ML17325B293

DC Cook Unit 2 Reactor Vessel Heatup & Cooldown Limit Curves for Normal Operation

ML17325B293
Person / Time
Site:	Cook
Issue date:	04/30/1989
From:	Meyer T, Ray N, Yanicko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML17325B291	List: ML17325B290 ML17325B292 ML17325B293
References
MT-SMART-090(89, MT-SMART-090(89), MT-SMART-90(89, MT-SMART-90(89), NUDOCS 8911060269
Download: ML17325B293 (174)

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{{#Wiki_filter:REMOVE INSTRUCTIONS FOR THE INSERTION OF REVISION 2 TO THE STEAM GENERATOR REPAIR REPORT S9110b02b9 891025 PDR ADOCK 050003ib ~NSERT Record of Revisions dated November 4, 1986 Page ii, Revision 1 Page v, Revision 0 Pages vii & viia, Revision 1 Page 44, Revision 0 Page 49a thru 49j, Revision 1 Page 55, Revision 0 Pages 61, Revision 1 & 62, Revision 0 Page 67, Revision 1 Page 75 & 76, Revision 1 Page 82a, Revision 1 Page 144, Revision 0 Pages

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& 165, Revision 0 Pages 170 & 171, Revision 0 Page 173, Revision 0 Page 183, Revision 0 Pages 202 & 203, Revision 0 Page 1-2, Revision 1 Page 1-4, Revision 1 Pages 1-6 thru 1-19, Revision 1 I Record of Revisiqns dated July 24, 1987 Page ii, Revision 2 Page v, Revision 2 Pages vii & viia, Revision.2 Page 44, Revision 2 Pages 49a thru 49q, Revision 2 Pages 55,

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RECORD OF REVISIONS REVISION NO. REVISION DATE DATE ISSUED ISSUED BY March 30, 1987 July 24, 1987 March 30, 1987 July 24, 1987 November 4, 1986 November 4, 1986 T. G. Harshbarger T. G. Harshbarger T. G. Harshbarger

Section ~Pa e 2.2.2 2.2.3 2.3 Parametric Comparison Materials Comparison Component Design Improvements 28 28 32 2.3.1 Design Improvements to Minimize Potential for Tube Degradation 32 2.3.2 Design Improvements to Increase Performance 37 2.3.3 Design Improvements to Enhance Maintainability and Reliability 38 2.4 2.4.1 2'.2 2.5 Codes and Standards Industry Codes and Standards USNRC Regulatory Guides Shop Tests and Inspections 40 40 40b 42 3.1 3.2 3.3 3.3.1 3.3.2 3.3.3 3.3.4 3.4 3.4.1 SECTION 3 - REPAIR PROJECT Overview Guidelines and Criteria Preshutdown Activities Site Preparation Shipment and Storage of Replacement Components Modification to Auxiliary Building Structural Steel Polar Crane Power Circuit Relocation Post Shutdown Activities Containment Preparations 47 50 50 55 55a 55a 55a 3.4.2 Removal of Concrete, Structural and Equipment Interferences 57 3.5 Steam Generator Removal Activities 62 3.5.1 Steam Generator Cutting Methods and Locations 62 3.5.2 Removal and Handling of the Steam Generator Upper Assemblies 64 3.5.3 Removal and Handling of the Steam Generator Lower Assemblies 66 Revision 2

~Sectic ~Pa e 6.2.1 6.2.2 I 6.3 6.4 6.4.1 6.4.2 6.4.3 Handling of Heavy Loads Shared System Analysis Fire Protection Evaluation Analysis of Significant Hazards Criterion 1 Criterion 2 Criterion 3 155 162 163 163a 164 164 165 SECTION 7 - ENVIRONMENTAL REPORT 7.1 7.2 . 7.2.1 7.2.2 Purpose of the Environmental Report The Plant and Environmental Interfaces Geography and Demography Regional Historic, Archaeological, Architectural, Scenic, Cultural, and Natural Features 166 166 166 167 7.2.'3 7.2.4 7 ' 7.2.6 Hydrology Geology Ecology Noise 167 168 168 169 7.3 7.3.1 7.3.2 Non-Radiological Environmental Effects Geography and Demography Regional Historic, Archaeological, Architectural, Scenic, Cultural, and Natural Features 169 169 170 7.3.3 7.3.4 7.3.5 7.3.6 7.4.1 Hydrology Geology ,Ecology Noise Radiological Environmental Effects Occupational Exposure 170 171 171 172 172 172 Revision 2

LIST OF TABLES Table 1.1-1 Title D. C. Cook Nuclear Plant Secondary Side Water Chemistry Specification History-Steam Generator ~Pa e 2.2-1 Comparison Between the Original and Repaired Steam Generators 27 212-2 Comparison of Design Data Between the Original and Repaired Steam Generators 29 202-3 Comparison of Materials of Construction Between the Original and Repaired Steam Generators 31 3.2-1 Industry Codes and Standards Applicable to the Steam Generator Repair Project Field Work 49b 3.2-2 USNRC Regulatory Guides Applicable to the Steam Generator Repair Project Field Work 49i 3.2-3 D. C. Cook Unit 2 Technical Specifications Not Applicable During the Steam Generator Repair Project 49n

3. 6-.1 3 '-1 Steam Generator Repair Welds Repair Project Manrem Estimates 75 98 3,8-2 Projected Project Totals by Phase for Man-hours and Man>>rem 103 7.4-1 7.4-2 7.4-3 Donald C.

Cook Annual Man-rem Expenditures Steam Generator Man-rem Expenditure Comparison Gross Contamination Levels by Location in Piping and Steam Generator 173 174 180 7.4-4 Donald C. Cook Nuclear Plant Unit 2 Estimated Steam Generator Curie Content 181 7.4-5 Effluent Release Isotopic Distributors, Steam Generator Replacement

Project, Surry Power Station

- Unit No. 2 182 7.4-6 Comparison of Gaseous Effluent Releases from Donald C. Cook Nuclear Plant 183 7.4-7 Radionuclide Concentrations in Reactor Coolant 184 vii Revision 2

LIST OF TABLES cont'd. Table 7.4-8 7.4-9 Title ~Pa e Estimated Radionuclide Releases Due to Discharge of Reactor Coolant Water 186 Estimated Specific Activities of Laundry Waste Water 185 7.4-10 7.4-11 Estimated Radioactive Liquid Effluent Releases During the Donald C. Cook Unit 2 Steam Generator Repair Project Comparison of Radioactive Liquid Effluent Releases 188 189 Summary Cost-Benefit Analysis for the Unit 2 Steam Generator Repair Project 200. viia Revision 2

FIGURE 2.2-2 MODIFICATIONS TO UPPER ASSEMBLY INTERNALS UPPER ASSEMBLY NEW STEAM WATER DEFLECTORS (3 TOTAL) iciwNXcc/0 +5N+C(l..b 00 0 0 0 0 0 0 00 0 0 0 4o o44 40 0 0 0 0 0 0 0 0 0 0 4 0 0 00 0 EXISTING SECONDARY SEPARATOR ~SECONDARY MANWAY (2"180'PART) ~NEW DRYER DRAINS (8 TOTAL) NEW STEAM CHIMNEYS {3 TOTAL) EXISTING SWIRL VANE ASSEMBLY SHELL NEW FEEDWATER RING AND INCONEL J-NOZZLES "ltd~:%i:i) g4 %~%MA>s NEW LOWER ASSEMBLY 44 RE VISION 2

o Although there will be no fuel in the Unit 2 core, Unit 2 will be considered to be in Mode 6 during the Steam Generator Repair Project. Unit 2 Technical Specifications will be adhered to with the exception of those Technical Specifications listed in Table 3.2-3. The Technical Specifications listed in Table 3.2-3 will not be applicable during the Steam Generator Repair Project. For purposes of Technical Specification applicability, the Steam Generator Repair Project will begin when "he last fuel assembly from the Unit 2 core is placed in the spent fuel pool and will end when the first fuel assembly is removed from the spent fuel pool to refuel the Unit 2 core. -49a-Revision 2

TABLE -1 INDUSTRY CODES AND STANDARDS APPLICABLE TO THE STEAM GENERATOR REPAIR PROJECT CODE OR STANDARD ACI 301-84, "Specifications for Structural Concrete Buildings, Chapters 2 and 3." ACI 304-85, "Recommended Practices for Measuring, Mixing, Transporting, and Placing Concrete." ACI 315-80, "Details and Detailing of Concrete Reinforcement." ACI 308-81, "Recommended Practice for Curing Con-crete." ACI 318-83, "Building Code Requirements for Rein-forced Concrete, Chapters 3, 4, and 5." American Welding Society D.l.1-1986, "Structural Welding Code Steel." American Welding Society D.l.3.-1981, "Structural . Welding Code, Sheet Steel." ASME Boiler and Pressure Vessel

Code, Section II, "Material Specifications," edition and addenda in use at time of material procurement.

ADDITIONALINFORMATION EXCEPTION I ~Exec tion: Mix proportions shall be selected (1) utilizing laboratory or field trial batches, (2) previous satisfactory performance on similar work using the same or similar materials, or (3) prior experience with these or similar materials to provide concrete of the required strength, durability, work-ability, economy, etc. I I I I I I I ~Exec tion: Curing shall be for a period of seven (7) days or until standard cured cylinders reach a comp-rehensive strength of 3500 PSI, whichever is first. Adherence to this criteria shall be sufficient to preclude testing for "Evaluation of Procedures," "Curing Criteria Effectiveness" or "Maturity Factor Basis." I ~Exec tibn: Mix proportions shall be selected (1) utilizing laboratory or field trial batches, (2) previous satisfactory performance on similar work using the same or similar materials, or (3) prior experience with these or similar materials to provide concrete of the required strength, durability, work-ability, economy, etc. I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I II. I I

TABLE 3.2-ontinued) CODE OR STANDARD ASME Boiler and Pressure Vessel

Code, Section III, "Rules for Construction of Nuclear Vessels/Rules for Construction of Nuclear Power Plant Components,"

edition and addenda as discussed below. The original Construction code for D. C. Cook Unit 2 nuclear vessels is Section III, 1968 Edition plus Addenda through Winter 1968, and for piping components is ANSI B31.1-1967 and ANSI B31.7-1969. As allowed by ASME Section XI, Subarticle IWA-7210, selected portions of the original Construction Codes dealing with installation and testing will be updated to applicable portions of Section III, 1983 Edition plus Addenda through Summer 1984. ASME Boiler and Pressure Vessel

Code, Section IX, "Welding and Brazing Qualifications," edition and addenda in use at time of procedure qualification.

ASME Boiler and Pressure Vessel

Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components,"

1983 Edition plus Addenda through Summer 1983. ANSI B31.1, "Power Piping", edition and addenda in use at time of contract award for field piping services. ANSI N45.2 - 1977 Quality Assurance

Program,

'Requirements for Nuclear Facilities USAS (ANSI) B31.1-1967, "Power Piping". USAS (ANSI) B31.7-1969, "Nuclear Power Piping". I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I ADDITIONAL INFORMATION EXCEPTION Exce tions: - Consistent with the plant design basis, fracture toughness requirements will not apply. - N-stamping of fabricated piping components will not be required. ~Exec tion: - Consistent with the plant design basis, fracture toughness requirements will not apply. ~Exec tion: - As noted under ASME Boiler and Pressure Vessel

Code, Section III, "Rules for Construction of Nuclear Vessels/Rules for Construction of Nuclear Power Plant Components"

above, these codes represent the original Construction Code for

TABLE 3.2-1 ontinued) CODE OR STANDARD ADDITIONAL INFORMATION EXCEPTION nuclear piping components. Portions dealing with materials and fabrication for new nuclear pressure retaining components, and installation and testing of all nuclear pressure retaining components, will be updated to ASME Section III, with the exception that fracture toughness requirements will not apply. The piping design basis and any additional design. activities relating to nuclear piping systems will be in accordance with USAS (ANSI) B31.1-1967. ASTM C31 "Standard Method of Making and Curing Concrete Specimens in the Field". ASTM C33 "Standard Specification for Coarse Aggregates". ~Exec tions: - The average fineness modulus of the fine aggregate may be between 2.5 and 3.0, however individual samples shall not vary more than 0.20 from the average. Compliance with gradation and fineness modulus requirements for fine aggregate shall consist of 4 out of 5 successive test results meeting the specifications. Coarse aggregate gradation shall be Number 57, 1 inch x C4. Coarse aggregate sodium sulfate soundness loss shall be a 10-percent maximum at 5 cycles. Coarse aggregate Los Angeles Abrasion loss shall be a maximum of 40 percent at 500 revolutions. ASTM C39 "Test Method for Compressive Strength of Cylindrical Specimens". ASTM C40 "Test Method for Organic Impurities in Fine Aggregates for Concrete".

