ML20203D214
| ML20203D214 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/09/1997 |
| From: | Howell A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Ray H SOUTHERN CALIFORNIA EDISON CO. |
| Shared Package | |
| ML20203D220 | List: |
| References | |
| 50-361-97-15, 50-362-97-15, EA-97-414, NUDOCS 9712160136 | |
| Download: ML20203D214 (87) | |
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art.tNGTON. T E x AS 76011 8004 December 9, 1997 EA 97-414 t
Harold B. Ray, Executive Vice President y
Southern California Edison Co.
San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, California 92674-0128
SUBJECT:
PREDECISIONAL ENFORCEMENT CONFERENCE - REACTOR COOLANT SYSTEM INCONEL 600 NOZZLE PENETRATIONS
Dear Mr. Ray:
This refers to the F.adecisional Enforcement Conference conducted in the Region IV office on September 30,1997. This meeting related to the scope of licensee corrective actions in response to identified primary water stress conosion cracking in reactor coolant system L
Inconel 600 nozzle penetrations.
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy.of this letter will be placed in the NRC's Public Document Room.
Enclosures included are the attendance list, the licensee presentation material, your letter of
, October 3,1997, the condensed transcript of the predecisional enforcement conference, the predecisional enforcement conference agenda, the apparent violation, and your letter of October 31,1997.
Your letter of October 3,1997 (Enclosure 3) provided responses to questions raised in the September 30,1997, meeting, clarified the licensee presentation material, and provided information regarding licensee-perceived errors in NRC Inspection Report 50-361/97-15; 50-302/97-15. Your letter of October 31,1997 (Enclosure 7), transmitted an update of your inconel 600 program plan (i.e., Document 90622, " Susceptibility of Reactor Coolant System Alloy 600 Nozzles to Primary Water Stress Corrosion Cracking and Replacement Program Plan,
. Revision 2, dated October 31,1997).
We have reviewed the information provided in your letter of October 3,1997, pertaining to perceived errors in the inspection report and have the following response.
Severs' of your comments relate to disagreement with the. number of nozzles reported to have exhibited through-wall leakage. Some difficulties were experienced during the inspection in making a clear determination from the reviewed documentation on whether identified cracking was partial or through wall in nature. We accordingly accept the numerical differences indicated by your responses. We also acknowledge that our use I
of the term " failure" coukt have been clearer with respect to distinguishing between partial and through-wall cracking.
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' Southern Cahfornia Edison Co.- * -
With respect to your comment on the fifth paragraph of the report executive summary, we agree that Document 90022 Revision 1, does not specifically use the term "in-house capability,"
With respect to your comment on Table 1, Note 3, in the inspection report, we do not consider the text to be in error. You indicated your belief that the failure mechanism was primary water Stress corrosion cracking for Nozzle 3TE0121Y2, without any stated basis.
As noted in paragraph 2 of page 11 of the inspection report, we considered this failure mechanism to be unlikely for a cold leg nozzle with a low service temperature of approximately 553'F. In the absence of specifin laboratory examination results for Nozzle 3TE0121Y2, coupled with the laboratory identification of a fatigue failure in the thermowell in another cold leg nozz!e,3TE9179-3, we believe that a postulated fatigue failure mecaanism is more credible than primary water stress corrosion cracking.
You indicated disagreement with the text in the third paragraph on page 10 of the inspection report that indicated that the attachment weld for a partial inconel 690 replacement nozzle (for Nozzle 2PDT0978-1) was made to a weld buildup pad on the piping outside diameter surface. We accept your comment that a weld buildup pad was not used on thk piping, in that the inspection did not focus on repair techniques. We note, however, that Attachment 8.1, Table 1, of your updated inconel 600 program plan (Enclosure 7) indicates that a weld buildup pad was, in fact, used for this replacement.
Specifically, the history column in Table 1 states, in part, for the replacement of Nr zzle 2PDT0978-1, "1993-Replaced with A690. Replaced from outside, weld pad.."
With respect to your response to text in the second paragraph on page 11 of the
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inspection report, concerning Root Cause Evaluation Report 97-001, we do not consider the text to be in error. Our position on this root cause evaluation report is contained in the second paragraph of page 14 of the inspection report.
With respect to your comment on the first paragraph of page 14 of the inspection report, we do not consider the text to be in error. This paragraph contained an inspector view that the detection in a nozzle (manufactured from Heat NX7630) of the presence of shallow intergranular cracking, at a location remote from the more highly stressed material adjacent to the J-weld, was indicative that the Heat NX7630 material v&s highly susceptible to primary water stress corrosion cracking. Your response to this view indicates that you believe that while Heat NX7630 microstructure may increase susceptibility to primary water stress corrosion cracking, other factors (e.g.,- temperature, yield strength, environment, residual stresses, etc.) result in Heat NX7630 being no more susceptible _to primary water stress corrosion cracking than other heats. We do not consider that this response provides a sufficient technical basis for recharacterizing our origina! inspection observation.
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. Southern California Edison Co. "
j We accept your comment (in response to a reference in the first paragraph of page 15 of the inspection report to a review being made of Document 90022, Revision 1. to ascertain program requirements for addressing primary water stress corrosion cracking in inconel 600 components) that the program plan provided recommendations, not 4
requirements.
Please note that it is our intention to issue corrections to inspection Report 50-361/97-15;
!50-362/97-15 for accepted comments that are believed to be factually relevant.
Should you have any questions concerning this matter, we will be pleased to discuss them with you.
Sincerely,
/c 'llN zi rIhur T. Howel 111, Director Division of Reactor Safety
Enclosures:
- 1. Attendance List
- 2. Licensee Presentation
- 3. Licensee Octooer 3,1997,ietter
- 4. Condensed transcript of Predecisional Enforcement Conference
- 5. Predecisional Enforcement Conference Agenda
- 6. Apparent Violation
- 7. _ Licensee October 31,1997, letter Docket Nos.: 50-361;50-362 License Nos.: NPF-10, NPF-15 I
cc w/ Enclosures 1,4,5, and U:
Chairman, Board of Supervisors 1-County of San Diego i
1600 Pacific Highway, Room 335 San Diego, California 92101 Alan R Watts. Esq.
Woodruff, Spradlin & Smart 701 S. Parker St. Suite 7000 Orange. California 92868-4720 t-4
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- Southern California Edison Co..
- Sherwin Harris, Resource Project Manager Public Utilities Department -
City of Riverside -
3900 Main Street t
Riverside, California 92522-R. W. Krieger, Vice President ~
Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128
' San Clemente, California 92674-C,128 -
Dr. Harvey Collins, Chief
. Division of Drinking Water and
~ Environmental Management California Department of Health Services P.O. Box 942732 Sacramento, California 94234 7320 Terry Winter, Manager Power Operations San Diego Gas & Electric Company P,0. Box 1831 San Diego, Cahfornia 92112 Mr, Steve Hsu Radiological Health Branch State Department of Health Services P.O. Box 942732 Sacramento, California 94234 Ma, or City of San Clemente 100 Avenida Presidio San Clemente, California 92672 Mr. Truman Bums \\Mr. Robert Kinosian California Public Utilities Commission 505 Van Ness, Rm. 4102 San Francisco, California 94102
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I Southern California Edison Co.
