ML062840162

License Amendments 297 and 278 Regarding Reactor Trip System Instrumentation

ML062840162
Person / Time
Site:	Cook
Issue date:	10/30/2006
From:	Tam P NRC/NRR/ADRO/DORL/LPLIII-1
To:	Nazar M Indiana Michigan Power Co
D. Jaffe
References
TAC MD0496, TAC MD0497
Download: ML062840162 (17)

Text

October 30, 2006 Mr. Mano K. Nazar 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 DCCNP-2) - ISSUANCE OF AMENDMENTS REGARDING REACTOR TRIP SYSTEM INSTRUMENTATION (TAC NOS. MD0496 AND MD0497)

Dear Mr. Nazar:

The Commission has issued the enclosed Amendment No. 297 to Renewed Facility Operating License No. DPR-58 for DCCNP-1 and Amendment No. 278 to Renewed Facility Operating License No. DPR-74 for DCCNP-2. The amendment consists of changes to the Technical Specifications in response to your application dated March 7, 2006, as supplemented by letter dated August 3, 2006.

The amendment revises Section 3.3.1, Reactor Trip System (RTS) Instrumentation, of the DCCNP-1 and DCCNP-2 Technical Specifications, changing the reactor trip on turbine trip interlock from the P-7 setpoint (10 percent rated thermal power) to the P-8 setpoint (31 percent rated thermal power).

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 by D. Jaffe/

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 No. 297 to DPR-58

2. Amendment No. 278 to DPR-74

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

Mr. Mano K. Nazar October 30, 2006 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 DCCNP-2) - ISSUANCE OF AMENDMENTS REGARDING REACTOR TRIP SYSTEM INSTRUMENTATION (TAC NOS. MD0496 AND MD0497)

Dear Mr. Nazar:

The Commission has issued the enclosed Amendment No. 297 to Renewed Facility Operating License No. DPR-58 for DCCNP-1 and Amendment No. 278 to Renewed Facility Operating License No. DPR-74 for DCCNP-2. The amendment consists of changes to the Technical Specifications in response to your application dated March 7, 2006, as supplemented by letter dated August 3, 2006.

The amendment revises Section 3.3.1, Reactor Trip System (RTS) Instrumentation, of the DCCNP-1 and DCCNP-2 Technical Specifications, changing the reactor trip on turbine trip interlock from the P-7 setpoint (10 percent rated thermal power) to the P-8 setpoint (31 percent rated thermal power).

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 by D. Jaffe/

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 No. 297 to DPR-58

2. Amendment No. 278 to DPR-74

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION PUBLIC GHill, OIS LPL3-1 R/F RidsOGCRp RidsNrrLATHarris RidsNrrPMPTam RidsNrrDrlLple RidsDorlDpr RidsNrrDirsltsb RidsAcrsAcnwMailCenter SMiranda RidsRgn3MailCenter Package Accession Number: ML063030058 Amendment Accession Number: ML062840162 TS Page Accession Number: ML063050086 OFFICE NRR:LPL3-1/PM NRR:LPL3-1/LA NRR:SPWB/BC OGC NRR:LPL3-1/BC NAME Jaffe for PTam THarris JNakoski* TCampbell** LRaghavan DATE 10/ 11/06 10/18/06 10/ 11/06 10/27/06 10/30/06

• Safety evaluation transmitted by memo of 10/11/06. **NLO subject to edits.

OFFICIAL RECORD COPY

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

Regional Administrator, Region III Michigan Department of Environmental U.S. Nuclear Regulatory Commission Quality Suite 210 Waste and Hazardous Materials Div.

