ML022680754
| ML022680754 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/24/2002 |
| From: | Alexion T NRC/NRR/DLPM/LPD4 |
| To: | Anderson C Entergy Operations |
| Alexion T W,NRR/DLPM,415-1326 | |
| References | |
| TAC MB3941 | |
| Download: ML022680754 (12) | |
Text
September 24, 2002 Mr. Craig G. Anderson Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:
ONE TIME EXTENSION OF THE INTEGRATED LEAK RATE TEST (TAC NO. MB3941)
Dear Mr. Anderson:
The Commission has issued the enclosed Amendment No. 219 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated January 31, 2002, as supplemented by letter dated September 9, 2002.
The amendment changes administrative Technical Specification 5.5.16 regarding the Containment Integrated Leak Rate Testing (ILRT) to allow a one-time extension of the interval (to 15 years) for performance of the next ILRT.
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/
Thomas W. Alexion, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-313
Enclosures:
- 1. Amendment No. 219 to DPR-51
- 2. Safety Evaluation cc w/encls: See next page
September 24, 2002 Mr. Craig G. Anderson Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:
ONE TIME EXTENSION OF THE INTEGRATED LEAK RATE TEST (TAC NO. MB3941)
Dear Mr. Anderson:
The Commission has issued the enclosed Amendment No. 219 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated January 31, 2002, as supplemented by letter dated September 9, 2002.
The amendment changes administrative Technical Specification 5.5.16 regarding the Containment Integrated Leak Rate Testing (ILRT) to allow a one-time extension of the interval (to 15 years) for performance of the next ILRT.
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/
Thomas W. Alexion, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-313
Enclosures:
- 1. Amendment No. 219 to DPR-51
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
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MSnodderly Accession No.: ML022680754
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AHodgdon RGramm DATE 09/18/02 9/19/02 09/18/02 9/18/02 09/17/02 9/20/02 9/25/02 OFFICIAL RECORD COPY
ENTERGY OPERATIONS INC.
DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 219 Renewed License No. DPR-51 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated January 31, 2002, as supplemented by letter dated September 9, 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, 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 Renewed Facility Operating License No. DPR-51 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 219, are hereby incorporated in the renewed license.
EOI shall operate the facility in accordance with the Technical Specifications.
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 25, 2002
ATTACHMENT TO LICENSE AMENDMENT NO. 219 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 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 5.0-25 5.0-25
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT NO. 1 DOCKET NO. 50-313
1.0 INTRODUCTION
By letter dated January 31, 2002, as supplemented by letter dated September 9, 2002, Entergy Operations, Inc. (the licensee), submitted a request for changes to the Arkansas Nuclear One, Unit No. 1 (ANO-1), Technical Specifications (TSs). The requested changes would revise administrative TS 5.5.16 regarding the Containment Integrated Leak Rate Testing (ILRT) to allow a one-time extension of the interval (to 15 years) for performance of the next ILRT.
The September 9, 2002, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice (67 FR 7417, dated February 19, 2002) or the original no significant hazards consideration determination.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR Part 50, Appendix J, Option B, a Type A test is required to be conducted at a periodic interval based on historical performance of the overall containment system. ANO-1 TS 5.5.16 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) 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.
As the most recent two Type A tests at ANO-1 have been successful, the current interval requirement is 10 years.
The licensee is requesting additions to TS 5.5.16, Reactor Building Leakage Rate Testing Program, which would indicate that it is permitted to take an exception from the guidelines of RG 1.163 regarding the Type A test interval. Specifically, the proposed TS states that the next ANO-1 Type A test performed after the April 16, 1992, Type A test shall be performed no later than April 15, 2007.
3.0 TECHNICAL EVALUATION
3.1 Containment Systems The licensee has performed a risk impact assessment of extending the Type A test interval to 15 years. The assessment was provided to the staff in the January 31, 2002, application for license amendment. Additional analysis and information were provided by the licensee in a supplemental letter dated September 9, 2002. 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 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 U. S. Nuclear Regulatory Commissions (NRC) 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.
Relaxing the test frequency from 3 in 10 years to 1 in 10 years increased 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), NUREG-1493 estimated 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 (PWR) and boiling water reactor 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 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 0.04 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 0.09 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.
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 1.5 x 10-7/year. However, there is some likelihood that the undetected flaw in the containment liner, estimated as part of the Class 3b frequency, would be detected as part of the IWE visual examination process of the containment liner. The initial IWE examination of the liner was completed in 2001. An additional liner examination is scheduled to occur no later than 2004, with follow-on examinations occurring once every three or four years in accordance with American Society of Mechanical Engineers (ASME) requirements. Eighty percent of the inner containment liner can be visually inspected. Assuming the visual inspections are 90 percent effective in detecting large flaws in the visible regions of the containment, the increase in LERF would go from 1.5 x 10-7/year to 4.2 x 10-8/year. Therefore, increasing the Type A interval to 15 years is considered to be a very small change in LERF when using the guidelines of RG 1.174.
