ML17334B344

Application for Amends to Licenses DPR-58 & DPR-74,changing Westinghouse Fuel & Reload Analysis Methodology

ML17334B344
Person / Time
Site:	Cook
Issue date:	02/06/1990
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
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ML17328A558	List: ML17328A557 ML17334B344
References
AEP:NRC:1071E, NUDOCS 9002120256
Download: ML17334B344 (184)

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ACCELERATED DI UTION DEMONS TION SYSTEM 4'

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ESSION NBR:9002120256 DOC.DATE: 90/02/06 NOTARIZED: NO DOCKET g FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana & 05000315 50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana & 05000316 AUTH. NAME AUTHOR AFFILIATXON ALEXICH,M.P. Xndiana Michigan Power Co. (formerly Indiana & Michigan Ele RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

S UBJECT: Application for amends to Licenses DPR-58 & DPR-74,changing I Westinghouse fuel & reload analysis methodology.

D DISTRIBUTION CODE: AOOID COPIES RECEIVED:LTR J ENCL I SIZE: 3 Z + tgO TITLE: OR Submittal: General Distribu on S NOTES: /

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL XD CODE/NAME LTTR ENCL PD3-1 LA 1 1 PD3-1 PD 1 1 D

GIITTER,J. 5 5 INTERNAL: NRR/DET/ECMB 9H 1 1 NRR/DOEA/OTS B1 1 1 1 D NRR/DST 8E2 1 1 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR/DST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 OC 1 0 OGC/HDS1 1 0 G FI E 01 1 1 RES/DSIR/EIB 1 1 RNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 lVQ.S-(C hag~r l~

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D NOTE TO ALL "RIDS" RECIPIENTS:,,

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOC/MENT CONTROL DESK, ROOM Pl-37 (EXT: 20079) TO ELIMINATEYOUR NAME FROQ DISI'RIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TAL NUMBER OF COPIES REQUIRED: LTTR ENCL

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Indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 NHEMSlMSL uacuanmu PQWM AEP:NRC:1071E Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 UNIT NO. 2 CYCLE 8 RELOAD LICENSING, PROPOSED TECHNICAL SPECIFICATIONS FOR UNIT 2 CYCLE 8, AND RELATED UNIT 1 PROPOSALS U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Attn: T. E. Murley February 6, 1990

Dear Dr. Murley:

This letter and its attachments constitute an application for amendment to the Technical Specifications (T/Ss) for Donald C. Cook Nuclear Plant Units 1 and 2. This amendment is requested to support the Cycle 8 reload of Unit 2. Indiana Michigan Power Company will reload the Donald C. Cook Nuclear Plant Unit No. 2, Cycle 8 with Westinghouse Vantage 5 (V5) fuel assemblies. Westinghouse has replaced Advanced Nuclear Fuels Corporation (ANF) as the fuel supplier for Unit 2. The majority of these proposed T/S changes are related to the change to Westinghouse fuel and reload analysis methodology. As discussed below, certain related Unit 1 T/S changes are also proposed. Entry into Mode 4 for Cycle 8 is anticipated to occur on or about August 24; 1990.

Content of the Submittal This submittal addresses two issues in addition to the proposed T/S changes for the Unit 2 core for Cycle 8. These are:

If) The Unit Licensing Basis

1) 2 K~V)A i

'OOQ INO When Unit 2 was relicensed for Cycle 6 operation, a newly OO revised Exxon Nuclear Company, Inc. (now Advanced Nuclear 0IO 0O Fuels Corporation) methodology was employed. This WhC methodology was based on the NRC's Standard Review Plan lA0

'blO (NUREG-0800). As a result, seven events not in the Unit 2 o~

licensing basis were analyzed. These seven events will N

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Dr. T. E. Murley AEP:NRC:1071E not be analyzed for Unit 2 Cycle 8 or subsequent cycles.

This issue was discussed with the staff on June 12, 1989.

It is addressed in more detail in Attachment 8.

2) Proposed Changes to the Unit 1 T/Ss There are a few changes to the Unit 1 T/Ss being proposed.

These occur where the justification for a proposed Unit 1 T/S change is essentially identical to the justification for a similar change to the Unit 2 T/Ss. By proposing the change for both units, efficiency is achieved in the review effort. In addition, the T/Ss for the units are maintained more nearly alike.

The Generic Letter 88-16 Submittal In parallel with this submittal, we are submitting our proposed T/S changes for Unit 2 in response to Generic Letter 88-16, "Removal of Cycle Specific Parameter Limits from Technical Specifications." Our identifier for the Generic Letter submittal is AEP:NRC:1077A. In AEP:NRC:1077A we propose a Core Operating Limits Report (COLR) for Donald C. Cook Nuclear Plant Unit 2. We will need your response to this submittal as soon as possible. If the staff cannot approve our Generic Letter 88-16 submittal, it will be necessary to propose additional T/S changes for Cycle 8 that would have been submitted in the COLR document. These proposed T/S changes would include:

o moderator temperature coefficient (MTC),

MTC 300 ppm MTC surveillance acceptance criterion, all rods out (ARO) position and control rod insertion limits, axial flux difference allowable deviation, and axial flux difference target band, o F and K(Z)

N o F H

and F H

slope The values of most of these parameters are different from those for Cycle 7. Attachment 7 contains the values of the above parameters currently planned for Cycle 8.

Please advise us before March 31, 1990 of your intentions regarding our COLR submittal so that we can take appropriate action to ensure timely approval of this submittal.

Dr. T. E. Murley AEP:NRC:1071E Or anization of the Submittal This submittal is organized to facilitate the reviewer's task. A detailed description of the organization of the submittal is found at the beginning of Attachment 1. This description will direct the reviewer to the locations of the significant hazards consideration analysis, proposed T/S changes, and supporting documentation.

Other Licensin Considerations

1) Environmental Aspects of Extended Burnup Fuel The Unit 2 Cycle 8 fresh fuel assemblies will be limited to 4.2 weight percent U-235 and at discharge will not exceed 56,000 MWD/MTU. The environmental aspects of extended burnup fuel were addressed in a previous submittal identified as AEP:NRC:1071F.

2) Feedwater System Malfunctions Causing an Increase in Feedwater Flow Review of this non-LOCA accident is continuing. If this review requires any change to the information supplied in Attachment 4, Appendix B, Section B.3.8a.2, the staff will be notified prior to April 15, 1990. We do not anticipate that any change will be required.

These proposed T/S changes have been reviewed by the Plant Nuclear Safety Review Committee and by the Nuclear Safety and Design Review Committee.

In compliance with the requirements of 10 CFR 50.91(b)(10), copies of this letter and its attachments have been transmitted to Mr. R. C. Callen of the Michigan Public Service Commission and the Michigan Department of Public Health.

This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.

Sincerely, M. P. Alexich Vice President ldp Attachments

Dr. T. E. Murley AEP:NRC:1071E cc: D. H. Williams, Jr.

A. A. Blind - Bridgman R. C. Callen G. Charnoff A. B. Davis - Region III NRC Resident Inspector - Bridgman NFEM Section Chief

SAFETY EVALUATION FOR THE DONALD C. COOK NUCLEAR PLANT UNIT 2 TRANSITION TO WESTINGHOUSE 17X17 VANTAGE 5 FUEL

TABLE OF CONTENTS Section ~Pa e

1.0 INTRODUCTION

AND CONCLUSIONS 2.0 MECHANICALEVALUATION 3.0 NUCLEAR EVALUATION 13 4.0 THERMALAND HYDRAULICEVALUATION 15, 5.0 ACCIDENT EVALUATION 20 6.0

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES 43

7.0 REFERENCES

50 APPENDIX A - TECHNICAL SPECIFICATIONS CHANGE PAGES APPENDIX B - NON-LOCA ANALYSES

. APPENDIX C - LOCA ANALYSES LIST OF TABLES Table No. Title ~Pa e 2.1 Comparison of 17x17 LOPAR and 17x17 VANTAGE 5 Fuel Assembly Design Parameters 4.1 Thermal and Hydraulic Design Parameters 17 6.1 Summary of Technical Specifications Changes 44 LIST OF FIGURES ure No. Title ~Pa e

'Fi 2-1 Comparison of Westinghouse VANTAGE 5 and ANF 17x17 Fuel Assembly Dimensions 12 5.1 Design Axial Power Distribution for non-OTbT Transients 41 Most Negative Moderator Temperature Coefficient Limit 42

fC

1.0 INTRODUCTION

AND CONCLUSIONS Donald C. Cook Nuclear Plant Unit 2 (Cook Nuclear Plant Units 2) is currently operating in Cycle 7 with an Advanced Nuclear Fuels (ANF) core. Beginning with Cycle 8, it is planned to refuel and operate with the Westinghouse VANTAGE 5 improved fuel design except for the inclusion of a Debris Filter Bottom Nozzle instead of the VANTAGE 5 bottom nozzle. As a result, future transition core loadings would range from approximately 40% VANTAGE 5 and 60% ANF to eventually an all VANTAGE 5 fueled core. The VANTAGE 5 fuel assembly was designed as a modification to the current Westinghouse Optimized Fuel Assembly (OFA) design (Reference 1).

The VANTAGE 5 design features were conceptually packaged to be licensed as a single entity.

This was accomplished via the NRC review and approval of the "VANTAGE 5 Fuel Assembly Reference Core Report," WCAP-10444-P-A (Reference 2). The initial irradiation of a fuel region containing all the VANTAGE5 design features occurred in the Callaway Plant in November 1987.

The Callaway VANTAGE 5 licensing submittal was made to the NRC on March 31, 1987 (ULNRC-1470, Docket No. 50-483). NRC approval was received in October 1987. Several of the VANTAGE5 design features, such as axial blankets, Reconstitutable Top Nozzles, extended burnup modified fuel assemblies and Integral Fuel Burnable Absorbers have been successfully licensed as individual design features and are currently in operating Westinghouse plants.

A brief summary of the VANTAGE 5 design features and major advantages of the improved fuel design and the Debris Filter Bottom Nozzle are given below. These features and figures illustrating the design are presented in more detail in Section 2.0.

Inte ral Fuel Burnable Absorber IFBA - The IFBA features a zirconium diboride coating on the fuel pellet surface on the central portion of the enriched UO2 pellet stack. In a typical reload core, approximately one third of the fuel rods in the feed region are expected to include IFBAs.

IFBAs provide power peaking and moderator temperature coefficient control.

Intermediate Flow Mixer IFM Grids - Three IFM grids located between, the three uppermost Zircaloy-H grids provide increased DNB margin. Increased margin permits an increase in the design basis F ~ and Fg.

Reconstitutable To Nozzle RTN - A mechanical disconnect feature facilitates the top nozzle removal. Changes in the design of both the top and bottom nozzles increase burnup margins by providing additional plenum space for fission gas accommodation and room for fuel rod growth.

Extended Burnu - The VANTAGE 5 fuel design will be capable of achieving extended burnups.

The basis for designing to extended burnup is contained in the approved Westinghouse extended burnup topical WCAP-10125-P-A (Reference 3).

Blankets - The axial blankets consist of a nominal six inches of natural UO2 pellets at each end of the fuel stack to reduce neutron leakage. Loading patterns utilizing radial blankets are'shown to further improve uranium utilization and provide additional pressurized thermal shock margin.

Debris Filter Bottom Nozzle DFB - This bottom nozzle is designed to inhibit debris from entering the active fuel'region of the core and thereby improves fuel performance by minimizing debris related fuel failures. The DFBN is a low profile bottom nozzle design made of stainless steel, with reduced plate thickness and leg height. The DFBN is structurally and hydraulically equivalent to the existing low profile bottom nozzle.

This submittal is to serve as a reference safety evaluation/analysis report for the region-by-region reload transition from the present Donald C. Cook Unit 2 ANF fueled core to an all VANTAGE 5 fueled core. This submittal examines the differences between the VANTAGE 5 and the ANF ~

fuel assembly desings andevaluates the effect of these differences on the cores during the transition" to an all VANTAGE 5 core. Although it is anticipated that the Cook Nuclear Plant Unit 2 will be'initially operated in Cycle 8 at the currently licensed core power level of 3411 MWt, unless specifically indicated, the VANTAGE 5 core evaluations and analyses were performed to support an uprate to a core thermal power level of 3588 MWt. The following assumptions made in the safety evaluations and analyses: a full power of FNgH of 1.62 for the VANTAGE5 fuel and 1.55 for the ANF fuel, maximum Fg of 2.22 for the VANTAGE 5 fuel and 2.10 for the ANF fuel and a 15% peak and 10% average steam gerierator tube plugging level.

2

The approved Westinghouse Revised Thermal Design Procedure (RTDP) is used in the DNB analyses of both VANTAGE 5 and ANF fuel assemblies for all DNB related accidents, excluding transients such as the hypothetical steamline break where RTDP methodology is not applicable.

For such transients standard DNB design methods are used. The WRB-2 DNBR correlation is used in the VANTAGE 5 DNB analyses. The ANF fuel is analyzed by using the W-3 DNB

'orrelation.

The standard reload design methods described in Reference 4 and will be used as a basic reference document in support of future Cook Nuclear Plant Unit 2 Reload Safety Evaluations (RSEs) for VANTAGE5 fuel reloads. Sections 2.0 through 5.0 summarize the Mechanical, Nuclear, Thermal and Hydraulic, and Accident Evaluations, respectively. Section 6.0 gives a summary of the changed needed to the Technical Specifications. Appendices A and B contain the Technical Specification change pages and non-LOCA safety analyses results, respectively. Appendix C contains the large and small break LOCA safety analyses.

l Consistent with the Westinghouse standard reload methodology, Reference 4 parameters are chosen to maximize the applicability of the safety evaluations for future cycles. The objective of subsequent cycle specific RSEs will be to verify that applicable safety limits are satisfied based on the reference evaluation/analyses established in this RTSR.

In order to demonstrate early performance of the VANTAGE 5 design product features in a commercial reactor, four VANTAGE 5 demonstration assemblies (17x17) were loaded into the V. C. Summer Unit 1 Cycle 2 and began power production in December of 1984. These assemblies completed one cycle of irradiation in October of 1985 with an average burnup of 11,357 MWD/MTU. Post-irradiation examinations showed all 4 demonstration assemblies were of good mechanical integrity. No mechanical damage or wear was evident on any of the VANTAGE 5 components. Likewise, the IFM grids on the VANTAGE5 demonstration assemblies had no effect on the adjacent fuel assemblies. All four demonstration assemblies were reinserted into V. C.

Summer 1 for a second cycle of irradiation. This cycle was completed in March of 1987, at which time the demonstration assemblies achieved an average burnup of about 30,000 MWD/MTU. The observed behavior of the four demonstration assemblies at the end of 2 cycles of irradiation was as good as that observed at the end of the first cycle of irradiation. The four assemblies were inserted for a third cycle of irradiation which was completed in November of 1988 (EOC burnup

46,000 MWD/MTU). Post-irradiation examinations showed all four assemblies were still in good mechanical condition.

In addition to V. C. Summer, individual VANTAGE 5 product features have been demonstrated at other nuclear plants. IFBA demonstration fuel rods have been irradiated in Turkey Point Units 3 and 4 for two reactor cycles. Unit 4 contains 112 fuel rods equally distributed in four demonstration assemblies. The IFBA coating performed well with no loss of coating integrity or adherence. The IFM grid feature has been demonstrated at McGuire Unit 1. The demonstration assembly at McGuire was irradiated for three reactor cycles and showed good mechanical integrity.

Several full regions of VANTAGE 5 fuel are currently in operation.

The results of the safety evaluations and analyses described herein lead to the following conclusions:

1. The Westinghouse VANTAGE5 reload fuel assemblies for the Cook Nuclear Plant Unit 2 are mechanically and hydraulically compatible with the current ANF fuel assemblies, control rods, and reactor internals interfaces. The VANTAGE5 fuel assemblies satisfy the current design bases.

2. The VANTAGE 5 fuel assembly responses under seismic and LOCA excitations were determined using the analytical model representation of the reactor core. Analysis of the 17x17 VANTAGE 5 fuel assembly component stresses and grid impact. forces due to postulated faulted condition accidents veriGed that the VANTAGE 5 fuel assembly design is structurally acceptable.

3. Changes in the nuclear characteristics due to the transition from ANF to VANTAGE5 fuel will be within the range normally seen from cycle to cycle due to fuel management effects.

4. Plant operating limitations given in the Technical Specifications will be satisfied with the.

proposed changes noted in Section 6.0 of this report. The plant can safety operate at its

'current licensed power of 3411 MWt with average steam generator tube plugging levels up to 10% and peak plugging up to 15%. A reference is established upon which to base Westinghouse reload safety evaluations for future reloads with VANTAGE 5 fuel.

2.0 MECHANICALEVALUATION This Section evaluates the mechanical design and the compatibility of the 17x17 VANTAGE 5 fuel assembly with the current 17x17 ANF fuel assemblies during the transition through mixed-fuel cores to all VANTAGE5 fuel cores at the Cook Nuclear Plant Unit 2. The VANTAGE5 fuel assembly has been designed to be compatible with Westinghouse designed LOPAR and Optimized Fuel Assemblies (OFA), reactor internals interfaces, the fuel handling equipment, and refueling equipment.

The'ANTAGE 5 design is compatible with and is an acceptable replacement for the Cook Nuclear Plant Unit 2 containing fuel of the ANF 17x17 design. The VANTAGE 5 design dimensions, as shown in Figure 2.1, are essential equivalent to the ANF 17x17 design from an exterior assembly envelope and reactor internals interface standpoint. Table 2.1 provides a comparison of the VANTAGE 5 and ANF 17x17 fuel assembly design parameters. The design basis and design limits for VANTAGE 5 are essentially the same as those for the Westinghouse LOPAR design.

The significant new mechanical features of. the VANTAGE 5 design relative to the current ANF 17x17 fuel assembly include the following:

Integral Fuel Burnable Absorber (IFBA)

Intermediate Flow Mixer (IFM) Grids Reconstitutable Top Nozzle (RTN)

Extended Burnup Capability Axial Blankets The VANTAGE 5 fuel assembly design for Cook Nuclear Plant Unit 2 cycle 8 operation will also include the Debris Filter Bottom Nozzle (DFBN). The debris filter feature will reduce the possibility of fuel rod damage due to debris-induced fretting.

Fuel Rod Performance Fuel rod design evaluations for the VANTAGE 5 fuel are performed using the NRC approved models in References 5 and 6 and the NRC approved extended burnup design methods in Reference 3 to demonstrate that all fuel rod design bases are satisfied.

There is no effect from a full rod design standpoint due to having fuel with more than one type of geometry simultaneously residing in the core during the transition cycles. The mechanical fuel rod design evaluation for each region incorporates all appropriate design features of the region, including any changes to the fuel rod or pellet geometry from that of previous fuel regions.

Analysis of IFBA rods includes any geometry changes necessary to model the presence of the burnable absorber, and conservatively models the helium gas release from the ZrB2 coating.

Fuel performance evaluations are completed for each fuel region to demonstrate that the design criteria will be satisfled for all fuel rod types in the core under the planned operating conditions.

Any changes from the plant operating conditions originally evaluated for the mechanical design of a fuel region (for example, a power uprating or an increase in the peaking factors) are addressed for all affected fuel regions when the plant change is to be implemented.

Grid and Guide Thimble Assemblies VANTAGE 5 top and bottom grids are fabricated from Inconel with intermediate structural grids being fabricated from Zircaloy-4. The ANF spacer grids are bi-metallic and are constructed from Zircaloy< with Inconel springs. The VANTAGE5 top and'bottom Inconel grids (non-mixing vane type) have a snag-resistant design feature which minimizes assembly interaction during core loading/unloading. The VANTAGE 5 Inconel grids are also similar in design to the Inconel grids of the Westinghouse LOPAR fuel assemblies. Design differences between Westinghouse VANTAGE5 and LOPAR fuel assemblies include 1) the grid spring and dimple heights have been modified to accommodate a reduced diameter fuel rod, 2) the grid spring force has been reduced in the top grid and 3) grid straps are somewhat thicker and higher.

The Intermediate Flow Mixer (IFM) grids shown in Figure 2 are located in the three uppermost spans between the Zircaloy-4 mixing vane structural grids and incorporate a similar mixing vane array. The primary function of the IFM grid is to provide mid-span flow mixing in the hottest fuel assembly spans. Each IFM grid cell contains four dimples which are designed to prevent mid-span channel closure in the spans containing IFMs and fuel rod contact with the IFM mixing vanes..

/

This simplified cell arrangement allows for short grid cells so that the IFM grid can accomplish its flow mixing objective with minimal pressure drop.

The IFM grids are not intended to be structural members. The outer strap configuration was designed similar to current fuel designs to preclude grid hang-up and damage during fuel handling.

Additionally, the grid envelope is smaller which further minimizes the potential for damage and reduces calculated forces during seismic/LOCA events. A eoolable geometry is, therefore, assured at the IFM grid elevation, as well as at the structural grid elevation.

The VANTAGE 5. guide thimble ID provides an adequate nominal diametral clearance of 0.061 inch for the control rods. For accident analyses, a 2.7 seconds scram time to the dashpot is used for the VANTAGE 5 assembly. The 0.5 second incrase in rod drop time for VANTAGE 5 as compared to LOPAR is due mainly to larger fuel assembly pressure drop attributed to VANTAGE 5 IFM grids. The increase in pressure drop results in increased RCCA resistance during rod drop excursions. Using conservative analytical techniques, the results of rod drop time calculations for Cook Nuclear Plant Unit 2 indicate that the specific scram time to the VANTAGE 5 dashpot is within the 2.7 seconds Technical Specifications limit. Thus, all safety limits associated with RCCA scram are satisfied. The VANTAGE5 thimble tube ID provides sufficient diametral clearance for burnable absorber and source rods.

Based on evaluations of design differences, it is concluded that VANTAGE 5 is mechanically compatible with both ANF and Westinghouse LOPAR fuel assemblies. The VANTAGE5 fuel rod mechanical design bases remain unchanged from the Westinghouse LOPAR fuel assemblies used previously in Cook Nuclear Plant Unit 2.

Rod Bow It is predicted that the 17x17 VANTAGE 5 fuel rod bow magnitudes will be bounded by by Westinghouse 17x17 LOPAR assembly rod bow data. The current NRC approval methodology for comparing rod bow for different assembly designs is given in Reference 7.

Rod bow in the VANTAGE 5 fuel rods containing IFBAs is not expected to differ in magnitude or frequency from that currently observed in both Westinghouse LOPAR and OFA fuel rods under similar operating conditions. No indications of abnormal rod bow have been observed during visual or dimensional inspections performed on test IFBA rods. Rod growth measurements were also within predicted bounds.

Fuel Rod Wear Fuel rod wear is dependent on both the support provided by the fuel assembly and the flow environment to which it is subjected. Due to the VANTAGE 5 fuel assembly design employing different guide thimble tube diameter as compared to the ANF 17x17 design in addition to intermediate flow mixer (IFM) grids, an unequal axial pressure distribution results between the ANF and VANTAGE5 fuel assembly designs. Because of the major hardware differences between ANF 17x17'and VANTAGE5 design, evaluations were performed to evaluate hydraulically compatability of the two designs.

Hydraulic compatibility of VANTAGE 5 and ANF 17x17 fuel assembly designs was demonstrated by 1) showing that the ANF 17x17 fuel assembly was hydraulically compatible to the Westinghosue 17x17 OFA fuel assemblies, 2) referring to the study that showed Westinghouse 17x17 OFA fuel assemblies are hydraulically compatible with VANTAGE 5 fuel assemblies and 3) maMng direct analyses of hydraulic compatibility of the ANF 17x17 fuel assemblies to the VANTAGE 5 fuel assemblies.

The aforementioned evaluation demonstrated the ANF 17x17 fuel assembly design to be hydraulically compatible with the Westinghouse 17x17 OFA design. Evaluations have also been performed to demonstrate compatibility of the Westinghouse VANTAGE 5 and LOPAR fuel assembly designs. VANTAGE 5 fuel rod wear predictions were extrapolated from full scale hydraulic test of a VANTAGE 5 assembly adjacent to a 17x17 OFA assembly since vibration test results indicated that the crossflow effects produced by this fuel assembly combination would have the most detrimental effect on fuel rod wear.

Results of wear inspection and analysis discussed in Reference 2, Appendix A.1A revealed that the VANTAGE 5 fuel assembly wear characteristic was similar to that of the 17x17 OFA when both sets of data were normalized to the same test duration time. It was concluded that the VANTAGE 5 fuel rod wear would be less that the maximum wear depth established, Reference 8, for the 17x17 OFA at end-of-life condition.

.In the hydraulic test of the 17x17 Optimized Fuel Assembly, some grid cell sizes were set such that small gaps existed between the grid support points and fuel rod clad. Other cells were set with

various values of spring preload. These grid/clad support conditions compare favorably with those used in the fretting wear test perform'ed on the ANF 17x17 proof-of-fabrication fuel assembly. The clad wear results indicative of hydraulic testing of the 17x17 Optimized Fuel Assembly with gaps and minimum preload is a conserv'ative prediction of the 17x17 ANF wear during transition.

Seismic OCA Im act on Fuel Assemblies An evaluation of the VANTAGE 5 fuel assembly structural integrity considering the lateral effect of LOCA and seismic loading has been performed.

The VANTAGE 5 fuel assembly is structurally equivalent to the LOPAR and ANF fuel designs.

The main differences between these designs are six Zircaloy-4 grids, three additional IFM grids, and optimized fuel rods. The load bearing capability for the Zircaloy< grids and flow mixers under the faulted condition loadings has been analyzed. The results indicated that 17x17 VANTAGE 5 grid loads are well below the grid strengths.

Based on the grid load results, the 17x17 VANTAGE 5 Zircaloy-4 grid is capable of maintaining the core eoolable geometry under the combined Design Basis Earthquake and asymmetric pipe rupture transients in either all VANTAGE 5 or transition core operations. The 17x17 VANTAGE 5 fuel assembly is structurally acceptable for Cook Nuclear Plant Unit 2. This is also true for a transition core composed of VANTAGE 5, ANF and LOPAR fuel assembly core configurations. The grids of either fuel type will not buckle due to combined impact loads of seismic and LOCA events. There is no flow channel reduction during a LOCA; thus, the eoolable geometry requirement is met. The stresses in the fuel assembly components resulting from seismic and LOCA induced deflections are well within acceptable limits.

2.8 Core Components The core components for Cook Nuclear Plant Unit 2 are designed to be compatible with both LOPAR and VANTAGE5 fuel assemblies. The LOPAR and VANTAGE5 thimble tubes provide sufficient clearance for insertion of control rods and thimble plugging devices to assure proper operation of these components and fuel assembly. During Cycle 8 operation of Cook Nuclear Plant Unit 2, core components containing secondary source assemblies are restricted to locations consistent with ANF fuel assemblies.

The thimble plugs included in the plugging devices for Cook Nuclear Plant Unit 2 have been designed to be compatible with both LOPAR and VANTAGE 5 designs from a mechanical and thermal/hydraulic perspective. The ANF thimble tube ID is enveloped by both LOPAR and VANTAGE 5 designs; thus, the thimble plugs are also compatible with ANF fuel assemblies.

10

TABLE 2.1 Comparison of ANF 17x17 and W 17x17 VANTAGE 5 Assembly Design Parameters ANF 17x17 W 17x17 PARAMETER DESIGN VANTAGE5 DESIGN Fuel Assy Length, in 159.710 159.975 Fuel Rod Length, in 152.065 152.285 Assembly Envelope, (width), in 8.426 8.426 Compatible with Core Internals Fuel Rod Pitch, in .496 .496 Number of Fuel Rods/Assy Number/Guide Thimble Tubes/Assy 24 24 Number/Instrumentation Tube/Assy Fuel Tube Material Zircaloy-4 Zircaloy-4 Fuel Rod Clad OD, in 0.360 0.360 Fuel Rod Clad Thickness, in 0.0250 0.0225 Fuel/Clad Radial Gap, mil 3.5 3.1 Fuel Pellet Diameter, in .3030 .3088 Fuel Pellet Length Enriched Fuel, in .348 .370 Unenriched Fuel, in N/A .500 Guide Thimble Material Zircaloy-4 Zircaloy-4 Guide Thimble OD .480 , .474 (above dashpot), in 11

SCHEMATIC OF WESTINGHOUSE 17xl7 VANTAGE 5 FUEL ASSEMBLY 1.745 W 159.975 W 1 660 ANF 159.710 ANF 152.285 W 2.385 W 152.065 ANF 2.720 ANF 5 LEAF SPRINC 152.6SW 152WW 122.07W 111.6OW 101.52W 91.2SW 60.97W 70.70W 50.1SW 29.10W 6,07WQ 5.475 W 152 74 142.70 112.15 9L60 71.05 50.50 29.95 S.91 JNF 1 X580 ANF ANF ANF ANF JNF ANF ANF ANF W - WESTINGHOUSE 17x17 VANTAGE 5 FUEL ASSEMBLY DIMENSION+

ANF ADVANCED NUCLEAR FUEL 17x17 FUEL ASSEMBLY DIMENSION+

ANF GRID WIDTH

- 2.50 (vaney to valley)

- 2.98 (peek to peek)

WESTINGHOUSE GRID WIDTH

- top and bottom grid I.52 ' Cook Nuclear Plant Unit 2 rnid grlds 2.25 ~

Hrn grlds .475 ~

FIGURE 2.1

~ based on inner strap width 4 dimensions are ln inches (nominal) COMPARISON OF WESTINGHOUSE VANTAGE 5 and ANF 17x17 FUEL ASSEMBLY DIMENSIONS

3.0 NUCLEAR EVALUATION The evaluation of the transition and equilibrium cycle VANTAGE 5 cores presented in Reference 2, as well as the Cook Nuclear Plant Unit 2 specific transition core evaluations, demonstrate that the impact of implementing VANTAGE5 does not cause a significant change to the physics characteristics of the Cook Nuclear Plant Unit 2 cor'es beyond the normal range of variations seen from cycle to cycle.

