ML050340624

License Amendments, Modification of TS 5.5.12 for Extension of Type a Leakage Rate Test (Tac No. MC3745, MC3746)

ML050340624
Person / Time
Site:	Browns Ferry
Issue date:	03/09/2005
From:	Ellen Brown NRC/NRR/DLPM/LPD2
To:	Singer K Tennessee Valley Authority
Brown E, NRR/DLPM, 415-2315
Shared Package
ML050340627	List: ML050340624 ML050740291
References
TAC MC3745, TAC MC3746
Download: ML050340624 (23)

Text

March 9, 2005 Mr. Karl W. Singer Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 ISSUANCE OF AMENDMENTS REGARDING ONE-TIME FREQUENCY EXTENSION FOR CONTAINMENT INTEGRATED LEAKAGE RATE TEST INTERVAL (TAC NOS. MC3745 AND MC3746)

Dear Mr. Singer:

The Commission has issued the enclosed Amendment Nos. 293 and 252 to Facility Operating Licenses Nos. DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Units 2 and 3, respectively. These amendments are in response to your application dated July 8, 2004, as supplemented by your letter of November 24, 2004, which provided additional information.

These amendments modify Technical Specification Section 5.5.12, Primary Containment Leakage Rate Testing Program, to allow a one-time 5-year extension to the 10-year frequency of the performance-based leakage rate testing program for Type A tests. The local leakage rate tests (Type B and Type C), including their schedules, are not affected by this approval.

The changes were submitted on a risk-informed basis as described in Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.

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

Sincerely,

/RA/

Eva A. Brown, Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-260 and 50-296

Enclosures:

1. Amendment No. 293 to License No. DPR-52

2. Amendment No. 252 to License No. DPR-68

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

ML050340624

• No Legal Objection NRR-043 OFFICE PDII-2/PM PDII-2/LA SPSB/SC EMEB/SC OGC*

PDII-2/SC NAME EBrown BClayton RDennig by memo dated KManoly by memo dated SCole NLO with comments MMarshall DATE 3/2/05 3/2/05 1/6/05 1/21/05 2/23/05 3/9/05

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 ISSUANCE OF AMENDMENTS REGARDING ONE-TIME FREQUENCY EXTENSION FOR CONTAINMENT INTEGRATED LEAKAGE RATE TEST INTERVAL (TAC NOS. MC3745 AND MC3746)

Dated: March 9, 2005 DISTRIBUTION:

PUBLIC PDII-2 R/F RidsOgcRp RidsAcrsAcnwMailCenter RidsNrrDlpmLpd2 (EHackett)

RidsNrrDlpmLpdii-2 (MMarshall)

RidsRgn2MailCenter (SCahill)

BClayton (Hard Copy)

RidsNrrDlpmDpr RidsNrrPMEBrown RDennig KManoly TBoyce DJeng RPalla RGoel G. Hill (4 Hard Copies)

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 293 License No. DPR-52

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated July 8, 2004, as supplemented by letter dated November 24, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations 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 Commission's regulations; 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by 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-52 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 293, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Michael L. Marshall, Jr., Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 9, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 293 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 5.0-20 5.0-20 5.0-21 5.0-21

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 252 License No. DPR-68

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated July 8, 2004, as supplemented by letter dated November 24, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations 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 Commission's regulations; 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by 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-68 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 252, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Michael L. Marshall, Jr., Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 9, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 252 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 5.0-20 5.0-20 5.0-21 5.0-21

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 293 TO FACILITY OPERATING LICENSE NO. DPR-52 AND AMENDMENT NO. 252 TO FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 DOCKET NOS. 50-260 AND 50-296

1.0 INTRODUCTION

By letter dated July 8, 2004, as supplemented November 24, 2004, the Tennessee Valley Authority (TVA, the licensee) submitted a request for changes to the Browns Ferry Nuclear Plant (BFN), Units 2 and 3 Technical Specifications (TS). The requested changes modify TS Section 5.5.12, Primary Containment Leakage Rate Testing Program, to add an exception from the guidelines of Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, and Nuclear Energy Institute (NEI) report NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 26, 1995, regarding the Type A test interval.