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~ Regional Administrator.
Resident Inspector _
DRP Director DRS-PSB Branch Chief (DRP/F, WCFO)-
MIS System Senior Project inspector (DRP/F, WCFO)
RIV File Branch Chief (DRP/TSS)
M.' Hammond (PAO, WCFO)
WCFO File G. F. Sanborn, EO W. L. Brown, RC J. Lieberman, OE, MS: 7-HS OE:EA File, MS: 7-HS DOCUMENT NAME: G:\\TA\\SOMS715.ixb To receive copy of document. Indicate in box: "C" = Copy without enclosures "E" = Copy witn enclosures "N" = No copy D:DRS,1/
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Branch Chief (DRP/F, WCFO)
MIS System Senior Project inspector (DRP/F, WCFO)'
RIV File
- Branch Chief (DRP/TSS)
M. Hammond (PAO,- WCFO)--_
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ENCLOSURE 1 ATTENDANCE LIST
PREDECISIONAL ENFORCEMENT CONFERENCE ATTENDANCE LICENSEE / FACILITY Southern California Edison San Onofre Nuclear Generating Station DATE/ TIME September 30. 1997 1 p.m. COT CONFERENCE LOCATION Training Conference Room Arlington. TX EA NUMBER EA 97-414 LICENSEE REPRESENTATIVES NAME (PLEASE PRINT)
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EMCLOSURE 2 LICENSEE PRESENTATION
y 7
h SOUTHERN CAllf0RNIA EDISON An EDISON Ih TERNATIONAl. Compuny SOUTHERN CALIFORNIA EDISON SAN ONOFRE NUCLEAR GENERATING STATION ALLOY 600 RCS PENETRATION NOZZLES INDUSTRY /NRC CHRONOLOGY SUPPLEMENTAL INFORMATION SEPTEMBER 30,1997
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SUPPLEMENTAL INFORMATION - CHRONOLOGY 1.
1984 NRC SER for Nine Mile Point, Unit 1 (6/29/84) 2.
February 1986 (Forced Cycle 4 Outage) Pressurizer Steam Space One Unit 3 nozzle leak (~0.15 gpm) [ Heat 54318]
High temperature steam environment believed dominant factor for i
failure mechanism.
Since Alloy 600 RCS PWSCC not understood Ly industry, and no plan endorsed, committed to proactively replace four remaining steam space nozzles [all Heat 54318] (two steam space aru two water space) with Alloy 600. First nozzle replacement in Industry.
One nozzle replacement incurred ~4.3 Man-Rem of exposure.
LER 86-003 and 86-003 Rev 1 3.
SCE Joined PWSCC Industry Groups CEOG EPRI 4.
March 1987 Nine Mile Point Code Relief (3/25/87) 5.
January 1987 (Unit 3 Cycle 3) Pressurizer Steam Space Two steam space nozzles [all Heat 54318) were proactively replaced with I,lloy 600.
Replacement of nozzles incurred ~8.6 Man-Rem of exposure.
6.
September 1987 (Unit 2 Cycle 4) Pressurizer Water Space One water space nozzles (all Heat 54318] were proactively replaced with Alloy 600.
Replacement of nozzles incurred ~4.3 Man-Rem of exposure.
1
l l
SUPPLEMENTAL INFORMATION - CHRONOLOGY (Continued) 7.
May 1988 (Unit 3 Cycle 4) Pressurizer Water Space One water space nozzles (all Heat 54318) were proactively replaced with Alloy 600.
Replacement of nozzles incurred ~4.3 Man-Rem of exposure.
8.
1989 NRC Calvert Cliffs Confirmatory Action Letter Closure 9.
November 1989, CEN 393 P, NP (11/3/89) " Pressurizer Heater Sleeve Susceptibility to PWSCC" (Issued to NRC on 11/17/89).
i 10.
February 1990 NRC IN 9010 (2/23/90), " Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600" 11.
March 1990, CE NPSD 555 (3/2/90),"Pressuri7.er Heater Performance" 12.
March 1990, EPRl/CEOG PWSCC Meeting (3/14/90) Rockville, MD.
13.
August 1990, CE NPSD 632 (8/15/90), " Pressurizer Heater Sleeve Examinations" 14.
September 1990, PWSCC Coordinating Group Meeting (9/12/90)
Parsippany, N.J.
15.
November 1990, CE NPSD-618 (11/5/90), "Intraspect/ET20 Eddy Current Imaging Development for Pressurizer Heater Sleeve Inspection for FPL, St. Lucie Unit 2" 16.
January 1991, PWSCC Coordinating Group Meeting (1/8/91) Palo Alto, CA.
17.
c bruary 1991, CE NPSD 617 P (2/25/91), " Destructive Examination of o
Pressurizer Instrument Nozzles from Calvert Cliffs Unit 2 and Evaluation of Similar Nozzles" 18.
March 1991, CE NPSD-649-P (3/18/91), "Information Package on inconel 600 Primary Pressure Boundary Penetrations" 2
SUPPLEMENTALINFORMATION CHRONOLOGY (Continued)
Listing all Alloy 600 penetrations for RCS and Pressurizer (less Rx Vessel) for all CEOG members.
19.
March 1991, CE NPSD 632 Part 2 (3/28/91), " Residual Stress Measurements on Calvert Cliffs 2 Pressurizer Heater Sleeves" 20.
April 1991, CE NPSD-648 P (4/25/91), " Corrosion and Corrosion / Erosion Testing of Pressunzer Shell Material Exposed to Borated Water" 21.
June 1991, CE NPSD-646 (6/5/91), CEOG Pressurizer Heater Sleeve Thermal Analysis" 22.
June 1991, CEN-406-P (6/6/91), " Status Report on CEOG Activities Concerning Primary Water Stress Corrosion Cracking of Inconel-600 Penetrations" [Sent to NRC on 5/31/91 via CEOG-91-300) 23.
September 1991, CE NPSD-659 P (9/25/91), "Additior.at Pressurizer Heater Sleeve Examinations" 24.
November 1991,1" EPRI PWSCC Workshop (10/911/91), Charlotte, NC.
25.
November 1991, CEOG Letter to EPRI(11/12/91), "CEOG Task 692 Near Term Activities" 26.
January 1992, CE NPSD-690-P (1/20/92) " Evaluation of Pressurizer Penetrations and Evaluation of Corrosion After Unidentified Leakage Develops Pressurizer inspection Recommendations" [Provided to NRC on 2/26/92 via CEOG-92-052]
27.
January 1992 (Cycle 6) SONGS Unit 3 Pressurizer Steam Space One nozzle [ Heat 94758] boric acid traces (<1 Drop per Day)
Inspection reveaied surface indications (Og leakage) on 2 other nozzles (Heats NX7630, NX0571]
3 nozzles above + the remaining pressurizer steam space noz7' were replaced with Alloy 690. Code requirements limited w&
filler material to Alloy 600 3
SUPPLEMENTAL IN FORM ATION - CH RONOLOGY (Continued)
Replacement of nozzles incurred ~17,2 Man-Rem of exposure LER 2 92-004-00 (3/19/92); LER 2-92-004-01 (5/1 92)
Because several heats failed, belief was that dominant failure mechanism was steam space temperature environment 28.
