ML090430367
| ML090430367 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/11/2009 |
| From: | Richard Guzman Plant Licensing Branch 1 |
| To: | Polson K Nine Mile Point |
| Guzman R, NRR/DORL, 415-1030 | |
| References | |
| TAC MD9453 | |
| Download: ML090430367 (25) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 11, 2009 Mr. Keith J. Polson Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P. O. Box 63 Lycoming, NY 13093
SUBJECT:
NINE MILE POINT NUCLEAR STATION, UNIT NO.1 - ISSUANCE OF AMENDMENT REGARDING ONE-TIME EXTENSION OF PRIMARY CONTAINMENT INTEGRATED LEAKAGE RATE TEST INTERVAL (TAC NO.
MD9453)
Dear Mr. Polson:
The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 202 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1), in response to your application dated August 15, 2008 (Agencywide Documents Access Management System (ADAMS) Accession No. ML082330228), as supplemented on December 4, 2008 (ADAMS Accession No. ML083440059).
The amendment revises NMP1 Technical Specification (TS) 6.5.7, "10 CFR 50 [Part 50 of Title 10 of the Code of Federal Regulations] Appendix J Testing Program Plan," to allow a one-time extension of the Integrated Leak Rate Test (ILRT) interval for no more than 5 years. The proposed amendment would allow the next ILRT for NMP1 to be performed within 15 years from the last ILRT as opposed to the current 10-year interval.
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosures:
- 1. Amendment No. 202 to DPR-63
- 2. Safety Evaluation CG wfencls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WAc:.I-III\1~Tru.1
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'U"}'JOJ"uuu I NINE MILE POINT NUCLEAR STATION, LLC (NMPNS)
DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 202 Renewed License No. DPR-63
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Nine Mile Point Nuclear Station, LLC (the licensee) dated August 15, 2008, as supplemented on December 4,2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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 Commission's 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 Commission's 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-63 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 202, is hereby incorporated into this license.
Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications.
-2
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION Mark G. Kowal, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the License and Technical Specifications Date of Issuance: March 11, 2009
ATTACHMENT TO LICENSE AMENDMENT NO. 202 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page 3 3 Replace the following pages of 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 Pages Insert Pages 355 355 355a 355a
-3 (3) Pursuant to the Act and 10 CFR Parts 30. 40. and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components.
(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30; Section 40.41 of Part 40; Section 50.54 and 50.59 of Part 50; and Section 70.32 of Part 70. This renewed license is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect and is also subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1850 megawatts (thermal).
(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 202, is hereby incorporated into this license. Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications.
(3) Deleted Renewed Lioense No. DPR 63 Amendment No. 191, 192, 193, 194,196,196,197,198,199,200,201,202
6.5.7 10 CFR 50 Appendix J Testing Program Plan
- a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, entitled "Performance-Based Containment Leak-Test Program," dated September 1995 with the following exceptions:
- 1. Type A tests will be conducted in accordance with ANSI/ANS 56.8-1994 and/or Bechtel Topical Report BN-TOP~1, and
- 2. The first Type A test following approval of this Specification will be a full pressure test conducted approximately 70, rather than 48, months since the last low pressure Type A test.
- 3. Exception to NEI 94-01, Rev. 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Section 9.2.3: The first Type A test performed after the June 8, 1999 Type A test shall be performed no later than June 8, 2014.
- b. The peak calculated containment internal pressure (Pac) for the design basis loss of coolant accident is 35 psig.
- c. The maximum allowable primary containment leakage rate (La) at Pac shall be 1.5% of primary containment air weight per day.
- d. Leakage Rate Surveillance Test acceptance criteria are:
- 1. The as-found Primary Containment Integrated Leak Rate Test (Type A Test) acceptance criteria is less than 1.0 La.
- 2. The as-left Primary Containment Integrated Leak Rate Test (Type A Test) acceptance criteria is less than or equal to 0.75 La, prior to entering a mode of operation where containment integrity is required.
- 3. The combined Local Leak Rate Test (Type B & C Tests including airlocks) acceptance criteria is less than 0.6 La. calculated on a maximum pathway basis, prior to entering a mode of operation where containment integrity is required.
- 4. The combined Local Leak Rate Test (Type B & C Tests including airlocks) acceptance criteria is less than 0.6 La, calculated on a minimum pathway basis, at all times when containment integrity is required.
AMENDMENT NO. 142, 157, 162, 181, 182, 202 355
- e. The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the 10 CFR 50 Appendix J Testing Program Plan.
The provisions of Specification 4.0.3 are applicable to the 10 CFR 50 Appendix J Testing Program Plan.
6.5.8 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Air Treatment (CRAT) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
- b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
- c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testlnq methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision O.
- d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation of the CRAT System, operating at a flow rate of 2025-2475 cfm, at a Frequency of 24 months. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.
