ML050130317

Ltr, RR No. 38 (TAC No. MC2381, MC2382)

ML050130317
Person / Time
Site:	Hatch
Issue date:	01/28/2005
From:	John Nakoski NRC/NRR/DLPM/LPD2
To:	Sumner H Southern Nuclear Operating Co
Gratton C, NRR/DLPM, 415-1055
References
TAC MC2381, TAC MC2382
Download: ML050130317 (15)

Text

January 28, 2005 Mr. H. L. Sumner, Jr.

Vice President - Nuclear Hatch Project Southern Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201-1295

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2 - EVALUATION OF RELIEF REQUEST (RR) NUMBER 38 (TAC NOS. MC2381 AND MC2382)

Dear Mr. Sumner:

By letter dated March 29, 2004, as supplemented by letter dated September 13, 2004, Southern Nuclear Operating Company, Inc. (SNC or the licensee), submitted proposed alternatives to the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) under the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(a)(3)(ii) for the Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2. SNC proposed two additional relief requests, RR-39 and RR-40, in the March 29, 2004, letter. The Nuclear Regulatory Commission (NRC) staff reviewed RR-39 and RR-40, and provided the results of that review in a letter to you dated January 7, 2005.

The following paragraphs summarize the NRC staffs findings regarding RR-38:

In RR-38, the licensee proposed an alternative inspection program for the circumferential welds in the reactor vessels (RVs) of Hatch, Units 1 and 2. The alternative program applies a probabilistic fracture toughness analysis to justify eliminating volumetric examinations of the RV circumferential welds, as required in Table IWB-2500-1 of Section XI of the ASME Code, and the augmented volumetric inspections for these welds, as required in 10 CFR 50.55a(g)(6)(ii)(A)(2). The NRC staff completed their review and determined that the licensees proposed alternative provides an acceptable level of quality and safety. Therefore, the licensee's alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

The March 29, 2004, letter requested that RR-38 be approved for the remainder of the 40-year initial license. During review of the RR, the NRC staff and SNC agreed that the RR should be amended to include the period of extended operation (i.e., through its current 60-year renewed license). Upon further review, the NRC staff determined that the neutron fluence values used in the RR were based on a calculational code (i.e., RAMA) that had not been reviewed and approved by the NRC staff. The licensee had initially anticipated that the NRC staffs review and acceptance of the RAMA code for calculating reactor pressure vessel neutron fluence would be complete by the time this RR was needed. The NRC staff is currently reviewing, but has not yet approved the use of the RAMA code for neutron fluence calculations. As a result, the NRC staff reviewed the licensees current, NRC approved neutron fluence analyses and concluded that the fluence values calculated using this methodology are conservative through July 31, 2007. Therefore, the NRC staff limits the duration of RR-38 until July 31, 2007.

H.L. Sumner, Jr Please note that if you plan to seek approval for RR-38 beyond July 31, 2007, you must submit a revised RR based on an analysis that uses an NRC staff-approved neutron fluence calculation methodology (e.g., RAMA, if approved for use by the NRC staff at that time). This issue was discussed with SNC staff in a teleconference on September 20, 2004.

The NRC staff's Safety Evaluation is enclosed. If you have any questions, please contact Christopher Gratton at 301-415-1055.

Sincerely,

/RA/

John A. Nakoski, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-321 and 50-366

Enclosure:

As stated cc w/encl: See next page

ML050130317 NRR-028 OFFICE PDII-1/PM PDII-1/LA SRXB(B) EMCB/SC(A) OGC PDII-1/SC NAME CGratton CHawes JUhle BElliot MZobler JNakoski DATE 1/27/05 1/27/05 01/06/05 1/11/05 1/18/05 1/28/05

Edwin I. Hatch Nuclear Plant, Units 1 & 2 cc:

Laurence Bergen Chairman Oglethorpe Power Corporation Appling County Commissioners 2100 E. Exchange Place County Courthouse P.O. Box 1349 Baxley, GA 31513 Tucker, GA 30085-1349 Mr. Jeffrey T. Gasser Mr. R.D. Baker Executive Vice President Manager - Licensing Southern Nuclear Operating Company, Inc.

