ML021970215

Amendments 98 & 98, ILRT Extension from 10 to 15 Years

ML021970215
Person / Time
Site:	Comanche Peak
Issue date:	08/15/2002
From:	Jaffe D NRC/NRR/DLPM/LPD4
To:	Terry C TXU Energy
Jaffe D, NRR/DLPM, 415-1439
References
TAC MB3685, TAC MB3686
Download: ML021970215 (16)

Text

August 15, 2002 Mr. C. Lance Terry Senior Vice President

& Principal Nuclear Officer TXU Energy ATTN: Regulatory Affairs P. O. Box 1002 Glen Rose, TX 76043

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES), UNITS 1 AND 2 -

ISSUANCE OF AMENDMENTS RE: ONE TIME EXTENSION OF APPENDIX J, TYPE A, INTEGRATED LEAK RATE TEST INTERVAL FROM TEN TO FIFTEEN YEARS (TAC NOS. MB3685 AND MB3686)

Dear Mr. Terry:

The Commission has issued the enclosed Amendment No. 98 to Facility Operating License No. NPF-87 and Amendment No. 98 to Facility Operating License No. NPF-89 for CPSES, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated December 26, 2001, as supplemented by letters dated February 4 and June 12, 2002.

The amendments revise TS 5.5.16, Containment Leakage Rate Testing Program, to extend Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix J, Type A, Containment Integrated Leak Rate Test (ILRT) date for CPSES, Units 1 and 2. The CPSES Unit 1 date would be extended from the fall of 2002 to December 2008, and Unit 2 would be extended from the fall of 2006 to December 2012. The following phrase is added to TS 5.5.16.a: ... as modified by the following exception: 1. NEI [Nuclear Energy Institute] 94 1995, Section 9.2.3: The first Type A Test performed after the December 7, 1993 Type A Test (Unit 1) and the December 1, 1997 Type A Test (Unit 2) shall be performed no later than December 15, 2008 (Unit 1) and December 9, 2012 (Unit 2).

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

Sincerely,

/RA/

David H. Jaffe, Senior Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-445 and 50-446

Enclosures:

1. Amendment No. 98 to NPF-87

2. Amendment No. 98 to NPF-89

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

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

Sincerely,

/RA/

David H. Jaffe, Senior Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-445 and 50-446

Enclosures:

1. Amendment No. 98 to NPF-87

2. Amendment No. 98 to NPF-89

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

PUBLIC PDIV-1 Reading RidsNrrDlpmPdiv (WRuland)

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RDennig JPulsipher MSnodderly TCheng RidsRgn4MailCenter (KBrockman)

LHurley, RIV Accession No.: ML021970215 OFFICE PDIV-1/PM PDIV-1/LA EMEB/SC

• SPLB/SC** OGC NLO PDIV-1/SC w/comment NAME DJaffe DJohnson DTerao SWeerakody SBrock RGramm DATE 7/22/02 07/22/02 7/8/02 7/28/02 8/5/02 8/9/02

• No major changes to SE Input dated July 8, 2002 (ML021900102)

• • No major changes to SE Input dated June 28, 2002

TXU GENERATION COMPANY LP COMANCHE PEAK STEAM ELECTRIC STATION, UNIT NO. 1 DOCKET NO. 50-445 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 98 License No. NPF-87

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

A. The application for amendment by TXU Generation Company LP dated December 26, 2001, as supplemented by letters dated February 4 and June 12, 2002, 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, as amended, 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 license 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. NPF-87 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 98 , and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. TXU Generation Company LP shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The 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/

Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 15, 2002

TXU GENERATION COMPANY LP COMANCHE PEAK STEAM ELECTRIC STATION, UNIT NO. 2 DOCKET NO. 50-446 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 98 License No. NPF-89

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

A. The application for amendment by TXU Generation Company LP dated December 26, 2001, as supplemented by letters dated February 4 and June 12, 2002, 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, as amended, 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 license 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. NPF-89 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 98 , and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. TXU Generation Company LP 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 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

ATTACHMENT TO LICENSE AMENDMENT NO. 98 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 98 TO FACILITY OPERATING LICENSE NO. NPF-89 DOCKET NOS. 50-445 AND 50-446 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 5.0-27 5.0-27

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 98 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 98 TO FACILITY OPERATING LICENSE NO. NPF-89 TXU GENERATION COMPANY LP COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2 DOCKET NOS. 50-445 AND 50-446