Specifically, the exception states that the first Unit 2 Type A test performed after the November 6, 1994, Type A test shall be performed no later than November 6, 2009, and the first Unit 3 Type A test performed after the October 10, 1998, shall be performed no later than October 10, 2013. The local leakage rate tests (Type B and Type C), including their schedules, are not affected by this request. The changes were submitted on a risk-informed basis as described in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. The November 24, 2004, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

2.0 REGULATORY EVALUATION

Appendix J, Option B to Title 10 of the Code of Federal Regulations (CFR) Part 50, requires that a Type A test be conducted at periodic intervals based on historical performance of the overall containment system. RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, describes an acceptable approach for assessing the nature and impact of proposed licensing basis changes by considering engineering issues and applying risk insights. Assessments should consider relevant safety margins and defense-in-depth attributes, including consideration of success criteria as well as equipment functionality, reliability, and availability. The analyses should reflect the actual design, construction, and operational practices of the plant. Acceptance guidelines for evaluating the results of such assessments are provided. This guide also addresses implementation strategies and performance monitoring plans associated with licensing basis changes that will help ensure that assumptions and analyses supporting the change are verified. Consistency with the defense-in-depth philosophy is maintained if a reasonable balance is preserved between prevention of core damage, prevention of containment failure, and consequence mitigation.

TS 5.5.12 requires that leakage rate testing be performed as required by 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines contained in RG 1.163 with certain exceptions specified in the TS. This RG endorses, with certain exceptions NEI 94-01, Revision 0. The exceptions are as follows:

1.

NEI 94-01 references ANSI/ANS-56.8-1994, "Containment System Leakage Testing Requirements," for detailed descriptions of the technical methods and techniques for performing Types A, B, and C tests under the amendment of Appendix J to 10 CFR Part

50. However, as stated in NEI 94-01, the test intervals in ANSI/ANS-56.8-1994 are not performance-based. Therefore, licensees intending to comply with Option B in the amendment to Appendix J should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01, rather than using the test intervals specified in ANSI/ANS-56.8-1994. All other technical methods and techniques for performing Types A, B, and C tests contained in ANSI/ANS-56.8-1994 are acceptable to the U.S. Nuclear Regulatory Commission (NRC) staff.

2.

Section 11.3.2, "Programmatic Controls," of NEI 94-01 provides guidance for licensee selection of an extended interval greater than 60 months or three refueling cycles for a Type B or Type C tested component. Because of uncertainties (particularly unquantified leakage rates for test failures, repetitive/common mode failures, and aging effects) in historical Type C component performance data, and because of the indeterminate time period of three refueling cycles and insufficient precision of programmatic controls described in Section 11.3.2 to address these uncertainties, the guidance provided in Section 11.3.2 for selecting extended test intervals greater than 60 months for Type C tested components is not presently endorsed by the NRC staff. Further, the interval for Type C tests for main steam and feedwater isolation valves in boiling water reactors (BWRs), and containment purge and vent valves in pressurized water reactors [PWRs]

and BWRs, should be limited to 30 months as specified in Section 3.3.4 of ANSI/ANS-56.8-1994, with consideration given to operating experience and safety significance.

3.

Section 9.2.1, "Pretest Inspection and Test Methodology," of NEI 94-01 provides guidance for the visual examination of accessible interior and exterior surfaces of the containment system for structural problems. These examinations should be conducted prior to initiating a Type A test, and during two other refueling outages before the next Type A test if the interval for the Type A test has been extended to 10 years, in order to allow for early uncovering of evidence of structural deterioration.

4.

Section 10.2.3.3 states that an as-found Type C test or an alternative test or analysis is to be performed prior to any maintenance, repair, modification, or adjustment activity if it could affect a valve's leak-tightness. "Alternative test or analysis" is not endorsed as an appropriate substitute for an as-found test, because the as-found test provides clear and objective evidence of performance of isolation components.

A Type A test is an overall (integrated) leakage rate test of the containment structure.

NEI 94-01 specifies an initial test interval of 48 months, but allows an extended interval of 10 years, based upon two consecutive successful tests. There is also a provision for extending the test interval, an additional 15 months in certain circumstances. The two most recent Type A tests at both BFN units have been successful, so the current interval requirement is 10 years.

3.0 TECHNICAL EVALUATION

3.1 Risk Impact Assessment The licensee has performed a risk impact assessment of extending the Type A test interval to 15 years. The risk assessment was provided in the July 8, 2004, application for license amendment. In performing the risk assessment, the licensee considered the guidelines of NEI 94-01, the methodology used in Electric Power Research Institute (EPRI) TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing, and RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and was established in 1995 during the development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak-Test Program, provided the technical basis to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement this basis, industry undertook a similar study. The results of that study are documented in EPRI Research Project Report TR-104285.

