ML020570201

License Amendment Request, Relocation of TS 3/4.9.6 & TS 3/4.9.7 & Associated Bases to CNP Updated Final Safety Analysis Report

ML020570201
Person / Time
Site:	Cook
Issue date:	02/22/2002
From:	Bakken A American Electric Power Co, Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NUREG-1431 Rev 2
Download: ML020570201 (40)

Text

Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INDIANA MICHIGAN POWER February 22, 2002 AEP:NRC:2039 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P 1-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Docket Nos. 50-315 and 50-316 License Amendment Request for Technical Specification 3/4.9, Refueling Operations

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposes to amend Appendix A, Technical Specifications (TS), of Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to relocate TS 3/4.9.6, "Refueling Operations - Manipulator Crane Operability" and TS 3/4.9.7 "Refueling Operations - Crane Travel - Spent Fuel Storage Pool Building," with associated Bases to the CNP Updated Final Safety Analysis Report (UFSAR).

Relocation of these TS and associated Bases to CNP's UFSAR will provide additional operational flexibility during refueling outages.

The proposed changes are based on the criteria of 10 CFR 50.36(c)(2)(ii) for items requiring a TS limiting condition for operation. The proposed changes are consistent with NUREG-1431, "Standard Technical Specifications Westinghouse Plants," Revision 2, dated June 2001.

I&M also proposes format changes to the affected TS page that improve appearance but do not affect any requirements.

U.S. Nuclear Regulatory Commission AEP:NRC:2039 Page 2 Enclosure I provides an oath and affirmation affidavit. Enclosure 2 provides a detailed description and safety analysis to support the proposed changes, including the 10 CFR 50.92(c) evaluation, which concludes that no significant hazard is involved, and the environmental assessment. Attachments 1A and 1B provide marked up TS pages for Unit 1 and Unit 2, respectively. Attachments 2A and 2B provide the proposed TS pages with the changes incorporated for Unit 1 and Unit 2, respectively. Attachment 3 provides a summary of the regulatory commitments made in this submittal.

I&M requests approval of the proposed amendment by April 5, 2002, to support the Unit 1 refueling outage. Once approved, the amendment will be implemented within 30 days.

No previous submittals affect the TS pages that are submitted in this request. If any future submittals affect these TS pages, I&M will coordinate the changes to the pages with the NRC Project Manager to ensure proper TS page control when the associated license amendment requests are approved.

Should you have any questions, please contact Mr. Gordon P. Arent, Manager of Regulatory Affairs, at (616) 697-5553.

Sincerely, A. C. Bakken, III Senior Vice President, Nuclear Operations

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Enclosures:

1. Notarized oath and affirmation

2. Evaluation of the proposed changes Attachments:

1. Unit 1 and Unit 2 Marked-up Technical Specification pages

2. Unit 1 and Unit 2 Proposed Technical Specification pages

3. Summary of commitments made in this letter

U.S. Nuclear Regulatory Commission AEP:NRC:2039 Page 3 c: J. E. Dyer MDEQ - DW & RPD NRC Resident Inspector R. Whale

Enclosure 1 to AEP:NRC:2039 AFFIDAVIT I, A. Christopher Bakken, III, being duly sworn, state that I am Senior Vice President, Nuclear Operations of American Electric Power Service Corporation and Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

American Electric Power Service Corporation A. C. Bakken, III Senior Vice President, Nuclear Operations SWORN TO AND SUBSCRIBED BEFORE ME THIS _2 DAY OF 2002 MyComis Notary les _

MCommssion Epires / /

to AEP:NRC:2039 Page I Application for Amendment to Technical Specification (TS) 3/4.9, "Refueling Operations" 1.0 Description Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposes to amend Appendix A, Technical Specifications (TS), of Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to relocate TS 3/4.9.6, "Refueling Operations - Manipulator Crane Operability" and TS 3/4.9.7 "Refueling Operations - Crane Travel - Spent Fuel Storage Pool Building," with associated Bases to the CNP Updated Final Safety Analysis Report (UFSAR). Relocation of these TS and associated Bases to CNP's UFSAR will provide additional operational flexibility during refueling outages.

The proposed changes are based on the criteria of 10 CFR 50.36(c)(2)(ii) for items requiring a TS limiting condition for operation (LCO). The proposed changes are consistent with NUREG-1431, "Standard Technical Specifications Westinghouse Plants," Revision 2, dated June 2001.

I&M also proposes format changes to the affected TS page that improve appearance but do not affect any requirements.

2.0 Proposed Change I&M proposes that the following Unit 1 and Unit 2 TS and associated Bases be relocated from the CNP TS to the UFSAR:

TS 4/3.9.6, "Manipulator Crane Operability" TS 4/3.9.7, "Crane Travel - Spent Fuel Storage Pool Building" Upon relocation to the UFSAR, these requirements would be controlled by 10 CFR 50.59.

Additionally, references to these TS will be deleted from the TS Index.

I&M also proposes three types of format changes to the revised TS pages. The changes are:

(1) Reformatting of the headers to include numbered first and second-tier TS section titles and a full-width single line to separate the header section titles from the page text.

(2) Reformatting of the footers to include "Page (page number)" center page, "AMENDMENT (past amendment numbers, with strikethrough, and ending with the current amendment number)" on the right side of the page, and a full-width single line to separate the footer from the page text.

to AEP:NRC:2039 Page 2 (3) Fully justifying the text and changing the font.

