ML103210032
| ML103210032 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/14/2010 |
| From: | Tam P Plant Licensing Branch III |
| To: | Weber L Indiana Michigan Power Co |
| Tam P | |
| References | |
| TAC ME4709, TAC ME4710 | |
| Download: ML103210032 (18) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 14, 2010 Mr. Lawrence J. Weber Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
SUBJECT:
DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: ELIMINATION OF REQUIREMENTS FOR HYDROGEN RECOMBINERS AND HYDROGEN MONITORS (TAC NO. ME4709 AND ME4710)
Dear Mr. Weber:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 313 to Renewed Facility Operating License No. DPR-58 and Amendment No. 296 to Renewed Facility Operating License No. DPR-74 for the Donald C. Cook Nuclear Plant, Units 1 and 2. The amendments consist of changes to the Technical Specifications in response to your application dated September 8, 2010.
The amendments delete the Technical Specification requirements related to the containment hydrogen recombiners and the hydrogen monitors, in accordance with Nuclear Energy Institute Technical Specification Task Force (TSTF) initiative designated as TSTF-447.
A copy of our related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
~~t:
Peter S. Tam, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosures:
- 1. Amendment No. 313 to DPR-58
- 2. Amendment No. 296 to DPR-74
- 3. Safety Evaluation cc w/encls: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT, UNIT 1 AMENDMENT RENEWED FACILITY OPERATING LICENSE Amendment No. 313 License No. DPR-58
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana Michigan Power Company (the licensee) dated September 8,2010, 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-58 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix S, as revised through Amendment No. 313, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 2
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 14, 2010
ATTACHMENT TO LICENSE AMENDMENT NO. 313 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 DOCKET NO. 50-315 Replace the following page of Renewed Facility Operating License No. DPR-58 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
REMOVE INSERT - 3 Replace the following pages of 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 Table of contents Page 3 of 5 Table of contents Page 3 of 5 3.3.3-2 3.3.3-2 3.3.3-3 3.3.3-3 3.3.3-4 3.3.3-4 3.3.3-5 3.3.3-5 3.6.8-1 3.6.8-1 3.6.8-2 5.6-4 5.6-4
- 3 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; and (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 I: 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 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 The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3304 megawatts thermal in accordance with the conditions specified therein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix S, as revised through Amendment No. 313, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Less Than Four Loop Operation The licensee shall not operate the reactor at power levels above P-7 (as defined in Table 3.3.1-1 of Specification 3.3.1 of Appendix A to this renewed operating license) with less than four reactor coolant loops in operation until (a) safety analyses for less than four loop operation have been submitted, and (b) approval for less than found loop operation at power levels above P-7 has been granted by the Commission by amendment of this license.
(4)
Indiana Michigan Power Company shall implement and maintain, in effect, all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility and as approved in the SERs dated December 12, 1977, July 31, 1979, January 10, 1981, February 7, 1983, November 22, 1983, December 23, 1983, March 16, 1984, August 27, 1985 Renewed License No. DPR-58 Amendment No.1 through 312,313
UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators 3.5.1-1 3.5.2 ECCS - Operating 3.5.2-1 3.5.3 ECCS - Shutdown 3.5.3-1 3.5.4 Refueling Water Storage Tank (RWST) 3.5.4-1 3.5.5 Seal Injection Flow 3.5.5-1 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment.
