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| | number = ML14015A453 | | | number = ML14015A453 |
| | issue date = 01/30/2014 | | | issue date = 01/30/2014 |
| | title = Millstone Power Station Unit 2, Issuance of Amendment Revise Footnote in Technical Specification 3.1.3.7, Control Rod Drive Mechanism (TAC No. MF1172) | | | title = Issuance of Amendment Revise Footnote in Technical Specification 3.1.3.7, Control Rod Drive Mechanism |
| | author name = Kim J S | | | author name = Kim J |
| | author affiliation = NRC/NRR/DORL/LPLI-1 | | | author affiliation = NRC/NRR/DORL/LPLI-1 |
| | addressee name = Heacock D A | | | addressee name = Heacock D |
| | addressee affiliation = Dominion Nuclear Connecticut, Inc | | | addressee affiliation = Dominion Nuclear Connecticut, Inc |
| | docket = 05000336 | | | docket = 05000336 |
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| | page count = 11 | | | page count = 11 |
| | project = TAC:MF1172 | | | project = TAC:MF1172 |
| | stage = Issuance | | | stage = Approval |
| }} | | }} |
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| =Text= | | =Text= |
| {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 January 30, 2014 SUBJECT: MILLSTONE POWER STATION, UNIT NO.2 -ISSUANCE OF AMENDMENT RE: REVISE FOOTNOTE IN TECHNICAL SPECIFICATION 3.1.3.7, CONTROL ROD DRIVE MECHANISMS (TAC NO. MF1172) Dear Mr. Heacock: The Commission has issued the enclosed Amendment No. 317 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated March 21, 2013. The amendment revises Technical Specification (TS) 3.1.3.7-Control Rod Drive Mechanisms to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal for MPS2. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Docket No. 50-336 Enclosures: 1. Amendment No. 317 to DPR-65 2. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT, INC. DOCKET NO. 50-336 MILLSTONE POWER STATION, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 317 Renewed License No. DPR-65 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by the applicant dated March 21, 2013, 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. | | {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 30, 2014 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 |
| -2 -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-65 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 317, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of issuance, and shall be implemented within 60 days of issuance. Attachment: Changes to the License and Technical Specifications Date of Issuance: January 30, 2014 FOR THE NUCLEAR REGULA TORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 317 RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 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 3 Insert 3 Replace the following page of the Appendix A Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change. Remove 3/4 1-31 Insert 3/4 1-31
| | |
| -3-Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (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 1 0 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 1: 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; and is subject to all applicable provisions of the Act and 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 The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 317, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications. Renewed License No. DPR-65 Amendment No. 317 REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE MECHANISMS LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod drive mechanisms shall be de-energized. APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentration is less than refueling concentration of Specification 3. 9.1 . ACTION: With any of the control rod drive mechanisms energized, restore the mechanisms to their energized state within 2 hours or immediately open the reactor trip circuit breakers. SURVEILLANCE REQUIREMENTS 4.1.3.7 24 hours. The control rod drive mechanisms shall be verified to be de-energized at least once per
| | ==SUBJECT:== |
| * The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATING, the reactor coolant system temperature is greater than 500° F, the pressurizer pressure is greater than 2000 psi a and the requirements of Limiting Condition for Operation for Specification 3.3 .1.1, "Reactor Protective Instrumentation," are met. MILLSTONE -UNIT 2 3/41-31 Amendment No. 116,291 317 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 317 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65 1.0 INTRODUCTION DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION, UNIT NO.2 DOCKET NO. 50-336 By letter dated March 21, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13085A081), Dominion Nuclear Connecticut, Inc., (the licensee) requested changes to the Technical Specifications (TSs) for Millstone Power Station Unit 2 (MPS2). The proposed changes would revise TS 3.1.3.7-Control Rod Drive Mechanisms (CORMs) to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal. 2.0 REGULATORY EVALUATION In Title 10 of the Code of Federal Regulations (1 0 CFR) Section 50.36, the Commission 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 related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) Limiting Conditions for Operation (LCOs); (3) Surveillance Requirements (SRs); (4) design features; and (5) administrative controls. 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. | | MILLSTONE POWER STATION, UNIT NO.2 -ISSUANCE OF AMENDMENT RE: REVISE FOOTNOTE IN TECHNICAL SPECIFICATION 3.1.3.7, CONTROL ROD DRIVE MECHANISMS (TAC NO. MF1172) |
| -2 -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. 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 NRC 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 (STSs). For this review, the staff used NUREG-1432, Revision 4, "Standard Technical Specifications, Combustion Engineering Plants." The STSs reflect the general guidance and LCO seeping criteria provided by the Commission's Final Policy Statement. 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 staff's interpretation of 10 CFR 50.36, including judgments 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. Licensees may revise the remaining TSs to adopt current improved STS 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 licensee's current requirement; or (3) the change is less restrictive than the licensee's current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards. 3.0 TECHNICAL EVALUATION The Control Rod Drive Mechanism (CRDM) requirement of TS 3.1.3.7, Control Rod Drive Mechanisms, is provided to assure that the consequences of an uncontrolled control element assembly (CEA) withdrawal from a subcritical transient will stay within acceptable levels. This TS assures that reactor coolant system (RCS) conditions exist which are consistent with the plant safety analysis prior to energizing the CRDMs. The accident is precluded when CRDMs are de-energized and cannot withdraw a CEA. The CRDMs may be energized with the boron concentration greater than or equal to the refueling concentration since, under these conditions, adequate shutdown margin is maintained even if the CEAs are fully withdrawn from the core. The TS 3.1.3.7, Control Rod Drive Mechanisms, requires that CRDMs are de-energized in Modes 3*, 4, 5 and 6, whenever the RCS boron concentration is less than the refueling
| | |
| -3 -concentration specified in TS 3.9.1, Boron Concentrations. As described in the footnote (denoted by an asterisk in Mode 3) in TS 3.1.3.7, CRDMs may be energized in Mode 3 as long as the following conditions exist: (1) four reactor coolant pumps operating, (2) RCS temperature >500 F, (3) pressurizer pressure> 2000 psia, and (4) high power trip is operable. The TS Table 3.3-1, Reactor Protective Instrumentation, identifies, as a minimum, the reactor protective instrumentation channels and bypasses that must be operable under specific modes and conditions. For the condition where the protective system trip breakers are in the closed position and the CEA drive system is capable of CEA withdrawal (i.e., CRDMs energized), TS Table 3.3-1 requires the high power trip (Functional Unit 2), Reactor Protection System (RPS) logic matrices (Functional Unit 13), RPS logic matrix relays (Functional Unit 14) and reactor trip breakers (Functional Unit 15) to be operable. Currently, the footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, only requires the High Power Level Trip to be operable when CRDMs are energized and capable of CEA withdrawal. It does not include the additional operability requirements (i.e., RPS logic matrices, RPS logic matrix relays, and reactor trip breakers) that were added toTS Table 3.3-1, Reactor Protective Instrumentation under License Amendment 282 on September 25, 2003 (ADAMS Accession No. ML032270057). The proposed change would eliminate the inconsistency between TS 3.1.3.7 and TS Table 3.3-1 by revising the footnote in TS 3.1.3.7. Specifically, the footnote would be revised to delete "the high power trip is OPERABLE" and replace it with "the requirements of Limiting Condition for Operation for Specification 3.3.1.1, Reactor Protective Instrumentation, are met." The staff has determined that the current MPS2 TSs do not provide operability requirements for RPS logic matrices, RPS logic matrix relays, and reactor trip breakers in TS 3.1.3.7. The proposed change would provide these operability requirements. The staff finds that the proposed change is more restrictive than the current TS requirements and that the additional restrictions on plant operation would enhance safety. Therefore, the proposed change is acceptable. 4.0 STATE CONSULTATION In accordance with the Commission's regulations, the Connecticut 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 the 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 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 as published in the Federal Register on June 11, 2013(78 FR 35061 ). 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.