3.0 Background In February 1987, the NRC published an Interim Policy Statement on TS improvements for nuclear power reactors (Reference 1). This policy statement established a specific set of criteria for determining which regulatory requirements and operating restrictions should be included in TS. In November 1987, the Westinghouse Owners Group (WOG) published WCAP-1 1618 (Reference 2), in which the criteria contained in the Interim Policy Statement were applied to standard Westinghouse TS to determine whether individual specification should be removed and relocated to licensee controlled documents. As documented in WCAP-1 1618, the WOG determined that the requirements of TS 3/4.9.6 and TS 3/4.9.7 were among those that could be relocated to another controlled document. In a May 1988 letter (Reference 3), the NRC published its conclusions regarding the WOG determinations documented in WCAP-1 1618.

That letter documented NRC agreement with the WOG conclusions regarding relocation of TS 3/4.9.6 and TS 3/4.9.7.

In July 1993, the NRC published its Final Policy Statement (Reference 4) on TS improvements.

This policy statement provided four specific criteria for determining which design features and information should be located in the TS as LCOs. These four criteria were very similar to the criteria published in the Interim Policy Statement. The four criteria provided by the Final Policy Statement are as follows:

1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The Final Policy Statement noted that those LCO's that do not meet any of the four criteria may be proposed for removal from the TS and relocated to a licensee-controlled document, such as the UFSAR. The Policy Statement also noted that licensees submitting amendment requests should identify the location and controls for the relocated requirements. The four criteria provided in the Final Policy Statement were codified in 10 CFR 50.36(c)(2)(ii).

to AEP:NRC:2039 Page 3 4.0 Technical Analysis Provided below is an evaluation of the requirements of TS 3/4.9.6 and TS 3/4.9.7 against the four criteria defined in the NRC's Final Policy Statement and 10 CFR 50.36(c)(2)(ii).

TS 3/4.9.6 Refueling Operations - Manipulator Crane Operability TS 3/4.9.6 specifies the minimum capacity and overload cutoff limits for the manipulator crane used for the movement of fuel assemblies within the reactor pressure vessel, and specifies the minimum capacity and load indicator limits for the auxiliary hoist used for the movement of control rods within the reactor pressure vessel.

TS 3/4.9.6 ensures that: (1) the manipulator crane will be used for movement of control rods and fuel assemblies within the reactor pressure vessel; (2) each crane has sufficient load capacity to lift a control rod or fuel assembly; and (3) the core internals and pressure vessel are protected from excessive lifting force in the event that they are inadvertently engaged during lifting operations.

Evaluation against 10 CFR 50.36(c)(2)(ii) criteria:

1. The refueling equipment and associated instrumentation is not used to detect, or indicate in the control room, a significant degradation of the reactor coolant pressure boundary.

2. The refueling equipment and associated instrumentation are not process variables, design features, or operating restrictions that are initial conditions of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The applicable design basis accident is a fuel handling accident in which a fuel assembly is dropped, resulting in a release of radioactive material. The refueling equipment and associated instrumentation do not affect the assumptions or initial conditions in the analysis of that accident.

3. The refueling equipment and associated instrumentation are not part of the primary success path to mitigate the consequences of a design basis accident. These components serve no mitigative function.

4. As summarized in Appendix A to WCAP- 11618, the requirements of TS 3/4.9.6 were determined not to be risk dominant based on core melt and health effects screening criteria. Additionally, TS 3/4.9.6 does not govern a system, structure, or component requiring risk review/unavailability monitoring as described in CNP's Maintenance Rule Program. The components associated with TS 3/4.9.6 were not evaluated as risk contributors in the CNP Individual Plant Examination (IPE).

to AEP:NRC:2039 Page 4 Based on the above, the design features and information in TS 3/4.9.6 does not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion as a TS LCO and therefore may be relocated to the CNP UFSAR.

TS 3/4.9.7 Refueling Operations - Crane Travel - Spent Fuel Storage Pool Building TS 3/4.9.7 prohibits loads in excess of 2,500 pounds from traveling over fuel assemblies in the storage pool. The TS also limits loads carried over the spent fuel pool and the heights at which they may be carried over the racks containing irradiated fuel assemblies so as to prelude impact energies over 24,240 inch-pounds if the loads are dropped from the crane.

TS 3/4.9.7 ensures that, in the event of a dropped load, 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array.

Evaluation against 10 CFR 50.36(c)(2)(ii) criteria:

1. Crane travel limits are not used to detect, or indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

2. This TS applies to the crane and its interlocks which have both design features and operating restrictions in place to prevent dropping a load on racks containing irradiated fuel that is stored in the spent fuel pool. Criterion 2 requires the design features or operating restrictions to be initial conditions of the design basis accident.

The initial condition of the design basis fuel handling accident is the dropping of a single fuel assembly. The crane interlocks are design features that are in place to prevent exceeding the initial condition (i.e. damage to more than one fuel assembly),

not an initial condition of itself. These design features are not, in themselves, initial conditions of a design basis accident. Similarly, the load and impact energy limits are operational restrictions that are intended to prevent exceeding the initial condition of the design basis accident. Therefore, the crane, its interlocks, and the load and impact energy limits are provided to prevent operation in a condition that could lead to an unanalyzed load drop accident.

3. The load and impact energy limits and the crane travel interlocks are not part of the primary success path to mitigate the consequences of a design basis accident or transient. These components serve no mitigative function.

