ML022170769
| ML022170769 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/05/2002 |
| From: | Tam P NRC/NRR/DLPM/LPD1 |
| To: | Conway J Nine Mile Point |
| Tam P, NRR/DLPM, 415-1451 | |
| References | |
| TAC MA3208 | |
| Download: ML022170769 (16) | |
Text
August 5, 2002 Mr. John T. Conway Site Vice President Nine Mile Point Nuclear Station, LLC Lycoming, NY 13093
SUBJECT:
NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE: INSERVICE TESTING REQUIREMENTS IN THE TECHNICAL SPECIFICATIONS (TAC NO. MA3208)
Dear Mr. Conway:
The Commission has issued the enclosed Amendment No. 173 to Facility Operating License No. DPR-63 for Nine Mile Point Nuclear Station, Unit No. 1. The amendment consists of changes to the Technical Specifications in response to your application transmitted by letter dated November 26, 2001, as supplemented by letter on May 20, 2002.
The amendment deletes Section 3/4.2.6, Inservice Inspection and Testing, revises Section 4.2.7, Reactor Coolant System Isolation Valves, adds a new Section 6.17, Inservice Testing Program, and deletes several reporting requirements in Section 6.9.3, Special Reports.
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,
/RA/
Peter S. Tam, Senior Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosures:
- 1. Amendment No. 173 to DPR-63
- 2. Safety Evaluation cc w/encls: See next page
August 5, 2002 Mr. John T. Conway Site Vice President 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 RE: INSERVICE TESTING REQUIREMENTS IN THE TECHNICAL SPECIFICATIONS (TAC NO. MA3208)
Dear Mr. Conway:
The Commission has issued the enclosed Amendment No. 173 to Facility Operating License No. DPR-63 for Nine Mile Point Nuclear Station, Unit No. 1. The amendment consists of changes to the Technical Specifications in response to your application transmitted by letter dated November 26, 2001, as supplemented by letter on May 20, 2002.
The amendment deletes Section 3/4.2.6, Inservice Inspection and Testing, revises Section 4.2.7, Reactor Coolant System Isolation Valves, adds a new Section 6.17, Inservice Testing Program, and deletes several reporting requirements in Section 6.9.3, Special Reports.
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,
/RA/
Peter S. Tam, Senior Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosures:
- 1. Amendment No. 173 to DPR-63
- 2. Safety Evaluation cc w/encls: See next page Package Number:
Accession Number: ML022170769 TSs:
OFFICE PDI-1\\PM PDI-1\\LA RORP\\SC OGC PDI-1\\SC NAME PTam SLittle RDennig*
AHodgdon RLaufer DATE 7/30/02 7/30/02 7/11/02 8/2/02 8/5/02
- SE transmitted by memo of 7/11/02; SE used essentially as-is OFFICIAL RECORD COPY
DATED: August 5, 2002 AMENDMENT NO. 173 TO FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 PUBLIC PDI R/F RLaufer SLittle PTam OGC GHill (2)
PHearn ACRS BPlatchek, RI cc: Plant Service list
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. 173 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 November 26, 2001, as supplemented by letter dated May 20, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this 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. 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. 173, is hereby incorporated into this license.
Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Richard J. Laufer, Chief,Section I Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 5, 2002
ATTACHMENT TO LICENSE AMENDMENT NO. 173 TO FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 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 Pages Insert Pages ii ii vi vi 105 105 106 106 108 108 368 368 374 374
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 173 TO 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 November 26, 2001, as supplemented May 20, 2002, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.90, Nine Mile Point Nuclear Station, LLC (NMPNS, the license) applied for an amendment to Operating License DPR-63 for Nine Mile Point Nuclear Station, Unit No. 1 (NMP1). This application proposes the following changes to the NMP1 Technical Specifications (TSs): (1) delete Section 3/4.2.6, Inservice Inspection and Testing and the associated Bases, (2) revise Section 4.2.7, Reactor Coolant System Isolation Valves and the associated Bases, (3) add Section 6.17, Inservice Testing Program, and (4) delete several reporting requirements in Section 6.9.3, Special Reports. The May 20, 2002, supplemental letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
The proposed revisions to the NMP1 TSs are consistent with Nuclear Regulatory Commission (NRC) guidance and the Improved Standard Technical Specifications (ITS) for General Electric (GE) Boiling Water Reactor (BWR)/4 and BWR/6 plants (NUREG-1433 and NUREG-1434, respectively). Most of these changes reflect those already adopted for the Nine Mile Point Nuclear Station, Unit No. 2 (NMP2) TSs, which has previously adopted the ITS. Accordingly, these changes to NMP1 TS will provide uniformity in process and practice between NMP1 and NMP2, reduce duplicate and burdensome requirements and enhance plant operations.