3.6.1-1 3.6.2 Containment Air Locks 3.6.2-1 3.6.3 Containment Isolation Valves 3.6.3-1 3.6.4 Containment Pressure 3.6.4-1 3.6.5 Containment Air Temperature 3.6.5-1 3.6.6 Containment Spray System 3.6.6-1 3.6.7 Spray Additive System 3.6.7-1 3.6.8 Deleted 3.6.8-1 3.6.9 Distributed Ignition System (DIS) 3.6.9-1 3.6.10 Containment Air Recirculation/Hydrogen Skimmer (CEQ) System 3.6.10-1 3.6.11 Ice Bed 3.6.11-1 3.6.12 Ice Condenser Doors 3.6.12-1 3.6.13 Divider Barrier Integrity 3.6.13-1 3.6.14 Containment Recirculation Drains 3.6.14-1 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs) 3.7.1-1 Table 3.7.1-1, OPERABLE Main Steam Safety Valves versus Maximum Allowable Power 3.7.1-3 Table 3.7.1-2, Main Steam Safety Valve Lift Settings 3.7.1-4 3.7.2 Steam Generator Stop Valves (SGSVs) 3.7.2-1 3.7.3 Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves (MFRVs) 3.7.3-1 3.7.4 Steam Generator (SG) Power Operated Relief Valves (PORVs) 3.7.4-1 3.7.5 Auxiliary Feedwater (AFW) System 3.7.5-1 3.7.6 Condensate Storage Tank (CST) 3.7.6-1 3.7.7 Component Cooling Water (CCW) System 3.7.7-1 3.7.8 Essential Service Water (ESW) System 3.7.8-1 3.7.9 Ultimate Heat Sink (UHS) 3.7.9-1 3.7.10 Control Room Emergency Ventilation (CREV) System 3.7.10-1 3.7.11 Control Room Air Conditioning (CRAC) System 3.7.11-1 3.7.12 Engineered Safety Features (ESF) Ventilation System 3.7.12-1 3.7.13 Fuel Handling Area Exhaust Ventilation (FHAEV) System 3.7.13-1 Cook Nuclear Plant Unit 1 Page 3 of 5 Amendment No. ~ 313
ACTIONS (continued)
CONDITION D. One or more Functions with two or more required channels inoperable.
E. Required Action and associated Completion Time of Condition C or D not met.
F. As required by Required Action E.1 and referenced in Table 3.3.3-1.
G. As required by Required Action E.1 and referenced in Table 3.3.3-1.
PAM Instrumentation 3.3.3 REQUIRED ACTION COMPLETION TIME D.1 Restore all but one channel to OPERABLE status.
7 days E.1 Enter the Condition referenced in Table 3.3.3-1 for the channel.
Immediately F.1 AND F.2 Be in MODE 3.
Be in MODE 4.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours G.1 Initiate action in accordance with Specification 5.6.6.
Immediately Cook Nuclear Plant Unit 1 3.3.3-2 Amendment No. ~ 313
PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS
N()TE----------------------------------------------------------
These SRs apply to each PAM instrumentation Function in Table 3.3.3-1, except where identified in the SR.
SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.
31 days SR 3.3.3.2 Deleted.
N()TE-----------------------------
Neutron detectors are excluded from CHANNEL CALIBRATI()N.
Perform CHANNEL CALI BRATI()N.
24 months Cook Nuclear Plant Unit 1 3.3.3-3 Amendment No.2&+- 313
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 2)
Post Accident Monitoring Instrumentation CONDITION REFERENCED FROM REQUIRED FUNCTION REQUIRED CHANNELS ACTION E.1
- 1.
Neutron Flux 2
F
- 2.
Steam Generator Pressure (per steam generator) 2 F
- 3.
Reactor Coolant System (RCS) Hot Leg 2
F Temperature (Wide Range)
- 4.
RCS Cold Leg Temperature (Wide Range) 2 F
- 5.
RCS Pressure (Wide Range) 2 F
- 6.
Reactor Coolant Inventory Tracking System 2
G (Reactor Vessel Level Indication)
- 7.
Containment Water Level 2
F
- 8.
Containment Pressure (Narrow Range) 2 F
- 9.
Penetration Flow Path Containment Isolation Valve 2 per penetration flow F
Position path(a)(b)
- 10.
Containment Area Radiation (High Range) 2 G
- 11.
Deleted
- 12.
Pressurizer Level 2
F
- 13.