February 1992, PVNGS LER 1-92 001 (2/3/92), APS reports a Unit 1 pressurizer steam space instrument nozzle leak 23.
March 1992 (Unreiated Cycle 6 Outage) SONGS Unit 2 Pressurizer Steam Space Two nozzles (Heat NA4411] boric acid traces (<1 Drop per day)
The two steam space nozzles were replaced with Alloy 690.
i e
Code requirements limited weld filler material to Alloy FnO.
Because different heat failed, belief reinforced on steam space temperature environment Replacement of Nozzles incurred ~8.6 Man-Rem of Exposure LER 92-004; Commitment to Replace All Steam Space Nozzles during the Cycle 7 30, March 1992 NRC.'CEOG Meeting (03/25/92) 31.
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.
32.
1992 Inspection Report 92-12 (SONGS)
The inspectors reviewed work with associated Unit 2 nozzle repair, and questioned SCE's effort to aggressively complete the operability assessment and failure mechanism. IFl 9212-03 was opened.
4
SUPPLEMENTAL INFORMATION - CHRONOLOGY (Continued) 33.
July 1992, inspection Report 91-10 The inspector.e reviewed cnd closed out LER 2 92-004 revisions O and 1 on Pressuriz0r Nozzle Cracks.
34.
August 1992, PWSCC Coordinating Group Meeting (8/11/92), Juno Beach,FL.
35.
October 1992, NUMARC PWSCC Meeting (10/2/92), Washington, DC.
36.
October 1992, Nozzle Integrity Assessment Meeting (10/21/92),
Washington, DC.
37.
November 1992, NUMARC PWSCC Meeting (11/11/92), Washington, DC.
38.
November 1992 NRC/NUMARC Alloy 600 Nozzle Meeting (11/20/92),
Rockville, MD.
I 39.
December 1992, 2"d EPRI PWSCC Workshop (12/1-3/92), Orlando, FL.
40.
December 1992, Nozzle Integrity Assessment Meeting (12/2/92),
Orlando, FL.
41.
December 1992, PWSCC Integrity Assessment /EdF Meeting (12/4/92), Orlando, FL.
42.
December 1992 - NRC Inspection Report 92 29 (SONGS)
This report closed IFI 9212-03 related to Unit 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.
43.
January 1993, PWSCC Integrity Assessment Meeting (1/13/93), Juno Deach, FL.
44.
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.
5
SUPPLEMENTAL INFORM ATION - CHRONOLOGY (Continued) 45.
February 1993, NUMARC PWSCC Mocting (2/19/93), Washington, DC.
46.
March 1993, CE NPSD 903 P (3/22/93), *CEDM Phase 1, Nozzle Evaluation" This report provided data on nozzle material heats and configurations for each member plant.
47.
March 1993, NRCINUMARC Alloy 600 Nozzle Meeting (3/3/93),
Rockville, MD.
48.
March 1993, CE NPSD 904 P (3/22/93),"CEDM Phase 1, World Follow" Documented information from cracking at Bugey and status of other EdF and World Wide inspections through the beginning of 1993.
49.
April 1993 (4/13/93) NRC Inspection Report 93 08 (St. Lucie) 50.
April 1993, Nozzle Integrity Assessment Meeting (4/15/93), Charlotte, NC 51.
April 1993, EPRl/EdF PWSCC Meeting (5/6/93), Herndon, VA.
52.
May 1993 NRC Status Report to the Commission (5/12/93) 53.
May 1993, Nozzle Integrity Assessment Meeting (5/13/93), Charlotte, NC 54.
May 1993, CEN 607 (5/28/93) " Safety Evaluation For ID Axial Cracking" This report was developed 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.
55.
June 1993, del 357 (6/4/93), "Domin on Engineering Report on Stress Analysis" 6
l SUPPLEMENTAL INFORMATION - CHRONOLOGY (Continued)
This report documented the results of finite element analyses on CEOG CEDM penetrations.
56.
June 1993 (Cycle 7) SONGS Unit 2 Hot Leg Nozzle (Heat NX7630) Boric Acid Traces (<1 Drop per Day)
First Non-Pressurizer Steam Space Failure Nozzle Replaced with Alloy 690.
Proactively replaced all 4 pressurizer steam space nozzles with Alloy 690 to replace two with interim repairs and the two remaining Alloy 600 nozzles with permanent replacements No through-wall cracks were observed.
Replacement of nozzles Incurred ~19,4 Man-Rem of Exposure LER 92-004 Rev 1 57.
June 1993, NUMARC Letter to NRC (6/16/93) 3 PWR Owners Group's safety assessments provided addressing inconel 600 CRDM/CEDM VHP cracking issue.
NUMARC's conc!usion was, "The reports confirm that the potential for cracking does not pose an immediate safety concern."
58.
July 1993 NRCINUMARC Alloy 600 Nozzie Meeting (7/15/93),
Rockville, MD.
59, October 1993, Nozzle Integrity Assessment Meeting (10/01/93),
Charlotte, NC 60.
November 1993 NRC Letter to NUMARC (11/19/93) 61.
December 1993, CEN-614 (12/30/93) " Safety Evaluation For OD Circumferential Cracking' This report, like CEN-607, was issued via NUMARC to the NRC. It documented analyses showing that propagation of an OD crack in a 7
l l
I SUPPLEMENTAL IN"ORMATION - CHRONOLOGY (Continued)
CEDMllCI penetration would require from 80 to 100 years to grow to a point where structural integrity of the penetre' ion would be in jeopardy.
62.
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 this conclusion.
63.
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 CEDMllCl nozzles with nozzles from foreign plants which had experienced cracking.
64.
March 1994, CE NPSD 927-P (3/30/94)," Stress Analysis Sensitivity Study" Compared the results of eight and twenty noted 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.
65.
April 1994, CE NPSD-918-P (4/11/94), " Phase 2, inspection Timing Model" This report supersedes CE NPSD-905-P relative to individual nozzle timing for susceptibility to cracking and crack propagation.
66.
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 nozzle ID locations.
67.
April 1994 (4/28/94), NRC Inspection Report 94-10 (St. Lucle) 8
SUPPLEMENTAL INFORMATION - CHRONOLOGY (Continued) 68.
June 1994, CE NPSE 938 P, Revision 1 (6110/94)," Alloy 600 Bar Stock Procurement - Material Specifications, Certified Test Reports &
Inspection Certificates" 69.
June 1994, 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.
70.
July 1994, CE NPSD 947 P (7/13/94), "PWSCC Mitigation Me^ hods" Report evaluated soveral mitigation methods including weld overlay, shot peening, and nickel plating as mitigation methods for CEDMllCl cracking.
71.
