Text

Review of Draft Safety Evaluation for Measurement Uncertainty Recapture Power Uprate for Donald C. Cook, Unit 2
	ML031270262
Person / Time
Site:	Cook
Issue date:	04/25/2003
From:	Joseph E Pollock; Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP:NRC:3902-01, TAC MB6751
	Download: ML031270262 (45)
	v • d • e

Indiana Michigan Power Company 500 Circle Drive Buchanan, MI 49107 1395 INDIANA MICHIGAN POWER April 25, 2003 AEP:NRC:3902-01 10 CFR 50, Appendix K Docket Nos.: 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Unit 2 REVIEW OF DRAFT SAFETY EVALUATION FOR MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE (TAC NO. MB6751)

This letter documents the performance of a review of the draft safety evaluation (SE) for the Donald C. Cook Nuclear Plant (CNP) Unit 2 measurement uncertainty recapture (M[UR) power uprate license amendment. Indiana Michigan Power Company's (I&M's) review confirmed that, with the clarifications provided in Enclosure 2, the factual information in the draft SE is complete and accurate.

By Reference 1, I&M, the licensee for CNP Unit 2, proposed to amend Facility Operating License DPR-74, including Appendix A, Technical Specifications, to allow a 1.66-percent increase in the licensed core power. The Reference 1 amendment request was supplemented by a letter, dated February 24, 2003 (Reference 2), that provided a proprietary copy of I&M's power calorimetric accuracy calculation for Unit 2. The information provided in References 1 and 2 formed the technical basis for the Nuclear Regulatory Commission (NRC) review of the Unit 2 MUR power uprate request.

The Reference 1 license amendment request is similar to the MUR power uprate that had previously been reviewed and approved for CNP Unit 1, by License Amendment Number 273 to Facility Operating License DPR-58 (Reference 3).

However, the approach to developing and formatting the SE for the Unit 2 license amendment was different from that used for Unit 1. The NRC staff has requested, via Reference 4, that I&M review the draft SE to confirm the factual information is accurate and complete.

q0oci

U. S. Nuclear Regulatory Commission AEP:NRC:3902-01 Page 2 This letter documents I&M's review of the draft SE. Enclosure 1 provides the affirmation requested by the NRC. Enclosure 2 summarizes the methodology and assumptions used in the I&M review of the draft SE, and provides comments that are intended to clarify information in the draft SE. Enclosure 3 identifies the factual portions of the draft SE that were within the scope of I&M's review. Attachment 1 provides a revised Regulatory Commitment.

In addition to the draft SE review documented in this letter, I&M reviewed the CNP core reload evaluation process in response to questions raised during a telephone conference conducted with the NRC on April 23, 2003. As a result of this review, I&M affirms that the Anticipated Transients Without SCRAM analysis will continue to be considered as part of future core reload evaluations, including review of moderator temperature coefficient, which are performed in accordance with CNP's 10 CFR Part 50, Appendix B design control process.

Furthermore, core reload evaluations are currently, and will continue to be, reviewed/evaluated in accordance with 10 CFR 50.59, Changes, tests, and experiments.

Finally, I&M has reviewed the regulatory commitments made in the Reference I license amendment request for consistency with the current implementation plans for the MUR Power Uprate Program. This review identified a potential inconsistency in the first commitment, which addresses the activities associated with installing the Leading Edge Flow Meter (LEFMrm) CheckPlus"" system and implementing the requested license amendment. To resolve this inconsistency, the commitment is revised, as follows:

"I&M is installing an LEFM CheckPlus system at CNP Unit 2 in anticipation of approval of this proposed amendment. Installation of this system will begin prior to the Unit 2 Cycle 14 refueling outage and will be completed after receipt of the requested license amendment. The design change for the installation will include instrumentation resealing, UFSAR revision, maintenance and operational procedure impacts, training, monitoring iso-phase bus duct temperature, and implementation of the LEFM CheckPlus system out-of-service administrative technical requirements. The UFSAR revision for the Unit 2 MUR power uprate will be reflected in the next update of the UFSAR submitted to the NRC pursuant to 10 CFR 50.71(e)."

This revised commitment is reflected in Attachment 1, Regulatory Commitments, to this letter. There are no new commitments made in this submittal.

U. S. Nuclear Regulatory Commission AEP:NRC:3902-01 Page 3 Should you have any questions, please contact Mr. Brian A. McIntyre, Manager of Regulatory Affairs, at (269) 697-5806.

Sincerely,

.Pollock Site Vice President NH/rdw

References:

1. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 2, Docket No. 50-316, License Amendment Request for Appendix K Measurement Uncertainty Recapture - Power Uprate Request," AEP:NRC:2902, dated November 15, 2002

2. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 2, Submittal of Change Sheet 4 to Power Measurement Uncertainty Calculation in Support of License Amendment Request for Appendix K Measurement Uncertainty Recapture - Power Uprate Request," AEP:NRC:3902, dated February 24, 2003

3. Letter from J. F. Stang, NRC, to A. C. Bakken HI, I&M, "Donald C.

Cook Nuclear Plant, Unit 1 - Issuance of Amendment 273 Regarding Measurement Uncertainty Recapture Power Uprate," dated December 20, 2002

4. Letter from J. F. Stang, NRC, to A. C. Bakken HI, I&M, "Donald C.

Cook Nuclear Plant, Unit 2 - Review of Draft Safety Evaluation for Measurement Uncertainty Recapture Power Uprate (TAC No.

MB675 1)," dated April 18, 2003

U. S. Nuclear Regulatory Commission AEP:NRC:3902-01 Page 4

Enclosures:

1. Notarized Affirmation Affidavit

2. I&M Methodology, Assumptions, and Comments Pertaining to the Review of the Draft Safety Evaluation for D. C. Cook Unit 2 Measurement Uncertainty Recapture Power Uprate License Amendment

3. Marked-up Copy of Draft Safety Evaluation Indicating Areas that are Outside the Scope of the I&M Review Attachments

1. Regulatory Commitments c: H. C. Chemoff - NRC Washingtion DC K. D. Curry - AEP Ft. Wayne, w/o enclosures/attachment J. E. Dyer - NRC Region III J. T. King - MPSC, w/o enclosures/attachment MDEQ - DW & RPD, w/o enclosures/attachment NRC Resident Inspector J. F. Stang, Jr. - NRC Washington DC to AEP:NRC:3902-01 Page I AFFIRMATION I, Joseph E. Pollock, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M) and that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M. Knowledgeable I&M personnel have reviewed the draft safety evaluation enclosed in the letter from John F. Stang (NRC) to A. Christopher Bakken Iml (I&M), dated April 18, 2003. Enclosures 2 and 3 to I&M Correspondence Number AEP:NRC:3902-01 summarize the methodology used by I&M for review of the draft safety evaluation and identify the factual matters that were subject to this review. To the best of my knowledge, information, and belief, the statements made in Enclosures 2 and 3 regarding I&M's review of the draft SE are true and correct.

Indiana Michigan Power Company J. E. Pollock Site Vice President SWORN TO AND SUBSCRIBED BEFORE MEE DAYO ~ h If ,2003 JULIE E.NEWMIULER Notary Public, Berrien County, Ml My Commission Expires Aug 22, 2004

// Notary(Public MyCmmission Expires _g-c55 c-;"d /

to AEP:NRC:3902-01 Page I I&M Methodology, Assumptions, and Comments Pertaining to the Review of the Draft Safety Evaluation for D. C. Cook Unit 2 Measurement Uncertainty Recapture Power Uprate License Amendment Review Assumptions and Methodologv As requested by letter dated April 18, 2003 (Reference 1), Indiana Michigan Power Company (I&M) personnel have completed a review of the draft safety evaluation (SE) for the Donald C.

Cook Nuclear Plant (CNP) Unit 2 Measurement Uncertainty Recapture (MUR) Power Uprate.

Consistent with the Nuclear Regulatory Commission (NRC) staff request, and the NRC staff guidance provided in the Office of Nuclear Reactor Regulation's Office Instruction COM-203, "Informal Interfacing and Exchange of Information with Licensees and Applicants," the intent of the draft SE review was to verify that the factual information is accurate and complete. indicates the portions of the draft SE that were within the scope of I&M's review.

The review did not verify information included in the Regulatory Evaluation or Summary sections of the draft SE, nor did it address the acceptability of the NRC staff conclusion for the topics reviewed in the draft SE.

For each of the technical areas evaluated in the Unit 2 MUR power uprate application, the SE includes a table that provides reference to the discussion of the topic in the amendment request, the CNP Updated Final Safety Analysis Report, and NRC-issued documents that contain analyses that bound the proposed conditions. Additionally, the tables indicate whether the topic is similar to the previously-approved Unit 1 MUR power uprate (i.e., whether the Unit 1 and Unit 2 topics are bounded from a design and licensing basis). For each topic, the I&M review verified the following factual information:

I&M verified that the column entitled, "Unit 2 MUR Application Section" provides references to the appropriate section and pages from the CNP Unit 2 MUR license amendment request (Reference 2).

In Table 3.1.2, "Instrumentation and Controls," I&M verified that the information in the column entitled "Meets Criteria in NRC-approved Topical Reports ER-80P & ER-157P" accurately reflects the conclusion reached by I&M in References 2 and 3, and provides the appropriate reference to the NRC's approval for each topic.

In Tables 3.2.2, "Reactor Systems;" 3.3.2, "Electrical Systems;" 3.4.2, "Civil and Mechanical Engineering;" 3.6.2, "Materials and Chemical Engineering;" and 3.8.2, "Plant Systems,"

I&M verified that the information in the column entitled, "Bounded by NRC-approved analysis" accurately reflects the conclusion reached by I&M in Reference 2 and provides the appropriate reference to the NRC's approval for each topic. (In this column, a "Y" is assumed to indicate that the Reference 2 evaluation concluded that the current licensing basis for the topic remains bounding for the 1.66 percent power uprate. An "N" is assumed to to AEP:NRC:3902-01 Page 2 indicate that the current analysis of record for the topic did not bound the 1.66 percent power uprate.)

- I&M verified that the correct references to NRC approval of the current analysis, or analyses, are provided for each topic, if applicable. For I&M's review of the draft SE, it was assumed that references are not provided to each occurrence in which a particular topic is mentioned in NRC correspondence, but rather only to those occurrences in which an analysis, or portions of an analysis, were approved by the NRC.

In Tables 3.2.2 through 3.8.2, I&M verified that the information provided in the column entitled, "Similar to Unit 1 MUR" accurately reflects whether the review or evaluation performed by I&M, as summarized in Reference 2, produced similar (not necessarily identical) results to the results of the corresponding topic for the Unit 1 MUR license amendment. (In this column, a "Y" is assumed to indicate that the Reference 2 review or evaluation produced similar results to the corresponding Unit 1 review or analysis, and an "N" is assumed to indicate that it did not. In several cases, notes are provided to clarify dissimilarities.)

For the Notes to Tables 3.2.2 through 3.8.2, I&M verified that the factual information pertaining to the Reference 2 reviews and evaluations are complete and accurate.

For other factual statements provided in the Technical Evaluation sections of the draft SE, I&M verified that the factual information pertaining to the Reference 2 and 3 review and evaluations are complete and accurate.

Results of Review I&M has identified several clarifications to the the information provided in the draft SE. These clarifications are delineated in the Comments section of this Enclosure.

I&M's review confirmed that, with the clarifications provided in the Comments section of this Enclosure, the factual information in the draft SE is complete and accurate.

Comments SE Section or Table Comment General It is noted that some sections of the draft SE refer to 10 CFR Part 50, Appendix A, General Design Criteria (GDC), which may not be applicable to the design of CNP.

3.0 It is recommended that the second bullet be clarified to reflect that the table indicates the location in Attachment 3 to the licensee's November 15, 2002 application, unless indicated otherwise.

to AEP:NRC:3902-01 Page 3 SE Section or Table Comment 3.1.2 In the second paragraph, first sentence, the date of the supplemental letter should be February 24, 2003, not 2002.

Table 3.1.2 For the topic "Methodology used to calculate the uncertainty of the LEFM system," under the column entitled Unit 2 MUR Application Section, it is noted that the February 24, 2003, supplement is not referenced. The supplement submitted the Unit 2 specific power measurement uncertainty calculation to the NRC.

Table 3.2.2 For the topic, "Post-LOCA Long-Term Core Cooling," Page 37 is the correct page for Section 11.1.3.1 of the Unit 2 MUR Application.

Table 3.2.2 For the topic, "Rupture of a Control Rod Drive Mechanism Housing," it was noted that the addition of a reference to Note 7 to the column entitled, "Similar to Unit 1 MUR," would provide consistency with the existing Note 7 reference for the "Rupture of a CRDM housing - Mode 3" topic on Page 7.

Table 3.2.2 For the topic, "Chemical Volume and Control System Malfunction," it was noted that the CNP Unit 2 UFSAR refers to this event as "Uncontrolled Boron Dilution" (see Note 8).