TABLE 3.2-ontinued) CODE OR STANDARD ADDITIONAL INFORMATION EXCEPTION ASTM C31 "Standard Method of Making and Curing Concrete Specimens in the Field". I ASTM C33 "Standard Specification for Coarse Aggregates". for nuclear piping components. Portions dealing with materials and fabrication for new nuclear pressure retaining components, and installation and testing of all nuclear pressure retaining components, will be updated to ASME Section III, with the exception that fracture toughness requirements will not apply. The piping design basis and any additional design activities relating to nuclear piping systems will be in accordance with USAS (ANSI) B31.1-1967. Exce tions: - The average fineness modulus of the fine aggregate may be between 2.5 and 3.0, however individual samples shall not vary more than 0.20 from the average. - Compliance with gradation and fineness modulus requirements for fine aggregate shall consist of 4 out of 5 successive test results meeting the specifications. - Coarse aggregate gradation shall be Number 57, 1 inch x W. - Coarse aggregate sodium sulfate soundness loss shall be a 10 percent maximum at 5 cycles. - Coarse aggregate Los Angeles Abrasion loss shall be a maximum of 40 percent at 500 revolutions. ASTM C39 "Test Method for Compressive Strength of Cylindrical Specimens", ASTM C40 "Test Method for Organic Impurities in Fine Aggregates for Concrete".

TABLE 3.2-1 ntinued) CODE OR STANDARD ADDITIONAL INFORMATION EXCEPTION ASTM C88 "Test Method for Soundness of Aggregates by Use of Sodium Sulfate or Magnesium Sulfate". ASTM C94 "Standard Specification for Ready Mix Concrete". ASTM C117 "Test Method for Materials Finer Than No. 200 Sieve in Mineral Aggregates by Washing". ASTM C123 "Test Method for Lightweight Pieces in Aggregate". ASTM C127 "Test Method for Specific Gravity and Adsorption of Coarse Aggregate". ASTM C128 "Test Method for Specific Gravity and Adsorption for Fine Aggregate". ASTM C131 "Test Method of Resistance to Degradation of Small-Size Coarse Aggregate by Abrasion and Impact in the Los Angeles Machine". ASTM C136 "Method for Sieve Analysis of Fine and Coarse Aggregates". ASTM C138 "Test Method for Unit Weight, Yield, and Air Content (Gravimetric) of Concrete". Exce tions: - Except strike off bar utilized in lieu of glass plate for unit weight determination. - Except "Yield" and "Air Content (Gravimetric)" portions will not be utilized. ASTM C142 "Test Method for Clay Lumps and Friable Particles in Aggregate" ~ ASTM C143 "Test Method for Slump -of Portland Cement Concrete". ASTM C150 "Specification for Portland Cement". ~Exec c1ons: - Except cement shall be free of false set when tested in accordance with ASTM C451.

TABLE 3.2-ontinued) CODE OR STANDARD I I I I I ADDITIONALINFOMfATION EXCEPTION - Except total alkalies shall not exceed 0.60 percent by weight when calculated as the percentage of Na20 plus 0.658 times the percentage of K20. I ASTM C172 "Method of Sampling Freshly Mixed Concrete".[ ASTM C231 "Test Method for Air Content of,Freshly Mixed Concrete by the Pressure Method". ASTM C260 "Specifications for Air-Entrained Admixtures for Concrete". ASTM C289 "Test Method for Potential Reactivity of Aggregates (Chemical Method)". ~Exec tice: - Only the Type 3 Apparatus shall be utilized. ASTM C309 "Specifications for Liquid Membrane-Forming Compounds for Curing Concrete". ASTM C311 "Methods of Sampling and Testing Fly Ash or Natural Pozzolans for Use as a Mineral Admixture in Portland Cement Concrete". ASTM C494 "Specification for Chemical Admixtures in Concrete". ASTM C566 "Test Method for Total Moisture Content of Aggregate by Drying". I= ASTM C6'7 "Practice for Capping Cylindrical Concrete Specimens". ASTM C618 "Specification for Fly Ash and Rain or Calcined Natural Pozzolan for Use as a Mineral Admixture in Portland Cement Concrete". ASTM C702 "Methods for Reducing Field Samples of Aggregate to Testing Size". I I I = I I I I I I I I I I I I I I I I I I

TABLE 3.2-1 (Continued) CODE OR STANDARD ADDITIONALINFORMATION EXCEPTION I SSPC-SP1 through SP10 - 1982 Steel Structures Painting Council Specifications for Surface Preparation of Steel Surfaces Note: 1) All ASTMs are latest edition.

TABLE 2 USNRC REGULATORY GUIDES APPLICABLE TO THE STEAM GENERATOR REPAIR PROJECT FIELD WORK REGULATORY GUIDE NUMBER REGULATORY GUIDE TITLE REGULATORY GUIDE REVISION ADDITIONAL INFORMATION EXCEPTIONS 1.8 Personnel Selection and Training 1-R (9/75) Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. 1.26 Safety Guide 30 Quality Group Classification and Standards for Water,

Steam, and Rad-waste Containing Components of Nuclear Power Plants Quality Assurance Requirements for Installation, Inspection and Testing of Instrumentation and Electrical Equipment 3 (2/76)

(8/72) Classification of Class 2 and 3 components for the purpose of implementing ASME Section XI requirements was made in accordance with this guide. Committed to in UFSAR, Section 1.7, "QAPD", Appndix A. 1.31 Safety Guide 33 Control of Ferrite Content in Stainless Steel Weld Metal Quality Assurance Program Requirements (Operational) 3 (4/78) (11/72) The requirements of this guide are now covered by ASME'Section III. Field work relating to the steam generator repair project will be in compliance with this regulatory guide. Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. 1.37 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants 0 (3/73) Committed to in UFSAR, Section 1.7, "QAPD", Appendix A.

TABLE 3.2-2 ( inued) 0-REGULATORY GUIDE NUMBER REGULATORY GUIDE TITLE REGULATORY GUIDE REVISION ADDITIONAL INFORMATION EXCEPTIONS 1.38 Quality Assurance Requirements for Packing, Shipping, Receiving

Storage, and Handling of Items for Water-Cooled Nuclear Power Plants 1 (10/76)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. 1.39 1.44 Housekeeping Requirements for Water-Cooled Nuclear Power Plants Control of Sensitized Stainless Steel 1 (10/76) 0 (5/73) Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. If applicable to this repair project, the field work will comply to this guide. 1.48 1.50 1.54 Design Limits and Loading Combinations for Seismic Category I Fluid System Components Control of Preheat Temperature for Welding of Low-Alloy Steel Quality Assurance Requirements for Protective Coatings Applied 0 (5/73) 0 (6/73) This regulatory guide was withdrawn 3/4/85 (see 50FR9732). Project repair work will be performed in compliance with this regulatory guide. Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. Exception: Committed only to ANSI N101.4-1972. 1.58 Qualification of Nuclear Power Plant Inspection Examination and Testing Personnel 1 (9/80) Committed to in UFSAR, Section 1.7, "QAPD", Appendix A.

TABLE 3.2-2 ( inued) REGULATORY GUIDE NUMBER REGULATORY GUIDE TITLE REGULATORY GUIDE REVISION ADDITIONAL INFORMATION EXCEPTIONS

1. 64 Quality Assurance Requirements for the Design of Nuclear Power Plants 0 (10/73)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. 1.68 1.71 1.74 Initial Test Program for Water-Cooled Nuclear Power Plants Welder Qualifications for Areas of Limited Accessibility Quality Assurance Terms and Definitions 2 (8/78) I' (12/73) 0 (2/74) This regulatory guide will be used only for guidance in developing a test program for those components and systems affected by the Steam Generator Repair Project. Welders making welds in areas of restricted accessibility will be required to practice and qualify on a similar configuration to the weld being made. Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. 1.88 1.89 Collection,

Storage, and Maintenance of Nuclear Power Plants Quality Assurance Records Environmental Qualification of Certain Electrical Equipment Important to Safety for Nuclear Power Plants 2 (10/76) 1 (7/84)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. Project repair work will be performed in accordance with this regulatory guide.

TABLE 3.2-2 inued) 'EGULATORY GUIDE NUMBER REGULATORY GUIDE TITLE REGULATORY GUIDE REVISION ADDITIONAL INFORMATION EXCEPTIONS 1.94 Quality Assurance Requirements for Installation; Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 1 (4/76) Exce tions: - "Grout testing" (ASTM C109) included in Table B of ANSI N45.2.5-1974 is inappropriate for field testing as it is a sophisticated laboratory test utilized for cement evaluation. In lieu of daily tests, pre-packaged non-shrink grouts shall be accepted for use on the basis of manufacturer's certification or compressive strength tests made in the field. Confirmation compressive strength tests shall be made during the first day's production and thereafter on a basis of either once per day of every one-hundred (100) bags used, whichever is least. - Water and ice shall be sampled and tested to ensure either potability or certified to contain not more than 2,000 parts per million of chlorides as Cl, nor more than 1,500 parts per million of sulfates as S04. Acceptability of this water or ice shall be per this certification and preclude the ASTM's referenced in Table B of ANSI N45.2.5-1974. - The reference,, in Table B of ANSI N45.2.5-1974, to soft fragment testing per ASTM changed designations to ASTM C851 which was deleted in 1985. No testing for soft fragments is intended. ~Exes tice: Sister splices will be substituted for production splice required for tensile testing under Section 4.9 of ANSI N45.2.5-1974.

TABLE 3.2-2 ( inued) REGULATORY GUIDE NUMBER REGULATORY GUIDE TITLE REGULATORY GUIDE REVISION ADDITIONAL INFORMATION EXCEPTIONS 1.100 Seismic Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 1 (8/77) Project repair work will be performed in accordance with this regulatory guide. 1.116 Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems. 0-R (5/77) Exception: Committed to ANSI N45.2.8 (1975), "Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Power Plants" per UFSAR, Section 1.7, "QAPD", Appendix A. Not committed to this regulatory guide. 1.123 Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Plants 1 (7/77) Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. 1.131 Qualification Tests of Electric

Cables, Field Splices, and Connections for Light-Mater-Cooled Nuclear Power Plants 0 (8/77)

Project field work will be performed in accordance with this regulatory guide. 1.144

1. 146 Auditing of Quality Assurance Programs for Nuclear Power Plants Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants 0 (1/79) 0 (8/80)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A. Committed to in UFSAR, Section 1.7, "QAPD", Appendix A.