Southern Nuclear Operating Company, Inc. P.O. Box 1295 P.O. Box 1295 Birmingham, AL 35201-1295 Birmingham, AL 35201-1295 Mr. G. R. Frederick, General Manager Resident Inspector Edwin I. Hatch Nuclear Plant Plant Hatch Southern Nuclear Operating Company, Inc.

11030 Hatch Parkway N. U.S. Highway 1 North Baxley, GA 31531 P.O. Box 2010 Baxley, GA 31515 Harold Reheis, Director Department of Natural Resources Mr. K. Rosanski 205 Butler Street, SE., Suite 1252 Resident Manager Atlanta, GA 30334 Oglethorpe Power Corporation Edwin I. Hatch Nuclear Plant Steven M. Jackson P.O. Box 2010 Senior Engineer - Power Supply Baxley, GA 31515 Municipal Electric Authority of Georgia 1470 Riveredge Parkway, NW Atlanta, GA 30328-4684 Mr. Reece McAlister Executive Secretary Georgia Public Service Commission 244 Washington St., SW Atlanta, GA 30334 Arthur H. Domby, Esq.

Troutman Sanders Nations Bank Plaza 600 Peachtree St, NE, Suite 5200 Atlanta, GA 30308-2216

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST (RR) NO. RR-38 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-321 AND 50-366

1.0 INTRODUCTION

In Letter No. NL-04-0478 dated March 29, 2004, as supplemented by Serial Letter No.

NL-04-1764 dated September 13, 2004, Southern Nuclear Operating Company, Inc. (SNC, the licensee), proposed an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI and applicable Addenda (henceforth Section XI), regarding the volumetric examination requirements for reactor vessel (RV) circumferential shell welds at the Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2. The licensee also requested relief from the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g)(6)(ii)(A)(2), as they pertain to performing augmented volumetric inspections of the RV circumferential welds at Hatch, Units 1 and 2.

Nuclear Regulatory Commission (NRC, the Commission) staff approval of the RR would authorize the use of a proposed alternative to these volumetric examinations requirements in accordance with the alternative probabilistic fracture mechanics methods discussed in Electric Power Research Institute Topical Report, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), and with the NRCs guidelines for proposing these alternative programs, as established in NRC Generic Letter (GL) 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.

2.0 REGULATORY EVALUATION

2.1 Inservice Inspection Requirements Inservice inspection (ISI) of the ASME for Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as required by 10 CFR 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states that alternatives to the requirements of Enclosure

paragraph (g) may be used, when authorized by the NRC, if: (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection [ISI] of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable Code of record for the second 10-year ISI for Hatch, Units 1 and 2 is the 1989 Edition of the ASME Code,Section XI.

2.2 Augmented Inservice Inspections Requirements for RV Shell Welds 10 CFR 50.55a(g)(6)(ii)(A)(2) requires licensees to augment their reactor vessel examinations by implementing, as part of the ISI interval in effect on September 8, 1992, the examination requirements for RV shell welds specified in Item B1.10,Section XI, Table IWB-2500-1, Examination Category B-A, Pressure Retaining Welds in Reactor Vessel.Section XI Item B1.10 includes the volumetric examination requirements for both circumferential RV shell welds, as specified in Section XI Item B1.11, and longitudinal RV shell welds, as specified in Section XI Item B1.12. 10 CFR 50.55a(g)(6)(ii)(A)(2) defines "essentially 100% examination as covering 90 percent or more of the examination volume of each weld.

2.3 Additional Regulatory Guidance 2.3.1 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 Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP), a technical committee of the BWR Owners Group (BWROG), submitted proprietary report BWRVIP-05. The BWRVIP-05 report evaluates the current inspection requirements for RV shell welds in BWRs, formulates recommendations for alternative inspection requirements, and provides a technical basis for these recommended requirements. As modified, the BWRVIP-05 proposed to reduce the scope of inspection of BWR RV welds from essentially 100 percent of all RV shell welds to examination of 100 percent of the axial (i.e., longitudinal) welds and essentially zero percent of the circumferential RV 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 BWRVIP-05.

This evaluation concluded that the failure frequency of RV 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 will be performed if axial weld examinations reveal an active degradation mechanism. The NRC staff supplemented this SER in an SER to the BWRVIP dated March 7, 2000. In this SER, the NRC staff updated the interim probabilistic failure frequencies for RV axial shell welds and revised the Table 2.6-4 to correct a typographical error in the 32 effective full power years (EFPY) Mean RTNDT value cited for the limiting Chicago Bridge and Iron (CB&I) case study for circumferential welds. The correction changed the 32 EFPY Mean RTNDT value for the CB&I case study from 109.5 EF to 134.9 EF.

