in the SCP, the selection of coarse tuff as backfill material is based on numerical simulations performed by DOE which suggest that coarse tuff would be a more effective barrier to capillary flow in the backfill matrix than fine materials. DOE's design for the engineered barrier system submitted with the license application will have to contain information sufficient for NRC to reach a favorable conclusion regarding the overall system performance objective. Backfill or packing around waste containers is not required by NRC regulations if DOE can demonstrate that applicable performance objectives can be met without it. If, on the basis of testing and experiments during site characterization, DOE decided that backfill would enhance engineered barrier system performance, the design would have to reflect this conclusion. DOE has already conducted research on a wide variety of candidate materials for backfill around waste packages in a variety of geologic media. The Commission continues to have confidence that backfill or packing materials can be developed as needed for the underground facility and waste package to meet applicable NRC licensing criteria and performance objectives. I.A.3.b. Borehole and shaft seals. The engineered barrier system described above is limited to the waste package and the underground facility as defined in 10 CFR part 60. The underground facility refers to the underground structure, including openings and backfill materials, but excluding shafts, boreholes, and their seals. Containment and release-rate requirements are specified for the engineered barrier system, but not for the borehole and shaft seals. Seals are covered under 10 CFR section 60.112, the overall post-closure system performance objective for the repository. Among other things, this provision requires that shafts, boreholes and their seals be designed to assure that releases of radioactive materials to the accessible PART 51 STATEMENTS OF CONSIDERATION environment following permanent closure conform to EPA's generally applicable standards for radioactivity. Although the criteria for seals given in 10 CFR part 60 do not specifically mention seals in ramps and the underground facility, it is reasonable to consider them together with borehole and shaft sealants, because the seals and drainage design in ramps and the underground facility could also affect the overall system performance of the geologic repository. Construction of the exploratory shaft facility (ESF) will be the first major site characterization activity at the repository horizon. Currently, DOE is reviewing its plans for construction of exploratory shafts. According to the 1989 "Reassessment Report," DOE is reevaluating the "locations chosen for the two exploratory shafts, the method chosen (drilling and blasting) for the construction of the shafts, the means of access (ramps or shafts) to the repository horizon, the need for additional exploratory drifts, and the design of the shafts and other components of the exploratory shaft facility." This reevaluation of plans for the shaft facility is in response to concerns from the NRC staff and the Nuclear Waste Technical Review Board (NWTRB). When the repository is decommissioned, NRC expects that most, if not all, shafts, ramps, and boreholes will probably have to be sealed to reduce the possibility that they could provide preferential pathways for radionuclide migration from the underground facility to the accessible environment. DOE estimates that as many as 350 shallow and 70 deep exploratory boreholes may be emplaced by the time site characterization has been completed at the Yucca Mountain site. Decommissioning may not occur for up to 100 years after commencement of repository operations. Because the final design for seals will likely have been modified from the initial license application design (LAD), DOE is viewing the seal LAD as serving two primary functions. As set forth in DOE's SCP for the Yucca Mountain candidate site, the seal LAD is to establish that: (1) " ... technology for constructing seals is reasonably available;" and (2) " ... there is reasonable assurance that seals have been designed so that, following permanent closure, they do not become pathways that compromise the geologic repository's ability to meet the closure performance objectives." To establish the availability of technology for seal construction, DOE has identified at least 31 site properties that need to be characterized in determining necessary seal characteristics. These properties include saturated hydraulic conductivity of alluvium near shafts, the quantity of water reaching the seals due to flooding events, and erosion potential in the shaft vicinity. The SCP also discusses material properties that need to be identified to determine sealing components such as initial and altered hydrologic properties of materials. The SCP indicates that DOE is planning to use crushed luff and cements in the sealing program at the Yucca Mountain candidate site. The stated advantages of using tuff include minimizing degradation of seal material and avoiding disruption of ambient ground-water chemistry. DOE's current design concept for meeting the overall performance objectives includes a combination of sealing and drainage. Seal requirements may be reduced in part by: (1) limiting the amount of surface water that may enter boreholes, shafts, and ramps; (2) selecting borehole, shaft, and ramp locations and orientations that provide long flow paths from the emplaced waste to the accessible environment above the repository; and (3) maintaining a sufficient rate of drainage below the repository horizon level so that water can be shunted past the waste packages without contacting them. Although DOE's program is focusing on seals for the Yucca Mountain candidate site, the Commission finds no basis for diminished confidence that an acceptable seal can be developed for candidate sites in different geologic media. The Commission finds no evidence to suggest that it can not continue to have reasonable assurance that borehole, shaft, ramp, and repository seals can be developed to meet 10 CFR part 60 performance objectives. J.B. Relevant Issues That Have Arisen Since the Commission's Original Decision l.B.1. In support of its argument on technical feasibility, the Commission stated in its 1984 Waste Confidence Decision that " ... DOE's program is 51-SC-52 providing information on site characteristics at a sufficiently large number and variety of sites and geologic media to support the expectation that one or more technically acceptable sites will be identified." The NWPAA required, however, that DOE suspend site-specific site characterization activities under the Nuclear Waste Policy Act of 1982 at all sites other than the Yucca Mountain, NV site. Under the NWPAA, the DOE program has been redirected to characterize candidate repository sites in sequence rather than simultaneously. If the Yucca Mountain site is found to be unsuitable, DOE must terminate site characterization activities there and provide Congress with a recommendation for further action, such as the characterization of another site. Because characterization of multiple sites now appears to be more directly related to the timing of repository availability than to the technical feasibility of geologic disposal as a concept, consideration of the Commission's aforementioned 1984 statement in light of the NWPAA will be discussed under Finding 2. I.B.2. What is the relationship, if any, of the "S-3 Proceeding" to the current review of the Commission's 1984 Waste Confidence Findings? Would the planned revision of the S-3 rulemaking be affected if the Commission had to qualify its current confidence in the technical feasibility of safe disposal? In its decision to remand to NRC the questions of whether safe offsite storage would be available by 2007-2009, or, if not, whether spent fuel could be safely stored onsite past those dates, the U.S. Circuit Court of Appeals observed that the issues of storage and disposal of nuclear waste were being considered by the Commission in an ongoing generic proceeding known as the "S-3" Proceeding. The S-3 Proceeding was the outgrowth of efforts to address generically the NEPA requirement for an evaluation of the environmental impact of operation of a light water reactor (LWR). Table S-3 assigned numerical values for environmental costs resulting from uranium fuel cycle activities to support one year ofLWR operation. NRC promulgated the S-3 rule in April 1974. In July 1976, the U.S. Circuit Court of Appeals found that Table S-3 was inadequately supported by the record regarding reprocessing of spent fuel and radioactive waste management, in part because the Commission, in reaching its assessment, had relied heavily on PART 51 STATEMENTS OF CONSIDERATION testimony of NRC staff that the problem of waste disposal would be resolved. When the U.S. Circuit Court of Appeals issued the remand on what were to become the "Waste Confidence" issues in May 1979, NRC had pending before it the final amended S-3 rule. The Court regarded the resolution of the issue of waste disposal in the S-3 proceeding as being related to the issue raised by the petitioners in the appeals of the NRC decisions on the expansion of spent fuel storage capacity. The Court said that the " ... disposition of the S-3 proceeding, though it has a somewhat different focus, may have a bearing on the pending cases." The Commission approved the final S-3 rule in July 1979. In October 1979, the Commission issued a Notice of Proposed Rulemaking (NPR) on the Waste Confidence issues in response to the remand by the Court of Appeals. In the NPR, the Commission stated that the proceeding would " ... draw upon the record compiled in the Commission's recently concluded rulemaking on the environmental impacts of the nuclear fuel cycle, and that the record compiled herein will be available for use in the general fuel cycle rule update discussed in that rulemaking." In the final Table S-3 rule issued in 1979, the Commission had said that " ... bedded salt sites can be found which will provide effective isolation of radioactive waste from the biosphere." When the Commission issued the 1984 Waste Confidence Decision, part of the basis for the discussion of waste management and disposal in the August 1979 final S-3 rule had changed. For example, in 1984 the repository program was proceeding under the NWPA, which required that DOE recommend three sites for si le characterization. NRC is preparing to amend 10 CFR 51.51, adding new estimates for releases of Tc-99 and Rn-222, and a revised narrative explanation describing the basis for values contained in Table S-3. The amendment would also explain the environmental effects of potential releases from the light water reactor (LWRJ fuel cycle, and postulate the potential radiation doses, health effects, and environmental impacts of these releases. It is unlikely that the revision will have any impact on the Commission's generic findings in the Waste Confidence proceeding. Nor is it likely that this reexamination of the Waste Confidence findings will affect the S-3 rule; the Waste Confidence Proceeding is not intended to make quantitative judgments about the environmental costs of waste disposal. Unless the Commission, in a future review of the Waste Confidence decision, finds that it no longer has confidence in the technical feasibility of disposal in a mined geologic repository, the Commission will not consider it necessary to review the S-3 rule when it reexamines its Was.le Confidence findings in the future. I.B.3. To what extent do developments in spent fuel disposal technology outside of the United States (e.g., Swedish waste package designs) enhance NRC's confidence in the technical feasibility of disposal of level waste and spent fuel? Spent fuel disposal technology is the subject of extensive research investigation in both Europe and North America. Advances in this technology are being communicated to the NRC staff both through bilateral agreements and the presentation of research results at international meetings. Outside the U.S., studies of spent fuel as a waste form are now being conducted primarily in Canada and Sweden, although both France and West Germany have small programs in this area. The Swedish studies have been mainly concerned with boiling water reactor (BWR) spent fuel, whereas the Canadian studies focus on spent fuel from that country's CANDU reactors, which use unenriched uranium in a core immersed in "heavy" water made from deuterium. BWR and CANDU fuel, like pressurized water reactor [PWR) fuel, are uranium dioxide fuels clad in zircaloy. However, the burnup rates for these three fuel types vary considerably. Ongoing research studies on spent fuel include: work on the characterization of spent fuel as a waste form; the corrosion of spent fuel and its dissolution under oxidizing and reducing conditions; the radiolysis of ground water in the near vicinity of the spent fuel, and its effects on the dissolution of the fuel; and the development of models to predict the leaching of spent fuel over long time periods. The results of this work are steadily increasing our understanding of spent fuel as a waste form. High-level radioactive waste, whether it is spent reactor fuel or waste from reprocessing, must be enclosed in an outer canister as part of the waste package. The canister surrounding the waste is expected to prevent the release of radioactivity during its handling at the repository site before emplacement. After emplacement in the repository, it is expected to prevent the release of radioactivity. for a specified period of time after the repository is closed, by providing a barrier to protect the waste from coming into contact with ground water. 51-SC-53 For practical reasons, canister materials may be divided into the following classes: (1) completely or partially thermodynamically stable materials such as copper; (2) passive materials such as stainless steel, titanium, Hastelloy, Inconel, and aluminum; (3) corroding or sacrificial materials such as lead and steel; and (4) non-metallic materials such as alumina and titanium dioxide ceramics and cement. Sweden has been conducting an extensive canister research program over the past several years. The main canister material of interest is copper, but titanium, carbon steel, and alumina and titanium dioxide are also being studied as reasonable alternatives, should unexpected problems be discovered with using pure copper. One of the Swedish canister designs is a 0.1-m thick copper container (as described previously in section I.A.2.a.), which is claimed to provide containment, in conjunction with an appropriate backfill material, for a period on the order of one million years. The critical factors for the isolation period for copper canisters are: (1) the presence of corrosive substances such as sulphide ions in the ground water; (2) the possibility of these substances reaching the canister surface; and (3) the degree of inhomogeneity, or pitting, of the resulting corrosion. Studies are continuing to obtain more information on pitting corrosion of copper and on techniques for welding thick-walled copper containers. Several conceptual designs for canisters for the safe disposal of unreprocessed spent fuel have also been developed in Canada. One canister design option is the supported-shell, metal-matrix concept, which involves packing the spent fuel bundles into a thin corrosion-resistant shell and casting the remaining space with a low melting point metal or alloy. Structural support for the shell would be provided by the resulting metal matrix. Lead is a possible matrix material because of its favorable casting properties, cost, and low melting point. Other supported shell canister concepts include the packed-particulate and structurally-supported designs. In these designs, a thin outer shell is supported by a particulate material packed around a steel internal structure that contains the spent fuel bundles. Several materials have been identified for the fabrication of the corrosion resistant outer shell, including commercially pure and low-alloy titanium, high nickel-based alloys such as Inconel 625, and pure copper. PART 51 STATEMENTS OF CONSIDERATION Detailed designs have been produced for all three types of supported shell canisters incorporating either a titanium or nickel alloy shell less than 6-mm thick. A conceptual design has also been produced for a copper-shell supported canister and a metal-matrix container with a relatively thick (25-mm) copper shell and a lead matrix material. This last canister is intended to contain 72 used CANDU fuel bundles in four layers of 18 bundles each. Both the Canadian and Swedish conceptual designs for the disposal of spent fuel in canisters provide for surrounding the canister with backfill material as part of the waste package when it is emplaced in the repository. This backfill material would be packed around the canister to retard the movement of ground water and radionuclides. Investigations of backfill material at the Stripa mine in Sweden have shown that bentonite and silica sand can be employed successfully as backfill, both around the canister and in repository tunnels. A bentonite-silica mixture is the recommended backfill material on the basis of its thermal and mechanical properties. Bentonite backfills have been shown to produce hydraulic conductivities that are very similar to the surrounding granite at Stripa. Problems concerning the variability of bentonite samples from different geographic locations can be eliminated if material from a single source is used. The presence of sulfur and some organic material, including bacteria, in many bentonites poses some problems related to accelerated corrosion. Treatment with hydrogen peroxide may be used to oxidize these organics. Heating the bentonite to 400 degrees C can also be effective, although this may alter the crystal structure of the bentonite. Many countries intend to dispose of their high-level radioactive waste by first converting the wastes into a solid, vitrified form after reprocessing. Since the leaching of the waste form by circulating ground water after disposal is the most likely mechanism by which the radionuclides might be returned to the biosphere, the waste form must be composed of a highly stable material with an extremely low solubility in ground water. Thus, the waste form itself should function as an immobilization agent to prevent any significant release ofradionuclides to the biosphere over very long time periods. The two primary materials currently being considered for use as solidified waste forms are borosilicate glass and SYNROC, a man-made titanate ceramic material. SYNROC was initially developed in Australia as an alternative material to borosilicate glass. It is composed primarily of three minerals (hollandite, zirconolite, and perovskite) which collectively have the capacity to accept the great majority of radioactive level waste constituents into their crystal lattice structure. These three minerals, or closely related forms, occur naturally, and have been shown to have survived for many millions of years in a wide range of natural environments. SYNROC has the property of being extremely resistant to leaching by ground water, particularly at temperatures above 100 degrees C. In addition, the capacity of SYNROC to immobilize high-level wastes is not markedly impaired by high levels of radiation damage. The high leach-resistance of SYNROC at elevated temperatures increases the range of geologic environments in which it may be used, such as deep geologic repositories in both continental and marine environments. Research and development work on improving SYNROC production technology is currently being done jointly in Australia and Japan. New methods of using metal alkoxides in the fabrication of SYNROC to obtain high homogeneity and lowered leachability have recently been developed in Australia. The Japanese have recently developed a new method that uses titanium hydroxide, as a reducing agent to produce SYNROC with a high density and low leach rate. A pilot facility for the production of non-radioactive SYNROC is now in operation in Australia, and a small pilot facility for producing SYNROC with radioactive constituents is being completed in Japan. On the basis of current information from the foreign studies just described on canisters, spent fuel as a waste form, backfill materials, and alternatives to borosilicate glass waste forms, the Commission concludes that there is no basis for diminished confidence that an acceptable waste package can be developed for safe disposal of high-level waste and spent fuel. J.C. Conclusion on Finding 1 The Commission has reexamined the basis for its First Finding in the 1984 Waste Confidence Decision in light of subsequent program developments, and concludes that Finding 1 should be reaffirmed. The technical feasibility of a repository rests initially on identification of acceptable sites. At this time, the Commission is not aware of any evidence indicating that Yucca 51-SC-54 Mountain is not acceptable for site characterization. There are many outstanding questions regarding the licenseability of the site, however. and they must be answered satisfactorily in order for NRC to issue a construction authorization for that site. If data obtained during site characterization indicate that the Yucca Mountain site is not suitable for a repository, DOE is required by the NWPAA to terminate site characterization activities and report to Congress. Within six months of that determination, DOE must make a recommendation to Congress for further action to assure the safe, permanent disposal of spent fuel and high-level waste. DOE could recommend, for example, that Congress authorize site characterization at other sites. Considering DOE's investigations of other potentially acceptable sites before its exclusive focus on Yucca Mountain, the Commission has no reason to believe that, given adequate time and program resources, a technically acceptable site can not be found. The technical feasibility of geologic disposal also depends on the ability to develop effective engineered barriers, such as waste packages. DOE is currently evaluating six candidate materials for waste containers, including austenitic steel and copper-and based alloys, and is planning form testing based on both spent fuel and high-level waste in borosilicate glass. On the basis of DOE's program, and results from Swedish investigations of a copper waste container, the Commission is confident that, given a range of waste forms and conservative test conditions, the technology is available to design acceptable waste packages. In addition to the materials testing for the waste container and waste form, there may be additional measures that can be taken to improve the effectiveness of the engineered barriers. It is known, for example, that the loading characteristics of the wastes diminish with time. Also, the longer wastes are stored before disposal, the smaller will be the quantities of radionuclides available for transport to the accessible environment. It is also technically feasible to separate from radioactive wastes the radionuclides that constitute the principal source of heat from the nuclides of greatest long-term concern. The former radionuclides, mainly fission products such as cesium-137 and strontium-90, could then be stored for a period of years while the fission products decay to the point where they could be disposed of either in a manner PART 51 STATEMENTS OF CONSIDERATION that does not require the degree of confinement provided by a geologic repository, or in a repository with less concern for thermal disturbance of the host rock's expected waste isolation properties. Meantime, the longer-lived remaining radionuclides, such as transuranic wastes with elements heavier than uranium, could be disposed of in a repository away from the fission products and without the high thermal loadings that would otherwise have to be considered in predicting the term waste isolation performance of the geologic setting. France, Great Britain, and Japan are currently pursuing this waste management strategy or a variant of it. The Commission emphasizes here that it does not believe that recycling technologies are required for the safety or feasibility of deep geologic disposal in the United States. Other countries, such as Canada, the Federal Republic of Germany, and Sweden are pursuing disposal strategies based on a similar view. Reprocessing, if employed in its current stage of development, would result in additional exposures to radiation and volumes of radioactive wastes to be disposed of. For the purpose of finding reasonable assurance in the technical feasibility