Steam Generator Water Level (Wide Range) 4 F
- 14.
Condensate Storage Tank Level G
2(C)
- 15.
Core Exit Temperature - Quadrant 1 F
2(C)
- 16.
Core Exit Temperature - Quadrant 2 F
2(C)
- 17.
Core Exit Temperature - Quadrant 3 F
2(C)
- 18.
Core Exit Temperature - Quadrant 4 F
2(d)
- 19.
Secondary Heat Sink Indication F
(per steam generator)
(a) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
(b) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.
(c) A channel consists of one core exit thermocouple (CET).
(d) Any combination of two instruments per steam generator, including Steam Generator Water Level (Narrow Range) and Auxiliary Feedwater Flow, can be used to satisfy Function 19 OPERABILITY requirements.
Cook Nuclear Plant Unit 1 3.3.3-4 Amendment No. 2-&7 31 3
3.6.8 3.6 CONTAINMENT SYSTEMS 3.6.8 Deleted Cook Nuclear Plant Unit 1 3.6.8-1 Amendment No. 2&7 313
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Post Accident Monitoring Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:
- a.
The scope of inspections performed on each SG,
- b.
Active degradation mechanisms found,
- c.
Nondestructive examination techniques utilized for each degradation mechanism,
- d.
Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e.
Number of tubes plugged during the inspection outage for each active degradation mechanism,
- f.
Total number and percentage of tubes plugged to date, and
- g.
The results of condition monitoring, including the results of tube pulls and in situ testing.
Cook Nuclear Plant Unit 1 5:6-4 Amendment No. 2&7, ~,~ 313
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT 2 AMENDMENT RENEWED FACILITY OPERATING LICENSE Amendment No. 296 License No. DPR-74
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana Michigan Power Company (the licensee) dated September 8, 2010, 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-74 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 296, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 2
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 14, 2010
ATTACHMENT TO LICENSE AMENDMENT NO. 296 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Replace the following page of Renewed Facility Operating License No. DPR-74 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
REMOVE INSERT - 3 Replace the following pages of 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 Table of Contents Page 3 of 5 Table of Contents Page 3 of 5 3.3.3-2 3.3.3-2 3.3.3-3 3.3.3-3 3.3.3-4 3.3.3-4 3.3.3-5 3.3.3-5 3.6.8-1 3.6.8-1 3.6.8-2 5.6-4 5.6-4
- 3 radiation monitoring equipment calibration, and as fission detectors in amounts as required.
(4)
Pursuant to th~ 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; and (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 I: 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 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 The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3468 megawatts thermal in accordance with the conditions specified therein and in attachment 1 to the renewed operating license.
The preoperational tests, startup and other items identified in Attachment 1 to this renewed operating license shall be completed. Attachment 1 is an integral part of this renewed operating license.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix 8, as revised through Amendment No. 296, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Additional Conditions (a)
Deleted by Amendment No. 76 (b)
Deleted by Amendment NO.2 (c)
Leak Testing of Emergency Core cooling System Valves Indiana Michigan Power Company shall prior to completion of the first inservice testing interval test each of the two valves in series in the Renewed License No. DPR-74 Amendment No. 1 thrololgh 295,296
UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators 3.5.1-1 3.5.2 ECCS - Operating 3.5.2-1 3.5.3 ECCS - Shutdown 3.5.3-1 3.5.4 Refueling Water Storage Tank (RWST) 3.5.4-1 3.5.5 Seal Injection Flow 3.5.5-1 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment..