The nuclear design philosophy, methods and core models used in the Cook Nuclear Plant Unit 2 reload transition core evaluations are described in References 2, 4, 9, 10 and 11. These licensed methods and core models have been used for Donald C. Cook Unit 1 and other previous Westinghouse reload designs using the OFA and VANTAGE 5 fuel. No changes from the above reference to the nuclear design philosophy, methods, or core models are necessary because of the transition to VANTAGE 5 fuel.

Based on the nuclear evaluation, the following Cook Nuclear Plant'Unit 2 Technical Specifications changes are proposed:

1) Increased F gH limits. These higher limits will allow loading pattern designs with reduced neutron leakage which in turn will allow longer cycles.

If

2) Increased Fg limit. This increased limit will provide greater flexibilitywith regard to accommodating the axially heterogeneous cores (axial blankets and reduced length burnable absorbers)

Power distributions and peaking factors show slight changes as a result of the incorporation of axial blankets and reduced length IFBAs in addition to the normal variations experienced with different loading patterns. The usual methods of enrichment variation and burnable absorber usage can be employed in the transition and full VANTAGE 5 cores to ensure compliance with the peaking factor Technical Specifications.

Evaluation of the key safety analysis parameters for the Cook Nuclear Plant Unit 2 reactor as it transitions to an all VANTAGE 5 core shows that the changes in values of the key safety analysis 13

parameters are typical of the normal cycle-to-cycle variations experienced as loading patterns change. As is current practice, each reload core design will be evaluated to assure that design and safety limits are satisfied according to the reload methodology. The design and safety limits will be documented in each cycle specific Reload Safety Evaluation (RSE) report which serves as a basis for any significant changes which may require a future NRC review.

14

4.0 THERMAL AND HYDRAULICEVALUATION The analysis of the ANF and VANTAGE 5 fuel is based on the Revised Thermal Design Procedure (RTDP) described in Reference 12. The ANF fuel analysis uses the W-3 DNB correlation described in References 13 and 14 and the VANTAGE 5 fuel uses the WRB-2 DNB correlation described in Reference 2. A 0.88 multiplier is applied to the W-3 DNB correlation to account for the 17x17 fuel rod diameter effect. The WRB-2 DNB correlation takes credit for the VANTAGE 5 fuel assembly mixing vane design. In addition the W-3 DNB correlation is used where appropriate. Table 4.1 summarizes the pertinent thermal and hydraulic design parameters.

The design method employed to meet the DNB design basis is the RTDP which has been approved by the NRC, Reference 12. Uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes are statistically combined with the DNB correlation uncertainties such that there is at least a 95 percent probability at a 95% confidence level that DNB will not occur 'on the most limiting fuel rod during normal operation and operational transients and during transient conditions arising from faults of moderate frequency (Condition I and II events as defined in ANSI N18;2). This gives the design limit DNBRs., Since the parameter uncertainties are considered in determining the design limit DNBR values, then the plant safety analyses are performed using values of input parameters without uncertainties. The design limit DNBRs are 1.23 and 1.22 for the typical and thimble cells respectively for VANTAGE 5 fuel and 1.39 and 1.36 for the typical and thimble cells respectively for ANF fuel. Standard Thermal Design Procedure (STDP) is used where the RTDP methodology is not applicable. In the STDP method the parameters used in analysis are treated in a conservative way so as to give the lowest minimum DNBR.

In addition to the above considerations, a plant specific DNBR margin has been considered in the analysis. In particular, DNBR safety analysis limits of 1.43 and 1.40 for the typical and thimble cells respectively for ANF fuel, and 1.69 and 1.61 for the typical and thimble cells respectively for VANTAGE 5 fuel were employed in the safety analyses. The differences between the design and safety analysis limits result in DNBR margin. A fraction of the margin is utilized to accommodate the transition core penalty. For VANTAGE 5 fuel this transition core penalty is a function of the number of VANTAGE 5 fuel assemblies in the core as given in Reference 15 and is based on a maximum value of 12.5%. There is no transition core penalty for ANF fuel for analyses using 15

cosine or positive axial offset axial power shapes. The transition core penalties for ANF fuel that occur with power shapes having large negative axial offsets are accounted for in the speciTic analyses that use these shapes. Additional margin is used to counteract rod bow. The fuel rod bow DNBR penalty is equal to 1.3% for VANTAGE 5 fuel (Reference 7) in the 20 inch grid spans.

No rod bow penalty is required in the 10 inch grid spans. There is no rod bow penalty for ANF fuel (Reference 16). The remaining margin, after consideration of these penalties, is reserved for flexibilityin the design. The plant specific DNBR margin, discussed above for RTDP, is preserved whenever STDP is used.

Hydraulic compatibility tests were performed by Combustion Engineering for the ANF 17x17 proof of fab fuel assembly. The'results of these tests were compared to hydraulic test data for the VANTAGE 5 fuel assembly (Reference 2). The data show that the ANF 17x17 fuel assemblies and the VANTAGE 5 fuel assembies are hydraulically compatible.

The Westinghouse transition core DNB methodology is given in References 1 and 17 and has been approved by the NRC via Reference 18. Using this methodology, transition cores are analyzed as if they were full cores of one assembly type (full ANF or full VANTAGE 5), applying

'he applicable transition core penalties.

The fuel temperatures used in safety analysis calculations for the VANTAGE5 fuel were calculated with the PAD performance code (Reference 6). This code was used to perform both design and licensing calculations. These fuel temperatures were used a's initial conditions for LOCA and non-LOCA transients.

16

TABLE 4.1 THERMAL AND HYDRAULICDESIGN PARAMETERS Bounding Parameters Bounding Parameters for Mixed Cores for Homogeneous VANTAGE 5 Thermal and H draulic Desi n Parameters Cores cle 10 & Be ond (Using RTDP)

Reactor Core Heat Output, MWt 3588 3588 Reactor Core Heat Output, 106 BTU/Hr 12243 12243 Heat Generated in Fuel, % '97.4 97.4 Core Pressure, Nominal, psia 2280 2130 F5I Nuclear Eethalpy Rise (ANF) 1.49[1+.2(1-P)]

Hot Channel Factor (V-5) 1.59[1+.3(1-P)] 1.59[1+.3(1-P)]

Safety Analysis Limit DNBR Typical How Channel (ANF) 1.43 (V-5) 1.69 1.69 Thimble (Cold Wall) How Channel (ANF) 1.40 (V-5) 1.61 1.61 DNB Correlation (ANF) W-3 (V-5) WRB-2 WRB-2 The 4% radial power uncertainty has been removed for statistical combination with other uncertainties in the RTDP analysis.

17

TABLE 4.1 (cont)

THERMAL AND HYDRAULICDESIGN PARAMETERS Bounding Parameters Bounding Parameters for Mixed Cores for Homogeneous VANTAGE 5 HFP Nominal Coolant Conditions cles 8 &9 Cores cle 10 & Be ond Vessel Minimum Measured Flow Rate (including Bypass) 106 ibm/hr 139.1 137.8 GPM 366,400 366;400 Vessel Thermal Design Flow Rate (including Bypass) 106 ibm/hr 134.6 133.2 GPM 354,000 354,000 Core Flow Rate (excluding Bypass, based on Thermal Design Flow) 106 ibm/hr 127.7 126.4 GPM 335,900 335,900 Fuel Assembly Flow Area for Heat Transfer, ft (ANF) 53.98 (V-5) 54.10 54.10 Core Inlet Mass Velocity, 106 ibm/hr-ft (Based on TDF) (ANF) 2.366 (V-5) 2.359 2.336 18

TABLE 4.1 (cont)

THERMAL AND HYDRAULICDESIGN PARAMETERS Bounding Parameters Bounding Parameters for Mixed Cores for Homogeneous VANTAGE5 Thermal and H draulic Desi n Parameters cles8&9 Cores cle 10 Ec Be ond (Based on Thermal Design Flow)

Nominal Vessel/Core Inlet Temperature, F 541.8 547.6 Vessel Average Temperature, F 576.0 581.3, Core Average Temperature, F 579.5 584.9 Vessel Outlet Temperature, F 610.2 615.0 Average Temperature Rise in Vessel, F 68.4 67.4 Average Temperature Rise in Core, F 71.7 70.6 t Heat Transfer Active Heat Transfer Surface Area, ft2 Average Heat Flux, BTU/hr-ft Average Linear Power, kw/ft (ANF/V-5) 57,505 (ANF/V-5) 207,410 Peak Linear Power for Normal Operation, kw/ft 5.72 13.3 57,505 207,410 5.72 13.3 Assumes all ANF or VANTAGE 5 core Based on 2.32 Fg peaking factor 19

5.0 ACCIDENT EVALUATION 5.1 Non-LOCA 5.1.1 Introduction This section addresses the impact of the complete transition, of Cook Nuclear Plant Unit 2 from ANF 17x17 fuel to Westinghouse 17x17 VANTAGE5 fuel on the FSAR Chapter 14 Non-LOCA Accident Analyses. The methods used for accident evaluation are described in Reference 4 and are discussed in further detail in Section 5.1.4.

The Cook Nuclear Plant Unit 2 licensing basis, as reported in the original FSAR (Reference 19) includes analyses or evaluations of fifteen (15) Non-LOCA accidents. These accidents are:

Uncontrolled RCCA Bank Withdrawal From a Subcritical Condition

b. Uncontrolled RCCA Bank Withdrawal at Power C. Rod Cluster Control Assembly (RCCA) Misalignment

d. Rod Cluster Control Assembly (RCCA) Drop

e. Uncontrolled Boron Dilution Loss of Forced Reactor Coolant Flow Startup of an Inactive Reactor Coolant Loop

h. Loss of External Electric Load or Turbine Trip Loss of Normal Feedwater Excessive Heat Removal due to Feedwater System Malfunction

k. Excessive Load Increase Loss of Offsite Power (LOOP) to the Station Auxiliaries Rupture of a Steamline (Steamline Break)

n. Rupture of a Control Rod Drive Mechanism (CRDM) Housing (Rod Cluster Control Assembly Ejection)

o. Major Rupture of Main Feedwater Pipe (Feedline Break)

All of the above fifteen Non-LOCA accidents have been reviewed to address any impact resulting from the VANTAGE5 fuel reload. The specific design associated with the VANTAGE5 fuel and the modified safety analysis assumptions that were considered in the Non-LOCA safety analysis are described in the following sections.

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5.1.2 VANTAGE 5 Design Features The design features of this VANTAGE 5 fuel reload transition that were considered in the Non-LOCA analysis and evaluations are:

. Intermediate Flow Mixer (IFM) Grids Axial Blankets Integral Fuel Burnable Absorbers (IFBAs)

Debris Filter Bottom Nozzle Reconstitutable Top Nozzle A brief description of each of these and its consideration in the Non-LOCA safety analyses follows:

IFM Grids The IFM grid feature of the VANTAGE 5 fuel design increases DNB margin. The fuel safety analysis limit DNB margin was set to ensure that the core thermal safety limits for the VANTAGE 5 fuel with an FNgH of 1.65 are acceptable. 'However, for the transition cycles the ANF fuel core thermal safety limits with FNgH of'1.55 are more restrictive. Thus, the more restrictive core limits correspond to the ANF fuel design. Any transition core penalty is accounted for with the available DNB margin.

The IFM grid feature of the VANTAGE 5 fuel design increases the core pressure drop. One result is that the control rod scram time to the dashpot has been increased to 2.7 seconds. This increased drop time primarily affects the fast reactivity transients which were reanalyzed for this report. The revised control rod drop time was incorporated in all the reanalyzed events requiring this parameter change. The Startup of an Inactive Reactor Coolant Loop transient not analyzed for this report has been evaluated for this parameter change.

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Axial Blankets and IFBAs Axial blankets reduce power at the ends of the rod which increases axial peaking at the interior of the rod. This effect is offset by the presence of part length IFBAs which flatten the power distribution. The net effect on the axial shape is a function of the number and configuration of IFBAs in the core and the time in core life. The effects of axial blankets and IFBAs on the reload safety analysis parameters are taken into account in the reload design process. The axial power distribution assumption in the safety analyses kinetics calculations have been determined to be sufficiently conservative to accommodate the introduction of axial blankets in the Cook Nuclear Plant Unit 2. Figure 5.1 shows the axial power distribution assumed in the Non-LOCA safety analyses.

Reconstitutable To Nozzle R and Debris Filter Bottom Nozzle DFBN Reconstitutable Top Nozzles (RTN) and Debris Filter Bottom Nozzles (DFBN) have been used extensively in Westinghouse designs. Analysis was performed to confirm the hydraulic compatibility of the Westinghouse nozzle designs to the existing ANF designs and therefore, will not impact any parameters important to the Non-LOCA safety analyses.

5.1.3 Modified Safety Analysis Assumptions Listed below are the analysis assumptions which represent a departure from that currently used for Cook Nuclear Plant Unit 2.

Revised Maximum Moderator Density Coefficient Increased Design Enthalpy Rise Hot Channel Factors (FNgH) and Fg for the Westinghouse VANTAGE 5 fuel Increase F gH Part Power Multiplier on Westinghouse VANTAGE 5 fuel Decreased Shutdown Margin Revised Thermal Design Procedure (RTDP)

Increased Core Power Reduced Temperature and Pressure (RTP) Operation 0 ppm boron concentration in the Boron Injection Tank (BIT)

Constant Steam Generator Level Program System Performance Degradation A brief description of each of these assumptions follows:

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Revised Maximum Moderator Densit Coefficient The analyses consider an End-of-Cycle (EOC) Life most positive Moderator Density Coefficient (MDC) of 0.54 b,k/gm/cc. The Moderator Temperature Coefficient (MTC) as a function of vessel average temperature is shown in Figure 5.2.

Increased FNgH and Fg The design F gH for the ANF and VANTAGE 5 fuel is 1.55 and 1.65 respectively. The Non-LOCA calculations applicable for the VANTAGE5 core have assumed a full power F gH of 1.65.

This is a conservative safety analysis assumption for this report.

The increase in the Technical Specification maximum LOCA Fg from 2.1 to 2.22 is conservatively bounded in the Non-LOCA transients. A maximum F< of 2.5 was assumed in the Non-LOCA safety analyses.

Increased FNgH Part Power Multi liers The FNgH part power multipliers are 0.2 for ANF fuel and 0.3 for VANTAGE 5 fuel. These values have been considered in the generation, of the core thermal limits for both fuel types. The changes in the core thermal safety limits result in a change to the Overtemperature and Overpower bT (OTbT/OPbT) reactor protection trip setpoints. Two sets of OTBT/OPbT setpoints were calculated. The first set of these setpoints is calculated based on ANF core thermal limits and is applicable for transition cycles. The second set of these setpoints is calculated based on VANTAGE 5 core thermal limits and is applicable for full VANTAGE 5 core (Cycles 10 and beyond). DNB analyses which are performed using LOFTRAN (see Appendix B, Reference 5) alone were analyzed twice, once for mixed core cycles and once for full VANTAGE 5 core. The remaining DNB analyses have accounted for the variation in FNgH part power multipliers between a mixed core and a full VANTAGE 5 core.

Decreased Shutdown Mar in SDM A change in the shutdown margin from 2.0% b,k/k to 1.3% b,k/k was considered in the Non-LOCA safety analyses.

Revised Thermal Desi n Procedure RTDP The calculational method utilized to meet the DNB design basis is the RTDP, which is discussed in Reference 12. Uncertainties'in the plant operating parameters are statistically treated such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR will be greater than the applicable limits as discussed in Section 4.2. Since the parameter uncertainties are considered in determining the design DNBR value, the plant safety analyses are perfornied using nominal initial conditions without uncertainties. The ANF fuel analyses used the W-3 correlation, while the VANTAGE 5 fuel analyses use the WRB-2 correlation.

Increased Core Thermal Power An increase in the nominal core thermal power from 3411 MWt to 3588 MWt was considered in the Non-LOCA safety analyses for the potential rerating of the Cook Nuclear Plant Unit 2. The Non-LOCA safety analyses performed at 3588 MWt will conservatively bound the current nominal core thermal power level of 3411 MWt.

Reduced Tem erature and Pressure RTP 0 eration Reduced temperature and pressure operation for Cook Nuclear Plant Unit 2 was considered in the Non-LOCA safety analyses. The full power vessel average temperature range of 547 F to 581.3 F at either of two values of pressurizer pressure (2100 psia or 2250 psia) was considered. However, because of the DNB constraints associated with the presence of ANF fuel during transition cycles (Cycles 8 and 9), a limitation on pressure and temperature conditions will apply. These include a full power vessel average temperature range of 547 ~F to 576 F, and a pressurizer pressure of 2250 psia (see Table B.2-1, cases 2 and 3 in Appendix B). Generating an acceptable nominal setpoint for the OTbT reactor trip setpoint during transition cycles has resulted in this limitation.

This limitation will not apply when a full VANTAGE 5 core is in place. The Non-LOCA safety analyses presented in this report provide support for a "full window" (see Appendix B Table B.2-1, cases 4-7) of operation in the assumed range of RTDP operation when a full VANTAGE5 core is in place at Cook Nuclear Plant Unit 2.

BIT Boron Concentration A zero (0) ppm BIT boron concentration was assumed in the Non-LOCA analyses to support BIT removal at Cook Nuclear Plant Unit 2. This is a conservative safety analysis assumption for this report.

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Steam Generator Water Level Pro ram A change in the steam generator water level program was considered in the Non-LOCA safety analyses. The existing steam generator water level program is a ramp function from 33% narrow range span (NRS) to 44% NRS from 0% power to 20% power and a constant level at 44% NRS between 20% power and 100% power. The steam generator water level program to be implemented at the beginning of Cycle 8 is a constant level at 44% NRS between 0% power and 100% power.

S tern Performance De radation The system performance degradation assumptions made for the Non-LOCA safety analyses are as follows:

A 10% average steam generator tube plugging level. This is a conservative safety analysis assumption for the Non-LOCA analyses presented in this report and bounds a 0% tube plugging level.

An increase in the Main Steamline Isolation Valve (MSIV) closure time from 5 seconds to 8 seconds with a corresponding increase in total response times.

A 10% Safety Injection Flow degradation.

A minimum required auxiliary feedwater flow rate of 450 gpm corresponding to the steam generator safety valve set pressure of 1123 psia was assumed for the Loss of Normal Feedwater analysis. For Loss of Offsite power to the Station Auxiliaries, a minimum auxiliary feedwater flow of 430 gpm corresponding to the steam generator safety valve set pressure of 1133 psia was assumed. A minimum auxiliary feedwater flow of 600 gpm corresponding to the steam generator safety valve set pressure of 1133 psia was assumed for the Feedline Break analysis.

5.1.4 Non-LOCA Safety Evaluation Methodology The Non-LOCA safety evaluation process is described in Reference 4. The methodology determines if a core configuration is bounded by existing safety analyses in order to confirm that applicable safety criteria are satisfied. The methodology systematically identifies parameter changes

on a cycle-by-cycle basis which may exceed existing safety analysis assumptions and identifies the transients which require reevaluation. This methodology is applicable to the evaluation of VANTAGE 5 transition and full cores.

Any required reevaluation identified by the reload methodology is one of two types. If the identified parameters is only slightly out of bounds, or the transient is relatively insensitive to that parameter, a simple evaluation may be made which conservatively evaluates the magnitude of the effect and explains why the actual analysis of the event does not have to be repeated.

Alternatively, should the deviation be large and/or expected to have a significantly or not easily quantifiable effect on the transients, reanalyses are required.

The reanalysis approach will utilize Westinghouse codes and methods which have been accepted by the NRC, and have been used in previous submittals to the NRC. These methods are those which have been presented to the NRC for a specific plant, reference SARs 'or reports for NRC approval. The analysis methods and codes are described in Appendix B.

With the exception of the Startup of an Inactive Loop, all the Non-LOCA accidents listed in Section 5.1:1 have been reanalyzed for this report. In accordance with the Technical Specification 3/4.4.1 (Amendment 59), Cook Nuclear Plant Unit 2 operation during Modes 1 and 2 with less than four loops is not permitted. Since three loop operation during Modes 1 and 2 is prohibited, the Startup of an Inactive Reactor Coolant Loop event was not considered for the transition to VANTAGE 5 fueL The key safety parameters are documented in Reference 4. Values of these safety parameters which bound both fuel types (ANF and VANTAGE 5) were assumed in the Non-LOCA safety analyses. For subsequent fuel reloads, the key safety parameters will be evaluated to determine if violations of these bounding values exist. Reevaluation of the affected accidents will take place as described in Reference 4.

5.1.5 Conclusions Descriptions of the Non-LOCA accidents reanalyzed for this report, method of analysis, results, and conclusions are contained in Appendix B. Based on the plant operating limitations given in the Technical Specifications and the proposed Technical Specifications changes given in Section 6.0 of 26

this report, the results show that the transition from ANF to 17x17 VANTAGE 5 fuel, including the aforementioned modified safety analysis assumptions described in Section 5.1.3, can be accommodated with margin to the applicable UFSAR safety limits.

The impact of the transition to VANTAGE5 fuel on Steam Line Break Mass and Energy Releases for both inside and outside containment is addressed in Section 5.4.

5.2 LOCA 5.2.1 Large Break LOCA 5.2.1.1 Description of Analysis/Assumptions for 17X17 VANTAGE 5 Fuel The large break Loss-Of-Coolant Accident (LOCA) analysis for Cook Nuclear Plant Unit 2, applicable to a full core of VANTAGE 5 fuel assemblies, was performed to develop Cook specific peaking factor limits. This is consistent with the methodology employed in the Reference Core Report for 17X17 VANTAGE 5, Reference 2. The Westinghouse 1981 Evaluation Model with BASH, References 20 and 21, was utilized and a spectrum of cold leg breaks were analyzed for Cook Nuclear Plant Unit 2 that bounds high and low pressure and high and low temperature operation. Other pertinent analysis assumptions include: a core thermal power of 3588 MWt, 15%

steam generator tubes plugged in each of four steam generators (i.e. uniform among the loops); an Fg of 2.22, an F gH of 1.62, and fuel data based on the new fuel thermal model, Reference 6.

The most limiting break determined from the high temperature/high pressure analysis was reanalyzed at the reduced temperature and reduced pressure conditions. In addition a case was analyzed to consider the closure on the RHR crosstie valves. This case was at 3413 MWt with the 95% part-power values of 2.335 and 1.644 for Fg and F gH respectively. The analysis assumptions, results, tables and figures are presented in Appendix C.

Section 2.0, Mechanical Design, demonstrates that the ANF 17x17 fuel assemblies currently in operation in Cook Nuclear Plant Unit 2 are very similar to the Westinghouse 17x17 VANTAGE 5 5 fuel assemblies in terms of geometric characteristics. Section 4.3 demonstrates that the 17x17 ANF fuel assembly is nearly identical to the Westinghouse 17x17 OFA assembly in terms of hydraulic characteristics. Therefore, the analyses reported in Reference 2 which demonstrate that the 17x17 VANTAGE 5 fuel features result in a fuel assembly that is more limiting than a Westinghouse '17x17 OFA fuel assembly, with respect to large break LOCA ECCS performance, remain valid as applied at Cook Nuclear Plant Unit 2. The same large break LOCA transition core 27

penalty reported in Section 5.2.3 of Reference 2 will be applied to the transition from 17x17 ANF fuel assemblies to Westinghouse 17x17 VANTAGE 5 fuel assemblies.

In addition, those ANF assemblies which remain in the core during transition to a full core of Westinghouse 17x17 VANTAGE5 fuel have lower Fg and FNgH limits (as specified in the Core Operating Limits Report). This provides additional assurance that the computed Peak Clad Temperature (PCT) for an entire core of Westinghouse 17x17 VANTAGE 5 fuel assemblies, including an appropriate transition core penalty, constitute a bounding analysis for the Cook Nuclear Plant Unit 2. As such, VANTAGE 5 fuel has been analyzed herein.

5.2.1.2 Method of Analysis The methods used to analyze the large break LOCA accident for Cook Nuclear Plant Unit 2 for VANTAGE 5 fuel, including computer codes used and assumptions are described in detail in Appendix C, Section C.3.1.2.

5.2.1.3 Results The results of this analysis, including tabular and plotted results of the break spectrum analyzed are provided in Appendix C, Section C.3.1.2, which has been prepared using the NRC Standard Format and Content Guide, Regulatory Guide 1.70, Revision 2 for accidents applicable to Cook Nuclear Plant, Unit 2.

Reference 20 states three restrictions related to the use of the 1981 Evaluation Model (EM) with BASH calculational model. The application of these restrictions to the plant specific large break LOCA analysis was addressed with the following conclusions:

Cook Nuclear Plant Unit 2 is neither an Upper Head Injection (UHI) nor Upper Plenum Injection (UPI) plant so restriction 1 does not apply.

The Cook Nuclear Plant Unit 2 plant specific LOCA analysis analyzed both minimum and maximum ECCS cases to address restriction 2. The CD = 0.6 Double Ended Cold Leg Guillotine (DECLG) break with minimum ECCS flows was found to result in the most limiting consequences.

Generic sensitivity studies were performed by Westinghouse for a typical 4-loop plant using different power shapes. This sensitivity study demonstrated that the chopped cosine was the most limiting power shape, Reference 21. A chopped cosine power shape was used in the large break LOCA analysis for Cook Nuclear Plant Unit 2, thus satisfying restriction 3.

5.2.1.4 Conclusions The large break LOCA analysis performed for the Cook Nuclear Plant Unit 2 has demonstrated that. for breaks up to a double-ended severance of the reactor coolant piping, the Emergency Core Cooling System (ECCS) will meet the acceptance criteria of Title 10 CFR Part 50 Section 46. That 1s:

1. The calculated peak cladding temperature will remain below the required 2200 F.

2. The amount of fuel cladding that reacts chemically with the water or steam to generate hydrogen does not exceed 1% of the hypothetical amount that would be generated if all the zirconium metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

3. The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.

4. The core remains amenable to cooling during and after the LOCA;

5. The core temperature is reduced and decay heat is removed for an extended period of time. This is required to remove the heat produced by the long-lived radioactivity remaining in the core.

The time sequence of events for all breaks analyzed is shown in Table C.3.1-5 of Appendix C, Section C.3.1.2.

The large break LOCA analysis for Cook Nuclear Plant Unit 2 assuming a full core of VANTAGE 5 fuel; utilizing the 1981 EM with BASH calculational model, resulted in a peak cladding temperature of 2140~F for the limiting CD 0.6 DECLG break at a total peaking factor 29

of 2.22. The maximum local metal-water reaction was 6.80% and the total core wide metal-water reaction was less than 0.3% for all cases analyzed. Further, the clad temperature transients reached a maximum at a time when the core geometry was still amenable to cooling.

The effect of the transition core cycles is conservatively evaluated to be at most 50 F higher in calculated peak cladding temperature which would yield a transition core PCT of 2190 ~F. The transition core penalty can be accommodated by the margin to the 10 CFR 50.46, 2200 ~F limit.

It can~.be determined from the results contained in Appendix C, Section C.3.1.2 that the large breaklLOCA ECCS analysis for the Cook Nuclear Plant Unit 2 remains in compliance with the requirements of 10CFR50.46 including consideration for transition core conGgurations.

5.2.2 Small Break LOCA 5.2.2.1 Description of Analysis and Assumptions for 17X17 VANTAGE 5 Consistent with the logic presented in Section 5.2.1.1 for large break LOCAs, the small break loss-of-coolant accident (LOCA) was analyzed assuming-a full core of VANTAGE 5 fuel to determine the peak cladding temperature. 's with the large break LOCA, the methodology employed in WCAP-10444-P-A, Reference 2, for transitioning from Westinghouse 17x17 OFA to 17x17 VANTAGE 5 fuel was applied to the transition from 17x17 ANF fuel assemblies to Westinghouse 17x17 VANTAGE 5 fuel assemblies. The currently approved NOTRUMP Small Break ECCS Evaluation Model, Reference 22, was utilized for a spectrum of cold leg breaks.

Appendix C, Section C.3.2, includes a full description of the, analysis and assumptions utilized for the Westinghouse VANTAGE5 ECCS Small Break LOCA analysis. Pertinent assumptions include an F gH of 1.62 for a full core of 17x17 VANTAGE5 fuel assemblies in the Cook Nuclear Plant Unit 2 core, a total peaking factor corresponding to 2.32 at the core mid-plane, 15% steam generator tube plugging, and a core thermal power level of 3588 MWt. The most limiting small break LOCA was computed for the low pressure/high temperature case and the limiting break size was reanalyzed for two additional cases to cover the range of operating temperatures and pressures being considered. An additional small break LOCA calculation was made which assumed that the HHSI cross tie valves were closed. To compensate for the reduction in safety injection due to closure of the cross tie valves, reactor power was reduced to 3413 MWt.

