Text

Issuance of Amendments Regarding Incorporation of TSTF-448, Revision 3, Control Room Habitability
	ML083430469
Person / Time
Site:	Cook
Issue date:	12/30/2008
From:	Beltz T; Plant Licensing Branch III
To:	Rencheck M; Indiana & Michigan Electric Co
	beltz T, NRR/DORL/LPL3-1, 301-415-3049
References
	TAC MD7554, TAC MD7555
	Download: ML083430469 (40)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 December 30, 2008 Mr. Michael W. Rencheck 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 REGARDING INCORPORATION OF TSTF-448, REVISION 3, "CONTROL ROOM HABITABILITY" (TAC NOS. MD7554 AND MD7555)

Dear Mr. Rencheck:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 307 and 289 to Renewed Facility Operating License Nos. DPR-58 and DPR-74, for Units 1 and 2 of the Donald C. Cook Nuclear Plant. The amendments are in response to your application dated December 27,2007, as supplemented by letter dated July 28,2008, for implementation of the Technical Specification (TS) Task Force Traveler (TSTF)-448, Revision 3, "Control Room Habitability. "

The amendment establishes more effective and appropriate action, surveillance, and administrative requirements related to ensuring the habitability of the control room envelope in accordance with the NRC-approved TSTF-448, Revision 3, and changes the TSs related to the control room emergency ventilation system in TS Section 3.7.10, "Control Room Emergency Ventilation (CREV) System," and TS Section 5.5.16, "Control Room Envelope Habitability Program." The amendment also adds a license condition to support implementation of the TS changes.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Terry. Beltz, 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. 307 to DPR-58

2. Amendment No. 289 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 TO FACILITY OPERATING LICENSE Amendment No. 307 Renewed License No. DPR-58

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indian Michigan Power Company (the licensee) dated December 27,2007, as supplemented by letter dated July 28, 2008, 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 Appendix B, as revised through Amendment No. 307, is hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Further, Renewed Facility Operating License No. DPR-58 will be amended to add the following license condition 2.C.(18), to read as follows:

- 2 (18)

Upon implementation of Amendment No. 307 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by TS Surveillance Requirement (SR) 3.7.10.4, in accordance with TS 5.5.16.c.(i), the assessment of CRE habitability as required by TS 5.5.16.c.(ii),

and the measurement of CRE pressure as required by TS 5.5.16.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with TS 5.5.16.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from June 1999, the date of the most recent successful tracer gas test, as stated in the December 4, 2003, letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 6.5.8.c.(ii), shall be within 3 years, plus the 9-month allowance of TS 4.0.2, as measured from February 19, 2004, the date of the most recent tracer gas test, as stated in the January 31, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent tracer gas test is greater than 3 years.

(c) The first performance of the periodic assessment of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within the next 182 days if not performed previously.

4.

This license amendment is effective as of the date of its issuance and shall be implemented within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION

.~:J J~~:tCh~;f~J Plant Licensing Branch 3-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: December 30, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 307 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 DOCKET NO. 50-315 Replace the following pages of the Renewed Facility Operating License with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT

- 3

- 48

- 5

- 5 (new)

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 5 of 5 Table of Contents, Page 5 of 5 3.7.10-1 3.7.10-1 3.7.10-2 3.7.10-2 3.7.10-3 3.7.10-3 5.5-14 5.5-14 5.5-15 (new)

- 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, Sections 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 herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A and Appendix B, as revised through Amendment No. 307 are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(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 four 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 30, 1981, February 7, 1983, November 22,1983, December 23,1983, March 16,1984, August 27,1985, Renewed License No. DPR-58 Amendment No. we, JOO, 307

- 48 (c)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders (16) The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20,2006, except the last action that requires incorporation ofthe strategies into the site security plan, contingency plan, emergency plan arid/or guard training and qualification plan, as appropriate.

(17) Ice Condenser Ice Fusion Time Requirement The licensee is authorized to change the Updated Final Safety Analysis Report (UFSAR) to allow inspection of each ice condenser within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of experiencing a seismic event greater than or equal to an operating-basis earthquake within the 5-week period after ice basket replenishment has been completed to confirm that adverse ice fallout has not occurred which could impede the ability of the ice condenser lower inlet doors to open. This action would be taken, in lieu of requiring a 5-week waiting period following ice basket replenishment, prior to beginning ascension to power operations, as set forth in the application for amendment dated February 29, 2008, and evaluated in the safety evaluation accompanying Amendment No. 303. The licensee shall update the UFSAR by adding a description of this change, as authorized by this amendment, and in accordance with 10 CFR 50.71(e).

(18) Upon implementation of Amendment No. 307 adopting TSTF-448, Revision 3, the determination of CRE unfiltered air inleakage as required by SR 3.7.1OA, in accordance with TS 5.5.16.c.(i), the assessment of CRE habitability as required by TS 5.5.16.c.(ii), and the measurement of CRE pressure as required by TS 5.5.16.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.1OA, in accordance with TS 5.5.16.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from June 1999, the date of the most recent successful tracer gas test, as stated in the December 4, 2003, letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, TS 5.5.16.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from June 1999, the date of the most recent successful tracer gas test, as stated in the December 4, 2003, letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 182 days if not performed previously.

Renewed License No. DPR-58 Corrected by letter dated 4/21/06 12/14/06 Revised by letter dated August 9, 2007 Amendment No. 300, 307

- 5 D.

Physical Protection The Indiana Michigan Power Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revision to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans", which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Donald C. Cook Nuclear Plant Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 1,"

submitted by letter dated May 10, 2006.

E.

Deleted by Amendment No. 80 F.

Deleted by Amendment No. 80 G. In all places of this renewed operating license, the reference to the Indiana and Michigan Electric Company is amended to read Indiana Michigan Power Company.

H.

Deleted by Amendment No. 287 I.

Deleted by Amendment No. 287 J. The licensee is authorized to use digital signal processing instrumentation in the reactor protection system.

K. Updated Final Safety Analysis Report The Indiana Michigan Power Company Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21 (d), describes certain future activities to be completed prior to the period of extended operation. The Indiana Michigan Power Company shall complete these activities no later than October 25,2014, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4) following issuance of this renewed operating license. Until that update is complete, Indiana Michigan Power Company may make changes to the programs and activities described in the supplement without prior Commission approval, provided that Indiana Michigan Power Company evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-58 Corrected by letter dated 4/21/06 12/14/06 Revised by letter dated August 9, 2007

- 6 L. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.

3.

This renewed operating license is effective as of the date of issuance and shall expire at midnight, October 25, 2034.

