ML17291B170
| ML17291B170 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/15/1995 |
| From: | Parrish J WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GO2-95-275, NUDOCS 9512270170 | |
| Download: ML17291B170 (24) | |
Text
RIC) RITY PACCELERATED RIDS PROCESSING)
REGULATORY INFORMATION DXSTRIBUTION SYSTEM (RIDS)
ACCESSION NBR: 9512270170 DOC. DATE: 95/12/15 NOTARIZED: NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit. 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION PARRISH,J.V. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards proposed alternative exams for 10CFR50.55a(g)(6),
augmented exam of RV 6 requests that proposed alternate to 10CFR50.55a(g)(6) be accepted as revised relief request 2-ISI-001 for first 10-yr ISX interval.
DISTRIBUTION CODE: A047D TITLE: OR COPIES RECEIVED:LTR Submittal: Inservice/Testing/Relief f ENCL L SIZE:
om ASME Code I9 GL-89-04 NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD4-2 LA 1 1 PD4-2 PD 1 1 CLIFFORD,J 1 1 INTERNAL: ACRS 6 0 AEOD/S PD/RAB 1 1 CONTE 1 1 1 NRR/DE/ECGB 1 1 EMCB 1 1 NRR/DE/EMEB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 RES/DET/EMMEB 1 1 RES/DSIR/EIB 1 1 EXTERNAL: LITCO ANDERSON 1 1 NOAC NRC PDR 1 1 NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE iVASTE! CONTACTTHE DOCUMENT CONTROL DESK, ROOM PI -37 (EXT. 504-2083 ) TO ELlib!I.'PATE YOUR NAXIE P ROil DISTRIBUTION LISTS I'OR DOCUME. FTS YOU DON "I'LED!
TOTAL NUMBER OF COPIES REQUIRED: LTTR 21 ENCL 14
WASHINGTON PUBLIC POWER SUPPLY SYSTEM PO. Box 968 ~ 3000 George Washington Way ~ Richland, Washington 99352-0968 ~ (509) 372-5000 December 15, 1995 G02-95-275 Docket No. 50-397 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
WNP-2, OPERATING LICENSE NPF-21 ALTERNATIVETO 10 CFR 50.55a (g)(6)(ii)(A), AUGMENTED EXAMINATIONOF REACTOR VESSEL
References:
- 1) Letter G02-93-166, dated June 24, 1993, JV Parrish (SS) to NRC, "Inservice Inspection Program Plan Relief Requests 2-ISI-001 and 2-ISI-P1P"
"Inservice Inspection Relief Requests - WPPSS Nuclear Project No 2 (TAC No. M87253)"
- 3) NRC letter dated February 1, 1995, JW Clifford (NRC) to JV Parrish (SS), "Evaluation of the First 10-Year Interval Inservice Inspection Program Plan, Amendment 1, for Washington Public Power Supply System Nuclear Project No. 2 (TAC No. M87807)"
- 4) Letter GO2-85-274 dated May 29, 1985, GC Sorensen (SS) to WR Butler (NRC), "Nuclear Plant No. 2, Inservice Inspection Program Plan, Relief Requests"
- 5) Letter dated March 27, 1987, EG Adensam (NRC) to GC Sorens'en (SS),
"Safety Evaluation for 1st Ten-Year Interval Inspection Program and Requests for Relief from Certain Requirements" This letter submits for NRC review, the Supply System's proposed alternative examinations for 10 CFR 50.55a(g)(6). The Supply System also requests that the proposed alternate to 10 CFR 50.55a(g)(6) be accepted as the revised relief request 2-ISI-001 for the first 10-year inservice inspection interval. The alternative proposal is attached to this letter.
9512270170 951215 PDR ADOCK 05000397 PDR Wka Aa A t WAlko WAI
Page 2 ALTERNATIVETO 10 CRF 50.55a'g)(6)(ii)(A), AUGMENTED EXAiVHNATIONOF REACTOR -VESSEL By Reference 1, the Supply System submitted revised relief request 2-ISI-001 in response to 10 CFR 50.55a(g)(6) with the intent that this was for the augmented reactor pressure vessel (RPV) examinations. Reference 2 denied the relief request stating that it was not presented as an alternate to 10 CFR 50.55a(g)(6). Reference 3 stated that the matter was still open and was to be resolved by the end of the first inspection period of the second inspection interval which ends April, 1998.