1.0 INTRODUCTION

By application dated December 26, 2001, as supplemented by letters dated February 4 and June 12, 2002, TXU Generation Company LP (the licensee) requested changes to the Technical Specifications (TSs) for the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2. The proposed changes would revise TS 5.5.16, Containment Leakage Rate Testing Program, to extend Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix J, Type A, Containment Integrated Leak Rate Test (ILRT) for CPSES, Units 1 and 2. The CPSES Unit 1 date would be extended from the fall of 2002 to December of 2008, and Unit 2 would be extended from the fall of 2006 to December 2012. The following phrase would implement this change to TS 5.5.16.a: ... as modified by the following exception: 1. NEI [Nuclear Energy Institute] 94 1995, Section 9.2.3: The first Type A Test performed after the December 7, 1993 Type A Test (Unit 1) and the December 1, 1997 Type A Test (Unit 2) shall be performed no later than December 15, 2008 (Unit 1) and December 9, 2012 (Unit 2).

The February 4 and June 12, 2002, supplemental letters provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

2.0 BACKGROUND

The regulation at 10 CFR Part 50, Appendix J, Option B requires that Type A tests be conducted at a periodic interval based on the historical performance of the overall containment system. CPSES, Units 1 and 2, TS 5.5.16 requires that the ILRT frequency shall be performed in accordance with 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, 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 each unit at CPSES have been successful, so the current ILRT interval requirement is 10 years.

3.0 EVALUATION 3.1 RISK ASSESSMENT The licensee has performed a risk impact assessment of extending the Type A test interval to 15 years. The assessment was provided to the U. S. Nucler Regulatory Commission (NRC) staff in the December 26, 2001, 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) Research Project Report 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 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, (September 1995) provided the technical basis to support rulemaking 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 the NRCs rulemaking basis, NEI 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 EPRI study estimated that relaxing the test frequency from 3 in 10 years to 1 in 10 years will increase the average time that a leak, 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, in percent of person-rem/year, for the pressurized water reactor and boiling water reactor representative plants confirmed the NUREG-1493 conclusion that a reduction in the frequency of Type A tests from 3 in 10 years to 1 in 10 years leads to an imperceptible increase in risk ranging from 0.02 to 0.14 percent.

Building upon the methodology of the EPRI study, the licensee assessed the change in the predicted person-rem/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 staff has issued RG 1.174 on the use of probabilistic risk assessment in risk-informed changes to a plants licensing basis. The licensee has proposed using RG 1.174 to assess the acceptability of extending the Type A test interval beyond that established during the 10 CFR Part 50, Appendix J, 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 reactor year and increases in large early release frequency (LERF) less than 10-7 per reactor 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 3 in 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 an analysis which estimated all of these risk metrics and whose methodology is consistent with previously approved submittals. The following conclusions can be drawn from the analysis associated with extending the Type A test interval:

1. A slight increase in risk is predicted when compared to that estimated from current requirements. Given the change from a 10 year test interval to a 15 year test interval, the increase in the total integrated plant risk, in person-rem/year, is estimated to be 0.006 percent. The increase in the total integrated plant risk, given the change from a 3 in 10 year test interval to a 15 year test interval, was 0.018 percent. NUREG-1493 concluded that a reduction in the frequency of tests from 3 in 10 years to 1 in 10 years leads to an imperceptible increase in risk, ranging from 0.02 to 0.14 percent.

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

2. RG 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in CDF less than 10-6 per reactor year and increases in LERF less than 10-7 per reactor year. Since the Type A test does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A test interval from 1 in 10 years to 1 in 15 years is estimated to be 1.9 x 10-8/year. The increase in LERF resulting from a change in the Type A test interval from the original 3 in 10 years to 1 in 15 years is estimated to be 5.6 x 10-8/year. Increasing the Type A test interval to 15 years results in a very small change in LERF.

The licensee performed additional risk analysis to consider the impact of hypothetical corrosion in inaccessible areas of the containment liner on the proposed change. The inaccessible areas included the imbedded side of the containment liner. The risk analysis considered the likelihood of an age-adjusted liner flaw that would lead to a breach of the containment. The risk analysis also considered the likelihood that the flaw was not visually detected but could be detected by a Type A test. When possible corrosion of the containment liner is considered, the increase in LERF resulting from a change in the Type A test interval from the original 3 in 10 years to 1 in 15 years is estimated to be 5.9 x 10-8/year. This additional risk analysis provides added assurance that increasing the Type A test interval to 15 years results in a very small change in LERF.

3. RG 1.174 also encourages the use of risk analysis techniques to help demonstrate 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 among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in the conditional containment failure probability was estimated to increase by 0.0010 for the proposed change and 0.0031 for the cumulative change of going from a test interval of 3 in 10 years to 1 in 15 years. The NRC staff finds that the defense-in-depth philosophy is maintained based on the change in the conditional containment failure probability for the proposed amendment.

