ML20206B287
| ML20206B287 | |
| Person / Time | |
|---|---|
| Issue date: | 04/15/1999 |
| From: | Jerome Murphy Committee To Review Generic Requirements |
| To: | Travers W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| References | |
| GL-92-01, GL-92-1, NUDOCS 9904290202 | |
| Download: ML20206B287 (110) | |
Text
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***** April 15, 1999 MEMORANDUM TO: William D. Travers Executive Director for Operations FROM: Josoph A. Murphy, Chairman -
# fr77y Committee To Review Generic ui nts
/ ///
SUBJECT:
MINUTES OF THE CRGR MEETING NUMBER 314 The Committee To Review Generic Requirements (CRGR) met on Thursday, January 30,1998, from 9:00 a.m. to 12:30 p.m. Attachment 1 contains a list of attendees. G. Laines (NRR) briefed the CRGR on requests for additional information (RAls) concoming generic letter 92-01, " Reactor Pressure Vessel integrity,* Revision 1, Supplement 1. Although, the staff had not explicitly asked for CRGR endorsement, the Committee's recommendations and support for the RAls were sought. The Committee noted that Supplement 1 had been f closed out, and the staff was seeking new and additional information which would have to be generated using specific Owners' Group data, for which not all licensees have made I commitments. Following a discussion on whether to issue another supplement to the generic letter requesting the new infonwetion being sought, the Committee recommended that the staff (i) send a letter to the NEl attaching the proposed RAls; (ii) place the package in the Public Document Room; and following industry feedback at workshop to be held on February 12,1998 (iii) retum to the CRGR with the next course of action. Attachment 2-A through 2-B contain relevant material. G. Holahan and G. Lainas, both of NRR, presented for CRGR review and endorsement the proposed generic letter on degradation of emergency core cooling system and containment spray system after a LOCA because of construction and protective coating deficiencies and foreign material in the containment. The generic letter titled " Potential for Degradation of Emergency Core Cooling System and Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Materialin }# the Containment," was revised in response to resolution of the public comments. The CRGR endorsed the generic letter with very minor comments. Attachment 3 contains details. G. Holahan (NRR) presented for CRGR review and endorsement the proposed generic letter titled
- Loss of Reactor Coolant inventory and Associated Potential for loss of Emergency Mitigetion Functions While in a Shutdown Condition." This generic letter had been revised to address public comments. The original staff proposal was reviewed at CRGR Meeting No. 291, held on September 11,1996, and was endorsed for issuance, without public comments, subject to, among others, one major comment - in addition to the 50.54(f) information request, the staff also invoke a " conditional" compliance exception to the backfit rule. Specific reference to 10 CFR 50, Appendix B, was recommended. The Committee's objective was to make it a one-step process - ask the addressees to submit the required information.80sl to require licensees to 9904290202 990415 '5A g PDR REV0P l#0CROR t [
MEETING 314 PDR 'r
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[ William D. Travers resolve the issue, where appropriate, in compliance with 10 CFR Appendix B. However, the generic letter was published for a 30 day comment period, and revised to incorporate changes recommended by CRGR, the EDO and the Commission staff. The version of the generic letter submitted for this round of CRGR review, had retumed to the original stand - a 50.54(f) Information request without invoking compliance exception. The staff indicated that the plants with design vulnerabilities will be dealt with on a case by case basis. The CRGR endorsement was contingent upon some comments which the staff agreed to address. Attachment 4 contains details.
- In accordance with the EDO's July 18,1983 directive concoming " Feedback and Closure of CRGR Review", a written response is required from the cognizant omce to report agreement or disagreement with the CRGR recommendations in these minutes. The response is to be forwarded to the CRGR Chairman and if there is disagreement with the CRGR recommendations, to the EDO for decision making.
Questions concerning these meeting minutes should be referred to Raji Tripathi (415-7584). Attachments: As stated cc: Commission (5) SECY M. Knapp, DEDE F. Miraglia, DEDO J. Lieberman, OE H. Bell, OlG ' K. Cyr, OGC J. Larkins, ACRS H. Miller, R-l L. Reyes, R-Il J. Dyer, R-Ill E. Merschoff, R-IV C. Paperiello, NMSS A. Thadani, RES S. Collins, NRR W. Kane, NRR G. Holahan, NRR D. Matthews, NRR i
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fa: 8 4 UNITED STATES Gg g g/ y { NUCLEAR REGULATORY COMMISSION / S WAsMNGToN. D.C. 3000s,em /
...,, February 5,1998 1 i
David J. Modeon Director of Engineering i i Nuclear Generation Division Nuclear Engineering institute
~ 1776 l Street, NW, Suite 400 '
Washington DC,20006-3708
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING REACTOR PRESSURE VESSEL INTEGRITY
Dear Mr. Modeen:
The NRC issued Generic Letter (GL) 92-01, Revision 1, Supplement 1 (GL 92-01, Rev.1, Supp.1), " Reactor Vessel Structural integrity" in May 1995. The purpose of this GL was to request licensees to identify, collect, and report any new data pertinent to the analysis of the structural integrity of their reactor pressure vessels (RPVs) and to assess the impact of those , I data on their RPV integrity analyses relative to the requirements of Section 50.60 of Title 10 of
/
the CM of Federal Reaulations (10 CFR Part 50.60),10 CFR 50.61, Appendices G and H to 10 CFR Part 50 (which encompass pressurized thermal shock (PTS) and upper shelf energy (USE) evaluations), and any potential impact on low temperature over pressure (LTOP) limits or I pressure-temperature (PT) limits. Alllicensees responded to the letter. Most licensees' responses indicated that additional work was required to respond to all the requests in the letter and that they would provide their final response after completion of additional generic studies. In late 1996 and ear 1y 1997, letters were issued to all licensees acknowledging the receipt of the in!tial responses to the GL and closing the TAC numbers associated with review of 7L 92-01, Rev.1, Supp.1. For most licensees, these letters indicated that a new TAC numba: would be opened for the review of their final responses when they were received. Following the issuance of these letters, an inspection of Framatome Technologies, Inc. (FTI) was performed by the NRC in May 1997 which focused on obtaining all available RPV weld chemistry data for RPVs fabricated by B&W. As a result of this inspection, additional data were identified which may affect previous RPV integrity analyses supplied by licensees with B&W fabricated RPVs. In July 1997, the Combustion En9 1neering Owners Group (CEOG) provided a report with additional RPV weld chemistry data for RPVs fabricated by CE which may affect previous RPV integrity analyses supplied by licensees with CE fabncated RPVs. As a follow-up to these letters, the CEOG report, and the NRC inspection of FTl on RPV weld chemistry, the staff plans to open new TACs on certain pressurized-water reactors (PWRs) with CE or B&W fabricated RPVs and request those licensees to provide their final responses to GL 92-01, Supp.1. Enclosed is a typical draft cover letter to a licensee and a draft RAl. Cover letters and RAls will be issued to PWR licensees with CE and B&W fabricated RPVs that have not provided an assessment of the FTl or CEOG data in response to GL 92-01, Rev.1, Supp.1. The areas covered in the enclosed draft RAI correspond to those specifed in GL 92-01, Rev.1, Supp.1. Based on our review of the generic work referenced above, the RAls identify specific issues requested to be addressed in licensee responses. These RAls along with the minutes of NGfy ff ; ~ yo. e .*.....~.,,o. ~--,...e~..-. e-. - ~ - , +. ..-- ~<n- - - . - - - >
( David J. Modeon the November 12,1997, meeting that were transmitted to you previously are intended to facilitate licensee final responses to the GL. The BWR RPV chemistry data submitted in December 1997 in BWRVIP report, " Update of Bounding Assessment of BWR/24 Reactor Pressure Vessel Integrity issues (BWRVIP 46)" is under staff review. RAls to BWR licensees may be issued in the future. The NRC will be sponsoring a Reactor Vessel Integrity Workshop on February 12 and 13,1998. The staff will discuss technical issues related to the RAI at the workshop. NEl is encouraged to share the information enclosed with all licensees prior to the workshop. This letter will be placed in the Public Document room. Your cooperation on l'hIs subject is greatly appreciated. If you have any questions please feel free to contact Barry Elliot of my staff on 415-2709. Sincerely,
. JW Gus C. Lainas, Acting Director Division of Engineering Office of Nuclear Reactor Regulation
Enclosures:
As stated cc: Kurt Cozens, NEl , i l l l 1 i 9
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CRGR Meeting No. 314 Attendance List ~ . .. 4; . ' "
> , s.: . . January 30l1998 CRGR MEMBERS NRC STAFF
- 1. T. Martin 1. G. Lainas, NRR
- 2. F. Miraglia 2. G. Holahan, NRR
- 3. W. M. Hodges for D. Ross 3. J. Shapaker, NRR
- 4. W. Kane for M. Knapp 4. A. Lee, OE
- 5. D. C. Dambly 5. G. Vissing, NRR
- 6. J. E. Dyer, RIV 6. J. Rosenthal for A. Hiser, NRR
- 7. B. Elliot, NRR
- 8. M. Mitchell, NRR CRGR STAFF 9. K. Wichman, NRR
- 10. S. Singh Bajwa, NRR
- 1. R. Tripathi, AEOD 11. R. Lobel, NRR
- 12. J. Davis, NRR
- 13. J. Tsao, NRR {
- 14. D. Serkiz, RES
- 15. M. Marshall, RES
- 16. M. Razzaque, NRR
- 17. R. Carusso, NRR
- 18. T. Collins, NRR
- 19. J. Kauffman, AEOD
- 20. J. Rosenthal, AEOD ACRS STAFF
- 1. P. Boenhert ATTACHMENT 1 n... . . . . - . . . . . - . - ..-- -
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s 5' DRAFT l ATTACHMENT 1 Mr. John Doe l Utility Addreen
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING REACTOR PRESSURE VES5EL INTEGRITY AT PLANT (8)
DearMr. Doe:
Generic Letter (GL) 92-01, Revision 1, Supplement 1 (GL 92-01, Rev.1, Supp.1), " Reactor Vessel Structural Integrity" was issued in May 1995. This GL requested licensees to perform a review of their reactor pressure vessel (RPV) structural integrity assessments in order to identify, collect, and report any new data pertinent to the analysis of the structuralintegrity of their RPVs and to assess the impact of those data on their RPV integrity analyses relative to the requirements of Section 50.60 of Title 10 of the Code of Federal Reaulations (10 CFR Part 50.60),10 CFR 50.61, Appendices G and H to 10 CFR Part 50 (which encompass pressurized , thermal shock (PTS) and upper shelf energy (USE) evaluations), and any potential impact on l low temperature overpressure (LTOP) limits or pressure-temperature (PT) limits. After reviewing your response, the NRC issued you a letter datedDATE for Plant (s). In this letter we acknowledged receipt of your response, noted that additional RPV information may become available as a result of Owners Group efforts and requested that you provide us with i the results of the Owners Groups' programs relative to your plant. We further indicated that a plant-specific TAC Number may be opened to review this material. In July 1997, the ! Combustion Engineering Owners Group (CEOG) provided a report with additional RPV weld i chemistry data for RPVs fabricated by CE. This additional RPV weld chemistry data may affect ; previous RPV integrity analyses supplied by licensees with CE fabricated RPVs. As a follow-up ! to the letter and the CEOG report, and in order to provide a complete response to items 2,3 and ! 4 of the GL, the NRC requests that you provide a response to the enclosed request for i additional information within 90 days of receipt of this letter. If a question does not apply to your situation, please indicate this in your RAI response along with your technical basis and, per GL 92-01, Rev.1, Supp.1, provide a certification that previously submitted evaluations remain valid.
]
The information provided will be used in updating the Reactor Vessel integrity Database (RVID). Also, please note that RPV integrity analyses utilizing newly identife' d data could result in the i need for license amendments in order to maintain compliance with 10 CFR Part 50.60,10 CFR 50.61 (pressurized thermal shock, PTS), and Appendices G and H to 10 CFR Part 50, and to address any potential impact on low temperature overpressure (LTOP) limits or pressure- l temperature (PT) limits. If additionallicense amendments or assessments are necessary, the attached requests that you provide a schedule for such submittals. l DRAFT
1 if you should have any questions regarding this request, please contact PROJECT MANAGER AT PHONE NUMBER. Sincerely,
. Project Manager
Enclosure:
As stated 1 l p-,_,...........,.,.,,,. .... . . . . . . . . . . . - . . . . . -
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REQUEST FOR ADDITIONAL INFORMATION REACTOR PRESSURE VESSEL INTEGRITY Section 1.0: Assessment of Best-Estimate Chemistry The staff recently received additional information that may affect the determination of the best-estimate chemistry composition for your RPV welds or your surveillance weld material. This information was provided to the NRC by the Combustion Engineerlag Owners' Group in report CE NPSD-1039, Revision 02, *Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," dated June 1997. Based on this information, in accordance with the provisions of Generic Letter 92-01, Revision 1, Supplement 1, the NRC requests the following:
- 1. An evaluation of the information in the reference above and an assessment of its applicability to the determination of the best-estimate chemistry for all of your RPV beltline welds. Based upon this reevaluation, supply the information necessary to completely fill out the data requested in Table 1 for each RPV beltline weld material.
Also provide a discussion for the copper and nickel values chosen for each weld wire heat noting what heat-specific data were included and excluded from the analysis and the analysis method chosen for determining the best estimate. If the limiting material for your vessel's PTS /PT limits evaluation is not a weld, include the information requested in Table 1 for the limiting material also. Furthermore, you should consider the infomistiori provided in Section 2.0 of this RAI on the use of surveillance data when responding. With respect to your response to this question, the staff notes that some issues regarding the evaluation of the data were discussed in a public meeting between the staff, NEl, and industry representatives on November 12,1997. A summary of this meeting is documented in a meeting summary dated November 19,1997, " Meeting Summary for November 12,1997 Meeting with Owners Group Representatives and NEl Regarding Review of Responses to Generic Letter 92-01, Revision 1, Supplement 1 Responses" (Reference 1). The information in Reference 1 may be usefulin helping you to prepare your response, in addition to the issues discussed in the referenced meeting, you should also consider what method should be used for grouping sets of chemistry data (in particular, thoto from weld qualification tests) as being from "one weld" or from multiple welds. This is an important consideration when a mean-of-the-means or coll-weighted average approach is determined to be the appropriate method for determining the best-estimate chemistry. If a weld (or welds) were fabricated as weld qualification specimens by the same manufacturer, within a short time span, using similar welding int.ut parameters, and using the same coil (or coils in the case of tandem arc welds) of weki consumables, it may be appropriate to consider all chemistry samples from that Wold (or welds) as samples from "one weld" for the purposes of best-estimate chemistry determination. If information is not available to confirm the aforementioned details, but sufficient evidence exists to reasonably assume the details are the same, the best-estimate b 4 e, m. - aA ema - ==
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2 chemistry should be evaluated both by assuming the data came from "one weld" and by assuming that the data came from an appropriate number of " multiple welds". A justification should then be provided for which assumption was chosen when the best estimate chemistry was determined. Section 2.0. Evaluation and Use of Surveillance Data The chemical composition report referenced in Section 1.0 includes updated chemistry estimates for heats of weld metal. These reports not only provide a suggested best estimate value but also !nclude the source data used in estimating the chemical composition of the heat of material. This perm:ts the determination of the best estimate chemical composition for the various sources of data including surveillance welds. Since the evaluation of surveillance data rely on both the best estimate chemical composition of the RPV weld and the surveillance weld, the information in these reports may result in the need to revise previous evaluations of RPV integrity (including LTOP setpoints and PT limits) per the requirements of 10 CFR 50.60,10 CFR 50.61, and Appendices G and H to 10 CFR Part 50. Based on this information and consistent with the provisions of Generic Letter g2-01, Revision 1, Supplement 1, the NRC requests the following:
- 2. that (1) the information listed in Table 2, Table 3, and the chemistry factor from the surveillance data be provided for each heat of material for which surveillance weld data are available and a revision in the RPV integrity analyses (i.e., current licensing basis) is needed or (2) a certification that previously submitted evaluations remain valid. Separate tables should be used for each heat of material addressed. If the limiting material for your vessel's PTS /PT limits evaluation is not a weld, include the information requested in the tables for the limiting material (if surveillance data are available for this material).
The information discussed in Section 1.0 of this RAI regarding the chemistry reports should be considered in this response along with the following questions and comments. All surveillance program results for the heats of material in a RPV should be considered in evaluating its integrity regardiess of source per 10 CFR 50.61 (" Surveillance program results means any data that demonstrates the embrittlement trends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with or without surveillance program integrated per 10 CFR 50, Appendix H.'). If any of the data provided in Table 2 are not used in the calculation of the embrittlement trend for a particular RPV weld, the technical basis for not including /using the data should be provided. , When assessing credibility of surveillance data that come from more than one source, adjustments to the surveillance data may be needed to account for differences in the chemical
. composition and irradiation environment of the different sources consistent with the requirements in 10 CFR 50.61. A method for accounting for these differences is discussed in i Reference 1.
Based on the information provided in Table 2, the credibility of the surveillance data can be evaluated. The results of these analyses including the slope of the best fit line through the
~
- - ~-
l 3 surveillance data can be provided in a format similar to that of Table 3. If the method for adjusting and/or normalizing the surveillance data when assessing credibility differ from the methods documented in Reference 1, provide the technical basis for the adjustment and/or the normalization procedure. If the chemical composition of the surveillance weld is not determined in accordance with Reference 1 (i.e., the mean of all chemistry analyses performed on the surveillance weld), provide the technical basis for the estimate. When determining the chemistry factor for a RPV weld from surveillance data, adjustments to the surveillance data may be needed to account for differences in the chemical composition and irradiation environment between the surveillance specimens and the vessel being assessed consistent with the requirements in 10 CFR 50.61. A method to account for these differences is provided in Reference 1. In addition,10 CFR 50.61(c)(2) specifies that licensees shall consider plant-specific information I (e.g., operating temperature and surveillance data) to verify that the Rb for each vessel beltline material is a bounding value. Regulatory Guide 1.9g, Revision 2 describes two methods for determining the amount of margin and the chemistry factor used in determining Rb. Position 1.1 describes the use of the Generic Tables in the Regulatory Guide. Position 2.1 describes the use of credible surveillance data. If the surveillance data are credible, the q may be reduced in half to calculate the margin term and the chemistry factor is to be determined from the best-fit line of the surveillance data. If the evaluation of the surveillance data indicate that the surveillance data set is not credible and the measured values of ART are less than the projected mean from the Tables plus the generic 2q, the chemistry factor may be calculated using either Position 1.1 or Position 2.1; however, the full margin term must be applied. The l method chosen must bound all the surveillance data to be in compliance with 10 CFR 50.61(c)(2). Based on the information provided in Table 2 along with the best estimate chemical composition i of the heat of material and the irradiation temperature of the plant whose vesselis being i assessed, the chemistry factor of the RPV weld can be determined. Note that the adjusted ARTa for a particular surveillance data point may be one value when c'etermining credibility and another value when determining the chemistry factor as a result of the different normalization procedures. If the method for adjusting and/or normalizing the surveillance data when determining the chemistry factor differs from the methods documented in Reference 1, provide the technical basis for the adjustment and/or the normalization procedure. Section 3.0: PTS /PT Limit Evaluation
- 3. If the limiting material for your plant changes or if the adjusted reference temperature for the limiting materialincreases as a result of the above evaluations, provide the revised RTm value for the limiting material in accordance with 10 CFR 50.61. In addition, if the adjusted Rbrvalue increased, provide a schedule for revising the PT and LTOP limits.
