Text

Units I and 2, Conditional Exemption from Measurement of End of Life Moderator Temperature Coefficient
	ML050660130
Person / Time
Site:	Cook
Issue date:	02/25/2005
From:	Jensen J; Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP:NRC:5132
	Download: ML050660130 (45)
	v • d • e

Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INDIANA MICHIGAN POWER February 25, 2005 AEP:NRC:5132 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant Units I and 2 Docket Nos. 50-315 and 50-316 Conditional Exemption from Measurement of End of Life Moderator Temperature Coefficient

Reference:

1. Robert C. Jones (Nuclear Regulatory Commission (NRC))

letter to Nicholas J. Liparulo (Westinghouse), "Acceptance for Referencing of Licensing Topical Report WCAP- 13749-P,

'Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL [End-ofLife] Moderator Temperature Coefficient Measurement,"' October 9, 1996.

2. N. J. Liparulo (Westinghouse) letter to J. E. Lyons (NRC),

"Clarification of Individual Control Rod Bank Worth Benchmark Criteria in WCAP-13749, 'Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement,"'

March 18, 1997.

3. 'Donald C. Cook Nuclear Plant Units 1 and 2, License Amendment Request - Conversion of Current Technical Specifications (CTS) to Improved Technical Specifications (ITS)," letter AEP:NRC:4901, dated April 6, 2004.

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposes to amend Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to modify Technical Specifications (TS) to revise the near-end of life Moderator Temperature Coefficient (MTC) Surveillance Requirement by placing a set of bDO\

U. S. Nuclear Regulatory Comrnmission AEP:NRC:5132 Page 2 conditions on core performance, which if met, would allow conditional exemption from the required MTC measurement. The conditional exemption will be determined on a cycle-specific basis by considering the margin predicted to the surveillance requirement MTC limit and the performance of other core parameters, such as beginning of life MTC measurements and the critical boron concentration as a function of cycle average burnup. The conditional exemption would minimize disruptions to normal plant operations. Plant safety criteria will not be compromised by the conditional exemption of this one measurement. No changes to the TS Bases will be required as a result of the proposed amendment.

This method has been accepted by the Nuclear Regulatory Commission (NRC)

(Reference 1) and clarified by Westinghouse through amendment to WCAP-13749 (Reference 2). The NRC has previously approved similar TS amendments at South Texas Project and Virgil C. Summer Plants (Accession Numbers ML023400252 and ML042040460) for other utilities on November 26, 2002 and July 21, 2004, respectively.

By Reference 3, I&M proposed conversion of the CNP Current Technical Specifications (CTS) to the Improved Technical Specifications (ITS) specified in NUREG-1431, "Standard Technical Specifications - Westinghouse Plants,"

Revision 2. Implementation of the ITS is expected to occur prior to implementation of the changes proposed in this amendment request. I&M has therefore provided copies of both the CTS and the current draft ITS pages that are affected by this proposed amendment. I&M will coordinate with the NRC Project Manager to ensure that the appropriate pages are issued. provides an affirmation statement pertaining to this letter. provides I&M's evaluation of the proposed change.

Attachments IA and lB provide CTS pages marked to show changes for Unit 1 and Unit 2, respectively. Attachments 2A and 2B provide CTS pages with the proposed changes incorporated. Attachments 3A and 3B provide ITS pages marked to show changes for Unit 1 and Unit 2, respectively. Attachments 4A and 4B provide ITS pages with the proposed changes incorporated. Attachment 5 provides typical revised pages from a Core Operating Limits Report for information only.

I&M requests this amendment be granted no later than December 1, 2005. I&M requests a 15 day implementation period following approval.

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.

U. S. Nuclear Regulatory Commission AEP:NRC:5132 Page 3 There are no new commitments in this letter. Should you have any questions, please contact Mr. John A. Zwolinski, Safety Assurance Director at (269) 466-2428.