- e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
- f. The provisions of TS 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
AMENDMENT NO. +9a, 202 355a
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 202 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION, LLC NINE MILE POINT NUCLEAR STATION, UNIT NO.1 DOCKET NO. 50-220
1.0 INTRODUCTION
By letter dated August 15, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082330228), as supplemented by letter dated December 4,2008 (ADAMS Accession No. ML083440059), Nine Mile Point Nuclear Station, LLC (NMPNS or the licensee) submitted a license amendment request (LAR) for Nine Mile Point Nuclear Station, Unit NO.1 (NMP1). The amendment would revise NMP1 Technical Specification (TS) 6.5.7, "10 CFR 50 [Part 50 of Title 10 of the Code of Federal Regulations] Appendix J Testing Program Plan," to allow a one-time extension of the Integrated Leak Rate Test (ILRT) interval for no more than 5 years. The proposed amendment would allow the next ILRT for NMP1 to be performed within 15 years from the last ILRT as opposed to the current 1O-year interval.
The supplemental letter dated December 4, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's initial proposed no significant hazards consideration determination as published in the Federal Register (FR) on October 21, 2008 (73 FR 62566).
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. NMP1 TS 6.5.7, "10 CFR 50 Appendix J Testing Program Plan,"
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. RG 1.163 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.
RG 1.163, Section C, "Regulatory Position" states; "licensees intending to comply with the Option B in the amendment to Appendix J should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01 rather than using test intervals specified in ANSI/ANS-56.8-1994."
-2 The industry guidelines in NEI 94-01 state that Type A testing shall be performed at a frequency of at least once every 10 years. NMPNS's proposed TS amendment would change the 10-year ILRT interval to a 15-year interval based on the results of the earlier ILRT, the ongoing LLRT and the inservice inspection (lSI) programs supported by risk informed analysis performed in accordance with the NRC staff guidelines in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." There are no changes to any Code or regulatory requirement.
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 (NMP1 TS exemption allows 70 months test interval), 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 NMP1 (completed in 1993 and in 1999) have been successful, so the current interval requirement would normally be 10 years.
The licensee is requesting a change to TS 6.5.7, which would add an exception from the guidelines of RG 1.163 and NEI 94-01, Revision 0, regarding the Type A test interval.
Specifically, the exception states that "the first Type A test performed after the June 8. 1999 Type A test shall be performed no later than June 8, 2014."
The local leakage rate tests (Type B and Type C tests), including their schedules, are not affected by this request.
3.0 TECHNICAL EVALUATION
3.1 Background As described in NMPNS's application and the NMP1 Updated Final Safety Analysis Report (UFSAR), NMP1 is a General Electric Boiling-Water Reactor (BWR) contained in a Mark I free standing steel containment building. The Mark I pressure suppression containment system consists of a drywell, a pressure suppression chamber (torus), a vent system connecting the drywell and suppression chamber, a vacuum relief system, and a containment cooling system.
The drywell, which houses the reactor vessel and reactor coolant recirculation loops, is a free standing, low-leakage steel pressure vessel enclosed in a reinforced concrete structure with a 2-to 3-inch air gap between the shell plate and concrete. The suppression chamber (torus) is a free-standing, toroidal shaped steel pressure vessel designed to hold a large volume of water (suppression pool) for use as a heat sink for postulated transients and accident conditions.
The Mark I primary containment is penetrated by access hatches, piping, and electrical penetrations. The integrity of the penetrations and containment isolation valves is verified through Type B and Type C tests as required by 10 CFR Part 50, Appendix J, and the overall integrity of the containment structure is verified through a Type A test. These tests are performed to verify the leak-tight integrity of the containment structure at the design-basis accident (DBA) pressure. The leakage rate testing requirements of 10 CFR Part 50 Appendix J Option B (Type A, Type B, and Type C tests) and the Containment Inservice Inspection (CISI) requirements mandated by 10 CFR 50.55a, ensure the continued leak-tight and structural integrity of the containment during its service life.
-3 NMP1 performs various inspections and tests that are routinely performed to assure primary containment integrity in addition to periodic Type A testing. These include Type Band C testing performed in accordance with Appendix J, Option B; inspection activities performed as part of the American Society of Mechanical Engineers Boiler Pressure and Vessel Code (ASME Code),
Section XI (Subsection IWE) inspection program; inspection of drywell and torus surfaces and structural elements; inspections of torus exterior surfaces and supports; and inspections of drywell interior coatings. The aggregate results of these tests and inspections provide a high degree of assurance of continued primary containment integrity.
The NMP1 Appendix J, Type B and Type C testing program evaluates all but a small portion of potential containment leakage pathways including electrical penetrations, airlocks, hatches, flanges, and valves. The Type Band C test program consists of local leak rate testing of penetrations with a resilient seal, expansion bellows, double-gasketed manways, hatches and flanges, drywell airlocks, and containment isolation valves that serve as a barrier to the release of the post-accident primary containment atmosphere. The results of the test program are used to ensure that proper maintenance and repairs are made on the primary containment components over their service life. Type B and Type C testing provide a high degree of assurance that primary containment integrity is maintained.
10 CFR Part 50, Appendix J, Option B, Section 1I1.A, states: "A general visual inspection of the accessible interior and exterior surfaces of the containment system for structural deterioration which may affect the containment leak-tight integrity must be conducted prior to each test, and at a periodic interval between tests based on the performance of the containment system." This inspection is also conducted during two other refueling outages before the next Type A test if the interval for the Type A test has been previously extended to 10 years, in order to allow for early discovery of structural deterioration. Effective September 1996, the NRC amended 10 CFR 50.55a to endorse Subsections IWE and IWL of the ASME Code,Section XI, 1992 Edition including 1992 Addenda. These subsections contain lSI and repair/replacement rules for Class MC (metal containment) and Class CC (concrete components. The NMP1 primary containment is a free-standing steel containment, to which the requirements of Subsection IWE apply.