The schedule should ensure that compliance with 10 CFR 50 Appendix G is maintained.
i - 4 Reference
- 1. Memorandum from Keith R. Wichman to Edmund J. Sullivan, " Meeting Summary for November 12,1997 Meeting with Owners Group Representatives and NEl Regarding Review of Responses to Ger.oric Letter 92-01, Revision 1, Supplement 1 Responses",
dated November 19,1997. Attachments:
- 1. Table 1
- 2. Tables 2,3
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t From: RajiTripathi To: WND2.WNP4.FJM, WND1.WNP2.DCD, ARD1.ARP1.JED2, TWD2... Date: 3/10/98 5:14pm Sut$ect: / DISCUSSION ITEM CLOSED SESSION @ CRGR MTG. 317 (3/17) TO: CRGR Members Dennis Allison apprised me of the staffs taking an exception to the draft minutes of the CRGR Meeting No. 314. The endosed meg. was sent by him to the members soliciting feedback by the 10th. In the light of staffs disagrooment win the CRGR comments and recommendations, I have once again reviewed my notes from the CRGR Meeting No. 314. Here I specifically excerpt from my notes the CRGR Chairman's final re-cap of the Committee's comments and recommendations: 1 Place proposed communication in the PDR to NEl and to the licensees (send a letter to NEl that would give you a vehicle [to make it avaialble to overyone])
- 2. Solicit comments from the Workshop participants
- 3. Revise as either individual RAI or a GL supplement. or administratively open the previously closed out GL
!supplemeno
- 4. Retum to the CRGR 1 believe that the draft minutes appropriately reflect the meeting proceedings. However, I would be very interested in the members' thoughts on the matter, and would like to invite a descussion at the next mig.
Should there still be staffs disagreement with the meeting minutes, and therefore, with the CRGR recommendations, there is a due process. In accordance with the EDO's July 18,1983 directive concoming Teodback and Closure of CRGR Review,* a written response is required from the cognizant ofilce to report agreement or disagreement with the CRGR recommendations in these minutes. The response is to be forwarded to the CRGR Chairman and if there is disagreement with the CRGR recommendations, to the EDO for decision making. Raji CC: DPA
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-s From: Dennis Allison To: crge l Date: 3/6/98 7:32pm
Subject:
Request for Comment Attachvd is a request for your comments, if any, by March 16, on the requests I for additional information (RAIs) on reactor vessel integrity as currently proposed by the staff. CC: MAM4, RJE, hCS l l l l y- -- __s
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N 3/6/98 Note to: CRGR Members W. Hodges W. Kane From: Dennis Allison Subj: Requests for Additional Information (RAls) on Reactor Vessel integrity This is to request your comments, if any, by March 16,1998 on the RAls as currently proposed by the staff. The draft minutes of Meeting No. 314, copy attached, which were circulated for comment via e-mail on 2/11/98 indicate:
- 1. .... the staff was seeking new and additionalinformation which would have to be generated using specific Owners Group data, for which not all licensees have made commitments,"
- 2. .... the Committee recommended that the staff (i) .... (ii) .... and (iii) retum to the CRGR with the next course of action," and
- 3. "As proposed, the RAls are beyond the scope of the original generic letter, Supplement 1, and thus constitute backfits."
The staff believes that the CRGR was pretty well satisfied with the proposed RAls .nd merely wanted to be informed of any changes made as a result of public comments at the NRC/NEl RPV workshop in February. The staff also does not believe the CRGR considered the RAls to be backfits. The staff's thoughts are provided in the attached memorandum from Gus Lainas ! to me. The memo also indicates that only two minor changes have been made to the RAls: 1
- 1. extending the response time to 90 days, and
- 2. adding a statement to the effect that the ratio procedure described in the PTS rule and Regulatory Guide 1.99 must be used when there is clear evidence of a difference in the copper and nickel content between the surviellance weld and the vesselweld.
As also indicated in the attached memorandum, the staff wants to move forward expeditiously with these RAls. Accordingly, please let Raji Tripathi know by March 16 if you have any questions or comments or believe there should be any further discussion of the RAls so that, if the CRGR is satisfied, she can so inform the staff.
l'. 1 l March 4,1998 l MEMORANDUM TO: Dennis P. Allison, Acting Technical Assistant Committee to Review of Generic Requirements [ Original signed by:] FROM: Gus C. Lainas, Acting Director ' Division of Engineering Office of Nuclear Reactor Regulation
SUBJECT:
RESOLUTION OF ISSUES REGARDING STAFF FOLLOWUP TO GENERIC LETTER 92-01, REVISION 1, SUPPLEMENT 1 On January 30,1997, Barry Elliot of my staff and I briefed the Committee to Review of Generic Requirements (CRGR) regarding the staff's proposal to issue Requests for Additional Infcrmation (RAls) to pressurized water reactor (PWR) licensees as a followup action on Generic Letter 92-01, Revision 1, Supplement 1,
- Reactor Pressure Vessel Integrity." Based on our understanding of the recommendations provided by CRGR during the January 30 meeting, the staff has completed all actions requested by CRGR. The staff is providing this memorandum and enclosure to formally close this issue with CRGR and to request that the draft minutes from the January 30 meeting be revised before they are issued in final form to clarify the issues to which the staff was asked to respond.
The staff has developed these RAls to: (1) ensure that licensees review the appropriate Owner's Group activities and provide an updated RPV integrity evaluation to the staff in a timely fashion and in accordance with GL 92-01, Revision 1, Supplement 1 and; (2) identify issues and provide additional staff guidance to licensees to ensure that acceptable methodologies for the evaluation of chemistry and surveillance data are used in RPV integrity evaluations. CRGR agreed at the January 30 meeting that it was not necessary to issue these questions as another supplement to.GL 92-01 but that industry comment should be solicited
. prior to their issuance. Since at that time the staff was preparing for the upcoming NRC/NEl Reactor Pressure Vessel (RPV) Workshop (February 12-13,1998), CRGR suggested that the staff place an example RAI in the Public Document Room by forwarding it to NEl for distribution to licensees and that comments received at the workshop regarding the example RAI be addressed prior to their issuance. The staff issued a letter and example RAI to David J.
Modeon (NEl) on February 5,1998, (see Enclosure 1) and received two comments regarding the RAI at the RPV Workshop. These comments can be summarized as: (1) explain exactly when the ratio procedure, per Regulatory Guide 1.99, Revision 2, should be used and; (2) establish a 90 day rather than a 60 day response period. The staff concluded that these comments do not change the technical basis for the staff's request and were, in general, not significant. The staff agreed to make these changes and has revised the RAls accordingly. A redline version of the RAls reflecting these changes has been provided in Enclosure 2. CONTACT: Matthew A. Mitchell, NRR
.415-3303 6 n .-. # .... a.. . - -.~..,-, - .. . ...- -....---- .-._,. .
e . Dennis P. Allison j Contrary to the draft meeting minutes circulated for comment via e-mail, we do not believe that CRGR requested that we present this issue again for their review nor do we believe that any question regarding the need for a backfit analysis existed. The staff believes that the aforementioned actions fulfill all of CRGR's requests regarding the RAls. The staff also believes that this memorandum fulfils the staff's obligation to respond to CRGR regarding the outcome of these actions. Furthermore, based on CRGR's comments (which included considering issuing these questions as Supplement 2 to the GL) and acceptance of the proposed RAls and de staff's justification for pursuing them, it would seem that no backfit analysis was considered necessary. Please advise us as soon as possible but no later than seven days of the date of this letter of any comments so that we can expeditiously proceed and issue the RAls.
Enclosures:
As stated
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I. 4 One of the Example Letters included in the Enclosures (New paragraph on ratio procedure appears on the 3rd page of attachment, in redline) Mr. John Doe, Utility, Address
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING REACTOR PRESSURE VESSEL INTEGRITY AT PLANT (S)
Dear Mr. Doe:
Generic Letter (GL) 92-01, Revision 1, Supplement 1 (GL 92-01, Rev.1, Supp.1), " Reactor l Vessel Structural integrity" was issued in May 1995. This GL requested licensees to perform a l review of their reactor pressure vessel (RPV) structural integrity assessments in order to ! identify, collect, and report any new data pertinent to the analysis of the structural integrity of j their RPVs and to assess the impact of those data on their RPV integrity analyses relative to I the requirements of Sectior 50.60 of Title 10 of the Code of Federal Reaulations (10 CFR Part l 50.60),10 CFR 50.61, Appendices G and H to 10 CFR Part 50 (which encompass pressurized thermal shock (PTS) and upper shelf energy (USE) evaluations), and any potential impact on icw temperature overpressure (LTOP) limits or pressure-temperature (PT) limits. After reviewing your response, the NRC issued you a letter dated DATE for Plant (s). In this letter we acknowledged receipt of your response, noted that additional RPV information may become available as a result of Owners Group efforts and requested that you provide us with the results of the Owners Groups' programs relative to your plant. We further indicated that a plant-specific TAC Number may be opened to review this material. In July 1997, the Combustion Engineering Owners Group (CEOG) provided a report with additional RPV weld chemistry data for RPVs fabricated by CE. This additional RPV weld chemistry data may affect previous RPV integrity analyses supplied by licensees with CE fabricated RPVs. As a follow-up to the letter and the CEOG report, and in order to provide a complete response to items 2,3 and 4 of the GL, the NRC requests that you provide a response to the enclosed request for additionalinformation within 90 days of receipt of this letter. If a question does not ; apply to your situation, please indicate this in your RAI response along with your technical ; basis and, per GL 92-01, Rev.1, Supp.1, provide a certification that previously submitted ! evaluations remain valid. ! The information provided will be used in updating the Reactor Vessel Integrity Database (RVID). Also, please note that RPV integrity analyses utilizing newly identified data could result in the need for license amendments in order to maintain compliance with 10 CFR Part 50.60,10 CFR 50.61 (pressurized thermal shock, PTS), and Appendices G and H to 10 CFR Part 50, and to address any potential impact on low temperature overpressure (LTOP) limits or pressure-temperature (PT) limits. If additionallicense amendments or assessments are necessary, the attached requests that you provide a schedule for such submittals. if you should have any questions regarding this request, plaase contact PROJECT MANAGER AT PHONE NUMBER. Sincerely,
, Project Manager
Enclosure:
As Stated
l REQUEST FOR ADDITIONA'L INFORMATION REACTOR PRESSURE VESSEL INTEGRITY Section 1.0. Assessment of Best-Estimate Chemistry The staff recently received additumalinformation that may affect the determination of the best-estimate chemistry composition for your RPV welds or your surveillance weld material. This information was prove.d to the NRC by the Combustion Engineering Owners' Group in report CE NPSD-1039, Revision 02, "Best Estimate Copper and Nickel Values in CE i , Fabricated Reactor Vessel Welds," dated June 1997. ' l 1 Based on this information, in accordance with the provisions of Generic Letter 92-01, ! Revision 1, Supplement 1, the NRC requests the following:
- 1. An evaluation of the information in the reference above and an assessment of its I !
applicability to the determination of the best-estimate chemistry for all of your RPV beltline welds. Based upon this reevaluation, supply the information necessary to completely fill out the data requested in Table 1 for each RPV beltline weld material. Also provide a discussion for the copper and nickel values chosen for each weld wire heat noting what heat-specific data were included and excluded from the analysis and the analysis method chosen for determining the best-estimate. If the limiting material for your vessel's PTS /PT limits evaluation is not a weld, include the information requested in Table 1 for the limiting material also. Furthermore, you should consider the information provided in Section 2.0 of this RAI on the use of surveillance data when responding. With respect to your response to this question, the staff notes that some issues regarding the evaluation of the data were discussed in a public meeting between the staff, NEl, and industry representatives on November 12,1997. A summary of this meeting is documented in a meeting summary dated November 19,1997, " Meeting Summary for November 12,1997 ! Meeting with Owners Group Representatives and NEl Regarding Review of Responses to Generic Letter 92-01, Revision 1, Supplement 1 Responses" (Reference 1). The information in Reference 1 may be useful in helping you to prepare your response. ' in addition to the issues discussed in the referenced meeting, you should also consider v,tiat method should be used for grouping sets of chemistry data (in particular, those from weld
~
qualification tests) as being from 'one weld" or from multiple welds. This is an important consideration when a mean-of-the-means or coil-weighted average approach is determined to be the appropriate method for determining the best-estimate chemistry, if a weld (or welds) were fabricated as weld qualification specimens by the same manufacturer, within a short time span, using similar welding input parameters, and using the same coil (or coils in the case of - tandem arc welds) of weld consumables, it may be appropriate to consider all chemistry samples from that weld (or welds) as samples from *one weld" for the purposes of best-estimate chemistry determination. If informatior is not available to confirm the aforementioned details, but sufficient evidence exists to reasonably assume the details are the same, the best-estimate chemistry should be evaluated both by assuming the data came 1 from "one weld" and by assuming that the data came from an appropriate number of " multiple welds". A justification should then be provided for which assumption was chosen when the I best-estimate chemistry was determined. NO 4 %-% p' m e - e . g e-* 4 4 W eO M$ -*S 4W **
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Section 2.0: Evaluation and Use of Surveillance Data The chemical composition report referenced in Section 1.0 inciudes updated chemistry estimates for heats of weld metal. These reports not only provide a suggested best estimate value but also include the source. data used in estimating the chemical composition of the heat of material. This permits the determination of the best estimate chemical composition for the various sources of data including surveillance welds. Since the evaluation of surveillance data l' rely on both the best estimate chemical composition of the RPV weld and the surveillance weld, the information in these reports may result in the need to revise pr6tma evaluations of
. RPV integrity (including LTOP setpoints and PT limits) per :hs requirements of 10 CFit 50.60, ,
10 CFR 50.61, and Appendices G and H to 10 CFR Part 50 l
~
l Based on this information and consistent with the prov 4 ions of Generic Letter 92-01, ; Revision 1, Supplement 1, the NRC requests the fellaing: ) l 2. that (1) the information listed in Table 2, Table 3, and the chemistry factor from the surveillance data be provided for each heat of material for which surveillance weld data j l are available AQd a revision in the RPV integrity analyses (i.e., current licensing basis) i l is needed or (2) a certification that previously submitted evaluations remain valid. I Separate tables should be used for each heat of material addressed. If the limiting material for your vessel's PTS /PT limits evaluation is not a weld, include the information requested in the tables for the limiting material (if surveillance data are available for this material). The information discussed in Section 1.0 of this RAI regarding the chemistry reports should be considered in this response along with the following questions and comments. All surveillance program results for the heats of material in a RPV should be considered in evaluating its integrity regardless of source per 10 CFR 50.61 (" Surveillance program results means any data that demonstrates the embrittlement trends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with orwithout surveillance program integrated per 10 CFR 50, Appendix H?). If any of the data provided in Table 2 are not used in the calculation of the embrittlement trend for a particular RPV weld, the technical basis for not including /using the data should be provided, i l
- When assessing credibility of surveillance data that come from more than one source, l
adjustments to the surveillance data may be needed to account for differences in the chemical composition and irradiation environment of the different sources consistent with the ; requirements in 10 CFR 50.61, A method for accounting for these differences is discussed in Reference 1. i Based on the information provided in Table 2, the credibility of the surveillance data can be l evaluated. The results of these analyses including the slope of the best fit line through the l surveillance data can be provided in a format similar to that of Table 3. If the method for j adjusting and/or normalizing the surveillance data when assessing credibility differ from the l methods documented in Reference 1, provide the technical basis for the adjustment and/or the ; normalization procedure, if the chemical composition of the surveillance weld is not determined in accordance with Reference 1 (i.e., the mean of all chemistry analyses performed i L __ _. _. _.
l on the surveillance weld), provide the technical basis for the estimate. When determining the chemistry factor for a RPV weld from surveillance data, adjustments to the surveillance data may be needed to account for differences in the chemical composition and irradiation environment between the surveillance specimens and the vessel being assessed consistent with the requirements in 10 CFR 50.61. A method to account for these dif'erences is provided in Reference 1. In addition,10 CFR 50.61(c)(2) specifies that licensees shall consider plant-specific information (e.g., operating temperature and surveillance data) to verify that the RTNDT for each vessel beltline materialis a bounding value. Regulatory Guide 1.99, Revision 2 describes two methods for determining the amount of margin and the chemistry factor used in determining RTNDT. Position 1.1 describes the use of the Generic Tables in the Regulatory Guide. Position 2.1 describes the use of credible surveillance data, if the surveillance data are credible, the oA may be reduced in half to calculate the margin term and the chemistry factoris to be determined from the best-fit line of the surveillance data. If the evc!uation of the surveillance data indicate that the surveillance data set is not credible and the measured l values of ARTNDT are less than the projected mean from the Tables plus the generic 2aa, the chemistry factor may be calculated using either Position 1.1 or Position 2.1; however, the full margin term must be applied. The method chosen must bound all the surveillance data to be in compliance with 10 CFR 50.61(c)(2). Based on the information provided in Table 2 along with the best estimate chemical composition of the heat of material and the irradiation temperature of the plant whose vesselis being assessed, the chemistry factor of the RPV weld can be determined. Note that the ; adjusted ARTNDT for a particular surveillance data point may be one value when determining credibility and another value when determining the chemistry factor as a result of the different normalization procedures. If the method for adjusting and/or normalizing the surveillance data when determining the chemistry factor differs from the methods documented in Reference 1, provide the technical basis for the adjustment and/or the normalization procedure. In a meeting between the staff and industry representatives at the l)RC on February 12,1998, an industry representative requested a clarification as to when the ratio procedure should be used to evaluate surveillance data. The ratio procedure is described in the PTS rule and RG 1.99, Revision 2. The ratio procedure is used to adjust the measured value of ARTNDT to account for differences in the chemical composition between the surveillance weld and the vessel beltline weld. The PTS rule and RG 1.99, Revision 2 indicate that when there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e. differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the ratio procedure must be used. Section 3.0: PTS /PT Limit Evaluation
- 3. If the limiting material for your plant changes orif the adjusted reference tsmperature for the limiting material increases as a result of the above evaluations. provide the revised RTPTS value for the limiting materialin accordance with 10 CFR 50.61. In addition, if the adjusted RTNDT value increased, provide a schedule for revising the PT and LTOP limits. The schedule should ensure that compliance with 10 CFR 50 Appendix G is maintained.
