ML080100186
| ML080100186 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/28/2008 |
| From: | Mahesh Chawla NRC/NRR/ADRO/DORL/LPLIII-1 |
| To: | Wadley M Nuclear Management Co |
| Chawla M, NRR/DORL/LPL3-1, 301-415-8371 | |
| Shared Package | |
| ML080100073 | List: |
| References | |
| TAC MD4209, TAC MD4210 | |
| Download: ML080100186 (16) | |
Text
January 28, 2008 Mr. Michael D. Wadley Site Vice President Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: (TAC NOS. MD4209 AND MD4210)
Dear Mr. Wadley:
The Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 183 to Facility Operating License No. DPR-42 and Amendment No. 173 to Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP),
respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated January 29, 2007, supplemented by letters dated November 19, 2007, and December 13, 2007 (Agencywide Documents and Management System Accession Nos. ML070300113, ML073240077, and ML073470821 respectively).
The amendments revise the TS 3.5.3 ECCS [Emergency Core Cooling System] Shutdown for PINGP to change operability requirements for the safety injection (SI) subsystem by addition of a Note to the Limiting Condition for Operation 3.5.3 One ECCS train shall be OPERABLE.
The Note states An SI train may be considered OPERABLE when the pump is capable of being manually started from the control room.
A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosures:
- 1. Amendment No. 183 to DPR-42
- 2. Amendment No. 173 to DPR-60
- 3. Safety Evaluation cc w/encls: See next page
ML080100186, Package: ML080100073, TS: ML080140425 OFFICE NRR/LPL3-1/PM NRR/LPL3-1/LA NRR/LPL3-1/LA DSS/SPWB/BC DIRS/ITSB/BC OGC/NLO w/comts NRR/LPL3-1/(A)BC NAME MChawla BTully THarris GCranston GWaig MSmith JCushing for PMilano DATE 1/16/08 1/15/08 1/25/08 1/28/08 1/28/08 1/28/08 1/28/08
Prairie Island Nuclear Generating Plant, Units 1 and 2 cc:
Jonathan Rogoff, Esquire Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Manager, Regulatory Affairs Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089 Manager - Environmental Protection Division Minnesota Attorney General=s Office 445 Minnesota St., Suite 900 St. Paul, MN 55101-2127 U.S. Nuclear Regulatory Commission Resident Inspector's Office 1719 Wakonade Drive East Welch, MN 55089-9642 Administrator Goodhue County Courthouse Box 408 Red Wing, MN 55066-0408 Commissioner Minnesota Department of Commerce 85 7th Place East, Suite 500 St. Paul, MN 55101-2198 Tribal Council Prairie Island Indian Community ATTN: Environmental Department 5636 Sturgeon Lake Road Welch, MN 55089 Nuclear Asset Manager Xcel Energy, Inc.
414 Nicollet Mall, R.S. 8 Minneapolis, MN 55401 Michael B. Sellman President and Chief Executive Officer Nuclear Management Company, LLC 700 First Street Hudson, MI 54016 Dennis L. Koehl Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Joel P. Sorenson Director, Site Operations Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089 July 2006
NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.183 License No. DPR-42
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nuclear Management Company, LLC (the licensee), dated January 29, 2007, supplemented by letters dated November 19, 2007, and December 13, 2007, 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 Facility Operating License No. DPR-42 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 183, 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 the date of its issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/JCushing for/
Patrick D. Milano, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: January 28, 2008
NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.173 License No. DPR-60
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nuclear Management Company, LLC (the licensee), dated January 29, 2007, supplemented by letters dated November 19, 2007, and December 13, 2007, 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 Facility Operating License No. DPR-60 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 173, 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 the date of its issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/JCushing for/
Patrick D. Milano, Acting Branch Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: January 28, 2008
ATTACHMENT TO LICENSE AMENDMENT NOS. 183 AND 173 FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Facility Operating License No. DPR-42 and DPR-60 with the attached revised pages. The changed areas are identified by a marginal line.
REMOVE INSERT DPR-42, License Page 3 DPR-42, License Page 3 DPR-60, License Page 3 DPR-60, License Page 3 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 3.5.3-1 3.5.3-1
(4)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NMC 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, NMC to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6)
Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to transfer byproduct materials from other job sites owned by Northern States Power Company for the purpose of volume reduction and decontamination.
C.
This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l: Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level NMC is authorized to operate the facility at steady state reactor core power levels not in excess of 1650 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 183, are hereby incorporated in the license. NMC shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection NMC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Prairie Island Nuclear Generating Plant Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program," Revision 1, submitted by letters dated October 18, 2006, and January 10, 2007.