The Subsection IWE containment inspection requirements are implemented at NMP1 through the CISI Plan and Schedule (NMP1-IWE-003, referred to herein as the IWE lSI program). The program contains detailed lSI requirements for Class MC components in accordance with 10 CFR 50.55a (b)(2)(vi) and (ix), and the ASME Code,Section XI, 1998 Edition with no addenda. The general visual examination requirements specified in the IWE lSI program satisfy the visual examination requirements specified in Option B. The first interval containment lSI program became effective on December 26, 1999, and is scheduled to end December 25, 2009.
NMP1 containment shell acceptance criteria are based on the calculated corrosion rate. The corrosion rate criteria are based on maintaining a wall thickness of greater than the minimum design value. Acceptance criteria have been established and are documented in procedure NDEP-VT-2.05 and in the IWE lSI program. Flaws identified during inspections are described as nicks, gouges, arc strikes, cracking, rust, or pitting. For each flaw, varying levels of severity are described and are evaluated as acceptable, unacceptable, or requiring further evaluation.
-4 Examination of pressure-retaining bolted connections and evaluation of containment bolting flaws or degradation are performed in accordance with the requirements of 10 CFR 50.55a (b)(2)(ix)(G) and 10 CFR 50.55a(b)(2)(ix)(H).
NMPNS justifies the proposed change to extend the current ILRT interval from 10 to 15 years based on previous plant-specific Type A test performance, continuous Type Band C testing, and the CISI program results, supported by a risk-informed analysis.
In its supplemental letter dated December 4, 2008, NMPNS provided response to the NRC staff's request for additional information (RAI) to address issues related to the Type A test results, Type Band C tests, and CISI. The NRC staff reviewed NMPNS's technical analysis in support of its proposed TS amendment which is described in Attachment 1 of its application and its response to the staff's RAI.
3.2 NMP1 Type A Test NMPNS states in its application, that based on the 1993 and 1999 Type A tests at NMP1, the current Type A test interval is once every 10 years. With the requested 5-year extension of the ILRT interval, NMPNS proposed performing the next overall verification of the containment leak tight integrity no later than June 8, 2014.
NMPNS provided the leakage rates of 0.4634 weight percent (wt%)/day and 0.5045 wt%/day for the 1993 and 1999 Type A tests, respectively. As stated in the application, the NMP1 maximum allowable primary containment leakage rate (La) is 1.5% of primary containment air weight per day at a pressure of 35 pounds-per-square inch (psig).
The results of the Type A ILRT show containment leakage within the acceptance limit and an adequate margin indicating leak-tightness of NMP1 containment structure.
Regulatory Position C.3 of RG 1.163 recommends that a visual examination of accessible interior and exterior surfaces of the containment structure should be conducted prior to initiating a Type A test, and during two other refueling outages before the next Type A test based on a 1O-year ILRT interval. NMPNS states in its application that the general visual examination requirements specified in the IWE CISI program satisfy the visual examination requirements specified in 10 CFR Part 50, Appendix J, Option B. NMPNS further states that the first interval CISI became effective on December 26, 1999, and is scheduled to end December 25, 2009.
The NRC staff's RAI requested that the licensee describe the plan to supplement the 10-year interval-based visual inspection requirement to accommodate the requested 15-year ILRT interval. In response to the staff's RAI, NMPNS states that the containment inspections will continue to be performed during the proposed 5-year extension of the Type A test interval in accordance with the IWE CISI program. The IWE CISI program requires a minimum of one inspection during each inspection period of the inspection interval. The proposed 5-year extension will coincide with the first and second inspection periods of the second interval of the IWE CISI program.
In addition, NMPNS states that the following visual inspections of accessible interior surfaces of the primary containment are conducted each refueling outage in accordance with approved
-5 plant procedures to provide reasonable assurance that the effects of aging will be adequately managed:
Drywell and Drywell Head Interior
- Vicinity of drywell penetrations for obvious structural discontinuities (cracks).
- Support attachments and brackets for obvious defects (missing or broken bolts/nuts, bent rods, plate buckling, etc.).
- Internal surface area for gross signs of corrosion and deterioration (depth greater than approximately 1/16"; indications of leak).
- Internal coated surface area for any visible defects including blistering, cracking, flaking, peeling and physical or mechanical damage (area larger than approximately 6 square feet).
Suppression Chamber Interior
- Vicinity of any penetrations for obvious structural discontinuities (cracks).
- Vent pipe expansion joints, support structures, brackets and bolting for obvious defects (missing or broken nuts/bolts, bent rods, plate buckling, etc.).
- Internal surface area, including water line regions, for gross signs of corrosion or buckling.
Based on its review, the NRC staff finds the licensee's response acceptable and consistent with the intent of Regulatory Position C.3 of RG 1.163 because during the first IWE CISI interval, ending in December 2009, three visual inspections of containment structure will be performed.
