ML040300152
| ML040300152 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 01/28/2004 |
| From: | Bhalchandra Vaidya NRC/NRR/DLPM/LPD4 |
| To: | Gerald Williams Entergy Operations |
| vaidya B, NRR/DLPM, 415-3308 | |
| References | |
| TAC MB8940 | |
| Download: ML040300152 (18) | |
Text
January 28, 2004 Mr. George A. Williams Vice President, Operations GGNS Entergy Operations, Inc.
P. O. Box 756 Port Gibson, MS 39150
SUBJECT:
GRAND GULF NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: ONE-TIME EXTENSION OF THE INTEGRATED LEAK RATE TEST AND DRYWELL BYPASS TEST INTERVAL (TAC NO. MB8940)
Dear Mr. Williams:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 164 to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1. This amendment revises the Technical Specifications (TSs) in response to your application dated May 12, 2003, as supplemented by letter dated October 29, 2003.
The amendment changes the administrative TS 5.5.12 regarding containment integrated leakage rate testing (ILRT) and TS 3.6.5.1.1 regarding drywell bypass leak rate testing (DWBT). The change would allow for a one-time extension of the interval from 10 to 15 years for performance of the next ILRT and DWBT.
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/
Bhalchandra Vaidya, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-416
Enclosures:
- 1. Amendment No.164 to NPF-29
- 2. Safety Evaluation cc w/encls: See next page
January 28, 2004 Mr. George A. Williams Vice President, Operations GGNS Entergy Operations, Inc.
P. O. Box 756 Port Gibson, MS 39150
SUBJECT:
GRAND GULF NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: ONE-TIME EXTENSION OF THE INTEGRATED LEAK RATE TEST AND DRYWELL BYPASS TEST INTERVAL (TAC NO. MB8940)
Dear Mr. Williams:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 164 to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1. This amendment revises the Technical Specifications (TSs) in response to your application dated May 12, 2003, as supplemented by letter dated October 29, 2003.
The amendment changes the administrative TS 5.5.12 regarding containment integrated leakage rate testing (ILRT) and TS 3.6.5.1.1 regarding drywell bypass leak rate testing (DWBT). The change would allow for a one-time extension of the interval from 10 to 15 years for performance of the next ILRT and DWBT.
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/
Bhalchandra Vaidya, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-416 DISTRIBUTION:
PUBLIC KMartin PDIV-1 Reading RPalla RidsNrrDlpmPdiv (HBerkow)
JPulsipher
Enclosures:
RidsNrrDlpmPdivLpdiv1 (RGramm) RidsRgn4MailCenter (AHowell)
- 1. Amendment No. 164 to NPF-29 RidsNrrLADJohnson RidsAcrsAcnwMailCenter
- 2. Safety Evaluation RidsNrrPMBVaidya MRubin RidsOgcRp TBoyce cc w/encls: See next page GHill (2)
DTerao RDennig
- No significant change from SE Input Accession No.:ML040300152
MRubin*
RDennig*
TBoyce**
KKannler**/nlo RGramm**
DATE 1/27/04 1/27/04 10/8/03 11/20/03 11/20/03 1/16/04 1/14/04 1/27/04 DOCUMENT NAME: C:\\ORPCheckout\\FileNET\\ML040300152.wpd OFFICIAL RECORD COPY
ENTERGY OPERATIONS, INC.
SYSTEM ENERGY RESOURCES, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION ENTERGY MISSISSIPPI, INC.
DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 164 License No. NPF-29 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated May 12, 2003, as supplemented by letter dated October 29, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 164, are hereby incorporated into this license.
Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 28, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 164 FACILITY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 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 3.6-53 3.6-53 3.6-53a 5.0-16 5.0-16
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 164 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC., ET AL.
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416
1.0 INTRODUCTION
By application dated May 12, 2003, as supplemented by letter dated October 29, 2003, Entergy Operations, Inc., et al. (the licensee), requested changes to the Technical Specifications (TSs) for Grand Gulf Nuclear Station, Unit 1 (GGNS). The supplemental letter dated October 29, 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 June 10, 2003 (68 FR 34666).