Table 3.2.2 Reference 3 should include the date of the letter - December 13, 1999.

Table 3.2.2 Notes 1 and 2 are not used in the table.

Table 3.4.2 For the topic, "Main Steam System," in the "Bounded by NRC-approved analysis" column, a reference should be added to License Amendments 182 and 167, dated September 9, 1994. (See Main Steam System and Steam Dump System in Table 3.8.2.)

3.5.2 The second paragraph addresses License Amendment Nos. 271 and 252, dated November 14, 2002, which approved selective implementation of an alternative source term (AST). License Amendment Nos. 258 and 241, dated November 13, 2001, approved use of the AST for the Fuel Handling Accident, and should be addressed in this section as well.

Table 3.6.2 The topics, "Structural Integrity of Control Rod Drive Mechanisms Nozzles," and "Structural Integrity of RV Internals," should also include Reference 7 in the column entitled, "Bounded by NRC-approved analysis."

Table 3.6.2 For the topic, "Structural Integrity of Other Class 1 Reactor Coolant System Components," under the UFSAR Section column, separate sections "4.3" and "14.3.3." These are separate sections.

Table 3.8.2 For consistency with other topics (e.g., Table 3.2.2, "RHR System"), the "NRC Staff Conclusion" for the topic "Post-LOCA Containment Hydrogen Generation,"

should reference the discussion in Section 3.8.2.1.

to AEP:NRC:3902-01 Page 4 SE Section or Table Comment Table 3.8.2 Note 1 is not applicable for the topic, "Main Steam System and Steam Dump System." The correct note for this topic is already provided as Note 2 to Table 3.4.2.

Table 3.8.2 Reference 2 should include the date of the letter - December 13, 1999.

References:

1. Letter from J. F. Stang, NRC, to A. C. Bakken HI, I&M, "Donald C. Cook Nuclear Plant, Unit 2 - Review of Draft Safety Evaluation for Measurement Uncertainty Recapture Power Uprate (TAC No. MB6751)," dated April 18, 2003

2. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 2, Docket No. 50-316, License Amendment Request for Appendix K Measurement Uncertainty Recapture - Power Uprate Request," AEP:NRC:2902, dated November 15, 2002

3. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 2, Submittal of Change Sheet 4 to Power Measurement Uncertainty Calculation in Support of License Amendment Request for Appendix K Measurement Uncertainty Recapture - Power Uprate Request," AEP:NRC:3902, dated February 24, 2003

Enclosure 3 to AEP:NRC:3902-01 Marked-up Copy of Draft Safety Evaluation Indicating Areas that are Outside the Scope of the I&M Review

0onaId 01 Cok NUDE Plant1Unit a SIfeW Evluafon fr Aendmet No Mnsulen~tUncerlity Pow Upral Recplure ENCLOSURE

TABLE OF CONTENTS

1.0 INTRODUCTION

...................................... -1 -

2.0 BACKGROUND

...................................... 3.0 EVALUATION ...................................... 3.1 Instrumentation and Controls ...................................... 3.1.1 Regulatory Evaluation ....................-.................. . 3-3.1.2 Technical Evaluation ................ ...................... 3-3.1.3 Summar ................................................. 3.2 Reactor Systems .... 3.2.1 Reoulatorv Evaluation .................. .................... 3.2.2 Technical Evaluation ................. ..................... 3.2.2.1 RHR Cooldown ....................................... 10-3.2.2.2 Steam Generators ...................................... 3.2.2.3 Changes to Technical Specifications . ...................... 3.2.2.4 Commitments f-' .... ................................... 3.2.3 Summary ........... 3.3 Electrical Systems ...... .- 13 -

3.3.1 Reaulatory Evaluation . ..

. .- . 13 -

3.3.2 Technical Evaluation ............. ... l.... ..... 3.3.3 Summae . . . . . .. ' 3.4 Mechanicaland CivilEngineering. . ............................. 3.4.1 Reaulator Evaluation . . ..-..- 15-3.4.2 Technical Evaluation . ......... 3.4.3 Summarv ......... . . . . . 3.5 pose Conseauences Analysis .. .. . .............. ........... '3.5.1 Regulatory Evaluation . '.'.-...................,..- 18-3.5.2 Technical Evaluation .. . - 18 -

3.5.3 Summary ...... 3.6 Materials and Chemical Enaineerin ............................ . 3.6.1 Regulatory Evaluation . '--... ............-............-. 19-3.6.2 Technical Evaluation ......... ... .- 19-3.6.2.1 Pressurized Thermal Shock and Upper-Shelf Enerav Analyses .. 3.7 Human Factors ............. ....- 22 -

3.7.1 Regulatory Evaluation . . . .................... . 22 -

3.7.2 Technical Evaluation ....... . ..............-..... . 23-3.7.2.1 Operator Actions . . 3.7.2.2 Emergency and Abnormal Operating Procedures . .- 23 -

3.7.2.3 Control Room Controls. Displas. and Alarms .- 23 -

3.7.2.4 Control Room Plant Reference Simulator . . 24 -

3.7.2.5 0eratorTrainino'Program .. 24 -

3.7.3 Summar ...- 24 -

3.8 Plant Systems ...-.................... . 24 -

3.8.1 Reaulatory Evaluation . . . .24 -

3.8.2 Technical Evaluation .- 25-3.8.2.1 Post LOCA Containment Hdroae n Generation.- 29 -

3.8.3 Summary .. 29-

4.0 STATE CONSULTATION

.- 30-

5.0 ENVIRONMENTAL CONSIDERATION

.- 30-

6.0 CONCLUSION

.- 30-

Attachment:

List of Acronyms

1ndt'C^ e5 Arc&S InchO J 0c ion Ate sc5ope J T#Ail 'eview d&So A4 ',2o3 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. TO FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT. UNIT 2 DOCKET NO. 50-316

§.0 INTRODUCINi,< 0:

By application dated November 15,;2bb2, as su plemented February 24, 2003, the Indianam Michigan Power Company (I&Mi "the licensee) requested an amendment to the Technical\

Specifications (TSs) for the Donald C. iCook Nuclear Plnt, Unit 2. The proposed amendmen would increase the licensed reactor core power level by 1.6.6 percent from 3411 megawatts:

thermal (MWt) to 3468 MWt. The proposed increase is considered a measurement uncertainty recapture (MUR) power upratei Specifically, the proposed changes would revise:

1.. Paragraph 2.C.(1) in Facility Operating License DPR-74 to authorize operaton at a steady-state reactor core power level not in excess of 3468 MWt (100-percent power).

2. The definition of RATED THERMAL POWER (RTP) in TS 1.3 to reflect the increase from 3411 MWt to 3468 MWt.

3. The maximum allowed power level in TS 3.5.2, Action b, from 3250 MWt to 3304 MWt, to increase the maximum allowable core power level with a safety injection cross-tie valve closed.

4. TS Table 3.7-1, 'Maximum Allowable Power Range Neutron Flux High Setpoint with Inoperable Steam Line Safety Valves during 4 Loop Operation," to reflect the maximum allowed power for operation with inoperable main steam safety valves (MSSVs). With one inoperable MSSV per loop, the power reduction would be revised from 61.6 percent RTP to 60.4 percent RTP. With multiple inoperable safety valves per loop, the power reduction and associated reduction in high flux reactor trip setpoints would be revised to 43.0 percent (two inoperable MSSVs) and 25.7 percent (three inoperable MSSVs).

The February 24, 2003, supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the FederalRegisteron January 21, 2003 (6 FR 2805)

2.0 BACKGRU \

Nuclear power plants are licensed to operate at a specified core thermal power. Title 10 of te Code of Federal Regulation (10 CFR), Part 50, Appendix K, requires licensees to assume thatI the reactor has been operating continuously at a power level at least 1.02 Uimes the licensedl power level when performing loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) analyses. This requirement is included to ensure that instrumentation uncertainties are adequately accounted for in the analyses. Appendix K to 10 CFR Part 50 allows licensees to assume a power level lower than 1.02 times the licensed power level (but not less than the licensed power level), provided the licensee has demonstrated that the proposed value adequately accounts for instrumentation uncertainties. The licensee has proposed to use a value of 1.0034. To achieve this level of accuracy, the licensee will install the more accurate feedwater flow measurement meter described in NRC-approved Caldon, Inc.

(Caldon) Topical Report ER-80P' and its supplement, Topical Report ER-1 57P2. (The currently installed venturi flow meter will remain in place.) The NRC staff approved Caldon Topical i :-, ;, Report ER-80P by a safety evaluation report dated March 8, 1999. The NRC staff approved

£Caldon Topical Report ER-8OP for licensees' use in submitting licensing applications for power, level increases to 1 percent and for requesting exemptions from certain requirements of 10 CFR Part 50, Appendix K. The NRC staff approved Caldon Topical Report ER-I 57P by a safety evaluation report dated December 20, 2001. Caldon Topical Report ER-1 57P justified power level increases to 1.7 percent. -

The licensee proposed to increase the power output of the plant by the difference between the 1.02 multiplier used in the existing analyses of record and the 1.0034 multiplier proposed as a result of the installation of the more accurate flowmeter. Since the analyses of record for LOCA and ECCS assumed a power level of 1.02 times the licensed power level, a 1.66-percent increase in power could be achieved without necessitating reanalyses of these events. Other design-basis analyses are evaluated to ensure an appropriate accounting of power level uncertainties.

By application dated June 28, 2002, the licensee requested a similar 1.66 percent MUR power uprate for Unit 1. The NRC approved 1.66 percent MUR power uprate for D. C. Cook Unit 1 by License Amendment No. 273, dated December 20, 2002. Given the many commonalities between the D. C. Cook Unit 1 and Unit 2 design and' licensing bases, the licensee utilized a/

similar approach for reviewring and evaluating the Unit 2 MUR power uprate as that which was/

previously approved by the NRC staff for D. C. Cook Unit 1. _

3.0 EVALUATION The NRC staff's evaluation of the proposed D. C. Cook Unit 2 MUR power uprate is based on the guidance provided by Regulatory Issue Summary (RIS) 2002-03, NGuidance on the Content of Measurement Uncertainty Recapture Power Applications." RIS 2002-03 delineates the Caldon ER-80P, Revision 0, 'Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFMTW System, March 1997 2 Caldon ER-1 57P. Revision 5, 'Supplement to Topical Report ER-80P: Basis for a Power Uprate With h

the LEFM' or checkPluswM System," October 2001

appropriate scope and level of detail for the review and approval of an MUR power uprate application. For every technical area where the proposed MUR power uprate conditions are bounded by existing design and licensing bases analyses, the NRC staff has confirmed that the proposed conditions continue to be bounded and has provided a table which summarizes

the topics within each primary technical area

where the topic is addressed in the licensee's November 15, 2002, application

where the topic is addressed in the D. C. Cook Updated Final Safety Analysis Report

references to NRC documents which contain analyses that bound the proposed conditions

whether the topic is similar to the previously approved D. C. Cook Unit I MUR power uprate

the NRC's conclusion of acceptability The corresponding references and notes for each table immediately follow the table.

For situations where the proposed MLR power uprate conditions are not bounded by existing,

.I - design and licensing bases, the licensee has performed new analyses and the NRC staff has, conducted an independent evaluation. --

t3.1 Instrumentation and Co6ntrols) ;-

3.1.1 Regulatory Evaluation The NRC staff's review in the area of instrumentation and controls covers (1)the proposed plant-specific implementation of the feedwater flow measurement device and (2) the power uncertainty calculations (NRC RIS 2002-03, Attachment 1,Section I). The NRC staff's review is conducted to confirm that the licensee's application of Caldon Topical Report ER-80P, as supplemented by Caldon Topical Report ER-157P, is consistent with the NRC staff's approvals of these topical reports. The NRC staff also reviews the power uncertainty calculations to ensure that (1)the proposed uncertainty value of 0.34 percent correctly accounts for the uncertainties due to power level instrumentation error and (2) the calculations meet the relevant requirements of Appendix K to 10 CFR Part 50.

3.1.2 Technical Evaluation The generic bases for the proposed power uprate are provided in Caldon Topical Report ER-80P and its supplement, Topical Report ER-157P. These topical reports document the Caldon leading edge flowmeter check (LEFM /W) and LEFM check plus (LEFM ,,+TM) systems' abilities to achieve increased accuracy of flow and temperature measurement.

In its February 24, 2002, supplemental letter, the licensee submitted an uncertainty evaluation which evaluates the accuracy with which reactor core thermal power may be determined using the new flowmeter. The licensee asserts that the new flowmeter will be installed, calibrated, and maintained in accordance with the recommendations of Caldon. On the basis of the proposed installation and instrument application, the licensee anticipates a thermal power measurement uncertainty not in excess of 0.34 percent of RTP. This anticipated uncertainty limit is supported by testing of the LEFM in a piping geometry representative of the actual installed geometry, and will be reconfirmed during the commissioning process following installation. Therefore, the original 2-percent margin would be reduced to 0.34 percent, allowing for a power uprate of 1.66 percent (2 percent - 0.34 percent).

