ML17292A885
| ML17292A885 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 06/05/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17292A884 | List: |
| References | |
| NUDOCS 9706120145 | |
| Download: ML17292A885 (17) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION UGMENTED REACTOR VESSEL EXAMINATION WASHINGTON PUBLIC POWER SUPPLY SYSTEM WASHINGTON NUCLEAR PROJECT UNIT 2 DOCKET NO. 50-397 I.O RNTR OO I
N The Technical Specifications (TS) for Washington Nuclear Project, Unit 2 (WNP-2) state that the inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1,
2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (BKPV) Code and applicable addenda as required by 10 CFR 50.55a(g),
except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
Additionally, 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4),
ASME Code Class 1,
2 and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME
- Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components,"
to the extent practical within the limitations of design,
- geometry, and materials of construction of the components. 'he regulations require that inservice examination of component's and system pressure tests conducted during the first ten-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.
The applicable edition, of Section XI of the ASME Code for the WNP-2 first ten-year inservice inspection (ISI) interval is the 1980 Edition thr ough Winter '1980 Addenda.
Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement.
After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i),
the Commission may grant relief and may impose 9706120145 970605 PDR ADQCK 05000397 P
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alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
In a letter dated December 15,
- 1995, Washington Public Power Supply System (WPPSS),
submitted to the NRC its alternative to the 10 CFR 50.55a(g)(6)(ii)(A) augmented examination of reactor pressure vessel (RPV) for WNP-2.
A conference call was held with WPPSS and our contractor, the Idaho National Engineering and Environmental Laboratory (INEEL) on March 11,
- 1997, to clarify the December 15, 1995, request related to reliefs other than an alternative to the 10 CFR 50.55a(g)(6)(ii)(A) augmented reactor pressure vessel examination.
WPPSS clarified that its submittal also included a number of American Society for Mechanical Engineers (ASNE) Boiler and Pressure Vessel
- Code,Section XI, Examination Category B-A, Item B1.21 and Bl.22 welds.
Since 10 CFR 50.55a(g)(6)(ii) only addresses those welds specifically categorized as RPV shell welds (Item Bl. 10), the Item B1.21 and Bl.22 welds are outside the scope of the augmented reactor pressure vessel (RPV) examination.
WPPSS proposed to submit requests for relief for the B1.21 and B1.22 welds under 10 CFR 50.55a(g)(5)(iv) which will be evaluated separately.
2.0 EVALUATION The staff, with technical assistance from its contractor, INEEL, has evaluated the information provided by the licensee for its alternative to the 10 CFR 50.55a(g)(6)(ii)(A) augmented examination of RPV for WNP-2.
Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the attached Technical Letter Report (TLR).
The staff determined that the augmented coverage requirements cannot be met for two circumferential shell welds due to physical restrictions that limit scan coverage.
For Weld AB,,weld configuration resulted in loss of transducer contact and limited coverage to 79 percent of the required volume.
For Weld AD, RPV stabilizer lugs obstructed the examination areas and limited coverage to 83 percent of the required volume.
To achieve complete coverage for the subject welds, design modifications would be required to increase access from the outside surface (OD).
The licensee made a reasonable effort to maximize examination coverage of their reactor vessel.
For the WNP-2 reactor vessel, the licensee considered examining the inside surface and had General Electric perform an access study.
It was determined by the licensee that little or no increase in coverage could be achieved by performing an inside diameter (ID) examination of these welds.
Using OD examination techniques, the licensee has examined in excess of 91 percent of the cumulative length of the RPV shell welds.
This level of examination coverage is significant and should have detected inservice degradation, if present.
Therefore, the licensee's proposed alternative provides an acceptable level of quality and safety.
- 3. 0 CONCLUSIONS Based on the information submitted, the staff adopts the contractor's conclusions and recommendations.
The staff has concluded that the licensee has exami'ned a significant portion of the reactor vessel welds and that service-induced degradation, if present, would have been detected.
Thus, the licensee's proposed alternative provides an acceptable level of quality and
- safety, and the licensee's proposed alternative is authorized pursuant to, 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(ii)(A).
