ML23151A487

PR-050 - 54FR52946 - Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

ML23151A487
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Issue date:	12/26/1989
From:	Taylor J NRC/EDO
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PR-050, 54FR52946
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ADAMS Template: SECY-067 DOCUMENT DATE: 12/26/1989 TITLE: PR-050 - 54FR52946 - FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL SHOCK EVENTS CASE

REFERENCE:

PR-050 54FR52946 KEYWORD: RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

STATUS OF RULEMAKING PROPOSED RULE: PR-050 RULE NAME: FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGA INST PRESSURIZED THERMAL SHOCK EVENTS PROPOSED RULE FED REG CITE: 54FR52946 PROPOSED RULE PUBLICATION DATE: 12/26/89 NUMBER OF COMMENTS: 17 ORIGINAL DATE FOR COMMENTS: 03/12/90 EXTENSION DATE: I I FINAL RULE FED. REG. CITE: 56FR22300 FINAL RULE PUBLICATION DATE: 05/15/91 OTES ON FILE LOCATED ON Pl.

'TATUS OF RULE TO FIND THE STAFF CONTACT OR VIEW THE RULEMAKING HISTORY PRESS PAGE DOWN KEY HISTORY OF THE RULE PART AFFECTEO: PR-050 RULE TITLE: FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGA INST PRESSURIZED THERMAL SHOCK EVENTS PROPOSED RULE PROPOSED RULE DATE PROPOSED RULE SECY PAPER: SRM DATE: I I SIGNED BY SECRETARY: 12/12/89 FINAL RULE FINAL RULE DATE FINAL RULE SECY PAPER: 91-062 SRM DATE: I I SIGNED BY SECRETARY: 05/06/91 STAFF CONTACTS ON THE RULE CONTACTl: ALLEN L. HISER MAIL STOP: PHONE: 492-3988 CONTACT2: MAIL STOP: PHONE:

DOCKET NO. PR-05O (54FR52946)

In the Matter of FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGA INST PRESSURIZED THERMAL SHOCK EVENTS DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT

12/20/89 12/12/89 FEDERAL REGISTER NOTICE - PROPOSED RULE 02/26/90 02/19/90 COMMENT OF MARVIN I . LEWIS ( 1) 03/08/90 03/05/90 COMMENT OF NUMARC (JOE F. COLVIN, EXEC. VICE PRESIDENT) ( 2) 03/12/90 03/12/90 COMMENT OF NUCLEAR INFORMATION AND RESOURCES SER.

(JAMES P. RICCIO) ( 3) 03/12/90 03/12/90 COMMENT OF GEORGIA POWER COMPANY

( W. G. HA IRS TON, I II) ( 4) 03/12/90 03/12/90 COMMENT OF ALABAMA POWER COMPANY (W. G. HA IRS TON, I I I) ( 5) 03/13/90 02/22/90 COMMENT OF CEOG (EDWARD C. STERLING, III . , CHAIRMAN) ( 6)

- 03/13/90 03/09/90 COMMENT OF FPL (R. J . ACOSTA, ACTING VICE PRESIDENT) ( 7) 03/14/90 03/12/90 COMMENT OF NUCLEAR UTILITY BACKFITTING & REFORM GRP (NICHOLAS REYNOLDS & DANIEL STENGER) ( 8) 03/14/90 03/01/90 COMMENT OF WESTINGHOUSE ELECTRIC CORPORATION (W. J. JOHNSON) ( 9) 03/14/90 02/12/90 COMMENT OF DUKE POWER COMPANY (HAL B. TUCKER) ( 10) 03/14/90 03/12/90 COMMENT OF DUPLICATE OF #5 (W. G. HAIRSTON, III) ( 11) 03/14/90 03/12/90 COMMENT OF DUPLICATE OF COMMENT #5 (W. G. HAIRSTON, III) ( 12) 03/15/90 03/12/90 COMMENT OF OMAHA PUBLIC POWER DISTRICT (W. G. GATES) ( 13)

- 03/15/90 03/09/90 COMMENT OF PACIFIC GAS AND ELECTRIC COMPANY (J. D. SHIFFER) ( 14)

DOCKET NO. PR-050 (54FR52946)

DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 03/15/90 03/12/90 COMMENT OF TOLEDO EDISON (DONALD C. SHELTON, VICE PRESIDENT) ( 15) 03/15/90 03/09/90 COMMENT OF CONSOLIDATED EDISON COMPANY OF NY, INC.

(STEPHEN B. BRAM, VICE PRESIDENT) ( 16) 03/16/90 03/12/90 COMMENT OF WISCONSIN ELECTRIC POWER COMPANY (C. W. FAY, VICE PRESIDENT) ( 17) 05/08/91 05/06/91 FEDERAL REGISTER NOTICE - FINAL RULE

DOCKET NUMBER

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• 91 MAY -8 P 3 :49 NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN: 3150 - ADOl Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events AGENCY: Nuclear Regulatory Commission .

ACTION: Final rule .

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations for light-water nuclear power plants to change the procedure for calculating the amount of radiation embrittlement that a reactor ves-sel receives. The pressurized thermal shock rule (PTS rule) establishes a screening criterion. This criterion limits the amount of embrittlement of a reactor vessel beltline material beyond which the plant cannot con-tinue to operate without justification based on a plant-specific analysis.

The final amendment does not change the screening criterion . The PTS rule also prescribes the procedure that must be used for calculating the amount of embrittlement for comparison to the screening criterion. The final amendment updates the procedure and makes it consistent with the one given in Regulatory Guide 1.99, Revision 2, published in May 1988.

EFFECTIVE DATE: (30 days after publication in the Federal Register).

FOR FURTHER INFORMATION CONTACT: Allen L. Hiser, Jr., Division of Engineer-ing, Office of Nuclear Regulatory Research, U.S . Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301)492-3988 .

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SUPPLEMENTARY INFORMATION:

Background

Pressurized thermal shock events are system transients in a pressur-ized water reactor (PWR) tha~ can cause severe-overcooling followed by immediate repressurization to a high level. The thermal stresses caused by rapid cooling of the reactor vessel inside surface combine with the pressure stresses to increase the potential for fracture if an initiating

• flaw is present in low toughness material. This material may exist in the reactor vessel beltline, adjacent to the car~, where neutron radiation gradually embrittles the material during plant lifetime. The degree of embrittlement depends on the chemical composition of 'the steel, especially the copper and nickel contents.

The toughness of reactor vessel materials is characterized by a 11 reference temperature for nil ductility transition 11 (RTNDT), which is determined by destructive tests of material specimens. For many reactors now in operation, toughness of the beltline materials at*room temperature is low. As temperature is raised, toughness increases slowly at first; but at the temperature'defined as RTNDT' toughness begins to increase much more rapidly. The transition in toughness from low values to h.igh that takes place above RTNDT means that vessel materials are quite tough at normal operating temperatures. Radiation embrittlement moves RTNDT to higher temperatures. Correlations based on test results for unirradiated and irradiated specimens have been developed to calculate the shift in RTNDT as a function of neutron fluence for various material compositions.

The value of RTNDT at a given time in a vessel's life is used in fracture 2

mechanics calculations to determine whether assumed ,

pre-existing flaws would propagate as cracks when the vessel is stressed.

The Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61, adopted on July 23, 1985 (SO CFR 29937), establishes a screening criterion. This screening criterion establishes a limiting level of embrittlement peyond which operation cannot continue without further plant-specific evaluation.

The screening criterion. is given- fo terms of RTNDP calculated as a func-tion of the copper and nickel contents of the material and the neutron fluence according to the procedure given in-the PTS rule, and called RTPTS to distinguish it from other procedures_ for calculating RTNDT*

The PTS rule requires each PWR licen_see to report the results of the calculations of predicted RTPTS values. for each beltline material, (in-eluding the copper, nickel and fluence values that provided the basis for the calculations) from the time he submits his report to the expiration date of the operating license (EOL). The PTS rule further provides that if RTPTS for the controiling material is predicted* to exceed the screening criierion before EOL,-the licensee should submit plans and a schedule for flux reduction programs that are reasonably practicable to avoid reaching the screening criterion. Finally, the PTS rule requifes licensees of pla~ts that would rea~h the screening crite~ion before EOL despite the flux reduction program to submit a plant-specific safety analysis justify-ing operation beyond the screening criterion. The licensee must submit the analysis at least 3 years before the plant is predicted to reach that 1 i.mit. Regulatory Guide 1.154, 11 Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors 11 provides guidance for the pr~paration_ of the report and describes acceptance criteria that the NRC staff would use.

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In response to the PTS rule, the licensees of operating reactors have submitted the fluence predictions and material composition data and these have now been accepted. Of greater importance are the flux reduction programs that have been undertaken by licensees for those plants having high values of RTPTS. { SY .F"' ,S-2.Jfij On December 26, 198{the Commission published .the proposed rule to change the procedure for calculating RTPTS to reflect recent findings that embrittlement is occurring faster than.predicted by the PTS*rule for some reactor vessel materials. Although the PTS,rule was adopted on July 23, 1985, the procedure for calculating RTPTS was developed in 1981-1982 and not updated because a number of licensees were using the 1982-formulations as the basis for flux-reduction programs. Meanwhile, plant surveillance data were being added to the data base and there were extensive new and more accurate correlations made. These culminated in Revision 2 to Regulatory Guide 1.99, 11 Radiation Embrittlement of Reactor Vessel Materials,' 1 published in May 1988. Revision 2 provides the basis for pressure-temperature limit calculations. Peer review_of the new correlations was pr~vided by the public comments on Revision 2.

In the regulatory analysis prepared for Rev-ision 2, and repeated in the regulatory analy~is for this amendment, the NRC evaluated the impact of amending the PTS rule t9 be consistent with the Guide. Copper and nickel contents and fluence values for each PWR reactor vessel were tak~n from the PTS submittals from licensees. When the values of RTPTS ~ere recalculated using ihese quantities and the procedure developed for Revi-sion 2, the results were higher for approximately half the vessels, including three vessels where the value may be over 60°F higher than 4

previously thought. This would increase the probability of PTS-induced vessel failure by a factor of at least 30 for those plants.

The NRC believes these changes in the nonconservative direction are greater than can be absorbed by the uncertainties believed to exist and taken into account by the NRC when the RTPTS -based screening limit was set. (A margin of 48°F is added in the calculation of RTPTS to cover not only the uncertainty in the formula for embrittlement but also the uncertainties in the copper, nickel, and fluence values entered in the formula.) Based on this new information, the probability of reactor ves-sel failure by fracture during a PTS event is presently higher in some vessels than the probability based on the procedure for calculating RTPTS which is given. _in the present PTS rule. Moreover, a few of those reactor vessels will reach the screening criterion in the 1990 1 s. Thus, the current PTS rule needs to be amended.

A 75 day comment period expired on March 12, 1990. Comments were received from 15 respondents.

Summary of Public Comments The proposed amendments have been modified in response to the comments received and will be published in final form, as modified, to become effective 30 days after publication of this*final rule. Changes were made in response to the public comments to introduce flexibility and technical improvement in the calculation of RTPTS by requiring con-sideration of the plant-specific surveillance data and operating condi-tions when they would have a significant effect on the date the screening criterion would be reached. Another change was made to loosen the 5

reporting schedule for licensees whose reactor vessels will not become highly embrittled. A summary of the public comments and staff responses fol lows:

1. Validity of a Limited Revis-ion Several comments questioned broad issues in the PTS rule and _urged that a limited revision not be undertaken. Some comments said that the screening criterion should be raised (made less conservative) because they believed that the calculated probability of fracture would be reduced if the new embrittlement formula was substituted for the old in those calcu-lations. Other comments pointed out changes in the assumptions about flaw size and location, as well as updated information about the expected severi'ty and frequency of PTS transients as reasons to revisit the screen-ing criterion. Still other convnents questioned the use of a single para-meter, RTPTS' in the screening criterion and asked for consideration of a multiparameter criterion.

Staff Response A general response to the comments is as follows. First, the scope of the proposed amendment is narrow:* to make, technical corrections-in the embrittlement formula for calculation of RTPTS values to compare to the screening criterion. A general revision of the PTS rule must wait until further research is done. Second, the screening criterion is not a safety limit. It is a tripwire which triggers a plant-specific safety analysis, i.e., it defines which licensees need to do that analysis and wh~n it should be done. Third, the screening criterion is not linked directly to a predicted-frequency of through-wall cracking. Only when 6

the plant-specific analysis is done (using plant-specific systems and fracture parameters) is the criterion for continued operation based on a through-wall crack frequency of 5xl0-G per reactor year. It is Regulatory Guide 1.154, 11 Fonnat and Content of Plant-Specific Pres-surized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors 11 (not 10 CFR 50.61) which states that this frequency is the staff's primary acceptance criterion for continued operation.

In specific response to the issue of conservatism, the Regulatory Analysis for the proposed rule summarized the results of some studies of the effects on through-wall crack frequency when calculated using the proposed embrittlement formula instead of the one used in the original PTS rule and Monte Carlo analyses done earlier. These studies showed that the PTS rule is more conservative than previously thought for some accident scenarios, but not for all. The results did not justify raising the screening criterion.

2. Alternative Use of Plant-Specific Surveillance Data Eleven out of fifteen convnents urged the addition of this alterna-tive to the proposed RTPTS calculation method based on copper and nickel contents and fluence, noting that this alternative is allowed in calculat-ing pressure-temperature limits using R-. G. 1. 99. 1 The strongest need for this alternative is for plants nearing the screening criterion. In the plant-specific PRA (probabilistic risk analysis) required as the basis for allowing a plant to operate beyon~jscreening criterion, any 1Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Revision 2, May, 1988.

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embrittlement information may be used if justificati,on is given. Noting this, commenters said that the use of plant-specific surveillance data would in some cases make the PRA results favorable; therefore, it should be permissible to use such data in calculating RTPTS' thereby avoiding the time and expense of the PRA analysis.

Staff Response The proposed amendment to the PTS rule is prescriptive on the issue of calculating RTPTS' because not many plants meet the criteria for 11 cre-dible11 surveillance data given in R.G. 1.99 in all respects and because the criteria are somewhat subjective. Lengthy disputes over credibility are anticipated, based on experience in applying R.G. 1.99 elsewhere.

Moreover, in many cases there is a difficult choice to be made between reliance on a very.small amount of plant-specific surveillance data, or a calculated value based on a large data base of specimens most of which were irradiated in other reactors.

Nevertheless, in response to the widespread comments, it is agreed that there is need for some flexibility in the PTS rule to permit consi-deration ,of all available information. A new paragraph (b)(3) has been added and the existing paragraphs (b)(3) through (b)(G) have been renum-bered. The intent of the new paragraph (b)(3) is to provide flexibility.

for-use in two kinds of special situations. Commenters dwelt on the situation where surveillance data showed the vessel to be significantly less embrittled than indicated by the proposed embrittlement formula.

In the other situation, there is information from surveillance data or other information such as the operating temperature of the reactor vessel that shows the vessel may be significantly more embrittled than calculated

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by the proposed rule. 2 Thus, some flexibility has been added to the rule to ensure that significant information is not ignored.

Several of the commenters on this issue recommended that Position C 2, 11 Surveillance Data Available, 11 as well as the criteria for credibil-ity of the surveillance data, given in R.G. 1.99, be incorporated in the PTS rule in total. The staff has rejected this suggestion in an effort to keep the implementation of the PTS rule as simple as possible. It is anticipated that only those licensees whose vessels are approaching the screening criterion would make use of paragraph (b)(3). Its use requires review and approval by the staff, at which time the guidelines in R.G. 1.99 may be appropriate, but not necessarily so.

3. Use of Measured Values of RTNDT Several co11111ents said that the changes in wording of the requirement in paragraph (b)(2)(i) that "measured values must be used if available ... 11 represented a change in the rule which reduced its flexibility.

Staff Response There is no change in intent. The words were changed in the proposed rule to remove any ambi,guity. A further clarification was made in the final rule by adding the words 11 if credible values are available. 11 The intent is to allow a licensee to offer justification for not using a particular measured value if he does not have confidence in it.

2The irradiation temperatures represented in the data base that was correlated to obtain the formula in the*PTS rule ranged from 525 to 590 degrees Fahrenheit. Operation below that temperature range is considered to cause more embrittlement.

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4. Only a few plants are affected significantly, but the proposed rule adds a regulatory burden on all and a public relations burden also.

The proposed PTS rule 11 reorders 11 the list of reactor vessels in terms of their sensitivity to PTS events, and should be revised to reduce these impacts by increasing flexibility in the requirements or by a multiparameter approach.

Staff Response To limit the effort required by the industry, the PTS rule prescribes a screening criterion to separate out those plants that should do the PRA analysis, based on the level of embrittlement of the reactor vessels, i.e.,

the rule describes who should do the analysis, and when they should do it.

Yet, the foregoing connents request that either some kind of intermediate screening procedure be established that considers several parameters instead of only RTPTS' or that the objective should be accomplished by introducing flexibility into the rule.

The staff has rejected the suggestion of a 11 mini-PRA 11 as an inter-mediate procedure, because that opens the door to very misleading con-clusions. When the PTS rule was in the early formative stages, there were proposals for a deterministic criterion. However, it soon became clear that there was no way to choose the design transient from among the array of transients of increasing severity but lower frequency.

Extending this reasoning to PRAs, the staff concludes that a partial PRA is inappropriate. _These comients have been rejected, but paragraph (b)(l) has been modified to reduce the reporting burden for all plants except those expected to reach the screening criterion before the end of 10

their operating life. These modifications are in addition to the amendments to paragraph (b)(l) that were published in the proposed rule to simplify the reporting requirements.

5. Use of 11 Adeguate Protection 11 Exception to the Backfit Rule One comment said that flexibility in granting exemptions to the rule or exceptions to the required submittal schedules would be reduced if exception was taken to the backfit rule (10 CFR 50.109) on the basis that the amendments to the PTS rule were needed to provide adequate protectio~ to the health and safety of the public.

Staff Response The staff has continued to cite 11 adequate protection, 11 because it believes that the amendment to the PTS rule is necessary to assure that there is no undue risk to public health and safety from pressurized 11 thermal shock. Characterizing the amendment as necessary to assure adequate protection 11 does not preclude the NRC from granting exemptions to the rule, so long as licensees propose alternatives which assure ade-quate protection. The staff also notes that the PTS rule, paragraphs (b)(5), (b)(G) and (b)(7), provides procedures for the kind of case-by-case review that would normally be the basis for an exemption. There is even what amounts to an appeal procedure in paragraph (b)(7) whereby a licensee whose plant-specific analysis and proposed corrective actions are not approved can again request consideration of additional modifica-tions to equipment, systems and operation of the facility in addition to those previously proposed.

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Finding of No Significant Environmental Impact The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule is not a major Federal action significantly affecting the quality of the human environment and therefore an environ-mental impact statement is not required.

The PTS rule is one of several regulatory requirements the function of which is to ensure reactor vessel integrity. This amendment to the PTS rule updates the procedure for calculating the level of embrittlement of the reactor vessel beltline as a result of neutron radiation. Use of the updated procedure will not result in any adverse changes in power level, effluents, or other operational characteristics of a nuclear power reactor.

Therefore, this rule is not expected to have any significant effect on the environment. Moreover~ since the use of the updated procedure is likely to result in more accurate and conservative predictions of transition to nil ductility, the risk of an accident and attendant environmental consequences is likely to be reduced under the new amended rule.

The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.

Single copies of the environmental assessment and the finding of no sig-nificant impact are available from Allen L. Hiser, Jr., Division of Engineer-ing, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555, Telephone: (301) 492-3988.

Paperwork Reduction Act Statement This rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). These information collection requirements were approved by the Office of Management and Budget Approval No. 3150-0011.

Public reporting burden for t~is collection of information is esti-mated to average approximately 331 hours0.00383 days <br />0.0919 hours <br />5.472884e-4 weeks <br />1.259455e-4 months <br /> per response, including time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send connents regarding this burden estimate or any other aspect of this collection of information, including suggestions for re-ducing this burden, to the Information and Records Management Branch (MNBB-7714), Division of Information Support Services, Office of Information Resources Management, U.S. Nuclear Regulatory Cormnission, Washington, D.C. 20555; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-3019, (3150~0011), Office of Management and Budget, Washington, D.C. 20503.

Regulatory Analysis The NRC staff prepared a regulatory analysis for the final rule-,

which describes the factors and alternatives considered by the Commis-sion in deciding to propose this rule.

A copy of the regulatory analysis is available for inspection and copying for a fee at the NRC Public Document Room, 2120 L Street NW.

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(Lower Level), Washington, DC 20555 .. Single copies of the analysis may be obtained from Allen L. Hiser, Jr., Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone, (301)492-3988.

Regulatory Flexibility Act Certification As required by the Regulatory Flexibility Act, 5 U.S.C. 605(b), the Cooroission certifies that this rule does not have a significant economic impact on a substantial number of small entities. This rule specifies minimum fracture toughness properties of irradiated pressure vessel mate-rials to ameliorate the effects of PTS events on nuclear facilities licensed under the provision of 10 CFR 50.21(b) and 10 CFR 50.22. The companies that own these facilities do not fall within the scope of the definition of 11 small entities 11 as set forth in the Regulatory Flexibility Act or the Small Business Size Standards in regulations issued by the Small Business Administration at 10 CFR Part 121.

Backfit Analysis The NRC has- concluded, on the basis of the documented evaluation required by 10 CFR 50.109(a)(4), that the backfit requirements contained in this amendment are necessary to ensure that the facility provides ade-quate protection to the public health and safety, and, therefore, that a backfit analysis is not required and the cost-benefit standards of 10 CFR 50.109(a)(3) do not apply. The documented evaluation given in the regula-tory analysis includes a statement of the objectives of and reasons for 14

the backfits that would be required by the rule and sets forth the basis for the NRC 1 s conclusion that these backfits are not subject to the cost-benefit standards of 10 CFR 50.109(a)(3).

List of Subjects 1n 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire preven-tion, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, and 5 U.S.C. 553, the NRC is adopting the following amendments to 10 CFR Part 50.

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PART 50 -- DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

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1. The authority citation of Part 50 is revised to read as follows:

AUTHORITY: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat.

936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat.

1244, as amended (42 U.S.C. 2132, -2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended 1244, 1246, (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L.95-601, sec. 10, 92 Stat.

2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. 936, 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332) .. Sections 50.13, 50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C.

2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332).

Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat. 2073 (42 U~S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80-50-81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under sec. 187, 68 Stat 955 (42 U.S.C. 2237).

For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C.

2273), §§ 50.46(a) and (b), and 50.54(c) are issued under sec. 161b, 16li and 1610, 68 Stat. 948, as amended (42 U.S.C. 220l(b)); §§ 50.7(a),

50.lO(a)-(c),

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50.34(a) and (e), 50.44(a)-(c), 50.46(a) and (b), 50.47(b), 50.48(a),

(c), (d), and (e), 50.49(a), 50.54(a), (i), (i)(l), (1)-(n), (p), (q),

(t), (v), and (y), 50.55(f), 50.55a(a), (c)-(e), (g), and (h), 50.59(c),

50.60(a), 50.62(c), 50.64(b), and 50.SO(a) and (b) are issued under sec.

161i, 68 Stat. 949, as amended (42 U.S.C. 2201 (i)); and§§ 50.49(d),.

(h), and (j), 50.54(w),(z),(bb),(cc), and (dd), 50.55(e), 50.59(b),

50.6l(b), 50.62(b), 50.70(a), 50.71(a)-(c) and (e), 50.72(a), 50.73(a) and (b), 50.74, 50.78, and 50.90 are issued under sec. 161(0), 68 Stat.

