ML040280368
| ML040280368 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/08/2004 |
| From: | Rossbach L NRC/NRR/DLPM/LPD3 |
| To: | Crane C Exelon Generation Co |
| Rossbach L, 415-2863, NRR/DLPM | |
| Shared Package | |
| ML040700454 | List: |
| References | |
| TAC MB7861, TAC MB7862 | |
| Download: ML040280368 (16) | |
Text
March 8, 2004 Mr. Christopher M. Crane, President Exelon Nuclear Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING CONTAINMENT LEAKAGE RATE TESTING (TAC NOS. MB7861 AND MB7862)
Dear Mr. Crane:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 220 to Facility Operating License No. DPR-29 and Amendment No. 214 to Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively. The amendments are in response to your application dated February 27, 2003, as supplemented by letters dated April 11 and August 5, 2003.
The amendments revise the Technical Specifications to allow a one-time change in the containment Type A integrated leakage rate test interval that extends the test interval from 10 to 15 years.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Lawrence W. Rossbach, Project Manager, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos.: 50-254 and 50-265
Enclosures:
- 1. Amendment No. 220 to DPR-29
- 2. Amendment No. 214 to DPR-30
- 3. Safety Evaluation cc w/encls: See next page
March 8, 2004 Mr. Christopher M. Crane, President Exelon Nuclear Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING CONTAINMENT LEAKAGE RATE TESTING (TAC NOS. MB7861 AND MB7862)
Dear Mr. Crane:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 220 to Facility Operating License No. DPR-29 and Amendment No. 214 to Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively. The amendments are in response to your application dated February 27, 2003, as supplemented by letters dated April 11 and August 5, 2003.
The amendments revise the Technical Specifications to allow a one-time change in the containment Type A integrated leakage rate test interval that extends the test interval from 10 to 15 years.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Lawrence W. Rossbach, Project Manager, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265
Enclosures:
- 1. Amendment No. 220 to DPR-29
- 2. Amendment No. 214 to DPR-30
- 3. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
PUBLIC PD3-2 r/f WRuland AMendiola LRossbach PCoates OGC ACRS RDennig GHill (4) DTerao MRing, RIII ADAMS Accession Number: ML040700454 (Package) *SE dated 11/24/03 ADAMS Accession Number: ML040280368 (Amendment) **SE dated 4/22/03 ADAMS Accession Number: ML04 (TS) ***see previous concurrence OFFICE PM:LPD3-2 LA:LPD3-2 IROB-A OGC NAME LRossbach PCoates TBoyce*** CMarco***
DATE 02/09/04 02/09/04 02/02/04 03/03/04 OFFICE SC:SPSB SC:EMEB SC:LPD3-2 NAME RDennig* DTerao** DPickett for AMendiola DATE 11/24/03 4/22/03 03/05/04 OFFICIAL RECORD COPY
Quad Cities Nuclear Power Station Units 1 and 2 cc:
Site Vice President - Quad Cities Nuclear Power Station Illinois Emergency Management Exelon Generation Company, LLC Agency 22710 206th Avenue N. Division of Disaster Assistance &
Cordova, IL 61242-9740 Preparedness 110 East Adams Street Quad Cities Nuclear Power Station Plant Manager Springfield, IL 62701-1109 Exelon Generation Company, LLC 22710 206th Avenue N. Document Control Desk - Licensing Cordova, IL 61242-9740 Exelon Generation Company, LLC 4300 Winfield Road Regulatory Assurance Manager - Quad Cities Warrenville, IL 60555 Exelon Generation Company, LLC 22710 206th Avenue N. Senior Vice President - Nuclear Services Cordova, IL 61242-9740 Exelon Generation Company, LLC 4300 Winfield Road Quad Cities Resident Inspectors Office Warrenville, IL 60555 U.S. Nuclear Regulatory Commission 22712 206th Avenue N. Vice President of Operations - Mid-West Cordova, IL 61242 Boiling Water Reactor Exelon Generation Company, LLC David C. Tubbs 4300 Winfield Road MidAmerican Energy Company Warrenville, IL 60555 One River Center Place 106 E. Second, P.O. Box 4350 Vice President - Licensing and Regulatory Davenport, IA 52808-4350 Affairs Exelon Generation Company, LLC Vice President - Law and Regulatory Affairs 4300 Winfield Road MidAmerican Energy Company Warrenville, IL 60555 One River Center Place 106 E. Second Street Director - Licensing and Regulatory P.O. Box 4350 Affairs Davenport, IA 52808 Exelon Generation Company, LLC 4300 Winfield Road Chairman Warrenville, IL 60555 Rock Island County Board of Supervisors 1504 3rd Avenue Associate General Counsel Rock Island County Office Bldg. Exelon Generation Company, LLC Rock Island, IL 61201 4300 Winfield Road Warrenville, IL 60555 Regional Administrator U.S. NRC, Region III Manager Licensing - Dresden, Quad Cities 801 Warrenville Road and Clinton Lisle, IL 60532-4351 Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 220 License No. DPR-29
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 27, 2003, as supplemented by letters dated April 11 and August 5, 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 3.B. of Facility Operating License No. DPR-29 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 220, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA by DPickett for/
Anthony J. Mendiola, Chief, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: March 8, 2004
EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 214 License No. DPR-30
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 27, 2003, as supplemented by letters dated April 11 and August 5, 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 3.B. of Facility Operating License No. DPR-30 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 214, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA by DPickett for/
Anthony J. Mendiola, Chief, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: March 8, 2004
ATTACHMENT TO LICENSE AMENDMENT NOS. 220 AND 214 FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Replace the following page of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain a line in the margin indicating the area of change.
