ML20198F658
| ML20198F658 | |
| Person / Time | |
|---|---|
| Issue date: | 01/08/1998 |
| From: | Shapaker J NRC (Affiliation Not Assigned) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-05, GL-97-5, TAC-M96401, NUDOCS 9801120105 | |
| Download: ML20198F658 (1) | |
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\*****/ January 8, 1998 MEMORANDUM T0: Document Control Desk Information and Records Management Branch Information Management Division Office of the Chief Information Officer FROM: James W. Shapaker /
Events Assessment, gneric Commun ations and Special Inspections Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation
SUBJECT:
DOCUMENTS ASSOCIATED WITH NRC GENERIC LETTER 97-05 STEAM GENERATOR TUBE INSPECTION TECHNIQUES (TAC NO M96401)
The Materials and Chemical- Engineering Branch (EMCB) in the Divisison of Engineering (DE) prepared the subject generic letter, which was issued on December 17, 1997, and given accession number 9712120014. There is material related to the subject generic letter that should be placed in the NRC Public Document Room and made available to the public.' Therefore. by copy of this memorandum. I am providing the following documents to the NRC Public Document Room: (1) a copy of the published version of the subject generic letter, (2) a copy of the information paper (SECY-97-280) that was sent to'the Commission, (3) a copy of each letter received in response to the notice of oportunity for public comment on the proposed generic letter that was published in the Federal Register on December 31, 1996, and (4) a copy'of the CRGR review package.
I request that you provide me with the Nuclear Docwents System accession number for this memorandum.' This information;may be provided by telephone
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(415-1151) or by e-mail (JWS). In addition, please modify the appropriate NUDOCS entries.to reflect the fact that the documents identified herein are related to Generic Letter 97-05. ,
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. 1 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 December 17,1997 l
NRC GENERIC LETTER 97 05: STEAM GENERATOR TUBE INSPECTION TECHNIQUES l
Mdressess 1 All holders of operating licenses for pressurized water reactors (PWRs), except those who have permanently ceased operations and have certified that fus! has been permanently removed from the teactor vessel.
Puroose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1) emphasize to the addressees the importance of performing steam generator tube inservice inspections using qualified techniques in accordance wita the requirements of Appendix B to 10 CFR Part 50, and (2) require certain information from addressr w to debrmine whether they are in compliance with the current licensing basis for their respective facnities given thelc steam generator tube Inservice inspection practices.
DRCkoround Steam generator tubing constitutes a significant portion of the reactor coolant pressure boundary (RCPB). The design of the RCPB for structural and leakage integrity is addressed in either Title 10 of the Code of FederalRegulations, Part 50 (10 CFR Part 50), Appendix A or the licensing basis of a facility. The General Design Criteria (GDC) of Appendix A state that the RCPB siiall"have an extremely low probability of abnormal leakage"(GDC 14) "shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation" (GDC 15), and "shall be designed to permit periodic inspection and testing of important areas and features to assess their structural and leaktight integrity" (GDC 32).
Once a plant is in operation, licensees are required by their technical specifications te perform periodic !nservice inspections of the steam generator tubing and to repair or remove from service all tubes with degradation exceeding the tube repair Ilmits. Eddy-current inspection techniques are the primary means by which Ikonsees assess the condition of the steam generator tubes, Such inspections are an important component of the defense-in-depth measures to ensure the structural and leaktight integrity of the steam generator tubta.
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- j CL C7-05 December 17,1997
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The NRC issued Generic Letter (GL) 95-03,"Circumferential Cracking of Steam Generator Tubes," on April 28,1995. One of the purposes of GL 95-03 was to emphasize the importance of uHlizing quwified inspection techniques and equipment capable of reliably detecting steam generator tube degradation. ,
Cuterion IX,
- Control of Speciai Processes," contained in Appendix G to 10 CFR Part 50 states,'
in part, that " measures shall be atablished to assure that special processes, including ...
nondestructive testing, are controlled and accomp!Ished by qualified personnel using qualified p ocedures." Although the ma!n focus of GL 95-03 was to address circumferential steam generator tube cracking, the requirement of using qualified inspection techniques applies to all inspections for all forms of tube degradation.