of geologic disposal, however, it is worth noting that technology is currently available to permit additional engineering control of waste forms if, for reasons not now foreseen, such control were deemed desirable at some future time. Meanwhile, the Commission continues to have confidence that safe geologic disposal is technically feasible for both spent fuel and high-level waste. DOE's reference design for the waste package in the December 1988 Site Characterization Plan does not include backfill or packing around waste containers in the emplacement boreholes. Neither is required under NRC rules so long as DOE can show that applicable regulatory criteria and objectives will be met. An air gap between the container and the host rock is currently one of the barriers in DOE's design for meeting the performance objective. DOE has conducted investigations on a variety of candidate materials for backfill in a variety of geologic media, and the Commission finds no basis to qualify its past confidence that backfill materials can be developed, if needed, to meet applicable NRC requirements. The December 1988 reference design for sealing boreholes, shafts, ramps and the underground facility at the Yucca Mountain candidate site employs crushed tuff and cement. Regardless of the geologic medium of the candidate site, DOE will have to show that the license application design meets NRC post-closure performance objectives. The Commission continues to have reasonable assuranGe that DOE's program will lead to identification of acceptable sealant materials for meeting these objectives. No major breakthrough in technology is required to develop a mined geologic repository. NRC will not be able to license a repository at a particular site, however, until there is sufficient information available for that site. The information needed to license a site includes site characterization data, data on repository design, and waste package design sufficient for performance assessment of the entire waste disposal system. Further, the Commission recognizes the challenge posed by the need to predict impacts of a repository on human health and the environment over very long periods of time. It will not be possible to test the accuracy of term repository performance assessment models in an absolute sense. The NRC does believe that existing performance assessment models have the potential to provide a basis for deciding whether a system for geologic disposal of level waste is acceptable, and can provide a sufficient level of safety for present and future generations under certain conditions. These conditions include addressing uncertainties, and gathering data from specific sites. Overall, from its reexamination of issues related to the technical feasibility of geologic disposal, the Commission concludes that there is reasonable assurance that safe disposal of level waste and spent fuel in a mined geologic repository is technically feasible. Original Finding 2: The Commission finds reasonable assurance that one or more mined geologic repositories for commercial high-level waste and spent fuel will be available by the years 2007-2009, and that sufficient repository capacity will be available within 30 years beyond expiration of any reactor operating license to dispose of existing commercial high-level radioactive waste and spent fuel originating in that reactor and generated up to that time. Revised Finding 2: The Commission finds reasonable assurance that at least one mined geologic repository will be available within the first quarter of the twenty-first century, and that sufficient repository capacity will be available within 30 years beyond the licensed life for operation (which may include the term of a revised or renewed license] of any reactor to dispose of the commercial 51-SC-55 high-level radioactive waste and spent fuel originating in such reactor and generated up to that time. JI.A. Issues Considered in Commission's 1984 Decision on Finding 2 11.A.1. Finding Technically Acceptable Sites in a Timely Fashion In order for the Commission to find that any candidate site for a repository is technically acceptable (that is, in compliance with NRC licensing requirements], the site must undergo comprehensive site characterization to assess its hydrologic, geologic, geochemical, and rock mechanics properties. It is possible that a site may be found unacceptable on the basis of surface-based testing, early in-situ testing or other site characterization activities. It will not be possible, however, for the NRC staff to take a position before a licensing board that a site will meet NRC requirements for construction authorization until the results of all site characterization activities are available. Even then, the staff may conclude that the evidence from site characterization does not constitute reasonable assurance that NRC performance objectives will be met. Also, the results of the licensing hearings on construction authorization cannot be predicted. If construction is authorized and when it is substantially complete, DOE is required to obtain, in addition to the construction authorization permit, a license to receive and possess. waste at the geologic repository operations area in order to commence repository operations. These considerations argue for maintaining the ready availability of alternative sites if, after several years, site characterization or licensing activities bring to light difficulties at the leading candidate site. In support of its argument on technical feasibility, the Commission stated in its 1984 Waste Confidence Decision that " ... DO E's program is providing information on site characteristics at a sufficiently large number and variety of sites and geologic media to support the expectation that one or more technically acceptable sites will be identified." At the time, DOE was required under the NWPA to characterize three candidate repository sites. The NWPAA had a major impact on DOE's repository program, however. Under the NWPAA, DOE was required to suspend site-specific activities at the Hanford, WA and Deaf Smith County, TX sites, which had been approved by the President for site characterization for the first repository. Redirection of the repository program to single-site PART 51 STATEMENTS OF CONSIDERATION characterization (or, if necessary, sequential site characterization if the Yucca Mountain site is found to be unsuitable) will permit DOE to concentrate its efforts and resources on information gathering at a single site, as opposed to spreading out its efforts over a range of sites. The possible schedular benefits to single-site characterization, however, must be weighed for the purposes of this Finding against the potential for additional delays in repository availability if the Yucca Mountain site is found to be unsuitable. By focusing DOE site characterization activities on Yucca Mountain, the NWPAA has essentially made it necessary for that site to be found suitable if the 2007-2009 timeframe for repository availability in the Commission's 1984 Decision is to be met. Clearly, the Commission cannot be certain at this time that the Yucca Mountain site will be acceptable. Although the Commission has no reason to believe that another technically acceptable site can not be found if the Yucca Mountain site proves unsuitable, several factors raise reasonable doubts as to the availability of even one repository by 2007-2009. These include: (1) the current reliance on a single site with no concurrently available alternatives; (2) the probability that site characterization activities will not proceed entirely without problems; and (3) the history of schedular slippages since passage of the NWPA. For example, DOE's schedule for the first repository slipped five years (from 1998 to 2003) between January 1983, when the NWPA was enacted, and January 1987, when the first Draft Mission Plan Amendment was issued. The schedule for excavation of the exploratory shaft for the Yucca Mountain site has slipped by more than five years since the issuance of the PDS in March 1986. In the past several years, DOE has cited numerous reasons for program slippages, including the need for a consultation process with States and Tribes, Congressional actions (e.g., the barring of funds in the 1987 budget appropriation for drilling exploratory shafts], and DOE's recognition that the EIS and license application would require more technical information than previously planned. In the November 1989 *"Report to Congress on Reassessment of the Civilian Radioactive Waste Management Program," DOE announced a further extension of three years until 1992 for sinking the exploratory shaft, and extensions until 2001 for submittal of the license application and 2010 for repository availability. DOE attributes the causes for these delays to prolonging the schedule for site characterization and repository development activities, and to the unwillingness, to date, of the State of Nevada to issue the permits required for DOE to .begin testing. In the "Reassessment Report," DOE proposes to focus the repository program on the evaluation of features of the site that can be studied through surface-based testing, beginning in January 1991. The aim of this surface-based testing program is to make an early determination as to whether there are any features of the site that would render it unsuitable for development as a repository. Of course, the site may be found unsuitable or unlicenseable at any time during the site characterization or licensing process. The NRC supports DOE's efforts to reach an early determination that this may be the case. If the Yucca Mountain site is unsuitable, it will be necessary to begin work to identify and characterize another candidate site for a repository. The sooner this determination is made, the sooner DOE will have an alternative site available for disposal of high-level waste. The NRC had anticipated additional delays in repository program milestones when it issued its Proposed Waste Confidence Decision Review (54 FR 39767). One of the key issues in the repository program to date has been the need for DOE to develop a qualified quality assurance (QA] program. For example, DOE has taken the position, with which NRC agrees, that sinking of exploratory shafts should not occur before it has a qualified quality assurance (QA) program in place. The Commission believes that DOE's aggressive, success-oriented schedule for this milestone did not allow for unexpected developments. Indeed, the effort to develop an acceptable QA program has, in itself, identified problems in design control and other processes that must be resolved in order to establish a qualified program that addresses all applicable NRC licensing requirements. DOE has made progress in development of its QA program with seven contractor plans accepted in October and November 1989. NRC expects that DOE should be able to have the study plans and technical procedures which implement the contractor plans ready in time for surface-based testing at the Yucca Mountain site to begin by January 1991, consistent with the schedule for starting surface-based testing in the Reassessment Report. DOE's current schedule appears to be more realistic than previous schedules. 51-SC-56 Yet even this schedule could prove unattainable due to difficulties of a technical nature that are outside of DOE's control. for example litigation over gaining access to the Yucca Mountain site. Although the NWPAA is a clear and strong reaffirmation of Congressional support for the timely development of a repository, the Commission in this Waste Confidence review cannot ignore the potential for delay in repository availability if the Yucca Mountain site, or any other single site designated for site characterization, is found to be unsuitable. Without alternative sites undergoing simultaneous characterization or even surface-based testing, DOE will have to begin characterizing another site if the site currently selected for characterization proves unsuitable. The earlier a determination of unsuitability can be made, the smaller the impact of such a finding would be on the overall timing of repository availability. DOE has estimated conservatively that it would require approximately 25 years to begin site screening for a second repository, perform site characterization, submit an EIS and license applications, and await authorizations before the repository could be ready to receive waste. In its June 1987 Mission Plan amendment, DOE stated "It ... seems prudent to plan that site-specific screening leading to the identification of potentially acceptable sites should start about 25 years before the start of waste acceptance for disposal." DOE went on to say that it considered this estimate to be conservative because it does not account for expected schedular benefits from the first repository program, including improvements in such areas as site screening, site characterization, and performance assessment techniques. Although DOE's estimate was premised on the successful completion of a program for the first of two repositories, schedular benefits from improvements in the understanding of waste isolation processes would still be available. The glass waste form from the Defense Waste Processing Facility now under construction at Savannah River, SC, for example, will be available for testing under simulated repository conditions well before the turn of the century under current DOE schedules, and improvements in the modelling of spent fuel behavior within waste canisters can be applied in performance assessments largely irrespective of the geology of a site. It may also be pertinent that when DOE made its 25-year estimate for the second repository program in mid-1987, the law at the time PART 51 STATEMENTS OF CONSIDERATION required the simultaneous characterization of three sites, so that DOE could not proceed to develop one site for a repository until the completion of characterization at the site that required the most time. In view of DOE's new schedule, it no longer appears feasible for repository operation to commence prior to 2010. As stated in the Proposed Decision Review, the Commission does not believe it would be prudent to reaffirm the Agency's 1984 finding of reasonable assurance that the 2007-2009 timetable will be met. As the Court of Appeals noted in remanding this issue to NRC, the ultimate determination of whether a disposal facility will be available when needed " ... can never rise above a prediction." The Commission is in the position of having to reach a definitive finding on events which are approximately two decades away. We believe that the institutional timescale for this question can more realistically be framed in decades than in years. As the program proceeds into the next century, it will become easier for NRC to make more definitive assessments, if necessary, of the time a repository will be available. In light of all these considerations, the Commission believes it can have reasonable assurance that at least one repository will be available within the first quarter of the twenty-first century. This estimate is based on the time it would take for DOE to proceed from site screening to repository operation at a site other than Yucca Mountain, if this should prove necessary. Assuming for the sake of conservatism that Yucca Mountain would not be found suitable for repository development, it is reasonable to expect that DOE would be able to reach this conclusion by the year 2000. This would leave 25 years for the attainment of repository operations at another site. NRC will reassess progress towards attaining repository operation by 2025 prior to 2000 during its next scheduled review of its Waste Confidence Findings, if not sooner. DOE's current focus on surface-based testing as an early indicator of repository suitability should help provide a strong basis for evaluating the likelihood of meeting the 2025 estimate of repository availability. Jl.A.2. Timely Development of Waste Packages and Engineered Barriers. The November 1989 Reassessment Report announced that "major activities related to the design of a repository at the Yucca Mountain site and waste package are being deferred. They will be resumed when more information is available concerning the suitability of the site. This approach will conserve resources and allow the DOE to concentrate efforts on scientific investigations." Prior to the Reassessment Report, DOE's most recent conceptual design for the waste package was discussed in the Site Characterization Plan (SCP) for the Yucca Mountain site. As information is obtained from site characterization activities and laboratory studies, the conceptual design will evolve in successive stages into the Advanced Conceptual Design (ACD), the LAD, and the final procurement and construction design. DOE has identified four areas of investigation related to the waste package LAD: (1) waste package environment; (2) waste form and materials testing; (3) design, analysis, fabrication, and prototype testing; and (4) performance assessment. Numerous uncertainties exist in each of these areas. DOE's testing program will attempt to reduce uncertainties in these areas where possible. For example, situ testing is expected to decrease significantly uncertainties regarding the repository host rock mass in which the waste packages will be emplaced. In the area of performance assessment, however, where results of relatively short-term testing of complex waste-ground water interactions must be extrapolated over as many as 10,000 years, it may be necessary to rely more heavily on the use of simplifying assumptions and bounding conditions than in other areas of investigation. As discussed under Finding 1, the Commission continues to have reasonable assurance that waste packages and engineered barriers can be developed which will contribute to meeting NRC performance objectives for the repository. Development of acceptable waste packages and engineered barriers for a repository in the 2010 timeframe will depend on the overall acceptability of the Yucca Mountain site. If the site is found to be unsuitable, waste package and engineered barrier development will have to begin for a different site, because under the NWPAA, DOE may not carry out site characterization and waste package development work at sites other than the Yucca Mountain site. Although much of the work related to waste form, materials, and performance assessment for the waste package can proceed independently of in-situ testing, the investigations related to waste package environment depend on the schedule for this testing. The schedule for in-situ testing depends on when DOE is able to resolve outstanding issues which have impeded shaft sinking and 51-SC-57 in-situ testing, and on DOE's being granted access to the site to begin surface-based testing. In sum, the Commission is not aware of any scientific or technical problems so difficult as to preclude development of a waste package and engineered barrier for a repository at Yucca Mountain to be available within the first quarter of the twenty-first century. Moreover, even given the uncertainty regarding the ultimate finding of site acceptability, and the uncertainty concerning the range of site-related parameters for which the engineered facility and waste package will have to be designed, the Commission finds reasonable assurance that waste package and engineered barrier development can be completed on a schedule that would permit repository opera lion within the first quarter of the twenty-first century. If necessary (that is, if Yucca Mountain were found unsuitable by the turn of the century), DOE could initiate site characterization and develop waste packages and engineered barriers at another site or sites and still commence operation before the end of the first quarter of that century. 11.A.3. Institutional Uncertainties. II.A.3.a. Measures for dealing with Federal-State-local concerns. In its 1984 Waste Confidence Decision, the Commission found that the NWPA should help to minimize the potential that differences between the Federal Government and States and Indian tribes will substantially disrupt or delay the repository program. The Commission noted that the NWPA reduced uncertainties regarding the role of affected States and tribes in repository site selection and evaluation. The Commission also said that the decision-making process set up by the NWPA provides a detailed, step-by-step approach that builds in regulatory involvement, which should also provide confidence to States and tribes that the program will proceed on a technically sound and acceptable basis. Despite the expected and continuing State opposition to DOE siting activities, the Commission has found no institutional developments since that time that would fundamentally disturb its 1984 conclusions on this point. NRC regulatory involvement, for example, has indeed been built into the process. DOE has continued its interactions with NRC regarding repository program activities since the Commission's 1984 Waste Confidence decision was issued. NRC provided comments to DOE on major program PART 51 STATEMENTS OF CONSIDERATION documents such as the Siting Guidelines and the PDS as required by the NWPA, and NRC concurred on those documents. NRC also reviewed and provided comments to DOE on the DEAs and FEAs. In the December 22, 1986 letter to DOE on the FEAs, the NRC staff noted that " ... significant efforts were made by DOE to respond to each of the NRC staff major comments on the DEAs, and in fact, many of these comments have been resolved." NRC provided comments to DOE on the 1987 Draft Mission Plan Amendment, and DOE responded to most of these comments in the Final Mission Plan Amendment provided to Congress on June 9, 1987. Since enactment of the NWPAA in December 1987, DOE-NRC interactions have focused on the Yucca Mountain site. In January 1988, DOE issued the Consultation Draft Site Characterization Plan (CDSCP) for the Yucca Mountain site. The NRC staff provided comments in the form of draft and final "point papers" on the CDSCP. The NRC comments included several objections related to: (1) the failure to recognize the range of alternative conceptual models of the Yucca Mountain site; (2) the status of the quality assurance (QA) plans for site characterization activities; and (3) concerns related to the exploratory shaft facility. Although the December 1988 SCP shows improvement over the CDSCP, NRC continues to have an objection involving the need for implementing a baselined QA program before beginning site characterization and an objection involving the need for DOE to demonstrate the adequacy of both the ESF design and the design control process. Prior to the November 1989 Reassessment Report, DOE had committed to having a qualified QA program in place before sinking the exploratory shaft at the Yucca Mountain site. This commitment has not changed. However, in view of the extension in the schedule for shaft sinking from November 1989 to November 1992, qualified QA plans are needed in the near term for meeting the January 1991 schedule for surface-based testing. In addition to having a qualified QA program in place, DOE must also have issued the pertinent study plans for site characterization activities they wish to begin. DOE has taken measures to clarify and institutionalize the roles of other Federal agencies in addition to NRC. In the Draft 1988 Mission Plan Amendment, DOE described interactions with these agencies. DOE has a Memorandum of Understanding (MOU) with the Mine Safety and Health Administration of the Department of Labor for technical support and oversight for shaft construction and other site characterization activities, and with the Department of Transportation to define the respective responsibilities of the two agencies in the waste disposal program. DOE also has interagency agreements with the Bureau of Mines and the U.S. Geological Survey of the Department of the Interior. DOE's efforts to address the concerns of States, local governments, and Indian tribes have met with mixed results. For example, DOE has not succeeded in finalizing any consultation and cooperation (C&C) agreements as required under section 117(c) of the NWPA, as amended. These agreements were to help resolve State and Tribal concerns about public health and safety, environmental, and economic impacts of a repository. Publication of the Siting Guidelines under section 112(a) of the NWPA resulted in numerous lawsuits challenging the validity of the Guidelines. Similarly, the FEAs were challenged in the Ninth Circuit by affected States and tribes. The NWPAA did not curtail financial assistance to affected States and tribes, except to redefine and redistribute it if DOE and a State or tribe enter into a

