LR-N990250, Provides Proposed Alternative & Supporting Justification for Relief from Augmented Inservice Requirements of 10CFR50.55a(g) for Volumetric Exam of RPV Circumferential Welds
| ML20209G374 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/09/1999 |
| From: | Dawn Powell Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-98-05, GL-98-5, LR-N990250, TAC-MA4383, NUDOCS 9907190144 | |
| Download: ML20209G374 (12) | |
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D PSEG Public Service Electric and Gas Cornpany P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit JUL 9W LR-N990250 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 RELIEF REQUEST REGARDING AUGMENTED REACTOR VESSEL INSPECTION HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 Gentlemen:
Pursuant to 10CFR50.55a(a)(3) and 10CFR50.55a(g)(6)(ii)(A)(2), Public Service Electric and Gas Company (PSE&G) requests relief from the augmented inservice requirements of 10CFR50.55a(g) for the volumetric examination of reactor pressure vessel (RPV) circumferential welds (ASME Code Section XI, Table IWB-2500-1, Evaluation Category B-A, item 1.11, Circumferential Welds). NRC Generic Letter 98-05, " Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief From Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds", describes the conditions to be met when proposing relief from the augmented requirements.
The attachment to this letter includes the proposed alternative and supporting justification for the relief as well as the basis for concluding that Hope Creek meets the conditions described in Generic Letter 98-05. Based on the evaluation contained in the attachment, 1 PSE&G has concluded that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the proposed alternative satisfies the requirements of 10CFR50.55a(a)(3)(i).
Examination of the circumferential welds is required to be performed during the upcoming ninth refueling outage at Hope Creek (scheduled to begin in April 2000).
Since Hope Creek would realize immediate benefits from the proposed relief during the upcoming refueling outage, PSE&G is requesting that the NRC approve these changes by March 15,2000.
Changes similar to the changes proposed in this letter were approved by the NRC for Nine Mile Point Nuclear Station, Unit 1, in a safety evaluation report, dated April 7,1999 I (TAC NO. MA4383).
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Should you have any questions regarding this response, please contact Mr. C. Manges at 609-339-3234. 1 1
Sincerely,
- . .A D. R. Powell Director - Licensing / Regulation & Fuels l
Attachment:
Relief Request No. RR-B4 C Mr. H. Miller, Administrator - Region l U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission j l
One White Flint North Mail Stop 8B1 l 11555 Rockville Pike Rockville, MD 20852 i Mr. D. Orr (X24)
USNRC Senior Resident inspector- HC l
Mr. K. Tosen, Manager IV Bureau of Nuclear Engineering I P. O. Box 415 Trenton, NJ 08625 i E
Attachm:nt 1 H:pe Creek ISI Program - R:ll;f Reque:t N3. RR-B4 ,
HOPE CREEK GENERATING STATION ISI PROGRAM RELIEF REQUEST NO. RR-B4 I
- 1. COMPONENT DESCRIPTION Class 1, Category B-A, item No. B1.11 reactor pressure vessel pressure retaining circumferential shell welds. The subject welds are identified in the following table:
Reference No. Description 100010 W-4 Shell to Shell 100015 W-5 Shell to Shell i 100020 W-6 Shell to Shell I
i 100025 W-7 Shell to Shell 100030 W-8 Shell to Lower Head
- 11. ASME EXAMINATION REQUIREMENT A September 8,1992 revision to 10CFR50.55a(g)(6)(ii)(A) requires that a volumetric examination of reactor pressure vessel shell assembly welds be performed completely i once, as an augmented examination requirement. This new rule revoked previously )
granted licensee relief requests regarding the extent of volumetric examination on ASME i Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.10, ,
circumferential and longitudinal reactor pressure vessel shell welds. The augmented !
examinations are to be performed using the procedures specified in the ASME Section l XI Code Edition applicable to the inspection interval in which the augmented I examinations are performed.