In the BWRVIP-05 report, the BWRVIP committee concluded that the conditional probabilities of failure for BWR RV 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 NRC staffs assessment conservatively calculated the conditional probability of failure values for RV 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. The failure frequency 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 CB&I, Combustion Engineering (CE), and Babcock and Wilcox (B&W). The analysis identified a cold overpressure event that occurred in a foreign reactor as the limiting event for BWR RVs, with the pressure and temperature from this event used in the probabilistic fracture mechanics calculations. The NRC 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 NRC staffs SER of March 7, 2000, identifies the conditional failure probabilities for the plant-specific conditions with the highest projected reference temperature (for that fabricator) through the expiration of the initial 40-year license period. Table 2.6-5 of NRC staffs SER of July 28, 1998, identifies the conditional failure probabilities for the plant-specific conditions with the highest projected reference temperature (for that fabricator) through the expiration of an 80-year license period, which constitutes the licensing basis if two 20-year extended periods of operation have been granted for a BWR-designed nuclear power plant.

2.3.2 Generic Letter 98-05 On November 10, 1998, the NRC issued GL 98-05 that 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 reactor pressure vessel welds (ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, Circumferential Shell Welds) by demonstrating conformance with the following safety criteria:

(1) At the expiration of the operating license, the licensees will have demonstrated that limiting probability of failure for their limiting RV circumferential welds will continue to satisfy (i.e., be less than) the limiting conditional failure probability for circumferential weld assessed in the applicable BWRVIP-05 limiting case study.

(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.

In GL 98-05, the NRC staff stated that licensees applying the BWRVIP-05 criteria would need to continue performing the volumetric inspections of all axial RV shell welds that are required by the ASME Code,Section XI, Table IWB-2500-1, Inspection Category B-A, Item B1.12, and the augmented volumetric inspections of the RV axial shell welds that are required under 10 CFR 50.55a(g)(6)(ii)(A)(2). For plants that are currently licensed to operate in accordance with their initial 40-year operating licenses, the limiting case studies are provided in Table 2.6-4 of the revised SER on BWRVIP-05 dated March 7, 2000. For plants that have been granted operating licenses to operate for an extended period of operation, the limiting case studies are provided in Table 2.6-5 of the NRC staffs SER of July 28, 1998.

3.0 TECHNICAL EVALUATION

3.1 Code Requirement for Which Relief is Requested The licensee requested relief from the following requirements in the 1989 Edition of Section XI of the ASME Code (Section XI):

• Subarticle IWB-2500, Table IWB 2500-1, Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, No. B1.11, Circumferential Shell Welds.

3.2 Licensees Proposed Alternative to the ASME Code Using the guidelines of GL 98-05 and Topical Report BWRVIP-05 and the NRC staffs determination in its July 28, 1998, SER on BWRVIP-05, the licensee proposed to use a probabilistic fracture mechanics evaluation for the circumferential shell welds in the Hatch, Units 1 and 2 RVs as the basis for eliminating the required volumetric examinations and augmented volumetric examinations for the welds through the expiration of the extended periods of operation for Hatch, Units 1 and 2.

The licensee proposed the following alternative in lieu of performing the required volumetric examinations of the RV circumferential shell welds:

Axial welds and intersecting portions of circumferential welds will be examined to the extent practical, dependent upon interference by another component or restrictions due to the geometrical configuration. For those cases where the reduction in coverage is greater than 10%, relief will be requested pursuant to 10 CFR 50.55a requirements.

In SNC Letter No. NL-04-1764, dated September 13, 2004, the licensee clarified that the alternative inspection program in RR-38 is requested through the expiration of the periods of extended operation for Hatch, Units 1 and 2.

3.3 Licensees Bases for Alternative The licensee based RR-38 on the NRCs RR provisions of GL 98-05 and the guidelines of BWRVIP-05. The licensee cited the following acceptance criteria as the bases for evaluating the acceptability of RR-38.