During the fall of 1994, the BWR Vessel and Internals Project (BWRVIP) began work on an industry alternative to 10 CFR 50.55a(g)(6). As a subscriber of the BWRVIP, the Supply System decided to wait for the conclusion of the BWRVIP work before presenting our response to 10 CFR 50.55a(g)(6). On September 28, 1995 BWRVIP submitted their recommendations to the NRC as ERPI report TR-105697 (BWRVIP-05). It is understood by the Supply System that this report has not been endorsed by the NRC at this time. However certain information from the report has been referenced to provide additional support for the Supply System position for our alternate proposal.
It is also requested that Table 2 from the alternative proposal (attachment) be used as a replacement for the coverage obtained for all RPV welds in relief request 2-ISI-001. Table 2 provides for more coverage of the RPV than what was originally submitted to the NRC in revision 0 of the ISI Program Plan for the first inspection interval (Reference 4). This relief request, 2-ISI-001, was approved by Reference 5.
Should you have any questions or desire additional information regarding this matter, please call me or D.A. Swank at (509) 377-4563.
Sincerely,
. V. Parrish (Mail Drop 1023)
Vice President, Nuclear Operations DPR/lm Attachment CC: LJ Callan - NRC RIV KE Perkins, Jr. - NRC RIV, Walnut Creek Field Office NS Reynolds - Winston & Strawn JW Clifford - NRC DL Williams - BPA/399, NRC Sr. Resident Inspector - 927N
ALTERNATIVETO 10 CFR 50.55a (g)(6)(ii)(A) dated August 31, 1992 WNP-2 has determined that the augmented examinations required by 10 CFR 50.55a (g)(6)(ii)(A), dated August 31,1992', cannot be performed completely. This submittal proposes alternative examinations that would provide an acceptable level of quality and safety.
Required Examinations The examination requirements of 10 CFR 50.55a (g)(6)(ii)(A), dated August 31,1992~
require that-all licensees augment their 'reactor vessel (RPV) examinations for ASME Section XI Examination Category B-A, Item Number B1.10, by implementing once, as part of the inservice inspection interval in effect on September 8, 1992, the examination requirements for reactor vessel shell welds specified in Item B1.10 of Examination Category B-A of the 1989 Edition of ASME Section XI. These requirements specify that essentially 100% of the weld for each weld be examined. The regulations define "essentially 100%" as more than 90% of the examination volume of each weld [10 CFR 50.55a (g)(6)(ii)(A)(2)].
Proposed Alternative Examinations The alternative examinations proposed in place of the augmented examinations defined in 10 CFR 50.55a are to perform examinations of greater than 90% of the total weld volume in Examination Category B-A, Item Number Bl.10 and greater than 80% of the total weld volume in Examination Category B-A (Item Numbers B1.10, B1.21, B1.22, B1.30, and B1.40) during the first inspection interval, rather than 90% of the weld volume for each weld. It is requested that the examinations performed during the first inspection interval be regarded as the alternative examinations as allowed in 10 CFR 50.55a (g)(6)(ii)(A)(4)'.
Weld volume is defined as that area requiring examination per ASME Section XI, 1989 Edition, FiguresIWB-2500-1, IWB-2500-2, IWB-2500-3, IWB-2500-4, and IWB-2500-5.
Basis for Requesting Alternative Examinations During the performance of the examinations it was determined that two of. the 18 item number B1.10 welds could not be examined to the full ASME Section XI Code coverage required by 10 CFR 50.55a (g)(6)(ii)(A). Weld AB (see Figure 1, RPV-101) received a 79% examination volume coverage and weld AD (see Figure 1, RPV-101) received 83% examination volume coverage. The total weld volume coverage for all item B1.10 welds exceeded 90%. The total weld volume for all Examination Category B-A welds examined exceeded 84%. WNP-2 was in the first inspection interval when the augmented examinations became effective and was scheduled to perform examinations on essentially 100% of the welds volume per the requirements of the 1980 Edition, Winter 1980 Addenda of Section XI. The examination requirements for the first inspection interval from the 1980 Edition, Winter 1980 Addenda are the same as that required by the 1989
I I
i$
ATTACHMENT Page 2 Edition referenced in 10 CFR 50.55a.
