ML073231227

License Amendment, Issuance of Amendment RTS and ESFAS Bypass Testing Capability

ML073231227
Person / Time
Site:	Cook
Issue date:	12/17/2007
From:	Tam P NRC/NRR/ADRO/DORL/LPLIII-1
To:	Rencheck M Indiana Michigan Power Co
tam P, NRR/ADRO/DORL, 415-1451
Shared Package
ML073510557	List: ML073231227 ML073520403
References
TAC MD3159, TAC MD3160
Download: ML073231227 (16)

Text

December 17, 2007 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 (DCCNP-1 AND -2)

ISSUANCE OF AMENDMENT - REGARDING REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (TAC NOS. MD3159 AND MD3160)

Dear Mr. Rencheck:

The Commission has issued the enclosed Amendment Nos. 300 and 283 to Renewed Facility Operating License Nos. DPR-58 (for DCCNP-1) and DPR-74 (for DCCNP-2). The amendments consist of changes to the technical specifications in response to your application dated September 15, 2006, as supplemented on July 25 and October 9, 2007.

The amendments revise the DCCNP-1 and DCCNP-2 Technical Specifications (TS) to allow certain functions in the reactor protection system and engineered safety feature actuation system instrumentation which have installed bypass test capability to be tested in bypass.

Specifically, the amendments revise the note for Limiting Condition for Operation D in Technical Specification Section 3.3.1, AReactor Trip System Instrumentation@ and Section 3.3.2, Engineered Safety Feature Actuation System Instrumentation. The administrative error in TS 3.3.1 Condition Q is not addressed in the amendments.

A copy of our related safety evaluation is enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

/RA/

Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316

Enclosures:

1. Amendment Nos. 300 and 283

2. Safety Evaluation cc w/encls: See next page

December 17, 2007 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 (DCCNP-1 AND -2)

ISSUANCE OF AMENDMENT - REGARDING REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (TAC NOS. MD3159 AND MD3160)

Dear Mr. Rencheck:

The Commission has issued the enclosed Amendment Nos. 300 and 283 to Renewed Facility Operating License Nos. DPR-58 (for DCCNP-1) and DPR-74 (for DCCNP-2). The amendments consist of changes to the technical specifications in response to your application dated September 15, 2006, as supplemented on July 25 and October 9, 2007.

The amendments revise the DCCNP-1 and DCCNP-2 Technical Specifications (TS) to allow certain functions in the reactor protection system and engineered safety feature actuation system instrumentation which have installed bypass test capability to be tested in bypass.

Specifically, the amendments revise the note for Limiting Condition for Operation D in Technical Specification Section 3.3.1, AReactor Trip System Instrumentation@ and Section 3.3.2, Engineered Safety Feature Actuation System Instrumentation. The administrative error in TS 3.3.1 Condition Q is not addressed in the amendments.

A copy of our related safety evaluation is enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

/RA/

Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316

Enclosures:

1. Amendment Nos. 300 and 283

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION PUBLIC GHill, OIS LPL3-1 R/F RidsOGCRp RidsNrrLATHarris RidsNrrPMPTam RidsDorlDpr RidsNrrDirslts RidsAcrsAcnwMailCenter RidsRgn3MailCenter HGarg RidsNrrDorlLpl3-1 Package Accession Number: ML073510557 Amendment Accession Number: ML073231227 TS Page Accession Number: ML073520403 OFFICE LPL3-1/PM LPL3-1/LA EICB/BC OGC/NLO LPL3-1/BC(A)

NAME PTam THarris WKemper* JBiggins CMunson DATE 12/5/07 11/30/07 11/20/07* 12/14/07 12/17/07

• Safety evaluation transmitted by memo of 11/20/07. *w/corrections OFFICIAL RECORD COPY

INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 300 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 15, 2006, as supplemented by letters on July 25 and October 9, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of 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. 300 are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 45 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Clifford G. Munson, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Operating License and Appendix A Date of Issuance: December 17, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 300 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 change area is identified by a marginal line.

REMOVE INSERT 3 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The change areas are identified by marginal lines.

