ML20212E665
| ML20212E665 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/24/1997 |
| From: | Ray H SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-361-97-15, 50-362-97-15, NUDOCS 9711040015 | |
| Download: ML20212E665 (21) | |
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- M ik soulHI RN CATH okNtA C n l C g") liarold R. Rav I L. IJ l J \J Executive Vi[e President An tresov ixnawrow (ump.ar October 24,1997 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Docket Nos. 50-361 and 50-362 Reply to a Notice of Violation San Onofre Nuclear Generating Station, _Unita 2 and 3
References:
dated September 10,1997
dated July 22,1997
- 3) Meeting, NRC/SCE meeting on Maintenance Rule aspects of >
RCS nozzle PWSCC (Handouts), dated August 21,1997 i 4) Meeting, NRC Predecisional Enforcement Conference (Handouts),
l dated September 30,1997
dated October 3,1997 Reference 1 transmitted the results of NRC Inspection Report No. 50-361 and 50-362/97-15, which concerns an inspection conducted at the South 9rn California Edison l (SCE) San Oncfre Nuclear Generating Station, Units 2 and 3. The enclosure to the Reference 1 letter contained a Notice of Violation (362/971542) which states that,
, contrary to the requirements of 10 CFR 50.65, SCE failed to demonstrate that the i\
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_ condition of the reactor coolant system was being effectively contre' led 'brough the
{ ga.performance capable of performingof appropriate its intended preventive function. SCE does not agreemaintenance, or admit that this
'h such
/ tha violation occurred, as discussed below and in the enclosure to this letter.
S i
L ng_ The basis for our conclusion that no violation occurred is that the reactor coolant system demonstrably did rernaln capable of performing its intended function, l(I \
.4@2 notwithstanding the circumstances cited in the inspection report. The pre 3 maintenance performed to effectively control the condition of the reactor coolant system
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i Document Control Dask 2- October 24,1997 was in accordance with the guidance available from the NRC itself, and did assure that l l
the system remained capable of performing its intended function. Indeed, we continue to follow that guidance currently and believe it fully satisfies NRC requirements.
1 In order for us to conclude that we failed to demonstrate that the reactor coolant system remained capable of performing its intended function, we believe we would need to be able to identify corrective action, which, if taken, would achieve such demonstration.
We specifically do not believe that the establishment of goals in accordance with 10 CFR 50.05(a)(1), applicable to primary water stress corrosion cracking (PWSCC) of reactor coolant system Alloy 600 penetrations, would achieve this result, even if the establishment of such goals were practical. Thus, in the absence of the ability to set goals applicable to PWSCC of Alloy 600 penetrations which would achieve demonstration of intended system function, and given that we conclude that there was no loss of intended function resulting from the PWSCC of Alloy 600 penetrations, bm,ed upon the extensive inspection and preventive maintenance conducted which we believe meets and exceeds applicable NRC guidance, SCE concludes that no violation occurred.
An important consideration in reaching this conclusion is our understanding of the purpose of the requirement in Techtsical Specification 3.4.13 prohibiting leakage from the reactor coolant system pressure boundary. As noted in the bases for this Technical Specification, ieakage from joints and interfaces is anticipated during plant life, through either operational wear or mechanical deterioration. When even minute leakage is detected from the reactor coolant system pressure boundary, prompt shutdown and repair is required. However, total unidentified leakage of as much as 1 gpm is considered to not compromise safety. As amply described in the NRC guidance concerning PWSCC of Alloy 600 penetrations (which is described in Appendix A of the enclosure to this letter), inspection for evidence of leakage during plant shutdowns, and replacement or repair of penetrations found leaking, provides adequate assurance that there will be no loss of intended function of the reactor coolant systerr.. References (2) through (5) provide additional relevant information in this regard.
Finally, while we conclude that we continue to manage the consequences of PWSCC of Alloy 600 penetrations in full compliance with the Technical Specifications, and both NRC and industry guidance, we also independently conclude that the preventive maintenance we perform in this regard conservatively assures that the reactor coolant system remains capable of performing its intended function. Because of the operational inconvenience resulting from penetration leakage, we continue to aggressively examine strategies for replacement of penetrations prior to any leakage occurring, consistent with the need to maintain radiation exposure resulting from penetration replacement ALARA.
Document Control Desk .3 October 24,1997 j If you have any further questions, please contact me.
Sincerely, 4.
