ML020720594

Issuance of Amendments 230 & 171 to Licenses DPR-57 & DPR-5, Respectively, Revising Surveillance Requirement 3.6.1.3.8

ML020720594
Person / Time
Site:	Hatch
Issue date:	04/11/2002
From:	Olshan L NRC/NRR/DLPM/LPD2
To:	Sumner H Southern Nuclear Operating Co
Olshan L N, NRR/DLPM, 415-1419
References
TAC MB2976, TAC MB2977
Download: ML020720594 (16)

Text

April 11, 2002 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 RE: ISSUANCE OF AMENDMENTS (TAC NOS. MB2976 AND MB2977)

Dear Mr. Sumner:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 230 to Facility Operating License DPR-57 and Amendment No. 171 to Facility Operating License NPF-5 for the Edwin I. Hatch Nuclear Plant, Units 1 and 2. The amendments consist of changes to the Technical Specifications in response to your application dated September 19, 2001, as supplemented by letter dated March 11, 2002.

The amendments revise surveillance requirement (SR) 3.6.1.3.8 by relaxing the 18-month reactor instrumentation excess flow check valve (EFCV) surveillance frequency. The revised SR states that a representative sample of the EFCVs will be tested every 18 months such that each EFCV will be tested at least once every 10 years.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Leonard N. Olshan, Senior Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-321 and 50-366

Enclosures:

1. Amendment No. 230 to DPR-57

2. Amendment No. 171 to NPF-5

3. Safety Evaluation cc w/encls: See next page

ML020720594 TS: *No major changes to SE.

OFFICE PDII-1/PM PDII-1/LA SPSB/BC* OGC** PDII-1/SC NAME LOlshan CHawes RBarrett RHoefling JNakoski DATE 04 /09/02 04 /10 /02 03/18/02 04/04/02 04/10/02 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 230 License No. DPR-57

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 1 (the facility)

Facility Operating License No. DPR-57 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 19, 2001, as supplemented by letter dated March 11, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 230, are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

John A. Nakoski, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance: April 11, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 230 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Appendix A Technical Specifications and associated Bases with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3.6-14 3.6-14 B 3.6-27 B 3.6-27 B 3.6-28 B 3.6-28 B 3.6-28a B 3.6-28a B 3.6-28b B 3.6-28b

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 171 License No. NPF-5

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility)

Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 19, 2001, as supplemented by letter dated March 11, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 171, are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

John A. Nakoski, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance: April 11, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 171 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the Appendix A Technical Specifications and associated Bases with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3.6-14 3.6-14 B 3.6-27 B 3.6-27 B 3.6-28 B 3.6-28 B 3.6-29 B 3.6-29

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 230 TO FACILITY OPERATING LICENSE DPR-57 AND AMENDMENT NO. 171 TO FACILITY OPERATING LICENSE NPF-5 SOUTHERN NUCLEAR OPERATING COMPANY, INC., ET AL.

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

1.0 INTRODUCTION

By letter dated September 19, 2001, as supplemented by letter dated March 11, 2002, Southern Nuclear Operating Company, Inc. (Southern Nuclear, the licensee), et al., proposed license amendments to change the Technical Specifications (TS) for the Edwin I. Hatch Nuclear Plant, Units 1 and 2. The proposed changes would revise the surveillance test requirements for the excess flow check valves (EFCVs). The supplemental letter dated March 11, 2002, provided clarifying information that did not change the scope of the September 19, 2001, application nor the initial proposed no significant hazards consideration determination.

2.0 BACKGROUND

EFCVs are installed in boiling water reactor (BWR) instrument lines penetrating the primary containment boundary to limit the release of fluid in the event of an instrument line break.

Regulatory Guide (RG) 1.11, Instrument Lines Penetrating Primary Reactor Containment, provides guidance on the implementation of General Design Criteria (GDC) 55 and 56 for instrumentation lines that penetrate primary reactor containment and are part of the reactor coolant pressure boundary. As stated in RG 1.11, EFCVs in combination with flow restricting features (line size or orifice) satisfy the requirements of GDC 55 and 56 for automatic isolation capability, maintain the reliability of the connected instrumentation, and ensure the functional performance of secondary containment in the event of an instrumentation line rupture.