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Atta:hm:nt 1 H:pe Cre:k ISI Progrcm - R:li:f R:qunt No. RR-84 Table IWB-2500-1 of the 1989 Edition of ASME Section XI (Code of record for the current ISI interval) requires a volumetric examination of essentially 100% of the weld length of circumferential and longitudinal shell welds during the first and each successive ISI interval. The ISI examinations are to be performed in accordance with ASME Section XI, Figures IWB-2500-1 and 2 (as applicable) and the nondestructive
. examination requirements of ASME Section V, Article 4, Paragraph T-441.3.2. The ASME requirements are supplemented by Regulatory Guide 1.150, " Ultrasonic Testing
' of Reactor Vessel Welds during Preservice and Inservice Examinations."
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. ASME Code Case N-460, "Altemate Examination Coverage for Class 1 and Class 2 Welds,Section XI, Division 1", defines " essentially 100%" by permitting an additive limitation of up to 10% of the weld length. Regulatory Guide 1.147, " Inservice inspection Code Case Acceptability ASME Section XI Division 1", approved Code Case N-460 for use by licensees. The provisions of 10CFR50.55a(g)(6)(ii)(A) also define " essentially 100%" as any amount greater than 90% of the examination volume of each weld.
One circumferential weld (W8) was examined during the Hope Creek fourth refueling outage. The balance were required to be examined during the seventh refueling outage (the last refueling outage of the first ISI interval); however, on October 31,1997, Hope Creek was granted temporary relief from examination of the remaining circumferential welds for two refueling outages (the seventh and eighth refueling outages). Examination of the remaining circumferential welds is therefore required to be performed during the upcoming ninth refueling outage at Hope Creek (scheduled to begin in April 2000).
Relief is requested from the inservice inspection requirements of 10CFR50.55a(g),
including both the 10CFR50.55a(g)(4) ISI examination and the 10CFR50.55a(g)(6) augmented examination requirements for ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, item No. B1.11, circumferential reactor pressure vessel shell welds. In lieu of the existing requirements, PSE&G proposes to implement the alternate requirements described in Section IV of this relief request. This relief is requested for the remaining term of operation under the existing license for the Hope Creek Generating Station i
111.- BASIS FOR RELIEF . l A. Introduction The technical basis for this request for inspection relief is documented in the following reports: 1 i
l J
l
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Atta:hm:nt 1 H:pe Creek ISI Program - R: lief R: quest No. RR-B4
. "BWR Vessel and Internals Project BWR Reactor Pressure Vessel Shell Weld inspection Recommendations (BWRVIP-05)," dated September 1995
. NRC safety evaluation report (SER), " Evaluation by the Office of Nuclear Reactor Regulation Related to the Review of the Topical Report by the Boiling Water Reactor Vessel and Internals Project: BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations, (BWRVIP-05)", dated July 28,1998 B. Generic Letter 9845 Criteria On November 10,1998, NRC Generic Letter 98-05, " Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief From Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds" was issued.
The letter stated that BWR licensees may seek permanent relief from performing examinations of RPV circumferential shell welds for the duration of the original operating license. This determination was supported by the above referenced NRC Safety Evaluation Report. The generic letter requires that the following conditions be met in order to justify the relief:
e circumferential welds will continue to satisfy the limiting condition failure probability for circumferential welds in the NRC SER at the expiration of the plant license e the licensee has implemented operator training and established procedures that
- limit the frequency of cold over-pressure events to the amount specified in the NRC SER The generic letter also stated that licensees still need to perform the required inspections of " essentially 100%" of axial welds.
Each of the above conditions is discussed below. j Welds Will Satisfy the Limiting Condition Failure Probability in NRC SER The NRC staff conducted an independent safety assessment of the analysis contained in BWRVIP-05 that included a probabilistic fracture mechanics (PFM) analysis to l estimate RPV failure probabilities. < Three key assumptions in the PFM analysis are: I e neutron fluence was that estimated to be end-of-license mean fluence; e chemistry values are mean values based on vessel types; and e potential for beyond design basis events is considered.
Although BWRVIP-05 provides the technical basis supporting the relief request, the following information is provided to show the conservatisms of the NRC analysis for the Hope Creek RPV.