3.6.1-1 3.6.2 Containment Air Locks 3.6.2-1 3.6.3 Containment Isolation Valves 3.6.3-1 3.6.4 Containment Pressure 3.6.4-1 3.6.5 Containment Air Temperature 3.6.5-1 3.6.6 Containment Spray System 3.6.6-1 3.6.7 Spray Additive System 3.6.7-1 3.6.8 Deleted 3.6.8-1 3.6.9 Distributed Ignition System (DIS) 3.6.9-1 3.6.10 Containment Air Recirculation/Hydrogen Skimmer (CEQ) System 3.6.10-1 3.6.11 Ice Bed 3.6.11-1 3.6.12 Ice Condenser Doors 3.6.12-1 3.6.13 Divider Barrier Integrity 3.6.13-1 3.6.14 Containment Recirculation Drains 3.6.14-1 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs) 3.7.1-1 Table 3.7.1-1, OPERABLE Main Steam Safety Valves versus Maximum Allowable Power 3.7.1-3 Table 3.7.1-2, Main Steam Safety Valve Lift Settings 3.7.1-4 3.7.2 Steam Generator Stop Valves (SGSVs) 3.7.2-1 3.7.3 Main Feedwater Isolation Valves (IVIFIVs) and Main Feedwater Regulation Valves (MFRVs) 3.7.3-1 3.7.4 Steam Generator (SG) Power Operated Relief Valves (PORVs) 3.7.4-1 3.7.5 Auxiliary Feedwater (AFW) System 3.7.5-1 3.7.6 Condensate Storage Tank (CST) 3.7.6-1 3.7.7 Component Cooling Water (CCW) System 3.7.7-1 3.7.8 Essential Service Water (ESW) System 3.7.8-1 3.7.9 Ultimate Heat Sink (UHS) 3.7.9-1 3.7.10 Control Room Emergency Ventilation (CREV) System 3.7.10-1 3.7.11 Control Room Air Conditioning (CRAC) System 3.7.11-1 3.7.12 Engineered Safety Features (ESF) Ventilation System 3.7.12-1 3.7.13 Fuel Handling Area Exhaust Ventilation (FHAEV) System 3.7.13-1 Cook Nuclear Plant Unit 2 Page 3 of 5 Amendment No. 2e9 296
PAM Instrumentation 3.3.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. One or more Functions with two or more required channels inoperable.
D.1 Restore all but one channel to OPERABLE status.
7 days E. Required Action and associated Completion Time of Condition C or D not met.
E.1 Enter the Condition referenced in Table 3.3.3-1 for the channel.
Immediately F. As required by Required Action E.1 and referenced in Table 3.3.3-1.
F.1 AND Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> F.2 Be in MODE 4.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> G. As required by Required Action E.1 and referenced in Table 3.3.3-1.
G.1 Initiate action in accordance with Specification 5.6.6.
Immediately Cook Nuclear Plant Unit 2 3.3.3-2 Amendment No. ~ '296
PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS
N()TE----------------------------------------------------------
These SRs apply to each PAM instrumentation Function in Table 3.3.3-1, except where identified in the SR.
SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.
31 days SR 3.3.3.2 Deleted.
N()TE-----------------------------
Neutron detectors are excluded from CHANNEL CALI BRATI()N.
Perform CHANNEL CALIBRATI()N.
124 months Cook Nuclear Plant Unit 2 3.3.3-3 Amendment No. ~ 296
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 2)
Post Accident Monitoring Instrumentation CONDITION REFERENCED FROM REQUIRED FUNCTION REQUIRED CHANNELS ACTION E.1
- 1.
Neutron Flux 2
F
- 2.
Steam Generator Pressure (per steam generator) 2 F
- 3.
Reactor Coolant System (RCS) Hot Leg 2
F Temperature (Wide Range)
- 4.
RCS Cold Leg Temperature (Wide Range) 2 F
- 5.
RCS Pressure (Wide Range) 2 F
- 6.
Reactor Coolant Inventory Tracking System 2
G (Reactor Vessel Level Indication)
- 7.
Containment Water Level 2
F
- 8.
Containment Pressure (Narrow Range) 2 F
- 9.
Penetration Flow Path Containment Isolation Valve 2 per penetration flow F
Position path(8)(b)
- 10.
