ML20202H395
| ML20202H395 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 02/11/1998 |
| From: | Hughey W ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GNRO-98-00015, GNRO-98-15, IEIN-97-063, IEIN-97-63, NUDOCS 9802200377 | |
| Download: ML20202H395 (7) | |
Text
O Entstgy operstions, Inc.
hh PO Bor 7%
Ptri Gbson, MS 39150 Tel 001437 G470 N["",w#
February 11,1998 esu wn i Nuclear Regulatory Commission Mail Station P137 Washington, D.C 20555 Attention:
Document Control Desk
Subject:
Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 Request for Alternative to 50.55a Examination Requirements of Category B1.11 Reactor Vessel Welds, Relief Request 1-2 00001 GNRO-98/00015 l
Gentlemen:
l l
In 1995 the Boiling Water Reactor Vessel & Internals Project (BWRVIP) transmitted BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations" to the Nuclear Regulatory Commission (NRC). The NRC's evaluation prompted numerous industry /NRC meetings and the issuance of NRC Information Notice 97-63.
NRC issued IN 97-63 to inform addressees of the Status of NRC staffs review of "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP 05)." Because of additional analysis requested by the NRC, the staff indicated in the information notice that consideration will be given to technically-Justified requests for relief from the augmented examination in accordance with 10 CFR 50.55a(a)(3)(i),10 CFR 50.55a(a)(3)(ii), and 10 CFR 50.55a(g)(6)(ii)A(5) for BWR licensees who are scheduled to perfcrm inspections of the BWR RPV cicumferential shell weids during the Fall 1997 or Spring 1998 outage seasons. Further, acceptably-justified reliefs would be considered for inspection # lays of up to two operating cycles for BWR RPV circumferential shell welds only.
Grand Gulf Nuclear Station (GGNS) hcreby requests schedular relief from the afformentioned requirements. The attached relief request has been prepared in accordance with the guidance of IN 97-63 using the BWRVIP suggested format. This alternative is being requested under the provisions of 10 CFR 50.55a(a)(3)(i) and will 4/-
provide an acceptable level of quality and safety. The Inservice Inspection Program at /
\\
GGNS complies with the 1992 edition with portions of the 1993 addenda of the ASMF.
Boiler and Pressure Vessel Code,Section XI.
9802200377 480211 H,, H, M,,,H,! 1,1 I, ADOCK 0500 6
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3 GN'RO-98/00015 Page 2 of 2 GGNS requests your review and approval by April 3,1998 to support GGNS's refueling outage scheduled to commence Apri 11,1998. Thank you in advance for your prompt attention its this matter. Should you have any questions or need any additional information, please contact Bill Brice at 601-437-6556.
Yours truly,
. h WKH/WBB attachment:
Relief Request 1-2-00001 cc:
Ms. J. L. Dixon-Herrity, GGNS Senior Resident (w/a)
Mr. L. J. Smith (Wise Carter) (w/a)
Mr. N. S. Reynolds (w/a)
Mr. H. L. Thomas (w/o)
Mr, E. W. Merschoff (w/a)
Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 Mr. J. N. Donohew, Project Manager (w/2)
Office of Nuclear Reactor Regulation
= U.S. Nuclear Regulatory Commission Mail Stop 13H3 Washington, D.C. 20555
ENTERGY OPERATIONS, INC.
GRAND GULF NUCLEAR STATION 2 nd TEN YEAR INTERVAL REQUEST NO. 1-2-00001 1.
COMPONENT / EXAMINATION IDENTIFICATION:
Code Class:
1
References:
ASME Section XI,1992 Edition !"'B 2500; 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(ii)(A)(2);
BWRVIP-05,"BWR Reactor Pressure Vessel Shell Weld Inspection";
NRC Information Notice 97-63,' Status of NRC StatTs Review of BWRVIP-05";
BWRVIP Response to NRC RAI on BWRVIP-05, 12/22/97 Examination Category:
BA Item No.:
B l.11 Examination Required:
Volumetric Examination of Welds and Adjacent Base Materials
==
Description:==
Circumferential Shell Welds in Reactor Vessel Component Number:
QlB13D001 II.
REQUIREMENTS:
ASME Section XI,1992 Edition, IWB-2500 requires the subject welds and associated base material to be volumetrically examined once each interval. The examinations are to be dispersed over the three periods of the interval within the limits specified by IWB-2412-1. Deferral of the examinations until the end of the interval is permissible; however, the examinations during the initial interval were not deferred, and IWB-2420 requires the sequence of examinations established in the first interval to be repeated during subsequent intervals to the extent practical.