. _ . . . . ~ . _ , . . . . . . _ ~ . - . .. . .___.. -
Reference
- 1. Memorandum from Ke'th r R. Wehman to Edmund J. Sullivan, " Meeting Summary for November 12,1997 Meeting with Owners Group Representatives and NEl Regarding Review of Responses to Generic Letter 92-01, Revision 1, Supplement 1 Responses",
dated November 19,1997. Attachments: 1. Table 1, 2. Tables 2, 3 I w_. _ _ -- . . . ,
I Attachment 2-D From: Raji Tripathi To: WND2.WNP6.GCL 1 Date: 3/17/98 5:07pm
Subject:
CRGR RNDORSEMENT OF THE RAls - FOLIDWUP TO GL 92-01, REV. 1, SUPPL.1
]
TO: Gus Lainas, NRR l In a Closed Session at the CRGR meeting this morning, the Committee discussed the need for further review of the subject RAls forwarded via your memorandum to Dennis P. Allison, dated March 4,1998. At this point the CRGR sees no value in reviewing the revised RAIs and has no objection to the staff proceeding with the issuance of the same. Also briefly discussed in this Closed Session was the exception taken by the staff on the draft minutes of the CRGR Meeting No. 304, as delineated in the'above referenced memorandum. The draft mts. already alluded to the obligatory post-CRGR-review staff actions, namely, issuance of letter to NEI forwarding the RAls, placement of this correspondence in the PDR, incorporation of comments of the padcipants of the February NRC-Industry Workshop, and informing CRGR of the revised course, if any. The re-submittal of the modified RAIs to the CRGR, thus, fulfills the staff's obligations. Please note that no changes will be made to the draft minutes of the CRGR Meeting No. 314, as issued for comments, with the exception that these minutes will make a reference to the Committee's brief discussion on this matter today at the CRGR Meeting No. 317, and will r.lso document the aforementioned decision by the Committee. If you have any questions regarding this transmittal, please feel free to call me at 415-7584. Raji Tripathi ec: Berry Elliot, Dennis Allison, Mat Mitchell CC: WND2.WNP4.FJM, 'ITM, DPA, WND2.WNP6.MAM4, ARD1.ARPl... ; 1 i
-a i - - - -
PROPOSED GENERIC LETTER
" POTENTIAL FOR DEGRADATION !F EMERGENCY CORE COOLING SYSTEM AND CONTAINMENT SPRAY SYSTEM AFTER A LOSS-OF-COOLANT ACCIDENT BECAUSE OF CONSTRUCTION AND PROTECTIVE COATING DEFICIENCIES AND FORElGN MATERIAL IN THE CONTAINMENT" (CRGR Meeting No. 314 - January 30,1998)
TOPIC Staff request for CRGR review and endorsement of the proposed generic letter titled, " Potential for Degradation of Emergency Core Cooling System and Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in the Containment." This revision incorporates resolution of public comments. The generic letter applies to both the boiling water reactors and the pressurized water reactors. As recommended by the Committee at the previous rounds of CRGR review, the staff has included the argument related to the 50.54(f) information request to verify compliance with 10 CFR 50, Appendix B,10 CFR 50.46 (the ECCS Rule) and 10 CFR 50.65 (the Maintenance Rule), especially, in the context of protective coatings
" HISTORIC" PERSPECTIVE The Committee had reviewed two previous "incamations" of this generic letter. An earlier version, " Potential for Degradation of Emergency Core Cooling System Recirculation Due to Construction Deficiencies and Foreign Materialin the Containment Following a Loss-of-Coolant Accident," was presented for the CRGR review and endorsement on February 25,1997 at CRGR Meeting No. 302. At that meeting' the CRGR commented that the 19 or so relevant NRC generic communications issued over the past dozen years were not fully effective, as incidents involving debris and potential degradation of the ECCS continue to be reported. Therefore, the CRGR recommended that the NRC needs to take a stronger stance on the subject, such as targeted inspections and escalated enforcement. It was agreed that the staff will re-write the generic letter.
(1) with a greater emphasis on the fact that the licensees are required to ensure the operability of the structures and components in order to comply with the provisions of the ECCS Rule (10 CFR 50.46) and the Maintenance Rule (10 CFR 50.65)'; ATTACHMENT 3
' The minutes of the CRGR meeting No. 302 (Part II) were issued on April 4,1997, 2
cognizant staff, after discussion with the o*her staff in NRR. added for CRGR presentation and the ensuing
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l l 2 (2) to make it clear that NRC regards these lapses as serious and is willing to focus necessary inspechon resources to verify licensee compliance with the existing regulations; and (3) that escalated enforcement actions may be taken when warranted. The staff had agreed to re-submit the revised generic letter reflecting the CRGR recommendations. At the meeting, the Committee agreed that the revised letter will be circulated to the CRGR members for endorsement, by negative consent. Subsequently, the CRGR staff was informed that the staff was re-drafting the generic letter by expanding the scope of the originally proposed generic letter to also include problems encountered with the use of unqualified paints and coatings in the containment. Also, the revised generic letter had retained some 50.54(f) information request feature to enable the NRC staff to verify licensee compliance with existing regulations. However, since the scope, contents, and tone of the modified generic letter has apparently changed so much since its original version, at the staffs request, the CRGR agreed to a re-review. The purpose of the revised proposed tvsr.eric letter, reviewed by the Committee on April 22, 1997 at CRGR Meeting No. 304, was to address: (1) a significant concem that foreign material capable of damaging the emergency core cooling system (ECCS) and the safety-related containment spray system (CSS) equipment or interfering with its successful operation continues to be found inside the primary containment at operating nuclear power plants; (2) continuing identification of construction deficiencies that cause the ECCS and the CSS to be outside their design bases; (3) continuing problems with the material condition of ECCS systems, structures, and components (SSCs) inside containment; (4) problems with the material condition of protective coatings inside containment; l (5) determination of the NRC's strong expectations that licensees will ensure the ECCS and i' CSS will remain capable of performing their intended safety functions and that the NRC will conduct inspections to ensure compliance with the existing licensing basis and respond to discovered inadequacies with aggressive enforcement consistent with its i enforcement policy; and (6) necessary information requested in order to evaluate addressees' programs for ensuring the operability of ECCSs and safety-related CSSs with respect to coatings inside j containment. Following the CRGR endorsement, the generic letter was published in the Federal Reaister for a 60-day comment period. The version developed after resolution of public comments is being reviewed at CRGR Meeting No. 314. l
. 1 1
3 BACKGROUND Memorandum dated December 23,1997 from F. J. Miraglia to T. T. Martin, requesting CRGR review and endorsement of the proposed generic letter titled, " Potential for Degradation of Emergency Core Cooling System and Containment Spray System after a Loss-of-Coolant ! Accident Because of Construction and Protective Coating Deficiencies and Foreign Materialin l the Containment." The attachments included the following.
- 1. Proposed Generic Letter ;
- 2. Original CRGR Charter Review Package
- 3. Public Comment Resolution Attachment 3-A contains the presentation material used by the staff.
ISSUES / CONCERNS AND RECOMMENDATIONS The revised generic letter was endorsed by the CRGR with only minor editorial comments. Attachment 3-B contains the endorsed version. , BACKFIT CONSIDERATIONS i in this proposed generic letter, the information request to verify compliance is being made under the provisions of 10 CFR 50.54(f). l l 4 '*= . - - -
CKCF- IIT- '., I3s]98
- Arra md 2 4 NRC STAFF PRESENTATION TO THE CRGR MEETING NO. 314
SUBJECT:
PROPOSED GENERIC LETTER POTENTIAL FOR DEGRADATION OF THE EMERGENCY CORE COOLING AND CONTAINMENT SPRAY SYSTEMS BECAUSE OF CONSTRUCTION AND PROTECTIVE COATINGS DEFICIENCIES AND FOREIGN MATERIAL IN CONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT DATE: JANUARY 27,1998 SPONSORING DIVISIONS: DE AND DSSA PRESENTER: EDMUMD J. SULLIVAN JAMES A. DAVIS PRESENTER'S TITLE / BRANCH / DIVISION: MATERIALS ENGINEER NRR/DE PRESENTER'S NRC TELEPHONE NUMBER (301)415-2713 1
(. HISTORY OF THE GL e Original Letter to CRGR - April 1,1997 e Presentation to CRGR - April 22,1997 e GL issued for Public Comment on May 13,1997 e Briefed ACRS on the GL - June 11,1997 e Comment Period Closes - June 27,1997 e Letter to CRGR on GL Addressing Public Comments - December 23,1997 l i l 2 p .. . . . . - . . . . . . . . . . . , , _ . . . _ - _ . . - - . . . . . . - _ _ . . . . . .
q f t L OUTLINE e Purpose of the GL e Purpose of Containment Coatings e Coating Categories e Coating Definitions " e Applicable Regulations and Guidance e Recent Coatings Events e Requested Information e Response to Significant Public Comments on Draft GL e Conclusions e Attachment A - Public Comments Page NEl Comments A-1 BG&E Comments A-3 PECO Comments A-4 TVA Comments A-5 Entergy Comments A-6 I 3
PURPOSE OF THE GENERIC LETTER RELATIVE TO COATINGS e To Notify Addresses About Problems with the Material Condition of Protective Coatings inside Containment e To Request Information Necessary to Verify l Compliance with the Addressees' Programs for Ensuring the Operability of ECCSs and Safety-Related SSCs with Respect to Coatings inside Containment
- To inform Licensees that Enforcement Action will be j Pursued if Debris is Found in Containment 4
_ _ _ _ _ _ _ _ _ _ _ _ = _ _ _ _- _ _ _ _ _- . _ _ ____ = =
y PURPOSE OF CONTAINMENT COATINGS eProtect Steel or Aluminum Surfaces Against Corrosion eProtect Against Wear of Metallic, Concrete, or Masonry Surfaces During Plant Operation eEase of Decontamination From Containment Surfaces 6 h 5 I l 1
COATING CATEGORIES
- Service Level 1 - Coating on Exposed Surfaces Within Containment e Service Level 2 - Coating on any Exposed Surface Outside of Containment but Subject to Radiation and Decontamination i l
e Service Level 3 - Coating Outside of Containment Whose Failure Could Adversely Affect Normal Plant Operation or Orderly and Safe Plant Shutdown l l 9 6
p, - , c COATING DEFINITIONS , e Safety Related Coating Systems-Used inside or Outside of Containment-Detachment could Adversely ! Affect the Safety Function of Safety-Related Structures, Systems, and Components. l l e Qualified - Used inside of Containment, Tested for l Critical Characteristics including DBA Conditions, and l Applied Using Qualified Applicators and Procedures. ; e indeterminate - Safety-Related Coating System with < Reasonable Doubt about Suitability for intended Use. e Unqualified - Not Tested - Assumed to Fail During DBA
/
7
l APPLICABLE REGULATIONS AND GUIDANCE l 1 e 10 CFR Part 50, Appendix B - QA Requirements e 10 CFR Part 50.65 - The Maintenance Rule e 10 CFR Part 50.46 - Acceptance Criteria for ECCS for l Light Water Nuclear Power Reactors l l l e Regulatory Guide 1.54 " Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants" l l l 1 8
COATING EVENTS e Zion Unit 2, January 1997, About 100 Pounds of Degraded, Qualified Coating Removed. Much of Unqualified Coating Removed. Random Adhesion Tests Conducted. e Indian Point, Unit 2, March 1995, Paint Peeling Off of the Floor at the 46 Foot Floor Elevation - Improper . Application e Sequoyah Units 1 & 2, October 1993, RCP Coatings Not identified in Uncontrolled Coatings Log. Uncontrolled Coatings Limit Exceeded -Required Addition of" Catch" Screens Around RCPs e Susquehanna Units 1 & 2, July 1993 and Sept 1995, Reassessment of ECCS Performance During DB LOCA Did not Consider Unqualified Coatings e Clinton, Coating Degradation at Repair Locations 9 5 ma.~ m a e, = .m am ---=a * " ^ * ** " "^
[. . REQUESTED INFORMATION
- Requests Information on Protective Coatings under 10 CFR 50.54(f) to Evaluate Programs to Ensure Coatings do not Detach During a DB LOCA o Description of Licensing Basis and how Plant Specific Procedures meet Appendix B o How Qualified and Unqualified Coatings are Tracked o Date and Findings of Last Assessment of Coatings and Date of Next Planned Assessment o Applicable Standards if Commercial-Grade ,
Dedication Program used l o Limit for the Amount (e.g., Square Footage) of ' Unqualified Coatings and how this Limit is Determined (Model for Transporting to Screen or Sump) o if no Commercial-Grade Dedication Program: Regulatory / Safety Basis for not Controlling these Coatings and Explanation of why such a Program not Required 10
o Applicable Standards for' Qualifying Coatings l l l l l l
}
I I e 11 l . - - . . . . . . _ . _ . . . . . . _ . _ - . _ _ . . . . . . . . . . . . . . . _ _
RESPONSE TO SIGNIFICANT PUBLIC COMMENTS ON DRAFT GL e PECO - Older Plants may not be Required to Account i for Unqualified Coatings. The Accounting and Analysis Would be a Backfit Under 10 CFR 50.109 e NRC Staff Response - Licensees must Satisfy Existing Licensing Bases. Licensees must Ensure . Proper Functioning of ECCS and Safety-Related Containment Spray Systems During a DBA e PECO - Coatings Outside of Containment Covered by
-the Maintenance Rule Should be Specifically ,
Excluded from the Scope of the GL e NRC Staff Response - The GL has been Modified to make it Clear that Responses to the GL Apply only to Service Level 1 Coatings inside Containment. e Entergy - NRC Should Recognize that Coatings may ; not_ be included Under the Maintenance Rule if they do not fall into one of the Categories Specified in 10 CFR 50.65.
- NRC Staff Response - The NRC Staff Position is the Service Level 1 Coatings fall Under the Maintenance Rule.
12
I l CONCLUSIONS e The Purpose of the GL is to Find out What Coatings ; Programs are Actually Being Used by industry and to Discuss the use of Enforcement Action if Debris is Found in Containment e No Major Technical Public Comments on the GL e Most Significant Were on Older Plants not Required to l log Unqualified Coatings, Coatings Outside of Containment Covered by the Maintenance Rule, and Whether or not Coatings Should be Covered by the Maintenance Rule l i 13
E. . . l l i~ APPENDIX A ; RESPONSE TO PUBLIC COMMENTS l l ON DRAFT GL NEl j
- 1. Separate the GL into a GL on Coatings and a IN on Foreign Material Inside Containment. l l
Response: The two Documents were Separate and ) were Combined at the Request of CRGR. l
- 2. There is Typically no " Single Coatings Program." The GL Should be Modified to Recognize Multiple Plant- l Specific Activities Which Address Protective Coatings.
Response: The GL was modified, Requesting Information on Coating Program or Programs.
- 3. Instances of Localized Coating Failures does not mean Complete Failure of the Coating System and does not Indicate that the Containment Recirculation Systems were Challenged.
l Response: The staff Agrees that Localized Coating Failures do not mean Complete Failure of the Coating ) System. However, it is the Obligation of the Licensee to 1 A-1 l 1
t Establish that CRS are not Challenged.
- 4. The NRC Staff must Exercise Judgment when Evaluating the Acceptability of any Given Licensee's Coating Activities Relative to the Industry Standards. l Response: The NRC Staff will take Appropriate Actions l in Keeping with NRC Requirements and Licensee Commitments.
- 5. The GL should use " Service Level 1,2, and 3" rather than " Class I and 11."
Response: The Staff Agrees and the Changes have been made. '
- 6. Does the Requested Information Cover only Coatings inside Containment or all Safety-Related Coatings.
Response: Only inside Primary Containment.
- 7. Add the Sentence: "Once in Contact with the Sump Screens or Suction Strainers, Coating Chips may impact the net Positive Suction Head Available to the ECCS/ CSS pump."
Response: This Sentence has been Added to the GL. A-2
i .
- 8. In Appendix C, Reference is made to Latex and Polyurethane Coatings for the Drywell and Wetwell of BWRs and Containments of PWRs. Neither of these Materials have been DBA Qualified or Radiation Qualified.
To Avoid inadvertently Encouraging their use, the GL Should State that they are not Qualified. ' Response: References to Latex and Polyurethane have been Deleted from the GL. BG&E
- 1. Proposed GL Discusses Class I and ll. Industry uses Service Levels 1,2, and 3. -
Response: The GL has 'oeen revised.
- 2. Not Industry Practice to Coat Aluminum or Galvanized.
Response: RG-1.54 and Some ASTM Standards mention Aluminum and Galvanized Surfaces. Reference to Aluminum and Galvanized Surfaces retained.
- 3. Coatings are not Applied inside containment to Protect from Erosion and Wear.
6 A-3 i bu.
i Response: Coatings are useri ..; mvent Wear. The word Erosion has been Deleted. ; PECO l
- 1. Wait for the EPRI Guidance Document to be issued. l l
Response: The Staff met with NEl/EPRI in April, July, ' September, and November of 1997 to Review the Status of the Guidance. EPRI held the Final Meeting in January of 1998 and will issue the Final Guidelines in the near Future. The GL was modified, Stating that Industry may Find the Document Useful Wherc Evaluating Coatings. ; Consideration may be Given to Endorse the EPRI Guidance in the Revised RG 1.54. !
- 2. Older Plants may not be Required to Account for Unqualified Coatings. The Accounting and Analysis Would be a Backfit Under 10 CFR 50.109.
Response: Licensees Must Satisfy Existing Licensing Bases. Licensees must Ensure Proper Functioning of ECCS and Safety-Related Containment Spray Systems During DBA. If Licensee can Demonstrate Assurance < without Quantifying the Amount of Unqualified Coating, this is Acceptable. This Discussion was Added to the GL. I
- 3. Installation of Large Capacity Strainers will have A-4
Significant Impact on the Limit on Unqualified Coatings and may Affect the Strategy Regarding Coating Maintenance. Response: The Staff Agrees. However, BWROG Guidance Specifies a Nominal Amount of Failed Paint. It is the Licensee's Responsibility to Ensure that this Number is Applicable and Bounding. If not, the Bounding Amount must be Determined. This Comment does not require a Change to the GL. 4. Coatings Outside of Containment Covered by the Maintenance Rule Should be Specifically Excluded form the Scope of the GL. Response: The GL has been Modified to Make it Clear that Responses to the GL Apply only to Service Level 1 Coatings inside Containment. l TVA i
- 1. More time is Required to Respond to the GL-i.e. go ;
from 45 to 120 days. Response: The GL was Changed to Give 120 Days.
- 2. The Proposed GL Pertains only to Primary Containment.
A-5
Response: This Change has been made.
- 3. The GL Should Acknowledge the Potential for Limited Loss of Adherence for Qualified Coatings. Loss of Adherence is identified and Mitigated Through inspections and Testing to Ensure Quantities are not Large Enough to Affect Safety-Related System Operation.
Response: The Staff Had Additional Correspondence 1 with TVA to Clarify the Intent of Question 3. TVA Intended j to say that Localized Areas may Lose Adherence Over Time, Resulting in Loose or Peeling Coating. These Localized Areas do not Indicate that the Entire Coating System is Defective. The Staff Responded that Licensees Should Evaluate Their Coating System to Determine if the Loss of Adherence is Localized or Broad-Based. The GL has been Modified Accordingly. Entergy
- 1. Separate the first two issues from the Coatings issue in Order to Provide Proper Focus to each.
