ML060890388
| ML060890388 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/10/2006 |
| From: | Stewart Bailey NRC/NRR/ADRO/DORL/LPLB |
| To: | Levis W Public Service Enterprise Group |
| Bailey S N | |
| References | |
| TAC MD1090 | |
| Download: ML060890388 (9) | |
Text
April 10, 2006 Mr. William Levis Senior Vice President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT RE:
DELETION OF COMPONENT IDENTIFICATION FOR OVERCURRENT PROTECTIVE DEVICES (TAC NO. MD1090)
Dear Mr. Levis:
The Commission has issued the enclosed Amendment No. 167 to Facility Operating License No. NPF-57 for the Hope Creek Generating Station (Hope Creek). This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 11, 2005. The amendment relocates the component identification from the overcurrent protective device lists in TS 3/4.8.4.1, Primary Containment Penetration Conductor Overcurrent Protective Devices, and TS 3/4.8.4.5, Class 1E Isolation Breaker Overcurrent Protective Devices, to the Hope Creek Updated Final Safety Analysis Report (UFSAR).
The Nuclear Regulatory Commission (NRC) staff has expedited its approval of this portion of your request to support maintenance activities in the upcoming Hope Creek refueling outage.
The NRC staff will issue the remainder of your request, to relocate the contents of TS 3/4.8.4.1 and TS 3/4.8.4.5 in their entirety to the UFSAR, in a separate safety evaluation and license amendment, if granted.
A copy of our Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Stewart N. Bailey, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-354
Enclosures:
- 1. Amendment No. 167 to License No. NPF-57
- 2. Safety Evaluation cc w/encls: See next page
ML060890388 OFFICE LPL1-2/PM LPL1-2/LA NRR/ITSB OGC LPL1-2/BC NAME SBailey: CM CRaynor TTjader MWoods DRoberts DATE 4/10/06 3/31/06 3/31/06 4/6/06 4/10/06
Hope Creek Generating Station cc:
Mr. Michael P. Gallagher Vice President - Eng/Tech Support PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Dennis Winchester Vice President - Nuclear Assessments PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. George P. Barnes Site Vice President - Hope Creek PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. George H. Gellrich Plant Support Manager PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Michael J. Massaro Plant Manager - Hope Creek PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Darin Benyak Director - Regulatory Assurance PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Ms. R. A. Kankus Joint Owner Affairs Exelon Generation Company, LLC Nuclear Group Headquarters KSA1-E 200 Exelon Way Kennett Square, PA 19348 Lower Alloways Creek Township c/o Ms. Mary O. Henderson, Clerk Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Dr. Jill Lipoti, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415 Mr. Brian Beam Board of Public Utilities 2 Gateway Center, Tenth Floor Newark, NJ 07102 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector Hope Creek Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038
PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 167 License No. NPF-57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by PSEG Nuclear LLC dated October 11, 2005, 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 set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 167, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Darrell J. Roberts, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 10, 2006
ATTACHMENT TO LICENSE AMENDMENT NO. 167 FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3/4 8-26 3/4 8-26 3/4 8-27 3/4 8-27 3/4 8-28 3/4 8-28 3/4 8-29 3/4 8-29 3/4 8-42 3/4 8-42 3/4 8-43 3/4 8-43
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 167 TO FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354
1.0 INTRODUCTION
By letter dated October 11, 2005, PSEG Nuclear LLC (PSEG or the licensee) requested changes to the Hope Creek Generating Station (Hope Creek) Technical Specifications (TSs).
PSEG proposed to relocate the TS requirements in TS 3/4.8.4.1, Primary Containment Penetration Conductor Overcurrent Protective Devices, and TS 3/4.8.4.5, Class 1E Isolation Breaker Overcurrent Protective Devices, to the Hope Creek Updated Final Safety Analysis Report (UFSAR).
To support maintenance activities in the upcoming Hope Creek refueling outages, the Nuclear Regulatory Commission (NRC or the Commission) is issuing this amendment to remove the component identification from TS 3/4.8.4.1 and TS 3/4.8.4.5. The NRC staff is still reviewing the proposal relocation of the remainder of the TS 3/4.8.4.1 and TS 3/4.8.4.5 requirements in their entirety to the UFSAR. This amendment is within the scope addressed by the proposed no significant hazards consideration determination as published in the Federal Register on March 6, 2006 (71 FR 11233).
2.0 REGULATORY EVALUATION
Section 50.36(c)(2)(ii) of Title 10 of the Code of Federal Regulations (10 CFR) contains the requirements for items that must be included in the TSs. This regulation states that a TS limiting condition for operation must 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.
Items not meeting any of these four criteria can be relocated from the TSs to a licensee-controlled document, such as the UFSAR. The licensee can then change the relocated requirements, if necessary, in accordance with the requirements of 10 CFR 50.59, Changes, tests, and experiments. The criteria and the NRC staffs evaluation of the component identifications proposed for relocation are discussed below.
3.0 TECHNICAL EVALUATION
NRC and industry representatives have developed guidelines for improving the content and quality of TSs. The utility owners groups and the NRC staff developed Standard Technical Specifications for each primary reactor type that would comply with the Commissions policy. In addition, the NRC staff, licensees, and owners groups developed a writers guide containing generic administrative and editorial guidelines for preparing TSs. The writers guide emphasized human factors principles for establishing the types of information in the TSs.
When requirements were shown to give little or no safety benefit, their removal from the TSs was deemed appropriate. This included the removal of detailed information from individual specifications. The Commission issued the NUREG-1433, Standard Technical Specifications General Electric BWR [boiling water reactor]/4, as a model for developing improved TSs for General Electric plants.
Currently, TS 3/4.8.4.1 requires the operability of the containment penetration conductor overcurrent protective devices. These devices are installed to minimize the damage from a fault in a component inside containment, or in the cabling that penetrates containment, which could damage an electrical penetration in such a way that the containment structure could be breached. Similarly, TS 3/4.8.4.5 requires the operability of the Class 1E isolation breaker overcurrent protective devices, which are installed to protect the Class 1E buses from overcurrent conditions. This amendment does not change the above requirements. This amendment relocates the component identification from these TSs to the Hope Creek UFSAR.
The design of the facility is required to be described in the UFSAR by 10 CFR 50.34. In addition, the quality assurance (QA) requirements of Appendix B to 10 CFR Part 50 require that plant design be documented in controlled procedures and drawings, and maintained in accordance with an NRC-approved QA plan. In 10 CFR 50.59, controls are specified, in part, for changing the facility as described in the UFSAR, and in 10 CFR 50.54(a) criteria are specified for changing the QA plan.
Relocating the details of the system design from the TSs to the UFSAR is acceptable because the information being relocated does not satisfy any of the 10 CFR 50.36(c)(2)(ii) criteria for retention in the TSs, and the changes to the UFSAR will be controlled by provisions of 10 CFR 50.59. Accordingly, the NRC staff finds that the component identification for the overcurrent protective devices may be relocated to the UFSAR.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New Jersey 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. 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 such finding (71 FR 11233). 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 Contributors: T. Valentine S. Bailey Date: April 10, 2006