FOR THE NUCLEAR REGULATORY COMMISSION IRAJ J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1.

Appendix A - Technical Specifications

2.

Appendix 8 - Environmental Technical Specifications Date of Issuance: August 30, 2005 Renewed License No. DPR-58

UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1-1 5.2 Organization 5.2-1 5.2.1 Onsite and Offsite Organizations 5.2-1 5.2.2 Unit Staff 5.2-1 5.3 Unit Staff Qualifications 5.3-1 5.4 Procedures 5.4-1 5.5 Programs and Manuals 5.5-1 5.5.1 Offsite Dose Calculation Manual (ODCM) 5.5-1 5.5.2 Leakage Monitoring Program 5.5-2 5.5.3 Radioactive Effluent Controls Program 5.5-2 5.5.4 Component Cyclic or Transient Limits 5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program 5.5-4 5.5.6 Inservice Testing Program 5.5-4 5.5.7 Steam Generator (SG) Program 5.5-5 5.5.8 Secondary Water Chemistry Program 5.5-7 5.5.9 Ventilation Filter Testing Program (VFTP) 5.5-7 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5-10 5.5.11 Diesel Fuel Oil Testing Program 5.5-10 5.5.12 Technical Specifications (TS) Bases Control Program 5.5-11 5.5.13 Safety Function Determination Program (SFDP) 5.5-12 5.5.14 Containment Leakage Rate Testing Program 5.5-13 5.5.15 Battery Monitoring and Maintenance Program 5.5-14 5.5.16 Control Room Envelope Habitability Program 5.5-14 5.6 Reporting Requirements 5.6-1 5.6.1 Deleted 5.6-1 5.6.2 Annual Radiological Environmental Operating Report 5.6-1 5.6.3 Radioactive Effluent Release Report 5.6-2 5.6.4 Deleted 5.6-2 5.6.5 CORE OPERATING LIMITS REPORT (COLR) 5.6-2 5.6.6 Post Accident Monitoring Report 5.6-4 5.6.7 Steam Generator Tube Inspection Report 5.6-4 5.7 High Radiation Area 5.7-1 Cook Nuclear Plant Unit 1 Page 5 of 5 Amendment No. 237, 298, 307

CREV System 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Control Room Emergency Ventilation (CREV) System LCO 3.7.10 Two CREV trains shall be OPERABLE.

NOTE-------------------------------------------

The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY:

MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies in the containment, auxiliary building, and the Unit 2 containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREV train inoperable for reasons other than Condition B.

A.1 Restore CREV train to OPERABLE status.

7 days B. One or more CREV trains inoperable due to inoperable CRE boundary in MODE 1, 2, 3, or 4.

B.1 AND B.2 Initiate action to implement mitigating actions.

Verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits.

Immediately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND B.3 Restore CRE boundary to OPERABLE status.

90 days Cook Nuclear Plant Unit 1 3.7.10-1 Amendment No. 2-8+, 307

3.7.10 CREV System ACTIONS (continued)

CONDITION C. Two CREV trains inoperable due to inoperable filter unit in MODE 1, 2, 3, or 4.

D. Required Action and associated Completion Time of Condition A, B, or C not met in MODE 1, 2,3, or 4.

E. Required Action and associated Completion Time of Condition A not met during movement of irradiated fuel assemblies.

F. Two CREV trains inoperable during movement of irradiated fuel assemblies.

OR One or more CREV trains inoperable due to an inoperable CRE boundary during movement of irradiated fuel assemblies.

I C.1 0.1 AND 0.2 E.1 OR E.2 F.1 REQUIRED ACTION COMPLETION TIME Restore filter unit to OPERABLE status.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Be in MODE 3.

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Place OPERABLE CREV train in pressurization/

cleanup mode.

Suspend movement of irradiated fuel assemblies.

Immediately Immediately Suspend movement of irradiated fuel assemblies.

Immediately G. Two CREV trains Enter LCO 3.0.3.

Immediately t G1 inoperable in MODE 1, 2,3,or4forreasons other than Conditions B and C.

Cook Nuclear Plant Unit 1 3.7.10-2 Amendment 1\\10. 2S+, 307

CREV System 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 SR 3.7.10.2 Operate each CREV train for ~ 15 minutes.

Perform required CREV System filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

46 days on a STAGGERED TEST BASIS In accordance with the VFTP SR 3.7.10.3 SR 3.7.10.4

NOTE------------------------------

Only required to be met in MODES 1, 2, 3, and 4.

Verify each CREV System train actuates on an actual or simulated actuation signal.

Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

24 months In accordance with the Control Room Envelope Habitability Program Cook Nuclear Plant Unit 1 3.7.10-3 Amendment No. 2-37, 307

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Battery Monitoring and Maintenance Prog ram This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer including the following:

a.

Actions to restore battery cells with float voltage < 2.13 V; and

b.

Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

5.5.16 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation (CREV) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c.

Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision O.

The following is an exception to Section C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

The appropriate application of ASTM E741-00 required by C.1.1 may include minor exceptions to the test methodology. These exceptions shall be documented in the test report.

Cook Nuclear Plant Unit 1 5.5-14 Amendment No. 28+, 200, 307

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Control Room Envelope Habitability Program (continued)

d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREV System, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the periodic assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by testing described in Paragraph C. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by Paragraphs C and D, respectively.

Cook Nuclear Plant Unit 1 5.5-15 Amendment No. 307

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 TO FACILITY OPERATING LICENSE Amendment No. 289 Renewed License No. DPR-74

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indian Michigan Power Company (the licensee) dated December 27, 2007, as supplemented by letter dated July 28, 2008, 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 Appendix B, as revised through Amendment No. 289, is hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications..

3.

Further, Renewed Facility Operating License No. DPR-74 will be amended to add the following license condition 2.C.(3)(ff), to read as follows:

- 2 (ff)

Upon implementation of Amendment No. 289 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by TS Surveillance Requirement (SR) 3.7.10.4, in accordance with TS 5.5.16.c.(i), the assessment of CRE habitability as required by TS 5.5.16.c.(ii),

and the measurement of CRE pressure as required by TS 5.5.16.d, shall be considered met. Following implementation:

(I)

The first performance of SR 3.7.10.4, in accordance with TS 5.5.16.c.(i),

shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from June 1999, the date of the most recent successful tracer gas test, as stated in the December 4, 2003, letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent tracer gas test is greater than 6 years.