Unit 1 Amendment No. 183
(5)
Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to transfer byproduct materials from other job sites owned by Northern States Power Company for the purposes of volume reduction and decontamination.
C.
This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l: Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level NMC is authorized to operate the facility at steady state reactor core power levels not in excess of 1650 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 173, are hereby incorporated in the license. NMC shall operate the facility in accordance with the Technical Specifications.
(3)
Physical Protection NMC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Prairie Island Nuclear Generating Plant Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program," Revision 1, submitted by letters dated October 18, 2006, and January 10, 2007.
Unit 2 Amendment No. 173
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 183 TO FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 173 TO FACILITY OPERATION LICENSE NO. DPR-60 NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
By letter to the Nuclear Regulatory Commission (NRC, the Commission) dated January 29, 2007, (Agencywide Documents and Management System (ADAMS) Accession No. ML070300113), as supplemented by letters dated November 19, 2007, (ADAMS Accession No. ML073240077), and December 13, 2007 (ADAMS Accession No. ML073470821), Nuclear Management Company, LLC, (NMC, the licensee) requested to amend Technical Specification (TS) 3.5.3 ECCS [Emergency Core Cooling System] Shutdown for Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2.
The proposed changes would clarify TS 3.5.3, ECCS Shutdown operability requirements for the safety injection (SI) subsystem by addition of a Note to the Limiting Condition for Operation (LCO) 3.5.3 One ECCS train shall be OPERABLE. The Note states An SI train may be considered OPERABLE when the pump is capable of being manually started from the control room. Thus, it clarifies that positioning the main control room SI pump hand switch to other than normal Auto does not conflict with design assumptions or surveillance requirements during the TS 3.5.3 APPLICABILITY - MODE 4 when both RCS cold leg temperatures are > SI pump disable temperature specified in PTLR, (e.g. 218 F to 350 F). Thus the note clarifies that taking the above action does not require operators to declare the associated ECCS train INOPERABLE under TS 3.5.3.
During startup and shutdown transition operations, the requirements of TS 3.5.3 overlap the requirements of TS 3.4.12 Low Temperature Overpressure Protection (LTOP)-Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature and TS 3.4.13 Low Temperature Overpressure Protection (LTOP) - Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature.
Whereas TS 3.5.3 requires one (1) OPERABLE ECCS train, TS 3.4.12 requires, in part, a maximum of one SI pump capable of injecting. TS 3.4.13 requires, in part, no SI pumps capable of injecting. The licensee expressed intent to satisfy TS 3.4.12 and 3.4.13 by positioning the SI pump hand switch(es) to pull-out, which prevents automatic pump start, and by placing blocking device over the hand switch(es) which, combined with the above, prevents a single main control room operator action from initiating injection. The LCO 3.5.3 note supports taking the above action without declaring ECCS INOPERABLE under 3.5.3.
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations 50.36, Technical Specifications The license amendment request (LAR) proposed to amend TS in a manner that complies with 10 CFR 50.36. The LAR included a summary statement of the bases or reasons for the proposed amendment and was derived from the analyses and evaluation included in the safety analysis report. The LCO, as amended by the NOTE, continues to specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. The amendment involved no changes to existing SRs which continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
Atomic Energy Commission (AEC) General Design Criteria (GDC) 37 - Engineered Safety Features Basis for Design According to the licensee, the construction of PlNGP was significantly complete prior to issuance of 10 CFR 50, Appendix A, GDC. PINGP was designed and constructed to comply with the AEC GDC as proposed on July 10, 1967, and as described in the Updated Safety Analysis Report.
AEC GDC proposed criterion 37 provides design guidance for the operating capability of the SI system. Specifically:
Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems.
As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assuming unobstructed discharge from both ends.
The amendments do not alter the provided engineered safety features designed ability to cope with any size reactor coolant pressure boundary break.
Standard Technical Specifications, Westinghouse Plants (NUREG-1431) & PINGP TS NUREG-1431 provides format and content guidance for TSs for plants with Westinghouse NSSS and has been approved for use by the NRC.
NUREG-1431 and the PINGP TS provide the following common guidance to determine OPERABILITY of TS-scoped systems, subsystems, and components (SSCs).
2.1 Definitions (same for NUREG-1431 and PINGP)
OPERABLE - OPERABILITY - A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling-and seal water, lubrication, and other auxiliary equipment that are required for the SSC or device to perform its specified safety function(s) are also capable of performing their related support function(s).