- d. I
Enclosure:
. NOV Reply to Violation & Appondix A ,
i cc: E. W. Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Director, Walnut Creek Fiald Office, NRC Region IV M.- B Fields, NRR Project Manager, San Onofre Units 2 and 3 J. A. Sloan, NRC Senior Resident inspector, San Onofre Units 2 and 3 i
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ENCLOSURE REPLY TO A NOTICE OF VIOLATION The enclosure to Mr. A. T. Howell's letter dated September 10,1997 states, in part:
"During an NRC inspection conducted on June 30 through September 2, 1997, one violation of NRC requirements was identified. In accordance with the ' General Statement of Policy and Procedure for NRC Enforcement Actions,' NUREG-1600, !
the violation is listed below:
"10 CFR 50.65(a)(1) states, in part, that each holder of a license to operate a nuclear plant shall monitor the perforrrance of structures, t ystems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, as defined in paragraph (b), are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety and, where practical, take into account industry-wide operatir.J experience.
"10 CFR 50.65(a)(2) states, in part, that monitoring as specified in paragraph (a)(1) is not required where it has been demonstrated that the performance or condition of a structure, system or compon6nt is being effectively controlled through the performanew of appropriate preventive rnalntenance, such that the structure, system or component remains capable of performing its intended function.
"10 CFR 50.6S(c) states that the requirements of this section shall be implemented by each licensee no later than July 10,1996.
' Contrary to the above, as of July 10,1996, the time when the licensee elected to not monitor the performance or condition of the reactor coolant system against licensee-establisned goals pursuant to the requiremen4 of Section (a)(1), the licensee failed in demonstrate that the condition of this system was being effectively controlled through the performance of appropriate preventive maintenance, such that the system remained capable of performing its intended function. Specifically, the licensee inadequately evaluated the appropriateness of the performance of preventive maintenance prior to placing the Unit 3 reactor coolant system under a 10 CFR 50.65(a)(2) category (i.e., the licensee did not consider in its evaluation the identification in 1995 of through-wall cracking in four reactor coolant system nozzle penetrations, which tepresented multiple failures of the barrier function of the reactor coolant system).
"This is a Severity Level IV violation (Supplement 1) (50-362/9715-02)."
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Enclosure
- 1. Reply to the Violation ,
1 For convenlonce in review of the response, the violation is addressed in three parts, as f follows: (Note: The response considers primary water stress corrosion cracking .
[PWSCC) of Alloy 600 reactor coolant system nozzles only, and this is not repeated l throughout the response.)
Part A of the Violation: SCE did not consider in its evaluation the identification in 1995 of through-wall cracking in "four' [ sic) reactor coolant system nozzle penetrations which represented multiple failures of the barrier function of the reactor coolant system.
(Note: SCE review of data indicates that there was evidence of through wall leakage in three instances; not four.)
Response to Part A: The ' barrier function' of the reactor coolant system is not a defined term. It can only be understood by reference to the Technical Specifications and to General Design Criterion (GDC) 14. The bases of Technical Specification 3.4.13 includes the following:
" Component joints are made by welding [ emphasis added), botting... During plant life the joint and valvo interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operationa! LEAKAGE LCO [1 gpm unidentified leakage) is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety.' ,
Section 1.1 of the Technical Specifications defines Pressure Boundary LEAKAGE as, ' LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall or vessel wall." Technical Specification 3.4.13 itself also provides that there shall be "No Prescure Boundary LEAKAGE".
t GDC 14 states that, "The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage [ emphasis added), of rapidly propagating failure, and of gross rupture'.
Taken together, SCE understands the foregoing to require that any pressure boundary leakage which is identified must be promptly corrected. SCE has not only complied with this requirement at all times, but has implemented extensive and aggressive inspection measures - often involving significant maintenance 2
Enclosure activities - to assure that any indicat:on of Pressure Boundary LEAKAGE at penetrations is identified during plant shutdowns.
Extensive NRC guidance has been issued concerning PWSCC of Alloy 600 penetrations. At no time has this guidance indicated that measure.1 other than inspection and repair during outages is needed, although this possibility has been addressed. (Please refer to Appendix A to this Enclosure.) SCE concludes that the condition resulting in this instance is not considered by the NRC to be a ' failure of the carrier function" of the reactor coolant system, as stated in Part A of the Violation. Moreover, SCE believes this to be reasonable, in that - as discussed in the NRC guidance at length - the penetration will remain intact and will continue to provide a substantial barrier against leakage from the reactor coolant system, such that it continues to perform its intended function as described in the bases for the Technical Specifications and GDC 14, Accordingly, consistent with the referenced NRC guidance, SCE does not consider identification of evidence of penetration leakage due to PWSCC of Alloy 600, which may be identified du,-'ng inspections conducted for the purpose of early detection of such leakage, to represent functional failures of the reactor coolant system under 10 CFR 50.65, notwithstanding the fact that the Technical Specifications do not permit continued operation until such leakage is repaired.
(We are required to evaluate structures, systems, or components (SSCs) against the established performance critoria using historical plant data, and industry data
- where applicable, to determine if the SSCs met the performance criteria, l Performance criteria for the reactor coolant system consist of functional failures and system availability.)