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Attachmsnt 1 H::pe Creek ISI Progr:m - Rslisf RIqu;ct No. RR-84 Since Hope Creek's vessel was the only one fabricated by Hitachi and the NRC SER states in Section 2.1 that the Hitachi fabricated vessel is comparable to a Chicago
. Bridge & Iron (CB&l) vessel, Hope Creek will use the CB&l information as the reference case model. For Hope Creek, the circumferential weld joint between Shells 4 and 5 of the RPV would be the limiting circumferential weld within the vessel (i.e. relative to RTNOT). For plants fabricated by or similar to CB&l, the mean end-of-license (EOL) neutron fluence used in the NRC PFM analysis was 0.19E+19 n/cm and the mean RTNDT Was 44.5 'F. However, at Hope Creek, the highest fluence anticipated at EOL is 0.075E+19 n/cm2 at the vessel inside surface and the mean RTNor is 1.4 *F. Thus, embrittlement due to fluence effects is expected to be much lower for Hope Creek, and the NRC analysis as described in the NRC staff independent assessment is conservative for Hope Creek in this regard. Therefore, there is significant conservatism in the already low circumferential weld failure probabilities as related to Hope Creek.
Other Hope Creek RPV shell weld information is provided in the attached Table 1.
Operator Training implemented and Procedures Established to Limit the Frequency of Cold Over-Pressure Events Operating procedural restrictions, operator training, and work control processes for the Hope Creek Generating Station provide appropriate controls to minimize the potential for RPV cold over-pressurization events. The basis for this conclusion is provided below.
Pressure testing of the RPV at Hope Creek is classified as an " Infrequently Performed Test or Evolution." This classification ensures that special management oversight and procedural controls exist during this testing to maintain the plant's level of safety within
. acceptable limits. Hope Creek practice is to heat up the reactor to hydrostatic test temperature prior to increasing pressure. During performance of an RPV pressure test, level and pressure are controlled with the CRD and RWCU systems using a " feed and bleed" procedure. The Standby Liquid Control (SLC) system is normally isolated from the RPV during pressure testing. Increase in pressure is limited to 50 psi per minute.
.This practice minimizes the likelihood of exceeding the pressure-temperature limits during performance of the test.
The High Pressure Coolant injection (HPCI) and Reactor Core Isolation Cooling (RCIC)
- pumps at Hope Creek are steam driven and do not function when the plant is in ccU shutdown. Actuation of either of these systems would not lead to a cold over-pressuit event. The SLC system is another high-pressure water source to the RPV. The SLC system initiates automatically only if reactor power is not downscale after a low reactor water level.or high reactor pressure condition; so SLC system automatic initiation will not occur in cold shutdown. The injection rate of the two SLC pumps is approximately 46 gpm per pump' which would give the operator ample time to control reactor pressure in the case of an inadvertent manual injection.
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' ' Attachm:nt'1 H pe Creek ISI Progrcm - R lisf Rsqunt No. RR-B4 ,
During normal cold shutdown conditions, RPV level and pressure are controlled with the Control Rod Drive (CRD) and Reactor Water Cleanup (RWCU) systems using a " feed and bleed" process. The reactor is not taken solid during these times, and plant procedures
' require opening the head vent valves after the reactor has been cooled to less than l 212*F. If either of these systems were to fail, the Nuclear Control Operator (NCO) would adjust the other system to control level. Under these conditions, the CRD system typically injects water into the reactor at a rate less than 60 gpm. This slow injection rate allows
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the operator sufficient time to react to unanticipated level changes and, thus, significantly reduces the possibility of an event that would result in a violation of the pressure-temperature limits.
During normal cold shutdown conditions, reactor water level, pressure, and temperature are maintained within established bands in accordance with operating procedures. Operations procedures governing control room activities require that NCOs frequently monitor for indications and alarms to detect abnormalities as early as possible. These procedures also require that the Control Room Supervisor (CRS) be notified of any changes or abnormalities in indications. These procedures also require that changes that could affect reactor level, pressure, or temperature only be' performed with the knowledge and at the direction of the CRS. Therefore, any deviations in reactor water level or temperature from a specified band ;
will be promptly identified and corrected. Finally, the status of plant conditions, on-going '
activities which could affect critical plant parameters, and contingency planning are discussed by operators at each shift turnover. This ensures that on-coming operators are cognizant of activities that could adversely affect reactor level, pressure, or temperature.