Criterion XI, " Test Cor?ol," requires, in part, that a test program be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in r,ervice is Hentified and performed in accordance with written test procedures which incorporate the requiremcnts and acceptance limits contained in applicable design documents.
Licensees have traditionally relied upon eddf current inspection techniques to assess the condition of their steam generator tubes. Although the eddy current method is a proven technique for detecting tube degradation, the ability to depth size indications is possible only for specific modes of degradation. Specifically, tube degradation from intergranular attack (IGA) and stress corrosion cracking (SCC), major modes of steam generator tube degradation, are difficult to size with eddy-current inspection techniques because of a number of complicating variables. In one recent instance, a licensee employed a technique to size the depths of lGA tube degradation based on tube specimens removed from two plants. However, pulled tube data analyzed after the initial application of the technique indicated that the method did not adequately estimate the true depth of the indications consistent with the criteria established for qualifying the sizing techqlque.
DI:cussion (1) Evaluation of Recent inspection Experience in general, plant technical specifications require the removal from service or the repair of those steam generator tubec with degradation exceeding 40 percent of the nominal tubs wall thicknest Criterion IX in Appendix B to 10 CFR Part 50 requires that nondestructive testing be completen using qualified procedures. Therefore, licensees must be able to demonstrate through the qualification process that an inspection technique used for sizing steam generator tube indications can measure the through wall penetration of cracks and other forms of degradation with en accuracy commensurate with the "baris" of the tube opair limits in the technical specifica'..ons.
Theoretically, there is c relationship between the depth of penetration of a detect and the eddy-current signal response; in practice, however, the relationship between signal voltage or phase angle and the degradation depth is U.*1uenced by many other variabiss. Oxide deposits,
p GL 97-05
- December 17,1997 Page 3 of 6 variability of tube material properties and geometry, degradation morphology, human factors, and eddy-current data analysis and acquisition practices sa some of the factors that can significantly alter a depth estimation of steam generator tube degradation The NRC is aware that the depth of several specific forms of volumetric steam generator tube degradation can be sized with a reasonable degree of accuracy; however, qualifying techniques for sizing of some forms of degradation, e.g., IGA and SCC, has been problematic.
In order to successfully disposition steam generator tube degradation in accordance with the repair limhs in the technical specifications and Appendix B to 10 CFR Part 50, the inspection process must be capable of (1) detecting indications of tube degradation, (2) characterizing the Indications, e.g., c,acklike, IGA, manufacturing bumish mark, or wear and the orientation for cracklike degradation, and (3) accurately sizing the depth of degradation. The term " inspection process" refers to the use of one or a combination of nondestructive inspection techniques to evaluate a specific mode of steam generator tube degradation. This evaluation could potent!aily include three inspection methods (e.g., eddy-current probes)-one for detection, one for characterization,6nd a third to size the indication. However, the successful qual.ication of the inspection process requires a qualification of each method (i.e., probes, cables, software, etc.) for the mode of degradation being evaluated in the steam ')enerator tube examinations.
Experience has demonstrated that for effective qualification the data set demonstrating the capability of the inspection process should consist, to the extent practical, of service-degraded tube specimens (i.e., specimens removed from operating steam generators), supplemented, as necessary, by tube specimens containing flaws fabricated using alternative methods provided that the nondestructive examination parameter responses from these flaws are fully consistent with actualinservice degradation of the same flaw geometry.
(2) Safety Assessment Steam generator tube degradation is managed throut,h a combination of several defense in-depth measures including inservice inspection, tube repair criteria, primary to secondary leak rate monitoring, water chemistry, operator training, and analyses to ensure safety objectives are met in addition, on the basis of NRC conclusions regarding the potential consequences of steam generator tube failure events in NUREG 0844,"NRC Integrated Program for the Resolution of Unresolved Safety Issues A 3, A 4, and A 5 Regarding Steam Generator Tube IntecQ," the rhk from the potential rupture of one or more tubes is small. However, since tube ruptures represent a failure of one of the principal fission product bounda-ies and present a pathway for a release to the environment bypassing the containrNat, all reasonable precautions should be taken to prevent such at occurrence.