| | ==Dear Mr. Heacock:== |
| -4-6.0 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) there is reasonable assurance that 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. Principal Contributor: A. Guzzetta Date: January 30, 2014 Mr. David A Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 January 30, 2014 SUBJECT: MILLSTONE POWER STATION, UNIT NO.2-ISSUANCE OF AMENDMENT RE: REVISE FOOTNOTE IN TECHNICAL SPECIFICATION 3.1.3.7, CONTROL ROD DRIVE MECHANISMS (TAC NO. MF1172) Dear Mr. Heacock: The Commission has issued the enclosed Amendment No. 317 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated March 21, 2013. The amendment revises Technical Specification (TS) 3.1.3.7-Control Rod Drive Mechanisms to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal for MPS2. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Docket No. 50-336 Enclosures: 1. Amendment No. 317 to DPR-65 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: PUBLIC RidsNrrDorllpl1-1 Resource RidsNrrDoriDpr Resource RidsAcrsAcnw_MaiiCTR Resource A ccess1on N ML14015A453 o.: OFFICE LPL 1-1/PM LPL 1-1/LA NAME JKim KGoldstein DATE 1/16/14 1/16/14 Sincerely, Ira/ James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation LPLI-1 R/F RidsNrrLAKGoldstein RidsNrrDssStsb Resource R. McKinley, Rl RidsRgn1 MaiiCenter RidsNrrPMMillstone RidsNrrDssSrxbResource *S d t ee memo a ed Januarv 9, 2014 SRXB/BC STSB/BC OGC/NLO LPL 1-1/BC LPL 1-1/PM w/comments CJackson* REIIiott JWachutka BBeasley JKim 1/9/14 1/23/14 1/27/14 1/29/14 1/;jU/14 OFFfaiAL RECORD COPY
| | |
| }} | | The Commission has issued the enclosed Amendment No. 317 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated March 21, 2013. |
| | The amendment revises Technical Specification (TS) 3.1.3.7- Control Rod Drive Mechanisms to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal for MPS2. |
| | A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. |
| | Sincerely, James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336 |
| | |
| | ==Enclosures:== |
| | : 1. Amendment No. 317 to DPR-65 |
| | : 2. Safety Evaluation cc w/encls: Distribution via Listserv |
| | |
| | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT, INC. |
| | DOCKET NO. 50-336 MILLSTONE POWER STATION, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 317 Renewed License No. DPR-65 |
| | : 1. The Nuclear Regulatory Commission (the Commission) has found that: |
| | A. The application for amendment by the applicant dated March 21, 2013, 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-65 is hereby amended to read as follows: |
| | (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 317, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications. |
| | : 3. This license amendment is effective as of the date of issuance, and shall be implemented within 60 days of issuance. |
| | FOR THE NUCLEAR REGULA TORY COMMISSION Benjamin G. Beasley, 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: January 30, 2014 |
| | |
| | ATTACHMENT TO LICENSE AMENDMENT NO. 317 RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 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 Insert 3 3 Replace the following page of the Appendix A Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change. |
| | Remove Insert 3/4 1-31 3/4 1-31 |
| | |
| | Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (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 1: 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; and is subject to all applicable provisions of the Act and 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 The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal. |
| | (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 317, are hereby incorporated in the renewed license. |
| | The licensee shall operate the facility in accordance with the Technical Specifications. |
| | Renewed License No. DPR-65 Amendment No. 317 |
| | |
| | REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE MECHANISMS LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod drive mechanisms shall be de-energized. |
| | APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentration is less than refueling concentration of Specification 3. 9.1 . |
| | ACTION: |
| | With any of the control rod drive mechanisms energized, restore the mechanisms to their de-energized state within 2 hours or immediately open the reactor trip circuit breakers. |
| | SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod drive mechanisms shall be verified to be de-energized at least once per 24 hours. |
| | * The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATING, the reactor coolant system temperature is greater than 500° F, the pressurizer pressure is greater than 2000 psi a and the requirements of Limiting Condition for Operation for Specification 3.3 .1.1, "Reactor Protective Instrumentation," are met. |
| | MILLSTONE - UNIT 2 3/41-31 Amendment No. 116,291 317 |
| | |
| | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 317 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC. |
| | MILLSTONE POWER STATION, UNIT NO.2 DOCKET NO. 50-336 |
| | |
| | ==1.0 INTRODUCTION== |
| | |
| | By letter dated March 21, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13085A081), Dominion Nuclear Connecticut, Inc., (the licensee) requested changes to the Technical Specifications (TSs) for Millstone Power Station Unit 2 (MPS2). |
| | The proposed changes would revise TS 3.1.3.7- Control Rod Drive Mechanisms (CORMs) to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal. |
| | |
| | ==2.0 REGULATORY EVALUATION== |
| | |
| | In Title 10 of the Code of Federal Regulations (1 0 CFR) Section 50.36, the Commission 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 related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) |
| | Limiting Conditions for Operation (LCOs); (3) Surveillance Requirements (SRs); (4) design features; and (5) administrative controls. |
| | 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. |
| | 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 NRC 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 (STSs). For this review, the staff used NUREG-1432, Revision 4, "Standard Technical Specifications, Combustion Engineering Plants." The STSs reflect the general guidance and LCO seeping criteria provided by the Commission's Final Policy Statement. |
