ML040340168

Issuance of Amendment No. 173, Authorizing Changes to the Surveillance Requirements for Containment Integrated Leak Rate Testing in TS 4.4.a, Interval Leak Rate Test (Type A).

ML040340168
Person / Time
Site:	Kewaunee
Issue date:	04/06/2004
From:	John Lamb NRC/NRR/DLPM/LPD3
To:	Coutu T Nuclear Management Co
References
TAC MB9907
Download: ML040340168 (15)

Text

April 6, 2004 Mr. Thomas Coutu Site Vice President Kewaunee Nuclear Power Plant Nuclear Management Company, LLC N490 Highway 42 Kewaunee, WI 54216-9511

SUBJECT:

KEWAUNEE NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT (TAC NO. MB9907)

Dear Mr. Coutu:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 173 to Facility Operating License No. DPR-43 for the Kewaunee Nuclear Power Plant. This amendment consists of changes to the Technical Specification (TS) in response to your application dated June 20, 2003, as supplemented by letter dated December 12, 2003.

The amendment authorizes changes to the surveillance requirements for containment integrated leak rate testing in TS 4.4.a, Integrated Leak Rate Tests (Type A).

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely,

/RA/

John G. Lamb, Project Manager, Section 1, Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-305

Enclosures:

1. Amendment No. 173 to License No. DPR-43

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

April 6, 2004 Mr. Thomas Coutu Site Vice President Kewaunee Nuclear Power Plant Nuclear Management Company, LLC N490 Highway 42 Kewaunee, WI 54216-9511

SUBJECT:

KEWAUNEE NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT (TAC NO. MB9907)

Dear Mr. Coutu:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 173 to Facility Operating License No. DPR-43 for the Kewaunee Nuclear Power Plant. This amendment consists of changes to the Technical Specification (TS) in response to your application dated June 20, 2003, as supplemented by letter dated December 12, 2003.

The amendment authorizes changes to the surveillance requirements for containment integrated leak rate testing in TS 4.4.a, Integrated Leak Rate Tests (Type A).

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely,

/RA/

John G. Lamb, Project Manager, Section 1, Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-305

Enclosures:

1. Amendment No. 173 to License No. DPR-43

2. Safety Evaluation cc w/encls: See next page Distribution:

GHill (2) JLamb PUBLIC ACRS PD 3-1 r/f PLouden, RIII OGC THarris ADAMS ACCESSION NO. ML040340168 *Provided SE input by memo OFFICE PM:PD3-1 LA:PD3-1 SC:SPSB* SC: EMEB* SC:IROB OGC SC:PD3-1 NAME JLamb THarris RDennig DTerao TBoyce RHoefling LRaghavan DATE 02/18/04 02/12/04 01/19/04 01/06/04 03/23/04 04/01/04 04/05/04 OFFICIAL RECORD COPY

Kewaunee Nuclear Power Plant cc:

John Paul Cowan Jonathan Rogoff Executive Vice President & Vice President, Counsel & Secretary Chief Nuclear Officer Nuclear Management Company, LLC Nuclear Management Company, LLC 700 First Street 700 First Street Hudson, WI 54016 Hudson, MI 54016 Larry L. Weyers James McCarthy Chairman, President and CEO Plant Manager Wisconsin Public Service Corporation Kewaunee Nuclear Power Plant 600 North Adams Street N490 Highway 42 Greey Bay, WI 54307-9002 Kewaunee, WI 54216-9511 David Zellner Gerry Riste Chairman - Town of Carlton Manager, Regulatory Affairs N2164 County B Kewaunee Nuclear Power Plant Kewaunee, WI 54216 N490 Highway 42 Kewaunee, WI 54216-9511 Mr. Jeffery Kitsembel Electric Division David Molzahn Public Service Commission of Wisconsin Nuclear Asset Manager PO Box 7854 Wisconsin Public Service Corporation Madison, WI 53707-7854 600 N. Adams Street Green Bay, WI 54307-9002 Thomas Webb Nuclear Asset Manager Wisconsin Public Service Corporation 600 N. Adams Street Green Bay, WI 54307-9002 Resident Inspectors Office U. S. Nuclear Regulatory Commission N490 Hwy 42 Kewaunee, WI 54216-9510 Regional Administrator Region III U. S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351

NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-305 KEWAUNEE NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 173 License No. DPR-43

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

A. The application for amendment by Nuclear Management Company, LLC (the licensee), dated June 20, 2003, as supplemented by letter dated December 12, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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-43 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 173, are hereby incorporated in the license. The licensees shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance, and is to be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 6, 2004

ATTACHMENT TO LICENSE AMENDMENT NO. 173 FACILITY OPERATING LICENSE NO. DPR-43 DOCKET NO. 50-305 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT TS 4.4-1 TS 4.4-1

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO AMENDMENT NO. 173 TO FACILITY OPERATING LICENSE NO. DPR-43 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR POWER PLANT DOCKET NO. 50-305

1.0 INTRODUCTION

By application dated June 20, 2003, as supplemented by letter dated December 12, 2003, the Nuclear Management Company (NMC or the licensee) requested changes to the Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS). Specifically, the proposed changes would revise the surveillance requirements for containment integrated leak rate testing (ILRT) in TS 4.4.a, Integrated Leak Rate Tests (Type A), to allow a one-time extension of the interval between ILRTs from 10 years to 15 years.

The supplemental information dated December 12, 2003, contained clarifying information, did not change the scope of the June 20, 2003, application or the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix J, Option B requires that a Type A test be conducted at a periodic interval based on historical performance of the overall containment system. KNPP TS 4.4 requires that Type A leakage rate testing be performed in accordance with the Containment Leakage Rate Testing Program. KNPP TS 6.20, Containment Leakage Rate Testing Program, 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 Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995. This RG endorses, with certain exceptions, 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.

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 most recent two Type A tests at KNPP have been successful, so the current interval requirement is 10 years.

The licensee is requesting a change to TS 4.4, Containment Tests, which would add a

one-time change from the guidelines of RG 1.163 and NEI 94-01, Revision 0, regarding the Type A test interval. Specifically, the proposed TS states that the requirement to test at least once per 10 years would be changed to at least once per 15 years, but only for the interval following the Type A test performed in April 1994.

The local leakage rate tests (Type B and Type C tests), including their schedules, are not affected by this request.

3.0 TECHNICAL EVALUATION

The Nuclear Regulatory Commission (NRC) staff reviewed the licensees regulatory and technical analysis in support of its license amendment as described in the licensees submittal dated June 20, 2003, as supplemented in a letter dated December 12, 2003. The following evaluation addresses the acceptability of issuing amendments pursuant to 10 CFR 50.92.

3.1 TS Surveillance Requirement (SR) 4.4.a, Integrated Leak Rate Tests (Type A)

The licensee has proposed to revise SR in TS 4.4.a to add a paragraph after the end of the first sentence. The revised TS 4.4.a is as follows with the proposed changes underlined:

a. Integrated Leak Rate Tests (Type A)

Perform required visual examination and leakage rate testing in accordance with the Containment Leakage Rate Testing Program.

As a one-time change, the Type A test frequency specified in NEI 94-01, Revision 0, paragraph 9.2.3, as ...at least once per 10 years based on acceptable performance history is changed to ... at least once per 15 years based on acceptable performance history. This change applies only to the interval following the Type A test performed in April 1994.

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. The NEI guidelines also contain a provision for extending the test interval an additional 15 months under certain circumstances. The leak rate testing requirements of Option B in Appendix J, and the containment inservice inspection (ISI) requirements mandated by 10 CFR 50.55a complement each other in ensuring the leak-tightness and structure integrity of the containment. Therefore, a detailed evaluation related to the ISI of the containment and potential areas of degradation in the containment is performed in the following section.