To verify compliance with Appendix B to 10 CFR Part 50 and the technical specifications, and to maintain a reasonable level of assurance that structural and leakage integrity margins for steam generator tubes are satisfied, the NRC has concluded that il is appropriate for the addressees to review the types of steam generator tube indications that are being left in service based on sizing, the inspection method being used to perform the sizing for each type of indication, and the technical basis for the acceptability of each inspection method.
GL 97-05 December 17,1997 1
Page 4 of 6
, [hou! red Infotmation Within 90 days of the date of this generic letter, addressees are required to submit a wntten response that includes the following information:
whether it is their practice to leave steam generator tubes with indications in (1) service based on sizing, if the response to item (1)is affirmative, those licenst es should submit a written (2) report that includes, for each type of indication, a description of the associated nondestructive examinatk,n method being used and the technical basis for the acceptability of the technique used.
Address the required written informa!!on to the U.S. Nuclear Regulatory Commission, ATTN:
Doccment Control Desk, Washington, D.C. 20555-0001, under oath or affirmation under the provisions of Section 182a, Atomic Energy Act of 1954, as c:nended, and 10 CFR 50.5 This information will en .'.ile the Commission to determine whether a license should be mo suspended or revoked in addition, submit a copy of the written information to the appro regional administrator.
NRC staff will review the responses to this generic letter and if concems are identified, aff addressees will be notified.
Backfii Discussion This generic letter only requests information from the addressees under the provision Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). This generic leHer does not constitute a backfit as defined in 10 CFR 50.109(a)(1) since impose mooifications of or additions to structures, systems or components or to design o operation uf an addressee's facility, it also does not impose a therefore, nas not performed a backfit analysis.
Reason for Information Reauest This generic letter transmits an information request pursuant to the provis%ns of Sec of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f) for the purpose of verifying compliance with applicable regulatory requirements. Specifically, the request information will enable the NRC staff to determine whether eddressees, given their steam generator tube inspection practices, are in compliance with current licensing basi respective facilities, in particular, this information will help to ascertain whether e not th regulatory requirements pursuant to Appendix B to 10 CFR Part 50, namely, Criterion
" Control of Special Processes," and Criterion XI," Test Control," are met.
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GL 97-05 !
December 17,1997 Page 5 of 6 End.eral Register Notification A notice of opportunity for public comment was published in the FederalRegister(61 FR 69118) on December 31,1996. Comments were received from one industry organ!zation and one licensee. Copies of the stJff evaluation of these comments have been made available in the public document room.
Emperwork Reduction Act Statement This generic letter contains information collections that are subject to the Paperwork Reduction :
Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved by the Office of Management and Budget, approval number 3150-0011, which expires September 30, 2000.
The public reporting burden for this collection of information is estimated to average 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-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. The U.S. Nuclear Regulatory Commission is seeking public comment on the potentialimpact of the collection of ;nformation contained in the generic letter and on the following issues: ,
- 1. Is the proposed collection of information necessary for the proper performance of
' the functions of the NRC, including whether the information will have practical utility?
- 2. Is the estimate of burden accurate?
- 3. Is there a way to enhance the quality, utility, and clarity of the information to be collected?
- 4. How can the burden of the collection of information be minimized, including the use of automated collection techniques?
Send comments on any aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch, T-6 F33, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Intemet electronic mall at bjs1@nrc. gov; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB 10202 (3150-0011), Office of Mana0ement and Budget, Washingtem, DC 20503.
