ML042110243

Safety Evaluation, Relief Request for Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds in Accordance with BWRVIP-05 and GL 98-05

ML042110243
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	09/30/2004
From:	Howe A NRC/NRR/DLPM/LPD1
To:	Kansler M Entergy Nuclear Operations
Ennis R, NRR/DLPM, 415-1420
References
GL-98-005, TAC MC0848
Download: ML042110243 (20)

Text

September 30, 2004 Mr. Michael Kansler President Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

SAFETY EVALUATION OF RELIEF REQUEST ISI-06 FOR VOLUMETRIC EXAMINATION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL SHELL WELDS IN ACCORDANCE WITH BOILING WATER REACTOR VESSEL AND INTERNALS PROJECT-05 AND GENERIC LETTER 98-05, VERMONT YANKEE NUCLEAR POWER STATION (TAC NO. MC0848)

Dear Mr. Kansler:

By letter dated September 25, 2003, as supplemented by letter dated May 17, 2004, Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Entergy or the licensee),

submitted Relief Request ISI-06 requesting relief from American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, requirements related to examination of reactor pressure vessel (RPV) circumferential shell welds at Vermont Yankee Nuclear Power Station (VYNPS). The relief request would authorize the use of a proposed alternative in accordance with Boiling Water Reactor Vessel and Internals Project-05 (BWRVIP-05) and Generic Letter (GL) 98-05 to the RPV circumferential shell welds examination requirements of ASME Code,Section XI, for the remaining portion of the current license period.

The licensee requested relief from the following requirements of ASME Code,Section XI, 1998 Edition through 2000 Addenda:

Subarticle IWB-2500, Table IWB 2500-1, Examination Category B-A, Pressure Retaining Welds in Reactor Vessel.

This relief is requested for the following components:

ASME Code Section XI, Examination Category B-A, Code Item No. B1.11, Circumferential Shell Welds.

The U.S. Nuclear Regulatory Commission (NRC) staff has completed its review of relief request ISI-06 as documented in the enclosed safety evaluation (SE). The NRC staff has reviewed the licensees submittal and finds that the licensee has acceptably demonstrated that the appropriate criteria in GL 98-05 and the staffs evaluation of the BWRVIP-05 report have been satisfied regarding permanent relief from inservice inspection (ISI) requirements of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, for the volumetric examination of RPV circumferential welds.

The NRC staff concludes that the licensees permanent relief request from the examination requirements of circumferential shell welds for the remaining portion of the current license

M. Kansler period of 32 effective full-power years, pursuant to Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(i), is acceptable and is consistent with the information contained in GL 98-05. The staff has also determined that the alternative program provides an acceptable level of quality and safety.

Additional requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third-party reviews by the Authorized Nuclear Inservice Inspector.

If you have any questions regarding this matter, please contact the VYNPS Project Manager, Mr. Richard B. Ennis, at (301) 415-1420.

Sincerely,

/RA/

Allen G. Howe, Vermont Yankee Section Chief Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosure:

Safety Evaluation cc w/encl: See next page

M. Kansler period of 32 effective full-power years, pursuant to Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(i), is acceptable and is consistent with the information contained in GL 98-05. The staff has also determined that the alternative program provides an acceptable level of quality and safety.

Additional requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third-party reviews by the Authorized Nuclear Inservice Inspector.

If you have any questions regarding this matter, please contact the VYNPS Project Manager, Mr. Richard B. Ennis, at (301) 415-1420.

Sincerely,

/RA/

Allen G. Howe, Vermont Yankee Section Chief Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosure:

Safety Evaluation cc w/encl: See next page DISTRIBUTION PUBLIC CRaynor OGC CAnderson, RGN-I PDI-2 Reading REnnis ACRS KKarcagi CHolden MMitchell GHill (2) JUhle AHowe BFu JJolicoeur VRodriguez NRay LLois VBucci, OIG Accession No.: ML042110243

• Input provided by memo dated July 6, 2004, incorporated with no significant changes

• • Input provided by memo dated November 26, 2003, incorporated with no significant changes OFFICE PDI-1/PM PDI-VY/PM PDI-2/LA EMCB/SC(A) SRXB/SC OGC VY/SC NAME DSkay REnnis CRaynor MMitchell* JUhle** DReddick AHowe DATE 9/23/04 9/22/04 8/30/04 07/06/04 11/26/03 9/29/04 9/30/04 OFFICIAL RECORD COPY

Vermont Yankee Nuclear Power Station cc:

Regional Administrator, Region I Ms. Carla A. White, RRPT, CHP U. S. Nuclear Regulatory Commission Radiological Health 475 Allendale Road Vermont Department of Health King of Prussia, PA 19406-1415 P.O. Box 70, Drawer #43 108 Cherry Street Mr. David R. Lewis Burlington, VT 05402-0070 Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W. Mr. James M. DeVincentis Washington, DC 20037-1128 Manager, Licensing Vermont Yankee Nuclear Power Station Ms. Christine S. Salembier, Commissioner P.O. Box 0500 Vermont Department of Public Service 185 Old Ferry Road 112 State Street Brattleboro, VT 05302-0500 Montpelier, VT 05620-2601 Resident Inspector Mr. Michael H. Dworkin, Chairman Vermont Yankee Nuclear Power Station Public Service Board U. S. Nuclear Regulatory Commission State of Vermont P.O. Box 176 112 State Street Vernon, VT 05354 Montpelier, VT 05620-2701 Director, Massachusetts Emergency Chairman, Board of Selectmen Management Agency Town of Vernon ATTN: James Muckerheide P.O. Box 116 400 Worcester Rd.