REMOVE INSERT 3.3.1-2 3.3.1-2 3.3.1-3 3.3.1-3 3.3.1-4 3.3.1-4 3.3.2-2 3.3.2-2 3.3.2-3 3.3.2-3 3.3.2-4 3.3.2-4 3.3.2-5 3.3.2-5 3.3.2-6 3.3.2-6 3.3.2-7 3.3.2-7 3.3.2-8 3.3.2-8 3.3.2-9 3.3.2-9 3.3.2-10 3.3.2-10 3.3.2-11 3.3.2-11

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 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 Appendix B, as revised through Amendment No. 300, 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 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 299, 300

INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 283 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 15, 2006, as supplemented by letters on July 25 and October 9, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of 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. 283, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 45 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Clifford G. Munson, Acting Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Operating License and Appendix A Date of Issuance: December 17, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 283 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 change area is identified by a marginal line.

REMOVE INSERT 3 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The change areas are identified by marginal lines.

REMOVE INSERT 3.3.1-2 3.3.1-2 3.3.1-3 3.3.1-3 3.3.1-4 3.3.1-4 3.3.2-2 3.3.2-2 3.3.2-3 3.3.2-3 3.3.2-4 3.3.2-4 3.3.2-5 3.3.2-5 3.3.2-6 3.3.2-6 3.3.2-7 3.3.2-7 3.3.2-8 3.3.2-8 3.3.2-9 3.3.2-9 3.3.2-10 3.3.2-10 3.3.2-11 3.3.2-11

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 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 Appendix B, as revised through Amendment No. 283, 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 test each of the two valves in series in the Renewed License No. DPR-74 Amendment No. 1 through 282, 283

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 300 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AMENDMENT NO. 283 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 application to the U.S. Nuclear Regulatory Commission (NRC, Commission) dated September 15, 2006 (Accession No. ML062710309), as supplemented on July 25 (Accession No. ML072200260) and October 9, 2007 (Accession No. ML072900635), Indiana Michigan Power Company (I&M, or the licensee) requested an amendment to the Renewed Facility Operating License for Donald C. Cook Nuclear Plant, Units 1 and 2 (DCCNP-1 and DCCNP-2).

The proposed amendments would revise the DCCNP-1 and DCCNP-2 Technical Specifications (TS) to change Required Action Notes in TS 3.3.1, AReactor Trip System [RTS] Instrumentation,@

and TS 3.3.2, AEngineered Safety Features Actuation System [ESFAS] Instrumentation,@ to reflect installed bypass test capability. The changes to the Required Action Notes are consistent with wording in Standard Technical Specifications (NUREG-1431, Revision 3) for plants with installed bypass test capability.

The administrative error in TS 3.3.1 Condition Q is not addressed in the amendments.

The licensee=s supplements cited above provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff=s original proposed no significant hazards consideration determination as published in the Federal Register on November 21, 2006.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations, Part 50, Section 36 (10 CFR 50.36) requires that each license authorizing operation of a production or utilization facility to include TS. The TS are required to include surveillance requirements (SRs), which are requirements relating to testing, calibration, or inspection to assure that the necessary quality of systems and components are maintained, that the facility operation will be within safety limits, and that the limiting conditions for operation will be met. This requested DCCNP-1 and DCCNP-2 amendment changes the Required Action Notes to allow certain reactor trip system (RTS) and engineered safety feature actuation system (ESFAS) instrumentation SRs to be performed without entering a TS Action statement.

The NRCs requirement at 10 CFR 50.55a(h)(2) provides that each licensees protection system must meet the requirements stated in either Institute of Electrical and Electronics Engineer (IEEE) Standard 279, Criteria for Protection Systems for Nuclear Power Generating Stations, or IEEE Standard 603-1991, Criteria for Safety Systems for Nuclear Power Generating Stations, and the correction sheet dated January 30, 1995. For plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of IEEE Standard 603-1991, and the correction sheet dated January 30, 1995. For DCCNP-1 and DCCNP-2, the protection systems are designed in accordance with the requirements of the proposed IEEE Standard 279, dated August 1968. The proposed installed bypass test capability is designed to meet the intent of that standard.