950, as amended (42 U.S.C. 2201(0)) .

2.** In§ 50.8, paragraph (b) is revised to read as follows:

§ 50.8 Information collection requirements: 0MB approval.

(b) The approved information collection requirements contained in this part appear in§§ 50.30, 50.33, 50.33a, 50.34, 50.34a, 50.35, 50.36, 50.36a, 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 50.60,' 50.61, 50.63, 50.64, 50.71, 50*.72, 50.80, 50.82, 50.90, 50.91, and Appendixes A, B, E, G, H, I , J , K, M, N, 0, Q, and R.

1'- ~ ~

3. In§ 50.61, paragraph (b) is revised to read as follows:

§ 50.61 Fracture toughness requirements for protection against pressurized thermal shock events.

(b) Requirements.

(1) For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall submit projected values of RTPTS for reactor vessel beltline materials by giving values for the time of submittal, the expiration date of the operating license, 17

the projected expiration date if a change in the operating license has been requested, and the projected expirati_on date of a renewal term if a request for license renewal has been submitted. The assessment must use the calculative procedures given in paragraph (b)(2) of this section.

The assessment must specify the bases for the projection, including the

.assumptions regarding core loading patterns. The submittal must list the copper and nickel contents, and the fluence values used in the calcula-tion for each beltline material. If these quantities differ from those submitted in response to the original PTS rule and accepted by the NRC, justification must be provided. If the value of RTPTS for any material in the beltline is projected to exceed the PTS scr~ening criterion before the expiration date of the operating license or.the proposed expiration date if a change in the license has been requested, or the end of a renewal term if a request for license renewal has been submitted, this assessment must be submitted by (6 R10nths after the effective dat~of this section). *otherwise, this assessment must be submitted with the next update of the pressure-temperature l im*its, or the next reactor vessel material surveillance report, or 5 years from the effective date of this rule, whichever comes first. These submittals must be updated whenever there is a significant change in projected values of RTPTS' or upon a request for a* change in the expiration date for operation of the facility.

(2) The pressuriz.ed thermal shock* (PTS) screening criterion is 270°F for plates, forgings, and axial weld materials, or 300°F for cir-cumferential weld materials. For the purpose of comparison with this criterion, the va 1ue of RT PTS. for the reactor vessel must be calculated 18

1rf :f)i.1 ,rec f),n 0 as follows, except as provided in paragraph (b)(3)rThe calculation must be made for each weld and plate, or forging, in the reactor vessel beltline.

Equation 1: RTPTS =I+ M+ ~RTPTS (i) 11 111 means the initial reference temperature (RTNDT) of the unirradiated material measured as defined in the ASME Code, Paragraph NB-2331. Measured values must be used if credible values are available; if not, the following generic mean values must be used: 0°F for welds made with Linde 80 flux, and -56°F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.

(ii) 11 M' 1 means the margin to be added to cover uncertainties in the values of initial RTNDT' copper and nickel contents, fluence and the calculational procedures. In Equation 1, Mis 66°F for welds and 48°F for base metal if generic values of I are used, and Mis 56°F for welds and 34°F for base metal if measured values of I are used.

(iii) b.RTPTS is the mean value of the adjustment in reference tempera-ture caused by irradiation and should be calculated as follows:

Equation 2: aRT = (CF)f (0.28-0.10 log f)

PTS (iv) CF (°F) is the chemistry factor, a function of copper and nickel content. CF is given in Table 1 for welds and in Table 2 for base metal (plates and forgings). Linear interpolation is permitted. In Tables 1 and 2 "Wt-% copper 11 and "Wt-% nickel 11 are the best-estimate values for the material, which will normally be the mean of the measured values for a 19

plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld. If these values are not available, the upper limiting values given in the material specifications to which the vessel was built may be used. If not available, conservative estimates (mean plus one standard deviation) based on generic data 1 may be used if justification is provided. If none of these alternatives are available, 0.35% copper and 1.0% nickel must be assumed.

(v) 11 f 11 means the best estimate neutron fluence, in units of 1019

• n/cm 2 (E greater than 1 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service in question.

1oata from reactor vessels fabricated to the same material specification in the same shop as the vessel in question and in the same time period is an example of 11 generic data. 11 20

TABLE 1 CHEMISTRY FACTOR FOR WELDS, OF Copper, Nickel, Wt-%

Wt-% 0 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02 21 26 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 24 43 54 54 54 54 54 0.05 26 49 67 68 68 68 68 0.06 29 52 77 82 82 82 82 0.07 32 55 85 95 95 95 95 0.08 36 58 90 106 108 108 108 0.09 40 61 94 115 122 122 122 0.10 44 65 97 122 133 135 135 0.11 49 68 101 130 144 148 148 0.12 52 72 103 135 153 161 161 0.13 58 76 106 139 162 172 176 0.14 61 79 109 142 168 182 188 0.15 66 84 112 146 175 191 200 0.16 70 88 115 149 178 199 211 0.17 75 92 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191 220 238 0.20 88 104 129 160 194 223 245 0.21 92 108 133 164 197 229 252 0.22 97 112 137 167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 239 268 0.25 110 126 148 176 209 243 272 0.26 113 130 151 180 212 246 276 0.27 119 134 155 184 216 249 280 0.28 122 138 160 187 218 251 284 0.29 128 142 164 191 222 254 287 0.30 131 146 167 194 225 257 290 0.31 136 151 172 198 228 260 293 0.32 140 155 175 202 231 263 296 0.33 144 160 180 205 234 266 299 0.34 149 164 184 209 238 269 302 0.35 153 168 187 212 241 272 305 0.36 158 172 191 216 245 275 308 0.37 162 177 196 220 248 278 311 0.38 166 182 200 223 250 281 314 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 21

a

~

TABLE 2 CHEMISTRY FACTOR FOR BASE METAL, OF Copper, Nickel, Wt-%

Wt-% 0 0.20 0.40 0.60 0.80 1.00 1. 20 I

0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02 20 20 20 20 20 20 20 0.03 20 20 20 20 20 20 20 0.04 22 26 26 26 26 26 26 0.05 25 31 31 31 31 31 31 0.06 28 37 37 37 37 37 37 0.07 31 43 44 44 44 44 44 0.08 34 48 51 51 .51 51 51 0.09 37 53 58 58 58 58 58 0.10 41 58 65 65 67 67 67 0.11 45 62 72 74 77 77 77 0.12 49 67 79 83 86 86 86 0.13 53 71 85 91 96 96 96 0.14 57 75 91 100 105 106 106 0.15 61 80 99 110 115 117 117 0.16 65 84 104 118 123 125 125 0.17 69 88 110 127 132 135 135 0.18 73 92 115 134 141 144 144 0.19 78 97 120 142 150 154 154 0.20 82 102 125 149 159 164 165 0.21 86 . 107 129 155 167 172 174 0.22 91 112 134 161 176 181 184

- 0.23 0.24 0.25 0.26-0.27 0.28 95 100 104 109 114 119 117 121 126 130 134 138 138 143 148 151 155 160 167 172 176 180 184 187 184 191 199 205 211 216 190 199 208 216 225 233 194 204 214 221 230 239 0.29 124 142 164 191 221 241 248 0.30 129 146 167 194 225 249 257 0.31 134 151 172 198 228 255 266 0.32 139 155 175 202 231 260 274 0.33 144 160 180 205 234 264 282 0.34 149 164 184 209 238 268 290 0.35 153 168 187 212 241 272 298 0.36 158 173 191 216 245 275 303 0.37 162 177 196 220 248 278 308 0.38 166 182 200 223 250 281 313 0.39 171 185 203 227 254 285 317 0.40 l75 189 207 231 257 288 320 22

I (3) To verify that the values of RTPTS calculated as required by paragraph (b)(2) of this section are bounding values for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not 'limited to the reactor vessel operating temperature and surveillance results. Results from the plant-specific surveillance program shall be integrated into the embrittlement estimate if, (i) the plant-specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99 Revision 2, and (ii) the RTPTS value changes significantly2.

Any information that is believed to improve the accura_cy of the RTPTS value significantly shall be reported to the Director, Office of Nuclear Reactor Regulation. Values of RTPTS that have been modified using the pro-cedures of this paragraph are subject to the approval of the Director, Office of Nuclear Reactor Regulation when used as provided in this R~lw sedi,~

(4) For each pressurized water nuclear power reactor for which the value of RTPTS for any material in the beltline is projected to exceed the PTS screening criterion before the expiration date of the operating license, or the projected expiration data if a change in the license has been requested, or the end of a renewal term if a request for license renewal has been submitted, the licensee shall submit by (9 months after 2changes to RTPTS values are considered significant if either the value determined in paragraph (b)(2) of this section or the alternate value determined in paragraph (b)(3) of this section, or both values, exceed the screening criterion, prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.

23

the effective date of this section) an analysis and schedule for imple-mentation of such flux reduction programs as are reasonably practicable to avoid exceeding the PTS screening criterion set forth in paragraph (b)(2) of this section. The schedule for implementation of flux reduction measures may take into account the schedule for submittal and anticipated CoDHDission approval of detailed plant-specific analyses, submitted to demonstrate acceptable risk at values of RTPTS above the screening limit due to plant modifications, new information or new analysis techniques .

(5) For each pressurized water nuclear power reactor for which the analysis required by paragraph (b)(4) of this section indicates that no reasonably practicable flux reduction program will prevent the value of RTPTS from exceeding the PTS screening criterion before the expiration date of the operating license, or the projected expiration date if a change.in the operating license has been requested, or the end of a renewal term if a request for license renewal has been submitted, the licensee shall submit a safety analysis to detennine what, if any, modi-fications to equipment, systems, and operation are necessary to prevent potential failure of the reactor v~ssel as a result of postulated PTS events if continued operation beyond the screening criterion is allowed.

In the analysis, the licensee may determine reactor vessel materials properties based on available information, research results, and plant

~

surveillance data, and may use probabilistic fracture mechanics tech-niques. This analysis must be submitted at least 3 years before the value of RTPTS is projected to exceed the PTS screening criterion or by one year after the effective date of this amendment, whichever is later.

24

.I

\

(6) After consideration of the licensee's analyses (including effects of proposed corrective actions, if any) submitted in accordance with paragraphs (b)(4) and (b)(5) of this section, the Comrnissio~ may, J

on a case-by-case basis, approve operation of the facility at values of RTPTS in excess of the PTS screening criterion. The Commission will con-sider factors significantly affecting the potential for failure of the reactor vessel in reaching a decision.

(7) If the Commission concludes, pursuant to paragraph (b)(6) of this section, that operation of the facility at values of RTPTS -in excess of the PTS screening criterion cannot be approved on the basis of the licensee's analyses submitted in accordance with paragraphs (b)(4) and (b)(5) of this section, the licensee shall request and receive Collll1lission approval prior to any operation beyond the criterion. The request must be based upon modifications to equipment, systems, and operation o_f the facility in addition to those previously proposed in the submitted anal-yses that would reduce the potential for failure of the reac.tor vessel due to PTS events, or upon further analyses based upon new information

• or improved methodology.

-;£ Dated at Rockville MD, this , - day of ~ , 1991.

For the NucleafRegu~atory Co111J1ission.

es M. Taylo ecutive Director for Operations 25

DOCKET NUMBER PII o 5 PROPOSED RULE~-~~

( 54 FR 5:;;._q4rc,)

W1scons1n 00(.;KEiED Electnc USNRC POWER COMPA NY 231 W. Michiga n, PO. Box 2046, Milwaukee, WI 53201 -ro HAR 16 P4 :16 (414) 221 -2345 OFF!CE OF SECR£1ARY VPNPD-90-128 DOCKfi ING&. S[RV IC f.

NRC-90-026 BRAN CH March 12, 1990 Mr. Samuel J. Chilk, Secretary U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention: Docketing and Service Branch Gentlemen:

COMMENTS ON PROPOSED RULE §10 CFR 50.61 FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL SHOCK EVENTS 54 FED. REG. 52946-52950 (DECEMBER 26, 1989)

The NRC has proposed to amend the pressurized thermal shock (PTS) rule for determining allowable reactor vessel embrittlement to make it consistent with the procedure given in Regulatory Guide 1.99, Revision 2. We have reviewed the proposed rule and offer the following comments.

The original PTS rule established screening criteria for minimizing risk of vessel failure and provided a method of calculating a material's RTNor (termed RTPTs) for comparison to the criteria. Numerous deterministic and probabilistic studies clearly demonstrated that the RTPrs correlation and present PTS screening criteria (270°F and 300°F) are a linked, conservative index and limit, respectively, of vessel failure probability due to PTS events. The intent then of the PTS rule is to cause plants to perform a rigorous safety analysis and apply remedial measures before PTS/vessel failure risk (as indicated by RTPrs of vessel materials approaching the screening criteria) becomes critical. We believe the original PTS rule is still adequate for its intended screening purpose as discussed below.

Supplementary information in the subject Federal Register notice indicates that the RTNor correlations of Regulatory Guide 1.99, Revision 2, are needed in the PTS rule to account for findings that embrittlement is occurring faster in some materials than predicted by RTPrs due to the effects of nickel content. This regulatory analysis provides a perception that the current PTS Acknowledged by ca A subsidial}' of Wisconsin Enel!JY ColJ)Omtioo

U.S. NUCLEAR REGULATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETAR¥ OF THE COMMISSION Document Statistics Postmark Date _ :3+-rf _!3-+,{__1_6 Copies Received _ _._/_ _ _ _ __

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NRC Document Control Desk March 12, 1990 Page 2 rule is non-conservative. This focus may not be appropriate, since approximately one half of the plants will have a reduction in the calculated RTPTs values if Regulatory Guide 1.99, Revision 2, is incorporated into 10 CFR Part 50.61. Additionally, NRC's regulatory analysis of the proposed PTS rule change indicates that, if detailed risk analyses were performed for PWR's utilizing Regulatory Guide 1.99, Revision 2, then the screening limits in the rule should be increased to maintain equivalent levels of safety. Furthermore, the proposed change addresses only the RTNDT term; it neglects research related to pressurized thermal shock that indicates additional margins of safety due to improved understanding of crack arrest, flaw distributions, and mixing during SBLOCA events.

For the above reasons, we believe that the NRC's proposed change to the PTS rule is not an improvement on the margins of safety for PTS events. Rather, we believe that the change merely "shuffles the deck" of plants in relation to the PTS screening criteria. This upsets the planning of many licensees who are trying to manage vessel embrittlement and may reduce the overall industry goal of improving vessel integrity by creating indecision and delay.

Some have suggested that the NRC is making the PTS rule change to cause a very few plants, where the value of vessel RTNDT may be over 60°F higher than previously thought, to undertake flux reductions, safety analyses, and other remedial measures to reduce PTS risk. We submit that the NRC can do this without subjecting the entire industry to a destabilizing rule change that is not a proven, lasting, or obvious improvement to safety.

By utilizing all information and indicators at its disposal, not just the PTS trend correlation which is for screening purposes only, the NRC can ensure a licensee's vessel integrity program is adequate. Currently, the NRC utilizes information concerning plant transients, health physics, design basis, NDE results, operating procedures, operator training, and other factors to effectively regulate licensees in many areas and should do so in this case to improve vessel integrity for the aforementioned few plants without revising the PTS rule.

If Regulatory Guide 1.99, Revision 2, is ultimately implemented into 10 CFR Part 50.61, we have the following additional comment.

The current proposed approach for determination of vessel embrittlement is too narrow and prescriptive and ignores the most relevant data available to licensee. The approach in 10 CFR 50.61(b) (2) should also allow the use of credible surveillance data to determine the actual level of radiation damage for comparison to the screening criteria. Although there are numerous possibilities, two methods at least for

NRC Document Control Desk March 12, 1990 Page 3 incorporating surveillance data appear valid:

1. Allow use of Position 2.1 of Regulatory Guide 1.99, Revision 2.

2. When statistically significant populations of data exist, allow the development of material specific RTNDT trend correlations and uncertainty determinations.

These options should be explicitly permitted at all times in t~e proposed PTS rule and not just allowed when the licensee is approaching the screening criteria and is performing detailed plant-specific analyses to demonstrate acceptable risk at values of RTns above the screening criteria.

In summary, we oppose the change to the PTS rule because it does not improve the margins of safety for a majority of licensees.

Rather, the proposed change and future changes may have a destabilizing and possibly negative effect on overall vessel integrity in the industry. For those few plants which appear more susceptible to PTS due to increased embrittlement, increases in event frequency or severity, inspection results, etc.,

individual regulation--not industry regulation--is warranted.

Finally, methodology for the use of credible surveillance data must be allowed in the PTS rule for the licensee to compare that data to the screening criteria and to ultimately manage the effects of vessel embrittlement.

Very truly yours,

c. 4.u!

Vice President Nuclear Power

Stephen B. Bram DOCKET NUMBER PR S o Vice President PROPOSED RULE - J Consolidated Edison Company of New York, Inc. ( 5 Lf /-K .5Z 9 L.f~ DOCKETED Indian Point Station USNRC Broadway & Bleakley Avenue 1S@

March 9, 1990 Buchanan, NY 10511 Telephone (914) 737-8116 Re: Indian Point Unit No. 2 "St) MAR J9 P 3 :56 Docket No. 50- 247 JFF!CE: Of SECRrTARY OOCKfi ING e.. SEilVICF.

BRANC~

Secretary US Nuclear Regulatory Commission Washington, DC 20555 Attention: Docketing and Service Branch

SUBJECT:

NRC Proposed Rule Amending Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

Dear Sir:

Consolidated Edison Company of New York, Inc., as owner and operator of Indian Point Nuclear Generating Station Unit No. 2, provides the following comments on the subject proposed rule.

Regulatory Guide 1.99 provides guidance to licensees for complying with the requirements of 10 CFR 50 Appendix G, regarding normal heatup and cooldown limitations for iterations within the control of the operator. The Pressurized Thermal Shock (PTS) rule (10 CFR 50.61), however, is intended to address limits associated with hypothetical accident transient conditions.

Thus, the calculational methods and criteria differ significantly for the normal versus accident situations. Application of new, more conservative calculational techniques to the normal heatup and cooldown limitations addressed by Regulatory Guide 1.99, as contemplated by the proposed rule, wi ~l necessarily cause a further narrowing of the operating window, as recognized by Generic Letter 88-11. For this reason, any utilization of the NRC's proposed application of Regulatory Guide 1.99, Rev. 2 in the calculation of RTPTS for compliance with the PTS rule for hypothetical accidents would be inappropriate without a concurrent re-evaluation by the NRC of the cumulative and additive effects of the overall co_nservatisms underlying the analytical bases for the PTS rule. This re-evaluation should also consider appropriate rev1s1ons to the PTS screening criteria. Only then can a valid re- evaluation of remaining vessel service life be conducted.

We appreciate this opportunity to provide comments on the proposed rule.

Very truly yours, Acknowledged by card ..5.(t~.\.i~. . . . . . . .

U.S. NUCLEAR REGULATORY COMMISSION OOCKETINc'~ & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date _3-'/'-'-14-'f / .;_f_d_ ____

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cc: Document Control Desk US Nuclear Regulatory Commission Mail Station Pl-137 Washington, DC 20555 Mr. William Russell Regional Administrator - Region I US Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Donald S. Brinkman, Senior Project Manager Project Directorate I-1 Division of Reactor Projects I/II US Nuclear Regulatory Commission Mail Stop 14B-2 Washington, DC 20555 Senior Resident Inspector US Nuclear Regulatory Commission PO Box 38 Buchanan, NY 10511

POCKET NUMBER PR 50 ~

PROPOSED RULE

{5'-f FR 5r:2CJ 1/[p) &

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USNRC

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'90 t1AR 15 PS :02 A Centerior Energy Company OFFICE OF SECRETARY DONALD C. SHELTON Vice President- Nuclear DOCKETING I'. SEHVICL [419] 249-2300 ElRANCl-i Docket Number 50-346 License Number NPF-3 Serial Number 1785 March 12, 1990 Secretary, United States Nuclear Regulatory Commission Attention: Docketing and Service Branch Washington, D. C. 20555

Subject:

Proposed Rule, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 54 Fed. Reg. 52946 (December 26, 1989)

Gentlemen:

Toledo Edison has reviewed the subject proposed changes to 10CFR50.61, the Pressurized Thermal Shock (PTS) rule, and comments submitted by the Nuclear Management and Resources Council (NUMARC) and the Nuclear Utility Backfitting and Reform Group (NUBARG). Toledo Edison endorses the NUMARC and NUBARG comments.

If you have any questions concerning this matter, please contact Mr. R. W.

Schrauder, Manager - Nuclear Licensing, at (419) 249 - 2366.

Very truly yours, PWS/ssg cc: P. M. Byron, DB-1 NRC Senior Resident Inspector .

A. B. Davis, Regional Administrator, NRC Region III T. V. Wambach, DB-1 NRC Senior Project Manager cknow1 ed b

• card ...5.lt \i9.................

THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652

US. NUCLEAR REGULATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETAR¥ OF THE COMMISSION Document Statistics Postmark Date 3/ I3/ ~ D r I Cop:es Received. _ _,_ _ _ _ __

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UUliKt: I NUMBER PROPOSED RULE PR so U5J./. FR-5c:L 9'/-h)

Pacific Gas and Electric Company 77 Beale Street James D. Shiffer s1n Francisco CA 94106 , . E" Senior Vice President and

' DOCK£ T v General Manager 415/972-7000 USNRC Nuclear Power Genera ti on 415/973-4684 March 9, 1990 '90 HAR 15 PS :02 PG&E Letter No. DCL-90-069 Samuel J. Chilk, Secretary U.S. Nuclear Regulatory Commission ATTN: Docketing and Service Branch Washington, DC 20555 Re: Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Proposed Pressurized Ther.mal Shock Rule Change (10 CFR 50.61)

Dear Mr. Chilk:

This letter is in response to the announcement for comments in the Federal Register (54 FR 52946), dated December 26, 1989, to the proposed rule change for pressurized thermal shock (PTS) that specifically considers the procedure for calculating the amount of radiation embrittlement that a reactor pressure vessel receives during service and projected to the expiration date of the operating license. The following comments reflect the views of Pacific Gas and Electric Company (PG&E) to the proposed rule change to 10 CFR 50.61.

PG&E's comments can be divided into three separate concerns:

• Changing the calculational procedure for predicting the adjusted RTNDT (termed RTpTs) in the PTS Rule without adjusting the screening criteria;

• Not allowing plant-specific irradiation surveillance information to be credited in the determination of RTpTs; and

• Not allowing or defining a simplified plant-specific screening analysis instead of, or prior to the need for, an extensive plant-specific probabilistic analysis as prescribed in Regulatory Guide 1 .154.

The following discussion presents PG&E's position on each of these concerns.

The current PTS Rule contains a predictive calculational procedure for RTPTS which was used as the basis for the screening criteria limits established in 1982. The proposed PTS Rule requires a change in the method for calculating RTPTS without also adjusting the screening criteria limit. This does not take into account the interaction of the method for calculating RTPT$ and the method used for establishing the screening criteria limit.