Per the NRC SE dated July 28, 1998 and Generic Letter 98-05, BWR licensees may request relief from the inservice inspection requirements of 10 CFR 50.55a(g) for volumetric examination of circumferential reactor pressure welds (ASME Section XI Code, Table IWB-2500-1, Examination Category B-A, Item 1.11, Circumferential Shell Welds) by demonstrating:

1. At the expiration of their license, the circumferential welds satisfy the limiting conditional failure probability for circumferential welds in this evaluation . . . .[GL 98-05 Safety Criterion 1].

2. Licensees have implemented operator training and established procedures that limit the frequency of cold over pressure events to the amount specified in this report . . . .[GL 98-05 Safety Criterion 2].

3.3.1 License Basis for Conforming with GL 98-05 Safety Criterion 1 - Criterion for Conditional Probabilities of Failure In letter dated March 29, 2004, the licensee provided its 54 EFPY Mean RTNDT calculations for the limiting circumferential welds in the Hatch, Units 1 and 2 RVs (Refer to Enclosure 1 of RR-38) in order to support its basis for meeting GL 98-05 Safety Criterion 1 and to demonstrate that the 54 EFPY Mean RTNDT values for Hatch, Units 1 and 2 are bounded by the Mean 64 EFPY RTNDT value for the limiting CE-VIP case study.

3.3.2 License Basis for Conforming with GL 98-05 Safety Criterion 2 - Criterion on Mitigating the Probability of Cold Overpressurization Events The licensee provided the following technical basis for meeting GL 98-05 Safety Criterion 2:

SNC has previously demonstrated that operator training and established procedures limit the frequency of cold over pressure events. This information was supplied to the NRC in the December 2, 1998 Hatch Unit 1 submittal (for the permanent deferral of the augmented examination requirements), which was subsequently approved by the NRC in the March 11, 1999 SE. This information was later referenced by SNC in the January 31, 2001 response to License Renewal RAls, where, it was also noted that the operator training and procedures for Hatch Units 1 and 2 are the same. Extracts of this information are shown in Enclosure 3 [of SNC Letter NL-04-478, dated March 29, 2004].

4.0 NRC STAFF EVALUATION As discussed in Section 2.3.2 of this SE, GL 98-05 provides two criteria that BWR licensees requesting relief from ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential RV welds (ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, Circumferential Shell Welds) must satisfy. These criteria are intended to demonstrate that the conditions at the applicants plant are bounded by those in the safety evaluation. The licensee will still need to perform the required inspections of essentially 100 percent of all axial welds.

4.1 Neutron Fluence Calculation for RR-38 For any given RV circumferential or axial weld material, the conditional probability of failure increases with the materials neutron fluence value and mean RTNDT value, as projected to the expiration of the operating license for the facility. GL 98-05 stipulates that, at the expiration of the operating license, the mean RTNDT estimates for circumferential welds should satisfy the limiting conditional failure probability for the weld materials, as stated in the NRC staffs SER of July 28, 1998. The neutron fluence values for the RV circumferential welds at the inside surface of the RV are critical inputs to the mean RTNDT estimate calculations.

SNCs current method for calculating neutron fluence does not conform with the NRC staffs recommended methodology in Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure VEssel Neutron Fluence. SNC had committed to have a conforming neturon fluence calculational methodology (i.e., the RAMA code) for Hatch, Units 1 and 2 approved by December 15, 2004. However, the approval of the RAMA code has been delayed. In a letter dated July 13, 2004 (SNC Letter No. NL-04-1123), SNC requested to revise its commitment date for having a neutron fluence calculational methodology that is compliant with RG 1.190, from December 15, 2004, to July 31, 2007. SNC stated that the previous commitment date was selected arbitrarily (based on the expectation that RAMA would be approved by this date), and that there was sufficient conservatism in the pressure-temperature (P-T) curves calculated using the current fluence methodology to allow the plant to operate safely until July 31, 2007. The NRC approved the commitment change request in a letter to SNC dated September 9, 2004. The NRC staff is currently reviewing the RAMA code and anticipates that the review will be complete before July 31, 2007.

NRC staff experience has shown that neutron fluence values that are calculated using methods that do not conform with RG 1.190 are typically within +/- 40 percent of the values that would be obtained using the recommended methodology of RG 1.190. As stated previously, SNCs current fluence methodology does not conform with RG 1.190. SNC stated in a Letter No.