Sensitivity studies performed using the NOTRUMP small break evaluation model have demonstrated that VANTAGE 5 fuel is more limiting than OFA fuel in calculated ECCS 30

performance. It has been previously demonstrated that the 17x17 ANF fuel assemblies are essentially identical in both geometry and hydraulic characteristics to the Westinghouse 17x17 OFA fuel assembly. Therefore, the conclusion of Reference 2 that a small break LOCA analysis for a full core of Westinghouse 17x17 VANTAGE5 fuel is bounding, remains valid. On this basis, only VANTAGE 5 fuel was analyzed, since it is the most limiting of the two types of fuel (17x17 ANF and Westinghouse 17x17 VANTAGE-5) that would reside in the core for Cook Nuclear Plant Unit 2.

5.2.2.2 Method of Analysis The methods of analysis, including codes used and assumptions, are- described in detail in Appendix C, Section C.3.2.

5.2.2.3 Results The results of this analysis, including tabular and plotted results of the break spectrum analyzed, are provided in Appendix C, Section C.3.2.

5.2.2.4 Conclusions The small break VANTAGE 5 LOCA analysis for Cook Nuclear Plant Unit 2, utilizing the currently approved NOTRUMP Evaluation Model resulted in a peak cladding temperature (PCT) of 1357 F for the 4-inch diameter cold leg break at high temperature and low pressure. The 4-inch break size was then used for both a low temperature/high pressure and high temperature/high pressure analysis which resulted in PCTs of 1315 'F and 1325 F respectively.

The analysis assumed a limiting small break power shape consistent with a Fg(Z) envelope of 2.32 at the core midplane elevation and 2.15 at the top of the core. The maximum local metal-water reaction is 0.15%, and the total core metal-water reaction is less than 0.3 percent for all cases analyzed corresponding to less than 0.3 percent hydrogen generation. The clad temperature transients turn around at a time when the core geometry is still amenable to cooling.

Analyses presented in Appendix C, Section C.3.2 show that one high head charging pump and one safety injection pump, together with the accumulators, provide sufficient core flooding to keep the calculated peak clad temperature well below the required limit of 10 CFR 50.46 for the Cook Nuclear Plant Unit 2. It can also be seen that the ECCS analysis remains in compliance with.all other requirements of 10 CFR 50.46 and the peak cladding temperature results are below the peak 31

cladding temperatures calculated for the large break LOCA. Adequate protection is therefore afforded by the ECCS in the event of a small break LOCA.

5.2.3 Transition Core Effects on LOCA When assessing the effect of transition cores on the large break LOCA analysis, it must be determined whether the transition core can have a greater calculated peak cladding temperature (P~ than either a complete core of the 17x17 ANF assembly design or a complete core of the Westinghouse 17x17 VANTAGE5 design. For a given peaking factor, the only mechanism available to cause a transition core to have a greater calculated PCT than a full core of either fuel is the possibility of flow redistribution due to fuel assembly hydraulic resistance mismatch. Hydraulic resistance mismatch will exist only for a transition core and is the only unique difference between a complete core of either fuel type and the transition core.

In addition, all the various LOCA related analyses discussed below have beeri analyzed or evaluated to include a control rod drop time of 2.7 seconds, as is required for the 17x17 VANTAGE 5 fuel.

'5.2.3.1 Large Break LOCA The large break LOCA analysis was performed with a full core of VANTAGE 5 fuel and conservatively applies the blowdown transient results to transition cores. The VANTAGE 5 differs hydraulically from the 17x17 ANF assembly design it replaces. The difference in the total assembly hydraulic resistance between the two designs is approximately 10% higher for VANTAGE 5.

An evaluation of hydraulic mismatch of approximately 10% showed an insignificant effect on blowdown cooling duiing a LOCA. The SATAN-VIcomputer code models the crossflows between the average core flow cha'nnel (average of 192 fuel assemblies) and the hot assembly flow channel (one fuel assembly) during blowdown. To better understand the transition core large break LOCA blowdown transient phenomena, conservative blowdown fuel clad heatup calculations have been performed to determine the clad temperature effect on the new fuel design for mixed core configurations. The effect was determined by reducing the axial flow in the hot assembly at the appropriate elevations to simulate the effects of the transition core hydraulic resistance mismatch.

In addition,- the Westinghouse blowdown evaluation model was modified to account for grid heat transfer enhancement during blowdown for this evaluation. The results of this evaluation have shown that no peak cladding temperature penalty is observed during blowdown for the mixed core.

32

Therefore, it is not necessary to perform a blowdown calculation for the VANTAGE 5 transition core configuration because the evaluation model blowdown calculation performed for the full VANTAGE 5 core is conservative and bounding.

'ince the overall resistance of the two types of fuel is essentially the same, only the crossflows during core reflood due to Intermediate Flow Mixing grids need be evaluated. The LOCA analysis uses, the BASH computer code to calculate the reflood transient, Reference 20, which utilizes the BART code, Reference 23. A detailed description of the BASH code is given in Appendix C.

Fuel assembly design specific analyses have been performed with a version of the BART computer code, which accurately models mixed core configurations during reflood. Westinghouse transition core designs, including specific 17X17 OFA to VANTAGE 5 transition core cases, were analyzed.

For this case, BART modeled both fuel assembly types and predicted the reduction in axial flow rates at the appropriate elevations. As expected, the increase in hydraulic resistance for the VANTAGE 5 assembly was shown to produce a reduction in reflood steam flow rate for the VANTAGE5 fuel at mixing vane grid elevations for transition core configurations. This reduction in steam flow rate is partially offset by the fuel grid heat transfer enhancement predicted by the BART code during reflood. The various fuel assembly specific transition core analyses performed resulted in peak cladding temperature increases of up to 50 F for core axial elevations that bound the location of the PCT. Therefore, the maximum PCT penalty possible for VANTAGE 5 fuel residing in a transition core is 50 'F, Reference 2. As stated earlier, this transition core penalty continues to apply to the transition from 17x17 ANF fuel assemblies to Westinghouse 17x17 VANTAGE 5 fuel assemblies due to the near identical design of 17x17 ANF.and Westinghouse 17x17 OFA fuel assemblies. Once a full core of VANTAGE 5 fuel is achieved the. large break LOCA analysis will apply without the transition core penalty.

5.2.3.2 Small Break LOCA The NOTRUMP computer code, Reference 24, is used to model the core hydraulics during a small break LOCA event. Only one core flow channel is modeled in the NOTRUMP computer code, Reference 22, since the core flow rate during a small break LOCA is relatively slow, providing enough time to maintain flow equilibrium between fuel assemblies (i.e., no crossflow).

Therefore, hydraulic resistance mismatch is not a factor for small break LOCA. Thus, it is not necessary to perform a small break LOCA evaluation for transition cores, and it is sufficient to 33

reference the small break LOCA for the complete core 'of the VANTAGE 5 fuel design, as bounding for all transition cycles.

5.2.4 Post-LOCA Long-Term Core Cooling - ECCS flows, Core Subcriticality and Switchover of the ECCS to Hot Leg Recirculation The implementation of VANTAGE 5 fuel at the Cook Nuclear Plant Unit 2 does not affect the assumptions for decay heat, core reactivity or boron concentration for sources of water residing in the containment sump Post-LOCA. Thus, these licensing requirements associated with LOCA are not significantly affected by the implementation of VANTAGE 5 fuel.

Additionally, Westinghouse and-American Electric Power Service Corp. perform an independent check on core subcriticality for each fuel cycle operated at Cook Nuclear Plant Unit 2.

5.2.5 Short-Term Containment Analysis The containment building subcompartments are the fully or partially enclosed volumes within the containment which contain high energy lines. These subcompartments are designed to limit the adverse effects of a postulated high energy pipe rupture within them. The short term'containment integrity analysis is used to verify the adequacy of interior structures and walls by demonstrating that calculated differential pressures are less than design limits. The functioning of the ice condenser is demonstrated and containment integrity is also verified. The short-term containment integrity analysis is described in Section 14.3.4.3 of the Cook Nuclear Plant Unit 2 UFSAR.

The short-term containment analysis was recently performed to support operation of the Cook Nuclear Plant Unit 2 at an uprated NSSS power level of 3600 MWt, RCS average vessel temperatures over the range of 547 ~F to 581.3 ~F, and at RCS pressures of 2100 psia or 2250 psia.

This analysis is documented in Section 3.4.1 of WCAP-11902, Reference 25. Since the peak subcompartment pressures occur within a couple of seconds of transient initiation, the changes resulting from the VANTAGE 5 fuel reload do not affect the short-term containment analysis.

5.2.6 LOCA Containment Integrity The long term peak containment pressure calculation was recently performed to support operation of the Cook Nuclear Plant Unit 2 with the RHR crosstie valves closed at an NSSS power level of 3425 MWt. This analysis is documented in WCAP-11908, Reference 26. The analysis documented 34

in WCAP-11908 also provides justification for operation at 3425 MWt NSSS power, RCS vessel average temperatures of 547 'F to 578.7 F and RCS pressurizer's pressures of 2100,psia or 2250 psia. The analysis also considers and provides justification for operation with 10% average (15%

peak) steam generator tube plugging; 10% high head charging, safety injection, and residual heat removal pump degradation; initial accumulator volume of 946 + 25 cubic feet; 10%'ontainment spray flow rate degradation; and spray additive tank deletion. Other changes resulting from the VANTAGE 5 fuel reload do not affect the LOCA containment integrity analysis.

The effect that design changes to the reactor fuel assemblies can have on Containment Mass and Energy releases used to determine Containment Peak Pressure are dependent upon:

1) The change in core fluid volume as a result of the new fuel design.

2) Increase or decrease in core stored energy.

3) Effect of the new fuel design on reflood flooding rates as a result of core flow area or hydraulic resistance changes.

The VANTAGE 5 fuel design and the ANF 17x17 fuel design utilize a fuel rod smaller in diameter than the 15x15 OFA fuel which is modeled in the containment analysis documented in WCAP-11908. Therefore, the core stored energy is less than what is modeled in the WCAP-11908 analysis. The core volume is the same with 15x15 OFA fuel as with VANTAGE 5 and/or ANF fuel. The hydraulic resistance of the VANTAGE 5 fuel with the Intermediate Flow Mixing grids is larger than the hydraulic resistance of the 15x15 OFA fuel modeled in the analysis. The hydraulic resistance of the ANF 17x17 fuel is also larger than the hydraulic resistance of the 15x15 OFA fuel modeled in the analysis. The analysis, therefore, calculates conservatively high mass and energy releases to the containment. Thus, the containment analysis documented in WCAP-11908 bounds operation of Cook Nuclear Plant Unit 2 with a mixed ANF/VANTAGE 5 or full VANTAGE 5 core and the conclusions of WCAP-11908 remain valid.

5.2.7 Steam Generator Tube Rupture Analysis The analysis for a Steam Generator Tube Rupture accident (SGTR) presented in the Cook Nuclear Plant Unit 2 UFSAR was performed to ensure that the offsite radiation doses remain below the limits based on the 10CFR100 guidelines.

35

A subsequent evaluation was performed and is documented in WCAP-11902 (Reference 25),

Section 3.5, to determine the effect of increased power and revised temperature and pressure operation. This evaluation considered NSSS power levels up to 3600 MWt, a range of full power RCS vessel average temperatures between 547.0 ~F and 581.3 'F, and RCS pressurizer pressures of 2250 psia or 2100 psia.

The evaluation also considered 10% average (15% peak) steam generator tube plugging, 15%

auxiliary feedwater flow degradation, and 25 gpm charging flow imbalance. The other system

'performance degradation and fuel related changes considered in this report do not affect the SGTR accident analysis.

The primary thermal hydraulic parameters affecting the conclusion of the SGTR accident analysis are the extent of fuel failure, the primary to secondary break flow rate through the ruptured tube,

, and the mass released to the atmosphere from the steam generator with the ruptured tube. The UFSAR SGTR accident analysis and the WCAP-11902 evaluation are based on an assumption of 1% defective fuel, and an initial primary coolant activity corresponding to this amount of defective fuel. These assumptions will not be affected by the change to VANTAGE 5 fuel. The primary to secondary break flow'rate and the mass release to the atmosphere are dependent upon the initial

. reactor and steam generator conditions of power. Since the range of operating conditions at Cook Nuclear Plant Unit 2 has been considered in WCAP-11902 and will not change due to the implementation of VANTAGE 5 fuel, it is concluded that the primary to secondary break flow rate and atmospheric steam release will not change due to the implementation of VANTAGE 5 fuel.

Therefore, the consequences of a SGTR accident will not be increased by the implementation of VANTAGE 5 fuel and the SGTR accident evaluation in WCAP-11902 remains bounding.

5.3 LOCA Hydraulic Forces Analysis 5.3.1 Introduction The purpose of the LOCA hydraulic forces analysis was to provide LOCA hydraulic forcing functions which were used in conjunction with the seismic analysis to verify the structural integrity of the core components for the proposed 17x17 VANTAGE 5 fuel reload, including the rerating program and peak steam generator tube plugging to 15% for Cook Nuclear Plant Unit 2 at the limiting primary fluid temperatures and pressures. The LOCA hydraulic forcing functions were 36

generated for the accumulator injection line break in the cold leg. The LOCA hydraulic forces analysis takes advantage of the elimination of large primary pipe ruptures (Reference 27) to reduce some of the expected increase in the magnitude of the peak forces which may occur due to the rerating program.

5.3.2 Method of Analysis The method of analysis, to determine the LOCA hydraulic forcing functions, considers the accumulator injection line break at the reduced RCS primary temperatures, a core power of 3588 MWt, a peak steam generator tube plugging level of 15%, and a nominal RCS pressurizer pressure of 2250 psia. The computer codes that are used to evaluate the postulated LOCA are MULTIFLEX1.0, LATFORC, and FORCE2. MULTIFLEX(Reference 28) is used to calculate the thermal hydraulics of the rector coolant system due to a postulated LOCA. LATFORC uses the pressure distribution in the downcomer annulus region calculated by MULTIFLEXto determine the lateral hydraulic forcing functions on the reactor vessel, core barrel and the thermal shield.

FORCE2 uses the pressure transient in the reactor vessel calculated by MULTIFLEXto calculate the vertical forces on the vessel internals and core components.

5.3.3 Results Results of the LOCA hydraulic forces analyses have shown that eliminating large pipe ruptures and analyzing reactor coolant branch line breaks partially offset the expected increases in the LOCA hydraulic forcing functions due to the reduced reactor coolant temperatures as proposed for the rerating program. Evaluations have shown that the LOCA hydraulic forcing functions from a double-ended guillotine break or a limited displacement break in the reactor coolant piping used in the structural integrity analyses (Reference 25) at current thermal conditions are still more limiting than the branch line LOCA hydraulic forcing functions at the reduced temperature conditions. Specifically, Reference 25 concluded that the peak horizontal forces from a 100 square inch reactor vessel inlet nozzle break remain limiting when compared to an accumulator injection line break. On this basis it was also concluded that the LOCA hydraulic forcing functions which were used as the bases for the original qualification of the reactor vessel, internals and loops remain bounding.

I However, to specifically evaluate the structural integrity 37 of the 17x17 VANTAGE 5 fuel, LOCA hydraulic forcing functions have been generated for the accumulator injection line break for the h

rerating program to be used as input to determine the structural integrity of the core components.

The evaluation of structural integrity for the core components can be found in Section 2.7 of this report which addresses seismic and LOCA considerations. This section provdies the evaluation and conclusions on the structural integrity of the 17x17 VANTAGE 5 fuel as a result of the accumulator injection line break LOCA hydraulic forcing functions calculated for the rerating program at reduced temperature conditions.

5.4 Steamline Break Mass and Energy Releases This section addresses the impact of the complete transition of Cook Nuclear Plant Unit 2 from ANF 17x17 fuel to Westinghouse 17x17 VANTAGE 5 fuel on the Steamline Break Mass and Energy releases for both inside and outside containment.

5.4.1 Steamline Break Mass and Energy Releases Inside Containment The Steamline Break Mass and Energy releases inside containment have been calculated to bound both Cook Nuclear Plant Unit 1 with 15x15 fuel and Cook Nuclear Plant Unit 2 VANTAGE 5 fuel. WCAP-11902, Supplement 1, Section S-3.3.4.1 documented this analysis which supports the Cook Nuclear Plant Unit 2 transition to VANTAGE 5 fuel, and includes the modified safety analysis assumptions as discussed in Section 5.1.3. The RCCA insertion time to dashpot assumed in the analysis was 2A seconds, which does not bound the 2.7 second time conservatively assumed for the VANTAGE 5 fuel. Also, the analysis did not consider a 10% Safety Injection Flow degradation. An evaluation has been performed which concludes that these differences would have an insiginificant effect on the calculated Mass and Energy releases. Thus, the analysis supports the transition to VANTAGE 5 fuel.

5.4.2 Steamline Break Mass and Energy Releases Outside Containment The current Mass and Energy'releases applicable for use in outside containment equipment qualification evaluation for Cook Nuclear Plant Unit 2 are documented in Reference 29 (Category 1). These releases included the effect of superheated steam for use in evaluation of the outside containment equipment qualification issues.

The Mass and Energy releases of Reference 30 have been evaluated for their applicability to the Cook Nuclear Plant Unit 2 VANTAGE5 transition. This evaluation concludes that the Mass and Energy releases documented in Reference 30 will remain bounding for the transition of Cook 38

e Nuclear Plant Unit 2 to VANTAGE 5 fuel, provided the following Technical Specifications.and modified safety analysis assumptions/limitations are maintained. The outside containment Mass and Energy releases are insensitive to a 25 gpm charging flow imbalance.

Maximum allowable NSSS power no greater than 3425 MWt.

End-of-Cycle (EOC) Life most positive Moderator Density Coefficient (MDC) not more positive than 0.43 b,k/gm/cc. The Moderator Temperature Coefficient (MTC) as a function of vessel average temperature is shown in Figure 5.2 Minimum shutdown margin of 1.6% b,k/k.

Maximum allowable steamline isolation valve closure time no greater than.5.0 seconds (see NOTE below).

The compensated nominal setpoint for low steamline pressure no less than 520 psig.

This setpoint corresponds to the analysis setpoint of 379 psig.

NOTE: A safety evaluation independent of the Cook Nuclear Plant Unit 2 VANTAGE 5 program has been performed to support an increase of 3.0 seconds in the steamline isolation valve closure time and related steamline isolation Engineered Safety Feature (ESF) response time (Reference 30).

The new superheated Mass and Energy releases to bound both Cook Nuclear Plant Unit 1 with 15x15 fuel and Cook Nuclear Plant Unit 2 with VANTAGE 5 fuel, including the modified safety analysis assumptions as discussed in Section 5.1.3 have been calculated by Westinghouse and were provided to AEPSC. The evaluation for determining the acceptability of these new superheated Mass and Energy releases for outside containment equipment qualification has not been completed for this report. The above Technical Specifications and modified safety analysis assumptions could be removed and the modified safety analysis assumptions as discussed in Section 5.1.3 could be supported at a later time, provided the new Mass and Energy releases are determined by AEPSC to be acceptable for outside containment equipment qualification.

39

5.4.3 Conclusions The Cook Nuclear Plant Unit 2 transition to VANTAGE 5 fuel, including the modified safety analysis assumptions (Section 5.1.3) can be supported for the Mass and Energy releases inside containment.

The current Mass and Energy releases outside containment as documented in Reference 5.29 will remain,abounding for this report, provided the Technical Specifications limitations and modified safety. analysis assumptions as noted in Section 5.4.2 are maintained.

40

1.20 1.0 0.8

)

I

<<C 0.6 0.4 0.0 0.2 0.4 0.6 0.8 1.0 FRACTION OF CORE HEIGHT COOK NUCLEAR PLANT UNTl' HGURE S.l Design Axial Power Distribution for non-OTnT Transients (WCAP-9500)

-32 0 547.8 -55.8 CL I -34 CE LE LrJ CD

-35 Lt LtJ CD

-36 CL I

CCE PT LE CL -37 LIJ CL LtJ

-38 CL CD CL LtJ -39 CD CD 678.7 -5 I.8 45 550 555 560 565 570 575 580 YESSEL AYERAGE MODERATOR TEMPERATURE ~oF COOK NUCLEAR PLANT UNTI' FIGURE 52 Most Negative Moderator Tempenture Coefiicieat Limit 42

6.0

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES Table 6.1 presents a list of the Technical Specifications changes. The changes noted in Table 6.1 are given in the proposed Technical Specifications page changes in Appendix A.

43

TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES SECTION PAGE CHANGE REASON FOR CHANGE 1.0, Add COLR COLR implementation pgI to index 1.12a, Add COLR COLR implementation pg 1-3 Figure 2.1-1, Revised safety. limits ~ Reanalysis supports VANTAGE 5 reload pg 2-2 221 Design flow change Change in design flow due to VANTAGE 5 pg 2-5 & trip setpoint fuel reload, RTDP implementation Table 2.2-1, Revise Overtemperature Reanalysis supports VANTAGE 5 reload pg 2-7 & 2-8 hT limits Table 2.2-1, Revise Overpower Reanalysis supports VANTAGE 5 reload pg 2-9. b,T limits 2.1.1 Bases, Update to bases VANTAGE 5 fuel reload and COLR pgB2-1 & implementation (relocation of F gH)

B 2-2 2.1.1 Bases, Update to bases VANTAGE 5 fuel reload and delete pg B 2-4 Cycle 6 specific information 2.1.1 Bases, Revise bases Reanalysis supports VANTAGE 5 reload pg B 2-5 2.1.1 Bases, Revise bases circuit Reanalysis supports VANTAGE 5 reload pg B 2-7 breaker time 3/4.1.1.1, Decrease shutdown Reanalysis with reduced SDM pg 3/4 1-1 &, margin 1-2 3/4.1.1.2, Decrease shutdown Reanalysis with reduced SDM.

pg 3/4 1-3 & margin Change to Westinghouse dilution 1-3b accident methodology

TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)

SECTfON PAGE CHANGE REASON FOR CHANGE 3/4.1.1.4, MTC relocated to COLR VANTAGE 5 fuel reload and pg 3/4 1-5 & & revised EOL limit COLR implementation (relocation 3/4 1-6 of MTC) 3/4.1";1.5, Minimum temperature Reanalysis with reduced temp pg 3/4 1-7 for surveillance req.

3/4.1.2.3, Change ch. pump Make consistent with pg 3/4 1-11 discharge head the analysis 3/4.1.2.4, Change ch. pump Make consistent with pg 3/4 1-12 discharge head the analysis 3/4.1.2.7, Change 80 F to Make spec consistent with the

'pg 3/4 1-15 70 GF 'nalysis limit 3/4.1.2.8, Change volume from Make spec consistent, with the pg 3/4 1-16 5650 to 7715 gallons VANTAGE 5 reload analysis limit to

& change 80 F to 70 'F accommodate reduced rod worth and management flexibility 3/4.1.3.1, Delete reference COLR implementation pg 3/4 1-19 to F>g. 3.1-1 3/4.1.3.4, Change rod drop time Make spec consistent with the pg 3/4 1-23 from 2.2 to 2.7 sec analysis limit & COLR implementation Relocate steps withdrawn to COLR 3/4.1.3.5, Relocate shutdown rod COLR implementation (relocation pg 3/4 1-24 insertion limits to of shutdown rod insertion limits)

COLR 45

TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)

SECTION PAGE CHANGE REASON FOR CHANGE 3/4.1.3.6, Relocate control rod COLR implementation (relocation pg 3/4 1-25 insertion limits to of control rod insertion limits)

COLR 3/4.1.3.6, Delete figure 3.1-1 COLR implementation pg 3/4 1-26 3/4.3.2.1, Relocate axial flux COLR implementation (relocation pg 3/4 2-1 & difference limits to of AFD limits) 2-3 COLR 3/4.3.2.1,. Relocate axial flux COLR implementation (relocation pg 3/4 2-4 difference allowable of AFD allowable deviation) deviation Fig. to COLR 3/4.3.2.2, pg 3/4 2-5 to COLR

'OLR Relocate Fg limits implementation (relocation of Fg limit) 3/4.3.2.2, Relocate K(Z) & V(Z) COLR implementation (relocation pg 3/4 2-8, flgures to COLR of Fg limit) 2-8a & 2-8b 3/4.3.2.3, Relocate FNgH COLR implementation (relocation pg 3/4 2-9 limits to COLR of F gH limit) 3/4.2.5.1, Reformat DNB spec Adopt planned Cook Nuclear Plant pg 3/4 2-15 Change DNB parameter Unit 1 spec format consistent with values and add low VANTAGE 5 reload Tavg window 3/4.2.5.1, Delete tables 3.2-1 Adopt planned Cook Nuclear Plant Unit 1 pg 3/4 2-16 &, and 3.2-2 spec format 2-17 & 2-18 Delete 3.2.5.2 Not required

TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)

SECTION PAGE CHANGE REASON FOR CHANGE 3/4.3.2.6, Relocate Fg limits COLR implementation (relocation pg 3/4 2-19 to COLR of Fg limit)

Changed definition of Fg Westnghouse CAOC methodology Table 3.3-2, Changed and added RPS Make consistent with the analysis pg 3/4 3-9 & response times limits 3-10 Table 3.4-4, Change ESFAS setpoint Make consistent with analysis pg 3/4 3-25 Table 3.3-5, Changed ESF response time Make consistent with the analysis pg 3/4 3-26 & times limits

-3/4 3-27 & 3/4 3-28 3/4.4.1.2, Reduce number of RCPs Make consistent with the analysis pg 3/4 4-2 & required operable in limits 4-3 mode 3 3/4.4.4, Change water volume Make consistent with the analysis pg 3.4 4-6 from 62% to 92% limit 3/4.4.6.2, Controlled leakage Consistent with analysis pg 3/4 4-15 & in terms of resistance 3/4 4-16 3/4.5.1b, Revise minimum Make consistent with analysis pg 3/4 5-1 contained borated limits water volume & min/max cover-pressure 3/4.5.2.f, Revised SI pump Reanalysis with degraded SI pg 3/4 5-5 performance performance 47

TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)

SECIYON PAGE CHANGE REASON FOR CHANGE 3/4.5.2.h, Revised SI pump flow Adopt limits similar to Cook Nuclear Plant pg 3.4 5-6 balance limits Unit 1 3/4.5.5, " Reduce RWST min temp Make spec consistent with pg 3/4 5-11 to 70 'F analysis limit 3/4.1.1.1, Decrease shutdown Reanalysis with reduced pg B 3/4 1-1 margin shutdown margin B 3/4.1, Revise concentrations Make spec consistent with analysis pg B 3/4 1-3 and volumes limits B 3/4.2.1, Revise to reflect COLR implementation pg B'3/4 2-1 & COLR implementation (relocation of AFD limits) 2-2 & 2-3 Changed to WCAP-8385 Westinghouse methodology B 3/4.2.2 & 3, Revised to reflect VANTAGE 5 reload T-H analysis and pg B 3/4 2-4 COLR implementation COLR implementation (relocation of thru 2-4b & VANTAGE 5 reload Fg and F gH limits)

B 3/4.2.5, Revise to reflect Reanalysis with reduced temp pg B 3/4 2-5 reduced temp DNB limit B 3/4.2.6, Revise to reflect Make spec consistent with pg B 3/4 2-5 CAOC control analysis B 3/4.5.5, Reduce RWST temp to Make spec consistent with the pg B 3/4 5-3 70 GF analysis limit B 3/4.7.1, Reformat valve lift Make consistent with the pg B 3/4 7-1 criteria analysis limit 48

TABLE 6.1

SUMMARY

OF TECHNICAL SPECIFICATIONS CHANGES (continued)

SECTION PAGE CHANGE REASON FOR CHANGE 3.4.9.1, Delete reference to Reanalysis of refueling pg B 3/4 9-1 refueling reactivity reactivity at 2400 ppm boron calcs at 2000 ppm 6.9.1.11, Add COLR to section 6 COLR implementation pg 6-18 49

7.0 REFERENCES

1 Davidson, S. L, Iorii, J. A., "Reference Core Report - 17x17 Optimized Fuel Assembly,"

WCAP-9500-A, May 1982.

2 Davidson, S. L. and Kramer, W. R.; (Ed.) "Reference Core Report VANTAGE 5 Fuel Assembly," WCAP-10444-P-A, September 1985.

3 Davidson, S. L. (Ed.) et al., "Extended Burnup Evaluation of Westinghouse Fuel," WCAP-10125-P-A, December 1985.

4 Davidson, S. L (Ed.), et al., "Westinghouse Reload Safety Evaluation Methodology,"

WCAP-9272-P-A, July 1985.

5 Miller, J. V., "Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," WCAP-8720 (Proprietary), October 1976.

6 Weiner, R. A., et al., "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP-10851-P-A, August 1988.

7 Skaritka, J., (Ed.), "Fuel Rod Bow Evaluation," WCAP-8691, Revision1 (Proprietary), July 1979.

8 Davidson, S. L., Iorii, J. A. (Eds.), "Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly," WCAP-9401-P-A, August 1981.

9 Miller, R. W., et al., "Relaxation of Constant Axial Offset Control-Fg Surveillance Technical Specification," WCAP-10217-A, June 1983.

