ML060970173

Draft Regulatory Guide DG-1144, Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors

ML060970173
Person / Time
Issue date:	07/20/2006
From:	Hipolito Gonzalez NRC/RES/DFERR
To:
Gonzalez H, NRC/RES/DFERR, 415-0068
References
DG-1144
Download: ML060970173 (12)

Text

U.S. NUCLEAR REGULATORY COMMISSION July 2006 OFFICE OF NUCLEAR REGULATORY RESEARCH Division 1 DRAFT REGULATORY GUIDE Contact s: H.J. Gonzalez, (301) 415-0068 DRAFT REGULATORY GUIDE DG-1 1 4 4 GUIDELINES FOR EV ALUATING FATIGUE ANALYSES INCORPORATING THE LIFE REDUCTION OF M ETAL COM PONENTS DUE TO THE EFFECTS OF THE LIGHT-W ATER REACTOR ENV IRONM ENT FOR NEW REACTORS A. INTRODUCTION In Appendix A, General Design Criteria for Nuclear Pow er Plants, to Title 10 , Part 50 ,

of the Code of Federal Regulations (10 CFR Part 5 0), Domestic Licensing of Product ion and Utilization Facilities, General Design Criterion (GDC) 1, Quality Standards and Records, requires, in part, that st ruct ures, systems, and components t hat are import ant t o safety must be designed, fabricated, erected, and tested to quality standards commensurate w ith the importance of the safety function performed. In addition, GDC 30, Quality of Reactor Coolant Pressure Boundary, requires, in part, that components t hat are part of the react or coolant pressure boundary must be designed, f abricat ed, erect ed, and tested to the highest pract ical qualit y standards.

Augment ing t hose design criteria, 10 CFR 50.55a, Codes and Standards, endorses the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for design of safety-related syst ems and components. In particular, Sect ion 50 .5 5a(c), React or Coolant Pressure Boundary, requires, in part, t hat components of t he reactor coolant pressure boundary must meet the requirements for Class 1 components in Sect ion III, Rules for Construct ion of Nuclear Pow er Plant Components, of the ASME Boiler and Pressure Vessel Code. Specif ically, those Class 1 requirements contain provisions, including fatigue design curves, f or determining a components suitability for cyclic service. These fatigue design curves are based on strain-controlled test s performed on small polished specimens, at room temperat ure, in air environment s. Thus, these curves do not address the impact of the react or coolant system environment.

This regulatory guide is being issued in draft form to involve the public in the early stages of t he development of a regulatory position in this area. It has not received staff review or approval and does not represent an official NRC staff position.

Public comments are being solicit ed on this draft guide (including any implementation schedule) and its associated regulatory analysis or value/impact statement . Com ment s should be accompanied by appropriate support ing dat a. Writ ten comment s may be submit ted to the Rules and Directives Branch, Off ice of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555 -0001.

Comm ent s may be submit ted electronically through the NRCs int eract ive rulemaking Web page at htt p://w w w .nrc.gov/w hat-w e-do/regulatory/rulemaking.html. Copies of comments received may be examined at t he NRC Public Document Room, 11555 Rockville Pike, Rockville, MD. Comments w ill be most helpful if received by September 25, 2006 .

Requests f or single copies of draft or act ive regulatory guides (w hich may be reproduced) or for placement on an automatic dist ribution list f or single copies of f uture draft guides in specific divisions should be made to t he U.S. Nuclear Regulatory Commission, Washington, DC 20 55 5, At tention: Reproduction and Distribution Services Section, or by fax to (301)4 15 -22 89 ; or by email to Distribution@nrc.gov. Electronic copies of t his draft regulatory guide are available through the NRCs int eract ive rulemaking Web page (see above); the NRCs public Web sit e under Draft Regulatory Guides in t he Regulat ory Guides document collect ion of the NRCs Electronic Reading Room at ht tp:// w w w .nrc. gov/reading-rm/doc-collections/; and the NRCs Agencyw ide Documents Access and Management Syst em (A DAMS) at htt p://w w w .nrc.gov/reading-rm/adams.html, under A ccession No. ML060970173.