| | 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 staff's interpretation of 10 CFR 50.36, including judgments 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. Licensees may revise the remaining TSs to adopt current improved STS 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 licensee's current requirement; or (3) the change is less restrictive than the licensee's current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards. |
| | |
| | ==3.0 TECHNICAL EVALUATION== |
| | |
| | The Control Rod Drive Mechanism (CRDM) requirement of TS 3.1.3.7, Control Rod Drive Mechanisms, is provided to assure that the consequences of an uncontrolled control element assembly (CEA) withdrawal from a subcritical transient will stay within acceptable levels. This TS assures that reactor coolant system (RCS) conditions exist which are consistent with the plant safety analysis prior to energizing the CRDMs. The accident is precluded when CRDMs are de-energized and cannot withdraw a CEA. The CRDMs may be energized with the boron concentration greater than or equal to the refueling concentration since, under these conditions, adequate shutdown margin is maintained even if the CEAs are fully withdrawn from the core. |
| | The TS 3.1.3.7, Control Rod Drive Mechanisms, requires that CRDMs are de-energized in Modes 3*, 4, 5 and 6, whenever the RCS boron concentration is less than the refueling |
| | |
| | concentration specified in TS 3.9.1, Boron Concentrations. As described in the footnote (denoted by an asterisk in Mode 3) in TS 3.1.3.7, CRDMs may be energized in Mode 3 as long as the following conditions exist: (1) four reactor coolant pumps operating, (2) RCS temperature |
| | >500 F, (3) pressurizer pressure> 2000 psia, and (4) high power trip is operable. |
| | The TS Table 3.3-1, Reactor Protective Instrumentation, identifies, as a minimum, the reactor protective instrumentation channels and bypasses that must be operable under specific modes and conditions. For the condition where the protective system trip breakers are in the closed position and the CEA drive system is capable of CEA withdrawal (i.e., CRDMs energized), TS Table 3.3-1 requires the high power trip (Functional Unit 2), Reactor Protection System (RPS) logic matrices (Functional Unit 13), RPS logic matrix relays (Functional Unit 14) and reactor trip breakers (Functional Unit 15) to be operable. |
| | Currently, the footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, only requires the High Power Level Trip to be operable when CRDMs are energized and capable of CEA withdrawal. It does not include the additional operability requirements (i.e., RPS logic matrices, RPS logic matrix relays, and reactor trip breakers) that were added toTS Table 3.3-1, Reactor Protective Instrumentation under License Amendment 282 on September 25, 2003 (ADAMS Accession No. ML032270057). The proposed change would eliminate the inconsistency between TS 3.1.3.7 and TS Table 3.3-1 by revising the footnote in TS 3.1.3.7. Specifically, the footnote would be revised to delete "the high power trip is OPERABLE" and replace it with "the requirements of Limiting Condition for Operation for Specification 3.3.1.1, Reactor Protective Instrumentation, are met." |
| | The staff has determined that the current MPS2 TSs do not provide operability requirements for RPS logic matrices, RPS logic matrix relays, and reactor trip breakers in TS 3.1.3.7. The proposed change would provide these operability requirements. The staff finds that the proposed change is more restrictive than the current TS requirements and that the additional restrictions on plant operation would enhance safety. Therefore, the proposed change is acceptable. |
| | |
| | ==4.0 STATE CONSULTATION== |
| | |
| | In accordance with the Commission's regulations, the Connecticut 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 the 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 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 as published in the Federal Register on June 11, 2013(78 FR 35061 ). 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) there is reasonable assurance that 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. |
| | Principal Contributor: A. Guzzetta Date: January 30, 2014 |
| | |
| | January 30, 2014 Mr. David A Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 |
| | |
| | ==SUBJECT:== |
| | MILLSTONE POWER STATION, UNIT NO.2- ISSUANCE OF AMENDMENT RE: REVISE FOOTNOTE IN TECHNICAL SPECIFICATION 3.1.3.7, CONTROL ROD DRIVE MECHANISMS (TAC NO. MF1172) |
| | |
| | ==Dear Mr. Heacock:== |
| | |
| | The Commission has issued the enclosed Amendment No. 317 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated March 21, 2013. |
| | The amendment revises Technical Specification (TS) 3.1.3.7- Control Rod Drive Mechanisms to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal for MPS2. |
| | A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. |
| | Sincerely, Ira/ |
| | James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336 |
| | |
| | ==Enclosures:== |
| | : 1. Amendment No. 317 to DPR-65 |
| | : 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: |
| | PUBLIC LPLI-1 R/F RidsRgn1 MaiiCenter RidsNrrDorllpl1-1 Resource RidsNrrLAKGoldstein RidsNrrPMMillstone RidsNrrDoriDpr Resource RidsNrrDssStsb Resource RidsNrrDssSrxbResource RidsAcrsAcnw_MaiiCTR Resource R. McKinley, Rl A ccess1on No.: ML14015A453 *S ee memo dated Januarv 9, 2014 LPL 1-1/LA SRXB/BC STSB/BC OGC/NLO LPL 1-1/BC LPL 1-1/PM OFFICE LPL 1-1/PM w/comments NAME JKim KGoldstein CJackson* REIIiott JWachutka BBeasley JKim DATE 1/16/14 1/16/14 1/9/14 1/23/14 1/27/14 1/29/14 1/;jU/14 OFFfaiAL RECORD COPY}} |
Letter Sequence Approval |
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MONTHYEARML13085A0812013-03-21021 March 2013 License Amendment Request to Revise Footnote in Technical Specification 3.1.3.7-Control Rod Drive Mechanisms Project stage: Request ML13123A1812013-05-0202 May 2013 Acceptance Review Results Regarding Millstone 2 - Control Rod Drive Mechanisms License Amendment Request (MF1172) Project stage: Acceptance Review ML14015A4532014-01-30030 January 2014 Issuance of Amendment Revise Footnote in Technical Specification 3.1.3.7, Control Rod Drive Mechanism Project stage: Approval 2013-05-02