In addition, during the 5-year extension of the ILRT interval (from June 2009 to June 2014), two more visual examinations of the containment structure will be performed as part of the second interval of IWE CISI prior to pre-ILRT visual examination. Also, the visual examinations of accessible interior surfaces of the drywell and suppression chamber performed every refueling outage would provide additional assurance that the effects of aging are adequately managed.
3.3 NMP1 Type Band C Tests As stated in NMPNS's application, the proposed TS Amendment will not affect the scope, performance, or scheduling of Type B and Type C testing of containment penetrations and isolation valves. In response to the staff's RAI, NMPNS provided a table identifying Type Band Type C leak rate testing of penetrations that are planned to be performed during the requested 5-year extension of the ILRT interval. NMPNS states that Type B leak rate testing is on a maximum 10-year testing intervals. If the result of a Type B test exceeds the allowable limit, the penetration will be returned to the short interval (every outage). The Type C leak rate testing of containment isolation valves are on a maximum 60-month testing interval. If the result of a Type C test exceeds the allowable limit, the penetration will be returned to the short interval (every outage). NMPNS also indicates that during the spring 2009,2011, and 2013 outages, 261 planned Type B and Type C leak rate tests will be performed.
-6 NMPNS, in response to staff's RAI, provided the following results of last two maximum and minimum pathway local leak rate tests. The NMP1 acceptance criterion for combined LLRT (Type B and type C including airlocks) is 0.6 La (388.44 standard cubic feet per hour (scfh)).
efueling Outage Maximum Pathway Minimum Pathway I leakage (sefh) % of 0.6la leakage (sefh) % of 0.6la-2007 (N1R19) 240.38 61.9% 89.734 23.1%
2005 (N1R18) - 222.169 57.2% 83.492 21.5%
Furthermore, in response to staff's RAI, NMPNS provided a list of those penetrations with seals and gaskets and bolted connections. Those penetrations that are frequently disassembled (e.g, equipment hatch, drywell head) are on a 30-month (every refueling outage) Type B test interval. For those penetrations that are not routinely disassembled, the frequency of Type B testing is performance-based up to maximum of once every 120 months.
Based on the above, the NRC staff finds that the integrity of the containment pressure boundary penetrations (including equipment hatch, airlocks and drywell heads) and isolation valves are effectively monitored through Type Band C testing as required by 10 CFR Part 50, Appendix J and NMP1 TS.
3.4 Containment In-Service Inspection Program (ASME Code Section XI, Subsection IWE)
NMPNS states in its application, that the NMP1 primary containment is a free-standing steel containment, to which the requirements of SUbsection IWE of the ASME Code,Section XI, 1998 Edition apply. NMPNS also states that the first interval IWE CISI program became effective on December 26, 1999, and is scheduled to end December 25, 2009, and provided the following table indicating three inspection periods that are scheduled for the current inspection interval of IWE CISI.
NMP1 INSERVICE INSPECTION PERIODS INSPECTION PERIOD START PERIOD END REFUEL REFUEL PERIOD DATE DATE OUTAGE OUTAGE YEAR 1 December 26, 1999 December 25, 2002 RFO-15 1999 I
RFO-16 2001 2 December 26, 2002 December 25,2006 RFO-17 2003 RFO-18 2005 3 [ December 26,2006 December 25,2009 RFO-19 2007 Ii RFO-20 2009 The examinations are performed in accordance with non-destructive examination procedure NDEP-VT-2.05, "ASME Section XIIWE/IWL Visual Examination." The examination of pressure-retaining bolted connections and containment bolting flaws or degradation evaluations are performed in accordance with the requirements of 10 CFR 50.55a(b )(2)(ix)(G) and 10 CFR 50.55a(b )(2)(ix)( H).
- 7 Regarding the moisture barrier between the drywell shell and the concrete floor, the licensee states that the moisture barrier was inspected during the spring 2001 refueling outage in accordance with the IWE CISI program. Although during the inspection some degradation of the moisture barrier sealant was found, there was no indication of unacceptable degradation in the visible areas of the drywell shell adjacent to the moisture barrier. The licensee states that the moisture barrier was repaired during the spring 2001 refueling outage by applying new sealant where required.
NMPNS also states that in addition to the inspections performed in accordance with the IWE CISI program, periodic visual inspections of the coating on accessible interior surfaces of the drywell shell and drywell head are performed to identify evidence of deterioration. The inspections are performed every refueling outage to identify any visible defect including blistering, cracking, flaking, peeling, and physical or mechanical damage, and to evaluate, and determine any necessary corrective action.
NMPNS also described in its application, the following supplementary programs to monitor the corrosion rate of drywell and torus shell. These programs have been established to ensure timely action to correct degradations that could lead to loss of the intended function. As provided in NUREG-1900, "Safety Evaluation Report Related to the License Renewal of Nine Mile Point Nuclear Station, Units 1 and 2," September 2006, the NRC staff reviewed and accepted these supplementary programs, the projected drywell and torus shell thickness at the end of extended operation, and the established acceptance criteria.
3.5 Drywell IWE-1240 Augmented Inspection NMPNS states in its application that during the March 2007 NMP1 refueling outage, the drywell shell was inspected in accordance with the IWE CISI examination schedule, and an ultrasonic testing (UT) measurement found a minimum thickness of 1.089 inches for the drywell shell.