(II)

The first performance of the periodic assessment of CRE habitability, Specification 6.5.8.c.(ii), shall be within 3 years, plus the 9-month allowance of TS 4.0.2, as measured from February 19, 2004, the date of the most recent tracer gas test, as stated in the January 31, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent tracer gas test is greater than 3 years.

(III)

The first performance of the periodic assessment of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within the next 182 days if not performed previously.

4.

This license amendment is effective as of the date of its issuance and shall be implemented within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION

~c J~~:~et,c-J Plant Licensing Branch 3-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: December 30, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 289 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Replace the following pages of the Renewed Facility Operating License with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT

- 3

- 58

- 6

- 6 (new)

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 5 of 5 Table of Contents, Page 5 of 5 3.7.10-1 3.7.10-1 3.7.10-2 3.7.10-2 3.7.10-3 3.7.10-3 5.5-14 5.5-14 5.5-15 (new)

- 3 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, Sections 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 herein and in Attachment 1 to the renewed operating license.

The preoperational tests, startup tests 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 Appendix B, as revised through Amendment No. 289 are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(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 leak test each of the two valves in series in the Renewed License No. DPR-74 Amendment No. 288, 289

- 58

6.

Training on integrated fire response strategy

7.

Spent fuel pool mitigation measures (III)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders (dd) The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

(ee) Ice Condenser Ice Fusion Time Requirement The licensee is authorized to change the Updated Final Safety Analysis Report (UFSAR) to allow inspection of each ice condenser within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of experiencing a seismic event greater than or equal to an operating-basis earthquake within the 5-week period after ice basket replenishment has been completed to confirm that adverse ice fallout has not occurred which could impede the ability of the ice condenser lower inlet doors to open. This action would be taken, in lieu of requiring a 5-week waiting period following ice basket replenishment, prior to beginning ascension to power operations, as set forth in the application for amendment dated February 29, 2008, and evaluated in the safety evaluation accompanying Amendment No. 286. The licensee shall update the UFSAR by adding a description of this change, as authorized by this amendment, and in accordance with 10 CFR 50.71(e).

(ff) Upon implementation of Amendment No. 289 adopting TSTF-448, Revision 3, the determination of CRE unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.16.c.(i), the assessment of CRE habitability as required by TS 5.5.16.c.(ii), and the measurement of CRE pressure as required by TS 5.5.16.d, shall be considered met. Following implementation:

(I) The first performance of SR 3.7.10.4, in accordance with TS 5.5.16.c.(i),

shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from June 1999, the date of the most recent successful tracer gas test, as stated in the December 4, 2003, letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(II) The first performance of the periodic assessment of CRE habitability, TS 5.5.16.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from June 1999, the date of the most recent successful tracer gas test, as stated in the December 4, 2003, letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

Renewed License No. DPR-74 Corrected by letter dated 4/21/06 12/14/06 Revised by letter dated August 9, 2007 Amendment No. 28e, 289

- 6 (III) The first performance of the periodic measurement of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 182 days if not performed previously.

D.

Physical Protection The Indiana Michigan Power Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans', which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Donald C. Cook Nuclear Plant Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 1,"

submitted by letter dated May 10, 2006.

E.

Deleted by Amendment No. 63 F.

In all places of this renewed operating license, the reference to the Indiana and Michigan Electric Company is amended to read Indiana Michigan Power Company.

G. Deleted by Amendment No. 269 H.

Deleted by Amendment No. 2E39 I.

Deleted by Amendment No. 261 (1) Deleted by Amendment No. 261 (2) Deleted by Amendment No. 261 J.

The licensee is authorized to use digital signal processing instrumentation in the reactor protection system.

K.

Updated Final Safety Analysis Report The Indiana Michigan Power Company Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21 (d), describes certain future activities to be completed prior to the period of extended operation. The Indiana Michigan Power Company shall complete these activities no later than December 23,2017, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-74 Correoted by letter dated 4/21/06 12/14/06 Revised by letter dated August 9, 2007 Amendment No. 289

- 7 The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed operating license. Until that update is complete, Indiana Michigan Power Company may make changes to the programs and activities described in the supplement without prior Commission approval, provided that Indiana Michigan Power Company evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section..

L. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.

3.

This renewed operating license is effective as of the date of issuance and shall expire at midnight, December 23, 2037.

FOR THE NUCLEAR REGULATORY COMMISSION IRAJ J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1.

Preoperational Tests, Startup Tests and Other Items Which Must Be Completed Prior to Proceeding to Succeeding Operational Modes.

2.

Appendix A - Technical Specifications

3.

Appendix B - Environmental Technical Specifications Date of Issuance: August 30, 2005 Renewed License No. DPR-74

UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1-1 5.2 Organization 5.2-1 5.2.1 Onsite and Offsite Organizations 5.2-1 5.2.2 Unit Staff 5.2-1 5.3 Unit Staff Qualifications 5.3-1 5.4 Procedures 5.4-1 5.5 Programs and Manuals 5.5-1 5.5.1 Offsite Dose Calculation Manual (ODCM) 5.5-1 5.5.2 Leakage Monitoring Program 5.5-2 5.5.3 Radioactive Effluent Controls Program 5.5-2 5.5.4 Component Cyclic or Transient Limits 5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program 5.5-4 5.5.6 Inservice Testing Program 5.5-4 5.5.7 Steam Generator (SG) Program 5.5-5 5.5.8 Secondary Water Chemistry Program 5.5-7 5.5.9 Ventilation Filter Testing Program (VFTP) 5.5-7 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5-10 5.5.11 Diesel Fuel Oil Testing Program 5.5-10 5.5.12 Technical Specifications (TS) Bases Control Program 5.5-11 5.5.13 Safety Function Determination Program (SFDP) 5.5-12 5.5.14 Containment Leakage Rate Testing Program 5.5-13 5.5.15 Battery Monitoring and Maintenance Program 5.5-13 5.5.16 Control Room Envelope Habitability Program 5.5-14 5.6 Reporting Requirements 5.6-1 5.6.1 Deleted 5.6-1 5.6.2 Annual Radiological Environmental Operating Report 5.6-1 5.6.3 Radioactive Effluent Release Report 5.6-2 5.6.4 Deleted 5.6-2 5.6.5 CORE OPERATING LIMITS REPORT (COLR) 5.6-2 5.6.6 Post Accident Monitoring Report 5.6-4 5.6.7 Steam Generator Tube Inspection Report 5.6-4 5.7 High Radiation Area 5.7-1 Cook Nuclear Plant Unit 2 Page 5 of 5 Amendment No. ~, 2-79, 289

CREV System 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Control Room Emergency Ventilation (CREV) System LCO 3.7.10 Two CREV trains shall be OPERABLE.