The proposed changes will revise the Operating License to change Administrative TS 5.5.12 regarding containment integrated leak rate testing (ILRT) and TS 3.6.5.1.1 regarding drywell bypass leak rate testing (DWBT). The change would add an exception to the commitment to implement containment ILRT program in accordance with the Safety Evaluation (SE) issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087), as modified by the SE issued for Amendment No. 135 to the Operating License. The change would allow for a one-time extension of the interval from 10 to 15 years for performance of the next ILRT and the DWBT.
Specifically, GGNS proposes to revise TS 5.5.12 by adding to the end of the second sentence the following:
, except that the next Type A test performed after the November 24, 1993 Type A test shall be performed no later than November 23, 2008.
GGNS also proposes to revise TS 3.6.5.1.1 by adding an exception to the Frequency requirement of 120 months that states:
, except that the next drywell bypass leak rate test performed after the November 24, 1993 test shall be performed no later than November 23, 2008.
The proposed changes would represent a one-time deferral of the ILRT and the DWBT by up to five additional years.
2.0 REGULATORY EVALUATION
2.1 Type A Test Interval Extension Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix J, was revised in 1995 by the addition of Option B, Performance-Based Requirements, to the original requirements, which were then designated as Option A, Prescriptive Requirements. Option B requires that a Type A test be conducted at a periodic interval based on historical performance of the overall containment system. As part of the development of Option B, the NRC also developed Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, to specify a method acceptable to the NRC for complying with Option B.
Option B requires that the RG or other implementation document used by a licensee to develop a performance-based leakage rate testing program be included, by general reference, in the plant TSs. However, the licensee does not use the RG. Instead, the licensee references the NRC staff SE that was the basis for an earlier exemption from Appendix J, issued to GGNS on April 26, 1995. The exemption expired in 1998, so License Amendment No. 135 was issued to implement Option B at GGNS, using for guidance the SE written for the exemption and also the SE written for Amendment No. 135.
Accordingly, GGNS TS 5.5.12 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 SE issued by the Office of Nuclear Reactor Regulation dated April 26, 1995, as modified by the SE issued for Amendment No. 135 to the Operating License.
A Type A test is an overall (integrated) leakage rate test of the containment structure. The TS provisions require an initial test interval of approximately 3.33 years (3 tests in 10 years), but allow 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 GGNS have been successful, so the current interval requirement is 10 years.
The licensee is requesting an addition to TS 5.5.12, which would add an exception from the normal requirements regarding the Type A test interval. Specifically, the proposed TS states that the next Type A test performed after the November 24, 1993, Type A test (the date of the latest test) shall be performed no later than November 23, 2008.
The local leakage rate tests (Type B and Type C tests), including their schedules, are not affected by this request.
2.2 Drywell Bypass Leak Rate Test Interval Amendment No. 126 to the operating license for GGNS, was issued on August 1, 1996 (ADAMS Accession No. ML021480466). The amendment required that the DWBT be conducted at least once every 10 years on a performance-based frequency (the DWBT frequency had been once per 18 months). In the event that a test is performed with the bypass leakage greater than its limit, the test frequency becomes once every 48 months. Following two consecutive tests with bypass leakage greater than its limit, the test frequency is once every 24 months until two consecutive tests are less than or equal to the bypass leakage limit.
The last DWBT was successfully conducted in November 1993.
One purpose of the change was to make the DWBT frequency the same as the Appendix J Type A test frequency, because the two tests share test equipment and system lineups. Thus, the licensee has accompanied its request for a one-time Type A test interval extension to 15 years with a request for a one-time extension of the DWBT interval to 15 years.
In the NRC staffs SE for the amendment cited above, the staffs acceptance of the proposed 10-year test interval was based on the licensees capability to assure that the likelihood of significant bypass leakage is acceptably low. This was based on the design of the drywell and its penetrations, the TSs and administrative controls in place, and the results of previous leakage tests, as well as deterministic and risk calculations. The NRC staff gave considerable weight in its evaluation of the licensees commitment to assess the drywell leakage at least once per cycle to assure that the drywell remains operable.
During a small break loss-of-coolant accident (LOCA), potential leak paths between the drywell and containment airspace could result in excessive containment pressure, since the steam flow into the airspace would bypass the vapor suppression capabilities of the pool. The potential leakage paths between the drywell and the containment are: 1) piping and electrical penetrations; 2) the drywell equipment hatch; and 3) the drywell personnel air lock. The staff found that 1) the electrical penetrations are unlikely to leak significantly, and the drywell bypass leak rate assumed in design calculations, is so large that, even if the valves in many of the pipes were left open, the design limit would not be exceeded; and 2) both the equipment hatch and drywell air lock have double compression seals and are periodically leak tested.