Based on these conclusions, the 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 ASSESSMENT The primary reactor containment of each CPSES unit is a large, reinforced concrete vertical right cylinder with a flat base and a hemispherical dome. The containment pressure boundary consists of the steel liner, containment access penetrations, and process piping and electrical penetrations. The integrity of the penetrations is verified through Type B and Type C local leak rate tests (LLRT) as required by 10 CFR Part 50, Appendix J, and the overall integrity of the containment structure 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 (DBA) pressure. As stated in the December 26, 2001, application, only one Type A test has been performed for each unit of CPSES (December 7, 1993, for Unit 1 and December 1, 1997, for Unit 2) after the commercial operation (August 13, 1990, for Unit 1 and August 3, 1993, for Unit 2). Based on the guidelines provided in Section 9.2.3 of NEI 94-01, which is endorsed by Regulatory Guide 1.163, and the engineering evaluation performed to document the basis for treating the pre-operational test as one of the two Type A tests, the current interval requirement is 10 years for both units of CPSES. With the requested extension of the ILRT time interval, the licensee proposed that the next overall verification of the containment leak-tight integrity will be performed by December 15, 2008, for Unit 1 and December 9, 2012, for Unit 2. Because the leak rate testing requirements (ILRT and LLRTs) of 10 CFR Part 50, Appendix J, Option B and the containment inservice inspection (ISI) requirements mandated by 10 CFR Part 50, Section 50.55a, complement each other in ensuring the leak-tightness of the pressure boundary and the structural integrity of the containment, the licensee, in its request, provided information related to the ISI of the containment and potential areas of weakness in the containment that may not be apparent in the risk assessment. In addition, the licensee provided information in the December 26, 2001, application and the February 4, 2002, supplement, to explicitly address the five issues identified by the staff during its ILRT review of other plants. The staffs evaluation of the licensees response to these generic issues is discussed later in this safety evaluation.

In addressing its containment ISI program, the licensee stated that inspections of the CPSES containment buildings (concrete shell and liner) are conducted in accordance with the requirements of the 1998 edition of Subsections IWE and IWL of the American Society of Mechanical Enginers (ASME) Boilier and Pressure Vessel Code (Code)Section XI, supplemented by licensee commitments. According to a July 23, 1999, letter to the licensee concerning ASME Code inspections for the CPSES containments, NRC staff found the use of

the requirements of the 1998 edition of ASME Subsections IWE and IWL to perform the ISI for the CPSES containment acceptable. The licensee also indicated that CPSES completed the first interval inspections for CPSES, Units 1 and 2, in September 2001 with acceptable results.

For the issue related to the application of any augmented examination (required by IWE Table-2500-1, Examination Category E-C) and findings at CPSES, the licensee stated that, based on the results of the containment ISI examinations of 2001, there are no areas of the containment liner that required augmented examinations according to Subarticle IWE-1240 of Subsection IWE of ASME Section XI, 1998 Edition. The licensee also indicated that the examinations identified three indications of construction deficiencies in the Unit 1 containment liner insert plates containing full penetration attachment welds. An evaluation determined that insert plate welds were of an acceptable quality to adequately assure performance of the safety function of the containment liner.

From the discussion above, the staff finds that the licensees ISI program will provide reasonable assurance that the containment structural integrity will be maintained during the extended ILRT period.

With regard to ISI of seals, gaskets, and pressure retaining bolting, the licensee stated in the December 26, 2001, application and the February 4, 2002, supplement that with the NRCs authorization to use the 1998 Edition of the ASME Code,Section XI, Subsections IWE and IWL, supplemented by licensee commitments, there is no separately scheduled ISI on seals, gaskets, or bolting components that are Type B tested per Option B of Appendix J of 10 CFR Part 50. According to the licensee, in their February 4, 2002, supplement, Appendix J, Type B testing requirements are applied to the O-rings, electrical and blind flange penetrations, and the airlocks. The combination of the 1998 Code-required general visual examinations for the Examination Category E-A, and the Appendix J, Type B (Option B) test will provide reasonable assurance that the integrity of the containment boundary is maintained during the period of the extended Type A test frequency. On these bases, the staff finds that the licensee has established adequate procedures to ensure the integrity of the containment pressure boundary.

As for the integrity of two-ply stainless steel bellows, the licensee stated that CPSES, Units 1 and 2, do not contain such bellows that act as part of the containment boundary. Therefore, the concerns of Information Notice 92-20 are not applicable to CPSES, Units 1 and 2, and no ISI of these bellows is required.

