ML050420165
| ML050420165 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/01/2005 |
| From: | Richard Ennis NRC/NRR/DLPM/LPD1 |
| To: | Kansler M Entergy Nuclear Operations |
| Boska J, NRR, 301-415-2901 | |
| References | |
| TAC MC5256 | |
| Download: ML050420165 (13) | |
Text
June 1, 2005 Mr. Michael Kansler President Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
VERMONT YANKEE NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: RELOCATION OF INSERVICE TESTING REQUIREMENTS TO A LICENSEE-CONTROLLED PROGRAM (TAC NO. MC5256 )
Dear Mr. Kansler:
The Commission has issued the enclosed Amendment No. 224 to Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station (Vermont Yankee), in response to your application dated December 7, 2004.
The amendment revises the Technical Specifications (TSs) by removing Surveillance Requirement (SR) 4.4.A.3 for testing the setting of the standby liquid control system pressure relief valves. Also, SR 4.5.A.5 was revised to remove stroke time specifications for the recirculation pump discharge valves. The design details related to the safety functions of these systems will continue to be contained in the Vermont Yankee Updated Final Safety Analysis Report. Testing of the pressure relief valves and the recirculation pump discharge valves will be performed in accordance with SR 4.6.E, which invokes the inservice testing requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(f).
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Richard B. Ennis, Senior Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271
Enclosures:
- 1. Amendment No. 224 to License No. DPR-28
- 2. Safety Evaluation cc w/encls: See next page
Mr. Michael Kansler President Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601
SUBJECT:
VERMONT YANKEE NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: RELOCATION OF INSERVICE TESTING REQUIREMENTS TO A LICENSEE-CONTROLLED PROGRAM (TAC NO. MC5256 )
Dear Mr. Kansler:
The Commission has issued the enclosed Amendment No. 224 to Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station (Vermont Yankee), in response to your application dated December 7, 2004.
The amendment revises the Technical Specifications (TSs) by removing Surveillance Requirement (SR) 4.4.A.3 for testing the setting of the standby liquid control system pressure relief valves. Also, SR 4.5.A.5 was revised to remove stroke time specifications for the recirculation pump discharge valves. The design details related to the safety functions of these systems will continue to be contained in the Vermont Yankee Updated Final Safety Analysis Report. Testing of the pressure relief valves and the recirculation pump discharge valves will be performed in accordance with SR 4.6.E, which invokes the inservice testing requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(f).
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Richard B. Ennis, Senior Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271
Enclosures:
- 1. Amendment No. 224 to License No. DPR-28
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
PUBLIC CRaynor DRoberts CAnderson, RGN-I PDI-2 Reading REnnis GHill (2) BPoole, OGC CHolden OGC TBoyce VBucci, OIG DLPM DPR ACRS DTerao ACCESSION NO.: ML050420165 OFFICE PDI-2/PM PDI-2/PM PDI-2/LA EMEB/SC IROB/SC NAME JBoska REnnis CRaynor DTerao TBoyce DATE 5/3/05 5/3/05 5/3/05 5/4/05 5/10/05 OFFICE OGC PDI-2/SC NAME KKannler VNerses for DRoberts DATE 5/13/05 6/1/05 OFFICIAL RECORD COPY
Vermont Yankee Nuclear Power Station cc:
Regional Administrator, Region I Ms. Carla A. White, RRPT, CHP U. S. Nuclear Regulatory Commission Radiological Health 475 Allendale Road Vermont Department of Health King of Prussia, PA 19406-1415 P.O. Box 70, Drawer #43 108 Cherry Street Mr. David R. Lewis Burlington, VT 05402-0070 Pillsbury, Winthrop, Shaw, Pittman, LLP 2300 N Street, N.W. Mr. James M. DeVincentis Washington, DC 20037-1128 Manager, Licensing Vermont Yankee Nuclear Power Station Ms. Christine S. Salembier, Commissioner P.O. Box 0500 Vermont Department of Public Service 185 Old Ferry Road 112 State Street Brattleboro, VT 05302-0500 Montpelier, VT 05620-2601 Resident Inspector Mr. Michael H. Dworkin, Chairman Vermont Yankee Nuclear Power Station Public Service Board U. S. Nuclear Regulatory Commission State of Vermont P.O. Box 176 112 State Street Vernon, VT 05354 Montpelier, VT 05620-2701 Director, Massachusetts Emergency Chairman, Board of Selectmen Management Agency Town of Vernon ATTN: James Muckerheide P.O. Box 116 400 Worcester Rd.