July 1994, EPRI TR-103696, "PWSCC of Alloy 600 Materiais in PWR Primary System Penetrations" EPRI states, "It is important to note that none of the Alloy 600 penetratior PWSCC incidents which have occurred to date have posed a significant 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 concern for cracking of CRDM nozzles provided that visual inspections for boric acid leakage are performed per Generic Letter 88.05."
72.
October 1994 NUREGICR-6245 73, November 1994, 3'd EPRI PWSCC Workshop, Tampa, FL.
74.
November 1994, CE NPSD 949 P (11/28/94)," Evaluation of Boric Acid Corrc : ion 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 realistic case 9
SUPPLEMENTAL INFORMATION - CHRONOLOGY (Continued) showed more than 15 years of leakage could exist. Report justified that undetected leakage did not present an immediate safety concern.
75.
January 1995 Petition Denial DD 95-2 (1/26/95) 76.
February 1995 (GONGS Unit 2 Cycle 8 Outage) l No PWSCC detected in Pressurizer Steam Space or any other RCS nozzles Only One Alloy 600 RCS PWSCC in Last 36 Months Reinforces Belief That PWSCC Mechanism Was Limited to Steam Space, Not Material Heat 77.
July 1995 (SONCS Unit 3 Cycle 8 Outage) Precsurizer Steam Space Weld Material Crack at Butter to J-weld Replaced All Four Steam Space Nozzles e
Replacement of Nozzles incurred XX Man-Rem of Exposure 78.
July 1995 (Unit 3 Cycle 8 Outage) Hot Leg Nozzles Two Nozzles (Heats 9294 and NX7630] Boric Acid Traces (<1 Drop per Day)
Second Non-Pressurizer Sterm Space Failure Nozzles Replaced with Alloy 690 Replacement of Nozzles Incurred ~4.4 Man-Rem of Exposure e
LER 3-95-001 79 October 1995, CE NPSD-1028 (10/3/95), " Fabrication of Ten Pressurizer Nozzle Assemblies" 80.
October 1995, CE NPSD-1017 (10/06/95), " Assessment of Grain Boundary Carbide Distnbution in Alloy 600 CEDM and ICE Nozzles" 10
SUPPLEMENTALINFORMATION CHRONOLOGY (Continued) i 81.
November 1995, CEN 406 NP (11/2/95),"A Status Report On CEOG Activities Concerning Primary Wate* Stress Conosion Cracking of inconel-600 Penetrations"
[ report sent to NRC from Palisades]
82.
December 1995, CE NPSD-ir.19 (12/27/95), " Summary Report of Stress Evaluation for a Deep Crack Repair of Alloy 600 CEDM Penetrations" 83.
Summary of San Onofre Experience at End of 1995 Pressurizer Steam Space is susceptible region for PWSCC development - as evidenced by 4 nozzles with through-wall tracking, and 2 nozzles with surface indications (on leakage)
Numerous steam space nozzles replaced Multiple material heats involved - does not appear solely material e
heat driven (i.e., " Bad Heat")
NX7630 - 2 failures 9294 - 1 failure e
4411 - 2 failures Replacement of pressurizer steam space nozzles resulted in XX Man-Rem and hot leg nozzles resulted in XX Man-Rem, for a total of XX Man-Rem Option of alternate proactive replacement o
Replacement of all nozzles would incur XX Man-Rem n' exposure Would not pass ALARA cost-benefit analysis Management decision (ALARA Conservative) to await e
cycle 9 inspection data and subsequent industry /NRC experience, and prepare a that plan after the Cycle 9 Outage (Target: October 31,1997) 11
\\
i SUPPLEMENTAL INFCRMATION - CHRONOLOGY (Continued) 84.
April 1996 -Inspection Report 96 02 (SONGS)
Review and closure of LER 3-95-001-00 on RCS nozzle leakage.
Additionally, the inspector (s) evaluated the acceptability of welding materia ls used on repairs of RCS nozzles and identified inconsistencies with UFSAR tables. NCV on LER.
85.
July 1996, CE NPSD 1032 (7/15/96), "CEDM Repair Procedure" 86.
July 1996, CE NPSD 1013-P (7/19/96)," Development of a Deep Crack Repair Capability for Alloy 600 CEDM Penetr"'ons" 87.
October 1996, SONGS LER 3-96-004 (10/23/96)
Reports leakage indications on 3 pressurizer instrument nozzles found during a nozzle inspection at the beginning of the cycle 8 refueling outage. The cause was identified as PWSCC. All Unit 3 pressurizer rozzles were inspected and 4 nozzles were replaced. The outer portion of the Inconel 600 nozzle had been previously replaced with 690 material. but the weld filler material was equivalent to inconel 600.
When rqolaced, new filler material equivalent to inconel 690 was used.
88.
December 1996, WCAP 13929, Rev. 2 (12/9/96), " Crack Growth and Micros tructural Characterization of Alloy 600 Head Penetration Materials" 89.
February 1997,4'" EPRI PWSCC Workshop (2/25 27/97), Datona Bea;h, FL.
In St. Lucie's presentation," EXPERIENCE.WITH DETECTION AND REPAIR OF PWSCC FLAWS IN PWR PRESSURIZER AND RCS LOOP ALLOY 600 PENETRATIONS AT ST. LUCIE UNIT 2," St. Lucie c oncluded 1) the observed cracks were determined stable by fracture mechanics; 2) stress analysis shows cracking will be axial; and 3) ejection, confirmed by field observation, is unlikely. They also concluded the only safety concern was the boric acid corrosion from long term unidertified leaks which are being managed by inspection.
Therefore, PWSCC nozzle cracking is not a safety issue; however, there are economic concerns of unplanned repairs.
12 l
\\
1 SUPPLEMENTALINFORMATION CHRONOLOGY (Continued) 90.
March 1997, SECY 97-06; (3/18/97) 91.
April 1997 Generic Letter 97 01 (4/1/97) 92.
April 1997, NRC Inspection Report 97-05 (SONGS)
The inspectors observed work related to inconel nozzle replacement, and found the work thoroughly performed. The report discussed the details of the repair activities.
93.
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 walkdown as part of the units return t power following the cycle 9 refueling outagn. The outer portion of the inconel 600 nozzle was replaced with 690 material. PWSCC was identified as the cause.
94.
May 1997, SONGS LER 3 97-001 (5/9/97)
Reports leakage from 5 Unit 3 nozzles found as part of the initial walkdown at the beginning of the cycle 9 refueling outage. The outer portion of the inconel 600 nozzle was replaced with 690 material. The LER acknowledges PWSCC as the likely cause.
95.
June 1997, NRC Inspection Report 97-09 (SONGS)
Reports the results of resident inspector activities, including observations of nozzle replacement. The inspectors noted the licensee identified the potential leakage in accordance with established plans.
96.
July 1997, NRC Inspection Report 97-08 (SONGS)
ISI AND BORIC ACID INSPECTION - The inspectors noted the Boric Acid control program was being implernented in accordance with the established program. IFl 9501-01 related to containment inspections on Boric Acid was also closed out.
97.