TABLE 3.2-3 D. C. COOK UNIT 2 TECHNICAL SPECIFICATIONS NOT APPLICABLE DURING THE STEAN GENERATOR REPAIR PROJECT TECHNICAL SPECIFICATION NUNBER TITLE ADDITIONAL INFORHATION COWENT 3.'l.1.3 Reactivity Control Systems-Boron Dilution This Technical Specification ensures adequate mixing of coolant with the low boron concentration stream being introduced into the system. This mixing prevents a large concentration gradient in the core which would cause localized power excursions. Mith no fuel in the reactor vessel, there is no concern about decay heat removal or boron mixing. 3.1.2.1 Reactivity Control Systems-Boration Systems - Flow Paths-Shutdown This Technical Specification requires that one boron injection flow path remains operable. This ensures that negative reactivity control is available. With no fuel in the reactor vessel there is no need for negative reactivity control.. 3.1.2.5 Reactivity Control Systems-Boric Acid Transfer Punps-Shutdown This Technical Specification requires that at least one boric acid transfer punp remain operable. This ensures that negative reactivity control is available. With no fuel in the reactor vessel there is no need for negative reactivity control. 3.3.3.9 Instrunentation - Radioactive Liquid Effluent Instrunentation, specifically the following survei llance requirements: 4.3.3.9.2, 1b Steam Generator Slowdown Line (2-R-19) 4.3.3.9.2, 1c Steam Generator Slowdown Treatment Effluent (2-R-24) Because there will be no steam or steam generators these two monitors will not be maintained operable. -49n-Revision 2

TABLE 3.2-3 (Continued) TECKMICAL SPEC IF ICATIOM MUHBER TITLE ADDITIOMAL IMFORMATIOM COHMEMT 3.3.3.10 Instrunentaticn - Radioactive Gaseous Process and Effluent Honitoring Instrunentation, Specifically the following surveillance requirements: 4.3.3.10.2, Za Condenser Evacuation System Keble Gas ActivityHonitor (SRA-2905) 4.3.3.10.2, 2b Condenser Evacuation System Effluent FLow Rate (SFR-401, 2-HR-054, SRA-2910) 4.3.3.10.2, 6a Gland SeaL Exhaust Keble Gas Activity (SRA-2805) 4.3.3.10.2, 6b System Effluent Flow Rate (SFR.201, 2-HR-054, SRA.2810) Because there will be no steam or stean generators these eight monitors wiLL not be maintained operable. 3.4.7 Reactor Coolant System - Chemistry This technical specif ication provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. During the steam generator repair there will be a period of aFproximately six months when the Reactor Coolant System wiLL be drained to half-loop, the reactor vessel head wilL be fn place and the Residual Heat Removal Punps will be shutdown. During this portion of the outage it will not be possible to obtain a chemistry sac@le from the Reactor Coolant System. Therefore the Reactor Coolant System wilL be placed within speciffcaticn limits prior to this shutdown and isolation period. Once san@ling can be reestablished following the stean generator repair it will be verified that the Reactor Coolant System is still within the chemistry limits. If the Reactor Coolant System is not within the chemistry limits, the system wiLL be cleaned-~ prior to reloading fuel into the reactor. Our engineering -49o-Revision 2

TABLE 3.2-3 (Continued) TECHNICAL SPECIF ICATION NL%BER TITLE ADDITIOHAL INFORHATION COHHENT evaluation has determined that the structural integrity of the Reactor Coolant System will not be diminished by an unlikely increase in chlorides or fluorides above the technicaL specification limits of 0.16 ppm. This is based on the Reactor Coolant System being at anhient temperature during this period and that stress corrosion cracking (SCC) does not occur below 80 F and rarely at Less than 145 F.

Also, SCC does not occur mtiL the concentration of chloride and fluoride reaches several orders of magnitude above the technical specification limit of 0.15 ppm; the level below which the Reactor Coolant System will be Left at during the period of shutdown and isolation.

3.9.1 Refueling Operations - Boron Concentration Since there will be no fuel in the reactor vessel limitations cn reactivity conditions in the reactor vessel are no Longer a concern. 3.9.2 Refueling Operations-Instrunentati on Since there will be no fuel in the reactor vessel there will be no change in the reactivity condition of the core, therefore, the source range neutron flux monitors are not needed. 3.9.8.1 Refueling Operations - Residual Heat RemovaL and Coolant Circulation Mith no fuel in the reactor vessel there will be no residual heat to remove. Therefore, there is no need to maintain an operational residual heat removaL loop. 3.9.8.2 Refueling Operations - Low Mater Level Mith no fuel in the reactor vessel there will be no residual heat to remove. Therefore, there is no need to maintain an operationaL residual heat removal loop. -49p-Revision 2

TABLE 3.2-3 (Continued) TECHNICAL SPECI F ICATION NUNBER TITLE ADO IT IONAL INFORNATION COGENT 6.5.1.6(a) Adninistrative Controls - Plant Nuclear Safety Review Comnittee-Responsibilities The PNSRC will review the following steam generator repair project docLments: 1. The Steam Generator Repair Report 2. The Steam Generator Repair Ouality Assurance Program 3. Procedures covering return to service testing. 6.8.2 Adninistrative Controls-Procedures The PNSRC will review the procedures written covering return to service testing. 6.8.3 Adninistrative Controls-Procedures Temporary changes made to procedures covering return to service testing provided items a, b, and c of technical specification 6.8.3 are satisfied. 6.12.2 Administrative Controls - High Radiation Area The keys to those high radiation areas turned over to the steam generator project team shall be maintained under the adainistrative control of the Project Health Physicist. -49q-Revision 2

The replacement lower assemblies will be transported to the Donald C. Cook Plant by barge/railroad combination. They will be barged to Mt. Vernon,

Indiana, where they will be transferred to railroad cars for transportation by rail to the plant.

The lower assemblies will be drained, dried and sealed prior to shipment. A nitrogen blanket will be maintained on the primary and secondary side during shipment and storage. During transportation the assemblies will be supported on the barge/car deck on specially fabricated

saddles, tied down by cables and restrained by end braces secured to the deck.

3.3.3 Modification to Auxilia Buildin Structural Steel To handle the loads associated with the Steam Generator Repair Project, the existing Auxiliary Building overhead bridge crane will be upgraded to single-failure-proof status and a second 150/20 ton single-failure-proof overhead bridge crane will be installed in the Auxiliary Building. Both cranes will travel on the existing rails, which extend the length of the auxiliary building, while carrying loads approaching 250 tons (see Section 6.2.1 and Supplement 1 of this report for a detailed description of the cranes and the load handling methodologies). Each crane rail is supported by a crane rail girder which in turn transfers the crane load to the auxiliary building structural steel columns. An analysis was performed to ensure the integrity of the existing auxiliary building structural steel elements which support the crane loads. The analysis was performed assuming both cranes operating in tandem while moving a 300 ton load. The results of the analysis shows that the existing auxiliary building structural steel is adequate to support the crane loads with minor modifications. Revision 2

This Space Left Intentionall'y Blank 3.3.4 Polar Crane Powe Circuit Re ocation Approximately 200 feet of the polar crane power supply cable is located in the cut area of a Unit 2 steam generator doghouse enclosure wall. To eliminate this cable as a cut interference and at the same time provide maximum availability of the polar crane, the cable will be permanently rel'ocated prior to the start of the steam generator repair project. The entire cable, from the containment penetration connection up to the crane, will be replaced to avoid splicing. The rerouted new cable is of approximately the same length as the existing cable and therefore will not significantly increase the permanent combustible fire loading in the containment building. The rerouted cable will be mounted to the walls per Seismic Class I requirements. 3.4 Post Shutdown Activities 3.4.1 Conta nment Pre a ations 3.4.1.1 Reactor Vessel Prior to the start of repair project the reactor will be defueled. The upper internals wi'll be returned to the reactor vessel and the reactor vessel head reinstalled, The missile shields will be reinstalled and a heavy steel work

55a-Revision 2

platform will be assembled over the refueling cavity. Lay-up procedures to insure reactor vessel cleanliness, prevent foreign objects from entering the reactor vessel, and minimize corrosion of the reactor coolant system will be developed. 3.4.1.2 Polar Crane The polar crane is equipped with a 250-ton capacity main hoist and 35-ton auxiliary hoist mounted on a single trolley. The polar crane possesses sufficient capacity to handle all major lifting requirements for the steam generator project inside containment and can be rerated to a higher capacity as required; however, rerating of the hoists is not anticipated'. -55b-Revision 2

Some circuits of the following systems will be temporarily disconnected and/or removed: o Fire Detection Communication Steam Generator Process Instrumentation Containment Ventilation Fuel Handling Hydrogen Recombiner 600 V Non-Ess Dist. & 120/208 V Lighting o Seismic Instrumentation Equipment determined to be essential during the Steam Generator Repair Project will be relocated, and/or its cable,

conduit, and cable trays will be re-routed as required to maintain the equipment in proper operating condition.

3.4.2.7 Heating, Ventilation and Air Conditioning Ductwork Ductwork in the removal pathway will be removed or temporary relocated. Duct I pieces removed will be cleaned, marked snd placed in temporary storage outside containment until needed for reinstallation. 3.4.2.8 .Steam Generator Insulation The existing steam generator metallic insulation will be reused. The outer dimensions of the replacement steam generators duplicates the original steam generators, although some insulation sections will require modifications to accommodate the additional hand holes and inspection ports. Sections of insulation shall be removed,

cleaned, wrapped in plastic bags and stored in strong tight containers.

These containers will be stored outside containment off the ground and protected from the weather. Sequence of removal and storage location will be documented to facilitate installation. Those Revision 2

sections requiring modifications will be stored separately to allow rework prior to installation. The original equipment supplier, Diamond Power Speciality Corp., will provide procedures and technical supervision for insulation removal,

storage, modifications and installation.

3.4.2.9 Seismic Restraints Removal The steam generator snubbers will be removed to provide access for handling 1 and movement of the steam generators. In addition, the pipe whip restraint at the main steam pipe will also be removed. Removal and storage of the snubbers and restraints will be in accordance with approved procedures and/or specifications. Snubbers are periodically removed for, ISI testing and off-site disassembly and inspection by an independent laboratory. Removal and reinstallation procedures will be similar to those 3.4.2.10 Fire Sensors Thermistor cable tray fire sensors will be pulled back where they extend beyond removed cable tray sections. These sensor circuits will remain in service during the steam generator project and will be reinstalled in accordance with approved procedures. 3.5 Steam Generator Removal Activities 3.5.1 Steam Generato Cuttin ethods and Locations 3.5.1.1 Feedwater and Main Steam Line Piping Cuts The feedwater and main steam lines will be mechanically cut in two places. The location of the cuts, the equipment to be used, and the method of cutting Revision 2

After the lifting assembly is installed, the crane shall take the weight of the lower assembly while the lower assembly is still supported by the temporary lateral support and the steam generator support columns. The temporary lateral support will be removed and the lower assembly then lifted slightly off its support columns. The lower assembly shall be raised until the lifting assembly is approximately 2'-0" below the underside of the steam generator doghouse enclosure roof and then moved horizontally until it is within approximately 6 inches of the opening in the steam generator doghouse enclosure wall. It will be lifted again until the bottom of the lower assembly clears the horizontal wall cut. V It will then be moved horizontally out of the steam generator enclosure. After clearing the steam generator doghouse enclosure a downending fixture will be attached to the steam generator lower assembly and it will be lowered onto a set of low profile saddles. After the lower assembly has been secured to the saddles and the saddles have been placed on rollers, the upper assembly will be winched through the equipment hatch. Once the lower assembly is through the Unit 2 equipment hatch and. resting on the transport deck in the auxiliary building between the Unit 1 and Unit 2 equipment hatches, it will be attached to the tandem auxiliary building bridge cranes. The lower assembly will then be lifted, rotated and moved. in a southeast direction until it has passed the southwest corner of the spent fuel pool. After the lower assembly has passed by the southwest corner of the spent fuel pool it will be oriented in an east-west direction and moved to the eastern edge of the elevation 650'loor. At the eastern edge of the elevation 650'loor, the lower assembly will be moved out into the railroad bay and oriented in a north-south direction, lowered to the 609'levation and secured to a wheeled transporter. The lower assembly will then be transported Revision 2