Sincerely, Joseph N. Jensen Site Vice President KS/rdw

U. S. Nuclear Regulatory Commission AEP:NRC:5 132 Page 4

Enclosures:

1. Affirmation

2. Licensee's Evaluation Attachments:

IA. Donald C. Cook Nuclear Plant Unit 1 Current Technical Specification Pages Marked To Show Changes 1B. Donald C. Cook Nuclear Plant Unit 2 Current Technical Specification Pages Marked To Show Changes 2A. Donald C. Cook Nuclear Plant Unit 1 Current Teclmical Specification Pages With the Proposed Changes Incorporated 2B. Donald C. Cook Nuclear Plant Unit 2 Current Technical Specification Pages With the Proposed Changes Incorporated 3A. Donald C. Cook Nuclear Plant Unit 1 Improved Technical Specification Pages Marked To Show Changes 3B. Donald C. Cook Nuclear Plant Unit 2 Improved Technical Specification Pages Marked To Show Changes 4A. Donald C. Cook Nuclear Plant Unit 1 Improved Technical Specification Pages With the Proposed Changes Incorporated 4B. Donald C. Cook Nuclear Plant Unit 2 Improved Technical Specification Pages With the Proposed Changes Incorporated

5. Typical Revised Core Operating Limits Report Pages (For Information Only) c: J. L. Caldwell, NRC Region III K. D. Curry, Ft. Wayne AEP, w/o enclosures/attachments J. T. King, MPSC C. F. Lyon, NRC Washington, DC MDEQ - WHMD/HWRPS NRC Resident Inspector

Enclosure 1 to AEP:NRC:5132 AFFIRMATION 1, 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 Joseph N. Jensen Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIScP 6 _DAY OF , 2005 XU \Ndlary public My Commission Expires CAo1{ c7

__ t _

Enclosure 2 to AEP:NRC:5132 LICENSEE'S EVALUATION

Subject:

Conditional Exemption from Measurement of End of Life Moderator Temperature Coefficient

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements / Criteria

6.0 ENVIRONMENTAL CONSIDERATION

S

7.0 REFERENCES

8.0 PRECEDENT

r*: to AEP:NRC:5132 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 changes would revise the near-end of life (EOL) Moderator Temperature Coefficient (MTC) Surveillance Requirement (SR) by placing a set of conditions on' core performance, which if met, would allow conditional exemption from the required MTC measurement. The conditional exemption will be determined on a cycle-specific basis by considering the margin predicted to the surveillance requirement MTC limit and the performance of other core parameters, such as beginning of life (BOL) MTC measurements and the critical boron concentration as a function of cycle average burnup. The conditional exemption would minimize disruptions to normal plant operations. Plant safety criteria will not be compromised by the conditional exemption of this one measurement. No changes to the Technical Specification (TS)

Bases will be required as a result of the proposed amendment.

I&M requests approval of the proposed amendment by December 1, 2005. I&M requests that the amendment allow a 15 day implementation period.

2.0 PROPOSED CHANGE

By separate correspondence, I&M has previously proposed conversion of the CNP Current Technical Specifications (CTS) to the Improved Technical Specifications (ITS) specified in NUREG-1431. Implementation of the ITS is expected prior to implementation of the changes proposed in this amendment request. I&M has therefore provided copies of both the CTS and the current draft ITS pages that are affected by this proposed amendment.

CTS Changes SR 4.1.1.4.b is modified to suspend the MTC measurement if the model benchmark criteria and Revised Prediction specified in the Core Operating Limits Report (COLR) are satisfied.

WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," is added to the list of references for the COLR in TS 6.9.1.9.2.

ITS Changes SR 3.1.3.2 is modified to suspend the MTC measurement if the model benchmark criteria and Revised Prediction specified in the COLR are satisfied.

WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," is added to the list of references for the COLR in TS 5.6.5.

'I,- .. to AEP:NRC:5132 Page 3 In summary, the proposed change will revise TS governing near-EOL MTC by placing a set of conditions on core performance, which if met, would allow conditional exemption from the required MTC measurement. The conditional exemption will be determined on a cycle-specific basis by considering the margin predicted to the surveillance requirement MTC limit and the performance of other core parameters, such as beginning of life MTC measurements and the critical boron concentration as a function of cycle average burnup.

3.0 BACKGROUND

One of the controlling parameters for power and reactivity increases is the MTC. The requirements of TS 3.1.1.4 ensure that the MTC remains within the bounds used in the applicable Updated Final Safety Analysis Report Chapter 14 accident analysis. This, in turn, ensures inherently stable power operations during normal operation and accident conditions.