The LAR does not alter this definition nor does it alter the LCO requirements necessary to satisfy the definition of OPERABILITY in the TS 3.5.3 APPLICABILITY MODE.
NUREG-1431 and PINGP Guidance Specific to TS 3.5.3 ECCSShutdown NUREG-1431 and PINGP TS 3.5.3 guidance are very similar. The NUREG-1431 Bases state, in part, that Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic SI actuation is not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.
The PINGP Bases state, in part, that due to the lower pressure and temperatures in the RCS.
The probability of occurrence of a DBA is reduced. Therefore, the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic SI actuations are not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.
The LAR does not alter the applicable Bases or analyses beyond adding the further clarification nor does it alter the LCO requirements necessary to satisfy the definition of OPERABILITY in the TS 3.5.3 APPLICABILITY MODE.
NUREG-0800, Standard Review Plan (SRP)
PlNGP was not licensed to NUREG-0800, "Standard Review Plan (SRP)."
3.0 TECHNICAL EVALUATION
Prairie Island has two separate subsystems for ECCS, namely Safety Injection and residual heat removal. Both subsystems consist of two 100 percent capacity trains, which are interconnected and redundant, such that either train is capable of supplying 100 percent of the flow required to mitigate the accident consequences. Low temperature overpressure protection (LTOP) is provided with two pressurizer power operated relief valves with lift settings within the temperature range of 310 F to 218 F as specified in the pressure temperature limit report.
In the applicability for TS LCO 3.5.3, ECCS - Shutdown, both RCS cold leg temperatures are greater than the SI pump disable temperature specified in the PTLR. Currently this is between 218 F and 350 F.
The licensee proposed a Note stating, An SI train may be considered OPERABLE when the pump is capable of being manually started from the control room.
An SI pump auto-start function is required during Modes 1, 2, and 3 operation. TS SR 3.5.2.6 Requires verification that each ECCS pump starts automatically on an actual or simulated actuation signal. It is applicable to Modes 1, 2, and 3. LCO 3.5.3, ECCS Shutdown does not impose this SR.
The objective of adding the note to LCO 3.5.3 is to further clarify that placement of main control room SI pump hand-switches in pull-to-lock does not render ECCS INOPERABLE during TS 3.5.3 APPLICABILITY. Based on information provided by the licensee, the NRC staff
determined that, during LCO 3.5.3 APPLICABILITY, ECCS pump auto-start is:
Not a functional capability or performance level of equipment required for safe operation of the facility; Not assumed in the analyses and evaluation included in the safety analysis report; Not provided for in the current plant design (e.g. no auto-start signals are generated);
Not needed to satisfy Atomic AEC GDC 37 - Engineered Safety Features Design Basis, which applies to the current license; and Not required by SR 3.5.3.1 to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Specifically, SR 3.5.3.1 references TS 3.5.2 ECCS-Operating SRs applicable to LCO 3.5.3. but notably excludes SR 3.5.2.6 (verification that each ECCS pump starts automatically).
In their application, the licensee requested to amend LCO 3.5.3. The LCO 3.4.12 and 3.4.13 were also discussed in the application and during the course of the review, but neither an amendment nor NRC staff interpretation of LCO 3.4.12 or 3.4.13 was specifically requested by the licensee. Therefore, the issuance of this amendment does not convey a NRC staff position or endorsement of the licensees approach to meeting LCOs 3.4.12, 3.4.13, or any other Technical Specification.
Summary The NRC staff has reviewed the proposed TS change to add a Note to the TS 3.5.3, ECCS Shutdown, LCO. The licensee evaluation complies with 10 CFR 50.36 and AEC GDC 37 requirements. This TS change does not materially alter the existing LCO 3.5.3.
Therefore, the NRC staff has concluded that this proposed TS change is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change the surveillance requirements. The 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 (72 FR11392). 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
- 1.
Thomas J. Palmisano (NMC) letter to USNRC, License Amendment Request (LAR) to Revise Technical Specification (TS) 3.5.3 Operability Requirements for Safety Injection (SI) Subsystem, dated January 29, 2007.
- 2.
Michael D. Wadley (NMC) letter to USNRC, Supplement to License Amendment Request (LAR) to Revise Technical Specification (TS) 3.5.3 Operability Requirements for Safety Injection (SI) Subsystem, dated November 19, 2007.
- 3.
NUREG-1431, Revision 3, Standard Technical Specifications Westinghouse Plants, dated April 1995.
Principal Contributors: K. Desai, NRR R. Telson, NRR Date: January 28, 2008