[Table View] |
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Category:Letter
MONTHYEARML24240A1692024-09-18018 September 2024 Cy 2023 Summary of Decommissioning Trust Fund Status ML24260A2192024-09-16016 September 2024 Decommissioning Trust Fund Disbursement - Revision to Previous Thirty-Day Written Notification ML24260A1952024-09-16016 September 2024 Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel ML24248A2272024-09-0404 September 2024 Operator Licensing Examination Approval ML24240A1532024-09-0303 September 2024 Summary of Regulatory Audit Supporting the Review of License Amendment Request for Implementation of Framatome Gaia Fuel IR 05000336/20240052024-08-29029 August 2024 Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Reports 05000336/2024005 and 05000423/2024005 IR 05000336/20240022024-08-13013 August 2024 Integrated Inspection Report 05000336/2024002 and 05000423/2024002 ML24221A2872024-08-0808 August 2024 Independent Spent Fuel Storage Installation (ISFSI) - Submittal of Cask Registration for Spent Fuel Storage IR 05000336/20244412024-08-0606 August 2024 Supplemental Inspection Report 05000336/2024441 and 05000423/2024441 and Follow-Up Assessment Letter (Cover Letter Only) ML24212A0742024-08-0505 August 2024 Request for Withholding Information from Public Disclosure - Millstone Power Station, Unit No. 3, Proposed Alternative Request IR-4-13 to Support Steam Generator Channel Head Drain Modification ML24211A1712024-07-25025 July 2024 Associated Independent Spent Fuels Storage Installation, Revision to Emergency Plan - Report of Change IR 05000336/20244032024-07-22022 July 2024 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000336/2024403 and 05000423/2024403 IR 05000336/20245012024-07-0101 July 2024 Emergency Preparedness Biennial Exercise Inspection Report 05000336/2024501 and 05000423/2024501 ML24180A0932024-06-28028 June 2024 Readiness for Additional Inspection: EA-23-144 IR 05000336/20240102024-06-26026 June 2024 Biennial Problem Identification and Resolution Inspection Report 05000336/2024010 and 05000423/2024010 ML24178A2422024-06-25025 June 2024 2023 Annual Report of Emergency Core Cooling System (ECCS) Model, Changes Pursuant to the Requirements of 10 CFR 50.46 IR 05000336/20244402024-06-24024 June 2024 Final Significance Determination for Security-Related Greater than Green Finding(S) with Assessment Follow-up; IR 05000336/2024440 and 05000423/2024440 and Notice of Violation(S), NRC Investigation Rpt 1-2024-001 (Cvr Ltr Only) ML24177A2792024-06-20020 June 2024 Preparation and Scheduling of Operator Licensing Examinations ML24170B0532024-06-10010 June 2024 DOM-NAF-2-P/NP-A, Revision 0.5, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML24162A0882024-06-10010 June 2024 Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications ML24165A1292024-06-0505 June 2024 ISFSI, 10 CFR 50.59 Annual Change Report for 2023 Annual Regulatory Commitment Change Report for 2023 ML24128A2772024-06-0404 June 2024 Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary) ML24151A6482024-06-0303 June 2024 Changes in Reactor Decommissioning Branch Project Management Assignments for Some Decommissioning Facilities ML24110A0562024-05-21021 May 2024 Exemption from the Requirements of 10 CFR Part 50, Section 50.46, and Appendix K Regarding Use of M5 Cladding Material (EPID L-2023-LLE-0013) (Letter) ML24109A0032024-05-21021 May 2024 Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary) ML24141A2432024-05-20020 May 2024 Response to Request for Additional Information Regarding Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment No. 15 ML24141A1502024-05-20020 May 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24142A0952024-05-20020 May 2024 End of Cycle 22 Steam Generator Tube Inspection Report IR 05000336/20240012024-05-14014 May 2024 Integrated Inspection Report 05000336/2024001 and 05000423/2024001 and Apparent Violation ML24123A2272024-05-0202 May 2024 Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24123A2042024-05-0202 May 2024 Pre-Decisional Replay to EA-23-144 IR 05000336/20244012024-04-30030 April 2024 Security Baseline Inspection Report 05000336/2024401 and 05000423/2024401 (Cover Letter Only) ML24123A1222024-04-30030 April 2024 Inservice Inspection Program - Owners Activity Report, Refueling Outage 22 ML24116A0452024-04-25025 April 2024 Special Inspection Follow-Up Report 05000336/2024440 and 05000423/2024440 and Preliminary Finding(S) of Greater than Very Low Significance and NRC Investigation Report No. 1-2024-001 (Cover Letter Only) ML24116A1742024-04-24024 April 2024 Annual Radiological Environmental Operating Report ML24114A2662024-04-24024 April 2024 Submittal of 2023 Annual Radioactive Effluent Release Report ML24103A0202024-04-22022 April 2024 Summary of Regulatory Audit in Support of License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits ML24106A2032024-04-15015 April 2024 2023 Annual Environmental Operating Report ML24088A3302024-04-0404 April 2024 Regulatory Audit Plan in Support of License Amendment Request to Implement Framatome Gaia Fuel ML24093A1022024-04-0101 April 2024 Alternative Request IR-4-13, Proposed Alternative Request to Support Steam Genera Tor Channel Head Drain Modification ML24093A2162024-04-0101 April 2024 Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits IR 05000336/20240112024-04-0101 April 2024 Comprehensive Engineering Team Inspection - Inspection Report 05000336/2024011 and 05000423/2024011 ML24092A0752024-03-28028 March 2024 3R22 Refueling Outage Inservice Inspection (ISI) Owners Activity Report Extension ML24088A2352024-03-26026 March 2024 Decommissioning Funding Status Report ML24086A4762024-03-22022 March 2024 Application for Technical Specification Change to Extend the Inspection Interval for Reactor Coolant Pump Flywheels Using the Consolidated Line-Item Improvement Process ML24086A4802024-03-22022 March 2024 Alternative Request IR-4-14, Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examination for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles ML24051A1922024-03-0808 March 2024 – Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) IR 05000336/20230062024-02-28028 February 2024 Annual Assessment Letter for Millstone Power Station, Units 2 and 3, (Reports 05000336/2023006 and 05000423/2023006) ML24053A2632024-02-21021 February 2024 Unit 3, and Independent Spent Fuel Storage Installation, Notification Pursuant to 10 CFR 72.212(b)(1) Prior to First Storage of Spent Fuel Under a General License ML24057A0612024-02-19019 February 2024 and Virgil C. Summer Power Nuclear Stations - Nuclear Property Insurance Coverage 2024-09-04