NOTE-------------------------------------------

The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY:

MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies in the containment, auxiliary building, and the Unit 1 containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREV train inoperable for reasons other than Condition B.

A.1 Restore CREV train to OPERABLE status.

7 days B. One or more CREV trains inoperable due to inoperable CRE boundary in MODE 1, 2, 3, or 4.

B.1 AND B.2 AND B.3 Initiate action to implement mitigating actions.

Verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits.

Restore CRE boundary to OPERABLE status.

Immediately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 90 days Cook Nuclear Plant Unit 2 3.7.10-1 Amendment No. 2W, 289

3.7.10 CREV System ACTIONS (continued)

C. Two CREV trains inoperable due to inoperable filter unit in MODE 1, 2, 3, or4.

D. Required Action and associated Completion Time of Condition A, B, or C not met in MODE 1, 2,3, or 4.

E. Required Action and associated Completion Time of Condition A not met during movement of irradiated fuel assemblies.

F. Two CREV trains inoperable during movement of irradiated fuel assemblies.

OR One or more CREV trains inoperable due to an inoperable eRE boundary during movement of irradiated fuel assemblies.

G. Two CREV trains inoperable in MODE 1, 2,3,or4forreasons other than Conditions B and C.

C.1 Restore filter unit to OPERABLE status.

D.1 Be in MODE 3.

AND D.2 Be in MODE 5.

E.1 Place OPERABLE CREV train in pressurization/

cleanup mode.

OR E.2 Suspend movement of irradiated fuel assemblies.

F.1 Suspend movement of irradiated fuel assemblies.

G.1 Enter LCO 3.0.3.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Immediately Immediately Immediately Immediately Cook Nuclear Plant Unit 2 3.7.10-2 Amendment 1\\10. 2W, 289

CREV System 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CREV train for ~ 15 minutes.

46 days on a STAGGERED TEST BASIS SR 3.7.10.2 Perform required CREV System filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.7.10.3

NOTE------------------------------

Only required to be met in MODES 1, 2, 3, and 4.

Verify each CREV System train actuates on an actual or simulated actuation signal.

24 months SR 3.7.10.4 Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

In accordance with the Control Room Envelope Habitability Program Cook Nuclear Plant Unit 2 3.7.10-3 Amendment No. 299, 289

5.5 Programs and Manuals 5.5 Programs and Manuals 5.5.16 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation (CREV) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision O.

The following is an exception to Section C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

The appropriate application of ASTM E741-00 required by C.1.1 may include minor exceptions to the test methodology. These exceptions shall be documented in the test report.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREV System, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the periodic assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in Paragraph C. The unfiltered air inleakage limit for.radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

Cook Nuclear Plant Unit 2 5.5-14 Amendment No. 289

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Control Room Envelope Habitability Program (continued)

f.

The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by Paragraphs C and D, respectively.

Cook Nuclear Plant Unit 2 5.5-15 Amendment No. 289

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 307 TO FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 289 TO FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-315AND 50-316

1.0 INTRODUCTION

By letter dated December 27,2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080160132), as supported by letter dated December 4, 2003 (ADAMS Accession No. ML033460373) and supplemented by letter dated July 28,2008, (ADAMS Accession No. ML082321110), Indiana Michigan Power Company (I&M) (the licensee) requested changes to the Technical Specifications (TS) for Units 1 and 2 of the Donald C. Cook Nuclear Plant (CNP). The staffs proposed no significant hazards consideration determination was published in the Federal Register on January 29,2008 (73 FR 5224).

The proposed amendment would modify the CNP-1 and CNP-2 TS requirements related to control room envelope (CRE) habitability in TS Section 3.7.10, "Control Room Emergency Ventilation (CREV) System," and TS Section 5.5, "Programs and Manuals." The proposed changes are consistent with Nuclear RegUlatory Commission (NRC)-approved Technical Specification Task Force (TSTF)-448, Revision 3, "Control Room Habitability." The amendment would also add a license condition to support implementation of the TS changes.

The supplemental letter dated July 28, 2008, provided information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's initial proposed no significant hazards consideration determination as published in the Federal Register on January 28, 2008.

On August 8,2006, the commercial nuclear electrical power generation industry owners group TSTF submitted a proposed change, TSTF-448, Revision 3, to the improved standard technical specifications (STS) (NUREGs 1430-1434) on behalf of the industry (TSTF-448, Revisions 0, 1, and 2 were prior draft iterations). TSTF-448, Revision 3, proposed to establish more effective and appropriate action, surveillance, and administrative STS requirements related to ensuring the habitability within the CRE.

In NRC Generic Letter (GL) 2003-01 (Reference 1), licensees were alerted to findings at facilities that existing TS surveillance requirements (SR) for the control room habitability systems may be inadequate. Specifically, the results of ASTM E741 (Reference 2) tracer gas tests to measure CRE unfiltered inleakage at facilities indicated that the differential pressure surveillance is not a reliable method for demonstrating CRE boundary operability. Licensees were requested to address existing TS as follows:

- 2 Provide confirmation that your TSs verify the integrity [i.e., operability] of the CRE

[boundary], and the assumed [unfiltered] inleakage rates of potentially contaminated air.

If you currently have a differential pressure SR to demonstrate CRE [boundary] integrity, provide the basis for your conclusion that it remains adequate to demonstrate CRE integrity in light of the ASTM E741 testing results. If you conclude that your differential pressure SR is no longer adequate, provide a schedule for:

1) Revising the SR in your TS to reference an acceptable surveillance methodology (e.g., ASTM E741); and

2) Making any necessary modifications to your CRE boundary so that compliance with your new SR can be demonstrated.

If your facility does not currently have a TS SR for your CRE integrity, explain how and at what frequency you confirm your CRE integrity and why this is adequate to demonstrate CRE integrity.

To promote standardization and minimize the resources that would be needed to create and process plant-specific amendment appiications in response to the concerns described in the GL, the industry and the NRC proposed revisions to CRE habitability system requirements contained in the STS, using the STS change traveler process. This effort culminated in Revision 3 to traveler TSTF-448, "Control Room Habitability," which the NRC staff approved on January 17, 2007.

Consistent with the traveler as incorporated into I'JUREG-1431, the licensee proposed revising actions and SRs in Specification 3.7.10, "Control Room Emergency Ventilation System (CREV),"

and adding a new administrative controts program, Specification 5.5.16, "Control Room Envelope Habitability Program." The purpose of the changes is to ensure that CRE boundary operability is maintained and verified through effective surveillance and programmatic requirements, and that appropriate remedial actions are taken in the event of an inoperable CRE boundary.