July 1997, CE NPSD-1085 (7/20/97) "CEOG Response to NRC GENERIC LETTER 97-01,' Degradation of CEDM Nozzle And Other Vessel Closure Head Penetrations
13 I
SUPPLEMENTAL INFORM ATION - CHRONOLOGY (Continued) i 98.
Provided the CEOG responce to GL 97-01, 98.
July 1997, SONGS LER 3 97 002 (7/30/97)
Repods 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 portion of the inconel 600 nozzle was replaced with 690 material.
The LER acknowledges PWSCC as the cause and credits SCE's inspect and replace program for finding these nozzles that weren't found
]
at the beginning of the outage.
99.
September 1997, NRC Inspection Report 9715 (SONGS)
The repod -? o notes that though the cycle 9 RFO, Unit 2 has experienc' d 4 nozzle cracks and Unit 3,14 cracks, it was also noted that 2 heats experienced 4 cracks ea:h. The report also states there is i
no current nozzle replacement plan due to development of in-house capabilities, and that theee actions to develop the capabilities were not started until the 3rd quarter 1996. Also, an apparent violation of 10CFR 50 Appendix B, Criterion XVI for failuro to implement actions to preclude recurrence, was stated.
I
l b EUisi)]'
SOUTHERN CALIFORNIA EDISON SAN ONOFRE NUCLEAR GENERATING STATION l
- 4. w,s.x,mm rm u m,.,
I i
l PRE-DECISIONAL ENFORCEMENT CONFERENCE ALLOY 600 RCS PEXETRATION NOZZLES [ HEAT XX7630]
September 30,1997
Agenda l
I
+ Introduction
- Apparent Violation
- SCE Response l
+ Heat NX7630
+ Alloy 600 PWSCC Discussion
+ Alloy 600 RCS Penetration Nozzle Program
+ Regulatory
+ Conclusion Page 2 e.
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APPARENT VIOLATION NRC INSPECTION REPORT 97-15 COVER LETI'ER. PAGE 2, PARAGRAPil 2 "The apparent violation involved the failure to implement actions to preclude recurrence of Heat NX7630 nozzle penetration cracking in accordance with the requirements of Criterion XVI of Appendix B to 10 CFR Part 50."
Inspection Report Details, Page 15, Paragraph 3 - Discussion "Of specific concern... were the four confirmed failures discussed above in nozzle penetrations manufactured from Heat NX7630 that had occurred during 1992-1997. No actions had been taken by the licensee to preclude failure recurrence despite the high cracking susceptibility demonstrated by the performance history of the material heat."
i Page 4
APPARENT VIOLATION (con't)
+ It is unclear waen t 1e apparent violation occurred.
+ "Tae eva uation of missed opportunities s aoulc.
norma ly c epenc. on waether t:1e information avai ab.e to the licensee shoulc reasonably.aave causec. action t aat would have prevented t ae l
violation." (SRC Enforcement Po. icy Section VI) 4 1
I i
i Page5 i
l
APPARENT VIOLATION (con't; NRC INSPECTION REPORT 97-15 COVER LETTER, PAGE 2, PARAGRAPH I - Additional Concems
... we consider your overall response to be of concern.
Specifically, your current approach to this degradation issue allows the potential for reactor coolant pressure boundary leakage to occur during Mode I power operations, a condition not permitted by your Technical Specifications. We further note that repetitive failures have occurred in individual heats ofInconel 600 alloy nozzle penetrations, indicating certain heats of material are particularly susceptible to primary water stress corrosion eracking."
Page 6
l RESPONSE TO TEE APPARET VIOLATION
+ SCE POLICY:
- SCE will never operate the plant anticipating any specific pressure boundary component to leak.
Where SCE can anticipate a nozzle as being likely to leak during the next cycle, SCE will replace the nozzle.
+ SONGS has operated and will continue to operate, in full compliance with the Tec',ical Specifications.
+ There was no reason for special replacement activities on Heat NX7630 PWSCC at SONGS. Augmented inspections were performed in accordance with Generic Letter 88-05 (Boric Acid Corrosion) and Information Notice 90-10 (PWSCC ofInconel 600). SCE's program identified and corrected PWSCC in nozzles (including Heat NX7630),
by replacing with Alloy 690 to preclude recurrence.
^
Page 7
RESPONSE TO THE APPARENT VIOLATION (con't) 4
+ SCE did not violate 10 CFR 30 Appendix B, Criterion XVI.
+ SCE cannot identify corrective action in response to this apparent violation which would preclude recurrence.
Page3
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HEAT NX7630
+ Four NX7630 Cases Identified in IR 97-15
+ NX7630 Heat Susceptibility
+ NX7630 Replacement Approach
+ NX7630 and Technical Specifications
+ NX7630 and Criterion XVI Page 10
Four XX7630 Cases Ic entifiec in IR 97-15
+ First (February 1992) NX7630 PWSCC case was on the Pressurizer vapor space and was not through wall.
Proactively inspected nozzles in response to a vapor space nozzle leak (Heat 94758) and replaced.
+ Second (June 1993) NX7630 PWSCC case was not confirmed by Radio-Chemical analysis (boric acid crystals were four years old), but was acted upon conservatively.
l
+ The actual through wall weepage of the third (July 1995) and fourth (April 1997) NX7630 PWSCC cases was too small to be measured.
l l
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XX7630 Heat Susceptibility Nozzle PWSCC is a function of time, temperature, residual stress in the
+
nozzle, weld, micro-structure, and water chemistry.
Some NX7630 nozzles in the most aggressive environment
+
(Pressurizer), are still performing without visual indications of PWSCC.
NX7630 falls within the normal distribution of Alloy 600 PWSCC (see
+
charts).
Based on the information available in 1995 (failure history, metallurgy
+
evaluation, industry practices and standards), there was no way to predict or expect that NX7630 was at any more risk for PWSCC than any other heat of Alloy 600.
Page 13
ALLOY 600 NOZZLE PWSCC HISTORY (through 1995, percent of oopulation)
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i 14 1
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1 i
ALLOY 600 NOZZLE PWSCC HISTORY (by heat, through 1995, percent of population) 1
..,~~"
se #"
)
100.0% -T'.
,a&
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l 75.0 %
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.____..___.2___.u__
_ _.___ ___.__ __.,_____. _ __ _ _ _3 I
15 I
a i
!l l
f
1 5
(
h Alloy 600 Nozzle PWSCC Indications t
CE Plants (T > 600 F) s Heat NX7630 i
90 l
80 70 t j w/rr/insp/c-rmo#9 i - -
+
7-
+
_ :7 _ - -
60 a
1 50-}-
+
4 1
t 40 t
1' 30 L
20 e
r j
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ICE Plants (Thot > 600 Deg)!
-l MX7630 l--
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Eta Peta r^2 n/s 1997 147.7793 1.48 334/304 0927 l
.2 n'
110.9914 1.763 46/42 l
.1
/
i
.1 1
10 100 1000 Tinse to Indication (EFFH x 1000)
Nonnalized to 653*F r,
When 99. Confidence Limits are applied to the Industry data, the NX7630 data is completely bounded indicating that there is not a significant statistical difference between the two populations.