U.S. NUCLEAR REGULATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETAR¥ OF THE COMMISSION Document Statistics P stmark Date - -3 /1.2_/ 1o_ _ _ __

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Samuel J. Chilk, Secretary March 9. 1990 PG&E Letter No. DCL-90-069 As stated in the Federal Register notice and contained in the NRC Staff's regulatory analysis, conditional failure probabilities were calculated to be lower using the new method for calculating RTPTS* Other studies conducted by the nuclear industry have also shown this decrease in conditional failure probability, and yet these results are not being considered in formulation of the proposed rule change. PG&E believes that the NRC Staff should reconsider the proposed rule change and determine either an adjustment in the screening criteria limits for those plants that now have been put in the position of exceeding the screening criteria limits or develop other methods for flexibility such as adjustments that can be made in the critical values of RTpTs without performing a complete Regulatory Guide 1.154 analysis.

Considering the second item of concern, part of the impetus for changing the PTS Rule is to make it consistent with the newer embrittlement calculational procedure contained in Regulatory Guide 1.99, Revision 2 (published in 1988).

A comparison of the two documents reveals that the revised PTS Rule would still differ from the Regulatory Guide 1.99, Revision 2 method in that plant-specific irradiation surveillance data cannot be applied to the estimation of RTPTS* This inconsistency should be eliminated to permit "credible" surveillance program data to be used (as in Regulatory Guide 1.99, Revision 2) to adjust the material-specific chemistry factor and/or reduce the additional margin term. The use of "credible" surveillance data, which is the best measure of actual vessel material behavior, may allow some plants to meet the screening criteria limits without performing a complete Regulatory Guide

1. 154 analysis. Since the NRC has placed a strong emphasis on surveillance program data, utilities should be allowed the maximum degree of benefit from these data. This would encourage utilities to maintain and enhance their current irradiation surveillance programs to better manage and monitor the effects of irradiation embrittlement.

The last point is the need for simplified plant-specific screening analysis, similar to the surveillance program, in which utilities can gain a benefit for having plant-specific conditions or operating/maintenance history which can be factored into the screening process before performing a comprehensive Regulatory Guide 1. 154 risk assessment analysis. The costs and demands for performing a complete Regulatory Guide 1.154 analysis are significant, and no utility has performed and submitted such an analysis for the Staff's review and acceptance. PG&E believes that performance of such analysis should be one of the last steps taken during evaluation of reactor pressure vessel integrity. Accordingly, PG&E encourages the NRC to consider methods for gaining benefit from state-of-the-art inservice inspection techniques or applying known plant-specific design features.

PG&E is confident that the NRC will give serious consideration to the above concerns. In particular, the issue of not allowing the use of irradiation surveillance data in determining RTPTS should be addressed. PG&E believes that the prescriptive nature of the proposed PTS Rule, coupled with the costly

Samuel J. Chilk, Secretary March 9, 1990 PG&E Letter No. DCL-90-069 nature of a Regulatory Guide 1.154 analysis, leaves utilities with very few options.

Kindly acknowledge receipt of this material on the enclosed copy of this letter and return it in the enclosed addressed envelope.

Sincerely, aaCJ J. D. Shiffer

~r cc: A. P. Hodgdon J. B. Martin M. M. Mendonca P. P. Narbut H. Rood CPUC Diablo Distribution 3096S/0080K/DY/1249

'DOCKET NUMBER PRoPosEo RULE PR 5o ~

(5 1/ FR5~/1¥0) ~

Omaha Public Power District -W HM 15 PS :02 1623 Harney Omaha, Nebraska 68102 -2247 402/536-4000 OFFICE OF SECRJTAR,Y DOCKET ING & St.tl VILf.

March 12, 1990 BRANCH LIC-90-0202 Secretary U. S. Nuclear Regulatory Commission Attn: Docketing and Service Branch Washington, DC 20555

Reference:

Docket No. 50-285 Gentlemen:

SUBJECT:

Proposed Rule Change to the Title 10 Code of Federal Regulations Part 50.61; RIN 3150-ADOl Omaha Public Power District (OPPD) has reviewed the proposed rule change to 10CFR50.61 "Fracture Toughness Requirements for Protection Against Thermal Shock Events" as contained in Volume 54 Number 246 of the Federal Register dated December 26, 1989 pages 52946-52950, and has the following comments:

1. OPPD agrees that the Regulatory Guide 1.99 Revision 02 shift correlation is presently the most appropriate method for predicting irradiation damage, since it is both the best fit to power reactor surveillance data and contains separate correlations to account for varying embrittlement sensitivity in the base metal and weldments. Thus, the Regulatory Guide 1.99 Revision 02 shift correlation is the proper procedure for calculating irradiation damage and is appropriate for incorporation into the PTS rule.

2. The NRC states that the proposed amendment makes the procedure for calculating the amount of embrittlement for PTS consistent with the procedure given in Regulatory Guide 1.99 Revision 02 . However, the Regulatory Guide Revision 02 position on the use of available surveillance data is not presently included in the rule. This difference causes an inconsistency between the PTS rule and other evaluations based on the use of Regulatory Guide 1.99 Revision 02. Specifically, the inconsistency would not permit reduction of the applied margin term when two or more credible surveillance data sets are available. Thus, the improved accuracy of Regulatory Guide 1. 99 Revision 02 would not be fully available to OPPD for incorporation into the Fort Calhoun Station (FCS) RTeTS calculation.

This could result in an overly conservative prediction of the irradiation shift and unnecessarily high and artificial RTPTS values.

45 -5124 Employment with Equal Opportunity Acknowledged by c::afd ** ~11r19,g_,..""~

Mal e/Female --i.j, )',

U.S. NUCLEAR REGULATORY COMMISSION DOCttt:TING & SERVICE SECTION OFf-11., 1 r- TLlF. SECRETARY OF TrlE: v Mfv1:s ION Document Statistics Postmark Date ~ - - - --

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U. S. Nuclear Regulatory Commission LIC-90-0202 Page 2 It is therefore recommended that the use of credible surveillance data be allowed to establish a consistent calculational method with Regulatory Guide 1.99 Revision 02 and the calculation of the RTPTS value under the proposed amendment.

3. The reason for the change in the rule to require the use of measured values of the initial reference temperature, RTNoT, is not identified. The present rule states that if a measured value is not available then specified generic values for initial RTNDT must be used. The proposed rule changes the use of initial RTNDT values by requiring the use of measured values, if available; if not, then specific generic values must be used.

This proposed change in the PTS rule forces the use of a specific measured value when available. The proper use of initial RTNDT values should be based on a technical justification for using a measured or generic value on a case-by-case basis. This type of approach would be consistent with the philosophy used in determining residual element content of reactor pressure vessel steels. Therefore, it is recommended that the second sentence in paragraph 2(b)(2)(i) be revised to include the use of either a generic or measured mean value, whichever is justified.

4. A general comment is provided with respect to the Regulatory Analysis. The Regulatory Analysis presents the perception that the current PTS rule is nonconservative. This focus may not be appropriate for some nuclear plants, however for FCS it is appropriate because there will be an increase in the calculated RTprs values if Regulatory Guide 1.99 Revision 02 is incorporated into 10CFRS0.61. It is anticipated that if detailed risk analyses are performed for FCS, the integrated risk of vessel failure would be less than presently perceived and the true risk to the plant due to PTS events would be considerably below the risk value associated with the PTS screening criterion. It is perceived by OPPD that this lower risk is consistent with the PTS basis document SECY-82-465, "NRC Staff Evaluation of Pressurized Thermal Shock" and US NRC Regulatory Guide 1.154, "Format and Content of Plant Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors".

OPPD appreciates the opportunity to comment on the proposed rule change to 10CFRS0.61 "Fracture Toughness Requirements for Protection Against Thermal Shock Events". If you should have any questions, please contact me.

Sincerely, 4V.4~

W. G. Gates Division Manager Nuclear Operations WGG/pjc c: LeBoeuf, Lamb, Leiby &MacRae A. Bournia, NRC Project Manager R. D. Martin, NRC Regional Administrator, Region IV P. H. Harrell, NRC Senior Resident Inspector

DOCKET NUMBER II PROPOSE RULE PR 5o Duke Power Company HAL B. Tucker P.O. Box 33198 ( St/- FP-. 5cJ.<it/&) Vice President Charlotte, N.C. 28242 COChC1 ED Nuclear Production USNRC (704)373-4531 DUKEPOWER

'90 t1AR 14 P4 :21 March 12, 1990 The Secretary of the Commission U. S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Docketing and Service Branch

Subject:

Fracture Toughness Requirements For Protection Against Presurrized Thermal Shock Events Proposed Rule Duke Power Company Comments

Dear Sir:

In the Federal Register (54 FR 52946) dated December 26, 1989, the Nuclear Regulatory Commission published for comments a proposed rule that would amend regulations for light-water nuclear power plants to change the procedure for calculating the amount of radiation embrittlement that a reactor vessel receives.

Duke Power endorses those comments submitted by NUMARC and NUBARG. We would also like to emphasis that the subject calculations for our nuclear plants have been performed within the past few years. Our plants are not now in any danger zone as identified in the regulatory analysis for the subject proposed rule.

Duke requests flexibility in scheduling when the next calculations are to be performed, in that, we request they be performed in conjunction with the next scheduled capsule reports.

We appreciate the opportunity to comment on the proposed rule and will discuss our comments as you deem necessary.

__s;/truly y o ~ /

=r~4~*

Hal B. Tucker DM477 /td Acknowledged by card ..1.h.tl~.a____ _

U.S. NUCLEAR REGULATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date 3 "°- O Copies Received I __ _

Add'I Copies Reproduced ~ _

Special Distribution /2 lb.5 , Pf)I-;.

tt.AlcJ...tLf..L._,

DOCKET NUMBER PROPOSED RULE PR 60

(.s*t.j FIJ., So2 9' 'It,,)

B 0)

DOCKEi ED USNRC Westinghouse Energy Systems ~ OX 355 Electric Corporation *90 MAR 14 P4 :2 ~ittsburgh Pennsylvania 15230-0355 March 1, 1990 0FF!CE OF SECRETARY DOCKETING;, SEHV!Cf NS-NRC-90-3492 BRANCH

• Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attn: Secretary, Docketing and Service Branch

Subject:

Proposed Rule, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Volume 54 Federal Register 52946, December 26, 1989 Westinghouse has reviewed the NRC proposed amendment to 10CFR50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events " and offers the following comments.

1. There is still a technical discrepancy between the use of Regulatory Guide 1.99, Revision 2 and the deterministic and probabilistic analytical basis for the original PTS Rule and screening criteria. Westinghouse recognizes that the NRC has addressed this concern as documented in their Regulatory Analysis portion of the proposed new PTS Rule document.

However, the issue is still unresolved technically. The comments that were made in our response to the proposed incorporation of Regulatory Guide 1.99, Revision 2 methods into the original PTS Rule are still valid. The prior Westinghouse letter NS-NRC-86-3124, dated 5/1/86, and the Westinghouse Owners Group letter OG-183, dated 5/15/86 are attached (Attachments 1 and 2).

2. The proposed change to the PTS Rule will cause a reordering of reactor vessels relative to predicting when they will approach or exceed the PTS screening criteria. This reordering will result from changes in radiation embrittlement trend curve predictions rather than any new physical change in the reactor vessel. It is recognized that Regulatory Guide 1.99, Revision 2 may provide a more accurate prediction of radiation embrittlement to all vessels as a group. However , on a plant specific basis, a reordering of vessels in the industry and potential changes in utility plans for flux reduction or other mitigative actions will occur as a result of a change in the prediction of a single parameter - RTpls*

Future changes may also be expected in trend curves with additiona changes required in utility actions. The PTS Rule should consider some means to minimize these changes resulting from a change in a single parameter and for better addressing reactor vessel integrity . Existing information (most of which has been developed since the PTS Rule was initially formulated) could be used to provide a multi-parameter basis to assess PTS rather than the current single parameter of RTPTS*

Aclcnowledged by cam .. 2J1.t2J.1~...............

U.S. NUCLEAR REGULATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date 3 /~/'JD Copies Receive ___ / _ _ _ __

Add'I Copies Reproduced """"".3

=---

Special Distri tior _8.!_D5 PD/!..., J

~ -

NS-NRC-90-3492 Page 2

3. There is a techn i cal question regarding the use of Regulatory Guide 1.99, Revision 2 in the PTS Rule but not allowing the use of plant specific surveil l ance capsule data in predicting RTNDT which is sanctioned in the Regulatory Guide. Although the same technical concern expressed in Item 1 above would exist if the use of surveillance capsule data were allowed in the prediction of RTpTs, this technical question should be addressed.

4. The proposed PTS Rule should allow the use of new data in lieu of generic initial RTNDT values for weld materials as they become available since those data are not a function of predicted radiation embrittlement .

If there are any questions, please contact Mr. T. A. Lordi of my staff at 412-374-4311 .

Manager Department MAW/WH/0799A cc: Pryor N. Randall Director or Engineering Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington , D. C. 20555

ATTACHMENT 1 Westinghouse Water Reactor

• Nuclear Technology D1v1s1on Electric Corporation Divisions Box355 PfttsbLr£h Peoosylvania 15230 May 1, 1986 NS-NRC-86-3124

• Rules and Procedures Branch Division of Rules and Records Office of Administration U.S. Nuclear Regulatory Comnission 1717 H Street Washington, DC 20555

SUBJECT:

Westinghouse Conments on Draft Regulatory Guide 1.99 Revision 2 (TASK ME 305-4)

Dear Sir:

Westinghouse has reviewed the NRC Draft Regulatory Guide 1.99 Revision 2,

• Radiation Damage to Reactor Vessel Materials* (TASK ME 305-4). While Westinghouse supports the overall technical conclusions and reconmendations outlined in the proposed regulatory guide, it does take exception to certain specific items. Sunrnary comnents that follow are based upon a detailed review of the proposed regulatory guide which. 1s given in Attachment l.

l. Pages 2-5, Discussion New features in the proposed guide revision 2 which involve l) assumptions of a different relationship for data fitting, 2) division of the previous single data set into one data group for base materials and a separate data group for weld materials, and 3) e11mination of previously used data from research reactors are considered by Westinghouse to be reasonable in developing acceptable procedures for calculating the effects of neutron radiation damage to the low-alloy steels currently used for light-water-cooled reactor vessels.

2. Page 7, Equation 3 A mandatory change requested by Westinghouse deals with the attenuation formula to calculate 6RTNDT at any depth (Page 7

• Equation 3 of the proposed regulatory guide). This attenuation equation is an approximation and plants should have the option of calculating it. The formula should normalize curves where the spectrum in the vessel looks like the spectrum in the surveillance capsule. The option should be given for plants to use this equation or develop their own formula with their own attenuation.

Further discussions of this type are also included in Attachment l.

1272n:4/MAW/4-86

Page 2 NS-NRC-86-3124 May 1, 1986

3. Page B, Charpy Upper - Shelf Energy, (Page 14, Figure 2)

The data (reduced pl~nt surveillance data) given does not substantiate the curves (Figure 2, Page 14) of *Predicted Decrease in Shelf Energy as a Function 2f Copper Content and Fluence 8

• For fluence above 1 x 1019 n/cnf, there appears to be a steady state condition based on Westinghouse data. Westinghouse reconmends this Figure be deleted until a new set of curves is developed.

4. Pages 17-19, Consequences, Costs and Benefits

• It is not clear that the economic impact of Revision 2 to Regulatory Guide 1.99 has been fully established, either for the heatup and cooldown pressure - temperature limit curves or for the potential implications for the Pressurized Thermal Shock (PTS) Rule Requirements. For instance, 1f the effect of the shift in the 3/4T RTNDT on the heatup pressure temperature limits were underestimated, it is possible that the costs of implementing the new Revision were underestimated.

5. Page 19, Impact on Other Requirements The calculation procedures of this draft guide are not the same as those given in the Pressurized Thermal Shock (PTS) Rule for calculating RTprs values, the reference temperatures that are to be compared with the screening criteria given in the PTS Rule. If the proposed correlation would at some future time replace the RTprs correlation in the PTS Rule, 1t must be determined if the Screening Criteria of the Rule remain appropriate.

The proposed correlation is not representative of the reference temperature correlation that was used in the deterministic and probabilistic fracture mechanics analyses used to establish and support the screening criteria. In particular for the supporting probabilistic analyses, not only is the relation between adjusted reference temperature vs. fluence different, but the distributions that are used to reflect the uncertainties in the material chemistry, initial properties, and fluence values are no longer equivalent to the *margin* terms used in the proposed correlation to obtain conservative, upper bound values of adjusted reference temperature.

Therefore, 1f the proposed correlation would be used in the future for purposes of comparison against the screening criteria, the screening values need to be verified to insure that both sides of the equation (i.e. RTprs calculated~ RTprs screening) are equivalent.

We would expect the RTprs screening values to be different, particularly for vessels limited by circumferential welds and plate material. Determination of flaw arrest can play a significant role in substantiating the screening criterion for circumferential welds 1272n:5/MAW/5-86

Page 3 NS-NRC-86-3124 May 1, 1986 and, therefore, would be affected by the new correlation versus that which exists in the PTS rule for the adjusted reference temperature versus fluence relationship. The screening criterion for plates would be significantly different because the amargin* term, a 6 ,

used to determine upper-bound values for adjusted reference temperature, has been significantly reduced in the new correlation.

Use of the proposed correlation (or other defendable alternative correlations) for the purpose of calculating risk of vessel failure to be compared against PTS-related through-wall crack frequency goal (Ref. 1) in plant specific safety analyses for PTS is appropriate since the above discrepancy would not exist.

Therefore, it is reconmended that the PTS Rule should remain unchanged since the proposed correlation can be incorporated into the plant specific safety analyses for those vessels that are projected to exceed the screening criteria during their anticipated lifetime.

More importantly, the rule should remain fixed as is so that utilities do not have to work with a moving target in establishing goals for plant specific actions, such as flux reductions, to minimize PTS concerns.

If there are any questions, please contact Mr. C. W. Hirst of my staff at 412-374-4311.