Thus, the fact that SCE did act consider the 1995 instances of penetration leakage prior to placing the Unit 3 reactor coolent system under a 10 CFR 50.65(a)(2) category as of July 10,1996, is not a violation of requirements because those instances were not considered functional failures of the reactor coolant system.
l Part B of the Violation: SCE inadequately evaluated the appropriateness of the performance of preventive maintenance prior to placing the Unit 3 reactor coolant system under a 10 CFR 50.65(a)(2) category.
Response to Part B: In accordance with the Regulatory Guide 1.160 section titled, l 'Use of Existing Licensee Programs,' and NUMARC 93-01, Section 7.0,
- Utilization l of Existing Programs", SCE used axisting program results to support the determination that reactor coolant system performanen was being effectively 3
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Enclosure controlled through preventive maintenance. That is, SCE's existing program, which is documented in Program Plan 90022," Susceptibility of Reactor Coolant System Alloy 600 Nozzles to Primary Water Stress Corrosion Cracking and Replacement Program," provided an adequate evaluation of the appropriateness of preventive maintenance in this instance, and there was no need for another evaluation to be performed. Thus, the lack of another evaluation is not a violation of NRC requirements.
Part C of the Violation: The licensee failed to demonstrate that the condition of the reactor coolant system was being effectively controlled through the performance of appropriate preventive maintenance, such that the system remained capable of performing its intended function.
Response to Part C: SCE's preventive maintenance program for reactor coolant system penetrations is implemented in its Alloy 600 program referenced above.
The progr am includes requirements for identifying inspection frequencies and corrective actions to be taken when indications of leakage are detected. Most importantly, the program is based on extensive industry and NRC evaluations of the significance of PWSCC of Alloy 600 penetrations on the capability of the reactor coolant system to perform its intended function. As documented in Appendix A to this Enclosure, these evaluations support the use of inspection and repair of leakage as providing assurance that the intended function will be maintained.
SCE has considered what alternative action it might have taken. The NRC inspection report and Notice of Violation indicate that the reactor coolant system should have been placed in category 10 CFR 50.65 (a)(1) on July 10,1996, such that the performance of the penetrations would have been monitored against goals which we would have established commensurate with safety and industry-wide operating experience. We believe that use of the existing program, whici;is fully permitted by Regulatory Guide 1.160, entirely satisfied this purpose of category (a)(1). Thus, SCE did demonstrate that the system was being effectively controlled through the performance of appropriate maintenance, as permitted by category (a)(2) no further goal-setting was needed, and the fact that the leakage of the penetrations did not result in placement of the system in category (a)(1) is not a violation of NRC requirements.
Finally, SCE has considered what additional licensee-established goals it might establish under 10 CFR 50.65 (a)(1) for the penetrations, consistent with safety (as evaluated by the NRC), industry experience, and Al. ARA. We cannot identify such goals, but we would welcome NRC guidance in this regard, applied on an industry-wide basis.
Enclosure
- 2. Actions Taken As noted in the inspection report, the per.etratio'1 replacements were accomplished under construction work orders. TI 3 failure history screening performed as part of the Maintenance Rulo imptomentation .?ior to July 10,1996, did not includo review of construction work orders. SCE has completed review of construction work orders implemented since July 1993, and dolormined that this was an isolated occurrence.
SCE policy is to proactively replace any penetrations that can reasonably be predicted to leak, prior to such a leak developing. In addition, in order to minimize i'uture impacts to operational reliability, SCE will implement strategies to replace penetrations over time which are considered more susceptible to leakage than others. However, ALARA considerations require that this be dono sebetively.
As a result of soveral recent problems, including steam generator tube leakage, steam generator manway gasket leakage, and a shutdown cooling valvo plug leak, the reactor coolant systems of both units have boon placed in category (a)(1). The Alloy 600 penetrations will continue to be managed under updates of the program referenced above, and this program will be referenced in other appropriate documents.
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APPENDIX A ALLOY 600 RCS PENETRATION NOZZLES:
NRC AND INDUSTRY PERSPECTIVES As noted in the chronology below, Alloy 600 RCS penetration nozz!a PWSCC first occurred at San Onofre in 1986. Alloy 600 RCS penetration nozzle PWSCC has occurred at numerous facilities and has been subsequently observed in RCS head penetrations, RCS Pressurizer penetrations, and finally in RCS piping (both hot leg and cold leg) penetrations.
From a review of the regulatory record, the limiting case, from a safety significance perspective, is considered to be the RCS head penetrations. This is due to their larger physical size (worst case in the event of catastrophic failure) and the difficulties in performing visualinspections due to interferences. Therefore, the RCS head penetrations have been and continue to be considered the primary focus, and the bounding case for accident and safety significance analyses.