Some editorial and plant-specific changes were incorporated into this safety evaluation resulting in minor deviations from the model safety evaluation text in TSTF-448, Revision 3.

2.0 REGULATORY EVALUATION

The NRC staff finds that the CNP-1 and CNP-2 identified the applicable regulatory requirements in the December 27,2007, submittal. The regulatory requirements and guidance which the NRC staff considered in its review of the application are as follows:

2.1 Control Room and Control Room Envelope NRC Regulatory Guide (RG) 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors," Revision 0, May 2003, (Reference 4) uses the term "control room envelope" in addition to the term "control room" and defines each term as follows:

Control Room: The plant area, defined in the facility licensing basis, in which actions can be taken to operate the plant safely under normal conditions and to maintain the reactor

-3 in a safe condition during accident situations. It encompasses the instrumentation and controls necessary for a safe shutdown of the plant and typically includes the critical document reference file, computer room (if used as an integral part of the emergency response plan), shift supervisor's office, operator wash room and kitchen, and other critical areas to which frequent personnel access or continuous occupancy may be necessary in the event of an accident.

Control Room Envelope (CRE): The plant area, defined in the facility licensing basis that, in the event of an emergency, can be isolated from the plant areas and the environment external to the CRE. This area is served by an emergency ventilation system, with the intent of maintaining the habitability of the control room. This area encompasses the control room, and may encompass other non-critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident.

NRC RG 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003 (Reference 5), also contains these definitions, but uses the term CRE to mean both. This is because the protected environment provided for operators varies with the nuclear power facility. At some facilities this environment is limited to the control room; at others, it is the CRE. In this safety evaluation, consistent with the proposed changes to the STS, the CRE will be used to designate both environments as defined above. For consistency, facilities should use the term CRE with an appropriate facility-specific definition derived from the above CRE definition.

2.2 Control Room Emergency Ventilation (CREY) System The CREV System (the term used at CNP-1 and CNP-2 for the Control Room Envelope Emergency Ventilation System (CREEVS>> provides a protected environment from which operators can control the unit, during airborne challenges from radioactivity, hazardous chemicals, and fire byproducts, such as fire suppression agents and smoke, during both normal and accident conditions.

The CREV System is designed to maintain a habitable environment in the control room envelope for 30 days of continuous occupancy after a Design-Basis Accident (DBA) without exceeding a 5 Roentgen equivalent man (Rem) total effective dose equivalent (TEDE).

The CREV System consists of two trains that share a common high efficiency particulate air (HEPA) filter and charcoal adsorber. With the exception of the shared HEPA filter and charcoal adsorber, each CREV train is redundant and capable of maintaining the habitability of the CRE.

Each CREV train is considered operable when the individual components necessary to limit CRE occupant exposure are operable. A CREV train is considered operable when the following apply:

Fan is operable; HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration functions; Ductwork, valves, and dampers are operable, and air circulation can be maintained; and CRE boundary is operable (the single boundary supports both trains).

-4 The CRE boundary is considered operable when the measured unfiltered air inleakage is less than or equal to the inleakage value assumed by the licensing basis analyses of DBA consequences to CRE occupants.

2.3 Regulations Applicable to Control Room Habitability In a memorandum dated September 18, 1992, the Commission approved the staff proposal in SECY-92-223, "Resolution of Deviations Identified During the Systematic Evaluation Program,"

not to apply 10 CFR Part 50, Appendix A, "General Design Criteria [GOG] for Nuclear Power Plants," to plants with construction permits prior to May 21, 1971. CNP-1 and CNP-2 were licensed for construction prior to May 21, 1971, and at that time committed to the draft GDC.

These became know as the CNP Plant Specific Design Criteria (PSDC) when the CNP Updated Final Safety Analysis Report (UFSAR) was developed. The CNP PSDC, which are similar to the GDC for nuclear power plants described in Appendix A to 10 CFR Part 50, are contained in Section 1.4 of the CNP UFSAR (Reference 7).

In its response to NRC GL 2003-01, by letter dated December 4, 2003, the licensee listed the PSDC and GDC applicable to CRE habitability at CNP-1 and CNP-2. Specifically, PSDC 1,2,3, 4, 5, 11, 40 and GDC 19 are applicable to CNP CRE habitability. A summary of these design criteria follows:

PSDC 1, "Quality Standards," requires that those systems and components of reactor facilities which are essential to the prevention of accidents, or the mitigation of the consequences of nuclear accidents which could cause undue risk to the health and safety of the public, shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed.

PSDC 2, "Performance Standards," requires that those structures, systems and components of reactor facilities which are essential to the prevention, or to the mitigation of the consequences, of nuclear accidents which could cause undue risk to the health and safety of the public shall be designed, fabricated, and erected to performance standards that enable such structures, systems and components to withstand, without undue risk to the health and safety of the public, the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon such as earthquake, tornado, flooding condition, high wind or heavy ice.

PSDC 3, "Fire Protection," requires that the reactor facility shall be designed to ensure that the probability of events such as fires and explosions and the potential consequences of such events will not result in undue risk to the health and safety of the public. Non-combustible and fire resistant materials shall be used throughout the facility wherever necessary to preclude such risk, particularly in areas containing critical portions of the facility such as containment, control room, and components of engineered safety features.

PSDC 4, "Sharing of Systems," requires that reactor facilities not share systems or components unless it can be shown that such sharing will not result in undue risk to the health and safety of the public.

PSDC 5, "Records Requirements," requires the reactor licensee to be responsible for assuring the maintenance, throughout the life of the reactor, of records of the design,

- 5 fabrication, and construction of major components of the plant essential to avoid undue risk to the health and safety of the public.

PSDC 11, "Control Room," requires that the facility be provided with a control room from which actions to maintain safe operational status of the plant can be controlled.

Adequate radiation protection shall be provided to permit continuous occupancy of the control room under any credible post accident condition or as an alternative, access to other areas of the facility as necessary to shutdown and maintain safe control of the facility without excessive radiation exposures of personnel.

PSDC 40, "Missile Protection," requires that adequate protection for the engineered safety features, the failure of which would result in undue risk to the health and safety of the public, shall be provided against dynamic effects and missiles that might result from plant equipment failures.

GDC 19, "Control Room," requires that a control room be provided from which actions can be taken to operate the nuclear reactor safely under normal conditions and to maintain the reactor in a safe condition under accident conditions, including a loss-of coolant accident. Adequate radiation protection is to be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of specified values.