benefits agreement.

The State of Nevada and affected local governments are eligible to receive financial assistance. DOE has attempted to negotiate an agreement with the State of Nevada for monetary benefits under Section 170 of the NWPAA. This Section would provide for payments of $10 million per year before receipt of spent fuel, and $20 million per year after receipt of spent fuel until closure of the repository. These payments would be in addition to certain monetary benefits for which the State is eligible under the NWPA, as amended. Also under a benefits agreement, a Review Panel would be constituted for the purpose of advising DOE on matters related to the repository, and for assisting in the presentation of State, tribal, and local perspectives to DOE. The beneficiary to a benefits agreement must waive its right to disapprove the recommendation of the site for a repository and its rights to certain impact assistance under Sections 116 and 118 of the NWPA, as amended. To date, the State of Nevada has declined DOE's offer to negotiate a benefits agreement. In 1989, the State of Nevada requested $23 million for work on Yucca Mountain. Congress appropriated $5 million and authorized DOE to release an additional $6 million at the discretion of the Secretary on the basis of good faith efforts of the State to 51-SC-58 allow technical investigations to begin at the site. The NWPAA introduced several new organizational entities to the repository program with responsibilities that may contribute to resolving concerns of Federal, State, and local governments involved in the program. Under section 503 of the NWPAA, the Nuclear Waste Technical Review Board (NWTRB) is to evaluate the technical and scientific validity of DOE activities under the NWPAA, including site characterization and activities related to packaging or transportation of spent fuel. The NWPAA also established the Office of Nuclear Waste Negotiator, who is to seek to negotiate terms under which a State or Indian tribe would be willing to host a repository or MRS facility at a technically qualified site. Among the duties of the Negotiator is consultation with Federal agencies such as NRC on the suitability of any potential site for site characteriza lion. Secretary of Energy James Watkins has emphasized the importance of the Negotiator to the success of the program. A Negotiator could contribute to the timely success of the repository program by providing an alternative site to the Yucca Mountain site that would still have to be technically acceptable, but that would enjoy the advantage of reduced institutional uncertainties resulting from opposition of State or affected Indian tribes. The President nominated and the Senate recently confirmed David Leroy to be the Negotiator. An additional measure which may facilitate documentation and communication of concerns related to a repository is the Licensing Support System (LSSJ. The LSS is to provide full text search capability of and easy access to documents related to the licensing of the repository. Although the primary purpose of the LSS is to expedite NRC's review of the construction authorization application for a repository, it will be an effective mechanism by which all LSS participants, including the State and local governments, can acquire ear:ly access to documents relevant to a repository licensing decision. DOE is responsible for the design, development, procurement and testing of the LSS. LSS design and development must be consistent with objectives and requirements of the Commission's LSS rulemaking and must be carried out in consultation with the LSS Administrator and with the advice of the Licensing Support System Advisory Review Panel. NRC (LSS Administrator) is responsible for the management and operation of the PART 51 .STATEMENTS OF CONSIDERATION LSS after completion of the DOE design and development process. Procedures for the use of the LSS are part of revisions to 10 CFR part 2, NRC's Rules of Practice for the adjudicatory proceeding on the application to receive and possess waste at a repository. These revisions were the result of a "negotiated rulemaking" process in which affected parties meet to reach consensus on the proposed rule. The members of the negotiating committee included: DOE; NRG; State of Nevada; coalition of Nevada local governments; coalition of industry groups; and a coalition of national environmental groups. The coalition of industry groups dissented on the final text of the proposed rule, but the negotiating process enabled NRG to produce a proposed rule reflecting the consensus of most of the interested parties on an important repository licensing issue. NRG is committed to safe disposal of radioactive waste and the protection of public health and safety and the environment. Any State with a candidate site for a repository should be assured that a repository will not be licensed if it does not meet NRG criteria. NRG has its own program for interaction with the State of Nevada and affected units of local government, and will continue to provide information to Nevada and consider State concerns as requested. Given the difficult nature of siting a repository, the Commission believes that the NWPA, as amended, has achieved the proper balance between providing for participation by affected parties and providing for the exercise of Congressional authority to carry out the national program for waste disposal. The NWPAA provides adequate opportunity for interaction between DOE and other Federal agencies, States, tribes, and local governments such that concerns can be presented to DOE for appropriate action. Both the NRG and the Sta le or tribe can exercise considerable prerogative regarding repository development. The State or tribe may disapprove the recommendation that the site undergo repository development. This disapproval can be overridden only by vote of both houses of Congress within 90 days of continuous session. If the State disapproval is overridden, DOE may submit an application for authorization to construct the repository, and, if approved, a subsequent application to receive and possess waste for emplacement. NRG will make decisions on the license applications according to the requirements of its statutory mission. Despite the complexity of the overall process and the strong views of the participants in it, the Commission sees no compelling reason to conclude that current institutional arrangements are inadequate to the ta.sk of resolving State, Federal, and local concerns in time to permit a repository to be available within the first quarter of the twenty-first century. II.A.3.b. Continuity of the management of the waste program At the time the Commission issued its 1984 Waste Confidence Decision, the possibility that DOE functions would be transferred to another Federal agency was cited as the basis for concerns that the resolution of the radioactive waste disposal problem would likely undergo further delays. The Commission responded that in the years since the Administration had proposed to dismantle DOE in September 1981, Congress had not acted on the proposal. The Commission further stated that even if DOE were abolished, the nuclear waste program would simply be transferred to another agency. The Commission did not view the potential transfer in program management as resulting in a significant loss of momentum in the waste program. The Commission also concluded that the enactment of the NWPA, which gave DOE lead responsibility for repository development, further reduced uncertainties as to the continuity of management of the waste program. Section 303 of the NWPA did, however, require the Secretary of Energy to " ... undertake a study with respect to alternative approaches to managing the construction and operation of all civilian radioactive waste facilities, including the feasibility of establishing a private corporation for such purpose." To carry out this requirement, DOE established the Advisory Panel on Alternative Means of Financing and Managing Radioactive Waste Facilities, which came to be known as the "AMFM" Panel. The Panel's final report, issued in December 1984, concluded that several organizational forms are more suited than DOE for managing the waste program, including an independent Federal agency or commission, a public corporation, and a private corporation.

The report identified a public corporation as the preferred alternative on the basis of criteria developed by the Panel for an acceptable waste management organization.