[Table view] Category:License-Operating (New/Renewal/Amendments) DKT 50
MONTHYEARML24128A2772024-06-0404 June 2024 Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary) ML24109A0032024-05-21021 May 2024 Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary) ML23341A0172024-01-12012 January 2024 Issuance of Amendment No. 288 Revision to Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures ML23072A0892023-05-0101 May 2023 (Amendments 346 & 286), North Anna 1 & 2 (Amnds 294 & 277), Surry 1 & 2 (Amnds 311 & 311), and Summer 1 (Amd 225) - Issuance of Amendments to Revise TSs to Adopt TSTF-554 Revise Reactor Coolant Leakage Requirements ML23058A4542023-03-16016 March 2023 Issuance of Amendment Nos. 345 and 285 Regarding Adoption of Technical Specification Task Force-359, Increase Flexibility in Mode Restraints ML23005A1702023-01-17017 January 2023 Correction to Issuance of Amendment No. 283 to Revise Technical Specifications to Adopt TSTF-569, Revision of Response Time Testing Definition ML22095A1072022-07-11011 July 2022 Issuance of Amendment Nos. 120, 344, & 284, 293 & 276, & 307 & 307 to Relocate Requirements to the QAPD ML22041A0102022-03-0101 March 2022 V.C. Summer 1, Issuance of Amendment Nos. 283 (Millstone), 291 and 274 (North Anna), and 221 (Summer) to Revise TSs to Adopt TSTF-569 Revision of Response Time Testing Definition ML22007A1512022-02-16016 February 2022 Issuance of Amendment No. 282 Regarding Shutdown Bank Technical Specification Requirements and Alternate Control Rod Position Monitoring Requirements ML21326A0992022-01-0707 January 2022 Issuance of Amendment No. 281 Regarding Revised Reactor Core Safety Limit to Reflect Topical Report WCAP-177642-P-A, Revision 1 ML21262A0012021-11-0909 November 2021 Issuance of Amendment No. 280 Regarding Measurement Uncertainty Recapture Power Uprate ML21227A0002021-10-0505 October 2021 Issuance of Amendment No. 279 Regarding Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss-of-Coolant Accident ML21222A2302021-09-0909 September 2021 Issuance of Amendment No. 343 Revision to Technical Specifications for Steam Generator Inspection Frequency (L-2020-LLA-0227) ML21043A1622021-03-25025 March 2021 Issuance of Amendment No. 278 Regarding Revision to Battery Surveillance Requirements ML21026A1422021-02-23023 February 2021 Issuance of Amendment No. 342 Revision to Technical Specification Table 3.3-11, Accident Monitoring Instrumentation ML20275A0002020-10-14014 October 2020 Issuance of Amendment No. 277 to Revise Technical Specification 6.8.4.g to Allow a One-Time Deferral of the Steam Generator Inspections ML20237H9952020-09-29029 September 2020 Issuance of Amendment No. 341 Revision to Technical Specification 6.25, Pre-Stressed Concrete Containment Tendon Surveillance Program ML20191A0042020-08-0707 August 2020 Issuance of Amendment No. 340 Revised Technical Specification Limits for Primary and Secondary Coolant Activity ML20161A0002020-07-15015 July 2020 Issuance of Amendment No. 276 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals ML20140A3692020-06-24024 June 2020 Issuance of Amendment No. 339 Extension of Technical Specification 3.8.1.1, A.C. Sources - Operating, Allowed Outage Time ML19340A0252020-01-30030 January 2020 Issuance of Amendment No. 337 Regarding Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structure, Systems, and Components of Nuclear Power Reactors ML19305D2482019-12-31031 December 2019 Issuance of Amendments Adoption of Emergency Action Level Schemes Per NEI 99-01, Rev. 6 ML19150A3672019-05-30030 May 2019 Current Facility Operating License No. DPR-21, Tech Specs, Revised 5/30/2019 ML19126A0002019-05-28028 May 2019 Issuance of Amendment No. 273 Regarding Technical Specification Changes for Spent Fuel Storage and New Fuel Storage ML19042A2772019-03-21021 March 2019 Issuance of Amendment No. 272 Regarding Revision to Technical Specification Action Statement for Loss of Control Building Inlet Ventilation Radiation Monitor Instrumentation Channels ML18246A0072018-09-25025 September 2018 Issuance of Amendment No. 335 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals ML18038B2002018-02-26026 February 2018 Issuance of Amendment Nos. 118, 334, and 271 to Revise Licensee'S Name (CAC Nos. MF9844, MF9845, and MF9848; EPID L-2017-LLA-0245 and EPID L-2017-LLA-0346) ML17025A2182017-01-24024 January 2017 Issuance of Amendment Realistic Large Break Loss-of-Coolant Accident Analysis ML16308A4852016-12-22022 December 2016 Issuance of Amendment No. 331 Revision to Emergency Core Cooling System Technical Specification and Final Safety Analysis Report Ch. 14 to Remove Charging ML16249A0012016-09-30030 September 2016 Issuance of Amendments Small Break Loss of Coolant Accident Reanalysis ML16206A0012016-08-0404 August 2016 Issuance of Amendment No. 270 to Revise Technical Specification 5.6.3, Fuel Storage Capacity ML16193A0012016-07-28028 July 2016 Issuance of Amendments Removal of Severe Line Outage Detection from the Offsite Power System ML16131A7282016-07-28028 July 2016 Issuance of Amendment 268 Adopting Dominion Core Design and Safety Analysis Methods and Addressing the Issues Identified in Three Westinghouse Communication Documents ML16189A0762016-06-29029 June 2016 Supplement to Information Regarding License Amendment Request for Removal of Severe Line Outage Detection from the Offsite Power System ML16003A0082016-06-23023 June 2016 Issuance of Amendment No. 327 Proposed Technical Specification Changes for Spent Fuel Storage ML16068A3122016-03-31031 March 2016 Issuance of Amendment No. 326 to Revise Technical Specifications for Containment Leak Rate Testing ML16011A4002016-01-29029 January 2016 Issuance of Amendments Nos. 325 and 267 to Adopt TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation ML15288A0042015-11-30030 November 2015 Issuance of Amendment No. 266: Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System and Engineered Safety Feature Actuation System Instrumentation Test Times and Completion Times ML15280A2422015-10-29029 October 2015 Issuance of Amendment No. 324 Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, Adoption of TSTF-425,Rev 3 ML15245A4822015-10-0707 October 2015 Millstone Power Station, Surry Power Station, and North Anna Power Station - Issuance of Amendments to Revise the Cyber Security Milestone 8 Completion Date in the Renewed Facility Operating Licenses ML15225A0102015-08-28028 August 2015 Issuance of Amendment No. 264 Surveillance Requirement 4.4.4.2, Reactor Coolant System Relief Valves ML15246A1422015-08-27027 August 2015 Connecticut, Inc. Millstone Power Station Unit 3 Supplement to License Amendment Request to Revise Surveillance Requirement 4.4.4.2, Reactor Coolant System Relief Valves ML15225A0082015-08-26026 August 2015 Issuance of Amendment No. 322 Revision to the Final Safety Analysis Report - Examination Requirements for ANSI B31.1.0 Piping Welds ML15187A3262015-07-29029 July 2015 Issuance of Amendment No. 321, Adoption of TSTF-426, Revision 5, Revise or Add Actions to Preclude Entry Into LCO 3.0.3 - RITSTF Initiatives 6B & 6C ML15187A1862015-07-28028 July 2015 Issuance of Amendment No. 263, Regarding Technical Specification Changes to Auxiliary Feedwater System (MF4692) ML15187A0112015-07-27027 July 2015 Issuance of Amendment No. 262, Refueling Water Storage Tank Allowable Temperatures ML15093A0022015-05-20020 May 2015 Issuance of Amendment No. 320, Delete Technical Specification Index and Make Other Editorial and Administrative Changes ML15098A0342015-05-20020 May 2015 Issuance of Amendment No. 261, Delete Technical Specification Index and Make Other Editorial and Administrative Changes ML15093A4412015-05-18018 May 2015 Issuance of Amendment No. 319, Use of Areva M5 Alloy Clad Fuel Assemblies ML14178A5992014-07-11011 July 2014 Issuance of Amendment Proposed Changes to Technical Specification 3/4 7.5, Ultimate Heat Sink (Tac MF1780) 2024-06-04
[Table view] Category:Safety Evaluation
MONTHYEARML24128A2772024-06-0404 June 2024 Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary) ML24109A0032024-05-21021 May 2024 Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary) ML23341A0172024-01-12012 January 2024 Issuance of Amendment No. 288 Revision to Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures ML23283A3052023-12-20020 December 2023 Review of Appendix F to DOM-NAF2, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code (EPID L-2022-LLT-0003) (Nonproprietary) ML23230A0502023-10-0202 October 2023 5 of the Quality Assurance Topical Report - Review of Program Changes ML23226A0052023-09-26026 September 2023 Issuance of Amendment No. 287 Supplement to Spent Fuel Pool Criticality Safety Analysis ML23188A0432023-07-31031 July 2023 Authorization and Safety Evaluation for Alternative Request No. IR-04-11 ML23175A0052023-07-12012 July 2023 Alternative Request P-07 for Pump Periodic Verification Testing Program for Containment Recirculation Spray System Pumps ML23072A0892023-05-0101 May 2023 (Amendments 346 & 286), North Anna 1 & 2 (Amnds 294 & 277), Surry 1 & 2 (Amnds 311 & 311), and Summer 1 (Amd 225) - Issuance of Amendments to Revise TSs to Adopt TSTF-554 Revise Reactor Coolant Leakage Requirements ML23058A4542023-03-16016 March 2023 Issuance of Amendment Nos. 345 and 285 Regarding Adoption of Technical Specification Task Force-359, Increase Flexibility in Mode Restraints ML21320A0072022-09-0707 September 2022 Review of Appendix E to DOM-NAF-2, Qualification of the Framatome BWU-I CHF Correlation in the Dominion Energy VIPRE-D Computer Code (EPID L-2021-LLT-0000) (Non-Proprietary) ML22201A5082022-07-28028 July 2022 Authorization and Safety Evaluation for Alternative Request No. IR-04-09 ML22095A1072022-07-11011 July 2022 Issuance of Amendment Nos. 120, 344, & 284, 293 & 276, & 307 & 307 to Relocate Requirements to the QAPD ML22039A3392022-03-0303 March 2022 Request for Alternative Frequency to Supplemental Valve Position Verification Testing Requirements in the Fourth 10-year Valve Inservice Testing Program ML22041A0102022-03-0101 March 2022 V.C. Summer 1, Issuance of Amendment Nos. 283 (Millstone), 291 and 274 (North Anna), and 221 (Summer) to Revise TSs to Adopt TSTF-569 Revision of Response Time Testing Definition ML22007A1512022-02-16016 February 2022 Issuance of Amendment No. 282 Regarding Shutdown Bank Technical Specification Requirements and Alternate Control Rod Position Monitoring Requirements ML21326A0992022-01-0707 January 2022 Issuance of Amendment No. 281 Regarding Revised Reactor Core Safety Limit to Reflect Topical Report WCAP-177642-P-A, Revision 1 ML21262A0012021-11-0909 November 2021 Issuance of Amendment No. 280 Regarding Measurement Uncertainty Recapture Power Uprate ML21284A0062021-10-29029 October 2021 Authorization and Safety Evaluation for Alternative Request No. RR-05-04 and IR-4-02 ML21227A0002021-10-0505 October 2021 Issuance of Amendment No. 279 Regarding Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss-of-Coolant Accident ML21222A2302021-09-0909 September 2021 Issuance of Amendment No. 343 Revision to Technical Specifications for Steam Generator Inspection Frequency (L-2020-LLA-0227) ML21174A0202021-08-0202 August 2021 Relief Request for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval ML21167A3552021-07-16016 July 2021 Authorization and Safety Evaluation for Alternative Request No. RR-05-06 ML21167A2112021-06-30030 June 2021 Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0081 Through L-2020-LLR-0088) ML21075A0452021-03-26026 March 2021 Request to Utilize Code Case N-885 ML21043A1622021-03-25025 March 2021 Issuance of Amendment No. 278 Regarding Revision to Battery Surveillance Requirements ML21026A1422021-02-23023 February 2021 Issuance of Amendment No. 342 Revision to Technical Specification Table 3.3-11, Accident Monitoring Instrumentation ML20312A0022020-12-10010 December 2020 Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0027 Through L-2020-LLR-0032) ML20312A0012020-12-10010 December 2020 Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval ML20287A4712020-10-20020 October 2020 Proposed Alternative RR-05-05 to the Requirements of the ASME Code Containment Unbonded Post-Tensioning System Inservice Inspection Requirements ML20275A0002020-10-14014 October 2020 Issuance of Amendment No. 277 to Revise Technical Specification 6.8.4.g to Allow a One-Time Deferral of the Steam Generator Inspections ML20237H9952020-09-29029 September 2020 Issuance of Amendment No. 341 Revision to Technical Specification 6.25, Pre-Stressed Concrete Containment Tendon Surveillance Program ML20252A0072020-09-15015 September 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI ML20191A0042020-08-0707 August 2020 Issuance of Amendment No. 340 Revised Technical Specification Limits for Primary and Secondary Coolant Activity ML20189A2062020-07-16016 July 2020 Relief Request IR-4-03 Concerning Non-Code Methodology to Demonstrate Structural Integrity of Class 3 Moderate-Energy Piping ML20161A0002020-07-15015 July 2020 Issuance of Amendment No. 276 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals ML20140A3692020-06-24024 June 2020 Issuance of Amendment No. 339 Extension of Technical Specification 3.8.1.1, A.C. Sources - Operating, Allowed Outage Time ML20080K5082020-03-24024 March 2020 Alternative Request RR-05-03 for the Fifth 10-Year Inservice Inspection Interval ML19340A0252020-01-30030 January 2020 Issuance of Amendment No. 337 Regarding Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structure, Systems, and Components of Nuclear Power Reactors ML19305D2482019-12-31031 December 2019 Issuance of Amendments Adoption of Emergency Action Level Schemes Per NEI 99-01, Rev. 6 ML19340A0012019-12-18018 December 2019 Proposed Alternative Request IR-4-01 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography ML19338G3722019-12-18018 December 2019 Alternative Requests RR-05-01 and RR-05-02 for the Fifth 10-Year Inservice Inspection Interval ML19340A0002019-12-13013 December 2019 Relief Request IR-3-39, Proposed Alternative to ASME Code Weld Preheat Requirements ML19126A0002019-05-28028 May 2019 Issuance of Amendment No. 273 Regarding Technical Specification Changes for Spent Fuel Storage and New Fuel Storage ML19098A0342019-04-30030 April 2019 Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3, and Quad Cities, Units 1 and 2 - Revision to Approved Alternative to Use BWR Vessel and Internal Proj Guidelins ML19042A2772019-03-21021 March 2019 Issuance of Amendment No. 272 Regarding Revision to Technical Specification Action Statement for Loss of Control Building Inlet Ventilation Radiation Monitor Instrumentation Channels ML18290A6022018-11-13013 November 2018 Alternative Requests Related to the Fifth and Fourth 10-Year Interval Pump, Valve, and Snubber Inservice Testing Programs, Respectively (EPID L-2018-LLR- 0012 Through EPID L-2018-LLR-0022) ML18275A0122018-10-0404 October 2018 Alternative Request P-06 for the 'C' Charging Pump Test Frequency ML18246A0072018-09-25025 September 2018 Issuance of Amendment No. 335 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals ML18252A0032018-09-17017 September 2018 Alternative Requests RR-04-27 and IR-3-38 for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography 2024-06-04
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 30, 2014 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO.2 -ISSUANCE OF AMENDMENT RE: REVISE FOOTNOTE IN TECHNICAL SPECIFICATION 3.1.3.7, CONTROL ROD DRIVE MECHANISMS (TAC NO. MF1172)
Dear Mr. Heacock:
The Commission has issued the enclosed Amendment No. 317 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated March 21, 2013.
The amendment revises Technical Specification (TS) 3.1.3.7- Control Rod Drive Mechanisms to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal for MPS2.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosures:
- 1. Amendment No. 317 to DPR-65
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT, INC.
DOCKET NO. 50-336 MILLSTONE POWER STATION, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 317 Renewed License No. DPR-65
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by the applicant dated March 21, 2013, 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-65 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 317, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of issuance, and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULA TORY COMMISSION Benjamin G. Beasley, 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: January 30, 2014
ATTACHMENT TO LICENSE AMENDMENT NO. 317 RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 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 Insert 3 3 Replace the following page of the Appendix A Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3/4 1-31 3/4 1-31
Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (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 1: 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; and is subject to all applicable provisions of the Act and 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 The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 317, are hereby incorporated in the renewed license.
The licensee shall operate the facility in accordance with the Technical Specifications.
Renewed License No. DPR-65 Amendment No. 317
REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE MECHANISMS LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod drive mechanisms shall be de-energized.
APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentration is less than refueling concentration of Specification 3. 9.1 .
ACTION:
With any of the control rod drive mechanisms energized, restore the mechanisms to their de-energized state within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or immediately open the reactor trip circuit breakers.
SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod drive mechanisms shall be verified to be de-energized at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATING, the reactor coolant system temperature is greater than 500° F, the pressurizer pressure is greater than 2000 psi a and the requirements of Limiting Condition for Operation for Specification 3.3 .1.1, "Reactor Protective Instrumentation," are met.