Prior to incorporation of TSTF-448, Revision 3, the STS requirements addressing CRE boundary operability resided only in the following CRE ventilation system specifications described in NUREG-1431, TS 3.7.10, "Control Room Emergency Filtration System (CREFS)." In these specifications, the SR associated with demonstrating the operability of the CRE boundary requires verifying that one CREEVS train can maintain a positive pressure relative to the areas adjacent to the CRE during the pressurization mode of operation at a makeup flow rate.

Facilities that pressurize the CRE during the emergency mode of operation of the CREEVS have similar SRs. Regardless, the results of ASTM E741 (Reference 2) tracer gas tests to measure CRE unfiltered inleakage at facilities indicated that the differential pressure surveillance is not a reliable method for demonstrating CRE boundary operability. Licensees were able to obtain differential pressure and flow measurements satisfying the SR limits even though unfiltered inleakage was determined to exceed the value assumed in the safety analyses.

In addition to an inadequate SR, the action requirements of these specifications were ambiguous regarding CRE boundary operability in the event CRE unfiltered inleakage is found to exceed the analysis assumption. The ambiguity stemmed from the view that the CRE boundary may be considered operable but degraded in this condition, and that it would be deemed inoperable only if calculated radiological exposure limits for CRE occupants exceeded a licensing basis limit; e.g., as stated in GDC-19, even while crediting compensatory measures.

NRC Administrative Letter (AL) 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety," states that "the discovery of an improper or inadequate TS value or required action is considered a degraded or nonconforming condition," which is defined in NRC Inspection Manual Chapter 9900; see latest guidance in RIS 2005-20 (Reference 3).

Imposing administrative controls in response to improper or inadequate TSs is considered an acceptable short-term corrective action. The NRC staff expects that, following the imposition of administrative controls, an amendment to the inadequate TS, with appropriate justification and schedule, will be submitted in a timely fashion."

- 6 Licensees that have found unfiltered inleakage in excess of the limit assumed in the safety analyses and have yet to either reduce the inleakage below the limit or establish a higher bounding limit through re-analysis, have implemented compensatory actions to ensure the safety of CRE occupants, pending final resolution of the condition, consistent with RIS 2005-20.

However, based on GL 2003-01 and AL 98-10, the NRC staff expects each licensee to propose TS changes that include a surveillance to periodically measure CRE unfiltered inleakage in order to satisfy 10 CFR 50.36(d)(3), which requires a facility's TS to include SRs, which it defines as "requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that limiting conditions for operation will be met." (Emphasis added.)

The NRC staff also expects facilities to propose unambiguous remedial actions, consistent with 10 CFR 50.36(d)(2), for the condition of not meeting the limiting condition for operation (LCO) due to an inoperable CRE boundary. The action requirements should specify a reasonable completion time to restore conformance to the LCO before requiring a facility to be shut down.

This completion time should be based on the benefits of implementing mitigating actions to ensure CRE occupant safety and sufficient time to resolve most problems anticipated with the CRE boundary, while minimizing the chance that operators in the CRE will need to use mitigating actions during accident conditions.

2.4 Adoption of TSTF-448 Revision 3 Adoption of TSTF-448, Revision 3, will assure that the facility's TS LCO for the CREVS is met by demonstrating unfiltered leakage into the CRE is within limits (Le., demonstrating operability of the CRE boundary). In support of this, TSTF-448 also adds TS administrative controls to assure the habitability of the CRE between performances of the ASTM E741 test. In addition, adoption of TSTF-448 will establish clearly stated and reasonable required actions in the event CRE unfiltered inleakage is found to exceed the analysis assumption.

The changes made by TSTF-448 to the STS requirements for the CREVS and the CRE boundary, conform to 10 CFR 50.36(d)(2) and 10 CFR 50.36(d)(3). Their adoption will better assure that a plant's CRE will remain habitable during normal operation and DBA conditions.

The staff has therefore concluded that these changes, as applied to CNP-1 and CNP-2, are acceptable from a regulatory standpoint.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed changes against the corresponding changes made to the STS by TSTF-448, Revision 3, which the NRC staff has found to satisfy applicable regulatory requirements, as described above in Section 2.0. The emergency operational mode of the CREV System at CNP-1 and CNP-2 pressurizes the CRE to minimize unfiltered air inleakage.

The proposed changes are consistent with this design.

3.1 Proposed Changes The proposed amendment would strengthen CRE habitability TS requirements by changing TS 3.7.10, "Control Room Emergency Ventilation (CREV) System," and adding a new TS administrative controls program on CRE habitability. Accompanying the proposed TS changes are appropriate conforming technical changes to the TS Bases. The proposed revision to the Bases also includes editorial and administrative changes to reflect applicable changes to the corresponding STS Bases, which were made to improve clarity, conform to the latest information

- 7 and references, correct factual errors, and achieve more consistency among the STS NUREGs.

Except for plant-specific differences, all of these changes are consistent with STS as revised by TSTF-448, Revision 3.

The NRC staff compared the proposed TS changes to the STS and the STS markups and evaluations in TSTF-448. The staff verified that differences from the STS were adequately justified on the basis of plant-specific design or retention of current licensing basis. The NRC staff also reviewed the proposed changes to the TS Bases for consistency with the STS Bases and the plant-specific design and licensing bases. The proposed Bases for TS 3.7.10 refer to specific guidance in NEI 99-03, "Control Room Habitability Assessment GUidance," Revision 0, dated June 2001 (Reference 6), which the NRC staff has formally endorsed, with exceptions, through RG 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors," dated May 2003 (Reference 4).

3.2 Editorial Changes The licensee proposed editorial changes to TS 3.7.10, "Control Room Emergency Ventilation (CREV) System," to establish standard terminology, such as "control room envelope (CRE)" in place of "control room," except for the plant-specific name for the CREV System, and "radiological, chemical, and smoke hazards" in place of various phrases to describe the hazards that CRE occupants are protected from by the CREV System. These changes improve the usability and quality of the presentation of the TS, have no adverse impact on safety, and therefore, are acceptable.

3.3 TS 3.7.10, "Control Room Emergency Ventilation (CRE\\/) System" The licensee proposed to revise the action requirements of TS 3.7.10, to acknowledge that an inoperable CRE boundary, depending upon the location of the associated degradation, could cause just one, instead of both, CREV trains to be inoperable. This is accomplished by revising Condition A to exclude Condition B, and revising Condition B to address one or more CREV trains, as follows:

Condition A One CREV train inoperable for reasons other than Condition B.