Regarding testing history, the NRC staff found that the maximum observed leakage rate value of bypass leakage had been about seven percent of the design limit of 35,000 standard cubic feet per minute (scfm). Ten drywell bypass leakage rate tests have been performed at GGNS and there have been no test failures.
The NRC staff also reviewed the risk associated with the increase in the test interval from 18 months to 10 years, and found that there was only a small effect on risk. The NRC staff considered the increase in risk due to the increase in the test interval to be acceptable.
The NRC staff had also requested that the licensee propose a method of monitoring the drywell for significant leakage during operation. The licensee committed to assess drywell leaktightness at least once per operating cycle. The assessment is actually performed every quarter by running the drywell purge compressors to pressurize the drywell. The drywell purge compressors are part of an engineered safety system which forces air from the primary containment into the drywell. The compressors are required to be operated at least 15 minutes every quarter and an assessment is performed in approximately 1000 scfm. The assessment considers whether a compressor is capable of increasing the pressure in drywell. The NRC staff concluded that the proposed method provided reasonable assurance that the TS value of drywell bypass leakage would not be exceeded. The regular monitoring of drywell leakage helps to ensure that there is no significant undetected degradation of the drywell.
The NRC staffs SE for License Amendment No. 126 concluded that the proposal to change the DWBT interval from 18 months to 10 years (given good performance) was acceptable based on the low increase in risk, the large margin for leakage, the controls in the TSs to assure closure of the penetrations, if required, and the licensees commitment to assess the drywell bypass leakage, thereby assuring operability at least once every operating cycle.
The licensees current request is to modify the DWBT surveillance requirement, TS 3.6.5.1.1, to say that the next drywell bypass leak rate test performed after the November 24, 1993, test shall be performed no later than November 23, 2008.
The NRC staff finds that the licensee in Attachment 1, Section 10 of its May 12, 2003, submittal, as supplemented, identified the applicable regulatory requirements. The regulatory requirements and guidance documents on which the staff based its acceptance are: Appendix J of 10 CFR Part 50, 10 CFR 50.55a(g), RG 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, to specify a method acceptable to the NRC staff for complying with Option B of Appendix J, RG 1.174 An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, 10 CFR 50.90, "Application for amendment of license or construction permit," 10 CFR 50.91, "Notice for public comment; State consultation," and 10 CFR 50.92, "Issuance of amendment."
3.0 TECHNICAL EVALUATION
The NRC staff has reviewed the licensee's regulatory and technical analyses, in support of its proposed license amendment, which are described in Attachment 1 of the licensee's May 12, 2003, submittal, as supplemented. The detailed evaluation below supports the conclusion 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 Commission's 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.
3.1 Inservice Inspection (ISI) for Primary Containment Integrity GGNS is designed with a General Electric Company boiling water reactor (BWR) enclosed by a Mark III type containment. The drywell is enclosed within the primary containment and is designed to divert the energy released during a design-basis, large-break LOCA. The drywell communicates with the primary containment through a series of horizontal vents in the drywell wall. The vents are covered both inside and outside the drywell by water from the annular shaped suppression pool. The pool forms a seal between the drywell and the primary containment. The drywell contains the reactor coolant system and other high energy piping systems. The GGNS Updated Final Analysis Report, Section 6.2 describes the primary containment in detail.
Several tests are performed to ensure the integrity of the containment/drywell function, including both the ILRT and the DWBT. Testing frequencies for the ILRT are performance-based, as allowed by 10 CFR Part 50, Appendix J, Option B. Option B requires that RG 1.163 or another implementation document used by a licensee to develop a performance-based leakage rate testing program be incorporated by a general reference in the plant TSs.
Structural degradation of containment is a gradual process that occurs due to the effects of pressure, temperature, radiation, chemical, or other factors. Such effects are identified and corrected when the containment is periodically inspected to verify its structural integrity, in accordance with the requirements of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (Code),Section XI, Subsections IWE and IWL.
On August 1, 1996, the NRC issued an amendment to the GGNS Facility Operating License that revised the TSs to allow performance-based drywell bypass leakage surveillance tests.
