ML20141N585

Draft Major Issues in Reactivation of Nuclear Power Plant Const Projects

ML20141N585
Person / Time
Issue date:	02/28/1986
From:	Spangler M Office of Nuclear Reactor Regulation
To:
Shared Package
ML20141N584	List: ML20141N583 ML20141N585
References
NUDOCS 8603170198
Download: ML20141N585 (48)

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i MAJOR ISSUES IN REACTIVATION OF NUCLEAR POWER PLANT CONSTRUCTION PROJECT 5 l

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Miller 8. Spangler l-

! Special Assistant for Policy Development Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commtssion  ;

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SUMMARY

This policy study is both exploratory and limited in scope. Its primary objective is to identify the principal issues requiring Of fice-level consideration in the event of reat-tivation of the construction of one or more of the nuclear power plants falling into two categories:

(1) LWR units issued a Construction Permit whose construction has been cancelled, and (2) LWR units whose construction has been deferred. The study scope is limited to identifying regulatory issues or questions deserving analysis rather than providing, at this time, answers or recommended actions. Five tasks are addressed: a tabulation of the status of 35 cancelled and 8 deferred LWR units; an identification of safety and environmental issues to be dealt with; an identification of regulatory or policy issues and needed information to determine the desirability of revising certain rules and policies; an identification of regulatory options and decision criteria; and an identification of decision considerations in determining staff requirements and organizational coordination for LWR reactivation policy and implementation efforts.

Regarding the last of these tasks, it is shown that a very large network of analytical factors and decision considerations are involved that imply large uncertainties in estimating the level and timing of staff requirements. Indeed, there is an interlocking catch-22 aspect of such a task. The particular choices made by the NRC of its regulatory options could have either a discouraging or encouraging impact on utility decisions to reactivate or not their cancelled or deferred LWR project and other decisions related to equipment preservation programs and design change options. The latter might involve a switch to a new standard plant design for some of the 22 cancelled projects that, for example, have less than $100 million of equipment onsite. Likewise, the actual utility decisions that will ultimately be made regarding reactivation options--and especially staff estimates of the aggregate effect of these decisions on NRC resource requirements--

is a significant factor conventionally used in making decisions on regulatory options.

Another factor involved in staff requirement estimates is the number and type of safety and environmental ist,ues resulting from significant new information since the units were deferred or cancelled and the possibility of site-related developments not anticipated that would contribute to staf f re-review in the event of project reactivation.

On the other hand, utilities facing reactivation decisions must cope with the uncertain-ties of the cost impacts of dealing with these issues as well as the possible construc-tion delays and potential cost imp 1(cations of NRC's severe accident research programs and regulatory decisions yet to be made regarding policies, rules, and the resolutions of safety issues currently being addressed by the staff. Sections 6.3 and 6.4 address these utility concerns over the stability and predictability of NRC's regulatory frame-work and outcome of research programs as these might affect investor confidence that weighs heavily in any reactivation decisions.

Although there is a complexfty of regulatory options and research projects affecting these cost uncertainties, a conclusion of this study is that, on balance, these programs and options have substantial potential for irproving regulatory stability and predicta-bility, especially as these might hold implications for cost escalation of reactivated LWR projects. Many of these research offorts and regulatory developments and options are, or would be, directed to the objective of improving regulatory stability and pre-dictability as well as limiting future design changes to those which yield a substantial fil

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increase in the overall prctection of the public health and safety and that meet the cost-effectiveness Guidelines. Moreover,criteria of NRC's new Backfit Rule and revised Regulatory Analysis staff efforts are in progress to review technical specifications and other regulatory rules or procedures to eliminate or modify those which make only marginal, if any, contribution to safety.

The preliminary, reconnaissance-level information assembled on the status of deferred or cancelled plants indicates that all eight deferred units and thirteen cancelled plants have major amounts of equipment onsite (i.e., in excess of $100 million acquisition value per unit).

NRC needs to define its policies for equipment preservation programs for such equipment to maintain quality assurances in the event of their use in future pro-ject reactivations or resale to other project use. All of the sites for deferred sinits and a majority of sites of the cancelled units are still available for future construc-tion of a nuclear, coal, or other baseload unit, although some have undesirable features for a coal-fired plant.

Most importantly, the status review indicated that the owners of nearly all deferred and cancelled units are beset by serious financial difficulties and need-for plant issues that clearly must improve before more than a handful of these can reach a decision to reactivate these projects. Accordingly, it would be premature for NRC to make a survey of utility reactivation intentions as a basis for LWR reactivation policy decisions. ,

Improvements in the political and investment cifmate for utility reactivation decisions will depend, in large measure, on the cooperative and separate initiatives of industry and government, only some of which have been identified in this preliminary study effort.

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CONTENTS

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SUMMARY

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1. STUDY SCOPE AND 08JECTIVES.................................... 1 2.

DETERMINANTS OF STAFF REQUIREMENTS AND ORGANIZATIONAL C00RDINATION................................... 1 2.1 The Number of Reactivations and Their Timing.............

2. 2 2 Reactivation Options chosen and Plan Status Considerations................t 2 2.3 The Number and Difficulty of. Technical....................and Environmental 2.4 NRC's Choice of Issues.....................................

Regulatory Options for 2 Treating Reactivation Issues...............

2. 5 Organizational Coordination.............................. 2

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3. STATUS AND OUTLOOK FACTORS OF CANCE DEFERRED PLANTS.....................LLED .......................... OR 2 3.1 Status Summary of Units with Cancelled Construction......

3.2 4 Status Summary of Units with Deferred Construction........ 10 4.

IDENTIFICATION OF POSSIBLY RELEVANT SAFETY AND ENVIRONMENTAL ISSUES ARISING FROM SIGNIFICANT NEW INFORMATION.............. 16 4.1 Safety Issues...................... .................... 16 4.2 Environmental Issues............... ............. ...... 20 5.

IDENTIFICATION OF REGULATORY OR POLICY ISSUES OF POSSIBLE RELEVANCE TO THE ADEQUACY OF EXISTING RULES AND P THEIR NEED FOR REVISION..........................OLICIES ............ OR 22 6.

REGULATORY OPTIONS AND DECISION CRITERIA..................... 25 6.1 Decision Criteria and Regulatory Purposes Guiding LWR Reactivation Policy Development......................... 25 6.2 Regulatory Options for Dealing with Reactivation Issues. 29 6.3 Potential Impact of Research Pro of Regulatory 0ptions. . . . .. . . . . . . grams on the Choice

....................... 35 6.4 The Image Problem of Identifying and Choosing between Alternatives to Deal More Effectively with a Complex i ty o f Regula to ry I s s ue s. . . . . . . . . . . . . . . . . . . . . . . . . 38 7.

REFERENCES................................................... 42 v

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FIGURES Floure 1

h A conceptual model of interrelationships between factors and events impacting utility choice of LWR reactivation options and NRC choice of regulatory options and staffing requirements............................................. 3 2

A preliminary framework of industry and government initia-tives that would serve an' implicit national goal of improv-ing the U.S. outlook for maturation progress in nuclear reactor safety technolo industry growth........gy and the resumption of

.................................. 41 TABLES Table 1

Status of LWRs granted Construction Permits whose construction has been cance11ed..........................

5 2

Status of LWRs granted Construction Permits whose cons truction has been de ferred. . . . . . . . . . . . . . . . . . . . . . . . .11 3

Documentary sources of information to understand the nature and im experience... . .. .portance of new LWR safety

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MAJOR ISSUES IN REACTIVATION OF NUCLEAR POWER PLANT CONSTRUCTION PROJECi$ _

Miller B. Spangler Special Assistant for Policy Development Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

1. STUDY SCOPE AND OBJECTIVES This policy study is both exploratory and limited in scope. Its primary objective is to identify the principal issues requiring Office-level consideration in the event of reactivation two categories:of the construction of one or more of the nuclear power plants falling into cancelled, and (2)(1) LWR LWR units units issued whose a Construction construction has beenPermit whose construction has been deferred. The study scope is limited to identifying regulatory issues or questions deserving analysts rather than providing, at this time, tasks are regarded answers or recommended actions. Specifically, the following as appropriate:

(1) A preliminary tabulation of individual nuclear units whose construction has been deferred or cancelled, but with construction quality preserved, and their status relative to current NRC policies and regulations.

(2) An identification of special safety or envirormental issues that might require (updated) analyses and staff review (or re review), the need for which arises because of significant new information.

(3) An identification of regulatory or policy issues and related information useful to a determination of whether existing rules or policies are adequate or whether new or changed ones are desirable, especially regarding the dif ferential situation of a reactivated extension CP application or updated re-review.versus a still valid CP, but one possibly in need of (4) An identification of regulatory options for dealing with the above issues and de-cision criteria by which to judge their relative merits.

(5) An identification requirefrents of decision considerations and organizational coordination. essential to a determination of staff One important aspect of this study is the identification of those determinants of NRC staff requirements and organization coordination needs associated with the reactivation of nuclear power plant construction projects.

As will be seen in the discussions to follow, the level'of staff requirements is one of the decision criteria by which the regulatory options of item 4 above should be evaluated. Staff requirements, in turn, are related to many of the factors addressed in performing study tasks 1-3.

2.

DETERMINANTS OF STAFF REQUIREMENTS AND ORGANIZAfl0N COORDINATION Because of its interrelationships with other task elements, the identification of decision considerations essential to a determination of staff requirements and organizational coordination (task 5) occupies a focal role in the study design,

The following is an identification of the relevant decision considerations or deter-minant factors of staffing or organizational requirements essential to the effective performance of NRC's regulatory responsibilities associated with the reactivation of nuclear power plant construction of cancelled or deferred units.

'2.1 The Number of Reactivations and Their Timing To a significant degree, NRC stalf requirements will be greater (all other things being equal) the larger the number of LWR units being reactivated. The timing of any reac '

tivations is also important to NRC decisions on staffing requirements which must be allocated over future years and budgets. For example, if analysis should reveal that

.few reactivations will likely occur over the next 3 or 5 years among the population of 35 units with construction cancelled and the 8 units with deferred construction, this would hold considerable import for staffing decisions even though the number of reacti-vations over a longer period may prove to be far greater.

2.2 Reactivation Options Chosen and Plant Status Considerations The number and timing of LWR reactivations is determined by the utilities' choices of reactivation options. As seen in Figure 1, these choices, in turn, are impacted by a host of factors and events. Not the least of these is what choices NRC makes in the way of regulatory options that may tend to encourage or discourage a given utility's dect-sion to reactivate, or not, nuclear plant construction and the timing of such decisions.

Also important is the status of cancelled or deferred plants, including such investment-

.related considerations as financial factors, electricity demand growth (need for plant),

other regulatory factors at the state or federal (non-NRC) level, socio political fac-tors, and near- or long-term advances in technology and scientific knowledge. A number of these factors are discussed below in Section 3.

2.3 The Number and Difficulty of Safety and Environmental Issues In addition to the number and timing of LWR reactivations, NRC staff requirements (including types of necded skills) will be affected by the nueber and difficulty of safety and environmental issues requiring staff review in the event of LWR reactivation. This is discussed in Section 4.

2.4 NRC's Choice of Regulatory Options for Treating Reactivation Issues The choice of regulatory options for the of fective treatment of reactivation issues includes, among other cost-benefit considerations, the costs of regulatory staf f (or contractor technical assistance) associated with the various options (see Section 6).

By the t'.me token, the particular choice actually made will then impose the estimated staff requirements.

2.5 Organizational Coordination It is envisioned that most of the regulatory options will require coordinated efforts of 5ne Office of Nuclear Reactor Regulation, the Office of the Executive Legal Director, and the Office of Inspection and Enforcement. For those options involving rule changes or standards development, the Office of Nuclear Regulatory Research and several offices of the Commissioners' staff may also be involved.

3. STATUS AND OUTLOOK FACTORS OF CANCELLED OR DEFERRED PLANTS Direct contact was made with utility organizations that are the principal owners of the canceiled or deferred plants to obtain reconnaissance level information on 4,he status of i

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--A conceptual rnodel of interrelationships between factors and events impacting utility choice Of LWR reactivation options and NRC choice Of regulatory Options and staffing requirements.

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_these units and certain outlook factors of possible relevance to any reactivation dect-

sfons.

.quite clear there was zero or miniscule possibility of constructio undetermined) possibilities might be reflected by the factual data For example, negative fact' ors in the outlook for project reactivation would include:

the sale of a site or. its. commitment to use in the construction of other buildings or major or refusefacilities; the planned the installation conversion of peaking units of the' unit to other fuels such as coal, gas, of major a;nd minor equipment orand non-nuclear units owned by the company.

itstheuse cancellation, as pa;rts, sale,oror scrapping spare parts, fo equipment is still available for project reactivationHowever, in those cases where very ifmited the possibility might not be

, foreclosed that, depending on the strengthening energy, de Certification in accordance with NRC's new Severe Acciden Accordingly, preifminary information was obtained of the site's availability for a future plant installation,'either coal, nuclear, or some other type of unit.

3.1 Status Summary of Units with Cancelled Construction There are presently 35 LWR units on 21 different sites whose construction has been cancelled.

have Construction Permits.At least three of these units (Marble Hill 1 & 2 an

- limited) financial data on the status of these 35 units is shown in Table 1.

It should

'be noted that assembled which inhas thisnot exploratory effort basically reconnaissance level information was yet been validated.

construction progress was obtained initially from internal NRC sources. However,Data for on the firs

• the overallpurposes of this study, precision of data is not essential to provide a general perspective. ,

. imprecise since a variety of reference points were used for(1)different the units date of the. press announcement of construction cancellatior., (2) the date of the utility's NRC letter tonotification provided the NRC of requesting cancellation of the CP, or (3) the date at which CP cancellation.

ranged from a few days or months to upwards of a year or more.The separation of these resp

'The'plant's

.the principal relevance of the CP issue date is that it provides a crude indication of vintage.

the most recent is 1-4-79 The(Jamesport oldest CP1 & issue 2 . date for these units is 10-27-72 (Zimmer) and of units is January 1977, or nine years ago). The mean CP issue date for this population the engineering and safety design features is at least 9 and possibly 11 years old considering that the major items of equipment such as the NSSS and TG would have been ordered several years before the CP issue date.

The date of construction cancellation holds several possible levels of significance:

indicates the length of time up to the present during which mainte of concern in preserving quality assurance in the utilization of the remaining equipment in the event of project reactivation, and (3) in some cases of longer duration (since cancellation), it suggests an adverse financial impact on the utility in the delay of investment the recovery plant becomes in a rate base increase--an action which of ten does not begin until operational.

To serve these purposes, the earliest reference date of

-_ cancellation (i.e., the press announcement date) would be the most meaningful. For the population of 35 units, the oldest construction cancellation date is 9-12-74 (Vogtle 3 & 4) and the most recent is 8-29-84 (Yellow Creek 1 & 2 and Hartsville Al & A2 1The titles and dates of NUREG citations are found at the end of the paper .