Containment Area Radiation (High Range) 2 G
- 11.
Deleted
- 12.
Pressurizer Level 2
F
- 13.
Steam Generator Water Level (Wide Range) 4 F
- 14.
Condensate Storage Tank Level G
2(c)
- 15.
Core Exit Temperature - Quadrant 1 F
2(C)
- 16.
Core Exit Temperature - Quadrant 2 F
2(C)
- 17.
Core Exit Temperature - Quadrant 3 F
2(C)
- 18.
Core Exit Temperature - Quadrant 4 F
2(d)
- 19.
Secondary Heat Sink Indication F
(per steam generator)
(a) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
(b) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.
(c) A channel consists of one core exit thermocouple (CET).
(d) Any combination of two instruments per steam generator, including Steam Generator Water Level (Narrow Range) and Auxiliary Feedwater Flow, can be used to satisfy Function 19 OPERABILITY requirements.
Cook Nuclear Plant Unit 2 3.3.3-4 Amendment No. ~ 296
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2)
Post Accident Monitoring Instrumentation CONDITION REFERENCED FROM REQUIRED FUNCTION REQUIRED CHANNELS ACTION E.1 2(e)
- 20.
Emergency Core Cooling System Flow (per train)
F
- 21.
Containment Pressure (Wide Range) 2 F
- 22.
Refueling Water Storage Tank Level 2
F 1(I)
- 23.
RCS Subcooling Margin Monitor F
- 24.
Component Cooling Water Pump Circuit Breaker 2
G Status
- 25.
Containment Recirculation Sump Water Level 2
F (e) Any combination of two instruments per train, including Centrifugal Charging Pump Flow, Safety Injection Pump Flow, Centrifugal Charging Pump Circuit Breaker Status, and Safety Injection Pump Circuit Breaker Status, can be used to satisfy Function 20 OPERABILITY requirements.
(f) An OPERABLE plant process computer (PPC) subcooling margin readout can be used as a substitute for an inoperable Function 23, ReS Subcooling Margin Monitor.
Cook Nuclear Plant Unit 2 3.3.3-5 Amendment No. ~, 2-&2-296
3.6.8 3.6 CONTAINMENT SYSTEMS 3.6.8 Deleted Cook Nuclear Plant Unit 2 3.6.8-1 Amendment No. ~ 296
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Post Accident Monitoring Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:
- a.
The scope of inspections performed on each SG,
- b.
Active degradation mechanisms found,
- c.
Nondestructive examination techniques utilized for each degradation mechanism,
- d.
Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e.
Number of tubes plugged during the inspection outage for each active degradation mechanism,
- f.
Total number and percentage of tubes plugged to date, and
- g.
The results of condition monitoring, including the results of tube pulls and in situ testing.
Cook Nuclear Plant Unit 2 5.6-4 Amendment No. ~, 2+G, 2-7-9 296
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 313 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 296 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316
1.0 INTRODUCTION
By letter dated September 8, 2010, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102590353), Indiana Michigan Power Company (I&M, the licensee) submitted a license amendment request regarding the Donald C. Cook Nuclear Plant Unit 1 and Unit 2 (CNP) Renewed Facility Operating Licenses. The proposed amendment would delete the Technical Specification (TS) requirements related to the containment hydrogen recombiners and the hydrogen monitors.
The Nuclear Regulatory Commission (NRC) has revised Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, "Standards for Combustible Gas Control System in Light Water-Cooled Power Reactors." The amended standards eliminated the requirements for hydrogen recombiners and relaxed the requirements for hydrogen and oxygen monitoring. In letters dated December 17, 2002, and May 12, 2003, the Nuclear Energy Institute (NEI)
Technical Specification Task Force (TSTF) proposed to remove requirements for hydrogen recombiners and hydrogen and oxygen monitors from the standard technical specifications (STS) (NUREGs 1430 - 1434) on behalf of the industry to incorporate the amended standards.