4 10 Davidson, S. L. (Ed.), et al., "ANC: Westinghouse Advanced Nodal Computer Code,"

WCAP-10965-P-A, September 1986.

50

11 Nguyen, T. Q., et al., "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," WCAP-11596-P-A, June 1988.

12 Friedland, A. J., and Ray, S., "Revised Thermal Design Procedure," WCAP-11397-P-A, April 1989.

13 Tong, L. S., "Critical Heat Fluxes in Rod Bundles, Two Phase Flow and Heat Transfer in Rod Bundles," Annual Winter Meeting ASME, November 1968, p. 3146.

14 Tong, L S., "Boiling Crisis and Critical Heat Flux," AEC Office of Information Services, TID-25887, 1972.

15 Schueren, P., McAtee, K. R., "Extension f

of Methodology for Calculating Transition Core DNBR Penalties," WCAP-11837, May 1988.

16 Final Safety Analysis Report - Chapter 3 - Donald C. Cook Unit 2, Docket Number 50-316, July 1982.

17 Letter from E. P. Rahe (W) to Miller (NRC) dated March 19, 1982, NS-EPR-2573, WCAP-9500 and WCAPS 9401/9402 NRC SER Mixed Core Compatibility Items.

18 Letter from C. O. Thomas (NRC) to Rahe (W) - Supplemental Acceptance No. 2 for Referencing Topical Report WCAP-9500, January 1983.

19 Final Safety Analysis Report- Chapter 14 - D. C. Cook Unit 2, Docket Number 50-316, Amendment 75, April 1977.

20 Kabadi, J. N., et al., "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A Revision 2 with Addenda (Proprietary), March 1987.

51

21 Besspiata, J. J., et al., 'The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code, Power Shape Sensitivity Studies," WCAP-10266-P-A Revision 2 Addendum 1 (Proprietary), December 15, 1987.

22 Lee, N., et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary) and WCAP-10081-A(Non-Proprietary), August 1985.

23 Young, M. Y., et al., "BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients," WCAP-9561-P-A (Proprietary), March 1984.

24 Meyer, P. E., "NOTRUMP - A Nodal Transient Small Break And General Network Code,"

WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Non-Proprietary), August 1985.

25 "Reduced Temperature and Pressure Operation for Donald C. Cook Nuclear Plant Unit 1

'icensing Report," WCAP-11902, October 1988.

26 "Containment Integrity Analysis for Donald C. Cook Nuclear Plant Units 1 and 2," WCAP-11908, July 1988.

27 Eisenhut, D. C. (NRC) to Operating PWR Licensees, "Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops (Generic Letter 84-04)," February 1, 1984.

28 K. Takeuchi, et al., "MULTIFLEX1.0 - A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic Structure Systems Dynamics," WCAP-8708-P-A (Proprietary),

WCAP-8709-A (Non-Proprietary), Westinghouse Electric Corporation, September 1977.

29 Butler, J. C., and Love, D. S., "Steamline Break Mass/Energy Releases for Equipment Qualification Outside Containment," WCAP-10961, Rev. 1 (Proprietary) and WCAP-11184 (Non-Proprietary), October 1985.

30 Letter from J. C. Hoebel (W) to R. B. Bennett (AEPSC), "Evaluation of Increased Steam Generator Stop Valve Closure Time," AEP-90-123, January 19, 1990.

52

APPENDIX A TECHNICAL SPECIFICATIONS CHANGE PAGES FOR THE DONALD C. COOK NUCLEAR PLANT UNIT 2 TRANSITION TO 17x17 VANTAGE 5 FUEL

NOEX DEFINITIONS SECTION PAGE 1.0 IDN De fined Te%o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~~~ ~~~~~~~~~~ ~~~~ ~~~~~ ~~ ~~~

Thermal Powel ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

Ra ted Thermal Power ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ l-l 0 perational Nodeo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ 'o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-1 Action ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~

Operable - Operability......................................

R eportable Event..........................................;. 2 Containment Integrity........................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-2 Channel Calibration............................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-2 Channel Check............................................... ,1-2 Channel Functional Test..................................... 1-3 n...

Core Al Shut anted V

yg j @PC e~ri~~, g +OCT

~ ~ ~ ~

1-3 1-3 Identified Leakage..................-......... " ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-3 Unidentified Leakage.......'......."""""".. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-3 Pressure Boundary Leakage....................... 1-4 Control 1 ed Leakage......... ". " ". "" ". ""

. - ~ ~ ~ ~ 1-4 guadrant Power Tilt Ratio....................... ~ ~

jh

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-4 Dose Equivalent I-131........................,.............. 1-4 Staggered Test Basis............................ 1-5 F requency Notation.............................. ~ ~ ~ ~ 1-5 Reactor Trip System Response Tile.......,....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-5 Engineered Safety Feature Response Tiae......... ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-5 Axial Flux Difference........................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ '1 P hysics T'ests ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6 7-Average Disintegration Energy................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6

~

So Checko ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

'rce 1-6 Process Control Program (PCP)................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6 D., C. COOK UNIT 2 Amendment o AC 0 ~

F1N1T 1 ON 5

~ @To CP~~iidi'JCT10NAL TEST 1.11 A CHAN'lB. FUNCT10t4L TKIT shall be:

a. lnalog channels - the 5nJection of a s(aulated signal into the channel as close to the pr<aary sensor as practicable to verify OrKL'41L?Tf including alarn and/or trip functions.

6. Iistable channels - the inJection of a sieulated signal into tÃe channel sensor to verify tPKRASlLPf including alarm and/or trip functions.

CORE AL, ERAT'CN 1,12 CORE ALTERAT1ON shall be the II:verent or aanipulat<on of any com-pcnen. wM<n the reactor pressure vessel ~3th the vessel head'~oved and fuel in the vessel. Suspens)on of CORK ALTERATION shall mt preclude cc:plat<en cf move.".~nt of a component to a safe conservative pos)t)on, uS J 4 ff ggi 5"

~ oyeIOI gt 1 Qo 1 g

~4 I e la 8 ~ QA RIJ J il

~

4 ~

1,13 S~JTXiN Y~RG1N shall be the {nstantaneous imcun of reactivity by wh'(eh'he reac or is subcd ical or ~ould be subcrkt)cal fran $ a prese:t condition asswfng all full lergth rod cluster assemblies (shutdern and central J are fully inserted except for the single rod cluster asse=hly of highest reactivity ~or 5 which is assigned to be iully A 'utra~n.

lQ Ni ..-.ED L A'i"5i 1.14:"DPlFiD LD,XASK shall be:

Leakage,(OAcept CCtPiROLLKQ LE4XASK) into closed syitens, such as ~p seal or valve packing leaks eat are captured and conducted to a sup or collect)ng task, or

b. Leakage into the conta$ reent acsphew Wm sources that are both speckf)cally located and knern e( 'e. mt u <nter'.ere

~1th the ooerat<on of leakage de~tion systams or not to be pRKSRlRK IglJNPRv L~e'X'GK, Or c~ Reactor coolant sy's~ leaklge th~wgn a ste& gene.atcr to

~"oe seccncary systole ~

+ o 01ay + oego ~ s oeeo% cot ~

~ ~ ~

~ enc ant No

IwsrRT 8 CORE OPERATING The CORE OPERATING LIMITS REPORT (COLR) 1s the LINITS REPORT un1t-spec1f1c docuaent that prov1des core operat1ng 11m1ts for the current operat1ng reload cycle. These cycle-spec1f1c core operat1ng 11a1ts shall bc detena1ned for each reload cycle 1n accordance ~1th Spec1 f1cat1on lln1t operat1on e1th1n these operat1ng 11a1ts addressed $ n )nd1v1dual spec1f1cat1ons.

CW.LI(

VNACCEPTABLE OPERATION SSO PS'/go pSly

"'o PS s

loco PSlg'uge

~ SIO 'sst 8yo sly I

~ 580 AC PTASL OPE TI 570 gV I; sso 0.2 O.i Q. '.8 '.0 1.2 FFI TIUN GF RAT THERMAL PQNER Y

PRESSURE (PSI SREAKPOINTS (FRACTION RATED TNERNL ER, T AVG, DEG F aaaaaaaaaeae ee 1840 0.00,616.2) , 0.98,585.1), ( 1.20, 6.5) 1940 0.00,623.8 0,93,594.7 1.20,563.5) 2040 0.00,631.0) 0.88>603.8) 1.20, 569.6) 2250 0.00,645.9) 0.80,622.3) 1.20,580.9) 229 0.00,647.9 0.80,624.5) 1.20,586.5.)

24 (0.00,657. 0.77,635.6 1.20,597.

2.1-1 Reactor Core Safety Limits-Four Loops in Operation C. COO@, U;<n' F2 ~~HDTl'o 82~

pFSiOW 'Pic 4 Og ~f>b Q Pressure Power Lmial VraQ ~

Tavg Power Lfml ~

Tavg Power Lfrzl ~

Tavg Power Lf~ ~

Tavg 1775 0.00 615.4 0.98 583.8 1.02 580.9 1.2, 558.1 2000 0.00 631.8 0.86, . 605.8 0.96 597.5 1.2 568.5 2100 0.00 639.1 0.82 614.0 0.96 601.6 1.2 573.1 2250 0.00 649.2 0.72 628.6 0.98 605.2 1.2 580.4 2400 0.00 659.0 0.62 642.0 1.1 599.0 1.2 588.1 2400 PSIA 2250 PSIA 2100 PSIA em 2000 PSIA Ota gem 1775 PSIA

~ NO 5

NM 660 FRACTION OF THERNL POWER pes<P Figure 2.1-1 Reactor Core Safety Limits Four Loops in Operation

n 1 ~ Nanual Raaotor Trip Not Applicable Not hypltcable

2. Pcwer Range, Neutron Flux Kev Setpoint - < 25% of RATED Iae Setpoint - 5, 26% of THERMAL PCNER RATED TIIEih P0IIER

~

High Setpoint >> c 109% of RATED Nigh Setpoint c 110% of TIIERNLL PCNER RATED THERMAL POIIER

3. Sewer Range, Neutron Flux, < St of RATED THERMAL PINER with g 5.5\ of RATED THE+ah High Positive Rate a tlae constant > 2 seconds PINER eith a tiae constant seconds

i. lligh Peear 1ancN, Naut'lux, Negative Rate

< 5\ of RATED THEN6lL PCNER with a time constant > 2 seconds c 5.5% of RATED THEWS PtNER Itith a tine constatlt g 2 seconds

5. Interaediate Range, Neutron c 25% of RlLTED THERMAL PINER g 30\ of RlLTED THERMAL Flux ONER

5. Bourne Range, Neutron Flux < 10 counts per second c 1.3 a 10 counts per second

't. OverteIIperatme LT See Note 1 See Note

~e Ove+4IHr See Note 2 See Note i

9. pressudaer Pressure -he p 1950 ps' p 1%F0 ysig

10. Pressuriaer Pressme High 2385 psig g 2395 ps'

11. Pressuriier Xater Encl High'2.

< 92% of lnstruaent span ~

93'f

%0 instmsent span Kess of Floe > 90% of design flew per loop+ ~

p 89.1 of design flov per loopo

~Design floe is >~&0 pi per loop.

g/, 4.oo

Co ue U E 0 n

8 1+ wl HoTE 1! Overtexperature AT < AT (Kl K2 1+ (T-T )+K~(P P ) fl(hl))

9 I%ere! AT Indicated AT at RhTED TllERMhL POMER hverage texperature, P

~pc o'

~ - Indicated Tavg at RhTED THERHhL POMER c Pressurizer Pressure,. psig PI 2235 psig (indicated RCS noxinal operating pressure) i+; e function gen~r~t~d b~ the lead-lag controller for T d~nax 11+~

coxpensation Tl ~ $

2 Tixe onstants utilised in the lead>>lag controller for Tay 35 sees/ g2 4 secs.

Laplace transfora operator

co t ed Co t ue 0 e to Kl ~$ 80 I, of K> O. Ol gZ~

K> -e-.axe-:m o. oooXZ and f (N) is a function of the indicated difference betMeen top and bottoa detectors of ion chaibersg arith gains to be selected based on measured the plier-range nuclear instruaent response during plant startup tests such thatt I'e

~sr (i)'or q - q between ~percent and percent, f (AI) > 0 (vherII q End q are percent RATED TIIERNAI, pnnER ln tha tap and batten l

halves of the etre respectively, and qt + qb is total THEIL POWER in percent of MTED THERMhL POWER).

~73 (ii) the gT trip setpoint shall be autoaaticalky for each percent that the magnitude of (q - q ) exceeds - percent, reduced b percent oi its value at MTED THERMhL POWER. 3.

for each percent that the aagnitude of (q - q ) exceeds S'iii) percent, the hT trip setpoint shall be autoaaticalky reduced by percent oi its value at RhTED THERMhL POWER. z.O

ot ue)

S IJ E T IOMTRI 8 T

Note 2c QverpoMBr .AT < AT (K -K5 - K6 (T-T )-f2(AI))

1+i~

8'FC.O where< AT0 Indicated AT at rated pover 0F T hverage temperature, Tll Indicated T at RATED TIIERMAL POMER op, K4 ksOV8 l 0 K m. 0.02/ p ior increasing average temperature and 0 for decreasing average temperature 0.00197 for T >A T" I K6 0 for T < T" The function generated by the rate lag controller for Tavg 1+v~ q dynaaic compensation T3 Tiae constant utilized in the rate lag controller for Tavg T3 10 secs.

Laplace transfora operator c7.0 f2(AI)

Note 3s The channel~s aaxiwua trip point shall not exceed its computed trip point by sore than percent AT span.

/.9 d'or<<han i! trip point shall its trip point by Note

~

The channel~s aaxiaua 80' percent AT span.

not exceed computed

,, The restrictions of this safety 1iRit prevent over sating of the aad sible cladding perforation vhich vould result the release of fisston roducss to the reactor coolant. Ovarheatt of the fuel claddinj is prevent>> y restricting fuel operattoa to vtshtn ~ nucleate boiling reggae vhere the et transfer coefficient ts large the cladding surface aaaperacure slightly above the coolant sa ation temperature.

Operation ove the upper boundary of ~ nucleate boiling regtae codd

-esult tn excessi cladding temperature ecause of the onset of depa tuse froa nucleate boili (DNb) and the res tant sharp reduccion tn heat transfer coeffietent. DhS is not ~ 4 ectly measurable paraaecer during operation and therefor TPSR~AL, N and Reactor Coolant Tearperacure and tressure have been relet to DQ. This relation has been developed to predics the DQ flux and ch loe ion of DNb for axtally uniform and non-unifora heat flux distrib ons. The local DQ heat flux ratio, DQR, defined as che ratio of the f$ux chat vould cause ac a pac'cieular core loeacton to the local eac, a DNb ts indicative of- che aargta so DQ.

Ihe DNb design b s is as foll s: there aust be as lease a 95 percent probability that the atoum DNbR of ~ limctting rod during Condition I and II events ts great than or equal to DQR liats of the DNb correlation beiag used (the b correlacion in thts a lteacion). The correlation DNbR limit is estab shed based on the entire ap ieable experimental data sec such that the is a 95 percent probability ch 95 percent confidence chat DQ vill ao oecu" vhea the aintaum DQR ts at ~ DQR ltais.

The urves of Figure 2.1-1 shov she loci of p ~ts of THEL%pX, 2%ER.

Reactor olant 5yscea pressure and average tempera belov vhich the calo ted DQR is ao less than the correlacion DQR 1 aver ge eachalpy at the vessel exit ts 1ess than the aa t value or che y ef saturate lt 4. Oncertatnttes ta'rimp eystea pressure, core s rasure, eor rial pover, prfaary coolant flov rate, aa4 fua1 fabrtcast tolerane s been tacluded ta the aaalyses from vhteh Figure 2.1-1 ts tve4.

]r~W D. C. COQX 'mÃIT 2 l 2-1

J >,.j; W poet: 41'~+>(u'i

. L1 H i ~$ gPRaW~

The ourves are based on a nuclear ontha?yy rise hot channel factor, and a reference cosine vith a pe4k of Q$ 5 tor axial pover ahap ~ la allovance js incl ded for an increase < F ~ ~ at reduced yover based on the expression:

F N

zl ~ ~ [1 1 T4H (1-z) l e is the o'4 4 tio of fly THg~g, r>u4 VS These I WC ux oonditions are goer than those calcu>rted for the raage of all control rods fully vithdravn to the saxi'llovable control rod insertion assuing the axial pover iabalance is vithin the Xiaits of the

( 'I) function of the Overtaaperatur ~ trip. Shen the axial pover 1abalance 3 not vithin the tolerance, the axial pover iabalance effect on the Overtenperatur ~ 4 T trips vill reduce the setpoats to provide protection consistent vith core safety linits.

2.1.2 tv vet t'he restr'ction of the'Safety Limit yrotects the intel. ity of the Reactor Coolant Systec free overpressurixation and thereby prevents the release of radionuclides coat~ed in the reactor coclant frcn reschin the containnent atoms pher e.

The the lSHE reactor press' vessel aA yressuriser are designed !o Section Code for Nuclear tover tlant vhich perlits a aaxiaun transient III of press -e of l1OS (2v35 ps'g) o. desi': ressur ~, The Reactor Coc'an: Syste=

pipit, I valves and fittings> are desi~ed to LHSX 31.1 195? Edition, vhich yern'ts a aaxixun transient pressure of 1205 (2955 psjg) of conponent design

ress -e. ~e Safety <<=it of 2735 ysig 's therefor ~ corsistent vi

h the desi'r'ter'a aA associated code ra;areerts.

~e e= i-e >es-tor ".oo'ant yste- 's hy".roteste" at e10v "sig tR~S of desi~ yressu."e, to deoonstrate Lnt~-ity prwr to ~tial opera"on.

D.C. COI OPT 2 B 2-2 QKZ)YQHT NO

Tho Power Range Negative Race trfp provides, 'proceccf on to ensure cha the calculated DHRL is aafntafned above tho design DHSR value tor mht le coatrol rod drop accidents. The analysis of a single control rod dro t'op s~~

accideaC fndtcaces a return to full power Nay be initiated by Cho aut

~";

c control system in rospoa'se Co a contiauod Sall yover turbine demand or by the nega ive aodorator temperature feedb k s le contr rod d p aaa ys plan ill t c ro crated r tho 'rim t een Crt ia for forms d Cycl be le ed The Intermediate and Source Range, tucloar Hux trips provide reacco" core proceccion during reaccor scarcupo These crfps provide redundant proceccton to the lov secpoint trtp of 'che Power'Range, Neutron flux chayuls. Tho Source Raage Chanaels 10 vill fniciace a reactor Crtp at about counts per second unless manually blocked vhen p-6 becomes active. Th>>

Intermediate Range Channels vill iaittate a reactor trfp at a cuCrent level proportional to approx~ately 25 perceat of RATED THELM. N'ER unless manually blocked vhen P-10 becomes acctve. lo credit vas takea for operacfon of the trips associated vith either che Intermediate or Source Range Chanaels in che accident analyses; however, their functional capabiltty ac the specified trip settings is required 4y this specificatioa to enhance tho overall reliability of cho Reactor protection System.

The Overcemperacure 4T trtp provides core prococcfon co proven DiQ for all combfnat'ons of yressuze, pover, coolant temperacu=e, and axial power dtscrtbucton, yrovided chat tho transfeat ts slov vich respec Co ptpiag transit delays from che core to the temperature detectors (about 4 seconds),

and piessure is vtchtn tho range becwoea the Hfgh and Lo'v ?ressuCe reactor triys. This'ocpofat includes corrections for chaages ia daastty and heat capacity of vater vtch temperature and dynamic compensation for ptping delays from the coro to tho loop temperature detectors. This reactor crfp ltmft is alvays belov che coro safety lfait u shown fn ftguro 2.1-1. If axial peaks are mro aovoro the design, as fadtcated by the difference between top and botcoi power range nuclear detectors, the reactor trip ts autoaatfcally reduced according to che notactons fn Table 2.2 1.

~

D. C. COOK ~ NET 2 124

]v~

AS S The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to,within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature ie slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL pOWER and Reactor Coolant Temperature and Pressure have been related to DNB through the WRB-2 correlation and W-3 correlation for conditions outside the range of WRB-2. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio, is defined as the ratio of the heat flux that would cause DNB at a(DNBR),

particular core location to the local heat flux, and is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-2 correlation for Vantage-5 fuel, and the W-3 correlation for ANF fuel and conditions which fall outside the range of applicability of the WRB-2). The correlation DNBR limits are established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for WRB-2 and 1.3 for the W-3).

In meeting the DNB design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are statistically combined with the DNBR correlation statistics such that there is at least a 954 probability with a 954 confidence level that the minimum DNBR for the limiting rod is greater than or equal to a calculated design limit DNBR. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the DNBR correlation statistics establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. For D. C. Cook Unit 2, the design DNBR values are 1.23 and 1.22 for Vantage-5 fuel typical and

~C Jl thimble cells, respectively, and 1.39 and 1.36 for typical and thimble cells for the ANF fuel. Zn addition, margin has been maintained in both fuel types by performing safety analyses to a safety analysis limit DNBR.

The margin between the design and safety analysis limit DNBR is used to offset known DNBR penalties (i.e. transition core penalties, rod bow, etc.) and provide DNBR margin for operating and design flexibility.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR limit value or the average enthalpy at the vessel exit is less than the enthalpy of saturated licpxid.

The Overpover 4f reactor trip provides assurance of fuel iategricyy e.g., ao melting, under all possible ovezpowr conditions, Umits the required range for Overtemperacure < proceetion, and provides a backup to the High Neutron Flux trip. The setpoiat iacludes corrections for changes in density and heat capacity of vater vich temperature, and Optic compensation for piping delays from the core to the loop temperature detectors. No c"edit en for operation of this, trip tn che accident analyses; hovever, ics functional caps ac the specified trip setting is required by this specification to enhance the overall reliability of che Reactor Protect S stem. axis a are severe es gay cate the di rance b een to bottom po 'ange n ar detec , the etor rip i tomaci y reduc rding to aotatio n Tabl .2-1.

The Pressuri=er High and Lov Pressure trips are provided to limit the pressure range in vhieh reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure proteecion, and is therefore set lover than.the sec pressure for these vaLves (2485 psig). The High Pressu" ~ trip provides protection for a Loss of External Load evenc. The Lov Pressure trip provides protection by tripping the zeaetor ia the eveac of a loss of reactor coolant pressure.

The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the vater level to a volume sufficient to retain a ate bubb d revet vater zel ef throu 4 e pressurizer safecy va s No edit vas a oz ration o s cz p tri aec ettin nt ~ s; o r, its tional cificat to ility ~ spec requir y this e the of th actor ection em.

T4- p<>u~~ is~~ 4/ warp~ isre < 7~y pi r~uJ~

Ie 8 vA/< L pv-gp c g peer.

D. C. COOK QSZT 2 b 2-5 AMEND?KNT NO.

The Vndervoltaje and Vndezfrequency Reactor Coolant tunP bus trips PzcnCde reactor cor ~ Protection a jai".est D!lb as a result of loss Of voltaje or u~zfrequency to lore Sm one reactor coolant perp. The specified set points assure ~ reactor <<ip sijnsl is jenerated before the lov flov tzip set point ts reached. Thee delays ar ~ incorporated in the undezfzequency and unde+volta je trips to prevent spurious reactor <<jps fzoa SOIentazy e i

~ lectzical pover transients. for undervolt~a, the delay ts set ao chat the ceca reoodrad for a adgcal co reacS che reaccor crdr 'Oreetcere fallacies cha simultaneous tri tvo or lore reactor coolant pump bus circuit breakers shall not excee ~ conds. For underfrequency, the delay is set so that undezfrequency tri set point is reached shall not exceed 0.3 seconds.

/,2 A Turbine Trip causes a direct reactor <<ip vhen operatinj above P-7.

Each of the turbine trips provide tu=bine protection and reduce the severity of the ensuing tzansient. So credit vas taken in the accident analyses for operation of these trips. Theiz functional capability at the specified trip settinjs is required to enhance the overall reliability of the Reactor Protection System.

D. C. COOK Qi?T 2 j27 NjEHDNXHT SO.

3.l.l l The SHUTDOi8 NI51N shall he . jk/k.

/.4 INES I, P~P~

QQJglf:

Vith the SHUTOSO NRClN c l( ~

Jk/k, ianediately initiate and continue horation at p l0 cpm of 20,000 ppm boric acid solution or equivalent until the required SHUTDOWN NRClN is restored.

i.l.1.l.l The SHUTDON MARGIN shall be determined to be g . dQk:

Vithin one hour after detection of an inoperable control rod(s) and at least once per l2 hours thereafter ahi)e the rod(s) is inoperable. lf the Inoperable control rod fs Iaovable or untrippable, the above required SHUTRNH NRClN shall he increased hy an amount at least equal to ihe withdrawn cnorth of ihe Ianovable or untrippable control rod(s).

gpss Vhen in IRIDES l or 2~, at least once per l2 hou'rs hy verifying that control hank withdrawal is within the liaits of Specification 3.3.3.6.

co Vhen in NDE 2, erlthin I hours prior to achieving reactor criticality hy verifying that the predicted critical control rod position is within the liaits of Specification 3.1.3.5.

~e Special Test Exception $ .10.l Nlth K l.o H Nith N ff c l.0 N.C. COOK NIT 2 0/i l-l ~ccrc@ &. ik

d. ft$ or to $ nitial operat$ on abov~ $% RATEO THERNL PStER ~fter each fuel loadina, by consideration of the factors of e belli <ith the cation control banks at the eaxioua $ nsertion 1$ ait of Specif$

3.1.3.5.

e. Shen $ n NOE 8, at least once per tl hours by consideration of the following factors:

~

. l. Reactor coolant system boron concentration,

2. Control rod position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. , Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be coapar'ed to predicted i

values to demonstrate agreement within 1% kk/k at least once per 3]

Effective full Po~er Days (EFPO). This comparison shall consider at least those factors stated in Specification l.l.l.l.l.e, above. The predicted reactivity values shall be ad)usted (noraal1zed) to correspond to the actual core conditions. prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

4.1.1,1.3,Prior to blocking ES unctional Vnits $ n accordance with footnotes i and ii of Table -3, SNJTOON NRCIN shall be deterained to ik/k by consideration of the factors of be reater than or equal to

.l.l.lie above.

~ ~ ~ The Reac or Coolant Systea average temperature used $ n caking jhis SNJTNW NRCIN deterwination shall be less than or equal to 350'F. This SHUTONN NARGIN shall 4e aaintained at all t$ aes Ken the ESF functions are blocked $ nE 3.

D. C. COOK - LJNIT 2 3/i 1-2 4NXNDNXRT SCl o ~

~~ ~ ~

~ ~ ~

~ ~ ~ ~

~ ~

HO ~

n A

0 0

.Z lO 5.0 CI 4.0 Oj Z gS U goo z

3.0 I O o

CJ Z

O 5 2.0 SSOWOSSMODE i gP~

5.0 m

zO 0.0 0 5400 z

0 8ORON CONCENTRATION (PPM)

BIO C5l FIGURE 3.t-3 REQUSEO SHUTOOWN MARGN

3.1,1,+ The moderator temperature coefficient (ETC) shall beg ~<<~<~ ~~

4i~rtj slscefic/ IN fA CAtlrpsrip CI .N p fj4Mr+47C, g) 74< HAvrev~

d uj )gal C.i~(r gg,g. g gs'ye~ i<<C" < 1r 74m r~4cs i jy 444 S b ve h gc, ,~ sF Cycle C.) r 8'si) c.i~ r NODES 1 and 2+ only' NDES 1, 2 and 3 only>>

<<CycLg &Fr Cgoh)LI~ 7 gpgcIF rP se yA C0C.

Mi h the HTC more posi ive than th imit

1. Establish and maintain control rod vithdraval limits su!ficien: to restore th>> NC to vithin its limits vi:hin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STAÃDbY vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These vithdraval limits shall be in addition to the Lnseition limits of 5pecification 3.1.3.6.

2. Main ain the control rods vithin the vithdraval limits established above until subsequen: measurement verifies that the NTC has been restored to vitt..'".. its limit for the all rods vithdravn condition.

3. Prepare and submit a Special Report to the Commission pursuant to 5pecification 6.9.2 vithin 10 days describing the value of the Ieasured NTC, the interia control rod vithdraval limits and the predicted average core burnup necessary for restoring the positive NTC to vithin its limit for the all rods vithdravn condition.

ES gf'ccifirO a+ 7AC C+<+

b. M!th the NTC core negative than the imit be in NOT SHUTDOWN vithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• Mich than or equal to 1.0 X

ff greater

~ See Special Test Exception 3.10.3 D. C. COOK ~ VÃIT 2 3/4 1 5 ~~ gO, 4KiRP

CTI CONTROL 5YSTENS VK UJLHCE +Z UIRXNENTS 4.1.1.4.1 The NTC ahall be determined to be within Sta liaise by confismasory aeaeureaensa. Naaeured ETC valuaa ahall be extrapolated and/or compenaated to yoriit direct coapariaon vith the above liaise. 4 4.1.1.4.2 The NTC ahall be datelined to be within tta 1Lsite during each fuel cycle aa follove:

a) The ETC ahall be measured and c ared to the SOL li5lit ~ prior to ini operas on ove 5i of NhTKD tOVER, cer each fuel loading.

b) The HTC shall wag<<d be measured as any %tEINL tO vishin 7 EFPD after reaching an equilibrium boron concentration of 300 ppm and the value compared to the Xn the event this comparison indicates at the ETC vill, be mori negative than the EOL imit, the MTC ahall be remeasured at least once pe 14 EFPD during the remainder of the fuel cycle and the HTC value compared to the EOL Iiait.