Examples of EFCV installations include reactor pressure vessel level and pressure instrumentation, main steam line flow instrumentation, recirculation pump suction pressure, and reactor core isolation cooling steam line flow instrumentation. EFCVs are not required to close in response to a containment isolation signal and are not required to operate under post loss-of-coolant accident conditions.

Hatch TS Surveillance Requirement (SR) 3.6.1.3.8 currently requires verification of the actuation capability of each reactor instrumentation line EFCV every 18 months. The SR demonstrates that each reactor instrumentation line EFCV is operable by verifying that the valve actuates to restrict flow to within limits. The proposed change would revise TS SR 3.6.1.3.8 to relax the 18-month EFCV surveillance frequency by limiting the number of tests to a

representative sample every 18 months such that each EFCV will be tested at least once every 10 years. The representative sample consists of approximately equal number of EFCVs being tested every 18 months such that each EFCV is tested at least once every 10 years.

The basis for the request is the high degree of reliability shown by the EFCVs and the low consequences of an EFCV failure, as presented in General Electric Nuclear Energy (GENE)

Topical Report B21-00658-01, Excess Flow Check Valve Testing Relaxation, dated November 1998. The staff safety evaluation report (SER) dated March 14, 2000, approved this topical report. The supporting analysis for the licensees conclusion is based on Topical Report NEDO-32977-A, Excess Flow Check Valve Testing Relaxation, dated June 2000 which incorporated the staffs March 4, 2000, SER. The Topical Report NEDO-32977-A provided:

(1) an estimate of steam release frequency into the reactor building due to a break in an instrument line concurrent with an EFCV failure to close, and (2) an assessment of the radiological consequences of such a release. The Boiling Water Reactor Owners Group (BWROG) concluded that EFCV testing intervals could be extended up to 10 years based on the topical report reliability and consequence analysis without significantly affecting plant risk.

The BWROG suggested a staggered test interval based on actual valve performance with each valve being tested at least once every 10 years. The staff accepted the generic applicability of the topical report by SER dated March 14, 2000, and agreed that the EFCV test interval could be extended to as much as 10 years. The staff also noted that licensees adopting the topical report must have a failure feedback mechanism and corrective action program to ensure that EFCV performance continues to be bounded by the topical report results. Additionally, each licensee is required to perform a plant-specific radiological dose assessment, and EFCV failure rate and release frequency analysis, to confirm that its facility is bounded by the generic analysis of the topical report.

The proposed change adopts the staffs approved Technical Specification Task Force (TSTF)

Traveler TSTF-334, Revision 2, Relaxed Surveillance Frequency for Excess Flow Check Valves Testing. TSTF-334 was approved by the staff on October 31, 2000, by letter from W. D. Beckner to A. R. Pietrangelo, Nuclear Energy Institute. It proposed specific changes to the Standard Technical Specifications (STS) providing guidance for licensees implementing the extended EFCV surveillance test intervals proposed in the topical report. TSTF-334 is applicable only for those plants for which NEDO-32977-A is applicable and is subject to EFCV performance and corrective action criteria to be developed by the licensee.

3.0 EVALUATION The staff reviewed the licensees submittal for conformance to the March 14, 2000, staff SER for Topical Report NEDO-32977-A and the guidance of approved TSTF-334, Revision 2. The staffs evaluation concerned itself with the following areas: (1) EFCV failure rate and release frequency, (2) the licensees failure feedback mechanism and corrective action program, (3) radiological dose assessment, and (4) conformance of the revised TS to generic TS guidance.

3.1 EFCV Failure rate and release frequency In the topical report, EFCV reliability was evaluated based on testing experience provided by 12 different BWR plants. The composite data indicated that EFCVs are very reliable. The data represented 12,424.5 valve years of operation with a total of 11 failures noted. The EFCV

composite failure rate was 1.67E-07/ hour and was referenced as the upper limit failure rate in the topical report.

The staff noted in its review of the report that the BWROG assumed the EFCV failure rate was constant over time and did not account for potential age-related degradation in the EFCV failure rate. Additionally, the staff questioned the use of an instrument line break frequency based on WASH-1400 and not on more current data. To address this concern, the BWROG Request for Additional Information (RAI) response included an updated instrument line failure frequency of 3.52E-05 failures/ year based on the Electric Power Research Institutes Technical Report No. 100380, Pipe Failures in U.S. Commercial Nuclear Power Plants, dated July 1992. This value is 6.6 times greater than the value calculated in the topical report using WASH-1400 data.