This proposed change is designated TSTF-447.
The NRC staff prepared a model safety evaluation (SE) for the elimination of requirements regarding containment hydrogen recombiners and the removal of requirements from TS for containment hydrogen and oxygen monitors and solicited public comment (67 FR 50374, published August 2, 2002) in accordance with the Consolidated Line Item Improvement Process (CLlIP). The use of the CLlIP in this matter is intended to help the NRC to efficiently process amendments that propose to remove the hydrogen recombiner and hydrogen and oxygen monitor requirements from TS. Licensees of nuclear power reactors to which this model applies were informed (68 FR 55416; September 25,2003) that they could request amendments conforming to the model, and, in such requests, should confirm the applicability of this SE to their reactors and provide the requested plant-specific verifications and commitments.
Enclosure
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2.0 BACKGROUND
Regulatory Issue Summary 2000-06, "Consolidated Line Item Improvement Process for Adopting Standard Technical Specification Changes for Power Reactors," was issued on March 20, 2000. The CUIP is intended to improve the efficiency of NRC licensing processes.
This is accomplished by processing proposed changes to the STS in a manner that supports subsequent license amendment applications. The CUIP includes an opportunity for the public to comment on proposed changes to the STS following a preliminary assessment by the NRC staff and finding that the change will likely be offered for adoption by licensees. The NRC staff evaluates any comments received for a proposed change to the STS and either reconsiders the change or proceeds with announcing the availability of the change for proposed adoption by licensees. Those licensees opting to apply for the subject change to their plant TS are responsible for reviewing the NRC staffs evaluation, referencing the applicable technical justifications, and providing any necessary plant-specific information. Each amendment application made in response to the notice of availability would be processed and noticed in accordance with applicable rules and NRC procedures.
The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36. This regulation requires that the TSs include items in five specific categories.
These categories include (1) safety limits, limiting safety system settings and limiting control settings, (2) limiting conditions for operation (LCO), (3) surveillance requirements, (4) design features, and (5) administrative controls. However, the regulation does not specify the particular TSs to be included in a plant's license.
Additionally, 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether an LCO is required to be included in the TS. These criteria are as follows:
- 1.
Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- 2.
A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that assumes either the failure of or presents a challenge to the integrity of a fission product barrier.
- 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.
- 4.
A structure, system or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
Existing LCOs and related surveillances included as TS requirements which satisfy any of the criteria stated above must be retained in the TSs. Those TS requirements which do not satisfy these criteria may be relocated to other licensee-controlled documents.
As part of the rulemaking that revised 10 CFR 50.44, the Commission retained requirements for ensuring a mixed atmosphere, inerting Mark I and II containments, and providing hydrogen control systems capable of accommodating an amount of hydrogen generated from a metal water reaction involving 75 percent of the fuel cladding surrounding the active fuel region in Mark III and ice condenser containments. The Commission eliminated the design-basis
- 3 loss-of-coolant accident (LOCA) hydrogen release from 10 CFR 50.44 and consolidated the requirements for hydrogen and oxygen monitoring to 10 CFR 50.44 while relaxing safety classifications and licensee commitments to certain design and qualification criteria. The Commission also relocated without change the hydrogen control requirements in 10 CFR 50.34(f) to 10 CFR 50.44 and the high point vent requirements from 10 CFR 50.44 to 10 CFR 50.46a.
3.0 EVALUATION The ways in which the requirements and recommendations for combustible gas control were incorporated into the licensing bases of commercial nuclear power plants varied as a function of when a particular plant was licensed. Plants that were operating at the time of the Three Mile Island (TMI), Unit 2, accident are likely to have been the subject of confirmatory orders that imposed the combustible gas control functions described in NUREG-0737, "Clarification of TMI Action Plan Requirements," as obligations. The issuance of plant-specific amendments to adopt these changes, which would remove hydrogen recombiner and hydrogen and oxygen monitoring controls from the TS, supersede the combustible gas control specific requirements imposed by post-TMI confirmatory orders.