- 4. Justification for Using Alternate Examinations The Supply System has reviewed the Boiling Water Reactor Vessel and Internals Project report "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05) "4 submitted to the NRC September 28, 1995'. After reviewing BWR fabrication practices, in service inspection data, operational issues, degradation mechanisms, non-destructive examination capabilities and probabilistic fracture mechanics analysis results, the report concluded that 50% of the longitudinal shell welds (Category B-A, Item No. B1.12 should be volumetrically examined and that no volumetric examinations should be performed on the circumferential shell welds (Category B-A, Item No. B1.11). The alternative proposed by the Supply System exceeds this recommendation in that over 90% of the total weld volume of Category B-A, Item No.
B1.11 and B1.12 is examined.
Augmented RPV examinations were mandated by 10 CFR 50.55a (g)(6)(ii)(A) due to NRC concerns of RPV degradation due to irradiation, stress corrosion cracking, service induced cracking and early Editions of ASME Section XI only requiring a small percentage of the RPV welds to be examined in later inspection intervals'. The following sections address these concerns as they relate to WNP-2.
Irradiation Effects Radiation embrittlement (at the end of 40 years) is not a problem outside of the vessel beltline region because the irradiation of those areas is less than 1 X 10" nvt with neutron energies in excess of 1 MeV7. At the conclusion of cycle 11 in Spring, 1996, WNP-2 will have approximately 8 effective full power years. This corresponds to a fluence of less than 1 X 10" n/cm'ith neutron energies in excess of 1 MeV't the RPV. Therefore at the end of cycle 11, irradiation contribution to brittle fracture of the RPV including the beltline region is not a concern at WNP-2. In general, due to the design and operating conditions of BWRs, irradiation contributes little to the brittle fracture of the vessels'.
Stress Corrosion Cracking (SCC) at WNP-2 The Supply System has been committed to the BWR Water Chemistry Guidelines""
since their inception. These guidelines and other proactive efforts have enabled the Supply System to maintain a high level of water quality that helps mitigate SCC.
AITACHMENT Page 3 Yearly averages of reactor water conductivity during the first inservice inspection interval are12>>
Cycle Cond. pS/cm 1 - 1986 .344 2 - 1987 .237 3 - 1988 .241 4 - 1989 .218 5 - 1990 .170 6 - 1991 .196 7 - 1992 .192 8 - 1993 .154 9 - 1994 .162 10- 1995 175" The only instances of stress corrosion cracking in BWR vessel shells have been associated with attachments and cladding, and these are self limiting due to the lack of a sustained driving force'4. There has been no identified instances of stress corrosion cracking at WNP-2.
Service Induced Cracking WNP-2 has had three incidents of service induced (fatigue) cracking of internal components. Two are attributed to vibrational fatigue. The third is also thought to be due to vibrational fatigue. None of the cracking involved the RPV shell. The cracking has not been associated with vessel internals that are attached to the RPV shell.
Failure of the weld holding one of the surveillance specimen holders was found during refueling outage four (4). Failure was attributed to vibration fatigue and an under sized weld.
During refueling outage nine (9) a crack was discovered in one jet pump sensing line, The crack is attributed to vibration fatigue.
Two of the 80 tack welds on the jet pump adjusting screws were found cracked during refueling outage ten (10). The cracking is attributed to fatigue.
The significant inservice cracking which has occurred in large pressure vessels designed and fabricated to the ASME Code has been limited to PWRs. No instances of significant service induced cracking of BWR pressure vessel low alloy shell material have been identified'~. Instances of inservice cracking at WNP-2 have been associated with vessel internals and have not involved the vessel shell,
ATTACHMENT Page 4 Examination Volume Obtained Code Volume of B1.10 welds Examined During the first inservice inspection interval greater than 91% of the total ASME Code weld volume for item B1.10 welds received a complete Code examination (see Table I). Two welds, AB and AD, did not receive the 10 CFR 50.55a (g)(6)(ii)(A) required minimum examination coverage of greater than 90% of weld volume.
Circumferential weld AB, in the beltline region, was examined to 79% of the full ASME Code coverage requirements. Transducer contact was lost due to the transition from a 9-7/16" plate to a 6-7/16" plate. The coverage for this weld is diagramed in Figures 2 through 7.
Circumferential weld AD was examined to 83% of full ASME Code coverage.
The lack of coverage from one side is due to the RPV stabilizer lugs welded to the vessel wall which limited transducer access. The weld received 100% scan coverage (entire weld length) from one side and 83% scan coverage from the other side.