The NRC may not conduct or sponsor, and a person is not rsquired to respond to, a collection of information unless it displays a currently veN OM3 control number, 9 > ,e w '-r-*---sw--= wmew -wty yT--w7-*w-- - - - " '
6 GL 97-05
- December 17, ii 37 Page 6 of 6 If you have any questions about this matter, please contact the technical contact listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager,
- k W.b Roe, Acting Director j)ision of Reactor Program Management Office of Nuclear Reactor Regulation Technical contact
- Phillip Rush, NRR 301 415 2790 E mail: pjr1@nrc. gov Lead Project ManaDer: Alexander W. Dromenck, NRR 301-415 3473 E mail: awd@nrc. gov
Attachment:
List of Recently issued NRC Generic Letters
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'* Attachment GL 97-05 December 17,1997 Page 1 of 1 LLST OF RECENTLY IS1UED GENERIC LETTERS GENERIC DATE OF LETTER SUBJECT ISSUANCE ISSUED TO 96 06, Assurance of Equipment 11/13/97 All holders of OLs for nuclear Sup. 1 Operability and Containment power reactors except those Integrity Dunag Design Basis who have permanently Accident Conditions ceased operations and have certified that fuel has been permanently removed from the reactor vessel 91 18, Information to Licensees 10/08/97 All holders of OLs for nuclear R v.1 Regarding NRC Inspection power and NPRs, including Manual Section on Resolution those power reactor of Degraded and Nonconform- licensees who have per-ing Conditions manently ceased operations, and all holders of NPR licenses whose license no longer authorizes operation 97 04 Assurance of Sufficient Net 10/07/97 All holders of OLs for nuclear Positive Suction Head for power plants, except those Emergency Core Cooling who have permanently and Containment Heat ceased operations and have Removal Pumps certified that fuel has been permanently removed from the reactor vessel 97 03 Annual Financial Surety 07/09/97 Uranium recovery licensees Update Requirements for and state officials Uranium Recovery Licensees 97-02 Revised Contents of the 05/15/97 All holders of OLs for NPRs, Monthly Operating Report except those who have permanently ceased operations and have certified that fuel has been per-manently removed from the reactor vessel OP =_ Operating License CP = Construction Permit NPR = Nuclear Power Reactors
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POLICY ISSUE I i
(Information) l D.ttcember 3.1992 SECY-97-280 f_QB: The Commissioners EdQM: L Joseph Callan Executive Director for Operations
SUBJECT:
PROPOSED NRC GENERIC 1.ETTER, " STEAM GENERATOR TUBE INSPECTION TECHNIQUES" PURPOSE:
To inform the Commission, in accordance with the guidance in a memorandum dated December 20,1991, from Samuel J. Chilk to James M. Taylor regarding SECY.91 172,
- Regulatory impact Survey Report Final," of the staffs intent to issue the attached generic letter. The purpose of the generic letter it, to (1) emphasize to licensees the importance of performing steam generator tube inservice insoections using qualified techniques in accordance with the requirements of Appendix B to 10 CFR Part 50 and (2) require certain information from licensees to determine whether they are in compliance vith the current licensing basis for their respective facilities, given their steam generator tube inservice inspection practices.
DISCUSSION:
The structaral and leakage integrity of steam generator tubing is maintained through several defense-in-depth measurer, including inservice inspection, tube repair criteria, primary-to-secondary leak rate monitoring, water chemistry, operator training, and analysec to ensure that safety objectives are met. The degraded tubes must be removed or repaired if detected indications (flaws) exceed 40 percent of the nominal tube wall thickness as required in plant technical specifications. The indications are detected by periodic inspections using qualified nondestructive testing as required by Criterion IX in Appendix B to 10 CFR Part 50. Eddy current technology, one method of nondestructive testing, is the primary means used by the industry to assess the condition of steam generator tubing. '
The oddy current inspection technique correlates the depth and length of an indication to signal responses received by probes passing through the inside of the tube. Although the eddy current method is a proven technique for detecting the longth of indications, there has been limited success in demonstrating its capability to accurately measure the depth of certain types CONTACT: Phillip Rush, NRR 415-2790 SECY NOTE:
To Be Made Publicly Available In 5 Working Days From The Date Of This Paper
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The Commissioners 3 l
of steam generator tube indications. Specifically, indications caused by intergranular attack (IGA) and stress-corTosion cracking are difficult to size with eddy current techniques because of ,
a number of complicating variables, such as oxide deposits, material properties and geometry, crcck morphology, human factors, data analysis, and data acquisition practices, in one recent instance, a licensee sized the depths of lGA indications and removed from service those tubes with IGA Indications exceeding the 40 percent through wall repair limit. Data from subsequent destructive examinations of several degraded tube specimens removed from the licensee's steam generators durint,'he outage indicated that the estimated through wall extent of degradation in these specimens, based on eddy current, was significantly less than the true depth of the IGA indications.
i To verify corupliance with Appendix B to 10 CFR Part 50 and the plant technical specifications and to maintain a reasonable level of assurance that structural and leakage integrity margins for steam generator tubes are satisfied, the generic letter requests licensees to submit a written response stating whether they leave steam generator tubes with indications in service based on >
sizing the depth of confirmed indications. If a depth sizing is used, licensees should submit a description of the asrociated eddy current method being used and the technical basis for the acceptability of the technique used.