Remove Page Insert Pages 5.5-11 5.5-11 5.5-12
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 220 TO FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 214 TO FACILITY OPERATING LICENSE NO. DPR-30 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By application dated February 27, 2003, as supplemented by letters dated April 11 and August 5, 2003, Exelon Generation Company, LLC (the licensee) requested changes to the Technical Specifications (TS) for the Quad Cities Nuclear Power Station (QCNPS), Units 1 and
- 2. The supplements dated April 11 and August 5, 2003, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 1, 2003 (68 FR 15759).
The proposed changes would revise the TS to allow a one-time change in the containment Type A integrated leakage rate test interval that extends the test interval from 10 to 15 years.
Specifically, the proposed changes would revise TS 5.5.12(a) to add the following after the end of the second sentence:
, as modified by the following exceptions:
- 1. NEI 94-01-1995, Section 9.2.3: The first Unit 1 Type A test performed after the July 23, 1994, Type A test shall be performed no later than July 22, 2009.
- 2. NEI 94-01-1995, Section 9.2.3: The first Unit 2 Type A test performed after the May 17, 1993, Type A test shall be performed no later than May 16, 2008.
2.0 REGULATORY EVALUATION
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. QCNPS TS 5.5.12(a) 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 both Quad Cities units have been successful, so the current interval requirement is 10 years.
The licensee is requesting a change to TS 5.5.12, Primary Containment Leakage Rate Testing Program, which would add two exceptions from the guidelines of RG 1.163 regarding the Type A test interval. Specifically, the proposed TS states that the first Unit 1 Type A test performed after the date of the latest Type A test, July 23, 1994, shall be performed no later than July 22, 2009, and the first Unit 2 Type A test performed after the May 17, 1993, Type A test shall be performed no later than May 16, 2008.
The local leakage rate tests (Type B and Type C tests), including their schedules, are not affected by this request. Also, the vacuum breaker surveillance requirements of TS 3.6.1.7 and TS 3.6.1.8, including their schedules, are not affected by this request.
3.0 TECHNICAL EVALUATION
3.1 Probabilistic Risk Assessment Evaluation 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 February 27, 2003, application for license amendment. Additional analysis and information was provided by the licensee in its letter dated August 5, 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 Quad Cities 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 which would be detectable only by a Type A test, would go 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 pressurized water reactor 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 was completed in 1995, the 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. The staffs assessment of this change in LERF is discussed below in paragraph 2.
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 staffs assessment of the licensees demonstration that the defense-in-depth philosophy is met is discussed below in paragraph 3.
The staff has reviewed the licensees analyses for a change in test frequency from three tests in 10 years to one test in 15 years. The staffs assessment is presented in the following three paragraphs. The assessment in these three paragraphs bound the effect of going from the current one test in 10 years to one test in 15 years because they are based on a more restrictive three tests in 10 years rather than the licensees current requirement of one test in 10 years. The following assessment can be drawn from the analysis associated with extending the test frequency to a one in 15-year test frequency:
- 1. The increase in the total integrated plant risk as a result of a change from a three in 10-year test frequency to a one in 15-year test frequency is estimated to be less than 0.01 person-rem per year. This increase is comparable to that estimated in NUREG-1493, where it was concluded that a reduction in the frequency of tests from three in 10 years
to one in 20 years leads to an imperceptible increase in risk. Therefore, the staff finds that the increase in the total integrated plant risk for the proposed change is considered small.