Vernon, VT 05354-0116 Framingham, MA 01702-5399 Operating Experience Coordinator Jonathan M. Block, Esq.

Vermont Yankee Nuclear Power Station Main Street 320 Governor Hunt Road P.O. Box 566 Vernon, VT 05354 Putney, VT 05346-0566 G. Dana Bisbee, Esq. Mr. John F. McCann Deputy Attorney General Director, Nuclear Safety Assurance 33 Capitol Street Entergy Nuclear Operations, Inc.

Concord, NH 03301-6937 440 Hamilton Avenue White Plains, NY 10601 Chief, Safety Unit Office of the Attorney General Mr. Gary J. Taylor One Ashburton Place, 19th Floor Chief Executive Officer Boston, MA 02108 Entergy Operations 1340 Echelon Parkway Ms. Deborah B. Katz Jackson, MS 39213 Box 83 Shelburne Falls, MA 01370

Vermont Yankee Nuclear Power Station cc:

Mr. John T. Herron Mr. Ronald Toole Sr. VP and Chief Operating Officer 1282 Valley of Lakes Entergy Nuclear Operations, Inc. Box R-10 440 Hamilton Avenue Hazelton, PA 18202 White Plains, NY 10601 Ms. Stacey M. Lousteau Mr. Danny L. Pace Treasury Department Vice President, Engineering Entergy Services, Inc.

Entergy Nuclear Operations, Inc. 639 Loyola Avenue 440 Hamilton Avenue New Orleans, LA 70113 White Plains, NY 10601 Mr. Raymond Shadis Mr. Brian OGrady New England Coalition Vice President, Operations Support Post Office Box 98 Entergy Nuclear Operations, Inc. Edgecomb, ME 04556 440 Hamilton Avenue White Plains, NY 10601 Mr. James P. Matteau Executive Director Mr. Michael J. Colomb Windham Regional Commission Director of Oversight 139 Main Street, Suite 505 Entergy Nuclear Operations, Inc. Brattleboro, VT 05301 440 Hamilton Avenue White Plains, NY 10601 Mr. William K. Sherman Vermont Department of Public Service Mr. John M. Fulton 112 State Street Assistant General Counsel Drawer 20 Entergy Nuclear Operations, Inc. Montpelier, VT 05620-2601 440 Hamilton Avenue White Plains, NY 10601 Mr. Jay K. Thayer Site Vice President Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500 Mr. Kenneth L. Graesser 38832 N. Ashley Drive Lake Villa, IL 60046 Mr. James Sniezek 5486 Nithsdale Drive Salisbury, MD 21801

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST AND ALTERNATIVES FOR VOLUMETRIC EXAMINATION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL SHELL WELDS IN ACCORDANCE WITH BOILING WATER REACTOR VESSEL AND INTERNALS PROJECT-05 AND GENERIC LETTER 98-05 ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

1.0 INTRODUCTION

By letter dated September 25, 2003, as supplemented by letter dated May 17, 2004, Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Entergy or the licensee),

submitted Relief Request ISI-06 requesting relief from American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, requirements related to examination of reactor pressure vessel (RPV) circumferential shell welds at Vermont Yankee Nuclear Power Station (VYNPS). The relief request would authorize the use of a proposed alternative in accordance with Boiling Water Reactor Vessel and Internals Project-05 Report (BWRVIP-05) and Generic Letter (GL) 98-05 to the RPV circumferential shell welds examination requirements of ASME Code,Section XI, for the remaining portion of the current license period.

2.0 REGULATORY EVALUATION

Inservice inspection (ISI) of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable Addenda, as required by Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g), except where specific relief has been granted by the Nuclear Regulatory Commission (NRC or the Commission), pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) of 10 CFR states, in part, that proposed alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that the proposed alternatives would provide an acceptable level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), components (including supports) which are classified as

ASME Code Class 1, 2, and 3 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. Section 50.55a(g)(4)(i) of 10 CFR requires that inservice examination of components and system pressure tests conducted during the first 120-month 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) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.

The applicable ISI code of record for VYNPS is the 1998 Edition with Addenda through 2000 of Section XI of the ASME Code. VYNPS proposes to use alternatives in accordance with BWRVIP-05 and GL 98-05 for the duration of their fourth ten-year interval (September 1, 2003 through August 31, 2013).

2.1 Regulatory Background 2.1.1 BWRVIP-05 Report By letter dated September 28, 1995, as supplemented by letters dated June 24 and October 29, 1996, May 16, June 4, June 13, and December 18, 1997, and January 13, 1998, the BWRVIP, a technical committee of the Boiling Water Reactor Owners Group (BWROG),

submitted the proprietary report, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05). The BWRVIP-05 report evaluates the current inspection requirements for RPV shell welds in BWRs, formulates recommendations for alternative inspection requirements, and provides a technical basis for these recommended requirements. As modified, the BWRVIP-05 report proposed to reduce the scope of inspection of BWR RPV welds from essentially 100 percent of all RPV shell welds to examination of 100 percent of the axial (i.e., longitudinal) welds and essentially zero percent of the circumferential RPV shell welds, except for the intersections of the axial and circumferential welds. In addition, the report includes proposals to provide alternatives to ASME Code requirements for successive and additional examinations of circumferential welds, provided in paragraph IWB-2420 and IWB-2430, respectively, of Section XI of the ASME Code.