The NRC staff is not establishing a new position in this generic letter. The generic letter only $
requests information from licensees under the provisions of Section 182a of the Atomic Energy Act of 1954, as amonded, and 10 CFR 00J'A(f).
A notice of opportunity for public comment was published in the Federa/ Register (61 FR 69118) on December 31,1996. Comments were received from one industry organization and one licensee. The comments on the proposed generic letter focused on (1) the need for the generic letter, (2) the length of the period in which licensees could respond to the generic letter, and (3) editorial commente. The staff has evaluated these concems and made appropriate ,
changes to the generic letter. Copies of the comment letters that were received are available in the Public Document Room (PDR). A copy of the staffs evaluation of the comments is available in the NRC Central Files and will be made available in the PDR after the generic letter is issued.
The Committee To Review Generic Requirements (CRGR) reviewed the proposed generic letter during Meeting Number 296 on November 19,1996, and formally endorsed the final proposed generic letter in Meeting Number 309 on August 5,1997. The staff incorporated comments made by the CRGR in those meetings.
1 The Office of the General Counsel reviewed this generic letter and has no legal objection to its contents.
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The Commissioners 3 i The staff intends to issue this generic letter 5 working days after the date of this information paper.
L. J eph Callan Exec tive Director for Operations
Attachment:
Proposed NRC Generic Letter, ' Steam Generator Tube inspection Techniques
- DISTRI!3UTION:
Commiasioners OGC OCAA OIG OPA OCA CIO CFO EDO REGIONS St.CY
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January 28,1997 Mr. David L. Meyer, Chief Itules Iteview and Directives 13 ranch U.S. Nuclear llegulatory Commission Washington, DC 20555 0001 SUHJECT: Proposed Generic Communication: Steam Generator Tube Inspection Techniques - (61 Enl. Reg. 09118 -
December 31,1990)-Remiest for Public Comment These comments are submitted on behalf of the nuclear power industry by the Nuclear Energy Institute (NEI)1 in response to the December 31,1990, Federal Register notice of opportunity for public comment concerning the proposed generic letter," Steam Generator Tube Inspection Techniques."
The stated purposes of the proposed generic letter are to:
a Emphasize to addressees the importance of performing steam generator tube in service inspection using qualified techniques; and
- llequest information from addressees to verify ifin service inspection (ISI) practices comply and conform with their respective current licensing basis.
The industry believes that these purposes can be achieved more effectively through the guidance being developed as part of the NRC industry interactions on the steam generator rule, its associated regulatory guide, and industry guidance developed in response to the rule.
The issues ofISI qualification and safety assessment discussed in the proposed generic letter are captured in the NRC draft Steam Generator Tube 8
NEl is the organizanon respons ble for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including regulatory aspects of generic operational and technicalisst.es. NE!
members include all utihties beensed to operate commercial nuclear power plants in the United States, nuclear pl. int designern. major architect /engineermg firms, fuel fabrication facilities. materials bcensees, and other organizationa and individuals involved in the nuclear energy induntry.
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- Mr. David L. Meyer January 28,1997 Page 2 Integrity regulatory guide. As a result of discussions with NRC staff on that document, the industry through EPRI is revising the PWR Steam Generator Tube Examination Guidelines to clarify criteria and considerations for the qualification ofIGI techniques.
One item yet to be resolved is the issue of supplemental performance demonstration. The industry ISI community is working with the NRC staff to better understand performance demonstration issues as presented in the draft regulatory guide More effort is required to finalize the approach.
The industry through NEI is developing guidance to address steam generator programmatic issues, including safety assessments based on NRC approved performance criteria. The industry is working closely with the NRC staff on the develJpment and application of these criteria.