NL-04-1152, also dated July 13, 2004, that based on the current fluence methodology, the estimated fluence values on August 1, 2007, for Hatch Units 1 and 2, will be 24.2 and 22.1 EPFY, respectively. These values are 44.8 percent and 40.9 percent of the 54 EFPY neutron fluence values estimated by the current methodology for the Hatch, Units 1 and 2, respectively.

As a result, neutron fluence estimates for August 1, 2007, are more conservative than the values estimated for the fluence at 54 EFPY, even considering the 40 percent adjustment for the nonconforming methodology.

Therefore, the NRC staff finds the licensees neutron fluence estimates are acceptable to warrant approval of fluence values used in the 54 EFPY Mean RTNDT analyses. However, the NRC staff is limiting the acceptance of RR-38 to the period through July 31, 2007. SNC also stated in NL-04-1152 that the P-T limit curves will be evaluated and revised, if necessary at that time.

4.2 Circumferential Weld Conditional Failure Probability The NRC staffs SER for the BWRVIP-05 report evaluated the conditional failure probabilities for axial and circumferential shell welds in the limiting BWR RV designs manufactured by different vendors, including RVs manufactured using by CE, CB&I, and B&W. The SER also reported the Mean RTNDT calculations and values that were derived from the conditional failure probabilities for the limiting case studies. For a plant granted a renewed operating license, the evaluation criteria for the limiting conditional failure probabilities and Mean RTNDT values are those listed for the limiting case studies specified in Table 2.6-5 of the staffs SER of July 28, 1998. The associated limiting case studies, conditional failure probabilities, and Mean RTNDT values listed in Table 2.6-5 of the NRC staffs SER are limited to plants that have accumulated no more than 64 EFPY of power operation.

The renewed operating licenses for Hatch, Units 1 and 2 were approved and issued by the NRC on January 15, 2002. In the renewed operating licenses, the staff granted power operation through August 6, 2034, for Hatch, Unit 1 and June 13, 2038, for Hatch, Unit 2, which represent operations through 54 EFPY of power operation. The period of applicability in Table 2.6-5 of the SER on BWRVIP-05 is bounding for operations of the Hatch, Units 1 and 2 reactors to the expiration of the extended operating licenses and is representative of the evaluation for RR-38.

Since the Hatch, Units 1 and 2 RVs were fabricated by CE, the CE-VIP limiting case study in Table 2.6-5 provides the applicable case-study conditional probability of failure value and Mean RTNDT value criterion for the evaluation of RR-38.

In the license renewal application for the Hatch, Units 1 and 2 reactors, SNC identified the calculation of the Mean RTndt values for the Hatch, Units 1 and 2 RV circumferential welds as a time-limiting aging analysis (TLAA) for the application. In the staffs evaluation in Section 4.6 of NUREG-1803, Safety Evaluation Report Related to the License Renewal of the Edwin I. Hatch Nuclear Plant, Units 1 and 2 (December 2001), the NRC staff concluded that SNC had performed a valid TLAA analysis to justify re-submittal of the alternative inspection proposal for the Hatch, Units 1 and 2 RV circumferential welds to the expiration of the extend periods of operation for the reactor units. SNCs submittal of RR-38 on March 29, 2004, as amended in SNC Letter No. NL-04-1764, dated September 13, 2004, was performed to justify elimination of the volumetric examinations and augmented volumetric examinations for the RV circumferential welds through the expiration of the extended periods of operation for Hatch, Units 1 and 2.

The NRC staff performed an independent calculation of the Mean RTNDT values for the limiting Hatch, Units 1 and 2 RV circumferential welds through 54 EFPY. Table 4.1-1 on page 8 of this SE provides a summary of the Mean RTNDT values calculated by the staff for the Hatch, Units 1 and 2 RV through 54 EFPY and a comparison of the staffs Mean RTNDT values to both the corresponding Mean RTNDT values calculated by SNC and the Mean RTNDT value criterion for the limiting CE-VIP case study at 64 EFPY.

The results in Table 4.1-1 demonstrate that the Mean RTNDT values calculated by the licensee for the Hatch, Units 1 and 2 RV circumferential welds are less than that for the limiting CE-VIP case study and are in agreement with those calculated by the NRC staff. Based on this analysis, the NRC staff concludes that SNC has provided a valid basis for concluding that the conditional probability of failure values for the Hatch, Units 1 and 2 RV circumferential welds are sufficiently low to justify elimination of the volumetric examinations that are required for these welds through 54 EFPY. However, for the reasons stated in Sections 4.1 and 5.0 of this SE, the NRC staff is limiting its authorization for elimination of the required examinations until July 31, 2007.