In 1992. Title 10 of the Code of Federal Regulations (10 CFR) were amended with the addition of 50.55a(g)(6)(ii)(A)," Augmented Examination of Reactor Vessel." Section 50.55a(g)(6)(ii)(A)(2) requires licensees to augment their reactor vessel examinations 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 Bl.10 of Examination Category B-A," Pressure Retaining Welds in Reactor Vessel' in Table IWB-2500-1 of subsection lWB of the 1989 Edition of ASME Section XI, subject to the conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4). The augmented examination when s
not deferred in accordance with the provisions of 50.55a(g)(6)(ii)(A)(3), shall be perfomied in accordance with the related procedures specified in the Section XI Edition Page1of5
and Addenda applicable to the inservice inspection interval in effect on September 8, 1992. For the purpose of this augriented examination," essentially 100%", as used in Table IWB 2500-1, means more than 90% of the examination volume of each weld, whcre the reduction in coverage is due to interference by another component or part geometry.
Section 50.55a(g)(6)(ii)(A)(3) permits licensees with fewer than 40 months remaining in the inservice inspution interval in efTect en September 8,1992, to defer the augmented reactor vessel examination specified in 50.55a(g)(6)(ii)(A)(2) to the first period of the next inspection interval under certain conditions. Ilowever, if the augmented examinations are deferred to the first period of the next inspection interval, 50.55a(g)(6)(ii)(A)(3)(vi) requires the deferred examinations to be performed in accordance with the related procedures specified in the Section XI edition and addenda applicable to the inspection interval in which the augmented examination is perfonned.
Section 50.55a(g)(6)(ii)(A)(4) indicates that the requirement for augmented examination of the reactor vessel may be satisfied by an examination of essentially 100% of the reactor shell welds specified in 50.55a(g)(6)(ii)(A)(2) the have been completed, or are scheduled for implementation with a written commitment, or are required by 50.55a(g)(4)(i), during the inservice inspection interval in effect on September 8,1992.
111.
IIASIS FOR ALTERNATIVE Pursuant to the provisions of 10 CFR 50.55a(a)(3)(i), and consistent with information contained in NRC Infonnation Notice 97-63, an alternative is requested from the examination of RPV circumferential welds as required by ASME Section XI,IWB-2500, F.xamination Category B-A, item No. Bl.11, and 10 CFR 50.55a(g)(6)(ii)(A)(2) as described within. This proposed attemative is to postpone the examination of the RPV circumferential welds for two operating cycles, until Refueling Outage RFl1 that is presently scheduled to begin in approximately April,2001. The basis for this request for alternative is documented in the report "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)" that was transmitted to the NRC in September 1995 and BWRVIP Response to NRC RAI on BWRVIP-05 that was transmitted to the NRC on December 18,1997.
The BWRVIP-05 report provides the technical basis for eliminating examinations of BWR RPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. Additionally, the NRC as.sessment demonstrated that examination of BWR RPV circumferential shell welds does not measurably affect the probability of failure. Therefore the NRC evaluation appears to support the conclusions of BWRVIP-05.
The independent NRC assessment utilized the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are:
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(1) the. neutron fluence was that estimated to be end-of-license (EOL) mean fluence, (2) the chemistry values are mean values based on vessel types, and (3) the potential for beyond-design-basis events is considered.
Although BWRVIP-05 provides the technical basis supporting the request Ibr alternative, the following information is provided to show the conservatisms of the NRC analysis for the Grand Gulf Nuclear Station (GGNS). For plants with RPVs fabricated by Chicago Bridge & Iron (CB&l), the mean EOL neutron fluence used in the NRC PFM analysis 2
was 0.19E+19 rvem. Ilowev:r, the highest surface fluence for the GGNS RPV beltline region at the end of the requested alternative period is predicted to be 0.102E+19 n/ cm,
2 Thus the effect of 11uence on embrittlement is much lower, and the NRC analysis as described in the NRC independent assessment is conservative for GGNS in this regard.
1 herefore, there is significant conservatism in the already low circumferential-weld-failure probabilities as related to GGNS. Other GGNS RPV shell weld infonnation that the NRC staff has requested be included in this relief request is provided in the attached Table 1.
As shown in UFSAR Figure 5.3-9, GGNS does not have any circumferential welds in the beltline region. Ilowever, an evaluation showing the effects of radiation have been performed on the two circumferential welds that are closest to the core. The effects of irradiation depicted in this relief request are significantly exaggerated because:
the two welds are not located in the peak fluence region of the beltline, however peak beltline lluence values have been used in their evaluation (weld AB is approximately 5 inches below the core and weld AC is approximately 22 inches above the core),
the fluence used in this relief request represents surface fluence and not 1/4t e
fluence, and there is no credit taken for the attenuation caused by the RPV inner surface cladding.
The results of the evaluations are listed in Table 1. As shown in Table 1, the calculated embrittlement shift in RTsor (i.e., ARTsor) for the GGNS Unit i vessel is a maximum of 22.76* F at the end of the requested relief period. By comparison, using the mean values for fluence and weld chemistry assumed for CB&I reactor vessels in Table 7-5 of to the NRC independent assessment report, a ARTsor of 30.16* F would be derived. Therefore, the ca:culated ARTsor value for the GGNS vessel is less than, and thus bounded by, the embritticment shift assumed in the NRC's independent assessment.