We believe these issues can be better addressed after more progress is made in the steam generator rulemaking activities.
In tho event the generic letter is issued, we recommend:
. That the response period be increased to 90 days; and
- Delete the requirement for an " interim 30 day required response." The interim response places a needless burden on licensees without adding any value to the process. Licensees understand their obligation to respond to a 50.54(f) request for information within the time period allotted by the generic letter.
We appreciate the opportunity to comment on this proposed generic letter. Plc ase direct any questions on our comments to Clive Callaway at 202 739 8114.
Sincerely,
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Ralph E. Beedle E RCC/ec ..
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c: Gus Lainas, NRC F .:
Stewart L. Magruder, NRC ' M _$
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41i foyenentie Street Moll b Roteigh NC 27602 Serial: PE& RAS97-020 March 14,1997 Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, DC 20555 0001
Subject:
Columents on Two Proposed NRC Generic Letters:
Degradation of Steam Generator internals (61 FR 69116), and Steam Generator Tube Inspection Techniqaes (61 FR 69118) p
Dear Sir or Madasi:
Carolina Pow:r $ light Company (CP&L) offers the following specific comments on the proposed generic letter on degradation of steam generator internals:
v v The proposed generic letter requests the licensee to provide a ". [ discussion] of the program in place, if my, to detect degradation of steam generator internals and a description of the inspection plans, including the nspection scope, frequency, methods, equipmcnt and criteria, andplansfor co, rective action in the event degrajuion Isf(mnd."[ emphasis added] While CP&L certainly has no specific objection to providing a general description ofits inspection program, we question whether it is possible to specify a plan for corrective action prior to assessing any damage found during these inspections. CP&L recommend. that this aspect of the request be clarified.
CP&L ofTers the following specific comments on the proposed generic letter on steam generator tube inspection techn! ques:
. The requested information, as expressed in the proposed generic letter, is unclear. For example, the use of the word " defect" used in the Requested Information section of the proposed generic letter is inconsistent with the definitions in the Technical Specifications at CP&L plants. CP&L's Tc:hnical Specifications for the Robinson Nuclear Plant and the llanis Nuclear Plant define " defect" as "... an imperfection of such severity that it exceeds the plugging limit." The word " defect" should be
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e, k March 14,1997 Chief, Rules Review and Directives Branch 2 replaced with the word " indication" as used in EPRI Guideline NP 6201,"PWR Steam Generator Tube Examination Guidelines," Rev. 4.
- The proposed generic letter requests licensees to inform the NRC:
(1) Whether it is their practice to leave steam generator tubes with defects in service, based on sizing, and (2) if the response to item (1) is affirmative.
those licensees are requested to submit a written report that includes, for each type of steam generator degradation mechanism, a description of the associated nondestructive examination method being used and the technical basis for the acceptability of the technique used."
CPAC recommends for clarity that " degradation mechanism" ht replaced with " type ofindication," e.g., circumferential cracking, axial cracking, large volumetric indication (wastage), small volumetric indication (pitting), denting, primary versus secondary side indication.
If you have anyijuestions regarding these comments, please contact me at (919) 546 6901, or Mike Murdock at4919) 546 3193.
- Sincerely, Cb
- T.D. Walt Manager, Performance Evaluation
& Regulatory Affairs MLM/
cc: Mr. L. A. licyes, Regional Administrator - Regien 11 Mr. J. B. Brady, USNRC Resident inspector - IINP, Unit i Mr. B. B. Desai, USNRC Resident Inspector liBRSEP, Unit 2 Mr. N. B. Le, USNRC Project Manager - IINP, Unit 1 Ms. B. L. Mozafari, USNRC Project Manager - IIBRSEP, Unit 2 l
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ATTACHMENT 4 CRGR REVIEW PACKAGE PROPOSED ACTION: laoue a generic letter to PWR licensees (except those licenses that have been amended to possession-only status) on steam generator tube inspection techniques. Following a public comment period, the generic communicaticn will be put in fir.al form and issued for use.