This draft regulatory guide provides guidance for use in determining the acceptable fatigue lif e of ASME pressure boundary components, w it h consideration of the light -w ater reactor (LWR) environment . In so doing, this guide describes a methodology that the st aff of the U.S. Nuclear Regulat ory Commission (NRC) considers accept able to support review s of applications that the agency expects to receive for new nuclear react or construction permits or operating licenses under 10 CFR Part 50 , design certifications under 10 CFR Part 52 ,

and combined licenses under 10 CFR Part 52 t hat do not reference a standard design.

Because of signif icant conservatism in quant if ying other plant -related variables (such as cyclic behavior, including stress and loading rates) involved in cumulative fatigue life calculations, the design of the current fleet of reactors is sat isfactory, and the plants are safe to operate.

The NRC issues regulat ory guides t o describe t o t he public met hods that the st aff considers acceptable for use in implementing specif ic part s of the agency s regulations, to explain techniques that the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to applicant s. Regulatory guides are not substit ut es for regulations, and compliance w it h regulat ory guides is not required. The NRC issues regulat ory guides in draft form to solicit public comment and involve the public in developing the agencys regulatory positions. Draf t regulatory guides have not received complete st aff review and, t herefore, t hey do not represent official NRC staff posit ions.

This regulatory guide contains information collections that are covered by the requirements of 10 CFR Part s50 and 52, w hich t he Of fice of Management and Budget (OMB) approved under OMB control numbers 3150-0011 and 3150-0151, respectively. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information collection request or requirement unless t he request ing document displays a currently valid OM B control number.

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B. DISCUSSION The ASME Sect ion III design curves, developed in t he late 1960s and early 1970s, are based on tests conducted in laboratory air environments at ambient temperatures.

The original code developers applied margins of 2 on strain and 20 on cyclic lif e to account for variations in materials, surface finish, data scatter, and environmental effects (including temperature differences bet w een specimen test conditions and react or operating experience).

How ever, the developers lacked sufficient data to explicitly evaluate and account for the degradat ion at tributable to exposure t o aqueous coolants. More recent fatigue t est dat a from the Unit ed St ates, Japan, and elsew here show that the LWR environment can have a significant impact on the fatigue life of carbon and low -alloy steels, as w ell as austenitic stainless steel.

Tw o distinct methods can be used to incorporate LWR environmental eff ects into the fatigue analysis of ASME Class 1 components. The first method involves developing new fatigue curves that are applicable to LWR environments. Given that the fatigue life of ASME Class 1 components in LWR environment s is a funct ion of several parameters, this method w ould necessitate developing several fatigue curves t o address pot ent ial paramet er variations. An alternative w ould be t o develop a single bounding fatigue curve, w hich may be ov erly conservative for most applications. The second method involves using an environmental correct ion f act or (Fen ) to account for LWR environments by correcting the f atigue usage calculated w it h t he ASME air curves. This met hod af fords the designer greater flexibility to calculate the appropriate impacts for specific environmental parameters.

In addition, applicant s have already used this met hod in t heir license renew al applications.

The NRC st aff has selected t he Fen method, as described in NUREG/CR-6909, Effect of LWR Coolant Environments on the Fatigue Life of React or Materials. In particular, Appendix A to that report, Incorporating Environmental Eff ects into Fatigue Evaluations, describes a methodology that t he staff considers acceptable to incorporate the effects of reactor coolant environments on fatigue usage factor evaluations of metal components.

In addition, NUREG/CR-6909 provides a comprehensive review of, and technical basis for, the methodology proposed in this draft regulatory guide, including analysis of each paramet er aff ecting the fatigue evaluations. In developing the underlying statistical models, the researchers analyzed existing data to predict fatigue lives as a f unction of temperat ure, st rain rate, dissolved oxygen level in w ater, and sulfur content of the steel. The resultant met hod post ulates a strain threshold, below w hich environment al effects on fatigue life do not occur.

By definition, Fen is the ratio of f atigue life of the component material in a room temperature air environment to its f atigue lif e in LWR coolant at operat ing t emperat ure. To incorporat e environment al ef fects int o t he f atigue evaluation, the f atigue usage is calculated using ASME Sect ion III Code provisions, and the fatigue design curve is multiplied by the correction factor.