Table 4.1-1 Comparison of NRC and SNC 54 EFPY Mean RTNDT Calculations to the 64 EFPY Mean Calculations for the Limiting CE-VIP Case Study on BWRVIP-05 NRC 54 EFPY SNC 54 EFPY NRC 54 EFPY SNC 54 EFPY Limiting Mean RTNDT Mean RTNDT Mean RTNDT Mean RTNDT 64 EFPY Calculations Calculations Calculations Calculations CE-VIP for Hatch, for Hatch, for Hatch, for Hatch, Case Study Unit 1 Unit 1 Unit 2 Unit 2 (Note 1) (Note 1) (Note 1) (Note 1)

Alloy % Cu 0.13 0.197 0.197 0.047 0.047 Alloy % Ni 0.71 0.060 0.060 0.049 0.049 RTNDT(U) (EF) 0 -10.0 -10.0 -50.0 -50.0 Fluence 0.400 0.236 0.236 0.244 0.244 (1019 n/cm2)

Chemistry Factor 151.7 91.4 91.0 30.7 31.0 RTNDT (EF) 113.2 55.7 55.5 19.0 19.2 Mean RTNDT (EF) 113.2 45.7 45.5 -31.0 -30.8 NRC Established 1.99E-4 Mean RTNDT is Mean RTNDT is Mean RTNDT is Mean RTNDT is Probability of Failure (no failure: Lower than Lower than Lower than Lower than

[ P(F/E) ]Criterion for Note 2) Case Study Case Study Case Study Case Study Case / Result for Plant Mean RTNDT : Mean RTNDT : Mean RTNDT : Mean RTNDT :

Specific Calculation Criterion is Criterion is Criterion is Criterion is met. (Note 2) met. (Note 2) met. (Note 2) met. (Note 2)

Notes: 1. For the Hatch, Units 1 and 2 RVs, the limiting circumferential weld materials determined by the staff were equivalent to those determined by SNC. For Hatch-1, the limiting RV circumferential weld is 1-313A, which was fabricated from weld heat No. 90099. For Hatch-2, the limiting RV circumferential weld is 301-871, which was fabricated from weld heat No. 4P6052.

2. If the plant-specific Mean RTNDT is less than the Mean RTNDT associated with Limiting Case Study, the staff concludes that probability of failure for the plant-specific circumferential weld under review will be less that for the limiting circumferential weld in the Limiting Case Study. BWR plants that meet this criterion may conclude that the probability of failure for the limiting circumferential RV welds is sufficient to justify elimination of the volumetric examinations required by Section XI of the ASME Code (Examination Category B-A, Item B.1.11) and augmented volumetric examinations for the circumferential welds required by 10 CFR 50.55a(g)(6)(ii)(A)(2).

4.3 Minimizing the Possibility of Low Temperature Overpressurization The licensee stated that its bases for meeting Acceptance Criterion 2 of GL 98-05 and for demonstrating that the licensee has implemented acceptable procedures and controls for mitigating a low-temperature-overpressurization event are given in the Consideration of Low Temperature - Over Pressurization Events section of the licensees enclosure to SNC Serial Letter HL 5710, dated December 2, 1998, and are applicable to the evaluation of RR-38. In this letter, the licensee stated that the following operational controls, procedural controls, and staff training practices are in place to minimize the possibility of a low temperature overpressurization event.

4.3.1 Operational and System Design Considerations From an operational basis, the reactor feedwater system (RFS), high pressure coolant injection system (HPCI), reactor core isolation cooling system (RCIC), and standby liquid control system (SLC) are the high pressure systems that provide coolant or have the potential to provide coolant at high pressure into the RV. The HPCI and RCIC pumps are steam driven and cannot function during cold shutdown. The RFS pumps automatically trip during transient and postulated loss-of-coolant accidents (LOCA) conditions, and are manually tripped during routine reactor shutdowns. Since RFS pumps are steam driven, they cannot be operated during cold shutdown condition. Although not addressed in the licensee's submittal, the NRC staff noted that SLC is solely a manual injection system; there are no automatic starts associated with SLC at Hatch, Units 1 and 2. Operator initiation of the SLC occurs only in accordance with applicable Hatch, Units 1 and 2 emergency operating procedures and would not occur during normal utility planned shutdowns of the reactors or during transient operating conditions. SLC might be manually initiated during a postulated design basis LOCA; however, if manually initiated, the SLC injection rate of approximately 40 gallons per minute (gpm) would allow operators sufficient time to control RV water level and pressure during the postulated event.