CATEGORY: 2 RESPONSE TO REQUIREMENTS FOR CONTENT OF PACKAGE SUBMITTED FOR CRGR REVIEW 9
(1) The proposed genede reqdroment or staff position as it is proposed to be sont out to licensees. Where the objective or intended result of a proposed generic requirement or staff position can be achieved by setting a readily quantifiable standard that has an unambiguous relati" 'do to a readily measurable quantity and is enforceable, the proposed requirement W.Jd merely specify the objective or result to be attained, rather than prescribing to the licensee how the objective or result is to be attained.
Plant technical specifications require the removal from service or repair of those steam generator tubes with degradation exceeding 40 percent of the nominal tube wall thickness. Licensees must be able to demonstrate through the qualification process that an inspection technique used for depth sizing steam generator tube indications can measure the through wall penetratiun of cracks and other ferms of degradation with an accuracy commensurete with the " bases" of the tube repair limits in the technical specifications.
Therefore, the addressees who are leaving steam generator tubes with defects in service based on sizing are requested to submit, for each steam generator degradation mechanism, the associated nondestructive examination (NDE) method being used and the technical basis for the acceptability of each technique.
(ii) Draft staff papers or other underlying staff documents supporting the requirements or staff poaltjons. (A copy of all materials referenced in the document shall be made available upon request to the CRGR staff. Any Committee member may request CRGR staff to obtain a copy of any reference .natorial for his or her use.)
Title 10 of the Code of Federal Regulations Part 50 (10 CFR Part 50), Appendix B, Criterion IX, " Control of Special Processes," requires that measures shall be established to assure that special processes, including welding, heat treating, and nondestructive testing, are controlled and accomplished by qualified personnel using qualified orocedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements. Criterion XI, " Test Control," requires, in part, that a test program be established to assure that all testing required to demonstrate that structures, system.a. and components will perform satisfactorily in service is
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i identified and performed in accordance with written test procedures which incorporate the requirements and cceptance limits contained in applicable design documents. ;
- - On August 6,1990, the NRC staff issued Information Notice 90-48, " Stress ;
CorroWon Cracking in PWR Steam Generator Tubes.' in this information notice the ,
staff indicated that the effectiveness of oddy current testing for detecting and a! zing ;
- stress corrosion crocking could be enhanced through improved criteria for the qualification and performancs demonstration of the oddy current data acquisition equipment (including probes) and data analysis procedures. l This generic letter will enable the NRC to verify that the addressees are in compliance '
with existing regulatory requirements.
(iii) Each proposed requirement or staff poaltion shat contain the sponsoring offloe's t position as to whether ti.e proposal would increase requirements or staff poeltions, ,
implement existing requirements or staff poeltions, or wnuld relax or reduce existing requirements or staff poeltions.
The proposed generic letter requests information to be submitted so that the NRC can verify that the addressees are a compliance with existing regulatory requirements.
(iv) The proposed method of implementation with the concurrence (and any comments) of OGC on the method proposed. The concurrence of affected program offices or an !
explanation of any nonconcurrermes.
The proposed method of implementation is the development and issuance of the generic ietter.
(v) Regulatory analyses conforming to the directives and guidance of NUREG/BR-0058 and NUREG/CR 3568. (This does not apply for backfits that ensure compliance or ensure, define, or redefine adequate protection. in these cases a documented evaluation is required as discussed in IV.B.(lx).)
The proposed generic letter requests information to be subtritted so that the NRC can verify that the addressees are in compliance with axisting regulatory requirements; therefore, no value/ impact analysis was porformed.
(vil identification of the category of reactor plants to which the generic requirement or staff poaltion is to apply (that is, whether it is to apply to new plants only, new OLs only, OLs after a certain date, OLs before a certain date, all OLs, all plants under construction, all plants, all water reactors, all PWRs only, some vendor types, some vintage types auch as BWR 6 and 4, jet pump and noriet pump plants, etc.).
The g'eneric requirements will apply to all holders of oporating licenses or construction permits for , pressurized water reactors (PWRs), except those whose licenses have been amended to possesdon only status.