2443 Warrenville Road Hazardous Waste & Radiological Lisle, IL 60532-4351 Protection Section Nuclear Facilities Unit Attorney General Constitution Hall, Lower-Level North Department of Attorney General 525 West Allegan Street 525 West Ottawa Street P. O. Box 30241 Lansing, MI 48913 Lansing, MI 48909-7741 Township Supervisor Lawrence J. Weber, Plant Manager Lake Township Hall Indiana Michigan Power Company P.O. Box 818 Nuclear Generation Group Bridgman, MI 49106 One Cook Place Bridgman, MI 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office Mark A. Peifer, Site Vice President 7700 Red Arrow Highway Indiana Michigan Power Company Stevensville, MI 49127 Nuclear Generation Group One Cook Place James M. Petro, Jr., Esquire Bridgman, MI 49106 Indiana Michigan Power Company One Cook Place Bridgman, MI 49106 Mayor, City of Bridgman P.O. Box 366 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

INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 297 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 March 7, 2006, as supplemented by letter dated August 3, 2006, 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. 297, 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 prior to entry into Mode 1 following the Cycle 21 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

L. Raghavan, Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Operating License Date of Issuance: October 30, 2006

ATTACHMENT TO LICENSE AMENDMENT NO. 297 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 3.3.1-4 3.3.1-4 3.3.1-5 3.3.1-5 3.3.1-6 3.3.1-6 3.3.1-13 3.3.1-13

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. 297, are hereby incorporated in the renewed l 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 296, 297

INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 278 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 March 7, 2006, as supplemented by letter dated August 3, 2006, 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. 278, 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 prior to entry into Mode 1 following the Cycle 17 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

L. Raghavan, Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 30, 2006

ATTACHMENT TO LICENSE AMENDMENT NO. 278 RENEWED FACILITY OPERATING LICENSE NO. DPR-58 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 3.3.1-4 3.3.1-4 3.3.1-5 3.3.1-5 3.3.1-6 3.3.1-6 3.3.1-13 3.3.1-13

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. 278, are hereby incorporated in the renewed l 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 277, 278

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 297 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AMENDMENT NO. 278 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 (DCCNP-1 AND DCCNP-2)

DOCKET NOS. 50-315 AND 50-316

1.0 INTRODUCTION

By application to the U.S. Nuclear Regulatory Commission (NRC) dated March 7, 2006 (Accession No. ML060760532), as supplemented by letter dated August 3, 2006 (Accession No. ML062200241), Indiana Michigan Power Company (I&M, the licensee) proposed changes to the Technical Specifications (TSs) for DCCNP-1 and DCCNP-2. The changes would increase the power level required for a reactor trip following a turbine trip signal. The current TS requires a reactor trip when a turbine trip signal occurs and the reactor power is equal to or greater than 10 percent of the rated thermal power (RTP). It also requires that the automatic reactor trip on a turbine trip signal be blocked when the reactor power decreases below 10 percent RTP. The licensee proposed to change the requirement so that the anticipatory reactor trip on turbine trip shall be unblocked at 31 percent RTP (i.e., at the P-8 setpoint). The change would allow the licensee to implement an interlock system at the units to block reactor trips on turbine trips for reactor power levels below 31 percent RTP. Since many turbine trips occur at low power levels, the proposed TS would decrease the frequency of unnecessary challenges to the reactor protection system and thereby increase plant availability.

The licensees August 3, 2006, letter, contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

2.0 REGULATORY EVALUATION

The licensees proposal to interlock this automatic reactor trip on turbine trip function with P-8 (about 31 percent RTP) has been implemented at several other nuclear power plants. Four examples are listed in the table below:

Plant Name P-7 P-8 Byron/Braidwood Units 1 and 2 (1987) 10% RTP 30% RTP Salem (1988) 11% RTP 36% RTP North Anna Units 1 and 2 (1989) 8% RTP 30% RTP Indian Point Unit 3 (1999) 10% RTP 50% RTP The reactor trip, when it is demanded by a turbine trip, is not assumed to occur in accident analyses, since the turbine trip signal originates in a non-seismically qualified area (the turbine building). However, the load rejection or turbine trip event is evaluated in order to address Three Mile Island (TMI) Action Item II.K.3.10 in NUREG-0737 (Reference 7), which states that:

The anticipatory trip modification proposed by some licensees to confine the range of use to high-power levels should not be made until it has been shown on a plant-by-plant basis that the probability of a small-break loss-of-coolant accident (LOCA) resulting from a stuck-open power-operated relief valve (PORV) is substantially unaffected by the modification.