This value exceeded the established minimum wall thickness criteria of 1.049 inches, and, was therefore, considered acceptable. Grids overlapping the four areas measured in 2007 were marked on the shell using a permanent marking means so that these same points can be measured again in future refueling outages. The NRC staff's RAI requested NMPNS to provide a general description and correlation between the results of 2003 and 2007 examinations.
NMPNS, in response to the staffs RAI, the licensee stated the following:
Due to the radiological conditions existing in the drywell during the 2003 refueling outage, the investigation of the condition was limited to four areas of the drywell shell (around 3 of the area coolers) that were considered to represent the worst areas of major corrosion. A UT thickness reading was taken at each of these four identified locations. The thickness reading locations were defined by measured distances from the floor and nearby support beams, but no grids were applied to the shell to facilitate future location of the exact spots where the thickness readings were taken. The evaluation performed in 2003 evaluated the lowest readings found at each measured location against the minimum required wall thickness and concluded that the drywell shell was acceptable for continued service.
-8 UT thickness measurements were taken during the 2007 refueling outage at the reported locations where the 2003 measurements had been taken. It was anticipated that a corrosion rate could be determined from a comparison of the 2003 and 2007 readings; however, the corrosion rates derived from that limited set of data points were widely scattered, unrealistic (one location showed a gain in wall thickness) and inconsistent with the observed condition of the drywell shell. It was concluded that this limited data could not be used as the sole basis for determining a corrosion rate. This result was attributed to the likelihood that the exact same spots had not been measured in 2003 and in 2007. Therefore, actions were taken during the 2007 refueling outage to establish a more repeatable means of determining wall thickness so that a truly representative corrosion rate can be determined. Grids were painted on the drywell shell at the areas of interest and readings taken at multiple grid points. Measurements taken during the 2009 refueling outage at the same grid points will allow actual corrosion rates to be established and addressed in accordance with the drywell supplemental inspection program acceptance criteria, which have been added to the IWE Program.
In response to the staff's RAI to provide a discussion of the general corrosion condition in the monitored areas, NMPNS stated the following:
The areas of localized drywell shell corrosion were extensively inspected by the IWE Responsible Individual during the 2007 refueling outage. These areas were observed to have a generalized corroded surface, but no evidence of loose corrosion products was present. There were no rust flakes or blisters on the surfaces, no evidence of pitting, and no build up of rust flakes on the floor below the areas. If significant shell corrosion had taken place, corrosion products should have been observed in the areas since carbon steel corrosion products expand significantly. The absence of corrosion products was inconsistent with the corrosion rates that were indicated by comparing the first 2007 set of four UT thickness measurements with the 2003 UT thickness measurement data.
Also, the NRC staff requested the licensee to discuss the schedule for the next UT measurements, root cause determination, and any planned or already implemented corrective actions based on the results of 2007 examinations. In response to the staff's RAI, NMPNS states that UT thickness measurements will be taken during the 2009 refueling outage at the grid locations established in 2007. These 2009 measurements will be compared to the baseline data established in 2007 to determine a corrosion rate for the 2-year period. The corrosion rate determined from the UT measurement data and the remaining margin to the minimum required wall thickness will determine the subsequent frequency of performing UT thickness measurements as well as the need to implement mitigative strategies (e.g., application of protective coatings, repair, or replacement of affected sections of the shell).
In addition, NMPNS states, in its supplemental letter, that the apparent cause of the localized corrosion of the drywell shell in the area of each of the drywell area coolers was determined to be the cleaning practices for the area cooler coils. The procedure for cleaning the area coolers called for the coils to be rinsed with a cleaning agent. There were no protective measures for the liner and no requirement to rinse the liner after cleaning. As indicated in its supplemental
-9 letter, the procedure for cleaning the cooler coils was revised in 2003 to require the use of protection on the liner before cleaning of the coolers.
The staff also requested NMPNS to discuss whether there are other areas of the drywell shell requiring augmented examination. In response to the staff's RAI, NMPNS states that other than the drywell augmented inspection, there are no other areas requiring augmented examination in accordance with IWE-1240. NMPNS also states that during the license renewal application process, NMPNS committed to perform an augmented VT-1 visual inspection of the containment penetration stainless steel bellows using enhanced techniques qualified for detecting stress-corrosion cracking. NMPNS states that these inspections are beyond the scope of examinations required by Table IWE-2500-1 of the ASME Code,Section XI and thus are referred to as augmented examinations in the IWE containment inspection program plan.
However, they are not considered augmented examinations as defined in IWE-1240.
3.6 Torus Corrosion Monitoring Program NMPNS states in its application that the NMP1 torus corrosion monitoring program is based on a commitment to periodically monitor torus condition in the form of component inspections and analysis. This program includes the following inspection and analysis methods: (1) obtaining periodic torus wall UT thickness measurements over a pre-defined grid system, (2) periodic removal, analysis, and comparison of torus material coupons to the results of the UT measurements; and (3) performance of visual inspections of accessible external surfaces of the torus support structure.