- 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 1.3 x 10-8 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 [Code],
Section XI, Subsections IWE/IWL). Visual inspections are expected to be effective in detecting large flaws in the visible regions of containment, and this would reduce the impact of the extended test interval on LERF. The licensees risk analysis considered the potential impact of age-related corrosion/degradation in inaccessible areas of the containment liner on the proposed change. The increase in LERF associated with corrosion events is estimated to be less than 1 x 10-8 per year. The 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 0.6 percentage points for the cumulative change of going from a test frequency of three in 10 years to one in 15 years. The 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 staff finds that the increase in predicted risk due to the proposed change is within the acceptance guidelines of RG 1.174 and maintains the defense-in-depth philosophy specified therein. Therefore, the increase in predicted risk is acceptable.
3.2 Mechanical and Structural Integrity Evaluation 3.2.1 Inservice Inspection for Primary Containment Integrity QCNPS utilizes a General Electric boiling water reactor (BWR/3) enclosed within a Mark I-type primary containment. The containment designs of both units are similar; they include steel drywells and suppression chambers with interconnecting vent pipes with bellows, primary containment access penetrations, and other process piping and electrical penetrations. 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 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 containments at the design basis accident (DBA) pressure. The staff requested additional information regarding the containment inservice
inspection (CISI) program. The licensee addressed the staffs concerns in its supplemental responses as discussed in the following paragraphs.
The licensee stated that the Quad Cities CISI program was developed in accordance with Subsections IWE and IWL of the ASME Code Section XI, 1992 Edition with the 1992 Addenda, as modified by NRC final rulemaking to 10 CFR 50.55a published in the Federal Register on August 8, 1996. The Program consists of a CISI Basis Document, Program Plan and CISI Drawings. The Basis Document defines the scope of accessible and inaccessible areas and components. The Program Plan contains information such as the inspection schedules and program relief requests. The CISI drawings provide detailed information regarding type of components, as-built information, etc. The CISI Program is intended to provide controls necessary to assure detection of degradation affecting containment integrity.
The licensee stated that general visual inspections are performed on the Quad Cities containments at intervals described in their program. The first interval of the QCNPS CISI Program is effective from September 30, 1998, through September 9, 2008. The first period of the first CISI interval ended on September 9, 2001, and the CISI inspections for both Units 1 and 2 have been completed. The inspections performed by the licensee included the accessible containment surface areas, including structural attachments and penetrations, pressure-retaining bolting, and Class MC supports. The second period of the first CISI interval is scheduled to end on September 9, 2005.
The staff has concluded that the ISI program at Quad Cities has been established and is being conducted in accordance with Section XI of the ASME Code and that the methods and schedule employed are acceptable.
3.2.2 Implementing ASME Code Section XI, Subsection IWE-1240 at Quad Cities -
Augmented Examination Certain areas of containment may be more susceptible than others to corrosion. To address this potential problem, Subsection IWE-1240 of Section XI of the ASME Code (IWE-1240) requires licensees to identify any areas of the containment that might require augmented examination. Compliance with IWE-1240 is required by 10 CFR Part 50.55a(b)(2). In its February 27, 2003, submittal, the licensee stated that the Quad Cities containment ISI program has not identified any surface areas classified as augmented examination areas in accordance with Table IWE-2500-1, Examination Category E-C.
The licensee states that for CISI inspections performed at QCNPS, various indications were observed, documented, evaluated, and determined to be acceptable . . . With the exception of the degraded drywell moisture barrier on Units 1 and 2 . . . " Corrective actions were performed consisting of removal of the moisture barriers, examination of base metal surfaces and installation of new barriers. The inspection sample was expanded and no other areas were found degraded. This information was reported to NRC in letters dated June 3, 2002 and February 24, 2003.
The staff has concluded that the licensee has adequately applied the requirements of the ASME Code in making its determination that no areas of the containment require augmented inspection.
3.2.3 Schedule for Examination and Testing of Seals, Gaskets and Bolts Providing Containment Pressure Boundary Integrity Type A containment leak tests evaluate the integrity of the entire containment; however, the most likely source of a containment leak is through a penetration. To address this under the 10 CFR 50 Appendix J, Option B program, those Type B penetrations that use resilient seals, gaskets, etc., are tested within the guidelines provided by Option B and RG 1.163. The QCNPS testing program is set up so that 100 percent of all components are tested during each 10-year interval. Those components that do not fall under extended test frequencies are tested at least once every 30 months. Components that are not under extended test frequencies are either penetrations that are disassembled and reassembled each outage or are components that have not demonstrated acceptable performance history per the Primary Containment Leakage Rate Program. Components that are disassembled and reassembled during an outage receive an as-found test prior to any work and an as-left test after all work is completed.