Condition B One or more CREV trains inoperable due to inoperable CRE boundary in MODE 1,2,3, or4.

This change clarifies how to apply the action requirements in the event just one CREV train is unable to ensure CRE occupant safety within licensing basis limits because of an inoperable CRE boundary. It enhances the usability of Conditions A and B with a presentation that is more consistent with the intent of the existing requirements. This change is an administrative change because it neither reduces nor increases the existing action requirements, and, therefore, is acceptable.

New Required Action B.1 requires the licensee to immediately implement mitigating actions.

New Required Action B.2 requires the licensee to verify, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , that in the event of a DBA, CRE occupant radiological exposures will not exceed the calculated dose of the licensing basis analyses of DBA consequences and CRE occupant exposure to hazardous chemicals and smoke will not exceed limits. New Required Action B.3 requires the licensee to restore CRE boundary to operable status within 90 days.

- 8 The 24-hour Completion Time of new Required Action B.2 is reasonable based on the low probability of a DBA occurring during this time period, and the use of mitigating actions as directed by Required Action B.1. The 90-day Completion Time of new Required Action B.3 is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. The 90-day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most anticipated problems with the CRE boundary. Therefore, proposed Actions B.1, B.2, and B.3 are acceptable.

The licensee proposed to add a new condition to TS 3.7.10, Action F, which states, "One or more CREV trains inoperable due to an inoperable CRE boundary during movement of irradiated fuel assemblies." The specified Required Action proposed for this condition is the same as for the existing condition of Action F, which states, "Two CREV trains inoperable during movement of irradiated fuel assemblies." Accordingly, the new condition is stated with the other condition in Action F using the logical connector "OR." The practical result of this presentation in format is the same as specifying two separately numbered Actions, one for each condition. Its advantage is to make the TS Actions table easier to use by avoiding having an additional numbered row in the Actions table. The new condition in Action F is needed because proposed Action B will only apply in Modes 1, 2, 3, and 4. As such, this change will ensure that the Actions table continues to specify a condition for an inoperable CRE boundary during movement of irradiated fuel assemblies. Therefore, this change is administrative and acceptable.

In the pressurization/cleanup mode of operation, the CREV System isolates unfiltered ventilation air supply intakes, filters the emergency ventilation air supply to the CRE, and pressurizes the CRE to minimize unfiltered air inleakage past the CRE boundary. The licensee proposed to delete the CRE pressurization SR. This SR requires verifying that one CREV System train, operating in the pressurization/cleanup mode, can maintain a pressure ~0.0625 inches water gauge, relative to the outside atmosphere during the pressurization/cleanup mode of operation at a makeup flow rate s 1000cfm. The deletion of this SR is proposed because measurements of unfiltered air leakage into the CRE at numerous reactor facilities demonstrated that a basic assumption of this SR, an essentially leak-tight CRE boundary, was incorrect for most facilities.

Hence, meeting this SR by achieving the required CRE pressure is not necessarily a conclusive indication of CRE boundary leak tightness, Le., CRE boundary operability. In its response to GL 2003-01, dated December 4, 2003, the licensee committed to determine the applicability of the NRC approved TSTF-448 to CNP and submit a license amendment request to require periodic measurement of unfiltered CRE inleakage. Based on the adoption of TSTF-448, Revision 3, the licensee's proposal to delete SR 3.7.10.4 is acceptable.

The proposed CRE inleakage measurement SR states, "Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Habitability Program." The CRE Habitability Program TS, proposed TS 5.5.16, requires that the program include "Requirements for determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the frequencies specified in Sections C.1 and C.2 of RG 1.197, Revision 0 (Reference 5). This guidance references ASTM E741 (Reference 2) as an acceptable method for ascertaining the unfiltered leakage into the CRE. The licensee has proposed to follow this method. Therefore, the proposed CRE inleakage measurement SR is acceptable.

- 9 3.4 TS 5.5.16, "Control Room Envelope Habitability Program" The proposed administrative controls program TS is consistent with the model program TS in TSTF-448, Revision 3. In combination with SR 3.7.10.4, this program is intended to ensure the operability of the CRE boundary, which as part of an operable CREV System will ensure that CRE habitability is maintained such that CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under DBA conditions without personnel receiving radiation exposure in excess of 5 Rem TEDE for the duration of the accident.

A CRE Habitability Program TS acceptable to the NRC staff requires the program to contain the following elements:

Definitions of CRE and CRE boundary.

This element is intended to ensure that these definitions accurately describe the plant areas that are within the CRE, and also the interfaces that form the CRE boundary, and are consistent with the general definitions discussed in Section 2.1 of this safety evaluation. Establishing what is meant by the CRE and the CRE boundary will preclude ambiguity in the implementation of the program.

Configuration control and preventive maintenance of the CRE boundary.

This element is intended to ensure the CRE boundary is maintained in its design condition. Guidance for implementing this element is contained in RG 1.196 (Reference 4), which endorsed, with exceptions, NEI 99-03, Rev. 0 (Reference 6).

Maintaining the CRE boundary in its design condition provides assurance that its leak tightness will not significantly degrade between CRE inleakage determinations.

Assessment of CRE habitability at the frequencies stated in Sections C.1 and C.2 of RG 1.197, Revision 0 (Reference 5), and measurement of unfiltered air leakage into the CRE in accordance with the testing methods and at the frequencies stated in Sections C.1 and C.2 of RG 1.197.

This element is intended to ensure that the plant assesses CRE habitability consistent with Sections C.1 and C.2 of RG 1.197 and NRC approved exceptions. Assessing CRE habitability at the NRC accepted frequencies provides assurance that significant degradation of the CRE boundary will not go undetected between CRE inleakage determinations. Determination of CRE inleakage using test methods acceptable to the NRC staff assures that test results are reliable for ascertaining CRE boundary operability.

Determination of CRE inleakage at the NRC accepted frequencies provides assurance that significant degradation of the CRE boundary will not occur between CRE inleakage determinations.

The licensee proposed the following exception to Sections C.1 and C.2 of RG 1.197, to be listed in the TS with this program element:

- 10 Exception: The appropriate application of ASTM E741-00 required by C.1.1 may include taking minor exceptions to the test methodology. These exceptions shall be documented in the test report.

The licensee provided justification for the exception in the July 28, 2008, supplemental letter.