In the NRC safety evaluation reports that approved Caldon Topica Reports ERh80Pi nd ER-1 75P, the NRC requested that licensees address the following issues when applying for tde approval of an MUR power uprate: E

. Maintenance and calibration procedures that will be implemented with the incorporation of the the LEFM.

2. For plants that currently have LEFi installed, licensees should provide an evaluation of the operational and maintenanue history of the installation and confirm that the installed insteumentation is representative of the LEFI system and bounds the analysis andlt assumptions set forth in Topical Report ERe80.1

3. The methodology used to calculate the uncertainty of the LEFIV.

4. Where, the ultrasonic meter (including the LEFIV) was not installed with flow elements>

calibrated-to a site-specific piping configuration (flow profiles and meter factors not A,,--

representative of the plant-soecific installation), licensees should provide additional . -,

justification for use.v; -. L -L The licensee provided the information concemning ecofte above: issues in its applicto Land supple entJtT-hRC staff Eas reviewed the regulatory and technical analyses provided by-the licensee. -Hi L g Table 3.1.2 Instrumentation and Controls Unit 2 MUR Meets Criteria In Topic Application UFSAR NRC-approved Similar to NRC Staff Section Section Topical Reports Unit 1 MUR Conclusion ER-80P & ER-157P Compliance with 1.1, I.1.A - I.I.C n/a Y Y Acceptable Caldon Topical (pages 17 - 18) (References 3,4)

Reports -_ x _

Maintenance and I.1.D I.1.F, 1.1.G n/a yWl I yN&.

2 Acceptable Calibration (pages 18, 19, (Reference 2)

Procedures 22.25-26) Aid t Operational and 1.1, .1.D n/a iY" Acceptable A

Maintenance History (Criterion 2)t3 , (References 2, 3,4) of the LEFM (pages 17 - 19)

Installation I t

Methodology used to L.1.D n/a ysiel4 Acceptable calculate the (Criterion 3) (References 1, 2) uncertainty of the (pages 21 - 22)

LEFM system Ultrasonic Meter 1.1.D n/a y M cceptable Installation (Criterion 4) (References 2, 3, 4)

__ aaes 20 - 21) _

Tabe 31.2Refrences: A I.D C. Cook Units 1 and 2 License Amendment Nos. 148 and 134, dated August 27, 1990 [Approved the transition to Westinghouse 17x17 VANTAGE 5 fuel and the use of Westinghouse Licensing Topical Report)

< WCP-1397--Ag~evsedThemal esig Prcedues ated April 1989]

. C. Cook Unit 1 License Amendment No. 273, dated December 20, 2002 [Approved Measurement Uncertainty Recapture Power Uprate]

3. Letter from NRC, to C. L. Terry, TU Electric, Comanche Peak Steam Electric Station, Units 1 and 2 - Review of Caldon Engineering Topical Report ER 80P, 'Improving Thermal Power Accuracy and Plant Safety while Increasing Power Level Using the LEFM System' (TAC Nos. MA2298 and 2299), dated March 8, 1999

4. Letter from S. A. Richards, NRC, to M.A. Krupa, Entergy, Waterford Steam Electric Station, Unit 3; River Bend Station; and Grand Gulf Nuclear Station - Review of Caldon, Inc. Engineering Report ER-157P (TAC Nos. MB2397, MB2399 and MB2468), dated December 20, 2001 Table 3.1.2 Notes:

1. The maintenance and calibration procedures for the LEFM flow measurement system were addressed and found acceptable In the NRC staffs safety evaluation (SE) for D.C. Cook Unit I License Amendment No. 273.

The licensee will use the same maintenance and calibration procedures for the D. C. Unit 2 LEFM flow measurement system as those approved forD. C.Cook Unit 1.

2. Maintenance and calibration procedures will be developed as part of the implementation of the LEFM design change package specific for Unit 2. -

3. As noted in Section 3.1.2.2 of the NRC staffs SE for D. C. Cook Unit 1 License Amendment No. 273, the licensee has 'committed to confirm that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions In the Caldon Topical Report ER-80P.'

4. Methodology used to calculate the .uncertainty of the LEFM system for the proposed D. C. Cook Unit 2 MUR/

power uprate will be the same as that approved for D. C. Cook Unit 1. The licensee's overall statistica approach to combining uncertainties is in compliance with ANSI/ISA 67.04.01-2000, 'Set points forNula The NRC staff has reviewed the licensee's proposed plant-specific implementation of the feedwater flow measurement device and the power uncertainty calculations. The NRC staff concludes that the licensee's proposed application of Caldon Topical Report ER-80P, as supplemented by Caldon Topical Report ER-157P, is consistent with the NRC staff's approvals of these topical reports. The NRC staff also concludes that the licensee has adequately accounted for the uncertainties due to power level instrumentation error in their power level uncertainty calculations and demonstrated that the calculations meet the relevant requirements of 10 CFR Part 50, Appendix K. Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to instrumentation and controls.

3.2 Reactor Systems 3.2.1 Regulatory Evaluation The NRC staff review in the area of reactor systems covers the impact of the proposed MUR power uprate on (1) fuel design, (2) nuclear design, (3) thermal-hydraulic design, (4) performance of control and safety systems connected to the reactor and reactor coolant system, and (5) LOCA and non-LOCA transient analyses (NRC RIS 2002-03, Attachment 1, Sections II, III, and VI). The review is conducted to verify that the licensee's analyses bound plant operation at the loss-of-coolant accident MUR power level and that the results of the licensee's analyses related to the areas under review continue to meet the applicable acceptance criteria following implementation of the proposed MUR power uprate. Guidance

and acceptance criteria for the NRC staff's review of reactor systems are contained in '

Chapters 4, 5, 6, and 15 of NUREG-0800, Standard Review Plan for the Review of Safet Analysis Reports for Nuclear Power Plants LWR Edition.n 3.2.2 Technical Evaluation The NRC staff reviewed the licensee's application related to reactor systems performance and determined that existing analysis of record for many areas continue to bound operation of the plant at the proposed MUR power level. The results fthe NRC staffs review in the reactor svstems area are summarized in Table 3.2.2 belo o icneepe wTe hdm lew eidual )

rmvI (H)codwn analyses to support th~e proposed MUR power uprate becauseth existing analyses of record did not bound proposed plant operation. fThe NRC staff's review of ine licensee's RHR cooldown analyses is discussed in Section 3.2.2.1 of this SE. In addition, the NRC staff evaluated the impact of Several recent Westinghouse Nuclear Safety Advisory Letters (NSALs) on steam generator (SG) performance. This evaluation is provided in Section 3.2.2.2 below.,.

Table 3.2.2 Reactor Systems Unit 2 MUR Bounded by Topic Application UFSAR NRC-approved Similar to NRC Staff Section Section analysis Unit I MUR lConclusion Accidents and Transigents Analyses of Record Post-LOCA 111 1 : 14.3.1 j nYi Acceptable Long-Term Core (page 36) - (References 3. 4) Y Cooling : ______X______

Hot Leg Switchover III. 1. 3. 2 14.3.1 Y cceptable (page 37)- (References 3,4) Y _

SG Tube Rupture - 11.1.4 14.2.4 Y cceptable Thermal-Hydraulic (page 38) (References 2,5, 6)N ; yNotea Analysis : : - : ________.-

NonLOCA Analysis Single Reactor 11.3.6 14.1.6.2 - Y Wa4.s Acceptable Coolant Pump (page 46) (Reference 1)

Locked-Rotor Accident _ _ _ _ _

87 Loss of External 11.3.7 14.1.8 Yot 6 Acceptable Electrical Load - (page 47) (Reference 7)

Overpressure Analysis L; Loss of Normal 11.3.8 14.1.9 V V Acceptable Feedwater Flow and (page 47) 14.1.12 (Reference 1)

Loss of All AC Power Rupture of a Control 11.3.12 14.2.6 Y . Acceptable Rod Drive (page 49) (Reference 1)

Mechanism Housing RCCA Misalignment 11.3.1 14.1.3 Y . Acceptable and RCCA Drop (page 44) (Reference 1)

Table 3.2.2 Reactor Systems Unit 2 MUR Bounded by Topic Application UFSAR NRC-approved Similar to NRC Staff Section Section analysis Unit I MUR Conclusion Partial and 11.3.5 14.1.6.1 Y yNole Acceptable Complete Loss of (page 46) (Reference 1)

Forced Reactor Coolant Flow Uncontrolled RCCA 11.3.2 14.1.1 Y Y4o5 Acceptable Bank Withdrawal (page 44) (Reference 1) from a Subcritical Condition ;.

Chemical Volume 11.3.4 14.1.5W8 Y V Acceptable and Control System (page 46) (Reference 1)

Malfunction Excessive Heat 11.3.9 -14.1.10 .Y Oyne 5 7 Acceptable Removal Due to (page 47). (Reference 1)

Feedwater System Malfunctions .

Excessive Load 11.3.10 14.1.11 Y : Acceptable Increase Incident (page 48) (Reference 1)

Rupture of a Steam 11.3.11 14.2.5 VY . 5 Acceptable Pipe - Core (page 48) (Reference 1)

Response Analysis id;_.

Rupture of a Control 11.3.12 14.2.6 V y 7' Acceptable Rod Drive (page 49) .(Reference 1)

Mechanism Housing MODE 3 _ _f - _ ___f _ii Anticipated 11.3.13 3.3.1.7 Y Y V Acceptable Transients Without (page 49) (References 8, 9):

SCRAM Station Blackout 11.3.14 8.7 : V Acceptable (page 51) (References 10, 11)

Design Transients 11.4.1 4.1 9 YVO Acceptable (page 52) (References 1, 12)

Auxiliary Equipment 11.4.2 4.1 Y-We9 Acceptable Design Transients (page 54) (References 1, 12)

Feedwater System 11.3.9 14.1.10 NNo 6,7 Acceptable Malfunctions (page 47) (Reference 1)

(full-power case)

Loss of External 11.3.7 14.1.8 V No" [ 7 Acceptable Electrical Load - (page 47) (Reference 7)

DNB Case 9Uncontrolled RCCA Bank Withdrawal at Power 11.3.3 (page 45)

_____________ £ 14.1.2 I

Y (Reference 1)

Nbote6 I ___________

Acceptable

Table 3.2.2 Reactor Systems _

Unit 2 MUR Bounded by Topic lApplication UFSAR NRC-approved Similar to NRC Staff Section Section analysis Unit I MUR Conclusion Fuel Evaluation Nuclear Design IV.8.1 3.3 y yNO1 1o Acceptable (page 79) (References 1, 12, 13)

Fuel Rod Design IV.8.2 3.2.1 Y Y Acceptable (page 80) (References 1, 12, 13)

Core IV.8.3 .3.4 Y Y Acceptable Thermal-Hydraulic (page 80) (References 1,12, 13)

Design Fuel Structural IV.8.4 3.2.1 Y Y Acceptable Evaluation (page 81) _____ (References 1, 12,. 13) _____ _____

System Design _

RHR System VI.11.3 9.3 N t e1; NNOw11 Acceptable (page 87) (References 12, 14)- See Section 3.2.2.1

__________ _ X]Below)

Emergency Core VI.1.4 6.2 Y Y Acceptable Cooling System (page 88) (References 12,13) t NSSS Control VI.5 7.3 yNo el 2

Y Acceptable Systems (pages 95-98) (Reference 12)

NSSS Pressure VI.1 4.2.2.2.4.3.4 Y Y Acceptable Control Component (page 86) (References 12, 13) i___ ___

Sizing Low Temperature VI.5 4.2,4.2.2.8 Y Y Acceptable Overpressure (page 98) (References 13, 14, 15)

Protection Svstem _ _ .