.~J, ~ cDCQ pp~ 54rM I/< ~cQ

~

1 co ag.

D. C. COOK UNIT 2 3/4 1-4 ~ndment Ho. ~

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System lowest operating loop temperature (T ) shall be ~ 541'F. r APPLICABILITY: NODES 1 and 2 .

ACTION:

Mith a Reactor Coolant System operating loop temperature (T ) c 541'F, restore (T ) to within its limit within 15 minutes or be $ E HOT STANDBY within the $ 3xt 15 minutes.

SURVEILLANCE RE UIREMENTS 4.1.1.5 The Reactor Coolant System temperature (T ) shall be determined to be ~ 541'F:

a ~ Mithin 15 minutes prior to achieving reactor criticality, and At least once per 30 minutes when the reactor itical and the Reactor Coolant System T is less tha with the T -Tr reff Deviation Alarm not preset.

. avg s'w~/-

With Keff ~ 1.0.-

Q. C. COOK - UNIT 2 3/4 1 7

3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

MODES 5 and 6.

hQIQH:

r'.

Qith no charging pump OPERABLEsuspend alleoperations involving CORE ALTERATIONS or positive reactivity changes.*

b. With more than one charging pump OPERABLE or with a safety injection pump(s) OPERABLE0 when the temperature of any RCS cold leg'is less than or equal to 152 F, unless the reactor vessel head is removed, remove the additional charging pump(s) and the safety injection pump(s) motor circuit breakers from the electrical power circuit within one hour.

C~ The provisions of Specification 3.0.3 are not applicable.

4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying that, on recircu w the develops a discharge pressure of greater than'r equal o 4.0.5.

M~~o

+1QQ pSlCL pump n tested pursuant to Specification 4.1.2.3.2 All charging pumps .an sa ety injection pumps, excluding the above required OPERABLE charging pump, shall be demonstrated inoperable by verifying that the motor circuit breakers have been removed from their electrical power supply circuits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, except when:

a. The reactor vessel head is removed, or

b. The temperature of all RCS cold legs is greater than 152 F.

• For purposes of this specification, addition of water from the RVST does not constitute a positive reactivity addition provided the boron concentration in the RUST is greater than the minimum required by Specification 3.1.2.7.b.2.

D. C. COOK - UNIT 2 3/4 1-11 AMENDMENT NO.8 7

REACTIVITY CONTROL SYSTEMS CHARt'INC PtNPS ~ OPSQTIHC CONOITION FOR OPERATION 3.1.2.4 At least ha chargfny pumps shall he OPERABLE.

APPLICABILITY: NOES 1, 2, 3 and 4.

ACTIOII:

Nth only OPERABLE one. chargfng pump OPERABLE, restore at least ~ chargfng puaps to status vfthfn 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be fn at least HOT STANOSY and bor ated to a SHUTDOWN NRt IN equfvalent to at least 1% Sv'k at 204 F ~fthfn the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at 1east >o chargfng pumps to OPBABLE status Hthfn the nex 7 days or be fn COLO SHUTDOWN wfthfn the next 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />.

SURVEILLANCE RKOUIRBIEHTS 4.1.2.4 At 1eas t~o cha-. fry pumps sha11 he demonstrated OPERABLE by ve.f fying, that on recfrcu1atfon l'1m, each pump Ceve1ops a cfs-barge pressure of > ~P4S-t

~hen usted pursuantt 4 Specff'fcatfon 4.0.5. Z'250 gs]

O. C. COOK UNIT 2 3/4 tI2 QcL455t

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE:

a. A boric acid storage system and associated heat tractfng with:

A minimum usable borated water volume of 4300 gallons,

2. Between 20,000 and 22,500 ppm of boron, and

3. A minimum solution temperature of 145'F.

b. The refueling water storage tank with:

l. A minimum usable boratcd water volume of 90,000 gallons,

2. A minimum boron concentration of 24 0 ppm, and

3. A minimum solution temperature of APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE. suspend all operations involving CORE ALTERATIONS or posftive reactivity changes* until at least one borated water source fs restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

Verifying the boron concentration of thc water,

2. Verging the contained boratcd water volume, and

3. Verffyfng the boric acid storage tank solution temperature when ft fs the source of borated water.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying thc RWST temperature when ft is the 'source of boratcd water. I

• For purposes o this specification, addition of water from the RWST does not constitute a dilution activity provided the boron concentration fn the RWST fs greater than or equal to the minimum required by Specification 3.1.2.7.b.2.

D. C. COOK - UNIT 2 3/4 1-15 Amendment No. 82.

REACTiVITY CONTROL SYSTEMS BORATED MATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLK:

a. A boric acid storage system and associated he cfng with:

T7/

l. A minimum usable borated water volume o gallons,

2. Between 20,000 and 22,500 ppm of boron, and

3. A minimum solution temperature of 145'F.

b. The refueling water storage tank with:

1. A mfnfmum contained volume of 350,000 gallons of water,

2. Between 2400 and 2600 ppm of boron, and

3. h minimum solution temperature of dp APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equfvalent to at least lX ak/k at 200'F; restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOMN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. Nth the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be fn at 1east HOT.STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE:

D. C. COOK - UNIT 2 3/4 1-16 Amendment No.

c) A pover distribution map g,s obtained from the aovable incore detectors and F (Z) and T" are verified to be vithin eheir 1iwita vithin 7P houcs, aII d) Either the IHHNAL tOMER level is reduced to less than or equal to 750 of RATED TBEIQQL POtCR vithin one hour and viehin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutron flux trip seepoi'nt is re duced to less than or equal to $ 5l of RATED THERMAL 4'OAR, oc')

The remainder of the rods in ehe group vith the inoperable ':

rod are aligned to viehin g 12 seeps of the inoperable rod vithin one hour vhile maintaining the rod sequence and inser-tion limits  ; the THECAL ROVER level shall be restricted pu t to Specification 3.1.3.6 during sub-sequen eration.

P~P g -~ +

Ceres OtW~+i~~

5

c. y~ j PS EDP~~

4.1.3.1.1 The position of each full lengeh rod shall be determined to be viehin ehe group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals vhen the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

P 4.1.3.1.2 Each full length rod not fully inserted shall be determined eo be OPERABLE by movemene of at least 8 steps in any one direction at least once per 31 days.

D. C. COOK - VNIT 2 3/4 1 19 ANEHDMENT NO. JSe

J/gglfll/ i 74'gg, /pe Jl ] Cg 4 7l /py/Pr f (CIA 8) 3.1.3.4 The individual full e hutdovn and control) rod drop time from the fully yithdravn position shall be less than or equal t . p.7 seconds from beginning of decay o stationary gripper coil voltage to dashpot.

entry vith:

greater than or equal to 541 F, and

a. Tavg

b. hll reactor coolant pumps operating.

MODES 1 and 2.

hRGQE:

With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to vithin the above limit prior to proceeding to MODE 1 or 2.

a 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to entering MODE 2: J

a. For all rods folloving each removal of the reactor ve'ssel head,

b. For specifically affected individual rods folloving any maintenance on, or modification to the control rod drive system vhich could affect the drop time of those specific rods, and

c. ht least once per 18 months.

D. C. COOK - UNIT 2 3/4 1-23 AMENDMENT NON o

Cq~iYcP re PHySad 'H>F~~> + +" l 4' 3.1.3.5 hll hutdovn rods shall be 7 Ac cecr <r~riup iri'rrr<<Boa'pt7 ltaUUZI:

hfZIQH-'NS~~re P r Yi 1 IkfCPT7PN g,g~ iy one s utdovn rod except for surveil-With ~ aax lance t stin fpcJTaks o

ursuant gh t

g 7 7i S ecificat on 4....

tu Alit Y4/ Ih'R <

vi

>I 7 ~Jr F J '

T'A CO/.jp

~.

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

Yhi rer~zt.isa 4.1.3.5 c lwi7 a'p c c r ic.i r0'pg QD<p"

a. Wit n minutes pr or to vithdraval of any rods in control banks A, B, C or D during an approach to reactor criticality, and

b. ht least once per 12 hours thereafter.

• See Special Test Exceptions 3.10.2 and, 3.10.3

<<With K ff greater than or equal to KO D. C. COOK - UNIT 2 3/4 1-24 hMENDMENT NO.

3.1.3.6 The

~ Jf'rcr i

control banks shall be limited in hysical insertion as ehovn-~

ip yz< pNif pjpgp7/g c/~i ~ +pe p~ g Dc MODES 1+ and 2M.

MXIQE:

With the control banks inserted beyond the insertion limits, except for surveillance testing pursuant to Specification .1.3.1.2, either:

a. Restore the control banks to vithin the 1haits vithin tvo hours, or

b. Reduce THERMAL POWER vithin tvo hours to less than or equal to that fraction of RATED THERMAL POWER vhich is alloved b the rou position using the IJUFlflld& /rtt/7 J Jjcf~p pep lt Y~ CoC)g

c. Be in at least HOT STANDBY vithin 6 ours.

4.1.3.6 The position of each control bank shall be determined to be vithin the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals vhen the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• See Special Test Exceptions 3.10.2 and 3.10.3.

e With K ff greater than or equal to 1.0.

D. C. COOK - UNIT 2 3/4 1-25, AMENDMENT O. , ~

L 7 LIV tMfyWithdrawn) to.ot,me) 0},82,RR)

~ e gg

'0l ~

g,2t2) tl

~ >Q

~ ~ 10 ~

SANK ~~0~

200 (1.0 189)

SANK C a %00 NK (0 84 50 (0.19 0) 0 4L2 ~

4L4 4j yg

)

FRACTiON OF RATED THE IDEALFCREE

+

Sgue 3.1.1

. SOD SANK lNSEhTION 1.tMtTS f VE~

OUR LCKP CPERATlON THECAL yyyyEg

/L Q 1t 3.2.1 The indicat FLUX DIFFEICNCE (AFD) shall be aatntained vfthfn the target, band about a target flux difference. carr S~iJ p SrcciliiD iw T4~ grrerr~f gyes sgq+g gy~p(c4(

BUCLULUX:

hfZZQE:

a. Vtth the indicated AXIAL HID( DIFFERENCE outside of the target band about the target flux difference and vith THE1NAL POVER:

dv

1. Above 90% 0.9 x APL (vhtchever ts less) of RATED THER'. POVER,

~

vtthin 15 minutes; a) Either restore the tndfcated AFD to vtthin the target band 1tmfts, or b) Reduce THMAL POVER to less than 90% or 0.9 x.APL (vhichever ia less) of RATED THEINAL POVER.

2. Setveen 50i and 901 or 0.9 x APL (vhtchever ia less) of RATED THEINAL POVER:

a) POVER OPERATION aay continue provided:

1) Th>> indtcated AFD haa not been outside of the target band for aors than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation cumulative .durfng the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and Pea%'<J i~ 7<

2) Th>> indicated is vtthtn the lfaLita

~1. AFD Othervtae, reduce THE1NAL POVER to less than SOa of RATED THEINAL POVER vithtn $ 0 atnutes and reduce the Ponr Range Neutron Flux.-High Trtp Setpotnta to less than

~ r equal to 551 of RATED THERQL POQER vithin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b) Surveillance testing of the tover Range Neutron Flux Channels aay be performed pursuant to Specification 4.3.1. rv indtcated AHi fs aaintatned vtthin the liat f/4 ~ glib/

A total of 1C bours operation may be accuaula ~ vith e AFD outside of the target band during this testtng vithout penalty deviation.

+See Special Test Exception 3.10.2 D. C. COOK UNIT 2 3)4 2-1 AMENDHENT NO.H o

4.2.1.2 The in4tcate4 AFD shall be coast4ere4 outsi4e of tts target band Ken ac least 2 of 4 or 2 of S OPERAILE oxcore channels are indteattng the AF) ce be outside the tatget band.. Pena?ty deviation outside of the target band ahall be accumulated on a etae basta of:

A pena17 deviation of osLo %faute ter oath one %faute of ?SKL OPERATZON outside of the target band at THXLSLL 8Ã'ER levels.equal to or above 500 of RATED THERE'LL FABER, and

b. A penalty devtatton of one half atnute for each one abate of POVH OPXIATIOH outside of che target band at THEL%d. KEEL levels betveen 1Si and 501 of RATE) TBELSQ tNER.

4.2.1.3 The target axial flux 4ifference for the OPERABLE exeote channels applicable.

shall be detetmtned tn conjunction vith the aeasurenent of APL as defined in Specification 4.2.6.2. The provisions of Specification 4.0.4 ate not 4.2.1.4 The axial flux difference target band about the target, axial flux difference shall be datelined tn confuncttoa vith the aoasuteaent of APL as g)ceil ~ ~

JO fined tn S eciftcation 4 .6 The allovable value! o the tar et band a=e CoLR 0

applicable.

~ ptovtstons of Specifieat on .. ate not

~e D. C. CXX NIT 2 S/4 2 S Ammme IO.

FIGURE 3.2 1 ALLOWABLE DEVIATION FROM TARGET FLVX DIFFERENCE g) 20 100 ... I . I.

I

(->o. 0) (H->0,9O) Ulloc table, Operati I

~ 80 (-8 SO). (+8 90)

+3K T t+-. OfgC 8 fld ag

~ so tfon

(- S,5O) 25, jPm l 23.Q} (+23. )

20

)

0 0.

-10 0 $ 0 20 Devlatlon frne Target Flux Difference

3.2.2 F~(Z) ahall be liafted b e folloving relationships:

~Q'P F~(Z) 5 (zeal Fq(Z [K ] P % 0.5 F~<Z) S tK<Z)] F g .OJ ( P S O.S RATED THERNLL POVER

~ F (Z) ia the measured hot chaape1 factor including a 3X nanufac-

~turing tolerance uncertainty and a SX aLeaau?chant uncertainty

~ Z ia the

~ PP g1$ ev + C'le) g p (('-q~a7 ( (lv pp ('(dg 1 ~r'cw( ((r 0*

C'oc, a DE hEEE:

Vith F~(Z) exceeding ita linit:

a. Reduce THEL(A POUR at least 1% for each 1% F (Z) exceeds che liaLit vithin 15 ninutea an4 ainilarly reduce t9e Pover Range Neutron Flux-High Trip Setpointa vithin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; PS ER OPERATION nay proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; aubaequant POQZR OPEMION nay proceed provided the Overpover hT Trip getpointa have been reduce4 at leaat 1% for each 7.% F (Z) exceeds the limit.

b. 'dentify and correct the cawe of the out of limit condition prior

~

Co increasing THEL%LL?CNER above the reduced Ifnit reyCT'ed by a, above; THEL~ tOQER say then be tncreaae4 provided F (Z) ia denonatrated through incore aapping to be vithin ita Sist.

F g p- g.(m,

(((( ~ r a~ ghee 0 7 g-Eel 'P~

gk(gog~r(c~)

~p( c(F(sd )~ y' P D. C. COOK ~ UNIT 2 3/4 2-5

~

AHEND~ NO.

0

t I

~ ~

l

I ~ .'

~ ~

II~ I ~ I II

~

~

~

I ~ '(1 iiO+%37)

~ ~

I

~~

I Il ~

i

~

I ~

~ I I~

(12 o 0 I0o 710)

~

I

~ ~ I II',

Il

.I ~

~

I 2

ttcetE 3.2-2(a) 5. C. age( gNT 2, 6

COllt IIEICllt OF CA!IF. IIKICIIT RN EXXO!l NICLEhlt COi (III ~

K(L) NNNALlKED F (t) 10

%le 12 h5 h lOCTCOII

00 ~ 0 eee0 0

~ ~

~0

~ ~ r qJ,e

<<4 <<0 ~ ~ ~ ~

! ~ s i, OOO <<+~O ~ ~

~ ~

~ 00 ~ 00<<0 1.1d gt

~ ~ fy L. i::~ O~ ~ ~ ~

~ ~0 ~ Oe ~ 0 ~ \ ~ eg ~ 000

~ ~0 +5 Ttrgtj tend lk.'25, l . i2

(~.2S, 10>> )

'i

~ &00

~i+0'0 le g }

< te ~ ~ ~

~~

~0

~0 aS 1 Oi)

+

~

l y

) ~ ~ ~ ~

~

~

~ ~0f

~

~

~

~0 ~ 'P <<O0 0 O F0 ' ~ 0 ~ ~0

~ 0 'O&O <<0 ~ W 00 ~ O~ O~

1.04

<<<< I~0 ~ 0~. ~% &0

~

f

\ ~ ~

~ ~

f ~

0 <<0 0 0 1.0Z

~ <<<<, ~ 00<< ~ ~ ~ ~ ge

~ 00 ~ t ~00 gas

~ ~

OO~

~~ OO ~ ~0 ~ '

~

t

~

$~ ' 0 I .

1.00 D 4 4 I 0 lD

~a1 Hefgtn (t'eat)

'figgrg 3,2 3 Tt7) ls l Fun@ noh Of ~ Helps

H Vl~tl~

3.2.3

~e V~ s}all be limited by the folloving relationshi F~a S ~'tl + (1 P)]

vhere s the fraction of RATED T}KLIALPOLER NODE 1 V

Vith FH exceeding its limit:

gH

~. Reduce THERMAL PO'iMc. to less than 50% of RATED THERM. POMXR vithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and red ce the Pover Range Neutron F'ux-High Trip Setpoints to less than or equal to 55i of RATED THELSLL POVER vichin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />'.

Demonstrate through in-core mapping that P is vithin its limit vichin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit Q reduce THERMAL TSAR to less than Si of RATED THKh%Q. POHR vithin the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and

c. Identi y andcor.ect the cause of the out-of-limit condition prior to increasing THER'. POVKR; subsequent POt KR OPERATZON may proceed, provided that F'H is demonstrated through in-core mapping to be vichin its limit at a nominal 50l of RATED TBEL%Q. ?O'ER prior to

~ xceeding this THURS'NER, at a nominal 75i of RATED THKL%Q ?S ER prior to exceeding this THKLVL tOVKR and vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95'r greater RATED THER'. NKR.

nr~

is 7hr f'-imi oH AI

< ggzz p 7 Hrwe~~ I o gpgcgAid /H 1~+ PI rli.~russ r~ r p~ iver CC.Oid) piwre /~DE jul wiprAr~ Fi< I $ 'P~'l,~

PH C,oL)Z D. C. COOK VNIT 2 3/4 2-9 uaznNKNT NO

OW UT 0 B a av ERA ING A ERS GCTOOOPO 3.2.5 The following DNB and Tavg related parameters shall be maintained within the following'operational indicated limits:

a. DNB

1. Reactor Coolant System Tavg, 578.7 F*

2. Pressurizer Pressure ) 2194 psig*/**

3. Reactor Coolant System 366, 400 gpm***

Total Flow

b. Tavg

1. Reactor Coolant System Tavg ) 542.8 F*

IP "

~CTICN:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 54 of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIRE ENTS 4.2.5.1 Each of the parameters shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The indicators used to determine RCS total flow shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS total flow shall be determined by a power balance around the steam generators at least once per 18 months.

4.2.5.4 The provisions of Specification 4.0.4 shall not apply to primary flow surveillances.

Indicated average of at least three OPERABLE instrument loops.

• • Limit not applicable during either a thermal power ramp in excess of 54 of RATED THERMAL POWER per minute or a thermal power, step in excess of 104 of RTP.

• • • Indicated value 3/4 2-15

S.2.5.1 Waits ahevn on ol cnCng CNl le S.2 1:

related paraeeters shall aaintained vi~

Oo Reactor lant 5ystea T

b. ftessurixe pressure.

Co Reactor Coo t 5ystei Total Fiov te.

MODE 1 hfZZQE:

Vith any of the above parce ters ex odin'ts lait. restore the pareneter to vithin its liccit vtthin hours r reduce SiBUM. N:KR to less than 5% of RATXD IPSLJQ.?OVER vithin ne 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.2.5.1.1 Each of the par eters f Table S.2 1 shall be verified to be vithin their linits at 1e t once er 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.1.2 CALZbRhTION The RCS at least o ce per le aonths.

i tote flov rate dicators shall be subjected to a CHLOE 4.2.5.1.S The RCS t tal flov zat>> sha 1 be deterained by a pover balance around the steam p erators at least ence per 1$ aeths.

4.2.5.1.4 The p sicns of 5pecificat on 4.0.4 shall oot apply to pr flov eurreillanc s.

D. C. COOK IHZT 2 S/4 2-15 AKBCZKh O. I'

?aMKZEE Reactor Coolan System T I g 576. f. (indicated)

Pressurizer Press re 05 paid Reactor Coolant Sys em Total Flov Rate x 10 lbs/hr

'38.6 Limit not RATED TH RATED plicable during either a POWDER per Rtnute or ~

POWER.

~ rlHlp in excess PNKL step in 'excess f Se f 1 Os Indi ted 4veraie of at least three OPERhS Sna~nt loops.

~ ~ .3. 1 penalty for seasureaent uncertainty fncl d in thi alue D. C. COOK UNIT 2 3/4 2-16 AftEÃDMQC NO. 52

3.2.5.2 The tolloving MS related paraseters ahall be aain ined vithin the lfatts ahovn Table 3.2 2:

a. keact Coolant Iystea T

b. tnssur er Pressure.

MODES, 3, 4 and 5 dfZLK:

MODES 2 and 3 Uith any of the above pa eters exc edinI its 1Lait, restore the parameter to vithin its 1 it vithi 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or open the reactor trip system breakers vithin the xt h MODES 4 and 5 Uithin one hour either open i eactor trip systea breakers or render incapable of rod vithdraval.

the control rod drive syst 4.2.5.2 Each of the par stere of Table 3.2- shall be verified to be vithin their 1$ sLits at least o ce per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l I

• Qith the reactor trip system brea'kers in the closed position and the control rod drive systea capable of rod vithdraval.

D. C. COOK. - KT 2 3/4 2-17 'AÃBIDMBIT NO

IJHZT Reactor Coolan System T avg 549.2 F. (Reactor Subcritical)

Reactor Coolant Syst T avg g 516.3 F. (Reactor Critical)*

PzessQrizer PressUre > 2176 psig Reactor coolant loop op erationa iraq ements are contained in Specifications 3 ~ 4.1.1, 3.4.1,2.c and 3.4.1.3 c.

Indicated averag>> of at ast three OPELQ instnuaent loops.

D. C. COOK - UNIT 2 3/4 2 18 +mern'I NO52,

T Af~f iO

~ i'I ftA 3.2.6 RMhL POKR shall be less than or equal to hLiNhbLE POWER LEVEL (QL), given by the follovin elationships:

RS/

V p r'L

- min over Z of F (Z) V(Z)xF x 100', or 100%, vhichever is less.

hPL over Z of F x 100', 0%, vhich is s~

P

~ F (Z) is the measured hot channel factor, including a 3%

aRnufacturing tolerance uncertainty and a 50 aeasurement uncertainty.

TAc C04 R

~ V(Z) is the function defined in 8 ~

p (~)

max ovu~ ~ o+ <C~)

~ F 1.00 except vhen successive steady-state str u 0 eRps indicate an increase in , vith exposure.

Then either of the folloving penalties, F,p sh 1 be taken:

F 1.02 or, P

F 1.00 provided that Surveillance Requirement 4.2.6.2 is sltisfied once per 7 Effective Full Pover Days until 2 successive aaps indicate that the s not increasing. MAx ouCA W o f' ('~)i

\

~ The above. liiait is not applicable the folloving core reg ons.

1) Lover core region Oi to 10% inclusive.

2) Upper core region 90% to 100% inclusive.

MODE 1 HAP ls r]>> g ~,~,~< ggvFD 7]Jfg~A'L jdi /EfP CA Ops gr7r ~. ll,ri grip r(cot./?)

D. C. COOK - UNIT 2 3/4 2-19 urn@MENT SO.S2

A o Es s

1. Hanual Reactor Trip NOT APPLICABLE

2. Pover Range, Neutron Flux < 0.5 seconds*

3. Pover Range, Neutron Flux/

High Positive Rate NOT APPLICABLE

4. Pov r Range, Neutron Negative Rate Flux'igh

< 0.5 seconds+

5. Interaediate Range, Neutron Flux NOT APPLICABLE

6. Source Range, Neutron Flux NOT APPLICABLE

7. Overteiperature < 6.0 seconds*

8. Overpower hT NOT APPLICABLE

9. Pressuriaer Pressure Lov < +-.0 seconds

10. Pressurizer Pressure High < ~

4.0 seconds ll. Pressurizer Mater Level High

< g.o zseo~>>

~Neutron detectors ara exempt frox resPonse tive testing. Respon e he of the neutron flux signal portion of the channel shall be measured from detector ou"p"t or electronic coaponent in channel.

TABLE 3.3-2 Continued REAClOR TRlP SYSTFH 1NSTRINFNTATlON RESPONSE TINES FUNCTlONAL IJNIT RESPONSE TIRE

12. loss of Ftm - Single loop /,0 (Abave P-B) seconds

13. loss of Flat - Tm loops lo (Above P-7 and bein P-B) c seconds eO

14. Steae Generator Mater Level Lm-Let seconds

15. Steae/Feehater Fler HIsmatch and LoM Steam Generator Mater Level lCADLE

/.4

16. IJndervo1 tage-Reactor Coolant Pumps seconds

17. IJnderfrequency-Reactor Coolant Pumps c 0.6 seconds

18. Turbfne Trip A. Lnr Fluid Oll Pressure- NT APPLWCAhLK

0. Turbine Stop Valve NOT APPLlCABLE

19. Safety lnjectfon input from ESF HOT APPLiCABLK

20. Reactor Coo1ant Pump Areal'cr Position Trip N0T APPLICABLE

TABLE .3-4 Continued

~INEERE5 SAFETY FEJLTtNt ACTUATION SYSTEN INSTRUNRlITATION TRIP SEITOINTS UNCTIONAL TJNIT TRIP SETPOINT ALTANABHt VATIJES STEAM LINE ISOLATION

a. Manual Not Applicable Not Applicable

b. hutoaatie Actuation Logic Not Applicable Not, Applicable

c. Containment Pressure High-High < 2. 9 psig c 3o0 ysig

d. Steaa tlpr in Two Steaa Lines < h function defined as c A function def ned ae High Coincident vith T follows: hp cogrespond- follavss y copespond-Lcm-fur I,Q ng to . x 10 lbs/hr steaa ilo etveen Of and ing t x 10 lbs/hr ateax f eve bete@en Of and 20f load and then a hp 20f load and then a bp in-increasing linearly to creasing linearl hp ap6corresponding to 2 X copesponding to x 10 lbs/hr at full 1 10 lbs/hr at fu oad. 4 T > 541 F. Tavy- ~ 539 t.

e. Steaa Line Pressure>>-Lut > 600 psig steaa line a 585 yaig Iteaa line pressure yrasauro 5e TORBINI TRIP NlD PKENthTER ISOIATION

a. Steaa Cenerator Mater Level- < 67% of narrow range c 684 oC narra range High-High Instruaent span each Iteaa Instrument span each ateaa generator generator

TABLE 3.3-5 EHG?NE=RKD S>>FE ~ i'EA 'PES <ESPCNSE IN[T'.AT).'lG SHN'L>>N0 PJNCT:GN RESPONSE T..uE r i S:r't S I Manual Saf'ety lnjec icn (ECCS) Not Applicable F eedwa ter Isola tf on Not Applicable Reactor Trip (Sl) Not Applicable Contafr~ent isolation-Phase "A" Not Applicable Contafr~ent Purge and Exhaust isola tf cn Not Applicable Auxiliary Feedwa .ar Fumes Not ApplicabIe Esse.".'tia! Service '~ater Systen Not Applicable Ccntair..-..en Af ~ec r ula cn Fan Not ApplicabIe

b. Contair..-..ent S."ray Hot Applfc ble C=ntain.-..ent ! soIaticn-Phase "8" Not Applicable Con!afr..-.ent P r"e ard Exhaus Isola tf cn . Not Applicabla

c. Ccn.ainrent !solat cn-Phase " " Not Applicable Ccntai...-.ent Purge anc ~.'".aust iso la << I ch Not "pplc bl

d. S:e -., Line .sciatfon Not PppI '"*

e>>+ />>Sssure wfcg

a. Sa r e:y I r. ec .. cn ( =:"S)

b. Reactor rip (.rc' S:) 41,

c. Feedwate>> Isolatfcn d, Contafenert lsolatfcn-Phase "~" t1 '

tO <<

e. Ccntafr-..ent P.rge and ~exhaust isola tf cn Not Applicable Auxil fary Feecwa:er Pumps Not Applic ble

g. Essential Serice Ma e. System No. Apclfc bl 0 ~ C ~ CvllK <<<<Njl 2 3/>>'-Zo

34

a. Safety Injection (ECCS) g 24.0~/12.0f

b. Reactor Trip (from SI) g 2.0 C. Feedwater Isolation g 8.0 Containment Isolation-Phase 'A" c IB.Of

e. Containment Purge and Exhaust Isolation Not Applicable Rotor Driven Auxiliary Feedwater Pumps c'0.0 9, Essential Service Mater System c 48.0~/13.0N

4. iff r n i 1 Pr ur w n m

a. Safety Injection (ECCS) ~ 12.0$ /24.0NN

b. Reactor Trip (from SI) c 2.0 ce Feedwater Isolation < 8.0 Containment Isolation-Phase "A" g 18.0f/2&.Off

e. Containment Purge and Exhaust Isolation Hot'Applicable Hotor Driven Auxiliary Feedwater Pumps c 60.0

g. Essential Service Mater System c 13.0$ /4&.ONN am Flow in w t am in - h with Tay Low Low a~ Safety Injection (ECCS) Not Applicable

b. Reactor Trip (from SI) Not Applicable C~ Feedwater Isolation Hot Applicable Containment Isolation-Phase. Rot Applicable

e. Purge and Exhaust

'A':ontainment Isolation Hot Applicable Auxiliary Feedwater Pumps Not Applicable

g. Essential Service Mater System licable

h. Steam Line Isolation c+9- l3 D. C. COOK - UNIT 2 3/4 3-27 ~MENT No.