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Table 3. StatWs of lists granted Construction Permits whose construction has been cancelled * '

(8ased on reconnaissance level information; data not validated)

Reactor Reactor CP Cancel

(*"',. Oluip't Ordered Staeus of Onsite f quetament oc Mtis Reactor 1) nit r Disposed Site Disposi*n PUC or State-level issues / Actions Tyoe Vendor issued Date va luel '

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N155 l IG ] Det bMIIlions mintenance Pro-)rae itC Monit- or future Use-oring At Time of Cancel Vogtle 3 PWR We 6-2e-74 0 No equipment Status Change 9-42 74 NA NA NA Available vogtle 4 PWR We 0 WIP 6- 2s- 74 9-82-74 delivered NA NA Avallebte leone Surry 3 rwu saw u NA Nfr glor.e Sorry 4 tpro-33 3 48-,, No equip.ent NA NA PWR S&W 82-10-74 2-88-77 0 delivered IIA sea Available Nfr, finenesal None NA NA leerth Anna 3 PWR 56W 7 26-74 6-22-83 4 Equipment sold NA NA Austlable NFP. financ ial Nome flerth Anna 4 PWR B&W P ?b74 81-25-30 0 or scrapped lea NA Available lef P, finenClel Mone Jamesport I M4 NA rWR We 3-4-79 t to-se 0 sto equipment NA NA Available lefP. f inanc ial None

_Jamessort 2 NA PWR We 1-4-79 8 19-se 0 4,ll,,r,4 #A NA Available lef P. financial leone Nrris 2 PWR We t-27 70 82-te-a3 4C ten Avattable Nfr. financial stone N rris 3 FI Medium Soutine No PWR We 8 27 78 32-10-41 If y1 Medtwa Avellable.F2 Infr leone N rris 4 PWR We I-27-78 Soutine No Avellable,F2 Nfr II-88-al IC y1 Nedium tenne Cherokee I PWR C-E Soutine tio Available F2 NfP 82-30-77 4-29-43 18 Equipment sold lea NA Name Cherokee 2 PWR C-E 32 30-87 a l-2-s

• O 18 4 Sold NfP lea or scrapped #4 #4 Sold Cherokee 3 PWR C-t 82-36-77 33-2-a2 0 NA NA NfP NA laA NA Sold Nrble itill 1 PWR We 4-4-Pe 12 38-es SSc 90% of major Large Nfs tea Marble Mill 2 F3 F2 Undet'd, F4 PWR We 4-4-70 82-34-43 35C equip 8 e enette Larne F3 Financial rate, and F5 le tF-4 PWR 86W F3 Undet'd. F4 NfP tesues 2-24-F8 t-22-82 yg tarne FS Ulf-5 PWR C-E 7-3 1- 78 s-22-42 24 Soutine peo 00E t. ease '

ser & 88menetal toewes 46 76 large leone lyrone 1 FUR We #tateel Me F2 ofP & finemetal taseee ib27-77 P-24-F9 0 R R R NA NA Isone Sterling I PWR We e.26-tr S-*s-se O No equip't del, NA Avellable,FS NfP, F9 Isome WA NA tri

.forted River PWR C-E F-10-73 11-4-e0 5 R R R NA NA NA Avellable lef t FIO '

]' Callansay 2 PWR We 4-16 76 10-9-84 1 R R Medium NA evettable.Fil Financial /TNI-2.NfP leone Sailly 1 SWR CE s-2-74 R Routtee No Avellable NfP Nope Creet 2 s-en I B R R Small SP loose 3 BWR CE 8 8 74 82-29-88 18 iso Undeterdd,F12 Financial tesues Fll Citaton 2 y14 Small NA Iso Unavail..FIS i

BWR GE 2-24-76 10-14-43 gc R Fineaclet 6 rase seauce seer,e j River Send 2 R F16 small NA 90A Avellebte MfP SWR CE 3-25 77 t 29-se Ic F12 R F12 Medlue Nome 4 lisovr BWR CE 10-27 72 8 21-44 get Soutine les Avellebte NfP, financial 6 tete Nome Phipps Send I SWR FRS FIS FIS Large.Fl9 NA NA Coat convesales CE t-36-7e 2-toes 29 30 4 NA Phipps Send 2 R R F20 Large $P, soutine tha Up for sale NgP SWR CE l-36-Fo 2-16-83 5 R 3 F20 soone l Tellow Creek I BWR C-E larme SP. Routtn, geo Up for sale NfP None 84-29-74 s-29-84 35 Large

Teltow Creet ? BWR C-E 34-29-7e s-29-se 3 SP, Routine sao Available Nir leone N rtsville Al BWR CE s-9-33 6-29-44 44 hae Sr. Routine leo Available NfP None 1Arge Routine N rtsville A2 SWR CE s-9-pt 8-29-e4 34 Large peutSne

$so Available NfP None

, N rtsville SI SWR CE s.9-77 F22-4 3 12 Latre Iso Available IsfP leone j liertsville 82 SWR s-9-PF Reuttne geo Avellable NfP CE F22-83 2 IJ rge Isone Reuttne No Available sofP Name i Legend t NS$5 - Ihsclear Steen Supply System; TC - Turbine-Ceneraterg Oct - Construction lenteriale and tyulpmeet; lea - Isot Applicable; SP - Site- stab -

b 111:sttee Plas; 3 - Resold, cance!!ad before delivery, or scrapped; PUC - Public Utility Coasteeton; Nf P - Iseed for Plant tesue; F1 - Pootnote 1*

} Purchase pricel escludes installattom or estatenance costo (large - over $100 million; medium - $10 to 100 million; emell - less than $10 alliten).

OL applicattom received.

d Avattable means there are sie piene to sell the ette and na known constrainto to its ultimate use for a coat or nuclear plant addition .

F 5ee mest page for additional footnotes.

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Additional Notes for Table 1 FI-For Harris-2, 98% of NSSS delivered to site, but euch less for units 3 & 4 (reactor vessel and internals, piping, pumps, etc.); yard sale yielded little sales; TGs purchased for all units and stored offsite in Penna, and some piece of rotors sold.

F2- Site use for a possible coal plant addition may be feasible if plant would not exceed regional air pollution limits (already pressed) under EPA bubble concept; suitability of site for ash ponds not studied.

F3- About 10-15% of onsite equipment for Unit 2 sold and removed under Investment Recovery Program of PS1; protective maintenance of remaining equipment monitored by onsite !&E inspector until a year ago.

F4- One option under consideration is sale of site / equipment to a consortium for completion of Unit I under a plan requiring State legislative action; site unsuitable for baseload coal plant since rocky foundation infeasible for ash pond drainage; site could be used for peaking units (gas or oil).

FS- Investment recovery by PSI threaten (d by adverse Indiana Supreme Court ruttng agatnst rate-base decision by the PUC to permit investment recovery of cancelled Bailly plant, a ruling which may be challenged at the U.S. Supreme Court level.

F6-The combined acquisition value of delivered equipment (including some software servaces) for k%P-4 and kHP-5 is $558 million, of which $95 million (acquisition value) was sold at a recovery price of $20 million; the remaining equipment is onsite.

F7 A future coal-fired plant is not a practical option for the BHP-5 sate and is not being considered for the future use of this site; they do not foresee appreciable growth in electricity site.

demand over next 20 years; kNP-5 is a siamese twin to kVP-3 at the Satsop F8- Site for Tyrone-1 is regarded ideal for nuclear plant and not so good for coal with inadequate rail service.

F9- At time of cancellation, unable to obtain permit f rom State of Wisconsin because of PUC evaluation of inadequate need; part of Northern States Power Co. service area is in Wisconsic per a wholly owned subsidiary there.

F10- Constru: tion Permit for Sterling-1 was retracted by State of New York because of changed evaluation of need by the PUC; it is now forecasted that a baseload plant addition wall be needed by 1995.

FIl- A study in progress for installing either a coal plant or a waste recovery (refuse combustion) plant at the Forked River site with resolution undetermined arisang, in part, from NfP and financial issues.

F12- Bailly site is inadequate in size for adding a coal plant, but could accommodate peaking units in addition to the coal unit already onsite; no interest in reviving a r.uclear unit for the site.

F13- Possible appeal to U.S. Supreme Court for a reversal of Indiana Supreme Court decision denying investment recovery in the utility's rate base for abandoned nuclear construction work.

F14- Practically all of the equipment delivered to the Hope Creek-2 site was either scrapped, sold or used for spare parts with less than $5 million (acquisition value) remaining to be sold; the generator was used in the Sales plant sod the reactor pressure vessel was cut up and scrapped.

F15- The part of the site designated for Hope Creek-2 is now occupied by support service buildings and therefore site is unavailable for adding a nuclear plant in the future; a coal plant addition could be accommodated at the site if adjacent land is purchased.

F16- The TG and NSSS for Clinton-2 was ordered but cancelled and never delivered; such of tbc remaining equipment delivered to site is warehoused for use as spare parts for Unit !.

F17- Major components of the River Bend-2 NSSS being housed at the site (inerted reactor pressure vessel, pumps, motors, etc.) with no plan to sell; TG not fabricated; substantial quantities of CME warehoused both onsite and offsite.

Fig- Major equipment for Zimmer remains onsite; however, control rods and nuclear fuel removed from site and some equipment sold; the main steseline was cut and capped and the reactor building (with substantial equipment inside) was sealed and written off for tax purposes.

F19- Sealed off equipment in reactor building has large, but unrecoverable value since remainder of plant is being converted to coal use.

F20- The reactor vessel for Phipps Bend I & 2 was cut up and sold for scrap; other equipment delivered to site was either scrapped or will be scrapped.

The mean construction cancellation date for the 35 units is December 1982 or about 3 years ago. This is a conservative estimate since, as noted above, many of the

' cancellation dates recorded in Table 1 fall some months after the utility's board decision and pubile announcement. Moreover, it is known that, in some instances, the actual construction work -stoppage occurred months, or even years, before the board's

' decision to cancel. However, in at least one instance (WNP-3), construction was resumed af ter the formal decision to cancel was made in order to complete the enclosure of the containment building as a prudent financial decision in the protection from weather elements of delivered onsite equipment located in the containment building.

Moreover, the avoidance of cancellation penalties involved.with construction contracts contributed to this decision.

Regarding the data on percent construction progress, the. data of Table 1 represents the current status as obtained from utility sources. This percentage sometimes differed

! significantly (of ten higher) compared to other available data. For 17 of the 35 units,.

the construction progress was 0 to 04. The remaining 18 units have construction pro-gress ranging from 3% (WNP-5 and Yellow Creek 2) to 98% (Zimmer), with a mean completion rate of 23%. Eliminating thc Zimmer plant from these calculations (since it is being converted to a coal-fired plant) would yield a mean completion rate of 20% for the re-maining 17 units. It is also significant that 8 of the 35 units are noted in Table 1 as having OL appilcations received by the NRC. Five of these eight have less than 5%

construction progress.

The next set of information in Table 1 relates to the status of the equipment that was ordered for the 35 units and its disposal or retention. Both precise and detailed data on the status of this equipment is not readily available. However, some useful, though crude, insights were obtained from reconnaissance level information received from utility sources responsible for the equipment's care or disposal. At this preliminary  !

stage of investigation (and to facilitate response), only broadgage information was requested on three categories of equipment: the nuclear steam supply system (NSSS);

the turbine generator system (TG); and other construction materials and equipment (CME).

For seven of the units, the contracts were cancelled sufficiently early so that no equipment was delivered to the site. Moreover, such equipment as was delivered to the site was either resold, scrapped or used as parts, or spare parts, for other nuclear or non-nuclear plants owned by the utility for an additional seven units. Consequently, no equipment or materials assigned to the cancelled unit remains on the site for 14 of the 35 units.

For the remaining 21 units, the situations are so variable that their status is dif fi-cult to summarize. Indeed, for the 8 TVA units, only the sketchlest of information is available at this time. However, for some of their units with cancelled construction, it is known that substantial equipment remains onsite, while others have been

. scrapped, used as spare parts, and some was sold. Of the 13 remaining units having equipment onsite, perhaps a brief description of a few diverse situations will put into perspective the spectrum of status situations. Marble Hill units 1 and 2 (which are 55 and 35% complete) are reported to have close to 90% of the major equipment onsite that is.needed to complete the project. About 10 to 15% of the equipment (all from unit 2) has been sold and removed from the site.

Another interesting illustration is provided by Harris-2, 3 & 4. For the NSSS, 98% of this equipment for unit 2 (4% overall construction) was delivered to the site. Regarding i units 3 and 4 (each, 1% construction) delivery was made of the reactor pressure vessels and internals plus certain piping, safety injection pumps and fuel handling equipment.

The pressure vessels for all 3 units have been inerted using nitrogen. A yard sale was conducted, but not much equipment was sold. The turbine generators were purchased for all 3 units and are stored offsite in a Pennsylvania warehouse. Some pieces of the rotors have been sold.

A third example is the River Bend-2 unit whose construction was only 1*4 complete at the time of cancellation. Some major equipment was purchased and delivered to the site.

The inerted reactor pressure vessels are housed at the site. The turbine generator was not delivered to the site. Some other construction materials and equipment were delivered and housed at the site and others are being stored in the company's offsite warehouses. This includes a certain amount of pipe, reactor circulation pumps, emer-gency diesel generators, etc.

The next subject of reconnaissance level information is the maintenance status of onsite equipment or materials. Where there is no longer any materials onsite (or i

offsite, for that matter) still assigned to the unit whose construction was cancelled, concern for maintenance is mooted and, hence, designated in Table 1 as "not applicable" (NA). As noted above, 14 of the 35 units bear this designation. For the remaining 21 units, there is concern for the appropriateness of equipment maintenance to assure preservation of construction quality in the event construction is reactivated. A number of utilities are quite clear about their intenticns not to resume construction of the cancelled units. It is inappropriate in this status report to perform any analyses or speculations en reactivation likelihoods for each individual unit. Accordingly, only factual information on the historic or current status involving maintenance programs for these units is reported.

One crude indication of the magnitude of the maintenance problem is reflected in the acquisition value of equipment or materials assigned to the unit that still remains onsite or in offsite warehouses, as the case may be. Warehousing of this equipment is, of course, one type of important protective measure of value to the dimensioning of the concern for quality assurance of equipment maintenance, as noted above regarding the completion of the containment building for WNP-3. As noted in Table 1, only crude (order-of-atagnitude) categories were used to express the valse of equipment remaining onsite or warehoused offsite. The designation "large" was used for an acquisition value in excess of $100 million; " medium" was used for values between $10 and 100 million; and "small," for values less than 510 million. In all cases, the value is the purchase price or acquisition value and does not include installation or maintenance costs. In some cases, a significant, but not overwhelming, component of this purchase value includes scftware services such as engineering detail design work, equipment manuals or handbooks, etc. Of the 21 units for which maintenance information is relevant, the value of remaining equipment was categorized as small for 3 units, medium for 5 units, and large for 13 units.

As shown in Table 1, no I&E monitoring is done regarding the maintenance status of these 21 units having onsite equipment since their cancellation relieves NRC of any regulatory authority for such monitoring. However, reconnaissance level information reveals that routine maintenance practice of the utilities is being used to protect against deterioration of those onsite materials that have not been cutup for scrap.

Validation of the reliability of this information would, of course, require field visits to the pertinent sites and offsite warehouses.

Factual information is also relevant to a reactivation study regarding the site disposi-tion or known plans or constraints, if any, regarding the future use of the site for constructing new generating units (coal, nuclear, or other fuels) in lieu of reactivat-ing the presently cancelled units that would make use of the remaining equipment dis-cussed above. For 5 units, this latter possibility has been mooted by actions taken by the utility. The Cherokee site has already been sold for housing developments that negates quite definitely the possibility that the 3 cancelled units could be revitalized as a construction project at this site. A decision has been made to convert the 98*. com-pleted Zimmer plant to a coal-fired generating unit that would nullify a nuclear unit being completed at this site. The fifth unit for which a different commitment has bee 9

, t.

, p _

made affecting the future use of the site is Hope Creek-2. Here, the space originally dedicated for this unit adjacent to Unit-1 is now occupied by permanently constructed -

buildings that provide support services for the company's operation of Unit-1. It is >

'noted that the remaining open space on the site would be suitable for adding a coal-

fired plant only if adjacent land is purchased.