The NRC requested that GGNS monitor the drywell for significant leakage during operation.
GGNS committed to assess the leaktightness of the drywell at least once each operating cycle.
The assessment is actually performed every quarter by running the drywell purge compressors to pressurize the drywell. The drywell purge compressors are part of an engineered safety system which forces air from the primary containment into the drywell. The compressors are required to be operated at least 15 minutes every quarter and an assessment is performed to determine whether the compressors are capable of increasing the pressure in drywell. The NRC staff concluded in an August 1996 SE that the proposed method provided reasonable assurance that the TS value of drywell bypass leakage would not be exceeded. The regular monitoring of drywell leakage helps to ensure that there is no significant undetected degradation of the drywell.
The licensee provided information related to the ISI of the containment and discussed potential areas of weakness in the containment that may not be apparent in the risk assessment. These topics are discussed in the following paragraphs.
The licensee stated that general visual inspections are performed for accessible interior and exterior surfaces of the containment system of GGNS as required by the 10 CFR Part 50, Appendix J Program. These inspections are performed for structural problems that may affect either the containment structural leakage integrity or that might affect the performance of the ILRT. These examinations are currently required to be completed before the ILRT and during two other refueling outages before the next ILRT on a ten-year frequency. These requirements will not be changed through approval of this license amendment request.
Containment integrity is also verified through periodic ISIs conducted in accordance with the requirements of the ASME Code,Section XI. More specifically, the ASME Code,Section XI, Subsection IWE provides the rules and requirements for ISI of Class MC (metal containment) pressure-retaining components and their integral attachments in light-water cooled plants.
Subsection IWL provides the rules and requirements for ISI of Class CC (concrete containment) components. Furthermore, NRC regulations in 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct a general visual inspection of the containment in accordance with ASME Code,Section XI during each of the three inspection periods during the ten-year interval.
In addition, 10 CFR Part 50, Appendix J, Type B local leak rate tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are not affected by the proposed change to the Type A frequency.
The Containment ISI (CII) Program at GGNS is described in detail in CEP-CII-005, Grand Gulf Nuclear Station Containment Inservice Inspection (CII) Program Plan. This plan was developed in accordance with the requirements of the ASME Code,Section XI, Subsections IWE and IWL, 1992 edition with 1992 addenda, as modified by 10 CFR 50.55a.
The NRC staff has concluded that the ISI program at GGNS has been established and is being conducted in accordance with Section XI of the ASME Code, and that the methods and schedule employed meet the requirements of 10 CFR 50.55a and are, thus, acceptable.
Certain areas of containment may be more susceptible than others to degradation, such as corrosion. To address this potential problem, IWE-1240 requires licensees to identify any areas of the containment that might require augmented examination. The GGNS CII program currently requires augmented examinations in accordance with IWE-1240 for five areas as discussed in the following paragraphs.
Area EC-01 is a 2 x 2 area of the containment liner located directly below containment penetration #38. This area was discovered on April 21, 1998, and reported on condition report CR-GGNS-1998-1235 during a preliminary walk-down of the containment. The liner in this area was found to be heavily corroded and covered with a whitish deposit which appeared to have leached from the chill water supply piping insulation. Engineering evaluation determined that the area continued to perform its design function. The coating in this area had been damaged and was repaired. Subsequent Visual Testing (VT)-1 and Ultrasonic Testing (UT) examinations of the area occurred in 2001. These examinations showed no additional corrosion of the area.
This area was accepted by examination in accordance with the provisions of IWE-3122.1. This area remains an augmented inspection area due to the potential for wetting from condensation on the chilled water supply line running through penetration #38.
Area EC-02 covers all the class MC components on the auxiliary building side of penetration
- 38. This area was discovered during the initial IWE examinations on March 20, 2001, and reported on condition report CR-GGNS-2001-0467. UT examinations showed that there was no material loss (wall thinning) of the associated liner or piping materials affected, and the areas were accepted by examination in accordance with the provisions of IWE-3122.1. These examinations are documented in QIPN 0370-000-2001 and NDEN 0371-001-2001. The area was added as an area requiring augmented examination due to the potential for accelerated corrosion because of condensation on the chilled water supply line running through penetration
- 38.