LE 3.6-1 STEAH GENERATOR REPAIR MELDS MELO HATERIAL OUTSIDE MALL DIA.1 IN. IN. JOINT PROCESS2 F ILLER3 HINIICH PREHEAT POSTNEAT oF oF MELD F INISN NDE4 Feedwater Nozzle to elbow SASOS,C1-2 16 SA234,MPB 0.843 Single V 35-40O without backing ring (fIat root) GTAM-r SHAM-f ER70S-2 E7018 250 1100-1200 1 hr Above 600 heat & cool 400/hr Grind to remove weld ripple RT 1/3 fillplus RT final & HT Elbow to reducer or SA234,MPB 16 0.843 Single V with SHAM backing ring GTAM cap E7018 ER70S-2 50 1100-1200 1 hr Above 600 heat & coot 400/hr As welded RT, HT Reducer to Pipe SA234,MPB SA106,8 14 0.705 Single V with SHAM backing ring GTA'M cap E7018 ER70S-2 50 None As welded RT, HT Liner SA106,8 12-3/4 0.5 Single V GTAM-r SHAM-f ER70S-2 E7018 50 None As welded VT-1, RT 2 SG Vessel Transition cone to plate SA50S,C1-3 175-3/4 3.62 Double V to SA533, A, modified C1-1 backgouge SHAW-r &SAW, GHAM or FCAM E9018-H or 250 ESOIS-C3 & matching wire E81Ni1 1100-1200 2 hr 30 m Above 800 heat & cool 110/hr Grind for UT exam RT,UT, & PT or HT Mra~r plate and misc. non press comp SA285,C 124.25 3/8 Single V w/o SHAM or backing GTAM (flat root) E7018 ER70S-2 50 None Grind flush HT Hisc Coraections o Slowdown o Drain o Level SA508,C1-1a to SA105, red. to 2.5 to SA206,B pipe 2 SA106,B 1 SA106,8 3/4 >1-1/4 Socket SHAM or GTAM E7018 ER70S-2 50 As welded HT

'ABLE 3.6-1 (cont'd.) STEAH GENERATOR REPAIR MELDS MELD HATERIAI. CUTS IDE MALL DIA.1 IN. IN. ~OINT PROCESS2 FILLER3 HINIISH PREHEAT POSTKEAT OF OF MELD FINISH NDE4 Hain Steam o Nozzle to elbow or elbow to elbow o Reducer to elbow or pipe SA508, C1-2 32 to SA234, MPB

SA234, MPB or 32/30 SA155, C1-1, Gr-KC70 1-1/8 Single V 35 40o backing ring SHAM B/R with GTAM cap 1-1/8 Single V SHAM B/R 35.40o with GTAM backing ring cap E7018 E7018 250 250 1100.1200 1 hr 15 m Above 600 heat 8 cool 1100.1200 1hr15m Above 600 heat S cool 350/hr As welded RT, HT RT, HT Reactor Coolant Elbow to SG nozzle

SA351, CFBH to E308 weld overlay on carbon steel 31 ID 2.88 Single V flat root GTAM-r SHAM-f or Auto GTAM/GHA'M-f ER308 E308 ER308 50 None Grind 8 polish with 360 grit or finer RT,UT,PT 1.

Outside diameter except as noted. 2. Melds shall be made and qualified in accordance with the requirements of ASHE Code Sections III and IX. 3. Meld filler metals and electrodes to be ordered in accordance with ASHE Code Section II, part C. Austenitic stainless steel to meet delta ferrite requirements in ASHE Code Section Ill, NB-2433. Covered electrodes to meet analysis tests of ASHE Code Section III, NB2420. 4. NDE to be in accordance with ASHE Section V with acceptance standards in accordance with ASHE Code Section III.

In addition, a Plant/Project interface document shall be implemented to define areas of responsibility, communications,

control, and interface between the Project Radiation Protection/ALARA Group and the Plant Radiation Protection Section.

Regular'eetings between members of these two groups will be held to insure adequate communications and dissemination of information. -82a-Revision 2

o No changes are expected due to differences in initial conditions (zero load steam temperature and pressure are identical for the unit with repaired steam generators). The no load steam generator mass decreases insignificantly (-2.0 percent). Therefore the conclusions of the existing steam line break analyses remain valid for the repaired steam generators. 6.1.2.5 Steam System Piping Failures Refer to Section 6.1.2.4 for discussion that applies to this accident as well. 6.1.2.6 Loss of External Load Donald C. Cook Unit 2 is designed to have full load rejection capability, and a reactor trip may not occur following a loss of external load. It is expected that steam dump valves would open in such a load rejection, dumping steam directly to the condenser. Reactor coolant temperature and pressure do not significantly increase if the turbine bypass system and pressurizer pressure control system are functioning properly. If the steam dump valves do, not operate, the reactor will trip due to high pressurizer pressure

signal, high pressurizer level signal, or overtemperature T signal.

Primarily to show the adequacy of the pressure-relieving devices and to demonstrate core protection margins, the Donald C. Cook FSAR and analysis of record analyze cases where the steam dump valves do not operate, and there is no direct reactor trip due to a turbine trip. It is shown in the FSAR and the analysis of record that the accident criteria on system pressure and DNB are not violated in any of the loss-of-load cases. -144-Revision 2

An accident involving the dropping or tipping of the steam generators during the removal process is considered highly unlikely because of the strict controls which will be placed on the movement process. In the unlikely event that an accident involving the steam generators does occur, our reviews have determined that the only potential interactions with shared systems of significant concern involve the spent fuel pool cooling equipment located in the vicinity of the load path.

However, the slight potential for damaging spent fuel pool cooling equipment is not considered to represent an unreviewed safety question as defined in 10 CFR 50.59.

This conclusion is based on the various malfunction analyses presented in Chapter 9.4 of the FSAR. These analyses conclude that it is not possible for a piping failure to cause I drainage of the pool below the top of the stored fuel elements. In the event all. cooling for the pool is lost, it would take a minimum of 8 hours for the temperature in the pool to reach 180oF (which still allows 32oF margin to boiling). Thus, sufficient time exists to either restore cooling capability or replace water which could be lost through boiloff to prevent damage to the stored fuel elements. 6.3 Fire Protection Evaluation \\ The effect of a Unit 2 construction fire was evaluated by assuming that the equipment in the Unit 2 containment and Auxiliary Building fire areas directly affected by construction activities would be damaged. Loss of all equipment in the combined fire areas would not cause loss of Unit 1 safe shutdown capability. The fixed combustible loading of these fire areas will not be significantly affected by construction activities. Transient combustible loading in the construction areas will increase beyond the levels assessed in the Fire Hazards Analysis for normal conditions.-163-'evision 2

The Safe Shutdown Capability Assessment exemption requests and fire barrier evaluations for the affected fire areas were reviewed to assess the impact of increased fire loadings and fire hazards due to construction activities. Construction activities were determined not to impact the validity of these evaluations with respect to Unit 1 shutdown capability provided there is no continuity of combustibles, such as wooden temporary stairs and trash chutes, between the Crane Bay and the 650'levation of the Auxiliary Building which could promote rapid fire spread. Temporary stairs and other structures connecting these elevations will be made primarily of non-combustible materials or compensatory measures will be provided. The repair contractor will operate under existing plant procedures and administrative controls with the exception of areas turned over to his direct control for construction,

access, and laydown.

The repair contractor will prepare a fire protection program to govern work activities in the areas under his control which will be designed to minimize construction fire hazards. 6.4 Analysis of Significant. Hazards Consideration This section presents, pursuant to 10 CFR 50.91, the analysis which sets forth the determination that the Steam Generator Repair Project does not involve any Significant Hazard Consideration as defined by 10 CFR 50.92. In addition to the appraisal on the significant hazards issue using th' standards in 10 CFR 50.92, which are presented below, it is important to note that the Steam Generator Repair Project proposed by I&NECo involves practices that have been successfully implemented at two other commercial nuclear power

plants, namely, the steam generator repairs completed by the Virginia Electric

-163a-Revision 2

and Power Company for the Surry Power Station and by the Visconsin Electric Power Company for the Point Beach Nuclear Plant, Unit 1. The repair project is also similar to the repair projects conducted by the Carolina Power and Light Company for the H. B. Robinson Steam Electric Plant, Unit No. 2 and by the Florida Power and Light Company for the Turkey Point Plant Units 3 and 4. 6.4.1 Criterion 1 Involve a significant increase in the probability or consequences of an accident. The Steam Generator Repair Project does not affect the probability or consequence of an accident. The probability or consequence of an accident is determined by the design and operation of plant systems. The repair project involves the replacement of the Donald C. Cook Unit 2 Steam Generator Lower Assemblies. Due to the almost identical design of. the replacement lower assemblies the repair of the Donald C. Cook Unit 2 steam generators is a replacement in kind and will not change the design or operation of plant systems. Thus, this repair does not involve a significant increase in the probability or consequences of an accident previously evaluated. 6.4.2 Cr terion 2 Create the possibility of a new or different kind of accident from any accident previously evaluated. The possibility of a new or different kind of accident is not created by the repair to the Donald C. Cook Unit 2 steam generators. All components and piping will be reinstalled to meet the original design and configurations and installation requirements. Therefore, because there will be no changes to the plant and plant systems design no new or different accidents are created. Revision 2

6.4.3 Criterion 3 Involve a significant reduction in a margin of safety. Section 2.2 of this report illustrates that, although certain design enhancements have been

made, the steam generator repair will result in very little change to the original operating parameters.

Therefore, the impact on the accident analysis, as shown in Section 6.1 will be insignificant and there will be no significant resolution in the margin of safety. -165-Revision 2

7.3.2 Re ional Historic Archeolo cal A hitectural Scenic Cultural and Natural Features No known historic, archeological, architectural or natural resources exist on the portion of the plant site affected by the Steam Generator Repair Project. The access road used during plant construction parallels the beach and will be used for light construction traffic during the repair project. This traffic may pose an aesthetic impact to individuals using the beach for recreation, however, this is a temporary impact that will end with the completion of the repair project. 7.3.3 ~Hdrolo 7.3.3.1 Ground Water No impact to the site ground water is expected to occur as a result of the Steam Generator Repair Project. 7.=3.3.2 Surface Water No impact to the surface water associated with the plant site is expected to occur as a result of the construction phase of the Steam Generator Repair Project. In addition, the repaired steam generators will have essentially the same amount of blowdown discharged during operation as do the original steam generators and it is anticipated that there will be no changes to the plant NPDES permit. Revision 2

7.3.4 ~Geolo There will be no geological impacts as the result of the Steam Generator Repair Project. Excavation,

grading, and compaction will occur in limited amounts and these actions will occur in areas previously disturbed (i.e. parking lots,

roadways, and laydown areas).