3.2 ISI for Primary Containment Integrity KNPP is a 2-loop, Westinghouse pressurized-water reactor (PWR) with a large, dry (ambient) steel primary containment structure. The containment pressure boundary consists of the steel containment shell structure with a concrete enclosure building, containment access penetrations, and process piping and electrical penetrations. The primary containment is also called the Reactor Containment Vessel (RCV). The RCV is a cylindrical steel pressure vessel with a hemispherical dome and ellipsoidal bottom which houses the reactor vessel, steam

generators, reactor coolant pumps, reactor coolant loops, accumulators of safety injection system, reactor coolant pressurizer relief tank, and other branch connections of the reactor coolant system. The RCV is completely enclosed by the shield building. The RCV is supported on a grout base that was placed after vessel construction was completed and tested. Both the RCV and shield building are supported on a common foundation slab.

The integrity of the penetrations and isolation valves are verified through Type B and Type C local leak rate tests (LLRTs) as required by 10 CFR Part 50, Appendix J, and the overall leak-tight integrity of the primary containment is verified through an ILRT. These tests are performed to verify the essentially leak-tight characteristics of the containment structure at the design-basis accident pressure. The last ILRT for Kewaunee was performed in April 1994.

The current due date for the next ILRT is April 2004. Compliance with this due date would have required that KNPP perform the ILRT during the spring 2003 refueling outage; therefore, the licensee used a portion of the 15-month extension and the next ILRT is cheduled for the fall 2004 refueling outage. With the extension of the ILRT time interval, the next overall verification will be performed no later than April 2009. Because the amendment request did not provide any information on the condition of containment components, the NRC staff requested additional information from the licensee. The licensee provided additional information related to the ISI of the containment and discussed potential areas of weaknesses in the containment that may not be apparent in the risk assessment. The licensee provided responses to the NRC staffs request for additional information in a letter dated December 12, 2003. The NRC staffs evaluation of the licensees responses to the ISI issues is discussed in the following paragraphs.

3.2.1 ISI Program at Kewaunee The licensee stated that the containment ISI program was established in 1995 in accordance with Subsection IWE and IWL of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1992 Edition with the 1992 Addenda, to detect degradation affecting the containment pressure integrity. The first 10-year inspection interval has been established from September 9, 1996, to September 9, 2006. The containment inspection (Subsection IWE) is divided into three 40-month inspection periods as follows:

First Period: September 9, 1996 - January 9, 2000 Second Period January 9, 2000 - May 9, 2003 Third Period May 9, 2003 - September 9, 2006 The scope of the program includes all the containment surfaces, pressure-retaining welds, containment surfaces requiring augmented examination, seals, gaskets, moisture barriers, pressure-retaining dissimilar metal welds, pressure-retaining bolting and other pressure-retaining components that are required to be examined. The licensee states that no exceptions are taken to the ASME Code,Section XI requirements, and that it has not requested any alternatives to Subsections IWE and IWL requirements for KNPP that credit the performance of the integrated (Type A) or local (Type B and C) leak rate testing.

The licensee states that containment ISI program examinations performed since the most recent ILRT (April 1994) include visual examinations of the containment vessel pursuant to ASME Code,Section XI, Subsection IWE. These visual examinations were conducted between 1998 and 2001. With the exception of the conditions described below, all results were within the established Code acceptance criteria.

In 1998, an inspection of the KNPP containment identified the following:

• Two (2) visual recordable indications on Reactor Building Containment Vessel Plate 98 and Plate 107 were recorded during performance of general visual examinations. The recordable indications were removed by grinding. Ultrasonic, magnetic particle and general visual examinations were performed following repair with no indications recorded.

• Two (2) visual recordable indications on Reactor Building Containment Vessel Plate 120 and Plate 155 were recorded during performance of general visual examinations. The recordable indications were apparent gouges in the base metal. Supplemental ultrasonic examination determined that there was no violation of minimum wall and the gouges were accepted as is.

In 2000, an inspection of the KNPP containment identified the following:

• Visual recordable indications on Reactor Building Containment Vessel equipment door inner and outer gaskets were recorded during performance of visual testing (VT) VT-3 examinations. The recordable visual indications noted were damage to the gaskets and portion of the gaskets with tear. Both the inner and outer door gaskets were replaced.