NMPNS states that:
- To date, the coupon corrosion rate has been consistently lower than the corrosion rate determined via UT thickness measurement;
- The next coupon corrosion rate determination is scheduled for the spring 2009 refueling outage;
- Analysis of February 2007 UT thickness measurements at selected torus bottom mid-bay locations indicated that the smallest local/individual wall thickness obtained during this survey was 0.4452 inches with calibration adjustment applied. All wall thickness measurements exceeded the allowable average minimum wall thickness of 0.431 inches, and thus, were considered acceptable;
- The most recent inspections of the torus external structures, performed during the 2007 refueling outage, found the condition of the structures acceptable with no signs of missing or loose hardware, spalied concrete or major degradation that would impact the structural integrity of the torus structure.
NMPNS further states that the issue of a through-wall torus shell crack discovered at the James A. Fitzpatrick Nuclear Power Plant (JAF) on June 27,2005, was reviewed for applicability to NMP1. NMPNS concluded that the condition at JAF is not applicable to NMP1 due to
- 10 differences in system configurations, which include: (1) the JAF high-pressure coolant injection (HPCI) turbine exhaust line that discharges into the suppression pool is open ended and does not have an end cap or a sparger. In contrast, NMP1 does not have a steam turbine driven HPCI system; and (2) the Emergency Condenser (EC) system vent line and the Electromagnetic Relief Valve (ERV) discharge lines, which are the systems that may discharge steam into the suppression pool, both terminate below the torus water surface and the EC System has a sparger allowing for steam condensing.
The NRC staff finds that the torus corrosion monitoring program provides reasonable assurance that any degradation of the torus shell material and external structure which could result in a loss of its intended function will be detected and corrective actions will be taken to ensure its structural integrity.
3.7 Plant Operational Performance NMPNS states in its application that during power operation, the primary containment is inerted with nitrogen to maintain oxygen concentration within TS limits. As a result, the primary containment is maintained at a slightly positive pressure during power operation. Maintaining the containment pressurized at power assures that gross containment leakage that may develop during power operation will be detected. A drywell pressure alarm will alert operators that the drywell pressure is abnormal (greater than 2 psig or less than 1 psig).
3.8 NRC Information Notice 92-20 NRC Information Notice 92-20 was issued to alert licensees to problems with local leak rate testing of two-ply stainless steel bellows used on piping penetrations at some plants. In its application, NMPNS states that NMP1 fluid lines with temperatures above 150 of have a guard pipe between the hot fluid line and the penetration attachment to the drywell steel, in addition to the double-seal arrangement. This configuration protects the penetration against over pressurization should the hot line rupture inside the penetration. The hot fluid from a rupture of this type would be vented into the drywell by the guard pipe. The hot fluid penetrations have two expansion bellows, one inside the drywell and one outside. The main steam line penetration bellows have been modified by adding "clam-shell" bellows joints over the existing joints, which were removed, to accommodate the thermal expansion as well as any movement due to a line rupture. Inner bellows are designed for a lower pressure than the outside bellows to assure inward leakage in the event of failure. NMPNS states that Type B test results of the primary containment penetrations with expansion bellows have been satisfactory, and the test interval is 120 months for all penetration expansion bellows except the main steam and feedwater penetrations which have a fixed 30-month Type B test interval.
Considering the satisfactory results of Type B tests of primary containment penetrations with expansion bellows and NMPNS commitment to perform an augmented VT-1 visual inspection of the stainless steel bellows using enhanced techniques, the staff finds the licensee's inspection and testing program effective to ensure leak tightness and to detect degradation of stainless steel bellows.
- 11 3.9 Generic Letter 87-05 On March 12, 1987, the NRC issued Generic Letter (GL) 87-05 requesting information from BWR owners regarding their intended actions to determine if drywe lis at their facilities have degraded by the same corrosion mechanism observed in the sand cushion location of the drywell shell at Oyster Creek. In response to GL 87-05, as stated in staff's safety evaluation for NMP1 license renewal, NMPNS conducted several investigations and inspections which determined that water intrusion into the NMP1 sand cushion had not occurred. The NRC staff reviewed and found NMPNS's program to manage the reactor cavity to drywell refueling seal acceptable. In its application dated August 15, 2008, NMPNS further states that inspections performed during the spring 2007 outage, and review of the video recordings of the sand cushion area under the drywell indicated a dusty and dry environment, and that there were no observed indications of water leakage. NMPNS states that only an expected nominal general corrosion of the interior of the carbon steel drain line was observed and found to be acceptable.
In summary, on the basis of its review of the information provided in NMPNS's application and responses to the staff's RAI, the staff finds that: (1) the results of the past ILRT and LLRT demonstrate that the leak-tight integrity of the containment structure has been adequately managed; (2) the structural integrity of the containment vessel is verified through periodic lSI conducted as required by Subsections IWE of the ASME Code,Section XI; (3) the integrity of the penetrations and containment isolation valves are periodically verified through Type Band Type C tests as required by 10 CFR Part 50, Appendix J, and NMP1 TS; (4) the licensee is employing a CISI program, torus corrosion monitoring program, and drywell augmented inspection program that require evaluation of any potential degradation of accessible and inaccessible areas of the containments; and (5) the primary containment protective coating is inspected visually every refueling outage and repair of any identified damage is adequately managed. Accordingly, the NRC staff concludes that NMPNS has an adequate CISI program and procedures in place to examine, monitor and correct potential age-related and environmental degradations of the pressure retaining components of the NMP1 primary containment structure. Therefore, NMPNS's proposed one-time extension of the ILRT interval from 10 to 15 years is acceptable. In scheduling the next ILRT, the staff expects that NMPNS will also take into consideration the staff position described in the Regulatory Issue Summary 2008-27, "Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50," dated December 8,2008 (ADAMS Accession No. ML080020394).