4. As summarized in Appendix A to WCAP-1 1618, the requirements of TS 3/4.9.7 were determined not to be risk dominant based on core melt and health effects screening criteria. Additionally, TS 3/4.9.7 does not govern a system, structure, or component requiring risk review/unavailability monitoring as described in CNP's Maintenance to AEP:NRC:2039 Page 5 Rule Program. The components associated with TS 3/4.9.7 were not evaluated as risk contributors in the CNP IPE.

Based on the above, the design features and information in TS 3/4.9.7 does not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion as a TS LCO and therefore may be relocated to the CNP UFSAR.

The relocation of TS 3/4.9.6 and TS 3/4.9.7 is also consistent with NUREG-1431, "Standard Technical Specifications - Westinghouse plants" (Reference 5). NUREG 1431 does not include any specifications equivalent to TS 3/4.9.6 and TS 3/4.9.7. Therefore, relocation of the requirements of TS 3/4.9.6 and TS 3/4.9.7 is consistent with both past and current regulatory positions and requirements.

5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration I&M has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

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

Response: No The proposed changes are administrative in nature in that they result in relocation of requirements from TS 3/4.9.6 and 3/4.9.7, with associated Bases, to the CNP UFSAR. Changes to the UFSAR are controlled by 10 CFR 50.59. Regulation 10 CFR 50.59 requires that NRC approval be obtained prior to any change to the UFSAR that would result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated. Accordingly, the relocation of requirements from TS 3/4.9.6 and 3/4.9.7, with associated Bases to the CNP UFSAR provides continued protection from changes involving unapproved increases in the probability of occurrence of an accident. The relocation of the requirements of TS 3/4.9.6 and 3/4.9.7 would not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, configuration of CNP or the manner in which it is operated. Therefore, the proposed change does not significantly increase the probability of occurrence of an accident previously evaluated.

The proposed change to relocate TS 3/4.9.6 and 3/4.9.7, with associated Bases to the CNP UFSAR does not impact the consequences of an accident because there to AEP:NRC:2039 Page 6 is no effect on the structures, systems and components that mitigate the effects of an accident, or the manner in which they are operated. In accordance with 10 CFR 50.59, if any proposed change to the UFSAR results in more than a minimal increase in the consequences of an accident previously evaluated, NRC review and approval is required prior to the change being made. Accordingly, the relocation of requirements from TS 3/4.9.6 and 3/4.9.7, with associated Bases to the CNP UFSAR provides continued protection from changes involving unapproved increases in the probability of in the consequences of an accident.

Therefore, the relocation of requirements will not affect offsite doses, and the consequences of an accident previously evaluated are not significantly increased.

The format changes improve the appearance of the affected pages but do not affect any requirements.

Therefore, the probability of occurrence and the consequences of an accident previously evaluated are not significantly increased.

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

Response: No The proposed change to relocate TS 3/4.9.6 and 3/4.9.7, with associated Bases, to the CNP UFSAR does not create new accident causal mechanisms. Plant operation will not be affected by the proposed change and no new failure modes will be created. Regulation 10 CFR 50.59 requires that NRC approval be obtained prior to any change to the UFSAR that would create the possibility of a new or different kind of accident from any accident previously evaluated.

Accordingly, the relocation of requirements from TS 3/4.9.6 and 3/4.9.7, with associated Bases to the CNP UFSAR provides continued protection from unapproved changes involving new or different kinds of accidents.

The format changes improve the appearance of the affected pages but do not affect any requirements.

Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.

to AEP:NRC:2039 Page 7

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

Response: No The proposed change to relocate the requirements from the TS to the UFSAR does not impact equipment design or operation and no changes are being made to the TS required safety limits, safety system settings, or any safety margins associated with TS 3/4.9.6 and 3/4.9.7. Changes to the UFSAR are controlled under the 10 CFR 50.59 process, which requires a safety evaluation to be performed. If any proposed change to the UFSAR results in a design basis limit for a fission product barrier, as described in the UFSAR, being exceeded or altered or results in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses, NRC review and approval will be required prior to the change being made.

Accordingly, the relocation of requirements from TS 3/4.9.6 and 3/4.9.7, with associated Bases to the CNP UFSAR provides continued protection from changes involving a reduction in the margin of safety. The format changes improve the appearance of the affected pages but do not affect any requirements.

Therefore, there is no significant reduction in the margin of safety.

In summary, based upon the above evaluation, I&M has concluded that the proposed change involves no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria 5.2.1 TS and Regulations TS 3/4.9.6, TS 3/4.9.7 and their associated Bases are affected by the proposed change in that they would be deleted and their requirements relocated to the UFSAR. Relocation of these requirements is acceptable for the reasons described above.

10 CFR 50.36(c)(2)(ii) specifies the criteria for inclusion of requirements in Technical Specification LCOs. As described above, relocation of the TS 3/4.9.6, and TS 3/4.9.7 requirements to the UFSAR is consistent with 10 CFR 50.36(c)(2)(ii).

to AEP:NRC:2039 Page 8 5.2.2 Design Bases UFSAR 14.2.1, Fuel Handling Accident The UFSAR states that the possibility of a fuel handling accident is very remote because of the many administrative controls and physical limitations imposed on the fuel handling operations. Interlocks prevent movement of the crane hook over the spent fuel pool, except when it is necessary to service the pool and its equipment and instrumentation, and to add or remove any equipment associated with spent fuel handling, storage, or inspection. The crane hook is limited to the TS 3.9.7 value with the entire operation under strict administrative control.