Response: They will remain together. A-6
/
~
l
- 2. Add the Sentence: "Once in Contact with the Sump i Screens or Suction Strainers, Coating Chips may impact the net Positive Suction Head Available to the ECCS/ CSS pump."
Response: See NEl 7. This sentence was added. :
- 3. NRC Should Recognize that coatings may not be included Under the Maintenance Rule if they do not fall into one of the Categories specified in 10 CFR 50.65.
Response: The NRC Staff Position is that Service Level I Coatings fall under the Maintenance Rule.
- 4. The GL Implies that Violations and Their Associated Severity Levels are set and do not follow the Criteria in NUREG-1600 (Enforcement Policy).
i Response: The GL has been Modified to Clarify that the Staff does Follow NUREG 1600.
- 5. There is a long History of ECCS Problems Caused by Debris and Foreign Material; There does not Appear to be a long History of Problems with Coatings.
Response: There have been Recent incidents: Zion, Clinton, etc. There is increased Awareness by the NRC staff, EPRI, ASTM, and NEl. A-7
[;
- 6. Request for Longer time to Respond,120 days.
Response: The GL Extended the Response Time to 120 days.
- 7. The Expectations of" Maintenance Activities" Should be Specified in the GL to Avoid Confusion.
Response: The Following Sentence was Added:
" Maintenance Activities involve Reworking Degraded Coatings, Remwing Degraded Coatings to Sound Coatings, Preparing the Surfaces, Applying new Coatings, and Verifying the Coatings' Quality.
- 8. The definition of Service Level 1 Coatings is inconsistent with Industry Interpretation.
Response: This has been Corrected.
- 9. Footnote 1 of Appendix C Discusses Coatings on Small Components such as Small Lighting Fixtures, Non-Safety-Related Power Buses. The Presence of a Coating on a Component does Alter the Safety Classification of the Component. Safety-Related Components can be Coated with Unqualified Coatings and Non-Safety-Related Components can be Coated with Qualified Coatings. It is A-8
{. ] 4 Industry's Understanding that Small Components Should be Exempted Regardless of the Safety Classification of the Component. Response: The Paragraph that references the Footnote was Deleted and the Footnote is no Longer Needed. 1 l l l A-9
ATTACHMENT 3-B E-MAILS RELAYING THE CRGR ENDORSEMENT, RELAYING THE CRGR ACCEPTANCE OF THE CHANGES MADE, WITH COMMENTS, AND THE ENDORSED VERSION OF THE GENERIC LETTER l
From: Raji Tripathi To: WND2.WNP6.GMH, WND2.WNP6.GCL Date: 2/5/98 2:54pm
Subject:
CRGR ENDORSEMENT OF THE " GUNK-IN-THE-SUMP" GL TO: Gary Holahan and Gus Lainas The GL on paints / coatings and other foreign materialin the containment was reviewed by the CRGR at the Committee's 314th meeting, held on January 30,1998. The Committee made very minor comments and endorsed the GL for issuance. This e-mail formally relays the Committee's endorsement. RajiTripathi cc: CRGR Members and Jim Shepaker I CC: WND2.WNP6.RML, WND2.WNP6.JAD,' WND1.WNP2.DCD, WND2.... l i 9 e<awg,,.go,, , ,e *:
l From: RajiTripathi To: WND2.WNP4.FJM, WND1.WNP2.DCD, ARD1.ARP1.JED2, TWD2... , Date: 4/1/98 2:5 Bpm :
Subject:
GUNK-IN-THE-SUMP GL - MINOR REVISION ! TO: CRGR members The subject GL was reviewed and endorsed by the CRGR at CRGR meeting No. 314, with very minor comments. Since then, the staff has modified the " Required information" section at the advice of OGC. The CRGR Chairman has approved the change. For the members' information, the affected paragraph and its new version are enclosed. The CRGR meeting minutes, to be issued shortly, will include the modified text. i h* r p 4 6ee,
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_ t rii. .e_ _ _ er,.,.... uIW. ,_ _ L w - W _.~ _ _ A, ,w ,w.,w W _._. L.._.r_._._._._.__._.. . 1 Revised text at OGC advice and accepted by the CRGR with changes, I as noted: (2) Information demonstrating compliance with either: (1) your plant-specific licensing basis for tracking the amount of unqualified coatings inside the containment and for assessing the impact of potential coating debris on the operation of safety-related SSCs during a postulated DB LOCAtThe submitted information_should_ include;
1 1 (a) The date and findings of the last assessment of coatings, and the planned date of the next assessment of coatings. (b) The limit for the amount of unqualified protective coatings allowed in the containment and how this limit is determined. Discuss any conservatism in the method used to determine this limit. (c) If a commercial-grade dedication program is being used at your facility for dedicating commercial-grade coatings for Service Level 1 applications inside the containment, describe why the program adequately qualifies such a coating for Service Level 1 service. Identify what standards or other guidance currently are being used to dedicate containment coatings at your facility; or, (ii) for plants without such licensing basis requirements, the requirements of 10 CFR 50.46b(5), "Long-term cooling" and the functioning of the safety-related CSS as set forth in your licensing basis. If a licensee can demonstrate this assurance without quantifying the amount of unqualified coating, this is acceptable l. (a) lf-e commercial-grade coatings are being used at your facility for Service Level 1 applications, and such coatings are not dedicated or controlled under your Appenetodix B QA Program, provide the regulatory and safety basis for not controlling these coatings in accordance with such a 4 p'Mram. Additionally, explain why the facility's licensing basis does not require such a program. Raji cc: G. Holahan, G. Lainas, J. Shapaker CC: JWS, WND2.WNP6.GMH, WND2.WNP6.GCL
r .. . . UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 l Month XX,1998 NRC GENERIC LETTER 97-XX: POTENTIAL FOR DEGRADATION OF THE EMERGENCY CORE COOLING SYSTEM AND THE CONTAINMENT SPRAY SYSTEM AFTER A LOSS-OF-COOLANT ACCIDENT BECAUSE OF CONSTRUCTION AND PROTECTIVE COATING DEFICIENCIES AND FOREIGN l MATERIAL IN THE CONTAINMENT I l Addressees All holders of operating licenses for nuclear power reactors, except those who have i permanently ceased operations and have certified that fuel has been permanently removed from ! the reactor vessel. Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter for several reasons. It alerts addressees that foreign material continues to be found inside operating nuclear power plant containments. During a design basis loss-of-coolant accident (DB LOCA), this foreign material could block an emergency core cooling system (ECCS) or safety-related containment spray system (CSS) flow path or damage ECCS or safety-related CSS equipment.
- In addition, construction deficiencies and problems with the material condition of ECCS systems, structures, and components (SSCs) inside the containment continue to be found. Design deficiencies also have been found which could degrade the ECCS or safety-related CSS. No action or information is requested regarding these issues. The NRC has issued many previous generic communications on this subject, as discussed later in this generic letter, and expects addressees to have considered possible actions at their facilities to address these concems.
The NRC expects addressees to ensure that the ECCS and the safety-related CSS remain capable of performing their intended safety functions. The NRC will conduct inspections to ensure compliance with existing licensing bases and respond to discovered inadequacies with aggressive enforcement consistent with the NRC Enforcement Policy. The NRC is also issuing this generic letter to alert the addressees to the problems associated
. with the material condition of Service Level 1 protective coatings inside the containment and to request information under 10 CFR 50.54(f) to evaluate the addressees' programs for ensuring that Service Level 1 protective coatings inside containment do not detach from their substrate
l Generic Letter 97-xx Month, Date, Year Page 2 of 10 during a DB LOCA and interfere with the operation of the ECCS and the safety-related CSS. The NRC intends to use this information to assess whether current regulatory requirements are being correctly implemented and whether they should be revised. Backaround Foreign Material Exclusion, Construction Deficiencies, and Design Deficiencies in some recent events discussed in Appendix A to this generic letter, foreign material which could have affected the operation of the ECCS was discovered inside the containment. As part . of its review of these events, the NRC staff reviewed the history of such events and identified several related problems. ! A more complete list of the related events is provided in Appendix B. As discussed in Appendix A, almost all of these events have been the subject of previous NRC generic communications and licensee event reports (LERs). The following types of problems continue to occur-(1) Foreign material has been found in areas of the containment where it could be transported to the sump (s) or the suppression pool and potentially affect the operation of the ECCS or safety-related CSS. Such material has also been found in PWR sumps, in BWR suppression pools and downcomers, and in safety-related pumps and piping. (2) Deficiencies have been found in the construction of the ECCS cumps or strainers. These deficiencies, which could have impaired the operation of the ECCS or the safety- I related CSS, include missing screens, unintended openings in screens, and incorrectly sized screens. l l (3) Problems have also been found with the material condition of sumps or suction strainers. These problems, potentially impairing the operation of the ECCS or safety-related CSS, include deformed suction strainers and unintentional flow paths created by missing grout. (4) Design deficiencies have been found, including flow-line valves with clearances smaller than the sump screen mesh size and strainers with a flow area smaller than required. (5) There have been two incidents, described in LERs, in which doors to emergency sump structures were left open when ECCS and safety-related CSS operability was required by the technical specifications. The Discussion section of this generic letter discusses the regulatory and safety basis for these concems. It is evident that past NRC generic communications have not been completely effective in controlling these problems. Nevertheless, the NRC expects that licensees will ensure that the ECCS and safety-related CSS remain capable of performing their intended safety functions. n .-. .. n. -
r-Generic Letter 97-xx Month, Date, Year Page 3 of 10 The NRC plans to further emphasize this issue by conducting inspections to ensure compliance with existing plant licensing bases. The NRC staff will respond to discovered inadequacies with aggressive enforcement consistent with the NRC Enforcement Policy. Protective Coatings Protective coatings inside nuclear power plant containments serve three general purposes. Protective coatings are applied to carbon and low alloy steel and, less commonly, to aluminum and galvanized surfaces to control corrosion. (Although aluminum and galvanized surfaces are , not commonly coated, nothing in NRC rules and regulations or industry standards prevents ' these surfaces from being coated.) Protective coatings are applied to surfaces to control radioactive contamination levels. Protective coatings are also applied to protect surfaces from i wear. l Protective coatings inside the containment and the regulatory requirements and guidance for their use are discussed in Appendix C. Qualified protective coatings are capable of adhering to their substrate during a DB LOCA in order to minimize the amount of material which can reach the emergency sump screens or suctim strainers and clog them. Not all coatings inside the containment are qualified. The amount of unqualified coatings must be controlled since the unqualified coatings are assumed to detach from their substrates during a DB LOCA or steam line break and may be transported to the emergency sump screens or suction strainers. Once in contact with sump screens or suction strainers, coating chips may adversely impact the net positive suction head (NPSH) available to the ECCS or CSS pump. ! In some cases, coatings which should have been qualified failed during normal operation. Some , of these events are discussed in Appendix D. I Discussion NRC regulations in 10 CFR 50.46 require that licensees design their ECCS to provide long-term cooling capability so that the core temperature can be maintained at an acceptably low value and decay heat can be removed for the extended period required by the long-lived radioactivity remaining in the core. This must be demonstrated while assuming the most conservative single ! failure. Some addressees may credit CSSs in the licensing basis for radioms tive-source-term l and pressure reduction. These CSSs may also take suction from the suppression pools or i emergency sumps. Foreign materials, degraded coatings inside the containment that detach from their substrate, and ECCS components not consistent with their design basis, along with LOCA-generated i debris, are potential common-cause failure mechanisms which may clog suction strainers, sump screens, filters, nozzles, and small-clearance flow paths in the ECCS and safety-related CSS and thereby interfere with the long-term cooling function.
Generic Letter 97-xx Month, Date, Year Page 4 of 10 Qualified coatings used inside containment should be capable of withstanding the environmental conditions of a postulated DB LOCA without detaching from their substrates (detached coatings may be transported to the sumps or strainers and cause or contribute to flow blockage). Some small, localized areas of degradation of the coatings may not be indicative of widespread failure of the coatings. However, the condition of the coatings should be evaluated by suitable means. Although not endorsed by the staff, the Electric Power Research Institute (EPRI) is currently preparing a guidance document for containment coatings, " Guidelines on the Elements of a Nuclear Safety-Related Coatings Program." Licensees may find this document useful when , evaluating coatings. The LERs and NRC inspection reports described in Aopendix D of this I generic letter provide evidence of weaknesses in addressee programs with regard to applications of protective coatings for Service Level 1 (see definitions of Service Levels in Appendix C). These weaknesses include deficiencies in addressee programs to (1) control the preparation and cleanliness of the substrate before the coatings are applied, (2) control the preparation of paint before its application, (3) control the dry film thickness of coatings applied to the substrate, (4) monitor for, and control the use of, excessive amounts of unqualified coatings inside the containment, (5) monitor the status of " qualified" coatings already applied to the surftcces of the containment structure and to other equipment inside the containment, and (6) assess the safety significance of coatings inside containment that have been determined to detach from their substrate and to repair these coatings, if necessary. j The NRC has issued a number of generic communications on various aspects, including problems with protective coatings, or the potential for the loss of the ECCS and safety-related CSS as a result of strainer clogging and debris blockage. These generic communications are listed in Appendix E. The basic safety concem applies to both PWRs and BWRs. These events, discussed in these generic communications, as well as similar events described in LERs i and NRC inspection reports, demonstrate the need for a strong foreign material exclusion (FME) program in all areas of PWRs and BWRs that may contain materials that could interfere with the successful operation of the ECCS. Other events demonstrate the need to ensure the correct design and to maintain the material condition of emergency core cooling system and safety-related containment spray system SSCs, including the suppression pools, ECCS strainers and sumps, and the protective coatings inside containment. The requirements of 10 CFR Part 50, Appendix B, are germane to this issue. Maintenance rule,10 CFR 50.65, " Requirements for monitoring the effectiveness of maintenance at nuclear power plants," includes in its scope all safety-related SSCs and those non-safety-related SSCs that fall into the following categories: (1) those that are relied upon to mitigate accidents or transients or are used in plant emergency operating procedures; (2) those whose failure could prevent safety-related SSCs from fulfilling their safety-related function; and (3) those whose failure could cause a reactor scram or an actuation of a safety-related system. The PWR sumps and BWR strainers are included within the scope of the maintenance rule. To the extent that protective coatings meet these scoping criteria, they are within the scope of the maintenance rule. The maintenance rule requires that licensees monitor the effectiveness of
I Generic Letter 97-xx Month, Date, Year Page 5 of 10 maintenance for these protective coatings (as discrete systems or components or as part of any SSC) in accordance with paragraph (a)(1) or (a)(2) of 10 CFR 50.65, as appropriate. Although this generic letter concems coatings within the containment and requests information about coatings within containment, addressees should ensure that all coatings which meet the maintenance rule scoping criteria are included in the programs and procedures for implementing the maintenance rule. The NRC expects all addressees to have programs and procedures in place to ensure that the ECCS and the safety-related CSS are not degraded by foreign material in the containment, that the ECCS and the safety-related CSS are consistent with their design and licensing bases, and that sumps, strainers, and coatings are in good material condition. The staff may evaluate the condition of sumps, strainers and protective coatings during routine inspections as well as specialinspections. The NRC has conducted numerous inspections in the areas addressed by this generic letter; for ' example, the NRC issued Temporary Instruction 2515/125, " Foreign Material Exclusion Controls," on August 25,1994. Violations of this instruction have been identified and appropriate enforcement action has been taken in accordance with the NRC's Enforcement Policy (NUREG-1600, " General Statement of Policy a'nd Procedures for NRC Enforcement Actions: Enforcement Policy"). A list of significant enforcement actions is provided in Appendix F of this generic letter. The NRC will continue to conduct inspections in order to ensure _ compliance with the existing licensing bases and to respond to discovered inadequacies with I aggressive enforcement consistent with the NRC Enforcement Policy. I I The NRC may consider violations in this area as significant regulatory failures and will, accordingly, consider categorizing inadequacies as Severity Level lil violations. The NRC will also consider the long history of generic communications on the issues addressed by this generic letter as prior notice to licensees when the agency assesses civil penalties in accordance with Section VI.B.2 of the Enforcement Policy. Finally, notwithstanding the normal civil penalty assessment, the NRC will consider whether the circumstances of the case warrant escalation of enforcement sanctions in accordance with Section Vll.A.1 of the Enforcement Policy. If in the course of assessing the effectiveness of the plant-specific FME program or preparing a response to the requested information an addressee determines that its facility is not in compliance with the Commission's requirements, the addressee is expected to take whatever actions are deemed appropriate in accordance with requirements stated in Appendix B to 10 CFR 50 and as required by the plant technical specifications to restore the facility to compliance. Reauired Information Within 120 days of the date of this generic lettsr, addressees are required to submit a written ! response that includes the following information: { I
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Generic Letter 97-xx Month, Date, Year I Page 7 of 10 (1) your plant-specific heensing basis for tracidng the amount of unqualified coatings inside the containment and.for assessing the impact of potential coating debris on the operation of safety 4 elated. SSCs_during a postulated DB_LOCA6The submitted information.should include: (a) The date and findings of the last assessment of coatings, and the planned date of the next assessment of contings_. (b) The. limit for the amount of unqualified guave coatings. allowed in the containment and how this limit is_ determined. Discuss.any conservatism in the method used to determine this limit.j (c) if a commercial-grade _ dedication program. is being used at your l facility for dedicating commercial grade coatings for Service. Level i 1 applications inside the containment, describe why the program j adequately qualifies such a coating for Service Level .1 service l Identify what standards _or other guidance currently are being used to dedicate containment costings at your_ facility; ord (ii)j for plants _without such licensing basis._ requirements, the requiremen's of 10 CFR 50.46b(5),' "Long-term cooling" and the functioning of the, safety-related CSS as set forth in your licensing basis. Elf a licensee _can. demonstrate this assurance without quantifying the amount of unqualified coating, this is acceptable.' If. commercial-
. grade coatings are being used at your facility for Service Level.1f applications, and :
such coatings are not dedicated or controlled under your Appeneesdix B QA Program, provide the regulatory and safety basis for not controlling these.costings in accordance with such a program.-. Additionally, explain why the_f_scility's. licensing basis does.not require _such. a program. Address the required written information to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, under oath or affirmation pursuant to Section 182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). This information will enable the Commission to determine whether the I; cense should be modified, suspended, or revoked in addition, submit a copy of the written information to the appropriate regional administrator. Backfit Discusebn This generic letter requires information from the addressees under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). This generic letter does not constitute a backfit as defined in 10 CFR 50.109(a)(1) since it does not impose modifications of or additions to systems, structuras, and components or to the design or operation of an addressee's facility. It also does not impose an interpretation of the Commission's rules that is either new or different from a previous staff position. The staff has, therefore, not performed a backfit analysis. ~. __ .