3.10 Risk Analysis NMPNS has performed a risk impact assessment of extending the Type A test interval to 15 years. The risk assessment was provided in the licensee's August 15, 2008, application.
Additional analysis and information was provided by NMPNS in its supplemental letter dated December 4,2008. In performing the risk assessment, the licensee considered the guidelines of NE194-01, the methodology used in Electric Power Research Institute (EPRI) TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing," the NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Surveillance Intervals, and RG 1.174." NMPNS also provided results of a risk assessment based on the methodology in EPRI TR-1 009325, Revision 2, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals."
- 12 However, the NRC staff relied on the results using the NEI Interim Guidance because the latter methodology has not been approved.
The basis for the current 1O-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 NMP1 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, that was detectable only by a Type A test, goes undetected from 18 to 60 months. Since Type A tests only detect about 3 percent of leaks (the rest are identified during local leak rate tests based on industry leakage rate data gathered from 1987 to 1993), this results in a 10 percent increase in the overall probability of leakage. The risk contribution of pre-existing leakage for the PWR 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 a fraction of one person-rem per year in increased public dose.
Building upon the methodology of the EPRI study and the NEI Interim Guidance, NMPNS assessed the change in the predicted person-rem per year frequency. NMPNS 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 plant's licensing basis. NMPNS 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 states that a pRA used in risk-informed regulation should be performed in a manner that is consistent with accepted practices. In RIS 2007-06, "Regulatory Guide 1.200 Implementation," the NRC clarified that for all risk-informed applications received after December 2007, the NRC staff will use RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," to determine whether the technical adequacy of the PRA used to support a submittal is consistent with accepted practices. In the Final Safety Evaluation for NEI 94-01, Revision 2, (ML081140105) the NRC staff states that Capability Category I of the ASME PRA Standard shall be applied as the standard for assessing PRA quality for ILRT extension applications since approximate values of core damage frequency (CDF) and large early release frequency (LERF) and their contribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.
- 13 In accordance with this guidance, NMPNS addresses the technical adequacy of the PRA which forms the basis for the subject risk assessment in its application and supplemental letter. As described therein, a full update of the Level 1 and 2 internal events PRA model for NMP1 was completed in January 2008 with the objective of meeting RG 1.200 and ASME PRA Standard Capability Category II requirements. An industry peer review team reviewed the updated PRA model in February 2008, and the resulting facts and observations were reviewed for applicability to the ILRT extension. A summary of the findings from the peer review, and an assessment of the impact of the findings on the risk assessment for the ILRT extension are provided in NMPNS's December 4, 2008, letter. The licensee further states that most of the findings are related to documentation, and that an assessment of the model changes required to address the unresolved peer review findings determined that the changes would have a negligible, if any, impact on the results of the risk assessment. Given that the licensee has (1) evaluated its PRA against RG 1.200 and the ASME PRA Standard, (2) evaluated all of the findings developed during the reviews of its PRA for applicability to the ILRT extension, and (3) determined that any unresolved issues would not impact the conclusions of the ILRT risk assessment, the NRC staff concludes that the current NMP1 PRA model is of sufficient technical quality to support the evaluation of changes to ILRT frequencies.
RG 1.174 provides risk-acceptance guidelines for assessing the increases in CDF and LERF for risk-informed license amendment requests. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. NMPNS has estimated the change in LERF for the proposed change based on the cumulative change from the original frequency of three tests in a 1O-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. NMPNS estimated the change in the conditional containment failure probability for the proposed change to demonstrate that the defense-in-depth philosophy is met.
The following comparisons of risk are based on a change in test frequency from three tests in 10 years (the test frequency under Appendix J, Option A) to one test in 15 years. This bounds the impact of extending the test frequency from one test in 10 years to one test in 15 years. The following conclusions can be drawn from the analysis associated with extending the Type A test frequency:
- 1. Given the change from a three in 1O-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. 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 about 3.6 x 10-8 per year based on the plant-specific internal events PRA, and 2.1 x 10.7 per year when external events are included. 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
- 14 (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 licensee's risk analysis considered the potential impact of age-related corrosion/degradation in inaccessible areas of the containment shell on the proposed change. The increase in LERF associated with corrosion events is estimated to be about 5 x 10-8 per year.
Pursuant to RG 1.174, when the calculated increase in LERF is in the range of 7
10- per year to 10-6 per year, applications are considered if the total LERF is less than 10-5 per year. Based on information provided by the licensee, the total LERF for internal and external events, including the requested change, is about 1.9 x 10-6 per year, which meets the total LERF criteria. 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 approximately one percentage point 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 NRC 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.
3.11 Deterministic Analysis NMPNS provided the results of the last two NMP1 ILRT tests. These tests were satisfactory with leakage rates for the 1993 and 1999 Type A tests being 0.4634 wt%/day and 0.5045 wt%/day, respectively. These results are less than the maximum allowable containment leakage rate (La at Pa), of 1.5% containment air weight per day at a pressure of 35 psig.