As provided below, the NRC and industry's guidance on the safety significance of Alloy 600 RCS penetration PWSCC are consistent:
- PWSCC is a known phenomena. Alloy 600 RCS penetration nozzles are susceptible to PWSCC. There is no reliable predictive methodology for explicitly predicting individual nozzle susceptibility to PWSCC. PWSCC is a function of time, temperature, residual stress in the nozzle, weld, microstructure, and water chemistry.
- PWSCC in Alloy 600 RCS penetrations which are not roll expanded (EdF nozzles are roll expanded) results in only axial cracking due to the stresses involved.
- PWSCC axial cracks are short in length, and crack growth beyond the initial weld region is very slow since operating stresses in the region are low.
- Augmented visual inspections for cracks and indications (boric acid residue) are relied upon to identify PWSCC, and upon discovery repair and/or replacement is effected to the identified nozzle.
Appendix A The following six examples ara best illustrative of the NRC's independent review and guidan:e regarding Alloy 600 RCS penetration PWSCC:
Examp!31: January 1995 NRC Petition Denial D.D. 95 2 On January 26,1995, the Director, Office of Nuclear Reactor Regulation, denied a petition filed on behalf of Greenpeace international, to shutdown plants based on PWSCC. The NRC's denial states, in part:
"In 1990, the NRC Staff identified to the Commission primary water stress corrosion cracking (PWSCC) of Alloy 600 in components other than steam generator tubing as an emerging technical issue after cracking was noted in pressurizer heater sleeve penetrations at a domestic PWR facility. At that time, the Staff reviewed the safety significance of the cracking as well as the repair and replacement activities at the affected facility.
'The Staff determined that the safety significance of tne cracking was low because the cracks were axial, had a low growth rate, were in a material with an extremely high flaw tolerance (high fracture toughness) and, accordingly, were unlikely to propagate very far. These factors also demonstrate that any cracking would result in a detectable leak before a penetration broke."
" Based on the owners groups safety assessments, a leak in a VHP [ vessel head penetration) would be detected before significant damage could occur to the VHP or the reactor vessel. This would result in the deposition of boric acid crystals on the vessel head and surrounding area that would be detected during surveillance walkdowns. Consequently, the concerns raised by the Petitioner do not raise any immediate safety concerns... Immediate inspections are not required since there is no immediate safety concern.. .
"CEOG submitted the detailed findings of it's Alloy 600 component PWSCC program initiated in 1990 to the Staff in a proprietary report on February 26, 1992. The conclusions of the report, which focused primarily on pressurizer heater sleeves and Instrument nozzles [ emphasis added), in part, follow:
"1) Circumferential cracking of the heater sleeves and the instrumentation nozzles [ emphasis added) is not a credible failure modo...
1 .
Appendix A "3) Visual inspection is the best method for detecting a leaking sleeve or nozzle... (emphasis added]
"The Staff has reviewed the report, and finds that it's results and recommended inspections, coupled with field experience, provide a sufficient basis to conclude that loss of structural integrity and ejection of components with respect to pressurizers are highly unlikely."
Examole 2: SECY 97-063. March 19E Proposed NRC Generic Letter: " Degradation of Control Rod Drive Mechanism and Other Vessel Closure Head Penetrations" "Beginning in 1986, leaks havo been reported in several Alloy 600 pressurizer instrument nozzles [ emphasis added) at both domestic and foreign reactors...The NRC staff identified primary water stress corrosion cracking (PWSCC) as an emerging technical issue to the Commission in 1989, after cracking was noted in Alloy 600 pressurizer heater sleeve penetrations at a domestic PWR fac;lity. The NRC staff reviewed the safety significe of the cracking that occurred, as well as the repair and replace ..it activities at the affected facilities. The NRC staff determined that the cracking was not of immediate safety significance because the cracks were axial, had a low growth rate, were in a material with an extremely high flaw tolerance (high fracture toughness) and, accordingly, were unlikely to propagate very far. These factors also demonstrated that any cracking would result in detectable leakage and the opportunity to take corrective action before a penetration would fail."
Example 3: Generic Letter 97-01. Aoril 19E Generic Letter 97-01 addresses the issue of the potential for cracking in Alloy 600 CRDM nozzles and other vessel head closure penetrations (VHP).
"The NRC staff determined that the cracking was not of immediate safety significance because the cracks were axial, had a low growth rate, were in a material with an extremely high flaw tolerance (high fracture toughness), and accordingly, were unlikely to propagate very far. These factors also demonstrated that any cracking would result in detectable leakage and the opportunity to take corrective action before a penetration would fail."
Appendix A The Generic Letter also states:
"After considering this information, the NRC staff has concluded that VHP cracking does not pose an immediate or near term safety concern."