The core spray system (CS), low-pressure coolant injection/residual heat removal system (LPCI/RHR), control rod drive system (CRD), and reactor water cleanup system (RWCU) can also inject coolant into the reactor. CS and LPCI/RHR are low pressure emergency core cooling systems (ECCS), whose pumps create a shutoff head of 375 pounds per square inch differential (psid) for CS and 223 psid for LPCI/RHR. Should either of these systems be started (i.e., inject as designed) during cold shutdown, the resulting reactor pressure and temperature would be below the P-T limits. CRD and RWCU use a feed-and-bleed process to control RV level and pressure during normal cold shutdown conditions.

4.3.2 Procedural Considerations Plant-specific normal operating and transient operating procedures have been established to provide guidance to the operators regarding compliance with the Technical Specification P-T limits. These procedures direct operators to respond to any unexpected or unexplained rise in reactor water level, which could result from spurious actuation of an injection system. The procedural actions include preventing condensate pump injection, securing ECCS system injection, tripping CRD pumps, terminating all other injection sources, and lowering RV level via the RWCU system. In addition, plant-specific emergency operating procedures have been

established to ensure that proper operating actions are followed during postulated design-basis LOCAs. The emergency procedures include control of reactor water level, reactor pressure, and reactor temperature during these postulated events and instructions on how and when SLC should be manually initiated.

4.3.3 Operator Training The licensee emphasized that training and testing of control room operators is an integral part of ensuring the abilities of the operators to implement these procedures. On the basis of the P-T limits of the operating systems, operator training, and established plant-specific procedures, the licensee determined that a nondesign-basis cold overpressure transient is unlikely to occur.

4.3.4 Staff Determination on SNCs Basis for Meeting Acceptance Criterion 2 of GL 98-05 The staff concluded that, based on the licensees information provided about the systems that inject at high pressures, operator training, and plant-specific procedures at Hatch, Units 1 and 2, the possibility of a low temperature overpressurization event will be minimized, and thus, the licensee has provided a sufficient basis to support the NRC staffs approval of the alternative examination request for circumferential shell welds in the Hatch, Units 1 and 2 RVs.

5.0 CONCLUSION

FOR RR-38 The NRC staff has completed its review of the licensees submittal and determined that the licensee conforms to the applicable safety evaluation criteria in NRC GL 98-05 and in the BWRVIP-05 report. The NRC staff has also determined that the licensee has acceptably demonstrated that the conditional probability of failure values for the Hatch, Units 1 and 2 RV circumferential welds are sufficiently low to justify the elimination of the augmented volumetric examinations required by 10 CFR 50.55a(g)(6)(ii)(A)(2), and the volumetric examinations required by the ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11.

Based on this analysis, the NRC staff concludes that the licensees proposed alternative will provide an acceptable level of quality and safety in lieu of performing the required volumetric examinations. Therefore, the licensee's alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

SNCs letter dated March 29, 2004, requested that RR-38 be approved for the remainder of the 40-year initial license. During review of the RR, the NRC staff and SNC agreed that the RR should be amended to include the period of extended operation. Upon further review, the NRC staff determined that the neutron fluence values used in the RR were based on a calculational code (RAMA) that had not been review and approved by the NRC staff. The licensee had initially anticipated that the NRC staffs review and acceptance of the RAMA code for calculating reactor pressure vessel neutron fluence would be complete by the time this RR was needed. The NRC staff is currently reviewing, but has not yet approved the use of the RAMA code for neutron fluence calculations. As a result, the NRC staff reviewed the licensees current, NRC-approved neutron fluence analyses and concluded that the fluence values calculated using this methodology are conservative through July 31, 2007. Therefore, the NRC staff limits the duration of RR-38 until July 31, 2007.

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

Principal Contributor: J. Medoff, DE L. Lois, DSSA Date: January 28, 2005