Very truly yours,

~~~ anager Nuclear Safet;*~artment MAW/jag cc: P. N. Randall, NRC Ref. 1 Draft Regulatory Guide *format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors* TASK SI 502-4.

1272n:6/MAW/5-86

ATTACHMENT 1 DETAILED TECHNICAL REVIEW OF PROPOSED REGULATORY GUIDE 1.99 REVISION 2 The proposed Revision 2 to Regulatory Guide 1.99 is based primarily on an updated version of the Regulatory Guide 1.99 Revision l trend curve for predicting the shift in RTNDT. The trend curve currently used in Regulatory Guide 1.99 Revision l does not reflect the current understanding of radiation embrittlement and a revision is needed. The proposed revision recognizes both the effect of nickel on the radiation sensitivity of the steel and the tendency of the shift to saturate at high fluences. Both of these effects have been well substantiated by numerous studies. The proposed correlation

• provides more realistic predictions of shift for all materials at high fluences. The fact that the Regulatory Guide Revision l gives unrealistic predictions was recognized in the acceptance of an alternative trend curve for use in the PTS evaluation. The proposed revision is a significant improvement over Regulatory Guide Revision 1. The process of refining the trend curves is a continuing effort and improvements can always be made. For instance, additional data can be added, but any il'lll)rovements that can be inmediately forseen will be small in comparison. However, there are a few questions that may require careful consideration before adoption. This document will attempt to outline those concerns.

The form of the fluence term used in the proposed correlation produces an effective saturation in the shift at high fluences for all materials. The proposed correlation generally predicts lower shifts for all materials at high fluences than Regulatory Guide Revision l which does not contain this strong saturation effect. However, in some materials, particularily the high copper/

low nickel steels and high nickel weldments, the proposed trend curve predicts higher shifts at low fluences. This behavior is illustrated in Figure 1. The revised trend curves were based on statistical fits to the data, but the form of the fluence term was selected to match the high fluence behavior of the steels. There is a significant amount of scatter in the low fluence data and it is possible that alternative fluence terms might fit the data equally as -

well. In situations.where low fluence behavior becomes restricting, it may be important to re-examine this portion of the trend curve. Also, it should be 9SSSQ:1D/050186

noted that the shift for moderate copper/high nickel weldments is equivalent to high copper/moderate nickel weldments at all fluences, which would be unexpected based upon prior experience. Therefore, this consideration should be further evaluated.

The attenuation of the neutron fluence through the vessel wall produces both a reduction in the magnitude of the flux and a shift in the energy distribution of the neutrons. The proposed correlation accounts for this effect by using an attenuation coefficient for the predicted shift. Several assumptions are incorporated in the attenuation coefficient.

The first assumption in the attenuation coefficient is that the shift should be correlated with the number of displacements per atom (dpa), rather than

• with the high energy neutron flux. The energy distribution of the neutrons does not change dramatically between surveillance capsules, and therefore, the ratio of dpa to high energy neutron fluence is approximately constant.

However, through the vessel wall, the ratio of dpa to high energy neutron fluence increases due to the changes in the neutron energy distribution.

Basing the damage attenuation coefficient on the decrease in dpa through the wall rather than the decrease in high energy neutron fluence has the-effect of increasing the predicted shift on the vessel outer surface. The number of displacements per atom is a widely accepted measure of radiation damage.

Because there is only a limited amount of data comparing predictions based on dpa to predictions based on high energy neutron fluence, it seems reasonable to base the regulation on dpa as that is the measure of exposure most closely linked to the processes of radiation damage.

The second assumption underlying the attenuation coefficient is that the trend curve used to obtain the shift at the surface (i.e., full trend curve) can be

. approximated by a simple power law. This assumption is required to produce the simple attenuation coefficient given in the regulation. The power law assumption eliminates much of the saturation effect from the trend curve. It is possible to calculate the attenuation through 'the vessel wall without making this simplifying assumption. The implications of the assumption are a function of the act~al neutron fluence at the vessel surface. In Figures 2, 3 and 4, the attenuation through the wall as calculated by the procedure given 95550:lD/050186

1n the proposed correlation 1s compared to attenuation based on the full trend curve using both cor~elations with dpa and fluence. Given the foregoing discussion of dpa versus fluence as a measure of exposure, it must be assumed that the dpa curve gives the most exacting prediction. Also shown on these Figures is the prediction based on Regulatory Guide Revision l attenuated'on the basis of dpa. The trend curve used for these calculations is the same low copper high nickel trend curve presented in Figure 1. At lxlO19 n/cm2 on the vessel ID, the proposed Regulatory Guide procedure will overestimate the shift at the vessel OD. Note that at this low fluence, the shift predicted by Regulatory Guide Revision 1 for this material is lower throughout the vessel thickness. Th1s behavior can be understood by examining Figure 1 and noting that at fluences below 3xlO19 n/cm2, the proposed trend curve predicts higher shifts than the Regulatory Guide 1.99 Revision l trend curve. When the ID fluence is. increased to 4xlo19 n/cm2, the Regulatory Guide Revision 2 approximation is slightly lower than the actual calculation based on the attenuation of dpa. At this higher fluence, the saturation term becomes more important and the ratio of OD shift to ID shift increases for both the dpa attenuation curve and the fluence attenuation curve. In contrast, the proposed Regulatory Guide procedure assumes that the ratio of OD shift to ID shift is constant. This effect is even more pronounced when the ID fluence is increased to ex1O19 n/cm2* In this case, the gap between the proposed Regulatory Guide and the dpa based attenuation curve is increased. Through 110st of the vessel wall the proposed Regulatory Guide procedure actually falls below the fluence based attenuation curve. It is interesting to note that Revision 1 of the Regulatory Guide continues to predict a lower shift at the vessel OD than any of the methods based on the new trend curve. Even at the vessel ID fluence of ex1O19 n/cm2, the vessel OD fluence remains below 3xlO19 n/cm 2* For 110st materials, the Regulatory Guide Revision l curve exceeds the proposed Regulatory Guide_ curve over the entire fluence range and the OD shift would be greater for Regulatory Guide Revision 1. However, the lower power law used in the proposed procedure (fluence raised to the .28 power as opposed to the .5) does imply that the percentage decrease in the shift through the vessel wall will always be smaller than the percentage decrease based on the Regulatory Guide 1.99 Revision 1 trend curve. The change towards saturation in the trend curve, which has a large benefit at high fluences, necessarily implies that the attenuation of the shift is decreased.

9555Q:1D/O5O186

The third concern about the attenuation term is where to normalize the surveillance capsule data to the vessel fluence. Although the proposed correlation states that the neutron energy spectra at the vessel ID most closely approximates the surveillance capsule position, the energy spectra at the surveillance capsule actually resembles the spectra at a position one to two inches inside the vessel. If this new normalization point were chosen, there would be a slight reduction in the shift on the vessel OD.*

The final concern about the attenuation term is the manner in which the margin term is propagated through the wall. The statistical analysis indicates that the proper margins to add to the predicted shifts are 28°F for welds and 17°F for base metals regardless of the fluence on the vessel. Obviously this statement is not entirely true because 'the margin for the irradiation term for

• zero fluence is obviously zero. The proposed Regulatory Guide takes note of this fact by stating that the margin should not exceed one half of the predicted shift at the ID of the vessel. It would seem more reasonable to limit the margin on the basis of the shift at the position in the vessel where the shift is calculated. This would eliminate the possibilty of having a 111argin on the vessel OD which would exceed the predicted shift. This could be accomplished by striking the word *surface* from the end of Section 1.1 of proposed Regulatory Guide 1.99 Revision 2.

The questions about the effects of the power law approximation, the point of normalization and the propagation of the error tem could all be eliminated by

- basing the entire trend curve on a dpa calculation. The extra effort required to calculate the dpa once the neutron flux spectrum has been determined is minimal. Westinghouse currently calculates the spectra as part of the normal dosimetry program. The trend curve could easily be restated in terms of dpa by determining the ratio of dpa to high energy neutron fluence, R.

Alternatively, an effective fluence could be determined at any position in the vessel wall by dividing the dpa value by R. This effective fast neutron fluence (which would generally exceed the actual high energy neutron fluence)

\

could then be used in the proposed trend curve. This procedure would provide a method of determining a plant specific attenuation profile for the shift.

9555Q:1D/050186

In sunmary, the proposed Regulatory Guide Revision 2 represents a signlficant technical improvement over existing Regulatory Guide 1.99 Revision 1, with several exceptions. It would seem reasonable to allow prediction of the through wall shift attenuation on the basis of the dpa calculations performed as part of the dosi~etry program. In addition the word *surface* could be struck from the end of section 1.1. Although there are a few remaining questions that merit further inquiry, they do not affect the overall results of the revised trend curve.

9555Q:1D/050186

Figure 1

.15~ Cu .8" Nl .012~ P 450----------------------------------.

' 400 300 r,.

150 100 50 0-t------~-----------------------~

0 2 Pluence (10-19 n/cm-2)

Figure 2 Models for Attenuation of Shift Low Copper, Bi1h Nickel Weld

.ffl0-----------------------------.

m Fluence - 1:x:10-19 n/cm""2 350 300 t 250

'4 J:

IQ 200 150 100 Reg. Guide 1.99 Rev. 1 llf'ith dpa 0 --.....-------r-----,----,----,----,---~---r----1 0 2 8 8 10 Distance (inches)

Figure 3 Models for Attenuation o,f Shift Low Copper, Rl1h Nickel Weld 450-------------------------------1 ID Fluence = 4:x:10-19 n/cm--2 350 300 0 ~-----r------T-------------,r------,-----r--~

0 2 4 6 8 10 Distance (Inches)

• Figure 4 Models for Attenuation of Shift Low Copper, Hi1b Nickel Weld

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

400 350 300 Reg, Guide 1.99 Rev. 1 with dpa

\of

,i:: 200 IJ 100 o-1------------------------------------1 0 2 8 8 10 Dbtance (inches)

• NS-NRC-86-3124 KEY WORDS: M Conments Draft R.G. 1.99 Rev. 2 bee: E. P. Rahe, Jr. (MNC 4-12), ll, lA R. 6. Saint-Paul (Brussels), ll, lA J. Cobian (Madrid), 1L, lA J.M. Moore (PC 3-600), ll, lA M. D. Beaumont (BLO, Bethesda), ll, lA W. J. Johnson (MNC 4-16), ll, lA C. W. Hirst, (MNC 4-01), ll, lA T. A. Meyer (MNC 3-26), ll, lA F. L. Lau (MNC 4-36), lL, lA 1 D. Sharp ( Expo Mart), ll, lA

8. S. Monty (MNC 4-13), ll, lA T. A. Lordi (R&D 701-304), lL, lA E. J. Rusniea (MNC 3-23), lL, lA S.S. Palusamy (MNC 3-24), lL, lA W. C. Gangloff (MNC 3-09), lL, lA T. R. Mager (MNC 3-21), lL, lA S. E. Yaniehko (MNC 3-26), lL, lA R. L. Turner, (MNC 3-26), lL, lA K. R. Balkey (MNC 3-26), ll, lA R. Lott (R*D 302), ll, lA 1272n:7/MAW/4-86

ATTACHMENT 2 e Commonwealth Edlaon 72 WHt Adams Street, Chicago, Illinois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 OG-183 May 15, 1986 Rules and Procedures Branch Division of Rules and Records Office of Administration U.S. Nuclear Regulatory Collntlsslon 7735 Old Georgetown Rd.

Bethesda, Maryland

Subject:

Westinghouse Owners Group Transmittal of Comments on Draft Regulatory Guide 1.99 Revision 2 (TASK ME 305-4)

Dear Sir:

The Westinghouse Owners Group(~) has reviewed the NRC Draft Regulatory Gu Ide 1 . 99 Revis l on 2, "Radiation Damage to Reactor Vesse I Mater i a Is

(TASK ME 305-4). While the ~OG supports the overall technical conclusions and recommendations outlined in the proposed regulatory guide, it does take exception to certain specific items. SUmmary comments that follow are based upon a detailed review of the proposed regulatory guide which is given in Attachment I .

I. It is rec0fllll8nded that the word "damage" be changed to embrlttlement" in the title and throughout the body of proposed Regulatory Guide 1.99 Revision 2. This change will help to eliminate a "negative" iaage associated with the subject.

2. Pages 2-5, Discussion New features in the proposed guide revision 2 which Involve I) assumptions of a different relationship for data fitting, 2) division of the previous single data set into one data group for base materials and a separate data group for weld materials, and 3) all inatlon of previously used data from research reactors are considered by the Westinghouse Owners Group to be reasonable in developing acceptable procedures for calculating the effects of neutron radiation dalllage to the low-alloy steels currently used for 11 ght-water-coo Ied reactor vesse Is. _ ,,-

0144V:12

3. Page 7 1 Equation 3 The Westinghouse Owners Group is of the opinion that a change to equation 3 on page 7 of the draft is mandatory. This equation is the attenuation fornula to calculate .lRT~o~ at any depth. Since this attenuation

• equation Is an approximation, calculating It should be a plant-specific option. The fonaula should normalize curves where the spectr1.111 in the vessel resembles the spectrun In the survelllance capsule. Plants should be given the option to use the equation In the draft or to develop their own formula with their own attenuation.

Further discussions of this type are also included In Attachllent I.

4. Page a, Charpy Upper - Shelf Energy (Page 14 1 Figure 2)

The reduced plant surveillance data given does not substantiate the curves on Figure 2 "Predicted Decrease in Shelf Energy as a Function of Copper Content and Fluence". Based on Westinghouse data there appears to be a steady state condition for fluence values above 1x1o'"n/e11 1

• The Westinghouse Owners Group reconnends this Figure be deleted until a new set of curves Is developed.

We recommend uti Ities be pemltted to use plant specific survel I lance data or published data to predict the decrease (as a function of copper content) In the upper shelf energy.

5. Pages 17-18 1 Consequences, Costs and Benefits It is not clear that the eeon0111lc Impact of Revision 2 to Regulatory Guide I . 99 has been fu 11 y es tab I i shed I e I ther for the heatup and coo I down pressure - temperature I iII U curves, or for the potent I a I hip 11 cations for the Pressurized Thermal Shock (PTS) Rule Requirements.

For instance, the proposed regulatory guide correlation generally results in much higher shifts in RT~or at lower fluences than prior correlations that were used to generate current operating limits curves. In particular, the heatup curves, which are governed by a postulated external flaw at the 3/4 wall thickness location, have the potential to become 11uch more restrictive because of the above effect. This restriction has the potential to delay plant startup because of reactor coolant pump startup delay, thereby resulting in potential replacement power costs. Therefore, It must be insured that the proposed regulatory guide correlation does not Include any undue conservatisms, such as noted in C01111ent 3 above, because of thla potential econocnic Impact.

The economic Impact for pressurized thermal shock is discussed in the next C0111&nt.

0144V:12

6. Page 19 1 Impact on Other Requirements The calculation procedures of this draft guide are not the same as those given I~ the Pressurized Thermal Shock (PTS) Rule for calculating RTPrs va Iues. If the proposed eorre Iat ion wou Id at 801118 future t i,!18 rep Iace the RTPrs correlation In the PTS Rule, it aist be deter lned If the Screening Criteria of the Rule r811181n appropriate.

The proposed correlation is not representative of the reference tllllp8rature correlation that was used In the deterministic and probabilistic fracture mechanics analyses used to establish and support the screening criteria. In particular for the supporting probabilistic analyses, not only is the relation between adjusted reference tetnperature vs. fluence different, but the distributions that are used to reflect the uncertainties in the material chellistry, initial properties, and fluence values are no longer equivalent to the " argin" terms used in the proposed correlation to obtain conservative, upper bound values of adjusted reference temperature. Therefore, if the proposed correlation would be used In the future for purposes of comparison against the screening criteria, the screening values need to be verified to insure that both sides of the equation (i.e. RTPrs calculated i RT~Ts screening) are equivalent.

We would expect the RTPTs screening values to be higher, particularly

• for vessels limited by circumferential welds and plate material.

Determination of flaw arrest can play a significant role in substantiating the screening criterion for circumferential welds and, therefore, would be positively affected by the new correlation versus that which exists In the PTS Rule for the adjusted reference temperature versus fluence relationship. The screening criterion for plates would be significantly higher because the **aarg In" term, cr ~, used to determine upper-bound values for adjusted reference temperature, has been significantly reduced in the new correlation.

Use of the proposed correlation (or other defendable alternative correlations) for the purpose of calculating risk of vessel failure to be C0111p1red against PTS-related through-wall crack frequency goal (Ref. I) in plant specific safety analyses for PTS Is appropriate since the above discrepancy would not exist.

Because the proposed correlation is an advance1119nt over the RToTs correlation In the PTS Rule, it appears to be logical to include this latest Mthodology in the PTS requirements. Significant advancements, however, have also been recently ade In the event sequence, thermal-hydraulic, and probabilistic fracture 118Chanics analysis methods that were used to support the screening criteria. These analytical advancwnts would significantly increase the screening criteria, as evidenced by results from the A-49 program and other plant specific analyses (2). To be technically conpleta, al I advancements would have to be taken into account for the 111.1 It 1-d i sc Ip 11 ned subject of ..PTS if the proposed correlation Is considered In the future to replace the RT~7; correlation.

0144V: 12

If the RTg*, correlation Is replaced without taking al I advancements Into consideration, several reactor vessels would be unnecessarily shown to have a safety concern. Utilities would be required to take unnecessary actl ons to Just If y cont Inued ope rat Ion, Inc l_ud i ng the perf or.ping of plant-specific safety analyses, that could wel I exceed $1 ml 11 ion per plant. Thia figure does not include indirect costs from unwarranted Intervenor and pub 11 c 119d Ia attent Ion that wou Id be expected to occur.

If all of the advanceaents associated with pressurized ther111&I shock would be taken Into account In a lengthy process for revising the PTS Rule, we would expect that all vessels would have equivalent or possibly even greater urglns than those calculated using the current PTS Rule requir9118nts. Revising the PTS Rule would project an unw21rranted luge to

• lntervenors and the pub I le lledia that the regulators and industry do not have a firm understanding of what constitutes a safe reactor vessel. When In fact, the work perforllled by the Nuclear Regulatory Colllmisslon and the Industry groups* in 1982 and 1983, along wl th the detailed analyses recently completed by the National Labs, does constitute a fina basis for the current PTS Rule requirements t~t properly insure an acceptably low risk of vessel failure from PTS events.

• Therefore, it Is recommended that the PTS Rule should remain unchanged.

The rule should r8118in fixed in its current form so that utilities do not have to work with a 110ving target In establishing goals for plant specific actions, such as flux reductions, to minimize PTS concerns each time an advancement In the technology Is made.

Very truly yours, L. D. Butterfield, Chairman Westinghouse Owners Group cc: P.N. Randal I, USNRC

• H. Thollpson , Jr . , USNRC WOG Representatives Materials Subc01111ittee 0144V:12

Ref. 1: Dracft Regulatory Guide "Forwat and Content of Plant-Specific Pressurized Therul Shock Safety Analysis reports for Pressurized Water Reactors" TASK SI 502-4.

Ref. 2: Turner, R.L., Balkey, K.R., end Phillips, J.H., "A Plant Specific Risk Scoping Study of Reactor Vessel Pressurized Theral Shock," Advances In Proballstlc Fracture Mechanics, PVP-Vol. 92, Aller-lean Society of Mechanical Engineers, 1984, pp.87-103.

0144V:12

ATTACHMENT 1 DETAILED TECHNICAL REVIEW OF PROPOSED REGULATORY GUIDE 1.99 REVISION 2 The proposed Revision 2 to Regulatory Guide 1.99 1s based primarily on an updated version of the Regulatory Guide 1.99 Revision 1 trend cu~ve for predicting the shift 1n RTNDT. The trend curve currently used in Regulatory Guide 1.99 Revision 1 does not reflect the current understanding of radiation embrittlement and a revision is needed. The proposed revjsion recognizes both the effect of nickel on ~he radiation sensitivity of the steel and the tendency of the shift to saturate at high fluences. Both of these effects have been well substantiated by numerous studies. The proposed correlation provides more realistic predictions of shift for all materials at high fluences. The fact that the Regulatory Guide Revision 1 gives unrealistic predictions was recognized in the acceptance of an alternative trend curve for use in the PTS evaluation. The proposed revision 1s a significant improvement on the old Regulatory Guide. The process of refining the trend curves is a continuing effort and improvements can always be made. For instance, additional data can be added, but any improvements that can be i11111ediately forseen w1.ll be small in comparison. However, there are a few questions that may require careful consideration before adoption. This document will attempt to outline those concerns.

The form of the fluence term used in the proposed correlation produces an effective saturation in the shift at high fluences for all 111terials. The proposed correlation generally predicts lower shifts for all 1111terial~ at high fluences than the current Regulatory Guide which does not contain this strong saturation effect. However, in some materials, particularily the high copper/

low nickel steels and high nickel weldments, the proposed trend curve predicts higher shifts at low fluences. This behavior is illustrated in Figure 1. The revised trend curves were based on statistical fits to the data, but the form of the fluence term was selected to match the high fluence behavior of the steels. There is a significant amount of scatter in the low fluence data and 1t is possible that alternative fluence terms might fit the data-equally as well. In situations where low fluence behavior becOrAes restricting, it may be important to re-examine this portion of the trend curve. Also, it should be 95550:1D/042986

noted that the shift for 110derate copper/high nickel weldments is equivalent to high copper/moderate nickel weldments at all fluences, which would be unexpected based upon prior experience. Therefore, this consideration should be further evaluated.

The attenuation of the neutron fluence through the vessel wall produces both a reduction in the magnitude of the flux and a shift in the energy distribution of the neutrons. The proposed correlation accounts for this effect by using an attenuation coefficient for the predicted shift. Several assumptions are incorporated in the attenuation coefficient.

The first assumption in the attenuation coefficient is that the shift should be correlated with the nUllber of displacements per atom (dpa), rather than with the high energy neutron flux. The energy distribution of the neutrons does not change dramatically between surveillance capsules, and therefore, the ratio of dpa to high energy neutron fluence is approximately constant.

However, through the vessel wall, the ratio of dpa to high energy neutron fluence increases due to the changes i~ the neutron energy distribution.

Basing the damage attenuation coefficient on the decrease in dpa through the wall rather than the decrease in high energy neutron fluence has the effect of increasing the predicted shift on the vessel outer surface. There is only a limited amount of data comparing predictions based on dpa to predictions based on high energy neutron fluence. The change from fluence to dpa as a measure

• of exposure requires further study .