The EPRI study used an analytical approach similar to that presented in NUREG-1493 for evaluating the incremental risk associated with increasing the interval for Type A tests. The Appendix J, Option A, requirements that were in effect for BFN Units 2 and 3 early in the plants life, required a Type A test frequency of three tests in 10 years. The EPRI study estimated that relaxing the test frequency from three tests in 10 years to one test in 10 years would increase the average time that a leak that was detectable only by a Type A test goes undetected from 18 to 60 months. Since Type A tests only detect about 3 percent of leaks (the rest are identified during local leak rate tests based on industry leakage rate data gathered from 1987 to 1993), this results in a 10-percent increase in the overall probability of leakage. The risk contribution of pre-existing leakage for the PWR and BWR representative plants in the EPRI study confirmed the NUREG-1493 conclusion that a reduction in the frequency of Type A tests from three tests in 10 years to one test in 20 years leads to an imperceptible increase in risk that is on the order of 0.2 percent and a fraction of one person-rem per year in increased public dose.

Building upon the methodology of the EPRI study, the licensee assessed the change in the predicted person-rem per year frequency. The licensee quantified the risk from sequences that have the potential to result in large releases if a pre-existing leak were present. Since the Option B rulemaking in 1995, the NRC staff has issued RG 1.174 on the use of probabilistic risk assessment (PRA) in evaluating risk-informed changes to a plants licensing basis. The licensee has proposed using RG 1.174 guidance to assess the acceptability of extending the Type A test interval beyond that established during the Option B rulemaking.

RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in core damage frequency (CDF) less than 10-6 per year and increases in large early release frequency (LERF) less than 10-7 per year. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. The licensee has estimated the change in LERF for the proposed change and the cumulative change from the original frequency of three tests in a 10-year interval. RG 1.174 also discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. The licensee estimated the change in the conditional containment failure probability for the proposed change to demonstrate that the defense-in-depth philosophy is met.

The licensee provided analyses, as discussed below. The following comparisons of risk are based on a change in test frequency from three tests in 10 years (the test frequency under Appendix J, Option A) to one test in 15 years. This bounds the impact of extending the test frequency from one test in 10 years (the current test frequency for BFN Units 2 and 3 under Appendix J, Option B) to one test in 15 years. The following conclusions can be drawn from the analysis associated with extending the Type A test frequency:

1. Given the change from a three in 10-year test frequency to a one in 15-year test frequency, the increase in the total integrated plant risk is estimated to be less than 0.01 person-rem per year. This increase is comparable to that estimated in NUREG-1493, where it was concluded that a reduction in the frequency of tests from three in 10 years to one in 20 years leads to an imperceptible increase in risk.

Therefore, the increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

2. The increase in LERF resulting from a change in the Type A test frequency from the original three in 10 years to one in 15 years is estimated to be 3.4 x 10-8 per year based on the internal events PRA, and 6.2 x 10-8 per year including both internal and external events. However, there is some likelihood that the flaws in the containment estimated as part of the Class 3b frequency would be detected as part of the IWE/IWL visual examination of the containment surfaces (as identified in American Society of Mechanical Engineers [ASME] Boiler and Pressure Vessel Code,Section XI, Subsections IWE/IWL). Visual inspections are expected to be effective in detecting large flaws in the visible regions of containment, and this would reduce the impact of the extended test interval on LERF. The licensees risk analysis considered the potential impact of age-related corrosion/degradation in inaccessible areas of the containment shell on the proposed change. The increase in LERF associated with corrosion events is estimated to be less than 1 x 10-8 per year. The NRC staff concludes that increasing the Type A interval to 15 years results in a very small change in LERF and is consistent with the acceptance guidelines of RG 1.174.

3. RG 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy.

Consistency with the defense-in-depth philosophy is maintained if a reasonable balance is preserved between prevention of core damage, prevention of containment failure, and consequence mitigation. The licensee estimates the change in the conditional containment failure probability to be an increase of approximately 1 percentage point for the cumulative change of going from a test frequency of three in 10 years to one in 15 years. The NRC staff finds that the defense-in-depth philosophy is maintained based on the small magnitude of the change in the conditional containment failure probability for the proposed amendment.

Based on these conclusions, the NRC staff finds that the increase in predicted risk due to the proposed change is within the acceptance guidelines while maintaining the defense-in-depth philosophy of RG 1.174 and, therefore, is acceptable.