A different kind of situation is the Phipps Bend site which, according to a' media '

report.2 TVA' officials have decided to put up for sale. Should this sale be consummated,

' the site would.no longer be available for building either a coal or nuclear plant.' The f Bailly site on which a coal-fired plant has long been located is too small to accommodate i a baseload coal plant addition, but it would be suitable for the construction of peaking units. If the Marble Hill units are not reactivated as nuclear units, the option does not exist to convert them to coal-fired units. The principal reason is that the site l

'has rocky soil that is unsuitable for an ash pond construction with the essential perco- '

lation or' drainage factors needed for a coal-fired plant. The site'would, however, be suitable for use as a switchyard station and the addition of peaking units.

The Satsop site (WNP-3 and WNP-5) is located on a hilly location and is not regarded as

. sultanle for adding a coal-fired plant because of poor access to railroad transportation for the large amount of coal that-would need to be transported to the site plus uncer-tainty over the transferability of water rights and discharge permits. The Harris site '

'(units 2, 3 & 4) would be suitable for the addition of a coal-fired plant, but there is  :

some concern whether such a plant (even with SO: scrubbers) might unacceptably press r against regional air quality limits under the EPA bubble concept because of the number of coal-fired plants already in the region. Similar. problems may arise at present nuclear sites in other regions, especially if more stringent air pollution standards might be applied to deal with acid rain problems (50s and NO*) or rising concern over greenhouse effects (CO3 ).8 q

The final status element dealt with in Table 1 is to identify and note whether there l had been any signiffrant change in those issues or actions of Public Utility Commissions  !

or other State agencies that, at least in part, contributed to the cance11ation.3 This limitation of scope does not imply that other causal factors ano their recent or future -

status changes are not important to any future analysis that would impact utility choice l of reactivation options. With reference to Figure 1, also of potential importance are the changes in status or the number and type of technical and environmental issues  !

(item 6); the change since cancellation,in NRC's current rules, policies and practices l (item 5); and the past or future changed in various causal factors or developmental '

events (items 7-12). -Theserepresentalmoreambitiousscopeofinquiryandanalysis '

than is suited to this exploratory effor*.  :

An inspection of Table 1 shows that a major factor contributing to cancellation of a  ;

. majority of the 35 units were issues of need for plant (NfP) relating to a drastic '

reduction in electricity demand growth projections that occurred due to the severe rise in energy prices and other inducements to energy conservation following the Mid-East oil  ;

i 1"TVA to sell Phipps Bend plant site," News Sentinel, Knoxville, TN, November 19, 1985. l 2U.S. Environmental Protection Agency, Potential Climatic Impacts of Increasino  !

Atmospheric CO, with Emphasis on Water Availability and Hydrology in the United States '

(Washington, D.C., April 1984); and J. Hoffman, D. Keyes, and J. Titus, Pro.fecting Future Sea Level Rise: Methodoloay.

Estimates to the Year 2100, and Research Needs. U.S. Environmental Protection Agency (October 24, 1983). y sEnergy Information Adminis Wation, Nuclear Plant Cancellations: Causes, Costs, and Conscauences, r,')E/EIA-0392{ U.S. Department of Energy, April 1983. {i

n embargo of October 1973.

regions with interconnected grids. Excessive reserve margins developed for all U.S. contine Although these margins are being worked down to varying degrees as a result of recent strengthening of electricity demand, nevertheless or late 1990s and, in some cases, after the turn of the century.fo The use of the descriptor "none" in Table 1 for status change regarding need for plant reflects only

.some that improvement the NfP issue still exists for near-term reactivation of these units, even though in the longer-term outlook may have occurred.

This is an area requir-ing careful analysis for each region and service area of those utilities with cancelled etc. is to be properly assessed. construction of units, if the reactivation outlook affecti The reduced electricity demand growth and consequent stretched-out schedules of both coal and nuclear plant construction contributed to financial stress on many utilities.

Commissions (PUCs), to institute appropriate rate measures tributed to, affected or failed to adequately, relieve financial stress on a number of the utilities. While some improvements on this score may have occurred for some of the utilities, in a number of cases financial or rate issues still remain to some degree or other that would discourage project reactivation.

It must be acknowledged that effort.

tory a detailed investigation of these status elements was not possible in this explora-possible inconsistencies of treatment or interpretation.The assembled data co 3.2 Status Summary of Units with Deferred Construction 8 units with deferred construction:In contrast with the large number of LWRs with c Seabrook-2, and Limerick-2. Grand Gulf-2, Perry-2, WNP-1, WNP-3, Midland-1 & 2 deferred plant since the Philadelphia Electric Company decided on to 23, 1985 December PUC to hold project costs to $3.2 billion.1 resume work on the mothballed Limerick-2 However, the Limerick-2 unit is retained in the list of deferred units of this exploratory study to illustrate the kird of decision considerations and circumstances that could affect future decisions of other units As seen in Table 2, the Construction Permit issue dates for the deferred LWRs range from 12-15-72 for Midland-1 & 2 to 4-11-78 for WNP-3. The mean CP issue date for these 8 units is April 1975 or approximately 11 years ago. This implies a vintage of techno-logical design that is almost 2 years older than the mean vintage of the 35 units with cancelled construction and a mean CP issue date of January 1977. However, the 8 deferred units have a mean deferral date of September 1983 (about 2.4 years ago) which is somewhat more 1ationrecent than the units.

of cancelled mean cancellation date (December 1982) computed above for the popu-As previously noted, the length of time since deferral or cancellation of construction holds implications for such concerns as the length of the equipment maintenance program and the likelihood that the construction labor force has been disbanded and moved from the region.

Of considerable relevance to the outlook for project reactivation is that the average construction completion for the 8 deferred units is 55%. This is substantially higher than the mean completion rate of 20% computed above for the 17 units with cancelled construction having construction progress of 3% or more at the time of cancellation.2 In each of the cases, the deferred units have all the major equipment ordered and onsite 1"PE to Resume Limerick Work," Philadelphia Inquirer, December 24 , 1985.

2This excludes the Zimmer plant because of the coal conversion plan for this unit .

Table 2. Status of tuts granted Construction refutts tshose construction his been deferred.

(Sesed on reconnaissante level information; data not validated)

Reactor Reactor CP M er- Constr*n @e Ww hus M We WWat w M PUC or State-level Issues or Actions Reactor Unit or Resold , Site Dispost'n Vendor issued ra Progress Tyoe Walee b psaintenance I&[ #eonit- or future Use

, Date (1) un5] gg l gng H illions Program oring At Isme of Mmal N m in Status Grated Gulf 2 godt GE S-4-74 8 15-84 35 (All major large EPP Yes Undeter- Rate issues '

mined ng ssues r e (em IIst fnt M g,gg g Seabrook 2 PWt We 8-3-76 9 25-84 24 C y Large (PP Yes Unde er- "'

, les change h sttel Iteed issue i

Limerick 2 gist 6-It- 7* I*'If Mc (All major Rate issues Accepted PUC GE g,,9, gpp y,, ,g g

. (*" I 1962 equipment on fleed $$ sue affer to complete site) isater sunalv tssue Perry 2 with constraints gigg gg g.3 pp p.gy.gg g4 d j Large S (PP Yes II A

l's sist a

' l5 g, g, site) f r e am' * * * * **' -*' -** y W'P- 3 C IA3 "*l

PWI B&W i n-N- rs 63 targe EPP Yes Site leased IIeed issue '

neeP-3 1961 Q}ft on from DOE Financing issues hcW *'

PWI C.[ 4 s3.pe 16C IA 88$0' Large [PP N rch equipment on Yes Undeter- IIeed issue sto change

' 1983 site) mined Financing issues nidland 1 (All major PWI 34W 47-15-77 85 c Rate issues Hearing in T-15-84 equimment on large EPP Yes Undeter. Financing issues Spring 1986 j g si t,iT .a.ea r , t on fue t o. A on rat, u.-

Midland 2 " 85 c (All major j '#'"

equipment on large 7 15 + [PP Yes Undeter.

g sitell , mined Ditto Ditto i

• t egend: W%s . Esclear steen seeply system. M - fortle-Genereter; Os . seestrestles notertels and testament; 8 A - het applicable; ord - orareed; a-peseld; Ipp ggstperet prosecootsee engre; put-pebles Ottitty Centssles.

j Nthese prete; eacleers testellelsen er ee6eteneere costs steege - esse 8848 es!!seet median - Sie se lee e888Best emit - less then Ste saltten).

"Ot appItsettee etteteed.

l by-F constrectlee escludes fas tlettes le ceason with perry-l.

' fell setority eresores to place fw Perry-1 and Perry-7.

'(p espered IF 3-94; esteettee ressested so 5-74-84 fee complettee by 17-5-89.

~

he espleed F-5-64; estenstem regsessed se 9-II-84 for eseptet tee by 7-5-89.

• ce-stretttoe of soit matted se e.e, e4

'tesele es se set tee enere constercetten, i

1 4

L

~

l and therefore the acquisition value is large (i.e. in excess of $100 million per unit).

'By the same token, equipment preservation programs,have been in effect for each deferred '

unit and each unit has received periodic inspection by NRC regional I&E inspectors to '

ascertain whether construction quality is being preserved.

The detailed nature of the equipment preservation this exploratory effort.programs and findings of NRC inspectors is beyond the scope of Regarding the remaining subjects covered in Table 2 such as site disposition and the status of state-level issues or actions, these are sufficiently varied as to invite case-by-casa discussion of the reconnaissance level information. As noted above, an interesting development in this regard is the recent decision to reactive Limerick-2.

At the time of deferral of this unit in early 1982, state-level issues involved the need-for plant question, rate base treatment of both units 1 and 2, and water supply issues involving permits that would enable the transfer of water to a constructed reservoir from the Delaware River Basin during peak periods of water availability.1 In addition, cost overrunstheofPennsylvania PUC was concerned about financial matters due to the large plant construction. These were aggravated in large part by the con-struction schedule stretchouts due to lagging electricity demand growth and the backfits imposed by the NRC resulting from the TMI Action Plan. The PUC based its recent deci-sions permitting an optional arrangement for Limerick-2 to be completed as a nuclear unit on a consultant study that estimated costs of four different options for meeting the region's future power needs in the early 1990s: (1) finishing Limerick-2, (2) ex-tending the life of older, non-nuclear plants, (3) butiding a new coal-fired plant, or (4) increasing the use of conservation plus purchased power.2 It should be noted in this regard that a rear view perspective that compares the completed costs of nuclear units.in recent years with the original cost estimates is far more disparate than a forward-looking estimate that compares total nuclear generating costs with those of a new coal-fired plant. Not only have coal-fired plants experienced large capital cost increases in recent years due to some of the same reasons for rising nuclear capital costs such as higher costs of labors and construction materials plus interest rates on borrowed were funds higher than the early 1970s when many of the deferred nuclear plants planned.

Also, coal-fired plants have increased capital costs because of the need to add anti pollution equipment and the recent trend and future prospects of fuel cost rises that appear to be greater for delivered coal in a number of regions than uranium fuel.8'4 An important factor in the decision of the Philadelphia Electric Company to accept the PUC constraints of an investment cap of $3.2 billion (plus an average annual plant capacity factor of 60%) relate to a no-strike agreement for the completion of Limerick-2.

The agreement, signed by Philadelphia Electric, the construction unions, and the Bechtel Corporation which is managing the construction project also limits wage increases to 5% per annum. The OL licensing hearings for Unit 2 were completed concur-rently with Unit I which is currently operating in its testing phase.

1" Extension asked on Limerick water," Philadelphia Inquirer, December 18, 1985.

2"PUC offers deal to build Limerick-2," Philadelphia Inquirer, December 6,1985.

8H. Kahn and J. Simon, The Resourceful Earth (1984).

4A Comparison of Future Costs of Nuclear and Coal-Fired Electricity, Study Group Report of the Committee on Financial Considerations, Atomic Industrial Forus, Inc. ,

November 1984.

13-A'quite contrasting situation tolthat of Limerick-? Arises f rom the nature of issues surrounding'the possible reactivation ' abandonment, or conversion options of the LWRs at Midland-112. . Construction was canc,elled for these units on July 15,19M following

' a 'long history of ~ 1itigation, technical, financial and rate base issues. Unit I was designed as a co generation plant that would supply electricity for Consumers Power's

. service area.and process steam to a DOW chemical plant on an adjacent site. The

. delayed construction schedules and cost overruns led ~to court action involving 00W's suit to withdraw from its contractual obligations.' Quite significant technical issues

.also arose over the settling of the auxiliary building at the Midland site and what

' design' changes or other measures are required to deal satisfactorily with the attendant foundation problems.

Both the legal and technical issues have exacerbated the financial

' and rate base' issues that have also affected deferred units elsewhere.

y The' DOW Corporation backed out of the contract on July 15, 1983 and one year later, the Consortium of' Users. in the Detroit area (including the GM Corporation) opposed the Utility's request before the PUC for favorable financing terms. These actions had a:

negative effect on the utility's ability to raise funds for project completion which currently is estimated to be 85% complete for each unit. There have been two hearings

' going oi, in parallel with the same Atomic Safety.and Licensing Board (ASLB) focusing on

. soil settlement issues plus such OL-contested issues as generic safety issues, cooling pond environmental analysis, cost-benefit analysis, severe accident analysis, emergency planning, and Quality Assurance / Quality Control analysis.

Faced with the possibility of filing for bankruptcy if rate relief for a recovery of at

1 east some of the $4.1 billion thus far invested in the Midland units, Consumers Power is engaged in a cost study of converting the units to either a natural gas combined-cycle plant or.a coal-fired plant.1 In the past, the Public Service Commission of Michigan has'sometimes allowed utilities to charge customers' the cort of abandoned pro-

'jects depending on the outcome of a prudency review of project management.2 The Construction Permits for both units have expired during 1984 and the utility has applied to NRC for their extension. The NRC staff reviewing this request has decided to wait on the outcome of the utility's decision expected this year whether or not to complete construction of the nuclear units or to exercise one of the other options under consideration.

Both need-for plant and financial issues have led to the deferral of the WNP-1 and WNP-3 units of the Washington Pubile Power Supply System (WPPSS), a joint operating agency and municipal

.ing systems.corporation composed of 19 Public Utility Districts and three municipal light-The Northwest Power Planning Council (NWPPC) and the Bonneville Power Administration (BPA) share in the responsibilities for projecting power requirements i' of the Northwest Region. The BPA is reported to have said the earliest that power from the two WNP units would be needed is 1994, but there is a one-in-three chance that one

-of the two LWRs would be needed before that date.3 A question being addressed is

.which of the 2 units should be completed first and whether the other should be abandoned

- or preserved for later reactivation. Complicating these planning decisions is the ability of WPPSS to sell bonds to finance completion of one or both of these units be-cause of the bond default on the cancelled WNP-4 and WNP-5 projects.

1" Consumers Power To Take Write-Off," New York Times, December 10, 1985.

2" Plan Allows PSC Advance Approval on Power Plants," Grand Rapids Press, November 19, 1985.

. 3"Which WPPSS plant will be finished?" Olympian, December 15, 1985.

+-

- - _ _ _ - - , _ - - - - - - - - _ - . - - - - , , . . _ . _ ~ .

a

~The. WNP-1 unit located on the Hanford reservation of.the DOE is 63% complete with esti- '

mated costs of of $2 million $1.4 per billion(Ibid.

month. for completion and current imothballing (maintenance) costs 4.

ern Washington (Satsop siteTTs)76* complete requiring an estimated $1.4 pletion and current maintenance costs of $3 million per month. (Ibid.)