The second assumption underlying the attenuation coefficient is that the trend curve used to obtain the shift at the surface (i.e., full trend curve) can be approximated by a simple power law. This assumption-is required to produce the simple attenuation coefficient given in the regulation. The power law assumption eliminates much of the saturation effect from the trend curve. It is possible to calculate the attenuation through the vessel wall without making th1s s1mplify1ng assumption. The impl1cations of the assumption are a function of the actual neutron fluence at the vessel surface. In Figures 2, 3 and 4, the attenuation through the wall as calculated by the p_rpcedure given 9555Q:1D/051386

1n the proposed correlation is compared to attenuation based on the full trend curve using both correlations with dpa and fluence. Also shown on these Figures is the prediction based on Regulatory Guide Rev1s1on 1 attenuated on the basis of dpa. The trend curve used for these calculations 1s the same low copper high nickel trend curve presented 1n F1gure 1. At lxlO19 -n/cm2 on the vessel ID, the proposed Regulatory 6u1de procedure will overestimate the sh1ft at the vessel OD. Note that at th1s low fluence, the sh1ft predicted by Regulatory Guide Revision 1 for this aater1al 1s lower throughout the vessel thickness. This behavior can be understood by examining F1gure 1 and noting that at fluences below 3x1O19 n/cm2, the proposed trend curve predicts higher shifts than the Regulatory Gu1de 1.99 Revis1on l trend curve. When the ID fluence 1s increased to 4x1O19 n/cm2, the Regulatory Guide Rev1sion 2

• approxil'A8tion is slightly lower than the actual calculation based on the attenuation of dpa. At this higher fluence, the nturation tenn becomes more important and the ratio of OD shift to ID sh1ft increases for both the dpa attenuation curve and the fluence attenuation curve. In contrast, the proposed Regulatory Guide procedure assUTRes that the ratio of OD shift to ID shift is constant. This effect is even more pronounced when the ID fluence is increased to ex1O19 n/cm2* In this case, the gap between the proposed Regulatory Guide and the dpa based attenuation curve is increased. Through most of the vessel wall the proposed Regulatory Guide procedure actually falls below the fluence based attenuation curve. It is interesting to note that Revision 1 of the Regulatory Guide continues to predict a lower sh1ft at the

• vessel OD than any of the methods based on the new trend curve. Even at the vessel ID fluence of ex1O 19 n/cm2, the vessel OD fluence remains below 3xio19 n/cm2* For 110st naaterials, the Regulatory Gu1de Revision 1 curve exceeds the proposed Regulatory Guide curve over the ent1re fluence range and the OD sh1ft would be greater for Regulatory Guide Revision 1. However, the lower power law used in the proposed procedure (fluence raised to the .28 power as opposed to the .5) does imply that the percentage decrease in the sh1ft through the vessel wall will always be smaller than the percentage decrease based on the Regulatory Gu1de 1.99 Rev1sion 1 trend curve. The change towards saturation in the trend curve, which has a large benefit at h1gh fluences, necessarily implies that the attenuation of the,...,shift is decreased.

955SQ:1D/051386

The third concern about the attenuation term is where to nonnalize the surveillance capsule data to the vessel fluence. Although the proposed correlat1on*states that the neutron energy spectra at the vessel ID most closely approximates the surveillance capsule position, the energy spectra at the surveillance capsule actually resembles the spectra at a position one to two inches inside the vessel. If this new nonnalization point were chosen, there would be a slight reduction in the shift on the vessel OD.

The final concern about the attenuation term is the manner in which the margin tena 1s propagated through the wall. The statistical analysis indicates that the proper 1111rgins to add to the predicted shifts are 2e*F for welds and 17°F for base metals regardless of the fluence on the vessel. Obviously this statement is not entirely true because the 111rgin for the irradiation term for zero fluence is obviously zero. The proposed Regulatory Guide takes note of this fact by stating that the margin should not exceed one half of the predicted shift at the ID of the vessel. lt would seem more reasonable to limit the anargin on the basis of the shift at the position in the vessel where the shift is calculated. This would eliminate the possibilty of having a IDBrgin on the vessel OD which would exceed the predicted shift. This could be accomplished by striking the word asurface* from the end of Section 1.1 of proposed Regulatory Guide 1.99 Revision 2.

The questions about the effects of the power law approximation, the point of normalization and the propagation of the.error term could all be eliminated by basing the entire trend curve on a fluence or dpa calculation. The extra effort required to calculate the dpa once the neutron flux spectrum has been determined is minimal. Westinghouse currently calculates the spectra as part of the normal dosimetry program. The trend curve could easily be restated in terms of dpa by deteraiining the ratio of dpa to high energy neutron fluence, R. Alternatively. an effective fluence could be determined at any position 1n the vessel wall by dividing the dpa value by R. This effective fast neutron fluence (which would generally exceed the actual high energy neutron fluence) could then be used in the proposed trend curve. This procedure would provide a method of determining a plant specific attenuation profile for ,, the shift.

tSSSQ:lD/051386

In sunmary, the proposed Regulatory 6uide Revision 2 represents a significant technical 1111Prove11ent over existing Regulatory Guide 1.99 Revision l, w1th several exceptions. It would seem reasonable to allow predict1on of the through wall shift attenuation on the basis of fluence or dpa calculations performed as part of the dos111etry program. In addition the word *surface*

could be struck from the end of section 1.1. Although there are a few rema1ning questions that merit further 1nqu1ry, they do not affect the overall results of the revised trend curve.

9SSSQ:1D/051386

Figure 1

.15,t cu .a,c Ni .012x P 00 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

400 300

....d

\4 200 ll 150 100 50 Q-t--------------------,-------...-------1 0 2 Pluence (10-19 n/cm-2)

Figure 2 Models for Attenuation of Shift Low Copper, Bl1h Nickel Weld ID nuance =- 1x10-19 n/cm-2 350 300 a.

UMI 100 Reg. Gui de 1 *99 Rev. 1 llf'I th dpa 0 2 B 8 10 Distance (inches)

Figure 3 Models for Attenuation of Shift Low Copper, Ht1h Nickel Weld 400 ID nuence

• 4-:rl0... 19 n/cm ... 2 300

"'....i: 200 m

0 -l---------------,-----------------------------1 0 2 4 6 8 10 Distance (Inches)

-* Figure 4 Models for Attenuation of Shift Low Copper. Blah Rickel Weld 480 - - - - - - - - - - - - - - - - - - - - - - - - - - - -

m nuence

• lx10-19 n/cm-Z 300 Reg~ Guide 1.99 Rev. 1 with dpa a.

J:

I) 200 150 100 0----------.--------------------.-----1 0 2 4 8 8 10 Distance (inches)

.... OG-183 May 15, 1986 bee _..:A.c:-Weav.,.- ~ :MNC ,.._..,.

• T. Mager - MNC 3-21 T. Lordi - MNC 232W 0144V:12

NS-NRC-90-3492 bee: W. J. Johnson (ECE 4-10), lL, lA G. Desaedeleer (Brussels), lL, lA J. Cobian (Madrid), lL, lA J.M. Moore (EXPO 335), lL, lA M. D. Beaumont (Rockville), lL, lA A. T. Paterson (ECE 4-09), lL, lA P.A. Loftus (ECE 4-14), lL, lA T. A. Lordi (ECE 407A), lL, lA T. A. Meyer (STC/701-401), lL, lA K. R. Balkey (STC/701-402), IL, IA F. L. Lau (ECE-473), IL, lA S. L. Anderson (ECE-47), IL, IA D. R. Sharp (ECE-323), IL, IA B. D. Sloane (ECE-323), lL, lA A. C. Cheung (ECE-429A), IL, IA J. S. Ivey (ECE-427), IL, IA R. A. Wiesemann (ECE-415),IL, IA R. Sero (ECW-245), IL, IA T. R. Mager (STC/?OI/4-44), IL, IA B. A. Bishop (STC/701-404), IL, IA D. C. Adamonis (STC/70I/4-44), IL, IA

DOC KE NUMBER POPO ED RULF PR 5-0

( 5'tF~S~9¥-&-)

LAW OFFI CES BISHOP, COOK, PURCELL & REYNOLDS DOCK£iEO USNRC 1400 L STREET, N.W.

... WASHINGTON, D.C. 20005 -35 02 (202) 371-5 7 00

• 90 HAR 14 A11 :25 March 12, 1990 WR ITER' S D IRECT D IAL Mr. Samuel J. Chilk Secretary U.S. Nuclear Regulatory Commission Washington, o.c. 20555 Subj: Proposed Rule, Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events. 54 Fed. Rea. 52,946 . (Dec. 26, 1989)

Dear Mr. Chilk:

In accordance with the above-referenced Proposed Rule, the following comments are submitted on behalf of the Nuclear Utility Backfitting and Reform Group ("NUBARG").1/ Our comments are limited to a single issue: the application of the NRC's backfitting rule, 10 C.F.R. § 50.109, to the proposed changes to the Pressurized Thermal Shock ("PTS") rule.

Summary of Position NUBARG generally agrees with the need to modify the PTS calculation method as discussed in the proposed rule. NUBARG does not believe, however, that the NRC should invoke the "adequate protection" exception to the backfitting rule. Many plants are so far away from the PTS screening criterion that it cannot fairly be said that the new PTS calculations are necessary for those plants to continue to provide adequate protection of the public health and safety. NUBARG further believes that the rule should contain flexibility to allow such plants to justify exemptions where redoing the PTS calculations would not appreciably alter the conclusion with respect to vulnerability of the reactor vessel to pressurized thermal shock. The rule should also allow flexibility in the schedule for the new PTS calculations, especially for those licensees who just performed PTS calculations in the last few years and are not approaching the screening criterion.

1/ NUBARG, which consists of the 25 nuclear utilities listed in the attachment hereto, actively participated in the development of the NRC backfitting rule and has followed its implementation closely.

Ackn ledg by

U.S. NUCLEAR REGULATORY COMMl8SION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETAA¥ OF THE COMMISSION Document Statistic.,

Postmark Da 3/;-+-

Coples Received-_,~__ -=------*

2/10 Add'I Copies Reproduced _3_ _ __ _

RIDS, PPR )

Mr. Samuel J. Chilk March 12, 1990 Page 2 Discussion On December 26, 1989, the NRC proposed amendments to its regulations to change the method for calculating the amount of radiation embrittlement that a reactor vessel receives. The current PTS rule (10 C.F.R. § 50.61) sets up a screening criterion that establishes a limiting level of embrittlement beyond which the plant cannot continue to operate without justification based on a plant-specific analysis. The proposed amendments would update the method for calculating the amount of embrittlement for comparison to the screening criterion. As a result, the amended rule would require licensees to recalculate the amount of embrittlement for their plants.

NUBARG generally agrees with the reason advanced by the NRC to update the PTS calculations. The stated reason for the proposed change to the PTS rule is that the NRC believes that it has significant new information about radiation embrittlement.

In its Federal Register notice of the proposed amendments, the Commission stated that recent findings have shown that embrittlement is occurring faster than originally predicted by the PTS rule for some reactor vessel materials. The NRC has recognized that without the changes to the PTS rule, about half of the plants will be operating with a reduced margin of safety.y Accordingly, the NRC proposed to amend the PTS rule to incorporate new and more accurate correlations due to new plant surveillance data.

NUBARG also agrees with the NRC's determination that implementation of the proposed amendments is a backfit, since the amendments would modify the procedures required to operate a facility. 10 C.F.R. § 50.109(a) (1). Under the backfitting rule, where the Commission has determined that an amended provision of the NRC's rules constitutes a backfit, the Commission is required to perform a "systematic and documented analysis" to justify its requirement that licensees perform the backfit. 10 C.F.R. § 50.109(a)(2).

The Commission, however, in its analysis concerning the proposed rule, invoked the "adequate protection" exception to the backfitting rule, 10 C.F.R. § 50.109(a) (4) (ii). The Staff concluded that the backfit requirements contained in the proposed amendments "are necessary to ensure that the facility provides adequate protection to the public health and safety, and, therefore, that a backfit analysis is not required and the cost-y The NRC's Regulatory Analysis for the proposed changes to the PTS rule states that "(O]f the 61 PWR's tabulated, RTpts would actually decrease in 33 cases if the PTS rule were amended."

Mr. Samuel J. Chilk March 12, 1990 Page 3 benefit standards of 10 C.F.R. § 50.109(a) (3) do not apply." 54 Fed. Reg. at 52,948. As a result, the NRC did not perform a "systematic and documented analysis" to justify the imposition of the backfit on licensees.

For the following reasons, NUBARG believes that the use of the "adequate protection" exception to the backfitting rule is inappropriate. The NRC has acknowledged that certain plants are not in danger of reaching the screening criterion. Evidence indicates that new calculations may not be justified for certain classes of plants. In the Regulatory Analysis for Rev. 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," the NRC stated that "[I]f Revision 1 is retained as the basis for pressure-temperature limits, about half of the plants will be operating with limits that provide a reduced margin of safety against vessel fracture . . . . " (Emphasis added.) As this statement implies, even without a change in the PTS calculations, about half of the plants would not be operating with reduced safety limits.

Accordingly, we do not believe that it is fair to conclude across the board that plants will present an undue risk unless the new calculations are performed. For those plants that are well below the screening criterion, this is demonstrably not so.

Rather than making an across-the-board finding that the rule change is needed for "adequate protection," the better approach would be to perform a backfitting analysis, using existing information justifying the rule change. Our review of the existing information (g_.g., "Regulatory Analysis For Proposed Amendment of the PTS Rule, 10 C.F.R. § 50.61 11 ; Reg. Guide 1.99, Rev. 2, "Embrittlement of Reactor Vessel Materials," May 1988, and supporting Regulatory Analysis) indicates that the support for a backfitting analysis has already been assembled and that the analysis could be performed readily based on this information.l/

NUBARG further recommends that additional flexibility be built into the rule, especially for plants that do not present a l/ The NRC's backfitting analysis should take into account, among other considerations, "the potential impact of differences in facility type, design or age on the relevancy and practicality of the proposed backfit . " 10 C.F.R. § 50.109(c) (8). The backfitting analysis should therefore consider plant-specific differences in vulnerability to PTS based on facility design, age, etc. The analysis may be important in that it may show that new calculations are not justified for certain classes of plants -- or at least that exemptions to the new calculation requirement could be appropriate.

Mr. Samuel J. Chilk March 12, 1990 Page 4 significant PTS concern. For example, the NRC should incorporate as part of the rule, or at least as a part of the Statement of Considerations accompanying the rule, a mechanism by which licensees are permitted to perform a preliminary analysis to determine whether new PTS calculations are truly necessary for their plants. In this way, the revised rule could allow flexibility for plants that are well below the screening criterion to seek exemptions from the requirement to redo the PTS calculations. Some plants may be so far below the screening criterion that a change in calculations is not necessary for that plant to maintain adequate protection of the health and safety of the public. The costs of recalculation for these plants could far outweigh the benefits provided by new calculations. For such plants, the NRC should recognize that they have an opportunity to seek exemptions from the new rule under 10 C.F.R. § 50.12 . .ij In addition, the rule should allow flexibility in the scheduling for the new PTS calculations. Under the proposed revisions to the rule, licensees will be required to submit their new PTS calculations within six months after the effective date of the rule. 54 Fed. Reg. at 52,948. Such a short schedule is not necessary or appropriate for those licensees who have recently (within the past few years) performed PTS calculations which showed the reactor vessel to be well within the PTS screening criterion. For such plants, the rule should allow the licensee to be able to agree with its NRC project manager on an appropriate schedule for submittal of new PTS calculations . .2J if The rule, as proposed, however, may make it more difficult for these licensees to request justifiable exemptions from the PTS rule. By concluding that the rule change is necessary to provide the minimum level of "adequate protection" required by the Atomic Energy Act, it may be more difficult for the NRC to make the requisite findings under 10 C.F.R. § 50.12 to grant an exemption. To grant an exemption under 10 C.F.R. § 50.12, the NRC must find that the exemption will not result in "undue risk" to public health and safety

-- that is to say that "adequate protection" will be maintained. The NRC should therefore be careful in adopting the proposed rule not to suggest that exemptions cannot be justified .

.2./ In this regard, the rule should allow the schedule to be worked out with the project managers on a basis that is consistent with the schedule for capsule removal and analysis.

Mr. Samuel J. Chilk March 12, 1990 Page 5 Conclusion NUBARG generally agrees with the proposed rule, but believes that the NRC should not invoke the "adequate protection" exception of the backfitting rule. Further, the rule should allow for additional flexibility for plants well below the PTS screening criterion to seek exemptions from the requirement to redo the PTS calculations or at least to work out a more reasonable schedule for submittal of the new calculations.

Counsel o Nuc ear Utility Backfitting tl Reform Group Attachment

ATTACHMENT NUBARG Members Arkansas Power & Light Company Baltimore Gas & Electric Company Cleveland Electric Illuminating Company Commonwealth Edison Company Consolidated Edison Company of New York, Inc.

Duke Power Company Florida Power & Light Company Florida Power Corporation Nebraska Public Power District N~w York Power Authority Niagara Mohawk Power Corporation Northeast Utilities Pennsylvania Power & Light Company Philadelphia Electric Company Portland General Electric Company Rochester Gas and Electric Corporation System Energy Resources, Inc.

TU Electric Toledo Edison Company Washington Public Power Supply System Yankee Atomic Electric Company (representing also Public Service Company of New Hampshire, New Hampshire Yankee Division, Maine Yankee Atomic Power Company, and Vermont Nuclear Power Corporation).

DOCKET NUMBER PR 50 PROPOSED RULE ----':-"-----

(54 f{-<5;;).q4-<o")

(7J 7

P.O. Box 14000, Juno Beach, FL 33408-0420 LOCK[iED MARCH 0 9 ~

9 9 Mr. Samuel J. Chilk L- ~ 11AR 13 P3 :43 Secretary U. s. Nuclear Regulatory Commission ijF'F!C:.: OF SECit(_IA~Y Washington, D. C. 20555 o*ocKE1 ING & St t<Vlf.f 8RANC!1 Attention: Docketing and Service Branch

Dear Mr. Chilk:

Re: Proposed Rule Change to 10 CFR 50. 60 Fracture Toughness Requirement for Protection Against Pressurized Thermal Shock Events Florida Power & Light (FPL}has reviewed the proposed rule change to 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal-Shock Events." The following comments are offered:

The rule change incorporates the methodology of Regulatory Guide 1.99 Revision 2 Regulatory Position 1 for determination of charpy shift. This formulation uses copper and nickel content and fluence for the determination of shift.

The rule should contain the option of using Regulatory Position 2 which allows the use of credible surveillance data. The advantages of this approach to those licensees who qualify to use it are:

• Allow plant specific data to be the basis of flux reduction.

• Take into account secondary contributors to neutron damage such as weld flux, minor elements and heat treatment.

• Reduce the need to continually change the rule based on new data since the new data become automatically incorporated with each new capsule that is removed.

• Take into account outliers which behave significantly better or worse than the trend.

We believe that the Regulatory Guide 1.99 Revision 2 formulation generally works well to predict the effects of neutron damage on charpy energy 30 ft lb shift . However, the use of Regulatory Position 2, as an option, is more consistent with the current state of knowledge of the industry and Revision 2 of the Regulatory Guide.

FPL appreciates the opportunity to comment on the above rule.

-~~

R. J. Acosta Acting Vice President - Nuclear Energy RJA/JAD/gp c nowled d an FPL Group company

DOCKET OFFICE Of THE SKRET OF THE COMMISSION mark Date es Received _ ___,__ _ __

' I Copies Reproduced 3 ial Distribution f2 IDS f Dr:2 1

~

WNERS GROUP Arizona Public Service Co. Baltimo re Gas & Electric Co. Flo rida Power & Light Co . Maine Yankee Atomic Power Co. Omah~ * ,PQ¥ter-.District Palo Verde I , 2, 3 Calvert Cliffs I, 2 St. Lucie I, 2 Maine Yankee Ft. C:. I LL:

Arkansas Power & Light Co . Consumers Power Co. Louisiana Power & Light Co . Northeast Utilitie s Service Co . Southern

• Edison Co .

ANO 2 Palisades Waterford 3 Millstone 2 SONGS 2, 3 Edward C. Sterling, III, P.E., Chairman "Ql 1 .

c/o Arizona Public Service/ 11226 N . 23rd Avenue Phoenix, AZ 850291'5tan@fh '1~3J' 3 ,41 February 22, 1990

.~. ;* r !C[ CF 5ECR:.:. TARY CEOG-90-144 iji.iCKET 1HG *,. Sr flV !CF.

iRM Oi Mr. Samuel J. Chilk, Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention : Docketing and Service Branch

Subject:

Proposed Rule Change To The Code of Federal Regulations 10 CFR Part 50.61, RIN: 3150-AD0l Dear Mr. Chilk The purpose of this letter is to provide C-E Owners Group comments on the proposed rule change to 10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."

Based upon review of the proposed rule change and associated Regulatory Analysis, four comments are provided. First, the CEOG agrees that the Regulatory Guide 1.99 Revision 02 shift correlation is presently the most appropriate method for pred i cting irrad i ation damage, s i nce it i s the best fit to power reactor surveillance data and since separate correlations which account for varying embrittlement sensitivity are utilized for we l dments and base metal. Consequently, the Regulatory Guide 1.99 Revis i on 02 shift correlation is the proper procedure for calculating irradiation damage in power reactor environments and is approp ri ate for i ncorporation i nto the PTS rule.

Second, the NRC states that the proposed amendment makes the procedure for ca l culating the amount o f embrittlement for PTS consi ste n t wi t h the procedure g i v e n i n Regu lato r y Gu i de 1 . 99 Revi s i on 02 . Howe v er, the Regulato ry Guide 1 .99 Revision 0 2 4'Cl(nowled~ t,y caro. ~ / l ..f.1.<J.. _____

,uutAR REGULATORY COMMISSIOl't DOC KETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics ark Date s Received I Copies Reproduced al Distribution R IDS1 POR,

~~

Mr. Samuel J. Chilk February 22, 1990 Page 2 CEOG-90-144 position for the case when surveillance data is available is not presently included in the proposed rule change. This difference causes an inconsistency between the PTS rule and other evaluations which also use Regulatory Guide 1.99 Revision 02 (e.g., Pressure-Temperature limits and Low Temperature Overpressure Protection requirements). Specifically, the inconsistency would not allow reduction of the applied margin term when two or more credible surveillance data sets are available although it would be permitted by the Regulatory Guide.

Consequently, the improved accuracy of the Regulatory Guide 1.99 Revision 02 procedures is not made fully available and very conservative RTPTS values may be calculated for those utilities which have an accurate understanding of the irradiated properties of the limiting reactor vessel material. This could result in overly conservative predictions of the irradiation shift and unnecessarily high RTPTS values which may artificially exceed the applicable screening criterion. It is, therefore, recommended that use of credible surveillance data in accordance with Regulatory Position 2.1 be allowed in the calculation of RTPTS under the proposed amendment.

Third, and not identified as a change in the rule is the requirement for using measured values of initial reference temperature, RTNDT" The present rule states that if a measured value is not available then specified generic values for initial RTNDT must be used. However, the proposed rule change modifies the use of initial RTNDT values and states that measured values must be used if available; if not then specific generic values must be used. This proposed change in the PTS rule forces the use of a specific measured value when available. The proper use of initial RTNDT should be based upon a technical justification

Mr. Samuel J. Chilk February 22, 1990 Page 3 CEOG-90-144 for using a measured or generic value on a case-by-case basis.

This permits a utility to justify the use of a value of RTNDT which accounts for any plant specific material characteristic.