The BWROG RAI response also assumed the observed EFCV failures were five times the actual observed number (55 vs. 11) listed in the topical report. The additional impact of an increase in instrument line failure frequency and a fivefold increase in EFCV failures assumed by the BWROG RAI response demonstrated that release frequencies remained low, with limited impact on release frequency.

To estimate the release frequency initiated by an instrument line break, two factors are considered: (1) the instrument line break frequency downstream of the EFCV, and (2) the probability of the EFCV failing to close. Although Hatch was not an original member of the owners group EFCV committee, the Hatch data were found to be consistent in both the time sampled and EFCV reliability when compared to the topical report data. The Hatch plant-specific EFCV failure and release rates are comparable to industry data and consistent with the staff topical report SER conclusions. Based on the above, the staff does not consider the estimated increase in release frequency for Hatch to be significant.

3.2 Licensees Failure Feedback Mechanism and Corrective Action Program The staff noted that the topical report does not provide a specific failure feedback mechanism, but does state that a plants corrective action program must evaluate equipment failures and establish appropriate corrective actions. The BWROG responded to the staff RAI question concerning failure feedback by stating that each licensee that adopts the relaxed surveillance intervals recommended by the topical report should ensure that an appropriate feedback mechanism responsive to EFCV failure trends is in place.

The licensee stated that EFCV surveillance failures will be documented in the Hatch corrective action program as surveillance test failures. The EFCV failure will be evaluated and corrected.

Valves that are repaired and not replaced will be included in the next surveillance. To ensure EFCV performance remains consistent with the extended test interval, minimum performance criteria have been established by the licensee. The criteria for reactor instrument line EFCVs have been established for Hatch as less than or equal to two maintenance-preventable functional test failures per fuel cycle (18 or 24 months) to ensure that the EFCV performance remains consistent with the extended surveillance interval assumptions and adverse trends in EFCV performance are identified. Additional criteria of less than or equal to two consecutive test failures will also be implemented. Based on its review, the staff finds that the licensees program to account for potential changes in EFCV failure rates satisfies the TSTF-334 performance and corrective action criteria and is therefore acceptable.

3.3 Radiological Dose Assessment

The radiological consequences for an instrument line break have been previously evaluated by the licensee in the Hatch Updated Final Safety Analysis Report (UFSAR) Section 15.4.13. The analysis does not credit the EFCVs for isolating the break, but does assume the discharge of reactor water is through an instrument line with a 1/4 inch flow-restricting orifice for the duration of the event.

The topical report stated that the magnitude of release through an instrument line would be within the pressure control capacity of reactor building ventilation systems and that the integrity and functional performance of secondary containment and the standby gas treatment system, following an instrument line break, would continue to be met. The licensee confirmed that if an EFCV should fail, the restricting orifice or line restriction limits the steam release and the integrity and functional performance of secondary containment will be maintained. The Hatch UFSAR Section 15.4.13 notes that operator action would be required for plant shutdown and depressurization to terminate the event.

The resulting offsite exposures are a small fraction of the 10 CFR Part 100 limits. As a result, a failure of an EFCV to close is bounded by the licensees previous analysis. The radiation dose consequences for an instrument line break are therefore not impacted by the proposed change.

Based on the above, the staff agrees with the licensees determination that the current licensing basis remains applicable for the proposed EFCV surveillance interval with regard to the potential radiological consequences of an instrument line break with failure of the EFCV to isolate.

3.4 Conformance of the proposed TS to generic TSTF guidance The Hatch TS SR 3.6.1.3.8 currently requires verification that each reactor instrumentation line EFCV be demonstrated OPERABLE at least once every 18 months by verifying the valve actuates to restrict flow within limits. The current sentence in TS SR 3.6.1.3.8 will be revised to read, Verify each reactor instrumentation line EFCV (of a representative sample) actuates to restrict flow to within limits.