In particular, the report indicated that a public corporation would be stable, highly mission-oriented, able to maintain credibility with stakeholders, and more 51-SC-59 responsive to regulatory control than a Federal executive agency. Commenting on the AMFM Panel's report in April 1985, DOE recommended retaining the present management structure of the waste program at least through the siting and licensing phase of the program. Congress did not take action to implement the Panel's recommendations, and DOE's management of the waste program has remained uninterrupted. By enacting the NWPAA, Congress effectively reaffirmed DOE's continued management of the waste program. Congress did not revise DOE's role as the lead agency responsible for development of a repository and an MRS. Congress did establish several new entities for the purpose of advising DOE on matters related to the waste program, such as the NWTRB and the Review Panel, to be established if DOE and a State or tribe enter into a benefits agreement under Section 170 of the NWPAA. Congress provided further indication of its intent that DOE maintain management control of the waste program for the foreseeable future in requiring, under Section 161, that the Secretary of DOE " ... report to the President and to Congress on or after January 1, 2007, but not later than January 1, 2010, on the need for a second repository." This is not to say, however, that there have been no management problems in the DOE program. Since the enactment of the NWPA in 1983, only one of the five Directors of DOE's Office of Civilian Radioactive Waste Management (OCRWM] has held the position on a permanent basis. Inadequate progress toward an operating repository has concerned several Congressional observers, including Senator J. Bennett Johnston, Chairman of the Senate Energy and Natural Resources Committee. In February 1989 confirmation hearings for then-Secretary-of-Energy-designate James Watkins, Senator Johnston strongly criticized mounting cost projections and lack of progress in the program, and called for new and stronger management. In the November 1989 Reassessment Report, DOE discussed several new initiatives for improving its management of the repository program. The initiatives include "direct-line" reporting from the Yucca Mountain Project Office to the Office of Civilian Radioactive Waste Management (OCRWMJ, and an independent contractor review of OCRWM management structures, systems and procedures to identify program redundancies, gaps, and \' PART 51 STATEMENTS OF CONSIDERATION strengths. The OCRWM is also implementing improvements in the overall Program Management System, the QA program, and establishment of program cost and schedule baselines. Whether the management structure of the repository development program should in fact be changed is a decision best left to others. The Commission believes that a finding on the likely availability of a repository should take management problems into account, but finds no basis to diminish the degree of assurance in its 1984 conclusion on this issue. Events since the submission of the AMFM Panel report do not indicate that there will be a fundamental change in the continuity of the management structure of the program any time soon. In addition, it cannot be assumed that the program would encounter significantly less difficulty with a new management structure than it would continuing under the present one. Under either scenario, however, the Commission believes it would be more prudent to expect repository operations after the 2010 timeframe than before it. Neither the problems of a new management structure nor those of the existing one are likely to prevent the achievement of repository operations within the first quarter of the next century, however. II.A.3.c. Continued funding of the nuclear waste management program Section 302 of the NWPA authorized DOE to enter into contracts with generators of electricity from nuclear reactors for payment of 1.0 mill (0.1 cent) per kilowatt-hour of net electricity generated in exchange for a Federal Government commitment to take title to the spent fuel from those reactors. In the 1984 Waste Confidence Decision, the Commission noted that all such contracts with utilities had been executed. After the 1984 Decision, President Reagan decided that defense high-level wastes are to be collocated with civilian wastes from commercial nuclear power reactors. DOE's Office of Defense Programs is to pay the full cost of disposal of defense waste in the repository. DOE is required under Section 302(a)(4) of the NWPA, as amended, " ... annually [to] review the amount of the fees ... to evaluate whether collection of the fees will provide sufficient revenues to offset the costs .... " In the June 1987 Nuclear Waste Fund Fee Adequacy Report, DOE recommended that the 1.0 mill per kilowatt-hour fee remain unchanged. This assessment was based on the assumption that an MRS facility would open in 1998, the first repository would open in 2003, and the second repository in 2023. These assumptions do not reflect changes in the waste program brought about by the NWPAA enacted in December 1987. Two such changes with significant potential impacts were the suspension of site-specific activHies related to the second repository until at least 2007, and the linkage between MRS construction and operation and the granting of a repository construction authorization, which will probably occur no earlier than 1998. DOE has not issued a fee adequacy report since the June 1987 report. When the updated report is released, it is expected to reflect overall program cost savings to the utilities resulting from: (1) limiting site characterization activities to a single site at Yucca Mountain, NV; and (2) the DOE Office of Defense Programs' sharing other program costs with generators of electricity " ... on the basis of numbers of waste canisters handled, the portion of the repository used for civilian or defense wastes, and the use of various facilities at the repository," in addition to paying for activities solely for disposing of defense wastes. An additional factor which may eventually also contribute to the overall adequacy of Nuclear Waste Fund fees is the likelihood that a significant number of utilities will request renewals of reactor operating lifetimes beyond their current 01 expiration dates. 01 renewal would provide additional time during which Nuclear Waste Fund fees could be adjusted, if necessary, to cover any future increase in per-unit costs of waste management and disposal. It is expected that the new report may reflect a recent Court decision which found that fees paid into the Nuclear Waste Fund be adjusted to reflect transmission and distribution losses. The Commission recognizes the potential for program cost increases over estimates in the 1987 Nuclear Waste Fund Fee Adequacy Report. If there is a significant delay in repository construction, for example, it is reasonable to assume that construction costs will escalate. There may also be additional costs associated with reactor dry cask storage of spent fuel, if DOE does not have a facility available to begin accepting spent fuel by the 1998 date specified in the NWPA. These costs would be further increased if one or more licensee was to becqme insolvent and DOE was required to assume responsibility for storage at affected reactors before 1998. In the event of insolvency, DOE would still have sufficient funds to take over responsibility for managing spent fuel until a repository is available. Because spent fuel disposal costs are directly related to the amount of electricity 51-SC-60 generated, with contributions to the NWF based on a kilowatt-hour surcharge that must be paid in term installments, utilities can be presumed to be mostly up-to-date with their contributions. It is highly unlikely that a utility would jeopardize its contract for spent fuel disposal with DOE by defaulting on a periodic payment to save a few million dollars. Even if a utility were to default, it would not be much in arrears for its spent fuel before it would trigger close DOE scrutiny and mitigative action. Larger amounts in default could possibly occur with those relatively few utilities that have not paid their full share of pre-1983 collections. This issue arises because several utilities elected to defer payment for spent fuel generated prior to April 1983 into the fund and, instead, themselves hold the money that was collected from ratepayers for the one-time fee. DOE's Inspector General believes that some of those utilities may not be able to make their payments when due. The NRC understands from OCRWM staff that, if a nuclear utility licensee were to default on its one-time contribution to the NWF, DOE is not precluded from accepting for disposal all spent fuel from that utility. Thus, the NRC does not view this issue as affecting its confidence that the spent fuel will be disposed of. Rather, the issue is one of equity--that is, will a utility and its customers and investors or U.S. taxpayers and/or other utilities ultimately pay for disposal of spent fuel generated prior to April 1983. The Commission does not believe that a licensee's potential default has a direct bearing on the Commission's Waste Confidence Decision. The full impact of the program redirection resulting from the NWPAA and the outlook for the timing of repository availability will continue to be assessed annually. If it does appear that costs will exceed available funds, there is provision in the NWPA for DOE to request that Congress adjust the fee to ensure full-cost recovery. Thus, the Commission finds no reason for changing its basic conclusion that the long-term funding provisions of the Act should provide adequate financial support for the DOE program. II.A.3.d. DOE's schedule for repository development At the time that the 1984 Waste Confidence Decision was issued, the Nuclear Waste Policy Act of 1982, enacted in January 1983, had been in effect for less than 20 months. The NWPA had established numerous deadlines for various repository program milestones. Under section PART 51 STATEMENTS OF CONSIDERATION 112(b)(1)(B), the NWPA set the schedule for recommendation of sites for characterization no later than January 1, 1985. Section 114(a)(2) specified that no later than March 31, 1987, with provision for a 12-month extension of this deadline, the President was to recommend to Congress one of the three characterized sites qualified for an application for repository construction authorization. Under section 114(d), NRC was to issue its decision approving or disapproving the issuance of a construction authorization not later than January 1, 1989, or the expiration of three years after the date of submission of the application, whichever occurs later. Section 302(a)(5)(B) required that contracts between DOE and utilities for payments to the Waste Fund provide that DOE will begin disposing of spent fuel or high-level waste by January 31, 1998. In little more than a year after enactment, the schedule established by the NWPA began proving to be optimistic. In the reference schedule for the repository presented in the April 1984 Draft Mission Plan, for example, DOE showed a slip from January 1989 to August 1993 for the decision on construction authorization. In the 1984 Waste Confidence Decision, the Commission recognized the possibility of delay in repository availability beyond 1998, and did not define its task as finding confidence that a repository would be available by the 1998 milestone in the NWPA. The Commission focused instead on the question of whether a repository would be available by the years 2007-2009, the date cited in the court remand as the expiration of the OLs for the Vermont Yankee and Prairie Island reactors. The NRC believed that the NWPA increased the chances for repository availability within the first few years of the first century, by specifying the means for resolving the institutional and technical issues most likely to delay repository completion, by establishing the process for compliance with NEPA, and by setting requirements for Federal agencies to cooperate with DOE in meeting program milestones. Finding that no fundamental technical breakthroughs were necessary for the repository program, the Commission predicted that " ... selection and characterization of suitable sites and construction of repositories will be accomplished within the general time frame established by the Act [1998] or within a few years thereafter." In January 1987, DOE issued a Draft Mission Plan Amendment to apprise Congress of significant developments and proposed changes in the repository program. In the Draft Amendment, DOE announced a five-year delay in its schedule for repository availability from the first quarter of 1998 to the first quarter of 2003. DOE's reasons for the delay included the need for more time for consultation and interaction with States and Tribes, the requirement in DOE's 1987 budget that funds not be used for drilling exploratory shafts in 1987, and the need for more information than previously planned for site selection and the license application. The 1987 Draft Mission Plan Amendment set the second quarter of 1988 as the new date for exploratory shaft construction at the Yucca Mountain site. When the final 1987 Mission Plan Amendment was submitted to Congress in June 1987, the schedule for shaft sinking at the Yucca Mountain site had slipped six months to the fourth quarter of 1988. Congress did not take action to approve the June 1987 Mission Plan Amendment as DOE had requested. On December 22, 1987, the NWPAA was enacted. The NWPAA had its major impact on the repository program in suspending site characterization activities at the Hanford and Deaf Smith County sites and authorizing DOE to characterize the Yucca Mountain site for development of the first repository. DOE subsequently issued the Draft 1988 Mission Plan Amendment in June 1988, to apprise Congress of its plans for implementing the provisions of the NWPAA. In the Draft 1988 Mission Plan Amendment, DOE's schedule for shaft sinking at Yucca Mountain had slipped another six months to the second quarter of 1989. Since the NRC published the Proposed Waste Confidence Review (54 FR 39767) for comment, the schedule for shaft sinking has been changed from November 1989 to November 1992. Issues requiring DOE attention before site characterization can begin have been identified, and it is possible that additional issues affecting DOE's readiness will come to light. However, DOE has made progress in completing QA plans since September 1989, and it is reasonable to expect that study plans and technical procedures needed for surface-based testing will be ready in time for testing to begin by January 1991. Heretofore, the repository schedule has always been agressive and highly success-oriented. In comments on the Draft 1988 Mission Plan Amendment, the Commission noted that the schedule has not allowed adequately for contingencies, and that, given the compression in the schedule for near-51-SC-61 term program milestones, DOE had not shown how it would be able to meet the 2003 milestone for repository operation. The revised schedule announced in the November 1989 Reassessment Report includes a new reference schedule for the restructured repository, MRS, and transportation programs. Under the restructured program, the schedule for submittal of a construction authorization application to NRC has been extended from 1995 to 2001, and the schedule for repository operation at Yucca Mountain, if that site is found to be suitable, is 2010. DOE believes that this reference schedule is the first repository program schedule since passage of the NWPA that is based on a "realistic assessment of activity duration and past experience." The new schedule allows more time for scientific investigations than earlier schedules. NRC believes that the restructured program has been responsive to NRC concerns that the quality and completeness of site investigations were being compromised in order to satisfy unrealistic schedule requirements. Another potential source of delay in repository availability may arise from NRC regulations. Given the revised schedule, however, the NRC does not believe this is likely. The Commission believes that current NRC rules are fully adequate to permit DOE to proceed to develop and submit a repository license application, but further clarification of these rules is desirable to reduce the time needed to conduct the licensing proceeding itself. In order to meet the three-year schedule provided in the NWPA for a Commission decision on repository construction authorization, the NRC staff has undertaken to refine its regulatory framework on a schedule that would permit DOE to prepare and submit an application for repository construction authorization under its current schedule. The Commission fully intends to avoid delaying DOE's program, while working to reduce the uncertainties in NRC regulatory requirements that could become contentions in the licensing proceeding. Even if there are any delays resulting from a need for DOE to accommodate more specific regulatory requirements in its site characterization or waste package development programs, the Commission is confident that the time savings in the licensing proceeding will more than compensate for them. In view of the delays in exploratory shaft excavation since the 2003 date for repository availability was set, the Commission believed it was optimistic to expect that Phase 1 of repository operations would be able to begin by PART 51 STATEMENTS OF CONSIDERATION 2003. As DOE's schedule for repository availability has slipped a year and a half since the date was changed from 1998 to 2003, the earliest date for repository availability would probably be closer to 2005. Given additional delays in shaft sinking and DOE's revised program schedule, NRC believes that 2010 is the earliest date for repository availability at Yucca Mountain. Yet, the Commission recognizes that DOE is committed to improving the schedule where possible without sacrificing quality and completeness of scientific investigations. An institutional issue that may further affect DOE's schedule is the status of EPA standards for disposal of spent fuel and high-level waste. These standards are required under section 121(a) of the NWPA. Under 10 CFR section 60.112, NRC's overall postclosure system performance objective, the geologic setting shall be selected and the engineered barrier system, which includes the waste package, must be designed to assure that releases of radioactive materials to the accessible environment, following permanent closure, conform to EPA's standards. 40 CFR part 191, the EPA standards, first became effective in November 1985. In July 1987, the U.S. Court of Appeals for the First Circuit vacated and remanded to EPA for further proceedings subpart B of the high-level radioactive waste disposal standards. As noted under the aforementioned I.A.1., the standards have not been reissued. A significant modification in the reissued EPA standard may affect the schedule for completing the design of the waste package and engineered barrier to the extent that design testing is planned to demonstrate compliance with the standards. DOE's current site characterization plans for demonstrating compliance with 40 CFR part 191 are based on the standards as promulgated in 1985. DOE is proceeding to carry out its testing program developed for the original EPA standards. DOE has stated that if the EPA standards are changed significantly when they are reissued, DOE will reevaluate the adequacy of its testing program. The Commission believes that DOE's approach is reasonable. Much of the information required to demonstrate compliance with the EPA standards is expected to remain the same regardless of the numerical level at which each standard is set. Considering the importance of developing the repository for waste disposal as early as safely practicable, it would be inappropriate for DOE to suspend work on development of engineered barriers pending reissuance of the standards, unless EPA had given clear indications of major changes in, them. Another possibility is that, regardless of any changes in the repromulgated EPA standards, they will be litigated in Federal court. Even if this proves to be the case, however, the Commission believes that any such litigation will still permit EPA to promulgate final standards well within the time needed to enal;:ile DOE to begin repository operations at any site within the first quarter of the twenty-first century. Given the current DOE program schedule, and assuming that the QA program can be qualified and based testing begun within the next year, the Commission finds that although it is not impossible that a repository at Yucca Mountain will be available by 2007-2009, it is more likely that the earliest date for a repository there is 2010. If DOE determines Iha t the Yucca Mountain site is unsuitable, and if DOE makes this determination by the year 2000, the NRC believes that a repository at another site could be available within the first quarter of the next century. The Commission will reevaluate these dates during the next scheduled Waste Confidence Review in 1999. 11.B. Relevant Issues That Have Arisen since the Commission's Original Decision JI.B.1. NRG stated in 9-14-87 correspondence to Sen. Breaux on pending nuclear waste legislation that under a program of single site characterization, " ... there may be a greater potential for delay of ultimate operation of a repository than there is under the current regime where three sites will undergo at-depth characterization before a site is selected." To what extent does the NWPAA raise uncertainty about the identification of a technically acceptable site and potential delay in repository availability by limiting site characterization to a single candidate site (Yucca Mt.) and by raising the possibility that a negotiated agreement might influence repository site selection? Does this uncertainty affect confidence in the availability of a repository by 2007-2009? In providing comments to Congress on proposed amendments to the NWPA, NRC took the position that simultaneous site characterization of three sites, as required by the NWPA, was not 51-SC-62 necessary to protect public health and safety. NRC further stated that the adequacy of a site for construction authorization would ultimately be determined in a licensing proceeding, and that NRC would only license a site that satisfied NRC licensing requirements. As described next, the Commission believes that the NWPAA contains numerous provisions to ensure that a technically acceptable site will be identified. The NWPAA does not reduce the scope of site characterization activities that DOE is authorized to undertake. The Amendments Act establishes a Nuclear Waste Technical Review Board composed of individuals recommended by the National Academy of Sciences and appointed by the President to evaluate the scientific validity of DOE activities, including site characterization activities, and to report its findings at least semiannually to Congress and DOE. The Amendments Act also provides funding for technical assistance to States, tribes, and affected units of local government. Finally, section 160(1) of the NWPAA provides that "Nothing in this Act shall be construed to amend or otherwise detract from the licensing requirements of the NRC established in Title II of the Energy Reorganization Act of 1974 (42 U.S.C. 5841 et seq.)." In providing for these reviews and in reaffirming NRC's licensing authority, the NWPAA ensures that a candidate site for a repository must satisfy all NRC requirements and criteria for disposal of high-level radioactive wastes in licensed geologic repositories. Section 402 of the NWPAA establishes the Office of the Nuclear Waste Negotiator. The duty of the Negotiator is to attempt to find a State or tribe willing to host a repository or MRS at a technically qualified site. The Negotiator may solicit comments from NRC, or any other Federal agency, on the suitability of any potential site for site characterization. Section 403(d)[4) strengthens the Commission's confidence that a technically acceptable site will be identified by providing that DOE may construct a repository at a negotiated site only if authorized by NRC. Given these safeguards on selection of a technically acceptable site, the Commission does not consider that the possibility of a negotiated agreement reduces the likelihood of finding a technically qualified site. The Commission raised the concern as early as April 1987 that under a program of single-site characterization, there could be considerable delay while PART 51 STATEMENTS OF CONSIDERATION characterization was completed at another site or slate of sites if the initially chosen site was found inadequate. By terminating site characterization activities at alternative sites to the Yucca Mountain site, the NWPAA has had the effect of increasing the potential for delay in repository availability if the Yucca Mountain site proves unsuitable. The provision in the NWPAA for a Negotiator could reduce the uncertainty and associated delay in restarting the repository program by offering an alternate to the Yucca Mountain site; but at the time of this writing, a Negotiator has not been appointed. It should be noted here that the repository program redirection under the NWPAA does not, per se, have a significant impact on the Commission's assurance of repository availability by 2007-2009, the relevant dates in the original Waste Confidence Proceeding, or on availability by 2010, DOE's current date. The Commission's reservations about affirming this timeframe derive from other considerations, including delays in sinking shafts and the potential for other delays in meeting program milestones, that would have arisen without the NWPAA. The Amendments Act does, however, effectively make it necessary that Yucca Mountain be found suitable if the 2007-2009 or 2010 timeframe is to be met; this target period would almost certainly be unachievable if DOE had to begin screening to characterize and license another site. Thus, confidence in repository availability in this period would imply confidence in the suitability of Yucca Mountain. The Commission does not want its findings here to constrain in any way its regulatory discretion in a licensing proceeding. Therefore, the Commission declines to reaffirm the 2007-2009 timeframe in the original decision or to affirm the current 2010 date for repository operation. /1.B.2. In the Draft 1988 Mission Plan Amendment, DOE stated that " ... the data indicate that the Yucca Mountain site has the potential capacity to accept at least 70,000 MTHM [metric tons heavy metal equivalent] of waste, but only after site characterization will it be possible to determine the total quantity of waste that could be accommodated at this site." a. Do the issues of limited spent fuel capacity at Yucca Mountain, indefinite suspension of the second repository program, and the likelihood that no more than one repository will be available by 2007-2009 undermine the NRC's 1984 assurance that "sufficient repository capacity.will be available within 30 years beyond expiration of any reactor operating license to dispose of existing commercial high level radioactive waste and spent fuel originating in such reactor and generated up to that time?" b. Is there sufficient uncertainty in total spent fuel projections (e.g., from extension-of-life license amendments, renewal of operating licenses for an additional 20 to 30 years, or a new generation of reactor designs] that this Waste Confidence review should consider the institutional uncertainties arising from having to restart a second repository program? 11.B.2.a. Although it will not be possible to determine whether Yucca Mountain can accommodate 70,000 MTHM or more of spent fuel until after site characterization, the Commission does not believe that the question of repository capacity at the Yucca Mountain site should be a major factor in the analysis of Finding 2. This is because it cannot be assumed that Yucca Mountain will ultimately undergo development as a repository. The generic issue of repository capacity does add to the potential need for more than one repository, however. As noted earlier, the NWPA established deadlines for major milestones in the development of the first and the second repository programs. The Act also required NRC to issue a final decision on the construction authorization application by January 1, 1989 for the first repository, and January 1, 1992 for the second [or within three years of the date of submission of the applications, whichever occurred later). The July 1984 Draft DOE Mission Plan set January 1998 and October 2004 as the dates for commencement of waste emplacement in the first and second repositories, assuming that Congressional authorization was obtained to construct the second repository. Thus, at the time the 1984 Waste Confidence Decision was issued, DOE was authorized and directed to carry out two repository programs under a schedule to make both facilities operational by 2007-2009. DOE and NRC were also working under the constraint, still in force under the NWPA as amended, that no more than 70,000 MTHM may be emplaced in the first repository before the second is in '51-SC-63 operation. Because DOE estimated at the time that commercial U.S. nuclear power plants with operating licenses or construction permits would discharge a total 160,000 MTHM of spent fuel. it appeared that at least two repositories would be needed. In the 1984 Waste Confidence Decision, reactors were assumed to have a 40-year operating lifetime, and because the earliest licenses were issued in 1959 and the early 1960's, the oldest plants' licenses were due to expire as early as 1999 and 2000, as discussed in more detail below. Although it was expected that at least one repository would be available by this time, there was also a limit as to how quickly spent fuel could be accepted by the repository. DOE had estimated that waste acceptance rates of 3400 MTHM per year could be achieved after the completion of Phase 2 of the first repository. This rate could essentially double if two repositories were in operation. At 6000 MTHM/year, it was estimated that all the anticipated spent fuel could be emplaced in the two repositories by about the year 2026. This was the basis for the Commission's position that sufficient repository capacity would be available within 30 years beyond expiration of any reactor 01 to dispose of existing commercial high level waste and spent fuel originating in such reactor and generated up to that time. In May 1986, however, DOE announced an indefinite postponement of the second repository program. The reasons for the postponement included decreasing forecasts of spent fuel discharges, as well as estimates that a second repository would not be needed as soon as originally supposed. With enactment of the NWPAA in December 1987, DOE was required to terminate all site-specific activities with respect to a second repository unless such activities were specifically authorized and funded by Congress. The NWPAA reqv.ired DOE to report to Congress on the need for a second repository on or after January 1, 2007, but not later than January 1, 2010. Current DOE spent fuel projections, based on the assumption of no new reactor orders, call for 87,000 MTHM to have been generated by the year 2036, including approximately 9000 MTHM of defense high-level waste. With the likelihood that there will be reactor lifetime extensions and renewals, however, the no-new-orders case probably underestimates total spent fuel discharges. Also, the NWPAA did not change the requirement that no more PART 51 STATEMENTS OF CONSIDERATION than 70,000 MTHM could be emplaced in the first repository before operation of the second. It therefore appears likely that two repositories will be needed to dispose of all the spent fuel and level waste from the current generation of reactors, unless Congress provides statutory relief from the 70,000 MTHM limit, and the first site has adequate capacity to hold all of the spent fuel and high-level waste generated. The Commission believes that if the need for an additional repository is established, Congress will provide the needed institutional support and funding, as it has for the first repository. For all but a few licensed nuclear power reactors, OLs will not expire until some time in the first three decades of the twenty-first century. Several utilities are currently planning to have their OLs renewed for ten to 30 years beyond the original license expiration. At these reactors, currently available spent fuel storage alternatives effectively remove storage capacity as a potential restriction for safe operations. For these reasons, a repository is not needed by 2007-2009 to provide disposal capacity within 30 years beyond expiration of most OLs. If work is begun on the second repository program in 2010, the repository could be available by 2035, according to DOE's estimate of 25 years for the time it will take to carry out a program for the second repository. Two repositories available in approximately 2025 and 2035, each with acceptance rates of 3400 MTHM/year within several years after commencement of operations, would provide assurance that sufficient repository capacity will be available within 30 years of OL expiration for reactors to dispose of the spent fuel generated at their sites up to that time. There are several reactors, however, whose OLs have already expired or are due to expire within the next few years, and which are now licensed or will be licensed only to possess their spent fuel. If a repository is not available until about 2025, these reactors may be exceptions to the second part of the Commission's 1984 Finding 2, which was that sufficient repository capacity will be available within 30 years beyond the expiration of any reactor OL to dispose of the commercial high-level waste and spent fuel originating in such reactor and generated up to that time. The basis for this second part of Finding 2 has two components: (1) a technical or hardware component; and (2) an institutional component. The technical component relates to the reliability of storage hardware and engineered structures to provide for the safe storage of spent fuel. An example would be the ability of spent fuel assemblies to withstand corrosion within spent fuel storage pools, or the ability of concrete structures to maintain their integrity over lpng periods. In the 1984 Decision, the Commission found