The plant procedure for shutdown requires opening of the head vent valves after the .,
reactor has been cooled to less than 212*F. Procedural controls for reactor temperature,-ievel, and pressure are an integral part of operator training. !3pecifically, operators are trained in methods of controlling water level within specified limits, as well as responding to abnormal water level conditions outside the established limits.
Additionally, NCOs receive training on brittle fracture limits and compliance with the Technical Specification p'ressure-temperature limits curves. Plant-specific procedures have been developed to provide guidance to the operators regarding compliance with j the Technical Specification requirements on pressure-temperature limits. !
. The outage schedule and changes to the. schedule receive a thorough shutdown risk
-assessment review to ensure defense-in-depth is maintained. . Work is coordinated through the Work Control Center, which provides an additional level of Operations oversight. In the Main Control Room, the CRS is required to maintain cognizance of any activity that could potentially affect reactor level or decay heat removal during refueling outages. Pre-job briefings are conducted for work activities that have the potential to affect critical reactor parameters. These briefings are attended by the cognizant individuals involved in the work activity. Expected plant responses and contingency actions to address unexpected conditions or responses that may be encountered are included in the briefing discussion.
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' Atta:hm:nt 1 .
H:pe Cre:k ISI Progr:m - R:ll:f R: quest No. RR-B4 Inspections of Axial Welds Will Continue Axial welds will continue to be examined to the extent possible Essentially 100%
coverage could not be achieved for three axial welds during the first interval due to limitations. Although improved technology may allow additional coverage of these three welds during successive intervals, the minimum expected coverage for each axial weld is described in Table 2. Further details conceming these limitations are included as part
. of the first ten-year ISI interval requests for relief submitted on March 3,1998 and the subsequent response to the request for additional information submitted on December
.15,1998.
D. Conclusion Based on the documentation in BWRVIP-05, the risk-informed independent assessment performed by the NRC staff, and the discussion above, use of the attematives proposed in Section IV below provides an acceptable level of quality and safety. Accordingly, the proposed alternative satisfies the requirements of 10CFR50.55a(a)(3)(i).
IV. ALTERNATE PROVISIONS Pursuant to 10CFR50.55a(a)(i),10CFR50.55a(a)(ii), and 10CFR50.55a(g)(6)(ii)(A)(5),
PSE&G considers the following alternate provisions to be practical for the subject weld '
examinations.
The failure frequency for ASME Code Section XI, Table IWB-2500-1, Examination
. Category B-A ltem No. B1.11 circumferential reactor pressure vessel shell we!ds is sufficiently low to Justify elimination of the ISI and augmented examination requirements of 10CFR50.55a(g).
The ISI and augmented examination requirements of 10CFR50.55a(g) for ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, item No. B1.12 longitudinal (axial) reactor pressure vessel shell welds shall be performed for all axial welds, and shall include inspection of the circumferential welds only at the intersections of these welds with the axial welds, or approximately 2-3 percent of these welds.
The procedures for these examinations shall be qualified such that flaws reievant to reactor pressure vessel integrity can be reliably detected and sized, and the personnel implementing these procedures shall be qualified in the use of the procedures.
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Atta hment1 H:pe Creek ISI Program - R li;f Riqu:ct No. RR-B4 Successive Examications of Flaws For ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No.
B1.11 circumferential reactor pressure vessel shell welds, successive examinations of flaws per IWB-2420, are not required for non-threatening. flaws (e.g., such as embedded flaws from material manufacturing or vessel fabrication which experience negligible or no growth during the design life of the vessel), provided that the following conditions are met:
- 1. The flaw is characterized as subsurface in accordance with BWR Vessel and Internals Project Report, BWRVIP-05, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations.