Example 4: November 1991NRQ_ Letter in a November 19,1993, letter from William T. Russell to William Raisin, the 14RC responded to NUMARC's June 16,1993, letter regarding Alloy 600 CRDM/CEDM head penetrations. The NRC's conclusion is:
Based on the overseas inspection findings and the review of your analyses, the staff has concluded that there is no immediato safety concern for cracking of the CRDM/CEDM penetrations."
" Based upon information received from os rseas regulatory authorities, your analyses, and staff reviews, the staff be!! eves that catastrophic failure of a penetration is extremely unlikely. Rather, a flaw would leak before it -
reached the critical flaw size...."
Examole 5: NRC Information Notice IN 90-10 I t
~
"The cracking to date in the thermal sleeves and the instrument nozzles
[ emphasis added) of the domestic PWRs has been reported as being only axially oriented. The safety imolication of an axial crack is no' considered a significant threat to Ine structural Integrity of the components and most likely will result in a ,
small Ieak...Circumferential cracking poses a more serious safety concern becauso if it were to go undetected it could lead to a structural failure of a component rather than to a limite d leak."
" it may be prudent for licensees of all PWRs to review their Alloy 600 applications in the primary coolant pressure boundary, and when necessary, to implement an augmented inspection program."[ emphasis added)
Examole 6: NRC Status Reoort to the Commission .
On May 12,1993, the b'RC staff provided a status report to tne Ccmmission regarding PWSCC of Alloy 600 components. The NRC concluded the following at that time:
- - - - - -. _ _ . ~ _ _ _ _ _ _ _ _
Appendix A "Having reviewed the information to date, including the inspection results and findings, the staff maintains its view that this issue is of low safety significance since all cracks reported to date, with perhaps one exception (a.k.a. EdF French reactor), are short in length and axially oriented in an extremely flaw-tolerant mater lal."
1 Finally, the following is a compendium of the Alloy 600 RCS penetration PWSCC history, involving the NRC and the Industry (including SCE):
- 1. 1984 NRC SER or Nine Mile Point, Unit 1 (6/29/84) 1 2. November 1986, LER 86-003 and 86-003 Rev i
- 3. March 1987 Nine Mlle Point Code Relief (3/25/87)
. 4, 1989 NRC Calvert Cliffs Confirmatory Action Letter Closure 2
- 5. November 1989, CEN-393 P, NP (11/3/89)"Prescurizer Heater Sleeve Susceptibility to PWSCC" [lssued to NRC on 11/17/89) 3
~
- 6. February 1990 NRC IN 90-10 (2/23/90), " Primary Water Stress Corrosion ~
Cracking (PWSCC) of Alloy 600"
- 7. March 1990, CE NPSD 555 (3/2190),"Precsurizer Heater Performance"
- 10. September 1990, PWSCC Coordinating Group Meeting (9/12/90) Parsippany, NJ
- 11. November 1990, CE NPSD-618 (11/5/90),"!ntraspect/ET20 Eddy Current Imaging Development for Pressurizer Heater Sleeve Inspection for FPL, St. Lucie Uni' 2"
- 12. Janucry 1991, PWSCC Coordinating Group Meeting (1/8/91) Palo Alto, CA
- 13. February 1991, C5 NPSD-617-P (2/25/91), " Destructive Examination of Pressurizar Instrument Nozzles from Calvert Cliffs Unit 2 and Evaluation of Similar Nozzles'
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il T
Appendix A -
- 14. March 1991, CE NPSD-649-P (3/18/91),"Information Package on Alloy.600._
. Primary Pressure Boundary Penetrations" -Listing all Alloy 600 penetrat.as for RCS and Pressurizer (less Rx Vessel) for all CEOG members;
- 15. March 1991, CE NPSD432 Part 2 (3/28/91), " Residual Stress Measurements on Calvert Cliffs 2 Pressurizer Heater Sleeves"
- 16. April 1991, CE NPSD448-P (4/25/91), " Corrosion and Corrosion / Erosion Testing .
of Pressurizer Shell Material Exposed to Borated Water"-
Analysis"
- 18. June u91, CEN 406-P (6/6/91)," Status Report on CEOG Activities Concerning Primary Water Stress Corrosion Crackrq of incenel-600 Penetrations" (Sent to NRC on 5/31/91 via CEOG-91-300]
o 19. September 1991, CE NPSD459-P (9/25/91), " Additional Pressurizer Heater .
Sleeve Examinations"
- 22. January 1992, CE NPSD490 P (1/20/92)" Evaluation of Pressurizer Penatrations
' and Evaluation of Corrosion After Unidentified Leakage Develops Pressurizer -
Inspection Recommendat:ons" [Provided to NRC on 2/26/92 via CEOG-92-052] .