On July 28, 1998, the NRC staff issued a safety evaluation report (SER) on the BWRVIP-05 report1. This evaluation concluded that the failure frequency of RPV circumferential welds in BWRs was sufficiently low to justify elimination of ISI of these welds. In addition, the evaluation concluded that the BWRVIP proposals on successive and additional examinations of circumferential welds were acceptable. The evaluation indicated that examination of the circumferential welds shall be performed if axial weld examinations reveal an active degradation mechanism.

1 The NRC staff has identified that, in some instances, the NRC SER is referenced as dated on July 28, 1998, others on July 30, 1998. For clarification purposes, the staff wants to add that this SER is a letter addressed to Carl Terry, BWRVIP Chairman, dated July 28, 1998 and titled Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925).

In the BWRVIP-05 report, the BWRVIP committee concluded that the conditional probabilities of failure for BWR RPV circumferential welds are orders of magnitude lower than that of the axial welds. As a part of its review of the report, the NRC conducted an independent probabilistic fracture mechanics assessment of the results presented in the BWRVIP-05 report.

The staffs assessment conservatively calculated the conditional probability of failure from RPV axial and circumferential welds during the initial (current) 40-year license period and at conditions approximating an 80-year vessel lifetime for a BWR nuclear plant, as indicated respectively in Tables 2.6-4 and 2.6-5 of the staffs July 28, 1998, SER. The failure frequency for a RPV is calculated as the product of the frequency for the critical (limiting) transient event and the conditional probability of failure for the weld.

The NRC staff determined the conditional probability of failure for axial and circumferential welds in BWR vessels fabricated by Chicago Bridge and Iron (CB&I), Combustion Engineering, and Babcock and Wilcox. The analysis identified a cold overpressure event that occurred in a foreign reactor as the limiting event for BWR RPVs, with the pressure and temperature from this event used in the probabilistic fracture mechanics calculations. The staff estimated that the probability for the occurrence of the limiting overpressurization transient was 1 x 10-3 per reactor year. For each of the vessel fabricators, Table 2.6-4 of the staffs SER identifies the conditional failure probabilities for the plant-specific conditions with the highest projected reference temperature (for that fabricator) after the initial 40-year license period.

2.1.2 GL 98-05 On November 10, 1998, the NRC issued GL 98-05 which states that BWR licensees may request permanent (i.e., for the remaining term of operation under the existing, initial license) relief from the ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential RPV welds (ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, Circumferential Shell Welds) by demonstrating that:

(1) at the expiration of the license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the NRC staffs July 28, 1998, SER, and (2) licensees have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the NRC staff's July 28, 1998, SER.

Licensees will still need to perform the required inspections of "essentially 100 percent" of all axial welds.

2.1.3 Successive Examination of Flaws As specified in the staffs July 28, 1998, SER, for ASME Code Section Xl, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, RPV Circumferential Shell Welds (i.e., at intersections with axial welds), successive examinations per Subarticle IWB-2420 are not required for non-threatening flaws (e.g., 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 BWRVIP-05; (2) the non-destructive examination technique and evaluation that detected and characterized the flaw as originating from material manufacture or vessel fabrication is documented in a flaw evaluation report; and (3) the vessel containing the flaw is acceptable for continued service in accordance with Subarticle IWB-3600, "Analytical Evaluation of Flaws," and the flaw is demonstrated acceptable for the intended service life of the vessel.

For ASME Code Section Xl, Table IWB-2500-1, Examination Category B-A, Item No. B1.12, RPV Longitudinal [Axial] Shell Welds, all flaws shall be reinspected at successive intervals consistent with ASME Code and regulatory requirements.

2.1.4 Additional Examinations of Flaws As specified in the staffs July 28, 1998, SER, for ASME Code Section Xl, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, RPV Shell Circumferential Welds (i.e., at intersections with axial welds), additional examinations per Subarticle IWB-2430, "Additional Examinations," are not required for flaws provided the following conditions are met:

(1) If the flaw is characterized as subsurface in accordance with BWRVIP-05, then no additional examinations are required.

(2) If the flaw is not characterized as subsurface in accordance with BWRVIP-05, 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 Subarticle IWB-3600 demonstrating that the vessel is acceptable for continued service.

If the flaw meets the criteria of Subarticle IWB-3600 for the intended service life of the vessel, then additional examinations may be limited to those welds subject to the root cause conditions and failure mechanisms, up to the number of examinations required by Subarticle 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 Xl, Table IWB-2500-1, Examination Category B-A, Item No. B1.12, RPV Longitudinal [Axial] Shell Welds, additional examination for flaws shall be in accordance with Subarticle IWB-2430. All flaws in RPV shell axial welds shall require additional weld examinations consistent with ASME Code and regulatory requirements. Examinations of the RPV shell circumferential welds shall be performed if RPV axial welds reveal an active degradation mechanism.

3.0 RELIEF REQUEST ISI-06 3.1 Code Requirement for which Relief is Requested The licensee requested relief from the following requirements of ASME Code,Section XI, 1998 Edition through 2000 Addenda:

Subarticle IWB-2500, Table IWB 2500-1, Examination Category B-A, Pressure Retaining Welds in Reactor Vessel.