This type of approach would be consistent with the philosophy used in determining residual element content of reactor pressure vessel steels. It is therefore, recommended that the second sentence of paragraph 2(b) (2) (i) of the proposed amendment be revised as follows:

"Either a measured or generic mean value must be used, whichever is justified. If a generic value is justified, the following generic mean values must be used: o°F for ...

weld fluxes."

Finally, a general comment is provided with respect to the Regulatory Analysis. The Regulatory Analysis provides a perception of general nonconservatism in the current PTS rule.

This focus of the Regulatory Analysis may not be appropriate since thirty-three plants will have a reduction in the calculated RTPTS values if Regulatory Guide 1.99 Revision 02 is implemented into 10 CFR 50.61. This reduction in RTPTS may be as high as 37°F. Additionally, it is anticipated that if detailed risk analyses were performed for PWRs, the integrated risk of vessel failure would be less than presently perceived and that the true risk at any plant due to PTS events would, in all likelihood, be considerably below the risk value associated with the PTS screening criterion. This perception of lower risk is consistent with the PTS basis document SECY-82-465, "NRC Staff Evaluation of Pressurized Thermal Shock", and US NRC Regulatory Guide 1.154, "Format and Content of Plant Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors".

Mr. Samuel J. Chilk February 22, 1990 Page 4 CEOG-90-144 Should you have any questions concerning these comments please feel free to contact me.

Sincerely, E. C. Sterling Chairman, C-E Owners Group ECS/pwr cc: C-E Owners Group Mr. w. H. Rasin, NUMARC Mr. J. w. Pfeifer, C-E Mr. P. N. Randall, NRC Mr. B. Boger, NRC

Alabama Power Company 40 Inverness Center Parkway Post Office Box 1295 Birmingham , Alabama 35201 Telephone 205 868-5581 COCKETEO USNHC ~

W. G. Hairston, Ill

,z.ie:;;1 Senior Vice President Nuclear Operations '90 MAR J' P4 :22 Alabama Power the southern electric system March 12, 199(IDfFICE OF StCRETARY DOCKET !NG & SEflVIC[

BRANCH Docket Nos. 50 -348 50-364 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Docketing and Service Branch Comments on Proposed Rule "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" (54 Federal Register 52946 of December 26, 1989)

Dear Mr. Chilk:

Alabama Power Company has reviewed the proposed rule, 10 CFR Part 50, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" ,

published in the Federal Register on December 26 , 1989. In accordance with the request for comments, Alabama Power Company hereby is in total agreement with the NUMARC comments which were provided to the NRC on March 5, 1990.

Should you have any questions , please advise.

Respectfully submitted ,

vJ.,J-1 ~

W. G. Hairston , III WGH,III/JMG:kdc cc: Mr. S. D. Ebneter Mr. E. A. Reeves Mr . G. F. Maxwell Acknowledged by ean1 ... ?l \.?.~ :J.P.**~

EGULATORY COMMISSION I ~G & SERVICE SECTION

, FICE OF THE SECRETAR¥ OF THE COMMISSION Document Statistics Postmark Date 3 r

f; ,;2_/

CJ()

Copies Received ____ / _ _ _ __

Add'I Copies Reproduced . .0.,,.___ _ __

Special Distribution /2IDS PD R.

~

Georgia Power Company DOCKET NUMBER 333 Piedmont Avenue Atlanta, Georgia 30308 Telephone 404 526-3195 PROPOSED RULE Pl 50 (54Ff<5~qtf{p)

Mailing Address. '?>

40 Inverness Center Parkway Post Office Box 1295 B1rm1ngham, Alabama 35201 Telephone 205 868-5581 W. G. Hairston, Ill Senior Vice President Nuclear Operations March 12, 1990 Docket Nos. 50-321 50 -424 HL-995 50-366 50-425 ELV-01421 Mr . Samuel J . Chilk Secretary of the Commission U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Docket i ng and Service Branch Comments on Proposed Rule "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" (54 Federal Register 52946 of December 26, 1989)

Dear Mr. Chilk:

Georgia Power Company has reviewed the proposed rule , 10 CFR Part 50, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events ",

published in the Federal Register on December 26, 1989. In accordance with the request for comments , Georgia Power Company hereby is in total agreement with the NUMARC comments which were provided to the NRC on March 5, 1990.

Should you have any questions , please advise.

Respectfully submitted ,

w.).,~~

W. G. Hairston , III WGH, I I I/ JMG : kdc 1

6 10'-~---..-

Aclcn()Wledged by ~.., ..:.J-~

--.A

U.S. NUCLEAR REGULATORY COMMISSION DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date .3 1.:1 o Copies Received_-'I_ _ __

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ii Distrituti nJ ?tDS PDR 1 KIJ.A<..tl-tUL -

DOCKET NUMBER PROPOSED RULE PR SQ._-

( 54- FR 5d C/ 1/-ro)

DOCKETED USNRC

• 90 MAR 12 P5 :QS Nuclear Information and Resource service 1424 16th Street, N.W., Suite 601, Washington, D.C. 20036 (202) 32~q9~ t~.z;~,.s,Eq~f\~;f'rf t, * \

Board of Directors Kay Drey March 12, 1990 St. Louis, MO Anne Hess U.S. Nuclear Regulatory Commission New York, NY Joan Holt Office of the Secretary New York, NY 11555 Rockville Pike David H . Horowitz Rockville, MD New York, NY Tim Johnson AAt1anta, GA ATTN: Docketing & Service Branch 9 Bi11Jordan Akron, OH

Dear Sir or Madam:

Janet Lowenthal Chevy Chase, MD Mary Morgan Enclosed you'll find the comments of the Nuclea r New York, NY Information & Resource Service on the proposed rule Betsy Taylor entitled: Fracture To ~gh. ess Requirements for Protection Washington, DC Ellyn Weiss Against Pressurized Thermal Shock, dated March 12, 1990.

Washington, DC National Advisory Board Steve Aftergood Sincerely,

~-~

Committee to Bridge the Gap*

~

June Allen North Anna Environmental Coalition*

Robert Backus Riccio Backus, Shea & Meyer*

Dr. Rosalie Bertel!

_ alnstitute of Concern for 9 Public Health*

Barbara Bosson Actress Bruce Cockburn Musician Harlan Ellison Author Dr. Jack Geiger, M.D.

Prof. of Comm. Medicine, CUNY Medical School*

Marla Gibbs Actress Whoopi Goldberg Actress Janet Hoyle Blue Ridge Environmental Defense League*

Dr. H .W. lbser California State University*

Dr. Judith Johnsrud Environmental Coalition on Nuclear Power*

Charles Komanoff Komanoff Energy Associates

• Dr. Marvin Resnikoff Radioactive Waste Campaign*

Mary Sinclair Great Lakes Energy Alliance*

• Organizations listed for identification only dedicated to a sound non-nuclear energy policy. s-itr olq D

~nowledgad by catd * * * ***r * * * * = *

~ NUCLEAR REGULATORY COMMISSION DOCKET ING & SERVI CE SECT ION OFFICE OF THE SECRETARY OF THE COMM ISS ION Document Statistics Postmarlc Date D Copies Received

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UNITED STATES OF AMERICA NUCLEAR REGULATORY,COMMISSION Proposed Rule: 10 CFR Part 50 )

)

Fracture Toughness Requirements )

For Protection Against )

Pressurized "Thermal Shock Events )

)

54 F.R. 52946 )

NUCLEAR INFORMATION & RESOURCB SERVICE COMMENTS INTRODUCTION The hi~tory of the Nuclear Regulatory Commi~sion's ha~dling of _the pressurized thermal shock issue has been one of vac'illation and obfuscation. The proposed rule ,under consideration does little to b're~Jt from .. the Commission's past practices and faulty _logie*.

.While an attempt to better estimate the phenomenon*of embri,ttlement is laudable, the.Commission's regulatory analysis remains wanting.

The desire to bring regulations into a9cord with". t;.he curren.t understanding o*f ,embri ttlement seems to be the extent to which the proposed rule can stand the force of logic. Once again, the Commission staff is siding with industry and clouding the issue with considerations

• other than those pertaining to the safe operation of nuclear power plants.

I.THE PRESS~IZBD THERMAL SHOCK RULB SHOULD BB BROUGHT INTO ACCORD WITH CURRENT UNDERSTANDING.

OF REACTOR.PRESSURE VESSEL EHBRlTTLEMENT, The current pressurized thermal shock (PTS) rule, adopted in July 1985, was promulgated at a point in time where the Commission, staff and industry understanding of irradiation embrittlement was limited. The,method*for calcul~ting the toughness of the reactor p:r:essure vessel, the reference temperature for nil ductility transition *(RTNDT), failed to account for the irradiation embrittlement of'copper and nickel in-reactor pressure vessel (RPV) materials.

Revisio:r:i 2 to regulatory ,guide 1. 99, "Radiation Etnbrittlement to Reactor _Vessel Materials", corrects the shortcomings of the PTS rule by incorpo~ating the current, ~owledge of

• the effects* of radiation on copper, and nickel in :th;e vessel materials. However, the rev.ise:d r.egulatory *guide is j_ust th~t ,* a guide, it has n~ force of law.behind it. Thus .~he utilities that are operating reactors with RPV's which are mo:r:e severely embrittled, given the current

• stat~- of. kno,wl~dge, are n.ot necessarily. those that will_ come under the purview of the current Pl'S rule.

• Due.to the non-conservative ~ature of the Commission's other rules and regulations whic;::h impact upon the PTS rule ( see point IV.) it is imp-erative that current knowledge of radiation embrittlement *be folded into the PTS rule. To allow the current rule to.go unchanged would decrease the margin bf safety at nuclear power plants operating with increasingly embrittled reactor vessels. Furthermore* it would focus Commission a;ttentioh on plants

other than those which are the most severely embrittled and thus are at greater risk of experiencing a failure of the reactor vessel due to a PTS event.

II. THE PROPOSED RULE FAILS TO MAKE A CONCOMITANT CHANGE IN THE SCREENING CRITERION FOR THE PTS RULE, THE STAFF'S RATIONALE IS BASED ON CONSIDERATIONS OTHER THAN SAFETY AND IS BOTH ARBITRARY AND CAPRICIOUS.

While the staff's regulatory analysis attempts to rationalize

• its failure to change the screening criteria, the discussion before the Advisory Committee on Reactor Safeguards is much more telling.

The reasons to be gleaned from the discussions before the ACRS has

'little to do with the safe operation of.reactors with embrittled RPVs nor does i,t concern the "adequate protection II of the public health and safety., The staff's rationale is actually based on its desire to.avoid reopening the entire PTS issue. Regardless of the regulatory analysis presented ,in support of the proposed rule, the fact is that the NRC staff is not basing its decisions on the basis of safety considerations.but on administrative expedience.

On March 23, 1989, Dr. Neal Randall presented the Staff's case to the ACRS. Discussions between Dr. Seiss of the ACRS and Dr.

Randall*are most enlightening. In response to a question from Dr.

Seiss, Dr. Randall commented that the proposed rule will give more conservative results for a little more than half of the.plants under consideration. Dr. Randall later went on to attempt to address the question, "shouldn't we also change the screening criterion because the formula used to derive the probabilities in the original rule was the old PTS formula. And now we're going to

change the formula." . (Transcript of JOINT ACRS SUBCOMMITTEE MEETING MATERIALS AND METALLURGY/STRUCTURAL ENGINEERING ON REVISED PTS RULE AND EMBRITTLEMENT OF RPV SUPPORTS; p. 3 7, , (March 2 3 ,

1989). The discussion between Dr. Seiss and Dr. Randall continued as follows:

DR. RANDALL: The first reason I don't want to change the screening criterion is that if we do so,we will have to reopen the whole issue of PTS.

DR. SEISS: This is a non-technical reason.

DR. RANDALL: I can't DR. SEISS: I mean DR. RANDALL: Characterize it any way you like. I can't believe r. can sell the idea, you 'know, that this was a neat *mathematical process* where we've picked an*

accepta:qle probability and-- we've read off a screening criterion, and how if we calculated a different probability, we're just making a delta change in the screening criteria, and are happy.

(Id. at 39).

As the above discussion reveals, the staff's analysis and decision to change the metho,d for calculating RTPrs without making a concomitant change in the screening criterion is unscientific at

- best. It is,* in fact, both arbitrary and capricious. Dr.

Randal-1 's inability to present a cogent argument in support of the staff's position brings *into question the regulatory analysis sUbmi tted in support of the proposed rule. .Furthermore, it reveals a lack of co'ncern for the role of the NRC as protector of the public health and safety.

One has come to expect such an abrogation of duty by the NRC Commissioners (see for example: In the Matter of PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE, ET AJ.i. (Seabrook Station, Units 1 and 2)

CLI-90-02; CLI-90-03, (March 1, 1990) however, to find that such an abdication- of responsibility has similarly become a common practice of the NRC staff brings into question the integrity of the NRC as regulators of a most dangerous technology.

III. FAILURE TO MAKE A CONCOMITANT CHANGE IN

'l'HE PTS SCREENING CRITERION ACTUALLY DECREASES THE SAFETY MARGIN FOR PLANTS OPERATING WITH EMBRITTLED REACTOR PRESSURE VESSELS, As the regulatory analysis indicates, the 'screening criterion is a trip wire triggering the need for a plant specific probabilistic risk assessment (PRA). The staff then reasons that because it is merely an indicator of when such a PRA is necessary there is no great need to bring it into accord with the current knowledge of reactor pressure vessel embrittlement. This argument is both specious and logically inconsistent.

Let us - take Calvert Cliffs Unit 1 as, an example. The

• applicable-screening criteriqn for Calvert Cliffs Unit 1 is 270 degrees Fahrenheit. The reference temperat'-;lre for pres-surized thermal shock, the temperature at which PTS can be expected to occur, is characterized by RTPTs. Under the present rule RTPTS at the end of license life is 238 degrees Fahrenheit. Using regulatory guide 1. 99, revision 2 to calculate RTPTS' as is suggested by the proposed rule, shows RTPTS at the end of license life to be 301 degrees fahrenheit. There is an actual change in RTPTS of 63 degrees. Thus given the current knowledge of RPV embrittlement, the reactor vessel at Calvert Cliffs Unit 1 is losing its ductility at a greater rate than would be suggested under the present PTS

rule. To once again quote Dr. Randall, 11 (t) he state of embrittlement is worse as figured by the new rule. 11 (Transcript ACRS SUBCOMMITTEE MEETING ON REVISED PTS RULE, P. 4 7) To acknowledge that embrittlement is occurring at an accelerated rate thus increasing the RTPTs without making a concomitant increase in the screening criterion temperature flies in the face of logic. While the staff will argue that the :methods used in calculating RTPTS and the screening criterion are not directly analogous, one need

- not be a nuclear physicist to figure out that if the rate of embrittlement is accelerating and the screening criterion remains the same, the point at which a PRA of the plant is triggered is going to represent a decrease in the operating safety margin for that facility.

Again using Calvert Cliffs Unit 1 as an example, if under the proposed rule the RTPTS is 301 degrees fahrenheit, a difference of 63 degrees over the present PTS rule, the RPV in Unit 1 is losing its ductility at an accelerated rate. Under the current staff

proposal, a plant specific PRA is not triggered until the screening criterion of 270 degrees is reached. Given the current knowledge of irradiation embrittlement the PRA is not triggered under the proposed rule until the RPV is more severely embrittled. Thus the reactor will operate for a longer period with a more severely embrittled RPV. The increased embrittlement of the vessel thus increases the *risk that a less severe overcooling event could cause it to rupture. This translates into a decrease in the operating safety margin for the plant. If, as regulatory guide 1.99, revision 2 indicates, embrittlement is occurring at an accelerated rate, the

point at which a PRA is conducted must also be revised up ward to maintain the same operational safety margin. Increased knowledge of embrittlement should not translate into*a decrease in the amount of protection afforded the public. Yet, this seems to be precisely what the staff is proposing by failing to alter the screening criterion.

The NRC staff can bake-their numbers any way they like, but their calculations must stand the test of, log;ic. Science and logic

- should not be mutually exclusive. However, to listen to the NRC staff argu.e that RTPTS should be changed but the* screening criterion can r_emain the same ma_kes one feel as though he had fallen down the rabbit hole :and joined Alice-in her Wonderland. To continue to argue this proposition in the absence of any cogent arguments to support it reveals either *ignorance, intellectual dishonesty or misplaced priorit~es. Dr. Randall's testimony before the ACRS reveals that the NRC staff is placing administrative con~iderations before the safe operation of

• reactqrs with embri ttled

• pressure vessels. Therefore let us hope that only the latter is true.

IV. DUE TO THE NON-CONSERVATIVE NATtTRB OF OTHER FACTORS AFFECTING THE PTS RULE, 'l'HB DECREASED OPERATIONAL SAFETY MARGIN IMPLICIT IN THE STAFF'S PROPOSED RULE IS UNACCEPTABLE, In a recent case brought before the Atomic Safety and Licensing Board,* the Center for Nuclear Responsibility and Joette Lorion challenged the pressure/temperature limits for Florida Power and Light Company's* Turkey Point Plant Units 3 and 4. One of Ms.

Lorion's contentions challenged the facility's Integrated surveillance Program. Under this program, surveillance data from Unit 3 was used to* estimate the embri ttlement of the reactor pressure vessel in Unit 4. In Ms. Lorion's amended petition she cites a letter from Dr. George Sih, Director of Fracture Mechanics at Lehigh University. Dr. Sih states that:

The rate at which the bel tline weld material deteriorates and/or embrittle1:;1 depends on the combined effects of irradiation and pressurized thermal shock. It is .plant specific in the sense that the influence differs inherently from one unit to another. In other words, the metallurgical properties alone cannot determine the damage behavior of the welds. The loading history plays a ,major role. Unless the rates of irradiation, fluctuations in thermal gradients and time variation in pressure are exactly the same for both Units No. 3 and No. 4,

• one is not justified to assume that the data collected in Unit No. 3 could be applied to predict the behavior of Unit No. 4. Hence, conclusions drawn on RT.NOT for Unit No. 4 based on data from Unit No. 3 cannot be considered valid. (PETITIONERS' AMENDED. REQUEST FOR HEARING AND PETITION FOR LEAVE TO INTERVENE,February 17, 1989, In the Matter of Florida Power and Light Company, (Turkey Point Units 3 & 4) Docket Nos. 50-250 OLA, 50-251 OLA, ASLB No. 89-584-01 LA, (Pressure/Temperature Amendments)

• While Ms. torfon' s contention was thrown out "impermissible attack on Commission rules" and Dr. Sih's comments were dismissed due to the fact that they weren't presented in the as an form of an affidavit, the arguments raised are no less compelling.

The ASLB merely found procedural grounds for denying valid

• arguments. In fact the comments of Dr. Sih are confirmed by Dr.

Randall's testimony Before the ACRS. Embrittlement of the RPV is contingent upon the amount of copper and nickel in the metal and the extent of neutron exposure or fluence. In Dr. Randall's discussion before the ACRS he states that "(t)he fluence estimate is plant specific." (Transcript of JOINT ACRS SUBCOMMITTEE

MATERIALS AND METALLURGY/STRUCTURAL ENGINEERING: ON REVISED PTS RULE AND EMBRITTLEMENT OF RPV SUPPORTS, p. 80). Thus the use of surveillance data in the PRA triggered by the screening criterion is going to render non-conservative results.

Let us now turn to the non-conservatism allowed in the PRA itself. First, most PRAs only take into consideration design basis accidents. Unfortunately, this does not reflect reality. A second area of concern in regards to the accuracy of PRAs is the factoring of human error. The ACRS has attempted to address this issue by incorporating a "plant performance objective" in the PRA. However, due to the fact that neither the staff nor the ACRS was able to come up with a workable definition, the ACRS suggested an alternative, " .*. 'a prominent caveat, e.g. , a warning that PRA results do not tell the full story, should be made part of the (Safety* Goal) policy or of the implementation plan.' We recommend that such a statement be made an explicit part of the plan. 11 (ADEQUATE PROTECTION AS IT RELATES TO SAFETY GOALS: ACRS AND STAFF

• POSITIONS, SECY-89-375, Enclosure 1, p. 3 (December 14, 1989).

Commission p0licies and practices further exacerbate the problems posed in attempting to address the PTS issue through the use of PRAs. The refusal of the Commission and staff to recognize potential PI'S initiators, e.g. failure to acknowledge the potential for a multiple steam generator tube rupture or the use of the absurd "leak-before-break" theory, merely add to the* non-conservatism already present in the use of probabilistic risk assessment.

Given the uncertainties in the very basis of the PTS Rule, the

_., I I _,....

non-conservatism allowed in the proposed rule by maintaining the same screening criterion can not be tolerated.

CONCLUSION The staff's desire to better estimate the rate of embrittlement in the proposed rule should be supported. However, the logical inconsistencies, unfactored probabilities and the basic inability of PRAs to accurately represent reality undermine the

• good intentions of the staff. Administrative expedience is no~ a legitimate reason for the staff's failure to concomitantly address RTPTS and the screening cr_i terion. There is no "defense in depth"_

for the failure of a reactor pressure vessel and once you have core on the floor all the PRAs in the world are not going to make a bit of difference. Bringing the Pl'S rule into accord with our present knowledge of embrit.tlement is a good first step. And the good intentions of the staff in addressing RTPTS are laudable. However, to paraphrase Samuel Johnson, the road to meltdown is paved with good intentions.

~~-~

Nuclear Information & Resource Service 1424 16th Street NW, suite 601 Washington, D.C. 20036 c2oi) 328 - 0002 Dated: March 12, *1990

DOCKET NUMBER 50

~='='Ai~i=iii~ ( 64 PR 5J q f[p)

DOCKETED NUCLEAR MANAGEMENT AND RESOURCES COUNCIL USNHC 1776 Eye Street, NW.

• Suite 300

• Washington, DC 20006-2496 (202) 872-1280 Joe F. Colvin Executive Vice President & e,n 1cE: ::r s~ cRFTAR.,,

Chief Operating Officer March 5, 1990 ~~ .; ..; (\ ~ ~ I,-. G :'c r I** J !r :f

• !-~Nt.~~:

Mr. Samuel J. Chilk Secretary, U.S . Nuclear Regulatory Commission Washington, DC 20555 ATTN: Docketing Service Branch One White Flint North 11555 Rockville Pike

- Rockville , MD 20852 RE: Proposed Ru l e - "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events "

54 Fed. Reg. 52946 - 52950 (December 26 , 1989)

Request For Comments

Dear Mr. Chilk:

The enclosed comments are submitted on behalf of the Nuclear Management and Resources Council, Inc . ("NUMARC") in response to the proposed rule of the U.S. Nuclear Regulatory Commission ("NRC" or "Commission") entitled "Fractu r e Toughness Requirements for Protection Against Pressurized Thermal Shock Events," published on December 26, 1989 .

NUMARC is the organization of the nuclear power industry that is responsible for coordinating the combined efforts of all utilities licensed by the NRC to construct or operate nuclear power plants, and of other nuclear industry organizations , in all matters i nvolving generic regulatory policy issues and on the regulatory aspects of generic operat i onal and technical issues affecting the nuclear power industry. Every utility responsible for construction or operating a commercial nuclear power plant is a member of NUMARC . In addition, NUMARC's members include major architect-engineering firms and al l of the major nucl ear steam suppl y system vendors .

NUMARC supports the rule with the reservations expressed in the following enclosed comments. We would be pleased to discuss these comments in more detail with the NRC Commission and staff .

Sincerely ,

~ 1--k ~

F. Col vi n JFC\rs Enclosures Acknowledged by ca .. _ q fL':/1 O

• NUCLEAR REGU LA TORY COMMISSIU,,.

DOCKET ING & SERVI CE SECT ION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statist ics Postmark Date _3--J/-'0~/..1_

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ENCLOSURE "NUMARC INDUSTRY COMMENTS ON PTS RULE" NUMARC has collected and consolidated selected industry comments on the PTS rule revision by seeking the contributions of the Electric Power Research Institute (EPRI), including a select group of concerned utilities, and of the four nuclear steam supplier (NSSS) Owners Groups. '

I. Surveillance Samples:

Reactor vessel embrittlement and the issue of pressurized thermal shock

- ..... (PTS) are concerns that have warranted attention over the last ten years. In response to these concerns, the industry has taken significant steps to mitigate the effects of vessel embrittlement and reduce the risks due to a PTS transient.

The PTS Rule established the screening criteria limits for minimizing risk of vessel failure and provided a means for calculating the value of RTNDT (termed RTP. 8 ) for comparison with the screening criteria limits. Since the issuance of the original PTS Rule, the publication of Regulatory Guide 1.99, Revision 2 established an improved method for determining radiation embrittlement of reactor vessel materials. The stated reason for the proposed change to the PTS Rule is to make it consistent with the procedures given in Regulatory Guide 1.99, Revision 2.