3.2 Containment Leakage Testing Both BFN Units 2 and 3 are General Electric BWR/4 plants with Mark I primary containments.

The Mark I primary containment consists of a drywell, which encloses the reactor vessel, reactor coolant recirculation system and branch lines of the reactor coolant system, a toroidal-shaped pressure suppression chamber containing a large volume of water, and a vent system connecting the drywell to the water space of the suppression chamber. The primary containment is penetrated by personnel access hatches, piping, and electrical penetrations.

As stated in the licensees submittal, TVA performed the last two consecutive integrated leakage rate tests (ILRTs) on March 1991 and November 1994 for Unit 2, and on November 1995 and October 1998 for Unit 3. TVA stated the results of previous Type A tests demonstrated that the BFN Units 2 and 3 containment structures remain essentially leak-tight barriers and represent minimal risk to increased leakage.

Because the leak rate testing requirements (ILRT and local leak rate tests) of Option B of 10 CFR Part 50, Appendix J, and the containment inservice inspection (ISI) requirements mandated by 10 CFR 50.55a complement each other in ensuring the leak-tightness and structural integrity of the containment, the NRC staff, from its review of Type A test interval extension applications submitted by other licensees, has identified a number of general areas for which the licensee was requested to provide additional information in relation to the ISI of the TVA containments.

Type A Test Results With respect to the results of the previous Type A tests for BFN Units 2 and 3 presented in a table shown in page E1-6 of the submittal, the test data appear to indicate that the November 6, 1995, test for Unit 3 registered a leak rate almost three times the leak rates measured for the other three tests for both Units 2 and 3 reported in the table. TVA was requested to discuss plausible reasons for registering such large fluctuations in the above mentioned BFN Type A test results.

TVA responded that the table on page E1-6 of the submittal provides the Containment Integrated Leak Rate Test (CILRT) results for the two most recent CILRT () performances on Units 2 and 3. TVA confirmed the NRC staffs observation that the table shows the November 1995 test for Unit 3 revealed a leak rate which was approximately three times the leak rate measured for the other three tests for Units 2 and 3. The difference was primarily due to leakage identified at several instrument tubing compression fittings and packing leaks within the CILRT boundary. TVA further indicated that the relative magnitude of these leaks was small and did not preclude successful performance of the CILRT, so the leak locations were recorded and repairs were deferred until after the completion of the CILRT and verification tests. Repairs were subsequently performed during the containment depressurization period. Since no convenient provisions for quantifying these local leaks existed, no adjustments were made to the overall CILRT results to reflect improvements realized by these repairs. TVA noted that the subsequent Unit 3 CILRT, which was performed in 1998, was comparable with typical BFN test results.

The NRC staff finds that the above TVA justification is adequate and acceptable because the abnormal result of the November 1995 test was correctly attributed to leakage identified at several instrument tubing compression fittings and packing leaks within the CILRT boundary, and also because the October 1998 test results performed after the leakages were subsequently repaired, did show comparable results.

Visual Inspections Referring to the last paragraph of page E1-6 of the submittal, Appendix J Visual Inspections, the NRC staff requested the licensee to discuss how the proposed one-time extension of the integrated leak rate test interval from 10 years to 15 years would impact the visual inspection plans originally scheduled for BFN Units 2 and 3 containments. Specifically, TVA was asked to provide a description of its inservice inspection methods/plans for the additional 5-year extended period that would provide assurance that in the absence of an ILRT for 15 years, the containment structural and leak tight integrity will be maintained.

TVA indicated that 10 CFR Part 50 Appendix J program requires visual inspections to be performed of accessible interior and exterior surfaces of the containment system for structural problems that may affect either the containment structural leakage integrity or performance of the test. These examinations are conducted prior to initiating a test and during two other refueling outages before the next test to allow for early detection of evidence of structural deterioration. TVA noted that the examination of the primary containment structure is conducted in accordance with the schedule guidelines of RG 1.163. Additionally, the plant instruction, which implements this examination, reflects a required frequency in accordance with ASME Section XI, Subsection IWE, and 10 CFR 50.55a(b)(2)(ix)(E), such that one examination is scheduled during each inspection period. Acceptance criteria are in accordance with 10 CFR Part 50 Appendix J and ASME Section XI, IWE-3000. TVA asserted that these requirements will not be changed as a result of the proposed extended ILRT interval.