Attachment:
Technical Letter Report Principal Contributor:
T. HcLellan Date:
Oune 5, 1997
ECHNICAL LETTER REPORT LTERNATIVE TO 10 CFR 50.55a 6 ii A
UGMEN ED ACTOR PRESSUR VESS AMINAT 0 WASHINGTON PUBLIC POWER SUPP SYSTEM WPPSS NUC AR PROJECT UNIT 2 DOCK T NUMBER 50-397
1.0 INTRODUCTION
By letter dated December 15, 1995, the licensee, Washington Public Power Supply System (WPPSS),
proposed an alternative to the augmented examination of the reactor pressure vessel (RPV) required by 10 CFR 50.55a(g)(6)(ii)(A).
The licensee's submittal also included a number of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel
- Code,Section XI, Examination Category B-A, Item B1.21 and B1.22 welds.
10 CFR 50.55a(g)(6)(ii) only addresses those welds specifically categorized as RPV shell welds (Item Bl. 10).
Welds other than Item Bl. 10 are outside the scope of the augmented RPV examination and, therefore, not addressed in this report.
The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the information provided by the licensee regarding this alternative in the following section.
2.0 EVALUATION The information provided by the licensee in support of the proposed alternative has been evaluated and the basis for disposition is documented below.
Alternative to 10 CFR 50.55a 6 ii A
Au mented Reactor 'Pressure Vessel Examination Re ulator Re uirement:
In accordance with 10 CFR 50.55a(g)(6)(ii)(A), all licensees must implement once, as part of the inservice inspection interval in effect on September 8,
- 1992, an augmented volumetric examination of the RPV welds specified in Item Bl. 10 of Examination Category B-A of the 1989 Edition of the ASME Code,Section XI.
Examination Category B-A, Items Bl.ll and Bl.12 require volumetric examination of essentially 100X of the RPV circumferential and longitudinal shell welds, as defined by Figures IWB-2500-1 and -2, respectively.
Essentially
- 100X, as defined by 10 CFR 50.55a(g)(6)(ii)(A)(2),
is greater than 90X of the examination volume of each weld.
icensee's Pro osed Alternative:
The licensee requests approval of the limited examination of Welds AB and AD in lieu of examining essentially 100M of each RPV shell weld as required by 10 CFR 50.55a(g)(6)(ii)(A).
The augmented examination requirements have been satisfied for the other 16 RPV shell welds at WNP-2.
The coverages for all the welds are listed in the table below.
Circumferential Item Bl.ll Weld ID X Examined Longitudinal Item B1.12 Weld ID X Examined AA AB AC AD 99.8 79.7 92.4 83.6 BA BB BC BD BE BF BG BH BJ BK BH.
BN BP BR 90.8 91.9, 91.2 92.8 98.1 95.6 98.1 93.7 91.7 96.8 92.1 100 100 100 icensee's Basis for the Pro osed Alternative (as stated):
"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 ASHE Section XI Code coverage required by 10 CFR 50.55a(g)(6)(ii)(A).
RPV-101) received 79X examination volume coverage and Weld AD (see Figure 1,
RPV-101) received 83X examination volume coverage.
The total weld volume coverage for all item Bl.10 welds exceeded 90X.
The total weld volume for all Examination Category B-A welds exceeded 84X.
WNP-2 was in the first inspection interval when the augmented examinations became effective and was scheduled to perform examinations on essentially 100X of the weld 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 Edition reference'd in 10 CFR 50.55a.
Justification for Using Alternate Examinations "The Supply System has reviewed the Boiling Water Reactor Vessel and Internals Project report BMR Reactor Pressure Vessel Shell Meld inspection Recommendations (BWRVIP-05) submitted to the NRC September 28, 1995.
After reviewing the BWR fabrication practices, inservice inspection data, operation
- issues, degradation mechanisms, non-destructive examination capabilities and probabilistic fracture mechanics analysis results, the report concluded that 50X of the longitudinal shell welds (Category B-A, Item No. BI. 12) should be volumetrically examined and that no volumetric examinations should be performed on the circumferential shell welds (Category B-A, Item No. Bl.ll).
The alternative proposed by the Supply System exceeds this recommendation in that over 90X of the total weld volume of Category B-A, Item No. Bl. 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 ASHE 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 10'8 nvt with neutron energies in excess of 1 HeV.
At the conclusion of cycle 11 in Spring,
- 1996, WWP-2 will have approximately 8 effec)ive full power years.
This corresponds to a fluence of less than 1 x 10'/cm with neutron energies in excess of 1 HeV at the RPV.
Therefore at the end of cycle ll, irradiation contribution of brittle fracture of the RPV including the beltline Not included in this report.