The TS place both Limiting Condition for Operation (LCO) and SR constraints on the MTC, based on the accident analysis assumptions for the moderator density coefficient. A positive moderator density coefficient corresponds to a negative MTC. The most negative MTC LCO limit requires that the MTC be less negative than the specified limit for the all rods withdrawn, EOL, rated thermal power condition. To demonstrate compliance with the most negative MTC LCO, the surveillance requires verification of the MTC after 300 ppm equilibrium boron concentration is reached. Because the Hot Full Power (HFP) MTC value will gradually become more negative with further core burnup and boron concentration reduction, a 300 ppm MTC surveillance value should be less negative than the EOL LCO limit. To account for this effect, the 300 ppm MTC surveillance value is sufficiently less negative than the EOL LCO limit value, providing assurance that the LCO limit will be met as long as the 300 ppm MTC surveillance criterion is met.

Currently, the TS require measurement of MTC at BOL to verify the most positive MTC limit and near-EOL to verify the most negative MTC limit. At BOL, the measurement of the isothermal temperature coefficient is relatively simple to perform since it is done at hot zero power isothermal conditions and is not complicated by changes in the enthalpy rise or the presence of xenon. The measurement made near EOL is performed at or near HFP conditions.

MTC measurements at HFP are more difficult to perform due to 1) small variations in soluble boron concentration, 2) changes in xenon concentration and distribution, 3) changes in fuel temperature, and 4) changes in enthalpy rise created by small changes in the core average power during the measurement. Changes in each of these parameters must be accurately accounted for when reducing the measurement data, or additional measurement uncertainties will be introduced. Even though these additional uncertainties may be small, the total reactivity change associated with the swing in moderator temperature is also relatively small. The resulting MTC measurement uncertainty created by even a small change in power level can then become significant and, if improperly accounted for, can yield misleading measurement results.

Each MTC measurement presents a perturbation to normal operation and to the reactor itself, requiring dedicated time by control room operators for most of a shift and a significant amount of primary coolant water during the evolution. An alternate method is proposed for use at CNP to AEP:NRC:5132 Page 4 to minimize disruption to normal plant operations. The MTC measurement is replaced by a design calculation of the core MTC if predefined requirements are met.

The proposed change would allow modification of the EOL MTC SR by placing a set of conditions on core performance. If these conditions are met, i.e., the specified revised prediction of the MTC and limits for several core parameters measured during the cycle are within specified bounds, the surveillance measurement would not be required.

4.0 TECHNICAL ANALYSIS

The conditional exemption from the HFP near-EOL 300 ppm MTC measurement does not impact the safe operation of CNP. The safety analysis assumption of a constant moderator density coefficient and the actual value assumed will not change. The TS Bases for, and the values of, the most negative MTC LCO and SR are not altered. Instead, a revised prediction is compared to the surveillance MTC to determine if the limit is met. The method for calculating the revised prediction is consistent with the approved methodology of WCAP-13749-P-A (Reference 1).

The methodology for the proposed change was submitted to the U. S. Nuclear Regulatory Commission (NRC) as Westinghouse topical report WCAP-13749 in May 1993. In October 1996, the NRC determined the report to be acceptable for referencing in license applications to the extent specified and under the limitations stated in the Brookhaven technical evaluation report and the NRC staff's safety evaluation report. Reference 1 includes all of these documents.

The topical report was approved by the NRC with two requirements:

. only PHOENIX/ANC calculation methods are used for the individual plant analyses relevant to determinations for the EOL MTC plant methodology, and

. the predictive correction is reexamined if changes in core fuel designs or continued MTC calculation/measurement data show significant effect on the predictive correction.

CNP will meet both of these requirements. The PHOENIX/ANC calculation methods are used for the CNP core designs. Prior to use of the conditional elimination technique, CNP will confirm that core design changes and MTC calculation and measurement data do not show a significant effect on the predictive correction. If a significant effect is found, the use of the predictive correction will be reexamined.

All of the core performance benchmark criteria specified in WCAP-13749-P-A, which are confirmed from startup physics test results, from routine HFP boron concentration measurements, and from flux map surveillances performed during the cycle, must be met before the Revised Predicted MTC can be calculated per the prescribed algorithm in Reference 1.

Enhancement I&M is using NRC-approved WCAP-13749-P-A as the basis for this license amendment request.

I&M will meet all of the technical requirements in the approved WCAP, but proposes an enhancement to reduce regulatory burden for both the NRC and the licensee. I&M proposes not to submit a "Most Negative Moderator Temperature Coefficient Limit Report" (the Report) to to AEP:NRC:5132 Page 5 the NRC. There are two reasons for this. First, there is an inconsistency in the WCAP regarding the time frame of data collection and the submittal of the Report to the NRC. More importantly, the Report serves no apparent technical or business need. Each of these reasons is explained below. This enhancement has been accepted by the NRC through approval of similar TS amendments as listed in Section 8.0 of this amendment request.