Furthermore, it can be seen in the attached Table I that the calculateo Upper Bound RT ot value for the GGNS near-beltline welds is a maximum of 25.52
- F at the end of N
the requested relief period. For comparison, the Upper Bound RTsor value in Table 7-8 of Enclosure 1 to the NRC's independent assessment report of BWRVIP-05 is 32.7 F for fluence reference c.se 1. Again, the calculated Upper Bound RTsor values for the GGNS Page 3 of 5 f
l I
e
o.
vessel circumferential welds are clearly bounded by the limiting RTsor from Table 7-8 (CB&l. vessels) of the NRC independent assessment report, thus providing additional assurance that the GGNS vessel welds are also bounded by BWRVIP-05 report.
An added safety margin has been provided at GGNS by the nondestructive examination (NDE) of the vessel welds. A complete Preservice inspection (PSI) was performed on all of the RPV shell welds, both longitudinal and encumferential, to the maximum extent practical before GGNS initially loaded fuel. The same welds have also completed inservice Inspection (ISI) ultrasonic examinations required during the first 10-year inte val. The examination coverage for both PSI and ISI for all welds except for circumferential weld AA exceeded 90% coverage of the full volume. Weld AA has been examined over its complete length, but due to scanning limitations from the lower head i
side of the weld, it was only examined for approximately 67% of the Code required volume.
At the August 8,1997 meeting and in the NRC's independent assessment, the NRC staff indicated that the potential for, and consequences of, nondesign-basis events not addressed in the BWRVIP-05 report should be considered. In particular, the NRC staff stated that nondesign-basis cold over-pressure transients should be considered. It is highly unlikely that a BWR would experience a cold over-pressure transient. In fact, for a llWR to experience such an event would generally require several operator errors. At l
the August 8,1997 meeting, the NRC staff described several types of events that could be precursors to BWP. RPV cold over-pressure transients. These were identified as precursors because no cold over-pressure event has occurred at a U.S, BWR. Also at the 5
August 8 meeting, the NRC stalTidentified one actual cold over-pressure event that occurred during shutdown at a non-U. S. BWR. This event apparently included several operator errors that resulted in a maximum RPV pressure of i150 psi with a temperature c
range of 79"F to 88* F. As a result of the NRC's concerns, the BWRVIP has included in Attachment I to their response to the NRC's RAI on BWRVIP-05 significant discussion regarding BWR cold pressurization events. GGNS has reviewed the BWRVIP's response and concurs that the conditions and events are accurately depicted and that the procedures and personnel training at GGNS are comparable to those described by the BWRVIP and are adequate to prevent a cold over-pressure transient event. Consequently, the probability of a cold over-pressure transient is considered to be less than or equal to that used in the NRC analysis described in the NRC independent assessment and is conservative for GGNS.
IV.
CONCLUSION Based on BWRVIP-05, the risk-informed independent assessment performed by the NRC stalT, the BWRVIP's response to the NRC's RAI, and the discussion contained within, an alternative to the cited requirements under the provisions of 10 CFR 50.55a(a)(3)(i) to delay the described examinations until Refueling Outage RFI 1 is reasonable and will provide an acceptable level of quality and safety.
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C, Table 1 GGNS RPV Shell Weld Information Variable Variable Value by Weld Seam Identification Weld Seam'/
AH (Lower Cire.
AC (Upper Civ c.
AC (Upper Cire.
Welding Process Scam)
Scam)
Seam)
SAW SAW SMAW 2
Fluence @ 13.1 EFPY 0102 x 10 n/cm2 0.102 x 10 n/cm2 0.102 x 10 n/cm2 (end of relief request period)
Initial RTsm
-40.0 F
- 20.0 F
- 60.0 F Weld Chemistry Factor 41 54 27 Weld Copper Content 0.03 wt%
0.04 wt%
0.02 wt%
0.95 wt%
0.91 wt%
increase in Reference Temperature due to 17.28 F 22.76 F 11.38 F Irradiation (ARTS.or)
Margin Term 17.28 F 22.76
- F 11.38 F Mean Adju.cted Reference Temperature.
- 22.72
- F 2.76
- F
- 48.62 F (Mean ART)
Upper Bound Adjusted Reference Temperature
- 5.44
- F 25.52 F
-37.24 F (Upper Bound ART)
NOTES:-
- 1) GGNS RPV beltline does not contain circumferential welds. Figure 5.3-9 of the GGNS UFSAR shows that weld seam AB is approximately 5 inches below the core and weld seam AC is approximately 22 inches above the core.
- 2) The value is the peak fluence in the beltline region that was linearly interpolated to 13.1 EFPY The use of peak fluence from the beltline region to compute the shift due to irradiation of the circumferential welds (that are not actually in the beltline region) provides a conservative upper bound ARTsor.
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