7.3.5 ~Ecolo 7.3.5.1 Terrestrial Ecology There will be 'no impacts to the terrestrial ecology surrounding the - plant site for the following reasons: o No habitat will be removed as a result of the Steam Generator Repair Project since all activities related to the repair project will occur on previously disturbed area (i.e. existing access

roads, parking lots, laydown area.

o Since the area affected is already subjected to the intrusion of man and machinery (i.e. security patrols, existing security lights, and normal plant operations), animals residing in the areas adjacent to the construction related activities should not be disturbed by the increased activity. 7.3.5.2 Aquatic Ecology As discussed in Section 7.3.3.2 neither the construction phase of the Steam Generator Repair Program or the operation of the repaired steam generators will impact the.aquatic ecology associated with the plant site. -171-Revision 2

TABLE 7.4-1 DONALD C. COOK PER UNIT AVERAGE ANNUAL MAN-REM EXPENDITURES YEAR Exposure Man-rem 1980 246 1981 327 1982 321 1983 283 1984 1985 448 1986 336 -173-Revision 2

TABLE 7.4-6 COMPARISON OF GASEOUS EFFLUENT RELEASES FROM DONALD C. COOK NUCLEAR PLANT Radioactive ~seciee Noble gases Iodines Particulates Tritium Average 1985 Release/Unit C 2.47 x 103 6.46 x 10 2 3.72 x 10 2 10.8 Estimated Release During the SG Repair Effort C Negligible 6.9 x 10-6(1) 2.92 x 10 4 Negligible otes (1) Estimated from Surry Unit 2 Data. -183-Revision 2

7.9 Environmental Controls The following environmental controls shall be utilized to minimize the environmental impacts associated with the steam generator repair program. These environmental controls shall be reviewed by the contractor prior to the start of work. In addition; it is recommended that these environmental controls be included as part of the contractor work specifications. 7.9.1 Noise To reduce the impact of noise on the surrounding community, the majority of the construction activities involving the use of heavy machinery will take place only during the day shift. If second shift construction activity involving heavy machinery must occur, it will end by 9:00 p.m. Noise from internal combustion engines will be controlled by the use of exhaust mufflers ~ 7.9.2 Limitations o a h e Movement No machinery will be allowed to operate in areas not previously disturbed by construction activities. If areas not previously disturbed are inadvertently impacted by machinery, it will be the responsibility of the contractor operating the machinery to restore the disturbed area to its original state. 7.9.3 Handlin and Sto a e of Oil and Pollutin Materials The handling and storage of oil and polluting materials will be conducted in accordance with the D. C. Cook, "Oil Spill Prevention Control and Countermeasure Plan," and the D. C. Cook, "Pollution Incident Prevention Plan." -202-Revision 2

7.9.4 Environmental Monitorin Periodic inspections of the construction activities will be conducted. If any of the construction activities appear to be causing significant environmental

impacts, appropriate actions will be taken.

7.9.5 Permits A list of State and local permits needed to begin construction activities at D. C. Cook will be developed by the D. C. Cook Environmental Section and the AEPSC Radiological Support Section. The AEPSC Radiological Support Section will be responsible for obtaining the required permits. 7.10 Conclusion It is concluded that with the proper mitigation practices as outlined in the Environmental Controls Section of thi's report, no significant adverse environmental impact will result from the proposed activity, that there are no preferable alternatives to the proposed action and that the impacts associated with the repair program are outweighed by its benefits. It is further concluded that the site preparation work, as described in Section 3, does not involve an unreviewed environmental question pursuant to Part II, Section 3.1 of the Donald C. Cook Plant Environmental Technical Specifications. -203-Revision 2

D. C. COOK PLANT UNIT NO. 2 STEAM GENERATOR REPAIR REPORT SUPPLEMENT 1 TABLE OF CONTENTS SECT ON TITLE PAGE 1.2 GENERAL EVALUATIONS 1-3 1-3 1.2.1 1.2.2 1.2.3 1.2.4 1.2.5 1.2.6 Crane Manufacturer and Design-Rated Load Comparison to NUREG-0554 and NUREG-0612 Seismic Analysis Lifting Beams Interfacing Lift Points Monorail Hoist 1-3 1-3 1-15 1-18 1-19 1-19

1.3 CONCLUSION

1-19 TABLE 2.2-1 2.2-2 LIST OF TABLES ~TI LE 150-Ton Capacity Single-Failure-Proof Crane Design Factors Steam Generator Repair Project Auxiliary Building Crane'ifts Over 60 Tons PAGE 1-5 1-7 LIST OF FIGURES FIGURE TITLE PAG 1.2-1 Mathematical Model of Crane Trolley at Mid Span 1-20 1-2 Revision 2

design, fabrication, inspection, testing and operation as delineated in NUREG-0554 and supplemented by NUREG-0612. This evaluation is presented in the form of a point-by-point comparison to NUREG-0554'his point-by-point comparison was developed by AEPSC and Whiting Corporation. The new crane will meet all applicable sections of CMAA Specification ¹70, Revision 75 and ANSI B30.2.0 - 1967. For ease in making a point-by-point comparison the following section numbers correspond to the section numbers in NUREG-0554: 2. C T 2.1 Construct on and 0 eratin Per ods Since the Donald C. Cook Nuclear Plant is an operating

plant, the construction portion of this section is not applicable.

For the repair project and subsequent operating period the new crane will be designed per CMAA ¹70, Revision 75. Dynamic loads are considered due to load accelerations associated with a 150-ton load but not seismic loadings. Simultaneous static and dynamic loading will not stress the equipment beyond the material yield. 2.2 Maximum Crit cal Load Since the new crane will be operating indoors, degradation due to exposure will not be considered a factor in the crane design.

However, items subject to wear will have an additional design factor applied to them (see Table 2.2-1 of this supplement).

'1-4 Revision 2

2.2 Maximum Critical Loads (cont'd.) The crane is being designed per CMAA ¹70, Revision 75 for dynamic loads due to the load accelerations associated with 150 ton load. Considering dynamic loads due only to load acceleiations, the maximum critical load is 150 tons the same as the design rated load.

However, as presented in the preliminary seismic analysis discussion, Section 1.2.3, when dynamic loads due to a seismic event (safe shutdown earthquake) are applied to the crane the maximum critical load is 60 tons.

A maximum critical load of 60 tons is sufficient for all but 24 lifts associated with the repair project. Because these 24 lifts are one time only special lifts the provisions of NUREG-0612 Section 5.1.1(4) will apply. This section states that for special lifts, loads imposed by the safe shutdown earthquake need not be included in the dynamic loads imposed on the lifting device. Therefore, for these 24 special lifts the maximum critical load will be the same as the design rated load of 150 tons. The design rated load and the maximum critical load will be marked on the crane. 1-6 Revision 2

TABLE 2.2-2 STEAM GENERATOR REPAIR PROJECT AUXILIARYBUILDING CRANE LIFTS OVER 60 TONS Itery Est. Wt. ~owns Number ~Lff s Steam Generator Concrete Doghouse Front Roof Section 70 Steam Generator Concrete Doghouse Back Roof Section 60 Old Steam Generator Upper Assembly 112 Old Steam Generator

• Lower Assembly 247 New Steam Generator
• Lower Assembly 240 Refurbished Steam Generator Upper Assembly 112 24 Total
• These lifts will be made using the upgraded existing crane and the new crane in a tandem configuration.

1-7 Revision 2

TABLE 2.2-2 STEAM GENERATOR REPAIR PROJECT AUXILIARYBUILDING CRANE LIFTS OVER 60 TONS Item Est. Wt. ~Tons Number Lifts Steam Generator Concrete Doghouse Front Roof Section 70 Steam Generator Concrete Doghouse Back Roof Section 60 Old Steam Generator Upper Assembly 112 Old Steam Generator

• Lower Assembly 240 New Steam Generator
• Lower Assembly 247 Refurbished Steam Generator Upper Assembly 112 24 Total
• These lifts will be made using the upgraded existing crane and the new crane in a tandem configuration.

1-7 Revision 2

2.3 0 eratin Environment Since the crane will be operated in the auxiliary building the crane will not be subjected to design basis accident type changes in pressure, temperature, humidity or exposed to corrosive or hazardous conditions. Therefore, such considerations have not been included in the design of the crane. The ranges of temperature,

pressure, and humidity anticipated for crane usage are as follows:

Temperature: Ambient temperature inside the auxiliary building with seasonal variations between winter and summer. Pressure: Ambient pressure except during refueling outage activities, when slightly negative pressure ( ) 1/8 inch w.g.) will be maintained as required by Technical Specification 4.9.12.d.4. Humidity: This could range from a minimum of 0% to a maximum 100%. Material Pro erties 2.5 In addition to impact testing requirements on the main hook, structural members essential to structural integrity and greater in thickness than 5/8 inches are fabricated of impact tested material in accordance with the Section III of the ASME code. The minimum operating temperature of the crane will be established by the crane manufacturer. Any necessary steps to prevent operation of the crane below the minimum operating temperature will be taken. In

addition, low alloy steels are not used in the fabrication of the crane, and cast iron is restricted to non-load bearing components.

Seismic Desi n 2.6 See Section 1.2.3. Lamellar Tearin The main bridge girders and structural load support members of the trolley, specifically those members supporting the critical load, are fabricated from structural plate. Welded, rolled structural shapes are not used for these members. Moreover, weld joints associated with the structural members within the main hoist load path are typically oriented such that the induced stresses will not be manifested in lamellar tearing at the weld zone. All weld joints whose failure could result in the drop of a critical load will be nondestructively

examined, If any of these weld joint geometries would be susceptible to lamellar tearing, the base metal at the joints will be nondestructively examined.

1-8 Revision 2

2.7 St uctura ti ue As stated in Section 2.1, the crane will not be used for plant construction lifts. The allowable stress range for the fatigue design of this crane is higher than the normal design allowables of Crane Manufacturers Association of America (CMAA) Specification No. 70-1975. As a result, a fatigue analysis will not be performed, since it is not a governing factor in design of the crane. 2.8 Weldin Procedures Welding, welding procedures (pre heat, post weld heat treatments), and welder qualifications are in accordance with AWS Dl.l "Structural Welding Code."

Further, low-alloy materials will not be used in the main load support structure.

SAFETY FEATURES 3.2 Auxiliar S stems The auxiliary 20 ton hoist is of single-failure-proof design. Where dual components are not provided within either hoist mechanical load path, redundancy is provided through an increased design factor on such components as required per NUREG 0612. 3.3 Elect ic Cont ol S stems Limit controls are incorporated to minimize the likelihood of inflicting damage to the hoisting drive machinery and structure that otherwise might occur through inattentive and/or unskilled operator action. An emergency stop button will be added to the radio remote control unit.that will interrupt the power supply to the crane and stop all crane motion. 3.4 Eme enc Re a s This crane is designed so that, should a malfunction or failure of controls or components occur, it will be able to hold the load while repairs and adjustments are made. HOISTING MAC NERY 4.1 Reevin S stem The static-inertia design factor of the wire rope, with all parts in the dual system supporting the DRL is 11 to 1. Such conservative design more than surpasses requirements to sustain the dynamic effects of load transfer due to the loss of one of the two independent rope systems with an ample design margin remaining in the 1-9 Revision 2

six parts supporting the load. The maximum load (including static and inertia forces) on each individual wire rope in the dual reeving system with the MCL attached will not exceed 10$ of the manufacturer's published breaking strength. Compliance to this recommendation requires high alloy rope. By definition, reverse bends do not exist in the reeving system of the main hoist. Studies have been conducted to establish the effects of reverse bend on fatigue life. In consideration for the geometry of wire rope (helix) construction, unless the distance between th'e sheaves in the load block and head block are under one lead of the wire rope, a reverse bend cycle is not incurred.