For Unit 2, a VT-3 examination will be conducted in accordance with Subsection IWE during the Unit 2 Cycle 14 (2007) refueling outage, which is in the third period of the first IWE inspection interval. The Unit 2 Cycle 15 (2009) refueling outage will be in the first period of the second IWE inspection interval. TVA indicated that during this refueling outage, a General Visual examination will be performed as required prior to conducting the CILRT. For Unit 3, a VT-3 examination will be conducted in accordance with Subsection IWE during the Unit 3 Cycle 12 (2006) and/or Cycle 13 (2008) refueling outages, which are in the third period of the first IWE inspection interval. The general visual examination will be performed as required during the Unit 3 Cycle 14 (2010) refueling outage, which is in the first period of the second IWE interval.

During the Unit 3 Cycle 15 (2012) refueling outage, which is in the second period of the second IWE interval, a General Visual examination will also be performed as required prior to conducting the CILRT.

In addition to Containment ISI (CISI) examination described above, accessible portions of the drywell and torus surface are inspected during each refueling outage as required by Surveillance Instruction 0-SI-4.7.A.2.K, Primary Containment Drywell Surface Visual Inspection and by Technical Instruction 0-TI-417, Inspection of Service Level I, II, III Protective Coatings. These complement the periodic visual inspections of the interior and exterior of the containment structure and serve to provide added assurance of structural integrity between CISI examinations. TVA asserted that the inspections and examinations described above, provide an additional degree of assurance of continued containment structural integrity in support of requested CILRT extension.

The NRC staff finds that the above description of BFN plant-specific ISI methods/plans for the additional 5-year extended period acceptable, because they are consistent with applicable IWE requirements. Timely implementation of the ISI plans should provide a reasonable assurance that in the absence of an ILRT for 15 years, the containment structural and leak-tight integrity will be maintained.

Category E-A Examinations Regarding the Category E-A Examinations discussed on page E1-8 of Reference 5.1, TVA stated that the general visual examinations identified some mechanical damages such as pitting, gouges, dents, rust and arc strikes, which were evaluated by TVA and found acceptable. The NRC staff requested TVA to discuss the extent of the loss of the liner material that resulted from the observed mechanical damages and the technical basis for its determination that the incurred damages were acceptable.

TVA stated that two general visual examinations have been conducted on each unit. There was no significant metal loss in areas where dents, rust, and arc strikes were noted. These conditions were not considered evidence of degradation that could affect containment structural integrity or leak tightness described in IWE-3500.1, and were addressed in the BFNs 10 CFR Part 50 Appendix B corrective action program. TVA further stated that there were two conditions noted in general visual examinations that had the potential to exceed the acceptance standards of IWE-3500. These involved pitting and gouges. TVA indicated that areas of pitting have been identified in the liner in the moisture seal barrier region on both units. This pitting was found in the liner immediately above and below the seal where the seal was removed for replacement, and can be characterized as localized and sporadic pitting with pit depths ranging from less than 1/64 inch to 3/32 inch. Ultrasonic thickness measurements (UT) were obtained and used to evaluate the condition of the liner in these areas. This minor surface pitting was determined by TVA to not affect the structural integrity or leak tightness of the containment vessel. The thickness of the liner in this area is greater than the minimum required thickness of 1 inch as documented in the plant design basis calculations. A Notice of Indication was initiated during the performance of the general visual examination during the Unit 3 Cycle 10 (2002) refueling outage when several gouges in the Unit 3 drywell liner were identified. Six gouges of varying depths were found in the Unit 3 liner at the reactor building elevation 590 ft. TVA determined that these gouges were mechanical damage caused by a failed ventilation duct.

The gouges were blended-out and the base metal re-coated. The remaining liner thicknesses in all but one location were greater than 90 percent of the nominal liner thickness. The one location that contained the deepest gouge resulted in a minimum wall thickness of 0.723 inch after the discontinuity was removed by buffing. The resulting shell thickness of 0.723 inch is slightly below 90 percent of the nominal thickness, which is 0.7425 inch. TVA Engineering reviewed the results of the VT-3 and UT examinations performed after the discontinuity was removed and concluded the underthickness was acceptable based on the allowances in ASME Section III as documented in plant calculations.

The NRC staff finds the above TVA response adequate and acceptable because the localized and sporadic surface pitting were evaluated via use of the ultrasonic testing results and determined to not affect the structural integrity or leak tightness of the containment vessel, and the several gouges identified in the Unit 3 drywell liner were blended-out and evaluated by TVA Engineering as acceptable based on the allowances per ASME Section III Code.