Vernon, VT 05354-0116 Framingham, MA 01702-5399 Operating Experience Coordinator Jonathan M. Block, Esq.
Vermont Yankee Nuclear Power Station Main Street 320 Governor Hunt Road P.O. Box 566 Vernon, VT 05354 Putney, VT 05346-0566 G. Dana Bisbee, Esq. Mr. John F. McCann Deputy Attorney General Director, Nuclear Safety Assurance 33 Capitol Street Entergy Nuclear Operations, Inc.
Concord, NH 03301-6937 440 Hamilton Avenue White Plains, NY 10601 Chief, Safety Unit Office of the Attorney General Mr. Gary J. Taylor One Ashburton Place, 19th Floor Chief Executive Officer Boston, MA 02108 Entergy Operations 1340 Echelon Parkway Ms. Deborah B. Katz Jackson, MS 39213 Box 83 Shelburne Falls, MA 01370
Vermont Yankee Nuclear Power Station cc:
Mr. John T. Herron Mr. Ronald Toole Sr. VP and Chief Operating Officer 1282 Valley of Lakes Entergy Nuclear Operations, Inc. Box R-10 440 Hamilton Avenue Hazelton, PA 18202 White Plains, NY 10601 Ms. Stacey M. Lousteau Mr. Danny L. Pace Treasury Department Vice President, Engineering Entergy Services, Inc.
Entergy Nuclear Operations, Inc. 639 Loyola Avenue 440 Hamilton Avenue New Orleans, LA 70113 White Plains, NY 10601 Mr. Raymond Shadis Mr. Brian OGrady New England Coalition Vice President, Operations Support Post Office Box 98 Entergy Nuclear Operations, Inc. Edgecomb, ME 04556 440 Hamilton Avenue White Plains, NY 10601 Mr. James P. Matteau Executive Director Mr. Michael J. Colomb Windham Regional Commission Director of Oversight 139 Main Street, Suite 505 Entergy Nuclear Operations, Inc. Brattleboro, VT 05301 440 Hamilton Avenue White Plains, NY 10601 Mr. William K. Sherman Vermont Department of Public Service Mr. John M. Fulton 112 State Street Assistant General Counsel Drawer 20 Entergy Nuclear Operations, Inc. Montpelier, VT 05620-2601 440 Hamilton Avenue White Plains, NY 10601 Mr. Jay K. Thayer Site Vice President Entergy Nuclear Operations, Inc.
Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500 Mr. Kenneth L. Graesser 38832 N. Ashley Drive Lake Villa, IL 60046 Mr. James Sniezek 5486 Nithsdale Drive Salisbury, MD 21801
ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 224 License No. DPR-28
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (the licensee) dated December 7, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows:
(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 224, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA by VNerses for/
Darrell J. Roberts, Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 1, 2005
ATTACHMENT TO LICENSE AMENDMENT NO. 224 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 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 92 92 97 97 102 102
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. DPR-28 ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.
VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271
1.0 INTRODUCTION
By letter dated December 7, 2004, Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (the licensee) submitted a request to amend the Vermont Yankee Nuclear Power Station (VYNPS) Technical Specifications (TSs). The proposed changes would revise the TSs by removing Surveillance Requirement (SR) 4.4.A.3 for testing the setting of the standby liquid control system (SLCS) pressure relief valves. SR 4.5.A.5 would be revised to remove stroke time specifications for the recirculation pump discharge valves. The design details related to the safety functions of these systems will continue to be contained in the VYNPS Updated Final Safety Analysis Report (UFSAR). Testing of the pressure relief valves and the recirculation pump discharge valves would be performed in accordance with SR 4.6.E, which invokes the inservice testing requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(f). TS Bases Sections 3.4 and 4.4 would be revised to delete the discussion of testing of the SLCS pressure relief valves.
2.0 REGULATORY EVALUATION
The construction permit for VYNPS was issued by the Atomic Energy Commission (AEC) on December 11, 1967. The plant was designed and constructed based on the proposed General Design Criteria (GDC) published by the AEC in the Federal Register (32 FR 10213) on July 11, 1967 (hereinafter referred to as draft GDC). The AEC published the final rule that added Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants, in the Federal Register (36 FR 3255) on February 20, 1971 (hereinafter referred to as final GDC).