Generic Letter 97-xx Month, Date, Year Page 8 of 10 Reasons for information Requirement This generic letter transmits a requirement to submit information pursuant to the provisions of Section 182a of the Atomic Energy Act of 1954, as amecded, and 10 CFR 50.54(f) for the purpose of verifying compliance with applicable regulatory requirements. Specifically, the required information will enable the NRC staff to determine whether the addressees' protective coatings inside the containment comply and conform with the current licensing basis for their respective facilities and whether the regulatory requirements pursuant to 10 CFR 50.46 are being met. Protective coatings are necessary inside containment to control radioactive contamination and to protect surfaces from erosion and corrosion. Detachment of the coatings from the substrate may make the ECCS unable to satisfy the requirement of 10 CFR 50.46(b)(5) to provide long-term cooling and may make the safety-related CSS unable to satisfy the plant-specific licensing basis by controlling containment pressure and radioactivity following a LOCA. Paperwork 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 on September 30, 2000. The public reporting burden for this collection of information is estimated to average 400 hours per response, including the time for reviewing instructions, searching existing data sources, gathering and mair:taining the data needed, and completing and reviewing the collection of information. The NRC is seeking public comment on the potential impact of the collection of information contained in the generic letter and on the following issues: I (1) is the proposed collection of information necessary for the proper performance of the functions of the NRC, and will the information 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, can automated collection techniques be used? Send comments on any aspect of this collection of information, including suggestions for reducing this burden, to the information and Records Management Branch, T-6F33, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the Desk Officer, Office of information and Regulatory Affairs, NEOB-10202 (3150-0011), Office of Management and Budget, Washington, DC 20503. 4 m-%+. + ew - .
Generic Letter 97-xx Month, Date, Year l Page 9 of 10 The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. If you have any questions about this matter, please contact either of the technical contacts or the lead project manager listed below, or the appropriate Office of Nuclear Reactor Regulation project manager. Jack W. Roe, Acting Director Division of Reactor Program Management > Office of Nuclear Reactor Regulation I Technical Contacts: Richard Lobel, NRR (301) 415-2865 e-mail: RML@NRC. GOV James Davis, NRR (301)415-2713 e-mail: JAD@NRC. GOV Lead Project Manager: Guy Vissing, NRR (301) 415-1441 e-mail: GSV@NRC. GOV r.... . . . . ~ . . . .
Generic Letter 97-xx Month, Date, Year Page 10 of 10 Appendices: Appendix A, "ECCS SUMP AND STRAINER EVENTS INVOLVING FOREIGN MATERIAL INSIDE THE CONTAINMENT AND CONSTRUCTION AND DESIGN DEFICIENCIES" Appendix B, " OPERATIONAL EVENTS INVOLVING DEBRIS IN ECCS RECIRCULATION FLOW PATHS" Appendix C, " BACKGROUND ON REGULATORY BASIS FOR PROTECTIVE COATINGS" Appendix D, " CHRONOLOGY OF INCIDENTS AND ACTIVITIES RELATED TO PROTECTIVE COATINGS" Appendix E, " GENERIC COMMUNICATIONS ISSUED BY THE NRC ON THE SUBJECT OF ECCS AND SAFETY-RELATED l CSS SUMP AND STRAINER BLOCKAGE" J Appendix F, "NRC ENFORCEMENT ACTIONS DEALING WITH CONSTRUCTION AND PROTECTIVE COATINGS DEFICIENCIES AND FOREIGN MATERIAL EXCLUSION" i I
Appendix A ECCS SUMP AND STRAINER EVENTS INVOLVING FOREIGN MATERIAL INSIDE THE CONTAINMENT AND CONSTRUCTION AND DESIGN DEFICIENCIES On November 16,1988, the NRC issued Information Notice (IN) 88-87, " Pump Wear and Foreign Objects in Plant Piping Systems," concoming several incidents in which the potential existed for a flow reduction as a result of pump wear and foreign objects in plant piping systems. In one of these incdonts, the licensee found foreign objects in a temporary pump discharge cone strainer. The licensee investigated further and found foreign objects, dating to early construction modifications, in the sump. In addition, various deficiencies were found in the sump screens. On November 21,1989, the NRC issued IN 89-77, " Debris in Containment Emergency Sumps and Incorrect Screen Configurations," which discussed loose parts and debris in the containment sumps of three pressurized-water reactors (PWRs), Surry Units 1 and 2 and Trojan. At Surry Units 1 and 2, some of the debris was large enough to cause pump damage or flow degradation. In addition, some of the screens had gaps large enough to allow additional loose material to enter the sump. The licensee found that screens that separate the redundant trains of the inside recirculation spray system were missing at both units. At Trojan, the licensee discovered debris in the sump. Some debris was found after containment closeout. In addition, stil! later, before startup, the NRC identified missing portions of the sump top screen and inner screen. IN 89-77 also reported that in 1980 the Trojan licensee found a welding rod jammed between the impeller and the casing ring of a residual heat removal pump. On December 23,1992, the NRC issued IN 92-85, " Potential Failures of Emergency Core Cooling Systems Caused by Foreign Material Blockage," which alerted licensees to events at two PWRs. In these events, foreign material blocked flow paths within the ECCS safety injection and containment spray pumps so that the pumps could not produce adequate flow. On April 26,1993, and May 6,1993, the NRC issued IN 93-34, " Potential for Loss of Emergency Cooling Function Due to a Combination of Operational and Post-LOCA Debris in Containment," and its supplement. In these information notices, the NRC described several instances of clogged ECCS pump strainers, including two events at the Perry Nuclear Power Plant, a domestic boiling-water reactor (BWR). In the first Perry event, residual heat removal (RHR) strainers were clogged by operational debris consisting of " general maintenance-type material and a coating of fine dirt." After cleaning the strainers in January 1993, the licensee discovered that RHR A and B strainers were deformed. The strainers were replaced. The second Perry event involved an RHR pump test which was run after a plant transient in March 1993. Pump suction pressure dropped to 0 KPa (0 psig). No change in pump flow rate was observed. Material found on the strainer screen was analyzed and found to consist of glass fibers from temporary drywell cooling filters that had been inadvertently dropped into the suppression pool and corrosion products that had been filtered from the pool by the glass fibers adhering to the surface of the strainer. This significantly increased the pressure drop across the strainer. A-1 I
.= .- .. .
in response to these two events, the !icensee for Perry increased the suction strainer area, provided suction strainer backflush capability, and improved measures to keep the suppression pool clean. On May 11,1993, the NRC issued Bulletin 93-02, " Debris Plugging of Emergency Core Cooling Suction Strainers," which requested that both PWR and BWR addressees (1) identify fibrous air filters and other temporary sources of fibrous material in containment not designed to withstand , a loss-of-coolant accident (LOCA) and (2) take prompt action to remove the foreign matter and l ensure the functional capability of the ECCS. All addressees have responded to the bulletin, I and the NRC staff has completed its review of their responses, l I The licensee for Arkansas Nuclear One, Unit 2, reported by Licensee Event Report (LER) 93- l 002-00, dated November 22,1993, that the containment sump integrity was inadequate to keep ; foreign material out. Holes in the masonry grout below the sump screen assembly would have ; let water into the sump without being screened. The licensee attribuhd this condition to failure i to implement design basis requirements for the sump during initial plant construction. The holes were difficult to detect. The holes appeared to be part of the design because of their uniform spacing and because they were "somewhat recessed...such that to see the holes they must be viewed from near the floor or from a significant distance away from the sump." On August 12,1994, the NRC issued IN 94-57, " Debris in Containment and the Residual Heat Removal System," which alerted operating reactor licensees to additional instances of degradation of ECCS components because of debris. At River Bend Station, the licensee found a plastic bag on an RHR suction strainer. At Quad Cities Station, Unit 1, on July 14,1994, the remains of a plastic bag were found shredded and caught within the anti-cavitation trim of an RHR test retum valve. Subsequent to that event at Quad Cities, Unit 1, the licensee observed reduced flow from the C RHR pump and, upon further investigation, found a 10-cm (4-in.) diameter wire brush wheel and a piece of metal wrapped around a vane of the pump. On January 25,1995, the NRC issued IN 95-06, " Potential Blockage of Safety-Related Strainers by Material Brought inside Containment," which discussed a concem that plastic or fibrous material, brought inside the containment to reduce the spread of loose contamination, to identify equipment, or for cleaning purposes, may collect on screens and strainers and block core cooling systems. Several examples were cited. ; On October 4,1995, the NRC issued IN 95-47, " Unexpected Opening of a Safety / Relief Valve , and Complications involving Suppression Pool Cooling Strainer Blockage," which discussed an event on September 11,1995, at the Limerick Generating Station, Unit 1, during which a safety / relief valve discharged to the suppression pool. The operators started an RHR pump in the suppression pool cooling mode. After 30 minutes, fluctuating motor current and flow were observed. Subsequent inspection of the strainers found them covered with a " mat" of fibrous material and sludge (corrosion products) from the suppression pool. The licensee removed approximately 635 kg (1400 lb) of debris from the Unit 1 pool. A similar amount of debris had been removed earlier from the Unit 2 pool. A supplement to IN 95-47 was issued on November 30,1995. A-2
, - - , + =
On October 17,1995, the NRC issued NRC Bulletin 95-02, " Potential Clogging of a Residual Heat Removal (RHR) Pump Strainer While Operating in Suppression Pool Cooling Mode," which discussed the Limerick Unit 1 event and requested that BWR addressees review the operability of their ECCS pumps and other pumps that draw suction from the suppression pool while performing their safety function. The addressees' evaluations were to take into consideration suppression pool cleanliness, suction strainer cleanliness, and the effectiveness of the ; addressees' foreign material exclusion (FME) practices. In addition, BWR addressees were ! requested to implement appropriate procedural modifications and other actions (e.g., suppression pool cleaning), as necessary, in order to minimize the amounts of foreign material in the suppression pool, drywell, and containment. BWR addressees were also requested to verify their operability evaluation through appropriate testing and inspection. On February 10,1996, the NRC issued IN 96-10. " Potential Blockage by Debris of Safety > System Piping Which is Not Used During Normal Operation or Tested During Surveillances," which discussed debris blockage in ECCS lines taking suction from the containment sumps at a PWR in Spain. In one of the two partially blocked lines, almost half the flow area of the pipe was blocked; the other line was less blocked. Upon further investigation, Spanish regulators found that many sections of piping in both PWRs and BWRs are only called upon to function during accident conditions and are not used during normal operation or tested during functional surveillance tests. The licensee in this case concluded that the safety significance was low because the partial blockage of the lines would not have prevented the ECCS from providing sufficient core cooling. However, it was also noted that some of the debris could have been entrained in the water flow and could have detrimental effects on other parts of the system (e.g., pump and valve components and heat exchangers). In addition, in LER 96-005, the licensee for the H.B. Robinson Steam Electric Plant, Unit 2, reported finding in a pipe in the sump an item of debris larger than the 0.95 cm (3/8-inch) diameter of the holes in the containment spray nozzle. l In LER 96-007, the licensee for Diablo Canyon Nuclear Power Plant, Unit 1, reported a radiograph inspection finding that openings in the Diablo Canyon plant's 3.81 crn (1-1/2 in.) centrifugal-charging-pump runout-protection manual throttle valves and in the 5.08 cm. (2 in.) safety-injection (SI) to cold-leg manual throttle valves were less than the 0.673-cm (0.265-inch) diagonal opening in the containment recirculation sump debris screen. Therefore, debris could potentially block charging or SI flow through these throttle valves during the recirculation phase of a LOCA. The licensee concluded that even with a postulated blockage of the throttle valves, the RHR system flow by itself would be sufficient to maintain adequate core cooling during recirculation following a postulated accident. As a corrective action, the Diablo Canyon licensee stated in LER 96-007 that the system would be modified to ensure that the throttle valve clearance is greater than the maximum sump screen opening. After reviewing an Institute of Nuclear Power Operations (INPO) operational experience report on this event, the licensee for Millstone Nuclear Station, Unit 2, determined that eight throttle valves in the high-pressure safety-injection (HPSI) system injection lines were susceptible to the failure mechanism described in Diablo Canyon Nuclear Power Plant LER 96-007. This situation is discussed in NRC IN 96-27, " Potential Clogging of High Pressure Safety injection Throttle Valves During Recirculation," dated May 1,1996. The Millstone Unit 2 licensee concluded that A-3
the type of debris that would pass through the screen openings would tend to be of low density and low structural strength and that material of this type would be reduced in size as it passed through the HPSI and containment spray pumps. In addition, the differential pressure across the HPSI system injection valves and containment spray nozzles would tend to force through the valves or nozzles any material that is " marginally capable" of obstructing flow. These conclusions may be plant-specific and may not be applicable to other designs. The Millstone Unit 2 licensee committed to replace the sump screen with one that is consistent with the original design. On May 6,1996, the NRC issued Bulletin 96-03, " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors," which requested actions by BWR addressees to resolve the issue of BWR strainer blockage because of excessive buildup of debris from insulation, corrosion products, and other particulates, such as paint chips and concrete dust. The bulletin proposed four options for dealing with this issue: (1)installlarge-capacity passive strainers, (2) install self-cleaning strainers, (3) install a safety-related backflush system that relies on operator action to remove debris from the surface of the strainer to keep it from clogging, or (4) propose another approach that offers an equivalent level of assurance that the ECCS will be able to perform its safety function following a LOCA. BWR addressees were requested to implement the requested actions of Bulletin 96-03 by the end of the first refueling outage beginning after January 1,1997. On October 30,1996, the NRC issued IN 96-59, " Potential Degradation of Post Loss-of-Coolant Recirculation Capability as a Result of Debris," to alert addressees that the suppression pool and associated components of two BWRs, LaSalle County Station, Unit 2, and Nine Mile Point Nuclear Station, Unit 2, were found to contain foreign objects that could have impaired successful operation of emergency safety systems that used water from the suppression pool. In particular, debris was found in the downcomers (large-diameter pipes connecting the drywell to the suppression pool). Although the licensee for Nine Mile Point, Unit 2, had previously cleaned the suppression pool, the downcomers had not been inspected. In addition, the licensee found debris covers in place on seven of the eigh* downcomers located in the pedestal area directly under the reactor vessel. These debris covers had been in place since construction. LER 96-11-00 attributes this oversight to inadequate managerial methods and to environmental conditions, as the " accessibility of the pedestal area downcomers requires removal of grating in the undervessel area and climbing down to the dimly lit subpile floor. The plastic covers on the downcomers are not visible from the grating elevation because of the missile shield plates above the downcomer floor penetrations. Furthermore, since the first ; refueling outage, access to this area has been limited because of the high contamination levels l and general ALARA [as low as reasonably achievable radiation dose] considerations." Although the NRC has not previously discussed the subject in a generic communication, licensee event reports have been submitted regarding the loss of control of containment sump access hatches leaving them open during periods when ECCS sump integrity was required. For ; example, in LER 89-014-01, the licensee for Diablo Canyon Nuclear Power Plant, Unit 1, discussed the opening of the sump access hatch at various times at power "without adequate consideration of ECCS operability." in LER 96-006, Watts Bar Nuclear Plant, Unit 1, reported , that an operator observed a containment sump (trash screen) door open while ECCS operability I was required. 1' A-4
Appendix B OPERATIONAL EVENTS INVOLVING DEBRIS IN ECCS RECIRCULATION FLOW PLANT / REPORT PROBLEMS DISCUSSED Haddam Neck in July 1975, six 55 gallon drums of NRC Inspection Report 50- sludge with varying amounts of 213/96-08 debris removed from ECCS sump. North Anna Units % Galvanized ductwork painted with LER 84-006-00 unqualified paint. Millstone Unit 1 Existing suction strainers too small LER 88-004-00 when criteria of RG 1.82 Rev.1 applied. Strainers will be replaced with larger strainers if ISAP criteria met. 1 Surry Power Station Units 1. Foreign material from construction l
% , activities found in cone strainer of LER 88-017-01 recirculation spray system. Material IN 88-87 could have rendered system .
IN 89-77 inoperable. I
- 2. Gaps in sump screens since initial construction.
Trojan Nuclear Plant 1. Wire mesh screen on top of sump LER 89-016-01 trash rack not installed. IN 89-77 2. Screen damage. ,
- 3. Significant amount of debris '
discovered in the sump. Could have caused loss of part or all of ECCS. Diablo Canyon Unit 1 1. Debris in sump. 1 LER 89-014-01 2. As-built sump configuration not in IN 89-77 accordance with design.
- 3. Safety function would not have been impaired.
TMl Unit 1 1. Modification of sump access LER 90-002-00 hatches left holes in top of sump screen cage. Could damage pumps or clog spray nozzles. B-1
1 McGuire Unit 1 1. Loose material discovered in LER 90-0112-00 upper containment before entry into Mode 4. Items found would not have made ECCS inoperable. Calvert Cliffs Unit % 1. Unit 2 sump found to contain 11.3 NRC inspection Report kg (25 lb) dirt, weld slag, pebbles, i (March 5,1991) etc. Inspection of Unit 1 found less ! than 1 lb. debris. Possible minor damage to ECCS pumps. _ Diablo Canyon Unit 2 1. Numerous instances of material i LER 91-012-00 left unattended or abandoned in i sump level of containment (tools, l plastic tool bags, clothing, etc.).
- 2. Material would not have prevented ECCS recirculation function.
H.B. Robinson Unit 2 B safety injection pump flow reduced LER 92-013-00 due to blockage in minimum flow recirculation check valve and flow orifice on July 8,1992. A pump OK. 4 Foreign material also found in l RWST. H.B. Robinson Unit 2 On August 24,1992, following a LER 92-018-00 reactor trip, A and B safety injection puriips inoperable due to reduced flow. Found during unscheduled surveillance to demonstrate Si operability. Pt. Beach Unit 2 September 18,1992: During LER 92-003-01 Technical Specification (Inservice IN 92-85 Inspection Testing) testing of the A containment spray pump the pump was declared inoperable. A foam rubber plug was blocking pump suction. Plug removed and pump tested satisfactorily. One train of Unit 2 residual heat removal, safety injection and containment spray systems inoperable for entire operating cycle. Plug was part of a cleanliness barrier. B-2 p4y -+ - e - e .-*=.%wq,,e-,e < ~ -
.-e.e-ee- 4
Perry Nuclear Plant 5/92: During a refueling outage, l LER 93-011-00 foreign objects discovered in the containment side of the suppression pool. Fout:ng of RHR strainers found. Strainers not cleaned. 1/93: RHR A/B strainers found deformed (collapsed inward in the l direction of the fluid flow). Strainers l replaced. 3/93: RHR A/B operated in 1 suppression pool cooling node. Pump suction pressure deuessed. Could have compromised long-term RHR operation. Susquehanna Units % 1. Assessing impact of debris and i LER 93-007-00 corrosion products adhering to (voluntary) fibrous materials that may be dislodged by pipe break. .
- 2. Developing procedures to j backflush strainers. l Sequoyah Unit 2 Design basis limit for unqualifed l LER 93-026-00 coatings inside containment had i been exceeded. Additional quantity of unqualified coatings on RCP motor platform discovered. Path to ECCS sump. Screens will be installed before startup.
ANO Unit 2 7 unscreened holes found in LER 93-002-00 masonry grout below screen IN 89-77 Supplement 1 assembly of ECCS sump. Could potentially degrade both trains of HPSI and containment spray. Had previously inspected sump because of IN 89-77; did not discover problem. NRC estimate of incrementalincrease in core damage: 3 X10*. l 4 B-3 3
ANO Unit 1 1. 22 unscreened 15.2 cm x 7.6 cm LER g3-005-00 (6"X3") pipe openings at base of IN 89-77 Supplement 1 sump curb, the result of a modification beforeinitialoperation. !