VT-3 examinations were performed on the replacement equipment door, inner and outer gaskets, and found to be acceptable.

• Leak testing results obtained per Appendix J Type B test requirements on Penetration 41E (vacuum breaker) O-ring seals exceeded the administrative limits.

The condition was repaired and subsequent reexamination measured acceptable leakage.

In 2001, an inspection of the KNPP containment identified the following:

• General visual indication consisting of a 4 x 8 inch surface defect was recorded on Plate 155. This surface defect was recorded and accepted in 1998 and showed no change in dimension or surface condition during the 2001 refueling outage.

• General visual indication consisting of a slight inward bulge on Plates 74, 75 and 83 was recorded. The slight inward bulge was evaluated by KNPP Engineering and accepted under KNPP specification TS-1052, Addendum No. 4, Item No. 19, Section 10.3, Shell Tolerance.

• General visual indications consisting of a slight outward bulge on Plate 144, 145, 146, 147 and 148 were recorded. The slight outward bulge was evaluated and accepted by KNPP engineering analysis.

• General visual indication consisting of weld deposits on Plate 62 and arc strikes on Plate 99 were recorded and are acceptable per ASME Boiler and Pressure Vessel Code Section XI 1992 Edition including 1992 Addenda.

• VT-3 indications consisting of lack of bonding and tears were recorded in the moisture barriers on Plate 62, 64, 65, 66, and 67, and were repaired.

• A VT-1 indication on an Emergency Airlock bolt was recorded and was repaired.

The NRC staff finds that these conditions were adequately identified and the necessary corrective actions were taken.

3.2.2 Implementing IWE-1240 at Kewaunee The licensee states that ASME Code,Section XI, Subsection IWE inspection plan was implemented for KNPP on September 9, 1996. All inspections have been completed through the second period, first outage, of the first 10-year surveillance interval. There are currently no identified areas at KNPP that require augmented inspection in accordance with IWE-1240.

3.2.3 IWE Table-2500-1, Examination Categories E-D and E-G for seals and gaskets, and examination and testing of bolts The licensee states that the KNPP containment ISI program includes all the containment surfaces, pressure-retaining welds, containment surfaces requiring augmented examination, seals, gaskets, moisture barriers, pressure-retaining dissimilar metal welds, pressure-retaining bolting and other pressure-retaining components that are required to be examined. The first 10-year containment ISI interval has been established from September 9, 1996 to September 9, 2006. The containment ISI program is unaffected by the proposed amendment, and will continue to provide a high degree of assurance that any degradation of the containment pressure boundary will be detected and corrected before it can result in an unacceptable leakage path.

3.2.4 Integrity of stainless steel bellows In the past, the NRC staff has found that two-ply stainless steel bellows are susceptible to trans-granular stress corrosion cracking, and the leakage through them may not be detectable by Type B testing (see NRC Information Notice 92-20 for further information). The licensee states that KNPP has nine penetration assemblies that incorporate two-ply mechanical bellows, and one penetration assembly (Penetration 18) that contains three bellows with mesh inserts.

These are the two main feedwater, two main steam, two steam generator blowdown, two residual heat removal, one letdown, and fuel transfer tube (penetration 18) penetrations. Plant drawings indicate that wire mesh is installed between the two-plies of each bellows assembly, ensuring that an adequate gap exists to measure leakage when performing the required Type B tests. The ILRT administrative acceptance criterion for measured leakage through these penetrations is very low at 100 standard cubic centimeters per minute. These penetrations have been tested each outage per the KNPP Containment Leak Rate Testing Program with satisfactory results. The historical as-found LLRT results for each of the bellows are provided in Attachment 2, Table 1 of the licensees letter dated December 12, 2003.