Areas EC-03, EC-04, and EC-05 indicate three separate places where rust streaking was observed coming from behind expansion foam placed between the deck at elevation 1354" and the containment liner. These areas were examined using VT-1 and UT techniques. The initially inaccessible areas were accepted by examination in accordance with the provisions of IWE-3122.1. This examination and acceptance are documented in QIPN 0520-000-2001, with supplementary UT data collected and reported on NDEN 0521-001-2001. The areas covered in EC-03 through EC-05 may be subjected to accelerated corrosion due to the potential for these areas to become wetted. As a result, these are treated as augmented inspection areas.
The licensee developed a formal technical position to evaluate the results of the inspections that are conducted in accordance with the requirements of ASME Code,Section XI, Subsections IWE and IWL, 1992 edition with 1992 addenda, as modified by 10 CFR 50.55a.
The NRC staff concludes that the licensee has adequately applied the requirements of the ASME Code in making their determination of areas in the containment that require augmented inspection and the subsequent examinations of these areas.
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 Part 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 NRC authorized alternatives to the requirements of ASME Code,Section XI pursuant to 10 CFR 50.55a(a)(3) for CIIs for Arkansas Nuclear One, Units 1 and 2, GGNS, River Bend Station, and Waterford Steam Electric Station, Unit 3, on January 13, 2000. The authorizations allowed the use of existing 10 CFR Part 50, Appendix J, Type B testing as a verification method for ensuring containment integrity in lieu of disassembling the subject components for the sole purpose of examination. As stated in the alternatives (ISI Relief Request IWE-03 for seals and gaskets and ISI Relief Request IWE-02 for examination and testing of bolt torque and tensioning), the alternative examinations using Appendix J, Type B testing will be performed at least once during each CII interval. Thus, the proposed interval extension for the Type A test (ILRT) does not affect the frequency of these alternative examinations because they will be performed once in each ten-year inspection interval.
The Nuclear Energy Institute (NEI) provides guidelines in NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J," which describes the Type B testing frequencies in paragraphs 10.2.1 and 10.2.2. The extended test interval for Type B penetrations (except containment airlocks) is up to a maximum of once per 120 months.
The airlocks and penetrations with resilient seals are to be tested at a frequency of once per 30 months.
The NRC staff has concluded that the licensee has established a schedule for the examination of seals and gaskets and for the examination and testing of bolted connections that is consistent with the authorized relief for alternatives to the requirements in the applicable sections of both the Code and the regulations and is, therefore, acceptable.
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).
GGNS has only one bellows assembly that could be subject to the failure mechanism described in the IN. This is an expansion bellows assembly associated with the horizontal fuel transfer tube. GGNS conducted several tests to verify the adequacy of the local leak rate testing for this bellows assembly. The tests have always demonstrated zero leakage. No indications were found of cracks or gouges and the bellows assembly was described as being in good condition.
Tests were done to verify that air could pass through each of the bellow's halves from one test connection to the other and that there were no obstructions to the flow. During refueling outage 6 (1993), tests were done to confirm that the bellow's annulus is vented to the containment atmosphere. This ensured that the annulus was being subjected to ILRT test pressure (about 12 psig) and was the fourth ILRT with all results being well below the acceptance limits. In addition, a visual inspection using liquid leak detection fluid was done of the exterior of the bellows assembly while attempting to pressurize the bellow with air. All of this testing provides a high degree of confidence that the test methods currently being used are adequate to detect leakage across the bellow's assembly. It is also worthwhile to note that the bellow is not subjected to large or rapid temperature changes or other operationally induced stresses.
The staff has reviewed the licensee's submittal and has determined that, with respect to the leakage of two-ply bellows, the licensee is aware of the problem and is taking appropriate action to assure that any degradation to the bellow is detected. Under these circumstances, the staff does not believe that a Type A leak test is likely to identify any bellow leakage that would not be otherwise identified through inspections required by the Code.
There are inaccessible areas of the GGNS containment, including parts of the inner and outer surfaces, covered and blocked by concrete. There are no programs that monitor the condition of the inaccessible areas of the containment liner plate directly. When there is an indication of potential degradation of inaccessible areas of the containment liner plate, this finding is evaluated and appropriate actions are taken to assure the adequacy and integrity of the containment. Condition reports CR-GGN-2001-0120 and CR-GGN-2001-0309 document cases where indications of potential problems within inaccessible areas were noted. In both cases, foam material was removed to gain access to the areas in question for direct examination. In both cases, the initially suspect areas contained light corrosion that was within the acceptance standards for the containment. The areas were ultimately accepted by examination.