- 2. Tears in screen. !
- 3. Floor drains leading to sump not
- creened. )
- 4. Licensee estimated increase in core damage frequency SX10*.
San Onofre Units % 1. Irregular annular gap LER 93-010-00 (approximately 1G.2 cm [6"]) i (voluntary) surrounding 20.3 cm (8") LTOP discharge line penetrating horizontal steel cover plate.
- 2. Engineering analysis concluded both sump trains operable.
Vermont Yankee 1. LPCS suction strainers smaller LER g3-015-00 than calculations assumed. NPSH calculations performed in 1986 following change to NUKON" insulation invalid.
- 2. Strainers replaced with larger strainers. j Sout:1 Texas Units % 1. Sump screen openings from initial I LER 94-001-00 construction discovered. Frame plate at floor warped, creating soveral openings approximately 1.6 cm (5/8"). Additional 0.6 cm (1/4")
gaps discovered. Based on ECCS pump tests performed by the manufacturer, the licensee concluded the deficiencies had no safety significance. Point Beach Unit /1 NRC inspector found grout l NRC Inspection Report deterioration under sump screens. May 6,1904 Could result in flow bypass, or particles of grout COULD enter ECCS pumps. LaSalle Unit 1 April 26 and May 11,1994: Divers i IN 94-57 inspecting suppression pool during I outage found operational debris. I B-4
River Bend June 13,1994: Plant h refueling IN 94-57 outage. Foreign materias found in suppression pool. Plastic bag j removed from B RHR pump suction strainer. Other objects: tools, grinding wheel, scaffolding knuckle, stepoff pad Quad Cities Unit 1 July 14,1994, post-maintenance test IN 94-57 of A loop RHR indicated a plugged torus cooling / test retum valve. Inspection discovered remains of shredded plastic bag in anti-cavitation trim installed during a recent outage. July 23,1994: 4" diameter wire brush and a piece of metal found wrapped around a vane of the C RHR pump. Browns Ferry Units %/3 1. Unqualified coatings on T May 20,1994, Letter to quenchers in suppression pool. NRC 2. Continued operation acceptable.
- 3. Will remove coatings next refueling outage.
Palisades Plant Signs, adhesive tape, and labels with LER 94-014-00 potential to block the ECCS sump were found in containment. Containment spray and HPSI pumps declaredinoperable. Engineering analysis concluded that the sump screen would not be significantly blocked. Watts Bar Units % Screens installed around RCP NRC Inspection Report 50- motors to catt,h unqualified paint not 390 and 50-391/94-59 adequately located to contain all (September 28,1994) unqualified coatings. ; indian Point Unit 2 Licensee discovered portions of the l' LER 95-005-00 floor on Elevation 46 in containment had lifted and cracked. In other locations, floor coating cracked when i stepped on. Licensee concluded that j sump function would not be i compromised. i l B-5 . i u_ _ _ . . . _ _
Susquehanna Units % Licensee took actions to address LER 93-007-001 concem of clogging ECCS suction September 11,1995 strainers. Among these actions: removal of fibrous insulation from HELB areas, testing to determine whether the debris could block the strainer, quantification of corrosion products on structural steelin wetwell, a comprehensive analysis of containment debris effects. Coating and insulation procedures contain steps to reduce potential for strainer blockage. Prairie Island Unit 2 Broken labels for pipe hangers and NRC Inspection Report 50- labels affixed to wall with degrading 282/05-009 adhesive discovered by NRC inspector after licensee closeout inspection. Licensee concluded that . this potential debris would not affect l operability of ECCS. I Palisades Unsecured material stored on the I NRC Inspection Report 50- landings of stairways. Broken glass 225/95-008 and pieces of signboard and other
" unauthorized" material found in area ;
designated debris-free. l 1 Limerick Unit 1 Debris was allowed to collect in NRC Inspection Report 50- suppression pool rendering the A 352/96-04 RHR pump inoperable when safety / relief valve lifted on September 11,1995. l Duane Arnold FME controls inadequate in drywell. NRC Inspection Report 50- Hardhats and debris noted. 331/95-003 1 Foreign PWR 1. Operator found dobris in the sump. NRC IN 96-10 2. 2 of 4 ECCS lines taking suction from the sump were partially blocked by debris. Debris present since plant construction. Millstone Unit 2 10 locations inconsistent with the LER 96-008 screen mesh size were identified. Placed plant outside original design basis. Sump screen replaced. B-6
Watts Bar Unit 1 Operator observed containment LER 96-006-00 sump trash screen door was open when plant in MODE 4 and ECCS required to be operable. Calvert Cliffs Units % Several holes identified in each unit's LER 96-003-00 containment sump screen larger than described in the FSAR. Holes field-installed for transmitter tubing. Concluded not a threat to plant safety. Diablo Canyon Unit i Various debris that could pass LER 96-007-00 through the containment sump screen could be larger than openings in the 3.8 cm (1-1/2") centrifugal-charging pump runout-protection manual throttle valvas and 5.1 cm (2") SI-to-cold leg manual throttle valves. . Haddam Neck 1. Discrepancies in sump screen LER 96-014-00 mesh sizing, screen fitup, and NRC Inspection Report 50- method of attachment discovered. 213/96-08 Sump screen replaced. Sump will be inspected after every refueling outags. Licensee reported this as a condition which alone could have prevented the fulfillment of a safety function.
- 2. 5 208 liter (55-gallon) drums of sludge removed frem ECCS sump.
Also, plastic sheeti g, nuts, and bolts, tie wraps and pencils. Big Rock Point " Housekeeping in containment in the NRC inspection Report 50- area under the emergency 155/96-004 condenser and the reactor depressurization system isolation vdes was poor." B-7
Catawba Unit 1 6 floor drains inside crane wall were NRC Inspection Report 50- not mvered with screen that had a 413/96-11 fine' mesh than the sump screen. 0.6 nm (1/4") holes rather than 0.3 crn (1/8") holes. Crane wall pene' rations close to containment floor r.ould allow the transport of debris to the sump screen. Penetretions sealed. Millstone Unit 2 Contair, ment sump screens had LER 50-336/96-08 been ire,orrectly constructed so that NRC Inspection Report largerdebr:s than analyzed could 50-336/96-08 pass ti rough the ECCS. Vogtle Unit 2 Loose debris in "readily accessible NRC inspection Report 50- aress' identified by NRC inspectors 425/96-11 inside containment had the potential LER 96-007-00 to bic :,k emergency sump screens during accident conditions. Licensee's evaluation concluded that debris did not represent " substantial challenge" to ECCS. 0.6 m2 (6 ft 2) of debris estimated. Additionalitems identified by licensee and NRC inspector during startup while in MODE 3. Further evaluation by licensee concluded that RHR pump would not have had adequate NPSH because of debris. Nine Mile Point Unit 2 A significant amount of debris was NRC Inspection Report 50- found in the suppression pool and 410/96-11 downcomers during refueling outage NRC Event Report 31172 number 5. Licensee's preliminary > evaluation concluded that operability of ECCS could have been compromised. B-8 l
LaSalle Unit 2 Foreign material recovered from NRC Event Report 31159 suppression pool and downcomers. LER 96-009-00 This material would challenge the operability of the ECCS. Approximately 0.7 m2 7 ft2 per strainer removed from Suppression pool. Material most likely from construction or early outages. Specia! multiple ECCS pump runs perfonned with satisfactory results. No apparent transport of the foreign material discovered during this outage. , i Millstone Unit 3 1. Construction debris discovered in 1 LER 96-039-00 containment recirculation spray ; system (RSS) containment sump and I in RSS suction lines.
- 2. Gaps discovered in RSS sump cover plates. !
- 3. Later inspection found other sump enclosure gaps. ;
- 4. Bolts and clips missing from the vo.dex suppression grating
- 5. Debris found in all 4 RSS pump suction lines.
H.B. Robinson Unit 2 1. Openings found in sump screens. LER 96-005-00 They could have allowed debris above a certain size to enter the sump or prevented the screens from performing their design function.
- 2. An item of debris in excess of the 1 cm (3/8") diameter containment spray nozzles was found in 36 cm (14") sump drain pipe.
Zion Unit 1 Two 2.5 cm (1") holes detailed on LER 97-001-00 drawings, were not in the sump cover. Holes allow air to escape as sump fills. Potential to hinder flow to RHR pump suction during a LOCA. B-9
Zion Unit 2 1. Miscellaneous debris found NRC Inspection Report 50- throughout containment. 295/96-20 2. Containment recirculation sump 50-304/96-20 screen damage. March 24,1997 3. Peeling and flaking paint on containment surfaces. Sequoyah Unit 1 During shutdown on March 22,1997 10 CFR 50.72 an oil cloth was introduced to Report 32139 containment. If it had come free of its , (April 11,1997) restraints, it could have blocked one or both refueling drains so that water in upper containment might not have flowed freely to lowerlevel of i containment, where sump is located. Millstone Unit 1 Most of the coating in the torus is 10 CFR 50.72 unqualified, which could affect the Report 32161 operability of the low-pressure I (April 16,1997) coolant injection and core spray systems. ! Clinton Significant degradation in protective 50.72 Report 32633 coatings in the containment wetwell. July 15,1997 Some degradation in the drywell, i Licensee concluded that the amount of degraded coatings from the containment and the drywell could have exceeded the ECCS strainer loading under accident conditions. Oconee Unit 2 Degraded protective coatings in No written report. containment. St. Lucie Unit 2 Containment sump screens with to LER 50-389/97-002 gaps in screen enclosure contrary to design. B-10
DC Cook Units % A 1 cm (0.25") particulate retention 50.72 Report 32875 requirement for the containment recirculation sump was not proper 1y established following sump modifications. Inadvertent pathways with openings greater than 1 cm (0.25") were found, including 3 cm(0.75") vents in roof of sump (see following item). Licensee concluded that the ECCS was outside its design , basis. DC Cook Units % Licensee determined that inlet 50.72 R3oort 32903 venting requirement for the containment recirculation mismp: J..s 4 net - p.1y maintained following modifications to the sumps. Sump venting requirement was met through addition of 5 3 cm (0.75") holes drilled in the roof of the sump inlet. When these holes were discovered during refueling outages they were filled with concrete, eliminating the venting path. Turkey Point Units 3/4 Gaps greater than 1 cm (0.25") found 50.72 Report 32910 in screens for Unit 3 and 4 sumps. (September 11,1997) D.C. Cook Units % Enough fibrous material was found in 50.72 Report 32948 both Unit 1 and Unit 2 containments (September 17,1997) to potentially cause excessive blockage of the Containments recirculation rump screen during the recirculation phase of a LOCA. Both units were already shut down for other reasons. The material was removed from both units before startup. t B-11
DC Cook Units % A 1 cm (0.25") particulate retention 50.77 Report 32875 requirement for the containment (September 25,1997) recirculation sump was not properly established following sump modifications. Inadvertent pathways with openings greater than 1 cm (0.25") were found, including 3 cm (0.75") vents in roof of sump (see following item). Licensee concluded that the ECCS was outside its design basis. l l l 1 i I B-12
Appendix C' BACKGROUND ON REGUI.ATORY BASIS FOR PROTECTIVE COATINGS This appendix discusses the regulatory basis, including industry standards and regulatory guidance for protective costings inside the containment. However, this discussion is only for information. Addressees should continue to comply with the plant licensing basis. At nuclear power plants, coatings and paints serve to (1) protect carbon and low alloy steel, austenitic steel, and less commonly, galvanized steel, and aluminum surfaces against corrosive environments; (2) protect metallic, concrete, or masonry surfaces against wear during plant operation; and (3) allow for ease of decontamination of radioactive nuclides from the containment wall and floor surfaces. These coatings come in inorganic forms, such as zinc-based paints, and organic forms, such as epoxy coatings. ANSI Standards N101.2, " Protective Coatings (Paints) for Light Water Nuclear Reactor i Containment Facilities," and ANSI N101 A, " Quality Assurance for Protative Coatings Applied to l Nuclear Facilities," classify coatings as Service Level 1, Service Level 2, or Service Level 3. J A Service Level 1 coating is used on an exposed surface area within the containment. A Service Level 2 coating is used on any exposed surface area located outside containment but subject to radiation and decontamination. A Service Level 3 coating is used on any exposed surface area located outside containment whose failure could adversely affect normal plant operation or orderly and safe plant shutdown. l This generic letter concems the possible detrimental effects of failed coatings on a plant's ability to recirculate coolant following a LOCA. Therefore, this generic letter is concemed with Service Level 1 coatings. j Protective coatings applied to the interior surfaces of the containment structure and to SSCs inside the containment are considered qualified coatings if they have been subjected to physical property (adhesion) tests under conditions that simulate the projected environmental conditions I of a postulated design basic (DB) LOCA and have been demonstrated to maintain their adhesive properties under these simulated conditions. These tests are typically conducted in accordance with the guidelines, practices, test methods, and acceptance criteria specified in applicable industry standard procedures (such as those issued by the American National Standards institute, Inc. [ANSl], or the American Society for Testing and Materials [ ASTM]) for coatings applications. However, the licensing basis for Service Level I coating applications may contain exceptions to, or provide attemative means of meeting the intent of, the test methods in ! these standards, provided an adequate safety basis is given to and accepted by the NRC staff I as to why accepting the exceptions or altematives could affect the performance of the ECCS and safety-related CSS during a postulated DB LOCA. In regard to protective coatings used for Service Level I service applications inaide the containment, the staff normally concludes that a coating system is acceptable for service if it has been demonstrated that the coating system is C-1 ;
qualified to maintain its integrity during a postulated DB LOCA and if the programs for controlling applications of coating systems for Service Level I service applications are implemented in accordance with a quality assurance (QA) program that meets the requirements of Appendix B to 10 CFR Part 50. Protective coatings that have not been successfully tested in accordance with the provisions in the applicable ANSI or ASTM standards or have not met the acceptance criteria of the standards are considered to be " unqualified"; that is, they are assumed to be incapable of maintaining their adhesive properties during a postulated DB LOCA. The sta# normally assumes that " unqualified" coatings applied to the interior surfaces of the containment structure and to SSCs inside the containment structure will form solid debris products under DB LOCA conditions. These debris products should, therefore, be evaluated for their potential to clog ECCS sump screens and strainers and their effect on the operability of safety-related pumps taking suction from ECCS sumps and suppression pools during a postulated DB LOCA. The NRC issued Regulatory Guide (RG) 1.54-1973, " Quality Assurance Requirements for Protective Costings Applied to Water-Cooled Nuclear Power Plants," to give the industry an acceptable method for complying with the QA requirements of 10 CFR Part 50, Appendix B, as they relate to protective coating systems applied to cart >on and low alloy steel, austenitic stainless steel, aluminum, galvanized steel, and masonry surfaces of water-cooled nuclear power reactors. In RG 1.54-1973, the NRC stated that the guidelines for coating applications in ANSI Standard N101.4-1972, " Quality Assurance for Protective Coatings Applied to Nuclear Facilities," as supplemented in RG 1.54-1973, delineate acceptable QA criteria for providing confidence that " shop or field coating work [will] perform satisfactorily in service." The quality assurance provisions stated in ANSI Standard N101.4-1972, as endorsed by the staff in RG 1 1.54-1973, are considered by the staff to provide an adequate basis for complying with the i pertinent QA requirements of 10 CFR Part 50, Appendix B. These standaros delineate the type ! of tests to be performed to qualify a given coating for nuclear applications. However, how a licensee implements its progran: for controlling activities re!ated to protective coating applications at a particular nuclear plant depends on the plant's licensing basis. Neither RG 1.54-1973 nor the applicable ANSI standards are NRC requirements: they merely delineate acceptable programs and practices for controlling coatings application activities at nuclear power plants. ANSI Standard N101.4-1972 provides recommended guidelines for implementing QA programs regarding coating applications at domestic nuclear power plants. ANSI Standard N101.4-1972, as endorsed in RG 1.54-1973, delineates recommended guidelines and criteria for establishing QA and quality control programs for coating activities. Such a program should control work conditions, the ambient environmental conditions for coating applications, selection and pro.?urement activities for coatings, and preparation of substrate surfaces, establish QA procedures for coating applications, qualify personnel involved in coating preparation, application, and inspection activities, and establish coating inspection guidelines and acceptance criteria. ANSI Standard N101.4-1972, as endorsed by RG 1.54-1973, also recommends keeping certain QA records On coatings activities. . ANSI Standard N101.4-1972 states that ANSI Standard N5.9," Protective Coatings (Paints) for the Nuclear Industry"(later reissued as ANSI Standard N512), and ANSI Standard N101.2, C-2
" Protective Coatings (Paints) for Light-Water Nuclear Reactor Containment Facilities," are additional acceptable standards governing activities related to the selection and evaluation of protective coatings applied both in the shop (i.e., at vendor or manufacturer facilities) or in the feld.
1 RG 1.54 is currently undergoing a major revition (it was last revised in 1973). Many of the documents ref6renced in RG 1.54 are outdstad and have been replaced by newer ASTM or ANSI standards. ASTM Committee D-33, " Coatings for Power Generation Facilities," has developed the standards that replace many of the standards referenced in RG 1.54-1973. At the request of the NRC staff, this committee is currently developing a maintenance standard for qualified coatings. This standard will cover inspsetion of existing coatings, application of new coatings over the original substrate (steel, concrete, galvanized steel, aluminum), new coatings over a substrate-old coating interface, and new coatings over old qualified coatings. When this l standard is approved, RG 1.54-1973 will be revised to reflect current standards. Using more up- j to-date industry standards for protective coatings may require changing a plant's licensing ' basis. Use of these standards must conform with existing NRC requirements, including 10 CFR Part 50, Appendix B. l 1 C-3 L
Appendix D CHRONOLOGY OF INCIDENTS AND ACTIVITIES RELATED TO PROTECTIVE COATINGS in January 1997, Commonwealth Edison Company (Comed), the licensee for the Zion Nuclear l Plant, Unit 2, discovered flaking and unqualified paint applied to the containment surfaces (IN 97-13, " Deficient Conditions Associated With Protective Coatings At Nuclear Power Plants"). The peeling of the protective coatings was determined to occur at the horizontal junction lines located between the concrete shells that were used in construction of the Zion Unit 2 l containment structure. Comed estimated that the total weight of degraded coatings (peeling paint) was approximately 445 N (100 lb). Comed also initially estimated that an additional 557-650 m8(6000-7000 ft') of coatings on surfaces inside containment were not qualified to withstand the environmental conditions of a postulated DB LOCA, in accordance with the testing criteria of ANSI Standsid N512-1974. Comed determined that the peeling of the qualified coatings on the containment surfaces was due to improper surface preparation, resulting in j inadequate adhesion of the coating following application. ' Comed corrected the condition of the paint by removing all of the degraded " qualified" paint inside the Zion Unit 2 containment and by removing all of the additional" unqualified" paints that were determined to be located within the analytically determined zone of influence.8 Comed also performed 33 random adhesion or " pull" tests on the remaining, intact, " qualified" paint inside the containment structure. All of these tests were performed in accordance with the applicable testing requirements specified in ANSI Standa,d N512-1974. All of the tests exhibited " pulls" in excess of the 890 N (200 lb) required by the standard, thus demonstrating that the remaining qualified coatings were acceptable for service during the next operating cycle. On March 10,1995, Consolidated Edison Company (Coned), the licensee for Indian Point Station, Unit 2, reported in LER 95-005-00 that paint was peeling off the floor at the 14-meter (46-ft) elevation of the Indian Point Unit 2 containment structure. The paint was applied to the 14-meter (46-foot) floor elevation during the 1993 refueling outage as an interim measure for ! reducing personnel radiation exposures until a more permanent floor resurfacing could be accomplished. Coned determined that the following factors contributed to the cracking and delsmination of the paint: (1) in some areas, the paint had been applied in excess of the dry film thickness recommended by the manufacturer of the paint; (2) during preparation of the paint, too much paint thinner was added to the paint, which led to an excessive amount of coating shrinkage when the paint dried; (3) no scarification of the floor surface was performed before application of the paint to remove old coatings, greases, or silicone or wax buildups from the floor surface; and (4) the painters had not been trained to apply the particular brand of paint. Coned determined the root cause of the coatings event to be the painters' failure to follow controlled procedures for applying the particular brand of paint. To address the nonconforming condition of the paint, Coned removed all of the old paint from the 14-m (46-foot) floor elevation and repainted the floor elevation with a qualified coating in accordance with the station's
- All of the unqualified paint within the containment sump's zone of influence was removed, with the exception of approximately 1 m2 (12 ftd of unqualified paint applied to small components, such as lighting fixtures or name tags.