Page 16 3
o
XX7630 Heat Susceptibility (con't)
IR 97-15 reference to SCE Root Cause Evaluation RCE 92-019,
+
" SONGS 3 Pressurizer Level Instrument Nozzle Leakage" "The inspector considered that the presence of shallow intergranular cracking (remote from the more highly stressed material adjacent to the J-weld) indicated that the Heat NX7630 material was highly susceptible to primary water stress corrosion cracking.
The laboratory results were. thus, considered to provide some explanation for the subsequent failure history of this heat?'
RCE 92-019 provides no evidence tha. NX7630 was more susceptible
+
than other heats of Alloy 600:
"The presence of high density carbide bands may explain the cracking observed on this heat despite its considerable low yield strength."
RCE 92-019 does not conclude that NX7630 is at high risk, nor does it
+
recommend wholesale replacement.
Thisjudgement is confirmed by the statistical analysis on prior pages.
+
Page 17
XX7630 Replacement Approach
+ No reliable method exists to forecast nozzle failure, including heat of material. Lacking the means to discriminate, all nozzles would require replacement.
+ 1995 ALARA consequences for complete NX7630 nozzle replacement would be approximately 48 personrem (e.g.,
1994 site exposure was 40 personrem).
+ Visual inspection and replacement as necessary is entirely consistent with Industry practice and NRC acceptance of similar applications at other plants.
Page IR
NX7630 anc Tecanical SpeciTeations
+ SCE will never operate the plant anticipating any specific pressure boundary component to leak. Where SCE can anticipate a nozzle as being likely to leak during the next cycle, SCE will replace the nozzle.
+ Technical Specification LCO 3.4.13 limits RCS operational leakage:
No pressure boundary leakage, i gpm unidentified leakage Pagel'1
XX7630 and Tecanica Speci ~1 cations (con't)
+
Tech Spec Bases 3.4.13,"RCS Operational Leakage":
"Componentjoints are made by welding, bolting...
"During plant life thejoint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety."
Tech Specs require immediate shutdown if pressure
+
boundary leakage source is identified, which SCE has done. (Tech Spec LCO 3.4.13, Action B)
Page 20
XX7630 anc Criterion XVI
+ NX7630 PWSCC xvas properly identified, corrected by replacing nozzles with Alloy 690 to prevent recurrence, and augmented inspections were performed in accordance with Generic Letter 88-05 and Information Notice 90-10.
+ SCE could not have predicted which nozzles ivould develop PWSCC based on heat alone, and therefore, short of replacing all Alloy 600 nozzles, could not have taken action to prevent all failures.
+ SCE believes we acted appropriately based on all available information on RCS penetration PWSCC and ALARA.
Page 21
l e
i o
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Industry Experience with Alloy 600 RCS Penetration PWSCC
==~\\
n
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St Lude No SeetMc Dee e a f St f%k @
fckd Iage U
Typical Alloy 600 Nozzle
/
NOZZLE
/
]/
/
/
L I
K J-WELD RCS PIPING INSTRUMENT NOZZLE SHOWN i
Page 24 l
4
. ~,.
PWSCC in A oy 600 Instrument Nozz es 2 f4ezzles Wended M Vesset Shes
- 1. Nozzles Atedened on inses Surtace N
Tpeal Anas Crad
\\
Y i
f/////N///////////H//H//hNOh' Y
f//HH//N//H/h/HN/H'/HN /
y nnconeneunew g sta *rss swe end Wclded Pressuriser NotAe Residual stresses from cold working (boring operation) and weld
+
shrinkage lead to increased PWSCC susceptibility Cracking is axial since hoop stresses exceed axial stresses at high
+
stress locations
+
Axial cracks are short in length (high stress region at J-groove sveld)
Crack growth beyond J-groove weld region is very slow since
+
operating stresses in the region are low r...,3
Circumferential Cracking in Roll Expandec Ec.F Nozzles s umw. ume w,s n ssae.
2mm*. na E. =ne m ve,w sw g
\\
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s %uw wee.e = sm e srwe N
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s+
- l" rus:111//,/i:,,,,,,,,,,,,,,,,,
4A4
% sia wis sw.e o.e Itolksi & Weldeel Precurifer Nonle,
Some French nozzles have experienced shallow circumferential cracking
+
Circumferential cracking arose from roll expansion into the vessel hell prior
+
to welding
+
SONGS nozzles were not roll expanded and therefore not subject to circumferential cracking I'ase 26 i
Alloy 600 Nozzles Present in Pressurizer, Loop Piping, and SG Cold Leg Channel Heads
+
Selected for its close match of thermal expansion properties to base
+
materials and its corrosion resistance
+
Installed without stress relief Like other uses of Alloy 600 in RCS, these nozzles have proven
+
susceptible to PWSCC Through 1995, over 50 nozzles in 9 units have been found with
+
PWSCC in the U.S.; San Onofre Units 2 and 3 had 9
+
SCE has a history of managing Alloy 600 in a conservative manner consistent with safety, ALARA, operational significance, and industry practice Page 27
) i sou nt n ta no n a E
EDISON Alloy 600 RCS Penetration Nozzle Program Page 28
Alloy 600 Planning
+ Repair Feasibility Including ALARA
+ Inspection Capabilities
+ Confirming Suspect Leakage
+ Industry Experience and Recommended Approaches
+
1993 Plan l
+
1995 Plan
+ 1997 Refueling Planning
+ SONGS Experience With Alloy 600
+ Safety Engineering Assessment Report
+ Future Plans
+ Mechanical Nozzle Seal Assembly Page 29
\\.
Repair Feasibility /ALARA Pressurizer Example Shown External repair was essential
+
(pad not needed for loop installations)
Makes loop repairs l.easible
+
Vessel / loop entry not required 4
Controls foreign material entry into
[
RCS PRESSURE
~ 2.3 personrem per liot leg nozzle and
/
- BOUNDARY WELD
- 4.4 personrem per pressurizer nozzle i
.M[:
- .:. f -
il
/-
Degradation is abandoned in place
[:
- \\/:
+
Exposes loop piping and vessel to boric
+
-n-acid V-V l
~
+
Considered interim in '93, now
~
k qualified as a permanent repair gj NOZZLE With increasing use at SONGS, a
- l?
+
I monitoring program is being added to V
PRESSURIZER l
Cycle 10 outage e e.w n
l Inspection Capabih. ties l
+
Inspection methods consist of:
l bare metal and Radio-Chemistry i
f
+
Bare metal inspection essential to early detection ofinstrument l
F (~'
nozzle leakage 1
y;y
./
m ur -
I
,,?;j '
~~ j,39[ $jj Q t
+
San Onofre set the precedent for f
bg
'Q.
-24 bare metal inspection of RCS nw n
$yh-,..
bd.Q instrument nozzles and use of i auM i
inspection windows l
. v,(.
. L,,.
1 y' '%
. ) *,
l Further refinement in 1997
.' b.,
2 '.,.
+
4 37. 3
- ' ' r)!