2.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act (Act) requires a nuclear power plant operating license to include TSs as part of the license. The technical specifications ensure the operational capability of structures, systems and components that are required to protect the health and safety of the public. The Commissions regulatory requirements that are related to the content of the TSs are contained in 10 CFR 50.36, which includes the following categories: (1) safety limits, limiting safety systems settings and control settings; (2) limiting conditions for operation (LCO); (3) surveillance requirements; (4) design features; and (5) administrative controls.
In general, there are two classes of TSs changes: (1) changes needed to reflect modifications to the design basis (requirements are derived from the design basis), and (2) voluntary changes to take advantage of the evolution in policy and guidance as to the required content and preferred format of the TSs over time. This amendment deals with only the second class of changes. In determining the acceptability of such changes, the NRC staff interprets the requirements of the current version of 10 CFR 50.36, using as a model the accumulation of generically approved guidance in the ITS. For this review the NRC staff used NUREG-1433, Rev. 2, dated October 10, 2001.
NUREG-1433, Rev. 2, incorporates the general guidance and LCO scoping criteria provided by the Commissions Final Policy Statement on Technical Specification Improvement for Nuclear Power Reactors, published in the Federal Register on July 23, 1993 (58 FR 39132) and incorporated in 10 CFR 50.36 as of August 18, 1995.
Within this general framework, licensees may remove material from their TSs on two conditions: (1) the material is not required to be in the TSs based on the NRC staff interpretation of 10 CFR 50.36, including judgments about the level of detail required in the TSs, and (2) there exist suitable alternative regulatory controls for the material. The alternate regulatory controls that the licensee is proposing to provide specific details of the affected TS are the following references:
- a. 10 CFR 50.55a, Codes and standards;
- b.Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code);
c.
ASME Code Section XI/OM Part 10;
- e. Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995;
- f. 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors; and
- g. 10 CFR 50.73, Licensee Event Report System.
These references are acceptable regulatory controls since they are either from 10 CFR, NRC RGs or industry standards that are approved by the NRC staff.
Licensees may revise the TSs to adopt current ITS format and content provided that plant-specific review supports a finding of continued adequate safety, provided: (1) the change is editorial, administrative or provides clarification (i.e., no requirements are materially altered), (2) the change is more restrictive than the licensees current requirement, or (3) the change is less restrictive than the licensees current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards. The detailed application of this general framework, and additional specialized guidance, are discussed in Section 3.0 in the context of specific proposed changes.
3.0 TECHNICAL EVALUATION
3.1 Section 3/4.2.6, Inservice Inspection [ISI] and Testing The licensee proposed to delete Section 3/4.2.6, and add a new Section 6.17, to implement the inservice testing (IST) Program in accordance with 10 CFR 50.55a, and consistent with Section 5.5.7 of NUREG-1433. The licensee proposed these modifications in order to make the NMP1 TSs consistent with the improved STS and the NMP2 TSs.