This relief is requested for the following components:

ASME Code,Section XI, Class 1, Examination Category B-A, Code Item No.B1.11, Circumferential Shell Welds.

3.2 Licensees Proposed Alternative to the ASME Code In accordance with 10 CFR 50.55a(a)(3)(i), and consistent with information contained in GL 98-05, the licensee proposes to perform a volumetric examination of essentially 100 percent of the accessible RPV axial welds to the extent possible and the incidental portions of circumferential welds where they intersect the axial welds during the first period of the fourth ISI interval.

3.3 Licensees Basis for Alternative The BWRVIP-05 report provides the technical basis to justify relief from the examination requirements of RPV shell circumferential welds. The results of the NRC's evaluation of BWRVIP-05 are documented in the SER on BWRVIP-05. GL 98-05 states that BWR licensees may request permanent (i.e., for the remaining term of operation under the existing, initial license) relief from the ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential RPV welds (ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, Circumferential Shell Welds) by demonstrating that:

(1) at the expiration of the license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the NRC staffs July 28, 1998, SER; and (2) licensees have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the NRC staff's July 28, 1998, SER.

Licensees will still need to perform the required inspections of "essentially 100 percent" of all axial welds.

Entergy states that VYNPS complied with 10 CFR 50.55a(g)(6)(ii)(A)(2) required augmented examination of RPV circumferential and axial shell welds during refueling outage (RFO) 19 (1996) in the first period of the third ISI interval. VYNPS provided the results of these examinations to the NRC. The RFO 19 scope of examinations included essentially all of the accessible circumferential and axial shell welds.

The examinations detected seven indications in the vessel, one of which exceeded the ASME Code Section XI, IWB-3511 acceptance criteria. The flaw was accepted in an NRC SE dated October 11, 1996. Subsequently, VYNPS sought relief from successive examinations of this flaw under the provisions of BWRVIP-05, Section 9.2.1.2, with the stipulations that the flaw be acceptable for continued service in accordance with IWB-3600, and the flaw be demonstrated acceptable for the intended service life of the vessel. The staff accepted the BWRVIP-05 reports recommendations for successive additional examinations in the SER on BWRVIP-05, and the NRC authorized VYNPS's alternative on April 7, 1999.

On January 5, 2004, the NRC staff issued a request for additional information (RAI) regarding a flaw exceeding the acceptance criteria. The staff inquired about VYNPSs commitment to re-examine this particular flaw according to a letter dated January 28, 1999. The licensee responded to the RAI with a letter, dated May 17, 2004, which states that the subject indication was inspected in April 2004 during RFO 24 using a performance demonstration initiative qualified ultrasonic test sizing technique and showed no enlargement.

Entergy addressed the provisions in GL 98-05 for seeking relief from the requirements of examination of BWR RPV circumferential shell welds as recommended in BWRVIP-05 as discussed in Sections 3.3.1 and 3.3.2 of this SE.

3.3.1 GL 98-05, Provision 1 Entergy provided the following information to demonstrate that at the expiration of the VYNPS operating license in 2012, VYNPS RPV circumferential shell welds will continue to satisfy limiting conditional failure probability for the circumferential welds stated in the NRC staff's July 28, 1998, SER:

a. Neutron Fluence/Embrittlement The BWRVIP-05 report stated, "Embrittlement issues are addressed in 10 CFR 50 Appendix G through requirements associated with upper shelf energy (USE) and the reference temperature of nil-ductility transition (RTNDT). In order to account for the effects of embrittlement, adjusted reference temperatures (ARTs), defined as the initial RTNDT plus the irradiation shift for fluence, are determined. It is possible that ARTs may result in pressure-temperature testing criteria that are difficult to meet due to increased temperature requirements. However, due to low BWR fluence, an unacceptable ART will not be reached, even when extended life is planned." Also, the report states that, "In addition to increasing RTNDT the USE of low alloy steel materials decreases with neutron exposure. However, for the relatively low fluence BWR, maintaining a USE above 50 ft-lbs is not a concern. Also, Code margins required by Appendix G are satisfied at USE values as low as 35 ft-lbs and thus are not a safety concern. Based on the above, it can be seen that although irradiation embrittlement of materials can be a significant concern, its effect is minimal for the relatively low fluence environment of a BWR."

As documented in a letter to the NRC, dated March 26, 2003, the licensee projects that the decrease in USE data for the end of the current operating license is well within the limits provided in NEDO-32205, the equivalent margins topical report applicable to VYNPS. This topical report follows the methods provided in ASME Code Case N-512 and was accepted by the NRC. The projected decrease in limiting plate/weld projected USE decrease will be less than 13.5 percent/7.4 percent and well below the 21 percent/34 percent allowable decrease from NEDO-32205. Therefore, VYNPS remains in compliance with USE requirements of

10 CFR 50 Appendix G by demonstrating that the projected decrease in USE per the guidance of Regulatory Guide (RG) 1.99 meets bounding limits established in the topical report.

b. Probabilistic Fracture Mechanics (PFM) Analysis Although BWRVIP-05 provides a technical basis for this relief, an independent NRC assessment of the analysis contained in the BWRVIP-05 report was conducted. The independent NRC assessment used the FAVOR code to perform PFM analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are: the neutron fluence was estimated to be end-of-license mean fluence, the chemistry values are mean values based on vessel types, and the potential for beyond design basis events were considered.