A clamp asxmbly may have obstructed i
" +. !?
view of nozzle to shell interface enough Mk.., '
[
to obscure a very small hypothetical f
-,QL leak from view.
A:7.-
I f
~
^ ' &$
i
. &{C.
?'. ~, ^ ^ :.
l&
Page 31 l
Confirming Suspect Leakage June '97 Inspection - Leakage Suspected but cleared by Radio-Chemistry July '97 Inspection - Obvious Leakage i(L%g:p%g=1 e
a i$%T i
g wmw.y dl~? f*ttign s
D% a Gj
's,
c e:
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i Bgyqmg Aq g% Q1p, a;gigh, x..
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+
Without bare metal visual inspection techniques this leakage would not have been detected
+
Radio-chemistry analysis limitations have been incorporated into the inspection program
+
Material heat NX9915 example shus.n Page 32
Industry Experience anc.
Recommended Approaches
+ Accurate PWSCC prediction is not feasible. Instead, susceptibility is a function of a number of variables.
i
+
This response has been backed up by safety analysis and root cause failure analysis performed by utilities, owners groups, vendors, and EPRI.
+
Premature preventative replacement has lead, at times, to a second round of replacement activities in pressurizers.
+ Industry has largely handled leakage from small-diameter, non-rolled, Alloy 600 penetrations by visualinspection.
Page 33
e 1993 Alloy 500 Plan
+
San Onofre Experience Pressurizer nozzle leakage One Unit 2 hot leg nozzle
+
Premises Absolute predictions could not be made, instead only relative susceptibility could be assigned aller weighing a number of factors (temperature, yield strength, residual stresses, weld, product fonn, cold work, contaminants, microstructure)
Truly pennanent repairs may not be feasible Low safety significance
+
Recommendations and Conclusions Low safety significance premise was confinued Replacement recommendations contingent on development of exteriorjoint design Contingency repair plans were in place Inspect nozzles at the beginning of each refueling Modify insulation to facilitate this inspection Page 34
9 1993 Alloy 600 Plan (con't)
Location Susceptibility Recommendation Disposition Vapor Space Low (Alloy 690 Visual Inspection Implemented since 92/93)
Water Space High Replace in 1995 Deferred through 1997 refuelings refuelings Heater Sleeves Moderate Visual inspection Implemented - based on CEOG assessment Hot Leg Moderate Consider Deferred through 1997 l
replacement in 1997 refuelings l
refuelings CEDM/ICI Moderate Volumetric Implemented, inspection in 1995 scheduled for 1998-9 refuelings refuelings RCS Cold Leg Low Visual inspection Implemented Page 35
1995 Alloy 600 Plan
+
Updated prior to 1995 refuelings
+
Plan Conduct nozzle inspection early in refueling using the improved insulation access Visual. isotopic, and borie acid inspections Repair leaking and questionable nozzles
+
Implementation No Unit 2 repairs were needed 2 Unit 3 hot leg nozzles were identified and repaired using the external repair 4 Unit 3. vapor space nozzles replaced after finding weld filler metal cracking
+
Conclusions (After 1995 Refuelings)
Plan was managing nozzle cracking San Onofre trends consistent with industry In-house external repair capability desired, but not a precondition to repiacement activities e,y y,
1997 Refue:ing Planning
+
Continued with 1995 Plan
+
Initiating pro-active replacement was considered as a means of improving overall outage efficiency Decided to continue with visual inspection
+
Low safety significance assumptions remained valid Minimized unnecessary radiation exposure 1
Consistent with NRC's published comments SONGS nozzles svere performing consistent with industry experience Primary chemistry was well-managed within industry guidelines More appropriate uses for resources existed Requested Safety Engineering Assessment following Unit 2 Outage
+
Page 37
i San Onofre Experience with Al oy 600 PWSCC in RCS Penetrations r
SONGS Unit 2 SONGS Unit 3 l""'i "
Total Crack ed")
Cracked")
Tot al Cracked'U Cracked ("
Nozzles Through Through Nozzles Through Through Installed 1995 1997 Installed 1995 1997 Pressurizer lleater Sleeses 30 30 a
instrument Nouk s 7
2 (2) 3 (3) 7 4 (3) 4 (3)
Reactor Vessel ficad CEDM Noules 91 91 Incore Instrument Nozzles 10 10 IIcad Vent Nonle 1
I Steam Generators 8
8 Primary Loop Ilot Leg Instrument Noules 32 I (1) 1 (I) 32 2 (2) 10 (7)
Cold Leg Instrument Nonles 12 12 1 (0)
Totals 190 3
4 190 6
15 Nonles from lleat 7630 11 I (1)
I (1) 13 2 (I) 6 (3)
(1) Nonles found cracked during leakage inspections or during presentisc replacements NOT E: Nonic count in parentheses exhibited confirmed through-wallleakage. Three of the Unit 3 hot leg instrument noules, and the cald leg mstrument nonle had leakage that was not confirmed by radio-chemistry.
Page 38
Safety Engineering Assessment Report (SEA 97-02)
+ Requested by SCE Management in March 1997 in l
response to Unit 2 pressurizer water space nozzle leak (Heat K248) l
+ Multi-Disciplinary Team - led by Nuclear Oversite
+ Used hindsight to identify " lessons learned" for future improvement
+ The report did not attempt to make a balancedjudgement of the prudency of decisions made.
Pace 39 I
a
Safety Engineering Assessment Reaort 1
(SEA 97-02) (con' 0
+ Recommendations
- Reduce outage impact risk by procuring and qualifying Mechanical Nozzle Seal Assemblies (awaiting NRC approval)
- Revise inspection procedure to provide for removal of l
obstructions and more detailed visual inspections (completed)
- Reassess Alloy 600 nozzle program after each refueling (in progress F/C 10/31/97)
Note: Pressurizer water space and hot leg nozzle replacement in Unit 3 was not recommended.
I' age Ja l
Safety Engineering Assessment Report (SEA 97-02)(con'0 Missed opportunity defined as a decision point, in which, with the use of
+
hindsight, a different decision could have been made which would have prevented outage impact.
Primary Missed Opportunity
+
occurred during the inspection of 2TE101, at the start of the Cycle 9 refueling outage, when the presence of shims and a retaining clamp did not allow a clear view of the nozzle and may have hidden evidence ofleakage."
+
Response
l The primary missed opportunity was corrected via improvements to the l
existing proactive inspection program. Specifically, shims and clamps i
are removed from Alloy 600 nozzles and additional inspections are performed during startup operations.
l Page 41 i
l Safety Engineering Assessment Report (SEA 97-02) (con't)
+ Two Secondary Previous Missed Opportunities "The first... occurred when a technical recommendation was changed from a proactive nozzle replacement program to an inspect-and-replace as necessary strategy."
+ Response:
- This opportunity refers to the work scope for Units 2 and 3 Cycle 8 refueling outage (February 1995).
- Management considered the technical recommendation and elected to defer nozzle replacement plans based on reasons previously described.
