ML022410163
| ML022410163 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 09/17/2002 |
| From: | Thadani M NRC/NRR/DLPM/LPD4 |
| To: | Cottle W South Texas |
| Thadani M, NRR/DLPM, 415-1476 | |
| References | |
| TAC MB2897, TAC MB2901 | |
| Download: ML022410163 (17) | |
Text
September 17, 2002 Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 Wadsworth, TX 77483
SUBJECT:
SOUTH TEXAS PROJECT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: REVISING THE APPENDIX J INTEGRATED LEAK RATE TESTING INTERVAL (TAC NOS. MB2897 AND MB2901)
Dear Mr. Cottle:
The Commission has issued the enclosed Amendment No. 143 to Facility Operating License No. NPF-76 and Amendment No. 131 to Facility Operating License No. NPF-80 for the South Texas Project, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated August 2, 2001, as supplemented by letters dated March 6, April 2, and June 25, 2002.
The supplemental letters of March 6, April 2, and June 25, 2002, provided additional information of clarifying nature which did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission staffs original proposed no significant hazards consideration determination as published in the Federal Register (67 FR 50959 published August 6, 2002).
The amendments revise the TS permitting a one time extension of Title 10 of the Code of Federal Regulations, Part 50, Appendix J, Option B, Performance-Based Leakage-Test Requirements.
A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.
Sincerely,
/RA/
Mohan C. Thadani, Senior Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499
Enclosures:
- 1. Amendment No. 143 to NPF-76
- 2. Amendment No. 131 to NPF-80
- 3. Safety Evaluation cc w/encls: See next page
September 17, 2002 Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 Wadsworth, TX 77483
SUBJECT:
SOUTH TEXAS PROJECT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: REVISING THE APPENDIX J INTEGRATED LEAK RATE TESTING INTERVAL (TAC NOS. MB2897 AND MB2901)
Dear Mr. Cottle:
The Commission has issued the enclosed Amendment No. 143 to Facility Operating License No. NPF-76 and Amendment No. 131 to Facility Operating License No. NPF-80 for the South Texas Project, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated August 2, 2001, as supplemented by letters dated March 6, April 2, and June 25, 2002.
The supplemental letters of March 6, April 2, and June 25, 2002, provided additional information of clarifying nature which did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission staffs original proposed no significant hazards consideration determination as published in the Federal Register (67 FR 50959 published August 6, 2002).
The amendments revise the TS permitting a one time extension of Title 10 of the Code of Federal Regulations, Part 50, Appendix J, Option B, Performance-Based Leakage-Test Requirements.
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/
Mohan C. Thadani, Senior Project Manager, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499
Enclosures:
- 1. Amendment No. 143 to NPF-76
- 2. Amendment No. 131 to NPF-80
- 3. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
PUBLIC RidsNrrDlpmLpdiv (WRuland)
RidsNrrDlpmLpdiv1 (RGramm)
PDIV-1 Reading RidsNrrPMMThadani RidsNrrLAMMcAllister RidsOgcRp RidsAcrsAcnwMailCenter GHill (4)
EThrom MSnodderly TCheng RidsRgn4Mailcenter (KBrockman)
RDennig ACCESSION NO: ML022410163
- NLO w/ changes
- No change in the SE input OFFICE PDIV-1/PM PDIV-D/LA EMEB/BC*
SPLB/BC*
SPSB/BC*
OGC **
PDIV-1/SC NAME MThadani MMcAllister CBray RGramm DATE 9/9/02 9/3/02 08/23/02 07/19/02 07/19/02 9/12/02 9/12/02 OFFICIAL RECORD COPY
STP NUCLEAR OPERATING COMPANY DOCKET NO. 50-498 SOUTH TEXAS PROJECT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
License No. NPF-76 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by STP Nuclear Operating Company* acting on behalf of itself and for Texas Genco, LP, the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and the City of Austin, Texas (COA) (the licensees), dated August 2, 2001, as supplemented by letters dated March 6, April 2, and June 25, 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.
- STP Nuclear Operating Company is authorized to act for Texas Genco, LP, the City Public Service Board of San Antonio, Central Power and Light Company, and the City of Austin, Texas, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.
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-76 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 143, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The STP Nuclear Operating Company 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 30 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: September 17, 2002
STP NUCLEAR OPERATING COMPANY DOCKET NO. 50-499 SOUTH TEXAS PROJECT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
License No. NPF-80 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by STP Nuclear Operating Company* acting on behalf of itself and for Texas Genco, LP, the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and the City of Austin, Texas (COA) (the licensees), dated August 2, 2001, as supplemented by letters dated March 6, April 2, and June 25, 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.