~ ~ ~ ~ ~ ~ ~

~

g

6. ee w

a. Safety InJection (ECCS) g 12.0¹/24.0¹¹

b. Reactor Trip (from SI) g 2.0

c. Feedwater Isolation S 8.0 d.. Containment Isolation-Phase h" g 18.0¹/28.0~

e. Containment Purge and Exhaust Isolation Not Applicable

f. Notor Driven Auxiliary Feedwater Pumps c 60.0

g. 'ssential Service Mater System < 14.0¹/48.0¹¹

h. Steam Line Isolation ( 8.0 7.

b.

C d.

~

Containment Spray Containment Isolation-Phase Steam Line Containment Isolation hir Recirculation "8"

Fan

~

S 45.0 5

ll0 ble

a. Turbine Trip < 2.s

b. Feedwater Isolation 9.. V ~~ V V

a. Motor Driven Auxiliary Feedwater Pumps c 60.0

.urbine Driven Auxiliary Feedwater Pum < 60.0

10. 4

a. }iotor Driven Auxiliary Feedwater Pumps S 60.0

a. Motor Driven Auxiliary Feedwater Pumps S 60.0 12.

a. Turbine Driven Auxiliary Feedwater Pum S 60.0 D. C. COOK UNIT 2 3/4 3-28

3.4.1.2 a. The reactor coolant loops 1isted belov shall be OPERABLE and in operation as required by items b, c, and d:

1. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,

2. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,

3. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,

4. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump.

b. At least tvo of the above coolant loops shall be OPERABLE and at least one loop in operation if the reactor trip breakers are in the open position, or the control rod drive system is not capable of rod vithdrawal.*

7w>

C. ht least of the above coolant loops shall be OPERABLE and in operat on when the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal.

d. At least three of the above coolant loops shall, be OPERABLE and in operation above P-12. (Refer to Technical Specification 3.3.2.1, Table 3.3-3 for instrumentation requirements.)

• All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitteg that would cause dilution of the reactor coolant system boron 0

concentration *, and (2) core outlet temperature is maintained at least 10 F belov saturation temperature.

~ For purposes of this specification, addition of vater from the RUST does not constitute a dilution activity provided the boron concentration in the RWST is greater than or equal to the minimum required by specification .

3.1.2.8.b.2.

D. C. COOK - UNIT 2 3/4 4-2 hMENDMENT NO. III

3.4.1.3 ~. The coolant loops lfsted belov shall be OPERABLE and in operation as required by items b and c:

Reactor Coolant Loop 1 and fts associated steam generator and reactor coolant pump,*

I

2. Reactor Coolant Loop 2 and fts associated steam generator and reactor coolant pump,*

3. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,*

4. Reactor Coolant Loop 4 and fts associated steam generator and reactor coolant pump,*

5. Residual Heat Removal - East, ~

6. Residual Heat Removal Vest ~

b. ht least tvo of the above coolant loops shall be OPERABLE and at least one loop in operation if the reactor trip breakers are in the open position, or the control rod drive system is not capable of rod vithdraval.~

7Q'o c ~ ht least of the above reactor coolant loops shall be OPERABLE an in operation vhen the reactor trfp system breakers. are in the closed position and the control rod drive system is capable of rod vithdraval.

MODES 4 and 5

+ A reactor coolant pump shall not be started vith one or more of the RCS cold leg temperatures less than or equal to 152 F unless 1) the pressuriser vater volume is less than 62% of span or 2) the secondary vater temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures. Operability of a reactor coolant loop(s) does not require an OPERABLE auxiliary feedvater system.

~ The normal or emergency pover source may be inoperable fn MODE 5.

~ hll reactor coolant de-energized for up pumps and to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> residual heat removal pumps may be provided 1) no operations are permftted that vould cagg+dflutfon of the reactor coolant system boron concentratfon *, and 2) core outlet temperature is maintained at least 10 F belov saturation temperature.

~ For purposes not, of this specification, addition of vater from the RVST does constitute ~ dilution activity provided the boron concentration in the RVST is greater than or equal to the minfmum required by specification 3.1.2.8.b.2 (MODE 4) or 3.1.2.7.b.2 (MODE 5).

D. C. COOK - UNIT 2 3/4 4-3 AMENDMENT NO. g

~ ~

~ ~

~ ~

REACTOR COOLANT SYSTEM WRESSUR12ER It1ITING CONDITION FOR OPERATION 4.4.4 Th ressurizer shall be OpERABLE vith a water volume less than or equal to of span and at least 150 kM of pressurizer heaters.

gg ISDEI \, R. L iCTCOM:

Vith the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters, either restore the inoperable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the nert 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and in HDT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With the pressurizer otherwise inoperable, be in at 'least HOT SHUTDOWN with the reactor trip breakers open vithin l2 hours.

SURVEILLANCE RE UIRE!lENTS 4.4.4.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.4.2 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least ance per 18 months by transferring power fram the normal to the emergency power supply and energizing the required capacity of heaters.

O.C. COOK VHIT 2 3/44 5 lpandment Ho

REACTOR CAAKAHT SYSTEM<

OPERATIONAL LEAKAGE LINITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be lfmfted to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGEN

c. 1 GPH total pr fmary-to-secondary leakage through all steam gener-ators and 500 gallons per day through any one steam generator,

d. 10 GPH IDENTIFIED LEAKAGE from the Reactor Coolant S tern, and f ~sec, y

C NTROLLEO LEAKAG ('gAIgcpo~ APjnrg te cI~~ Acrsyie~cr C /k'~ ATe Pq AASA. fO O,PC C S 1 GPH age from any reactor coolant system pressure isolation valve specified in Table 3. 4-0.

APPLICABILITY: %DES 1, 2, 3'nd 4 ACTION:

agee

a. fifth any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and fn COLD SHUTDOWN within the'ollowfng 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. fifth any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be fn at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and fn COLO SHUTDOW within the followfng 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. Mfth any reactor coolant system pressur e isolation valve{s) leak-greater than the above limit, except when:

. 1. The leakage fs less than or equal to 5.0 gran, and 2, The most recent measured leakage does not exceed the previous measu'red leakage* by an amount that reduces the yl yy yy a y ly (as fran the performance of pressure indicators) ff accomplished fn accordance with approved procedures and supported by ccmputatfons sho~ing that the methad fS Capable Of demOnStrating ValVe CCeplfanCe wftn the leakaae criteria.

D.C. COOK - UNIT.Z 3/4 4-15 Order dated Aorfl 20, 1981

~0

~ ~

ol

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued margin between the most recent measured leakage and the maximum limit of 5.0 gpm by 50% or more, declare the leaking valve* inoperable and isolate the high pressure portion of the affected system from the low pressure portion by the use of at least two closed valves, one of which may be the OPERABLE check valve and the other a closed de-energized motor operated valve. Verify the isolated condition of the closed de-energized motor operated valve at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in at least HOT STAVDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. Monitoring the CONTROLLED LEAKAGE to the>reactor coolant pump seals at least once per 31 days, K

d. Performance of a Reactor Coolant System ~ater inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation, and

e. Monitoring the reactor head flange leakoff system at, least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.6.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4-0 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit prior to entering MODE 3:

a. After each refueling outage;

b. Whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if months; leakage testing has not been performed in the previous 9

• No Report required (6.9.1) unless the valve has been declared inoperable.

D. C. COOK - UNIT 2 3/4 4-16 Order dated April 20, 1981

The seal line resistance is equal to 2.31*Pd/Q, where Pd is the charging pump discharge pressure minus the RCS pressure in psi, and Q is the CONTROLLED LEAKAGE in gpm.

3 4.5 EHERGENCY CORE COOLING SYSTEMS ECCS ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:

a. The isolation valve open,

b. A contained borated water volume of betwee and 971 cubic feet.

c. A boron concentration between 2400 ppm and 2600 ppm, and

d. A nitrogen cover-pressure of between and sig.

APPLICABILITY: MODES l. 2 and 3*.

pe'CTION:

a., With one accumulator inoperable,'except as a result of,a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be kn HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With one accumulator inoperable due to the isolation valve being closed, either ienediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UIREHENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:

1. Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and

2. Verifying that each accumulator isolation valve is open.

• Pressurizer ressure above 1000 psig.

D. C. COOK - UNIT 2 3/4 5-1 Amendment No.

EHERCENCY CORE COOLINC SYSTEMS SURVEILLANCE RE UIREMENTS Continued

d. At least once per 18 months by:

1. ~

Verifying aucomac'c isolation and interlock action of the RHR system from the Reactor Coolant System Mhen the Reactor Coolant System pressure is above 600 psig.*

2. A visual inspection of the containmenc sump and vezifying chat the subsystem suction inlets are noc restricted by debris and tha" che sump components (trash racks, screens, etc.) shov no evidence of structural distress or corrosion.

e. At least once per 18 months, during shutdown, by:

1. Verifying chac each automatic valve in the floM path actuates to its correct position qn a Safety In]ection tesc signal.

2.. Verifying that each of the folloving pumps stazt automaticallv upon receipt of a safety An)ection test s'ignal: .

a) Centrifugal charging pump b) Safety injection pump c) s'dual heat r oval pump By verif 'in'.that J jg~fi4 each o the folio'~ing pumps develops the indicate ~Me~'cat'onssure on recircu'ation floe when tested pursuant t z 4.0.5:

1. Centrifugal charging pump > ' 22, tO PSid

2. Safety In)eccion pump /pe jul

3. Residual heat removil pump Qg f'5l g ~ By. verifying the correct position of eat mechanical s top for the folloMing Emergency Core Cooling Syscem throctle valves:

1. Vithin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking opezation oz maintenance on the valve shen the ECCS subsystems are required to be OPERABLE,

• The provisions of Specification 4.0.7 are applicable.

D. C. COOK UNIT 2 3/4 5-5 Amendment H . $ ~5

SUltYETLLAHCE NEOUIRBIBlTS Contf nuad Z. it least ence per Q mnths.

boron In)ectfon Safety Infectfon Thrott l e Ya1 yes Throttle Valves Valve lhaber Valve ~r 2 SI 141 LI I 2SI IXL N 2.'~$ I 141 L2 2SI QL $

2 SI 141 L3

4. 2 SI 141 L4 ly perforafng a flow balance test durfny shutdown followfng c~letfon of eodfffcatfons to the KCCS suhsystaa that altar the suOsys~ flow characterfstfcs and verffyfny the followfng f1ow rates:

boron In)ectfon Systaa Safety Infectfon System Sfn le Puen~ Sf 1 ~ 1uan~

Loop 1 boron Injectfon Loop 1 and 4 Cold Lag

'low 117.5. gapa Flow a 300 gpa Loop 2 boron Infectfon Loop 2 and 3 Cold Leg Flow 117.5 gpw F1ow i 300 gpe

~oebfned Loop 1,2,3 and 4 Cold Loop 3 boron In4ectfon Flow 117.5 ye ~ Flow Tota1 SIS

{sfngle puno) cC40 gxn.

(sfnyle numn) flowr Loop 4 boron Infectfon fncl.gdfng n{nfflow, shall not 5 exceed 700 ge.

The f1ow rate boron Ingectfon (lI) lfne should he to provfde D7.5 ape (neafnal ow fnto each loop. Voder thee tfons there fs Nero mfnf flow and b0 lated RCP seal f on lfne f1ow. The actual flow fn each bI lfne may devfate nal so long as the dffference hEOrotn the hfghest and lowest fl or less and the total f1ow to the four branch,lfnes does ceed 470 gpss. f1ow (total flow) reef red f s 345. d gapa three est conservatfve ow) lmrsjnch 1 f nes.

O. C. Ceo' NfIT 2

ZNSERT H The flow rate in each Boron Znjection (BZ) line should be adjusted to provide 117.5 gpm (nominal) flow in each loop. Under these conditions there is zero mini-flow and 80 gpm, plus or minus 5 gpm simulated RCP seal injection line flow. The actual flow rate in each BZ line may deviate from the nominal so long as:

a) the difference between the highest and lowest flow rate is 25 gpm or less.

b) the total flow rate to the four branch lines does not exceed 470 gpm.

c) the minimum flow rate through the three most conservative (lowest flow) branch lines must not be less than 300 gpm.

d) the charging pump discharge resistance (2.31*Pd/Qd"2) must not be less than 4.73E-3 ft/gpm"2 and must not be greater 9.27E-3 ft/gpm"2, (Pd is the pump discharge pressure atthanrunout; Qd is the total pump flow rate.

1

EMERGENCY CORE COCLING SYSTEMS REFUELING ltATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RMST) shall be OPERABLE with:

a. I minimum contained volume of 350,000 gallons of borated water,

b. Between 2400 and 2600 ppm of boron, and I

c. A minimum water temperature of APPLICABILITY: NODES 1, 2, 3 and 4.

ACTION:

Nth the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.5.5 The RMST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

l. Verifying the contained borated water volume in the tank, and

2. Verifying the boron concentration of the ~ater.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the EST temperature.

D. C. COOK - UNIT 2 3/4 5-11 Amendment No.

A suff1cient SHUTDON NRCIN ensures that 1) the reactor can he sade subcriiical from all operating conditions. 2) the reactivity iransients Issociated with postulated accident conditions are controllable within acceptable 11mits, and 3) the reactor sill he aa1ntained sufficiently subcr itical io preclude 1nadvertent cr1tical1iy 1n the shutdown condition.

SHUTNN NLRGIN requirements vary throughout core l1fe as a function of fuil depletion, RCS boron concentration, and RCS T . The oost restrictive condition for 1ncreased load events occurs at KOL, Nth T at no load operating temperature, and is associated with a postulate! steam line break accident and resulting uncontro11ed RCS oldown. 1n the analys1s of this accident, a minimum SHUTDOWN NRCIN o . bk/k 1s 1nitially required to control the reactivity transient and aut atic KSF 1s assumed to be available. /. Cg Technical Specification requirements call for verificat1on that the SHUTDOWN NRCIN is greater than or equal to that erhich would be required for the NODE 3 low temperature value, 350'F, prior io blocking safety 1ngection on either ihe P-II or P-l2 permissive interlocks. This assures 1n the event of an inadvertent opening of two cooldown steam dump valves that adequate shutdown reactivity is availab1e to allow the operator to identify and

'erminate the event.

@1th T c 200'F, the reactivity transients resulting from a postulated steam line SI Pak cooldown are minimal and a l% hk/k SHUTDOWN NRCIN provides adequate protection for this event.

i ln shutdown NDES and 5 shen heat ~val 4s provided hy the residual heat removal system, active reactor coolant system vol~ say he reduced.

increased SHUTDON NRCIN requirements shen operating ender these conditions 4s provided for high reactor coolant sysiew boron concentrat1ons to ensure sufficient t1ae for operator response 1n the event of a horon 41lut1on trans1ent.

The SHUTDOWN NRCIN requ1rments are based Ipon ihe 11I1t1ng conditions doser)bed above and are cons1steni Hth fSAR safety analysis assumptions.

i ainiaum flow rate of at least $ 000 CN prov1des adequate six1ng, prevents stratification and ensures that reactivity changes sill he gradual during boron concentration reductions 1n the Reactor Coolant System. A flow rate of at least 3000 CPN will c1rculate an equivalent Reactor Coolant System volume of ]2;612 cubic feet 1n approxiaately $ 0 minutes. The reactiv1ty change rate associated with boron reductions sill thi~4re he e1thin the capability for operator recognition and control.

D. C. COOK . NIT 2 I 3/1 1-] axacecue so. III ~

With the RCS average temperature above 200 0 F, a minimum of tvo separate and redundant boron fn)ection systems are provided to ensure single functional capability fn the event an assumed failure renders one of the systems inoperable. hllovable out-of-service periods ensure that minor component repafr or corrective action may be completed vithout undue risk to overall facility safety from fn]ectfon system failures during the repair period.

The limitation for a aaxiaua of ~e centrifugal charging pump to be OPERhbLE and the Survei,llance Requirement to verify all charging pumps and safety infection pumps, except the required OPERhbLE charging pump, to be inoperable belov 152 F, unless the reactor vessel head is removed, provides assurance that a aass addition pressure transient can be relieved by the operation of a single PORV.

The boration capability of either system is sufficient to provide the required SHUTDOWN MhRGIN from expected operating conditions after xenon decay and cooldovn to F. The aaxfaua expected boration capability usable volume requirement is stora e tanks storage ta . The or, gallons of 20,000 ppa borated vater from the boric acid gallons of borated vater froa the refueling vater quired RWST volume is based on an assumed boron concentration o ppa. The ainfaua RWST boron concentration required by e post- Ch long-tera cooling analysis is 2400 ppm. The ainfaum contained RWST volume s based on ECCS considerations. See Section b 3/4.5.5. The or tion urce 'ume a the ric a sto ge nse at ely en i eas to 565 gallo . Thi alue as c en t c is nt it Unit With the RCS temperature belov 200 0 F, one in)ection system is acceptable vithout single failure consideration on the basis of the stable reactivity condition of the reactor and the, additfonal restrictions prohibiting CORE hLTERhTIONS and positfve reactivity change in the event the single fn)ection system becomes inoperable.

The boron capability required belov 200 F is sufficient to provide the required 0

MODE 5 SHUTDOWN MhRGIN after xenon decay and cooldovn from 200 F to 140 F. This condition requires usable voluaes of either 4300 gallons of 20,000 ppm borated vater froa the boric acid storage tanks or 90,000 gallons of borated vater from the refueling vater storage tank. The value for the boric acid storage tank volume includes sufficient boric acid to borate to

m. Th>> required RWST voluae is based on an assumed boron concentration of ppa. The minimum RWST boron concentration required by the post-LOCh long- era cooling analysis fs 2400 ppa.

J The limits on contained vater voluae and boron concentration of the RWST also ensure a pH value of betveen 7.6 and 9.5 for the solution recfrculated vithin containaent after a LOCh. Thfs'pH band" afnfafzes the evolution of iodine and ainiaixes the effect of chloride and caustic stress corrosion on aechnical systems and components.

The OPERhbILITY of boron in)ection systea during REVELING ensures that this system is available for reactivity control vhile in MODE 6.

D. C. COOK - UNIT 2 b 3/4 1-3 JQKNDMENT NO. Sg ~ .7

IL TPf t j rc The specifica iona of Nls section pro&Ca assurance of fuel i=tegr'ty Cur'ng Condi-ion I (honaal Operaticn). aad I (Inc'Cents of moderate F=erueacy) even s by: (a) gaiataini g the calcuiated DQR in the core at or above design Curing aoz<<al operation aad in shor- tern transients ~ and (b) l&itiag the f'ssioa gas release, fuel pe'le teaperatu:e arA cla44ing nechanical properties to vithin assuned Cesign criteria. a add'ioai limiting the peak linear pover density Curing ConCition I events provides sssu ance tha the initial coaCitioas ass'~ed foz the LOCA analyses are net an" 6e ECCS acceptance criteria linit of 2200 F is not exceeded.

The def'aitiors cf certain hot channel aaC peakiag factors as used in rhese speckficat'oas are as follovs:

-"-)(2) Peat Flux Hot Chaaae Factor, is defined as the naxiav= local hes:

f"'x on the surface of a fue'oC at core eleva ion 2 Civ'ded by the average fue'od hea- flux. al'oving for caaufac uring tolerarces oa ue'el'ets ard rods, nuclear Erthalpy Rise Hot Channel Factor, is de i<<ed as the tat'o of the integral of lirear paver along che rod vith the highest ia-egrate4 pover to the average rod pove" Pfjjrp g~p 4a sC grpr~t (l4LrC)

The limits on F (I) aad F '. e 0 ~

hY.

st og o' ~ ~o s I o I ~

~ <<oe Ci C The lhasa ts on AXZhL FLUX Olm'RKHCZ assure that the T~(Z) upper bound eT&elope is not exceeded during ei,ther !NVLL1 operation or Ln the orat of

<<e redistzibution folloving povez chan es. The P (2) er bound el e is

]4'f~alii il- t <~ 8scr P/f/ITIC 4gree re g gyps f Q,gsc)

Target flux di fereace is detained at equ'librba <<enon conCitioas.

The full length rods tLay be pos'tioned vithin the core in accordarce vith their respective insertion limits and should be inserted near their noel position for steady state opezation at high pover levels. The value of the D. C. COOK NiT 2 j S/4 2-1 amema SO

POXR 01STRlRPlCa L,lyly iarett flux 41ffer ence obta1ned under tuse cond1 t1ons 41v14>>d by th>> fract1cn of AJLTGI THERMA. PteKR 1s the target flux 4)fftrence a. MATEO T%~<A. PChKR for the asscc1attd core hurnup cond1t1ons. Target flux 41ffarencas for other TNtlPQL MH levels are obta1ned by mlt101ylng the VTB TAKR!i'lX% va'ue by the apprOOrfate fraCt1Onal TNK~ POX% leVtl. The'er1%1 ~dat1ng Cf the target flux 41fftrence value 1s necessary to reflect core h~up cons1 4er at1ons.

Although 1t 1s 1ntaedtd Clat the plant ~111 he ~rat& ~1th the iXiA.

R.UX OlPEREltCK e1th<n the tar-tt band about tPe target flu 41fft. ance) eur(ne raeie e1ane TMRual. Joltà recce.1ona ~ cenerol ree aeaUen uttl causa the AFO tO dtV1a e Outs<C>> Of the tar t band at reduCtd THKia'QL K'aeD levt>s. Th1s dtv1at1cn ~$ 11 not affect the xenon rt41str1hut1on suff tc1trtly tO Change tht tnvelcpt of hei)L'$ng faCtorS rh1ch say bej reached cn a Substqutnt return to U EQ THERMA. 1CSKR (s1th t te ASD e1th1n the tarp>>t band) prov<did the t1w durat1on of the dtv1at1on 1s 11@1 M. Accordfrqly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ptnalty dtv1at1on 11a)t cualat1ve dur1nq the prtv1ous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

)s provided for operation outs1de of the tar;et hand hut HQ1n the 11a1t!

wh1le at TNUP%L PCiCR ltvtlS abave SCK Of UTKO THUJA. )

PlMN. Fcr THK KAl. KMD leve1s baton 155 and $ 75 of lATQ TNGJQL, ICE'U, dev1at1ons of the AH outs14o of the tar;tt hand ars less s1gn<f1cart. The penalty of 2 hcurs actual tiara reflects th1s reduc>>4 s<gn1f1cahca.

Provisions for ecnftorfnq the AFO cn an autceat1c bas1s art dtr<vtd I'res .ht p1ant precasS C-gute'!rovgh the AFO Igniter Ala<. The c= ta, ctt>>ravines the cnt a<nuta average of each of t!A CPKQLE axcore cotectct outputs and prov1dts an alarm eessagt 1f the AFO fe' least 2 of 4 cr 2 of 3 (}PKVSLE exccrt chanrals are outs1dt the tar;et band and t.'e THER~

PCliKR 1S greater than KX Or 1.$ X APt. Of LLTEO inKK~ PAUL (Ih1Chevtr 1s less), Our1ay ootrat1on at THERNL MU levtls betwtn N5 ard K5 or 4.9 x Atg of klTKO T<R~ fCSXR (rh)chevtr 1s le s) and helen l55 an4 50$ MTKO WUPQL IO'KR. the 'Ccaeute outou S an alen wssage rhtn i

the penalty dev1at1on ac +elates beyon4 the l<s1ts of hour and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, resptct<vtly.

e~-

T?w bas1s and te Nodcl y fcr es.ah11sh1 NeA l1a1ts fs presant .

1n topical report foe D. C. CCCX NIT 2 I 3/4 2 ? ~~xT NC

~ oo

~ ~ ~ ~

o ~

ot @~egg

SC

('4 i4~ spa. c v' p'a grater t

~ Q Qt f ~ pt&

The l~ts on hea f1m ho. camel factor and nuclear enthalpy r'se ho channel factor ensu. ~ that 1) the design 1wits on peak local yover density and minimum DÃhR are not exceeCed atd 2) )n the even>> of ~ LOCA the peak temperature vill not excaeC the 2200 f ECCS accepsarxe cr'teria lim't. ue'lad Each cf chose is measurable bus vill normally only be dete=ineC periodically as specified in Specifications 4.2.2.1, 4.2.2.2, 4.2.3, 4.2.6.i.

and 4.2.6.2. This yerioCic su.-~eillance is sufficient so erasure that the l.'+its are maintained proviCed:

a. Consro rods ir. a single grou cove toSether vith no 'nC" .dual roC

'nsertion dif ctiny by more than + 12 steps rom the group dew-..d pos'tion.

b. Consrol rod poups are sequenced v h ove spying $ o'ps as described in Specif ication 3.'.3.6.

c. The cortrol rod insertion limits of Specificatiora 3.1.3.5 anC 3.1.3.6 are ma'ntained.

d. The axial yover d'stribution, expressed in terms of AXLE F X DIP7GLEHCE, is maintained v't".in she 1im' e'en gs ops,r. ~ s< 7'l Qi>>df>> p cs + p"r~ G' vill be maintains ovided con ~ Qa

d. abo4~e are maintained. The POUR allovs changes in the raC'al yover shay&or relaxacion of i, as a function of TP:-KVJ.

all permissible rod insertion 1'=its. The for= of this relamsicn for LSSR lairs 's discussed ir. Sect'or. 2.1.1 of the basis.

%%en an F measuremers is taken, both exper'ensal erro and manu assuring tolerance muss be alloved for. 5% is the appropriate allovance for a full core map taken vith the tncore de secor flux mapping system and 3%

is the appropriate allovance for aanufacsuri tolerance.

'C A ur en or a11 ~ 4% on in ~

T a Sye a n so ur a e 1

D. C. COOK ~ CHIT 2 5 3/4 2-4

4~

ef ewr t$ gnre b+ned

~

4 a V4

&ed f1'l r CLlmtra igure 3.2 or F the e

t in 3.2 leeen

~

tio ting troa t f ue liiita a h otic for ich ~st b,'

Xiiiti i, a liaitigg obtained cause:

Vest ouse tuel n Iuc r Coapany Fue t Iel.>b Z flo4 2(1 0 ))y t 4 oa9 Z b Z f le o2(1 oO~P) j tOMER lthere:

b /t 2-2 Lsplays liaiti DNM F" aves fear Rorno clear Ccapan fuel f flovs o 36.77 Z ~a, act .63 Z 10 gpI Also'isplaye on F e b /a 2-2 is e lLIL oni F 4 'g ~L results fr LOCA fo exxon clear Co any fue 'ms aaintained lov t to t left f both ~ appli le DNM 4H 4 F

L(nest d

~

the LOCA F g

limit ov 0

and an 1 nit.

flIt or %asti house fu the~e Ls nly f." the ap Lcab'e r at'onship among b, F',

one K..

F,- and be obtaLn urement taken, bot experiaent error and ce c mus e al ove for. 5S is e the Ln re detector ~ux aapp'Lng system and 3$

appropt" e allovanc for ufactur'oler a full core aap @ken y h is the appropriate allovance for aanuf ur ~ o'e

%hen AS flov race iY and Fd> are measured, no additional allovances are necessary prior'to coop Lson sith the limits of Specification 3.2.s.

Measurement errors o for IC$ flov total flov race and a$ for f<~

been alloved for Ln eternmation of the design DNSR ralue and Ln the

'ave determination cf the LCCAJ CS limit.