Table 3.2.2

References:

1. D. C. Cook Units 1 and 2 License Amendment Nos.148 and 134, dated August 27,1990 [Approved the transition to Westinghouse 17x17 VANTAGE 5 fuel and the use of Westinghouse Lcensing Topical Report WCAP-1 1397-P-A, 'Revised Thermal Design Procedure,' dated April 1989]

2. D. C. Cook Unit 2 License Amendment No. 135, dated September 18, 1990 [Allowed Unit 2 SG stop valve closure within 8 seconds]

3. D. C. Cook Units I and 2 License Amendment Nos. 234 and 217 [Approved containment sump modification, as evaluated in Westinghouse Licensing Topical Report WCAP-15302, 'Donald C. Cook Nuclear Plant Units 1 and 2, Modifications to the Containment Systems, Westinghouse Safety Evaluation (SECL 99-076, Revision 3),

dated September 1999]

4. D. C. Cook Units 1 and 2 License Amendment Nos. 236 and 218, dated December 23,1999 [Rod cluster control assembly insertion credit following a large-break LOCA (LBLOCA)_

5 D.C.Cook Units 1and 2 License Amendment Nos. 256 and 239, dated October 24, 2001 [Analyses to address SG tube rupture overfill]

6. D. C. Cook Units I and 2 License Amendment Nos. 271 and 252, dated November 14,2002 [Alternative source term for control room habitability]

7. D. C.Cook Units I and 2 License Amendment Nos. 182 and 167, dated September 9, 1994 [Approved increase in main steam safety valve setpoint tolerances]

8. Letter from J. F. Stang, NRC, to M. P. Alexich, I&M, 'Donald C. Cook Nuclear Plant Nos. Units I and 2, Compliance with ATWS [Anticipated Transient Without Scram] Rule 10 CFR 50.62 (TAC Nos. 59082 and 59083),' dated April 14, 1989

9. Letter from J. Glitter, NRC, to M. P. Alexich, l&M, 'Safety Evaluation for Generic Letter 83-28, Item 4.5.3, Reactor Trip Reliability - On-Line Functional Testing of the Reactor Trip System (TAC Nos. 53971 and 53972),'

dated August 16, 1989

10. Letter from T. G. Colbum, NRC, to E. E. Fitzpatrick, I&M, 'Station Blackout Analysis, Donald C. Cook Nuclear Plant, Units I and 2 (TAC Nos. 68532168533),' dated October 31, 1991

11. Letter from J. F.Stang, NRC, to E. E. Fitzpatrick, l&M, 'Station Blackout Analysis, Donald C. Cook Nuclear Plant, Units 1and 2 (TAC Nos. 68532 and 68533),' dated April 23, 1992

12. Safety Evaluation Report, 'Safety Evaluation by the Directorate of Licensing U.S. Atomic Energy Commission in the Matter of Indiana & Michigan Electric Company and Indiana & Michigan Power Company Donald C. Cook Nuclear Plant- Units 1 and 2, Docket Nos. 50-315 and 50-0316, dated September 10, 1973

13. Letter from NRC to Indiana and Michigan Electric Company, 'Supplement 7 to Safety Evaluation Report,' dated December 23, 1977

14. D. C.Cook Units I and 2 License Amendment Nos. 219 and 203, dated December 10, 1997 [Approved changes to RHR automatic interlock surveillance requirements]

15. D. C.Cook Units I and 2 License Amendment Nos. 176 and 161, dated March 9, 1994 [Power-Operated Relief Valve and Block Valve Reliability, and Additional Low-Temperature Overpressure Protection in Response to NRC Generic Letter 90-06]

Table 3.2.2 Notes

1. Parameters for the proposed MUR power uprate remain bounded due to "trimming' the endpoints of the full-power T,, range. This approach was performed for both D. C. Cook Unit 1 and Unit 2 MUR power uprates.

2. D. C. Cook Unit 2 overtemperature delta T/overpower delta T setpoint coefficients remain unchanged for the Unit 2 MUR, whereas the D. C. Cook Unit 1 setpoint coefficients required restrictions to be applied to support the proposed MUR power uprate.

3. D.C. Cook Unit 2 steam generator tube rupture (SGTR) overfill analysis was performed at a core power of 3588 MWt, which bounds the proposed MUR power uprate; whereas, the Unit 1 SGTR overfill analysis was performed at 3250 MWt, and required a sensitivity analysis.

4. References 2 and 6 of the Table 3.2.2 above addressed radiological consequences of an SGTR; Reference 5 of Table 3.2.2 above approved the supplemental SGTR analysis.

5. The D. C.Cook Unit 2 analysis was performed with a core power of 3588 MWt, which bounds the proposed MUR power uprate conditions. For Unit 1, an evaluation of the DNB cases of this event was required.

6. The D. C. Cook Unit 2 analysis was performed with a core power of 3588 MWt, which bounds the proposed MUR power uprate conditions. The Unit 1 MUR power uprate required reanalysis of this event.

7. For the proposed D.C. Cook Unit 2 MUR power uprate, the analyses are bounding at the core power level o 3588 MWt, so each accident analysis is evaluated in one section. However, for the D. C. Cook Unit 1 MUR

power uprate, several accident analyses were divided into more than one section to clarify where certain cases were either evaluated differently, or reevaluated.

8. D.C. Cook Unit 2 UFSAR Section 14.1.5 is entitled, 'Uncontrolled Boron Dilution," whereas the D.C. Cook Unit I UFSAR Section 14.1.5 is entitled, 'Chemical and Volume Control System Malfunction."

9. The design transients for D. C. Cook Unit 2 were last evaluated for fuel Cycle 8 in the SE for D.C. Cook Unit I License Amendment No. 134, dated August 27, 1990, which approved the use of Westinghouse 17 x 17 VANTAGE 5 fuel.

10. The licensee proposed implementation for the D. C. Cook Unit 2 MUR power uprate at the beginning of core operating Cycle 14 (spring 2003), whereas the Unit I MUR was Implemented in mid-cycle.

11. The licensee re-performed the RHR cooldown analysis to support the proposed D.C. Cook Unit 2 MUR power uprate. The revised analysis, which considers a change to the plant's RTP only, demonstrates that the licensee will still be able to reach Mode 5 conditions within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months on a single train of RHR, and the time to cool down to <140 OF with two trains of RHR available has increased from less than 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months to less than 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months .

(See Section 3.2.2.1 below) For D.C. Cook Unit 1, the single-train cooldown analysis demonstrated that the plant would be able to reach Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months andtthe two-train analysis already assumed a bounding Initial power level of 3411 MWt.

12. The licensee is in the process of conducting steam dumplmargin-to-trip final analyses for D. C.Cook Unit 2.

(See Section 3.2.2.4 below) 3.2.2.1 RHR Cooldown Various D. C. Cook Unit 2 TSs require that the plant be capable of being placed in cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . In addition, the current licensing basis states that under normal operating conditions, the RHR system is capable of reducing RCS temperature to 140 °F within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months following a reactor shutdown. The licensee re-performed the RHR cooldown analysis for the single and two-train scenarios since the current analyses assumed a core power level of 3411 MWt. The licensee's reanalysis used the same input assumptions, methodology, and technique as the current analysis, with the exception of the core power level assumptions.

For the reanalysis, the licensee used a core power level of 3482 MWt, which bounds the proposed MUR power uprate level of 3468 MWt.

The licensee's reanalysis showed that for a single-train cooldown, the TS requirement of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is met. The results of the dual train cooldown demonstrated that the plant could be cooled down to 140 OF within 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months , which exceeds the 20-hour value currently reflected in the plant's current licensing basis. The 20-hour cooldown time for dual-train operation is based on economic considerations only (i.e., balancinglthe time required for cooldown against the size and cost of RHR and component cooling water system components, such as heat exchangers, Since the reanalysis bounds the proposed power level of 3468 MWt, the TS requirement of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is satisfied for the single-train cooldown, and the new dual-train cooldown time results will be incorporated in the UFSAR, the NRC staff finds the RHR system acceptable for operation at the proposed 3468 MWt power level.

3.2.2.2 Steam Generators The Westinghouse Model 51 designed SGs originally installed in D. C. Cook Unit 2 were modified in 1988. Specifically, the lower assembly (including the tube bundle) was replaced

with those of a Model 54F design while the upper shell and inmrnals remained the original Model 51 design with upgraded internals. The modified SGs have been analyzed to design specifications for 3425 M~Wt and 3600 MWt NSSS power operating conditions. The licensee performed a comparison of the applicable MUR power uprate design transient set to the set of values evaluated for the modified SGs 3600 MWt operating condition.

Westinghouse issued three NSALs (NSAL-02-3 and Revision 1, NSAL-02-4, and NSAL-02-5) to document potential problems with the Westinghouse-designed SG water level setpoint uncertainties. NSAL-02-3 and its revision, dated February 15 and April 8,2002, respectively, deal with the uncertainties caused by the mid-deck plate located between the upper and lower taps used for SG water level measurements. These uncertainties affect the low-low level trip setpoint. NSAL-02-4, dated February 19, 2002, deals with a potential indication inaccuracy with the SG water level high-high trip setpoint for water levels above the SG mid-deck plate.

NSAL-02-5, dated February 19, 2002, involves the potential effects of the pressure differential across the SG mid-deck plate, with the focus on the potential impact to the initial SG water level modeled in the accident analyses-due to increased water level uncertainty.

D. C. Cook Unit 2 SGs were affected by the issue identified in NSAL-02-03. The licensee performed an evaluation of this condition at the proposed uprated power level and determined that adequate margin is available in the SG water level low-low trip setpoint calculation to accommodate the effects of a differential pressure across the SG mid-deck plate. The licensee determined that the existing calculation' bounds the issue identified by NSAL-02-3 and the proposed MUR conditions and found the SG water level low-low trip setpoint remains unaffected. Consequently, there is no effect on 'the setpoint values used in the analyses of record for the LOCA, non-LOCA transients, and the anticipated transient without scram event.

The licensee determined that the D.: C. Cook Unit 2 water level low-tow trip setpoint would be reached before the SG water level would reach the mid-deck plate level. Thus, the indication inaccuracy for water levels above the mid-deck plate is not of concern for D. -C. Cook Unit 2, and the existing SG water level high-high trip setpoint remains acceptable. There is no effect on the setpoint values used in the analyses of record, and the current analyses remain conservative.

The evaluation of the NSAL-02-4 issue by the licensee determined that the Unit 2 trip setpoint would be reached before the SG water level would reach the mid-deck plate level. The indication inaccuracy for water levels above the mid-deck plate is not of concern for Unit 2, and the existing SG water level high-high trip setpoint remains acceptable. Thus, there is no effect on the setpoint values used in the analyses of record, and the current analyses remain conservative.

The NSAL-02-5 issue pertained to the potential impact to the initial SG water level modeled in the accident analyses due to increased water level uncertainty. The increased uncertainty is a possible result of postulated pressure differential effects across the SG mid-deck plate. The specific accident analyses of interest are (1) loss of normal feedwater/loss of all AC power to the station auxiliaries, (2)feedwater malfunction, (3) feedline break, (4)steamline break mass and energy release calculations, and (5)LOCA mass and energy release calculations. The licensee performed an evaluation of the postulated condition and determined, in all cases, that the conclusions of the current analyses remain applicable and bounding due to existing available margin. The licensee found that the current analyses of record continue to remain

bounding. Thus, the current analyses remain conservative and support the Unit 2 MUR powe) uprate., -_ _ _ _ _ _ _ _

The NRC staff reviewed the licensee's evaluations of the NSALs discussed above and finds them acceptable. The NRC staff finds that the current analyses remain conservative with respect to the proposed D. C. Cook Unit 2 MUR power uprate. The NRC staff condudes that the SG water level issues are adequately addressed for the uprated power.

""3.2.2.3 Changes to Technical Speacaion The existing analysis of record supporting the maximum allowable core power level wit a safety injection crossfpie valve closed (TS 3.5.2,pAction b) was performed for a nominal p e level of 3250 MWt and a power level uncertainty of 2 percent. The licensee's November 15 2002, application justifies a reduction in the power level uncertainty from 2 percent to 0.34 percent. As a result, the licensee Vproposed to lincrease the maximum allowable core power level in TS 3.5.2, Action bmrby 1.66 percent (L~b., the difference between the original assumption of 2 percent uncertainty and the proposedbtlue of 0.34 percent uncertahis Se, licensee's proposed change would result in an increaseof the maximum allowable core power level in TS 3.5.2, Action b from 3250 MWt to 3304 MWt.rFasedxon (1) the NRC stwt acceptance of the newvvaauesof 0.34 percent for total power uncertainty (See Section 3.1 above), (2) the fact that the existing analysis of record accounted for 2-percent uncertainty, and (3)the fact that this change merely recovers the difference between the 2 percent assumed in the analysis of record and the0.34 percent accepted by the NRC staff in Section 3.1 of this SE, the NRC staff finds the proposed change acceptable.

InS Tble3.7 1 F"Maximum Allowable Power Range Neutron Flux High Setpoint with Inoperable Steam Line Safety Valves During 4 opOperation the licensee proposed t insertion of new values for the setpoints with inoperable tsteamline safety valves to be consistent with the proposed power uprate. For; D. C. Cook Unit 2, with one, two, and three\

steamline safety valves inoperable, the licensee proposed to change the maximum allowabl I power levels from 61.'6 percent, 43.9 percent,; and 26.2 percent to 60.4 percent, 43.0 percent,\

and 25.7 percent, respectively. To calculate these values for the cproposed uprated power level, the licensee used the conservative heat balance calculation des.nbed in TS Ba Section 3/4-7-1-1-r - -i Since the licensee used a conservaier heat balance calculation to determine the new power range neutron flux high setpoints, the NRC staff finds them acceptable for the proposed power uprate to 3468 MWt.