Because ILRTs help to identify areas of through-wall degradations when the containment vessel is pressurized, the staff requested that the licensee address how the potential leakages due to age-related degradation in the uninspectable areas (areas that cannot be visually examined) were considered in risk assessment of the extended ILRT. In addressing this staff concern, the licensee stated in their February 4, 2002, supplement that the Calvert Cliffs Nuclear Power Plant (CCNPP) method1 was used to evaluate the risk of the CPSES 1

Letter from Donna Skay, NRC to Charles H. Cruse, CCNPP, Calvert Cliffs Nuclear Power Plant (CCNPP), Unit 1 - Amendment Re: One-Time Extension of Appendix J , Type A, Integrated Leak Rate Test Interval and Exception from Performing a Post Modification Type A Test, dated May 1, 2002.

containment associated with liner corrosion. The risk assessment was performed based on the following assumptions:

1. A half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures.

2. The success data was limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. To limit the aging impact of this corrosion issue, additional success data was not used even though inspections were being performed prior to this date. And there is no evidence that liner corrosion issues were identified.

3. The liner flaw likelihood is assumed to double every five years. This is based solely on judgement and is included in this analysis to address the increased likelihood of corrosion as the liner ages. Sensitivity studies are included that address both doubling this rate every 10 years and every two years.

4. The likelihood of the containment atmosphere reaching the outside atmosphere, given a liner flaw exists, is a function of the pressure inside the containment. Even without the liner, the containment is an excellent barrier. But as the pressure in containment increases, cracks will form. If a crack occurs in the same region as a liner flaw, then the containment atmosphere can communicate to the outside atmosphere. At low pressures, this crack formation is extremely unlikely. Near the point of containment failure, crack formation is virtually guaranteed. Anchored points of 0.1 percent at 20 psia and 100 percent at 128.7 psia were selected. Intermediate failure likelihoods are determined through logarithmic interpolation. Sensitivity studies are included that decrease and increase the 20 psia anchor point by a factor of 10.

5. The likelihood of leakage escape (due to crack formation) in the basemat region is considered to be 10 times less likely than the containment cylinder and dome region.

6. A 5 percent visual inspection detection failure likelihood, given the flaw is visible, and a total detection failure likelihood of 10 percent is used. To date, all liner corrosion events have been detected by visual inspection. Sensitivity studies are included that evaluate total detection failure likelihoods of 5 percent and 15 percent.

7. All non-detectable containment over-pressurization failures are assumed to be LERF.

This approach avoids a detail analysis of containment failure and operator recovery actions.

As discussed in the case of CCNPP (see Footnote 1), the risk assessment (assessment method, assumptions, and results) performed for the CCNPP containment vessel is acceptable to the staff. The assessment results of the CPSES containment show that the risk of extending the ILRT to one in 15 years is small.

Based on the information provided in the December 26, 2001, letter, as supplemented, the staff finds that (1) the structural integrity of the CPSES, Units 1 and 2, containment vessels are verified through the periodic ISIs conducted as required by Subsections IWE and IWL of the ASME Code,Section XI; (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 and CPSES TS; and (3) the potential for large leakage from the areas that cannot be examined by the ISI has been explicitly modeled in performing the risk assessment. 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. Serious degradation of the primary containment pressure boundary is required to be reported under 10 CFR 50.72 and 10 CFR 50.73.

3.3 CHANGES TO THE TECHNICAL SPECIFICATIONS Based upon the NRC staff conclusions regarding the acceptability of the licensees containment inspection and testing programs for CPSES, Units 1 and 2, and the low increase in risk associated with extending the ILRT interval from 10 to 15 years, the NRC staff concludes that it is acceptable to change TS 5.5.16a. The change, as proposed by the licensee adds the following: ... as modified by the following exception: 1. NEI 94 1995, Section 9.2.3: The first Type A Test performed after the December 7, 1993 Type A Test (Unit 1) and the December 1, 1997 Type A Test (Unit 2) shall be performed no later than December 15, 2008 (Unit 1) and December 9, 2012 (Unit 2). NEI-94-01 includes the criterion that 10 CFR Part 50, Appendix J, Type A (ILRT), Option B testing be performed at a frequency of at least once per 10 years. The proposed change to TS 5.5.16a provides a suitable exception that extends the interval from 10 to 15 years. Accordingly, the NRC staff concludes that the proposed change to TS 5.5.16a is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Texas 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 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 (67 FR 5340, published February 5, 2002). 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 Contributors: J. Pulsipher M. Snodderly T. Cheng Date: August 15, 2002