The proposed changes do not alter these requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the 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.

6.0 Environmental Considerations I&M has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. I&M has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 Precedent Licensing Actions In a letter dated January 17, 2002, the NRC issued Amendment No. 137 (Reference 6) to Facility Operating License No. NPF-457 for the Hope Creek Generating Station. This amendment included relocation of TS 3/4.9.6, "Refueling Operations, Refueling Platform," TS 3/4.9.7, "Refueling Operations, Crane Travel - Spent Fuel Storage Pool," and the associated Bases, to the plant's UFSAR.

to AEP:NRC:2039 Page 9 In a letter dated February 10, 2000, the NRC issued Amendment No. 240 (Reference 7) to Facility Operating Licenses No. DPR-65 for the Millstone Nuclear Power Station, Unit 2. This amendment included relocation of TS 3/4.9.6, "Refueling Operations, Crane Operability Containment Building," TS 3/4.9.7, "Refueling Operations, Crane Travel - Spent Fuel Storage Building," and the associated Bases, to the plant's Technical Requirements Manual.

Similar to Hope Creek and Millstone Unit 2, I&M proposes to relocate Unit 1 and Unit 2 TS 3/4.9.6 and 3/4.9.7 and their associated Bases to a licensee controlled document. Based on issuance of Hope Creek Amendment No. 137 and Millstone 2 Amendment No. 240, the NRC has determined that the requested change is acceptable.

8.0 References

1. 52 FR 3788, "Nuclear Regulatory Commission - Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors," dated February 6, 1987.

2. WCAP-1 1618, "Methodically Engineered, Restructured and Improved, Technical Specifications," dated November 1987.

3. Letter from T. E. Murley, NRC, to R. A. Newton, WOG, dated May, 9, 1988

4. "Nuclear Regulatory Commission - Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," dated July 22, 1993.

5. NUREG 1431, "Standard Technical Specifications Westinghouse Plants," Revision 2, dated June, 2001

6. PSE&G Nuclear, Hope Creek Generating Station, License Amendment No. 137, dated January 17, 2002.

7. Northeast Nuclear Energy Company, Millstone Nuclear Power Plant, Unit 2, License Amendment No. 240, dated February 10, 2000.

Attachment 1A to AEP:NRC:2039 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW PROPOSED CHANGES REVISED PAGES UNIT 1 x

XIII 3/4 9-6 3/4 9-7 3/4 9-8 B3/4 9-2

INDEX DEFINITIONS SECTION PAGE 3/4.9 REFUELING OPERATIONS (Continued) 3/4.9.6 MANIPULATOR CRANE

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,y........................... .. 3/49-6 3/4.9.7 D A NE TRAVEL SPENT FUEL STORAIE POOL BILDIN .......... 3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ....................... 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ...................... 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL ......................................................... 3/4 9-11 3/4.9.11 STORAGE POOL WATER LEVEL .............................................................. 3/4 9-12 3/4.9.12 STORAGE POOL VENTILATION SYSTEM ................................................... 3/4 9-13 3/4.9.13 SPENT FUEL CASK MOVEMENT .............................................................. 3/4 9-17 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM ....................................... 3/4 9-18 314.9.15 STORAGE POOL BORON CONCENTRATION ............................................... 3/4 9-19 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN .......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS .............. 3/4 10-2 3/4.10.3 PRESSURE/TEMPERATURE LIMITATION-REACTOR CRITICALITY ............... 3/4 10-3 3/4.10.4 PHYSICS TESTS ..................................................................................... 3/4 10-5 3/4.10.5 NATURAL CIRCULATION TESTS .............................................................. 3/4 10-6 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID HOLDUP TANKS ......................................................................... 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas M ixture ............................................................................... 3/4 11-2 Gas Storage Tanks .................................................................................... 3/4 11-3 X AMENDMENT 4-O, 189 COOK NUCLEAR PLANT-UNIT 1

INDEX DEFINITIONS PAGE SECTION 3/4.7 PLANT SYSTEMS (Continued) 3/4.7.8 HYDRAULIC SNUBBERS ............................................................................... B 3/4 7-5a 3/4.8 ELECTRICAL POWER SYSTEMS ........................................................................ B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ............................................................................ B 3/4 9-1 3/4.9.2 INSTRUMENTATION ................................................... B 3/4 9-1 3/4.9.3 DECAY TIME ............................................................. B 3/4 9-1 CONTAINMENT BUILDING PENETRATIONS .................................................... B 3/4 9-1 3/4.9.4 3/4.9.5 COM M UNICATIONS ..................................................................................... B 3/4 9-1 3/4.9.6 MANIPULA T  :

OR CRANE O*D*.IT: .* tr. ...... .................................... B 3/4 9-2 CANtraEL SEtrFUEL STORAGE BUILDt...4............... B 3/4 9-2 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ............................. B 3/4 9-2 3/4.9.8 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ............................ B 3/4 9-2 3/4.9. 10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL ..................................................................... B 3/4 9-3 B 3/4 9-3 3/4.9.12 STORAGE POOL VENTILATION SYSTEM ........................................................

B 3/4 9-4 3/4.9.13 SPENT FUEL CASK MOVEMENT ....................................................................

B 3/4 9-4 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM .............................................