Moreover, the ratio of rope to sheave diameter in the only qualifying area of the hoist mechanism is related to the drum, which is 30 to 1; 125%

of minimum requirement per CMAA Spec. ¹70, Rev. 75. The pitch diameter of running sheaves and drums shall be in accordance with CMAA Spec. ¹70, Rev. 75. All fleet angles within the main hoist reeving are within the recommended 3 1/2 degrees. The crane is equipped with an equalizer beam/fixed sheave arrangement that provides two separate and complete reeving systems. Protection against excessive wire rope wear and fatigue damage will be ensured through periodic inspection and maintenance. 4.2 Drum Su ort The indicated drum support provisions are included in the design which, as required, would insure against disengagement of the drum from its braking control system. 4.3 Head and Load Blocks Both reeving systems associated with this crane are -designed with dual reeving. This design will ensure the vertical load balance is maintained. Each load-attaching point (sister hook and eye bolt) is amply designed to sustain 200% of the 150-ton DRL. The overhead crane shall be load tested at 125% of the 150-ton DRL. Nondestructive examination of the sister hook and eye bolt will be performed. After successful completion of the load test, a complete inspection of the crane, including a nondestructive examination of the sister hook and eye bolt, will be performed. 4.4 Hoistin S eed The main hoist full rated load speed of approximately 4.5 FPM is less than the suggested operating speed in the "slow" column of Figure 70-6 of CMAA specification ¹70. Revision 2

Further, the rope line speed at the drum at approximately 27 FPM is considered to be conservative.

Desi n A ainst Two-B ockin The main hoist is equipped with two independent travel limit control devices in addition to a load sensing

system, as suggested, to insure against two-blocking.

Actuation of hoist travel limit switches or load sensing devices will deenergize the hoist drive. In addition, the mechanical holding brake will have the capability to withstand the maximum torque of the driving motor, 4.6 Liftin Device The lifting beams and other devices attached to the crane hook block will be designed to have factors of safety based on guidelines noted in NUREG-0612 and NUREG-0554. Each device will be able to support a load of three times the load (static and dynamic) being handled without permanent deformation as recommended in Section 4,6 of NUREG-0554. 4.7 Wire Ro e Protection Operation of the hoist is only to be attempted with the trolley and block aligned over the center of the load for a vertical lift. 4.8 Mach ne Ali nment The provisions of this paragraph are incorporated in the design of the overhead crane. 4.9 Hoist B akin S stem The provisions of this paragraph are incorporated in the design of the overhead crane. 5. 5.1 BRIDGE AND TROLLEY Brakin Ca ac t The bridge and trolley drives will each be provided with an appropriately sized electric holding brake which, upon interruption of power, is applied whether through operator action or violation of travel limit provisions on the trolley and restrict area limit controls for the bridge.

Further, these brakes are capable of being operated manually.

The AC induction-motors and magnetic controls utilized for these drives are not prone to an overspeed condition, which is attributed to inherent operating characteristics. Therefore, overspeed limit controls for the bridge and 1-11 Revision 2

trolley motion equipped with this type of drive would represent a needless feature. Moreover, the motor controls are provided with adequate overload protection. The mechanical drive components are designed to sustain maximum peak loadings capable of being transmitted by either the motor or brake under all attitudes of normal crane operation. All other recommendations of this section are compatible with the design of the crane. 5.2 Safet Sto s As stated in Section 5.1, an overspeed condition considering the type of drive used for the bridge and trolley is not a concern with this equipment. Appropriately designed and sized bumpers and stops are provided in accordance with CMAA Spec. @70 Rev. 75 and are adequate to absorb the energy of the trolley and bridge in the event of limit switch malfunction. 6. R V AN CO TROLS 6.1 Driver Selection The main hoist motor was selected on the basis of hoisting the design-rated load (150 tons) at the design hoisting speed. Further, all proper and due consideration was given to the design of related-mechanical and structural components to adequately resist peak torques transmitted by this motor within normal design limits. Hoist overspeed and overload sensing-limit control provisions have been incorporated to guard against, such occurrences. Additionally, the hoist holding brakes are capable of controlling the design rated load within the 3 inches.(8 cm) specified stopping distance. In addition, emergency power disconnect switches will be located at operating floor level to interrupt power to-the crane independent of the crane controls. Since the MCL is less than the DRL, administrative controls will be established to reset the overloading sensing device. 6.2 Drive Control S stems The design considerations discussed in this section have been addressed and incorporated as appropriate except for the restriction of simultaneous operation of motions. The crane is not used to handle spent fuel assemblies. 6.3 Malfunction Protectio Features to sense, respond to, and secure the load in the event of hoist overspeed, overcurrent,

overload, over Revision 2

travel, and loss of one rope of the dual reeving system have been incorporated.

6.4 Slow S eed D ives Features recommended in this paragraph will be incorporated as part of the motion control circuitry. 6.5 Safet Devices Each hoist is equipped with two independent hoist overtravel limit controls. 6.6 Control Stations Since this crane is not equipped with a cab, the complete operating control system and emergency controls for the crane will be located on a radio remote control unit. In

addition, as stated earlier emergency power disconnect switches will be located at operating floor level to interrupt power to the crane independent of the radio remote control unit.

Since the design rated load is greater than the maximum critical load, administrative controls will be established to ensure that the resetting of the overload sensing device is properly conducted. 7. NSTAL TION INSTRUCTIONS 7.1 Genera Complete operation, maintenance, installation and testing instructions will be provided for the overhead crane by the crane manufacturer. 7.2 Const uction and 0 eratin Per ods As discussed in Section 2.1 this crane will not be used for plant construction. The crane will be designed for Class A-1 service as defined in CMAA Specification ¹70, Revision 75. The allowable design stress limits will not be exceeded during the repair project. During and after installation of the crane, the proper assembly 'of electrical and structural components should be verified. 8.- TES ING A D REVENTIV IN N CE 8.1 General A complete check will be made of all the crane's mechanical and electrical systems to verify the proper installation and to prepare the crane for testing. 1-13 Re~ision 2

Proof-testing of a subcomponent is an independent verification of the subcomponent's ability to perform. The main hook block and eye bolt of the hook block assembly will be tested at 200% of the design-rated load (DRL). Before and after this test, the hook and eye bolt will be subject to nondestructive examinations. The wire rope supplier will test a section of wire rope by subjecting it to an overload condition until breaking occurs. No other components of the crane shall be proof-tested. Upon successful completion of the above proof tests, the overhead crane will be tested at 125% of the DRL. This test will ensure the ability of the crane and its subcomponents to perform their intended function. 8 ' Static and D amic Load Tests The overhead crane will be tested after installation by means of a no-load test and a 125% capacity load test. The no-load test consists of operating each crane motion to its extreme travel limit without a load on the hook. During the no-load test, the crane bridge shall travel the entire length of the runway, the top-running trolley shall traverse the crane bridge, and the hook block shall be operated through its complete vertical travel limits. Upon successful completion of the no-load test, the 125% capacity DRL test will be conducted. Each crane motion shall be engaged with the 125% DRL test load suspended from the hook.

However, due to the physical restrictions of the plant, each motion will not be operated to its full travel limit during the 125%

DRL load test. 8.3 Two-Block Test. Although the hoist is equipped with an overload sensing

device, load-anchor testing is not recommended by the crane manufacturer (Whiting Corporation).

Since Whiting customers, have followed the recommendation, there is no available information on, past load-anchor tests. The overload-sensing device will be preset and tested using a load higher than the preset load. The last sentence of Section 8.3 of NUREG-0554 states: "The crane manufacturer may suggest additional or substitute test procedures that will ensure the proper functioning of protective overload devices." Based on that provision, and per crane manufacturers'ecommendations, we are planning to perform the overload testing rather than the load-anchor test. 8.4 0 erat on Tests Whiting's standard procedures require a no-load running test before shipment. Calibration and adjustments for hoist overload and overspeed will be done after installation. Revision 2

8.5 Mai tenance A maintenance program including periodic inspections of the crane will be developed. This maintenance program will ensure that the crane is maintained at the design rated load. Both the maximum critical load and the design rated load will be plainly marked on each side of the crane. OPERATING MANUAL The operating manual supplied by the crane manufacturer will comply with Section 9.0 in its entirety, including details on preventive maintenance program items noted in the first paragraph of Section 9,0 of NUREG-0554. The existing plant procedures on the preventive maintenance program will be revised to address the above-noted items. 10. UALITY ASSURANCE The Whiting Corporation is on the Donald C. Cook Nuclear Plant Qualified Suppliers List for spare and replacement crane parts. Whiting has a QA program that complies with ANSI N.45.2-1971/NRC Regulatory Guide 1.28. This program applies also to the fabrication of new cranes for nuclear power plants. Whiting will be audited for QSL recertification in April 1987 'onald C. Cook Nuclear Procedure MHI-2071, "Qualification and Training of Crane Operators," covers qualification requirements of crane operators and will be revised as necessary to reflect the single-failure-proof features of the new crane. 1.2.3 Seismic Analysis This section presents the preliminary seismic analysis conducted to demonstrate the largest load the new crane can stop and hold during a safe shutdown earthquake. The following information provides a description of the method of analysis, the assumptions

used, and the mathematical model evaluated in the analysis.

1.2.3.1 Analysis Description The crane was analyzed to determine the effect of seismic excitations. For this analysis, the matrix displacement method was used based upon finite element techniques. The crane was mathematically modeled as a syst: em of node points interconnected by various finite elements representing straight beams. All masses and inertias were distributed among the nodes whose degrees of freedom characterize the response of the structure. The interconnecting finite elements were assigned stiffness equivalent to that of the actual structure. 1-15 Revision 2

The mathematical model represents as accurately as possible the flexibilityof the bridge girders, hoist

rope, and girder end connection.

The trolley, the drive units and the bridge trucks were represented as rigid bodies. The crane was analyzed with the trolley positioned at mid-span. This was done with loads of 50 and 60 tons in the down position. Preliminary calculations showed that this condition wculd produce the maximum girder stress for a given load. The dynamic analysis was of the mode frequency (MODAL) type, solving for the resonant frequencies and the mode shapes that characterize the crane. The modes with meaningful participation in a given direction are directly expanded by the computer program to yield the expanded mode shapes, the element stresses and the reaction values. This type of analysis is linear and plastic deformation, sliding, friction, and slack rope are not taken into account. The normal mode approach was employed for the analysis of the components. All significant eigen-values and eigen-vectors were extracted, and these modes were combined by the method specified by the U. S. Nuclear Regulatory Commission, Regulatory Guide 1.29, Rev. 1, Section 1.2.2 (Combination of Modal Responses with Closely Spaced Modes by the 108 Method). Those modes with mode coefficient ratios less than 1% in the x direction or 0.5% in the y and z directions were dropped because their contribution is proportionally small when compared to the largest mode coefficient of the related directional excitation. The results of the three orthogonal dynamic excitations were combined by the square root of the sum of the squares method (SRSS) and then absolutely added to the results of the static condition. Because the y reaction exceeds the frictional resistance of those bridge wheels that are braked, slip will occur. The maximum acceleration in the y direction will be reduced from that predicted by the modal analysis. The primary y mode was therefore reduced by a scale factor such that the resulting y reaction approaches the maximum that could be sustained before slip. The results were then resummed as previously described. In order to assure structural integrity, the job specificat'ion requires that the maximum stresses'ot exceed the minimum yield strength of the material divided by 1.5 for the OBE and 1.1 for the SSE. The crane is constructed of ASTM A36 structural steel except for components which are specifically noted in the report. A36 material has a specified minimum yield 1-16 Revision 2

strength of 36 ksi. The combined bending and axial stresses are limited to 24 ksi for the OBE and 32.7 ksi for the SSE. The actual properties of the specified materials show a great deal of variation and are generally considerably higher than the minimum required by the material specification. Also the maximum stresses occur only at a point on a section and cannot be themselves be indicative of the tendency of the section to permanently

deform, especially when the nominal stresses on the extreme fibers of the adjoining faces are significantly lower. It is therefore conservative to compare the combined bending and axial stresses at the corners with the specified allowables to assure structural integrity.

Impact factors for wheel flange to rail contact, etc., have been consider negligible. The state of the art is such that these impacts cannot rigorously be studied;

however, independent time history analyses have been run in many cases, all indicating slow relative motion between the rail and the wheel.