. ment is working.on an equipment preservation program that would provide adequate main-The tenance

.for WNP-3.1for both units within an annual budget of $10 million for WNP-1 and $14 million

. Project WNP-3 is jointly owned by 4 private utilities (30%) and 8PA (70%),

whereas WNP-1~is controlled totally by 8PA.

.in.1983 challenging the mothballing of this unit.The ownership mix at WNP-3 led to a lawsuit An out-of-court settlement subs'e-

- quently guarantees the investor-owned utilities a fraction 1987 in exchange.for dropping the lawsuit.8 Socio political of 8PA power beginning in factors are quite different in the .immediate region of the two sites and may also affect any reactivation decisions that units.could follow 8PA's review in 1986 of the cost effectiveness of completing the two The bonds _for WNP-1 and WNP-3 are ensured by 8PA.

A substantial. fraction of the con-struction labor employed at both sites has dissipated and project reactivation at either site would require a build-up phase for recruiting and retraining of workers. Construc-tion unions have indicated a willingness to cooperate in actions essential to maintain improved control over construction costs in the event of project reactivation. The SPA

, forecasts of regional total electricity loads for the period 1985-2005 cover a wide range of estimates.. The low forecast is essentially zero growth, the medium forecast averages 1.31/yr. (compounded annually), and the high forecast averages 2.8K/yr.8- In meeting this growing demand from alternative energy sources (including conservation),'the WNP-1 and WNP-3 alternatives are assumed by the 8PA to " float," with construction occurr-ing when needed, with a 10% probability of involuntary termination (Ibid.) Further, the 18PA notes that although WNP-1 and -3 are " valuable options, legal and institutional 1 obstacles could impede their completion. Some obstacles have been cleared, but others still remain. SPA and others will continue efforts to remove those obstacles. However, because these efforts are not yet complete, the possibility that'these obstacles could prevent completion was directly factored into 8PA's analysis" (Ibid.).

Another of the deferred units is Seabrook-2 which was 24% complete at the time of defer-ral on September 25, 1984. The construction status report of September 30, 1985 indi-cated that 69% of the structural concrete was in place, the reactor pressure vessel was 2005restraints, ers, installed plus and 4% of the large bore process pipe and 35 of the large bore pipe hang-snubbers. The equipment preservation program is in effect which in-cludes long-term corrosion monitoring. At the time of deferral, Seabrook-2's outlook for completion was entangled in a number of issues involving PUC areas of responsibility or concern: rate base issues for investment recovery for unit I with long-term implications for unit 2; financial issues; and need-for plant issues. Currently, this list of concerns remains unchanged although some improvement in outlook may have occurred for each of these factors. The Chairman of Northeast Utilities has recently predicted that the New England Region may face a power shortfall (without new units being added by 1996), assuming 2.1%

annual growth in electricity demand nd with M111 stone-3, Seabrook-1 and some utility-owned

" hydroelectric projects all on Ifne

• If demand growth is as high as 35 per year, then he feels New England may have a shortage of baseload capacity as early as 1992, 1

1Information received from WPPSS by letter of January 30, 1986 from G. C. Sorensen to Miller Spangler.

a01vooian, December 15, 1985, o,p. c_it.

a1986 Resource Strateay, Draft Report, Bonneville Power Administration, U.S.

Department of Energy, November 1985.

• "New England Coming Up Short," The Enerny Daily, December 13, 1985, p. 4.

e Still another view is a prediction by John Sillin that New England demand will likely grow by 4-5% leading to additional new capacity requirements of 2600 MWe by 1993 if a 20% reserve margin is to be maintained. (Ibid.) Thus, a decision is possible within the next yearRegion England or so should as to whether be met the perceived needs for baseload capacity in the New for example, by reactivating Seabrook-2 or some other alternative supply sources po,ssibly including importation of energy from Canada.

The Perry-2 to that extends unit, November for which 30,construction 1991. is 44% complete, holds a Construction Permit were instrumental in the decision to defer construction.The Rate issues and lagging demand g revised completion date for Perry-2 is as yet undetermined but no other use is contemplated for the site. The PUC of Ohio has in progress an audit or prudency review of the utility's project manage-ment responsibilities as related to project cost escalations.1 A favorable outcome of this review could contribute significantly to a decision to reactivate construction of Perry-2 along with continued strengthening of demand and need for plant. In addition, there currently exists a transmission right-of-way issue, the details of which and cir-cuestances in this preliminary surrounding survey theofoutlook issues.for early resolution have not presently been obtained The final unit among the list of deferred LWRs is Grand Gulf-2. As may be expected, the completion of a second (or third) unit of a multi-unit site depends in part on the nature of the resolution of rate-base and other issues affecting the investment recovery for raisingand capital profitforrate outlook of the a subsequent first unit being completed with implications unit. Another kind of interrelationship between

. multiple-unit construction is that completion of the first unit often requires a substantial increase in consumer rates fer several years after the start of operations affecting the political climate for impesing further rate increases in a relatively few years for subsequent units. Thus, a longer construction delay eases somewhat this problem, a strategy that works at cross purposes for maintaining cost-control over time- ,

related cost overruns such as interest charges during construction (IOC) and escalation, or inflationary, costs for labor and construction materials (EOC). Moreover, the coming on line of a large block of baseload capacity of 1000 MWe or more with the first unit increases reserve margins for several years thus weakening the near-term need-for-growth.case for subsequent units unless there is a fairly good rate of regional demand plant These sorts of decisional considerations and uncertainties not only surround the question of whether or when Grand Gulf-2 will be reactivated but also a number of the other deferred units discussed above.

It was reported that the Boa *.1 of Middle South Utilities, Inc. , would meet on December 19, 1985 to reach a decision whether to cancel the Grand Gulf-2 unit and to initiate actions necessary before the Federal Energy Regulatory Commission and the courts to attempt to recover $927 million expenditures on unit 2 through rate relief over a period of years by means of charges to the system operating companies.2 Similar rate relief proceedings before state and local regulatory authorities to recover Grand Gulf-1 costs proved to be long and protracted. At its December meeting, the MSU Board decided to postpone for another year its decision regarding cancellation /

reactivated construction of Grand Gulf-2. On a positive note, it was reported that in

• December 1985 the Securities and Exchange Commission authorized Middle South Utilities, In ,, and its utility units to resume issuing bonds and short-term notes because the i u

1"PUC0 hires firm to study Perry cost," The Telearaph (Painesville, Ohio),

December 11, 1985.

k 8" Middle South may cut unit, make subsidiaries pay loss," Arkansas Democrat i (Little Rock), December 5, 1985.

company's financialfor significant concern prospects have improved.1 The need-for plant issue remains of the region.

4.

IDENTIFICATION OF POSSIBLY RELEVANT SAFETY AND ENVIRON SIGNIFICANT NEW INFORMATION 4.1 Safety Issues As noted in the discussione of Section 3, the mean CP issue date for the 35 LWR units ago.

with to arise.cancelled construction is about 9 years ago and for the There are basically two different sources of new safety information: (1) re-search programs directed to improved scientific information regarding safety phenomena i

analysis of the safety implications af ft.ilure/ systems reliab reactor experience in the United States and abroad. Regarding the second of these sou:ces, Table 3 (reproduced from NUREG-1070) exposes the wide variety of documen

'he nature of sources

. information and from reactor safety importance operating events.

of the documented failures of significance to understanding R?garding new safety information resulting from research programs, of considerable significance to the identification and assessment of importance of technical issues in the reactivation of LWR projects is the Severe Accident Research Program (SARP) . This ongoing program (NUREG-0900) has a? ready developed a large body of information to improve understanding generation of light water of the severe accident characteristics and risks of the current reactors.

The largest part of the SARP effort is dedicated to the better understanding of the physical phenomena of severe accidents and the staff's ability to model these phenomena in estimating severe accident behavior. This improved modeling capability is used in a number of ways, most notably in revised estimates of what radioactive materials are actually released to the environment in any identified severe accident sequence. The SARP also contains a substantial effort to examine all available data sources, especially the many detailed PRAs now available, to identify the important accident sequences for each class of reactor.

Especially since the THI-2 accident, there is a common interest in the international with a view to reducing severe accident risk. sharing of safety information obtain of Nuclear Installations (CSNI) of the Nuclear Energy Agency (NEA) has undertaken program to promote the sharing of technical information in this field. Areas of special concern include the thermal hydraulic behavior (both in-vessel and ex-vessel) of severe accident sequences, source term and fission product behavior, hydrogen and other gases, steam explosions, containment response, emergency instrumentation and equipment, and various aspects of short-term and long-term accident management.

Most of the backfits to operating plants or those engaged in the near-term operating licensing process (NTOLs) have resulted frna the THI-Action Plan (NUREG-0737). Over 85% of the Commission-approved TMI Action Plan items have now been completed. Thus, any reactivated construction of cancelled or deferred units could benefit from insights on technical issues involving desirable or required backfits made on behalf of the TMI Action Plan or other NRC requirements or notifications from the backfitting or design amendment experience of other units of similar vintage and design. In a number of instances for deferred or cancelled plants, the relevant experience has already been gained from the construction of a twin unit on the same site. On some of the more 1"SEC Authorizes Middle South Utilities and Units to Resume Financing Efforts "

Wall Street Journal, Eastern Edition, December 20, 1985. ,

Table 3. Documentary sources of information to understand the nature and importance of new LWR safety experience

• e Operating Reactors Licensing Actions Summary (NUREG-0748) e IE Bulletins (8 in 1983) e IE Information Notices (84 in 1983) e NRR Generic Letters (41 in 1983) e AE00 - review licensee event reports (about 4500 per year) e AE00 published case studies (several per year) e AE00 published engineering evaluations (30 in 1983) e AE00 published techical review reports (41 in 1983) e AE00 published Power Reactor Events Reports (6 per year, NUREG/8R-0051) e Report to Congress on Abnormal Occurrences, NUREG-0090 (12 per year) e NRC monthly status report to Congress (Sevill report) e Miscellaneous NUREGs; case related hearing testimonies, transcripts, etc.

e Plant-Specific PRAs e Foreign event information e INP0 SEE-IN Program (56 O&M reminders, 87 SERs, and 9 SOERs in 1983) e_ INPO NPRD system (40,000 component reports in 1983)

• Resulting from experience of failures in equipment and procedures during nuclear power plant construction, operation and maintenance.

Legend:

AE00- - The NRC Office of Analysis and Evaluation of Operational Data IE The NRC Office of Inspection and Enforcement INPO - Institute of Nuclear Power Operations NPRD - Nuclear Plant Reliability Data (System)

NRC -

Nuclear Regulatory Commission PRA -

Probabilistic Risk Assessment that mathematically quantifies an expected (or average) risk based on observed and calculated component and human failure rates and the anticipated consequences associated with these failures, which may occur either singly or in co.wbination SER -

Significant Event Report 50ER - Significant Operating Experience Report

r t

have been made or planned for.recently cancelled or deferred plants, some of the TMI It should be noted that many of the TMI Action Plan items involve changes in operating, maintenance, and accident management procedures TMI action items (NUREG-1070, p. 100).rather than hardware change which com Moreover, it should not be assumed that the outcomes of the Severe Accident Research Program will necessarily have negative safety implications requiring backfits. In the face of broad ranges of uncertainty surrounding PRAs or other risk estimates, there sometimes has been a tendency to use, in the interest of safety prudence, conservative modelling assumptions or sizeable margins of conservatism in equipment design.Thus, some of design the research requirements results that have may been lead to new requirements while still others may reduce overstated. Information regarding the Unresolved Safety Issue (A-49) involving the phenomena of Pressurized Thermal Shock (PTS) for certain PWR reactor vessels has had downgraded significance as a result of research effort, provided that the affected types of reactor vessels can be demonstrated to meet a so-called " Pressurized Thermal Shock Screening Criterion" which is related to the fracture resistance of these reactor vessels. Another important set of safety issues involvesprogression accident piping cracks, piping supports, and leak-before-break criteria in severe phenomena.

Recent operating events and research data in both the United States and foreign countries had led the NRC to initiate a comprehensive review of NRC requirements in the area of nuclear piping.

determine whether our current requirements should be modified.The goal of such a review is to This could lead to increasing requirements in some areas and decreasing requirements in others . The NRC Piping Review Committee work, covering the areas of BWR Stress Corrosion Cracking, Seismic Design, Potential for Pipe Breaks, and Dynamic Loads and Load Combinations, has been completed and reports on each major area have been pub 1fshed.8 Still another example of how research progress has aided NRC's regulatory approach in dealing with generic operating problems relates to General Design Criterion 4 as implemented in conjunction with the definition of a Loss-of-Coolant-Accident (LOCA).

In applying this regulation, a double ended guillotine break (DEGB) of the largest Core Cooling System (ECCS) reactor coolant pipe has been postulated as design basis and equipment qualification, as well as requiring protective devices (e.g., p,ipe whip restraints, and jet impingement shields) against the dynamic effects of such a postulated break. As a result of the advancement of the fracture mechanics technology, we are able to demonstrate by analyses that the probability of DEGB is extremely Irw for PWR main-loop piping and that the most likely failure mode of these pipes is " leek-before-break"; i.e., there will be a period of stable crack growth, not a sudden, " double-ended-break."

With regard to BWR pipe intergranular stress corrosion cracking (IGSCC), the AEC/Reguia-tory staff shut down all operating BWR plants in 1975 because of cracks discovered in some of the 4-inch by pass lines in recirculation loops. The same phenomenon was discovered in large recirculation pipe (up to 28-inch lines) in 1982 and 1983. The NRC mandated augmented inspections of plants scheduled for fueling in 1982 and 1983. The contrast here is that the NRC did not immediately shut down all plants for more severe, extensive cracking by the same degradation phenomenon. This is principally due to better understanding of the flawed pipe behavior under design basis loadings. However, IGSCC remains as one of the more salient BWR issues affecting piping and other components of the reactor internals as are certain issues in the halting of ATWS events (Anticipated Transients Without Scram). Diversity and reliability of Decay Heat Removal Systems and related Station Blackout issues (USIs A-45 and A-44) remain of Ilieport of the U.S. Regulatory Commission Pi 'ino Review Committee, NUREG-1061, U.S Nuclear Regulatory Commission (April 198 .

BWRs or PWRs with further research and analysis. troublesome conce Sabotage and fire protection plus core-melt accidents--are alst, of continuing concern for rese On the other hand, there has been a downgrading of the perceived importance or need further design changes beyond those already in place in the rules regarding such issu as steam explosions and turbine missiles, certain aspects of hydrogen control, and a number of other generic safety issues and unresolved safety issues that have been investigated.8 Improvements in PRA methodology and the data inputs to these quantitative risk asse ments tainty are a major focur of the Severe Accident Research Program since the wide uncer-relaxed requirements for design features or operating and main LWRs.

There are a number of reasons for this broad ranga frequency estimations, arising both from the uncertainties in the assessment of riskof un to other plants similar in design.for a specific plant and the uncertainties in culties in generically predicting the probability of a severe reactortheaccident:T considerable variability in the design features of existing LWRs; quantification of human error frequencies; common-cause failure mechanisms of multiple safety fea incompleteness in describing accident initiators (e.g., difficulty in including sabotage); assumptions made for success / failure criteria and for recovery actions; an the estimation intensity of thefires, earthquake, recurrence frequency hurricanes, of external events such as very high and floods.

With this backdrop, the SARP integrating elements--the Accident Sequence Evaluatfor.

Program (ASEP) and the Severe Accident Risk Reduction Program (SARRP)--have and extending available PRA data to assess present LWR risk from severe accidents, the impact to reduce ofuncertainty uncertainties, and in risk the cost-effectiveness of plant changes to reduce risk (or estimations).