.............................. B 3/49-4 3/4.9.15 STORAGE POOL BORON CONCENTRATION 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOW N MARGIN ................................................ . . .................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIM ITS ............................................................................................ B 3/4 10-1 COOK NUCLEAR PLANT-UNIT I xilI AMENDMENT 22, InQ, 173, -89, 208

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3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS CRANE TRAVEL SPENT FUEL STORAGE POOL BUILDING' LIMITING CONDITIN OPERATION

'FOR 3.9.7 Loads in exceess of 2,500 pounds shall be pr-ohibitecd fr:om travel over- fuel assemblies in the stor.age pool.

Leadsaried over the spent fuel pool and the heights at which they may be carried over-rck containing fuelI shall be limifted in such a way as to pr-eclude impact ener-gies over- 21,210 in. lbs., if th-e loads aredroped from the cane APPLiCABILII.ITTYV. Wihfail assemblies m me storage pool.

AgTION 7 .

With the requirements of the above specification net satisfied, place he cae ladd in1a safe condition. TPhe pr-ovisions of Specification 3.0.3 are not applicable.

CTTDVDTTANC A=

TEUE--- lylTfllV -9' 4.9.7.1 Crn interlock whc rvent crane trave with loads' i excess of 2,500 pounds over- fuel assemblies-shall be ddemonstratd OPERABLE within 7 days prior- to cr-ane use andatlstocpe 7 days thereafter- during cr-ane operation.

This Surveillance Requirement is not requird during the movement of steam generator- sections i the auxjiliay building for the Unit 1 steam generator- replacement project. Whlen crane travel inter-locks arc disengaged, administrative contr-ols shal be in place to prevent loads from passing over- the spent fuel pool.

4.9.7.2-; The potential impact energy due to dropin the cr-ane's-load shall be determined to e 44-242240 in. lbs. prior to moving eac load ovear rakscntaining fuel Shared system with Cook Nuclear Plant Unit COOK NUCLEAR PLANT-UNIT 1 Page 3/4 9-8 AMENDMENT 10", 143, 486, 233 I

3/4 BASES 3/4.9 REFUELING OPERATIONS 3/4.9.6 MANIPULA-TOR CRANE OPERRA3IIT T mE W0 The OPERABILITY requirements for the maiuto caes ensure that: 1) manipulator cranes will be used for movement of-control rods and fuel assemblies 2)each cr-ane has sufficient load c-apacity to lift a controel rod or fuel assembl and 3) the cor-e internals and pressure vessel are proetected from excessive lifting forcee in the evn they are inadvrtnty enagaged during lifting operations,.

3/4.9.7 CAN TRAEL SPENT FUEL STOAEBIDN I5E T~

The r-estriction on movement of loads in excess of 2,500 lbs. over other- fuel assemablies; in'toe storage poo ensures that, in the event of a droepped load, 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the stor-age racks will not r-esult in a cer-it-ical aarray. The 2,;50 lb. load restriction is based on the combined nominal weigh of a fuel assembly, a contrzol rod assembly, and an associated fuel handlfing tool. Release of activity from a single fuel assembly is coensistent with toe assumptionfo

- ... accident.

the analysis for a fehading The restriction on movements of loads in excess of toe impact energy limit, which is based on the kinetic energy of a dr-opped-fuel assembly and control rod assembly from 15" above the fuel storage r-ack, is to bound oter leads.

Prohibiting loads greater- than 2,500 pounds or- loads at heights that would exceed toe kinetic energy impacat limi allows flexibility in the movements of fue and other- relatively l-Ight loads, while providing r-easonable assurance tIlnt the cneu nce f A fue handlino accident will not be exceeeded.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140OF as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 9-2 AMENDMENT -8 Revised 7/2/99

Attachment 1B to AEP:NRC:2039 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW PROPOSED CHANGES REVISED PAGES UNIT 2 x

XIII 3/4 9-6 3/4 9-7 B3/4 9-2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 23/4.9 REFUELING OPERATIONS (Continued) 3/4.9.6 AU4^TA,1 CRT rr CG=ru...... ANE ........................................................ 349-6 3/4.9.7 CRAN SPE FL SOR.AG. P B .................... 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION .................................... 3/4 9-8 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM .................................. 3/4 9-9 3/4.9.10 WATER LEVEL-REACTOR VESSEL .................................................................................. 3/4 9-10 3/4.9.11 STORAGE POOL WATER LEVEL ....................................................................................... 3/4 9-11 3/4.9.12 STORAGE POOL VENTILATION SYSTEM ...................................................................... 3/4 9-12 3/4.9.13 SPENT FUEL CASK MOVEMENT ...................................................................................... 3/4 9-16 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM ......................................................... 3/4 9-17 3/4.9.15 STORAGE POOL BORON CONCENTRATION .................................................................. 3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ......................................................................................................... 3/410-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS .................................................................................................................................... 3/4 10-2 3/4.10.3 PHY SICS TESTS .................................................................................................................... 3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS .............................................................................................. 3/4 10-4 3/4.10.5 POSITION INDICATOR CHANNELS - SHUTDOWN ........................................................ 3/4 10-5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID HOLDUP TANKS ........................................... *........................................................ 3/411-1 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas M ixture ............................................................................................................ 3/4 11-2 G as Storage Tanks .................................................................................................................. 3/4 11-3 X AMENDMENT UP0-, 175 COOK NUCLEAR PLANT-UNIT 2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS (Continued) 3/4.7.8 SEALED SOURCE CONTAMINATION ............................................................................... B 3/4 7-6 3/4.8 ELECTRICAL POWER SYSTEMS .......................................................................................... B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ................................................................................................ B 3/4 9-1 3/4.9.2 INSTRUM ENTATION .......................................................................................................... B 3/4 9-1 3/4.9.3 D ECAY TIM E ......................................................................................................................... B 3/4 9-1 CONTAINMENT BUILDING PENETRATIONS ................................................................. B 3/4 9-1 3/4.9.4 COM MUNICATIONS .......................................................................................................... B 3/4 9-1 3/4.9.5 3/4.9.6 MA1N n111TO CRANE.QN CPED_ ART nr y ...................... .............. B 3/4 9-2 3/4.9.7 CRANE *^xrEt SPEN t - L