1
- 23. February 1992, PVNGS LER 1-92-001 (2/3/92), APS reports a Unit 1 pressurizer steam space instrument nozzle leak
.24. March 1992, San Onofre LER 2-92-004-00 (3/19/?2); LER 2-92-004-01 ($/18/92)
- 25. March 1992, NRC/CEOG Meeting (03/25/92)
. 26. April 1992, NRC Inspection Report 92-06 (SONGS) - Noted that SCE identified three Unit 3 nozzle leaks in the pressurizer, resulting from PWSCC. Noted SCE's
- effort to resolve the problems were professional and effective.. Also discussed _
was the meeting held in Walnut Creek where SCE presented information on the nozzle replacement to the NRC.
Appendix A
- 27. May 1992, NRC Inspection Report 9212 (SONGS) - The inspectors reviewed work with associated Unit 2 nozzle repair, and questioned SCE's effort to eggressively complete the operability assessment and failure mechanism.
IFl 9212 03 was opened.
- 28. July 1992, NRC Inspection Report 92 The inspectors reviewed and closed out LER 2-92-004 revisions O and 1 on Pressurizer Nozzle Cracks.
- 29. August 1992, PWSCC Coordinating Group Meeting (8/11/92), Juno Beach. FL
- 30. October 1992, NUMARC PWSCC Meeting (10/2/92), Washington, DC
- 31. October 1992, Nozzle Integrity Assessment Meeting (10/21/92). Washington, DC
- 32. November 1992, NUMARC PWSCC Meeting (11/11/92), Washington, DC
- 33. November 1992 NRC/NUMARC Alloy 600 Nozzle Meeting (11/20/92),
Rockville, MD d
- 35. December 1992, Nozzle Integrity Assessment Meeting (12/2/92), Orlando, Fl.
i
- 36. December 1992, FWSCC Integrity Assessment /EdF Meeting (12/4/92),
Orlando, FL
- 37. December 1992 - NRC Inspection Report 92-29 (SONGS) - This report closed IFl 9212-03 related to Ur.it 2 pressurizer nozzle repair. The inspector (s) identified concerns with timeliness in completing the assessment of the impact of the leakage. The CEOG evaluation was discussed with SCE and the inspector closed the IFl.
- 38. January 1993, PWSCC Integrity Assessment Meeting (1/13/93),
Juno Beach, FL
- 33. February 1993 - NRC Inspection Report 92 SONGS SALP - Stated in general maintenance and surveillance activities conducted more effectively, citing SCE's effort to repair Unit 3 pressurizer nozzles.
- 40. February 1993, NUMARC PWSCC Meeting (2/19/93), Washington, DC l
Appendix A
This report provided data on nozzle material heats and configurations for ecch member plant.
- 42. March 1993, NRC/NUMARC Alloy 600 Nozzle Meeting (3/3/93), Rockville, MD
Documented information from cracking at Bugey and status of other EdF and World Wide inspections through the beginning of 1993.
- 44. April 1993 (4/13/93) NRC Inspection Report 93-08 (St. Lucle)
- 45. April 1993, Nonle Integrity Assessment Meetle,g (4/15/93), Charlotte, NC
- 46. April 1993, EPRl/EdF PWSCC Meeting (5/6/93), Herndon, VA 4'/. May 1993, NRC Status Report to the Commission (5/12/93)
- 48. May 1993, Nozzle Integrity Assessment Meeting (5/13/93), Charlotte, NC
- 49. May 1993, CEN 607 (5/28/93)" Safety Evaluation For ID Axial Cracking" - This report was developted and issued to the NRC (via NUMARC) in May,1993. It concluded that ID axial cracking of CEDMllCl penetrations was not an immediate safety concern. Results documented in this report were largely based on conclusions from the Dominion Engineering Report (del-357) also funded under Task 744.
- 50. June 1'393, del 357 (6/4/93), " Dominion Engineering Report on Stress Analysis" -
This report documented the results of finite element analyse '. . CEOG CEDM penetrations.
- 51. June 1993, NUMARC Letter to NRC (6/16/93)-Three PWR Owners Group's safety assessments provided addressing Alloy 600 CRDM/CEDM VHP cracking issue. NUMARC's conclusion was,"The reports confirm that the potential for cracking does not pose an immediate safety concern."
- 52. July 1993, NRCINUMARC Alloy 600 Nozzle Meeting (7/15/93), Rockville, MD
- 53. October 1993, Nozzle intogrity Assessment Meeting (10/01/93), Charlotte, NC
Appendix A
- 54. November 1993 NRC Letter to NUMARC (11/19/93)
- 55. December 1993, CEN-614 (12/30/93)" Safety Evaluation For OD Circumferential .