confidence that available technology could in effect provide for safe storage of spent fuel for at least 70 years. The Commission's use of the expression "30 years beyond expiration of any reactor operating license" in the 1984 Finding was based on the understanding that the license expiration date referred to the scheduled expiration date at the time the license was issued. It was also based on the understanding that, in order to refuel the reactor, some spent fuel would be discharged from the reactor within twelve to eighteen months after the start of full power operation. Thus, the Commission understood that, depending on the date of the first reactor outage for refueling, some spent fuel would be stored at the reactor site for most of the 40-year term of the typical OL. In finding that spent fuel could be safely stored at any reactor site for at least 30 years after expiration of the OL for that reactor, the Commission indicated its expectation that the total duration of spent fuel storage at any reactor would be about 70 years. Taking the earliest licensed power reactor, the Dresden 1 facility licensed in 1959, and adding the full 40-year operating license duration for a scheduled license expiration in the year 1999, the Commission's finding would therefore entail removal of all spent fuel from that reactor to a repository within the succeeding 30 years, or by 2029. Even if a repository were not available until the end of the first quarter of the twenty-first century, DOE would have at least four years to ship the reactor's 683 spent fuel assemblies, totalling 70 metric tons initial heavy metal (MTIHMJ, from Dresden 1 without exceeding the Commission's 30-year estimate of the maximum time it would take to dispose of the spent fuel generated in that reactor up to the time its OL expired. (MTIHM is a measure of the mass of the uranium in the fuel (or uranium and plutonium if it is a mixed oxide fuel) at the time the fuel is placed in the reactor for irradiation.) Considering the experience from the 1984 and 1985 campaigns to return spent fuel from the defunct West Valley reprocessing facility, to the reactors of origin, 70 metric tons of BWR spent fuel can easily be shipped within four years. The first campaign, involving truck 51-SC-64 shipments of 20 metric tons from West Valley, NY, to Dresden 1 in Morris, IL, took eleven months. The second, involving truck shipments of 43 tons from West Valley to the Oyster Creek reactor in Toms River, NJ, took six months. (See Case Histories of West Valley Spent Fuel Shipments, Final Report, NUREG/CR-4847 WPR-86(6811)- 1, p. 2-2.) This estimate assumes, moreover, that no new transportation casks, designed to ship larger quantities of older, cooler spent fuel, for example, would be available by 2025. The institutional part of the question concerning the availability of sufficient repository capacity required the Commission to make a finding as to whether spent fuel in at-reactor storage would be safely maintained after the expiration of the facility OL. This question related to the financial and managerial capability for continued safe storage and monitoring of spent fuel, rather than to the capability of the hardware involved. The Commission determined, in Finding 3 of its 1984 Decision, that spent fuel will be managed in a safe manner until sufficient repository capacity is available to assure safe disposal, which was expected under Finding 2 to be about 30 years after the expiration of any reactor OL. (See discussion of Finding 3 below for additional discussion of the institutional aspects of spent fuel storage pending the availability of sufficient disposal capacity.) The availability of a repository within the first quarter of the twenty-first century holds no significant adverse implications for the Commission's institutional concern that there be an organization with adequate will and wherewithal to provide continued term storage after reactor operation. This could be a concern if a significant number of reactors with significant quantities of spent fuel onsite were to discontinue operations indefinitely between now and 1995, and the owners of these reactors did not appear to have the resources to manage them safely for up to 30 years pending the assumed availability of a repository in 2025. No such development is likely. No licenses for currently operating commercial nuclear reactors are scheduled to expire until the year 2000, and most such licenses will expire during the first two decades after 2006. (See Nuclear Regulatory Commission 1989 Information Digest, NUREG-1350, Vol. 1, p. 33.) The availability of the first repository by 2025, and of a second repository within one or two decades PART 51 STATEMENTS OF CONSIDERATION thereafter, would provide adequate disposal capacity for timely removal of the spent fuel generated at these reactors. There are several licensees, however, whose authority to operate their commercial reactors has already been terminated. These are Indian Point 1, Dresden 1, Humboldt Bay, and Lacrosse. They are also the only licensed power reactors that are retired with spent fuel being stored onsite. Assuming conservatively that a repository does not become operational until 2025, it appears likely that spent fuel will remain at these sites for more than 30 years beyond the time their reactors were indefinitely shut down, at which point their operating licenses could be considered to have effectively expired, although they will continue to hold a possession license for the storage of the spent fuel. In considering the means and motivation of the owner of an indefinitely retired reactor to provide safe long-term storage, the Commission believes it is useful to distinguish between the owner with only one reactor, and the owner of a reactor at a multi-unit site or an owner with operating reactors at other sites. In the case of a retired reactor at a multi-unit site, the owner would have a clear need to maintain the safety of storage at the r_etired reactor sufficiently to permit continued generation at the site. If the owner of the retired reactor also owned other reactors at other sites, the spent fuel at the retired reactor could be transferred, if necessary, to the storage facilities of other units still under active management. Of the four reactors just cited, Indian Point 1 and Dresden 1 fit this description, and the sibling reactors at their sites are operating under licenses that do not expire until well beyond the year 2000--that is, well within the post-OL period during which the Commission has found that spent fuel could be safely stored pending the availability of a repository. For the Lacrosse and Humboldt Bay reactors, the Commission is confident that, even if a repository is not available within 30 years following their retirement, the overall safety and environmental acceptability of extended spent fuel storage will also be maintained for these exceptional cases. Because there will still be an NRC possession license for the spent fuel at these facilities, the Commission will retain ample regulatory authority to require any measures, such as removal of the spent fuel remaining in storage pools to passive dry storage casks, that might become necessary until the time that DOE assumes title to the spent fuel under contracts pursuant to the NWPA. It should also be borne in mind that Humboldt Bay and Lacrosse are both small early reactors, and their combined spent fuel inventory totals 67 metric tons of initial heavy metal. (See Spent Fuel Storage Requirements (DOE/RL 88-34) October 1988, Table A.3b., pp. A.15-A.17.) If for any reason not now foreseen, this spent fuel can no longer be managed by the owners of these reactors, and DOE must assume responsibility for its management earlier than currently planned, this quantity of spent fuel is well within the capability of DOE to manage onsite or offsite with available technology. Nor does the Commission see a significant safety or environmental problem with premature retirements of additional reactors. In the Commission's original Waste Confidence Decision, it found reasonable assurance that spent fuel would have to spend no more than 30 years in post-operational storage pending the availability of a repository. For a repository conservatively assumed to be available in 2025, this expected 30-year maximum storage duration remains valid for most reactors, and would be true for all reactors that were prematurely retired after 1995. Based on the past history of premature shutdowns, the Commission has reason to believe that their likely incidence during the next six years will be small as a proportion of total reactor-years of operation. Historically, 14 of the 125 power reactors that have operated in the U.S. over the past 30 years have been retired before the expiration of their operating licenses. These early retirements included many low-power developmental reactors, which may make the ratio of 14 to 125 disproportionately high as a basis for projecting future premature shutdowns. The Commission is aware of currently operating reactors that may be retired before the expiration of their OLs, including: the recently-licensed Shoreham reactor, which has generated very little spent fuel; the Fort St. Vrain high-temperature gas-cooled reactor, which its owner plans to decommission; and the Rancho Seco reactor, which has operated.for the past 12 years and may or may not be retired. Assuming that these and perhaps a few more reactors do retire in the next several years, their total spent fuel storage requirements would not impose an unacceptable safety or environmental problem, even in the unlikely event that all these reactors' owners were rendered financially or otherwise unable to 51-SC-65 provide adequate care, and DOE were required to assume custody earlier than currently envisioned under the NWPA. Licensed non-power research reactors provide an even more manageable case. DOE owns the fuel for almost all of these reactors, many of which have been designed with lifetime cores that do not require periodic refueling. For those reactors that do discharge spent fuel, DOE accepts it for storage or reprocessing, and not more than an estimated 50 kilograms of such spent fuel are generated annually. Thus, given these worst-case projections, which are not expectations but bounding estimates, the Commission finds that a delay in repository availability to 2025 will not result in significant safety or environmental impacts due to extended operational spent fuel storage. To put it another way, the Commission is confident that, even if a repository were not available within 30 years after the effective expiration of the OLs for both currently retired reactors and potential future reactor retirements through 1995, the overall safety and environmental impacts of extended spent fuel storage would be insignificant. II.B.2.b. Although it is clear that there is uncertainty in projections of total future spent fuel discharges, it is not clear that the institutional uncertainties arising from having to restart a second repository program should be considered in detail in the current Waste Confidence Decision review. License renewals would have the effect of increasing requirements for spent fuel storage. The Commission understands that some utilities are currently planning to seek renewals for 30 years. Assuming for the sake of establishing a conservative upper bound that the Commission does grant 30-year license renewals, the total operating life of some reactors would be 70 years, so that the spent fuel initially generated in them would have to be stored for about 100 years if a repository were not available until 30 years after the expiration of their last OLs. Even under the conservative bounding assumption of 30-year license renewals for all reactors, however, if a repository were available within the first quarter of the twenty-first century, the oldest spent fuel could be shipped off the sites of all currently operating reactors well before the spent fuel initially generated in them reached the age of 100 years. Thus, a second repository, or additional capacity at the first, would be needed only to accommodate the additional quantity of spent fuel generated during the later years of these reactors' PART 51 STATEMENTS OF CONSIDERATION operating lives. The availability of a second repository would permit spent fuel to be shipped offsite well within 30 years after expiration of these reactors' OLs. The same would be true of the spent fuel discharged from any new generation of reactor designs. In sum, although some uncertainty in total spent fuel projections does arise from such developments as utilities' planning renewal of OLs for an additional 20 to 30 years, the Commission believes that this Waste Confidence review need not at this time consider the institutional uncertainties arising from having to restart a second repository program. Even if work on the second repository program is not begun until 2010 as contemplated under current law, there is sufficient assurance that a second repository will be available in a timeframe that would not constrain the removal of spent fuel from any reactor within 30 years of its licensed life for operation. IlB.3. Are early slippages in the DOE repository program milestones significant enough to affect the Commission's confidence that a repository will be available when needed for health and safety reasons? The 2007-2009 timeframe imposed on the Commission by the May 23, 1979 remand by the Court of Appeals was based on the scheduled expiration of the OLs for the Vermont Yankee and Prairie Island nuclear reactors. The specific issues remanded to the Commission were: (1) whether there is reasonable assurance that an offsite storage solution will be available by the years 2007-2009 (the expiration of the plants' operating licenses); and, if not, (2) whether there is reasonable assurance that the fuel can be stored safely at the sites beyond those dates. There was no finding by the Court that public health and safety required offsite storage or disposal by 2007-2009. In directing the Commission to address the safety of at-reactor storage beyond 2007-2009, the Court recognized the possibility that an offsite storage or disposal facility might not be available by then. The Commission has not identified a date by which a repository must be available for health and safety reasons. Taking into account institutional requirements for spent fuel storage, the Commission found, under Finding 3 in the 1984 Waste Confidence Decision, that spent fuel would be safely managed until sufficient repository capacity is available. The Commission also found, however, that in effect, under the second part of Finding 2, safe management would not need to continue for more than 30 years beyond expiration of any reactor's OL, because sufficient repository capacity was expected to become available within those 30 years. Considering that spent fuel would not have to be stored mpre than 30 years after any reactor's 40-year OL expiration, and taking into account the technical requirements for such storage, the Commission went on to determine under Finding 4 that, in effect, spent fuel could be safely stored for at least 70 years after discharge from a reactor. Thus, the Commission's 1984 Decision did not establish a time when sufficient repository capacity would be required; it established a minimum period during which storage would continue to be safe and environmentally acceptable pending the expected availability of sufficient repository capacity. Bearing in mind that reactor facilities were originally designed and OLs issued for a licensed life for opera lion of 40 years, the Commission is proposing elsewhere in this Federal Register notice a clarifying revision of Finding 4 to say that spent fuel can be safely stored at a reactor for at least 30 years after the "licensed life for operation" of that reactor. Implicitly, the proposed use of the phrase "licensed life for operation" clarifies that the Commission found in 1984 that NRC licensing requirements for reactor facility design, construction, and operation provide reasonable assurance that spent fuel can be stored safely and without significant environmental impacts for at least the first 40 years of the reactor's life. The Commission's proposed finding also implies that, barring any significant and pertinent unexpected developments, neither technical nor institutional constraints would adversely affect this assurance for at least another 30 years after that first 40 years. Another implication of this revised finding is that, where a utility is able to meet NRC requirements to extend that reactor's operating lifetime by license renewal, spent fuel storage for at least 30 years beyond the end of the period of extended life will also be safe and without significant environmental impacts. In assessing the effect of early slippages in DOE repository program milestones, therefore, the most important consideration is not the earliest date that an operating license actually expired, but the earliest date that an OL was issued. The earliest OL to be issued was for Dresden 1 in 1959, followed by a number of reactors licensed for operation in 1962. The OLs for all of the 111 power reactors now licensed to operate are currently scheduled to expire sometime within the 51-SC-66 first three decades of the twenty-first century, which is also the period in which their currently licensed life for operation would end. (See Nuclear Regulatory Commission 1989 Information Digest, NUREG-1350, Vol. 1, p. 33.) Thus, conservatively assuming here that there will be no license renewals, the earliest timeframe when a repository might be needed to dispose of spent fuel from the majority of reactors is 2029-2050. As proposed in the first part of Finding 2, the Commission has reasonable assurance that a repository will be available within the first quarter of the twenty-first century. Even if a repository were not available until 2025; this would be several years before the beginning of the earliest timeframe within which, based on an assumed 30-year storage after an assumed 40-year licensed life of reactor opera lion, a repository might be needed for spent fuel disposal. Thus, early slippages in DOE's program milestones do not affect the Commission's confidence that a repository will be available within that timeframe. IJ.B.4. NRG has stated that the 3-to 4-year license application review schedule is optimistic, and that for NRG to meet this schedule, DOE must submit a complete and high-quality license application. In the September 16, 1988 NRG comments to DOE on the Draft 1988 Mission Plan Amendment, the Commission requested that DOE acknowledge its commitment to develop this complete and high-quality application, "even 1f this would result in longer times to collect the necessary information and subsequent delays in submitting the license application." Will NRC's emphasis on the completeness and quality of the license application have a significant effect on the timing of the submittal of the license application and subsequent licensing proceeding to grant construction authorization in time for repository availability by 2007-2009? As the NRC indicated to DOE in NRC's October 25, 1985 comments on the draft PDS, the three-year statutory schedule for the NRC licensing proceeding on the application for construction authorization is optimistic. The Commission has sought ways to improve the prospects for meeting this schedule, for example by developing the LSS for expedited document discovery during the licensing proceeding. In the same correspondence on the PDS, NRC also stated that the adequacy of the three-year review period depends PART 51 STATEMENTS OF CONSIDERATION on DOE's submittal of a complete and high-quality application. A license application supported by inadequate data may lead to findings during the licensing proceeding that the results of certain tests cannot be admitted as part of the license application. If it is not possible to repeat the tests in question, NRC may have no alternative but to deny the application--with a consequent loss of program momentum and considerable financial cost. In the November 1989 Reassessment Report, DOE announced extensions in all major repository program milestones. The current target date for repository availability is 2010. In a speech before the 1989 Nuclear Energy Forum. W. Henson Moore, Deputy Secretary of Energy, stated that a permanent repository at Yucca Mountain could not be operational before 2010, under optimum circumstances. The 2010 earliest timeframe falls outside of the 2007-2009 timeframe for an "offsite storage solution" in the 1979 Court remand which precipitated the NRC's Waste Confidence Proceeding. In the Reassessment Report, DOE noted that in developing its current schedule, certain activities, one of which was NRC's review of the license application, were outside of DOE's control. However, DOE also stated that it would continue its ongoing interactions with NRC and EPA "to reduce the number of unresolved issues remaining at the time of licensing, which should enhance confidence that the license application can be reviewed in three years, as called for in the Nuclear Waste Policy Act." The NRC does not believe that it is likely that NRC's emphasis on completeness and quality of the license application will contribute to substantial delays in submitting the license application and in the licensing proceeding that would delay repository availability much beyond 2010 at the Yucca Mountain site. In any case, the Commission remains convinced that the benefits to the repository program of submitting a quality license application would outweigh the cost of delay in preparing the application. NRC has always placed great emphasis on early resolution of potential licensing issues in the interest of expeditious review of the license application and timely repository availability. It is in the same spirit of timely repository operation that the Commission is urging greater attention to quality than to meeting the schedule for submittal of the license application. NRC believes that a complete and quality license application offers the best available assurance that timely repository licensing and operation can be achieved. In addition to expediting the review of the application, a high-quality license application and site characterization program should enhance overall confidence that any site granted a construction authorization will prove to be reliable during the period of performance confirmation. It will also increase public confidence that the program is being carried out in a thorough and technically sound manner. II. C. Conclusion on Finding 2 In reexamining the technical and institutional uncertainties surrounding the timely development of a geologic repository since the 1984 Waste Confidence Decision, the Commission has been led to question the conservatism of its expectation that a repository would be available by 2007-2009. At the time of the 1984 Decision, the Commission said that timely attainment of a repository did not require DOE to adhere strictly to the milestones set out in the NWPA, and there would be delays in some milestones. It did not &ppear to the Commission at the time that delays of a year or so in meeting any of the milestones would delay the date of repository availability by more than a few years beyond the 1998 deadline specified in the Act. Since then, however, several developments have made it apparent that delays of more than a few years are to be the norm rather than the exception in the early years of this program. There has been a twelve-year slip in DOE's estimate of repository availability from 1998 to 2010, and DOE has been unable to meet such near-term repository program milestones as excavation of the exploratory shaft and the start of in-situ testing. There remains the possibility that potential repository availability at the Yucca Mountain site will be further delayed due to unforeseen problems during site characterization. In predicting the timing of repository availability, the suitability of Yucca Mountain should not be assumed. Yucca Mountain is now the only candidate site available; the NWPAA required that DOE terminate site characterization activities at all sites other than the Yucca Mountain site. In effect, the 2007-09 schedule for repository availability in the original Waste Confidence Decision could have been met only if Yucca Mountain survived the repository development process as a licensed site without major delays in site characterization and licensing. If this site were found to be unlicenseable or otherwise unsuitable, characterization 51-SC-67 would have to begin at another site or suite of sites, with consequent further delay in repository availability. The final decision on the suitability of the site to proceed to licensing and repository development will rest with DOE, but the position of the NRC staff will figure in that decision. The staff will not be able to make a recommendation to a licensing board to authorize repository construction at Yucca Mountain until all site characterization activities have been completed. DOE might thus be unable for several more years to determine whether there will in fact have to be a delay to find and characterize another site. Another reason the Commission is unwilling to assume the suitability of Yucca Mountain is that NRC must be mindful of preserving all its regulatory options--including a recommendation of license application denial--to assure adequate protection of public health and safety from radiological risk. In our view, it is essential to dispel the notion that for schedular reasons there is no alternative to the currently preferred site. This view is consistent with past Commission statements that the quality of DOE's preparations for a license application should take precedence over timeliness where the two conflict. It is also consistent with the view that because we are making predictions about completion dates for a unique and complex enterprise at least some 20 years hence, it is more reasonable to express the timescale for completion in decades rather than years. In order to obtain a conservative upper bound for the timing of repository availability, the Commission has made the assumption that the Yucca Mountain site will be found to be unsuitable. If DOE were authorized to initiate site screening for a repository at a different site in the year 2000, the Commission believes it reasonable to expect that a repository would be available by the year 2025. This estimate is based on the DOE position that site screening for a second repository should begin 25 years before the start of waste acceptance. The consideration of technical and institutional issues presented here has found none that would preclude the availability of a repository within this timeframe. Given DOE's revised schedule, which provides 11 years for site characterization activities instead of six, it is possible that the Yucca Mountain site could be found unsuitable after the year 2000. In this case, DOE would have fewer than 25 years to initiate site screening and develop a repository for availability by 2025. The NRC will evaluate the likelihood of this j PART 51 STATEMENTS OF CONSIDERATION development during the next scheduled review of the Waste Confidence Decision in 1999. For the second part of its 1984 finding on repository availability, the Commission found reasonable assurance that sufficient repository capacity will be available within 30 years beyond expiration of any reactor OL to dispose of existing commercial high level waste and spent fuel originating in that reactor and generated up to that time. The Commission believes that this finding should also be modified in light of developments since 1984. When the Commission made this finding, it took into consideration both technical and institutional concerns. The technical concern centered on the ability of the spent fuel and the engineered reactor storage facilities to meet the requirements for extended operational storage before shipment for disposal. The institutional question concerned whether the utility currently responsible for post-operational reactor storage, or some substitute organization, would be able to assure the continued safety of this storage. The principal new developments since 1984 that bear on these questions are: (1) that dry spent fuel storage technologies have become operational on a commercial scale; and (2) that several utilities are proceeding with plans to seek renewals of their OLs, with appropriate plant upgrading, for an additional period up to 30 years beyond the 40-year term of their current licenses. The accumulation of operating experience with dry-cask storage, a technology requiring little active term maintenance, provides additional assurance that both the technical and institutional requirements for extended post-operational spent fuel storage will be met. License renewals, however, would have the effect of increasing requirements for both the quantity and possibly the duration of storage. If the Commission were to grant 30-year license renewals, the total operating life of some reactors could be 70 years, so that the spent fuel initially generated in such reactors would have to be stored for about 100 years, if a repository were not available until 30 years after the expiration of their last OLs. This raises the question as to whether that spent fuel, and the hardware and civil engineering structures for storing it, can continue to meet NRC requirements for an additional 30 years beyond the period the Commission supported in 1984. For all the reasons cited in the discussion of Finding 4, the Commission believes there is ample technical basis for confidence that spent fuel can be stored safely and without significant environmental impact at these reactors for at least 100 years. If a repository were available within the first quarter of the twenty-first cent\lrY, the oldest spent fuel could be shipped off the sites of all currently operating reactors well before the spent fuel initially generated in them reached the age of 100 years. The need to consider the institutional aspects of storage beyond 30 years after OL expiration was not in evidence in 1984 because the Commission was confident that at least one repository would be available by 2007-2009. On that schedule, waste acceptance of spent fuel from the first reactor whose operating license had expired [Indian Point 1, terminated in 1980) could have begun within 30 years of expiration of that license. If a repository does not prove to be available until 2025, however, it would not be available within 30 years of the time that OLs could be considered effectively to have expired for Indian Point 1 and the three other plants with spent fuel onsite that were retired before the end of their licensed life for reactor operation. The same would be true of any additional reactors prematurely retired between now and 1995, when the 30-year clock starts for the availability of a repository by 2025. Premature shutdowns notwithstanding, the Commission has reasons to be assured that the spent fuel at all of these reactors will be stored safely and without significant environmental impact until sufficient repository capacity becomes available. Considering first the technical reasons for this assurance, it is important to recognize that each of these reactors and its spent fuel storage installation were originally licensed in part on the strength of the applicant's showing that the systems and components of concern were designed and built to assure safe operation for 40 years under expected normal and transient severe conditions. All of the currently retired reactors have a significant portion of that 40-year expected life remaining, and all have only small quantities of spent fuel onsite in storage installations that were licensed to withstand considerably larger thermal and radiation loadings from much greater quantities of spent fuel. Of the four reactors currently retired with spent fuel onsite, the two with far the longest terms of operation, Lacrosse and Dresden, were operated for 19 and 18 years, respectively. For the continued safe management of the spent fuel in storage installations at any existing or potential prematurely retired plant, the Commission believes it can reasonably rely on the continued 51-SC-68 structural and functional integrity of the plant's engineered storage installations for at least the balance of its originally licensed life as if the OL were still in effect. This is to say that for the purposes of Finding 2, no foreseeable technical constraints have arisen to disturb the Commission's assurance that spent fuel storage at any reactor will remain safe and environmentally acceptable for at least 30 years after its licensed life for operation, regardless of whether its OL has been terminated at an earlier date. The Commission also sees no insurmountable institutional obstacles to the continued safe management of spent fuel during the remainder of any shutdown reactor's initially licensed life for operation, or for at least 30 years thereafter. Because there will still be an NRC possession license for the spent fuel at any reactor that has indefinitely suspended operations, the Commission will retain ample regulatory authority to require any measures, such as removal of the spent fuel remaining in storage pools to passive dry storage casks, that might appear necessary after an OL expires. Even if a licensed utility were to become insolvent, and responsibility for spent fuel management were transferred to DOE earlier than is currently planned, the Commission has no reason to believe that DOE would be unable to carry out any safety-related measures NRC considers necessary. Thus, in the case of a premature reactor retirement, the Commission has an adequate basis, on both technical and institutional grounds, for reasonable assurance that spent fuel can be stored safely and without significant environmental impacts for at least 30 years beyond not only the actual end of that reactor's OL, but the end of its originally licensed life for operation. In sum, considering developments since 1984 in the repository development program, in the operating performance of U.S. power reactors, and in spent fuel storage technology, the Commission finds that: (1) the overall public health, safety, and environmental impacts of the possible unavailability of a repository by 2007-2009 would be insignificant; and (2) neither 30-year renewals of reactor licenses nor a delay in repository availability to 2025 will result in significant safety or