First, Section 3.3.3 of the WCAP states:

The Technical Specification Bases of the most negative MTC LCO and SR and the values of these limits are not altered. Instead, a revised prediction is compared to the SR MTC to determine if the SR limit is met. The revised prediction is simply the sum of the predicted HFP 300 ppm SR MTC plus an AFD correction factor plus a predictive correction term. This algorithm is summarized in Table 3-3.

Table D-2 of the WCAP states that the algorithm for determining the revised predicted near-EOL 300 ppm MTC is (emphasis added):

The Revised Predicted MTC = Predicted MTC + AFD Correction +

Predicted Correction where:

Predicted MTC is calculated from Figure 1 [Predicted HFP ARO 300 ppm MTC Versus Cycle Burnup] at the burnup corresponding to the measurement of 300 ppm at RTP conditions...

Table D-3 of the WCAP provides an example worksheet for calculating the revised predicted near-EOL 300 ppm MTC. Two of the required data inputs for the worksheet (B. 1 and B.2) are used to calculate the AFD correction term in the algorithm (emphasis added):

B. 1 Burnup of most recent HFP, equilibrium MWD/MTU conditions incore flux map B.2 Measured HFP AFD at burnup (B. 1) % AFD Reference incore flux map I.D. Date:

However, Appendix A to the WCAP states that a new TS 6.9.1.7 is to be added (emphasis added):

6.9.1.7 The most negative MTC limits shall be provided to the NRC Regional Administrator with a copy to the Director of Nuclear Reactor Regulation, Attention:

Chief, Core Performance Branch, U. S. Nuclear Regulatory Commission, Washington, D.

C. 20555, at least 60 days prior to the date the limit would become effective unless otherwise approved by the Commission by letter. This report wtill include the data reauiredfor the determinationof the Revised Prediction of the 300 ppm/ARO/RTP MTC per WCAP-13749, "Safety Evaluation Supporting the Conditional Elimination of the Most Negative EOL Moderator Tempaerature Coefficient Measurement", May, 1993 (Westinghouse Proprietary).

to AEP:NRC:5132 Page 6 Because the Report would have to be submitted at least 60 days before reaching 300 ppm boron concentration, it cannot include the 300 ppm data required for determining the Revised Prediction. To meet the Report submittal requirement, the data to be used in calculating the revised predicted MTC may have to be taken 60 to 90 days prior to reaching 300 ppm boron.

The WCAP does not provide any method for adjusting the revised predicted MTC to account for data collected 60 to 90 days prior to 300 ppm, nor does it provide justification for using such early data in the calculation. Therefore, the requirement to submit the Report and the requirements for the data that go into the report are inconsistent.

More importantly, the Report serves no apparent technical or business need. The applicability restrictions in the WCAP, the algorithm, and the acceptance criteria of the proposed Report would be included in the station procedure governing the EOL MTC surveillance. There is no compelling reason that this particular surveillance should require notifying the NRC prior to performing the surveillance procedure.

5.0 REGULATORY SAFETY ANALYSIS 5.1 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 probability or consequences of accidents previously evaluated in the Updated Final Safety Analysis Report (UFSAR) are unaffected by this proposed change because there is no change to any equipment response or accident mitigation scenario. 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 new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system.

! ,, Z. to AEP:NRC:5 132 Page 7 Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No The margin of safety associated with the acceptance criteria of any accident is unchanged.

The proposed change will have no affect on the availability, operability, or performance of the safety-related systems and components. A change to a surveillance requirement is proposed, but the limiting conditions for operation required by the Teclmical Specifications (TS) are not changed.

The Technical Specifications Bases are founded in part on the ability of the regulatory criteria to be satisfied assuming the limiting conditions for operation are met for the various systems. Conformance to the regulatory criteria for operation with the conditional exemption from the near-end of life moderator temperature coefficient (MTC) measurement is demonstrated and the regulatory limits are not exceeded. Therefore, the margin of safety as defined in the TS is not reduced.

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

In summary, based upon the above evaluation, I&M has concluded that the proposed amendment involves 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.

5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36 (c) (3), "Surveillance Requirements" stipulates that surveillances be performed to assure the necessary quality of systems and components be maintained, the facility operations will be within safety limits, and that the limiting condition for operations will be met.