The licensee is required to ensure the operational capability of Quality Group A, B, and C components by Sections 3.2.6 and 4.2.6. The Quality A, B, and C components include safety-related component supports, pressure vessels, piping, pumps and valves. The regulatory basis for these requirements is 10 CFR 50.55a, which requires the licensee to provide an IST Program that complies with Section XI of the ASME Code and applicable addenda in order to demonstrate the operational capability of Quality Group A, B, and C components. Furthermore, 10 CFR 50.55a requires that the licensee also conduct inservice inspections (ISI) in accordance with Section XI of the ASME Code and applicable addenda in order to ensure the operational capability of ASME Code Class 1, Class 2, and Class 3 classified components.
The proposed Section 6.17 establishes programmatic requirements for the 10 CFR 50.55a IST Program. The new Section 6.17 reads as follows:
This program provides controls for inservice testing of Quality Group A, B, and C pumps and valves.
- a. Inservice testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with the requirements for American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components specified in Section XI of the applicable ASME Boiler and Pressure Vessel Code Edition and Addenda, subject to the applicable provisions of 10CFR50.55a;
- b. The provisions of Specification 4.0.1 are applicable to the normal and accelerated testing frequencies for performing inservice testing activities; c.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
The ISI and IST requirements in the NMP1 TSs duplicate requirements of 10 CFR 50.55a. The NRC staff has reviewed the licensees proposal to remove Section 3/4.2.6 from the TSs and concludes that this portion of the amendment request is acceptable because the essential requirements are retained in the proposed Section 6.17. The NRC staff concludes that the proposed Section 6.17 is acceptable because it implements the requirements of 10 CFR 50.55a. Furthermore, the above provisions and clarifications are consistent with those in Section 5.5.7 Inservice Testing Program of the ITS (i.e., NUREG-1433).
3.2 Section 4.2.7, Reactor Coolant System Isolation Valves The licensee proposed to replace the quarterly testing requirements in Section 4.2.7 with periodic testing requirements of the proposed Section 6.17. Section 4.2.7 requires, at least once per quarter, all normally open power-operated isolation valves, excluding those in the feedwater and main steam line systems, be fully closed and reopened. The feedwater and main steam line valves must remain open and, therefore, testing during normal plant operation is impractical.
The requirements of the IST Program, required by 10 CFR 50.55a to follow the ASME Code,Section XI, include periodic testing and test frequencies for all reactor coolant system isolation valves that are necessary to assure their continued operability and isolation capability. This applies to all valves currently covered by Section 4.2.7. ASME Code Section XI/OM Part 10 permits valves to be tested quarterly, or during cold shutdown and refueling outages when it is not practical to perform valve testing during plant operation.
ASME Code, Section XI/OM Part 10 and 10 CFR 50.55a consider periodic testing without quarterly testing to be sufficient to assure that the reactor coolant system isolation valves are capable of mitigating the effects of reactor coolant loss by isolating the reactor coolant system in the event of a line rupture.
The NRC staff has reviewed the licensees proposal to delete Section 4.2.7, and concludes that this portion of the amendment request is acceptable since the requirements for periodic testing in the proposed Section 6.17 provide adequate testing procedures to assure the operability of the reactor coolant isolation valves.
3.3 Section 6.9.3, Special Reports
- a. Safety Class 1, 2, and 3 Reports Items b, c, and d require the licensee to submit, within 3 months, those ISI reports for Quality Group A, B, and C components, which are inspected pursuant to Section 4.2.6 and ASME Code,Section XI. NUREG-1433 does not contain guidance for such requirements corresponding to Section 6.9.3 of the NMP1 TSs. The licensee proposed to delete items b, c, and d from Section 6.9.3.
As discussed in 3.1 above, Section 4.2.6 will be removed from the TSs and reliance placed on ASME Code,Section XI as invoked in the new Section 6.17. ASME CodesSection XI requires preparation and submittal of ISI summary reports for Class 1 and Class 2 pressure retaining components and their supports. In a letter dated October 5, 2000, the NRC authorized the licensee to use ASME Code Case N-532 as an alternative to the ASME Code,Section XI for satisfying ISI summary report preparation and submittal requirements. ASME Code Case N-532 requires the preparation of an Owners Activity Report following each refueling outage and the submission of the Owners Activity Reports following the end of the inspection period to the NRC. The NRC staff concludes the proposed deletion of items b, c, and d from Section 6.9.3 is acceptable since the requirements of ASME Code Case N-532 will provide essential information for the NRC staff. In addition, should the NRC staff need more timely information, the NRC staff can inspect the information maintained onsite.