3.1 Hydrogen Recombiners The revised 10 CFR 50.44 no longer defines a design-basis LOCA hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant beyond design-basis accidents. Therefore, the Commission eliminated the hydrogen release associated with a design-basis LOCA from 10 CFR 50.44 and the associated requirements that necessitated the need for the hydrogen recombiners and the backup hydrogen vent and purge systems. As a result, the NRC staff finds that requirements related to hydrogen recombiners no longer meet any of the four criteria in 10 CFR 50.36(c)(2)(ii) for retention in the TS and may be relocated to other licensee-controlled documents for all plants.
Hydrogen recombiners are no longer required by 10 CFR 50.44 and no longer meet the criteria in 10 CFR 50.36(c)(2)(ii). Therefore, removal of the requirements for hydrogen recombiners from the Donald C. Cook TS and relocation to other licensee-controlled documents is an acceptable change.
3.2 Hydrogen Monitoring Equipment Section 50.44(b)(1), the STS, and plant-specific TS currently contain requirements for monitoring hydrogen. Licensees have also made commitments to design and qualification criteria for hydrogen monitors in Item II.F.1, Attachment 6 of NUREG-0737 and Regulatory Guide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident." The hydrogen monitors are required to assess the degree of core damage during a beyond-design-basis accident and confirm that
- 4 random or deliberate ignition has taken place. If an explosive mixture that could threaten containment integrity exists during a beyond-design-basis accident, then other severe accident management strategies, such as purging and/or venting, would need to be considered. The hydrogen monitors are needed to implement these severe accident management strategies.
With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. RG 1.97 recommends classifying the hydrogen monitors as Category 1. RG 1.97 Category 1 is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events and, therefore, are items usually addressed within TS. As part of the rulemaking to revise 10 CFR 50.44, the Commission found that the hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. The Commission concluded that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond-design-basis accidents.
Hydrogen monitoring is not the primary means of indicating a significant abnormal degradation of the reactor coolant pressure boundary. Section 4 of Attachment 2 to SECY-00-0198, "Status Report on Study of Risk-Informed Changes to the Technical Requirements of 10 CFR Part 50 (Option 3) and Recommendations on Risk-Informed Changes to 10 CFR 50.44 (Combustible Gas Control)," found that the hydrogen monitors were not risk-significant. Therefore, the NRC staff finds that hydrogen monitoring equipment requirements no longer meet any of the four criteria in 10 CFR 50.36(c)(2)(ii) for retention in the TS and, therefore, may be relocated to other licensee-controlled documents.
The elimination of Post-Accident Sampling System requirements from some plant-specific TS (and associated CUIP notices) indicated that during the early phases of an accident, safety-grade hydrogen monitors provide an adequate capability for monitoring containment hydrogen concentration. The NRC staff has subsequently concluded that Category 3 hydrogen monitors also provide an adequate capability for monitoring containment hydrogen concentration during the early phases of an accident.
However, because the monitors are required to diagnose the course of beyond-design-basis accidents, each licensee should verify that it has, and make a regulatory commitment to maintain, a hydrogen monitoring system capable of diagnosing beyond-design-basis accidents.
The licensee has verified that a hydrogen monitoring system capable of diagnosing beyond design-basis accidents is installed at Donald C. Cook and is making a regulatory commitment to maintain that capability. The licensee stated that the hydrogen monitoring system will be included in the CNP Unit 1 and Unit 2 Technical Requirements Manuals (TRMs). Therefore, removal of the requirements for hydrogen monitoring equipment from TS and placement in the TRMs is an acceptable change.
The deletion of the requirements for the hydrogen recombiner and hydrogen monitors resulted in deletion of TS Bases content for hydrogen recombiners and monitors. The NRC staff has confirmed that the related changes are appropriate and do not affect the technical requirements.