Category E-C Examinations With respect to the Category E-C Examinations discussed in the first paragraph of page E1-9 of the licensees submittal, TVA indicated that VT-1 visual examinations of the interior surfaces of Units 2 and 3's torus waterline region identified minor localized corrosion and pitting and these degradations were evaluated to be acceptable. TVA was requested to discuss the extent of the loss of material that were associated with the localized corrosion and pitting observed and the technical basis for the TVA determination that the observed containment shell degradation was acceptable.

TVA responded that the extent of material loss associated with the localized corrosion and pitting observed is not significant. The most recent VT-1 examination data obtained during Unit 3 Cycle 11 (2004) outage showed small indications characterized as scratches sporadically throughout the waterline region both above and below the waterline. These indications exhibited rust bleed through and were 1/4 inch to 1/2 inch in length with virtually no metal loss.

TVA asserted that these isolated areas of pinpoint rusting and minor localized corrosion did not represent any appreciable metal loss, and most metal loss in the waterline region due to pitting is less than 3 mils, with no areas of pitting or metal loss greater than 20 mils in depth. On Unit 2, indications were found ranging from discoloration to minor localized corrosion, and most metal loss in waterline region due to pitting is less than 4 mils, with no areas of pitting or metal loss greater than 20 mils in depth. The waterline region is coated, and there were no signs of flaking, peeling, blistering, cracking or other signs of distress indicative of structural degradation found on Units 2 or 3. TVA indicated that the conditions noted above were not considered to be defective conditions with respect to the acceptance criteria of IWE-3512.1.

The NRC staff finds the above TVA justification acceptable because TVA has shown that the results of the Category E-C examinations complied with the applicable IWE acceptance criteria.

Category E-D Examinations Referring to the Category E-D Examinations (Seals, Gaskets, and Moisture Barriers) discussed on page E1-9 of Reference 5.1, TVA was requested to discuss past BFN operating experience regarding the results of its VT-3 visual examinations of seals, gaskets and moisture barriers implemented once each inspection interval, and to describe the extent of defects found regarding BFNs moisture seal barrier and the defective portions replaced. TVA was also asked to discuss the potential negative impact of the proposed one-time ILRT interval extension upon TVAs continued ability to timely identify and dispose containment degradation and reasonably assure the leak tightness and structural integrity of the BFN containment.

TVA responded that prior to approval of Relief Request CISI-1, BFN conducted VT-3 examinations of seals and gaskets as required by IWE-3513.1 during the first IWE Interval.

These inspections identified various defects ranging from rust, pinches, and depressions of the o-rings; none of which resulted in a failure to meet containment leak rate requirements.

Following approval of Relief Request CISI-1, the only remaining Category E-D examination is the Moisture Barrier (MSB). The MSB is inspected once each operating cycle by either the CISI program or plant Surveillance Instruction 0-SI-4.7.A.2.K. Defects noted in the seal have ranged from wear, mechanical damage, punctures, age-related separation of the old seal, to separation of the new seal from the liner due to application problems. TVA indicated that when defects in the seal that requires repair are noted, the affected portion of the seal is removed and the liner inspected. Previous visual examinations of the liner in the area beneath the MSB have uncovered some random areas of surface pitting in the area immediately above the seal and continuing below the seal. The pitting can be characterized as localized with pit depths ranging from less than 1/64 inch to approximately 3/32 inch in depth. Improved MSB application techniques have been subsequently incorporated and no seal defects were noted during the Unit 3 Cycle 11(2004) refueling outage, which indicates these improved methods of preparing the surfaces for re-application of the seal have been effective. The improved performance of the MSB will prevent water from entering the concrete to steel interface and should preclude further pitting in these areas.

The routine scheduled inspections of the MSB each outage, by either the CISI program or Surveillance Instruction 0-SI-4.7.A.2.K is unaffected by the proposed one-time CILRT interval extension. These routine inspections will identify any seal degradation and evaluate the effect on the liner. These periodic inspections will continue to provide the ability to identify and evaluate containment degradation in a timely manner, and reasonably assure that the leak tightness and structural integrity of the BFN containment are maintained. Therefore, the proposed one-time CILRT interval extension will not adversely affect TVAs ability to promptly identify and address degradation in these areas.