The following is a statement contained in the "Executive Summary" of the NRC SE of the BWRVIP-05 Report: "It should be noted that the failure frequency for axial welds cited above are relatively high, but that there are known conservatism in these estimates. For example, these analyses were based on the assumption that the flaws in axial weld with the limiting material properties and chemistry are all located at the inside surface of the BWR RPV and at the location of peak end-of-license (EOL) azimuth fluence. Since flaws are distributed throughout the weld and EOL neutron fluence will not occur for many years, the staff has concluded that the present RPV failure frequency is substantially below that reported by the BWRVIP, and independently calculated by the staff and is not a near-term safety concern."

The following information is provided to show the conservatism of the NRC analysis with respect to the VYNPS. Changes in RTNDT may be used as one of the means for monitoring radiation embrittlement of reactor vessel materials. For plants with RPVs fabricated by CB&I, the mean end-of-license neutron fluence for circumferential welds used in the NRC staff and BWRVIP Limiting Plant-Specific Analysis (32 effective full-power years (EFPYs)), Table 2.6-4 of the SE for BWRVIP-05, was 5.1 x 1018 n/cm2. However the peak EOL fast fluence of 2.99 x 1017 n/cm2 (E> 1.0 MeV) for VYNPSs entire beltline is far less than that used in the NRC analysis. Therefore, there is significant conservatism with regard to the effect of fluence on embrittlement in the already low circumferential weld failure probabilities as related to the VYNPS RPV.

Table 1 provides a comparison between the NRC final evaluation of the BWRVIP-05 limiting plant-specific analysis data and VYNPS-specific data for weld chemistry factor and initial RTNDT.

As shown on Table 1, the impact of irradiation results in a lower plant-specific mean RTNDT for the VYNPS circumferential weld material, as compared to that for any of the staffs plant-specific analyses that were performed for the CB&I fabricated RPVs with the highest adjusted reference temperatures. Therefore, based on plant specific data, there is a lower conditional probability of failure for circumferential welds at VYNPS than that stated in the NRCs SE of the BWRVIP-05.

3.3.2 GL 98-05, Provision 2 Entergy provided the following information to demonstrate that VYNPS has implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the NRC staffs July 28, 1998, SER:

At an industry meeting on August 8, 1997, the NRC indicated that the potential for, and consequences of, non-design basis events not addressed in the BWRVIP-05 report should be considered. Later, in an RAI to the BWRVIP, the NRC requested that the BWRVIP evaluate the potential for non-design basis cold overpressure transients and responded to in BWRVIP letter to NRC dated December 18, 1997. The NRC also considered beyond design basis events, such as low temperature overpressure events in their PFM analysis. In the BWRVIP responses to the RAI the total probability of an occurrence of cold overpressure for other than BWR-4s was reported as 9.00 x 10-4. It was concluded that it is highly unlikely that a BWR would experience a cold overpressure transient. In fact, for a BWR to experience such an event would generally require several operator errors. The NRC described several types of events that could be precursors to BWR RPV cold overpressure transients.

These were identified as precursors because no cold overpressure event has occurred at a U.S. BWR. Also, the NRC identified one actual cold overpressure event that occurred during shutdown at a non-U.S. BWR. This event apparently included several operational errors that resulted in a maximum RPV pressure of 1150 psi with a temperature range of 79 to 88 EF.

The following addresses the high-pressure injection sources, administrative controls, and operator training regarding a cold overpressure event for the VYNPS plant:

a. Review of Potential High Pressure Injection Sources High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems The HPCI and RCIC systems use steam driven turbines to pump cold water into the VYNPS vessel. During reactor cold shutdown conditions, there is no steam available to operate these systems, making a cold overpressurization event impossible as the result of operation of these systems.

Feedwater/Condensate Systems The feedwater/condensate systems are potential high-pressure injection water sources into the reactor vessel. The condensate pumps provide water sources to the reactor feed pumps. The feed pumps provide water to the vessel. The normal reactor feed pump discharge pressure is approximately 1300 psig (with condensate pumps running) and the shutoff head pressure is approximately 1680 psig (assuming condensate pumps at shutoff head pressure). The normal condensate pump discharge pressure is approximately 420 psig and the shutoff head is approximately 550 psig. A system design feature of the reactor feed pumps is an automatic trip of all feed pumps on high vessel water level.

The startup procedure requires monitoring of reactor vessel temperatures and pressures. The condensate and feed water pumps are used to control the vessel level during startup. The reactor head vents are not closed until the coolant temperature is greater than 200 EF. This administrative action for head vent closure serves as a mechanism to reduce the likelihood of pressurization above 250 psig. When shutting down, VYNPS procedures require securing reactor feed pumps in sequence depending on the reactor power levels. Monitoring of reactor temperature, pressure, and cool down rates, are prescribed in procedures and Technical Specifications (TSs). During RFOs, the feedwater lines are isolated by closing block valves. At low power (approximately 5 to 10 percent), the lines are secured by removing both main feed regulating valves from service and manually controlling feed water with the auxiliary feed

regulating valve.

Reactor overpressurization by the feedwater/condensate systems is very unlikely since strict controls on temperature and pressure are imposed below 450 EF and the capacity of the systems to inject water is limited by using the auxiliary feed regulating valve. Any unexpected change in reactor water level would allow for operator action. Therefore, these systems do not present a significant potential for overpressurization.