3.0 TECHNICAL EVALUATION

3.1 Evaluation of the Licensees Analyses DCCNP-1 and DCCNP-2 are four-loop, Westinghouse-designed pressurized-water reactors that were licensed to operate on October 25, 1974, and December 23, 1977, respectively. The proposed change would raise the threshold power level above which the reactor will automatically be tripped in the event of a turbine trip. The licensee proposed to raise this threshold from about 10 percent of RTP to about 31 percent of of RTP. The licensee believes this change would result in a reduction in the frequency of unnecessary reactor trips when operating in the 10 percent to 31 percent RTP range.

To address TMI Action Item II.K.3.10, the licensee analyzed the turbine trip without a reactor trip transient, initiated from 31-percent power, using Westinghouses NRC-approved LOFTRAN code (Reference 8). The NRC previously approved LOFTRAN to allow Westinghouse to analyze system responses to non-LOCA events for conventional Westinghouse PWRs.

LOFTRAN simulates a multiloop system using a model containing a reactor vessel, hot- and cold-leg piping, steam generators, and a pressurizer. The use of LOFTRAN is conservative because comparison studies have shown that LOFTRAN overpredicts pressurizer pressure during pressure-increase transients. This turbine trip scenario is the event that is most affected by the proposed modification.

The licensees analyses indicate that the reactor coolant system (RCS) pressure increase, caused by the load rejection or turbine trip, would not be high enough to cause any of the PORVs to open. Therefore, the probability of a small-break LOCA resulting from a stuck-open PORV is not affected by the proposed modification.

The licensee also verified that the proposed modification has no effect upon the following events:

1. Uncontrolled rod cluster control assembly (RCCA) bank withdrawal from a subcritical condition

2. Uncontrolled RCCA bank withdrawal at power

3. RCCA misalignment

4. Chemical and volume control system (CVCS) malfunction

5. Loss of reactor coolant flow

6. Startup of inactive reactor coolant loop

7. Loss of external electrical load (from full power)

8. Loss of normal feedwater flow

9. Excessive heat removal due to feedwater system malfunctions

10. Excessive load increase incident

11. Loss of all alternating current power to the plant auxiliaries

12. Steam generator tube rupture

13. Rupture of a steam pipe

14. Rupture of a control rod mechanism housing (RCCA ejection)

15. Major rupture of main feedwater pipe (Unit 2 only)

16. LOCA and LOCA-related analyses

17. Containment integrity evaluation (short-term and long-term)

18. Main steamline break mass and energy release analyses The licensee did not evaluate the effect of the proposed modification upon the following events because these events are not in the licensing basis for DCCNP-2 (References 9 and 10)1:

a. Reactor coolant pump shaft break

b. Single RCCA withdrawal

c. Inadvertent loading and operation of a fuel assembly in an improper position

d. Inadvertent actuation of the emergency core cooling system that increases RCS inventory

e. Inadvertent actuation of the CVCS that increases RCS inventory

f. Inadvertent opening of a pressurizer PORV

g. Radiological consequences of failure of small lines carrying primary coolant outside containment These events, like the events that have been considered above, do not rely upon the reactor trip on turbine trip for protection. Therefore, they are unaffected by the proposed modification.

Although the two mass addition events (d and e above) could lead to the opening of one or more PORVs with the pressurizer in a water-solid condition, they would not be affected by the proposed change in the reactor trip on turbine trip interlock setting.

Based upon the NRC staffs previous evaluation of the plants referenced in Section 2.0 herein, which are similar in design to the D. C. Cook units, the NRC staff concluded that the above referenced transients are not affected by the proposed increase in power level required for a reactor trip following a turbine signal for DCCNP Units 1 and 2.

1 Because the above referenced transients were not part of the DCCNP-2 design-basis, they were also not considered for DCCNP-1.

3.2 Evaluation of Proposed TS Changes The licensee proposes to modify Table 3.3.1-1 of TS 3.3.1, Reactor Trip System Instrumentation, for both DCCNP-1 and DCCNP-2, changing turbine trip from low fluid oil pressure (Function 16.a) or from closure of the turbine stop valve (Function 16.b), to be applicable at Mode 1 when power levels are above the P-8 setting, rather than the current P-7 setting. This change is acceptable, since it is consistent with the evaluation in Section 3.1 above.