A comparison of the proposed change to 10 C.F.R. Part 61 and the existing Revision 2 to Regulatory Guide 1.99 is presented in the attachment to this enclosure. These two documents, which define the means for calculating RTNDT' are consistent only in the case where surveillance data for a given plant are not available. For plants with two or more credible surveillance data sets, use of such data is explicitly permitted in Regulatory Guide 1.99, Revision 2, for plant operating limits, but is not allowed for comparison with the screening criteria limits under the proposed PTS Rule change.

For plants that are expected to exceed the screening criteria limits the PTS Rule requires that a detailed plant-specific analysis be performed.

Regulatory Guide 1.154 defines the scope of an acceptable plant-specific analysis using a risk-based approach that takes into account the full range of uncertainties in all areas that *may relate to the PTS scenario. However, by using credible surveillance data, and other relevant information, it may be possible to demonstrate acceptability without having to perform a complete Reg. Guide 1.154 analysis.

1

The industry believes that the approach outlined for determination of vessel embrittlement in the PTS Rule ignores the very data that may be most relevant in determining the actual level of radiation damage. For plants nearing the screening criteria limits, use of plant-specific surveillance data should be permitted. This additional option would provide utilities with the incentive to update their existing surveillance program, or perhaps begin a supplemental surveillance program, to more accurately measure and predict actual vessel material properties along with their plans to manage the effects of embrittlement by implementing flux reduction or considering other remedial measures.

II. Measured Values:

One item, not identified as a change in the rule, is the requirement for using measured values of initial reference temperature, RT11>r* The present rule states that if a measured value is not available then specified generic values for initial RTND filY.il be used. However, the proposed rule change modifies the use of inftial RTIJ!IT values and states that measured values must be used if available; if not, tnen specific generic values must be used. In our opinion the proper use of initial RTNDT should be based upon a technical justification for using a measured or generic value on a case-by-case basis.

This would permit a utility to justify the use of a value of RT or which accounts for any plant specific material characteristic. This fype of approach would be consistent with the philosophy used in determining residual element content of reactor pressure vessel steels. It is therefore, recommended that the second sentence of paragraph 2(b) (2) (1) of the proposed amendment be revised as follows:

"Either a measured or generic mean value may be used, whichever is justified. If a generic value is justified the following generic mean values shall be used: 0° F for Linde 80 weld fluxes and

-56° F for welds made with Linde 0091, 1092 and 124; and ARCOS B-5 weld fluxes.*

III. Credence and Conservatism:

Recent NRC claims indicate that the risk of reactor vessel failure due to pressurized thermal shock under the present PTS Rule may be significantly higher than was previously thought and that some plants may not be safe under the present PTS Rule. The industry believes that the Rule is adequate to ensure plant safety for the following reasons:

1) The current PTS Rule was determined to be conservative for regulating vessels with PTS concerns through a significant amount of deterministic and probabilistic fracture mechanics and risk analyses.

2

2) Since the issuance of the rule in 1982, research related to the PTS issue has demonstrated additional margins of safety due to improved understanding of crack arrest and overly conservative cooldown rates used in previous PTS calculations for small break LOCA. These effects were not considered in the present rule, nor are they considered in the proposed rule.

3) A change 1n the trend curve formula in the proposed PTS Rule without changing the screening criteria limits essentially adds more, and perhaps unnecessary, conservatism to the proposed regulation for plants experiencing an increase in the calculated value of RTPrs because of shortened vessel life.

For these reasons, and without performing additional risk analyses, it should not be stated that the PTS Rule is being changed to correct a potential safety concern.

Finally, a general comment is provided with respect to the Regulatory Analysis. The Regulatory Analysis provides a perception of general nonconservatism in the current PTS rule. This focus of the Regulatory Analysis may not be appropriate since thirty-three plants will have a reduction in the calculated RTPts values if Regulatory Guide 1.99, Revision 02, is implemented into 10 C.F.K. Part 50.61. This reduction in RTPrs may be as high as 37° F. Additionally, it is anticipated that if detailed risk analyses were performed for PWR's, the integrated risk of vessel failure would be less than presently perceived and that the risk at any plant due to PTS events would, in all likelihood be considerably below the risk value associated with the PTS screening criterion. This perception of lower risk is consistent with the PTS basis document SECY-82-465, *NRC Staff Evaluation of Pressurized Thermal Shock," and US NRC Regulatory Guide 1.154, Format and 8

Content of Plant Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors."

Attachment:

Excerpts from the Proposed Rule and Reg. Guide 1.99 Rev. 2.

3

AT!!'ACHMENT E L

• XCeV""p'L Proposed Rules A!GULATORV au,01 1.91 (Tau ME~,

Am114on 2 Thil Mciion of ~ FEDERAL. REGISTER contai,- nooc. ID h Pl.tile of ltle May1*

popoMd . . .,..,. of""" and rwgutationa. The puri,oN of INN l'IOtal II to g!W Interested peno,w an RADIATION EMBRITTLEMENT

~ to pa.r11ap&te ~ I'll rule ffl&M'IQ pno, to h aCSop&,n of h lnll OF REACTOR VESSEL MATERIALS NIN.

NUClLlA REQULATORY COMMISSION A. INTRODUCTION 10 CFR Part IO CicnenJ Desip Critcrioa 31. Frxrurc Prnenaoa of Reactor Coolant P'rmun Boundary, of Apptndla A, **aa.ra.t Oaap IUNl1IO-AD01 Cnw,a for Nuclear POWff Planll. ** to 10 CPR Pin ,o, **0onmac:

L!Cetlllftl of Produc:noa and Unlizaion Faallda." l'llqUua. m Fracturt Toughneu R~ulrernenta Fof pan. wi me reaaor coolant prasu.n bounduy be daiped wtdl ProtKtlon Against Preuurtzld sufficicm matJUl 10 1n1W"1 Chai. wha IU'eUed llndu opcflWII, TbermaJShockEventa mamcenw:e. tmms, and po,n* 1ered acadcm c:onditiom, (I) tbe AQINCY: Nuclear Re,u)atory boundary behncs Ill I D0llbrm1e 1D1nD1'r Md (2) Iba prooebility cJ"'nmu11lon. of rapidly pr'Oplpcina ft-lcmra II lftinunired. Oamai 0aip Cmcnoa 31 also reqwrm chat die daip ranc die WIClftlmClm AcnoN:PropoMdnal1.

111 ~ die dfeca of imdilaoll oa ma&cnll propaua.

IUIIIUJrr. 11:tt Nuclear Re,ulatory Appendi& 0, Praczure Toustmaa Requirmiaa," and Appendia Comm.i11ion (NRC) it propoatn, to H, ltca=rVmelW....Sw-w:illmPropwn~**

amend Ill regulation.a for liaht-water wtuc:h imp,,Nm, 1.11 pat, Critenaa 31. --...- dll calaalation nuclear power planta to chan,t the of dwlps i.a frxmtl aastiw of l"elCIOr vaat nmcnats CIIIMd proeedW"I for calculttin, the amount of by Dellm'0D radiadaa lbn:lup:xa ltle ~ la T1lia pm clllalbw radiation embrittlement that a reactor &tnffll ptocedwa ~ 10 die N1lC air for cm,lm,. 'M ve11el receivea. The pre11uriztd thermal elfeca of neua-oe ndlacioa embrizdarwu .JI the low-alloy DIii 1bock nalt (PT'S nile) 11t1bliah11 a CIUftlldy med far li&bl*warar<DOled rcai:=r Yelllla.

1C1HDJ.nc criterion. Th.ii criterion limit, the amount or embrittlement or* reactor . The ca!cwaciYt procedurea siva ill R t ~ P0woa 1.1 of ve11el beltline material beyond which dlil sum.,. DIX Iba am .. ltac IMII ill ltle PresalNlld n.rm.

the plant cannot continue to operate Sbock naJe Cf 50.61, Fracmra Tou&hnaa Rcqwnmema ror ~

without tu,tification baaed OD* plant- tectJoa Apma Prcaw.ed Tbermli Sboct !vcra. ol 10 CR iped!lc analy1lt. Tbt propoaed Pan 50) lbr calc,gJecma RTP'TS, !bl mam ....... aa..* dlll

• amendment doe, not cbanp the 10 be compwd IO !bl ICl"laliq crill:rioll lf¥a ia die Nie. 111a acreen.tna criterion. 'l'bt PTS rwe alao mfucw oa wllidl dlil ltffllioa 2 la bUld may Ibo aft'lcl dlil pruaibt1 th1 procedure that muat ba bllia fGr dll PrS Nie. 111a rlalf Iii pswady CC11DC1eriDa whldm 111td lor calculatina the amount of co propm a cb11111

• I '°-61.

embrittlement lor comparieon to the ec:rtenina criterion. Tbt propoud amendment would update the procedure and make It conalatmt with tbe one sfYID In Rqulato17 Gmdt t.19. Rntaioa Z. publitbed in May 1818.

~

PNt,W'ized thmn&J lbock naita .,.

l)'ltem trantimta ma preuW"iud water nactor (PW'R) that can cauae Nvere OVl1"COOlina followed by imm.diate nprn1UN1tion to a biah lneL n.

thermal 1tm111 caund by rapid oooliq of tht Nactor YHHl inl.ide 1wface combine with the pl'tllm'l ltlelHI lo blC"laH the poteDtw lot fractvt if IA

&nidatm, flaw ii prttent iD low tou&hne** matenal. Tht, mattrial may niltin tht Nactornutlbeltlint, adjacnt lo the COl"I, wbert Dntroll

,.di,aon F1dually embrtttlet tbe material durin& plant life a.me. nt dqNt of tmbnttlemait dtpenda OIi the chemical composition of the 1tetL nptdally the copper and llicbl CODtenta.

Tbt touahnn1 of nactor YNNl material, ii characterized by 1

-r.rtrRCt tempcratmt for DiJ ductWty tranaltJon* CRTeri which cu be dtftntd I I foDowa. For many ructon DOW ID op.ration. to\llMIII of the behliDt materi&la 1t room tamperatare la too low to permJI fuD pmlurizatloll of the wunl With adequate aafety N,.tna.

A. temperatun ii rtiaed. lollghnna lzlcrtlHI ,Jowly at ftnt but II the temperatun defined I I RT..., toupnnl hqin, to 1Dcrt1N mud! mON npklly.

The tramitioii ID toqbn1 from low m* to hiah that taln place abcm aTDI occun over* temperatww -

IDterYal of about HO "F, nua It DOnDl1 operat1n1 temp1r1tmn. ntttl mattrtala an quilt toup. RTDI ii determined by de1tnactivt teall of material * ~

a.diaUOD tmbrHtltmtnl movtt aT-. lo

~ ltznperatw'tt. Comlation, baNd on tel1 relUlta for anirndiattd and

&rrtdiated ~ I D I have bNa developed to calculate tht lluft iD RT...,

u

• fimction of neutron Dunce ,_

wlriou malarial compoaltioaa. 11ae nJue of RTnt It I sf ven tiJU Ill I wnnr, lift fl utd ID frlct1n IDKhaniCI calcuJ1Uona to clttmmne whether 111i&mtcS pre-uilttna na..

WOll!d proptplt u lftCU wba tM wen! ll 11n1Md. .

'lblPrtNuriudflMirmallbodl(fflJ nit. 20 en 10.12, adoptad OD Jal, II.

1185 (50a°R Zll37). lltlbliabee I 1Crteftlqc:riwioa. ftilacna!Da criteriaa ntabliab.n I limltfnl 1rie1 el embnttlemnt beyond wbJda opml:ia cannot continul wtlboat further plut-lpeclBc IYaJaltSOIL 'fta tcl'tlnfzll .

crilt.nall II~- ID tarma al ay_,_

caJCllfated U I function Df tu oopplr ud IIJcbi COlltet, oltht material ud

._ DtUtrDII Dl&IDCI *ccotdina to the proc.dlft sfve ID lbe ffl Nie. ad oall.cS llT,,. to dJ1Uncwth It from olber procedW"N lor caleaJatiaa aT.,,

C. RIGULATORY POSmON

• * * *

• l. SlJRVllLLANCE DATA SOT AV.\JLUU (b) /wquiremenu. (1) For each pre11wued water nuclear power Nactor Wbn c:ndible ~ dm from tbe l'-=r Ill qu111DOG lor which an operaun, licente ba1 been art DOC ivaalabk, cak:ILIIDOG of llCIW'IXI rw1ia0oa embnmcmem of latutd. tht licenJN hall submit ltlc bmdim of J'aCEDf'~olU&bl*--- ra=n shouJd bs bued projected valuu of RT"' for reactor YHHI beltlint mat1rlal1 by IIWII oa cbc proctdura ID Jlel"WOfY POIJDOIII I.I m11.: W1U11D me value for the time ol 1ubmittaL the limmaom .a RqvJalDry Pmmoa 1.3.

explrttion date or tht Opel'ltina licenN.

the projected expiration date l/ a chanp In tht opel'ltlna license b11 been 1,1 Adjlll&ed Rd'ennce Tamperuun nquctttd. and the projected expiration date of a renewal term ii a reque1t for The ldJU.Slld rafertnea iempenrura I AR°T' for eac:n ma1crw ,n lietaM renewal ba1 bttn 1ubmltttd. Tbe tbs blllllne 11 11v1n by die follow1n1 uprUSIOA:

as1111ment mu1t UH the calculative procedW"tl ,;ven in parasrapb (b)(2) or ART

• lnwal RTp,;oT - lRT~oT - 'wfarJtn 11 thi1 Met.ion. The HHHD'lfflt muat lpteify th, ba1e1 fOI' the projtetlon.

includlni the HtumptionJ 1'981J'din, Iruuai RT-..OT rs the reference rempcmurc !or tne unundiatcd core lo1cfinl pattama. The 1ubm11tal rnatenaJ u defined an Parasn4)h SB-2331 of S<<uon W of tl'le muat lilt the copper and nickel content,, ASME BotJcr and PreuuR VCSIOI Code 1Ref. ':'). If measul"9d va.lues and the Ouenct valun u1ed in the o1 uuual RTNDT for die nwcna! 1n quauoa are noc ava1iacle.

calculation for each beltline material U Jencnc fflCIA vaJua for d\ll class* or nwenaJ may be used af lherc tbe11 quant:ftiet differ from thou arc lllfficlCftl lell l'CIWCI IO estabjiJh I mean and standard devll*

111bmitttd in re1ponH to the oriJinal UOII for die elm.

PTS rule and accepted by th* NRC.

jultillc.ation miat bt provided. nu, HN11ment muat bt 1ubm.itted by (e ilTNOT IS tbs mean value of !he adJUSuncnt zn reference montht after the effective data or tha lampCf'UW'9 causad by ,mcuauon and should be i:alcuwed u HCtlon), and muat be opdattd whenever followl:

there ill a 1ignil5cant chqe in r,rojecttd nJUH or RT"" or upon a NqUHt for a chqe in the expiration data for operation ol the racility. CP ( ~ is 1M chaNaty faaor. a funcuon of copper and iuckel (Z) 1'1le prtUwized thmnal 1bock COlllelll. CF is pcn ia Table I for wefds and an Table 2 for base (PI'S) ICneD.Ull c:r1tarlOD ii %70 -p fot rntW (p(alca and forJinp). unar UHCri,c>iMloa II pmnlllld. In plate* forginp. and uial weld Tabla I and 2 "wetpt*percsnl Cot'PW" and "wc11hl-pucen1 m11erlah, or Xl0 -P for circwnferentl&l meal are chi bac..amase Yalua for die maianal, wlldl wdl weld matlrlala. For tb1 pwi,oae of nonmUy be the man ol lhl masund vaJua for a plate or forpl1 compari1on witb tbit c:ritlrioa. the nlue of RT"' for tbe rtactor naaaJ mut bt or for weJd sampla made wldl lbc weld ware heal aumotr dlal calculated II followa. The calc:ulat:fon awi::bll me cntical v..i weld. If llldl va.lua an 1101 avaalable.

mUlt bt made for uch weld and plate, dlluppsliml:ina ¥UWV41 ID miamenal sptedirm IOwtudl or forsta,. In tba rtactor \'HNI beltlina. ma vCIIII wu ~ may be utd. ll not available, cormrvaave lqutk121 t; RT,..* I+ M + 6.11',. eaim1111 (rm plaa om SWldard dnimrloft) bUld on pncrie elm may be med if ju,al',cadoa ii prowled. II ihcra IS no uuormMJOII (I) T mtlN tbt IDJtiaJ rtfU'IDCI available, 0.35S COIIIICI' and 1.0~ ~ shouJd bl asawned.

temperature (RTIID'f) of tbt IIDifflldiated material meuW'9d u dtftned mlht n. n.a raca. ,0.21 - 0.10 lot,. ii dmrmlfted b1 c:alcala-ASME Code. Pa:qraph NB-mt.

• or from Pl,-. I.

MeHu:.d valun must be aed ii AYailable: It not. tbt followina ,a1ric "Maraia" II dll quMaY. -,, dla ii IO be tdded toobcaiD coo-IDllll vaJim mUlt be ued: 0 'F for ~ld¥t. 111)1111'-boaad Yl.lua of adjmled reference ~

weld, mada witb UDde

• Dux. ad for die caieallbOCII reqwred by A ~ 0 10 10 CR Pan 50.

-se 7 for weld, made witb Linde at.

1082 and W IDd ,UCOS M weld th1xa.

(U) -W- IDlaDI 1M mUliD lo be added ID cover anctrtaJntitl mthe *alul of lnttiaJ RTl!P" copper and alcbl conttnta. Duenc:e and the calculatloaal

-n.ctw111r---.1M111IToe0T11..--,- =- ._ .....

proceduru. In Equation 1. M i i

• 7 for Wilda and .. ., lot baN metal II anenc ,.,.,......,...,._,__.,. * ...,_e1.....,..,_,~IO*-i:

..... - - . . , . . AITll,.__.l,wft nlun of I are uHd. and M ii II 1' for Wilda and M 'F fot baN metal ii JDHIW"ed YalUII of I &l'I IINd.

(lU) ART"' ii tbt mean nJue of the adfu,tment In rtfernce temperature cauud by in'tdiaU011 and abowd be caJculatad II followa:

lquticml: ..,.,. * (CF)re&***-*

(1,*) CF rFl U the chemJ1t,y factor.

• Here.~ II r.nc SWl4ard Clc'l'IIIIOfl for ::,c .natW RT-.cT ,f.

function or copper and Dickel contmt -nalllJ'IO 'l'alUI of uuual RT'IIOT for :nc :'!'lltcnai ,n '1ucsuon is a II lfvaa In Table 1 ror weJda and in ~~wablt. at 1110 oc CSlJIIWa:I from lhe prs-.110fl of the taz :MU\Od.

Table J ror bHt mete! (platn and If not. and pncnc mean values for :naa clw of matmal ire used.

foJ'linp). Unear lnlef1)0lltiOD ii at II IJ\C stanaan1 de'l'IIUOfl obwned from I.he set of daU UICG tO permilled. ID Tablu 1 and Z '°WI* ~Sta.Dlas!I the man percent copper" and "Wl*pm:eftl nickel" are the be,r-ntimate w l1att for the material which will normalJy bt tbt ~.e nanc1lrC de'l'1auon for ~T'IICT * ~ A .s :s 91= !or -..elds and mean of the eiututed vah111 for a plate  : 7 "F for 1:1asc mew. ucqx :IW i,A :'lelC noc ucocd O 50 11mcs or fOl"li"I or for weld ,ample, made the mca va.luc of ~RTNOT with the weld wire hear Dumber tbt matchea the cn1Jc.1J ve11el weld. U these val1111 are not 1uilabl1, tht upper 2. Sl,'RVEJLLANCI DATA AVl\ll.AIU Wllliliia valuu SiYln In the malarial 1ped&ationa to which the 'rfllel wa1 Whn two or mon credlblt IUIWlllancc data sm (u ddlDld built may bt uaed. U DOt available. ID die DISCUSSIOII) become l'l&liabil from die reactor ut qowaaoa.

conaervalfve 11timalH (mean plut one Ibey may be llsed lO dllsnnlDI die ad}mred rcfcma lempcncun 1tandard deviation) bued on aenmc .s w aw,, appcr-shllr emro o1 &bl beltlim macanaia u data may bt IINd ii julbfication ii Jescnbod ill Rcplalory l'olulOnl 2. I and 2.l. rapecn¥1ly.

p,avided. U there ii no inlormation ewailabla. 0.15 pel"Cfflt copper and 1.0 l.l AdJumd ..,.._ Tempenlm'I

- pucanl nlchl mu11 be 111umed.

(v) *r meam tht bnt ntimate Tbt ldjasl,d referaa tctnpcn111t1 should be obcaUllld u follow* Finl. If tben ii d e a r ~ dial die ~ or llicksl neutron fhaence. In m:uta of 10 1* afm 1 (E sreater th&D 1 Mt'V). at tht clad-baff. COIIICllloltheau"leillaDceMtdditfustramcha&ollbaV--..S..

- , metal interface on tbt ln.lidt lllrlace of i.e., diJfen from dll averap for ltlt IWdd WU"& Ilea IIUl'Dber 1ht n11eJ at the location when tht moc1Dd ,.,.. cha ¥t:Ue1 wdd and me survailata we1c1. me material in qunt:IOD nteaiVH tht hipeat rncuunid valua of UTNOT should be ldju.aed by ~

flutDct for tht ptrfod of aervice in Nm by die rabO oldie dltm&my faaor for Iba ¥esael wtlcl 10 dial quntaoa. . for die satmllance weld. Second, 11'11 ~ dill snauld be (3) For ttcb preuurlnd water udtar fitted Ullftl !qaaaott 2 IO OOC&UI me relaaonallql ol AI.TNDT ID power reactor for which tba nlut of nuaa. To do IO, ca1aa.1.a d i e ~ raaor. c,, ror die baa RTm for az,y matmal fa tht beltliDt II projected to axceed the PT'S tcrNDiDI nc bf mu1up1y1111 adl ldjuad ARTNDT by 1a COffclf nuaa t'aclor. SIUMUftl lht produca. am! dmdiaC by the . . ot

• • int criterion btfol"t tht txplratioa date of ttie sqaua ot 1111 nNIICI raaon. nc rauJtma \'WI ot a.,-

th. operatifta licenH. or the protected expiration datt if a cbaqe in tht liceDN flllSNd Ill EqUlbOII 2 ril ln'IW rela&IOIIIAipolilTNOTID ha1 beta nqu11ted. or the end of a OllCft9C !NI fa die pin AlrWlllua dlsl aa IUCb I way 11 10 renewal tenn ti a nqunt for lJceme MINfflllt dll SIUII ol cbt . . , _ ol die anon.

nnewal ha* been submitted. tbe licenHe lhall tubmJt by (t moathl after To calcaJm dll marp1 la du cue, 1111 !q111bOD 4; !he vuua A the tff'edin date of thit NCtioa) ID fM9dllNfaroA1111Jbeeaillhali.

• analy1l1 ed IChedu1t for lmplemeatat:loa of tuch ftwi: nduct1aD U dlil proc:adurl pa a lu&flcr value ol ad,uszad refarenca pro,ram, a* &r1 reuonabJy precticable temper1111rt dlU dlar fivcll by usana die proc9dures of 1leplatMy to avoid txcaedin& tht PT'S ec:nenlq Poncioa I. I. die tuneallancll dlca sboaJd be mc. Umaa ~

criterion Mt forth in parasraph (b)(Z) of thi* NCtlon. n, echtdult for P11alowr\'Ult,lilblrmaybe11111L lmplementatioa of flux nductiaa meUW"tt may tab Into aoooaat tbl Forplalll fllV'ill SIIMillaaoldaa cbar ancndlblt maU raspa:a tchtdult for nbm.ittal ud atidpatld ftCtllt dllr Iba marena! doa DOI l'qllwnl die cntJCII rwna! 111 Commiulon approval of cllWJecl" plam- die ¥11111. die calaalatM Pl"OC8duta m cbil pde mould be IIMd spedftc anal)'HI. tubmltled ID roobcam aw valaeloldd, 4ATNDT.11tca1m,..... lfw11111p.

demon,tnte acceptabla ml al nlaN of dltvaJueoto,ma,beredacadfroaicbevaluapvailmelaa RTrn 1bovt the tc:reen1111 llmlt DI to ,.,..,.olleplaaryPos,ra 1.1 by&111m01am10bedleaded plant modiftcatfona. uw lalormatiae ar OIi a C:W~-CW baail dependfJ'I OD WMr1 Iba aneuund Yaiw newaJW)'llttecbnlqall. fall relacm IO die W calcalmd lbr die auneillua amana1a, (f) Far uda DreUWised water_..,

power reector tor wblch the uat,m required bJ parasr1pla (bXSJ ol lllll HCUon lndicatu that no re11onablr practicable Du reductfon prosram will prevent tbt n.lue of RTm from exceedina tht PTS ICl'ffninl crtttrian befcn the nptntioD date o1 tbt Net PTS

operaun, Ucen,e, or the projected J expiration date II a cban,e in the opera ti.cl licerue b11 been requ11ted. or the end ol a renewal term II a requut for Uc.enu renewal bu been tubmitted.

the 1ictn111 ,hall tubmit a tafety an1ly11t to determine what. II any, modlf!cation, to eqwpment. 1y1tem,,

and operation 11"1 necHHry to prevent potentlal failure or the l"llctor YetHl ..

a ntult or po,tulated PI'S nenta if continued operation beyond the acnerun, criterion It allow~ In the analy1l1. tht licensee may determine reactor ve11el En1terial1 propertin band on available Information.

rneucb re1ulta. and plant 1urv1illance data. and may uu probabili,tic fraeture mechan.lca tecluuqun. nit analyata mutt bt 1ubmitted at leaat I yaan btfol"I the value of RTm H defined In parasnpb (b)(2) of thi1 aectton la projected to exceed tbe PTS IC'efflina critlrion or by on* year after the effective date of thi1 etendment.

-tuchever la later.

(5) Altar contlderation of the liceruH't analy1t1 (includina efl'ectl of propoNd comctiv1 action-. II any) 1ubmitted in accordance with or paragraph, (b)(3) and (b)(4) thit NCtlon. tht Commi11lon may, on a cau-by-au ba1i1. approva operatiOD of the or facility at valuH RTrn In excen of the PTS acreenina criterion. Tht Commiuion will contidtr facton al,ni.ficantly affectinl the potential for fai!W't of the reactor veuel in reachina a deciaion.

(8) If the Comminion concllldn, or punuilflt to para,rapb (bH5) thit 1ection. that operation of tha facility at valu11 of RTm in Ueetl of the PJ'S la'ffDina c:riterloii cannot bt approved on th, ba1i1 of the Ucemn't analyu, tubmitted maccorc!ance with parqraplu Cb)(3l and ('oH4) of thlt HCtion. the UcentH ahall requ11t and receive Commiaalon approval prior to any operation beyond the criterion. Tbt nqunt mutt bt band apon modificatiom to equipment. l)'ltema.

and operation of tht facWty in addition

• to thOH prnioualr propoHd ID dw tubmftttd analyaea tut would reduce

• tht potential for failurt of tht reactor YHHI dut to m nenta. or apon further tJW)'HI baMd apOD DtW lmonnation or improved methodolol)'. ,

Pft DOCKET NUMBER 56 PROPOSED RtlLE ( 6 Lt f R 5;)..q '+ ~) CD l'0 (;l'i£1EO U5NRC Mar v in I . L.e wis "SU FEB 26 P4 :19 7801 Roosevelt Boule v a r d Suite 62 OfFfCf OF SECRETARY Phi.la., PA 19152 OOCKfilNG & ,..EtlVICf BRANCH (215)624-1574 Secretar y USNRC Washington, D. r. 20585 Dear Mr. Secretar y ;

Federal Register Notice, "Fracture Toughness Requirements for tection Against Pressurized Thermal Shock Events, Proposed

• e," inv ited comments from the public. I am responding herein to the Notice .

My particular comment is that the Proposed Ru le and the present screeni n g criterion fail to address the question of whethe r embrittled steel will safel y meet the challenge of a pressurized the r mal shock event in a pressurized light wate r r eacto r . fhe ne x t question is how and wh y does the proposed ru le fail to answe r the question of whether embrittled steel will safel y meet the challenge of a pressuri z ed thermal shock event .

The reason that the proposed rule does not answer the question of wheth e r embrittled steel will safely meet the challenge of a p r essurized thermal shock event is that the embrittlement is analyzed as a single , unrelated issue. Th e loss of toughness due to embrittlement is analyzed apart from othe r A g r adation processes occurring at the same time and in the s ame

~ nditions. Much of the data used to analyze the degradation or loss of toughness comes from coupons welded to the reactor vessel. These coupons are of the same material as the reacto r v essel . These coupons also experience r oughl y the sa me irradiation and corrosion. These coupons are very different in the s t re ss pi c tu r e which they e x perience. These co u pon s ma y expe r ience very little stress while the reactor wall may be highl y st r essed.

I he diffe rence in the stre s s p ic tw- e and other uncont r o 11 e d vai- i able s c a s t s a s hadow of suspicion on analysis using data fro m these c o u pons. ll n ti l actual data from the reactor pressure vJal 1 is analyzed and agrees with the data from these coupons, dat a f r om t hese c 11 p o n s is suspect and may not be applicable. Unti) actual data f r om ir radiated reacto r wall material has been analyzed , the s af es t approach is to c l ose down al) reactors no w .

I hope that y o u wi ll accept this suggestion and close do wn all n uc l ear r eact or s n ow.

Respectfully s u bmitted,

~~ / ~;), J?.?E 2 90 .

-ledged r,, carri..:1/.t.~ .. _ ,

u. S. NUCLEAR REGULA lORY COMMI !>>L DOCmlNG & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Dete Copies Received I I I Add' I Copies Reprod uced ..3 Special Distribution (2 I /J :51 P DR 1 e~

..,, ~-~ - t .,s. ,.~*

., '->'Ot,Y iO t;\,J ... *,.. .

psgo 01 J Original sent to the '-----

- Office of the Federal Regi~

DOCKET NUMBER PR SO - for publ!cation PROPOSED RULE 1) /(ukmaHh>; ..£~em

( .5"f/ ( /(_ §" J 1'-Jb

'89 DEC 20 P3 :32 NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN: 3150 - ADOl Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is proposing to amend its regulations for light-water nuclear power plants to change the procedure for calculating the amount of radiation embrittlement that a reactor vessel receives. The pressurized thermal shock rule (PTS rule) establishes a screening criterion. This criterion limits the amount of embrittlement of a reactor vessel beltline material beyond which the 9 plant cannot continue to operate without justification based on a plant-specific analysis. The proposed amendment does not change the screening criterion. The PTS rule also prescribes the procedure that ust be used for calculating the amount of embrittlement for comparison to the screening criterion. The proposed amendment would update the procedure and make it consistent with the one given in Regulatory Guide 1.99, Revision 2, published in May 1988.

