ML20195J229
| ML20195J229 | |
| Person / Time | |
|---|---|
| Issue date: | 06/14/1999 |
| From: | Anand R NRC (Affiliation Not Assigned) |
| To: | Newton R WESTINGHOUSE OPERATING PLANTS OWNERS GROUP, WISCONSIN ELECTRIC POWER CO. |
| References | |
| PROJECT-686 NUDOCS 9906180144 | |
| Download: ML20195J229 (9) | |
Text
p June.' 14, 1999 lr
- Mr. Roger A. Newton, Chairman L
LCM /LR Working Group Westinghouse Owners Group Wisconsin Electric Power Company -
231 West Michigan Milwaukee, WI 53201 i
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE l
. WESTINGHOUSE OWNERS GROUP GENERIC Ll"5NSE RENEWAL
' PROGRAM TOPICAL REPORT ENTITLED, " LICENSE RENEWAL EVALUATION: AGING MANAGEMENT FOR REACTOR VBSSEL INTERNALS,"
i WCAP-14577, JUNE 1997
Dear Mr. Newton:
l By letter dated September 2,1997, the Westinghouse Owners Group (WOG) submitted Topical Report WCAP-14577, " License Renewal: Aging Management for Reactor Vessel internals,"
l June 1997, requesting the U.S. Nuclear Regulatory Commission staff's review and issuance of a l'
safety evaluation report.
Based on tne review of the information submitted, the staff has identified, in the enclosure, areas where additional information is needed to complete the review.
Please provide a schedule for the submittal of your response within 30 days of the receipt of this letter. Additionally, the staffis willing to meet with the WOG before you submit your response to clarify the staffs request for additional information.
l-Sincerely, OrigirialSigriet!% -
l Raj K. Anand, Project Manager License Renewal and Standardization Branch Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation c
~ Project No. 686 k
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i WESTINGHOUSE OWNERS GROUP (WOG)
Project No. 686 cc:
Mr. Gregory D. Robison Ad Hoc Technical Group Coordinator LCM /LR Working Group Duke Power Company -
Westinghouse Owners Group P. O. Box 1006 Charlotte, NC 28201 Mr. Summer R. Bemis Westinghouse Owners Group Project Office Westinghouse Electric Corporation, ECE 5-16 P. O. Box 355 Pittsburgh, PA 15230-0355 Mr. Theodore A. Meyer Westinghouse Program Manager for WOG LCM /LR Program Westinghouse Electric Corporation, ECE 4-22 P. O. Box 355 htisburgh, PA 15230-0355 Mr. Charlie Meyer Westinghouse Lead Engineer for WOC LCWLR Prv, Westinghouse Electric Corporation, ECE 4-8 P. O. Box 355 Pittsburgh, PA 15230-0355 Douglas J. Walters Nuclear Energy Institute 776 i Street, NW Suite 400 Washington, DC 20006-3708 i
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REQUEST FOR ADDITIONAL INFORMATlON WCAP-14577 l
LICENSE RENEWAL EVALUATION:
AGING MANAGEMENT FOR REACTOR VESSEL INTERNALS (RVI) 1.
Industry Plans Since the submission of the topical report, the industry has consolidated efforts by the various owner's and other groups, e.g., the PWR Materials Reliability Project (MRP). What is the scope and nature of industry efforts addressing aging management issues related to RVi? What are the schedules for these activities, and how will the results of these industry efforts affect the conclusions and plans addressed in the topical report?
2.
Technical Prooress Since almost two years have elapsed from the date that the topical report was submitted, what changes would be made to the report considering technical progress during that time, with particular emphasis on the report sections addressing aging effects and aging management programs (AMP)?
3.
Baffle-Former Bolts in Sections 3.0 and 4.0 of the subject report, WOG, in part, addresses aging management review, aging effects evaluation, and proposed generic aging effects management activities and programs with regard to aging-related degradation of baffle bolts. Subsequent to the submittal of the subject report, WOG had periodic meetings and interactions with the staff from 1997 to the present regarding its ongoing programs and activities to resolve the baffle bolt cracking issues.
The ongoing programs and activities include: (1) development and approval of a prescribed analytical methodology for evaluating the acceptability of baffle bolting distributions under faulted conditions; (2) assessment of the safety significance of potentially degraded baffle bolting; (3) performance of baffle bolting inspections / replacements and testing on lead plants; and (4) development of an inspection monitoring and aging management program.
The staff requests that WOG describe their plam, and schedules for including the results of the above programs and activities in the aging management of baffle-former bolts.
4.
