ML071420071

Technical Specification Change of Conditional Surveillance Frequency

ML071420071
Person / Time
Site:	Cook
Issue date:	05/11/2007
From:	Jensen J Indiana Michigan Power Co
To:	Document Control Desk, NRC/NRR/ADRO
References
AEP:NRC:6331-06, TAC MD0496, TAC MD0497
Download: ML071420071 (16)

Text

Indiana Michigan Power INDIANA Cook Nuclear Plant One Cook Place MICHIGAN Bridgman, Ml 49106 POWlER' AEP.com A unit of American Electric Power May 11, 2007 AEP:NRC:6331-06 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-PI-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 Technical Specification Change of Conditional Surveillance Frequency

Reference:

Letter from Peter S. Tam, Nuclear Regulatory Commission, to Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Units 1 and 2 (DCCNP-l and DCCNP-2) - Issuance of Amendments Regarding Reactor Trip System Instrumentation (TAC Nos. MD0496 and MD0497)," Accession Number ML062840162, dated October 30, 2006.

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant Units 1 and 2, proposes to amend Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to modify Technical Specifications (TS) to change a Surveillance Requirement (SR) frequency associated with a reactor trip on turbine trip from being met prior to the P-7 interlock (approximately 10 percent power) to being met prior to the P-8 interlock (approximately 31 percent power).

By the referenced letter, I&M received approval of a license amendment request to modify the reactor trip on turbine trip mode of applicability from Mode 1, above the P-7 interlock, to Mode 1, above the P-8 interlock. During implementation of the amendment I&M determined that an SR for the reactor trip on turbine trip function should also have been changed. Existing SR 3.3.1.18 requires that a Trip Actuating Device Operational Test be performed prior to exceeding the P-7 interlock whenever the unit has been in Mode 3, if not performed within the previous 31 days. The proposed change will replace the "P-7" with "P-8" in this SR to provide consistency with the mode of applicability for the reactor trip on turbine trip function.

Enclosure I provides an affirmation statement pertaining to this letter. Enclosure 2 provides I&M's evaluation of the proposed change. Attachments IA and lB provide TS pages marked to show changes for Unit 1 and Unit 2, respectively. Attachments 2A and 2B provide TS pages with the proposed changes incorporated.

U. S. Nuclear Regulatory Commission AEP:NRC:6331-06 Page 2 The license amendment approved by the referenced letter has been implemented in Unit 1. Power operation in Unit 1 with the existing SR is conservative since the frequency of the existing SR is overly restrictive and requires performance of the SR prior to entering the mode of applicability. A plant modification is planned for Unit 2 during the Fall 2007 outage to implement the amendment as required by the referenced letter. I&M requests approval of the proposed amendment in accordance with the normal Nuclear Regulatory Commission review schedule.

Copies of this letter and its attachments are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.

There are no commitments made in this letter. Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.

Sincerely, Jos Jensen Oice President KAS/rdw

Enclosures:

l. Affirmation

2. Indiana Michigan Power Company's Evaluation Attachments: IA. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked To Show Changes lB. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked To Show Changes 2A. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages With the Proposed Changes Incorporated 2B. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages With the Proposed Changes Incorporated c: J. L. Caldwell, NRC Region III K. D. Curry, Ft. Wayne AEP, w/o enclosures/attachments J. T. King, MPSC MDEQ - WHMD/RPMWS NRC Resident Inspector P. S. Tam, NRC Washington, DC

Enclosure 1 to AEP:NRC:6331-06 AFFIRMATION I, Joseph N. Jensen, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Lih N. Jensen Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS ___ DAY OF 2007 J N, a* Public My. Commission Expires

Enclosure 2 to AEP:NRC:6331-06 INDIANA MICHIGAN POWER COMPANY'S EVALUATION

Subject:

Technical Specification Change of Conditional Surveillance Frequency

3.0 TECHNICAL ANALYSIS

4.0 REGULATORY SAFETY ANALYSIS 4.1 No Significant Hazards Consideration 4.2 Applicable Regulatory Requirements / Criteria

5.0 ENVIRONMENTAL CONSIDERATION

S

6.0 REFERENCES

to AEP:NRC:6331-06 Page 2

1.0 DESCRIPTION

This letter is a request by Indiana Michigan Power Company (I&M) to amend Facility Operating Licenses DPR-58 and DPR-74 for the Donald C. Cook Nuclear Plant (CNP) Units 1 and 2. The proposed change modifies a Technical Specification (TS) Surveillance Requirement (SR)

Frequency for the reactor trip on turbine trip function. The existing SR requires that the SR be met prior to reaching the P-7 interlock (approximately 10 percent power). The proposed change to the SR will require that the SR be met prior to reaching the P-8 interlock (approximately 31 percent power). The proposed change will ensure consistency between the TS SR Frequency and the mode of applicability for the reactor trip on turbine trip function.

2.0 PROPOSED CHANGE

I&M proposes to change TS 3.3.1, Reactor Trip System Instrumentation, SR 3.3.1.18 Frequency from "Prior to exceeding the P-7 interlock..." to "Prior to exceeding the P-8 interlock..."

Changes to TS Bases 3.3.1 are required to reflect the change to the SR Frequency. These changes will be made in accordance with the CNP Technical Specification Bases Control Program.