D-1
procedural requirements and the manufacturer's recommendations for the paint. Coned also retrained the paint specialists to reindoctrinate them regarding the importance of complying with the station's procedures and standards for coating applications. On October 18,1993, the Tennessee Valley Authority (TVA) reported in LER 93-026 the use of unidentified coatings on the surfaces of the No. 4 reactor coolant pump (RCP) motor housings at the Sequoyah Nuclear Plant, Units 1 and 2. These coatings were not accounted for in the licensee's QA Uncontrolled Costings Log. TVA determined that the No. 4 RCP motor housings were completely within the zones ofinfluence of the containment sumps at both Sequoyah units. The unqualified coating on each No. 4 RCP motor housing amounted to an additional 13.3 m2 (143 ft'); this amount was not accounted for by TVA in its 1986 assessment of unqualified coatings on the RCP motor i housings. The omission is significant because the maximum amount of uncontrolled coatings allowed by the Uncontrolled Coatings Logs for the Sequoyah units is 5.3 m2 (56.5 ft 2); this is the maximum amount of uncontrolled coatings that can be in the zone of influence of the i containment sump without having the potential to affect the operability of the ECCS and safety-related CSS.
)
The NRC summarized its review of the safety significance of the amount of unqualified paint on the No. 4 RCP motor housings in Inspection Reports (IR) Nos. 50-327/93-42 and 50-328/93-42 and in IR Nos. 50-327/94-25 and 50-328/94-25, dated November 9,1993, and September 12,1994, respectively. In IR Nos. 50-327/94-25 and 50-328/94-25, the NRC concluded that if the unqualified coatings on or within the RCP motor housings failed, they could potentially migrate to the containment sump during a postulated DB LOCA and impair the performance of the containment ECCS and the containment spray system during the event. TVA addressed this issue by modifying the RCP motor housings to include " catch" screens designed to prevent coating material on the motor housings from reaching the strainers in the containment sumps. I On July 2,1993, and September 11,1995, the Pennsylvania Power and Light Company (PP&L) issued LERs 93-007-00 and 93-007-01, respectively, to summarize its reassessment of ECCS performance at Susquehanna Steam Electric Station, Units 1 and 2, during a postulated DB LOCA. In its initial analysis of ECCS performance during a postulated DB LOCA, PP&L i determined that sources of fibrous insulating materials could not impair the operability of the ECCS at Susquehanna Units 1 and 2. However, PP&L's initial analysis did not account for " unqualified" coatings as potential sources of debris. In LER 93-007-00, PP&L discussed the effect of debris on the performance of the ECCS during a postulated DB LOCA. In the LER, PP&L stated that its increased awareness of the quantity of unqualified coatings and corrosion products ("other material") inside the containment was a key factor in deciding to reassess the sources of debris inside the Susquehanna Units 1 and 2 containments during a postulated DB LOCA. PP&L considered fibrous insulation material, unqualified coatings, and corrosion products as the sources of debris. PP&L's evaluation of the debris during the postulated event contained the following uncertainties: (1) uncertainty in qualify., J the sources of debris within the containment, (2) uncertainty in determining the amount of debris that could be dislodged during a postulated DB LOCA, and l D-2
j l (3) uncertainty in establishing exactly how the debris would be transported from its source to the ECCS strainers during the postulated event. Because of these uncertainties, PP&L stated in the licensee event report that if unqualified coatings and corrosion products were included among the materials that could become sources of debris, some potential existed for complete blockage of the suppression pool strainers during the event. PP&L addressed this issue, in part, by requiring that DE LOCA qualification testing be performed on all inorganic zinc paints inside the Susquehanna containments. PP&L also implemented improved administrative housekeeping and inventory controls and issued an administrative coating specification that restricted any coatings applied inside the containment structures to qualified coatings. On April 16,1997, the licensee for Millstone Nuclear Power Station Unit 1, a BWR-3 with a Mark I containment, reported to the NRC that a significant amount of coating work inside the Millstone Unit 1 torus (suppression pool) was unqualified. Millstone Unit 1 LER 97-026 stated that a number of different coating materials had been used inside the torus, but the locations and extent of various coating systems was unclear. On July 15,1997, the licensee for Clinton Nuclear Power Station, a BWR4 with a Mark lli containment, reported to the NRC that a significant quantity of degraded protective coatings was j removed from the primary containment and the drywell. The licensee stated that due to the 1 indeterminate condition of these degraded coatings, reasonable assurance could not be given that they would not have disbonded from their substrates enough to clog the ECCS suction I strainers during accident conditions. D-3
Appendix E GENERIC COMMUNICATIONS ISSUED BY THE NRC ON THE SUBJECT OF ECCS AND SAFETY-RELATED CSS SUMP AND STRAINER BLOCKAGE Generic Letter 85-22," Potential for Loss of Post Loss of Coolant Accident Recirculation Capability Due to insulation Debris Blockage," December 3,1985. IN 88-28, " Potential for Loss of Post Loss of Coolant Accident Recirculation Capability Due to insulation Debris Blockage," May 19,1988. IN 89-77, " Debris in Containment Emergency Sumps and Incorrect Screen Configurations," November 21,1989. IN 92-71, " Partial Blockage of Suppression Pool Strainers at a Foreign BWR," September 30,1992. IN 92-85, " Potential Failures c,f Emergency Core Cooling Systems by Foreign Material Blockage," December 23,1992. IN 93-34, " Potential for Loss of Emergency Core Cooling Function Due to a Combination of Operational and Post Loss of Coolant Accident Debris in Containment," April 26,1993. IN 93-34, Supplement 1, " Potential for Loss of Emergency Cooling Function Due to a Combination of Operational and Post Loss of Coolant Accident Debris in Containment," May 6,1993. Bulletin 93-02, " Debris Plugging of Emergency Core Cooling Suction Strainers," May 11,1993. NRC Bulletin 93-02, Supplement 1, " Debris Plugging of Emergency Core Cooling Suction Strainers," February 18,1994. IN 94-57, " Debris in Containment end the Residual Heat Removal System," August ' 12,1994. IN 95-06, " Potential Blockage of Safety Related Strainers by Material Brought inside
' Containment," January 25,1995.
IN 95-47, " Unexpected Opening of a Safety / Relief Valve and Complications involving Suppression Pool Cooling Strainer Blockage," October 4,1995. t E-1 I
r Bulletin 95-02, " Unexpected Clogging of a Residual Heat Removal (RHR) Pump Strainer While Operating in the Suppression Pool Coo!ing Mode," October 17,1995. IN 95-47, Revision 1," Unexpected Opening of a Safety / Relief Valve and Complications involving Suppression Pool Cooling Strainer Blockage," November 30, i 1995. IN 96-10, " Potential Blockage by Debris of Safety System Piping Which is Not Used During Normal Operation or Tested During Surveillances," February 13,1996. Bulletin 96-03, " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling Water Reactors," May 6,1996. j IN 96-27, " Potential Clogging of High Pressure Safety injection Throttle Valves During Recirculation," May 1,1996. IN 96-55, " inadequate Net Positive Suction Head of Emergency Core Cooling and Containment Heat Removal Pumps Under Design Basis Accident Conditions," ; October 22,1996. 1 IN 96-59," Potential Degradation of Post Loss of Coolant Accident Recirculation Capability as a Result of Debris," October 30,1996 IN 97-13, " Deficient Conditions Associated Wdh Protective Coatings at Nucisar Power Plants", March 24,1997. i E-2
Appendix F ENFORCEMENT ACTIONS TAKEN BY THE NRC DEALING WITH CONSTRUCTION AND PROTECTIVE COATINGS DEFICIENCIES AND FOREIGN MATERIAL EXCLUSION Pt. ANT DATE OF SEVERITY DESCRIPTION INSPECTION LEVEL / CIVIL PENALTY Surry Unit 1 7/30/88 3 Debris in containment
$50,000 sump Trojan 8/8/89 2 Inoperable recirculation
$280,000 sump Diablo Canyon 12/8/89 3 o Gaps in sump
$50,000 screens e Opening sump access hatches when sump operability is required e Debris in sump Perry 6/23/93 3 Clogged RHR strainers
$200,000 Arkansas Nuclear 10/25/93 3 Degradation of One $0 containment sump Unit 1 screens Browns Ferry 5/17/94 4 Unqualified protective Unit 1 $0 coatings applied to safety /reliefvalve discharge quenchers Point Beach 10/12/92 3 Foreign materialin Unit 2 _
$75,000 containment spray Sequoyah 9/3/94 4 Unqualified coatings on Units 1 and 2 $0 RCP motor stand F-1 m - _ _ . _ . . . . . _
PLANT DATE OF SEVERITY DESCRIPTION INSPECTION LEVEU CML PENALTY Nine Mile Point 4/10/97* 3 Debris in suppression Unit 2 $200,000 pool and downcomers Limerick Unit 1 7/3/96 3 RHR strainer was
$0 clogged, resulting in pump cavitation due to inadequate controls on suppression pool cleanliness H.B. Robinson 9/4/92 - 3 Foreign material
$50,000 exclusion controls inadequate during a modification of the RHR system LaSalle Unit 2 10/16/96 3 Foreign material found
$50,000* in downcomers from drywell to suppression pool and in the suppression pool.
Nine Mile Point 10/14/96 3 Foreign material found Unit 2 $50,000* in downcomers from drywell to suppression pool. St. Lucie Unit 2 5/18/97 4 Gaps in screen dividing
$0 the sumps for each train. Boots not installed at sump penetrations to keep debris from l entering the sump j through penetrations. i Sequoyah 9/3/94 4 Unqualified coatings on Unit 1 $0 the RCP motor stand
- Included with other enforcement actions !
l i F-2
PROPOSED GENERIC LETTER TITLED
" LOSS OF COOLANT INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS OF EMERGENCY MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION."
(CRGR Meeting No. 314 - January 30,1998) TOPIC Staff request for CRGR review and endorsement of the proposed generic letter titled, ' Loss of Coolant inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition.' HISTORIC PERSPECTIVE The original staff proposal was reviewed at CRGR Meeting No. 291, held on September 11, 1996, and was endorsed for issuance, without public comments, subject to, among others, one major comment -in addition to the 50.54(f) Information request, the staff also invoke a " conditional" compliance exception to the backfit rule. Specific reference to Appendix B was recommended. Attachment 4-A contains a summary of the Committee's specific recommendations. The Committee's objective was to make it a one-step process - ask the addressees to submit the required information. gad to require I;censees to resolve the issue, where appropriate, in compliance with 10 CFR Appendix B. However, the generic letter was published for a 30-day comment period, and revised to incorporate changes recommended by CRGR, the EDO and the Commission staff. BACKGROUND (1) Memorandum dated January 14,1998 from F. J. Miraglia to T. T. Martin, requesting CRGR review and endorsement of the proposed generic letter titled, " Loss of Reactor Coolant inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition"(CRGR ltem No.178)
- a. Proposed Generic Letter
- b. CRGR Charter Review Package
- c. Public Comment Resolution (2) E-mail from R. Tripathi to the CRGR members, dated January 28,1998, forwarding the draft issue Sheet.
ATTACHMENT 4
. . .. l
2 (3) SRM on SECY-96-231, dated January 22,1997, approving the proposed generic letter for l issuance for comments. ; I (4) FederalRegisterNotice, Vol. 62, No. 31,7.075-7077. ! (5) " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994," , AEOD/S95-01, dated March 1995. Items (3), (4) and (5) were hand-carried to the headquarters members on January 29,1998. (6) E-mail from R. Tripathi to the CRGR members, dated January 29,1998, forwarding the finallasue Sheet. Attachment 4-B contains the presentation material used by the staff. ISSUES, CONCERNS AND RECOMMENDATIONS This generic letter had been revised to address public comments. The version of the generic letter submitted for this round of CRGR review, had retumed to the original stand - a 50.54(f) Information request without invoking compliance exception. The staff indicated that the plants with design vulnerabilities will be dealt with on a case-by-case basis. The CRGR endorsement was contingent upon some comments which the staff agreed to address. Subsequently, the revised draft submitted by the staff, with a Division Director level consensus, on February 10, 1998, appeared to have addressed the CRGR comments (Attachment 4-C). The CRGR endorsement was relayed to the staff on February 12,1998.
I l l CRGR RECOMMENDATIONS ON THE PROPOSED GENERIC LETTER PRIOR TO THE ISSUANCE FOR COMMENTS (CRGR Meeting No. 291) l l
"...The Committee recommended that the draft generic letter be revised (1) to include explicitly (in addition to the analyses and information already requested in the draft letter) a provision that licensees are expected to take appropriate corrective action in accordance with 10 CFR 50, Appendix B, if the evaluation requested identifies a susceptibility to common cause failure resulting from events similar to those described in this generic letter (in order to ensure compliance with GDC 34 and 35, as applicable); (2) to reflect (in the "Backfit Discussion" section) that the generic letter includes both a request for information and backfitting (if corrective action must be taken to address common cause susceptibilities identified), which are justified under the provisions of 10 CFR 50.54(f) and 10 CFR 50.109(a)(4)(l) in order to verify / ensure compliance with applicable regulations; and (3) to follow more closely the approved format for this type of generic communication, in order to delineate more clearly the specific actions that are being requested, the specific information (and submittal schedule) being requested, the required response (and submittal schedule) to be provided by the licensee..."
)
l 1 I I ATTACHMENT 4-A
CR G J2 36 tjas;98 AMa du w J # f LOSS OF REACTOR COOLANT INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS OF EMERGENCY MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION WOLF CREEK EVENT OF SEPT.17, 1994 REACTOR COOLANT SYSTEM (RCS) BLOWDOWN THROUGH THE RESIDUAL HEAT REMOVAL (RHR) SYSTEM TO THE REFUELING WATER STORAGE TANK (RWST).
- APPROXIMATELY 9,000 GALLONS OF RCS INVENTORY BLOWN DOWN IN ABOUT 1 MINUTE.
POTENTIAL FOR 90% VOIDING IN RWST HEADER THAT IS SUCTION FOR ALL EMERGENCY CORE COOLING SYSTEM (ECCS) PUMPS IF BLOWDOWN IS NOT TERMINATED IN 3-5 MINUTES. MITIGATION OF AN EXTENDED BLOWDOWN MAY BE DIFFICULT BECAUSE OF POTENTIAL COMMON-CAUSE FAILURE OF ALL ECCS PUMPS. COULD RESULT IN LOSS OF BOTH RHR & ECCS FUNCTIONS. CAUSE OVERLAPPING OPERATIONS AND MAINTENANCE ACTIVITIES.
- FAILURE TO FOLLOW PROCEDURE.
f. s SAFETY SIGNIFICANCE PWRs: . l COMMON-CAUSE FAILURE OF THE ECCS PUMPS. SEQUENCE INCLUDES CONTAINMENT BYPASS. RELATIVELY SHORT TIME AVAILABLE FOR OPERATOR ACTION. CONDITIONAL CORE DAMAGE PROBABILITY ESTIMATED IS 3.0E-3. AEOD CHARACTERIZED THE EVENT AS THE MOST SIGNIFICANT PRECURSOR EVENT OF 1994. l BWRs:
- NO COMMON SUCTION LINE.
- SUPPRESSION POOL SIZED TO ACCOMMODATE HEAT FROM BLOWDOWN.
1
*- NOT VULNERABLE TO COMMON-CAUSE FAILURE. l l
l
y . HISTORY STAFF ISSUED IN 95-03, ALERTING LICENSEES TO CAUSES OF EVENT & POTENTIAL VULNERABILITIES. AEOD RELEASED REPORT (MARCH 1995) DESCRIBING WOLF CREEK EVENT & CATALOGING A TOTAL OF 19 EVENTS INVOLVING LOSS-OF-COOLANT WHILE ON RHR COOLING. SUPPLEMENT 1 TO IN 95-03 ISSUED (MARCH 26, 1997) PROVIDING FURTHER INFORMATION ABOUT EVENT & DISCUSSING AEOD REPORT. STAFF HAS BEEN WORKING ON GL TO REQUEST INFORMATION FROM LICENSEES WHO MAY BE VULNERABLE TO SIMILAR TYPES OF EVENTS: 0 INITIAL VERSION REQUESTED INFORMATION UNDER 10 CFR
- 50. 54 i,f) TO DETERMINE VULNERABILITIES, WITH POSSIBLE FUTURE ACTIONS NECESSARY TO CORRECT PROBLEMS WHERE IDENTIFIED; REGULATORY BASES CITED WERE GDC 34 & 35 REQUIREMENTS FOR RHR & ECCS PERFORMANCE. j 1
0 AFTER DISCUSSION WITH CRGR (SEPT.1996), STAFF MODIFIED l GL TO REQUEST THAT LICENSEES CORRECT PROBLEMS WHEN IDENTIFIED & NOTIFY NRC WHAT HAS BEEN DONE (BACKFITS AS NECESSARY). O COMMISSION DIRECTED STAFF TO MODIFY REFERENCES TO GDC 34 & 35, AND PUBLISH DRAFT GL IN FEDERAL REGISTER FOR ) PUBLIC COMMENT; GL PUBLISHED IN FEBRUARY / MARCH 1997. l
g . HISTORY (CONT' D) l l PUBLIC COMMENTS FROM INDUSTRY TOOK EXCEPTION TO:
- 1) CITATION OF GDC 34 & 35 AS REGULATORY BASIS, AND COMPLIANCE BACKFITS REQUIRED ON THOSE BASES.
- 2) LACK OF REGULATORY ANALYSIS JUSTIFYING BACKFITS.