The Type B and C testing requirements will not be changed as a result of the proposed extended ILRT interval.
The staff has evaluated the licensees application for amendment and has concluded that the licensees schedule for the examination of seals and gaskets and for the examination and testing of bolted connections is acceptable because it meets the requirements of Subsection IWE-2400 of Section XI of the ASME Code and 10 CFR 50.55a.
3.2.4 Integrity of Stainless Steel Bellows The staff has determined that two-ply stainless steel bellows can be susceptible to trans-granular stress corrosion cracking, and that the leakage through them may not be detectable by Type B testing (see NRC Information Notice (IN) 92-20, Inadequate Local Leak Rate Testing).
NRC had previously reviewed the Quad Cities bellows testing program as documented in a letter from B A Boger (USNRC) to T J Kovach, Exemption from the Testing Requirements of Appendix J to 10 CFR Part 50 for Dresden and Quad Cities Nuclear Power Stations dated February 6, 1992, and letter from R M Pulsifer (USNRC) to D L Farrar, Revision to Exemption from the Testing Requirements of Appendix J to 10 CFR Part 50 for Dresden and Quad Cities Nuclear Power Stations dated February 9, 1995. NRC concluded that the current testing program will detect bellows assemblies with significant flaws and result in the replacement of flawed assemblies within one operating cycle, or be tested with a Type B test to ensure that license limits are met during which period there is reasonable assurance that the bellows assemblies will not suffer excessive degradation.
The proposed change to the CISI test interval does not impact the bellows testing program.
The previous NRC conclusions regarding the adequacy of the Quad Cities bellows testing program are not changed by the proposed change in the CISI test interval.
3.2.5 Degradation of Inaccessible Side of the Containment Steel Shell Under its IWE program, the licensee evaluates the acceptability of inaccessible areas of the containment steel shell if conditions exist in the accessible areas that could indicate the presence of, or result in, degradation to the inaccessible area. Section 50.55a(b)(2)(viii)(E) of 10 CFR requires that this evaluation include the description of the type, estimated extent, and
cause of the degradation as well as the evaluation results of each area and a description of necessary corrective actions. The staff requested additional information regarding the inaccessible side of the containment steel shell. The licensee addressed the staffs concerns in its supplemental responses as discussed in the following paragraph.
The licensees April 11, 2003, submittal provided drawings showing the inaccessible areas of containment and the areas of augmented examinations. The inaccessible areas comprise a very small portion of the containment surface area. The critical areas such as penetrations and moisture barrier interfaces are encompassed by the QCNPS CISI Program. In addition, during power operation, the primary containment is inerted and maintained at a positive pressure.
Drywell and suppression chamber pressures are continuously recorded in the control rooms and monitored via alarm and shiftly surveillances. Continuous monitoring of slight positive pressure ensures that areas of containment degradation are detected before they could result in large leakage rates.
The staff finds that the licensee is employing an IWE program that requires it to estimate any potential degradation of inaccessible areas of the containments. In addition, the licensee performed risk analysis to consider the potential risk impact of corrosion in inaccessible areas of the containment shell. The evaluation concluded that the accessible areas examined are representative of the inaccessible areas. As described in Section 3.1.2 of this safety evaluation, the risk impact associated with an extension of the test interval is small, is consistent with the acceptance guidelines of RG 1.174, and is acceptable.
3.2.6 Mechanical and Structural Integrity Conclusion Based on the assessment described in the preceding paragraphs, the 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 no augmented inspection is required; (2) the integrity of the penetrations, containment isolation valves, and bellows is periodically verified through Type B and Type C tests as required by 10 CFR Part 50, Appendix J, and Quad Cities TS; (3) the licensee is aware of the potential problems associated with the degradation of two-ply stainless steel bellows and is taking prudent action, and that a Type A leak test is not likely to detect problems with these bellows that would otherwise go undetected; and (4) the licensee is employing an IWE program that requires it to estimate any potential degradation of inaccessible areas of the containments; furthermore, the licensee made conservative risk assumptions regarding such degradation. The staff concludes, therefore, that the licensees ISI program, as currently implemented, supports the one-time extension of the ILRT from 10 to 15 years.
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 Quad Cities Units 1 and 2 primary containments. Thus, granting a one-time extension of performing the integrated leak rate testing (ILRT) as proposed by the licensee in Section 5.5.12(a) of the proposed technical specification (TS) change request is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (68 FR 15759). 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, SPSB-C R. Palla, SPSB-A S. M. Unikewicz, EMEB Date: March 8, 2004