Portions of the GGNS liner are submerged in the suppression pool. The submerged surfaces are accessible and are examined at the end of the CII interval in accordance with ASME Code,Section XI requirements. It should be noted that the submerged portions of the liner are stainless steel. The potential for leakage under high pressure during core damage accidents is explicitly included in the risk analysis. The acceptance of the licensees risk assessment is discussed elsewhere in this SE.
Based on its review of the information provided in the licensees amendment request, the staff finds that: (1) the structural degradation of the accessible areas of GGNS containment will be adequately monitored through the periodic ISI conducted as required by Subsection IWE of Section XI of the ASME Code, and (2) the integrity of the penetrations and containment isolation valves will be 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 in accordance with Subarticle IWE-5000 of Section XI of the ASME Code. Significant degradation of the primary containment pressure boundary is required to be reported under 10 CFR 50.72 or 10 CFR 50.73.
3.2 Risk Assessment The licensee has performed a risk impact assessment of extending the test interval for the Type A test and the DWBT from 10 years to 15 years. The risk assessment was provided in the May 12, 2003, application for license amendment. Additional analysis and information was provided by the licensee in its supplemental letter dated October 29, 2003. In performing the risk assessment, the licensee considered the guidelines of NEI 94-01, the methodology used in Electric Power Research Institute (EPRI) Topical Report TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing," and RG 1.174.
The basis for the current 10-year Type A test interval is provided in Section 11.0 of NEI 94-01 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 GGNS 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 from 18 to 60 months that a leak, which is detectable only by a Type A test, goes undetected. 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 BWR 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 an increased public dose on the order of a fraction of one person-rem per year.
Building upon the methodology of the EPRI study, the licensee assessed the risk increase associated with extending the Type A test and the DWBT from 10 years to 15 years. The licensee quantified the risk from sequences that have the potential to result in large releases if a pre-existing containment leak or drywell bypass leaks were present. Since the Option B rule-making 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 the estimated risk increase. 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 and drywell bypass leak rate tests do not impact CDF, the relevant criterion is the change in LERF. The licensee has estimated the change in LERF for the proposed changes relative to the original frequency of three tests in 10 years. 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 changes to demonstrate that the defense-in-depth philosophy is met.
In assessing the risk impacts associated with the DWBT interval extension, the licensee applied the same basic approach as embodied in the EPRI methodology for Type A test interval extensions. The primary difference in the methodology used to evaluate the DWBT extension is in the determination of the conditional probability of an existing drywell leak, and in the assignment of various drywell and containment leakage combinations to appropriate containment failure categories. In a Mark III containment, the drywell is completely enclosed by the primary containment. As such, drywell leakage does not leak directly to the environment, but is further mitigated by the primary containment. Because of this dual structure, the licensee considered the probability of various drywell and containment leakage combinations. Similar to the EPRI methodology for Type A test frequency extensions, the drywell was considered either to be intact (base leakage assumed), to have a small pre-existing failure (10 times the base leakage), or to have a large pre-existing failure (35 times the base leakage). The base drywell leakage rate (900 scfm) was established through review of the "as-found" DWBT results from the previous DWBTs at GGNS. The probability of each of the drywell failure categories (intact, small leak, and large leak) was assumed to be the same as the equivalent categories for the Type A evaluations. The three drywell leakage levels were considered in combination with the three different containment leakage levels in the EPRI methodology, resulting in nine combinations of drywell and containment leakage sizes. For leakage combinations involving drywell leakage, the availability of containment sprays was also evaluated. For events in which containment sprays operate, drywell leakage was assumed to have no impact on the containments existing leakage category, since the sprays would condense any steam that bypasses the suppression pool. For events in which containment sprays do not operate, any increased drywell leakage was assumed to lead to containment failure. Each of the leakage combinations was assigned to one of the EPRI containment failure categories based on consideration of the availability of containment sprays. The remaining portions of the DWBT methodology are identical to that used for the Type A test frequency extension.