3.2.2.4 Conmites I To support the proposed D. C. Cook Unit 2 MUR power uprate, the licensee made the following) commitments:

APrior to implementing this uprate, a engineering/reload safety evaluation will be (

performed to ensure that the core design bounds the uprated condition. The UFSA twill be updated to reflect the safety evaluation. w-

Prior to implementing this uprate, the licensee will perform an analysis of the steamA dump valve flow capacity at the uprated power level and implement changes/

adjustments as required to ensure the valves have sufficient capacity. The UFSAR will

_be updated to reflect the analysis and/or commitments. -

Both of these commitments will be incorporated into Facility Operating License DPR-74 as license conditions.

3.2.3 Summary The NRC staff has reviewed the licensee's safety analyses of the impact of the proposed MUR power uprate on (1)fuel design, (2) nuclear design, (3) thermal-hydraulic design, (4) performance of control and safety systems connected to the NSSS, and (5) LOCA and non-LOCA transient analyses. The NRC staff concludes that the results of licensee's analyses related to these areas continue to meet the applicable acceptance criteria following implementation of the proposed MUR power uprate. Where additional evaluations/analyses were necessary, the NRC staff has reviewed these evaluations and analyses and finds that the licensee has satisfactorily addressed the areas discussed above, the supporting safety analyses were performed using NRC-approved methods, the input parameters of the analyses adequately represent the plant conditions at the proposed uprated power level, and the analytical results meet the applicable acceptance criteria. Based on the above, the NRC staff finds the proposed MUR 1.66-percent power uprate acceptable with respect reactor systems performance.'

3.3 Electrical Sstm 3.3.1 Regulatory Evaluation The NRC staff review in the area of electrical engineering covers the Impact of the proposed MUR power uprate on (1)grid stability, including performance of the main generator, main transformer, isophase bus, and unit auxiliary transformer/reserve auxiliary transformer, (2)emergency diesel generator loading, (3) station blackout, and (4)environmental qualification of electrical equipment (NRC RIS 2002-03, Attachment 1,Section V). This review is conducted to verify that the results of licensee analyses related to these areas continue to meet the requirements of 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 17, 10 CFR 50.63, and 10 CFR 50.49 following implementation of the proposed MUR power uprate.

3.3.2 Technical Evaluation The NRC staff has reviewed the licensee's application in relation to electrical system performance and determined that existing analyses of record for electrical systems bound the proposed operation of the plant at the uprated power level. The results of the NRC staff's review in the electrical engineering area are summarized in Table 3.3.2 below.

Table 3.3.2 Electrical Systems Unit 2 MUR Bounded by Application UFSAR NRC-approved Similar to NRC Staff Topic Section Section analysis Unit 1 MUR Conclusion Grid Stability V 8.5 Y Y Acceptable (page 85) (References 1, 2)

Main Generator V 8.0 . Yod l Acceptable (pages 83, 84) 10.3 (References 1,2, 3)

Main Transformer V 8.2 V Y Acceptable (page 84) (References 1, 2, 4, 5)

. .9 .

Isophase Bus VI.4, VII.3 8.1.2 ' Not. Acceptable (pages 95, 101) 10.7 (References 1. 2)

Unit Auxiliary Table V-1 8.0 Y V Acceptable Transformer I (page 83) 8.1.2 (References 1, 2, 4, 5)

Reserve Auxiliary Transformer , ____X______

Emergency Diesel V 8.5 V y* Y Acceptable Generators (page 85) 9.8.3 (References 1, 2, 6)

Station Blackout 11.3.14, V 8.7 Y Y Acceptable (page 51, 86) (References 3, 7)

Environmental V, V16.1 14.4 VyNoa2 Y Acceptable Qualification of (page 86) (References 1, 8)

Electrical Equipment Mi0 Table 3.3.2

References:

1. Safety Evaluation Report, 'Safety Evaluation by the Directorate of Licensing U. S. Atomic Energy Commission in the Matter of Indiana & Michigan Electric Company and Indiana & Michigan Power Company Donald C. Cook Nuclear Plant- Units I and 2, Docket Nos. 50-315 and 50-316,' dated September 10, 1973

2. Letter from NRC to Indiana and Michigan Electric Company, 'Supplement 7 to Safety Evaluation Report," dated December 23, 1977

3. Letter from T. G. Colburn, NRC, to E. E. Fitzpatrick, l&M, 'Station Blackout Analysis, Donald C. Cook Nuclear Plant, Units I and 2 (TAC Nos. 68532168533)," dated October 31, 1991

4. D. C. Cook Unit 2 License Amendment No. 22, dated July 10, 1980 [Approved changes to surveillance and monitoring requirements for degraded voltage]

5. D. C. Cook Units I and 2 License Amendment Nos. 137 and 124, dated May 25,1990 [Approved changes to allowable values for 4KV bus degraded voltage]

6. D. C. Cook Units I and 2 License Amendment Nos. 214 and 199, dated March 13, 1997 [Approved an increase

7. Letter from J. F. Stang, NRC, to E. E. Fitzpatrick, I&M, Station Blackout Analysis, Donald C. Cook Nuclear Plant, Units 1 and 2 (TAC Nos. M68532 and 68533),' dated April 23, 1992

8. Letter from S. A. Varga, NRC, to J. Dolan, I&M, Safety Evaluation Regarding Environmental Qualification of Electric Equipment Important to Safety,' dated January 11, 1985 Table 3.3.2 Notes:

1. Turbine Auxiliary Cooling Water (TACW) has been determined to have adequate margin to support power uprate requirements. However, similar to the D. C. Cook Unit I MUR power uprate evaluation, TACW flow to the iso-phase bus duct cooling system and stator water coolers will be monitored and adjusted during post-modification system operation to accommodate additional heat generated at the uprated power level.

2 The environmental qualification of electrical equipment is based on the results of accident analyses which assumed oore power levels that have been adjusted Ta~r a 2-percent calorimetric uncertainty.

3.3.3 Summary The NRC staff has reviewed the licensee's safety analyses of the impact of the proposed MUR power uprate on (1) grid stability, including performance of the main generator, main transformer, isophase bus, and unit auxiliary transformer/reserve auxiliary transformer, (2) emergency diesel generators, (3) station blackout, and (4) environmental qualification of electrical equipment. The NRC staff concludes that the results of licensee's analyses related to these areas continue to meet the requirements of 10 CFR Part 50, Appendix A, GDC-17, 10 CFR 50.63, and 10 CFR 50.49 following implementation of the proposed MUR power uprate.

Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to electrical engineering.

T3.4 Mechanical and Civil Engineerina 3.4.1 Regulatory Evaluation The NRC staff review in the area of mechanical and civil engineering covers the structural and pressure boundary integrity of NSSS and balance-of-plant (BOP) systems and components (NRC RIS 2002-03, Attachment 1,Section IV, Items 1.A, 1.B, and t.D). The NRC staff's review focuses on the impact of the proposed MUR power uprate on NSSS piping, components, and supports; BOP piping, components, and supports; reactor vessel (RV) and supports; control rod drive mechanism; SG and supports. reactor coolant pumps and supports; pressurizer and supports; reactor pressure vessel and supports, reactor internals and core supports; and safety-related valves. Technical areas covered by this review include stresses, cumulative usage factors, flow-induced vibration, high-energy line break locations, jet impingement and thrust forces, and safety-related valve programs. The review is conducted to confirm that (1) the results of the analyses continue to meet code allowable limits of the American Sodety of Mechanical Engineers (ASME) code of record for the plant, (2) the safety-related valves will continue to perform acceptably, and (3) the safety-related valve programs will continue to be adequate. The NRC staff's review is performed to verify compliance with 10 CFR 50.55a and 10 CFR Part 50, Appendix A, GDCs-1, 2, 4, 10, 14, 15, 30, 37, 40, 43, 46, and 54 following implementation of the proposed MUR power uprate.

F3.4.2 Technical Evaluation 3 The NRC staff has reviewed the licensee's application as related to the mechanical and civil engineering areas discussed above and determined that existing analyses of record bound plant operation at the proposed uprated power level. The results of the NRC staff's review in the civil and mechanical engineering area are summarized in Table 3.4.2 below.

I. Topic Unit 2 MUR Application Section Table 3.4.2 Civil and Mechanical Engineering UFSAR Section Bounded by NRC-approved analysis Similar to Unit I MUR NRC Staff Conclusion RV Structural IV.1, IV.1.1 4.2.2.1 yNIAcceptable Y Evaluation (pages 56, 57) 4.4 (References 1, 2)

Reactor Internals IV.1.2 3.2.2 Y Y Acceptable (page 57) 4.2.2.1 (References 1, 2)

Piping and Supports IV.2 4.2.2.7 - V Acceptable (page 62) ii 4.2.2.9 (References 1, 2)

Control Rod Drive IV.3 3.2.3.1.4. V Y Acceptable Mechanisms (page 64) - 3.2.3.2.2 (References 1, 2)

Reactor Coolant iV.4 4.2.2.5 Y Y Acceptable Pumps and Motors (page 65) (References 1,2)

SGs IV.5.2, IV.5.4 4.2.2.4 D Y Acceptable (pages 69,72) (References 1,2)

Pressurizer. IV.6 4.2.2.2 Y Y Acceptable (page 78) (References 1, 2)

NSSS Auxiliary IV.7 4.2.2.3 y Y Acceptable Equipment (page 79) 4.2.2.8 (References 1, 2)

__ fChapter

_ _ _ 9 L Balance of Plant Main Steam System Vl.2.1 10.2 *.Y Y Acceptable (page 89-90) - (References 1, 2)

Steam Dump VI.2.1 7.3.2 yNoO 2 V Acceptable System (page 89-90) 10.2 (References 1, 2)

Condensate and VI.2.2 10.5.1 V V Acceptable Feedwater System (page 90-91) (References 1, 2)

Auxiliary Feedwater VI.2.3 10.5.2 V V Acceptable System (page 91-92) (References 1, 3)

SG Blowdown VI.2.5 10.11 V Y Acceptable System (page 92-93) (References 1, 2)

<'IO

(

Table 3.4.2 Civil and Mechanical Engineering Unit 2 MUR l Bounded by Application FSAR NRC-approved Similar to NRC Staff Topic Section Section l analysis Unit I MUR Conclusion Programs High-Energy Line V-.6.5 5.2.2.7 Acceptable Break Program (page 106) 14A.11.2 (References 4, 5)l Motor-Operated VII.6.2 8.1.2 MyNote 4, 5 Acceptable Valve Program 3 (page 103-104) (References 6, 7, 8)

Air and Hydraulic Vl1.6.3 nl/a !-^~ Y N6Acetal Operated Valve (page 104) ax ;--:(Reference 6)

Program Xi Table 3.4.2

References:

1. Safety Evaluation Report, "Safety Evaluation by the Directorate of Licensing U. S. Atomic Energy Commission in the Matter of Indiana & Michigan Electric Company and Indiana & Michigan Power Company Donald C. Cook Nuclear Plant - Units I and 2, Docket Nos. 50-315 and 50-0316,' dated September 10, 1973

2. Letter from NRC to Indiana and Michigan Electric Company, 'Supplement 7 to Safety Evaluation Report,' dated December.23, 1977

3. D. C. Cook Units I and 2 License Amendment Nos. 214 and 199, dated March 13,1997 [Approved an increase in SG plugging limit) -

4. D. C. Cook Units I and 2 Ucense Amendment Nos. 244 and 225,' dated April 25, 2000 [Approved modification to turbine-driven auxiliary feedwater pumpomm cooler plant)

5. D. C. Cook Units I and 2 License Amendment Nos. 249 and 230,' dated November 21, 2000 (Approved changes for high-energy line break methodology)

6. D. C. Cook Unit I License Amendment No. 273, dated December 20,2002 [Approved Measurement Uncertainty Recapture Power Upratel

7. Letter from M. W. Rencheck, I&M, to NRC Document Control Desk, 'Donald C. Cook Nuclear Plant Units I and 2 Completion of Generic Letter (GL) 88-10 Motor-Operated Valve (MOV) Program Implementation and Description of Generic Letter 96-05 MOV Periodic Verification Program, (C1200-09[,' dated December 15, 2000

8. Letter from J. F. Stang, NRC, to R. P. Powers, I&M, 1Donald C. Cook Nuclear Plant, Units 1 and 2- Closeout of Licensing Action for Generic Letter 96-05, "Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves (TAC Nos. M97037 and M97038)' dated August 8, 2001 Table 3.4.2 Notes:

1. The operating envelope (pressure-temperature (P-T)) evaluated for the D. C. Cook Unit 2 MUR power uprate is consistent with the envelope evaluated for fuel Cycle 8 (D. C. Cook License Amendment No. 134, dated August 27, 1990). Therefore, the RV structural analyses and evaluations that demonstrate compliance with applicable limits of Section III of the ASME Boiler and Pressure Vessel Code remain valid.

2. The licensee is in the process of conducting steam dump/margin-to-trip final analyses for D. C. Cook Unit 2.

(See Attachment 5, 'Regulatory Commitments,' of November 15, 2002, application.)