Page 42
Safety Engineering Assessment Report (SEA 97-02} (con't l
+ "The second... occurred when a decision was made to continue L
with the inspect-and-replace strategy in lieu of a plan to perforrn anticipated repairs as well as proactively replace some nozzles during Cycle 9 refueling."
+ Response:
- In the Fall of 1995, Management reconsidered additional proactive nozzle replacements.
- The development ofin-house nozzle replacement capabilities was desired and completed in December 1996.
{
Because of favorable Cycle 8 results in managing Alloy 600, proactive nozzle replacements were considered not required in the Cycle 9 refueling.
This decision was to be re-considered after Cycle 9.
rye.o
-....,.. =.......;
Future Plans
+ Edison is finalizing plans for:
Exterior half-nozzle weld repairs Cycle 9 mid-cycles and Cycle 10 Refueling Mechanical nozzle seal assembly Pressurizer water space Contingency plan for hot leg use Volumetric inspection of CEDM/lCI's during refuelings U3 scheduled for Cycle 10 U2 considering inspection in Cycle 10
+ Nozzle plan update will be issued by October 31,1997 Page4J
Mechanical Nozzle Seal Assemiy - MKSA O
^7^
Custom design for each nozzle
+
7h {])
orientation and type I i
+
Install in Mode 5 Avoids core off-load required
+
for bottom hot leg no7.7les i
Il\\
(1
]J,), O
+
More costly than replacement i
gc 7
for hot leg nozzles F/
~
/j
+
Less costly for steam generator i
V and pressurizer nozzles y
~-
+
On-site demonstration j
i g completed for pressurizer and j[
core-offload designs (HL RTD)
+
Code relief submitted to NRC Page J3
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an a
s.%-~---na s+~-*+vn s', an s a
.- *. ass >
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s~,*-**<a sa I
wO
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60 O-M l
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to 5 s cr>
l s: - s sO E
$ LLJ 3 l
l
l REGELATORY NX7630 PWSCC was properly identified, corrected by replacing nozzles with
+
Alloy 690 to prevent recurrence, and augmented inspections were preformed in accordance with Generic Letter 88-05 and Information Notice 90-10.
SCE could not have predicted which nozzles would develop PWSCC based on
+
heat alone, and therefore, short of replacing all Alloy 600 nozzles, could not l
have taken action to prevent all failures.
l The NRC and the Industry have concluded PWSCC has low safety significance,
+
and inspect and repair is an acceptable response to PWSCC (see examples).
- SCE confirmed it was bounded by the NRC assessments. The SONGS PRA estimated instrument nozzle PWSCC to be very low risk, CDF<lE-6, consistent with the definition given in DG-1061.
hge47 g
g
Example 1: January 1995 NRC Petition Denial DD-95-2 On January 26,1995, the Director, Office ofNuclear 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 Staffidentified to the Conunission 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. [ continued next page]
l rage.as l
l l
Example 1: DD-95-2 (Cont)
"The Staff determined that the safety significance of the 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 l
occur to the VHP or the reactor vessel. This would result in the deposition of boric acid crysta!s 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."
j Page 89 l
1 Example 1: DD-95-2 (Cont)
"CEOG submitted the detailed findings ofit's Alloy 600 component PWSCC-L
(
program initiated in 1990 to the Staffin a proprietary report on February 26,1992.
The conclusions of the report, which focused primarily on pressurizer heater sleeves and instrument nozzles, in part, follow:
l "l) Circumferential cracking of the heater sleeves and the instrumentation nozzles is not a credible failure mode...
"3) Visual inspection is the best method for detecting a leaking sleeve or nozzle...
"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 l
with respect to pressurizers are highly unlikely."
l Page50
[
?
Example 2: SECY 97-063, March 1997 Proposed NRC Generic Letter:" Degradation of Control Rod Drive Mechanism and Other Vessel Closure Head Penetrations" "Beginning in 1986, leaks have been reported in several Alloy 600 pressurizer instrument nozzles at both domestic and foreign reactors...The NRC staffidentified 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 facility.
The NRC staff reviewed the safety significance of the cracking that occurred, as well as the repair and replacement activities at the affected facilities. The NRC staff determined that the cracking was not ofimmediate safety significance because the cracks l
were axial, had a low growth rate, were in a material with an extremely high l
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 l
before a penetration would fail."
PageSI
Example 3: Generic Letter 97-01, April 1997 Generic Letter 97-01 addresses the issue of the potential for cracking in Inconel 600 CRDM nozzles and other ' vessel head closure penetrations (VHP).
"The NRC staff determined that the cracking was not ofimmediate 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."
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."
nyu
Example 4: November 1993 NRC Letter In a November 19,1993 letter from William T. Russell to William Raisin, the NRC responded to NUMARC's June 16,1993 letter regarding Inconel 600 CRDM/CEDM head penetrations. The NRC's conclusion is:
" Based on the overseas inspection findings and the review ofyour analyses, the staff has concluded that there is no immediate safety concern for cracking of the CRDM/CEDM penetrations.
" Based upon information received from overseas regulatory authorities, your analyses, and staff reviews, the staff believes that catastrophic failure of a penetration is extremely unlikely. Rather, a flaw would leak before it reached the critical flaw size...."
hge 53
Example 5: NRC IN 90-10 "The cracking to date in the thennal sleeves and the instrument nozzles of the domestic PWRs has been reported as being only axially oriented. The safety implication of an axial crack is not considered a significant threat to the structural integrity of the components and most likely will result in a small leak...Circumferential cracking poses a more serious safety concern because ifit were to go undetected it could lead to a structural failure of a component rather than to a limited leak."
2 may be prudent for licensees of all PWRs to review their Alloy 600 cnpiie' ions in the primary coolant pressure boundary, and when nctessa.y, to implement an augmented inspection program."
Pace $4
i
)
i i
Example 6: NRC Status Report to the Commission On May 12,1993, the NRC staff provided a status report to the Commission regarding PWSCC of Alloy 600 components. The NRC concluded the following at that time:
l "Having reviewed the information to date, including the inspection results ar'l findings, the staff maintains its view that this issue is oflow safety significance since all cracks reported to date, with perhaps one exception a.k.a. French reactor, are short in length and axially oriented in an extremely flaw-tolerant material."
Page 55
U TLME LINE The time line (separate) contains entries on: NRC docketed material regarding l
PWSCC, CEOG and EPRI reports on PWSCC, NRC inspections at SONGS of PWSCC and closeout of LERs; and PWSCC events.
1 Pace %
CCNCLUSION Criterion XVI
Response
Nonconfonnance is promptly identified Each NX7630 Nozzle with PWSCC was and corrected.
replaced with Alloy 690 to preclude repetition.
l Cause of condition is determined and Each NX7630 PWSCC leak was corrective action taken to preclude promptly identified and corrected by repetition.
replacement with Alloy 690 to l
preclude repeat leakage.
I Heat NX7630 history of PWSCC is comparable with other heats.
SCE finds no reason to single out Heat NX7630, for special replacement decisions, as compared with all Alloy 600 nozzles.
Page 57
ENCLOSURE 3 LICENSEE LETTER OCTOBLA 3,1997 k
1
-. -..