The regulatory basis for TS 4.1.1.4, 'Moderator Temperature Coefficient," is to ensure that the value of the MTC remains within the limiting condition assumed in the CNP UFSAR accident and transient analyses.

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.

f - to AEP:NRC:5132 Page 8

6.0 ENVIRONMENTAL CONSIDERATION

S 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 surveillance requirement. 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.

7.0 REFERENCES

1. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1997.

8.0 PRECEDENT The NRC has approved similar submittals at plants implementing WCAP-13749-P-A.

South Texas Project Accession No. ML023400252 Virgil C. Summer Accession No. ML042040460

Attachment IA to AEP:NRC:5132 DONALD C. COOK NUCLEAR PLANT UNIT 1 CURRENT TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3/4 1-5a 6-12

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS 4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a) The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b) The MTC shall be measured at any THERMAL POWER within 7 EFPD after reaching an equilibrium boron concentration of 300 ppmri The measured value shall be compared to the 300 ppm surveillance limit specified in the COLR. In the event this comparison indicates that the MTC will be more negative than the EOL limit, the MTC shall be remeasured at least once per 14 EFPD during the remainder of the fuel cycle and the MTC value compared to the EOL limit.

Fef the MTC in accordanM fkI May

-4.mi be suspended, provided that the benichmaiaklcriteria in CP P-Aand the Rvisd eic s i satisfielj COOK NUCLEAR PLANT-UNIT I Page 3/4 1-5a AMENDMENT444,146

6.0 ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the U.S.

Nuclear Regulatory Commission (Attn: Document Control Desk), Washington, D.C. 20555, with a copy to the Regional Office no 'later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.9.1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

a. Moderator Temperature Coefficient Limits for Specification 3/4.1.1.4,

b. Rod Drop Time Limits for Specification 3/4.1.3.3,

c. Shutdown Rod Insertion Limits for Specification 3/4.1.3.4,

d. Control Rod Insertion Limits for Specification 3/4.1.3.5,

e. Axial Flux Difference for Specification 3/4.2.1,

f. Heat Flux Hot Channel Factor for Specification 3/4.2.2,

g. Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3, and

h. Allowable Power Level for Specification 3/4.2.6.

6.9.1.9.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

a. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 (Westinghouse Proprietary),

b. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," September 1974 (Westinghouse Proprietary),

c. WCAP-10216-P-A, Revision IA, "Relaxation of Constant Axial Offset Control/FQ Surveillance Technical Specification," February 1994 (Westinghouse Proprietary),

d. WCAP-10266-P-A Rev. 2, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," March 1987 (Westinghouse Proprietary).

e. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," July 1991 (Westinghouse Proprietary).

~WAP7 13749-P.A ."aet Baiation~Supportn h odtoa Eepino h ost Negative oCoef n as !ar9 hEL 22 tWestinghosePrt~ietz)

COOK NUCLEAR PLANT-UNIT I Page 6-12 AMENDM ENT 69,4-54, 44, 489,26 26,279

Attachment lB to AEP:NRC:5132 DONALD C. COOK NUCLEAR PLANT UNIT 2 CURRENT TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3/4 1-6 6-12

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a) The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b) The MTC shall be measured at any THERMAL POWER within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. The measured value shall be compared to the 300 ppm surveillance limit specified in the COLR. In the event this comparison indicates that the MTC will be more negative than the EOL limit, the MTC shall be remeasured at least once per 14 EFPD during the remainder of the fuel cycle and the MTC value compared to the EOL limit.

L Measureme'nt'of th&e MTC- in accordance with SR-A.I.1 Ab may be suspendeid, C a t R P sntesatisfied that iprpied the benchmark criteria ii COOK NUCLEAR PLANT-UNIT 2 Page 3/4 1-6 AM ENDM ENT V,40,4, 133

6.0 ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission (Attn: Document Control Desk), Washington, D.C. 20555, with a copy to the Regional Office no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.9.1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

a. Moderator Temperature Coefficient Limits for Specification 3/4.1.1.4,

b. Rod Drop Time Limits for Specification 3/4.1.3.4,

c. Shutdown Rod Insertion Limits for Specification 3/4.1.3.5,

d. Control Rod Insertion Limits for Specification 3/4.1.3.6,

e. Axial Flux Difference for Specification 3/4.2.1,

f. Heat Flux Hot Channel Factor for Specification 3/4.2.2,

g. Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3, and

h. Allowable Power Level for Specification 3/4.2.6.