This is because of the time dependency of the forcing function coming from the building into the crane. Note that the only coupling through which these forces can be transmitted is dynamic friction. Upon reaching the rail the wheel will first rise through the corner radius and then contact the rail. During this period, the structure is starting to deflect as the end of the crane in this direction is flexible. The computer analysis was performed using ANSYS, a large scale finite element program. Summary of Results The crane was mathematically modeled using finite elements. On the basis of preliminary runs, the number of degrees of freedom and the significance criteria for modal expansion were adjusted. Static and three load step reduced modal runs were made and the results summed. Because slip occurs, the y excitation was proportioned and these results resummed. The crane was analyzed with the main trolley at mid-span (see Figure 1.2-1). For this position the analysis was done with 50 and 60 ton loads on the main hook in the low position. From preliminary studies, the load case considered should yield the maximum stresses in the girders. Because of the seismic acceleration a slack rope condition was found to exist under certain conditions. This cannot be truly simulated with a linear modal analysis. However our experience with time history analyses shows that a 1-17 Revision 2

modal analysis tends to produce conservative results. The rope load predicated by the modal analysis is well below the allowable rope load. When the excess dynamic rope load (that which produces a slack rope) is deducted, a small upkick is produced by the loading conditions examined. When the wheel loads parallel to the runway are compared with the vertical wheel load times the coefficient of friction, it is found that the crane bridge will tend to slide under certain loading conditions examined. This sliding is oscillatory in nature and the loadings predicted by a modal analysis are conservative. The wheel loads have been adjusted to account for frictional effects. Although some non-linearities are produced by the specified excitations the specified linear analysis will conservatively predict the behavior of the crane during a seismic excitation. The crane was found to meet the requirements for a seismic excitation with a 60 ton load on the main hook. Lifting Beams Stress levels of all load-bearing members of the lifting beam will not exceed 6,000 psi under rated load. This low stress level meets requirements of NUREG-0612 and ANSI N14.6 specifications for increased design factors for single-load-path components. Further, this design stress level qualifies for material test exemptions per Paragraph AM 218 of the ASME Boiler and Pressure Vessel

Code, Section III, Division 2, as referenced in Paragraph 3.3.6 of ANSI N14.6-1978.

Proposed lifting beam will not be subject to high amounts of radiation, 200 mili-rem/hour maximum, nor willit be submerged at any time. Based on this criteria the ~ proposed lifting beam design will not be subject to any sections of ANSI N14.6-1978 which refers to submerged duty, decontamination or radiation degradation. Application of any coating system onto the lifting beam must not violate E.P.A. codes. Under Section 6 of ANSI N14 ~ 6-1978 the main beam section and the hooks swivel are single path designed with stress levels below 6,000 psi. Since the materials for these items will have mill certification and that 100% of critical welds will undergo nondestructive examination to ensure structural integrity, these two items will not be subject to load test of three times their rated capacity. These two items will however be subjected to a 1508 load test. 1-18 Revision 2

Interfacing Lift Points Interfacing liftpoints will be dual-load-path and will be designed to shear stress levels not to exceed 4,500 psi under rated load. This design stress levels qualifies for material test exemptions per Paragraph AM 218 of the ASME Boiler and-Pressure Vessel'ode,,Section III, Division 2 as referenced in Paragraph 3.2.6 of ANSI N14.6-1978. Monorail Hoist Due to the configuration of the two cranes in the auxiliary building there will be some areas in the auxiliary building that cannot be reached by either crane. To provide access to these areas the new crane will be equipped with a 2,500 lb. capacity, fully electric hoist mounted on a fixed monorail suspended from the idler girder of the new crane. The hoist will weigh approximately 1,200 lbs. and will have a vertical liftof approximately 122 feet. Control of the hoist will be by radio remote control. The hoist is being designed to ANSI/ASME HST-4M-1985, "Performance Standards for Overhead Electric Wire Rope Hoist." CONCLUSION The new crane being purchased by the Indiana 6 Michigan Electric Company for use during the Steam Generator Repair Project has been evaluated against the criteria of NUREG-0554 and NUREG-0612. Results of this evaluation have shown that the crane being purchased meets the guidelines and criteria of NUREG-0554 and NUREG-0612 and therefore will be classified and used as a single-failure-proof crane. 1-19 Revision 2

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ATTACHMENT NO. 3 TO AEP:NRC:0894L i Letter Report MT/SMART-090(89) D. C. COOK UNIT 2 REACTOR VESSEL HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION Apri 1 1989 ~~~ ~t4-qji l~ ~ Z~~ 1 q ~Jl C~ C) !- o / WJ tel ~ LV VJ r".. -"=- e.', ~ 1 C 1 ~ 3 I ( I ' ~ 1 UJ A 0'ii () L"< r.= t."J Prepared by: o o ay Verified by: Approved by: 7 1c 0

eyer, a ager Structural Ma'terials Engineering Prepared for American Electric Power Company Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 3716s/0421N:10

ATTACHMENT NO. 3 TO AEP: NRC: 0894L

1g 0

TABLE OF CONTENTS Section Titie Page 1.0 2.0

3.0 INTRODUCTION

FRACTURE TOUGHNESS PROPERTIES 'ADJUSTED REFERENCE TEMPERATURE 4.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3 5.0 6.0 HEATUP AND COOLDOWN LIMIT CURVES REFERENCES APPENDIX A HEATUP AND COOLDOWN DATA - WITHOUT MARGINS FOR INSTRUMENTATION ERROR A-1 APPENDIX B HEATUP AND COOLDOWN DATA - WITH MARGINS FOR INSTRUMENTATION ERROR B-1 37lds/0427dQ:10

IPt

LIST OF TABLES Table Title Page D. C. Cook Unit 2 Reactor Vessel Fracture Toughness Properties Calculation of Adjusted Reference Temperatures for Limiting 0. C. Cook Unit 2 Reactor Vessel Material - Intermediate Shell Plate - C5556-2 10 Calculation of Adjusted Reference Temperatures for 0. C. Cook Unit 2 Reactor Vessel Material - Intermediate Shell Plate - C5521-2 LIST OF FIGURES Figure 1 Titie Fluence Factor for Use in the Expression for ARTNDT Page 12 D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 12 EFPY (Without Margins) 13 0. C. Cook Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 12 EFPY (Without Margins) 14 D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 32 EFPY (Without Margins) 15 0. C. Cook Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 32 EFPY (Without Margins) 37)6@/Oi2760:10 ill

~ ( ~ LIST OF FIGURES (cont) I Figure Title Page D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 12 EFPY (With Margins) 7 D. C. Cook Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 12 EFPY (With Margins) 18 D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 32 EFPY (With Margins) 19 D. C. Cook Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 32 EFPY (With Margins) 20 3716e/041169:10 iv

1 P

HEATUP AND COOLDOWN LIMIT CURVES FOR NORMA OPERATION

1. 0 INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value.

of RTNDT (reference nil-ductility temperature) for the reactor vessel. The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture tough-ness properties and estimating the radiation-induced hRTNDT. RTNDT is designated as the higher of either the drop weight nil-ductility transition .. temperature (NDTT) or the temperature at which the material exhibits at least tt t 50 ft-lb of impact energy'and 35-mil lateral expansion (normal to the major working direction) minus 60'F. RTNDT increases as the material i s exposed to fast-neutron radi ation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, hRTNDT due to the radiation exposure associated with that ~ ~ ~ ~ ~ time period must be added to the original unirradiated RTNDT. The extent of the shift in RTNDT is enhanced by, certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)~ 2.0 FRACTURE TOUGHNESS PROPERTIES t The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC'Regulatory Standard Review Plan ,The pre-irradiation fracture-toughness properties for the materials in the D. C. Cook Unit 2 reactor vessel are presented in , table 1. 37'I es/0421bQ:10

From Regulatory Guide 1.99 Rev. 2 [1] the adjusted reference temperature (ART) for each material in the beltline is given by the following expression: ART = Initial RTNDT + ~RTNDT + M rg'n Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may ~ be used if there are sufficient test results to establish a mean and standard deviation for the class. hRTNDT is the mean val ue of the adjustment in reference temperature caused by irradiation and should be-calculated as follows: [CF]f(0. 28-0 ~ 10 1 og f) [CF] [ff] NDT (2) The value, "f", used in equation (2) is the calculated value of the neutron fluence at the location in the vessel at the location of the postulated

defect, n/cm (E >

1 MeV) divided by 10 The fluence factor, "ff" is shown in figure 1. To calculate bRTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth. -,24x (depth X) surface (3) where x (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface. The attenuated fluence is then used in equation (2) to calculate hRTNDT at the specific depth. ~ ~ CF ('F) is the chemistry factor, obtained from reference 1 for the beltline region materials of the D. C. Cook Unit 2 reactor pressure vessel. The 3716s/042160:10

limiting material was found to be the intermediate shell plate C5556-2 for 0. C. Cook Unit 2 for 12 EFPY and 32 EFPY. The calculation of ART for this limiting material is shown in table 2. The ART values at 1/4T and 3/4T locations will be used to develop the reactor pressure vessel heatup and cooldown curves as described in the following sections. 4.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time. KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code The KIR curve is given by the following eqqation: KIR = 26.78 + 1.223 exp [0.0145 (T-RTNDT" + 160)] (4) where KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT* P Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code ] as follows: (5)

• NOTE:

RTN>T as used in the ASME Code [3] is in fact the adjusted reference temperature (ART) as defined in NRC Regulatory Guide 1.99, Rev. 2 [1] and calculated in section 3.0. 371bs/Oi2189:l0

f 4

where K>M = stress intensi-ty factor caused by membrane (pressure) stress K>T = stress intensity factor caused by the thermal gradients K>R = function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, K>R 'is determined by the metal temperature at the tip of 'the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calcu-lated'and then the corresponding (thermal) stress intensity factors, K<T, for the reference flaw are computed. From equation (5), the pressure stress intensity factors are obtained

and, from these, the allowable pressures are ca 1 cul ated.

For the calculation of the allowable pressure ver'sus coolant temperature during cooldo'wn, the reference flaw of Appendix G to the ASME Code [3] is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall, because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown r ates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown. rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. 37 l6s/Oi21d9: lO

, Vfy I, jY '1 '4 >5 '"0' ir> pI

During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent t'o the vessel inside surface. This condition, of course, is not true for the steady->tate situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR at the 1/4 T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the incr ease in KIR exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIR for the 1/4 T crack during heatup is lower than the KIR for the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIR's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents q lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4 T deep outside 3710s/Oll 169:l0

t I ri I

surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolan't temperature') along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate, must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup

ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the 1983 Amendment to 10CFR50 has a rule which addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RTNpT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure. 5.0 HEATUP AND COOLOOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods described in section 4.0, and the Westinghouse procedure of reference 5. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in figures 2

3, 6 and 7 for 12,EFPY and in figures 4, 5, 8 and 9 for 32 EFPY. Figures 2 ~ ~ ~ ~ ~ ~ ~ ~ ~ through 5 do not have any margins for instrumentation error and figures 6 f through 9 contain margins for instrumentation error. This is in addition to other criteria which must be met before the reactor is made critical. The leak limit curve shown in figures 2, 4, 6 and 8 represent the minimum temperature requirements at the leak test pressure specified by applicable codes The leak test limit curves were determined by the methods of references 2 and 4. Finally, table 1 indicates that the limiting flange RTNDT of 30'F occurs in the vessel flange so the minimum allowable temperature of this region is 150'F per reference 4. These limits are less restrictive than the limits shown on figures 2 through 9. Figures 2 through 9 define the limits for ensuring prevention of nonducti le failure for the D. C. Cook Unit 2 Primary Reactor Coolant System.