,/cp Margin between the safety analysis limit DNBRs (1.69 and 1.61 for the Vantage 5 typical and thimble cells, respectively and 1.43 and 1.40 for the ANF fuel typical and thimble cells) and the design limit DNBRs (1.23 and 1.22 for the Vantage 5 typical and thimble cells, and 1.39 and 1.36 for the ANF fuel typical and thimble cells ) is maintained. A fraction of this margin is

'tilized to accommodate applicable transition core penalties and the appropriate fuel rod bow DNBR penalty for the Vantage 5 fuel (equal to 1.34 per MCAP-8691, Rev. 1). The remainder of the margin between design and safety analysis DNBR limits can be used for plant design flexibility.

a 0/4 J~ ~mneme Io 64

%~

~ ~

~ ~

0 ~

~ ~

~ ~

4 HP 1.55 'L ~ ~

~ ~:Le 0~ ~

1. 55 Hl 0 ~

~o

~ I

~ f 0

~ ~0 OO f

~ 0 1.45

~ ~ ~1

~ t 0%'0 0

' ~

0

.- ~

~0 as fs m je 1.35 O 20 40 dO SO tERCENT OF %TED THUNAl. KSN fISlJRE I 3/4 2 2 ILLUSTMTIYE EQW'LE OF F~ LIXIT YEaSuS a TANT tNER FN EXXON FuEt.

l4 The quadrant pover tilt ratio ltmfs assuresSn c?'As Ctssr'buston satisfies the destgn va'ues used the radfal pcvez ehe pover capability arslysts. RaCial povez distribution measu enenss.ar ~ maCe Curing staztup testing arA periodically during povez opezatton.

The Bait of 1.02 at vhtch corrective action ts required provtCes DQ and linear hea- pneraston rate protection vfeh x-y plane povez eflux.

The evo hou ttne -a'lovance for operation vtth a silt cordttton grea er t!un 1.02 but less than 1.09 ts provtCed to allov ident'ession and correction of a dropped or mtsaltgaed rod. In the evens such action does no-correce the tile, the margin for uncerea'nty on F ts refns aced by reCucfng the pover by 3 percent ron MrD THRM. FOUR fSr each percere o tls in excess of 1.0.'il xM . o Scg. qAyu~t- ~

he O'$3 rela ed parameters fn .fOD- 1 assu e that each of she parens=ere are main... ishin the normal s- . se envelope of operation assv..eC Sn she transient . s aralvses . The ltnies aze corsissens vtsh she -'ass s - e safety ana een analyefcally decor f'o c ude allovances for inst T,

aCequase eo mainsafn design DNR throughout analyze" -- . er.=. Ti.h.e 'nCScased':alues of LV pressu"Seer pressu"e, and The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perfodfc su~efllance of these parameters th"ough tnserumers readout Ss sufftc'ent eo ensure that the parameters are restored vithin their lfmtss follovtng load changes and other expected tzansien operation. The 12-hou" suveillance of the RCS flov measurement ts aCeruate to Ce <<cs flov degradation. The CHP'lNKL CALZ3RATION.performed after refueling ensu=es the accuracy o. the shifely flov measuremene. The total flov fs aeasureC after

~ ach refueBng based on a seconCary side calorfneeric and measuremenss of primary loop te=~erames.

The limits on pressurizer pressure and T tn ltODKS 2 and 3 provide protection against DN3 resulting from an uncoFBolled zod vishCraval from a subcrtttcal condition. The tnCtcated values of Tavj and pressurizer pressure include allovances for tnserumens errors.

S<a InSerk

~ pover tr t on contro procedu=e, FDC-II. aanages core pover dtstrtb such that Technical Specification ltmtts on F (X) -e not violated dur rmal operation and ltzLtts on CShR a=s o aced du=tng steady-state, load- v. and arttcipatad canst . e V(X) factor given tn the Technical- Specifics provides ans for predicting the aaxtmu=

F '(2) dtssrtbuston aneictpased opezatton under ehe FDC-II procedure thing into account the e aeasu"e tbrtum pover distribution. h comparison oi th F (2) vtth the Tec ectficaston lfmfs deter@i e pover level )APL) belov vhich the Te $ pectftcatton 1 can be protected by FDC-II. This comparison ts done 'ulastnZ AK, as defined tn spectftcatton $ .2.6.

D. C. COOK UhiT 2 5 $ /4 2-5 ggÃDKNi 10.

INSERT E to page B 3/4 2-5 The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The Tavg

< 578.7 Degrees F and Pressurizer Pressure > 2194 psig (indicated value) are consistent with the UFSAR assumptions and have been analytically demonstrated adequate to maintain the core at or above the design DNBR throughout each analyzed transient with allowance for measurement uncertainty. The Tavg > 542.8 Degrees F is consistent with a safety analysis performed to demonstrate that the plant may operate on a linear control program where the analytical limit of Tavg at 1004 RATED THERMAL POWER may range from 541.4 Degrees F to 580.1 Degrees F. The core may be operated with indicated vessel average temperature at any value between the u per and lower limits.

orth xn The limits are consistent with the UFSAR assumptions and have been analytically demonstrated to be adequate to maintain the core at or above the applicable design limit DNBR values for each fuel type (which are listed in the bases for Section 2.1.1) throughout each analyzed transient.

Constant Axial Offset Control (CAOC) operation manages core power distributions such that Technical Specification limits on Fn(Z) are not violated, during normal operation and limits on MDNBR are not violated during steady-state, load-follow, and anticipated transients. The V(Z) factor given in the Peaking Factor Limit Report and applied by the Technical Specifications provides the means for predicting the maximum

" FO(Z) distribution anticipated during operation using CAOC taking into account the incore measured equilibrium power distribution. A comparison

=

of the maximum Fn(Z) with the Technical. Specification limit determines the power, level /APL) below which the Technical Specification limit can be protected by CAOC. This comparison is done by calculating APL, as defined in specification 3.2.6.

The OPERhSILITY of the RVST as part of the ECCS ensures that a sufficient supply of borated vater is available for in]ection by the ECCS in the event of a LOCA+ The limits of RVST sinimmum volume and boron concentration ensure that 1) sufficient vater is available vithin containment to permit recirculation cooling flov to the core, and 2) the reactor vill remain subcritical in the cold condition folloving mixing of the RVST and the RCS vater volumes vith all control rods inserted except for the aost reactive control assembly. These assumptions ar ~ consistent vith the LOCA analyses.

The contained vater volume limit includes an allovance for vater not usable because of tank discharge line location or other physical characteristics.

The limits on contained vater volume and boron concentration of the RVST also ensure a pH value of betveen 7.6 and 9.5 for the solution recirculated vithin containment after a LOCA. This pH band ainhaizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The ECCS analyses to determine F limits in Specifications 3.2.2 and 3.2.6 '

assumed a RVST vater temperature of This temperature value oi the R4S vater determines that of the spray va r initially delivered to the containment follovt.ng LOCh. It is on of the factors vhich determines the containment back-pressure in the ECCS analyses, performed in accordance vith the provisions of 10 CFR 50.46 and h endix K to 10 CFR SO.

(- D. C. COOK - UHIT 2 3 3/4 5-3 ANENDMENT NO. 7

114 7 1 8 TQ ( PP't?'4 7 ejs+v gg t c The OPMILIiZ of the main steam line coCe safety valves ensures that the secondary system pressure vill be 1'ced to vi Sin 110l of its design pressu"e of lOSS psig during the most severe anticipated system The maximum relic ring capacity is assoc'aced vith a turbine tr'p operatiora'ra~iert.

from 100% RA.ED THERM. POVER coinc'Cent vith an assumed loss o conCenser heat s'nk (i.e., no steam biyass to che condenser).'he spec'ed va've li - se-'ngs arC reliev'rg capacities are in accordanc>>

Pressur vith the requirements of Section Iii of the hS.""- 3oiler anC

CoCe, 1971 EC'-'o cape oz a

]r" a c eco . ry steam

~ re of 17,, 001s/hr .rvhoh 1',000 at i. RATE

~17pe-re% .04 miiieaval oa a es er steLl generator ensures t at ~

suf c e re ev ng capacity is atailao e+ for the allovable TALMJQe

~ ~

PC+ye.

estrict'on ir. able 3.7-1.

STARTUP and/or ?04M 0?%ATION is al'ovable v'th safety valves inoperable vithin the limitations of the ACTION rec irements on the basis o the reCuction in secondary system steam flov and T~M, POLER.requ':ed by the reduced reactor tr'p sett'ngs uf the Pover Range !neutron Flux charnels.

Tne reactor r'p setpo'nt reC ctions are derived on the folloving bases:

For 4 loop operat'on v (v) ( )

~~

%here:

SP reduced reactor trip aecpoint in percent of RhTQ ~QL P04 V max~ number of inoperable aa ety valves per steam lire X~ total relieving capacity of a3.l.safe~ valves pe steam 1're in lbs./hou"s ~ 4,2SS,450 Y ~ maxhcch relieving capaci JJ 0 L.) one sa etv,valve in lbs./hour ~ b57,690 109 Pover Range Reutror. Flux-High Tr'p Setpo't fo" 4 loop operation

The total rated relieving capacity of all valves on all of the steam lines is 17,153,800 lbs/hr which is at least 10SS of the maximum secondary steam flow rate at 100% RATED THERMAL POWER.

The limitations on reactivity conditions during kEFQELINC ensure that:

1) the reactor vill remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the Water volume having direct access to the reactor vissel. These limitations are consistent vith the initial conditions assumed for the boron dilution incident in the accident analyses. The value of 0.9S or less for K includes a 1 percent delta k/k conservative a ovance for uncertainfHs.

Similarly, th>> boron concentration value of ppm or greater includes a conservative .uncertainty allovance of 50 0 oron entre requ ment o cat o . . .b een co ativ ncreas o ppm t ree vi ~ min oncentra of @ST.

The OPERASILITY of the source range neutron flux monitors ensures that redundant monitoring capability ts available to detect changes in the reactivity condition of the core.

The minimum requirement for reactor subcriticality prior to aovemen of irradiated fuel assemblies tn the reactor pressure vessel ensures that sufficient time has elapsed to allow th>> radioactive decay of the short lived fission products. This decay time is consistent vith the assumptions used in the accident analyses.

~

~

The requirements on containment building penetration closure and OPERABILITY ensure that a 'release of radioactive material vithin containment vill be restricced from leakage to the environment. The OPERASILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of contai aent pressurization, otential awhile in the REFUELINC NODE.

The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the.

facility status or core reactivity conditions during CORE ALTERATIONS.

5 3/4 9-1 AMENDMENT NO.

AERIVISThP. VE CCN.ROLS The radioactive ef luent release xeport to he auhmi>>ted 60 days af>>er anuazy 1 of each yeax shall also include an asaeaat.'.et'.t ot radiation doses>>o the lfkely moat expcsed member of the public frcn reactor releases arC e Her nearby uranium fuel cycle sources (including doses trom primar: etfluert pathways anC direct raCiation) for .he previous 12 consecutive mon>>hs to ahov confcrmat.ce vith 40 CFR 190, Envfroteental Radiatior. Prctee.'or.

StanCazda for Nuclear Pover Opezat'on. Acceptable methods for caleulatirg the dose contribution from liouid anC gaseous effluenta are given fn Pegulatory Quide 1.109 Pev. 1 The radfcactive efflust t release report shel) include the follovfng ir.forms 'cn oz each type of solid vaste shipped offaite Curing the report period:

a. Volume (cubic mete s),

h. Total curie quantity (speci y whether determined by measurement or es>>fmate),

c. Prine'pal radionuelides (spec'fy whether determined hy measure.-ent or estimate) g

d. Type of waste .(e.g., apex>> resin, ecmpacted dzy vaste, evaporator bot.orna),

e. Type cf contairer te.g., LSA, Type A, Ty.= . 8, I.arge guartxty),

ar.d

f. Sol'difcatior. agee. !e.g.. cement).

.he zadioae>>ive effluent release repoz>> shall ircluCe unplanned re'eases

='zrr the si>>e .o unrestric>>ed areas of raCioact'v~ -.aterials in gaseous anC 1'quid effluent on a quarterly basis.

.he radfoac>>fve effluent release reports shall include any change to the PROCESS CCN.ROL PPCQRAM (PCP) and the CFFS=T DCSE C"CLM.ICN i'd%.'A (CDQ:) made du ing the reporting period.

YCt~LY REA~R OPERATINC REPORT 6.9.1.10 Routine reports of operating stat'atfca and shutdown experience shall be submitted cn a monthly basis to>>Pe Director, Cf ice Of Ãanaget.-.ent ard Pxogram At.alyafs, Q.S. Nuclear Regulatory Ccmmission, Vaahfng orb.C.

2(555, vith a copy to the Reg'eral Ctffce no later>>har. the 15th of each ttonth follcvirg>>J:e cale>>daz north covered bv the repez .

~C)~(- yttt~~~Wt M C L-iW ( 7$ C~@MT

<. L I-(1 Q~c Atkv42 CA ~ 'i I'(

~ I

b. C. COOK UN~ 2 6 18 Amendmert No. 8

6.9.~

(. LL Core operating lfmfts shall be established and documented fn the CORE OPERATING LIlfITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1. Noderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance lfaft for Specification 3/4.1.1& 't

2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

3. Control Bank Insertion Limits for Specification 3/4.1-3.63 awe

4. Axial Flux Difference llnltsi target hand, nndJSP for Specification 3/4.2.1,

5. Heat Flux Hot Channel Factor, K(Z), er,

~>,

gg cf goal, J5% and ggihgt for SPeclflcatlon 3/4.2.2e cf j

6. Nuclear Enthalpy Rfse Hot Channel Factory Power Factor Nultfplfer r Specification 3/4.2.3, cad.

Q~~b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC fn:

1. WCAP-9272-P-A, 'WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July 1985 (M Proprietary).

(Hethodology for Specifications 3.l.l.P'~ moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Fact 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor and RR4 n.)

2a. QCAP-8385, 'PNER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES - TOPICAL REPORT', September 1974 (M Proprietary).

(Methodology for Specification 3.2.1 - Axial Flux Dfffcrcnce (Constant Axial Offset Control).)

2b. T. N. Anderson to K. Knfel (Chief of Core Performance Branch, NRC) January 31, 1980 --

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package.

(methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)

(continued) 2c. KUREG-0800, Standard Revier Plan, lJ. S. Kuclear Regulatory Coeeission, Section 4.3, Kuclear Design, July 1981. Sranch Technical Position CPB l.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev.2, July 198l.

Nethodology for Specification 3.2.1 - Axial Flux ifference (Constant Axial Offset Control).)

~

3. 1021 A, I 'OF CON I '

L OFFS RVEI CE TEC I IFICA 19 M Pr ieiar (Ne for e ic ions P. - Ax lu Diff ence la i 1 fse o rol e2 e xHo a ctor )

illa ireaents or Fq Neth gy .)

a. MCA -P-A, . I, ESTINGHOU CCS EVALUATIO L-l ON", ruary l et ology for p ation 3.2. - H nel Factor.)

4b MCAP 9561 -A) ADD 3s v I e BART A-I: A COHPUTER CODE FOR S ES IRATE IS OF REF SIENTS-SP EPORT: E NO ING M E ATION NO L,"

ly, (g:P iet r (N tho gy fo if .2 - t Flux Ho a actor.)

Ic. itCAA-10266-P-A Rev.2, 'THE 1981 VERSION OF WESTINGHOUSE EVALUATION NOQEL USING SASH COOE'I March 1987, (M Proprietary).

(Nethodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

The core operating liaits shall be deterained so that all applicable liaits (e.g., fuel theraal-aechanical liaits, core theraal-hydraulic liaits, ECCS liaits, nuclear liaits such as shutdown aargin, and transient and accident analysis liaits) of the safety analysis are aet.

The CORE OPERATING LINITS REPORT, including any aid-cycle revisions or supplenents thereto, shall be provided upon issuance, for each reload cycle, to the NRC Docenent Control Desk with copies to the Regional Administrator and Resident Inspector.

ATTACHMENT 1 TO AEP:NRC:1071E REASONS AND 10 CFR 50.92 ANALYSES FOR PROPOSED CHANGES TO THE DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 TECHNICAL SPECIFICATIONS to AEP:NRC:1071E Page 1 Or anization of the Submittal The primary purpose of this submittal is to obtain approval to operate Unit 2 for Cycles 8 and 9 with a mixed core of Vantage 5 and ANF fuel. To facilitate the reviewer's task, we have organized the submission in several attachments listed below.

One of these, Attachment 3, "Donald C. Cook Nuclear Plant Units 1 and 2, Summary of Proposed Technical Specifications Changes," identifies each proposed change by page and T/S number, briefly describes the change an'd the reason for the proposed change, and directs the reviewer to the appropriate supporting analysis or evaluation documentation and to the appropriate section of the 10 CFR 50.92 significant hazards consideration analysis.

The attachments are as follows:

This attachment, Attachment 1, contains a description of the organization of the submittal and our analysis of significant hazards considerations for the proposed T/S changes.

Attachment 2 contains the proposed T/S changes.

Attachment 3 contains a summary of the Unit 1 and Unit 2 proposed T/S changes with' description and the reasons for each proposed change.

Attachment 4 contains evaluations, T/S mark-ups, non-LOCA accident analyses, and LOCA analyses performed by Westinghouse in support of the Unit 2 Cycle 8 reload.

Attachment 5 contains the results of the calculation of mass and energy releases inside containment. This discussion has been taken from WCAP 11902, Supplement 1. Portions of this document which are needed to support operation of Unit 2 in Cycle 8 are included in this attachment.

Attachment 6 contains the justification for operation with pressurizer level increased from 67% to 92%. The material included in this attachment is part of a previous submittal to the NRC, WCAP 11902, "Pressurizer Operability Level Justification." WCAP 11902 was submitted on October 14, 1988, in AEP:NRC:1067. It is included with this submittal to facilitate the reviewer's task.

Attachment 7 is a copy of a,sample COLR with expected Unit 2 Cycle 8 values.

to AEP:NRC:1071E Page 2 Attachment 8 contains our analysis of significant hazards considerations for not analyzing seven events not in the Donald C. Cook Nuclear Plant Uni.t 2 licensing basis.

Attachment 9 is a copy of a letter from Joseph G. Giitter, NRC staff, of August 3, 1989, to Milton P. Alexich, AEPSC. This letter states that the seven events identified in Attachment 8 are not part of the Unit 2 licensing basis. The letter also requires a significant hazards consideration analysis of these seven events.

Attachment 10 is a copy of Part B of WCAP 10217-A. This WCAP is included as justification for a proposed T/S change in the power distribution limits section 'of the T/Ss. Westinghouse power distribution methodologies will be used in support of Unit 2 Cycle 8 in lieu of the ANF methodologies used in Cycles 4 through 7.

Attachment 11 is a copy of Section 1 2, "Major Analytical

~

Assumptions," from WCAP 11908, "Containment Integrity Analysis for Donald C. Cook Nuclear Plant Units 1 and 2." WCAP 11908 was submitted on August 22, 1988 in AEP:NRC:1024D.

Documentation of this analysis is provided to support our proposed residual heat removal pump surveillance requirement based on 10% degradation. It is included with this submittal to facilitate the reviewer's task.

II. Pur ose of Pro osed T S Chan es I

The proposed T/S changes included in this submittal are intended to accomplish six purposes. In the following discussion, the proposed T/S changes are grouped according to these purposes. These groups are:

Group 1) Make changes that result from the analyses performed by our contractor, Westinghouse, to support operation of the Donald C. Cook Nuclear Plant, Unit 2 in Cycles 8 and 9 with a mixed core of V5 fuel and ANF fuel.

Group 2) Remove certain T/S requirements that are part of the Unit 2, Cycle 6, Amendment No. 82. These T/Ss all address concerns in transition modes of operation, Modes 3 and 4. They are not standard in the sense of the Standard Technical Specifications (STS). We believe those concerns can be safely addressed administratively as they are at other nuclear A facilities.

to AEP:NRC:1071E Page 3 Group 3) Make the Unit 2 T/Ss more nearly like the Unit 1 T/Ss. There is one change in this category. It is a proposal to increase the pressurizer water volume limit. This change was submitted for Unit 1 on October 14, 1988, with our submittal AEP:NRC:1067.

It was approved in Amendment No. 126 to the Unit 1 operating license. For the convenience of the reviewer, the supporting evaluation is resubmitted in Attachment 6.

Group 4) Make administrative changes. Where substantive changes are proposed on a T/S page, some changes to enhance the clarity of the T/S are also proposed. In addition, some administrative changes result naturally from reformatting, page removals, and text movement.

Administrative changes are proposed for both units.

Group 5) Modify a surveillance requirement for axial flux difference (AFD).

Group 6) Make a number of Unit 1 changes where the justifications are identical to or essentially identical to those for proposed Unit 2 changes.

These proposed changes help to make the T/Ss for the two units more nearly alike.

Three of the Group 6 proposals correspond to proposed Unit 2 changes in Group (2) above, one to Group (5),

and two to Group (1). The Group (1) changes are a proposal to change the discharge pressure requirement for the residual heat removal and safety injection pumps. Reference to the supporting analysis is included in Attachment 5.

III. Overview of the Pro osed T S Chan es and 10 CFR 50.92 Evaluations A summary of the proposed T/S and Bases changes has been included as Attachment 3 to this letter. Attachment 3 contains the following information for each proposed T/S change:

a) Reference to T/S page and section.

') Reference to the Significant Hazards Consideration Analysis group.

c) Sequential identifier number.

d) Brief description.

e) Remarks which provide the reason for the change and a reference to safety analyses and evaluations as appropriate.

0 to AEP:NRC:1071E Page 4 This attachment (Attachment 1) includes an overview of the proposed T/S changes and our,10 CFR 50.92 analysis for no significant hazards consideration.

We have grouped the proposed T/S changes in this discussion according to the six purposes described above.

1) T S Chan e Grou 1 Changes based on analyses performed to support Unit 2 Cycle 8 operation.

The changes in this category are based on Westinghouse analyses and evaluations performed using NRC approved methodologies.

They are as follows:

a) Reactor Core Safety Limit Curves and OTdelta-T/OPdelta-T Reactor Trip Setpoints, T/S Figure 2.1-1 and T/S Table 2 '-1, Functional Units 7 and 8.

These changes are numbered 001, 007, 008, 011, 012, 013, 014, 015, 016, 019, 020, 021, 022, and 023 in Attachment 3. A new thermal design was performed for Cycle 8 operation. The results are described in Appendix B of Attachment 4, Section B.2.F 1. The proposed new safety limit curves are calculated for 3588 Mwt core power. They are conservatively applied to 3411 Mwt operation. The revised OTdelta-T/OPdelta-T setpoints which protect this thermal design are also discussed in Section B.2.2.1.

The revised setpoints are provided in Attachment 4, Appendix A.

b) Design Flow, Footnote *, T/S Table 2.2-1 This change is numbered 002 in Attachment 3. Design flow is minimum measured flow (MMF) divided by four.

MMF - 366,400 gpm. Design flow - 91,600. Analysis of DNB events with MMF using the Revised Thermal Design Procedure (RTDP) is discussed in Attachment 4, Appendix B, Section B.2.3. Table B.2-4 shows a loss of forced reactor coolant flow event to be a RTDP event. Use of the low flow trip is discussed in Attachment 4, Appendix B, Section B'.3.5.1. The analysis value of the low flow trip is given in Attachment 4, Appendix B, Table B.2-2 ~ Westinghouse has provided the design flow in Attachment 4, Appendix A.

to AEP:NRC:1071E Page 5 c) Shutdown Margin (SDM), T/S 3/4.1.1.1 and 3/4.1.1.2 This change is numbered 026 in Attachment 3. The analyses performed for Cycle 8 assumed a SDM of 1.3%

as discussed in Appendix B of Attachment 4, Section B.3.11.2. This value was also assumed for the mass and energy releases inside containment as discussed on Page S-3.3-12 of Attachment 5. The analysis of record for the mass and energy releases outside containment is discussed in Attachment 4, Section 5.4.2. This analysis assumed a SDM of 1.6%.

Based on this limiting analysis, we propose to reduce the SDM for Unit 2 from the current 2.0% to 1.6%.

The SDM requirement of 2.0% resulted from an ANF analysis which will be superceded by the Westinghouse analysis in Cycle 8.

d) Borated Water Sources, T/S 3/4.1.2.8 This change is numbered 034 in Attachment 3. The new bbric acid storage tank required volume calculated for Unit 2 is a bounding value which is expected to accommodate uprating to core thermal power of 3588 Mwt, fuel of increased enrichment for increased cycle length, and changes in vendor methodology. This change is indicated in Attachment 4, Table 6.1 and Attachment 4, Appendix A.

e) Rod Drop Time, T/S 3/4.1.3.4 This change is numbered 036 in Attachment 3. The intermediate flow mixer grid feature of the V5 fuel design increases the core pressure drop. Therefore, the control rod scram time to the dashpot has been increased to 2.7 seconds. This is discussed in Section 5.1.2 of Attachment 4 and in Section B.2.4 of Appendix B of Attachment 4.

Minimum Measured Flow, T/S 3/4.2.5 This change is numbered 043 in Attachment 3. A small change was made in the analysis assumption for primary flow. The analysis value for events not analyzed using the Revised Thermal Design Procedure (RTDP) is 354,000 gpm. The minimum measured flow (MMF) is 3.5% larger than this value conservatively allowing for measurement errors. The MMF is also the value used for flow in the DNB events analyzed with RTDP. This is discussed in Appendix B of Attachment 4, Section B.2.3.

to AEP:NRC:1071E Page 6 g) Tavg Window, T/S 3/4.2 '

This change is numbered 041 in Attachment 3. Unit 2 has been analyzed for operation over a range of primary temperatures'his is reviewed in general in Appendix B of Attachment 4, Section B.2.1. The limiting temperature assumptions are addressed throughout the discussions of the transients and accidents when limiting assumptions are discussed.

The values in T/S 3.2.5 are obtained from the analysis values shown in Table B.2-1 and Section B.2.1 of Attachment 4, Appendix B as follows:

Lower Limit High Limit OF F Analysis Value 547 576.0 Controller Allowance -5.6 +4.1 Readability Allowance +1.4 -1 4~

T/S Limit 542.8 578.7 The analysis values are obtained from cases 2 and 3 of Table B.2-1 which apply to Cycles 8 and 9. The T/S limits are 'provided in Attachment 4, Appendix'.

Our proposed T/S upper temperature limit is 578.7 F' and our proposed T/S lower temperature limit is set conservatively at 543.9 F. The lower limit is the value we plan to submit for Unit 1 in the future and was selected to make the T/Ss of the two units more consistent.

h) Minimum Pressurizer Pressure for Operation This change is numbered 042 in Attachment 3. A new pressure allowance for control and readability was established for Unit 2. See Attachment 4,

~

Appendix B, Section B.2.3, and Appendix C, Table C.3.1-1. SBLOCA used the same pressure assumptions shown for LBLOCA. The total allowance"used in the analysis was:

Controller Allowance 35 psi Readability Allowance -22 psi Margin ~6s i Total Allowance 63 psi to AEP

NRC:1071E Page 7 The T/S pressure limit is obtained analogously to obtaining the temperature limits above.

Analysis Value 2235 psig Total Allowance -63 psi Readability Allowance ~+22" ai T/S Limit 2194 psig This value is supported in Attachment 4, Table 6' and Appendix A.

The proposed T/S lower pressure limit is set conservatively at 2200 psig. This is the value we plan to submit for Unit 1 in the future and was selected to make the T/Ss of the two units more consistent.

i) F P

Penalty, T/S 3/4.2.6 This change is numbered 049 in Attachment 3 The ~

allowable power level (APL) limiting condition for operation will be supported in Cycle 8 by Part B of WCAP 10217-A. This results in a change in the penalty for increasing peaking factgrs. The current specification requires monitoring F~ for increases in successive steady state flux maps. This submittal proposes to monitor instead the maximum over Z of (F (Z)/K(Z)}. This change is based on p. B-4 of WCAP 10917-A. Part B of WCAP 10217-A is included in this submittal as Attachment 10 to facilitate the reviewer's task.

j) Pressurizer Level Reactor Trip Response Time, T/S Table 3.3-2 This change is numbered 050 in Attachment 3. This protection function is needed to prevent pressurizer fill for certain cases of uncontrolled control rod assembly bank withdrawal at power. This is discussed in Sections B.3,2A.2, B.3.2A.3, and B.3.2A.4 of, Attachment 4, Appendix B and is supported by Table 6.1 and Appendix A of Attachment 4.

k) Changes to Protection Response Times.

These changes are numbered 051 and 054 in Attachment

3. We propose to relax six protection response times listed in T/S Table 3.3-2 from present values to values used in the most recent analyses. The proposed changes are:

to AEP:NRC:1071E Page 8 Present Proposed Functional Unit Res onse Time Res onse Time Pressurizer Pressure-Low 1.0 sec 2.0 sec Pressurizer Pressure-High 1.0 sec 2.0 sec Loss of Flow-Single Loop 0.6 sec 1.0 sec Loss of Flow-Two Loops 0.6 sec 1.0 sec Steam Generator Water Level Low-Low 1.5 sec 2.0 sec Undervoltage - Reactor Coolant Pumps 1.2 sec 1.5 sec The new values are listed in Table B.2-2 and C.3.1-3 of Appendix B of Attachment 4 and are supported by Table 6.1 and Appendix A of Attachment 4.

High Steam Flow Setpoints, T/S Table 3.3-4, Functional Unit 4,d This change is numbered 066 in Attachment 3. New values for high steam flow setpoints are proposed.

This functional unit was not assumed in any safety analyses for Cook Nuclear Plant Unit 2. The setpoints are based on the mass and energy release inside containment which conservatively assumed safeguards actuation on low steam line pressure coincident with high steam flow in order to bound Cook Nuclear Plant Unit 1. This analysis is described in Attachment 5, particularly page S-3.3-11. The setpoints,are provided in Attachment 4, Table 6.1 and Appendix A.