6.9.1.9.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

a. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 (Westinghouse Proprietary),

b. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report,"

September 1974 (Westinghouse Proprietary),

c. WCAP-10216-P-A, Revision IA, "Relaxation of Constant Axial Offset ControVFQ Surveillance Technical Specification," February 1994 (Westinghouse Proprietary),

d. WCAP-10266-P-A Rev. 2, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," March 1987 (Westinghouse Proprietary).

e. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," July 1991 (Westinghouse Proprietary).

ngtv Oioeratoic it Measurement,,.Mr_9, Wsifghouse COOK NUCLEAR PLANT-UNIT 2 Page 6-12 AMENDMENT A1, 4138, 4-i, 4AL, 490, X210,261

Attachment 2A to AEP:NRC:5132 DONALD C. COOK NUCLEAR PLANT UNIT 1 CURRENT TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED 3/4 1-5a 6-12

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.1 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS 4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:

i a) The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading. i b) The MTC shall be measured at any THERMAL POWER within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.* The measured value shall be compared to the 300 I ii ppm surveillance limit specified in the COLR. In the event this comparison indicates that the MTC will be more negative than the EOL limit, the MTC shall be remeasured at least once per 14 EFPD during the remainder of the fuel cycle and the MTC value compared to the EOL limit.

Measurement of the MTC in accordance with SR 4.1.1 .4.b may be suspended, provided that the benchmark criteria in WCAP-1 3749-P-A and the Revised Prediction specified in the COLR are satisfied.

COOK NUCLEAR PLANT-UNIT I Page 3/4 1-5a AMENDMENT444,446,

6.0 ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the U.S.

Nuclear Regulatory Commission (Attn: Document Control Desk), Washington, D.C. 20555, with a copy to the Regional Office no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.9.1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

a. Moderator Temperature Coefficient Limits for Specification 3/4.1.1.4,

b. Rod Drop Time Limits for Specification 3/4.1.3.3,

c. Shutdown Rod Insertion Limits for Specification 3/4.1.3.4,

d. Control Rod Insertion Limits for Specification 3/4.1.3.5,

e. Axial Flux Difference for Specification 3/4.2.1,

f. Heat Flux Hot Channel Factor for Specification 3/4.2.2,

g. Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3, and

h. Allowable Power Level for Specification 3/4.2.6.

6.9.1.9.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

a. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 (Westinghouse Proprietary),

b. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," September 1974 (Westinghouse Proprietary),

c. WCAP-10216-P-A, Revision IA, "Relaxation of Constant Axial Offset Control/FQ Surveillance Technical Specification," February 1994 (Westinghouse Proprietary),

d. WCAP-10266-P-A Rev. 2, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," March 1987 (Westinghouse Proprietary).

e. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," July 1991 (Westinghouse Proprietary).

f. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement,' March 1997, (Westinghouse Proprietary).

COOK NUCLEAR PLANT-UNIT I Page 6-12 AM ENDM ENT 69, 4-4, 474,489, 2O"

-26,2,9,

Attachment 2B to AEP:NRC:5132 DONALD C. COOK NUCLEAR PLANT UNIT 2 CURRENT TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED 3/4 1-6 6-12

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.1 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a) The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b) The MTC shall be measured at any THERMAL POWER within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.* The measured value shall be compared to the I 300 ppm surveillance limit specified in the COLR. In the event this comparison indicates that the MTC will be more negative than the EOL limit, the MTC shall be remeasured at least once per 14 EFPD during the remainder of the fuel cycle and the MTC value compared to the EOL limit.

Measurement of the MTC in accordance with SR 4.1.1 .4.b may be suspended, provided that the benchmark criteria in WCAP-13749-P-A and the Revised Prediction specified in the COLR are satisfied.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 1-6 AMENDMENT N, 407.,408,4-33,

6.0 ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission (Attn: Document Control Desk), Washington, D.C. 20555, with a copy to the Regional Office no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.9.1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

a. Moderator Temperature Coefficient Limits for Specification 3/4.1.1.4,

b. Rod Drop Time Limits for Specification 3/4.1.3.4,

c. Shutdown Rod Insertion Limits for Specification 3/4.1.3.5,

d. Control Rod Insertion Limits for Specification 3/4.1.3.6,

e. Axial Flux Difference for Specification 3/4.2.1,

f. Heat Flux Hot Channel Factor for Specification 3/4.2.2,

g. Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3, and

h. Allowable Power Level for Specification 3/4.2.6.