The licensee proposes to modify Condition N of TS 3.3.1, deleting references to Functions 16.a and 16.b from the list of functions which require entry into Condition N (i.e., reduce power to less than P-7) if the required action and associated completion time of Condition D is not met.

As stated in the above paragraph, Functions 16.a and 16.b are being modified to refer to P-8 instead of P-7. This change is acceptable, since it is consistent with the evaluation in Section 3.1 above.

The licensee proposes to add a new Condition O, which specifies that if [r]equired Action and associated Completion Time of condition D not met for Function 16.a and 16.b, reduce thermal power to less than P-8. This change is acceptable, since it is consistent with the evaluation in Section 3.1 above.

The licensee proposed to re-number existing Conditions O, P, and Q to P, Q, and R due to introduction of the new Condition O above. This change is purely editorial and acceptable.

3.3 Summary of Technical Review The licensee has shown, to the satisfaction of the NRC staff, that shifting the reactor trip on turbine trip interlock from P-7 to P-8 will not impact any of the accident analyses in the DCCNP-1 and DCCNP-2 licensing bases. Furthermore, the change will not impact any of the seven events/analyses (i.e., a - g above) that are identified as being absent from the units licensing bases. The licensee has also evaluated the effect of the change upon TMI action Item II.K.3.10, regarding a stuck open PORV, and found acceptable results. Therefore, the licensees proposal to change the reactor trip on turbine trip interlock from P-7 to P-8 is acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in 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 NRC has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (71 FR 23956). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant

to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

(1) Indiana Michigan Power Company, letter AEP:NRC 6331, "Donald C. Cook Nuclear Plant, Units 1 and 2, Docket Nos. 50-315 and 50-316, Response to Request for Additional Information and Supplement Regarding Technical Specification Change of Interlock for a Reactor Trip on Turbine Trip, dated August 3, 2006 (Accession No. ML062200241).

(2) Indiana Michigan Power Company, letter AEP:NRC 6331-03, "Donald C. Cook Nuclear Plant, Units 1 and 2, Docket Nos. 50-315 and 50-316, Technical Specification Change of Interlock for a Reactor Trip on Turbine Trip, dated March 7, 2006 (Accession No. ML060760532).

(3) Amendment Nos. 13 and 3, Byron and Braidwood Units 1 and 2 - Issuance of Amendments to Allow Deletion of Reactor Trip on Turbine Trip Below 30 percent of Rated Thermal Power, dated December 8, 1987 (Accession No. ML020850675).

(4) Amendment Nos. 85 and 58, Technical Specification Changes - Trip Reduction P-7 Permissive to P-9 Permissive (TAC Nos. 66039 and 66040), Re: Salem Nuclear Generating Station, Unit Nos. 1 and 2, dated June 27, 1988 (Accession No. ML011690022).

(5) Amendment Nos. 119 and 103, North Anna Units 1 and 2 - Issuance of Amendments Re: Direct Reactor Trip on Turbine Trip Blocked Below 30 percent of Rated Thermal Power (TAC Nos. 69800 and 69801), dated July 18, 1989 (Accession No. ML013460457).

(6) Amendment No. 192, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Reactor Trip on Turbine Trip (TAC No. MA4696), dated September 8, 1999 (Accession No. ML003780834).

(7) Clarification of TMI Action Plan Requirements, NUREG-0737, dated November, 1980.

(8) LOFTRAN Code Description, WCAP-7907-P-A, April, 1984.

(9) Letter from Joseph G. Giitter, NRC, to Milton P. Alexich, Indiana Michigan Power Company, Analysis of D.C. Cook Unit 2, Cycle 8 Reload, dated August 3, 1989 (Accession No. 8908140265).

(10) Donald C. Cook Nuclear Plant, Unit 1, Update Final Safety Analysis Report, Rev 19.

(11) NUREG-0800, Standard Review Plan, Revision 2, July, 1981.

(12) Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, RG 1.70, Rev 3, November, 1978.

(13) ANS 51.1, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants (replaces ANSI N18.2, 1973), 1983.

Principal Contributor: S. Miranda Date: October 30, 2006