The NRC staff finds the above TVA discussion covering its operating experience of Category E-D Examinations (Seals, Gaskets, and Moisture Barriers) acceptable because TVAs scheduled Category E-D Examination activities provide a basis for the NRC staff to conclude that its periodic inspections will reasonably assure the leak tightness and structural integrity of the BFN containment.

Generic Letter 87-05 Inspections With respect to the Generic Letter 87-05 Inspections discussed on page E1-10 of the licensees submittal, TVA stated that no significant corrosion had occurred in the sand region but there has been evidence of water leaking from the sand bed drains on both Units 2 and 3 since the 1987 inspections. TVA was requested to discuss the results of the ultrasonic thickness measurements at the sand bed region performed in September 1998 for Unit 3 and April 1999 for Unit 2, including the extent of liner thickness reductions observed and the basis for TVAs determination that the inspections verified the integrity of the liner. TVA was also asked to discuss the potential negative impact of the proposed one-time ILRT interval extension upon TVAs continued ability to timely identify and dispose containment degradation and reasonably assure the leak tightness and structural integrity of the BFN containment.

TVA stated that, UT examinations in 1998 on Unit 3 and in 1999 on Unit 2 indicated no significant corrosion had occurred in the sand bed region. The UT examinations conducted were from the floor to the horizontal weld between the first and second courses of drywell plates. The drywell floor is at elevation 549.92 ft and the sand bed region extends from elevation 548.79 ft to 550.29 ft; therefore, approximately 0.37 ft of the sand bed region extends above the floor. TVA indicated that minor localized surface pitting has been observed in this area and in the liner immediately below the MSB, however, this localized pitting does not affect the structural integrity or leak tightness of the containment vessel. The thickness of the liner in this area is greater than the minimum required thickness of 1 inch as documented in the design basis calculations. TVA further noted that when portions of the moisture seal barrier are removed for repair, the liner surfaces below the MSB are accessible for inspection. Damaged or defective portions of the MSB have been replaced to provide an appropriate level of protection from moisture intrusion in the area below the seal. There has been no evidence of propagation of iron oxide to the concrete surface noted that would be indicative of liner corrosion below the concrete and there is no significant thinning of the liner in this region. TVA asserted that routine inspections of the sand bed region are conducted by either the CISI program, the coatings program in procedure 0-TI-417, and Surveillance Instruction 0-SI-4.7.A.2.K, which will not be affected by the proposed CILRT request. Hence, these routine inspections will continue to identify any degradation and evaluate the effect on the liner. These inspections continue to provide the ability to timely identify and evaluate containment degradation and reasonably assure the leak tightness and structural integrity of the BFN containment are maintained. In the above discussion, TVA asserted that the proposed one-time ILRT interval extension will not affect its ability to promptly detect and address degradation in these areas. Based on NRC staffs assessment of past TVA containment performance experience as indicated in the licensees submittal, the NRC staff concurs with this assertion.

Monitoring of Nitrogen Makeup The NRC staff requested TVA to discuss its past operating experience regarding its use of the monitoring of the nitrogen makeup to containment as a complimentary means for assuring the leak tightness and structural integrity of the BFN containment, and to elaborate as to why the use of this supplementary approach strengthens its case for requesting a one-time extension of the BFN containment ILRT interval.

TVA responded that during plant operation, the BFN containment is inerted with nitrogen.

Drywell pressure is maintained in the range of 1.1 to 1.35 psid positive pressure with respect to the suppression pool (torus) and consequently positive with respect to the outside atmosphere.

The torus air space pressure is typically slightly positive with respect to atmosphere (about 0.1 psig). Although normal operating pressures in the drywell and torus atmosphere are less than that resulting from a Design Basis Accident, the fact that the containment is pressurized provides a reliable means of verifying that no large leak paths exist in the containment structure. Specifically, any substantial containment leak path will cause operational difficulties in maintaining positive pressure in the containment and the condition will manifest itself in an excessive nitrogen make-up rate. Monitoring for containment leakage is accomplished by monitoring the average daily nitrogen consumption used by the containment inerting system and is determined daily by the performance of Surveillance Instruction, SI-4.7.A.2.a, Primary Containment Nitrogen Consumption and Leakage. Significant containment leakage would be identified through increased nitrogen usage needed to maintain the required TS pressure, and would be investigated promptly and addressed within the scope of the plant Corrective Action Program. Excess nitrogen consumption will also be observed if the nitrogen supply system external to the containment is not tight. There is recent operating experience following a refueling outage where excess nitrogen make-up use was documented and promptly investigated, which was then determined to be caused by supply side leaks. This demonstrated the efficacy of the plant Corrective Action Program in flagging excess nitrogen use and investigating the cause. Therefore, the nitrogen make-up monitoring procedure supports TVAs position for a one-time extension of the ILRT interval in that it provides an additional means of identifying conditions that would be indicative of containment leakage, and that corrective action would be initiated in a timely manner. The NRC staff finds the above BFN operating experience-based response adequate and acceptable.