Response Times for Containment Pressure-High Safeguards Actuation, T/S Table 3.3-5, Initiating Signal 2 These changes are numbered 069, 070, and 071 in Attachment 3. The addition of response times for Engineered Safety Features Actuation on containment pressure high are proposed based on the mass and energy releases inside containment. This analysis is discussed in WCAP 11902, Supplement 1. The description is included in this submission as Attachment 5. The safeguards employed for these transients are discussed on page S-3 '-12 of that attachment. The response times are provided in Attachment 4, Table 6.1 and Appendix A.

to AEP:NRC:1071E Page 9 n) Response Times for Steam Generator Water Level High-High Safeguards Actuation, T/S Table 3.3-5, Initiating Signal 8 These changes are numbered 072 and 073 in Attachment

3. This safeguards function is required to terminate main feedwater flow in events where the main feedwater system malfunctions causing an increase in feedwater flow. This is discussed in Section B.3.8A.2.3 of Appendix B to Attachment 4. The response times are provided in Table B.2-3 of Appendix B and in Appendix A of Attachment 4.

o) Pressure and Volume Limits for Accumulators, T/S 3/4.5.1 These changes are numbered 084 and 085 in Attachment 3.

New values for the accumulator volume and pressure ranges are proposed. They are based on the values used in the new LOCA analyses. The values used in the analyses are indicated in Tables C.3.1-2 and C.3.2-1 of Attachment 4, Appendix C. The applicability of the analyses to mixed core configuration is discussed in Attachment 4 Section 5.2 and in Attachment 4, Appendix C. The limit values for the pressure and volume limits are found in Attachment 4, Table 6.1 and Appendix A.

p) Safety Injection and RHR Pump Degradation, T/S 4.5.2.f These changes are numbered 086 and 087 in Attachment

3. AEP proposes new surveillance acceptance criteria for the safety injection (SI) and residual heat removal (RHR) pumps. These changes are based on the analyses performed for Unit 2 Cycle 8 and the analyses performed for both units as a part of the reduced temperature and pressure and the uprate programs. The section of WCAP 11908, "Containment Integrity Analysis" which describes the major analytic assumptions is included as Attachment 11 for the reviewer's convenience. WCAP 11908 was submitted on August 22, 1988 in AEP:NRC:1024D. The differential pressures, calculated by our contractor Westinghouse, are found in Attachment 4, Table 6.1 and Appendix A.

to AEP:NRC:1071E Page 10 These differential pressures are converted to discharge pressures as follows:

~32 Pum ~RHR Pum Differential Pressure 1385 psid 160 psid Suction Pressure ~24 si ~30 si Discharge Pressure 1409 psig 190 psig The suction pressure values are conservatively based on the maximum elevation of water possible in the refueling water storage tank (RWST) (limited by the elevation of the tank overflow piping), and the T/S minimum RWST tank temperature.

We are not proposing to change the discharge pressure for the centrifugal charging pump because our work on the new mass and energy releases analysis outside containment is incomplete.

However, we propose to revise the Unit 1 discharge pressures for the RHR and SI pumps. The present and proposed T/S minimum discharge pressure requirements are as follows:

Present Proposed Re uirement si Re uirements si SI 1345 1409 RHR 165 190 The revised requirements result from our discovery that the T/S requirements, which were recently approved (Amendment 126) for the Unit 1 Reduced Temperature and Pressure (RTP) Program, are inconsistent with the assumptions used by Westinghouse in performing the RTP analyses.

In WCAP 11902, which contains the RTP analyses, Westinghouse specified the required pump pressures as 1345 psi differential for the SI pumps and 165 psi differential for the RHR pumps. The Cook Nuclear Plant T/Ss are written in terms of pump discharge pressure, however, and the conversion from differential pressure to discharge pressure was inadvertently neglected. In addition, Westinghouse recently informed us that the value of 1345 psid differential pressure that was supplied in WCAP 11902 was in error and should actually be 1385 psi differential.

to AEP:NRC:1071E Page ll The correct differential pressures are found in WCAP 11902, Supplement Table S-3.13-2 and are identical to the values applicable to Unit 2. Therefore, the correct values for the Unit 1 discharge pressures are also identical to those calculated above for Unit 2.

Table S-3.13-2 is included in Attachment 5 for the reviewer's convenience.

Upon discovery of the discrepancies, the more stringent requirements for the SI and RHR pumps were implemented administratively. Surveillance test results have demonstrated that the pumps met the more stringent requirements at all times since the issuance of Amendment 126 and therefore the discrepancies did not impact safety.

q) Boron Injection System Flow Imbalance, T/S 4.5.2.h This change is numbered 089 in Attachment 3. SBLOCA was analyzed assuming a SI flow consistent with the proposed T/Ss. This is discussed in Attachment 4, Appendix C, Section C.3.2. The other analyses are not significantly affected by the proposed T/S. This parameter is also provided in Table 6.1 and Appendix A of Attachment 4 ~

10 CFR 50.92 Evaluation for T S Chan e Grou 1 Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

1) Involve a significant increase in the probability or consequences of an accident previously analyzed,

2) Create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or

3) Involve a significant reduction in a margin of safety.

Criterion 1 The proposed T/S changes to support Cycle 8 operations are accompanied by extensive analyses and evaluations which indicate that they will not result in an unsafe condition at the plant. The analyses and evaluations support our conclusion that the proposed T/S changes will not involve a significant increase in the probability or consequences of any accident previously analyzed,

0 to AEP:NRC:1071E Page 12 Criterion 2 The Cycle 8 analyses and evaluations comply with the.

licensing basis of the plant. Thus, the proposed T/S changes, should not create the possibility of a new or different kind of accident from any previously analyzed or evaluated.

Criterion 3 The proposed T/S changes to support Cycle 8 operations are accompanied by extensive analyses and evaluations which indicate that'hey will not result in an unsafe condition at the plant. The analyses and evaluations support our conclusion that the proposed T/S changes will not involve a significant reduction in any margin of safety.

The conclusions which our contractor, Westinghouse, drew from the safety evaluations and analyses are found on page 4 of Attachment 4. Westinghouse in part based their conclusions on the implementation of their proposed T/S changes in Table 6.1 and Appendix A of Attachment 4. We have proposed those changes unless the change is in another submittal, a more conservative change is proposed, or no change is more conservative than the Westinghouse proposal. Therefore, the Westinghouse conclusions support our 10 CFR 50.92 analysis.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing examples (48 FR 14870) of amendments considered not likely to involve a significant hazards consideration.

The sixth of these examples refers to changes that may result in some increase to the probability or consequences of a previously analyzed accident, but the results of which are within limits established as acceptable.

The analyses performed by Westinghouse in support of Unit 2 Cycle 8 operation comply with the licensing basis of the unit. Thus, we believe the example cited is applicable and that the changes should not involve a significant hazards consideration.

2) T S Chan e Grou 2 Removal of transition mode T/Ss associated with the Unit 2 Cycle 6 reanalysis.

a) Background In our March 14, 1986, submittal to the NRC (AEP:NRC:0916I) containing the proposed T/S changes to AEP:NRC:1071E Page 13 for the Unit 2 Cycle 6 reload, we included several T/Ss which are not in the Westinghouse Standard T/Ss (STS). They were incorporated in the Unit 2 T/Ss by Amendment 82 to Operating License DPR-74. These T/Ss resulted from a review of abnormal operating occurrences (AOO) and postulated accidents (PA) in the transition, shutdown, and refueling modes of operation (Modes 2,3,4,5 and 6). Reviews of AOOs and PAs in Modes 2 through 6. were conducted in conjunction with analyses and evaluations performed by Exxon Nuclear Company (now Advanced Nuclear Fuels Corporation) at the request of the NRC.

One non-standard T/S was proposed for Unit 1 in our submittal AEP:NRC:0916W. It is the shutdown margin in shutdown Modes 4 and 5. This was included in the Unit 1 T/Ss by Amendment 120 to Operating License DPR-58. We propose to revise the Unit 1 shutdown margin T/Ss in the same fashion that we propose to revise the Unit 2 shutdown margin T/Ss ~

AEP proposes to remove a number of these requirements from the T/Ss, thereby creating a structure more like that of the STS. The issues that are addressed by the T/Ss which we propose to delete will be addressed in the future by appro'priate administrative controls.

We believe this will simplify our T/Ss in conformance with industry practice and ensure the continued safe operation of the Cook Nuclear Plant.

Our plan was discussed with the NRC staff in a meeting on June 12, 1989 at NRC headquarters on Rockville Pike in Maryland. The staff expressed no concern with this aspect of our discussion. A summary of meeting topics and issues of concern to the staff are documented in a letter to M. P.

Alexich, Vice President of Indiana Michigan Power Company with its attachments from Joseph G. Giitter, Project Manager of the NRC staff, dated August 3, 1989. For the convenience of the reviewer, this letter is included in this submittal as Attachment 9.

The T/Ss which are impacted by the proposed changes in this group are:

(1) 3/4.1.1.1 Shutdown Margin, Standby, Startup, (Unit 1, and Power Operations Unit 2)

(2) 4.1.1.1.3 Shutdown Margin Surveillance (Unit 2)

to AEP:NRC:1071E Page 14 (3) 3/4.1.1.2 Shutdown Margin, Shutdown (Unit 1, Unit 2)

(4) 3/4.2.5.2 DNB Parameters, Modes 2,3,4,5 (Unit 2)

(5) Table 3.3-3 ESF Actuation System Instrumentation (Unit 2) b) Shutdown Margin in Shutdown Modes 4 and 5 These changes are numbered 027, 032 and 033 in Attachment 3. AEP proposes to rearrange T/Ss 3/4.1.1.1 and 3/4.1.1.2 into STS format by moving the Mode 4 requirement from 3.1.1.2 to 3.1.1.1 and deleting Figure 3.1-3 and related references. In the Cycle 6 submittal the Mode 4 shutdown margin was moved from 3/4.1.1.1 to 3/4.1.1.2. The Cycle 6 change to T/S 3/4.1.1.2, which then included Mode 4 and Mode 5 in the same T/S, was a revision that reflected the results of the boron dilution accident analysis when heat removal is via the residual heat removal (RHR) system. This analysis was performed by ANF.

AEP proposes to return 3/4.1.1.1 and 3/4.1.1 2 to STS

~

format which will be based on SDM requirements for analyses of the plant in Modes 1 and 2 ~ Protection for boron dilution accidents when the plant is operated on RHR will be addressed by administrative controls. This proposal applies to both units.

AEP presently plans to use the Westinghouse developed methods to ensure adequate operator response time in the event that either unit is subject to a dilution incident when operating in the various RHR cooling configurations. The methodology will address the limiting case of operation at reduced coolant inventory in the primary system. This methodology is a modification of Attachment 1 of the July 8, 1980, letter from T. M. Anderson of Westinghouse Electric Corporation to Victor Stello of the NRC. The identifier of Mr. Anderson's letter is NS-TMS-2273.

The modified methodology is similar to the original methodology. The NS-TMA-2273 methodology, or a

F I'

to AEP:NRC:1071E Page 15 similar methodology which is currently reflected in T/S 3/4.1.1.2 of Cook Nuclear Plant Units 1 and 2, has been in use on Unit 1 since the beginning of Cycle 6 and on Unit 2 since the beginning of Cycle 3.

The RHR function is a generic function. STS T/S 3.1.1.2 does not include unique SDM requirements for RHR operation. Furthermore, adequate response time for the dilution incident when cooling on RHR was ensured by administrative controls prior to Cycle 6 of Unit 2 and Cycle 10 of Unit 1. AEP will continue to ensure adequate operator response time by methodologies discussed above, or other effective methodologies.

c) Shutdown Margin Surveillance with ESF Actuations Blocked in Mode 3 These changes are numbered 031 and 058 in Attachment 3. The surveillance requirement 4.1.1.1.3 in the currently approved T/Ss is to be removed to further bring the Unit 2 T/Ss into conformance with the STS. AEP also proposes to revise footnotes ¹ and

¹¹ of Table 3.3-3. These footnotes are currently designed to be used with surveillance 4.1.1.1.3. The two footnotes, ¹ and ¹¹, will be returned to their original content, which is consistent with the STS.

Surveillance 4.1.1.1.3 and the elaboration of footnotes ¹ and ¹¹ were proposed as part of AEP:NRC:0916I for Cycle 6. As was indicated in XN-NF-85-28 (P), Disposition of Standard Review Plan Chapter 15 Events, these requirements were instituted to protect against a failure in the steam dump to condenser system below P-12. They were designed to ensure that in the event of a steam dump to condenser failure, the core would not become critical. A single failure in the steam dump to condenser below P-12 could potentially open three steam dump valves.

These valves would remain open until the operator was able to take action. Since this event is symmetric, ESF actuation on differential pressure between steam lines would not be expected. Safeguards actuations that would protect against symmetric events are blocked at P-ll and P-12.

Blocking of safeguards actuation on P-ll and P-12 are generic functions. These actions are required to cooldown and depressurize the units without a consequential safeguards actuation. The bases of the setpoints are generic. P-12 permits blocking of safeguards actuation on low secondary pressure when to AEP:NRC:1071E Page 16 0

Tavg is below 541 F. P-11 is above the low pressure SI setpoint and permits blocking of safeguards actuation on low pressurizer pressure. Blocking safeguards avoids undesired actuations during heatup and cooldown.

To protect against increased steam loads with safeguards blocked below P-11 and P-12, AEP plans to ensure cold shutdown SDM is available prior to blocking safeguards. Under certain circumstances in which cooldown is urgent, verification of SDM may not be complete prior to initializing of cooldown. An example of such a circumstance is a steam generator tube leak. Although safeguards will not be actuated by a tube leak, it is nevertheless important to promptly cooldown and depressurize. Procedures have been implemented for both units to address this possibility.

As noted above, the STS do not include surveillance 4.1.1.1.3 or the elaborations of footnotes ¹ and ¹¹.

AEP will ensure that an increase in steam load below P-11 and P-12 will not result in re-criticality by the methods discussed above or by other effective methods.

d) DNB Parameters - Modes 2, 3, 4 and 5 These changes are numbered 047 and 048 in Attachment 3. AEP proposes to delete the Modes 2, 3, 4 and 5 DNB requirements of T/S 3/4.2.5.2. AEP also proposes to remove Table 3.2-2 which is associated with specification 3.2.5.2. This specification was proposed as part of AEP:NRC:09161 for Cycle 6. The intent of this T/S is to prevent DNB from occurring should an uncontrolled rod withdrawal accident from subcritical occur.

AEP believes that other T/S requirements, which will be left in place, in conjunction with administrative controls are adequate to ensure that the assumptions of the analysis are met. T/Ss 3.4.1.2.c and 3.4.1.3.c which require three reactor coolant loops in operation when control rods are capable of withdrawal will be left in place. This provision is conservative relative to the analysis that was performed by Westinghouse for this submission which required only two operating reactor coolant loops.

T/S 3.4.1 1 requires that all reactor coolant pumps

~

be operable in Modes 1 and 2. Administrative controls 0

require all pumps when operating above 541 F. Therefore, in practice all pumps are in operation during startup ~ The operability to AEP:NRC:1071E Page 17 of T/S Table 3.3-1 for the nuclear 'equirements instrumentation high flux low setpoint reactor trip will also be left in place. In addition, the non-safety grade intermediate range and source range, trips are required to be operable by T/S Table 3.3-1 when control rods are capable of withdrawal.

Therefore, the source range high neutron flux trip will be available to terminate the event by tripping any withdrawn and withdrawing rods before any significant power level could be attained.

With the source range trip effective, DNB and primary system flow rate need not be considered. Also, the reactivity insertion rate would be slower when in any of the subcritical modes because the rod control system in Modes 2-5 is'expected to be in bank selects Therefore, a single failure in the rod control system could cause the withdrawal of only one bank, and its withdrawal rate would be expected to be slower than the maximum rod speed that is possible when in automatic rod control (and is assumed in the UFSAR analysis).

Administrative controls will ensure that the unit will be sufficiently shutdown such that criticality cannot occur with any bank fully withdrawn if the control rods are capable of withdrawal and the reactor is not at hot zero power (HZP) conditions.

Administrative controls also ensure that the upper temperature is limited at the HZP conditions by the setpoint of the non-safety grade steam generator power operated relief valves. The setpoint for these valves is 1040 psia limiting RCS temperature to approximately 549 F.

The analysis performed for the Donald C. Cook Nuclear Plant is similar to that of other plants.

Furthermore, the generic STS do not include DNB specifications for Modes 2, 3, 4 and 5. AEP believes that the requirements and controls discussed above will ensure that an inadvertent criticality wi.ll not occur.

e) Footnote $ to Applicable Mode for Turbine Trip and Feedwater Isolation in Table 3.3-3 and Table 4.3.2 These changes are numbered 057, 059, 078 and 082 in Attachment 3. The $ footnote was added to the Applicable Modes column of T/S Table 3.3-3 and the Modes in Which Surveillance Required of Table 4.3-2 as result of an evaluation of the increased feedwater event in Mode 4 performed by Exxon Nuclear Company (now Advanced Nuclear Fuels Corporation, ANF). AEP proposes 'to delete this footnote from Tables 3.3-3 and 4.3-2. The ANF calculation requires the steam

0 h

to AEP:NRC:1071E Page 18 generator water level--high-high functional unit to be operable when operating with the main feedpumps feeding the steam generators in Modes 3 and 4. The availability of feedwater isolation on high-high steam generator level limits the volume of cold water that can be added to the steam generators in any feedwater malfunction. This limits reactivity addition to the core.

Preheated feedwater is used frequently in Mode 3 operation. AEP plans to leave the Mode 3 requirement in place. The Mode 3 requirement has historically been in the T/Ss for both Cook Nuclear Plant units although it is not included in STS, Rev. 4. However, the likelihood of operating the main feedwater system in any mode other than 1, 2 or 3 is extremely remote.

The circumstances associated with this contingency will be addressed administratively as needed.

f) 10 CFR 50.92 Evaluation for T S Chan e Grou 2 Per 10 CFR 50,92, a proposed amendment to an operating license will not involve a significant hazards if the proposed amendment satisfies the following criteria:

1) Does not involve a significant increase in the probability or consequence of an accident previously analyzed,

2) Does not create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or

3) Does not involve a significant reduction in a margin of safety.

Criterion 1 The proposed T/S changes do not involve a physical change to the plant. The procedures and administrative controls for the plant will either remain in place as described above, or in some cases be replaced by controls which we believe are of comparable effectiveness. Therefore, we conclude that the proposed T/S changes will not result in a significant increase in the probability or consequences of any accident previously analyzed.

to AEP:NRC:1071E Page 19 Criterion 2 The proposed T/S changes do not involve a physical change to the plant. The procedures and administrative controls for the plant will either remain in place as described above, or in some cases be replaced by controls which we believe are of comparable effectiveness'herefore, we conclude that the proposed changes will not create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3 The proposed T/S changes do not involve a physical change to the plant. The procedures and administrative controls for the plant will either remain in place as described above or, in some cases, be replaced by controls which we believe are of comparable effectiveness. Therefore, we conclude that the proposed T/S changes will not involve a significant reduction in any margin of safety.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration. The sixth of these refers to changes that may result in some increase to the probability or consequences of a previously analyzed accident, but the results 'of which are within limits established as acceptable.

The proposed T/S changes in Change Group 2 remove certain activities from T/S control. However, since operation of the plant will be governed by the continuation of existing controls or controls of comparable effectiveness, the results of the underlying evaluation should remain within limits established as acceptable. For this reason we believe the sixth example bounds the proposed Change Group 2 changes.

3) TS Chan e Grou 3 Changes to Unit 2 T/Ss for consistency with Unit 1 T/Ss a) Discussion There is one change in this category. This change is numbered 083 in Attachment 3. It is a proposal to increase the water level for operability in the to AEP:NRC:1071E Page 20 pressurizer. This change was submitted for Unit 1 in our reduced temperature and pressure (RTP) program submittal, AEP:NRC:1067 dated October 14, 1988. It was approved in Amendment No. 126 to Operating Licensing DPR-58.

We propose to increase the maximum pressurizer level to 92% of span. The purpose of the maximum pressurizer level limit, as described in the Bases, is to ensure that a bubble can exist in the pressurizer. Westinghouse has determined that a bubble can be maintained at the 92% level. The change is described in detail in Section 3.13 of WCAP 11902 which is included in this submittal as Attachment 6 for the reviewer's convenience. The change will allow operational flexibility at the higher end of the Tavg spectrum analyzed for the RTP program.

b) 10 CFR 50.92 Evaluation for T S Chan e Grou 3 Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

1) Involve a'significant increase in the probability or consequences of an accident previously analyzed,

2) Create the possibility of a new or different kind of accident from any previously analyzed or evaluated, or

3) Involve a significant reduction in a margin of safety.

Criterion 1 The proposed T/S change proposed is accompanied by an evaluation which indicates that it will not result in an unsafe condition at the plant. The evaluations support our conclusion that the change, which has already been approved for Unit 1, will not involve a significant increase in the probability or consequences of any accident previously analyzed.

0 to AEP:NRC:1071E Page 21 Criterion 2 The evaluation of t'e proposed T/S change complies with the licensing basis of the plant. Thus, this change should not create the possibility of a new or different kind of accident from any previously analyzed or evaluated.

Criterion 3 The proposed T/S change is accompanied by an evaluation which indicates that it will not result in an unsafe condition at the plant. The evaluation supports our conclusion that the proposed T/S change, which has already been approved for Unit 1, will not involve a significant reduction in any margin of safety.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration. The sixth of these examples refers to changes that may result in some increase to the probability or consequences of a previously analyzed accident, but the results of which are within limits established as acceptable.

The evaluation for this proposed T/S change complies with the licensing basis of the plant.

Thus, we believe the example cited is applicable and that the change should not involve a significant hazards consideration.

a) Discussion This T/S change group consists of changes that are purely editorial in nature. The Attachment 3 identification numbers for these proposed changes are:

Unit 1 116, 118, 123 Unit 2 003, 004, 005, 006, 009, 010, 017, 018, 024, 025, 028, 029, 030, 035, 037, 038, 040, 044, 045, 046, 052, 053, 055, 056, 060, 061, 062, 063, 064, 065, 067, 068, 074, 075, 076, 077, 079, 080, 081, 088, 090.

to AEP:NRC:1071E Page 22 The changes in this group include proposals to enhance the readability of the T/Ss, to move existing text, or to perform other non-substantive changes as described in Attachment 3.

b) 10 CFR 50.92 Evaluation for T S Chan e Grou 4 Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration amendment does not:

if the proposed (1) Involve a significant increase in the probability or consequences of an accident previously evaluated, (2) Create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) Involve a significant reduction in a margin of safety.

Criterion 1 These proposed T/S changes, being editorial in nature and intended to improve the readability of the T/Ss, will not reduce in any way requirements or

'commitments in the existing T/Ss. Thus, we believe that no increase in the probability or consequences of a previously evaluated accident would be expected as a result of these proposed T/S changes.

Criterion 2 We believe that these purely editorial proposed T/S changes will not create the possibility of a new or different kind of accident from any previously evaluated, because no change to the plant or plant operations will results Criterion 3 We believe that, the proposed T/S changes will not involve a significant reduction in any margin of safety, because all accident analyses and nuclear design bases remain unchanged.

to AEP:NRC:1071E Page 23 Lastly, we note that the Commission has provided guidance concerning, the determination of significant hazards by providing certain examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration. The first of these examples refers to changes that are purely administrative in nature: for example, changes to achieve consistency throughout the T/Ss, correction of an error, or a change in nomenclature. This group of proposed changes is intended to achieve consistency between the Unit 1 and' T/Ss, or to improve the overall readability of the T/S document.

As these changes are purely editorial and do not impact safety in any way, we believe the Federal Register example cited is applicable and that the changes involve no significant hazards consideration.

5) T S Chan e Grou 5 Changes to the surveillance requirement for power distribution limits/axial flux difference for Units 1 and 2.

(a) Discussion This change is numbered 039 in Attachment 3.

The proposed change will eliminate unnecessary surveillances of the indicated axial flux difference (AFD) during power operation above 15% of rated thermal power. The current T/Ss for both units require in 4.2 alai.a.2 that the indicated AFD for each operable excore channel be monitored "at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD monitor alarm to operable status." The proposed T/S change will add a caveat. This surveillance will only be required "if the AFD had been outside of the target band for any period of time in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation."

The requirement to monitor the indicated AFD for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD monitor alarm to operable status results from the fact that the current computer program that monitors the AFD has no provision for updating the cumulative time out of the target band. The cumulative time out of the target band is reset to zero when the AFD monitor alarm is restored to operable status. In the case where it can be demonstrated that the AFD indeed has not been outside the target band in the previous 24-hour period, zero is the proper time for cumulative to AEP:NRC:1071E Page 24 time out of the target band. In this case, no further monitoring is needed. In the other case, where the AFD has been out of the target band in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, zero is non-conservative and is not acceptable for the cumulative time out of the target band.

Therefore, the AFD must be manually monitored for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to ensure that sometime in the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation that limit on cumulative time out of the target band is not exceeded. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the rolling log of cumulative time out of target band will start to overwrite itself and complete information for the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is contained in the log.

b) 10 CFR 50.92 Evaluation for T S Chan e Grou 5 Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:

1) Involve a significant increase in the probability or consequences of an accident previously analyzed,

2) Create the possibility of a new or different kind of accident from any previously analyzed or evaluated, or

3) Involve a significant reduction in a margin of safety.

Criterion 1 This proposed T/S change will not result in an increase in the probability or consequences of an accident previously analyzed. The requirement the AFD for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after to-'onitor restoring the AFD to operable status will be eliminated only when the axial power distribution has been within the target band for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to restoring the alarm to operable status. This will

,not impact the safety of the plant because the'larm can accurately monitor the cumulative time out of the target band in this case. If the AFD has been outside of the target band at any time in the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to restoring the alarm to operable status, then the surveillance of monitoring the AFD for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is still required.

to AEP:NRC:1071E Page 25 Criterion 2 We believe that adding the caveat to the additional surveillance requirement will not result in a new or different kind of accident from any previously evaluated.

The modified surveillance will ensure that the cumulative time out of the target band will be conservatively calculated for all circumstances either by the plant process computer or by manual monitoring.

Criterion 3 The requirement to monitor the AFD for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD to operable status will be eliminated only when the axial power distribution has been within the target band for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to restoring the alarm to operable status. This will not impact the safety of the plant because the alarm can accurately monitor the cumulative time out of the target band in this case. If the AFD has been outside of the target band at any time in the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to restoring the alarm to operable status, then the surveillance of monitoring the AFD for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is still required. We conclude that this proposed T/S change will not involve a significant reduction in any margin of safety.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration. The sixth of these refers to changes that may result in some increase to the probability or consequences of a previously analyzed accident, but the results of which are within limits established as acceptable.

We believe this example is applicable to the proposed Change Group 5 T/S change. As indicated above, we anticipate no consequential impact on delta flux monitoring and control to result from this change.

Therefore, the underlying evaluations and analyses for delta flux control should clearly remain within limits established as acceptable.

to AEP:NRC:1071E Page 26

6) T S Chan e Grou 6 Changes to Unit 1 T/Ss for consistency with Unit 2 T/Ss. These changes are, numbered 115, 117, 119, 120, 121 and 122. The corresponding Unit 2 changes are numbered 027, 032, 033, 039, 086 and 087 in Attachment 3.

These changes to the Unit 1 T/Ss are being proposed to make the Unit 1 T/Ss more like the Unit 2 T/Ss.,

The particular Unit 1 changes being proposed have an identical or essentially identical justification for both units and the Unit 2 justification is part of this submittal. The proposal to remove special shutdown margin requirements when operating on RHR, T/S Sections 3.1.1.1 and 3.1.1.2 and Figure 3.1-3, discussed in T/S Change Group 2b, is proposed for both units. The modification to the axial flux difference (AFD) surveillance 4 '.1.1.a.2, discussed in T/S Change Group 5 is proposed for both units.

Finally, the new acceptance criteria for safety injection and residual heat removal pumps discussed in T/S Change Group lp is proposed for both units for T/S Section 4.5.2.

A -10 CFR 50.92 evaluation is not included for Change Group 6 because we believe the evaluations performed for the corresponding Unit 2 changes in T/S Change Groups lp, 2b, and 5 apply to both Unit 1 and Unit 2.

IV. Pro osed Chan es to the Bases In addition to the changes to the T/Ss described above, we have also proposed changes to the Bases section to reflect in the safety analyses and changes in the T/Ss. both'hanges Descriptions of these changes have been included in Attachment 3.

Conclusion We believe that the proposed changes do not involve a significant hazards consideration because operation of Cook Nuclear Plant Unit 2 in accordance with these changes would not:

k (1) Involve a significant increase in the probability of occurrence or consequences of an accident previously analyzed, (2) Create the possibility of a new or different kind of accident from any accident previously evaluated, or to AEP:NRC:1071E Page 27 (3) Involve a significant reduction in a margin of safety.

This conclusion is based on our evaluation of the changes, which has determined that all proposed changes that are not administrative in nature, consistent with the STS, or consistent with the design basis of the plant are clearly traceable to the various supporting safety analyses, as referenced by Attachment 3. Assuming Commission acceptance of these analyses, it is our belief that they successfully demonstrate that applicable safety limits and margins to safety will be maintained.