- 3) EXCESSIVE BURDEN ON LICENSEES. !
l STAFF REVISED GL IN RESPONSE TO PUBLIC COMMENTS. l ACRS WAS BRIEFED ON 11/6/97. ACRS LETTER RECOMMENDED PROMPT ISSUANCE OF CURRENT VERSION OF GL.
'I 1
1
m 1
~ - l 1
CURRENT VERSION OF GL FOCUS OF GL SHIFTED FROM DESIGN ISSUE TO OPERATIONAL ISSUE.
"INFORMATION ONLY" FORMAT, UNDER 10 CFR 50.54(f).
- REGULATORY BASIS: 10 CFR 50, APPENDIX B, CRITERION V, " INSTRUCTIONS, PROCEDURES, AND DRAWINGS," STATES: l 0 "Ac.IVITIES AFFECTING QUALITY SHALL BE PRESCRIBED BY L. ' JMENTED INSTRUCTIONS, PROCEDURES, OR DRAWINGS OF A TYPE APPROPRIATE TO THE CIRCUMSTANCES AND SHALL BE ACCOMPLISHED IN ACCORDANCE WITH THESE INSTRUCTIONS, PROCEDURES, OR DRAWINGS."
l I 1
~ ~ l SPECIFIC INFORMATION REQUESTED i
DESCRIPTION OF ECCS FEATURES THAT COULD RENDER' SYSTEM VULNERABLE TO COMMON-CAUSE FAILURES, AS OCCURRED AT WOLF CREEK (E.G., COMMON SUCTION HEADER). ' DESCRIPTION OF FEATURES OF APPENDIX B QA PROGRAM, WITH CONSIDERATION OF PLANT-SPECIFIC DESIGN FEATURES, THAT PROVIDE l ASSURANCE THAT SAFETY-RELATED FUNCTIONS OF RHR AND ECC SYSTEMS WILL NOT BE ADVERSELY AFFECTED BY ACTIVITIES AT HOT SHUTDOWN. INCLUDES ADEQUACY OF TRAINING FOR CARRYING OUT THESE TYPES OF EVOLUTIONS, INDEPENDENT VERIFICATION OF VALVE POSITIONS, & CONTROLS IN PLACE TO ASSURE COMPLIANCE WITH PLANT OPERATING PROCEDURES. l I l 1 D s es e =4 e mo+ me e me s . wgra . n w w ~ - q 4
OTHER MODIFICATIONS
*. SCOPE NARROWED TO FOCUS ON PLANTS WITH:
- 1) COMMON SUCTION HEADER, AND
- 2) TWO OR FEWER ISOLATION VALVES IN LINE, AND
- 3) PIPE DIAMETER 2 INCHES OR MORE, AND
- 4) MODE 4 OPERATION.
ll l
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y n -. ~ . ,. ~ . .n- n. ,e. , ww e aw+.- e e. m a - - o- , , - ~ '
~
HOW DOES STAFF PLAN TO UTILIZE THE REQUESTED INFORMATION TO CLASSIFY PLANTS IN EITHER OF THE FOLLOWING TWO CATEGORIES: CATEGORY 1: DESIGN FEATURES (HARDWARE MODIFICATIONS) ALONE WILL PREVENT WOLF CREEK LIKE EVENT. NO FURTHER ACTION PLANNED. CATEGORY 2: ADMINISTRATIVE CONTROLS ARE NEEDED TO PREVENT SUCH EVENTS. CATEGORY 2 PLANTS WILL REQUIRE FURTHER STAFF EVALUATION TO DETERMINE ADEQUACY OF ADMINISTRATIVE CONTROLS IN PLACE. ASSESS NEED FOR FOLLOW UP ACTIONS. 9 l 'r "f
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UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 2055G-0001 January xx,1998 NRC GENERIC LETTER 97-xx: LOSS OF REACTOR COOLANT INVENTORY AND ASSOCl-ATED POTENTIAL FOR LOSS OF EMERGENCY MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION Addressees All holders of operating licenses for pressurized-water reactors (PWRs), except those who have permanently ceased operations, and have certified that fuel has been permanently removed from the reactor vessel. Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to request that addressees (1) assess the susceptibility of their [ residual. heat removal (RHR) and) emergency core cooling [(ECC)] systems {(E000;)} to common-cause failure as a result of reactor coolant system (RCS) draindown while in a shutdown condition, and (2) submit certain information, pursuant to Section 50.54(f) of Title 10 of the Code of Federa/ Regulations (10 CFR 50.54(f)), concoming their findings regarding potential pathways for inadv6rtent RCS drain <fown and the , suitability of {enf .. %a ea;re',} [ surveillance,' maintenance,~. modification and) operating I practices and {mdatenere;} procedures [regarding configuration control] during reac'.or shutdown cooling. fftwe} [The requested] information will enable NRC staff to verify whether addressees comply with NRC regulatory requirements and conform with current licensing bases for their facilities, {;ae'ading ta ed;bll;hm:n; c', end anded c'};[with regard to prescribing and accomplishing] activities effecting quality {eward;ng to decumented i,reedu ;;,} per Criterion V of Appendix B to 10 CFR Part 50. The staff is specifically concemed about@ qu;T.ty an;.e ;f edia:O; (}[ addressees' controls over the conduct of activities during hot shutdown conditions that _may affect the_ safety-related functions of the RHR system.and.the ECCS,) for example, the methods used to verify valve position, the controls in place to assure compliance with {te g/ed }[ plant surveillance, maintenanceimodification'and] operating procedures, and the adequacy of operator training for such activitie:{} waduced dur:ng hd
;hd%en sad l%a; "ed. .s 2; ; !di rd;;;d fund;en; ;' Oe r;;; dud h;d remenl ("J C)
;,0;m end th; 5000, ;; d;O;d la 10 0. "R" . er; 50, /ggadi /., 0; nerd 0;dia OTL-;e (000) 04 end 05, rapdt;dy}. All addressees are required to submit a written response to the (NRG)[ requested,information,] relative to {2; requc ded lafe.ne%a and} the requirements of theirlicenses.
ATTACHMENT 4-C
Generic Letter 97-xx Month, Day, Year Page 2 of 6 Discuss 6on The NRC issued information Notice (IN) 95-03, " Loss of Reactor Coolant inventory and Potential Loss of Emergency Mitigation Functions While in a Shutdown Condition," on January 12,1995, to alert addressees to an incident at the Wolf Creek Plant involving the loss of reactor coolant inventory while the reactor was in a hot shutdown condition. In that event, operators were attempting to reborate RHR train B, while at the same time maintenance personnel were repacking an RHR train A-to-train B crossover isolation valve. Train B is reborated by recirculating water through a loop that contains the RHR system piping, the refueling water storage tank (RWST), a containment spray pump, a manual RWST isolation valve, and an RHR system crossover line. When the RWST isolation valve was opened for the reboration process and the train A-to-train B crossover isolation valve was opened for stroke testing, a drain-down path was inadvertently created from the RCS to the RWST. At Wolf Creek, all {CCOO} [RHR.and ECC_ system) pump suction lines are tied into a common suction header. When the draindown event occurred, hot RCS water was introduced into this common suction header between the RWST and the {EOOO} [RHR and.ECC system] pumps. , This hot water flashed to steam, resulting in a steam / water mixture in the header. {'n ic cant l ef)[Had) an ECCS actuation [ occurred), this mixture would have been introduced into the i suction of the ECCS pumps. If operators had not been able to terminate the event, the hot I water in the RWST suction piping might have led to steam binding, which could have [ adversely] I affected-{ ell)[the] pumps in both ECCS trains. In addition, water flashing to steam in the header and the RWST could have caused serious mechanical damage to the RHR piping and the RWST as a result of water hammer. Finally, steaming through the RWST establishes a l containment bypass path. , The licensee estimated (using actual plant conditions) that for an unmitigated event, the reactor vessel water level could have drained to the bottom of the hot leg within 5 minutes and, as a j consequence, RHR pump A would have lost suction, cavitated, and failed. Shortly thereafter, j the common ECCS suction header could have reached a 90-percent steam / water ratio. The licensee also estimated that continued boil-off could have caused the pressure vessel water level to drop to the point of core uncovery in less than 1 hour. Events of this nature are considered particularly significant because they can result in loss of emergency core cooling capability and involve the potential for containment bypass. On March 25,1996, the staff issued a supplement to IN 95-03 that further analyzed the event. The NRC has also issued a number of other communications describing events at reactor facilities involving inadvertent loss of reactor coolant inventory while the reactor was in a shutdown condition. The Office for Analysis and Evaluation of Operational Data (AEOD) published AEOD/E704, " Discharge of Primary Coolant Outside of Containment at PWRs While on RHR Cooling," in March 1987, which documented six events involving RCS backflow into the RWST. In Generic Letter 88-17, " Loss of Decay Heat Removal (DHR) 10 CFR 50.54(f)," dated October 17,1988, the NRC requested several actions to address loss-of-DHR events that occurred while reactors were in a shutdown condition. In IN 91-42, " Plant Outage Events involving Poor Coordination Between Operations and Maintenance Personnel During Valve Testing and
Generic Letter g7-xx Month, Day, Year Page 3 of 6 Manipulations," dated June 27,1991, the NRC discussed inadvertent loss-of-inventory events. The AEOD report " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994," (AEOD/S95-01), dated March 1995, noted 19 events in which RCS water was transferred to the RWST. These events were primarily caused by personnel errors, poor coordination between operations and maintenance personnel, and {;n;." dent} } inadequate) procedures associated with the operation of the RHR system in the shutdown cooling mode. The personnel errors were primarily caused by inattention or lack of training; while the procedural deficiencies were related to omissions or lack of specificity in sequential valve operations when conducting tests on the RHR system. On the basis of this history and the potential for containment bypass, the staff has concluded that additional information is required to confirm the adequacy of existing configuration control, operating practices, and training for assuring the safety function capability of the i;n:da: hn: rein;;;l} [RHR] and ECC systems. Reauired Information Within 120 days of the date of this generic letter, addressees are required to submit a written response that includes the following information: (1) describe whether your {EOO ;y;;;.T. irdds) ,[ emergency. core cooling systems include] a feature such as a common suction header (which can render the {TOOO} [ systems) susceptible to common-cause failure, as a result of events similar to the Wolf Creek RCS drain-down event of September 17,1994); and (2) describe, with consideration of plant-specific design attributes, the features of your Appendix B quality assurance program (for example, the methods used to verify valve position, the controls in place to assure compliance with {the p ;n; }[ plant surveillance, maintenance, modification and] operating procedures, and the adequacy of operator training for such activities) that provide assurance that the safety-related functions of the RHR system and ECCS will not be adversely affected by activities conducted at hot shutdown (such as occurred at Wolf Creek). Addressees may limit their attention to thosel[ surveillance,) maintenance [, modification) and operational activities at hot shutdown during which it is feasible to divert RCS fluid to the RWST, resQing in simultaneous drain-down of the RCS and voiding in the {EOOO} suction header [for the RHR and ECC system pumps). Addressees may further limit their {;;;enfa to )[ response to the consideration of potential] configurations and conditions that {re .11
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-... ;rdd; *'OOE 4 eper;;;x; r;;;ts:r.; p;ter.;;;: few pethe thet lrd.d;};[ involve flow paths with] pipe diameters equal to or greater than 2 inche;{, ad ;;;;h te;; ;; ';;;r recTit,-;'end ;; ten}.
Addressees shall submit the required written responses, pursuant to 10 CFR 50.54(f) and 10 CFR 50.4, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, signed under oath or affirmation under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, with a copy to the appropriate regional administrator and the appropnate NRC resident inspector. Backfit Discussion This generic letter only requests information from the addressees under the provisions of-{40 OTC '^.W; f de;; ne;:n;;ts; my te1' :;ng}.[Soction 1.82a..of the Atomic Energy Act of
Generic Letter 97-xx Month, Day, Year Page 4 of 6
!I954, as amended, and .10 CFR 50.54(f), to verify addressee compliance.with the Commission's regulations and conformance with the current licensing-basis of their respectwe facilities relative to the safety-related functions of the RHR and ECC systems, including 10 CFR 50.46 and General Design Criteria 34 and 35 of Appendix A to 10.CF_R Part 50,;as appropriate, and the foquirements of Appendix B to 10 CFR Part 50. With respect to Appendix B to 10 CFR Part 50, the requested information will enable the NRC staff to determine whether adequate. control is being exercised over surveillance, maintenance, modification and operational actwitnes conducted at hot shutdown which can adversely affect the. safety-related functions of the RHR and ECC systems.1 No backfit is either. intended or approved in the context of issuance of this Beneric letter). Therefore, the staff has not performed a backfit analysis.
{!n vb; c' N "/d' O =h didadaan ;;st, th ; :n'ermd:en l; naded te vefi unace' eemp%nce ;;nh NCO regulderi r;qubmente end current uslng bus fer N 7 fed l:s a; rdd;d te h requhment; c' Odeden V c' Appadi S t; 10 Oin Ier: 5^, epd';;dly a; regard; h qudny wat.e: c' edidh ;;h eh an advernly ;"ed % sidia;';kd 'undene of th; nlln end ECO ;yd;me, J.hea aquhmerde ere de'.ned ln 10 Orn Pert 50, Appadi A. 000 34 ad 05, ad in 10 Orn 50.40.} [ Quality. Assurance) Criterion V of Appendix B to 10 CFR Part 50 requires that " activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings." Furthermore, licensees' {'eehn:cd Opd';;d:en; ('O)} ;[ technic 61 specifications) include requirements to {mirada RllC end E00 ;yd;m; egreLle du.ing cpre :s; ;n MODE 4} [ establish,' implement and_ maintain written administrative procedures to address startup, operation and shutdown of a shutdown cooling system). Maintenance and testing activities at Wolf Creek during hot shutdown were carried out contrary to documented procedures and the technical specifications, resulting in {the) RCS drain-down and the potential for common-cause failure of the {EOOO}.[RHR and ECC eystem) pumps, which could have compromised the ability of the RHR and ECC systems to fulfill {the)[their] safety functions {;pd';;d in 000 04 ad 05, reepd v;ly}. Furthermore, the staff has determined that cimilar {nera; rd;^;d ic} loss of-coolant [ events) while on RHR cooling have occurred at over 19 plant;{, me lng ' rem fdlure t; }llTThese' events were due to the failure on the part_of licensees to.either) establish adequate procedures or {fdlum ic} follow procedures and applicable technied specifications. Both of these condition;{, i.e., fdlure ta edebOh precedure; and fdlum to fd: eve praedure; end epptble teda:eel ; ped';;hne, anddut; les of cemp:nx} [ involve noncompliance) with the requirements of Criterion V of Appendix B to 10 CFR Part 50, and, {eensqu;rdly} [therefore), non-compliance with (M .nn'} [the) current licensing {besee) [ basis for,a. facility). Since, a relatively largefrechen) [ number) of the operating PWRs have experienced similar events, the staff believes that additionalinformation is required to confirm the adequacy of existing configuration control [ practices), operating practices, and training for assuring the safety function capability of the {rs due: bd re as;l} [RHR] and ECC systems. In accordance with the provisions of 10 CFR 50.54(f), {h desme;td adud:en preer;Md in Oed a 50.54(') van ad prepred hau;e N aquhd :n'e md:s i; nes:::ri to vef xmp%nce f v;Rh Odeden V c' 10 OIR rest 50, AppadL S} [an~ approved _ evaluation of the rationale for the information request contained herein is.not a prerequisite to issuance of the. generic letter.because,the_information being l
" * *V & 9 W
Generic Letter 97-xx Month, Day, Year Page 5 of 6 pequested is needed by the 'NRC staff to verify addressee compliance with theOurrent licensing l >ases of their respective. facilities). FederalRegister Nottfication A notice of opportunity for public comment was published in the FederaiRegister(62 FR 7075) on February 14,1997. Comments were received from four nuclear utility companies, the Nuclear Energy Institute, and the Nuclear Utility Backfitting and Reform Group. The staff's evaluation of the comments is available from the NRC Public Document Room. The generic letter has been appropriately revised to reflect the comments received. Paoerwork 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 mandatory information collection is estimated to average XXX hours per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information collection. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collections contained in the generic letter and on the following issues:
- 1. Is the proposed information collection 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 information collection be minimized, inch. ding the use of automated collection techniques?
Send comments on any aspect of this information collection, including suggestions for reducing the burden, to the Information and Records Management Branch (T 6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet electronic mail at BJS1@NRC. GOV; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management and Budget, Washington, DC 20503. Public Protection Notification if an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
Generic Lett er 97-xx l Month, Day, Year Page 6 of 6 l _, [tf you have any questions about this matter, please contact the technical contact listed below or the_ appropriate Office. of. Nuclear Re_ actor Regulation (NRR) project manager.) l l Jack W. Roe, Acting Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contact: M. M. Razzaque, NRR (301) 415-2882 E-mail: mmr1@nrc. gov Lead Project Manager: Kristine Thomas, NRR l (301) 415-1362 ' E-mail: kmt@nrc. gov l
Attachment:
List of Recently issued NRC Generic Letters
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i I 1 William D. Travers W1 15, 1m process - ask the addressees to submit the required information.30d to require licensees to resolve the issue, where appropriate, in compliance with 10 CFR Appendix B. However, th'e generic letter was published for a 30-day comment period, and revised to incorporate changes recommended by CRGR, the EDO and the Commission staff. The version of the generic letter submitted for this round of CRGR review, had retumed to the original stand - a 50.54(f) information request without invoking compliance exception. The staff inoicated that the plants with design vulnerabilities will be dealt with on a case by case basis. The CRGR endorsement was contingent upon some comments which the staff agreed to address. Attachment 4 contains details. In accordance with the EDO's July 18,1983 directive conceming " Feedback and Closure of CRGR Review", a written response is required from the cognizant office to report agreement or disagreement with the CRGR recommendations in these minutes. The response is to be forwarded to the CRGR Chairman and if there is disagreement with the CRGR recommendations, to the EDO for decision making. Questions concerning these meeting minutes should be referred to Raji Tripathi (415-7584). Attachments: As stated cc: Commission (5) SECY M. Knapp, DEDE F. Miraglia, DEDO J. Lieberman, OE H. Bell, OlG K. Cyr, OGC J. Larkins, ACRS H. Miller, R-l L. Reyes, R-Il J. Dyer, R-Ill E. Merschoff, R-IV C. Paperiello, NMSS A. Thadani, RES S. Collins, NRR W. Kane, NRR G. Holahan, NRR. D. Matthews, NRR
. Attachments: As stated
. Distribution (w/atts.):
s File Center PDR (NRC/CRGR) CRGR SF CRGR CF STreby Distribution (w/atts.) via e-mail.: JMitchell JShapaker MMitchell DAllison TCollins MRazzaque BElliot . ESullivan DISK / DOCUMENT NAME: C:\CRGR98\ MINUTES \ MINUTES.314 To rictive a copy of this document, inycate in the box: "C" = Copy w/o attachment, "E" = Copy w/ attachment, "N" = No copy OFC CRGR o y C:CPQ5lij NAME RTripat JI h DATE 4/l/99 4/[f99 ~ OFFIClAL RECORD COPY
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