3.0 TECHNICAL EVALUATION

3.1 The Licensees Proposed TS Changes The licensees proposed TS change would revise the reviewers note for Limiting Condition for Operation (LCO) in Section 3.3.1, Reactor Trip Instrumentation and Section 3.3.2, Engineered Safety Features Actuation System Instrumentation. The proposed amendment will allow the following functions in the RTS and ESFAS instrumentation, which have installed bypass capability to be tested in bypass:

TS Section 3.3.1 Function 6 Overtemperature Differential Temperature Function 7 Overpower Differential Temperature Function 8 Pressurizer Pressure Function 9 Pressurizer Water Level - High Function 10 Reactor Coolant Flow - Low Function 14 Steam Generator Water Level - Low Low Function 15 Steam Generator Water Level Coincident with Steam Flow/Feedwater Flow Mismatch TS Section 3.3.2 Function 1.c Safety Injection - Containment Pressure - High Function 1.d Safety Injection - Pressurizer Pressure - Low Function 1.e Safety Injection - Steam Line Pressure Function 4.d Steam Line Isolation - Steam Line Pressure - Low Function 4.e Steam Line Isolation - High Steam Flow in Two Steam Lines coincident with Tavg - Low Low Function 5.b Turbine Trip and Feedwater Isolation - Steam Generator Water Level -

High High Function 6.c Auxiliary Feedwater - Steam Generator Water Level - Low Low Function 7.c Containment Air Recirculation/Hydrogen Skimmer System -

Containment Pressure - High In its previous review of Westinghouse Topical Report WCAP - 10271, Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection System

Instrumentation, and its various revisions and supplements, the NRC staff established that bypass testing was an acceptable method for conducting periodic surveillance of the RTS and ESFAS as long as the channel did not fail to the bypass position, and did not require lifting of leads or installing jumpers. Placing a channel in bypass at DCCNP-1 or DCCNP-2 currently requires, in many cases, the use of jumpers or lifted leads. The licensee is modifying the circuitry to facilitate placing these channels in bypass without the use of jumpers or lifting leads. This modification will allow the licensee to test or maintain a channel without placing it in a tripped state. The December 20, 2002, acceptance letter for WCAP-15376-P, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Interval and Reactor Trip Breaker Test and Completion Times, noted that this topical report was built on the foundation established by WCAP-10271-P and WCAP-14333, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times. The NRC staff previously reviewed and approved the licensees application and verified that the applicable implementation requirements associated with the NRC staffs acceptance of WCAPs-10271-P, -14333-P and -15376-P (Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times) were also adequately addressed by the licensee. WCAP-14333-P has incorporated the Reviewers Note to the TS that allowed testing in bypass.

3.2 Evaluation of the Proposed Changes By the September 15, 2006, application, the licensee proposed to revise the Required Action Notes C.1, D.1, and D.2 for Section 3.3.1, and Required Action Notes D.1 and D.2 for Section 3.3.2. This revision would allow the surveillance testing be done in bypass and included nuclear instrumentation system (NIS) functions also. By letter dated June 18, 2007 (Accession No. ML071650508), the NRC staff requested the licensee to provide the design basis for the change instead of simply using the provisions of 10 CFR 50.59. By letter dated July 25, 2007, the licensee provided the requested information. However, the licensee did not justify the basis for the NIS functions. In a conference call, the NRC staff informed the licensee that since the licensee has not even purchased the bypass circuitry for NIS functions, and no design has been developed, it can not approve the TS changes for the NIS functions. Accordingly, by letter dated October 9, 2007, the licensee removed the NIS functions from the amendment application.

The RTS and ESFAS use 2-out-of-3 and 2-out-of-4 coincident logic, respectively, from redundant channels to initiate protective actions. Removal of one actuation channel for testing is accomplished by either placing that channel in a tripped state, where a 2-out-of-3 logic becomes a 1-out-of-2 logic or using bypass capability, where a 2-out-of-4 logic becomes a 2-out-of-3 logic. Within these systems, analog and digital channel comparators are placed in the tripped state in response to an inoperable channel. With an inoperable channel in the tripped state, maintenance or testing cannot be performed on a redundant channel unless one of the channels is bypassed. In addition, with a channel in the tripped condition, a second comparator trip in a redundant channel caused by human error, a spurious transient, or channel failure would initiate a reactor trip or ESFAS actuation. With bypass test capability, a channel may be bypassed for surveillance testing with an inoperable channel in the tripped state.