ASEP work to date indicates that a However, it is very difficult to quantifrelatively small set of important accident considerable variation in plant design. y their frequencies generically because of Risk studies performed to date indicate that the risk of any particular plant not yet explicitly studied could deviate significantly from the design andestimated operatingrisk of plants of similar design because of unique plant-specific characteristics.

The conclusion drawn from these research findings is that the safety technical issues for each of the cancelled or deferred LWR units of Tables I and 2 cannot be d plant specific, systematic safety examir.ation to determine what vulnerablif them. ties are present and what cost effective changes are desirable to reduce This approach has become a cornerstone of NRC's Severe Accident Policy for dealing with any severe accident issues relating to the wide variety of LWR designs among operating reactors (NUREG-1070, p.19):

Recognizing that plant-specific PRAs have yielded valuable insights to unique plant vulnerabilities to severe accidents leading to low-cost modifications, licensees of esch operating reactor will be expected to perform a limited-scope, accident safety analysis designed to discover instances (i.e. outliers) of particular mance, vulnerability given to core melt or to unusually poor con,tainment perfor-core-melt accidents. These plant-specific studies will serve to verify that conclusions developed from intensive severe accident safety analyses 15ee, A Prioritization of Generic Safety Issues, NUREC-0933, December 1983 , et. s_eg.

of reference plants. or surrogate plants can be applied to each of the individual operating During the next two years, the Commission will formulate a systematic approach, including the development of guidelines and procedural criteria, with an expectation that such an approach will be implemented by licensees of the remaining operating reactors not yet systematically analyzed in an equivalent or superior manner.

Many of the plant-specific PRAs have stimulated the licensees performing them to take corrective actions, either during the study or after the results were evaluated. To a large extent these modifications wert implemented _ voluntarily by the utilities in an effort to correct weaknesses in their plants. Moreover, the resulting changes in equipment design and operating and maintenance procedures to reduce the uncovered severe accident vulnerabilities to an acceptable level were accomplished with relatively minor cost (NUREG-1070, pp.123-131).

Some of the new safety information' of possible relevance since the CP was issued for cancelled se. or deferred plants may involve environmental rather than design factors per '

This could--but probably infrequently will--include significant site-related safety information regarding: earthquake faults or new data bearing on the likelihood of high-intensity seismic events, meterological or flood data, and newly developed (or discovered) external hazards relating to proximal transportation systems involving t

' hazardous chemicals or explosive substances, etc. The safety significance of any such newly developed information will need to be evaluated to determine what cost-ef fective changes, if any, are desirable.

4.2 Environmental Issues Significant new information bearing on environmental issues will have varying levels of importance in accordance with the situation of whether the Construction Permit is still valid or not. This matter primarily affects those LWR units on the list of plants with cancelled construction since a large majority of these have had their cps cancelled.

The expectation is that plants having cancelled cps would be required to reapply for a

. new CP, an action that would require updated environmental information and new '

environmental hearings on any contested issues.

Contested issues can almost assuredly be expected to arise in any new environmental hearings for any such reactivated projects regarding such issues as need-for plant 8 and the cost-benefit analysis of alternative sources of energy for generating the proposed capacity addition of the reactivated unit. Both of these cou1J prove to be very dif fi-cult issues in light of the changed trends in electricity demand growth and construction and fuel costs including factors of uncertainty in their future projection. Regulatory uncertainty, as it could affect estimated construction completion schedules, would likely be an important issue related to the cost-benefit analysis of energy options.

Any future developments in reducing significantly the cost of presently non-conventional energy sources could also become an issue for some sites, depending on the inherent 1"Need-for-plant" is quite similar to the term "need-for power" in that both terms encompass issues of projecting electricity demand growth. However, need-for plant i is technically broader in scope in that it includes the prospect that a significant component of need might arise from the retirement from the utility's generating capacity of obsolescent or high-operating-cost units. Need-for plant also focuses more on need for baseload capacity as distinct from peaking capacity. However, as terms of art, the two terms are often used synonomously as implying the fuller scope of analytical and forecasting issues dealt with in the question of the need t for added baseload capacity of x amount by y years in the future including the option of purchasing power.

t

-auncertainties utfifty mightof such technological progress and the length of the future date at which resubmit a CP application.

It-is to be noted that a Commission Final

. Rule on "Need for Power and Alternative Energy Issues in Operating Licensing Proceed-ings" (10 CFR Part 51, September 1,1982) forecloses the reopening of both need-for-plant and alternative energy issues at the OL stage of licensing.

In the case of LWR reactivations with cancelled cps, it appears likely that it will be easier to defend against need-for-plant hearing issues the longer the delay in-the reactivation decision as reserve margins dwindle and if, as some believe, elec-tricity demand growth rates of 3 or 4% per year is sustained over a longer period, .

thus weakening the supportability of growth rates of 1 to 25 per year that are now rather commonplace assumptions.

Likewise in 5 or 10 years from now, it may be easier -

to defend the credibility of regulatory stability and the realism of shortened con-struction schedules and more effective management of construction quality and cost,

. assuming progress on these accounts by plants still'under construction or possibly a new plant with a certified destgo, perhaps a new standard design using the Advanced LWR envelope' design criteria developed by the current ALWR project of EPRI.2 There are, of course,.other possible environmental issues that may need to be treated-whether a reactivated unit has a cancelled CP or not. These could arise from sig-nificant new information or changes in local, State or other Federal (i.e., non-NRC) regulations.

The latter might involve additions to the lists of endangered species, historic landmarks, wild or scenic rivers, etc., or possibly changes in water or air quality control regulations or regulations affecting noise, dust, aesthetic intru-sions, etc. Any new court rulings on the validity or implementation adequacy of old regulations or review practices could portend a significant change in the treatment of environmental. issues, but this is a questionable prospect. Significant changes in area demography and land or water uses beyond those projected at the time of CP issue could yield increases in environmental impacts.

There is also the possibility that alternative siting issues might be reinstated as hearing contentions for at least some of the possible LWR project reactivations where the CP has been cancelled. As in the case of need-for plant and alternative energy issues, the Commission has adopted a Final Rule on " Alternative Site Issues in Operating Licensing Proceedings" (10 CFR Part 51, September 1, 1982) that provides for NEPA purposes alternative sites will not be considered in operating license reviews for nuclear powr plants and need not be addressed by operating Ifcense appli-cents stage. in their environmental reports submitted to the NRC at the operating license The treatment of alternative siting issues at the stage of reapplying for a new CP (as is expected to be required for all LWRs with cancelled CP's) will be more tract-able or more difficult in accordance with the kinds and importance of any significant new information noted above. Included in the rationale.for imposing the Commission's Final Rule on' alternative siting issues at the OL lievnsing stage is the following statement (46 FR 28630, June 29, 1981):

Construction is usually about 35-65 percent complete at this time (depend-

" ing upon the number of units to be built at the site) and a corresponding portion of the total construction costs have already been incurred. Major construction related environmental impacts t. ave already occurred at the site and by the time the staff has completed its review of the appifcation and the operating license hearings begin, the plan will be even further along.

'Given this factual background, the Commission cannot readily conceive of a 1 John J. Taylor, " Reopening the Nuclear Future", EPRI Journal (March 1985), pp. 2-3.

~22-O situation where new information concerning the proposed site could be of such significance as to tilt the cost-benefit balance in favor of an alternative site. (In any event,10 CFR 2.758 of the Commission's regulations would permit an exception to or waiver of the rule in particular cases if special circumstances are shown.)

It is noted that none of the deferred units have cancelled cps (although 2 have less have 35%

than construction) construction and progress of only 35% five of the units with cancelled construction (Table 1) or more: Marble Hill-1&2, Zimmer (now mooted),

Yellow Creek 1, and Hartsville-A1.

in excess of 20% (Phipps Bend-1, 29% and Hartsville-A2,These 34%).Two others on this list two units and eight additional units with construction progress between 3 and 17% (WNP-4&5, Phipps Bend-2, Yellow Creek-2 and Hartsville-81 & 82) still have onsite equipment with acquisition value in excess of $100 million. It should be noted that in the long number of years substantially since this equipment its replacement value. was purchased, inflation will have increased quite In support of the Final Rule on removing siting issues at the OL Ifcensing stage, it was further stated that "the Ccmmission finds that new information at the operating license stage is very unlikely to upset the prior conclusions concerning alternative sites."

There is some basis for believing this conclusion may still prove relatively valid for many of the LWRs with cancelled construction in the event that interventions on alternative among siting issues are introduced in new CP hearings for reactivated projects this group.

The staff has successfully dealt with alternative siting contentions and related environmental impact issues for a large number of licensing actions with a strengthened methodological approach (NUREGs-0398, -0499, -0625, and -0701). Moreover, the construction and operating environmental impacts (with appropriate mitigative measures) are observed to have been relatively minimal.

5.

IDENTIFICATION OF REGULATORY OR POLICY ISSUES OF POSSIBLE RELEVANCE TO ADEQUACY OF EXISTING RULES AND POLICIES OR THEIR NEED FOR REVISION The discussion in Section 4 suggests that existing rules, policies, and staff review practices would appear to be adequate to deal with those safety or environmental issues arising from significant new information, insofar as it is possible to foresee such esercent issues using curFe'ntly available inrormation. However, the proper regulatory  !

issue in these cases is not so much a matter of whether present rules and policies will prove to be adequate, but what modifications of these could improve the cost effective-ness in the use of NRC and industry resources in treating these issues if, and whenever, these arise in the course of LWR reactivations and their treatment. Also at issue is whether a more stable regulatory outlook can be achieved that would encourage or facili-tate LWR reactivation in the public interest 8 rather than certain features of-the pre-sent regulatory regime which might inadvertently serve as unnecessary deterrents or encumberments to reactivation decisions and related implementation processes. The dis-cussion to follow in Section 6 will identify candidate regulatory options relating to these two regulatory issues and propose decision criteria for their evaluation. ,

It is noted that Section 4 only examined the outlook for reactivation issues arising from significant new information affecting safety or environmental impact concerns. In this section, it is important to identify the family of technical and regulatory issues not necessarily involving new information but which relate to regulatory concerns over 8See " Declaration of Purpose." Energy Reorganization Act of 1974, (Sec. 2a); and the thoughtful discussinn of the public interest by Lynn E. Weaver, "The Outlook for Nuclear Power: Is it still an energy option in the United States?"

Nuclear News, August 1984, pp. 108-112.

o the adequacy of " routine" maintenance procedures for onsite (and offsite) equipment still assigned to plants with cancelled construction (Table 1) having no current I&E inspection, and the adequacy of Equipment Preservation Programs of the utilities and the inspection procedures of I&E for equipment assigned to deferred plants (Table 2).

In this exploratory effort, no attempt was made to obtain detailed information on the maintenance or equipment preservation programs of the applicable utilities nor of I&E inspection procedures for the equipment of deferred units. Nevertheless, past experi-ence is sufficient to identify certain maintenance issues relating to the storage or preservation of both nuclear and non-nuclear equipment under a range of environmental conditions and preservation practices.

Although there have been NRC sponsored workshops on nuclear power plant aging and contracted research or research plans dealing with nuclear plant aging (NUREG/CP-0036, NUREG/CR-4144, and NUREG-1144), the main thrust of this effort is toward aging and related wear issues of equipment of operating plants under more or less normal operating environtrents. There does not appear to have been any similarly focused research studies by NRC or other institutions or agencies devoted to the objective, for example, of assigning prioritizations in the cost-effective appli-cetion of maintenance and inspection resources to pre-operational storage conditions (including the variations of temperature, moisture, dust, and possible corrosive environments!) associated with past and present warehousing or mothballing practices of the cancelled or deferred plants included in this study scope.

Quality Assurance (QA) criteria are provided for in Appendix 8 of 10 CFR Part 50. As defined in these rules, quality assurance comprises all those planned and systematic actions will penarmnecessary to provideinadequate satisfactorily service.confidence that a structure, system, or component Qualfty assurance ir.cludes quality control, which comprises those quality assurance actions related to the physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component, or system to predetermined requirements.

The applicant is required to establish a quality assurance program that documents by written policies, procedures, or instructions and is to be carried out throughout plant life in accordance with those policies, procedt.res, or instructions. The applicant shall identify the structures, systems, and comoonents to be covered by the quality assurance program and the major organizations participating in the program, together with the designated functions of these organizations. The quality assurance program shall provide control over activities affecting the quality of the identified struc-tures, systems, and components, to an extent consistent with their importance to safety.

Activities affecting quality shall be accomplished under suitably controlled conditions.

Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanness; and assurance that all prerequisites for the given activity have been satisfied. The program shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of quality by in-spection and test.

The program shall provide for indoctrination and training of per-sonnel performing activities affecting quality as necessarv to assure that suitable proficiency is achieved and maintained. The applicant shall regularly review the status and adequacy of the quality assurance program. Management of other organizations parti-cipating in the quality assurance program shall regularly review the status and adequacy of that part of the quality assurance program which they are executing.

1Research on air pollutants causing acid rain, for example, might suggest that, although regional pollutants of these kinds (being quantitatively small) might not have any significant corrosive impact on warehoused or outdoor mothballed equipment over only a few years, these corrosive elements possibly could have significant impact over an extended period such as 8 to 12 years or more.

Moreover, the quality assurance rules require that measures shall be established to control the handling, storage, shipping, cleaning and preservation of material and equipment deterioration. in accordance with work and inspection instructions to prevent damage or When necessary for particular products, special protective environments, shall be specified atmosphere, such as inert gas and provided.specific moisture content levels, and temperature levels, Additional requirements of current NRC requirements for quality assurance and control procedures include organization; design control; test control; inspection programs; audits; corrective actions; quality assurance records; and document control relating to procurement and documented instructions, procedures, or drawings appropriate to activities affecting quality including appropriate quantitative or qualitativeaccomplished.

satisfactorily acceptance criteria for determining that important activities have been An examination of Appendix B reveals that the quality assurance and quality control pro-cedural criteria are cast in very general terms and make no specific references to any special QA/QC circumstances that might arise from unusually prolonged periods of main-tenance or equipment preservation that are being encountered for deferred LWRs or the possible reactivation of cancelled units. Indeed, the QA regulations were drawn up well before such cancellations and deferred construction of such large numbers of LWRs could have been anticipated. Thus, there is cause to suspect the adequacy of these current QA regulations to deal with the sorts of safety issues that seem pertinent to the current situations of maintenance, inspection, and other measures of QA/QC that could affect the regulatory reviews of any reactivations of deferred or cancelled LWR units and the safety acceptability of this equipment or corrective measures.

What, then, are the regulatory questions or issues of possible significance to the adequacy cf existing rules or practices or the need for modified rules and practices to achieve cost-effective measures for equipment maintenance and quality preservation?

A preliminary list mignt include:

(1) Which equipment (e.g., as determined by PRAs or other risk assessment methods identifying the more dominant severe accident risk scenarios) is in greatest need of maintenance preservation aeasures and which is of low or insignificant priority?

(2) What ranges of environmental factors (e.g., moisture, temperature, mildew, dust or other air pollutants) that may be encountered in mothballing or warehousing prac-tices are the most threatening to quality assurance of the priority equipment identified in item I?

(3) Does the lengthening years of equipment storage have a magnifying adverse effect on the QA/QC issues identified in item 2?

(4) What porsible stochastic events (e.g., rodent activity, windblown missiles, vagrant activities, fire, flooding, etc.) could violate protective coverings of stored equipment exposing them to deleterious environmental conditions?

(5) What kinds of equipment maintenance or preservation p tices provide the most reliable protection at reasonable cast to deal with ths above issues?

(6) Are presently used testing, QA auditing, and inspection techniques the most cost-effective for uncovering undesirable maintenance practices and the state of cor-rosion, damage, or other quality deterioration of the safety prioritized equipment identified in item I?