At the NRC staffs request, the licensee provided the results of a sensitivity analysis in which the probability of each of the drywell failure categories is based on consideration of an expanded data set consisting of all "as-found" DWBT results for all Mark III containments. The licensee estimated the failure probability for the small drywell leakage category using a mean value derived from the available data, and estimated the failure probability for the large drywell leakage category using the Jeffreys non-informative prior value, since there have been no occurrences of large drywell leakage within the available data. The staff considers use of these failure probabilities appropriate for a realistic evaluation. The licensee also provided a sensitivity analysis in which credit is taken for the ability to depressurize the reactor coolant system via release of steam to the suppression pool. Drywell bypass is not a concern for such sequences as there is no steam release into the drywell early in an event (prior to reactor vessel breach).
Based on the analyses provided by the licensee in its supplement dated October 29, 2003, the following risk comparisons and conclusions can be drawn. The comparisons of risk from a change in test frequency from three tests in 10 years to one test in 15 years are considered to be bounding for GGNS comparative frequencies of one test in 10 years to one test in 15 years:
1.
Given the change from a three in 10-year test frequency to a one in 15-year test frequency, the increase in the total integrated plant risk is estimated to be less than 0.1 person-rem per year. This increase is comparable to that estimated in NUREG-1493, where it was concluded that a reduction in the frequency of tests from three in 10 years to one in 20 years leads to an "imperceptible" increase in risk. Therefore, the increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.
2.
As shown in Table 3-12 of the supplemental letter, the increase in LERF resulting from a change in the Type A and drywell bypass leak rate test frequencies from the original three in 10 years to one in 15 years is estimated to be 7.7 x 10-8 per year in the licensee's baseline analysis. If the probability of the drywell failure categories is based on "as-found" DWBT results for all Mark III containments, the increase in LERF is 3.8 x 10-7 per year assuming credit for drywell leakage mitigation by containment sprays (as in the baseline analysis), and 2.4 x 10-8 per year assuming credit for drywell leakage mitigation by either containment sprays or reactor coolant system depressurization.
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 ASME 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. Based on the review of data in Table 1-3 of the supplemental letter, the increase in LERF associated with corrosion events is estimated to be less than 1 x 10-8 per year. The NRC staff concludes that increasing the Type A and DWBT intervals 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. Based on information provided by the licensee, the change in the test frequency from three in 10 years to one in 15 years would increase the conditional containment failure probability by about one percentage point. 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, while maintaining the defense-in-depth philosophy, of RG 1.174 and, therefore, is acceptable.
Based on the foregoing evaluation, the staff finds that the interval until the next Type A and drywell bypass leak rate tests at GGNS may be extended to 15 years, and that the proposed changes to TS Sections 3.6.5.1.1 and 5.5.12 are acceptable.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Mississippi State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 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. 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 34666). 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.
7.0 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 the operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of amendments will not be inimical to the common defense and security or to the health and safety of the public. Therefore, a one-time 5-year extension to the current 10-year test interval for containment ILRT and DWBT, as requested by the licensee, is acceptable.
Principal Contributors: R. Palla J. Pulsipher K. Martin Date: January 28, 2004
Grand Gulf Nuclear Station cc:
Executive Vice President
& Chief Operating Officer Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205 Winston & Strawn 1400 L Street, N.W. - 12th Floor Washington, DC 20005-3502 Chief Energy and Transportation Branch Environmental Compliance and Enforcement Division Mississippi Department of Environmental Quality P. O. Box 10385 Jackson, MS 39289-0385 President Claiborne County Board of Supervisors P. O. Box 339 Port Gibson, MS 39150 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 Senior Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 399 Port Gibson, MS 39150 General Manager, GGNS Entergy Operations, Inc.
P. O. Box 756 Port Gibson, MS 39150 Attorney General Department of Justice State of Louisiana P. O. Box 94005 Baton Rouge, LA 70804-9005 State Health Officer State Board of Health P. O. Box 1700 Jackson, MS 39205 Office of the Governor State of Mississippi Jackson, MS 39201 Attorney General Asst. Attorney General State of Mississippi P. O. Box 22947 Jackson, MS 39225 Vice President, Operations Support Entergy Operations, Inc.
P.O. Box 31995 Jackson, MS 39286-1995 Director Nuclear Safety Assurance Entergy Operations, Inc.
P. O. Box 756 Port Gibson, MS 39150 Mr. Michael A. Krupa, Director Nuclear Safety & Licensing Entergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213-8298