Concrete and Steel Containment Degradation The NRC staff noted a concern that inspections of some reinforced concrete and steel containment structures have identified degradation on the un-inspectable (embedded) side of the drywell steel shell and steel liner of the primary containment. TVA was requested to address how these undetected potential leakages under high pressure during core-damage accidents are adequately factored into the risk assessment implemented for justifying the proposed one-time ILRT interval extension.

TVA responded that an assessment of the potential impact on LERF due to age-related degradation of non-inspectable areas of the containment is presented in Enclosure 3 to the licensees submittal. The analysis in section 6.11 of the Enclosure uses the quantitative approach used by other industry plants (e.g., Dresden). The industry data set and methodology were used to determine the corrosion induced non-detectible containment leakage probabilities used in the section 6.11 assessment. TVA indicated that BFN has not experienced any degradation of the steel shell and liner of primary containment which would affect the structural integrity or the leak tightness of the containment vessel, therefore, the use of the industry data set for BFN is judged to be both appropriate and conservative.

The NRC staff finds the containment degradation evaluation provided in Section 6.11 of is an adequate and acceptable way of addressing the NRC staffs concern and, therefore, considers the issues related to the degradation on the un-inspectable (embedded) side of the drywell steel shell and steel liner of the primary containment to be resolved.

On the basis of its review of the information provided by the licensee in the TS amendment request and the TVAs response to the request for additional information, the NRC staff finds that (1) the structural integrity of the containment vessel is verified through periodic ISIs as required by Subsection IWE of the ASME Code,Section XI, and (2) the integrity of the penetrations and containment isolation valves are periodically verified through Type B and Type C tests as required by 10 CFR Part 50, Appendix J. In addition, the system pressure tests for containment pressure boundary (i.e., Appendix J tests, as applicable) are required to be performed following repair and replacement activities, if any, in accordance with Article IWE-5000 of the ASME Code,Section XI. The findings and approval presented in this safety evaluation are based on the NRCs staffs review of plant specific information, evaluations, operating experience, and analyses provided and are applicable to BFN Units 2 and 3 only.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 (69 FR 46592). 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 Commission's 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.

7.0 REFERENCES

7.1 Letter from T. E. Abney of TVA to NRC, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specifications (TS) Change 448 - One-Time Frequency Extension for Containment Integrated Leakage Rate Test (ILRT) Interval, July 8, 2004.

7.2 Regulatory Guide 1.174, Rev. 1, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, June 2002.

7.3 Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program.

7.4 NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J.

7.5 ANSI/ANS 56.8 - 1994, Containment System Leakage Testing Requirements.

7.6 Letter from T. E. Abney, TVA to NRC, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specifications (TS) Change 448 - One-Time Frequency Extension for Containment Integrated Leakage Rate Test (ILRT) Interval - Response to Request for Additional Information (TAC Nos. MC3745 and MC3746), November 24, 2004.

Principal Contributors: David Jeng Robert Palla Raj Goel Date: March 9, 2005

Mr. Karl W. Singer BROWNS FERRY NUCLEAR PLANT Tennessee Valley Authority cc:

Mr. Ashok S. Bhatnagar, Senior Vice President Nuclear Operations Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Larry S. Bryant, General Manager Nuclear Engineering Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Michael D. Skaggs Site Vice President Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 General Counsel Tennessee Valley Authority ET 11A 400 West Summit Hill Drive Knoxville, TN 37902 Mr. John C. Fornicola, Manager Nuclear Assurance and Licensing Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Kurt L. Krueger, Plant Manager Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 Mr. Fredrick C. Mashburn Senior Program Manager Nuclear Licensing Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. Timothy E. Abney, Manager Licensing and Industry Affairs Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 Senior Resident Inspector U.S. Nuclear Regulatory Commission Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, AL 36130-3017 Chairman Limestone County Commission 310 West Washington Street Athens, AL 35611