Baffle-Former Bolts in Section 4.2.2 of WCAP-14577, WOG describes the AMP for baffle-former bolts (AMP-4.6),
which recommends continued use of the present surveillance techniques. The present surveillance techniques include: (1) visual (VT-3) examination; (2) loose parts detection monitoring; and (3) reactor coolant chemistry monitoring. AMP-4.6 provides options for correcting relevant conditions detected by the VT-3 examination. Further, WOG indicates that baffle-former bolt cracking has not been observed in Westinghouse domestic plants. However, l
baffle-former bolt cracking has been observed in French and Belgian plants and more recently in Westinghouse domestic plants using volumetric (UT) examination techniques.
l
- Based on the recent experience of volumetric inspection of baffle-former bolts at Ginna and Point Beach Unit 2, the staff requests that WOG propose an alternative program. In lieu of the proposed VT-3 examination, the WOG should consider volumetric inspection.
5.
Fatioue - Time-Limited Aoina Analysis in Section 1.0 of WCAP-14577, WOG indicates that one objective of the report is to identify and evaluate time-limited aging analyses (TLAA). In Section 2.5 of the report, WOG identifies fatigue as the only TLAA related to the RVI, and that the results from current TLAAs have been projected to an extended period of operation. In Section 3.0, WOG provides a summary list (Table 3 3) of fatigue-sensitive RVI components that could reach the fatigue usage limit within a 40-to-60-year time period. WOG indicates that the listed components were identified based on a review of calculated fatigue usage factors for internals components designed to American
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Society of Mechanical Engineers (ASME) Section lit, Subsection NG, hot functional test data, and a comparison of geometric and operating similarities, j
The staff requests WOG to provide a list of the TLAAs and a brief summary description of each of the listed analyses used to identify the fatigue-sensitive reactor vessel components. The staff I
requests WOG to clarify whether the fatigue-sensitive components listed in Table 3-3 apply to all Westinghouse-designed RV1 or only to those designed to ASME B&PV Code, Section lil, Subsection NG as described in Section 2.5.1 of the report.
For RVI designed prior to the ASME Code adoption of Subsection NG in Section Ill, what requirements were used for fatigue analysis of the RVI components for the initial operating period? How were these analyses updated to account for the license renewal period?
6.
Manaaement of Crackina and Neutron Irradiation Embrittlement The topical report indicates that effects of cracking due to irradiation embrittlement and IASCC are managed by AMP-4.1 through visual examination, loose parts monitoring and supplemental examination. VT-3 visual examination as required by Examination Category B-N 2/B-N-3 of Subsection IWB of ASME Code Section XI is not adequate for detecting lASCC. The activities for managing lASCC and irradiation embrittlement should be revised to provide a more effective management program. One acceptable program for managing these aging effects is outlined in the draft SER for the Calvert Cliffs license renewal application (Ref.1).
As an altemative AMP for IASCC and neutron irradiation embrittlement, the draft SER for the 4
Calvert Cliffs license renewal application (Ref.1) indicates that the applicant has cc mmitted to a two-part approach for managing lASCC and neutron embrittlement of RVI components. The first part of this approach is the use of supplemental (enhanced VT-1) examination of RVI components as part of the 10-year ISI program. This supplemental (enhanced VT-1) examination would be performed on the RVI components believed to be the limiting components for cracking, considering both the susceptibility of the component to the aging mechanism, as well as the material properties (in particular the fracture toughness) and the operating stresses on the component. These examinations would apply to all RVI components except for bolting.
j The second part of this approach involves consideration of data and evaluations from industry research activities to determine the susceptibility of RVI components to IASCC and neutron embrittlement. Should these data or evaluations indicate that the supplemental (enhanced W1) examinations can be modified or possibly eliminated, the applicant would be required to provide
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plant-specificjustification to demonstrate tne basis for the modification or elimination.
The topical report should be revised to provide a more effective aging management program for IASCC and neutron irradiation embrittlement. An acceptable alternative is the program j
committed to by the applicant for license renewal of the Calvert Cliffs plant.
i 7.
Aoina Effects and Manaaement for Cast Austenitic Stainless Steel (CASS)
The RVI components fabricated from CASS are potentially subject to a synergistic loss of fracture toughness due to the combination of thermal and neutron irradiation embrittlement. This enhanced loss of fracture toughness is not accounted for within the topical report nor in guidance in revisions to EPRI TR-106092 (Ref. 2). Further, the topical report rules out consideration of thermal embrittlement of RVI CASS components based upon tne lack of molybdenum in the materials. The NRC staff does not find this position of considering only thermal embrittlement to be acceptable. A modified screening approach should be used that is similar to that proposed in EPRI TR-106092 (Ref. 2), but also reflecting the potential synergistic effects of neutron irradiation and thermal embrittlement. One acceptable program is outlined below, consistent with the draft SER for the Calvert Cliffs license renewal application (Ref.1).
The modified approach described in the draft SER for the Calvert Cliffs license renewal application (Ref.1) consists of either a supplemental (enhanced VT-1) examination of the affected components as part of the applicant's 10-year ISI program during the license renewal term, or a component-specific evaluation to determine the susceptibility to loss of fracture toughness. The proposed evaluation willlook first at the neutron fluence of the component. If the neutron fluence is greater than 1 x 10" n/cm (E > 1 MeV), a mechanical loading 2
assessment would be conducted for the component. This assessment will determine the maximum tensile loading on the component during ASME Code Level A, B, C, and D conditions.