^ELrS...........TEB G.............. B 3/4 9-2 3/4.9.8 RESIDUAL HEATREMOVAL AND COOLANT CIRCULATION .................................... B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM .................................. B 3/4 9-3 9-3 3/4.9.10 and 3/4 9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL ....................... B 3/4 B 3/4 9-3 3/4.9.12 STORAGE POOL VENTILATION SYSTEM .......................................................................

B 3/4 9-4 3/4.9,13 SPENT FUEL CASK MOVEMENT .......................................................................................

B 3/4 9-4 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM .........................................................

B 3/4 9-4 3/4.9.15 STORAGE POOL BORON CONCENTRATION ..................................................................

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ................................................................................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS .......................... B 3/4 10-1 COOK NUCLEAR PLANT-UNIT 2 XIII AMENDMENT 4407,1t-56,1t-7-5,192

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS FTANTTDTTTATf*OR CANTE-GRERI'DABILTT 1* Y LIMITING CONDITION FOR OPERATION 3.9.6 The manipulator crane and auxiliary hoist s.ed r mve. ment of c.ntrol rods or- fue assembl and shall be OPERABLEAwt: LLttý{$S1A

a. The manipulator crane used for movement of fue~l assemblies having:

4.b-- A mkinimum capacity of 3250 pounds, an's 2- An everload cut off limit Cg2850 pounds.

b. The auxiliary heist used for- movement of control r-ods having:

4- A minimum capacity of 700 pounds, an.

A load indicat*r* which shal be used to prevent lifting loads in excess of 600 pounds.

ACh 7 e"eTtT-ION v Wjth the requirements fo-r crE-ane and/orf hoist QPERABILIT-Y net satisfied, suspend use of any inoperabl manipulator- crae and/or auxiliar-y hoist from oper-ations involving the movement of control r-ods and fuel assemblies within the r-eactor pressur-e vessel. The pr-ovisions of Specification 3.0.3 are not applicable.

SUVT ELLTTTTANrCE REQUT CIREMETS 4.9.6.1 Each manipulator- crane used for- mevement of fue assemblies within the rcactor- pressure vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior- to the start of such oper-ations by performing a load test of at least 32L50 pounds and demonstrating an automatic load cut off when the craane load exceeds 2850 pounds.

1.9.6.2 Each auxiliary hoist and associated load indicator- used- for- movement of control roeds withinth reactor- pr-essure vessel shall be demonstrated OPERAý.BLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior- to the start ot such operations by performing a load test of at least:700 pounds.-

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 9-6

314 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS C-.R ATNE TRAVEL SPENT FUEL STORAGE POOL BULDTINTG*

3.9.7 Loads in eces of 2,500 pounds shall be prohibited froem tral o1ver fue.asemli..i.th.. *v * * -peeb Loads carried over- the spent fuel pool and the heights- at. which they may be carri-ed overF rack containinig- fue sall be limited in such a way as to pr-eclude limpact ener-gies ovxer 21,210 in. lbs., if the loads are droepped from the cae.-

.1 APPLICABIL.TY: -Wwa mPi assemones m mec storage pel..

AcWTIONI With the requircments of the above se-cifiaton not satisfied, plae the crane load in a safe cendition. The provisions .f Specification3.03acn icable.

CLTTWLLT TTANURE'vE D lT.DRNTVT9 4.9.7.1 Crneinerlocks whic prevent crane travel with loads in exceess of 2,500 pounds over- fuc assemblies sh"ll be demonstrated OPER.A..BLE within 7 days prior to c.rane use and at least one. per

7 days thereafter- during cr-ane operation.

ThisSureillnceRequirement is not r-equired during .the movement of steam generator- sections in the auxiliar-y bilding for- the Unit 1 steam gener.atorrplacement pro.ject. When crane travel interlcks are disengaged, admiistrative contls shall be in placo to prevent leads from passing Over- the spent fuel poo.

1.The2 petentia impact energy due to droping the crane's lad shall be determined to be *921,210 in.

lbs. prior- to moving each load over racsks containing fuel.

Shared systAem-with Cook Nucalear- Plant Unit 1.

COOK NUCLEAR PLANT-UNIT 2 Page3/4 9-7 AMENDMENT 8=7, 96, 4-72, 216

3/4 BASES 3/4.9 REFUELING OPERATIONS T t 3/4.9.6 MANIPULA OtRCRANE Q t rrvt'rvI The OPERABILITY r-equirements for the manipulator- crans ensure that: I) manipulatort cranes mill be used for movement of control rods and fuel assemblies, 2) each cr-ane has sufficient load capacity te lift a control! rod or fuel assembly, and 3) the core internals and pr-essure vessel arc protected from excessive lifting forcse in the even they are inavertenly engaged during lifting operations.