A second concern regarding the ASME fatigue design curves involves nonconservatism of the current ASME stainless st eel air design curve. More recent evaluations of stainless st eel test data indicate that the ASME curve is inconsistent w ith the appropriate test materials and conduct of the fatigue t est. Consequently, through t his draft regulat ory guide, t he NRC staff endorses a new stainless st eel air design curve. Sect ion 5 .1.8 of NUREG/CR-6909 provides a comprehensive review of , and technical basis f or, that new design curve. The Fen def ined f or st ainless st eel in NUREG/CR-6909 should be used in conjunction w it h t he new stainless steel air design curve w hen evaluating the fatigue usage of ASME Class 1 components.

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In addition, Sect ion 6 of NUREG/CR-6909 includes an evaluation of the ASME design curve margins. In conduct ing that evaluation, the researchers review ed data available in the literature to assess the subfactors (excluding environment) that are needed to account for the ef fects of various uncert ainties and differences bet w een actual components and laboratory t est specimens. The researchers also performed statist ical analyses using Monte Carlo simulations to develop f atigue design curves, using the 95/95 criterion that the curves should provide 95% confidence, and 95% of the population w ill have a great er fatigue life than predicted by t he design curves. This criterion w as deemed acceptable because the fatigue design curves are based on crack initiation, rat her than component failure and, therefore, there is addit ional margin betw een crack initiation and actual component failure.

This conclusion is support ed by a risk st udy of fatigue crack init iat ion and grow th in actual LWR components, as documented in NUREG/CR-6674, Fatigue Analysis of Components for 60-Year Plant Lif e, w hich t he NRC published in June 2000. That risk st udy det ermined that t he estimated core damage frequency is low for components t hat have a relatively high probability of fatigue crack init iation.

The results of the Monte Carlo simulations indicate that for both carbon and low -alloy steels and austenitic stainless steels, t he current ASME Code procedure of adjusting the mean test data by a factor of 20 f or life is conservative compared to the 95/95 criterion.

The results also indicate that a minimum factor of 12 for cyclic life of both carbon and low -alloy st eels and aust enitic st ainless st eels w ill satisfy the 95/95 criterion. The result ant new air design curves, using margins of 1 2 f or life and 2 for stress, are show n in Figures 9, 10, and 37 of NUREG/CR-6909 for carbon steel, low -alloy steel, and austenitic stainless steel, respectively. These new air design curves are used in this draft regulatory guide; thus, if an applicant chooses to use the procedure discussed in this guide to determine the fatigue life of st ainless st eels, these air design curves should be used. How ever, the existing A SME air design curves for carbon and low -alloy steels may also be used w ith the procedure in this guide to determine the fatigue life of those materials, since their use w ill yield conservative results.

Several methods for calculating Fen w ere review ed and found acceptable.

Only the types of stress cycles or load set pairs that exceed strain threshold criteria for carbon and low -alloy steels and austenitic stainless steels need to be considered for Fen calculations.

The evaluation options depend on the complexit y of the analyzed transient condition and the detail of the evaluation. For example, in an evaluation w here the results of det ailed transient analyses are available to determine the necessary parameters (strain rat e, temperature, and others), the modified rate approach (presented and referenced in Section 4.2.14 of NUREG/CR-6909) is an accept able met hodology for det ermining t he Fen values.

This methodology involves a strain-based integral for evaluating conditions for w hich temperature and strain rate change, resulting in variation of Fen over t ime. This det ailed approach calculates t he Fen values based on the strain history f or each load set in t he fatigue analysis evaluation, considering the effects of strain rate and temperature variations for each incremental segment in the strain history. Such results may be used to reduce the conservatism in the calculated Fen values. For a simplified calculation yielding a more conservative result for a complex or poorly defined set of transients, the temperature is equal to the average temperat ure in the t ransient or segment . The calculated Fen values are t hen used to incorporat e environmental eff ects int o ASME fatigue usage factor evaluat ions.