DATE: Comment period expires {75days after publication in the Federal Register). Comments received after this date will be considered if it is

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practical to do so, but assurance of consideration cannot be given except for connents received on or before this date.

ADDRESSES: Mail written comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Service Branch. Deliver comments to: 11555 Rockville Pike, Rockville, Maryland between 7:30 am and 4:15 pm Federal workdays. Copies of comments received may be examined at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT: Pryor N. Randall, Division of Engineer-ing, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory COIIWli~sion, Washington, DC 20555, Telephone: (301)492-3842.

SUPPLEMENTARY INFORMATION:

Background

Pressurized thermal shock events are system transients in a pressurized water reactor (PWR) that can cause severe overcooling followed by immediate repressurization to a high level. The thermal stresses caused by rapid cooling of the reactor vessel inside surface combine with the pressure stresses to increase the potential for fracture if an initiating flaw is present in low toughness material. This material may exist in the reac-tor vessel be~tline, adjacent to the core, where neutron radiation gradually embrittles the material during plant lifetime. The degree of embri~tlement depends on the chemical composition of the steel, especially the copper and nickel contents.

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• The toughness of reactor vessel materials is characterized by a "reference temperature for nil ductility transition" (RTNDT), which can be defined as follows. For many reactors now in operation, toughness of the beltline materials at room temperature is too low to permit full pressurization of the vessel with adequate safety margins. As temperature is raised, toughness increases slowly at first; but at the temperature defined as RTNDT'_toughness begins to increase much more rapidly. The transition in toughness from low values to high that takes place above RTNDT occurs over a temperature interval of about 150°F. Thus at nonnal operating temperatures, vessel materials are quite tough. RTNDT is deter-mined by destructive tests of material specimens. Radiation embrittlement moves RTNDT to higher temperatures. Correlations based on test results for unirradiated and irradiated specimens have been developed to calcu-late the shift in RTNDT as a function of neutron fluence for various material compositions. The value of RTNDT at a_given time in a vessel's life is used in fracture mechanics calculations to determine whether assumed pre-existing flaws would propagate as cracks when the vessel is stressed.

The Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61, adopted on July 23, 1985 (50 CFR 29937), establishes a screening criterion~

This screening criterion establishes a limiting level of embrittlement beyond which operation cannot continue without further plant-specific evaluation. The screening criterion is given in terms of RTNDT' calculated as a function of the copper and nickel contents of the material and the neutron fluence according to the procedure given in the PTS rule, and called RTPTS to distinguish it from other procedures for calculating RTNDT" 3

The PTS rule requires each PWR licensee to report the results of the calculations of predicted RTPTS values for each beltline aterial, (including the copper, nickel and fluence values that provided the basis for the calculations) from the time he submits his report to the expira-tion date of the operating license (EOL) .. The PTS rule further provides that if RTPTS for the controlling material is predicted to exceed the screen-ing criterion before EOL, the licensee should submit plans and a schedule for flux reduction programs that are reasonably practicable to avoid reaching the screening criterion. Finally, the PTS rule requires licensees of plants that would reach the screening criterion before EOL despite the flux reduction program to submit a plant-specific safety analysis justify-ing operation beyond the screening criterion. The licensee must submit the analysis at least 3 years before the plant is predicted to reach that limit. Regulatory Guide 1.154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors" provides guidance for the preparation of the report and describes acceptance criteria that the NRC staff would use.

In response to the PTS rule, the licensees of operating reactors have submitted the fluence predictions and aterial composition data and

{with 2 or 3 exceptions) these have now been accepted. Of greater importance are the flux reduction programs that have been undertaken by licensees for those plants having high values of RTPTS.

Need for the Proposed Amendment The primary purpose of the proposed amendment is to change the procedure for calculating RTPTS to reflect recent findings that embrittle-4

ment is occurring faster than predicted by the PTS rule for some reactor vessel materials. Although the PTS rule was adopted on July 23, 1985, the procedure for calculating RTPTS was developed in 1981-1982 and not updated because a number of licensees were using the 1982 formulations as the basis for flux reduction programs. Meanwhile, plant surveillance data were being added to the data base and there were extensive new and more accurate correlations made. These culminated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Mate-rials," published in May 1988. Revision 2 provides the basis for pressure-temperature limit calculations. Peer review of the new correlations was provided by the public c01111Dents on Revision 2.

In the regulatory analysis prepared for Revision 2, and repeated in the regulatory analysis for this proposed amendment, the NRC evaluated the impact of amending tbe PTS rule to be consistent with the Guide.

Copper and nickel contents and fluence values f~r each PWR reactor vessel were taken from the PTS submittals from licensees. When the values of

- RTPTS were recalculated using these quantities and the procedure developed for Revision 2, the results were higher for approximately half the vessels, including three vessels where the value aay be over 60°F higher than previously thought. This would increase the probability of PTS-induced vessel failure by a factor of at least 30 for those plants.

The NRC believes these changes in the nonconservative direction are greater than can be absorbed by the uncertainties believed to exist and taken into account by the NRC when the RTPTS -based screening limit was set. (A margin of 48°F is added in the calculation of RTPTS to cover not only the uncertainty in the formula for embrittlement but also the uncertainties in the copper, nickel, and fluence values entered in the 5

formula.) Based on this new information, the probability of reactor vessel failure by fracture during a PTS event is presently higher in some vessels than the probability based on the procedure for calculating RTPTS which is given in the present PTS rule. Moreover, a few of those reactor vessels will rea~h the screening criterion in the 1990 1 s. Thus, the current PTS rule needs to be amended.

Explanation of the Proposed New Requirements The proposed amendment changes the procedure for calculating RTPTS and requires all licensees of operating PWR's to resubmit projected values of RTPTS using the new procedure. If the copper and nickel contents and fluence projections are the same as in the previous sub-mittal, they need only be listed. If there are changes in these projections, justification for the changes must be provided. If a licensee has already submitted the infonnation required by paragraph (b)(l) of this proposed amendment, the licensee may simply reference the earlier submittal.

The proposed amendment modifies the requirement for fluence projec-tions in the calculation of RTPTS*to take into consideration the poten-tial for a request for change in the expiration date for operation of the facility. This applies to requests to change the end of licensed life from 40 calendar years after the date of the construction permit to 40 years after the date of the operating license. It also applies to requests for license renewal and the need to consider projected values of RTPTS ~t the end of a renewal term.

An additional change is proposed to be made in paragraph (b)(4) with regard to the schedule for submittal of a safety analysis justifying 6

operation beyond the screening criterion. In the present PTS rule, this analysis must be submitted at least 3 years before reaching the screening criterion or by one year after issuance of Co111T1ission guidance and acceptance criteria, whichever is later. Regulatory Guide 1.154, which contains the necessary guidance and criteria was issued in January, 1987.

Therefore, this alternative schedule was omitted in the proposed amendment.

However, because one or tw plants might reach the screening criterion in less than 3 years after publication, when RTPTS is recalculated using the amended rule, the subllittal will be required at least three years before reaching the screening criterion or by one year after the effective date of the amended rule, whichever is later. The safety implications of this change in the schedule requirement are considered to be acceptably small, because RTPTS increases very ~lowly near the-screening criterion .

Environmental Impact: Categorical Exclusion The NRC has determined that this proposed rule is the type of action described in categorical exclusion 10 CFR 51.22(c)(3)(ii) and (iii).

Therefore, neither an environmental impact statement nor an environmental assessment has been prepared for this proposed rule.

Paperwork Reduction Act Statement

~his proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.).

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This rule has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.

Public reporting burden for this collection of infor11ation is estimated to average 254 hours0.00294 days <br />0.0706 hours <br />4.199735e-4 weeks <br />9.6647e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments regarding this burden estimate or any

- other aspect of this collection of information, including suggestions for reducing this burden, to the Records and Reports Manage111ent Branch (P-530), U.S. Nuclear Regulatory Cormnission, Washington, DC 20555; and to the Paperwork Reduction Project (3150-0011), Office of Management and Budget; Washington, DC 205--03.-

Regulatory Analysis The NRC staff has prepared a regulatory analysis for this proposed amendment, which describes the factors and alternatives considered by the Coinmission in deciding to propose -this rule.

The regulatory analysis for the proposed U1endment also discusses why the screening criterion is not being changed when the procedures for calculating RTPTS are changed. An anticipated public comment is that because the probabilistic fracture mechanics calculat1ons used in establishing the screening criterion ade use of the formula for RTPTS given in the PTS rule, the proposed change in the formula must change

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the calculated probabilities and, in turn, change the screening crite-rion. As shown in the regulatory analysis, failure probabilities at the same RTPTS screening criterion for the ~ost critical accident scenarios in three plants, when recalculated using the new embrittlement estimates, were somewhat lower, but the differences were quite dependent on the plant configuration and the scenario chosen. Because of the apparent plant-to-plant differences, it is better to trigger plant-specific analyses with a "trip wire that is believed to generically bound all plants. Furthermore, as described in the regulatory analysis, the screening criterion was based on a variety of considerations besides the probabilistic analysis.

A copy of the regulatory analysis is available for inspection and copying for a fee at the NRC Public Document Room, 2120 L Street NW.

(Lower Level), Washington, DC 20555. Single copies of the analysis may be obtained from Pryor N. Randall, Office of Nuclecµ- Regulatory Research, U.S. Nuclear Regulatory C0111Dission, Washington, DC 20555, Telephone, (301)492-3842.

Regulatory Flexibility Act Certification As required by the Regulatory Flexibility Act, 5 U.S.C. 605(b), the Commission certifies that this proposed rule does not have a significant economic illJ)act on a substantial number of small entities. This rule specifies minimum fracture toughness properties of irradiated pressure vessel materials to ameliorate the effects of PTS events on nuclear facilities licensed under the provision of 10 CFR 50.2l(b) and 10 CFR 50.22. The companies that own these facilities do not fall within the 9

scope of the definition of "small entities" as set forth in the Regula-tory Flexibility*Act or the Slllall Business Size Standards in regulations issued by the small Business Administration at 10 CFR Part 121.

Backfit Analysis The NRC has concluded, on the basis of the documented evaluation required by 10 CFR 50.109(a)(4), that the backfit requirements contained in this proposed amendment are necessary to ensure that the facility pro-vides adequate protection to the public health and safety, and, therefore, that a backfit analysis is not required and the cost-benefit standards of 10 CFR 50.109(a)(3) do not apply. The documented evaluation given in the regulatory analysis includes a statement of the objectives of and reasons for the backfits that would be required by the proposed rule and sets forth the basis for the NRC's conclusion that these backfits are not sub-ject to the cost-benefit standards of 10 CFR 50.109(a)(3).

List of Subjects in 10 CFR Part 50 _

Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and

- reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act 10

of 1974, and 5 U.S.C. 553, the NRC is proposing to adopt the following amendments to 10 CFR Part SO.

PART SO -- DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

1. The authority citation of Part 50 is revised to read as follows:

AUTHORITY: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat.

936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat.

1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as mnended 1244, 1246, (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also i~sued under Pub. L.95-601, sec. 10, 92 Stat.

2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. 936, 955 as a11ended (42 U.S.C. 2131, 2235), sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C.

2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C.

4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat.

1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80-50-81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234).

Appendix Falso issued under sec. 187, 68 Stat 955 (42 U.S.C. 2237).

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For the purposes of sec. 223, 68 Stat. 958, as utended (42 U.S.C.

2273), §§ 50.46(a) and (b), and 50.54(c) are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); §§ 50.7(a), 50.lO(a)-(c),

50.34(a) and (e), 50.44(a)-(c), 50.46(a) and (b), 50.47(b), 50.48(a),

(c), (d), and (e), 50.49(a), 50.54(a), (i), (i)(l), (1)-(n), (p), (q),

(t), (v), and (y), 50.SS(f), 50.SSa(a), (c)-(e), (g), and (h), 50.59(c),

50.60(a), 50.62(c), 50.64(b), and 50.80(a) and (b) are issued under sec.

161i, 68 Stat. 949, as amended (42 U.S.C. 2201 (i)); and§§ 50.49(d),

(h), and (j), 50.54(w),(z),(bb),(cc), and (dd), 50.SS(e), 50.59(b),

50.6l(b), 50.62(~), 50.70(a), 50.71(a)-(c) and (e), 50.72(a), 50.73(a) and (b), 50.74, 50.78, and 50.90 are issued under sec. 161(0), 68 Stat.

950, as* 81lended (42 U.S.C. 2201(0)).

2. In§ 50.61, paragraph (b) is revised to read as follows:

§ 50.61 Fracture toughness requirements for protection against pressurized thermal shock events.

(b) Requirements.

(1) For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall submit projected values of RTPTS for reactor vessei beltline materials by giving values for the time of submittal, the expiration date of the operating license, the projected expiration date if a change in the operating license has been requested, and the projected expiration date of a renewal ter11 if a request for license renewal has been submitted. The assessment ust use the calculative procedures given in paragraph (b)(2) of this section.

The assessment must specify the bases for the projection, including the assumptions regarding core loading patterns. The submittal must 11st the ll

copper and nickel contents, and the fluence values used in the calcula-tion for each beltline material. If these quantities differ from those submitted in response to the original PTS rule and accepted by the NRC, justification J>>LJst be provided. This assessment must be submitted by (6 months after the effective date of this section), and must be updated whenever there is a significant change in projected values of RTPTS' or upon a request for a change in the expiration date for operation of the facility.

(2) The pressurized thermal shock (PTS) screening criterion is 270°F for plates, forgings, and axial weld materials, or 300°F for circumferential weld materials. For the purpose of c0111parison with this criterion, the value of RTPTS for the reactor vessel must be calculated as follows. The calculation must be made for each weld and plate, or forging, in the reactor vessel beltline.

Equation 1: RTPTS =I+ M + ARTPTS 11 (i) I 11 means the initial reference temperature (RTNDT) of the unirradiated material 111easured as defined in the ASHE Code, Paragraph NB-2331. Measured values must be used if available; if not, the follow-ing generic mean values must be used: 0°F for welds ade with Linde 80 flux, and -56°F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.

(ii) 11 M11 means the margin to be added to cover uncertainties in the values of initial RTNDT' copper and nickel contents, fluence and the calculational procedures. In Equation 1, Mis 66°F for welds and 48°F 13

for base metal if generic values of I are used, and Mis 56°F for welds and 34°F for base metal if measured values of I are used.

(iii} ARTPTS is the aean value of the adjustment in reference tempera-ture caused by irradiation and should be calculated as follows:

Equation 2: ~RTPTS = (CF)f (0.28-0.10 log f)

(iv) CF (°F) is the chemistry factor, a function of copper and nickel content. CF is given in Table 1 for welds and in Table 2 for base metal (plates and forgings). Linear interpolation is permitted. In Tables 1 and 2 "Wt-% copper11 and "Wt-% nickel" are the best-estimate values for the aaterial, which will normally be the mean of the aeasured values for a plate or forging or for weld samples 111ade with the weld wire heat nllllber that matches the critical vessel weld. If these values are not available, the upper limiting values given in the material specifications to which the vessel was built may*be used. If not avail-able, conservative estimates (mean plus one standard deviation) based on generic data may be used if justification is provided. If there is no information available, 0.35% copper and 1.0% nickel must be assumed.

19 (v) ur means the best estiinate neutron fluence, in units of 10 n/cm2- (E greater than 1 HeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the mterial in ques-tion receives the highest fluence for the period of service in question.

(3) For each pressurized water nuclear power reactor for which the value of RTPTS for any material in the beltline is projected to exceed the PTS screening criterion before the expiration date of the operating renewal has been submitted, the licensee shall submit by (9 aonths after 14

TABLE 1 CHEMISTRY FACTOR FOR WELDS, °F Cop~er, Ni eke 1, Wt-%

Wt- 0 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02 21 26 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 24 43 54 54 54 54 54 0.05 26 49 67 68 68 68 68 0.06 29 52 77 82 82 82 82 0.07 32 55 85 95 95 95 95 0.08 36 58 90 106 108 108 108 0.09 40 61 94 115 122 122 122 0.10 44 65 97 122 133 135 135 0.11 49 68 101 130 144 148 148 0.12 52 72 103 135 153 161 161 0.13 58 76 106 139 162 172 176 0.14 61 79 109 142 168 182 188 0.15 66 84 112 146 175 191 200 0.16 70 88 115 149 178 199 211 0.17 75 92 . 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191. 220 238 0.20 88 104 129 160 194" 223 245 0.21 92 108 133 164 197 229 252 0.22 97 112 137 167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 239 268 0.25 110 126 148 176 209 243 272 0.26 113 130 151 180 212 246 276 0.27 119 134 155 184 216 249 280 0.28 122 138 160 187 218 251 284 0.29 128 142 164 191 222 254 287 0.30 131 146 167 194 225 257 290 0.31 136 151 172 198 228 260 293 0.32 140 155 175 202 231 263 296 0.33 144 160 180 205 234 266 299 0.34 149 164 184 209 238 269 302 0.35 153 168 187 212 241 272 305 0.36 158 172 191 216 245 275 308 0.37 162 177 196 220 248 278 311 0.38 166 182 200 223 250 281 314 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 15

TABLE 2 CHEMISTRY FACTOR FOR BASE METAL, °F Cop~er, Nickel, Wt-%

Wt- 0 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02 20 20 20 20 20 20 20 0.03 20 20 20 20 20 20 20 0.04 22 26 26 26 26 26 26 0.05 25 31 31 31 31 31 31 0.06 28 37 37 37 37 37 37 0.07 31 43 44 44 44 44 44

- 0.08 0.09 0.10 0.11 0.12 0.13 34 37 41 45 49 53 48 53 58 62 67 71 51 58 65 72 79 85 51 58 65 74 83 91 51 58 67 77 86 96 51 58 67 77 86 96 51 58 67 77 86 96 0.14 57 75 91 100 105 106 106 0.15 61 80 99 110 115 117 117 0.16 65 84 104 118 123 125 125 0.17 69 88 110 127 132 135 135 0.18 73 92 115 134 141 144 144 0.19 78 97 120 142 150 154 154 0.20 82 102 125 149 159- 164 165 0.21 86 107 129 155 167 172 174 0.22 91 112 134 161 176 181 184 0.23 95 117 138 167 184 190 194 0.24 100 121 143 172 191 199 204 0.25 104 126 148 176 199 208 214 0.26 109 130 151 180 205 216 221 0.27 114 134 155 184 211 225 230 0.28 119 138 160 187 216 233 239 0.29 124 142 164 191 221 241 248 0.30 129 146 167 194 225 249 257 0.31 134 151 172 198 228 255 266 0.32 139 155 175 202 231 260 274 0.33 144 160 180 205 234 264 282 0.34 149 164 184 209 238 268 290 0.35 1~.3 168 187 212 241 272 298 0.36 158 173 191 216 245 275 303 0.37 162 177 196 220 248 278 308 0.38 166 182 200 223 250 281 313 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 16

renewal has been submitted, the licensee shall submit by (9 months after the effective date of this section) an analysis and schedule for imple-mentation of such flux reduction progra11ts as are reasonably practicable to avoid exceeding the PTS screening criterion set forth in para-graph (b)(2) of this section. The schedule for implementation of flux reduction measures may take into account the schedule for submittal and anticipated Commission approval of detailed plant-specific analyses, sub-mitted to demonstrate acceptable risk at values of RTPTS above the screening limit due to plant modifications, new information or new anal-ysis techniques.

(4) For each pressurized water nuclear power reactor for which the analysis required by paragraph (b)(3) of this section indicates that no reasonably practicable flux reduction program will prevent the value of I

RTPTS from exceeding the ,PTS screening criterion before the expiration date of the operating license, or the projected_expiration date if a change in the operating license has been requested, or the end of a renewal term if a request for license renewal has been submitted, the licensee shall submit a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent potential failure of the reactor vessel as a result of postu-lated PTS events if con~inued operation beyond the screening criterion is allowed. In the analysis, the licensee may determine reactor vessel materials properties based on available information, research results, and plant surveillance data, and may use probabilistic fracture mechanics techniques. This analysis must be submitted at least 3 years before the value pf RTPTS as defined in paragraph (b)(2) of this section is pro-jected to exceed the PTS screening criterion or by one year after the effective date of this amendment, whichever is later.

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(5) After consideration of the licensee's analyses (including effects of proposed corrective actions, if any) submitted in accordance with paragraphs (b)(3) and (b)(4) of this section, the Commission may, on a case-by-case basis, approve operation of the facility at values of RTPTS in excess of the PTS screening criterion. The Commi5sion will consider factors significantly affecting the potential for failure of the reactor vessel in reaching a decision.

(6) If the Comission concludes, pursuant to paragraph (b)(5) of re this section, that operation of the facility at values of RTPTS in excess of the PTS screening criterion cannot be approved on the basis of the licensee's analyses submitted in accordance with paragraphs (b)(3) and (b){4) of this section, the licensee shall request and receive Comission approval prior to any operation beyond the criterion. The request must be based upon modifications to equipment, systems, and operation of the facility in addition to those previously proposed in the submitted anal-yses that would reduce the potential for failure* of the reactor vessel due to PTS events, or upon further analyses based upon new information or improved methodology.

Dated at Rockville, MD this *;:;~ day o f ~ , 1989.

For the Nuclear Regulatory Co111111ission.

or for Operations 18