6.9.1.9.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

a. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 (Westinghouse Proprietary),

b. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report,"

September 1974 (Westinghouse Proprietary),

c. WCAP-10216-P-A, Revision IA, "Relaxation of Constant Axial Offset ControlIFQ Surveillance Technical Specification," February 1994 (Westinghouse Proprietary),

d. WCAP-10266-P-A Rev. 2, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," March 1987 (Westinghouse Proprietary).

e. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," July 1991 (Westinghouse Proprietary).

f. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1997, (Westinghouse Proprietary).

COOK NUCLEAR PLANT-UNIT 2 Page 6-12 AMENDMENT 4, 48, 4-5, 5, 490, 24,26

Attachment 3A to AEP:NRC:5132 DONALD C. COOK NUCLEAR PLANT UNIT 1 IMPROVED TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3.1.3-2 5.6-3 5.6-4

MTC 3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.3.2 -- NU7NTE' Not required to be' performed provided that the Abe nrchmark criteria in WCAP-1 3749-P-A and the Revised Pedi ction ,speified .in jhe COLR 'are satisfied.

~

- - - - r-o- x- - . - - - -

Verify MTC is within lower limit. Once each cycle within 7 effective full power days (EFPD) after reaching an equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm AND 14 EFPD thereafter if MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR until the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of s 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR Cook Nuclear Plant Unit 1 3.1 .3-2 Amendment No. 2-85

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

5. LCO 3.1.6, "Control Bank Insertion Limits";

6. LCO 3.2.1, "Heat Flux Hot Channel Factor (Fa(Z))";

7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FNH )";

8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

9. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," Functions 6 and 7 (Overtemperature AT and Overpower AT, respectively)

Allowable Value parameter values;

10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and

11. LCO 3.9.1, "Boron Concentration."

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, 'Westinghouse Reload Safety Evaluation Methodology," (Westinghouse Proprietary);

2. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," (Westinghouse Proprietary);

3. WCAP-1 0216-P-A, "Relaxation of Constant Axial Offset Control/Fa Surveillance Technical Specification," (Westinghouse Proprietary);

4. WCAP-10266-P-A, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," (Westinghouse Proprietary); and

5. WCAP-1 2610-P-A, 'VANTAGE+ Fuel Assembly Reference Core Report," (Westinghouse Proprietary).

r WCAP-1 374 ' W1ty Evaluation Supportinthe Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measuremrent,'jW estinghouse Proprietary).i

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

5.6-3 Amendment No. 285 Cook Nuclear Cook Unit I1 Plant Unit Nuclear Plant 5.6-3 Amendment No. 2-M

Il X Reporting Requirements 5.6 5.6 Reporting Requirements

.IMITS REPORT(COLR) (continued)

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring Report When a report is required by Condition B or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation,".a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Steam Generator Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the NRC.

b. The complete results of the steam generator tube inservice inspection shall be submitted to the NRC prior to March 1 for the inspection that was completed in the previous calendar year. This report shall include:

1. Number and extent of tubes inspected;

2. Location and percent of wall-thickness penetration for each indication of an imperfection; and

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the NRC in accordance with 10 CFR 50.72. A Licensee Event Report shall be submitted in accordance with 10 CFR 50.73 and shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

Cook Nuclear Plant Unit 1 5.6-4 Amendment No. 285

Attachment 3B to AEP:NRC:5132 DONALD C. COOK NUCLEAR PLANT UNIT 2 IMPROVED TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3.1.3-2 5.6-3 5.6-4

MTC 3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.3.2 -N O,~7~-NTE-,",- - ,

Not required to be performed provided that the

'benchmark 'criteria'"in WCAP,-1 3749-P-A and the Revised Prediction specified in the COLR are satisfied:_______________

. .. ).:J.:: t __i _....>_, .. _._ih........ _ " .. ' o.5......s_"'.;

Verify MTC is within lower limit. Once each cycle within 7 effective full power days (EFPD) after reaching an equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm AND 14 EFPD thereafter if MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR until the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of

60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR Cook Nuclear Plant Unit 2 3.1 .3-2 Amendment No. 2-68

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

5. LCO 3.1.6, "Control Bank Insertion Limits";

6. LCO 3.2.1, "Heat Flux Hot Channel Factor (Fa(Z))";

7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FANH)";

8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

9. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," Functions 6 and 7 (Overtemperature AT and Overpower AT, respectively)