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C. REGULATORY POSITION This sect ion describes the methods t hat the staff considers acceptable for use in performing fatigue evaluations, considering the effects of LWR environments on carbon and low -alloy st eels, as w ell as austenitic stainless steels. Specif ically, these methods include calculating t he f atigue usage in air using ASME Code analysis procedures, and then employing t he environmental correct ion f act or (Fen ), as described in NUREG/CR-6909, Effect of LWR Coolant Environments on the Fatigue Life of React or Materials. In particular, Appendix A to that report, Incorporating Environmental Eff ects into Fatigue Evaluations, includes detailed descriptions and additional guidance concerning the overall methodology and all equations referred to in this sect ion.

1. Carbon and Low-Alloy Steels The follow ing procedure should be used to calculate the environmental fatigue usage of carbon and low -alloy steel components in LWR environment s.

1.1 Fatigue Usage in Air Calculate the fatigue usage in air using ASME Code analysis procedures and the fatigue curves provided in NUREG/CR-6909, Sect ion 4.1.9 , Figures 9 and 10 .

1.2 Environmental Correction Factor (Fen)

Calculate the environmental correction factor, Fen , using Equation A .2 of NUREG/CR-6909 for carbon steels, or Equation A.3 of NUREG/CR-6909 for low -alloy steels. The respect ive parameters should be calculated using Equations A.4 t hrough A.7 of NUREG/CR-6909 .

The strain t hreshold is show n in Equation A .8 of NUREG/CR-6909.

1.3 Environmental Fatigue Usage Calculat e the environmental fatigue usage using Equation A .14 of NUREG/CR-6909.

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2. Austenitic Stainless Steels The follow ing procedure should be used to calculate the environmental fatigue usage of austenitic st ainless steel components in LWR environment s.

2.1 Fatigue Usage in Air Calculate the f atigue usage in air using ASME Code analysis procedures and the new fatigue curve provided in NUREG/CR-6909, Sect ion 5.1.8, Figure 37.

2.2 Environmental Correction Factor (Fen)

For all t ypes of austenit ic stainless st eels (e.g., Types 304, 310, 316, 347, and 348),

calculate Fen using Equation A .9 of NUREG/CR-6909. The respective paramet ers are def ined in Equations A.10 through A.12 of NUREG/CR-6909. The st rain t hreshold is provided in Equation A .13 of NUREG/CR-6909.

2.3 Environmental Fatigue Usage Calculat e the environmental fatigue usage using Equation A .14 of NUREG/CR-6909.

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D. IMPLEMENTATION The purpose of this section is t o provide information t o applicant s and licensees regarding the NRC staff s plans for using this draft regulatory guide. This draft regulatory guide only applies to new plants and no backf itting is intended or approved in connect ion w it h its issuance.

The NRC has issued this draft guide to encourage public participation in its development.

Except in those cases in w hich an applicant or licensee proposes or has previously established an acceptable alternative method for complying with specified portions of the NRCs regulations, the methods to be described in the f inal guide w ill reflect public comment s and w ill be used in evaluating submitt als in connection with applications for construction permit s, standard plant design cert if icat ions, operating licenses, early site permits, and combined licenses.

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REFERENCES ASME Boiler and Pressure Vessel Code, Sect ion III, Rules f or Const ruction of Nuclear Pow er Plant Components, American Society of M echanical Engineers, New York, NY, 1992. 1 Chopra, O.K., Eff ect of LWR Coolant Environments on Fatigue Life of Reactor Materials, NUREG/CR-6909 (draf t), ANL-0 6/08, U.S. Nuclear Regulat ory Commission, Washingt on, DC, April 2006.2 Khaleel, M.A., et al., Fatigue Analysis of Components for 60-Year Plant Life, NUREG/CR-6674, U.S. Nuclear Regulat ory Commission, Washington, DC, June 2000. 2 VanDerSluys, W. Alan, PVRC Position on Environmental Eff ects on Fatigue Life in LWR Applications , Welding Research Council Bulletin 4 87, Welding Research Council, Inc.,

New York, NY. 3 1

Copies may be purchased f rom t he American Society of Mechanical Engineers, Three Park A venue, New York, NY 10016-5990; phone (212) 591-8500; f ax (212) 591-8501; w w w .asme.org.