The BWRVIP-05 report concluded that the probability of circumferential welds developing unacceptable flaws was very low'~. Both AB and AD are circumferential welds.
Volume of RPV welds covered Total RPV weld volume (Examination Category B-A, item numbers B1.10, B1.21, B1.22, B1.30 and B1.40) examined is greater than 84% of the Code required volume. As can be seen in Table II, they are representative of all locations within the vessel. In particular, 4 of the 5 beltline region welds have received coverage of over 90% of the Code required volume. The fifth weld, AB, received coverage of 79%. The major repair area within the beltline region received a 100% Code volume examination. The repair area measures 15 inches wide, 30 inches high and ranged from 2-3/4 to 3-7/8 inches deep. Greater than 86% of the beltline weld volume has been examined.
Results of First Inspection Interval RPV ISI Examinations No indications that exceed ASME Section XI acceptance criteria were found during the first inspection interval.
A%I'ACHMENT Page 5 Results of RPV Preservice Inspection Examination (PSI)
A manual preservice UT examination of essentially 100% of the RPV circumferential and longitudinal welds was performed prior to RPV installation' Results of PSI examination found no unacceptable indications".
Examinations from Internal Surface (ID) of RPV To assess what examination volume coverage was possible by doing the examinations using a technique that examined the welds from the inside (ID) of the RPV, General Electric performed an access study" from the inside of the RPV. The study estimated that approximately 84% of the RPV shell welds (B1.10) can be fully examined per ASME Code requirements using a combination of ID and OD techniques". This coverage was approximately the same percentage coverage that could be obtained using only the OD technique with some improvements to the tooling to increase coverage.
When the OD tooling was used with the improved tooling greater than 90% of the shell welds volume received a full Code examination.
When the OD Technique is supplemented by the ID tooling, the only definitive benefit from using ID tooling would be to increase coverage of weld AD by 7%. Adding ID tooling may increase total coverage of shell welds by less than 1% based on the access study. There is no assurance from the study that the coverage of weld AB would be increased by using the ID technique.
Plant design provided for access from the OD not from the ID. It was therefore prudent to perform the examinations from the OD where there was a known coverage, a known baseline, and known access to the welds.
Examination history of the BWR fleet As part of BWRVIP efforts a survey was conducted of the 37 domestic BWRs requesting information on their vessel weld examination history. Twenty-nine responded. Based on these responses 5,236 feet of vessel weld have been examined with 16 indications exceeding Section XI acceptance criteria detected. All 16 flaws were determined to be subsurface and acceptable per IWB-3600 analysis. The survey concluded that the BWRs were free of unacceptable fabrication defects and no flaw has developed during operation.
Evaluation supports the conclusion that there are no degradation mechanisms that have affected the BWR fleet's shell welds. (BWRVIP-05")
- 4. Conclusions
- 1. Irradiation embrittlement is minimal for the relatively low fluence environment of a BWR and the low EFPY that WNP-2 has experienced.
ATTACHMENT Page 6
- 2. WNP-2 has had good water chemistry from start-up so stress corrosion cracking concerns are minimal. The only instances of SCC in BWRs has been associated with attachments and cladding which these augmented examinations would not detect. WNP-2 has not experienced any instances of SCC.
- 3. Greater than 91% of the total weld volume of Examination Category B-A, Item Number B1.10 is receiving a full ASME Section XI Code volume examination.
- 4. Greater than 86% of the total weld volume in the beltline region, including 100% of the repair area, is receiving a full ASME Section XI Code volume examination.
- 5. Greater than 84% of the entire RPV weld volume is receiving a full ASME Section XI'Code volume examination.
- 6. A large percentage of the RPV shell weld (B1.10) volume is being examined.
- 7. A large percentage of the RPV weld volume as a whole is being examined. This volume represents every weld in the RPV.
- 8. Results from the first interval examinations found no unacceptable indications. This parallels the findings of the BWR fleet.
- 9. Use of ID tooling would not significantly increase the examination volume covered over that achieved only with OD tooling.
- 10. WNP-2 was in its first inspection interval when the augmented requirements were issued and was therefore required by Section XI Code to perform 100% examination of Category B-A welds.