3. A description of the D. C. Cook MOV Program was provided to the NRC in a letter dated December 15, 2000 (Reference 7 of Table 3.4.2 above).

4. Impacts to the D. C. Cook MOV Program were addressed in the SE for D. C. Cook Unit 1 License Amendment No. 273 (Reference 6 of Table 3A.2 above). This program is common to both D.C. Cook Unit 1 and Unit 2.

5. Reference 8 of Table 3.4.2 above is the NRC's closeout document for the MOV Program (GL 96-05), which documents the acceptance of the D. C. Cook MOV Program, based on NRC review andlor inspection.

6. Impacts to D.C. Cook Air and Hydraulic-Operated Valve Program were first addressed in the SE for D.C. Cook Unit 1 License Amendment No. 273 (Reference 6 of Table 3.4.2 above). This program is common to both C. Cook Unit 1 and Unit 2.

D.

3.4.3 Summary The NRC staff has reviewed the licensee's evaluation of the impact of the proposed MUR power uprate on NSSS and BOP systems and components with regard to stresses, cumulative usage factors, flow induced vibration, high-energy-line break locations, jet impingement and thrust forces, and safety-related valve programs and concludes that the these areas will continue to be acceptable following implementation of the proposed MUR power uprate. Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to the areas of civil and mechanical engineering.

3.5 Dose Consequences Analysis )

3.5.1 Reaulatory Evaluation The NRC staff review covers the impact of the proposed MUR power uprate on the results of dose consequence analyses (NRC RIS 2002-03, Attachment 1, Sections II and l1l). The review is conducted to verify that the results of the licensee's dose consequence analyses continue to meet the acceptance criteria in 10 C FR Part 100, 110 CFR 50.67, and/or 10 CFR Part 50, Appendix A, GDC-19, as applicable, following implementation of the proposed MUR power uprate.

[3.5.2 Technical Evaluation The NRC staff reviewed the impact of the proposed MUR power uprate changes on design-basis accident (DBA) radiological anal=ses as documented in Chapter 14 of the D. C. Cook UFSAR.

In its November 15, 2002, application, the licensee stated that the current radiological analyses of record for D. C. Cook Unit 2 were unaffected by the proposed power uprate because they were performed assuming a nominal core wer of 3588 MWt, which bounds the conditions for he pros 16 e power uprate. Usin t ok UFSAR documentation in addition to information in the Novem er 15, 2002, application, the NRC staff verified that the existing D. C. Cook Unit 2 UFSAR Chapter 14 radiological analyses source term and steam release assumptions, as appropriate, bound the proposed 1.66-percent power uprate conditions for analyses of the offsite radiological consequences of DBAs.

By D. C. Cook Units 1and 2 License Amendment Nos. 271 and 252, dated November 14, 2002, the NRC staff approved selective implementation of an alternative source term in accordance with 10 CFR 50.67. These amendments addressed control room dose only. In the analyses for these amendments, the licensee assumed a core power level of 102 percent of 3588 MWt (or 3660 MWt) for the revised -- analyses, which bounds the conditions for the proposed 1.66-percent

I power uprate for D. C. Cook Unit 2 for control room doses. The NRC staff found these analyses to be acceptable, as stated in the SE for D. C. Cook Ucense Amendment Nos. 271 and 252. ]

I 3.5.3 Summary The NRC staff has reviewed the licensee's assessment of the impact of the proposed MUR power uprate on dose consequence analyses. The NRC staff concludes that the results of licensee's analyses related to these areas continue to meet the applicable acceptance criteria following implementation of the proposed MUR power uprate. Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to dose consequence analyses.

3j..6 Materials and Chemical EngineerinqJ9g"'

3.6.1 Regulatory Evaluation The NRC staff review in the area of Materials and chemical engineering covers the effects that the proposed MUR power uprate will have on (1) the structural integrity evaluations for the RV, (2) SG tube integrity, and (3) erosion-corrosion programs (NRC RIS 2002-03, Attachment 1-Section IV, Items 1.C through 1.F). The NRC staff's review in this area focuses on the impact of proposed MUR power uprate on (1) the P-T limits for the RV and reactor coolant pressure boundary, (2) evaluations for ensuring the integrity of the RV and reactor coolant pressure boundary against pressurized thermal shock (PTS), (3) evaluations for ensuring that the RV materials have sufficient levels of upper-shelf energy (USE), (4)surveillance capsule withdrawal schedules, (5) licensee programs for addressing SG tube degradation mechanisms, and (6) erosion/corrosion. This review is conducted to verify that the results of licensee analyses related to these areas continue to meet the requirements of 10 CFR 50.60, -

10 CFR 50.61, and 10 CFR 50.55a;t 10 CFR Part 50, Appendix A, GDCs-1, 4, 14, 31; and 10 CFR Part 50, Appendices G and H,following implementation of the proposed MUR power uprate.

[ 3.6.2 Technical Evaluation)} 0at i; The NRC staff has reviewed the licensee's application as related to the material and chemical engineering areas discussed above and determined that, with the exception of the structural integrity evaluations for PTS and RV USE, the existing analyses of record bound proposed operation of the plant at the uprated power level. The NRC staffs evaluation of the effects of the proposed MUR power uprate on the PTS and RV USE analyses is given in Section 3.6.2.1 of this SE. The results of the NRC staff's review for the remaining areas within the materials and chemical engineering scope are summarized in Table 3.6.2 below.

Table 3.6.2 l_ Materials and Chemical Engineering Unit 2 MUR Bounded by Application UFSAR NRC-approved Similar to lNRC Staff Topic Section Section analysis Unit I MUR oncluslon SG Tube Vibration and IV.5.4 -Wa'm Y y"ie Acceptable Wear and Other Modes (pages 72-76) (References 2, 3) of Tube Degradation Regulatory Guide 1.121 IV.5.5 n/aN01 I nla y**t03 Acceptable Analysis (pages 76-78)

Flow-Accelerated VII.6.4 4 Y y /a 5 Acceptable Corrosion (pages 104-106) _ (References 4, 5)

Structural Integrity and Metallurgy 10 CFR Part 50 IV.1.1 4.2.5 yNfte6 yolae6 . Acceptable Appendix G - P-T 4.2.6 (Reference 6)

Limits 4.4.1

1. '.I 10 CFR Part 50 IV.1.1 3.3.2.8 N 7l6 Acceptable Appendix G - USEI 4.2.2.8 (See Section 13.62.1 below 10 CFR 50.61 PTS Enclosure 2, 3.3.2.8 N - g. AAcceptable Events Section 5.2 See Section (4.2.2.8 4.4.2 3.6.2.1 below 14.3.7 _ X _ _

10CFR Part 50 IV.1.1 4.5.1.1 e A cceptable Appendix H RPV (Reference 1)

Surveillance Program Leak-Before-Break t IV.2.3 4.3.1 Y Y ptable Analyses 5.2.2.7 (References 7, 8) 6.1 14.3.3.1 14.3.3.4

_ __ _ 14.3.3.6 Structural Integrity of IV.3 3.2.3.1.4 Y. V Acceptable Control Rod Drive 3.2.3.2.2 (References 1,2)

Mechanisms Nozzles 4.3.1 14.3.3 Structural Integrity of IV.1.2 3.2.2 Y V Acceptable RV Internals 4.2.2.1 (References 1, 2,9)

Structural Integrity of IV.4 4.2.2.5 Y Y Acceptable the Reactor Coolant (References 1,2)

Pump Flywheels Structural Integrity of IV.1.2, IV.2.3, 3.2.2 cceptable Other Class 1 Reactor IV.3, IV.4 3.2.3.1.4 (References 1,2,7.8)

Coolant System 3.2.3.2.2 Components 4.2.2.5, 4.3 14.3.3..

Table 3.6.2

References:

1. Safety Evaluation Report, Safety Evaluation by the Directorate of Licensing U.S. Atomic Energy Commission in the Matter of Indiana & Michigan Electric Company and Indiana & Michigan Power Company Donald C. Cook i Nuclear Plant - Units 1 and 2, Docket Nos.50-315 and 50-0316," dated September 10,1973

2. Letter from NRC to Indiana and Michigan Electric Company, 'Supplement 7 to Safety Evaluation Report,' dated December 23, 1977 I I

3. D. C. Cook Unit 2 License Amendment No. 100, dated March 8,1988 [Approved changes for the Steam 4.

Generator Repair Program]

D.C. Cook Unit I License Amendment No. 273, dated December 20, 2002 [Approved Measurement Uncertainty Recapture Power Uprate]

5. Letter from J. A. Grobe, NRC, to E. E. Fitzpatrick, l&M, 'NRC Inspection Report 50-315/97006 (DRS)' dated I

July 2, 1997

6. D. C.Cook Unit 2 License Amendment No. 255, dated March 20, 2003 [Approved revisions to P-T limits]

7. D. C.Cook Units I and 2 License Amendment Nos. 236 and 218, dated December 23, 1999 [Rod duster control assembly insertion credit following a large-break LocA (LBLOCA)]

8. Letter from J. F. Stang, NRC, to R. P. Powers, i&M, 'Donald C. Cook Nuclear Plant, Units 1 and 2 - Review of Leak-Before-Break for the Pressurizer Surge Line Piping-as Provided by 10 CFR Part 50, Appendix A, GDC-4 (TAC Nos. MA7834 and MA7835),,dated November 8,2000

9. D. C. Cook Units I and 2 License Amendment Nos.'148 and 134, dated August 27, 1990 [Approved the transition to Westinghouse 17x17 VANTAGE 5 fuel and the use of Westinghouse Licensing Topical Report WCAP-1 1397-P-A, 'Revised Thermal Design Procedure,' dated April 19891 Table 3.6.2 Notes:

1. The detailed SG component integrity analyses and evaluations are beyond the level of detail presented in the D, C. Cook UFSAR.

2. The D. C. Cook Unit 2 SG tube vibration and wear evaluation quantifies the results in terms of the fluidelastic stability ratio, tube amplitudes of vibration, and tube wear; whereas the D. C. Cook Unit 1 evaluations used the fretting wear damage parameter to quantify the results.

3. The D. C. Cook Unit 2 analyses consider a maximum level of SG tube plugging of 10 percent; whereas the D. C. Unit I analyses consider a 30-percent level of SG tube plugging.

4. Prior to submittal of the D. C. Cook Unit 1 MUR power uprate application, the Flow-Accelerated Corrosion (FAC)

Program was not discussed on the docket for either D.C. Cook unit.

5. The FAC Program and evaluation of that program for the MUR power uprates are common to both Cook units.

Reference 5 of Table 3.6.2 above is the NRC inspection report that documents the NRC's review of the licensee's implementation of the D. C. Cook FAC Program.

6. For D. C. Cook Unit 2, the proposed MUR uprate Is based on new P-T curves, which were approved by D. C. Cook Unit 2 License Amendment No. 255, dated March 20, 2003 (Reference 6 of Table 3.6.2 above). The new P-T curves are supported by revised Unit 2 RV integrity analyses that used revised neutron fluence calculations, which follow the guidance in RG 1.190, 'Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. The updated P-T curves used neutron fluence projections that correspond to 'I 3800 MWt, and thus bound the proposed MUR power uprate.

3.6.2.1 Pressurized Thermal Shock and Upper-Shelf Eneray Analyses The licensee evaluated the effect that the proposed MUR power uprate will have on the structural integrity evaluations for the RV in Section IV.1.1 of the November 15, 2002, application. These structural integrity evaluations included the evaluation of RV materials relative to PTS and USE concerns. The licensee concluded that the proposed 1.66-power uprate will not have a significant effect on the structural integrity evaluations for the

D. C. Cook Unit 2 RV. For D. C. Cook Unit 2, the projected end-of-license (EOL) neutron fluences for the RV are based on 32 effective full power years (EFPYs) of operation and a cr thermal power level of 3800 MWt.

The NRC staff performed an independentcalculation of the material property values (i.e., RTpTs values) for the RV beltline using the uprated neutron fluences for the RV in order to assess what effect the proposed uprated power conditions would have on the PTS evaluations for the plant and the validity of the licensee's conclusio.Sor the ev-aluation of PMS, the befltine of the K 'D.TC.Cook Uit 2 RV is limited by the amount of embrittlement that is Droiected to occur nR intermediate-shell plate 10-1 (material heat No. C5556-2) at EOLJ The NRC staff projected the RTpTs value for intermediate shell plate 10-1 to be 215 OF, as based on an uprated 32 EFPY neutron fluence of 1.625 x 1019 n/cm2. This meets the screening criterion in 10 CFR 50.61 for RV base metal materials (i.e., RTpTsg 270OF). Based on the above, the NRC staff concludes that RV beltline materials for D. C. Cook Unit 2 will continue to have a sufficient safety margin against the impacts of PTS events and finds the uprated PTS assessment for the D. C. Cook Unit 2 RV to be acceptable.