2 Copies are available at current rates from the U.S. Government Printing Off ice, P.O. Box 3708 2, Washington, DC 20 40 2-9 32 8 (t elephone 20 2-5 12 -18 00 ); or from the National Technical Information Service (NTIS) by w riting NTIS at 528 5 Port Royal Road, Springfield, VA 22161; http://w w w .nt is.gov; telephone 703-487-4650. Copies are available for inspection or copying f or a fee from t he NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; t he PDR s mailing address is USNRC PDR, Washington, DC 20555 (telephone: 301-415-4737 or 800-397-4209; fax: 301-415-3548; email: PDR@nrc.gov). Draft NUREG-series reports for public comment are also available elect ronically through the NRC s public Web site at ht tp://w w w .nrc.gov/reading-rm/doc-collections/nuregs/

docs4comment.ht ml. NUREG/CR-66 74 is also available through the NRCs Agencyw ide Documents Access and Management System (ADAMS) at htt p://w w w .nrc.gov/reading-rm/adams.html, under Accession No. M L0037 24 21 5, and NUREG/CR-69 09 is available in ADAMS under Accession No. ML061650347.

3 Welding Research Council Bulletin 487 is available for purchase from Welding Research Council, Inc.,

PO Box 201 54 7, Shaker Heights, Ohio (telephone: 216-6 58 -38 47 ). Purchase information is available online at http://w w w .f orengineers.org/cgi-bin/w rcbulletin/bulletin.pl?action= view ;id= 49 7.

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REGULATORY ANALYSIS

1. Issue The staff of the U.S. Nuclear Regulatory Commission (NRC) proposes to develop and issue a new regulatory guide, entit led Guidelines for Evaluating Fatigue Analyses Incorporat ing t he Life Reduction of Metal Components Due to t he Effects of the Light-Water Reactor Environment f or New Reactors. This new guide is a means to provide guidance to support license applications for new nuclear react or construct ion by determining accept able fatigue lif e assessments of reactor vessel pressure boundary components, w it h consideration f or the ef fects of a light-w ater react or (LWR) environment . The st aff proposes t o issue a draf t guide for public review and comment , resolve any st akeholder comment s, and then finalize and implement the guide.

On September 25, 1995, t he NRC staff submitted for Commission approval a Fatigue Act ion Plan (SECY-95-245), w hich addressed issues associated w ith assessing fatigue performance of structural components in LWR environments. The related Generic Safety Issue (GSI) 166, Adequacy of the Fatigue Life of Metal Components, evaluated concerns regarding the conservatism of t he fatigue curves used in designing existing LWR components.

Both SECY-95-245 and GSI-166 concluded that t he NRC need not take any major act ions regarding t he env ironment al ef fects f or current nuclear pow er plants, and the st aff resolved the issue addressed in GSI-166 f or the initial 40-year design life of operating components.

Nonetheless, t o address license renewal, the staff subsequent ly ident if ied GSI-190, Fatigue Evaluation of M etal Components for 60-Year Plant Life.

The NRC closed GSI-190 in December 1999, concluding that no generic regulatory action w as required. The st aff based this conclusion primarily on the negligible calculated increases in core damage frequency in extending a plant s operating life from 40 to 60 years. However, the calculations supporting the resolution of this issue, w hich included consideration of environmental eff ects and the nature of age-related degradation indicate the potential for an increase in t he f requency of pipe leaks as plant s continue t o operate. Thus, the st aff concluded that, consistent w ith exist ing requirements in 10 CFR 54.2 1, Contents of Applications Technical Information, licensees should address the eff ects of the coolant environment on component f atigue life as they f ormulate their aging management programs in support of license renew al.

The evaluations used in resolving GSI-16 6 and GSI-190 relied on conservatism in the existing component f atigue analyses. How ever, fatigue analyses for components of new reactors may not contain the same degree of conservatism. By let ter to the Chairman of the ASME Board of Nuclear Codes and Standards, dat ed December 1, 1999, the NRC st aff request ed t hat ASME modify it s Boiler and Pressure Vessel Code to include env ironment al eff ects in the fatigue design of components. In response, A SME initiated the PVRC Steering Committee on Cyclic Life and Environmental Effects, w hich recommended revising the Code fatigue design curv es (Welding Research Council Bulletin 4 87, PVRC Position on Environmental Effects on Fatigue Life in LWR Applications ); how ever, despite years of deliberation concerning the recommended methods and approaches to resolve concerns regarding environmental effects on fatigue life under LWR conditions, the ASME Subcommit tee on Environmental Fatigue has not reached a decision. Consequently, to move ahead, the NRC staff needs to develop a regulat ory position f or use in review ing applications for new plant construct ion.