4.0 VERIFICATION AND COMMITMENTS As requested by the NRC staff in the notice of availability for this TS improvement, the licensee has addressed the following plant-specific verifications and commitments.
- 5 4.1 Each licensee should verify that it has, and make a regulatorv commitment to maintain. a hydrogen monitoring system capable of diagnosing beyond design-basis accidents.
The licensee has verified that it has a hydrogen monitoring system capable of diagnosing beyond-design-basis accidents. The licensee has committed to include requirements for the hydrogen monitors within the CNP Units 1 and 2 TRMs. The licensee will implement this commitment as part of the implementation of the amendment.
4.2 For plant designs with an inerted containment. each licensee should verify that it has, and make a regulatorv commitment to maintain. an oxygen monitoring system capable of verifying the status of the inert containment.
Verification and commitment 4.2 is not applicable to CNP Units 1 and 2, which do not have inerted containments.
The NRC staff finds that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitments are provided by the licensee's administrative processes, including its commitment management program.
Should the licensee choose to incorporate a regulatory commitment into the emergency plan, final safety analysis report, or other document with established regulatory controls, the associated regulations would define the appropriate change-control and reporting requirements.
The NRC staff has determined that the commitments do not warrant the creation of regulatory requirements which would require prior NRC approval of subsequent changes. The NRC staff has agreed that NEI 99-04, Revision 0, "Guidelines for Managing NRC Commitment Changes,"
provides reasonable guidance for the control of regulatory commitments made to the NRC staff.
(see Regulatory Issue Summary 2000-17, "Managing Regulatory Commitments Made by Power Reactor Licensees to the NRC Staff," dated September 21,2000). The commitments should be controlled in accordance with the industry guidance or comparable criteria employed by a specific licensee. The NRC staff may choose to verify the implementation and maintenance of these commitments in a future inspection or audit.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
These amendments change requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluent 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 amendments involve no significant hazards consideration and there has been no public comment on such finding (75 FR 63209). Accordingly, the amendments meet 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 amendments.
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7.0 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Matthew Hamm, NRR Date:
December 14, 2010
December 14,2010 Mr. Lawrence J. Weber Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 SUB~IECT:
DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: ELIMINATION OF REQUIREMENTS FOR HYDROGEN RECOMBINERS AND HYDROGEN MONITORS (TAC NO. ME4709 AND ME4710)
Dear Mr. Weber:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 313 to Renewed Facility Operating License No. DPR-58 and Amendment No. 296 to Renewed Facility Operating License No. DPR-74 for the Donald C. Cook Nuclear Plant, Units 1 and 2. The amendments consist of changes to the Technical Specifications in response to your application dated September 8, 2010.
The amendments delete the Technical Specification requirements related to the containment hydrogen recombiners and the hydrogen monitors, in accordance with Nuclear Energy Institute Technical Specification Task Force (TSTF) initiative designated as TSTF-447.
A copy of our related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/RAJ Peter S. Tam, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosures:
- 1. Amendment No. 313 to DPR-58
- 2. Amendment No. 296 to DPR-74
- 3. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION PUBLIC LPL3-1 rtf RidsNrrDorlLpl3-1 Resource RidsNrrPMDCCook Resource RidsNrrLABTully Resource RidsOgcRp Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDirsltsb Resource RidsNrrDorlDprResource RidsRgn3MailCenter Resource M. Hamm, NRR OFFICE NRR/LPL3-1/PM t\\lRR/LPL3-1/LA TECH BR*
OGC**
NRR/LPL3-1/BC NAME PTam BTuily RElliott*
RPascarelli DATE 11/23/10 11/19/10 10/14/10*
12/14/10 Accession No' ML103210032
- Memo transmitting SE (Accession No. ML102670538 "OGe review was waived (see Accession No. ML072980209).
OFFICIAL RECORD COPY