- STP Nuclear Operating Company is authorized to act for Texas Genco, LP, the City Public Service Board of San Antonio, Central Power and Light Company, and the City of Austin, Texas, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.
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-80 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 143, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The STP Nuclear Operating Company 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 30 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: September 17, 2002
ATTACHMENT TO LICENSE AMENDMENT NOS. 143 AND 131 FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-80 DOCKET NOS. 50-498 AND 50-499 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 6-18a 6-18a
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 143 AND 131 TO FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-80 STP NUCLEAR OPERATING COMPANY, ET AL.
SOUTH TEXAS PROJECT, UNITS 1 AND 2 DOCKET NOS. 50-498 AND 50-499
1.0 INTRODUCTION
By application dated August 2, 2001, as supplemented by letters dated March 6, April 2, and June 25, 2002, STP Nuclear Operating Company (the licensee), requested changes to the South Texas Project (STP), Units 1 and 2, Technical Specifications (TSs). The proposed changes would revise the TS of STP, Units 1 and 2, by adding the following sentence to the end of the first paragraph of TS Subsection 6.8.3.j: The current ten-year interval between performance of the integrated leakage rate (Type A) test, beginning September 24, 1991, for Unit 2 and March 10, 1995, for Unit 1, has been extended to 15 years (a one-time change).
Without an extension, the licensee would have to perform Type A tests at both units in the near future.
The supplemental letters of March 6, April 2, and June 25, 2002, provided additional information of clarifying nature which did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on (67 FR 50959, published August 6, 2002).
The TS change is requested based on risk informed approach. In performing the risk assessment, the licensee considered the guidelines of Nuclear Energy Institute (NEI) 94-01, Nuclear Energy Institute Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 0; the methodology used in Electric Power Research Institute (EPRI) TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing; and Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.
The licensee will continue to inspect the reactor containment under the requirements of the American Society of Mechanical Engineers (ASME)Section XI Subsections IWE and IWL. The licensee is not asking for modifications to existing Type B and C testing programs.
2.0 REGULATORY EVALUATION
The regulation at Title 10 of the Code of Federal Regulations (10 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. STP, Units 1 and 2 TS 6.8.3.j requires that a program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions. It further requires that this program shall be in accordance with the guidelines contained in RG 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by exceptions set forth in the TS. This RG endorses, with certain exceptions, NEI 94-01, Revision 0, dated July 26, 1995.
A Type A test is an overall integrated leakage rate test (ILRT) 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 2 consecutive successful tests. There is also a provision for extending the test interval an additional 15 months in certain circumstances. The most recent 2 Type A tests at STP, Units 1 and 2 have been successful, so their current interval requirement is 10 years.
The licensee is requesting an addition to TS 6.8.3.j, Containment Leakage Rate Testing Program, which would indicate that they are allowed to take an exception to the guidelines of RG 1.163 regarding the Type A test interval. Specifically, the proposed TS states The current ten-year interval between performance of the integrated leakage rate (Type A) test, beginning September 24, 1991, for Unit 2 and March 10, 1995, for Unit 1, has been extended to 15 years (a one-time change).
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 assessment was provided to the NRC staff in the August 2, 2001, application for license amendment. Additional analysis and information were provided by the licensee in letters dated March 6, April 2, and June 25, 2002. In performing the risk assessment, the licensee considered the guidelines of NEI 94-01, the methodology used in EPRI TR-104285, and RG 1.174.
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 dated 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 (LLRTs) 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 reactors and boiling water reactors representative plants confirmed the NUREG-1493 conclusion that a reduction in the frequency of Type A tests from 3 per 10 years to 1 per 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 NRC 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 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 frequency:
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 0.002 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, is 0.006 percent. NUREG-1493 concluded that a reduction in the frequency of tests from 3 per 10 years to 1 per 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 5.1 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 1.2 x 10-7/year.
When the calculated increase in LERF is in the range of 10-7 per reactor year to 10-6 per reactor year, applications are generally considered if the total LERF is less than 10-5 per reactor year. The licensees evaluation considered both internal and external events, including internal fires, seismic, and severe weather events. An estimate of the total LERF can be made by summing the Class 2, 3b, 7, and 8 sequence frequencies from Table 3 of the licensees April 2, 2002 submittal. The summation of these frequencies is 6.1 x 10-6/year. Increasing the Type A interval to 15 years is considered to be a small change in LERF and meets the acceptance guidelines of RG 1.174 as supported by the estimate that the total LERF including external events is less than 1.0 x 10-5.