environmental impacts from extended post-operational spent fuel storage. The Commission finds ample grounds for its proposed revised findings on the expected availability of a repository.

The institutional support for the repository program is well-established. A mechanism for funding repository PART 51 STATEMENTS OF CONSIDERATION program activities is in place, and there is a provision in the NWP A for adjusting, if necessary, the fee paid by utilities into this fund. Congress has continued to provide support for the repository program in setting milestones, delineating responsibilities, establishing advisory bodies, and providing a mechanism for dealing with the concerns of States and affected Indian tribes. Technical support for extended spent fuel storage has improved since 1984. Considering the growing availability, reasonable cost, and accumulated operating experience with new dry cask spent fuel storage technology since then, the Commission now has even greater assurance that spent fuel can be stored safely and without significant environmental impact for at least 30 years after the expected expiration of any reactor's

01. Where a reactor's 01 has been terminated before the expected expiration date, the Commission has an adequate basis to reaffirm what was implicit in its initial concept, namely: that regardless of the actual date when the reactor's operating authority effectively ended, spent fuel can be stored safely and without significant environmental impacts for at least 30 years beyond that reactor's licensed life for operation.

There is thus no foreseeable health and safety or environmental requirement that a repository be made available within the 2007-2009 timeframe at issue in the Commission's original proceeding. Indeed, the Commission sees important NRC mission-related grounds for avoiding any statement that repository operation by 2007-2009 is required. Geologic disposal of high-level radioactive wastes is an unprecedented endeavor. It requires reliable projections of the waste isolation performance of natural and engineered barriers over millennia. After the repository is sealed, retrieval of the emplaced wastes will no longer be practicable, and the commitment of wastes to that site will, by design, be irreversible. In DOE's testing, both in the laboratory and at the candidate repository site, in its development of facility and package designs, and in all other work to demonstrate that NRC requirements will be met for a repository at Yucca Mountain, the Commission believes that the confidence of both NRC and the public depends less on meeting the schedule for repository operation than on meeting safety requirements and doing the job right the first time. Thus, given the Commission's assurance that spent fuel can safely be stored for at least 100 years if necessary, it appears prudent for all concerned to prepare for the better-understood and more manageable problems of storage for a few more years in order to provide additional time to assure the success of permanent geologic disposal. This is not to say that the Commission is unsympathetic to the need for timely progress toward an operational repository. It is precisely because NRC is so confident of the national commitment to achieve early repository operation that the Commission believes it no longer need add its weight to the considerable pressures already bearing on the DOE program. There is ample institutional impetus on the part of others, including Congress, the nuclear power industry, State utility rate regulatory bodies, and consumers of nuclear-generated power, toward DOE achievement of scheduled program milestones. With continuing confidence in the technical feasibility of geologic disposal, the Commission has no reason to doubt the institutional commitment to achieve it in a timeframe well before it might become necessary for safety or environmental reasons. Indeed, the Commission believes it advisable not to attempt in this review a more precise NRC estimate of the point at which a repository will be needed for radiological safety or environmental reasons, lest this estimate itself undermine the commitment to earlier achievement of repository operations. To find reasonable assurance that a repository will be available by 2007-2009, however, is a different and more consequential proposition in the context of this review. In light of the delays the program has encountered since its inception, and the regulatory need to avoid a premature commitment to the Yucca Mountain site, the Commission could not prudently describe a basis for assurance that the previous DOE schedule for repository operation in 2003 would not slip another four to six years under any reasonably foreseeable circumstances. The NRC believes it is more realistic to expect that a repository at the Yucca Mountain site could be available by the year 2010 or a few years thereafter, if the Yucca Mountain site is found to be suitable. This revised estimate, however, could too easily be misinterpreted as an NRC estimate of the time at which continued spent fuel storage at these sites would be unsafe or environmentally significant. The Commission's enhanced confidence in the safety of extended spent fuel storage provides adequate grounds for the view that NRC need not at this time define more precisely the period when, for 51-SC-69 reasons related to NRC's mission, a permanent alternative to operational spent fuel storage will be needed. The Commission therefore proposes the following revision of its original Finding on when sufficient repository capacity will be available: The Commission finds reasonable assurance that at least one mined geologic repository will be available within the first quarter of the twenty-first century, and sufficient repository capacity will be available within 30 years beyond the licensed life for operation [which may include the term of a revised or renewed license)' of any reactor to dispose of the commercial level radioactive waste and spent fuel originating in such reactor and generated up to that time. Reaffirmed Finding 3: The Commission finds reasonable assurance that high-level radioactive waste and spent fuel will be managed, in a safe manner until sufficient repository capacity is available to assure the safe disposal of all high-level waste and spent fuel. Ill.A. Issues Considered in Commission's 1984 Decision on Finding 3 In the Commission's discussion of Finding 3 in its Waste Confidence Decision (49 FR 34658, August 31, 1984), in Section 2.3 ;;,Third Commission Finding,' the Commission stated, Nuclear power plants whose operating licenses expire after the years 2007-09 will be subject to NRC regulation during the entire period between their initial operation and the availability of a waste repository. The Commission has reasonable assurance that the spent fuel generated by these licensed plants will be managed by the licensees in a safe manner. Compliance with the NRC regulations and any specific license conditions that may be imposed on the licensees will assure adequate protection of the public health and safety. Regulations primarily addressing spent fuel storage include 10 CFR Part 50 for storage at the reactor facility and 10 CFR Part 72 for storage in independent spent fuel storage installations [ISFSis]. Safety and environmental issues involving such storage are addressed in licensing reviews under both Parts 50 and 72, and continued storage operations are audited and inspected by NRC. NRC's experience in more than 80 individual evaluations of the safety of spent fuel storage shows that significant releases of radioactivity from spent fuel under licensed storage conditions are extremely remote. Some nuclear power plant operating licenses expire before the years 2007-09. For technical, economic or other reasons, other plants may choose, or be forced to terminate opera lion prior to 2007-09 even though their *The parenthetical phrase "which may include the term of a revised or renewed license" has been added to revised Finding 2 to make it consistent with revised Finding 4. PART 51 STATEMENTS OF CONSIDERATION opera ting licenses have not expired. For example, the existence of a safety problem for a particular plant could prevent further operation of the plant or could require plant modifications that make continued plant operation uneconomic. The licensee, upon expiration or termination of its license, may be granted [under 10 CFR Part 50 or Part 72) a license to retain custody of the spent fuel for a specified term (until repository capacity is available and the spent fuel can be transferred to DOE under Sec. 123 of the Nuclear Waste Policy Act of 1982) subject to NRC regulations and license conditions needed to assure adequate protection of the public. Alternatively, the owner of the spent fuel, as a last resort, may apply for an interim storage contract with DOE, under Sec. 135(b] of the Act, until not later than 3 years after a repository or monitored retrievable storage facility is available for spent fuel. For the reasons discussed above, the Commission is confident that in every case the spent fuel generated by those plants will be managed safely during the period between license expiration or termination and the availability of a mined waste repository for disposal. Even if a repository does not become available until 2025, nothing has occurred during the five years since its original Decision to diminish the Commission's confidence that high-level waste and spent fuel will be managed in a safe manner until a repository is available. The same logic just stated continues to apply through the first quarter of the twenty-first century. NRC l regulations remain adequate to assure safe storage of spent fuel and radioactive high-level waste at reactors, at independent spent fuel storage installations (ISFSis), and in an MRS until sufficient repository capacity is available. 10 CFR subsection 72.42(a) provides for renewal of licensed storage at ISFSis for additional 20-year periods for interim storage, or for additional 40-year periods for monitored retrievable storage of spent fuel and solidified radioactive high-level waste if an MRS facility is constructed, licensed, and operated. This would ensure that spent fuel and solidified high-level waste, if any were to be delivered to an MRS facility, would remain in safe storage under NRC regulation throughout its storage. The Commission has also published for public comment a proposed amendment to part 72 to issue a general license to reactor licensees to use approved spent fuel storage casks at reactor sites. Currently, the Commission is considering the draft final amendment for this rulemaking action. If this amendment is promulgated, no specific part 72 license would be required. Operating license holders would register with NRC to use approyed casks on their sites. Spent fuel may continue to be stored in the reactor spent fuel pool under a part 50 "possession only" license after the reactor has ceased operating. In addition, DOE's policy of disposing of the oldest fuel first, a.s set forth in its Annual Capacity Report, makes it unlikely that any significant fraction of total spent fuel generated will be stored for longer than the 30 years beyond the expiration of any operating reac~or . license. This expectation, established m the Commission's original proceeding, continues to be reasonable, even in the event that a repository is not available until some time during the first quarter of the twenty-first century. Even in the case of premature shutdowns, where spent fuel is most likely to remain at a site for 30 years or longer beyond 01 expiration (see Finding 2, previously discussed), the Commission has confidence that spent fuel will be safely managed until safe disposal is available. Until the reactor site has been fully decommissioned, and spent fuel has been transferred from the utility to DOE as required by NRC regulations, the licensee remains responsible to NRC. Furthermore, under 10 CFR subsection 50.54bb, originally issued in final form by the Commission with its 1984. Waste Confidence Decision, a reactor licensee must provide to NRC, five years before expiration of an OL, notice of plans for spent fuel disposition. Accordingly, the Commission concludes that nothing has changed since the enactment of the Nuclear Waste Policy Act of 1982 and the Waste Confidence Decision in August 1984 to diminish the Commission's " ... reasonable assurance that high-level radioactive waste and spent fuel will be managed in a safe manner until sufficient repository capacity is available .... " Pursuant to the NWPA, the Commission issued in final form 10 CFR part 53, "Criteria and Procedures for Determining Adequacy of Available Spent Nuclear Fuel Storage Capacity," addressing the determination of need, if any, for DOE interim storage. No applications were received by the June 30, 1989 NWPA deadline incorporated into the Commission's rule, and it seems unlikely that any applications will be made to NRC for interim storage by DOE. Even if NRC had made an exception for a late application, a determination would have to have been made before January 1, 1990 to comply with the NWPA. JII.B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 3 Although a DOE facility may not be available to enable the Department to 51-SC-70 begin accepting spent fuel in 1998, as currently provided in the contracts under the NWPA, the Commission's confidence in safe storage is unaffected by any potential contractual dispute between DOE and spent fuel generators and owners as to responsibility for spent fuel storage. In the event that DOE does not take title to spent fuel by this date, a licensee under either 10 CFR part 50 or part 72 cannot abandon spent fuel in its possession. The Commission recognizes that the NWPA limitation of 70,000 MTHM for the first repository will not provide adequate capacity for the total amount of spent fuel projected to be generated by all currently operating licensed reactors. The NWPAA effectively places a moratorium on a second repository program until 2007-2010. Either the first repository must be authorized and able to provide expanded capacity sufficient to accommodate the spent fuel generated, or there must be more than one repository. Since Congress specifically provided in the NWPAA for a first repository, and required DOE to return for legislative authorization for a second repository, the Commission believes that Congress will continue to provide institutional support for adequate repository capacity. The Commission's confidence about the availability of repository capacity is not affected by the possibility that some existing reactor licenses might be renewed to permit continued generation of spent fuel at these sites. Because only two reactor licenses are scheduled to expire before 2003, the impact of license renewals (a matter not considered in the Commission's 1984 Decision) will have no significant effect within the first quarter of the twenty-first century on scheduling requirements for a second repository. Renewals may slightly alleviate the need for a second repository in the short term, because spent fuel storage capacity will be expanded for extended storage at these reactor sites. Over the longer term, renewals might increase spent fuel generation well into the latter half of the twenty-first century. Nonetheless, nothing in this situation diminishes the Commission's assurance that safe storage will be made available as needed. In summary, the Commission finds no basis for changing the Third Finding in its Waste Confidence Decision. The Commission continues to find " ... reasonable assurance that high-level radioactive waste and spent fuel will be managed in a safe manner until sufficient repository capacity is PART 51 STATEMENTS OF CONSIDERATION available to assure the 'safe disposal of all high-level waste and spent fuel." Original Finding 4: The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the expiration of that reactor's operating license at that reactor's spent fuel storage basin, or at either onsi le or off site independent spent fuel storage installations. Revised Finding 4: The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the licensed life for operation (which may include the term of a revised or renewed license) of that reactor at its spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations. IV.A. Issues Considered in Commission's 1984 Decision on Finding 4 In the Commission's discussion of Finding 4 in its Waste Confidence Decision (49 FR 34658; August 31, 1984) section 2.4 "Fourth Commission Finding," the Commission said that: Although the Commission has reasonable assurance that at least one mined geologic repository will be available by the years 2007-09, the Commission also realizes that for various reasons, including insufficient capacity to immediately dispose of all existing spent fuel, spent fuel may be stored in existing or new storage facilities for some periods beyond 2007-09, The Commission believes that this extended storage will not be necessary for any period longer than 30 years beyond the term of an operating license. For this reason, the Commission has addressed on a generic basis in this decision the safety and environmental impacts of extended spent fuel storage at reactor spent fuel basins or at either onsite or offsite spent fuel storage installations. The Commission finds that spent fuel can be stored safely and without significant environmental impacts for at least 30 years beyond the expiration of reactor operating licenses. To ensure that spent fuel which remains in storage will be managed properly until transferred to DOE for disposal, the Commission is proposing an amendment to its regulations (10 CFR Part 50). The amendment will require the licensee to notify the Commission, five years prior to expiration of its reactor operating license, how the spent fuel will be manag'ed until disposal. The Commission's finding is based on the record of this proceeding which indicates that significant releases of radioactivity from spent fuel under licensed storage conditions are highly unlikely. It is also supported by the Commission's experience in conducting more than 80 individual safety evaluations of storage facilities. The safety of prolonged spent fuel storage can be considered in terms of four major issues: [a) The long-term integrity of spent fuel under water pool storage conditions, (b) structure and component safety for extended facility operation, (c) the safety of dry storage, and (d) potential risks of accidents and acts of sabotage at spent fuel storage facilities.

For reasons discussed above, the Commission arrived at a provisional figure of 70 years or more for storage (i.e., a 40-year reactor OL span, plus 30 years or more). The 70-year-plus estimate is supported by oral testimony from the nuclear industry to the Commission in the Waste Confidence Proceeding. (See Transcript of Commission Meeting, "In the Matter of: Meeting on Waste Confidence Proceeding," January 11, 1982, Washington, DC, pp.148-160).