3/4.9.7 C--R-ANE TRAVEL SPENT FUEL STORAGE BUILDING'25 M The r-estriction on movement of loads in excess of 2,500 lbs. over other- fuel assemblies in the stor-age pool ensures that, in the-event of a dropped load, 1) the activity release will be li-mited to that contained in a single fuel assembly, and 2) any possile dýýistrin of fujel in the stor ageraks will not result in a critical array. The 2,50 lb. load r-estiction is based on the cembi-ned nomna wegh of a-fel assembly, a controel rod assembly, and an associated fuel handling tool. Release of activity from a single fuel assembly is consistent with the assumptionfo the analysis for a fuel handling accident.

The restiction on movements of loads in-ex.ess of the impact energy limit, which is based on the kinetic. ener of a droppedf assembly and control rod assembly from. 15" above the fuel stor.age r-ack, is to bound other leads.

Prohibkitig loads greater- than 2,500 pounds or- loads at height tha woldexed thle kinetic energy impact limit allows flexibility in the movements of fuel and other reaieylgtloads, while prov~iding reasonable assuranc&

that the consequences of a fuel handling accident w.ill not be exceeded-.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140OF as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

COOK NUCLEAR PLANT-UNIT 2 Page B 3/4 9-2 AMENDMENT 59 Revised 7/2/99

Attachment 2A to AEP:NRC:2039 PROPOSED TECHNICAL SPECIFICATION PAGES REVISED PAGES UNIT I x

XIII 3/4 9-6 3/4 9-7 3/4 9-8 B3/4 9-2

INDEX DEFINITIONS PAGE SECTION 3/4.9 REFUELING OPERATIONS (Continued)

DELETED ............................................................................................ 3/4 9-6 3/4.9.6 3/4.9.7 DELETED ............................................................................................. 3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ....................... 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ...................... 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL ......................................................... 3/4 9-11 3/4.9.11 STORAGE POOL WATER LEVEL .............................................................. 3/4 9-12 3/4.9.12 STORAGE POOL VENTILATION SYSTEM ................................................... 3/4 9-13 3/4.9.13 SPENT FUEL CASK MOVEMENT .............................................................. 3/4 9-17 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM ....................................... 3/4 9-18 STORAGE POOL BORON CONCENTRATION ............................................... 3/4 9-19 3/4.9.15 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOW N M ARGIN ............................................................................. 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS ................ 3/4 10-2 3/4.10.3 PRESSURE/TEMPERATURE LIMITATION-REACTOR CRITICALITY ................ 3/4 10-3 PHYSICS TESTS ..................................................................................... 3/4 10-5 3/4.10.4 3/4.10.5 NATURAL CIRCULATION TESTS ............................................................. 3/4 10-6 3/4.11 RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS ........................................................................ 3/4 11-1 3/4.11.1 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas M ixture .............................................................. 3/4 11-2 Gas Storage Tanks .................................................................................... 3/4 11-3 X AMENDMENT 10, 189 COOK NUCLEAR PLANT-UNIT 1

INDEX DEFINITIONS PAGE SECTION 3/4.7 PLANT SYSTEMS (Continued)

B 3/4 7-5a 3/4.7.8 HYDRAULIC SNUBBERS ...............................................................................

B 3/4 8-1 3/4.8 ELECTRICAL POWER SYSTEMS ........................................................................

3/4.9 REFUELING OPERATIONS B 3/4 9-1 3/4.9.1 BORON CONCENTRATION ...........................................................................

B 3/4 9-1 3/4.9.2 INSTRUMENTATION ....................................................................................

DECAY TIME ............................................................................................ B 3/4 9-1 3/4.9.3 B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ....................................................

B 3/4 9-1 3/4.9.5 COMMUNICATIONS .....................................................................................

............................................................................................ B 3/4 9-2 3/4.9.6 DELETED

............................................................................................ B 3/4 9-2 3/4.9.7 DELETED B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION .............................

B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ............................

3/4.9. 10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL ........... ..................................... B 3/4 9-3 B 3/4 9-3 3/4.9.12 STORAGE POOL VENTILATION SYSTEM ........................................................

............... B 3/4 9-4 3/4.9.13 SPENT FUEL CASK MOVEMENT ..................................................

B 3/4 9-4 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM .............................................

B 3/4 9-4 3/4.9.15 STORAGE POOL BORON CONCENTRATION ....................................................

3/4.10 SPECIAL TEST EXCEPTIONS

. . ................... B 3/4 10-1 3/4.10.1 SHUTDOWN MARGIN ....................................................

3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS ............................................................................................ B 3/4 10-1 XIII AMENDMENT 22, 120, 173, M89, 208 COOK NUCLEAR PLANT-UNIT 1

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS 3.9.6 DELETED COOK NUCLEAR PLANT-UNIT 1 Page 3/4 9-6

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK COOK NUCLEAR PLANT-UNIT 1 Page 3/4 9-7

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS 3.9.7 DELETED COOK NUCLEAR PLANT-UNIT 1 Page 3/4 9-8 AMENDMENT M10,-143, -186, 233

3/4 BASES 3/4.9 REFUELING OPERATIONS 3/4.9.6 DELETED 3/4.9.7 DELETED 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION operation ensures that (1) sufficient The requirement that at least one residual heat removal (RHR) loop be in the reactor pressure vessel below cooling capacity is available to remove decay heat and maintain the water in circulation is maintained through 140OF as required during the REFUELING MODE, and (2) sufficient coolant boron stratification.