The licensee performed an 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 backside 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 ILRT. 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 still be 1.2 x 10-7/year. This additional risk analysis provides added assurance that increasing the Type A interval to 15 years causes a small change in LERF. The details of the NRC staff's evaluation of the impact of the hypothetical corrosion are discussed below under Section 3.2, Containment Structural Integrity.
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 among prevention of core damage, prevention of containment failure, and consequence mitigation. The conditional containment failure probability was estimated to increase by 0.0045 for the proposed change and 0.011 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 amendments.
Based on these conclusions, the NRC staff finds that the increase in predicted risk due to the proposed change is within the acceptance guidelines, the defense-in-depth philosophy of RG 1.174 is maintained, and the predicted risk is not significant.
3.2 Containment Structural Integrity 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 LLRTs 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 at or near the design-basis accident pressure to ensure that: (a) leakage through containment vessel or systems and components penetrating containment does not exceed allowable leakage rates specified in the TS; and (b) structural integrity of the containment vessel during its service life will be maintained.
As stated in Attachment 1 of the licensees submittal, STP, Unit 1 has performed three Type A tests, pre-operational and operational, and the completion dates of these tests are: March 25, 1987; January 10, 1991; and March 10, 1995. STP, Unit 2 has performed two Type A tests, pre-operational and operational, and the completion dates are: September 27, 1988; and September 23, 1991. Based on these successful Type A tests performed at STP, Units 1 and 2 and the requirements of 10 CFR Part 50, Appendix J, Option B, the current interval requirement is 10 years. 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 March 2010 on Unit 1 and September 2006 on Unit 2. The licensee also stated, that the existing Appendix J, Type B and Type C testing programs will not be modified under this request, and will continue to be performed in accordance with Appendix J and the associated TS. Because the leak rate testing requirements (ILRT and LLRTs) 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 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 response to the letters requesting additional information (RAIs) raised during a telephone conference on March 6, 2002, the licensee provided information to explicitly address the NRC staffs five generic issues developed during its ILRT review of other plants.
The staffs evaluation of the licensees response to these generic issues is discussed in the following paragraphs.
In addressing its containment ISI program, the licensee stated that ISIs of the STP containment building (concrete shell and liner) are conducted in accordance with the requirements of 1992 Edition with the 1992 Addenda of ASME Section XI, Subsections IWE and IWL supplemented by licensees commitments. The licensee provided a detailed description of its ISI program including inspection methods and acceptance criteria which meet the requirements of ASME Section XI, Subsection IWE and IWL. The licensee also indicated that there is no change to the schedule for the ISI as a result of the ILRT interval extension.
For the issue related to the application of any augmented examination (required by IWE Table-2500-1, Examination Category E-C) and findings at STP, the licensee stated that based on the results of the previous containment ISI examinations, there are no areas of the containment liner that required augmented examinations according to Subarticle IWE-1240 of the 1992 edition of ASME Section XI, Subsection IWE.
With regard to the issue related to the ISI of seals, gaskets, and pressure retaining bolted connections, the licensee stated that with the approved requests for relief for these areas (Relief Request RR-ENG-IWE-01 proposed to perform Type B test instead of VT-3 visual examinations for seals and gaskets; and RR-ENG-IWE-05 proposed to perform Type B testing and VT-1 visual examination once each inspection interval for the pressure retaining bolted connections), under 10 CFR Part 50, Appendix J, the containment leak-tight integrity will be pressure tested periodically during Type B test. In addition, the licensee stated that the sequence periods following September 9, 2001, will comply with IWE-2412, Inspection Program B. The minimum and maximum examination percentage to be completed for the applicable examination categories are determined based on Table IWE-2412-1. The NRC staff finds that the licensees ISI program applied to seals, gaskets, and pressure retaining bolted connections provides reasonable assurance that the integrity of the containment pressure boundary will be maintained during the period of the ILRT extension.
As for the integrity of stainless steel bellows (two-ply), the licensee stated that STP has only one bellowed penetration, and this bellows is single-ply. Therefore, the concerns of Information Notice 92-20, Inadequate Local Leak Rate Testing, are not applicable to the STP and no ISI of this bellows is required.