(vill For backfits other than compliance or adequate protection backfits, a backfit analysis as defined in 10 CFR 50.10s. The backfit analysis shall include, for each category
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of reactor Wants, an evaluaticn which demonstrates how the action should be '
priodtired and scheduled in Eght of other ongoing regulatory activities. The backftt !
analyala shau document for consideration information available concerning any of the l l fodowing factors as may be appropriate and any other information relevant and l
metodel to the propae3 action:
(a) Statement of the apoolfic etMwes that the proposed action la dealened to r i achieve; '
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(b) General deacdption of the activity that wedd be required by the licensee or applicant in order to complete the acdon; (c) Potential change in the risk to the public from the scoldental release of radosctive '
metodel; j
, (d) Potentialimpact on redological exposure of facility employees and other onalte i workers; (e) Installadon and continuing costs sanociated with the action, including the cost of i
facility downtime or the cost of construction delay; (f) The potential safety impact of changes in plant or operational complex!ty, i includng the relaticnship of proposed and existing regulatory requirements and }
- staff poaltions; (g) The estimated resource burden on the NRC associated with the proposed action >
j and the availability of resources; '
(h) The potendallmpact of dfferences in facility type, dealen, or age on the relevency
, and practicality of the proposed action; (i) Whether the proposed action is interim or final, and if interim, the justification for !
Imposig the proposed action on an interim baals; (il How the action should be prioritized and scheduled in light of other ongoing regulatory activities. The following information may be appropriate in this regard:
- 1. The proposed pdority or schedule,
- 2. A sue.imary of the current backlog of existing requirements swalting implementation,
- 3. An assessment of whether implementation of existing requirements should be deferred as a result, and
, 4. Any other information that may be conaldered appropriate with regard to i pdority, schedule, or cumulative impact. For example, could implementation be delayed pendng public comment?
This item is not applicable to the generic letter.
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(viii) For each backfit analysed pursuant to 10 CFR 50.109(e)(2) (i.e., not adequate j protection backfits and not compilance backfits), the propoolns Office Director's !
determination, together with the rationale for the determinatlon based on the conalderadon of paragraphs (l) and (vil) above, that:
1 (a) There le a substandelincrease in the overaN protection of public health and safety or the common defense and accurity to be derived from the proposal; and ;
(b) The droct and Indrect costs of implementation, for the facilities affected, are justifled in view of this increased protection.
The item is not applicable to the generic letter, (la) For adequate protection or compliance backfits evaluated pursuant to 10 CFR 50.10g(a)(4)
(a) e documented evaluation conaleting of:
(1) the objectives of the modfication (2) the reasons for the modfication (3) the baals for invoking the compliane:e or adequate protection exemption.
(b) in addtion, for actions that were immedately effective (and, therefore, lasued without prior CROR review as docussed in Ill.C), the evaluation shall document the safety significance and appropriateness of the action taken an:1(if applicable) consideration of how costs contributed to selecting the solution among various acceptable alternatives.
This item is not applicable to the generic letter.
(x) For each evaluation conducted for proposed relaxations or decreases in current requirements or staff positions, the proposing Office Director'c determination, together with the rationale for the determination based on the considerationi; or paragraphs (1) through (vil) above, that:
(a) The public health and safety and the common defense and security would be adequately protected if the proposed reduction in requirements or positions
- tere implemented, arvi (b) The cost savings attributed to the action would be substantial enough to justify taking the action.
This item is not applicable to the proposed generic letter since there is no relaxation
, or decrease in the current requirements.
(all For each request for information under 10 CFR 50.54(f) (which is not subject to exception as docussed in lil.A) en evaluation that includes at least the following elements:
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(a) A problem statement that descdbes du need for the informadon in terms of 1 potendal safety benefit.
(b) The Econsee actions required and the cost to develop a response to the information request.
(c) An anticipated schedule for NRC use of the informadon.
(d) A statement affirming that the requ6st does not impose new requirements on the licensee, other than for the requested informadon.
This part is not applicable beca'use the request for information under 10 CFP. 50.54(f) ,
that is described in this proposed generic letter is to verify compliance with existing requirements.
(alil An assessment of how the proposed acdon relates to the Commiselon's Safety Goal Policy Statement.
The NRC staff believes that the proposed generic lett3r has no impact on the Commission's Safety Goal Policy Statement since the requested information is considered necessary to verify compliance with existing regulations.
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