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In 1 0 CFR 50.55a, Codes and Standards, the NRC endorses t he ASME Boiler and Pressure Vessel Code for design of safety-related systems and components. Sect ion III, Subsect ion NB, of the Code contains guidance for the design of Class 1 nuclear pow er plant components, as w ell as criteria for determining a component s suitability for cyclic service.

Figures I-9.1 through I-9.6 of Appendix I to Section III specif y Code fatigue design curves that are used in making this determination. These fatigue design curves, w hich w ere developed in the late 19 60 s and early 197 0s, are based on strain-controlled tests performed on small polished specimens, at room temperature, in air environments. The test specimen best-fit dat a curves w ere adjusted for mean stress effects and then low ered by a factor of 2 for stress or 20 f or cycles (w hichever was more conservative) to establish the design curves.

These factors of 2 and 20 do not constitute safety margins; these factors w ere simply applied to the experimental data, in order to estimate the fatigue life of actual reactor components.

Moreover,Section III, Subsection NB-3121, of the ASME Code specif ies that experimental data used to develop the fatigue design curves do not include tests of specimens exposed to simulated LWR coolant, w hich might accelerate fatigue failure. (At the time, it w as not possible to account for t he degradation attribut able to exposure to aqueous coolant s.)

After about 20 years of research effort addressing the environmental degradation of fatigue crack nucleation, it has become apparent that exposure to LWR environments has a detrimental eff ect on the fatigue life of metal components, w hich affects the major categories of structural materials (i.e., carbon st eel, low -alloy steel, and austenitic st ainless steel). On the basis of t hat revelation, the NRC completed a multi-year study t o develop and evaluate the environment al fatigue test dat a. In conduct ing this study, Argonne National Laboratory (ANL) developed statist ical correlations that can be used to evaluate the fatigue life of ASME Code components in LWR environments. The results of t his study appear in NUREG/CR-6583, Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low -Alloy Steels, dated February 1 998, and NUREG/CR-5704, Effects of LWR Coolant Environment s on Fatigue Design Curves of Austenitic Stainless St eels, dat ed April 1999. In general, t he study results show that degradation is exacerbated by increasing temperat ure, decreasing loading rat e, increasing sulfur content of the materials, and oxygen cont ent of the coolant (for carbon and low -alloy st eels), and decreasing oxygen cont ent of the coolant (for aust enit ic stainless st eels). Not ably, the models developed by ANL are w ell-recognized in the international communit y, although ot her researchers in both the domest ic and international communit ies (Japan) have developed ot her approaches and methodologies for assessing t he environmental eff ect on fatigue analyses.

The NRC st aff has not previously issued a regulatory guide on the matter of acceptable fatigue lif e assessment s of ASME pressure boundary components, w it h consideration for the effects of an LWR environment. The staff anticipates that this regulatory guidance w ill be applicable to future applicants for new nuclear reactor construction permits or operating licenses under 10 CFR Part 50 , design certifications under 10 CFR Part 52 ,

and combined licenses under 10 CFR Part 5 2 that do not reference a standard design.

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2. Alternative Approaches The NRC staff considered the follow ing alt ernative approaches:

(1) Do not provide guidance.

(2) Endorse the ASME Code Case Standard initiative addressing the environmental effect on f atigue lif e reduction of metal components.

(3) Issue a new regulatory guide.

2.1 Alternative 1: Do Not Provide Guidance Under this no action alternative, t he staff w ould not issue regulatory guidance regarding the assessment of ASME pressure boundary components, w ith consideration for the ef fects of an LWR environment . This alt ernative w ill result in unnecessary burden for NRC staff and licensees, in connect ion w it h preparing and responding t o requests for addit ional information (RAIs), as w ell as re-analyses and supplement ation of license amendment applications. This alt ernative does not support any of the NRCs saf ety performance goals.