3.0 TECHNICAL ANALYSIS

The proposed change to SR 3.3.1.18 aligns the SR Frequency with the mode of applicability specified in TS 3.3.1, Table 3.3.1-1, for Function 16, Turbine Trip. By the Reference 1 Safety Evaluation, Nuclear Regulatory Commission (NRC) approved a change to the mode of applicability for Function 16 from Mode 1, above the P-7 interlock, to Mode 1, above the P-8 interlock. The basis for NRC approval was a plant-specific analysis verifying that the pressurizer power-operated relief valves (PORV) would not be challenged by increasing the power level at which a turbine trip would occur without a reactor trip. This analysis was required by an NRC position addressed in NUREG-0737, Item ll.K.3.10, following the Three Mile Island accident which stated that any modifications to anticipatory trips, such as the reactor trip on turbine trip, should not be made until it has been shown that the probability of a small break loss-of-coolant accident resulting from a stuck-open PORV is substantially unaffected by the modification. In addition, I&M confirmed, by review of the Updated Final Safety Analysis Report (UFSAR) Chapter 14 safety analyses, that the safety analyses results are not adversely affected by the modification. Details of the evaluation verifying pressurizer PORVs are not challenged and the review of the UFSAR Chapter 14 safety analyses are provided in Reference 2.

to AEP:NRC:6331-06 Page 3 4.0 REGULATORY SAFETY ANALYSIS 4.1 No Significant Hazards Consideration Indiana Michigan Power Company (I&M) has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No The proposed change revises a Technical Specification (TS) Surveillance Requirement (SR)

Frequency associated with the reactor trip on turbine trip function to be consistent with the mode of applicability for the function. The change to the frequency from prior to exceeding the P-7 interlock to prior to exceeding the P-8 interlock does not create any new credible single failure. The P-7 and P-8 interlocks are not accident initiators. The reactor trip on turbine trip function is an anticipatory trip, and the safety analysis does not credit this trip for protecting the reactor core. The consequences of accidents previously evaluated are unaffected by this change because no change to any accident mitigation scenario has resulted and there are no additional challenges to fission product barrier integrity.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No No changes are being made to the plant that would introduce any new accident causal mechanisms. The proposed change to the interlock at which the surveillance is performed in support of a reactor trip on turbine trip does not adversely affect previously identified accident initiators and does not create any new accident initiators. The change does not affect how the associated trip function operates. No new single failures or accident scenarios are created by the proposed change and the proposed change does not result in any event previously deemed incredible being made credible.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

to AEP:NRC:6331-06 Page 4

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No No safety analyses were changed or modified as a result of the proposed change in the surveillance frequency. All margins associated with the current safety analyses acceptance criteria are unaffected. The current safety analyses remain bounding. The safety systems credited in the safety analyses will continue to be available to perform their mitigation functions. The proposed change does not affect the availability or operability of safety-related systems and components.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Based on the above, I&M concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36 requires that each license authorizing operation of a' production or utilization facility include TS. The TS are required to include SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that the facility operation will be within safety limits, and that the limiting conditions for operation will be met. This amendment revises an SR to be consistent with the mode of applicability.

In conclusion, based on the considerations discussed above, (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 Nuclear Regulatory Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health or safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or SR. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

to AEP:NRC:6331-06 Page 5

6.0 REFERENCES

1. Letter from Peter S. Tam, NRC, to I&M, "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)," Accession Number ML062840162, dated October 30, 2006.

2. Letter from Joseph N. Jensen, I&M, to NRC Document Control Desk, "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," AEP:NRC:6331, Accession Number ML060760532, dated March 7, 2006.

Attachment 1A to AEP:NRC:6331-06 DONALD C. COOK NUCLEAR PLANT UNIT I TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3.3.1-10

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.18 Viao fe iNOTEq Verification of setpoint is not required.

Perform TADOT. Prior to exceeding the P-4 interlock whenever the unit has been in MODE 3, if not performed within the previous 31 days SR 3.3.1.19 ----- NOTE . -----

Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS

.1.

Cook Nuclear Plant Unit I 3.3.1-10 Amendment No. 287

Attachment 1B to AEP:NRC:6331-06 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3.3.1-10

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.18 Viaof iNOTEq.

Verification of sepoint is not rquired.

Perform TADOT. Prior to exceeding the P-7-0 interlock whenever the unit has been in MODE 3, if not performed within the previous 31 days SR 3.3.1.19 --- ---

Neutron detectors NOTE--- from are excluded --... response time testing.

Verify RTS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS Cook Nuclear Plant Unit 2 3.3.1-10 Amendment No. 269

Attachment 2A to AEP:NRC:6331-06 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED 3.3.1-10

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.18 ViciostiiNOTEquird..

Verification of setpoint is not required.

Perform TADOT. Prior to exceeding the P-8 interlock I whenever the unit has been in MODE 3, if not performed within the previous 31 days SR 3.3.1.19 Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS

.1.

Cook Nuclear Plant Unit I 3.3.1-10 Amendment No. 287,

Attachment 2B to AEP:NRC:6331-06 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED 3.3.1-10

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.18 sV c nnNOTEquied Verification of setpoint is not required.

Perform TADOT. Prior to exceeding the P-8 interlock whenever the unit I

has been in MODE 3, if not performed within the previous 31 days SR 3.3-1.19 ----- NOTE Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS Cook Nuclear Plant Unit 2 3.3.1-10 Amendment No. 269,