Information that would shed adequate light on the above six QA/QC issues might not yet exist in organized and readily available form to resolve the regulatory issue of whether existing NRC rules, policies, or review practices are appropriate in dealing most

h o-

! effectively with these issues or whether.an improved regulatory regime is desirable and what specific form the improvements should take.

6.

?.

REGULATORY OPTIONS AND DECISION CRITERIA 6.1 Decision Criteria and Regulatory Purposes Guiding LWR Reactivation Policy Development In the period since the Construction Permits were issued for the LWRs listed in Tables 1 and 2 with deferred or cancelled construction, a great deal of' reactor operating experi-ence and research findings have been generated increasing our knowledge of the safety

' tingsenvironmental and with diverse effects of numerous plant designs in a variety of environmental set-site characteristics. Substantial progress has been made in the

. assimilation of these findings in improved risk assessment and environmental 11 pact p

methodologies and further advances can be expected in the ne efforts of industry and other individsel and programmatic efforts.

At the same time, the NRC has sought to improve its regulatory effectiveness through policy developments and rule changes, some of which have recently been accomp11shed or that may be expected to be completed in the next several years. Holding special significance for policy development regarding LdA reactivation are the recently' Issued j Commission Severe Accident Policy Statement (NUREG-1070 and 50 FR 32138) and the i

revision of the rule on backfitting (10 CFR Part 50.109). Further policy developments j

are .in progress relating to safety goals (NUREG-0880, Rev.1), reassessment of methods

j. ~for estimating severe accident source terms (NUREG-0956), and probabilistic risk l

assessment of severe accident consequences (NUREG-1050 and NUREG-11501).

Because of these recent and pending developments of policies, rules and safety review i

procedures, transition. In it is fair to state that our regulatory regimen is in a state of general, these changes are intended to reflect:

e l

A shift away from detailed prescriptive requirements toward performance criteria

. consistent with NRC's primary responsibilities to assure protection of public health and safety as well as to protect environmental values; e

The carrying out of regulatory activities in a manner which recognizes that

' licensees have the primary responsibility for achievina safe. operation of nuclear facilities; and e

The principle that, in conformance with NRC's extant rules, policies, and i

understanding of safety vulnerabilities gained through operational experience data i

and research, the vendors and permittees or licensees are permitted a substantial flexibility to search for and select the most cost-effective approaches in the l

i design, NRC safety construction, goals 8 operation and maintenance of nuclear power plants to satisfy I and requirements and tc, protect environmental values subject to NRC regulatory review and dispository action consistent with our statutory l

t authorities and responsibilities, 1This study (in draft) will develop base-line assessments of severe accident risk for a spectrum of typical LWR plant and containment designs for regulatory applications including implementation of the Severe Accident Policy.

l . :The Commission is currently reviewing a safety Doel implementation plan in concert with other strategic goals and policies under development.

L 1

l

?_ _ _ _ _ _ _ _ _ __ - _ - _ . _ _ . - ___ w

y ,

e , '

! ' Within this developing framework of regulatory philosophies and principles, the .

.following purposes or objectives are. proposed to guide the formulation of an LWR Reactivation Pnlicy:

.(1) To achieve improved stability and predictability of reactor regulation in a manner that would encourage optimal utility choices of LWR reactivation options in keeping with the public interest; (2) -To clarify the regulatory procedures and requirements for cost-ef fective equipment preservation and site stabilization programs, as appropriate, for those LWR units with cancelled or deferred construction where the possibility of LWR reactivation cannot wholly be ruled out, especially where significant (onsite) equipment still i

remains that could be used in reactivation projects at the same or'other sites; (3) -To develop a strategic, integrated plan to improve and implement a regulatory pro-cess that would deal most effectively with the variety of regulatory issues envisioned for the differing circumstances and conditions of any reactivated LWR projects either in the near- or long-term regarding the CP or OL licensing phases of project reactivation; (4) To avoid imposing any unnecessary burden of information, analytical effort, or commitment.of other industry or regulatory resources on behalf of any regulatory objectives that have only marginal or dubious value for protecting the public health and safety, the common defense and security, and the environment; and (5) To ensure that NRC's review process will continue to provide appropriate access for the expression of the public's rafety and environmental concerns associated with LWR project reactivations.

The above regulatory philosophies, purpcses, and objectives would serve to guide the formulation of a reasonable set of regulatory options in the initial phase of an orderly decision process. Even before the analysis of these options should proceed, it is also desirable to establish the decision criteria by which the regulatory effective-ness of the options will be evaluated. For the identification of an appropriate set of decision criteria will estabitsh the kinds of information that needs to be assembled for the analysis. For example, some advance thought as to what decision criteria would be appropriate was instrumental to the formulation of the conceptual model of Figure 1 and the matrix content of Tables 1 and 2.

However, the parameters encompassed in these tables and chart do not alone suggest a sufficient set of information in meeting the needs of decisionmaking relative to a choice of regulatory option *. As discussed in Section 4, there is a need, in the reactivation of LWR projects, to consider whether certain backfits might be required as a result of significant new safety or environmental information that developed subsequent to the CP issue date. Unless these backfits have been mandated by subsequent rule changes, a determination of their requirement is subject to the procedures and criteria established in NRC's new Backfit Rule (10 CFR Part 50.109). In this Rule, backfitting is defined as the modification of or addition to systems, structures, com-ponents, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organ;2ation required to design, construct or operate a facility; any of which may result from a new or amended provision in the Comission

^

rules or the imposition of a regulatory staf f position interpreting the Commission rules that is either new or dif ferent from a previously applicable staff position af ter:

(1) The date of issuance of the construction permit for the facility for facilities having construction permits issued after October 21,1985; or (11) Six months before the date of docketing of the operating Ifcense application for the facility for facilities having construction permits issued before October 21,1985; or (111) The

c ,

~

.- j t

date of issuance of the operating license'.for the facility for facilities having operat-

'0ing licenses; of_this part.or. (iv) The date of issuance of the design approval under Appendix M

, N or different ways to the lists of LWRs in Tables 1 and 2 lieving d construction.

since all of these would come after the cut-off date21,of1985. OctoberThose requir having still-valid cps issued before October 21, 1985 Those LWRs

~

applications (i.e., eight LWRs with cancelled constructTon and all of the deferredand h LWRs), would benefit to varying degrees from the exemption dates of provision (ii).

However.. some of the other LWRs might benefit from reduced backfitting requirements if they are under duplicate, provision replicate, or standard reference designs that would qualify them (iv).2 i of these categories under the Standardization Policy.One deferred LWR (WNP-  !

For those future LWR reactivation projects to which the Back specific terms in Section (c) as follows:

'(a)(3) 1 The Commission shall require the backfitting of a facility only when it  !

determines, based on the analysis described in paragraph (c) of this section, 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 the backfit and that the direct and indirect costs of implementation for that facility are '

justified in view of this increased protection.

(c) i In reaching the determination required by paragraph (a) of this section, the Commission will consider how the backfit should be prioritized and scheduled in  !

Ifght of other regulatory activities ongoing at the facility and, in addition, will ,

-consider information available concerning any of the following factors as may be appropriate and any other information relevant and material to the proposed backfit: -

(1) to achieve;Statement of the specific objectives that the proposed backfit is designed (2) General description of the activity that would be required by the licensee  ;

or appifcant in order to complete the backfit; i (3) Potential change in the risk to the public from the accidental off-site ,

release of radioactive material; I i

(4) Potential impact on radiological exposure of faciifty employees; (5) Installation and continuing costs associated with the backfit, including  !

the cost of facility downtime or the cost of construction delay; [

(6) The potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing regulatory requirerents.  ;

(7) The estimated resource burden on the NRC associated with the proposed back-fit and the avaliability of such resources; i i

(8) The potential impact of differences in facility type, design or age on the '

relevancy and practicality of the proposed backfit; ,

i (9) Whether the proposed backfit is interim or final and, if interim, the justification for imposing the proposed backfit on an interim basis. i 1This discussion does not encompass the full provisions of the 8ackfit Rule, in the fnterest of brevity; the complete set of provisions should be carefully examined for [

relevancy to specific LWR reactivation situations. l

{

L__ - - - - - -

28 In addition to regulatory options involving possible future backfits to deal with any i safety or technical issues resulting from significant new information affecting reacti-vated LWRs, NRC has developed Regulatory Analysis Guidelines (NUREG/8R-0058, Rev.1) that provide decision procedures and criteria for dealing with certain proposed regula-tory actions ments including rulemaking and the possible establishment of other NRC require-and guidance.

'to easily determine: The intent is to allow decisionmakers, and other interested parties, (1) the conclusions reached, (2) the bases for conclusions, (3) the type, magnitude and source of uncertainties which might affect the conclusions, (4) the sensitivity of any conclusion to variation in the important input parameters affecting the tc theconclusions, conclusionsand (5) the analytical methods used and the logic which led the analyst presented.

A Regulatory Analysis of a proposed action, including a discussion of any reasonable alternatives to the proposed action, shall be prepared for each proposed rule and final rule that, in the determination of the responsible office director or the Executive Director for Operations, will likely result in the following:

e An annual effect on the economy of $100,000,000 or more in direct and indirect costs, or e

A significant impact on health, safety or the environment, or e

A substantial increase in the cost to NRC licensees, permit holders or applicants, to Federal, state or local governments, and geographical regions.

Appendix D to NUREG/8R-0058 (Rev. 1) provides the following specific examples of effects that could result in a cost or benefit of a proposed regulatory action or its alternatives (including the no-action alternative), thus constituting decision criteria:

1. RADIOLOGICAL SAFETY CONSEQUENCES (a) Change in accident probabilities; specify the accidents (old, new probabilities)

(b) Change in failure probabilities; describe the equipment directly and indirectly affected by the proposed action (old, new probabilities)

(c) Change in population at risk (percent and absolute)

(d) Change in occupational exposure; during installation, operation or maintenance (rem)

(e) Change in unplanned radioactise releases offsite (curies)

(f) Change in routine radioactive effluent releases (curies)

(g) Change in operator response times (seconds / minutes)

(h) Change in maintenance capabfifty (yes/no) (explain)

(1) Change in NRC's inspection and enforcement capabilities (yes/no)

2. SAFEGUARDS IMPACTS (a) Change in facility security (yes/no) (explain) l (b) Change in materials control and accountability (yes/no) (explain)

(c) Change in transportation security (yes/no) (explain)

3. OPERATIONAL IMPACTS I

(a) Change in reactor availability (hours / days)

(b) Change in facility downtime beyond that normally scheduled (hours / days)

(c) Change in allowable reactor rating (percent and absolute) i t

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4. ECONOMIC IMPACTS (a) Construction cost change (dollars) 1 (b) Operating cost changes (dollars) '

(c) Retrofit costs (dollars)

(d) Recordkeeping and reporting cost changes (staff-hour; dollars)

(e) Change in onsite personnel reqM rements (staff-hours)

(f) NRC costs change; include contractor technical assistance costs (staff-hours or dollars)

(g) Other increases in applicant expenditures for compliance with regulatory requirements. (staff-hours or dollars)

(h) Change in expected direct cost of an accident (dollars)

5. ENVIRONMENTAL IMPACTS ,

(a) Change in water quality [

(b) Change in air quality -

6. INFORMATION COLLECTION IMPACf5 (Resulting from appifcation, reporting or recordkeeping requirements)

(a) ' Annual licensee / applicant staff hours (hours)

(b) Annual lice see/ applicant cost (dollars)

(c) Annual cost to the NRC (hours / dollars)

7. .OTHER IMPACTS (for example)

(a) Consequences for small business (dollars / hours)

(b) Significant impacts on vendors and equipment suppliers (yes/no)

(c) Anti-competitive consequences (, impact on viability of existing firms to complete or provide equfpment)

(d) Availability of skilled-labor / professional assistance (regional employment figures by a relevant cate' gory) f ,.

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(e) Number of ifcensees affected . ,, f, f Itisnotedthattheaboveexamp$esofpossiblecostsandbeneff'esto'includeinthe scope of decision criteria include qualitative assessmenti as well as quantitative estimates of effects and their probabilities not necessarily expressed in commensurable units. The " apples and orangts" nature of these. decision criteria is dealt with by an overall subjective balancing judgment of beneficial and adverse effects. In addition to the aforecited examples of cests and benefits, it is proposed that the decision criteria include a determination of how well the regulatory options for an improved treatment of regulatory issues in LWR react'Ivation projects are in agreement with the three regulatory philosophies and five purposes or objectives set forth at the.>

beginning of this section, some of which are reflected, at least in part, in the above examples of decision criteria., '

6.2 Regulatory Options for Dealing with Reactivation Issues One alternative to the development of a policy for dealing with issues associated with the reactivation of nuclear plants with deferred or cancelled construction is to have no policy at all. That is to say, if there is little prospect for more than a few such plants to be reactivated, an economical use of NRC resources might be to treat such reactivations on a case-by-case ad hoc basis rather than through policy formulation.

The exercise of the ad hoc option, however, involves a catch-22 situation. This option

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30-is likely to have a discouracina effect on utility decisions to reactivate construction such units, given that:

(1) Most of the involved utilities have a weak financial situation compounded by an uncertain outlook for demand growth and the rate at which excess reserve margins will shrink; (2) The lack of NRC policy development would create a compounding of uncertainties over the stability and predictability of NRC regulation in dealing with reacti-vation issues and what this course of action might mean for licensing and con-struction delays and the need to demonstrate adequate management in cost control to obtain the confidence of investors, Public Utility Commissions, and the consumer public; and (3) The no policy option would forego opportunities to improve the current regu-latory regime at possibly modest cost in ways described below that would tend to encourage utility decisions on reactivation options which would best serve the public interest 1 (see above statement of proposed objectives of LWR reactivation policy).

Thus, the no policy option, if based on the assumption of a small number of reactivations, would be a self-fulfilling prophecy since it is a course of action that would likely exert a negative impact on reactivation decisions by utilities. Accordingly, the remain-ing discussion will deal with sub-options within a framework of LWR reactivation policy development.

It must be recognized at the outset that some policy options will pertain only to plants with cancelled construction (and especially those with a withdrawn Construction Permit),

while others will apply to both cancelled and deferred units, albeit with different weightings of decision criteria. It is expected, for example, that the reactivation of units with a withdrawn CF would require a re-application for a CP involving new (or amended) safety and environmental reports, staff re-reviews, hearings, and board dect-sions. Thus, policy options exist whether staff review for selected safety and environ-mental issues might best be accomplished through ceneric rulemakino covering some of these issues or whether these should receive, as before, case-by-case review for a CP reapplication.

An initiative to improve the regulatory efficiency and predictability of the treatment of certain environmental and safety issues through generic rulemaking is described in a 1978 NRC report on " Preliminary Statement on General Policy for Rulemaking To Improve Nuclear Power Plant Licensing" (NUREG-0499). In this report the staff porposed ten issues for generic rulemaking:

(1) Future availability and price of uranium; (2) Alternative energy sources to the nuclear option; (3) Need for adding baseload generating capacity; (4) Methodological and information requirements in the analysis of alternative sites; IIn the case of the Limerick-2 reactivation decision, it was laudably the perception by investors and the Pennsylvania Public Utility Commission, and not the NRC, as to what would best serve the public interest in the affected region in the face of a controver-sfal decision climate.

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. i g, (5) Criteria for the assessment of nuclear plant impacts and mitigative measures; (6) Genericproceduralcriter.iatodefinemoreconcrete1[NRCresponsibilityin assssments and decisionst regarding certain water-rel.ated impacts in relation to the, statutory authorities of EPA and permitting states;

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~ (7) NEPA 11, decision criteria for Operating License (OL) rehews; (8) Odeupational radiation expesure' control; '

(9) Generic radiological impact for normal LWR radionuclide releases; and 4< m ,

, (10) Threshold limits for generic disposition or cooling tower effects.