NMPNS states that containment inspections will continue to be performed during the proposed 5-year extension of the Type A test interval (June 2009 through June 2014), in accordance with the IWE lSI program. The IWE lSI program requires a minimum of one inspection during each inspection period of the inspection interval. This extension will coincide with the first and second inspection periods of the second interval of the IWE lSI program.
ASME Code,Section XI, Subsection IWE-1240 requires that surface areas likely to experience accelerated degradation and aging require augmented examinations. NMP1 has identified such
- 15 areas on the drywell shell and has performed augmented inspections in accordance with IWE 1240.
The drywell augmented inspection/monitoring plan is more effective than Appendix J Type A tests for identifying degrading minimum wall conditions, since the Type A test will only identify an actual breach in the pressure boundary.
NMP1 monitors torus wall thickness and corrosion rate to provide reasonable assurance that the minimum wall thickness is not reached. Determination of torus corrosion rates is an ongoing activity that considers inspection results and the remaining corrosion allowance. The NMPI Torus Corrosion Monitoring Program includes the following elements: Periodic torus wall UT thickness measurements are obtained over a pre-defined grid system, torus material coupons are periodically removed, analyzed, and compared to the results of the UT measurements, visual inspections of accessible external surfaces of the torus support structure are performed.
The Torus Corrosion Monitoring Program is more effective than Appendix J Type A tests for identifying degrading minimum wall conditions, since the Type A test will only identify an actual breach in the pressure boundary.
During power operation, the NMP I primary containment is inerted with nitrogen to maintain oxygen concentration within TS limits. As a result, the primary containment is maintained at a slightly positive pressure during power operation. Maintaining the containment pressurized at power assures that gross containment leakage that may develop during power operation will be detected. A drywell pressure alarm will alert operators that the drywell pressure is abnormal (greater than 2 psig or less than 1 psig). Surveillance Requirement 4.3.2 "Pressure Suppression System Pressure And Suppression Chamber Water Temperature And Level" provides for monitoring the pressure.
NMP1 has reviewed GL 87-05, "Request for Additional Information Assessment of Licensee Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells." GL 87-05 described drywell shell degradation that occurred at Oyster Creek Nuclear Generating Station because of water intrusion into the air gap between the outer drywell surface and the surrounding concrete and subsequent wetting of the sand cushion at the bottom of the air gap.
This area of the drywell is not preserved with a protective coating as the wall surface above it is.
In response to this GL, NMP1 performed additional inspections. Five of the 10 drains were inspected by inserting a fiber optical probe from the torus room open ended drain line all the way up to the sand cushion area under the drywell shell. There were no observed indications of water leakage into the sand cushion area.
In addition to the inspections performed in accordance with the IWE lSI program, NMP1 performs periodic visual inspections of the coating on accessible interior surfaces of the drywell shell and drywell head are performed to identify evidence of deterioration. The inspections are performed every refueling outage to identify any visible defects including blistering, cracking, flaking, peeling, and physical or mechanical damage. When degraded coatings are identified, evaluations are performed to determine any necessary actions (e.g., repair, removal, or replacement).
- 16 3.12 NRC Staff Conclusion On the basis of technical evaluation above, the NRC staff concludes that the proposed TS amendment request for a one-time 5-year extension to Type A containment integrated leak rate test interval for Nine Mile Point Unit 1 is acceptable. According to Section 6.5.7 of the proposed TS Amendment, the Type A test shall be performed no later than June 8, 2014.
The existing 10 CFR Part 50 Appendix J, Type B and Type C tests will continue to be performed at the scheduled frequency. Those areas which are most likely to develop leaks (penetrations, primary containment isolation valves, containment hatch, etc.) will be identified through the Type B and Type C testing.
NMP1 has established procedures for performing non-destructive examination of the accessible surfaces of the containment for detection of structural problems and corrosion. The licensee has indicated that they will continue to perform the examinations in the extended interval for the Type A test.
Based on the foregoing evaluation, the NRC staff finds that the interval until the next Type A test at NMP1 may be extended to 15 years, and that the proposed change to Section 6.5.7 of the NMP1 TSs is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official provided comments via e-mail and were considered in the staff's review of the licensee's application. The comments can be viewed in ADAMS (Accession No. ML090340296).
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 and changes SRs.
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 the finding issued on October 21,2008 (73 FR 62566). 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.
-17
6.0 CONCLUSION
The Commission has concluded, on the basis of 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 Commission's 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: F. Farzam B. Heida R. Palla Date: March 11, 2009
The amendment revises NMP1 Technical Specification (TS) 6.5.7, "10 CFR 50 [Part 50 of Title 10 of the Code of Federal Regulations] Appendix J Testing Program Plan," to allow a one-time extension of the Integrated Leak Rate Test (ILRT) interval for no more than 5 years. The proposed amendment would allow the next ILRT for NMP1 to be performed within 15 years from the last ILRT as opposed to the current 1O-year interval.
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely,
/raJ Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosures:
- 1. Amendment No. 202 to DPR-63
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
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ADAMS Accession No.: ML090430367