Cracking" - This report, like CEN-607, wat issued via NUMARC to the NRC. It documented analyses showing that propagation of an OD crack in a CEDMllCI penetration would require from 80 to 100 years to grow to a point where structural integrity of the penetration would be in jeopardy.
- 56. January 1994, NUMARC Letter to NRC (1/31/94)-The conclusion of this letter was that "neither the potential for circumferential cracking nor the existence of circumferential cracks pose an unreviewed or immediate safety issue." This letter included revised safety assessments from each of the 3 PWR Owners Groups in support of thic conclusion.
- 57. February 1994, CE NPSD 905 P, Revision 1 (2/15/94), "CEDM Phase I, Susceptibility Ranking" - Compared the properties, fabrication processes and environmental conditions of CEOG CEDM'iCl nonles with nonles from foreign plants which had experienced cracking.
- 58. March 1394, CE NPSD-927-P (3/30/94)," Stress Analysis Sensitivity Stus / -
Compared the results of analyses with both nugget cooling and heat transfer models of welding to address differences between WOG and CEOG safety analyses. Concluded that CEOG method was appropriate and that the results reached in CEN-607 were valid.
- 59. April 1994, CE NPSD-918 P (4/11/94), ' Phase 2, inspection Timing Model' - This report supersedes CE NPSD-905-P relative tc individual nonle timing for susceptibility to cracking and crack propagation.
- 60. April 1994, CE NPSD 919P (4/11/94), " Phase 2, inspection Strategy and Repair Report" - Report identified an inspection strategy for CEOG member vessel head penetrations, and repair requirements for shallow and deep cracks initiated from nonle ID locations.
- 61. April 1994 (4/28/94), NRC 1,npection Report 94-10 (St. Lucle)
- 62. June 1994, CE NPSE-938 P, Revision 1 (6/10/94)," Alloy 600 Bar Stock '
Procurement - Material Specifications, Certified Test Reports & Inspection Certificates" 9
Appendix A
- 63. June 1984, CE NPSE 948 (6/23/94), ' Leak Detection Methods Evaluation" -
Documented ABB review of available literature on leak detection methods, including a report made available by the B&WOG on the same subject. Reported that Nitrogen-13 detection systems showed the most promise.
- 64. July 1994, CE NPSD 947-P (7/13/94),"PWSCC Mitigation Methods"- Report evaluated several mitigation methods including weld overlay, shot peening, and nickel plating as mitigation methods for CEDM/ICI cracking.
4
- 65. July 1994, EPRI TR-103696, 'PWSCC of Alloy 600 Materials in PWR Primary System Penetrations" - F.PRI states, "It is important to note that none of thts Alloy 600 penetration PWSCC incidents which have occurred to date have posed a sigrqccat safety problem at the plants involved. This is because most of the cracks have been short and axial, and the leakage rates from the cracks have been very low...In summary,... cracking of Alloy 600 primary loop penetrations does not pose a significant safety problem...The NRC has concurred with the Industry position that there is no immediate safety concem for cracking of CRDM nozzles provided that visual inspections for boric acid leakage are performed per Generic Letter 88.05." [ emphasis added)
- 66. October 1994 NUREGICR-6245
- 68. November 1994, CE NPSD 949 P (11/28/94), " Evaluation of Boric Acid Corrosion of RV Heads Resulting from Leaking CEDM Nozzles"- Concluded that undetected leakage from cracks in adjacent CEDM nozzles could exist for almost nine years before ASME code requirements for reinforcement would be violated. A more rea'istic case showed more than 15 years of leakage could exist. Report justified that undetected leakage did not present an immediate safety concem.
- 69. January 1995 Petition Denial D.D.-95 2 (1/26/95)
- 70. August 1995, SONGS LER 3-95-001
- 71. October 1995, CE NPSD-1028 (1013/95), " Fabrication of Ten Pressurizer Nozzle
~ Assemblies"
- 72. October 1955, CE NPSD-1017 (10/06/95), " Assessment of Grain Boundary Carbide Distribution in Alloy 600 CEDM and ICE Nozzles"
- - - . . - . - . . . ~ .
. . . -._. - --. .-.--.- . -- ~
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Appendix A:
_.73. November 1995, CEN-406-NP (11/2/96), "A Status Report On CEOG Activities
_ Concerning Primary Water Stress Corrosion Cracking of inconel 600 i b Penetrations * [ report sent to'NRC from Palisades] _
- 74. - December 1995, CE NPSD-1019 (12/21/95), " Summary Report of Stress -
Evaluation for a Deep Crack Repair of Alloy 600 CEDM Penetrations"
- 75. April 1996 - NRC lospection Report 9642 (SONGS) - Review and closure of LER 3-95-001-00 on RCS nozzle leakage. Additionally, the inspector (s) evaluated "
the acceptability of welding materials'used on repairs of RCS nozzles and -
F identified inconsistencies with UFSAR tables. NCV on LER.