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As stated in NUREG-0499, the principal benefits sought in generic rulemaking for these K issues are: s ,

n' e Achievement of more effective public input and improved public understanding of NRC's analytical proce'dures and decision criteria in treating potential environmental and safety issues in the licensing process for nuclear power plants; ,

y e Imp'rovement of the stability and predictability of the licensing process, including t%e provision of orderly and clear procedures for State-Federal cooperation in treating generic licensing issues; and e Accomplishment of an overall savings of manpower and financial reheurces of the NRC, the public, the utility < industry, and other local, State, and Federal agencies involved in the nuclear ifcensing process.

U Themaindrawbackstotherulemakintioptionwerepercededas: (1) the short-term A. increase in dollar costs of the various participants fec the rulemaking action, includ-ing contractual support; and (2) the additional impacts (i.e., opportunity costs) of diverting manpower and other resources to the rulemaking process and away.from other ky to productive uses for a temporary period, W On December 14,1978. the Commission published for public ' comment an Interim Policy p Statement on " Generic Rulemaking To Improve Nuclear Power P1knt Licensing" (43 FR 58377)

However, due to the absence of new appifcations for Construction Permits and the diver-sion of staff effort to the development of a TMR Action Plan following the TMI accident of March 28, 1979, only Issue-7 (NEPA decision criteria for OL reviews) of the above ten issues was subsequently dealt with through generic rulemaking. This action led to two final rules in 10 CFR Part 51: (1) Alternative Site Issues in Operating License Proceedings (46 FR 28630), and (2) Need for Power and Alternative Energy !ssues in Opera-ting License Proceedings (47 FR 12940). Both of these rules preclude the subject issues from consideration in OL proceedings--an advantage to utilities reactivating deferred plants and to those few plants with cancelled construction that still have an active CP. Options also exist to improve the cost effectiveness, stability and predictability of the licensing process through generic rulemaking of<the remaining nine (and possibly additional safety and environmental) issues that would apply at the CP licensing stage.

This rulemaking would apply to any new CP application as well as a reactivated LWR pro-ject with a cancelled CP. It is also possible that rulemaking on some of these issues s

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could have provisions for closer cooperation or a shif t of regulatory responsibilities

-between the NRC and States in which the units ~are sited.8 A different kind of regulatory option of possible benefit to the reactivation of plants with cancelled. cps would be to establish through rulemaking a one-step CP/0L ifcensing and hearing process on the grounds that the bulk of the site-related and design-related environmental safety issues will have already been dealt with in the original CP hearings.

Any environmental and safety issues arising from significant new information since the initial CP hearing could thus be dealt with in a combined CP/0L hearing for any reac-tivated LWR project falling into the category of cancelled cps.

-Other options involving new or changed rules, policies, review procedures, etc. would appear to hold potential for improvec regulatory cost effectiveness and stability for LWR reactivation of both deferred and cancelled plants. These include:

(1) Policies and review procedures for standard plant designs (repilcate plants, duplicate plants, FDAs, or Design certifications of reference designs through rulemaking vs a custom plant; (2) Policies and review procedures involving a relatively complete design with a full scope PRA vs a design and PRA of limited scope; and (3) Options for. revised rules, regulatory guidelines, policies, technical specif f-cations, standard review plans, hearing guidance, legislative or ac.ninistrative reforms, etc. , to improve regulatory efficiency vs the current regulatory regine.

The regulatory benefits associated with standard plant designs (item 1) would, of course, only apply to those deferred or cancelled plants that qualify. For example, only one deferred plant is a standard design. However,19 of the 35 cancelled LWR units are some type of standard plant: duplicate (3), replicate (4), and reference design (12).

While some of these units had cancelled equipment deliveries, by the same token, it "

would appear that all of the 22 units (see Table 1) having no or less than $100 million aquisition value of onsite equipment might be candidates for a new unit of advanced design without economic penalty. For example, it is possible that some of the Balance-of-Plant equipment might suitably be matched with a new reactor design. But the biggest economic gains.and safety review benefits would likely accrue to a new standard plant

.with Design Certification through rulemaking as provided for in the new Severe Accident ,

Policy (NUREG-1070). The essentially complete design and the fulfillment of the follow-ing criteria and procedural requirements of a Design Certification (and other new plant designs) will increase stability and predictability of the regulatory process and pro-vide for a shortened licensing and construction schedule with its attendent cost savings

.and safety assurance benefits (NUREG-1070, p. 10):  ;

a. Demonstration of compliance with the procedural requirements and criteria of the current Commission regulations, including the Three Mlle Island require-ments for new plants as reflected in the CP Rule (10 CFR 50.34(f));

b. Demonstration of technical resolution of all applicable Unresolved Safety Issues and the medium- and high priority Generic Safety Issues, including a IFor example, see " Federal-State Cooperation in Nuclear Power Plant Licensing" (NUREG-0398) and various reports of the NRC Of fice of State Programs (NUREGs-0195 through -0204) and NUREG/CR-0022.

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special focus on assuring the reliability of decay heat removal systems and the reliability of both AC and DC electrical supply systems; c.

Completion of a Probabilistic Risk Assessment (PRA) and consideration of the severe accident vulnerabilities the PRA exposes along with the insights that it may add to the assurance of no undue risk to public health and safety; and d.

Completion of a staff review of the design with a conclusion of safety accept-ability using an approach that stresses deterministic engineering analysis and judgment complemented by PRA.

In addition, the Commission has proposed to the Congress its draft " Nuclear Power Plant Licensing and Standardization Act of 1985" which would provide for the issuance of com-bined Construction site approvals Permits and standard and Operating design approvals. Licenses in a one-step licensing process, early It is recognized that the above criteria and procedural requirements for new plant designs (both standard and custom) represent changes to the Commission's 1978 Standardi-zation Policy Statement and to 10 CFR 50, Appendix 0, which deals with standard reference designs.

Accordingly, the Commission has requested the staf f to prepare a revision of the standardization regulation and policy (NUREG-1070, p. 31). The staff has considered a number of options for this revision (including those provided by a Study Group of the Atomic Industrial Forum) and on Decen.ber 4,1985 proposed to the Commission a draf t

" Standardization Policy Statement" (SECY-85-382).

sfons to the 1978 standardization policy statement are:The most significant proposed revi-(1) The four licensing requirements for new plant designs as set forth in the Com-mission's severe accident policy statement have been added.

(2) Provisions for Design Certification through rulemaking have been added.

(3) The ability of the staff and Commission to make changes to approved or certi-fied designs has been made more restrictive. An explanation of the applica-bility of the backfitting rule (10 CFR 50.109) to each of the options has been added. The ability of holders of design approvals or certifications to make such changes has been made less restrictive.

(4) The overlap in the reference periods between the duplicate and replicate design concepts has been eliminated.

(5) Fees required of reference design appifcants will be allocated among the applicants for permits and licenses which propose to use the reference design.

Enactment of the draft " Nuclear Power Plant Licensing and Standardization Act of 1985" would be necessary prior to adopting this approach.

(6) Final design approvals and design certifications can be renewed once for a duration up to the original approval period. Preliminary design approvals can only be renewed on a finding of good cause.

Regarding both cancelled and deferred plants of custom or standard design, options are being explored by the staff in cooperation with IOCOR and other industry representatives or groups in the development of policies and review procedures (supra) involving a rela-tively complete design with a full-scope PRA y a design and PRA (or other systematic safety examination) of limited scope. It is recognized that there are a diversity of PRA methods, which, in accordance with Severe Accident Policy, are to be used to

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<0 complement deterministic engineering analysis and judgment.2 'These PRA methods will ,

. continue to undergo evolutionary development as the results of research programs and reliability data from operating reactors become available and as innovative uses of PRA in safety decision contexts suggest better ways to achieve the benefits of these (

methods while guarding against their limitations or improper uses. While learning curves of these kinds will likely continue for a decade or more, it would nevertheless i be constructive to consolidate this experience at various stages of PRA development and

utfifzation.~ At the present stage of development, a number of positive uses of PRAs l have been demonstrated, especially in identifying: (1) those contributors to severe

-accident risk that are clearly dominant and hence need to be examined for cost-effective f risk reduction measures and (2) those accident sequences that are clearly insignificant "

risk contributors and can therefore be prudently dississed. In-between cases are more

~ problematic.

The use of PRA and related systems reliability analysis methods are being applied to both new plants and existing plants. In the Severe Accident Policy, existing plants l

are defined as operating reactors and plants under construction for which an operating '

license has been applied. Regarding such methods of safety analysis and staff review, the Severe Accident Policy states (NUREG-1070, p. 8.):

One important source of new information is the experience of NRC and the nuclear 5 industry with plant-specific probabilistic risk assessments. Each of these analyses, which provide a more detailed assessment of possible accident scenarios, has exposed relatively unique vulnerabilities to severe accidents. Generally, the 3 undesirable risk from these unique features has been reduced to an acceptable  !

level by low-cost changes fr: procedures or minor design modifications. Accordingly,  !

. when NRC and industry interactions on severe accident issues have progressed suf- i ficiently to define the methods of anlysis, the Commission plans to formulate an l integrated systematic approach to an examination of each nuclear power plant now I operating or under construction for possible significant risk contributors (some-times called " outliers") that might be plant-specific and might be missed absent [

t a systematic search. Following the development of such an approach, an analysis  ;

will be made of any plant that has not yet undergone an appropriate examination. ,

The examination will include specific attention to containment performance in '

striking a balance between accident prevention and consequence mitigation. In implementing such a systematic approach,' plants under construction that have not .l yet received an Operating License will be treated essentially the same as the  :

manner by which operating reactors are dealt with. That is to say, a plant-specific review of severe accident vulnerabliftfes using this approach is not considered to (

be necessary to determine adequate safety or compliance with NRC safety regulations t under the Atomic Energy Act, or to be a necessary or routine part of an Operating [

License review for this class of plants. -i The Severe Accident Research Program (see below) as well as NRC's extensive severe acci-dont studies of certain fridividual plants will aid in determining the extent to which '

carefully analyzed reference plants can appropriately serve as surrogates for a class of similar plants as the basis for any generic conclusions. These studies will also aid in identifying the desirable scope and approach for follow-up safety studies of individual  :

plants. Any generic design changes that are identiffed as necessary for public health I and safety will be required through rulemaking and will be consistent with the Commis- i sion's backfit policy. e i

3For a fuller discussion, see NUREG-1070, Part IV, Severe Accident Program (pp. 21-36). I i

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9 t 6.3 Potentia 1' Impact of Research Programs on the Choice of Regulatory Options It is apparent from the preliminary status information regarding the continuing presence of financial and need-for plant issues of cancelled and deferred plants (see Tables 1 and 2) that relatively few LWR project reactivations can be expected in the next year or so. Accordingly, it is desirable that a LWR Reactivation Policy be formulated in such a way as to flexibly permit the use of the near-term results of NRC's research programs.

These results hold promise, among other objectives, of aiding the definitization of policy options safety review procedures, etc. that could measurably improve regulatory stability and the cost effective treatment of issues in LWR reactivation.

A number of these research programs are either part of the Severe Accident Research Program (SARP) described in NUREG-0900 or are follow-up programs or contracted research efforts that further improve PRA and related systems reliability analysis methods and their useful implementation in safety management.8 One such program is the Accident Sequence Evaluation Program (ASEP) begun as part of the SARP. The dominant accident sequences from some of the 20 completed plant PRAs in the United States are being used in the ASEP research in order to obtain risk profiles of the approximately 100 near-term and presently operating plants. In the absence of PRAs on each plant, the ASEP approach is the only known means for obtaining risk statements for all plants. A report describ-ing this industry-wide accident sequence likelihood information data base has been delayed to address the needs of the severe accident research program, but it will be resumed when the demands caused by the NRC reference plant analysis permit (NUREG-1150, forth-coming). In addition to plant-specific PRAs and ASEP-dominant accident sequences, operational occurrences as reported in licensee event reports can be analyzed in generic event trees to obtain a gross assessment of nuclear indu:try risk over periods of time.

The NRC Accident Sequence Precursor Program has completed an assessment of the 1969-1979 period (NUREG/CR-2497) and also the period 1980-1981 (NUREG/CR-3591).

Information on containment systems performance and failure modes, fission product source terms, and health consequences is being developed under the Severe Accident Risk Reduc-tion Program (SARRP) scheduled for completion in the summer of 1986. Another research program is aimed at the development of a Systems Analysis and Risk Assessment (SARA) system. The direct purpose of the program is to develop a capability for computation and analysis of nuclear power plant risk characteristics, using state-of-the-art, user-friendly and modularized computer sof tware and existing nuclear power plant risk infor-mation developed under two other key current research programs. The SARA system will enable a time-dependent display of cumulative costs and risk reduction benefits result-ing from past or future implementation of the resolution of a number of generic issues.

A research program is being mounted to improve regulatory ef fectiveness in treating a number of problems that have evolved over the years regarding technical specifications (hearafter, tech specs). Today, the compilation of tech specs has grown to over 500 pages and several thousand survelliance requirements. The absence of specific criteria as to the content of tech specs has resulted in numerous items of vastly differing levels of importance being included, as well as requirements that are occasionally inconsistent. This situation tends to divert attention from the principal safety parameters while focusing attention on detailed surveillance of lower-importance 3Much of the discussion of research programs in this section is adapted from:

Gary R. Burdick and Carl E. Johnson," Relevance of PRA to Regulatory Development," NRC Of fice of Nuclear Regulatory Research, a paper presented at the WAf Tec 13th Annual Energy Conference, Knoxville, TN, February 13, 1986.

o systems. The voluminous technical specifications have also become burdensome and costly to utilities, yet do not contribute corresponding benefits to safety. Further, some tech specs are complex and difficult for the control room operators to implement, and others may actually be adverse to safety (e.g. , certain forced shutdowns when, in fact, continued steady-state operation is the safest plant condition). Other concerns have been expressed regarding: excessive testing contributing to component wear, added maintenance downtimes resulting from component wear, unnecessary test downtimes, intro-duction of human errors, and the potential for common-cause failures. Finally, the NRC has not in the past discriminated between utilities with excellent preventive maintenance programs, who may not need prescriptive technical specifications, and utilities with poor preventive maintenance programs who may need them. Rather, the NRC and industry have concentrated on standardized specifications which would be applied to utilities uniformly to protect against the worst performers. This practice tends to place unnecessary burdens upon the good performers.

AccorJingly, a contract study has been recently authorized to provide a reliability and risk basis for evaluating technical specifications and exemption requests and to support the Office of Nuclear Reactor Regulation efforts to improve technical specifications, especially the identification of tech specs not important to reliability /rf sk and to identify dominant failure modes of safety equipment. While this program is principally directed to providing cost-effective methods of achieving reactor operation safety, its features would also aid in determining the dominant accident sequences of individual plant designs and, hence, it would be relevant to a determination of which onsite equip-ment of cancelled or deferred plants requires the greatest care for maintenance and pre-servation and which are insignificantly safety-related, requiring minimal preservation measures from a regulatory standpoint.

Another research program serving similar purposes is directed to the use of PRA in a Plant Reliability Program. Following the Three Mile Island accident and the Salem reactor trip failures, the NRC and others sponsored surveys of reliability techniques used by the aerospace, defense and other industries that might be applicable to LWRs.