2.2 Alternative 2: Endorse the ASME Code Case Standard Initiative Under t his alt ernative, the st aff w ould not develop its ow n regulat ory guidance, but w ould endorse an accept able indust ry st andard. The ASME Board of Nuclear Codes and Standards, Subcommit tee on Environmental Fatigue, is still developing a Code Case and non-mandatory procedure to provide guidance regarding the application of an environmental correction factor for fatigue analyses. This task w as assigned to the PVRC Steering Commit tee on Cyclic Life and Environmental Effects, w hich recommended revising the Code fatigue design curves (Welding Research Council Bulletin 4 87, PVRC Posit ion on Environment al Effects on Fatigue Life in LWR Applications ); how ever, despite years of deliberation, the ASME Subcommit tee on Environment al Fatigue has not yet approved t his proposal and has not reached a consensus regarding the approach or methodology that w ill be used for guidance.

The NRC staff, w ith support from ANL, has review ed these proposed methodologies.

Alt hough some aspects (e.g., the Fen approach) are considered accept able, the st aff st ill has concerns regarding t he bases and adequacy of ot her aspect s (e.g., the Z fact or).

The st aff does not ant icipate imminent development or consensus to f inalize the industry-st andard guidance; therefore, this alt ernative is no longer viable.

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2.2 Alternative 3: Issue a New Regulatory Guide Under this alternative, t he staff w ould develop a new regulatory guide as a means to provide guidance to support license applications for new nuclear react or construct ion by determining acceptable fatigue life assessments of react or vessel pressure boundary components, w ith consideration for the effects of an LWR environment. As such, t his alternative supports three of the NRCs five nuclear react or safety performance goals.4 (The tw o remaining goals, namely Security and Management, are not applicable to this guide.)

Specifically, issuing a new regulatory guide w ould (1) ensure protection of public health and safety and the environment by ensuring that safety analyses use appropriate analysis assumptions and methods, (2) ensure openness by involving the public in the development of this regulatory guidance t hrough t he public comment period, and (3) improve efficiency and effectiveness by providing licensees w it h t he staff s regulat ory posit ion, thereby minimizing RAIs and resubmitt als, and ensuring the adequacy of safety analyses. Also, this alternative w ill ensure availability of guidance for indust ry and NRC st aff review s upon the ant icipated receipt of new reactor const ruction license applications. Thus, the st aff has determined t hat this alternative is the most advantageous w ay t o address the need for regulatory guidance to enable both the industry and NRC staff to perform adequate fatigue analyses that appropriately incorporat e the reduced fatigue life of metal components att ributable to an LWR environment.

In developing this regulatory guidance, the st aff w ill allow a public comment period to resolve ongoing issues betw een the staff and industry /public st akeholders on the methodology and approach endorsed by the guidance.

3. Values and Impacts The proposed act ion is to issue a new regulatory guide. Therefore, compliance with the regulat ory position set forth in t he guide w ill be voluntary for new reactor const ruction license applicants. As w ith all regulatory guides, an applicant may propose alternative approaches to demonst rate compliance w it h the NRCs regulations.

This guidance w ill complement and be consist ent w it h current est ablished practices applied throughout the commercial nuclear power industry for license renew al evaluations.

Therefore, costs associated w ith implementing this guidance are expected to be minimal.

This guidance w ill apply to new nuclear pow er plants.

4. Conclusion Experience in review ing licensee renew al applications has demonstrat ed t he need for guidance in performing adequate fatigue analyses t hat incorporat e the reduced f atigue lif e of metal components attribut able to an LWR environment. Recent expressions of int erest related to future licensing of new reactors also indicate a need for new regulatory guidance.

Based on this regulat ory analysis, the st aff recommends t hat the NRC (1) prepare a new regulatory guide to support license applications for new nuclear react or construct ion by determining acceptable fatigue life assessments of react or vessel pressure boundary components, w it h consideration f or the ef fects of an LWR environment , (2) issue the draf t regulatory guide for public comment, and (3) finalize the regulatory guide upon resolution of public comments.

4 This alternative is not relevant to the NRCs tw o remaining performance goals of Security and Management.

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