Because ILRTs help to identify areas of through-wall degradations of the steel liner when the containment vessel is pressurized, the NRC staff requested that the licensee addresses how the potential leakage 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 that the approach described below was used to determine the change in likelihood, due to extending the ILRT interval, of detecting liner corrosion. This likelihood was then used to determine the resulting change in risk. The following issues are addressed:
Differences between the containment basement and the containment cylinder and dome;
The historical liner flaw likelihood due to concealed corrosion;
The impact of aging;
The liner corrosion leakage dependency on containment pressure; and
The likelihood that visual inspection will be effective at detecting a flaw.
A risk assessment was performed based on the following assumptions:
1.
Basemat concealed liner corrosion due to the lack of identified failure is a half failure.
2.
The success data are limited to those taken since September 1996 when visual inspection was required under 10 CFR 50.55a. Additional success data were not used to limit the impact of corrosion on aging even though inspections were performed prior to this date and there is no evidence that liner corrosion issues were identified.
3.
The likelihood of a liner flaw is assumed to double every five years. This is included to address the increased likelihood of corrosion due to aging.
4.
The likelihood of releasing containment atmosphere to the outside given a liner flaw is a function of pressure inside the containment. Anchored points of 0.1 percent at 20 psia and 100 percent at 150 psia were selected, with immediate failure likelihood determined through logarithmic interpolation.
5.
The likelihood of leakage escape due to crack formation in the basemat region is considered 10 times less likely than for the containment cylinder and dome regions.
6.
The likelihood of visual inspection detection failure is 5 percent given that the flaw is visible. The total detection failure likelihood is 10 percent.
7.
All non-detectable containment over-pressurization leakage events are assumed to be large early release.
Based on its assessment, the licensee concluded that considering increased frequency of visual inspections and the benefit of improved visual inspections under the 1996 inspection criteria, the increase in risk is less than 1.0E-07 for LERF which is within the guideline specified in RG 1.174. Our evaluation summarized under Section 3.1, Risk Impact Assessment, concludes that the additional risk analysis added assurance that increasing the Type A testing interval to 15 years causes a small change in LERF.
The NRC staff finds that, based on the information provided in the TS change request and the information to address the five general issues as discussed above, (1) the structural integrity of the containment vessel is 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 STP 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.
The NRC staff concludes that the revision to TS permitting a one time increase of ILRT interval is supported by the licensees analysis and is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions 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 a surveillance requirement. 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 cummulative 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 50959 dated August 6, 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:
E. Throm M. Snodderly T. Cheng Date: September 17, 2002
July 2002 South Texas, Units 1 & 2 cc:
Mr. Cornelius F. OKeefe Senior Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 910 Bay City, TX 77414 A. Ramirez/C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 Mr. M. T. Hardt Mr. W. C. Gunst City Public Service Board P. O. Box 1771 San Antonio, TX 78296 Mr. C. A. Johnson/R. P. Powers AEP - Central Power and Light Company P. O. Box 289 Mail Code: N5022 Wadsworth, TX 77483 INPO Records Center 700 Galleria Parkway Atlanta, GA 30339-3064 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 D. G. Tees/R. L. Balcom Texas Genco, LP P. O. Box 1700 Houston, TX 77251 Judge, Matagorda County Matagorda County Courthouse 1700 Seventh Street Bay City, TX 77414 A. H. Gutterman, Esq.
Morgan, Lewis & Bockius 1111 Pennsylvania Avenue, NW Washington, DC 20004 Mr. T. J. Jordan, Vice President Engineering & Technical Services STP Nuclear Operating Company P. O. Box 289 Wadsworth, TX 77483 S. M. Head, Manager, Licensing Nuclear Quality & Licensing Department STP Nuclear Operating Company P. O. Box 289, Mail Code: N5014 Wadsworth, TX 77483 Environmental and Natural Resources Policy Director P. O. Box 12428 Austin, TX 78711-3189 Jon C. Wood Matthews & Branscomb 112 East Pecan, Suite 1100 San Antonio, TX 78205 Arthur C. Tate, Director Division of Compliance & Inspection Bureau of Radiation Control Texas Department of Health 1100 West 49th Street Austin, TX 78756 Brian Almon Public Utility Commission William B. Travis Building P. O. Box 13326 1701 North Congress Avenue Austin, TX 78701-3326 Susan M. Jablonski Office of Permitting, Remediation and Registration Texas Natural Resource Conservation Commission MC-122 P.O. Box 13087 Austin, TX 78711-3087 G. R. Bynog, Program Manager/
Chief Inspector Texas Department of Licensing and Regulation Boiler Division P. O. Box 12157, Capitol Station Austin, TX 78711 Mr. Ted Enos 4200 South Hulen Suite 630 Ft. Worth, Texas 76109