In 1984, Argonne National Laboratory coalesced the results into a set of reliability elements that appear applicable to LWR safety. Brookhaven National Laboratory was then engaged to evaluate the effectiveness of these elements.1 The results of this planned evaluation of the effectiveness of reliability elements will serve two purposes. One is to serve as part of the technical basis for staff evaluations of tradeoffs that licensees may propose to substitute reliability program elements in place of specific prescriptive requirements. The second purpose of this research is to help achieve NRC goals to shift its regulatory emphasis away from detailed prescriptive requirements toward performance criteria (NUHEG-0885, Issue 4) with significant implications for improved safety and regulatory cost-ef fectiveness.

In order to develop methods for applying PRA results to manpower allocation decisions made by NRC inspectors, NRC has established a "PRA Application to NRC Inspection Program". One feature of this pregram is the development of a Plant Risk Status Information Management System (PRISIM), which is a decision-oriented, user-friendly, menu-driven program that contains data base management and interactive routine that will aid NRC regional inspectors in allocating their efforts toward those areas where they will have the greatest impact on safety. By the same token, it would be a 3 Carl E. Johnson, " Operational Safety Reliability Research," NRC Of fice of Nuclear Regulatory Research, a paper presented at the International Conference on Nuclear Power Plant Aging, Availability Factor and Reliability, American Society of Metals, San Diego, CA, July 8-12, 1985.

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-possibility for utility management to adapt this system (perhaps in conjunction with I .

SARA) for use in developing cost effective approaches for equipment preservation of

! cancelled or deferred plants as well as subsequent use in reactivated construction quality assurance programs.

Because some decisions made by~ inspectors depend on the current status of the plant,

( PRI5!M contains an interactive routine that allows the user to specify components or

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' subsystems that are out of service (or could fail in the future if equipment is not adequately preserved and maintained during a delayed construction phase). The user is then apprised of the impact the specified condition (s) places on instantaneous risk and the components that are most critical to maintaining plant safety under the specified i

condition (s). 'Thus~, inspectors can plan their actions using PRA-based information j

integrated with plant status information. A computer program was chosen to catalog and present the PRA information because the total amount of information is large, but the amount needed for any particular decision is relatively small. PRISIM allows the user to quickly and enormous logically quantities access the desired information without being overwhelmed by of data.

Moreover, the user does not need to have a background in computer operation it. presents. or PRA to use the program or understand and employ the information One final example of a research program with " piggy-back" imp 1tcations for an LWR Reactivation Policy that would improve the outionk for greater regulatory stability and cost-effective

" approaches for treating reactivation issues is that of the NRC Research

. Program Plan on Effectiveness of LWR Regulatory Requirements in Limiting Risk" (ELR).

On October 3,1984 the Commission issued a public notice of the availability of the ELR' Program Plan and invited pubite comment (49 FR 39066). This program is being initiated i

i to identify current regulatory requirements which, if deleted or appropriately modified, would improve th" efficiency or effectiveness of NRC's regulatory program for

! nuclear power plants without adversely affecting safety. Initially, thfs program will

! systematically assess the risk inrortance of selected current regulations in 10 CFR i Part 50 and related regulatory requirements.

! A number of existing programs 8 assess the adequacy of present regulations. However, i these programs are not specifically designed to weed out existing regulations or

• regulatory requirements which do not reduce risk significantly. Initially, this program is designed to (1) systematically screen all current regulatory requirements associated with 10 CFR Part 50 and to assess the importance of selected requirements based first on their contribution to assuring that nuclear power plants are safely designed, constructed, and operated and second on their impact on licensee applicant, o

and NRC resources and (2) identify and propose appropriate modifications to eliminate

' duplication., inconsistency or unnecessary requirements and thus focus available NRC and industry resources more directly and precisely on the significant safety areas and issues.

I l Prime candidates for modification will be (1) old regulatory requirements which in

! light of present knowledge may no longer be considered risk important or whose risk l;

importance may have been reduced substantially by the implementation of newer

" Examples include (1) the Generic Issue and Unresolved Safety Issue programs; (2) programs and tasks that would be guided by the Severe Accident Policy Statement; (3) the Integrated Safety Assessment Program for operating reactors; (4) the

{ operating experience review by the Office for Analysis and Evaluation of Operational Data; and (5) the many studies, analyses, test and experiments supported by the

{ Office of Research.

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O requirements and (2) areas in which there are large safety margins or conservatisms which can be reduced without measurably increasing the level of risk. In such cases, modification could produce a significant safety benefit, since the attention and resources of licensees, applicants, and the NRC that are now directed to these areas could be redirected to other areas of greater safety significance.

6.4 The Image Problem of Identifying and Choosing between Alternatives to Deal More Effectively with a Complexity of Regulatory Issues It would be unfortunate, indeed, if the above (and future) discussions of regulatory options to deal with the complexities and risk uncertainties of nuclear power plant technology created an impression that major or costly safety and environmental improve-ments are Itkely to be needed in LWR reactivations or that would yield an image of regu-latory instability with adverse impact on investor confidence. It is an inescapable fact that nuclear technologies have system complexities and that assessments of severe accident risks have broad ranges of un:ertainty in the face of numerous accident scenar-ios, each with a relatively low probability and, hence, an insufficient (but growing) data base of failure frequencies of a wide variety of custom designs. However, the large program of severe accident refeuch that the NRC launched several years ago has not been , Justified on the basis of expectations that major changes would be required in plant designs to make them acceptably safe. Rather, the prime objective of this program is to reduce the uncertainties surrounding the level of risk posed by the possibility of severe reactor accidents. We are not seeking risk reduction except in those cases in which the new information would suggest undue risk. Said in another way, it is our judgment that existing plants are safe enough, but we seek to narrow the window of uncertainty.

Moreover, an extensive appendix of NUREG-1070 on Severe Accident Policy dealt with current information bearing on the need for generic design changes of LWRs. The principal conclusion reached was that it is believed that the large body of currently available information summarized in this Appendix provides substantial support to the notions that existing plants pose no undue risk to public health and safety and few.

if any, generic design changes imposina major costs are likely to be required by new safety information that is prospective of development in the next several years. During the two years since this conclusion was first reached, no significant new information has been developed from SARP project completions or interim results, as well as other sources, to significantly undermine this conclusion. Although research results on the Mark I/BWR containment design did suggest that certain substantial modifications might be cost-effective in reducing severe accident risk for this particular design, none of the deferred or cancelled plant designs involve a Mark I containment.

Indeed, there are reasons for believing that regulatory changes already made or options for change under consideration (some of which have been discussed above) would lend sub-stantial support to the conclusion that a fairly stable regulatory climate is being achieved with a relatively favorable outlook for a shortened licensing and construction schedule for both new plant applications or the reactivation of cancelled or deferred plants with relatively minor changes affecting overall costs. These include:

(1) Staff efforts to re-examine more carefully what constitutes an appropriate course (including modifications in present rules, policies, guidelines, review procedures, etc.) that would avoid the disbenefits of both over-regulation and under-regulation.

(2) On balance, it is believed that the design improvements cost-effectively justified as a result of new safety or environmental information to avolo

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-39 under-regulation might', in a majority of plant-specific cases, be roughly compensated cost-wise by a relaxation of certain tech specs or other design criteria or requirements as a result of new information revealing that safety

- margins earlier assumed for certain systems or components are excessive or that certain systems or components have only marginal importance to severe accident risk.

(3) About 855 of the TMI action plans backfit requirements (NUREG-0737) are now completed.

(4) -A final technical resolution has been achieved for 15 of the 27 identified Unre-solved Safety Issues (USIs) and substantial progress has been made on the remaining USIs. .Likewise, substantial progress has been made in the analysis / resolution of the hi h- and medium- priority Generic Safety Issues that have been prioritized in NUREG- 33.

-(5) Regulatory stability is promoted for both existing and future plant designs by the issuance of the Severe Accident Policy. Statement (NUREG-1070): (a) the inherent flexibility _of the Statement permits risk- risk tradeoffs in systems and sub-systems design and encourages thereby innovative ways of achieving an improved overall systems reliability at a reasonable cost; (b) the statement that operating nuclear power plants require no further' regulatory action to deal with severe accident issues unless significant new safety information arises to question whether there is adequate assurance of no undue risk to public health and safety; (c) in the latter event, a careful assessment shall be made of the severe accident' vulnerability posed by the issue and whether this vulnerability is plant- or site-specific, or of generic importance; and (d) a requirement that the most cost-effective options for reducing this vulnerability shall be identified and a decision shall be reached consistent with the cost-effectiveness criteria of the Commission's backfit policy as to which option or set of options (if any) are justifiable and required to be implemented.

(6) The establishment of the Committee to Review Generic Requirements and the issuance of Regulatory Analysis Guidelines (NUREG/8R-0058, Rev. 1) and the new Sackfit Rule (10 CFR 50.109) to ensure, among other purposes, that there is a substantial increase in the overall protection of the public health and safety to be derived from any proposed backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection.

(7) The achievement of substantial progress in PRA methods and implementation techniques to reduce uncertainty in regulatory decisionmaking along with growing -

results from the Severe Accident Research Program and related research by U.S.

industry and foreign sources.

(8) Cooperative efforts between NRC and such industry groups as IDCOR, EPRI, INPO, and various study groups of the AIF to share factual information and insights on the importance of safety and environmental issues and possible cost-effective modifications of design or operating and maintenance procedures deserving of staff '

review or further research effort.

Regarding the last item, it is clear that, if thers is to be an improved and stable outlook for new or reactivated nuclear power plant construction in the United States, there are certain initiatives that need be the primary responsibility of industry, other initiatives that must be the primary responsibility of government, and still others that need to involve the ::coperative or joint efforts of industry and government. A preliminary rationalization of government and industry roles regarding

a J

such initiatives is shown in Ficure 2.3 It should be noted that, in regard to government initiatives, a number of the items involving an integrated plan of policy development have already been accomplished and several others are expected to be o

completed in the next few years.

In sum, an extensive discussion of regulatory options involving change in regulatory policies, rules, review practices, etc. , might, at first blush, imply regulatory instability. However, a closer examination of the changes recently achieved or being considered or developed, plus the increasing volume of supporting data being provided by the Severe Accident Research Program and other research efforts that are

! collectively and purposely directed to improving regulatory stability and cost-effective methods for dealing with safety and environmental issues, provides an l

opposite perspective. Hence, it is believed that the culmination of these efforts toward regulatory improvements will be successful in promoting a stable regulatory outlook which, among other benefits, would portend better control over construction schedules and costs for any new plant applications or LWR project reactivations. Yet, it is recognized that the actual achievement of improved cost control over plant construction will depend also on improved project management and possibly cooperative agreements with construction unions as was consummated recently in the Limerick-2 project reactivation.

i t

3As adapted from Harold R. Denton, "The Maturing of the U.S. Nuclear Industry,"

Nuclear Engineerina and Desion, Vol. 92 (1986), p. 91.

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e a

7. REFERENCES

• NUREG-0195, " Improving Regulatory Effectiveness in Federal / State Siting Action,"

May 1977.

NUREG-0196, " Success Factor Evaluation Panel," March 1977.

MUREG-0197, " State Regulatory Activity Involved in Need for Power," April 1977.

NUREG-0198, " State Perspectives on Energy Facility Siting," March 1978.

NUREG-0199 " Environmental Planning and the Siting of Nuclear Facilities: The Integra-tion of Water, Air, Coastal, and Comprehensive Planning into the Nuclear Siting Process," February 1977.

NUREG-0200, " Federal / State Regulatory Permitting Actions in Selected Nuclear Power Station Licensing Cases," June 1977.

NUREG-0201, " Water Supplies and the Nuclear Licensing Process," July 1977.

NUREG-0202, " Nuclear Power Plant Licensing: A New England Perspective," March 1977.

NUREG-0203, " State and Local Planning Procedures Dealing with Social and Economic Impacts from Nuclear Power Plants," January 1977.

NUREG-0204, " Alternative Financing Methods," March 1977.

NUREG-0398, " Federal-State Cooperation in Nuclear Power Plant Licensing," March 1980.

NUREG-0499, " Preliminary Statement on General Policy for Rulemaking To Improve Nuclear Power Plant Licensing," December 1978.

NUREG-0625, " Report of the Siting Policy Task Force," August 1979.

NUREG-0701, " United States Experience in Environmental Cost-Benefit Analysis for Nuclear Power Plants," August 1980.

NUREG-0718, Rev. 2 " Licensing Requirements for Pending App 1tcations for Construction Permits and Manufacturing License," January 1982.

NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

NUREG-0880, Revision I for Comment, " Safety Goals for Nuclear Power Plant Operations,"

May 1983.

NUREG-0885, Issue 4, "U.S. Nuclear Regulatory Commission Policy and Planning Guidance 1985," February 1985.

• Copies of these pubitcations by the U.S. Nuclear Regulatory Commission may be purchased by calling (202) 275-2060 or (202) 275-2171 or by writing to the Superintendent of Documents, U.S. Government Printing Office, P.O. box 37082, Washington. 0.C. 20013-l 7082 or the National Technical Information Service, Department of Commerce, 5285 Port Royal Road, Springfield, VA 22161.

l l

a t '  ;

. NUREG-0900, " Nuclear Power Plant Severe Accident Research Plan," January 1983.

NUREG-0933. "A Prioritization of Generic Safety Issues," December 1983.

NUREG-0956, " Reassessment of The Technical Bases for Estimating Source Terms,"

July.1985.

L NUREG-1024, " Technical Specifications--Enhancing the Safety Impact," November 1983.

NUREG-1050, September"Protsabilistic 1984. Risk Assessment (PRA) - Reference Document," Final Report, NUREG-1061, " Report of the U.S. Nuclear Regulatory Commission Piping Peview Committee,"

April 1985. l i

NUREG-1070, "NRC Policy on Future Reactor Designs: Decisions on Severe Accident Issues in Nuclear Power Plant Regulation," July 1985. ,

NUREG-1144, " Nuclear Plant Aging Research (NPAR) Program Plan," July 1985.  ;

NUREG-1150, " Nuclear Power Plant Risks and Regulatory App 1tcations," (in preparation).

NUREG/8R-00$8, Rev.1, " Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commissfon," May 1984. -

NUREG/CP-0036, " Proceedings of the Workshop on Nuclear Power Plant Aging," Sandia National Laboratories November 1982.

NUREG/CR-0022, "Need for Power: Determinants in the State Decisionmaking Processes,"

Center for Natural Areas March 1978.

NUREG/CR-0153, " Insights into Improving The Efficacy of Nuclear Power Plant Inspection Procedures Based Upon Risk Analysis," Batte1<e Columbus Laboratories, June 1978.

NUREG/CR-2040, "A Study of the Imp 1tcations of Applying Quantitative Risk Criteria in the Licensing of Nuclear Power Plants in the United States," Brookhaven National Laboratories, May 1981.

l NUREG/CR-2497, " Precursors to Potential Severe Core Damage Accidents: 1969-79, A Status Report," Dak Ridge National Laboratories, June 1982.

NUREG/CR-2631, "A Logical Framework For Identifying Equipment important to Safety in Nuclear Power Plants," Sandia National Laboratories, July 1983.

NUREG/CR-3591 " Precursors to Potential Severe Core Damage Accidents: 1980-1981. A Status Report," Dak Ridge National Laboratory, July 1984.

NUREG/CR-3762, " Identification of Equipment and components Predicted es 5fgnificant Contributors to Severe Core Damage," Idaho National Engineering Laboratory, May 1984. '

NUREG/CR-3818, " Report of Results of Nuclear Power Plant Aging Workshops," Sandia

, National Laboratories, May 1984.

NUREG/CR-4144, "Importance Ranking 8ased on Aging Considerations of Components included in Probabilistic Risk Assessments," Pactfic Northwest Laboratory, April 1985.

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