- b. Primary Containment Leakage Testing Report Item e requires reporting, within 3 months, the results of primary containment leakage testing conducted pursuant to Section 3.3.3, Leakage Rate. The licensee proposed to delete this reporting requirement since it is inconsistent with more recent requirements that govern leak rate testing, and which do not require a report to the NRC staff.
Section 6.16 of the TSs requires satisfying Option B of Appendix J, 10 CFR Part 50, by an Appendix J Testing Program Plan. Section 6.16 requires this testing program plan, with certain exceptions, to be in accordance with the guidelines contained in RG 1.163, Performance-Based Containment Leak-Test Program, dated September 1995. RG 1.163 endorses Revision 0 of Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J, dated July 26, 1995, with a few restrictions.
Section 12, Record Keeping, of NEI 94-01, Revision 0, requires that the licensee prepare and provide a post-outage inspection report onsite for internal and external review. Section 12 does not require the submission to NRC of the post-outage inspection report.
Based on the above review, the NRC staff concludes that the licensees proposal to delete Item e acceptable. The NRC staffs conclusion is based on the licensee conforming to the requirements of Section 12 of NEI 94-01, which is endorsed by RG 1.163.
- c. Secondary Containment Leakage Testing The licensee proposed to delete Item f, which requires reporting, within 3 months, the results of secondary containment leakage testing conducted pursuant to Section 3.4.1, Leakage Rate.
Appendix J to 10 CFR Part 50 was modified in 1995 to eliminate the general requirement for routine reporting of containment leak rate test results. Such reporting is not needed because routine test results are available onsite for NRC inspection, and test failures must be evaluated for reportability under 10 CFR 50.72 and 50.73. Serious or significant failures would be reportable under one or more of the reporting criteria in 10 CFR 50.72 and 50.73, such as: (1) condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; (2) the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety; (3) operation or condition that was prohibited by the plant's technical specifications; and (4) event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to shut down, remove residual heat, control the release of radioactive material, or mitigate an accident. This provides adequate notice to the NRC staff regarding containment leak rate test results.
Based on the above, the NRC staff concludes that the licensees proposal to delete Item f is acceptable. The NRC staffs conclusion is based on the licensee following the requirements of 10 CFR 50.72 and 10 CFR 50.73, which provide sufficient notification to the NRC staff of the leak tight integrity of the containment.
3.4 Conclusion of Technical Evaluation The NRC staff reviewed the licensees proposed changes for compliance with 10 CFR 50.36 and agreement with the guidelines set forth in NUREG-1433. As delineated above, the NRC staff concludes that the proposed changes to the NMP1 TSs are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment relates to changes in recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impacts statement or environmental assessment need be prepared in connection with the issuance of the amendment(s).
6.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Peter Hearn Date: August 5, 2002
Nine Mile Point Nuclear Station Unit No. 1 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 126 Lycoming, NY 13093 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul D. Eddy Electric Division NYS Department of Public Service Agency Building 3 Empire State Plaza Albany, NY 12223 Mr. William M. Flynn, President New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 Mark J. Wetterhahn, Esquire Winston & Strawn 1400 L Street, NW Washington, DC 20005-3502 Supervisor Town of Scriba Route 8, Box 382 Oswego, NY 13126 Mr. Michael J. Wallace President Nine Mile Point Nuclear Station, LLC c/o Constellation Energy Group 250 W. Pratt Street - 24th Floor Baltimore, MD 21201-2437 Mr. Raymond L. Wenderlich Senior Constellation Officer Responsible for Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093 Mr. James M. Petro, Jr., Esquire Counsel Constellation Power Source, Inc.
111 Market Place Suite 500 Baltimore, MD 21202