Standby Liquid Control (SLC) System The SLC system is a potential source of high-pressure water into the RPV during cold shutdown conditions. A key lock switch in the control room is required to operate the system since it does not have any automatic start capabilities. As a result, operation of the SLC system is a deliberate act strictly controlled by plant procedures and training. Even if the system was activated, the maximum SLC flow is 40 gpm, a rate that would allow time to control RPV pressure. Therefore, this system does not present a significant potential for overpressurization.

The Low-Pressure Coolant Injection (LPCI), Core Spray (CS), and Residual Heat Removal (RHR) Systems The LPCI, CS, and RHR systems inadvertent operation do not present a significant potential for overpressurization. The pressure-temperature curves for the VYNPS RPV permit increasing system pressures up to 250 psig over the temperature range of 80 to 110 EF and rapidly increase to 810 psig at 110 EF. Shutoff head pressures for CS and RHR pumps are approximately 300 psig. Since these shut-off head pressures could pressurize the reactor vessel to greater than 250 psig, a review of LPCI, RHR and CS operating modes is provided.

The reactor vessel is not likely to pressurize under loss-of-coolant accident (LOCA) conditions.

The LPCI and CS systems would be actuated with the reactor in a depressurized but metal hot condition during a LOCA from full power, which rapidly depressurizes the reactor vessel to a depressurized but metal hot condition. The metal temperature lags pressure substantially and would be greater than 100 EF.

In an emergency, following a loss of shutdown cooling, an alternate shutdown-cooling mode is permitted. This mode of shutdown cooling uses available pump(s) to circulate water from the torus, using a flow path through the reactor vessel and safety relief valves (SRV) discharge lines. The SRV control switches are placed in the "OPEN" position. With the SRV control solenoids energized, when pressure at the SRV main discharge reaches approximately 50 psig, the SRV will open allowing coolant to exit the reactor vessel and flow to the torus via the SRV discharge lines. In this situation, the open SRVs prevent reactor pressure from exceeding 250 psig. This mode of shutdown cooling would only be utilized in emergency situations.

The normal testing of LPCI and CS systems at power and shutdown uses a flow path of suction from the torus with return to the torus. The testing flow path will not enter the reactor vessel because there are two valves in-series that are interlocked to prevent simultaneous opening with reactor pressure above 350 psig. In a shutdown condition with either system under test lineup, an inadvertent injection would be detected by operations and the injection would be terminated before exceeding 250 psig based on observation and alarm of reactor vessel level.

A CS pump may be used for reactor vessel and cavity-fill during the outage. The full flow capability of the CS injection check valves are checked during this evolution. Under these conditions the reactor vessel head is removed, which will prevent overpressurization.

The shutdown-cooling mode of the RHR system prevents overpressurization since the RHR suction valves from the reactor will close at pressure greater than 150 psig, and closure of the suction valves will then trip the running RHR pump preventing vessel pressures from approaching 250 psig.

The reactor head vents are not closed until the coolant temperature is greater than 200 EF.

This administrative action for head vent closure serves as a mechanism to reduce the likelihood of pressurization above 250 psig.

Control Rod Drive (CRD) and Reactor Water Clean-up (RWCU) Systems The CRD and RWCU systems are used to control RPV water level during cold shutdown conditions using a feed and bleed process. Reactor pressure is controlled by venting the reactor to the main condenser via the main steam lines, main steam line vents, or HPCI or RCIC steam line drains when the reactor coolant temperature is less than 200 EF. The low flow rate of these pumps (CRD 55 gpm, RWCU 130 gpm) allows sufficient time for operator action to react to unanticipated level changes, and thus pressure changes. Therefore, these systems do not present a significant potential for overpressurization.

Class 1 Pressure Test The procedure used at VYNPS to perform hydrostatic testing incorporates controls, limitations, and precautions that are reviewed by all personnel involved with maintaining RPV pressure-temperature control. Rigorous abort criteria provide for immediate actions when limits are approached. Senior Operations Management provides management oversight during the evolution and a Senior License Operator is assigned as a Test Director to assure the procedure is coordinated from start to finish. The procedure is only initiated after a detailed pre-job briefing of all affected people at which time all procedural precautions, limitations, and abort criteria are presented. RPV temperature and pressure are monitored throughout the test to ensure compliance with required pressure-temperature limits. Pressurization rates are controlled throughout the test and direction is provided for control using the RWCU or the CRD pumps. Therefore, this system test does not present a significant potential for overpressurization.

In conclusion, the review of potential high-pressure injection sources confirms that a cold overpressure event at VYNPS is extremely unlikely.

b. Review of Operator Training and Work Control Process Reactor Operator Training Licensed operator training provides another method to control reactor water level, temperature, and pressure, in addition to the design and procedural barriers discussed above. Simulator training for start-up and shut-down scenarios provides an opportunity to perform the operator actions to maintain reactor pressure-temperature limits.

Procedural controls for reactor temperature, water level, and pressure are an integral part of operator training. Specifically, operators are trained in methods of controlling RPV water level within specified limits, as well as responding to abnormal RPV water level conditions outside the established limits. Additionally, control room operators receive training on compliance with the

TS pressure-temperature limits curves and the basis for these limitations and curves.

Plant-specific procedures have been developed to provide guidance to the operators regarding compliance with the TS requirements on pressure-temperature limits.