3.2.5 Inspection of embedded side of the containment steel shell The NRC staff expressed concern that inspections of some reinforced concrete and steel containment structures have identified degradation on the uninspectable (embedded) side of the containment steel shell of the primary containment. The licensee states that, as part of the scheduled containment ISI inspections, the moisture barrier areas on the periphery of the inaccessible portion of the containment vessel are inspected. Appropriate action would be taken for any indications of degradation in these areas. There are no other established programs for monitoring the condition of inaccessible portions of the steel containment vessel.

The KNPP containment vessel is freestanding, and the fraction of the surface area that is inaccessible for inspection is significantly smaller than at plants with steel-lined concrete containment. As a result, the potential contribution to risk from corrosion in uninspectable areas for an extended ILRT interval is less.

With regard to the issue related to the inaccessible areas of the steel containment vessel for which degradations cannot be found by visual examinations, the licensee, as discussed in the June 20, 2003, submittal and the supplement dated December 12, 2003, performed an ILRT extension risk assessment considering the potential age related corrosion effects on the containment integrity and a series of parametric sensitivity studies. The results of the risk assessment indicated that the ILRT interval extension has a minimal impact on the plant risk.

The details of the NRC staffs evaluation regarding the risk assessment performed by the licensee is described below in Section 3.4.

From the discussion above, the NRC staff finds that implementation of the licensees containment ISI program, including the areas covered by augmented inspections, will provide adequate assurance that the containment structural integrity will be maintained during the extended ILRT period.

Based on its review of the information provided in the licensees TS change request submittal dated June 20, 2003, and responses to the NRC staffs questions dated December 12, 2003, the NRC staff finds that (1) the structural integrity of the containment vessel is verified through the periodic ISI conducted as required by Subsections IWE and IWL of the ASME Code,Section XI; and (2) the integrity of the penetrations, containment isolation valves and mechanical bellows is periodically verified through Type B and Type C tests as required by 10 CFR Part 50, Appendix J and Kewaunee TS. 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 in accordance with Subarticle IWE-5000 of the ASME Code,Section XI. Significant degradation of the primary containment pressure boundary is required to be reported under 10 CFR 50.72 or 10 CFR 50.73.

3.2.6 ISI for Primary Containment Integrity Conclusion The NRC staff concludes, based on the considerations discussed above, that the licensee has adequate procedures to examine and monitor potential age-related and environmental degradations of the pressure retaining components of the Kewaunee primary containment.

Thus, the NRC staff finds that granting a one-time extension of performing the ILRT as proposed by the licensee in TS 4.4.a is acceptable.

3.3 Probabilistic Safety 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 June 20, 2003, application. Additional analysis and information was provided by the licensee in its letter dated December 12, 2003. 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 Kewaunee 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 the 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 boiling-water reactor 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 its analyses for a change in test frequency from three tests in 10 years to one test in 15 years, as discussed below. These comparisons bound the effect of going from the current one test in 10 years to one test in 15 years. The following conclusions can be drawn from the analysis associated with extending the test frequency to a one in 15-year 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 about 0.03 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 7.6 x 10-7 per year based on the internal events PRA. 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 examinations 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 liner on the proposed change. The increase in LERF associated with corrosion events is estimated to be about 2 x 10-8 per year.

When the calculated increase in LERF is in the range of 10-7 per year to 10-6 per year, applications are considered if the total LERF is less than 10-5 per year. The licensee estimates that the total LERF for internal and external events, including the impact of extending the Type A test interval, is approximately 8 x 10-6 per year. The NRC staff concludes that increasing the Type A interval to 15 years results in only a 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 0.4 percentage points 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 consistent with the acceptance guidelines while maintaining the defense-in-depth philosophy of RG 1.174 and, therefore, is acceptable. Based on the above evaluation, the NRC staff finds that the interval until the next Type A test at Kewaunee Nuclear Power Plant may be extended to 15 years, and that the proposed change to TS 4.4 is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Wisconsin State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

This amendment changes 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 amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent 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 this amendment involves no significant hazards consideration and there has been no public comment on such finding (68 FR 43391). Accordingly, this amendment meets 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 this amendment.

6.0 CONCLUSION

The NRC staff 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: G. Bedi R. Pulsipher R. Palla Date: April 6, 2004