The proposed alternative examinations meet the technical objective of the augmented examinations defined in 10 CFR 50.55a (g)(6)(ii)(A), dated August 31,1992 in that a large percentage of the RPV is being examined. The proposed alternative, performed within the time frame called out in the regulation, demonstrated that there are no unacceptable indications within the volume examined which provides assurance of RPV shell weld integrity. The large volume of weld that has been examined without any unacceptable indications demonstrates that the RPV weld integrity has been maintained. The alternative examinations proposed provide an acceptable alternative to 10 CFR 50.55a (g)(6)(ii)(A) for managing the integrity of the RPV welds.
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60'"scan Frori bottom side c/s area examined 65.3 I(i
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ATI'AC HMENT Page 14 Table I Examination Coverage Obtained During First Inspection Interval for Examination Category B-A, Item Bl.10 Welds Code Code Weld length Weld Area Weld Volwe Percent Volwe Scanned Ident. Ho. Description (inches) (square inch) (cubic inches) Examined (cubic inches)
AA BTH HD-SC¹1 WD 840 71.8 60312 99.8 60191 AB ¹1-¹2 SC CRC WD 840 70.5 59220 79.7 47198 AC ¹2-¹3 SC CRC WD 840 57.9 48636 92.4 44940 AD ¹3-¹4 SC CRC WD 840 60.7 50988 83.6 42626 BA ¹1 SC VRT WDQ45 130 119.8 15574 90.8 14141 BB ¹1 SC VRT WQ135 130 119.8 15574 91.9 14313 BC ¹1 SC VRT WQ225 130 119.8 15574 91.2 14203 BD ¹1 SC VRT WQ315 130 119.8 15574 92.8 14453 BE ¹2 SC VRT WQ 10 175 53.2 9310 98.1 9133 BF ¹2 SC VRT WQ100 175 53.2 93'10 95.6 8900 BG ¹2 SC VRT WQ190 175 53.2 9310 98.1 9133 BH ¹2 SC VRT WQ280 175 53.2 9310 93.7 8723 BJ ¹3 SC VRT WQ 50 169 61.9 10461 91.7 9593 BK ¹3 SC VRT WQ170 169 61.9 10461 96.8 10126 BH ¹3 SC VRT WQ290 169 61.9 10461 92.1 9635 BH ¹4 SC VRT WQ330 146 66.8 9746 100.0 9746 BP ¹4 SC VRT WQ 90 146 66.8 9746 100.0 9746 BR ¹4 SC VRT WQ210 146 66.8 9746 100.0 9746 Total Weld Length Total Weld Volwe Total Code 'Weld VolwL Scanned 5,525 inches 379,314 cubic inches 346,547 cubic inches Percent of Total Code Meld Volwe Examined = 91.4
ATTACHMENT Page 15 Table II Examination Coverage Obtained During First Inspection Interval Examination Category B-A RPV COVERAGE Code Code Iden t. Item 'Meld length Weld Area Weld Volune Percent Volune Scanned Ho. Ho. Description (inches) (square inch) (cubic inches) Examined (cubic inches),
AA 81.10 BTM HD-SC¹1*WD 840 71.8 60312 99.8 60191 AB 81.10 ¹1-¹2 SC CRC WD 840 70.5 59220 79.7 47198 AC 81 10 ¹2-¹3
~ SC CRC WD 840 57.9 48636 92.4 44940 AD 81.10 ¹3-¹4 SC CRC WD 840 60.7 50988 83.6 42626 AE 81.30 ¹4 SC-FL CRC WD 792 74.3 58846 49.4 29070 AG 81.40 TOP HD-FLG WELD 783 36.6 28658 95.1 27254 AH 81 ~ 21 TOP HD DOL PLT 516 31.4 16202 100 ' 16202 AJ 81.21 BOT HD DOL WELD 684 70.4 48154 100.0 48154 BA 81.10 ¹1 SC VRT W0945 130 119.8 15574 90.8 14141 BB 81.10 ¹1 SC VRT WQ135 130 119.8 15574 91.9 14313 BC 81.10 ¹1 SC VRT IS225 130 119.8 15574 91 ~ 2 14203 BD 81 ~ 10 ¹1 SC VRT IS315 130 119.8 15574 92.8 14453 BE 81 ~ 10 ¹2 SC VRT IS 10 175 53.2 9310 98.1 9133 BF 81 ~ 10 ¹2 SC VRT WQ100 175 53.2 9310 95.6 8900 BG 81.10 ¹2 SC VRT W9190 175 53.2 