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.
Yearly averages of reactor water conductivity during the first inservice inspection interval are:
~Cc1 e 1
- 1986
.334 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 instance of stress corrosion cracking in BWR vessel shells have been associated with attachment and cladding, and these are self limiting due to the lack of a sustained driving force.
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.
Code Volume of Bl. 10 Welds Examined "During the first inservice inspection interval greater than 91X of the total ASME Code weld volume for item Bl. 10 welds received a complete Code examination.
Two welds, AB and AD, did not receive the 10 CFR 50.55a(g)(6)(ii)(A) required minimum examination coverage of greater than 90X of the weld volume.
"Circumferential weld AB, in the beltline region, was examined to 79X 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 83X 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 100X scan coverage (entire weld length) from one side and 83X 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 Bl. 10, B1.21, B1.22, B1.30 and B1.40) examined isgreater than 84X 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 5 beltline region welds have received coverage of over 90X of the Code required volume.
The fifth weld, AB, received coverage of 79X. The major repair area within the beltline region received a
100X 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 86X 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.
- Figures provided by the licensee are not included in this report.
Results of RPV Preservice Inspection Examination (PSI)
"A manual preservice UT examination of essentially 100X 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 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 84X of the RPV shell welds (Bl. 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 90X of the shell welds volume received a full Code e'xamination.
"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 7X.
Adding ID tooling may increase total coverage of shell welds by less than 1X 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, 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).
Conclusions l.
"Irradiation embrittlement is minimal for the relatively low fluence environment of a BWR and the low EFPY that WNP-2 has experienced.
2.
3.
"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.
J "Greater than 91X of the total weld volume of Examination Category B-A, Item Number Bl.10 is receiving a full ASHE Section XI Code volume examination.
4.
"Greater than 86X of the total weld volume in the beltline region, including 100X of the repair area, is receiving a full ASHE Section XI Code volume examination.
5.
"Greater than 84X of the entire RPV weld volume is receiving a full ASHE Section XI Code volume examination.
6.
"A large percentage of RPV shell weld (Bl. 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 100X 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 not 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."
Licensee's Pro osed Alternative Examination (as stated):
"The alternative examinations proposed in place of the augmented examinations defined in 10 CFR 50.55a are to perform examinations of greater than 90X of the total weld volume in Examination Category B-A, Item Number B1. 10 and greater than BOX of the total weld volume in Examination Category B-A (Item Numbers Bl. 10, 81.21, B1.22, B1.30, and B1.40) during the first inspection interval, rather than 90X 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)."
Evaluation:
To comply with the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100X of each of the Item Bl. 10 shell welds.
In accordance with the regulations, essentially 100X is defined as greater than 90X of the examination volume of each weld.
As an alternative to the regulations, the licensee proposes that the examination of essentially 100X (>90X) of the cumulative weld volume be found acceptable in lieu of examining at least 90X of each weld.
At WNP-2, the augmented coverage requirements cannot be met for two circumferential shell welds due to physical restrictions that limit scan coverage.
For Meld AB, weld configuration resul'ted in loss of transducer contact and limited coverage to 79X of the required volume.
For Weld AD, RPV stabilizer lugs obstructed the examination areas and limited coverage to 83X of the required volume.
To achieve complete coverage for the subject welds, design modifications would be required to increase access from the outside surface (OD).
As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize examination coverage of their reactor vessels.
In cases where examination coverage from the OD is inadequate, alternative examinations from the inside surface (ID) using automated inspection tooling is a potential option.
For the WNP-2 reactor vessel, the licensee considered this option and had General Electric perform an access study.
Based on this access study, it was determined that little or no increase in coverage could be achieved by performing an ID examination of these welds.
Using OD examination techniques, the licensee has examined a substantial portion
(~79X) of each weld and has examined in excess of 91X of the cumulative length of the RPV shell welds.
This level of examination coverage is significant and should have detected inservice degradation, if present.
Therefore, the licensee's proposed alternative provides an acceptable level of'uality and safety.
3.0 CONCLUSION
The INEEL staff has reviewed the licensee's submittal and concludes that the licensee has examined a significant portion of the reactor vessel welds and that service-induced degradation, if present, would have been detected.
- Thus, the licensee's proposed alternative provides an acceptable level of quality and safety.
It is further concluded that additional examinations from the ID would not increase coverage substantially.
Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a,(g)(6)(ii)(A).
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