With the implementation of testing in bypass, additional time will be available to perform surveillance testing while preventing a spurious reactor trip or ESFAS actuation since the partial trip conditions that would have been present are eliminated. The logic requiring signals from two additional channels to actuate the protective function is maintained. This provides the benefits of reducing challenges to the plant safety systems that may result from spurious actuations and

potentially increasing plant availability. Administrative controls are provided to prevent the simultaneous bypassing of more that one redundant protection set at any one time, and to restore the system to normal operation.

In order to allow for testing in bypass, the licensee has proposed to revise Required Action Notes 1 and 2 for LCO Condition D for Sections 3.3.1 and 3.3.2. The testing in bypass can only be enabled by a key lock switch and the channel switch in test position. Therefore, it will require two independent manual switch actuations to place a channel in bypass. There is no automatic actuation associated with this bypass circuit. The existing design has been modified such that, when a test switch is placed in test, one channel function output to the solid state protection system (SSPS) trains would be connected to the bypass switch, and when the locking bypass switch is placed in bypass (by a key controlled by the Operation Department) an energized signal is applied to the input of the SSPS trains, thereby continuously providing power to the channel function and maintaining the channel in a non-tripped condition. The locking bypass switch is connected to a status light that provides indication both locally and in the control room when a channel bypass switch is armed. The licensee plans to use test switches capable of make-before-break, which will allow for seamlessly switching a channel bypass without incurring a trip signal.

In response to the NRC staff=s request for additional information, the licensee stated that its design modification process ensures that quality components are used, single-failure considerations are accounted for, and diversity and testability of components are provided.

Administrative controls are in place to ensure that failed channels are placed in the tripped condition following completion of surveillance testing. It would take a simultaneous failure of two switches to inadvertently cause the bypass of just one channel. Therefore, the single-failure criterion is maintained for the protection channel operation. The licensee stated that the components used for this modification to install bypass test capability are qualified commensurate with their use as nuclear grade equipment. Both the protection system bypass switch and the make-before-break switches were verified to be seismically qualified by the performance of seismic qualification tests as documented in the manufacturers seismic qualification test reports. The licensee has verified that bypass switches for each channel are completely independent from the other channels and redundancy and diversity of safety functions remain unchanged. Based on the above, the NRC staff has concluded that the licensee has adequately assured the design modification meets the plant licensing basis and surveillance testing of the function identified previously can be safely conducted.

3.3 Summary of Technical Evaluation Based on the above review, the staff concludes that the DCCNP-1 and DCCNP-2 bypass testing design meets the applicable NRC requirements in 10 CFR 50.55a(h)(2), and that sufficient controls are provided to preclude improper bypass of a channel. Also, the configuration of bypass testing is consistent with the previously NRC-accepted topical reports and the NRC staffs implementation requirements associated with the acceptance of these topical reports.

The NRC staff, therefore, finds the proposed changes to the RTS and ESFAS TS to permit bypass testing to be acceptable.

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 installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 consideration, and there has been no public comment on such finding (71 FR 67396). 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.

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.

Principal Contributor: H. Garg Date: December 17, 2007

Donald C. Cook Nuclear Plant, Units 1 and 2 cc:

Attorney General Michigan Department of Environmental Department of Attorney General Quality 525 West Ottawa Street Waste and Hazardous Materials Div.

Lansing, MI 48913 Hazardous Waste & Radiological Protection Section Township Supervisor Nuclear Facilities Unit Lake Township Hall Constitution Hall, Lower-Level North P.O. Box 818 525 West Allegan Street Bridgman, MI 49106 P. O. Box 30241 Lansing, MI 48909-7741 U.S. Nuclear Regulatory Commission Resident Inspector's Office Lawrence J. Weber, Plant Manager 7700 Red Arrow Highway Indiana Michigan Power Company Stevensville, MI 49127 Nuclear Generation Group One Cook Place Kimberly Harshaw, Esquire Bridgman, MI 49106 Indiana Michigan Power Company One Cook Place Joseph Jensen, Site Vice President Bridgman, MI 49106 Indiana Michigan Power Company Nuclear Generation Group Mayor, City of Bridgman One Cook Place P.O. Box 366 Bridgman, MI 49106 Bridgman, MI 49106 Special Assistant to the Governor Room 1 - State Capitol Lansing, MI 48909 Susan D. Simpson Regulatory Affairs Manager Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106