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 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 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 4.6 x 10-8/year. This additional risk analysis provides added assurance that increasing the Type A interval to 15 years is a very small change in LERF.
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 licensee estimates the change in the conditional containment failure probability to be an increase of 0.1 percent for the proposed change and 0.3 percent for the cumulative change of going from a test interval of 3 in 10 years to 1 in 15 years. The 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 Integrity ANO-1 is a Babcock & Wilcox-designed PWR. The primary reactor containment is a large, pre-stressed, concrete, vertical right cylinder with a flat base and a shallow spherical 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 essentially to verify the leak-tight characteristics of the containment structure at the design basis accident pressure. As stated in the January 31, 2002, application, the licensee has performed five ILRTs at ANO-1 during the period of its operating license. The two most recent ILRTs were performed in 1988 and 1992. These two tests were successful. On this basis, the current interval requirement is 10 years and the next ILRT is scheduled during the outage in October 2002. 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 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. The licensee also provided information to explicitly address several issues identified by the staff during its review of other nuclear plants. In the September 9, 2002, supplemental letter, the licensee provided additional information to verify its response to the first and third issues. The staffs evaluation of the information provided by the licensee to address these questions is discussed in the following paragraphs.
In addressing the issue related to its containment ISI program, the licensee stated (in the January 31, 2002, application and the September 9, 2002, supplemental letter) that ISI methods at ANO-1 are described in plant procedures in which the containment ISI portion is developed in accordance with the requirements of 1992 Edition with 1992 Addenda of ASME Boiler and Pressure Vessel Code (Code),Section XI, Subsections IWE and IWL, supplemented by licensees commitments. Any signs of degradation found as a result of general visual examination in accordance with the plant procedures that exceed the Code requirements are documented in the corrective action program and dispositioned in accordance with the ASME Code requirements. As a result of the implementation of its ISI program, the licensee has not identified any augmented examinations required for ANO-1. From the discussion above, the staff finds that the licensees ISI program, including areas of augmented inspections, will provide assurance that the containment structural integrity and leak-tight integrity will be maintained during the extended ILRT period.
With regard to the issue related to the ISI of seals, gaskets, and pressure-retaining bolted connections, the licensee stated (in the January 31, 2002, application) that with the approved requests for relief in these areas, the containment penetrations will be pressure tested periodically using a Type B test under Option B of Appendix J. As stated in the approved relief requests, the alternative examinations of Appendix J testing will be performed at least once during each containment inspection interval. (The Type B test interval for penetrations is a maximum of 10 years. The test frequency for air-locks, door seals, and penetrations with resilient seals is once per 30 months.) Thus, the extension requested for Type A testing does not affect the interval of these alternative examinations because they will be performed at least once in each 10-year inspection interval. On this basis, the staff finds that the schedule for examination of the seals, gaskets, and bolts will continue to provide reasonable assurance that the integrity of the containment pressure boundary will be maintained.
As for the issue related to the integrity of stainless steel bellows, the licensee stated that ANO-1 does not have bellows in any of the containment penetration configurations. Therefore, the concerns of Information Notice 92-20, "Inadequate Local Leak Rate Testing," are not applicable to ANO-1.
On the basis of its review of the information provided by the licensee in its TS amendment request and its response to the staffs questions, the staff finds that (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, and (2) the integrity of the penetrations and containment isolation valves are periodically verified through Type B and Type C tests as required by 10 CFR Part 50, Appendix J. In addition, the system pressure tests for containment pressure boundary (i.e., Appendix J tests, as applicable) are required to be performed following repair and replacement activities, if any, in accordance with Article IWE-5000 of the ASME Code,Section XI. 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 Evaluation Summary Based on the foregoing evaluation, the staff finds that the interval until the next Type A tests at ANO-1 may be extended to 15 years, and that the proposed changes to TS 5.5.16 are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The 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 and changes surveillance requirements. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (67 FR 7417, dated February 19, 2002). Accordingly, the 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 the amendment.
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 amendment 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: September 24, 2002
March 2001 Arkansas Nuclear One cc:
Executive Vice President
& Chief Operating Officer Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Director, Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 Mr. Mike Schoppman Framatome ANP, Richland, Inc.
Suite 705 1911 North Fort Myer Drive Rosslyn, VA 22209 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 310 London, AR 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 County Judge of Pope County Pope County Courthouse Russellville, AR 72801 Vice President, Operations Support Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205