the reactor core to minimize the effect of a boron dilution incident and prevent feet of water above the reactor The requirement to have two RHR loops OPERABLE when there is less than 23 not result in a complete loss of pressure vessel flange ensures that a single failure of the operating RHR loop will 23 feet of water above the reactor residual heat removal capability. With the reactor vessel head removed and in the event of a failure of the pressure vessel flange, a large heat sink is available for core cooling. Thus, to cool the core.

operating RHR loop, adequate time is provided to initiate emergency procedures 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM will be automatically The OPERABILITY of this system ensures that the containment vent and purge penetrations OPERABILITY of this system is isolated upon detection of high radiation levels within the containment. The to the environment.

required to restrict the release of radioactive material from the containment atmosphere AMENDMENT 78 COOK NUCLEAR PLANT-UNIT 1 Page B 3/4 9-2 Revised 7/2/99

Attachment 2B to AEP:NRC:2039 PROPOSED TECHNICAL SPECIFICATION PAGES REVISED PAGES UNIT 2 x

XIII 3/4 9-6 3/4 9-7 B3/4 9-2

INDEX REQUIREMENTS LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE PAGE SECTION 23/4.9 REFUELING OPERATIONS ContiLueM) 3/4 9-6 3/4.9.6 DELETED ...............................................................................................................................

3/4 9-7 3/4.9.7 DELETED ..............................................................................................................................

3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ....................................

3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ..................................

3/4 9-10 3/4.9.10 WATER LEVEL-REACTOR VESSEL ..................................................................................

3/4 9-11 3/4.9.11 STORAGE POOL WATER LEVEL .......................................................................................

3/4 9-12 3/4.9.12 STORAGE POOL VENTILATION SYSTEM ......................................................................

3/4 9-16 3/4.9.13 SPENT FUEL CASK MOVEMENT .......................................................................................

3/4 9-17 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM .........................................................

3/4 9-18 3/4.9.15 STORAGE POOL BORON CONCENTRATION ..................................................................

3/4.10 SPECIAL TEST EXCEPTIONS 3/4 104 3/4.10.1 SHUTDOWN MARGIN .......................................................................................................

3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION 3/4 10-2 LIMITS ....................................................................................................................................

3/4 10-3 3/4.10.3 PHYSICS TESTS ...................................................................................................................

3/4 10-4 3/4.10.4 REACTOR COOLANT LOOPS ..............................................................................................

3/4 10-5 3/4.10.5 POSITION INDICATOR CHANNELS - SHUTDOWN .............. *.........................................

3/4.11 RADIOACTIVE EFFLUENTS 3/411-1 3/4.11.1 LIQUID HOLDUP TANKS ....................................................................................................

3/4.11.2 GASEOUS EFFLUENTS Explosive Gas Mixture ............................................................................................................ 3/4 11-2 Gas Storage Tanks ................................................................................................................. 3/4 11-3 X AMENDMENT 07, 175 COOK NUCLEAR PLANT-UNIT 2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS (Continued) 3/4.7.8 SEALED SOURCE CONTAMINATION ............................................................................... B 3/4 7-6 3/4.8 ELECTRICAL POWER SYSTEMS ........................................................................................... B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ................................................................................................. B 3/4 9-1 3/4.9.2 INSTRUMENTATION .......................................................................................................... B 3/4 9-1 3/4.9.3 DECAY TIME ......................................................................................................................... B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ................................................................. B 3/4 9-1 3/4.9.5 COMMUNICATIONS ........................................................................................................... B 3/4 9-1 3/4.9.6 DELETED ............................................................................................................................. B 3/4 9-2 3/4.9.7 DELETED .......................................................................................................................... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION .................................... B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM .................................. B 3/4 9-3 3/4.9.10 and 3/4 9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL ....................... B 3/4 9-3 3/4.9.12 STORAGE POOL VENTILATION SYSTEM ....................................................................... B 3/4 9-3 3/4.9.13 SPENT FUEL CASK MOVEMENT ....................................................................................... B 3/4 9-4 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM ......................................................... B 3/4 9-4 3/4.9.15 STORAGE POOL BORON CONCENTRATION .................................................................. B 3/4 9-4 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ................................................................................................ B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS .......................... B 3/4 10-1 COOK NUCLEAR PLANT-UNIT 2 XIIII AMENDMENT IP9, 4--56,1t-75, 192

314 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS 3.9.6 DELETED COOK NUCLEAR PLANT-UNIT 2 Page 3/4 9-6

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS 3.9.7 DELETED COOK NUCLEAR PLANT-UNIT 2 Page 3/4 9-7 AMENDMENT 8-7, 96, =7-, 216

3/4 BASES 3/4.9 REFUELING OPERATIONS 3/4.9.6 DELETED 3/4.9.7 DELETED 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140OF as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

reactor The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the loop will not result in a complete loss of pressure vessel flange ensures that a single failure of the operating RHR and 23 feet of water above the reactor residual heat removal capability. With the reactor vessel head removed of the pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

Page B 3/4 9-2 AMENDMENT 59 COOK NUCLEAR PLANT-UNIT 2 Revised 7/2/99

Attachment 3 to AEP:NRC:2039 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 I&M will revise the appropriate sections of the UFSAR to include Within 30 days from requirements and information from TS 3/4.9.6, 3/4.9.7 and associated date of approval of Bases. license amendment request.