This testimony specifically addressed safety issues related to water pool storage of spent fuel and supported the position that spent fuel could be stored for an indefinite period, citing the industry's written submittal to the Commission in the proceeding. (See "The Capability for the Safe Interim Storage of Spent Fuel" (Document 4 of 4), Utility Nuclear Waste Management Group and Edison Electric Institute, July 1980). Some of this material alluded to in the oral testimony was subsequently referenced by the Commission in its discussion of water pool storage issues and its Fourth Finding of reasonable assurance that spent fuel and high level waste " ... will be managed in a safe manner." (See 49 FR 34658 at pp. 34681-2, August 31, 1984). If a reactor with a 40-year initial license were to have that license renewed for another 30 years, the Commission believes that the spent fuel generated at that reactor can be safely stored for at least several decades past the end of the 70-year operating period. Adding to these 70 years the expected 30-year post-OL period during which the Commission believes, under Finding 2, that sufficient repository capacity will be made available for any reactor's spent fuel, the total storage time would be about 100 years. In making the original Fourth Finding, the Commission did not determine that for technical or regulatory reasons, storage would have to be limited to 70 years. This is apparent from the Commission's use of the words " ... for at least 30 years beyond the expiration of that reactor's operating license ... [emphasis added]." Similarly, in using the words "at least" in its revised Finding Four, the Commission is not suggesting 30 years beyond the licensed life for operation [which may include the term of a revised or renewed license) represents any technical limitation for 51-SC-71 safe and environmentally benign storage. Degradation rates of spent fuel in storage, for example, are slow enough that it is hard to distinguish by degradation alone between spent fuel in storage for less than a decade and spent fuel stored for several decades. The Commission's revised Finding here is meant to apply_ both to wet storage in reactor pools and dry storage in engineered facilities outside the reactor containment building. Both dry and wet storage will be discussed in detail next. Since the original Waste Confidence Decision, which found that material degradation processes in dry storage were well-understood, and that storage systems were simple, passive, and easily maintained, NRC and ISFSI operators have gained experience with dry storage which confirms the Commission's 1984 conclusions. NRC staff safety reviews of topical reports on storage-system designs, the licensing and inspection of storage at two reactor sites, and NRC promulgation of the part 72 amendment for MRS, have significantly increased the agency's understanding of and confidence in dry storage. Under NWPA Section 218(a), DOE has carried out spent fuel storage research and development as well as demonstration of dry cask storage at its Idaho National Engineering Laboratory. Demonstration has been carried out for metal casks under review or previously reviewed by NRC staff. DOE has also provided support to utilities in dry storage licensing actions (see Godlewski, N.Z., "Spent Fuel An Update," Nuclear News, Vol. 30, No. 3, March 1987, pp.47-52). Dry storage of spent fuel has become an available option for utilities, with reactor dry storage licensed and underway at three sites: the H.B. Robinson Steam Electric Plant, Unit 2, in South Carolina, and the Surry Nuclear Station in Virginia. A license was recently granted for a modular system at Duke Power Company's Oconee Nuclear Station site. New applications have been received in 1989 for CP&L's Brunswick site, for the Baltimore Gas and Electric Company's Calvert Cliffs site, and in 1990 for Consumer Power Company's Palisades site. Based on utility statements of intent, and projections of need for additional storage capacity at reactor sites, the NRC staff expects numerous applications from utilities over the next decade (see "Final Version Dry Cask Storage Study," DOE/RW-0220, February 1989). PART 51 STATEMENTS OF CONSIDERATION Since the original Waste Confidence finding, the Commission has reexamined long-term spent fuel storage in issuing an amendment to 10 CFR part 72 to address the storage of spent fuel and high-level radioactive waste in an MRS, as envisioned by Congress in Section 141 of the NWPA. Under this rule, storage in an MRS is to be licensed for a period of 40 years, with the possibility for renewal. The Commission determined not to prepare an environmental impact statement for the proposed amendments to 10 CFR part 72, however. (See 53 FR 31651, p. 31657; August 19, 1988.J An environmental assessment and finding of no significant impact were issued because the Commission found that the consequences of long-term storage are not significant. The environmental assessment for 10 CFR part 72, "Licensing Requirements for the Independent Storage of Spent Fuel and High-Level Radioactive Waste," NUREG-1092, assessed dry storage of spent fuel for a period of 70 years after receipt of spent fuel from a reactor: The basis chosen for evaluating license requirements for the long-term storage of spent nuclear fuel and high-level radioactive waste in an MRS is an installation having a 70-year design lifetime and a 70,000 MTU storage capability. This assessment focuses on the potential environmental consequences for a long-term storage period, a period for which the Commission needs to assure itself of the continued safe storage of spent fuel and high-level radioactive waste and the performance of materials of construction. This means the reliability of systems important to safety needs to be established to ensure that long-term storage of spent fuel and HL W does not adversely impact the environment. For example, the staff needs to establish that systems, such as concrete shielding, have been evaluated to determine how their physical properties withstand the consequences of irradiation and heat flux for about a 70-year period. The Commission addressed structure and component safety for extended operation for storage of spent fuel in reactor water pools in the matter of waste confidence rulemaking proceeding. The Commission's preliminary conclusion is that experience with spent fuel storage provides an a de qua te basis for confidence in the continued safe storage of spent fuel for at least 30 years after expiration of a plant's license. The Commission is therefore confident of the safe storage of spent fuel for at least 70 years in water pools at facilities designed for a 40-year lifetime. The Commission also stated that its authority to require continued safe management of spent fuel generated by licensed plants protects the public and assures them the risks remain acceptable. In consideration of the safety of dry storage of spent fuel, the Commission's preliminary conclusions were that [its] confidence in the extended dry storage of spent fuel is based on a reasonable understanding of the material degradation processes, together with the recognition that dry storage systems are simpler and more readily maintained. In response to Nuclear Waste Policy Act of 19B2 authorizations, the Commission noted; ;, ... the Commission believes the information above [on dry spent fuel storage research and demonstration] is sufficient to reach a conclusion on the safety and environmental effects of extended dry storage. All areas of safety and environmental concern [e.g., maintenance of systems and components, prevention of material degradation, protection against accidents and sabotage) have been addressed and shown to present no more potential for adverse impact on the environmental and the public health and safety than storage of spent fuel in water pools.' At this time, the Commission is confident it can evaluate the long-term integrity of material for constructing an installation and provide the needed assurance for safe storage of spent fuel and HLW to establish the licensibility of an MRS over extended periods of time. The MRS fuel storage concepts discussed here for revision of 10 CFR Part 72 covers only dry storage concepts. [References omitted] The Commission believes that its 1984 Fourth Finding should be changed to reflect the environmental assessment in the 10 CFR part 72 MRS rulemaking and other evidence that spent fuel can be stored, safely and without significant environmental impact, for extended periods. Although the Commission does not believe storage in excess of a century to be likely, with or without an MRS, there is the potential for storage of spent fuel for times longer than 30 years beyond the expiration of an initial, extended, or renewed reactor OL, if a reactor operating under such a license were prematurely shut down. The Commission does not, however, see any significant safety or environmental problems associated with storage for at least 30 years after the licensed life for operation of any reactor, even if this effectively means storage for at least 100 years, in the case of a reactor with a 70-year licensed life for operation. Under the environmental assessment for the MRS rule, the Commission has found confidence in the safety and environmental insignificance of dry storage of spent fuel for 70 years following a period of 70 years of storage in spent fuel storage pools. Thus, this environmental assessment supports the proposition that spent fuel may be stored safely and without significant environmen ta! impact for a period of up to 140 years if storage in spent fuel pools occurs first and the period of dry storage does not exceed 70 years. The Commission has also found that experience with water-pool storage of spent fuel continues to confirm that pool storage is a benign environment for spent fuel that does not lead to 51-SC-72 significant degradation of spent fuel integrity. Since 1984, utilities have continued to provide safe additional reactor pool storage capacity through reracking, with over 110 such actions now completed. The safety of storage in pools is widely recognized among cognizant professionals. Specifically, the Commission notes one expert's view that: During the last 40 years there has been very positive experience with the handling and storing of irradiated fuel in water: thus wet storage is now considered a proved technology. There is a substantial technical basis for allowing spent fuel to remain in wet storage for several decades. For the past two decades, irradiated Zircaloy-clad fuel has been handled and stored in water. There continues to be no evidence that clad fuel degrades significantly during wet storage--this includes: fuel with burnups as high as 41,000 MWd/MTU; continuous storage of low-burn up fuel for as long as 25 years; and irradiation of fuel in reactors for periods up to 22 years. Cladding defects have had little impact during wet storage, even if the fuel is uncanned. [References omitted.] [See Bailey, W.J. and Johnston, Jr. A.B .. et al., "Surveillance of LWR Spent Fuel in Wet Storage," NP-3765, Electric Power Research Institute [EPRI), October 1984, pp. 2-10.] This last conclusion has been reaffirmed by the same authors, who recently wrote: "There continues to be no evidence that LWR spent fuel with Zircaloy or stainless steel cladding degrades significantly during wet storage [EPRI 1986; International Atomic Energy Agency (IAEAJ 1982]." (See "Results of Studies on the Behavior of Spent Fuel in Storage," Journal of the Institute of Nuclear Materials Management, Vol. XVI, No. 3, April 1988, p. 27.IV A). In addition to the confidence that the spent fuel assemblies themselves will not degrade significantly in wet storage, there is confidence that the water pools in which the assemblies are stored will remain safe for extended periods: As noted in the recent IAEA world survey, the 40 years of positive experience with wet storage illustrates that it is a fully-developed technology with no associated major technological problems. Spent fuel storage pools are operated without substantial risk to the public or the plant personnel. There is substantial technical basis for allowing spent fuel to remain in wet storage for several decades. Minor, but repairable, problems have occurred with spent fuel storage pool components such as liners, racks, and piping. [See Bailey, W.J., and Johnson, Jr., A.B., et al., "Surveillance of LWR Spent Fuel in Wet Storage," EPRI NP-3765, prepared by Battelle Pacific Northwest Laboratories, Final Report, October 1984, p. 6-1.] The studies just cited also support the view that rates of uniform corrosion of spent fuel cladding in storage pools are low over time. Localized corrosion on PART 51 STATEMENTS OF CONSIDERATION cladding surfaces has also been gradual and can be expected to remain so. Cladding that has undergone damage while in the reactor core has not resulted in significant releases of radioactivity when stored in pools. Furthermore, the operational experience accumulated since the 1984 Waste Confidence Decision and NRC experience in licensing and inspection reinforce the conclusions in that Decision that wet storage involves a relatively benign environment. There are no driving mechanisms, such as temperature and pressure, to degrade storage structures or components or the fuel itself, or to spread contamination. Degradation mechanisms are gradual and well understood; they allow ample time for remedial action, including repair or replacement of any failing systems. This extensive experience adequately supports predictions of term integrity of storage basins. The Commission also notes the endorsement of this basic confidence by cognizant professional organizations: The American Nuclear Society issued a policy statement [ANS 1986] in 1986 regarding storage of spent nuclear fuel. The statement indicates that continued wet storage of spent fuel at nuclear power plant sites until the federal government accepts it under existing contracts with the utilities is safe, economical and environmentally acceptable. [See Gilbert, E.R., Bailey, W.J., and Johnson, A.B., "Results of Studies on the Behavior of Spent Fuel in Storage," Journal of the Institute of Nuclear Materials Management, Vol. XVI, No. 3, April 1988, p. 27.IVAJ.] The Commission is aware that in December 1986 at the Hatch nuclear power plant, radioactive water leaked out of a spent fuel transfer canal between spent fuel pools. Contaminated water drained into a swamp and from there into the Altamaha River. Also, more recently, on August 16, 1988, a spent fuel pool cooling pump failed at the Turkey Point nuclear power plant, causing about 3000 gallons of radioactive water to leak into the spent fuel pool heat exchanger room. Approximately 1500 gallons leaked from that room to adjacent areas. Approximately six to seven gallons entered the plant intake canal via storm drains. There was no radiation release offsite in this event. However, the shoes and clothing of approximately 15 workers were contaminated. ' The occurrence of operational events like these have been addressed by the NRC staff at the plants listed. The staff has taken inspection and enforcement actions to reduce the potential for such operational occurrences in the future. The NRC staff has spent several years studying in detail catastrophic loss of reactor spent fuel pool water possibly resulting in a fuel fire in a dry pool, and recently participated in litigation over this issue relative to Vermont Yankee. The 1987 report, "Severe Accidents in Spent Fuel Pools in.Support of Generic Safety Issue 82" [NUREG/CR-4982), referred to in Public Citizen's comment represents an early part of the NRC's study. Subsequent study of the consequences and risks due to a loss of coolant water from spent fuel pools was conducted by the NRC, and the results were published in NUREG/CR-5176, "Seismic Failure and Cask Drop Analysis of the Spent Fuel Pools at Two Representative Nuclear Power Plants," January 1989, and NUREG-1353, "Regulatory Analysis for the Resolution of Generic Issue 82, >Beyond Design Basis Accidents in Spent Fuel Pools'," April 1989. These reports were cited in the Commission's Proposed Waste Confidence Decision Review (54 FR 39767-39797, at p.39795, September 28, 1989). Also issued in 1989, as part of the NRC staffs study, was "Value/Impact Analyses of Accident Preventive and Mitigative Options for Spent Fuel Pools" (NUREG/CR-5281). The primary concern regarding accidents in spent fuel pools is the loss of water and its capability to cool the radioactive fuel. Without sufficient water cooling, some performance assessment models suggest that the fuel's zircaloy cladding may initiate and sustain rapid oxidation (fire) that may spread to adjacent fuel assemblies, with the potential of releasing large amounts of radioactivity. The analyses reported in these NUREGs indicate that the dominant accident sequence which contributes to risk in a spent fuel pool is gross structural failure of the pool due to seismic events. Risks due to other accident scenarios (such as pneumatic seal failures, inadvertent drainage, loss of cooling or make-up water, and structural failures due to missiles, aircraft crashes and heavy load drops) are at least an order of magnitude smaller. For this study, older nuclear power plants were selected, since the older plants are more vulnerable to seismic-induced failures. The selected plants included the Vermont Yankee and the H.B. Robinson plants. Although these studies conclude that most of the spent fuel pool risk is derived from beyond design basis earthquakes, this risk is no greater than the risk from core damage accidents due to seismic events beyond the shutdown earthquake. Because of the large inherent safety margins in the design and construction of the spent fuel pool analyzed, it was determined that 51-SC-73 no action was justified to further reduce the risk (NUREG-1353). As stated in the Preface to NUREG-1353: This report presents the regulatory analysis, including decision rationale, for the resolution of Generic Issue 82, ;,Beyond Design Basis Accidents in Spent Fuel Pools.' The object of this regulatory analysis is to determine whether the use of high density storage racks for the storage of spent fuel poses an unacceptable risk to the health and safety of the public. As part of this effort, the seismic hazards for two older spent fuel pools were evaluated. The risk change estimates, value/impact and cost-benefit analyses, and other insights gained during this effort, have shown that no new regulatory requirements are warranted in relation to this generic issue. Thus, supported by the consistency of NRC experience with that of others, the Commission has concluded that spent fuel can be stored safely and without significant environmental impact, in either wet storage or in wet storage followed by dry storage, for at least 100 years. The Commission considers it unlikely, however, that any fuel will actually remain in wet storage for 100 years or even for 70 years. We anticipate that, consistent with the currently developing trend, utilities will move fuel rods out of spent fuel pools and into dry storage to make room in pools for freshly-discharged spent fuel. Although the Commission has concluded that reactor spent fuel pools can safely be used to store spent fuel for 100 years, there is no technically compelling reason to use them that long. If reactor licenses are renewed for as long as 30 years, making a total of 70 years of operation, it will be necessary to store the spent fuel discharged at the end of the reactor's operation in a spent fuel pool for several years to allow for radioactive decay and thermal cooling. After this period, the fuel could be placed in dry storage and the spent fuel pool decommissioned. Thus, for most reactors, the most likely maximum period of storage will be well within the extended 30-year post-operational period under the Commission's proposed revision to Finding 4. Moreover, considering that under certain conditions spent fuel can be stored safely and without significant environmental impacts for up to 140 years, the Commission believes there is ample basis for confidence in storage for at least 100 years. In its 1984 Waste Confidence Decision, the Commission also concluded that "there are no significant additional non-radiological impacts which could adversely affect the environment if spent fuel is stored beyond the expiration of operating PART 51 STATEMENTS OF CONSIDERATION licenses for reactors" [see 49 FR 34658 at p. 34686, August 31, 1984). The Commission did not find anything to contradict this conclusion in its 1988 rulemaking amending 10 CFR part 72 for long-term spent fuel and high-level waste storage at an MRS: In August 1984, the NRC published an environmental assessment for this proposed revision of Part 72 NUREG-1092, ;;,Environmental Assessment for 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Fuel and Level Radioactive Waste.' NUREG-1092 discusses the rriajor issues of the rule and the potential impact on the environment. The findings of the environmental assessment are ;;,(1) past experience with water pool storage of spent fuel establishes the technology for long-term storage of spent fuel without affecting the health and safety of the public, (2) the proposed rulemaking to include the criteria of 10 CFR Part 72 for storing spent nuclear fuel and high-level radioactive waste does not significantly affect the environment, (3) solid high-level waste is comparable to spent fuel in its heat generation and in its radioactive material content on a per metric ton basis, and (4) knowledge of material degradation mechanisms under dry storage conditions and the ability to institute repairs in a reasonable manner without endangering the health [and safety] of the public shows dry storage technology options do not significantly impact the environment.' The assessment concludes that, among other things, there are no significant environmental impacts as a result of promulgation of these revisions of 10 CFR Part 72. Based on the above assessment, the Commission concludes that the rulemaking action will not have a significant incremental environmental impact on the quality of the human environment. [53 FR 31651 at pp. 31657-31658; August 19, 1988.) Thus, the 1988 amendments to 10 CFR part 72 provide the basis for the Commission to conclude that the environmental consequences of term spent fuel storage, including radiological impacts, are not significant. Finally, no considerations have arisen to affect the Commission's confidence since 1984 that the possibility of a major accident or sabotage with offsite radiological impacts at a spent-fuel storage facility is extremely remote. NRC has recently reexamined reactor pool storage safety in two studies, "Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two Representative Nuclear Power Plants" (NUREG/CR-5176) and "Beyond Design Basis Accidents in Spent Fuel Pools" (NUREG-1353). These studies reaffirmed that there are no safety considerations that justify changes in regulatory requirements for pool storage. Both and dry-storage activities have continued to be licensed by the Commission. In its recent rulemaking amending 10 CFR part 72 to establish licensing requirements for an MRS, the Commission did choose to eliminate an exemption regarding tornado missile impact " ... to assure designs continue to address maintaining confinement of particulate material;' (53 FR 31651, p. 31655, August 19, 1988). However, NRC staff had previously considered tornado missile impacts in safety reviews of design topical reports and in licensing reviews under 10 CFR part 72. JV.B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 4 In its original Finding 4, the Commission found reasonable assurance of safe storage without significant environmental impacts for at least 30 years beyond reactor OL expiration. Delays and uncertainties in the schedule for repository availability since the 1984 Decision have convinced the Commission to allow some margin beyond the scheduled date for repository opening currently cited by DOE. As noted in Finding 2, the Commission has reasonable assurance that at least one repository will be available within the first quarter of the twenty-first century. For all currently operating reactors, this would still be within the period of 30 years from expiration of their OLs, which the Commission previously found to be the minimum period for which spent fuel storage could be considered safe and without significant environmental impact. Under the NWPA as amended, DOE is authorized to dispose of up to 70,000 MTHM in the first repository before granting a construction authorization for a second. Under existing licenses, projected spent fuel generation could exceed 70,000 MTHM as early as the year 2010. Possible extensions or renewals of OLs also need to be considered in assessing the need for and scheduling the second repository. It now appears that unless Congress lifts the capacity limit on the first and unless this repository has the physical capacity to dispose of all spent fuel generated under both the original and extended or renewed licenses--it will be necessary to have at least one additional repository. Assuming here that the first repository is available by 2025 and has a capacity on the order of 70,000 MTHM, additional disposal capacity would probably not be needed before about the year 2040 to avoid storing spent fuel at a reactor for more than 30 years after expiration of reactor OLs. Although action on a second repository before the year 2007 would require Congressional approval, the 51-SC-74 Commission believes that Congress will take the necessary action if it becomes clear that the first repository site will not have the capacity likely to be needed. If DOE were able to address the need for a second repository earlier, for example by initiating a survey for a second repository site by the year 2000, DOE might be able to reduce the potential requirement for extended spent fuel storage in the twenty-first century. The Commission does not, however, find such action necessary to conclude that spent fuel can be stored safely and without significant environmental impact for extended periods. The potential for generation and onsite storage of a greater amount of spent fuel as a result of the renewal of existing OLs does not affect the Commission's findings on environmental impacts. In Finding 4, the Commission did not base its determination on a specific number of reactors and amount of spent fuel generated. Rather, the Commission took note of the safety of spent fuel storage and lack of environmental impacts overall, noting that individual actions involving such storage would be reviewed. In the event there were applications for renewal of existing reactor OLs, each of these actions would be subject to safety and environmental reviews, with subsequent issuance of an environmental assessment or environmental impact statement, which would cover storage of spent fuel at each reactor site during the period of the renewed license. The Commission also notes that the amount of spent fuel expected to be discharged by reactors has continued to decline significantly, a trend already noted in the Commission's discussion of its Finding 5 (49 FR 34658 at p. 34687, August 31, 1984). At the time of the Commission's decision, " ... the cumulative amount of spent fuel to be disposed of in the year 2000 [was] expected to be 58,000 metric tons of uranium" (see "Spent Fuel Storage Requirements" (Update of DOE/RL 17) DOE/RL-83-1, January, 1983). Today, that figure has declined to 40,200 metric tons, the lower reference case which represents the conservative upper bound of commercial nuclear power growth (see "Integrated Data Base for 1989: Spent Fuel and Radioactive Waste Inventories, Projections, and Characteristics," DOE/RW-0006, Rev. 5, November 1989). The amount of spent fuel considered likely to be discharged by the year 2000 in the Commission's 1984 decision will not be attained until the end of calendar year 2010, if then. PART 51 STATEMENTS OF CONSIDERATION The Commission believes that its 1984 Finding 4 should be revised to acknowledge the possibility and assess the safety and environmental impacts of extended storage for periods longer than 70 years. The principal reasons for this proposed revision are that: (1) the term material and system degradation effects are well understood and known to be minor; (2) the ability to maintain the system is assured; and (3) the Commission maintains regulatory authority over any spent fuel storage installation. On the basis of experience with wet and dry spent fuel storage and related rulemaking and licensing actions, the Commission concludes that spent fuel can be safely stored without significant environmental impact for at least 100 years, if necessary. Therefore, the Commission is revising its original Fourth Finding thus: "The Commission finds reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the licensed life for operation (which may include the term of a revised or renewed license) of that reactor at its spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations." Reaffirmed Finding 5: The Commission finds reasonable assurance that safe independent onsite spent fuel storage or offsite spent fuel storage will be made available if such storage capacity is needed. V.A. Issues Considered in Commission's 1984 Decision on Finding 5 In its discussion of Finding 5 of its Waste Confidence Decision (49 FR 34658; August 31, 1984), the Commission said that: The technology for independent spent fuel storage installations, as discussed under the fourth Commission Finding, is available and demonstrated. The regulations and licensing procedures are in place. Such installations can be constructed and licensed within a five-year time interval. Before passage of the Nuclear Waste Policy Act of 1982 the Commission was concerned about who, if anyone, would take responsibility for providing such installations on a timely basis. While the industry was hoping for a government commitment, the Administration had discontinued efforts to provide those storage facilities .... The Nuclear Waste Policy Act of 1982 establishes a national policy for providing storage facilities and thus helps to resolve this issue and assure that storage capacity will be available. Prior to March 1981, the DOE was pursuing a program to provide temporary storage in off-site, or away-from-reactor (AFR), storage installations. The intent of the program was to provide flexibility in the national waste disposal program and an alternative for those utilities unable to expand their own storage capacities. Consequently, the participants in this proceeding assumed that, prior to the availability of a repository, the Federal government would provide for storage of spent fuel in excess of that which could be stored at reactor sites. Thus, it is not surprising that the record of this proceeding prior to the DOE policy change did not indicate any direct commitment by the utilities to provide AFR storage. On March 27, 1981, DOE placed in the record a letter to the Commission stating its decision ;;,to discontinue its efforts to provide Federal government-owned or controlled reactor storage facilities.' The primary reasons for the change in policy were cited as new and lower projections of storage requirements and lack of Congressional authority to fully implement the original policy. The record of this proceeding indicates a general commitment on the part of industry to do whatever is necessary to avoid shutting down reactors o* derating them because of filled spent fuel storage pools. While industry's incentive for keeping a reactor in operation no longer applies after expiration of its operating license, utilities possessing spent fuel are required to be licensed and to maintain the fuel in safe storage until removed from the site. Industry's response to the change in DOE's policy on sponsored away-from-reactor (AFR] storage was basically a commitment to do what is required of it, with a plea for a clear unequivocal Federal policy .... The Nuclear Waste Policy Act of1982 has now provided that policy. The Nuclear Waste Policy Act defines public and private responsibilities for spent fuel storage and provides for a limited amount of federally-supported interim storage capacity. The Act also includes provisions for monitored retrievable storage facilities and for a research development and demonstration program for dry storage. The Commission believes that these provisions provide added assurance that safe independent onsite or offsite spent fuel storage will be available if needed. [References omitted] The policy set forth in the NWPA regarding interim storage remains in place. Therefore, the Commission's confidence remains unchanged. The only policy change affecting storage involves long-term storage in an MRS. The NWPAA sets schedule restrictions on an MRS by tying it to the repository siting and licensing schedule. These restrictions effectively delay implementation of an MRS. Consequently, its usefulness in providing storage capacity relief to utilities is likely to be lost. The NWPAA established a Monitored Retrievable Storage Review Commission tasked with preparing a report on the need for an MRS facility as part of the national nuclear waste management system (section 143(a)). In its November 51-SC-75 1989 report "Nuclear Waste: Is There a Need for Federal Interim Storage?", the MRS Commission reached the following conclusion: An MRS linked as provided in current law would not be justified, especially in light of uncertainties in the completion time for the repository. Consequently, the Commission does not recommend a linked MRS as required by current law and as proposed by DOE. In the November 1989 Reassessment Report, DOE stated that current linkages between the repository and MRS program make it impossible for the DOE to accept waste at an MRS facility on a schedule that is independent from that of the repository. Therefore, the DOE plans to work with the Congress to modify the current linkages between the repository and the MRS facility and to embark on an aggressive program to develop an integrated MRS facility for spent fuel. The DOE believes that if the linkages are modified, it is likely that waste acceptance at an MRS facility could begin by 1998 or soon thereafter.