- 2. The NDE technique and evaluation that detected and characterized the flaw as originating from material manufacture or vessel fabrication is documented in a flaw evaluation report,
- 3. The vessel containing the flaw is acceptable for continued service in accordance with IWB-3600 and the flaw is demonstrated acceptable for the intended service life of the vessel.
For ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, item No.
B1.12 longitudinal (axial) reactor pressure vessel shell welds, successive examinations of flaws shall be in accordance with IWB-2420. All flaws in longitudinal shell welds shall be reinspected at successive intervals consistent with the ASME Code and regulatory requirements.
Additional Examinations of Flaws For ASME Code Section XI, Table IWB-2500-1, Examination Catpgory B-A, item No.
B1.11 circumferential reactor pressure vessel shell welds, additional examinations per IWB-2430, are not required for flaws provided that the following conditions are met:
- 1. If the detected flaw is characterized as subsurface, then no additional examinations are required.
'2. ~ lf the flaw is not characterized as subsurface, then an engineering evaluation shall be performed, addressing the following (at a minimum):
A determination of the root cause of the flaw An evaluation of any potential failure mechanisms An evaluation of service conditions which could cause subsequent failure An evaluation per IW8-3600 demonstrating that the vessel is acceptable for cor,tinued service 1
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Attachm:nt 1 Hope Creek ISI Progrr.m - Rrligf R:qu!";t No. RR-B4
- 3. If the flaw meets the criteria of IWB-3600 for the intended service life of the vessel, then additional examinations may be limited to those welds subject to the same root cause conditions and failure mechanisms, up to the number of examinations required by IWB-2430(a). If the engineering evaluation concludes that there are no additional welds subject to the same root cause conditions, or if no failure mechanism exists, then no additional examinations are required.
For ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, item No.
B1.12 longitudinal (axial) reactor pressure vessel shell welds, additional examinations for flaws shall be in accordance with IWB-2430. All flaws in longitudinal shell welds shall require additional examinations consistent with the ASME Code and regulatory requirements.
Examination of the circumferential shell welds shall be performed if longitudinal (axial) weld examinations reveal an active, mechanistic mode of degradation exists.
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Atta:hm:nt 1 -
H:pe Creek ISI Progr:m - R:ll f R:qu: t N2. RR-B4 Table 1 Comparative Parameters at 32EFPY Parameter Comparative Parameters NRC Limiting Plant at 32EFPY for the Specific Analyses Bounding Circumferential Parameters at 32 EFPY Weld SER Table 2.6-4 Peak EOL ID neutron 0.075E+19 0.51 E+19 2
fluence (n/cm )
{
\
I Initial (unirradiated) -30 -65 reference temperature ( F)
Weld chemistry factor 105 109.5 Weld copper content (%) 0.08 0.10 Weld nickel content (%) 0.59 0.99 increase in reference 31.4 109.5 temperature due to irradiation (ARTuor)( F)
)
Mean RTwor(oF) 1.4 44.5 I
l l
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Attachm:nt 1 H:pe Creek ISI Progr:m - R:ll:f Requ:et N2. RR-B4 Table 2 Weld Coverage for Axial Welds Weld ID Code First 10 Year Interval Successive Interval Minimum item Composite Coverage Pienned Examination Coverage 11-1 B 1.12 94.41 % 94.41 %
11-2 B 1.12 92.2 % 92.2 %
a.
11-3 B 1.12 95.8 % 95.8%
12-1 B 1.12 92.1 % 92.1 %
12-2 8 1.12 79.0 % 79.0 %
12-3 B 1.12 78.7 % 78.7 %
13-1 B 1.12 92.40 % 92.40 %
13-2 B 1.12 89.04 % 89.04 % l l
13-3 8 1.12 95.98 % 95.98%
14-1 B 1.12 95.98% 95.98 %
l 14-2 B 1.12 92.37 % 92.37 %
14-3 B 1.12 95.81 % 95.81 % ;
15-1 B 1.12 100 % 100 %
15-2 B 1.12 100 % 100%
15-3 B 1.12 100 % 100 %
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