9310 98.1 9133 BH 81.10 ¹2 SC VRT W9280 175 53.2 9310 93.7 8723 BJ 81.10 ¹3 SC VRT WQ 50 169 61.9 10461 91 ~ 7 9593 BK 81.10 ¹3 SC VRT WQ'170 169 61.9 10461 96.8 10126 BH 81.10 ¹3 SC VRT IS290 169 61.9 10461 92.1 9635 BH 81.10 ¹4 SC VRT W9330 146 66.8 9746 '100.0 9746 BP B1.10 ¹4 SC VRT WQ 90 146 66.8 9746 100.0 9746 BR 81.10 ¹4 SC VRT IS210 146 66.8 9746 100.0 9746 DA 81.22 BOT HD HRD Q272 56 62.7 3499 78.6 2750 DB 81.22 BOT HD HRD 9332 56 62.7 3499 78.6 2750 DC 81.22 BOT HD MRD 9 32 56 62.7 3499 78.6 2750 DD 81.22 BOT HD HRD 9 92 56 62.7 3499 78.6 2750 DE 81.22 BOT HD HRD 9152 56 62.7 3499 78.6 2750 DF 81 ~ 22 BOT KD HRD 9212, 56 , 62.7 3499 78.6 2750 DG 81.21 BOT KD DOL /270 219 77.3 16944 16.7 2830 DH 81.22 TOP KD MRD Q15 78 41.0 3198 100.0 3198 DJ 81 ~ 22 TOP HD HRD 975 78 41.0 3198 100 ' 3198 DK 81.22 TOP HD MRD 9135 78 41.0 3198 100.0 3198 DH 81 ~ 22 TOP HD MRD 9195 78 41.0 3198 100.0 3198 DH 81.22 TOP HD MRD Q255 78 41.0 3198 100.0 3198 DP 81.22 TOP HD MRD 9315 78 41.0 3198 100.0 3198 DR 81.21 BOT HD DOL / 90 219 77.3 16944 16.7 2830 Total Weld Length Total Weld Volune Total, Code Weld Volune Scanned 9,541 inches 605,241 cubic inches 508,574 cubic inches Percent of Total Code Weld Voluae Examined = 84.0
ATTACHMENT Page 16 References
- 1. Part 10 Code of. Federal Regulations 50.55a, published August 31, 1992
- 2. ibid.
- 3. ibid.
- 4. "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)", EPRI TR-105697, September 1995
- 5. Letter dated September 28, 1995, JT Beckham, Jr, BWR Vessel & Internals Project, to NRC, Attention: CE Carpenter, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05) ", EPRI TR-105697, September 1995
- 6. Federal Register, Volume 56, No. 21, Published January 31, 1991
- 7. WNP-2 FSAR section 5.3.3.1.1.1, pg 5.3-14
- 8. Calculation ME-02-89-58, Revision 1, dated November 8, 1993
- 9. EPRI Report TR-105697, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)", September 1995, pg 1-2
- 10. EPRI TR-103515, "BWR Water Chemistry Guidelines - 1993 Revision; Normal and Hydrogen Water Chemistry, February, 1994
- 11. Guidelines for Chemistry at Nuclear Power Stations, INPO 88-021 Revision 1, September 1991
- 12. Data for cycles 1 to 9 from Supply System Letter GO2-94-202, JV Parrish to NRC, "WNP-2, Operating License NPF-21 Response to Generic Letter 94-03, "Intergranular Stress Corrosion Cracking of Core Shroud", dated August 24, 1994
- 13. Data for cycle 10 is from Statistic Report CM-RP-ST-Rl (internal Supply System document)
- 14. EPRI Report TR-105697, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)", September 1995, pg 1-3
- 15. ibid., pg 1-3
- 16. "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)", EPRI TR-105697, September 1995, pg 9-2
ATTACHMENT Page 17
- 17. "Manual Ultrasonic Pre-Service Examination of WNP-2 Reactor Pressure Vessel",
Nuclear Energy Services, Inc., February, 1977
- 18. ibid., page 3 of 3
- 19. RDE No. 54-0991 Rev 0, Accessibility of the Hanford Unit 2 Reactor Pressure Vessel Seam Welds for an Ultrasonic Examination, General Electric Nuclear Energy, November 1, 1991.
- 20. ibid, page 3
- 21. EPRI Report TR-105697, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)", September 1995, pg 4-9