Work Control Process During plant outages, work control procedures require that the outage schedule, and changes to the schedule, receive a risk assessment review commensurate with their safety significance.

Senior operations personnel provide input to the outage schedule to avoid conditions that could adversely impact reactor water level, pressure, or temperature. Schedules are issued listing the work activities to be performed.

During RFOs, work is coordinated through the outage control center. In the control room, the shift manager is required, by procedure, to maintain cognizance of any activity that could potentially affect reactor water level or decay heat removal. The control room operator is required to provide positive control of reactor water level and pressure within the specified bands, and promptly report when operating outside the specified band, including restoration actions being taken. Cognizant individuals involved in the work activity attend pre-job briefings. Expected plant responses and contingency actions to address unexpected conditions, or responses that may be encountered, are included in the briefing discussion.

Based upon the above, the probability of a low temperature RPV overpressure event at VYNPS is considered to be less than, or equal to, that used in the NRCs SE.

4.0 STAFF EVALUATION As described previously, GL 98-05 lists two provisions that BWR licensees requesting relief from ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential RPV welds (ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item No.

B1.11, Circumferential Shell Welds) must satisfy. These provisions are intended to demonstrate that the conditions at the applicants plant are bounded by those in the SE. The licensee will still need to perform the required inspections of essentially 100 percent of all axial welds.

4.1 Circumferential Weld Conditional Failure Probability The staffs SER for the BWRVIP-05 report evaluated the conditional failure probability of circumferential welds for the limiting plant-specific case of BWR RPVs manufactured by different vendors, including CB&I, using the highest mean irradiated RTNDT to determine the limiting case. Since the VYNPS RPV was fabricated by CB&I, the licensee compared the mean irradiated RTNDT for VYNPS to that for the limiting CB&I case described in Table 2.6-4 of the staffs SER for the BWRVIP-05 report. As indicated in the licensees evaluation, the mean RTNDT for VYNPS is lower than that for the limiting CB&I case; therefore, the licensee concluded that the conditional failure probability for the VYNPS circumferential welds is bounded by the conditional failure probabilities in the staffs SER for the BWRVIP-05 report through the end of the current license period.

One of the provisions in GL 98-05 is to demonstrate that at the expiration of the license, the RPV shell circumferential welds continue to satisfy the limiting conditional failure probability for RPV shell circumferential welds that is established in the July 30, 1998 Safety Evaluation. The

first requirement for the conditional vessel failure probability is that the peak vessel fluence at the end of the current license remain equal to or less than the value assumed in the staff evaluation.

The staff issued Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, in March 2001 (Ref. 6). The RG provides guidance regarding acceptable methods for the benchmarking of vessel fluence methodologies based on the requirements of General Design Criterion (GDC) 31 and in part on GDCs 14 and

30. Therefore, the staff is basing the review of the peak vessel fluence evaluation for the Vermont Yankee plant on the adherence of the calculational method to the guidance in RG 1.190.

The NRC staff had previously evaluated the peak vessel fluence value for VYNPS as part of its evaluation of a proposed Technical Specification change to revise the pressure-temperature limits for the reactor pressure vessel. The staff approved the TS change by letter dated March 29, 2004. In its SE, the staff concluded that the licensees proposed fluence value of 2.99x1017 n/cm2 was estimated using a methodology consistent with the guidance in RG 1.190.

The calculated value accounts for past fuel loading patterns and projected loadings adjusted to the 24 month fuel cycle. Therefore, the staff concluded that the peak vessel fluence value of 2.99x1017 n/cm2 is acceptable through the current license period of 32 EFPY.

Comparison of the accepted value for 32 EFPYs and the limiting value indicates that the accepted value is significantly lower than the limiting value of 5.1x1018 n/cm2 used in the staff evaluation of July 30, 1998, and therefore, is acceptable.

The staffs SER for the BWRVIP-05 report provides a limiting conditional failure probability of 2.00 x 10-7 per reactor year for a limiting plant-specific mean RTNDT of 44.5 EF for CB&I fabricated RPVs. Comparing the information in the NRC Reactor Vessel Integrity Database with that submitted in the proposed relief request, the staff confirmed that the mean RTNDT of the circumferential welds at VYNPS is projected to be -58.2 EF at the end of the current license.

In this evaluation, the chemistry factor, RTNDT, and mean RTNDT were calculated consistent with the guidelines of RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials. The calculated value of mean RTNDT for the circumferential welds at VYNPS is significantly lower than that for the limiting plant-specific case for CB&I fabricated RPVs, indicating that the conditional failure probability of the VYNPS circumferential welds is much less than 2.00 x 10-7 per reactor year.

Previous examinations at VYNPS detected seven indications in the vessel, one of which exceeded the ASME Code Section XI, IBW-3511 acceptance criteria. The flaw in question is a circumferential weld and was characterized as a subsurface flaw. This flaw met all three criteria of BWRVIP-05 for the elimination of the successive re-examination requirements of the ASME Code and was approved for continued service. This flaw was accepted in an NRC SE dated October 11, 1996. Subsequently, on January 28, 2004, the staff requested additional information regarding VYNPSs re-examination commitment for this flaw. In a letter dated May 17, 2004, Entergy informed the NRC staff that the subject indication was inspected on April 2004 during VYNPSs most recent refueling outage, RFO 24. The inspection results indicate that the flaw dimensions have not enlarged and they remain bounded by that previously categorized. This provides additional assurance that an acceptable level of quality and safety will be maintained for the alternative inspection requested by the licensee.