MILLSTONE - UNIT 2 3/41-31 Amendment No. 116,291 317
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 317 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION, UNIT NO.2 DOCKET NO. 50-336
1.0 INTRODUCTION
By letter dated March 21, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13085A081), Dominion Nuclear Connecticut, Inc., (the licensee) requested changes to the Technical Specifications (TSs) for Millstone Power Station Unit 2 (MPS2).
The proposed changes would revise TS 3.1.3.7- Control Rod Drive Mechanisms (CORMs) to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal.
2.0 REGULATORY EVALUATION
In Title 10 of the Code of Federal Regulations (1 0 CFR) Section 50.36, the Commission 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 related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2)
Limiting Conditions for Operation (LCOs); (3) Surveillance Requirements (SRs); (4) design features; and (5) administrative controls.
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.
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 NRC 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 (STSs). For this review, the staff used NUREG-1432, Revision 4, "Standard Technical Specifications, Combustion Engineering Plants." The STSs reflect the general guidance and LCO seeping criteria provided by the Commission's Final Policy Statement.
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 staff's interpretation of 10 CFR 50.36, including judgments 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. Licensees may revise the remaining TSs to adopt current improved STS 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 licensee's current requirement; or (3) the change is less restrictive than the licensee's current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards.
3.0 TECHNICAL EVALUATION
The Control Rod Drive Mechanism (CRDM) requirement of TS 3.1.3.7, Control Rod Drive Mechanisms, is provided to assure that the consequences of an uncontrolled control element assembly (CEA) withdrawal from a subcritical transient will stay within acceptable levels. This TS assures that reactor coolant system (RCS) conditions exist which are consistent with the plant safety analysis prior to energizing the CRDMs. The accident is precluded when CRDMs are de-energized and cannot withdraw a CEA. The CRDMs may be energized with the boron concentration greater than or equal to the refueling concentration since, under these conditions, adequate shutdown margin is maintained even if the CEAs are fully withdrawn from the core.
The TS 3.1.3.7, Control Rod Drive Mechanisms, requires that CRDMs are de-energized in Modes 3*, 4, 5 and 6, whenever the RCS boron concentration is less than the refueling
concentration specified in TS 3.9.1, Boron Concentrations. As described in the footnote (denoted by an asterisk in Mode 3) in TS 3.1.3.7, CRDMs may be energized in Mode 3 as long as the following conditions exist: (1) four reactor coolant pumps operating, (2) RCS temperature
>500 F, (3) pressurizer pressure> 2000 psia, and (4) high power trip is operable.
The TS Table 3.3-1, Reactor Protective Instrumentation, identifies, as a minimum, the reactor protective instrumentation channels and bypasses that must be operable under specific modes and conditions. For the condition where the protective system trip breakers are in the closed position and the CEA drive system is capable of CEA withdrawal (i.e., CRDMs energized), TS Table 3.3-1 requires the high power trip (Functional Unit 2), Reactor Protection System (RPS) logic matrices (Functional Unit 13), RPS logic matrix relays (Functional Unit 14) and reactor trip breakers (Functional Unit 15) to be operable.
Currently, the footnote in TS 3.1.3.7, Control Rod Drive Mechanisms, only requires the High Power Level Trip to be operable when CRDMs are energized and capable of CEA withdrawal. It does not include the additional operability requirements (i.e., RPS logic matrices, RPS logic matrix relays, and reactor trip breakers) that were added toTS Table 3.3-1, Reactor Protective Instrumentation under License Amendment 282 on September 25, 2003 (ADAMS Accession No. ML032270057). The proposed change would eliminate the inconsistency between TS 3.1.3.7 and TS Table 3.3-1 by revising the footnote in TS 3.1.3.7. Specifically, the footnote would be revised to delete "the high power trip is OPERABLE" and replace it with "the requirements of Limiting Condition for Operation for Specification 3.3.1.1, Reactor Protective Instrumentation, are met."
The staff has determined that the current MPS2 TSs do not provide operability requirements for RPS logic matrices, RPS logic matrix relays, and reactor trip breakers in TS 3.1.3.7. The proposed change would provide these operability requirements. The staff finds that the proposed change is more restrictive than the current TS requirements and that the additional restrictions on plant operation would enhance safety. Therefore, the proposed change is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Connecticut 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 the 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 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 as published in the Federal Register on June 11, 2013(78 FR 35061 ). 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) there is reasonable assurance that 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.
Principal Contributor: A. Guzzetta Date: January 30, 2014
January 30, 2014 Mr. David A Heacock President and Chief Nuclear Officer Dominion Nuclear lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO.2- ISSUANCE OF AMENDMENT RE: REVISE FOOTNOTE IN TECHNICAL SPECIFICATION 3.1.3.7, CONTROL ROD DRIVE MECHANISMS (TAC NO. MF1172)
Dear Mr. Heacock:
The Commission has issued the enclosed Amendment No. 317 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated March 21, 2013.
The amendment revises Technical Specification (TS) 3.1.3.7- Control Rod Drive Mechanisms to provide consistency with the operability requirements of TS Table 3.3-1, Reactor Protective Instrumentation, when control rod drive mechanisms are energized and capable of withdrawal for MPS2.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, Ira/
James Kim, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosures:
- 1. Amendment No. 317 to DPR-65
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC LPLI-1 R/F RidsRgn1 MaiiCenter RidsNrrDorllpl1-1 Resource RidsNrrLAKGoldstein RidsNrrPMMillstone RidsNrrDoriDpr Resource RidsNrrDssStsb Resource RidsNrrDssSrxbResource RidsAcrsAcnw_MaiiCTR Resource R. McKinley, Rl A ccess1on No.: ML14015A453 *S ee memo dated Januarv 9, 2014 LPL 1-1/LA SRXB/BC STSB/BC OGC/NLO LPL 1-1/BC LPL 1-1/PM OFFICE LPL 1-1/PM w/comments NAME JKim KGoldstein CJackson* REIIiott JWachutka BBeasley JKim DATE 1/16/14 1/16/14 1/9/14 1/23/14 1/27/14 1/29/14 1/;jU/14 OFFfaiAL RECORD COPY