Differences between the draft GDC and final GDC included a consolidation from 70 to 64 criteria. As discussed in the Nuclear Regulatory Commission (NRC or the Commission) Staff Requirements Memorandum for SECY-92-223 dated September 18, 1992 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971. At the time of promulgation of Appendix A to 10 CFR Part 50, the Commission stressed that the final GDC were not new requirements and were promulgated to more clearly articulate the licensing requirements and practice in effect at that time. Each plant
licensed before the final GDC were formally adopted had been evaluated on a plant-specific basis, determined to be safe, and licensed by the Commission.
As discussed in Appendix F of the VYNPS UFSAR, the licensees for VYNPS have made changes to the facility over the life of the plant that may have invoked the final GDC. The extent to which the final GDC have been invoked can be found in specific sections of the UFSAR and in other VYNPS design and licensing basis documentation.
For the SLCS pressure relief valves, based on a review of UFSAR Section 3.8, "Standby Liquid Control System," NUREG-0800, Standard Review Plan, Section 9.3.5, and the licensees submittal dated December 7, 2004, the staff identified the following GDC as being applicable to the proposed amendment:
! Draft GDC 27, Redundancy of Reactivity Control, which requires that at least two independent reactivity control systems shall be provided.
Also, 10 CFR 50.62, "Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants," applies to the SLCS.
For the recirculation pump discharge valves, based on a review of UFSAR Section 14.6.3, "Loss of Coolant Accident," NUREG-0800, Standard Review Plan, Section 15.6.5, and the licensees submittal dated December 7, 2004, the staff identified the following GDC as being applicable to the proposed amendment:
! Draft GDC 44, "Emergency Core Cooling Systems Capability," which requires that at least two emergency core cooling systems, preferably of different design principles, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction to negligible amounts for all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe.
In 10 CFR 50.36, the NRC established its regulatory requirements related to the content of TSs.
Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) SRs; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plants TSs.
On July 22, 1993 (58 FR 39132), the Commission published a Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (Final Policy Statement) which discussed the criteria to determine which items are required to be included in the TSs as LCOs. The criteria were subsequently incorporated into the regulations by an amendment to 10 CFR 50.36 (60 FR 36953). Specifically, 10 CFR 50.36(c)(2)(ii) requires that a TS LCO be established for each item meeting one or more of the following criteria:
Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
Existing LCOs and related surveillances included as TS requirements which satisfy any of the criteria stated above must be retained in the TSs. Those TS requirements which do not satisfy these criteria may be relocated to other licensee-controlled documents.
In general, there are two classes of changes to TSs: (1) changes needed to reflect modifications to the design basis (TSs are derived from the design basis), and (2) changes to take advantage of the evolution in policy and guidance as to the required content and preferred format of TSs over time. In determining the acceptability of such changes, the staff interprets the requirements of 10 CFR 50.36, using as a model the accumulation of generically approved guidance in the Improved Standard Technical Specifications (ISTS). For this review, the staff used NUREG-1433, Revision 3, Standard Technical Specifications, General Electric Plants BWR [boiling-water reactor]/4.
Within this general framework, licensees may remove material from their TSs if the material is not required to be in the TSs based on the NRC staffs interpretation of 10 CFR 50.36, including judgements about the level of detail required in the TSs. As discussed in the Final Policy Statement, the NRC staff reviews, on a case-by-case basis, whether enforceable regulatory controls are needed for the relocated material (e.g., 10 CFR 50.59). Licensees may revise the remaining TSs to adopt current ISTS format and content provided that plant-specific review supports a finding of continued adequate safety because: (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.
3.0 TECHNICAL EVALUATION
3.1 SLCS Pressure Relief Valves The VYNPS UFSAR states, "The safety objective of the Standby Liquid Control System is to provide a method, independent of the control rods, to shut down the reactor from the full power condition and maintain the reactor subcritical during cooldown." The SLCS is designed to allow a control room operator to manually initiate an SLCS pump to inject a sufficient quantity of neutron absorber solution into the reactor vessel to shut down the reactor at a steady rate. The SLCS is also designed to maintain the reactor subcritical during a cooldown to cold shutdown at a cooldown rate within the capacity of the shutdown cooling systems.
There are two positive-displacement pumps in the SLCS. Either one of the two pumps is capable of meeting the required neutron absorber injection rates. These pumps are capable of reaching very high discharge pressures, albeit at a fairly low flow rate. The pumps are not capable of overpressurizing the reactor vessel or the reactor coolant system (RCS) as the RCS safety and relief valves are capable of relieving the full flow from both pumps. One SLCS pressure relief valve is installed downstream of each pump to protect the SLCS piping and components from overpressure. The SLCS relief valves must be set high enough to ensure the design flow rate of neutron absorber solution will inject into the reactor vessel under the design transients, and low enough to protect the SLCS piping and components from overpressure; for example, in the event the pumps are actuated with the SLCS downstream valves closed. To prevent bypass flow from one pump in the event that a relief valve fails open on the other pump, a check valve is installed downstream of each relief valve in each pump discharge line.