4.2 Minimizing the Possibility of Low Temperature Overpressurization VYNPS has demonstrated that, through design of the systems and the implementation of administrative controls, it is highly unlikely that the activation of high pressure injection will occur during low reactor coolant system (RCS) temperatures.

HPCI and RCIC Both systems are steam operated and are inoperable during reactor cold shutdown conditions.

Feedwater/Condensate Systems The feedwater/condensate system has a discharge pressure of 1300 psig. The system has an automatic trip of all feed pumps on a high water level signal. During startup, the feedwater and condensate pumps control the water level in the vessel. The vessel head vents are open until the RCS temperature rises above 200 EF and is used to reduce the likelihood of cold pressurization. When shutting down, RCS temperature, pressure and cooldown rate are monitored to comply with TSs. The feedwater pumps are secured sequentially and feed valves are manually operated to control feedwater and water level. Cold overpressurization is very unlikely, either during startup or shutting down.

SLC System The SLC system is activated from a key lock at the control room and is operated under plant procedures. However, the maximum flow is 40 gpm, thus, overpressurization with the standby liquid system is highly unlikely.

LPCI, CS, and RHR Systems The pressure-temperature curves for VYNPS allow pressures up to 250 psig for temperatures between 80 to 110 EF. Above 110 EF, the acceptable pressure rises rapidly to 810 psig. The LPCI, CS and RHR systems have a maximum injection pressure of 300 psig. Therefore, overpressurization is possible between 250 and 300 psig at temperatures below 110EF.

The LPCI and the CS systems are activated at low pressures following a LOCA or loss of shutdown cooling. Following a LOCA, the vessel depressurizes quickly, metal temperature is high and able to sustain the low injection pressures. Following loss-of-shutdown cooling, available pumps are used to circulate water from the torus through the vessel and out through the SRVs. In this condition, the relief valve control is set manually to open, which actually opens the valves at 50 psig, preventing vessel overpressure. Inadvertent low pressure injection under test conditions will be detected by the operator and terminated before it reaches 250 psig.

CRD and RWCU Systems Both systems are used to control vessel water level under cold shutdown conditions. Vessel pressure is controlled by venting to the condenser through the steam lines, for coolant temperatures under 200 EF. The low flow of both systems allows time for operator intervention if water level or pressure shows inadvertent increase. Therefore, these systems do not present a significant risk for cold overpressurization.

Vessel Pressure Testing Plant procedures for vessel hydrostatic testing is performed under controls and limitations involving senior reactor operations personnel. Vessel temperature and pressure have rigorous

abort criteria to assure compliance with pressure and temperature limits. Therefore, the vessel hydrostatic testing does not present a significant risk for cold overpressurization.

Operator Training and Work Control Processes Operators are specifically trained in the simulator on compliance with TS pressure-temperature limits. Plant-specific procedures have been developed to provide operator guidance regarding TS compliance.

During plant outages, all reactor procedures are assessed for their risk significance. In particular, procedures and evolutions which affect water level, vessel pressure and decay heat removal are under constant shift manager cognizance. Expected plant responses and contingency options are always available.

Based on the above, the NRC staff finds that the risk of cold overpressurization at the VYNPS plant is less or equal than that used in the NRC SE.

5.0 CONCLUSION

The NRC staff has reviewed the licensees submittal and finds that the licensee has acceptably demonstrated that the appropriate provisions in GL 98-05 and the staffs evaluation of the BWRVIP-05 report have been satisfied regarding permanent relief (i.e., for the remaining portion of the current license period of 32 EFPYs) from ISI requirements of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, for the volumetric examination of RPV circumferential welds.

The NRC staff concludes that the licensees permanent relief request from the examination requirements of circumferential shell welds for the remaining portion of the current license period of 32 EFPYs, pursuant to 10 CFR 50.55a(a)(3)(i), is acceptable and is consistent with the information contained in NRC GL 98-05. The staff has also determined that the alternative program provides an acceptable level of quality and safety.

Additional requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third party reviews by the Authorized Nuclear Inservice Inspector.

Principal Contributors: L. Lois V. Rodriguez N. Ray Date: September 30, 2004

TABLE 1 Vermont Yankee Nuclear Power Station Reactor Pressure Vessel (RPV) Shell Weld Information Bounding Circumferential Weld Vermont Yankee US NRC Shell Bounding Beltline Limiting 32 EFPYs Parameter 32 EFPYs Bounding CB&I Vessel Description Bounding Comparative Parameters Parameters SER Table 2.6-4 (Bounding Circ. Welds)

Neutron fluence at the end of the requested relief period 2.99 x 1017 n/cm2 5.1 x 1018 n/cm2 (Peak Surface Fluence Entire Beltline)

Initial (unirradiated) reference temperature -70 -65 (RTNDT), EF Weld Chemistry Factor (CF),

54 134.9 EF Weld Copper content % 0.04 0.10 Weld Nickel content % 1 0.99 Increase in reference temperature ( RTNDT),

11.8 109.5 EF=CF*(Fluence/10^19)^(0.2 8-0.1*LOG(Fluence/10^19))

Mean adjusted reference temperature (ART), -58.2 44.5 EF=(RTNDT(u) + RTNDT)