The term representative sample, as proposed by the topical report and TSTF-334 is not defined in the TS itself. However, the BWROG, in response to the staff RAI, stated that the term representative sample, with an accompanying explanation in the TS Bases, is identical to the current usage in the STS, NUREG-1433, Revision 1. Specifically, NUREG-1433 uses the term representative in TS SR 3.8.6.3 in reference to battery cell testing, and representative sample in SR 3.1.4.2 for verification of control rod scram times. The criterion for representative sample and the basis for the nominal 10-year testing interval are provided in the licensees submittal, and are similar to Insert 1 and Insert 2 in the staffs approved TSTF-334, Revision 2. The application of a representative sample for the EFCV testing SR, with an accompanying explanation in the TS Bases, is consistent with TSTF-334, Revision 2, and is therefore, acceptable to the staff.

The staff reviewed the revised TS wording in SR 3.6.1.3.8 and finds the proposed revision to be consistent with TSTF-334 and TS generic guidance. The licensee also included in its submittal a revised Bases for SR 3.6.1.3.8 that includes a discussion of the EFCV test frequency and the term representative sample.

3.5 Staff Conclusion As demonstrated in GENE Topical Report NEDO-32977-A, the impact of an increase in the EFCV surveillance test interval to 10 years results in an instrument line release frequency considered by the staff to be sufficiently low, especially since the consequences of an EFCV failure are bounded by previous licensee analysis; therefore, it is highly unlikely that this will lead to core damage. Additionally, the licensees evaluation results (including the plant-specific EFCV failure data and release frequency) are consistent with the topical report results. The staff concludes that the release frequency associated with the Hatch request for relaxation of ECFV surveillance testing is sufficiently low and therefore acceptable.

The consequences of steam release from the failure of the EFCVs are not significant, as shown by the previous licensee analysis. Based on the acceptability of the methods applied to estimate the release frequency, the licensees relatively low release frequency estimate, the negligible consequence of a release in the reactor building, in conjunction with a highly unlikely impact on core damage, the staff concludes that the impact on risk associated with the Hatch request for relaxation of ECFV surveillance testing is sufficiently low and is acceptable.

The topical report established that each plants corrective action program must evaluate equipment failures and establish appropriate corrective actions. These programs ensure that meaningful feedback data are acquired so that appropriate corrective action may be taken with regard to EFCV performance. The licensee provided information to the staff regarding EFCV performance criteria and the EFCV corrective action program. The staff finds the licensees program to be in conformance with TSTF-334, Revision 2, and the topical report; thus, it is acceptable to the staff.

Based on the above, the staff finds the proposed change to relax the Hatch instrument line EFCV surveillance frequency by allowing a representative sample of EFCVs to be tested every 18 months, with all EFCVs being tested at least once every 10 years, to be consistent with TSTF-334 generic guidance, Topical Report NEDO-32977-A, and the staffs March 14, 2000, SER and is therefore acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (66 FR 57125). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR

51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: C. Doutt N. Le Date: April 11, 2002

Edwin I. Hatch Nuclear Plant cc:

Mr. Ernest L. Blake, Jr. Charles A. Patrizia, Esquire Shaw, Pittman, Potts Paul, Hastings, Janofsky & Walker and Trowbridge 10th Floor 2300 N Street, NW. 1299 Pennsylvania Avenue Washington, DC 20037 Washington, DC 20004-9500 Mr. D. M. Crowe Chairman Manager, Licensing Appling County Commissioners Southern Nuclear Operating County Courthouse Company, Inc. Baxley, Georgia 31513 P. O. Box 1295 Birmingham, Alabama 35201-1295 Mr. J. D. Woodard Executive Vice President Resident Inspector Southern Nuclear Operating Plant Hatch Company, Inc.

11030 Hatch Parkway N. P. O. Box 1295 Baxley, Georgia 31531 Birmingham, Alabama 35201-1295 Mr. Charles H. Badger Mr. P. W. Wells Office of Planning and Budget General Manager, Edwin I. Hatch Room 610 Nuclear Plant 270 Washington Street, SW. Southern Nuclear Operating Atlanta, Georgia 30334 Company, Inc.

U.S. Highway 1 North Harold Reheis, Director P. O. Box 2010 Department of Natural Resources Baxley, Georgia 31515 205 Butler Street, SE., Suite 1252 Atlanta, Georgia 30334 Mr. L. M. Bergen Resident Manager Steven M. Jackson Oglethorpe Power Corporation Senior Engineer - Power Supply Edwin I. Hatch Nuclear Plant Municipal Electric Authority P. O. Box 2010 of Georgia Baxley, Georgia 31515 1470 Riveredge Parkway, NW Atlanta, Georgia 30328-4684