The NRC staff performed an independent calculation of the' USE values for the RV beftline materials using the uprated neutron fluences for the 1/4T location RV at EOL.F the evaluation o concerns, the beltline of the D. C. Cook Unit 2 RV is limited by the USE drop that is Droiected to occur in the RV intermediate shell plate 10-2 (material heat No. C5521-2).(

The NRC staff projected the EOL USE value for this material2 to be 67 11-bs, as based on an uprated 32 EFPY 1/4T neutron fluence of 0.968 x 1019 n/cm . This meets the screening criterion in Appendix G to 10 CFR Part 50 of 50 ft-lbs for RV beltline materials in the irradiated condition.

Based on the above, the NRC staff concludes that RV beltline materials for D. C. Cook Unit 2 will continue to comply with the USE requirements in Appendix G to 10 CFR Part 50.

3.6.3 Summar-The NRC staff has reviewed the licensee's evaluation of the impact of the proposed MUR power uprate on RV integrity, SG tube integrity, and erosion corrosion programs. The technical areas reviewed by the NRC staff are those discussed in Section 3.6.1 of this SE. Based on the above, the NRC staff concludes that the licensee has adequately addressed these impacts and has demonstrated that the plant will continue to meet the applicable requirements following implementation of the proposed MUR power uprate. Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to materials and chemical engineering.

3.7 Human Factors 3.7.1 Regulatory Evaluation The area of human factors deals with programs, procedures, training, and plant design features related to operator performance during normal and accident conditions (NRC RIS 2002-03, Attachment 1,Section VII, Items 1 through 4). The NRC staffs human factors evaluation is conducted to confirm that operator performance will not be adversely affected as a result of system changes required for the proposed MUR power uprate. The NRC staff's review covers licensee's plans for addressing changes to operator actions, human-system interfaces, and procedures and training required for the proposed MUR power uprate. The NRC's acceptance

criteria for human factors are based on GDC-19, 10 CFR 50.54(i) and (m), 10 CFR 50.59, 10 CFR 50.120, and 10 CFR 55.59.

3.7.2 Tchnical Evaluatin The NRC staff has developed a standard set of questions for the review of the human fctors) area. The licensee has addressed these questions in its November 15,2003, application./

Following is a summary of the licensee's responses and the NRC staff's conclusions./

3.7.2.1 Operator Actions The licensee indicated that the proposed MUR power uprate is not expected to have any significant affect on the manner in which the operators control the plant during normal operations or transient conditions. The licensee also indicated that all operator actions that were taken credit for in the safety analysis would still be valid following implemention of the proposed MUR power uprateJhe NRC staff finds the implementation of the proposed MUR power uprate a . . n1Will not have an adverse affect on operator actions.

.. .t EO rtenyEa onot oeanAbnormal Operating e Procedure The licensee indicated that there are currently no Emergency Operating Procedures ta) tE~s reference use of the LEFM. Specific procedures within the EOP program may require review /

and revision based upon the proposed power uprate plant parameters for thermal power, temperature, and pressure values. These changes were identified and will be implemented uender the de-sign change process to implement the proposed power uprate. Specifically, values in the EOPs, the EOP Footnotes document, and the Abnormnal Operating Procedures (AOPs)/

will be revised based upon the proposed 1.66-percent power uprate levels. Any changes to values that are referenced in the EOPs or AOPs will be revised by the EOP/AOP control program to fully implement the proposed MUR power uprate. In addition, impacts to the D. C. Cook Emergency and Abnormal Operating Procedures were addressed in the SE for D. C. Cook License Amendment No. 273. This program is common to both D. C. Cook Unit I and Unit 2. Based on the above, the NRC sta inds that necessary procedures will be changed or updated prior to the implementation of the license and TSs changes associated with the proposed MUR power uprate. The NRC staff finds this acceptable.

s.7.23 awntrol Room Controlsf Dispain. aTnd Alarars The licensee stated that the notification of the operators of the LEFM CheckPlus systems condition will be through the plant process computer (PPC). Alarms and annotation of tme LEF system status will be through the computer display PPC. The alarm will use the existing Computer Priority Alarm. This alarm functions to alert the operators of PPC points being out of service, as well as a PPC malfunction. The annunciator position on the control boards would not change. There are no new controls for the operator to manipulate. Response to this computer alarm will be proceduralized. The Alarm Response Manual would be updated accordingly. The licensee indicated that reactor operators would be trained on the changes in the PPC, alarms associated with the LEFM, and the changes in the Alarm Response Manual in a mannerD consistent with the design modification process. Changes to control room controls, displays, and alarms, the control room plant simulator, and the operator training program will be developed as part of the implementation of the LEFM design change package. (See/

Attachment 5, 'Regulatory Commitments," of the November 15, 2002, aliaion.This will be) finalized prior to implementing the proposed MUR power uprate.fT eNRC staff finds this acceptable.

3.7.2.4 Control Room Plant Reference Siml _

The D. C. Cook Nuclear Plant Simulator Certification was submitted in a letter fo M. P. Alexich, I&M, to T. E. Murley, NRC, dated August 24, 1990, pursuant to 10 CFR 55.45(b)(5). The proposed MUR power uprate is not expected to have a significan effect on any simulated systems and the simulator is not expected to be modified. If changes to the simulator are necessary, the licensee indicated that changes to the simulator associated with the MUR power uprate would be treated in a manner consistent with an other lant modification, and would be tested and documented accordingly. The NRC staff finds this

-acceptable.

31.7.2.5 Operator Training Program: 4 The installation of the LEFM and implementation of the proposed 1.66-percent MUR power) uprate would require procedure and training changes. Actions'would be added to the/

appropriate operating procedures and the development of an Admrinistrative Technical Requirement ine the event the LEFM system becoames utiavailable. Operations training ~

concerning the use of the LEFM, the associated procedures, and the Administrative Technical e uirement chan es will be completed prior to implementation of the MUR power uprate. he IsRC staff finds this accep a le.; --

3.7.3 Summary The NRC staff has reviewed the licensee's planned actions related to the human factors area and concludes that licensee has adequately considered the impact of the proposed MUR power uprate on changes to operator actions, procedures, plant hardware, and associated training programs to ensure that operators' performance is not adversely affected by the proposed MUR power uprate. The NRC staff further concludes that the licensee will continue to meet the requirements of 10 CFR 50.54(i) and (m),; 10 CFR 50.59, 10 CFR 50.120, and 10 CFR 55.59 following implementation of the proposed MUR power uprate. Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to the human factors aspects of required system changes.

[3-8 PlantSystems 3.8.1 Regulatory Evaluation The NRC staff review in the area of plant systems covers the impact of the proposed MUR power uprate on (1)containment performance analyses and containment systems, (2) safe shutdown fire analyses and required systems, (3)spent fuel pool cooling analyses and systems, (4) flooding analyses, (5) NSSS interface systems, (6)radioactive waste systems, and (7) engineered safety feature (ESF) heating, ventilation, and air conditioning systems (NRC RIS 2002-03, Attachment 1, Sections II, III, and VI). The review is conducted to verify that the licensee's analyses bound the proposed plant operation at the MUR power level and that the results of licensee analyses related to the areas under review continue to meet the applicable

acceptance criteria following implementation of the proposed MUR power uprate. Guidance and acceptance criteria for the NRC staffs review of reactor systems are contained in Chapters 3, 6, 9, 10, and 11 of NUREG-0800.

3.8.2 Technical Evaluation The NRC staff has reviewed the licensee's application as related to the plant systems areas discussed above and has determined that for most areas, existing analyses of record bound plant operation at the proposed uprated power level. Ther sults of the NRC staffs review in the esummarized in Table 3.8.2 below.The licensee performed new analyses or ost-LOCA containment hydrogen generationas evauaiono ese ana yses Is included in-Section 3.8.2.1 below.

ATable I

C 3.8.2 l - Plant Systems l.Unit 2 MUR Bounded by Application UFSAR NRC-approved :'Similar to lNRC Staff TpcSection Section ; analysis Unit I MUR lConclusion Post-LOCA 11.2.2 : X14.3.6 Nw~e 1 N Acceptable Containment Hydrogen (page 40) (References 3, 4, 5, 6)i G eneration Fi_ i__,______

Long-Term LOCA 11.2.3.1 14.3.4.3.1.2 Y Y Acceptable Mass and Energy (page 41) (References 2, 7)

Release Analysis Short-Term LOCA 11.2.3.2 14.3.4.5.1 . y Acceptable Mass and Energy (page 41) -(References 7,-8, 9)

Release Analyses Fire Protection Systems Fire Protection JV1I.6.6 I 1.0.1: 1V y l Acceptable Evaluation (page 106) ( 9Referencesj9,10,11, I _

Power/Steam Systems Main Steam System VI.2.1 7.3.2 NN~'I , Acceptable and Steam Dump (pages 89, 90) 7.3.3 (References 8, 9, 26)

System : 10.2 _,.

Condensate and VI.2.2 10.5.1 Y Y Acceptable Feedwater Systems (pages 90, 91 (References 8, 9)

Auxiliary Feedwater VI.2.3 10.5.2 V , Acceptable System and (pages 91,92 (References 8, 27)

Condensate Storage System Feedwater Heaters and VI.2.4 10.5.1 V V Acceptable Drains (page 92) (References 8, 9)

SG Blowdown System VI.2.5 10.11 Acceptable (pages 92,93) (References 8, 9)

I Cooling and Support Systems Component Cooling VI.3 9.5 Y 1.2 Y Acceptable Water System (page 93 ) (Reference 8)

Table 3.8.2 Plant Systems Unit 2 MUR Bounded by I' Application UFSAR NRC-approved Similar to NRC Staff Topic Section Section analysis Unit I MUR Conclusion Essential Service V1.3.2 9.8.3 Y Y Acceptable Water System (page 93) (References 8, 17,18,

_ __ _ __ __ __ _ 1 9 , 20 )_ _ _ _ _ _

Non-Essential Service VI.3.3 9.8.3 V Y Acceptable Water (page 93 ) (References 8, 21, 22) l Turbine Auxiliary Vi.3.4 10.7 , V Acceptable Cooling Water System (page 94 ) m (Reference 8)

Emergency Diesel VI.3.5 8.4 Y Y Acceptable Generator Aftercooler, (page 94) A (Reference 8)

Lube Oil, and Jacket Cooling Water System ______-_A___

Circulating Water V1.3.6 - 10.6 Y Y. Acceptable System (page 94) VReference 8)

Spent Fuel Pool VI.3.7 9.4 Y Y Acceptable Cooling System (page 94) (References 8,23, 24,

_ _ _ _ _X: _ _ X25 ) _

Heating, Ventilation, and Air Conditioning Systems Auxiliary Building Vl.4 9.9 V V Accepta le Ventilation (Page 95) (References 8, 9, 29, Systems" 3 31 _____i__31)

Engineered Safety' VIA 9.9 y Y Acceptable Features Ventilation (Page 95) (References 8, 9,29, System ad__

Ai:,

31) -

Containment VI.4 - 5.5 y Y Acceptable Ventilation System (Page 95) (References 2,8,9, 28, 30, 32)

Auxiliary Feedwater VIA 9.8.3, 9.9.3, V - Y Acceptable Pump Room Coolers (Page 95) 14.4.9 (Reference 18)

Control Room VI.3.2 9.10 V Y Acceptable Ventilation Systemt °'e4 VI.5(iii) (References 33, 34)

VII.6.11 V I.6.110_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Table 3.8.2

References:

1. D. C. Cook Unit 2 License Amendment No. 135, dated September 18, 1990 [Allowed Unit 2 SG stop valve closure within 8 seconds]

2. D. C. Cook Units 1 and 2 License Amendment Nos. 234 and 217 [Approved containment sump modification, as evaluated in Westinghouse Licensing Topical Report WCAP-15302, 'Donald C. Cook Nuclear Plant Units 1 and 2, Modifications to the Containment Systems, Westinghouse Safety Evaluation (SECL 99-076, Revision 3),-

dated September 1999]

3. D.C. Cook Units 1 and 2 License Amendment Nos. 148 and 134, dated August 27,1990 [Approved the transition to Westinghouse 17x17 VANTAGE 5 fuel and the use of Westinghouse Licensing Topical Report WCAP- 1397-P-A, ~Revised Thermal Design Procedure," dated April 1989]

4. D. C. Cook Units 1 and 2 License Amendment Nos. 214 and 199, dated March 13, 1997 [Approved an increase in SG plugging limit]

5. D.C. Cook Unit I License Amendment No. 252, dated March 29, 2001 [Approved changes to TSs for spray additive tank (the analyses covered both units but only resulted in changes to Unit 1)]

6. Letter from R. L. Baer, NRC, to J.Tillinghast, I&M, 'Order for Modification of License (Donald C, Cook Nuclear Plant Unit 2)," dated June 6,1978 [Modifies TS limit for total nuclear peaking factor (Fj)]