Although the Commission's confidence in its 1984 Decision did not depend on the availability of an MRS facility, the possibility of such a facility, as provided for in the NWPA, was one way in which needed storage could be made available.

The NWPAA makes an MRS facility less likely by linking it to repository development, unless Congress is willing to modify these linkages. The potential impact of the uncertainty surrounding an MRS on the Commission's confidence is, however, more than compensated for by operational and planned spent fuel pool expansions and dry-storage investments by utilities themselves--developments that had not been made operational at the time of the original Waste Confidence Decision. Consequently, the current statutory restrictions that may make an MRS ineffective for timely storage capacity relief are of no consequence for the Commission's finding of confidence that adequate storage capacity will be made available if needed. Although the NWPAA limits the usefulness of an MRS by linking its availability to repository development, the Act does provide authorization for an MRS facility. The Commission has remained neutral since its 1984 Waste Confidence Decision with respect to the need for authorization of an MRS facility. The Commission does not consider the MRS essential to protect public health and safety. If any offsite storage capacity is required, utilities may make application for a license to store spent fuel at a new site. Consequently, while the NWPAA provision does affect MRS development and therefore can be said to be limiting, PART 51

STATEMENTS OF CONSIDERATION the Commission believes this should not affect its confidence in the availability of safe storage capacity.

V.B. Relevant Issues That Have Arisen since the Commission's Original Decision on Finding 5 DOE will probably not be able to begin operation of a repository before 2010 under current plans, and operation might begin somewhat later. Given progress to date on an MRS, the link between MRS facility construction and repository construction authorization established by the NWPAA, and the absence of other concrete DOE plans to store the spent fuel, it seems unlikely that DOE will meet the 1998 deadline for taking title to spent fuel. unless DOE is successful in its efforts to work with Congress to modify the linkages. (Under section 302(a)(5)(B) of the NWPA, " ... the Secretary, beginning not later than January 31, 1998, will dispose of the high-level radioactive waste or spent nuclear fuel [subject to disposal contracts].") This potential problem does not, however, affect the Commission's confidence that storage capacity will be made available as needed. The possibility of a dispute between DOE and utilities over the responsibility for providing spent fuel storage will not affect the public health and safety or the environment. Uncertainty as to contractual responsibilities raises questions concerning: (1) who will be responsible; (2) at what point in time responsibility for the spent fuel will be transferred; (3) how the fuel will be managed; (4) how the transfer of management responsibility from the utilities to DOE will take place; and (5) how the cost of DOE storage might differ, if at all, from utility storage. Utilities possessing spent fuel in storage under NRC licenses cannot abrogate their safety responsibilities, however. Until DOE can safely accept spent fuel, utilities or some other licensed entity will remain responsible for it. Estimates of the amount of spent fuel generated have continued to decline. At the time of the Commission's Decision, the Commission cited in Finding 5 the cumulative figure of 58,000 metric tons uranium of spent fuel generated in the year 2000 [See 49 FR 34658, p. 34697, August 31, 1984.) More recently, DOE estimated 40,200 metric tons the lower reference case which represents the conservative upper bound of commercial nuclear power growth [see "Integrated Data Base for 1989: Spent Fuel and Radioac.tive Waste Inventories, Projections, and Characteristics," DOE/ RW-0006, Rev. 5, November 1989). Although estimates may show an increase at some date well into the twenty-first century if licenses of some reactors are renewed or extended, this possibility does not affect t~e Commission's confidence in the availability of safe storage capacity until a repository is operational. The industry has made a general commitment to provide storage capacity, which could include away-from-reactor [AFR) storage capacity. To date, however, utilities have sought to meet storage capacity needs at their respective reactor sites. Thus, a new industry application for AFR storage remains only a potential option, which currently seems unnecessary and unlikely. Utilities have continued to add storage capacity by reracking spent fuel pools, and NRC expects continued reracking where it is physically possible and represents the least costly alternative. Advances in dry-storage technologies and utility plans both have a positive effect on NRC's confidence. At the time the Commission reached its original findings, dry storage of LWR spent fuel was, as yet, unlicensed under 10 CFR part 72, and DOE's dry-storage demonstrations in support of dry-cask ; storage were in progress at the Idaho National Engineering Laboratory (INEL). Today, DOE's demonstration efforts have been successful (See Godlewski, N. Z., "Spent Fuel Storage-An Update," Nuclear News, Vol. 30, No. 3, March 1987, pp. 47-52, at p. 47.) Dry storage has been licensed at three reactor sites, and three new applications are under review. Dry cask storage is licensed at Virginia Electric Power Company's Surry Power Station site (see License, SNM 2501 under Docket No. 72-2), and dry-concrete module and stainless-steel canister storage is licensed at Carolina Power and Light Company's (CP&L's) H. B. Robinson, Unit 2, site [see License SNM 2502, under Docket No. 72-3). A license was recently granted for a similar modular system at Duke Power Company's Oconee Nuclear Station site. New applications have been received in 1989 for CP&L's Brunswick site, the Baltimore Gas and Electric Company's Calvert Cliffs site, and in 1990 for Consumer Power Company's Palisades site. Applications are also expected for CP&L's Robinson 2 site [at another onsite location to allow for greater storage capacity) and Wisconsin Electric Power Company's Point Beach site. The Tennessee Valley Authority has indicated that it will apply for a licensed dry storage installation at its Sequoyah plant site. Thus, the successful demonstration by DOE of dry cask technology for various cask types at INEL, utilities' actions to forestall spent fuel storage capacity shortfalls, and the continuing sufficiency 51-SC-76 of the licensing record for the Commission to authorize increases in reactor storage capacity all strengthen the Commission's confidence in the availability of safe and environmentally sound spent fuel storage capacity. Renewal of reactor OLs will involve consideration of how additional spent fuel generated during the extended term of the license will be stored onsite or offsite. There will be sufficient time for construction and licensing of any additional storage capacity needed. In summary, the Commission finds no basis to change the Fifth Finding in its Waste Confidence Decision. Changes by the NWPAA, which may lessen the likelihood of an MRS facility, and the potential for some slippage in repository availability to the first quarter of the twenty-first century (see our discussion of Finding 2) are more than offset by the continued success of utilities in providing safe at-reactor-site storage capacity in reactor pools and their progress in providing independent onsite storage. Therefore, the Commission continues to find " ... reasonable assurance that safe independent onsite spent fuel storage or offsite spent fuel storage will be made available if such storage is needed." Dated at Rockville, Maryland, this 11th day of September 1990. For the Nuclear Regulatory Commission. Samuel J. Chilk, Secretary of the Commission. 57 FR 18388 Published 4/30/92 Effective 6/1/92 Uranium Enrichment Regulations See Part 40 Statements of Consideration 58 FR 7715 Published 2/9/93 Effective 7 /1 /93 Licenses and Radiation Safety Requirements for lrradiators See Part 36 Statements of Consideration 59 FR 48944 Published 9/23/94 Effective 10/24/94 Certification of Gaseous Diffusion Plants See Part 76 Statements of Consideration PART 51

STATEMENTS OF CONSIDERATION 60 FR 22461 Published 5/8/95 Effective 6/7 /95 Nuclear Power Plant License Renewal; Revisions See Part 54 Statements of Consideration 60 FR 24549 Published 5/9/95 Effective 5/9/95 Changes to NRG Addresses and Telephone Numbers See Part 2 Statements of Consideration 61 FR 9901 Published 3/12/96 Effective 3/12/96 Minor Correcting Amendments See Part 19 Statements of Consideration 61 FR 28467 Published 6/5/96 Effective date delayed to 9/5/96 Comment period extended to 8/5/96 10 CFR Part 51 RIN 3150-AD63 Environmental Review for Renewal of Nuclear Power Plant Operating Licenses
AGENCY: Nu~lear R~gulatory Commission.

ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRG) is amending its regulations regarding environmental protection regulations for domestic licensing and related regulatory functions to establish new requirements for the *environmental review of applications to renew the operating licenses of nuclear power plants. The amendment defines those* environmental impacts for which a generic analysis has been performed that will be adopted in plant-specific reviews for license renewal and those environmental impacts for which specific analyses are to be performed.

The amendment improves regulatory efficiency in environmental reviews for license renewal by drawing on the considerable experience of operating nuclear power reactors to generically assess many of the environmental impacts that are likely to be associated with license renewal. The amendment also eliminates consideration of the need for generating capacity and of utility economics'from the environmental reviews because these matters are under the regulatory jurisdiction of the States and are not necessary for the NRC's understanding of the environmental consequences of a license renewal decision. The increased regulatory efficiency will result in lower costs to both the

applicant in preparing a renewal application and to the NRG for_ reviewing plant-specific applications and better focus ofreview resources on significant case specific concerns.

The results should be a more focused and therefore a more effective NEPA review for each license renewal. The amendment will also provide the NRC with the flexibility to address unreviewed impacts at the site-specific stage of review and allow full consideration of the environmental impacts of license renewal. The NRC is soliciting public comment on this rule for a period of 30 days. In developing any comment specific attention should be given to the treatment of low-level waste storage and disposal impacts, the cumulative radiological .effects from the uranium fuel cycle, and the effects from the disposal of high-level waste and spent fuel. DATES: Absent a determination by*the NRC that the rule should.be modified, based on comments received, the final rule shall be effective on August 5,

1996. The comment period expires on July 5, 1996. ADDRESSES:

Send comments to: The Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555--0001, Attention: Docketing and Services Branch, or hand deliver comments to the Office of the Secretary, One White Flint North, 11555. Rockville Pike, Rockville, Maryland between 7:30 a.m. and 4:15 p.m. on Federal workdays. Copies of comments received and all documents cited in the supplementary information may be

examined at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC between.the hours of
7:45 a.m. and 4:15 p.ni. on Federal workdays.

FOR FURTHER INFORMATION CONTACT: Donald P. Cleary, Office of Nuclear Regulatory Research; U.S. Nuclear Regulatory Commission, Washington, DC 20555--0001, telephone: (301) 415-6263; e-mail DPC@nrc.gov. SUPPLEMENTARY INFORMATION: 51-SC-77 I. Introduction II. Rulemaking History Ill. Analysis of Public Comments A. Commenters B. Procedural Concerns 1. Public Participation and the Periodic Assessment of the Rule and GEIS 2. Economic Costs arid-Cost-Benefit Balancing

3. Need for Generating Capacity and Alternative Energy Sources C. Technical Concerns 1. Category and Impact Magnitude Definitions

2. Surface Water Quality 3. Aquatic Ecology 4. Groundwater Use and Quality 5. Terrestrial Ecology 6. Human Health 7. Socioeconomics

8. The Uranium Fuel Cycle and Solid Waste Management

9. Accidents.

10. Decommissioning

11. Need for Generating Capacity 12. Alternatives to Liconso Ronewnl 13. License Renewal Scenario 14. Environmental Justice

IV. Discussion ofR~gulatory Requirements A. General Requirements B. The Environmental.Report

1. Environmental Impacts of License Renewal 2. Consideration of Alternatives C. Supplemental Environmental Impact Statement

1. Public Scoping and Public Comments on the SEIS 2. Cornmission"s Analysis and Preliminary Recommendation

3. Final Supplemental Environmental Impaci Statement i:J. NEPA Review for Acti}}

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