Allowable Value parameter values;

10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and

11. LCO 3.9.1, "Boron Concentration."

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, 'Westinghouse Reload Safety Evaluation Methodology," (Westinghouse Proprietary);

2. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," (Westinghouse Proprietary);

3. WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control/F 0 Surveillance Technical Specification," (Westinghouse Proprietary);

4. WCAP-10266-P-A, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," (Westinghouse Proprietary); and

5. WCAP-1 2610-P-A, 'VANTAGE+ Fuel Assembly Reference Core Report," (Westinghouse Proprietary).

6TTWCAP-1 aluation Supporting the oditoi~ii Ev9--,-"

Exemption of the Most Negative EOL Moderator Temperature Coefficient Measuremient; (Westinghouse Proprietary)*

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling'Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

5.6-3 Amendment No. 268 Cook Nuclear Plant Cook Nuclear Unit 2 Plant Unit 2 5.6-3 Amendment No. 268

Reporting Requirements 5.6 5.6 Reporting Requirements 5.. -CORE bPERTNGOiMiTS'REPORTTCOL~Rnictinued)

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring Report When a report is required by Condition B or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Steam Generator Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the NRC.

b. The complete results of the steam generator tube inservice inspection shall be submitted to the NRC prior to March 1 for the inspection that was completed in the previous calendar year. This report shall include:

1. Number and extent of tubes inspected;

2. Location and percent of wall-thickness penetration for each indication of an imperfection; and

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the NRC in accordance with 10 CFR 50.72. A Licensee Event Report shall be submitted in accordance with 10 CFR 50.73 and shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

Cook Nuclear Plant Unit 2 5.6-4 Amendment No. 2-68

Attachment 4A to AEP:NRC:5132 DONALD C. COOK NUCLEAR PLANT UNIT 1 IMPROVED TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED

- 3.1.3-2 5.6-3 5.6-4

MTC 3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.3.2 Not required to be performed provided that the benchmark criteria in WCAP-13749-P-A and the Revised Prediction specified in the COLR are satisfied.

Verify MTC is within lower limit. Once each cycle within 7 effective full power days (EFPD) after reaching an equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm AND 14 EFPD thereafter if MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR until the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of

< 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR Cook Nuclear Plant Unit 1 3.1 .3-2 Amendment No. 285 l

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLRY (continued)

5. LCO 3.1.6, "Control Bank Insertion Limits";

6. LCO 3.2.1, "Heat Flux Hot Channel Factor (Fa(Z))";

7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FN )";

8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

9. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," Functions 6 and 7 (Overtemperature AT and Overpower AT, respectively)

Allowable Value parameter values;

10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and

11. LCO 3.9.1, "Boron Concentration."

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," (Westinghouse Proprietary);

2. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," (Westinghouse Proprietary);

3. WCAP-1 0216-P-A, "Relaxation of Constant Axial Offset Control/F0 Surveillance Technical Specification," (Westinghouse Proprietary);

4. WCAP-10266-P-A, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," (Westinghouse Proprietary); and

5. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," (Westinghouse Proprietary).

6. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," (Westinghouse Proprietary).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

Cook Nuclear Plant Unit 1 5.6-3 Amendment No. 285 l

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring ReDort When a report is required by Condition B or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shalI outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Steam Generator Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the NRC.

b. The complete results of the steam generator tube inservice inspection shall be submitted to the NRC prior to March 1for the inspection that was completed in the previous calendar year. This report shall include:

1. Number and extent of tubes inspected;

2. Location and percent of wall-thickness penetration for each indication of an imperfection; and

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the NRC in accordance with 10 CFR 50.72. A Licensee Event Report shall be submitted in accordance with 10 CFR 50.73 and shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

Cook Nuclear Plant Unit 1 5.6-4 Amendment No. 285 l

Attachment 4B to AEP:NRC:5132 DONALD C. COOK NUCLEAR PLANT UNIT 2 IMPROVED TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED 3.1.3-2 5.6-3 5.6-4

MTC 3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.3.2 ~~~-NOTE----------

Not required to be performed provided that the benchmark criteria in WCAP-13749-P-A and the Revised Prediction specified in the COLR are satisfied.

Verify MTC is within lower limit. Once each cycle within 7 effective full power days (EFPD) after reaching an equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm AND 14 EFPD thereafter if MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR until the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of s 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR Cook Nuclear Plant Unit 2 3.1 .3-2 Amendment No. 268 l