The licensee stated that for the required testing methodology, ASTM E741-00 was not originally intended for nuclear power plant control room envelope testing, and that minor exceptions are necessary. The licensee also stated that exceptions to the test methodology will be documented in the individual test report to allow confirmation that the testing is in accordance with the requirements of proposed TS 5.5.16.c. The staff reviewed the justifications for the exception and found it to be adequate on the basis that the exception is necessary to meet the intent of ASTM E741 methodology when applied to nuclear power plant CRE testing,

Measurement of CRE pressure with respect to all areas adjacent to the CRE boundary at designated locations at a frequency of 24 months on a STAGGARED TEST BASIS.

Measurement results will be used in periodic assessments of the CRE boundary.

This element is intended to ensure that CRE differential pressure is regularly measured to identify changes in pressure warranting evaluation of the condition of the CRE boundary. Obtaining and trending pressure data provides additional assurance that significant degradation of the CRE boundary will not go undetected between CRE inleakage determinations.

Quantitative limits on unfiltered inleakage.

This element is intended to establish the CRE inleakage limit as the CRE unfiltered infiltration rate assumed in the CRE occupant radiological consequence analyses of DBAs. Unambiguous criterion for the CRE boundary to be considered operable in order to meet LCO 3.7.10 will ensure that associated action requirements will be consistently applied in the event of CRE degradation resulting in inleakage exceeding the limit.

Consistent with TSTF-448, Revision 3, the program states that the provisions of SR 3.0.2 are applicable to the program frequencies for performing the activities required by program paragraph number c, parts (i) and (ii) (assessment of CRE habitability and measurement of CRE inleakage), and paragraph number d (measurement of CRE differential pressure). This statement is needed to avoid confusion. SR 3.0.2 is applicable to the surveillance that references the testing in the CRE Habitability Program. However, SR 3.0.2 is not applicable to Administrative Controls unless specifically invoked. Providing this statement in the program eliminates any confusion regarding whether SR 3.0.2 is applicable, and is acceptable.

Consistent with TSTF-448, Revision 3, proposed TS 5.5.16 states that (1) a CRE Habitability Program shall be established and implemented, (2) the program shall include all of the NRC-staff required elements, as described above, and (3) the provisions of SR 3.0.2 shall apply to program frequencies. Therefore, TS 5.5.16, which is consistent with the model program TS approved by the NRC staff in TSTF-448, Revision 3, is acceptable.

- 11 3.5 Implementation of New Surveillance and Assessment Requirements by the Licensee The licensee has proposed license conditions regarding the initial performance of the new surveillance and assessment requirements. The new license conditions adopted the conditions in Section 2.3 of the model application published in the Federal Register on January 17, 2007 (72 FR 2022). Plant-specific changes were made to these proposed license conditions. The proposed plant-specific license conditions are consistent with the model application, and are acceptable.

3.6 Adoption of TSTF-448 Revision 3 by CNP-1 and CNP-2 The changes made by TSTF-448 to theSTS requirements for the CREV System and the CRE boundary conform to 10 CFR 50.36(d)(2) and 10 CFR 50.36(d)(3). The proposed plant-specific adoption of the changes also conform to regulatory requirements of 10 CFR 50.36(d)(2) and 10 CFR 50.36(d)(3) and will better assure that the CREs of CNP-1 and CNP-2 will remain habitable during normal operation and DBA conditions. The staff has therefore concluded that these changes are acceptable for adoption by CNP.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan state official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change 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 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 amendments involve no significant hazards considerations, and there has been no public comment on such finding (73 FR 5224). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) [and (c)(10)]. 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.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1.

NRC Generic Letter 2003-01, "Control Room Habitability," dated June 12, 2003.

2.

ASTM E 741 - 00, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution," 2000, (ASTM E741).

- 12

3.

NRC Regulatory Issue Summary 2005-20: Revision to Guidance Formerly Contained in NRC Generic Letter 91-18," Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability," dated September 26,2005.

4.

Regulatory Guide 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors," Revision 0, dated May 2003.

5.

Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003.

6.

NEI 99-03,"Control Room Habitability Assessment Guidance," Revision 0, dated June 2001.

7.

"D. C. Cook Nuclear Plant, Updated Final Safety Analysis Report, Revision 21," dated April 5, 2007 Principal Contributor: M. Hamm Date: December 30, 2008

December 30, 2008 Mr. Michael W. Rencheck 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 REGARDING INCORPORATION OF TSTF-448, REVISION 3, "CONTROL ROOM HABITABILITY" (TAC NOS. MD7554 AND MD7555)

Dear Mr. Rencheck:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 307 and 289 to Renewed Facility Operating License Nos. DPR-58 and DPR-74, for Units 1 and 2 of the Donald C. Cook Nuclear Plant. The amendments are in response to your application dated December 27,2007, as supplemented by letter dated July 28,2008, for implementation of the Technical Specification (TS) Task Force Traveler (TSTF)-448, Revision 3, "Control Room Habitability."

The amendment establishes more effective and appropriate action, surveillance, and administrative requirements related to ensuring the habitability of the control room envelope in accordance with the NRC-approved TSTF-448, Revision 3, and changes the TSs related to the control room emergency ventilation system in TS Section 3.7.10, "Control Room Emergency Ventilation (CREV) System," and TS Section 5.5.16, "Control Room Envelope Habitability Program." The amendment also adds a license condition to support implementation of the TS changes.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, IRAJ Terry A. Beltz, 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. 307 to DPR-58

2. Amendment No. 289 to DPR-74

3. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION:

PUBLIC RidsNrrAcrsAcnw_MailCTR Resource RidsNrrDirsltsb Resource J. Geissner, Rill RidsRgn3MailCenter Resource GHili (2)

LPL3-1 R/F RidsNrrDorlDpr Resource M. Hamm, NRR RidsNrrDorlLpl3-1 Resource RidsOgcRp Resource RidsNrrLATHarris Resource RidsNrrPMTBeltz Resource AmendmentAccessron N0.: ML083430469

SE provi'dedb>y memo 0 f 12/08/2008 OFFICE LPL3-1/PM LPL3-1/PM LPL3-1/LA SCVB ITSB/BC LPL3-1/BC NAME TBeltz PTam THarris JRaval REIliott

LJames DATE 12/15/08 12/19/08 12/17/08 12/17/08 12/08/08 12/30/08 OFFICIAL RECORD COpy

v t e Letter Sequence Approval
TAC:MD7554, Control Room Habitability (Approved, Closed) TAC:MD7555, Control Room Habitability (Approved, Closed)
Initiation Acceptance... Supplement Administration Questions Answers Results Approval Other: ML090160087
MONTHYEAR2008-07-28 [Table View]

ML083430469

Text

SUBJECT:

Dear Mr. Rencheck:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

SUBJECT:

Dear Mr. Rencheck:

Enclosures:

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