If the loading is compressive or low enough to preclude fracture of the component, then the component would not require supplementalinspection. Failure to meet this criterion would require continued use of the supplemental (enhanced VT-1) inspection. If the neutron fluence is less than 1 x 10" n/cm (E > 1 MeV), an assessment would be made to determine if the affected 2
component (s) are bounded by the screening criteria in EPRI TR-106092 (Ref. 2), modified as described below, in order to demonstrate that the screening criteria in EPRI TR-106092 (Ref. 2) are applicable to RVI components, a flaw tolerance evaluation specific to the RVI would be performed. If the screening criteria are not satisfied, then a supplemental (enhanced VT-1) inspection will be performed on the component.
i The CASS components should be evaluated to the criteria in EPRI TR-106092 (Ref. 2) with the e
following additional criteria:
Statically cast components with a molybdenum content meeting the requirements of SA 351 Grades CF3 and CF8 and with a delta ferrite content less than 10 percent will not need supplemental examination.
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Ferrite levels will be calculated using Hull's equivalent factors or a method producing an equivalent level of accuracy (t6 percent deviation between measured and calculated values).
Cast austenitic stainless steel components containing niobium are' subject to supplemental examination.
Flaws in CASS with ferrite levels less than 25 percent and no niobium may be evaluated using ASME Code IWB-3640 procedures.
Flaws in CASS with ferrite levels exceeding 25 percent or nicWm will be evaluated using ASME Code IWB-3640 procedures. If this occurs, fracture toughness data will be provided on a case-by-case basis.
Components that have delta ferrite levels below the screening criteria have adequate fracture toughness and do not require supplemental inspection. Components that have delta ferrite levels exceeding the screening criteria may not have adequate fracture toughness, as a result of thermal embrittlement, and do require supplemental inspection.
The topical report should be revised to provide a more effective aging management program for cast austenitic stainless steel. An acceptable alternative is the program committed to by the applicant for license renewal of the Calvert Cliffs plant.
8.
Sionificance of Void Swellina The topical report dismisses change of dimension of the RVi components due to void swelling as a significant aging effect due to (1) core management reducing neutron exposure levels such that the effects of swelling are either not significant or are limited to a small number of baffle-barrel region bolts, and (2) no degradation in ability of the structures in this region from meeting their intended functions. The NRC staff finds this evaluation of void swelling to be inadequate EPRI TR-107521 (Ref. 3) cites one source which predicts swelling as great as 14 percent for PWR baffle-former assemblies over a 40-year plant lifetime. The issue of concern is the impact of change of dimension due to void swelling on the ability of the RVI to perform their intended function.
The WOG should address the following:
How much of a change in dimension would be required before the intemals would not be able to meet their intended function?
1 l
What programs are the WOG parti 6ipating in that will evaluate the impact of the void swelling on the intended function of the internals?
When will these programs provide data to determine whether void swelling could impact the intended function of the internals?
1 I
5 Should it be determined that change of dimension by void swelling can impede the ability of the RVI to perform its intended function, then an appropriate aging management program would be required to assure that the need for corrective actions can be prcperly identified.
9.
ASME Code Limitations on Stresses or Deformations Section 2.4.1.2 of the topical report describes ASME Code limitations on stresses or deformations required to ensure a safe and orderly reactor shutdown in the event of an earthquake and major loss of-coolant incident loading conditions. Describe the specific current i
licensing basis limitations, and demonstrate that the material properties of the RVI components will continue to meet these limits under the neutron irradiation embrittlement conditions which will exist at the end of the license renewal period.
10.
Intended Functions of the Reactor \\/esselInternals Section 2.2 of the topical report describes the intended functions of the reactor vesselinternals on system level. The staff believes that the rule [10 CFR 54.21(a)(3)] requires that a renewal applicant demonstrate that the intended functions are maintained at the basic structure or component level. The report should, therefore, include RVI component-level intended functions which may include, but not be limited, to the following intended functions-1 Provide support and orientation of the reactor core (i.e., the fuel assemblies).
Provide support, orientation, guidance, and protection of the control rod assemblies.
Provide a passageway for the distribution of the reactor cociant flow to the reactor core.
e Provide a passageway for support, guidance, and protection for incore instrumentation.
Provide a secondary core support for limiting the core support structure downward displacement, Provide gamma and neutron shielding for the reactor pressure vessel.
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6-l Referen':es 1.
'" Safety Evaluation Report Related *.o the License Renewal of Calvert Cliffs Nuclear l
Power Plant, Units 1 and 2," dated March 1999.
2.
EPRI Technical Report TR-106092, " Evaluation of Thermal Aging Embrittlement for Cast Austenitic Stainless Steel Components in LWR Reactor Coolant Systems," Electric Power Research Institute, September 1997.
3.
EPRI Technical Report TR-107521, " Generic License Renewal Technical issues Summary," Electric Power Research Institute, April 1998.
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