VYNPS TS 4.4.A.3 currently specifies testing of the pressure relief valves to verify the setting is between a minimum (1400 pounds per square inch gauge [psig]) and a maximum (1490 psig) pressure setpoint. These settings are also specified in UFSAR Section 3.8.3. These settings do not meet any of the four criterion for LCOs that need to be included in the TSs, as stated in 10 CFR 50.36(c)(2)(ii). Refer to Section 2.0 of this Safety Evaluation (SE) for a detailed discussion of the four criterion. The fact that these settings are not included in NUREG-1433, Revision 3, is further evidence that the NRC staff does not believe these settings need to be in the TSs. Testing of the pressure relief valves will be performed in accordance with SR 4.6.E, which invokes the inservice testing requirements specified in 10 CFR 50.55a(f). The relief valve settings will be controlled in the UFSAR, where changes are subject to the requirements of 10 CFR 50.59, "Changes, tests, and experiments." Therefore, the NRC staff concludes that the deletion of TS 4.4.A.3 is acceptable because (1) its inclusion in the TSs is not specifically required by 10 CFR 50.36 or other NRC regulations, (2) the pressure relief valves continue to have a TS requirement for periodic testing (SR 4.6.E), and (3) the settings for the pressure relief valves are in the UFSAR and will be adequately controlled under the provisions of 10 CFR 50.59. The licensee has provided proposed revisions to the TS Bases which remove references to the SLCS pressure relief valve settings. The NRC staff concludes that the changes to the TS Bases are consistent with the changes to the TSs.
3.2 Recirculation Pump Discharge Valves The UFSAR describes the injection path for low-pressure coolant injection (LPCI) during a design basis loss-of-coolant accident (LOCA). The injection lines connect to the recirculation loops downstream of the recirculation pump discharge valves. The recirculation pump discharge valves and associated bypass valves are signaled to shut on a LPCI initiation signal and low reactor pressure. LPCI flow is thus directed to the reactor vessel through the normal recirculation flow path. Therefore, the safety function of the recirculation pump discharge valves is to prevent flow from the LPCI system from being diverted from the design flow path to the reactor vessel. This helps ensure the ability of the LPCI system to perform its safety function of providing core cooling following a LOCA.
VYNPS TS SR 4.5.A.5 is being revised to remove the stroke time requirements for the recirculation pump discharge valves. TS SR 4.5.A.5 states that the discharge valves shall be tested to verify full open to full closed in 27 to 33 seconds. The limit of 27 seconds has no safety significance, as the safety analysis has to assume the longest stroke time in the analysis.
UFSAR Section 7.4.3.5.4 discusses the valves controlled by the LPCI control circuitry. The time limit for the recirculation pump discharge valves to stroke from full open to full closed is discussed. These valve stroke times do not meet any of the four criterion for LCOs that need to be included in the TSs as stated in 10 CFR 50.36(c)(2)(ii). Refer to Section 2.0 of this SE for a detailed discussion of the four criterion. The fact that these valve stroke times are not included in NUREG-1433, Revision 3, is further evidence that the NRC staff does not believe these valve stroke times need to be in the TSs. Testing of the recirculation pump discharge valves will be performed in accordance with SR 4.6.E, which invokes the inservice testing requirements specified in 10 CFR 50.55a(f). The valve stroke times will be controlled in the inservice testing procedures or in the UFSAR, where changes are subject to the requirements of 10 CFR 50.59.
Therefore, the NRC staff concludes that the revision of TS 4.5.A.5 is acceptable because (1) the inclusion in TSs of recirculation pump discharge valve stroke times is not specifically required by 10 CFR 50.36 or other NRC regulations, (2) the recirculation pump discharge valves continue to have a TS requirement for periodic testing (SR 4.6.E), and (3) the valve stroke times will be adequately controlled under the provisions of 10 CFR 50.59 in the inservice testing procedures or the UFSAR.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Vermont State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.
The NRC staff has determined that the amendment involves no significant increase in amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (70 FR 2889). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement
or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: J. Boska Date: June 1, 2005