76; July 1996, CE NPSD 1032 (7/15/96), "CEDM Repair Procedure" E - 77) July 1996, CE NPSD-1013-P (7/19/96), " Development of a Deep Crack Repair
_ Capability for Alloy 600 CEDM Penetrations" l 78.' October 1996, SONGS LER 3 96-004 (10/23/96) - Reports leakage indications on _
L three pressurizer instrument nozzles found during a nozzle inspection et the
' beginning of the Cycle 8 refueling outage. The cause was identified as PWSCC.
All Unit 3 pressurizer nozzles were inspected and four nozzles were replaced.
The outer portion of the Alloy 600 nozzle had been previously replaced with Alloy 690 material, but the weld filler material was equivalent to Alloy 600. When replaced, new filler material equivalent to Alloy 690 was used.
- 79. December 1996, WCAP 13929, Rev. 2 (12/9/96), " Crack Growth and
- Microstructural Characterization of Alloy 600 Head Penetration Materials"
- 80. - February 1997,4* EPRI PWSCC Workshop (2/25 27/97), Datona Beach, FL - In St. Lucie's presentation," EXPERIENCE WITH DETECTION AND REPAIR OF PWSCC FLAWS IN PWR PRESSURIZER AND RCS LOOP ALLOY 600 F PENETRATIONS AT ST. LUCIE UNIT 2," St. Lucie concluded: 1) the observed F cracks were determined stable by fracture mechanics; 2) stress analysis shows
- . cracking will be axial; and 3) ejection, confirmed by fieki observation, is unlikely.
They_ also concluded the only safety concern was the boric acid corrosion from long term unidentified leaks which are being managed by inspection. Therefore, ,
PWSCC nozzle cracking is not a safety issue; however, there are economic H concerns of unplanned repairs.
1
--81. March 1997, SECY 97-063 (3/18/97) l 9
4
Appendix A
~ 82. April 1997 Generic Letter 97-01 (4/1/97)
- 83. April 1997, NRC Inspection Report 97-05 (SONGS) - The inspectors observed work related to Alloy nozzle replacement, and found the work thoroughly performed, The report discussed the details of the repair activities.
- 84. April 1997, SONGS LER 2 97-004 (4/2/97) Reports leakage from the Unit 2 pressurizer. This leakage was found during a mode 4 walk down as part of the unit's returri to power following the Cycle 9 refueling outage. The outer portion of the Alloy 600 nozzle was replaced with Alloy 690 material. PWSCC was identified as the cause.
- 85. May 1997, SONGS LER 3 97-001 (5/9/97)- Reports leakage from five Unit 3 nozzles found as part of the initial walk down at the beginning of the Cycle 9 refueling outage. The outer portion of the Alloy 600 nozzle was replaced with Alloy 690imaterial. The LER acknowledges PWSCC as the likely cause.
- 86. June 1997, NRC Insp ,ction Report 97-09 (SONGS)- Reports the results of resident inspector activities, including observations of nczzle replacement. The inspectors noted the licensee identified the potential leabge in accordance with established plans.
- 87. July 1997, NRC inspection Report 97 08 (SONGS)-ISI AND BORIC ACID INSPECTION - The inspectors noted the Boric Acid control program was being implemented in accordance with the established program. IFl 9501-01 related to containment inspections on Boric Acid was also closed out.
- 88. July 1997, CE NPSD 1085 (7/20/97)"CEOG Response to NRC GENERIC LETTER E7-01, ' Degradation of CEDM Nozzle And Other Vessel Closure Head Penetrations" - Provided the CEOG response to GL 97-01,
- 89. July 1997, SONGS LER 3-97-002 (7/30/97)- Reports leakage from four Unit 3 nozzles during the planned inspections as part of the units return to power at the end of Cycle 9 refueling. The outer porticn of the Alloy 600 nozzle was replaced with Alloy 690 material. The LER acknowledges PWSCC as the cause and
. credits SCE's inspe :t and replace program for finding these nozzles that waren't found at the beginning of the outage.-
Appendix A
-90. September 1997, NRC Inspection Report 97-15 (SONGS)- The report also notes that though the Cycle 9 RFO, Unit 2 has experienced 4 nozzle cracks and Unit 3,14 cracks, it was also noted that 2 heats experienced 4 cracks each. The report also states there is no current nozzle replacement plan due to development of in-house capabilities, and that these actions to develop the capabilities were not started until the 3rd quarter 1996. Also, an apparent violation of 10 CFR 50 Appendix B, Criterion XVI for failure to implement actions to preclude recurrence, was stated.
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