7. Supplement to Safety Evaluation Report, 'Supplement No. 3 to Safety Evaluation by the Directorate of Licensing U. S. Atomic Energy Commission in the Matter of Indiana & Michigan Electric Company and Indiana & Michigan Power Company Donald C. Cook Nuclear Plant Units 1 and 2, Docket Nos. 50-315 and 50-316," dated December 12,1974

8. Safety Evaluation Report, 'Safety Evaluation by the Directorate of Licensing U.S. Atomic Energy Commission In the Matter of Indiana & Michigan Electric Company and Indiana & Michigan Power Company Donald C. Cook Nuclear Plant - Units I and 2, Docket Nos. 50-315 and 50-316," dated September 10, 1973

9. Letter from NRC to Indiana and Michigan Electric Company, 'Supplement 7 to Safety Evaluation Report," dated December 23, 1977

10. D. C. Cook Units 1 and 2 License Amendment Nos. 31 and 12, dated July 31, 1979 [Added license conditions for the Fire Protection Program]

11. Letter from S. A. Varga, NRC, to J. Dolan, l&M, 'Safety Evaluation on Altemative Shutdown Capability," dated November 22, 1983 [Complies with Sections III.G and Ill.L of Appendix R]

12. Letter from S. A. Varga, NRC, to J. Dolan, I&M, Acceptance of Technical Exemptions from 10 CFR [Part] 50, Appendix R." dated August 27, 1985

13. Letter from B. J. Youngblood, NRC, to J. Dolan, l&M, "Safety Evaluation Report Regarding Altemative Shutdown Procedures Inthe Event of Fire at D.C. Cook Units I and 2," dated January 28, 1987

14. Letter from J. F. Stang, NRC, to M. P. Alexich, l&M, 'Unrated Fire Hatches In Fire Area Boundaries (TAC Nos. 61690/61691),' dated June 17, 1988

15. Letter from R. S. Boyd, NRC, to J. Tillinghast, I&M, "Issuance of Facility Operating License No. DPR-74 (Donald C. Cook Nuclear Plant, Unit No. 2)," dated December23,1977

16. Letter from J. B. Hickman. NRC, to E. E. Fitzpatrick, i&M, 'Revision to Technical Specification Bases Reflecting Change to Fire Suppression Backup Water Source (TAC Nos. M90177 & M90178)," dated December 14, 1994

17. D. C. Cook Units I and 2 License Amendment Nos. 164 and 149, dated April 22,1992 [Approved changes to make TSs more consistent with ASME Code Requirements]

18. D. C. Cook Units 1 and 2 Ucense Amendment Nos. 244 and 225," dated April 25, 2000 [Approved modification to turbine-driven auxiliary feedwater pump room cooler plant]

19. D. C. Cook Units 1 and 2 License Amendment Nos. 253 and 235, dated August 3, 2001 [Added requirement for essential service water cross-tie to opposite unit]

20. D.C. Cook Units 1 and 2 License Amendment Nos. 270 and 251, dated September 9, 2002 [Approved changes to allow one-time extended allowed outage time for essential service water pump replacement]

21. D.C. Cook Units I and 2 License Amendment Nos. 59 and 42, dated September 9, 1982 [Approved TS changes to reflect replacement of containment isolation valves]

22. D. C. Cook Units I and 2 License Amendment Nos. 95 and 81, dated April 23, 1986 [Approved changes to containment isolation valve testing requirements]

23. D. C.Cook Units 1 and 2 License Amendment Nos. 32 and 13, dated October 16, 1979 [Approved increased storage capacity in spent fuel pool]

24. D. C. Cook Units I and 2 License Amendment Nos. 169 and 152, dated January 14,1993 [Approved changes for spent fuel pool re-racking]

25. D.C. Cook Units 1and 2 License Amendment Nos. 260 and 243, dated November 30, 2001 [Approved revision to

'decay time' to allow start of core offload at 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months s]

26. D.C. Cook Units 1and 2 License Amendment Nos. 182 and 167, dated September 9, 1994 [Approved an increase in MSSV setpoint tolerances]

27. D.C. Cook Units I and 2 License Amendment Nos. 214 and 199, dated March 13,1997 [Approved an Increase in SG plugging limit]

28. Letter from D. L. Wigginton, NRC, to J. Dolan, Indiana and Michigan Electric Company, 'Amendment No.66 to DPR-58 and Amendment No.47 to DPR-74,' dated December 8, 1982

29. Letter from J. F. Stang, NRC, to M.P. Alexich, Indiana and Michigan Electric Company, 'Amendments Nos.124 and 111 to Facility Operating Licenses Nos. DPR-58 and DPR-74: Technical Specification Changes for the Engineered Safety Features and Storage Pool Ventilation System (TAC Nos. 65559 and 65560),' dated May 19, 1989

30. Letter from J. B. Hickman, NRC, to E. E. Fitzpatrick, Indiana and Michigan Electric Company, 'Donald C. Cook Nuclear Plant, Unit Nos. I and 2- Issuance of Amendments Re: Containment Purge (TAC Nos. M91956 and M91957)," dated June 23, 1995 [Amendment No. 195 to Facility Operating License No. DPR-58 and Amendment No. 181 to Facility Operating License No. DPR-74]

31. Letter from J. F.Stang, NRC, to R. P. Powers, Indiana and Michigan Electric Company, 'Donald C. Cook Nuclear Plant, Units I and 2 - Issuance of Amendments (TAC Nos.:MA9394 and MA9395),- dated October 24,2001

[Amendment 257 to DPR-58 and Amendment 240 to DPR-74]

32. Letter from J. F.Stang, NRC, to R. P. Powers, Indiana and Michigan Electric Company, 'Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments (TAC Nos. MB1973 and MB1974)," dated November 21, 2001

[Amendment 259 to DPR-58 and Amendment 242 to DPR-741

33. D.C. Cook Units 1 and 2 License Amendment Nos. 258 and 241, dated November 13, 2001 [Partial alternative source term]

34. D. C.Cook Units I and 2 License Amendment Nos. 271 and 252, dated November 14,2002 [Alternative source term for control room habitability]

Table 3.8.2 Notes:

1. To support the proposed D. C. Cook Unit 2 MUR power uprate, the licensee performed an evaluation to demonstrate that the post-LOCA hydrogen generation at the uprated power level remain within acceptance criteria (See Section 11.2.2 of the licensee's November 15, 2002, application and Section 3.8.2.1 below for the NRC staffs evaluation). For Unit 1, the existing post-LOCA hydrogen analysis was based upon a core power of 3411 MWtM which bounds the proposed D. C. Cook Unit I MUR power uprate.

2. The licensee re-performed the RHR cooldown analysis to support the proposed D. C. Cook Unit 2 MUR power uprate. The revised analysis, which considers a change to the plant's RTP only, demonstrates that the licensee will still be able to reach Mode 5 conditions within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months on a single train of RHR, and the time to cool down to

<140 "F with two trains of RHR available has increased from less than 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months to less than 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months .

3. The auxiliary building ventilation systems at D. C. Cook Include the engineered safety features ventilation system, fuel handling area ventilation system, general ventilation systems, and general supply system.

4. The control room ventilation system was assessed as part of the on-site radiological dose consequences assessment, the heat load assessment for the essential service water system, and the temperature, pressure, and radiation levels used in the environmental qualification of equipment analyses.

3.8.2.1 Post LOCA Containment Hydrogen Generain The licensee's review determined thatthe analysis of record for post-LOCA hydrogen generation) was performed for core thermal power of 3411 MWt. This analysis was performed to bound/

both units and is presented in Section 14.3.6 of the D. C. Cook Unit I FSAR. The proposed/

uprated power level for D. C. Cook Unit 2 is 3468 MWt. Therefore, the existing analysis of record does not bound proposed operation of the plant at the uprated power level. To support the power uprate application, the licensee performed evaluations for the post-LOCA hydrogen generation analysis. The licensee's evaluation covered operation up to 3588 MWt. Since the calculated hydrogen produced by radiolysis in the core and sump is a function of the ionizing radiation flux, the licensee assumed that the hydrogen produced by radiolysis is directly proportional to the core power level. The licensee assumed a one-to-one correlation and increased the hydrogen produced by radiolysis in the core and sump by 5 percent, which corresponds to an increase of 5-percent power, The licensee determined that the hydrogen generation from sources other than tadiolysis is not affected by the power uprate. The licensee's evaluation was based -upon the application of a conservative and bounding power increase of 5 percent compared to the requested power increase of 1.66 percent. The I .i licensee's evaluation concluded that hydrogen production from all sources increases by only 1 percent during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and by 2 percent at the end of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months as a result of the increase in power level. Further, the calculations show that if recombiners are started at or before the time at which the containment hydrogen concentration reaches 3.5 percent volume, the resulting hydrogen concentrations remain below the lower flammability limit of 4.0 percent.

Similar evaluations were also performed for containment subcompartment hydrogen concentrations. The licensee again increased the hydrogen produced by radiolysis in the core and sump by 5 percent, corresponding to an increase of 5-percent power. The licensee did not increase hydrogen generation sources other than radiolysis because it determined that hydrogen generation from such sources is not affected by the power uprate. The licensee's evaluations for the hydrogen concentrations in containment subcompartments concluded that an increase in power up to 3588 MWt would result in an increase of 0.1 percent in the short-term peak subcompartment hydrogen concentration following a LBLOCA and an increase of 1.6 percent in the long-term (i.e., final analysis time of S-10 hours following a LBLOCA and -14 hours following a small-break LOCA) peak subcompartment hydrogen concentrations. Further, the calculated values for the short-term and long-term hydrogen concentrations remain below the flammability The NRC staff reviewed the licensee's applications related to post-LOCA hydrogen generation in containment and determined that (1) the evaluations were performed in an acceptable manner to bound the proposed operation of the plant at the uprated power level and (2)the resulting hydrogen concentrations remain below the flammability limit of 4.0 percent. Based on the above, the NRC staff condudes that the proposed MUR power uprate is acceptable with respect to post-LOCA containment hydrogen generation.

3.8.3 Summary The NRC staff has reviewed the licensee's safety analyses of the impact of the proposed MUR power uprate on (1) containment performance analyses and containment systems, (2) safe shutdown fire analyses and required systems, (3)spent fuel pool cooling analyses and systems, (4)flooding analyses, (5) NSSS interface systems, (6) radioactive waste systems, and

(7) ESF heating, ventilation, and air conditioning systems. The NRC staff concludes that the results of licensee's analyses related to these areas would continue to meet the applicable acceptance criteria following implementation of the proposed MUR power uprate. Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to plant systems.

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 the requirements With respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes the surveillance requirements. The 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 Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (68 FR 2805). Accordingly, the amendment meets the eligibility criteria for categorical exclusion setforth 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

i The 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.

Attachment:

List of Acronyms Principal Contributors: J. Stang M. Shuaibi Date:

LIST OF ACRONYMS

[ I_ n AUP Abnormal uperating Procedures ASME American Society of Mechanical Engineers BOP balance-of-plant CFR Code of Federal Regulations DBA design-basis accident DNB departure from nucleate boiling ECCS emergency core cooling system EOL end of license EOP emergency operating procedure ESF engineered safety feature FAC flow-accelerated corrosion LEFM 0 t ileading edge flowmeter LOCA loss-of-coolant accident MOV motor-operated valve MSSV main steam safety vaves MUR measurement uncertainty recaputure MWt megawatts thermal NRC Nuclear Regulatory Commission NSAL Nuclear Safety Advisory Letters NSSS Nuclear Steam Supply System PPC plant process computer P-T pressure-temperature PTS pressurized thermal shock RCCA rod cluster control assembly RCS reactor coolant system RHR residual heat removal ATTACHMENT

I RIS Regulatory Issue Summary RTP rated thermal power RV reactor vessel SE safety evaluation SG steam generator SGTR steam generator tube rupture UFSAR Updated Final Safety Analysis Report

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I to AEP:NRC:3902-01 Page I REGULATORY COMMITMENTS The following table identifies those actions committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.

Commitment Date 1-I&M is installing an LEFM CheckPlus system at CNP Unit 2 in Prior to raising power anticipation of approval of this proposed amendment. Installation of above 3411 MWt this system will begin prior to the Unit 2 Cycle 14 refueling outage and will be completed after receipt of the requested license amendment. The design change for the installation will include instrumentation resealing, UFSAR revision, maintenance and operational procedure impacts, training, monitoring iso-phase bus duct temperature, and implementation of the LEFM CheckPlus system out-of-service administrative technical requirements. The UFSAR revision for the Unit 2 MUR power uprate will be reflected in the next update of the UFSAR submitted to the NRC pursuant to 10 CFR 50.71(e).

v t e Letter Sequence Draft Other
TAC:MB6751, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Withholding Request Acceptance Questions Answers Results Approval, Approval Other: ML031260667, ML032671104
MONTHYEAR2003-04-25 [Table View]

ML031270262

Text

References:

Enclosures:

References:

1.0 INTRODUCTION

2.0 BACKGROUND

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

Attachment:

References:

References:

References:

References:

References:

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

Attachment:

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