6.0 REFERENCES

1. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission,

May, 1988.

2. "Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981. 3. ASME Boiler and Pressure Vessel Code, Section III, Division 1-Appendixes, "Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonducti le Failure," pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York, 1986. 4. Code of Federal Regulations,

10CFR50, Appendix G, "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal

Register, Vol. 48 No. 104, May 27, 1983.

371St/Ol2760:10

5. "Procedure for Developing Heatup and Cooldown Curves," Westinghouse Electric Cor'poration, Generation Technology Systems Division Procedure GTSO-A-1.12 (Rev. 0)., July 13, 1988. 6. "Assessment of Regulatory Guide 1.99, Revision 2 Upon D. C. Cook Units 1 and 2 Adjusted Reference Temperatures and Pressure-Temperature Curves" SWRI Project 17-2544. 3716s/04216Q:10

TABLE 1 D. C. COOK UNIT 2 REACTOR VESSEL TOUGHNESS PROPERTIES (UHIRRADIATED) Component Code Ho. Material Type Cu 50 ft-lb 35 Hil T Temp(a) HDT ( F) ( F) USE RT NDT NHWD(a) ( F) (ft-lb) (ft-tb) 0 Closure Head Dome Closure Head Segment Closure Head Segment Closure Head Flange Vessel Flange Inlet Nozzle Inlet Hozzle Inlet Nozzle Inlet Nozzle Outlet Nozzle Outlet Nozzle Outlet Nozzle Outlet Nozzle Upper Shell Upper Shell Upper Shell Inter. Shell Inter Shell. Lower Shell Lower Shell Bottom Head Segment Bottom Head Segment Bottom Head Dome Inter. 8 Lower Shell Long. and Girth Held Seam 80048-2 89M3-2 A5189-2 4437-V-1 4436-V-2 269T-2 270T-1 269T-1 270T-2 271T-1 271T-2 272T-1 272T-2 C5518-2 C5521-1 C5518-1 C5556-2 C5521-2 C5540-2 C5592-1 C5823-2 A4957-3 B0018-18 (HT S3986 8 Linde 124 Flux Lot No. 0934) A533B Cl. 1 A533B CI. 1 A533B Cl. 1 A508 Cl ~ 2 A508 Cl. 2 A508 Cl. 2 A508 Cl. 2 A508 CI. 2 A508 Cl. 2 A508 Cl. 2 A508 Cl. 2 A508 Cl. 2 A508 Cl ~ 2 A533B Cl. 1 A533B CI. 1 A533B CI. 1 A533B CI. 1 A533B CI. 1 A533B Cl. 1 ~ A533B Cl. 1 A5338 Cl. 1 A5338 CI. 1 A533B CI. 1 SAH NA 0.64 -20 HA 0.66 -20 NA 0.63 10 HA 0.70 -20 NA 0.70 30 NA 0.85 -20 HA 0.9l -20 HA HA -'IO HA HA -10 NA 0.80 0 NA 0.80 0 NA NA -10 NA HA 0 0.12 0.61 10 0.14 0.59 0 0.12 0.57 10 0.15 0.57 0 0.14 0.58 10 0.11 0.64 -20 0.14 0.59 -20 NA 0.57 -10 NA 0.51 -10 NA 0.61 -50 0.06 0.97 -40 30 -3 72 5 15 -15 - 3 NA NA 12 -15 NA NA 88 93 66 118(b) 98(b) 35(b) 25(b) 45 20 9 25 -20 -20 12 -20 30 -20 -20 -10. -10 0 0 -'IO 0 28 33 10 58(b) 38(b)

-20(b)

-20(b) -10 -10 -50 -35(b) 148 143.5 140.5 239 161 201 ~ 5 239.5 NA HA >179 181 HA NA 107.5 112 > 82.5 109.5 111.5 113 107 129 149 177 NA 96 93 91 155 105 131 156 HA HA NA 117.5 NA NA 70 73 HA 90(b) 86(b) 110(b) 103(b) 84 97 115 97(b) Plan as appropriate a) Estimated per NRC Standard Review b) Actual values HA - Hot available or not applicable, KWD - Hajor Working Direction NHWD - Normal to NMD

~ fC ' f All I

TABLE 2 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING D. C. COOK UNIT 2 REACTOR VESSEL MATERIAL-INTERMEDIATE SHELL PLATE C5556-2 Parameter Re ulator Guide 1.99 - Revision 2 Chemistry Factor, CF ('F) Fluence, f (10 n/cm ) 19 2 a Fluence Factor, ff 108.4 0.47 0.79 . 108.4 0.17 0.53 108.4 1.24 1.06 108.4 0.44 0.78 ~RTNDT CF x ff ( F) Initial RTNDT, I ('F)

Margin, M ('F) 86 58 34 58 58 34 115 58 34 85 58 34

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • k'k********)t'k*'k***4**5t'kk'k)kAk'kk%

Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 178 ART = Initial RTNDT + ARTNDT + Margin 150 207 177 (a) Fluence, f, is based uPon fsurf (10 n/cm E)1 Mev) = 2.1 at 32 EFPY 19 [6] at vessel inner surface. The D. C. Cook Unit 2 reactor vessel wall thickness is 8.63 inches at the bel tl inc region. (b)

Margin, M = 2[vI

+ a> ] ', oI is the standard deviation for initial 2 2 0.5 RTNDT and a< is the standard deviation for ARTNDT. aI is assumed to be O'F for measured values of initial RTNDT and o< is 17'F for base metal [1]. 3715s/082ib9:10 10

~ \\ tk-pVi fr 4

TABLE 3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR D. C. COOK UNIT 2 REACTOR VESSEL MATERIAL-INTERMEDIATE SHELL PLATE C5521-2 Parameter Re ulator Guide 1.99 - Revision 2 Chemistry Factor, CF ('F) Fluence, f (10 n/cm ) 19 2 b Fluence Factor, ff 110 0.47 0.79 110 0.17 0.53 110 1.24 1.06 110 0.44 0.78 hRTNDT CF x ff ( F) Initial RTNDT, I ('F)

Margin, M ('F) 87 38 17 58 38 17 117 38 17 86 38 17

• • • • • • • • • • • • • • • • • • • • • • • • • )tk**********'O'Jt'****4k***A****'k'kkk*******'k**k***

Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 142 ART = Initial RTNDT + ARTNDT + Margin 113 172 141 (a) Based on surveillance capsule data. (b) Fluence, f, is based upon f urf (10 n/cm E>1 Mev) = 2. 1 at 32 EFPY '6] at vessel inner surface. The D. C. Cook Unit 2 reactor vessel wall thickness is 8.63 inches at the beltline region. (c)

Margin, M = 2[ol

+ o< ]', aI is the standard deviation for initial 2 2.5 RTNDT and a< i s the standard devi at ion for b RTNDT ~ a I i s assumed to be O'F for measured values of initial RTNDT and a< is 17'F for base metal. a< is cut=in half to 8.5'F, since surveillance capsule data is used [1]. 371St/082i69:10

1 h W i4 P I. I N V t

IIII ~ R illI illl IIIRIIRIIII IIIIIIiiiI IIIIhllll llllllll IIIIIII II IIIII> I~Iiil IIII <<re II IIII

I P, t ' Il gf r I l Ql

o MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE C5556-2 RTNDT AFTER 12 EFPY: 1/4T, 178'F 3/4T, 150'F CURYES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERYICE PERIOD UP TO 12

EFPY, DOES NOT CONTAIN MARGIN FOR POSSIBLE INSTRUMENT ERRORS.

2500 2250 Lea'est Lim)t 2000 1750 1500 Unacceptable Operation Acceptable Operation 1250 a. 100 0 4V I 7c0 4Z o 500 250 HeatIIp Rates 60'F/Hr Criticality Limit Based on Inservice Hydrostat)c Test Temperature (311 F) for the Service I poHod Up oo 12 EppY L 0 50 100 150 200 250 300 350 400 450 INDICATED TEMPERATURE (DEC.F) o ~ ~ ~ ~ ~ Figure 2. D. C Cook Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 12 EFPY (Without Margins) 3T10ss041089.10 13

A (p f

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE C5556"2 RTNDT AFTER 12 EFPY 1/4T 178 F 3/4T, 150'F CURVES APPLICABLE FOR COOLOOWN RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 12 EFPY. DOES NOT CONTAIN MARGIN FOR POSSIBLE INSTRUMENT ERRORS. 2500 2250 200G 1750 ~00 f2=0 C1 LJ n. I 000 Unacceptable Operation Acceptab'le Operation I ~ Cool Cown Rates F/Hr 0 20 ~40 60 l00 ( a S ~ a I ~ C 50 100 s I I I ~ 5'0 4GG I I ~ I 50 200 250 "OG 25G INDICATED TEMPERATURE (DEC.F') Figure 3. D. C. Cook Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 12 EFPY (Without Margins) 3110j/Oi)N9:10

C I P ~ I P t i'g x '4'

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE C5556"2 RTNDT AFTER 32 EFPY: 1/4T, 207'F 3/4T, 177'F CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. DOES NOT CONTAIN MARGIN FOR POSSIBLE INSTRUMENT ERRORS. "OG 2250 Leak Test Limit 20GC I750 Q ~av unacceptab1e Operation I2~0 Ill i C1V 0 I000 CI uJ Clz 500 250,, Hoatuo Rates 00F/Hr ~ i ~ + I I I I I Acceptab>e Operation Critica1$ ty Limit Based on Inservfce Hydrostat)c Test Tenperature (341'F) for the Service I Period Up to 32 EFPY I 0 I i I 50 l00 150 200 250 300 350 400 450 500 INOICATEO TEllPERATURE (OEG.F) ~ ~ ~ ~ ~ ~ Figure 4. D. C. Cook Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 32 EFPY (Mithout Margins) 3710e/041000; l0 15

'I wr WI l~ 'k I.?

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE e5556-2 RTNDT AFTER 32 EFPY: 1/4T, 207'F 3/4T, 177'F CURVES APPLICABLE FOR COOLDOMN RATES UP TO 60'F/HR FOR THE. SERVICE PERIOD UP TO 32 EFPY. DOES NOT CONTAIN MARGIN FOR POSSIBLE INSTRUMENT ERRORS. 2500 2250 2000 1750 1500 1250 Unacceptable Operatfon a. 1000 QtJ I 750 Oz 500 250 Cooldown Races 'F/Hr 0 20 40 60 100 Acceptable Operatfon 0 50 100 150 200 250 300 350 400 450 500 INOICATKO TEMPERATURE (OEG.F') Figure 5. D. C. Cook Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 32 EFPY (Nithout Margins) 311'/MIN10

\\ C~ 4

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE C5556-2 ~ ~ RTNDT AFTER 12 EFPY: 1/4T, 178'F 3/4T, 150'F CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 12 EFPY. CONTAIN MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRlNENT ERRORS. 2500 Leek Test Limit 2250 2000 1750 1500 IIeetup Iotas Nef~r lh LJ 10hl O LJ I O s ~ s s ~ 1 s I 1 I Unacceptabl ~ 0 ration 500 2=0 Acceptable Operation baaed on Insane ca srerestet14 Test T~rstere 529F) for the Service Fertee Ue te 12 EFFT~ 50 100 150 200 25p 3pp 35p 400 450 5p INDICATED TEMPERATURE (DEG.F') Figure 6. D. C. Cook Unit 2 Reactor Coolant System Heatup ~ ~ Limitations Applicable for the First 12 EFPY (With Margins) 3z>asiocisasno 17

(4 W I I't IP 1 AWL ~I n

MATERIAL'ROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE C 5556-2 RTNDT AFTER 12 EFPY 1/4T 178 F 3/4T, 150'F CURVES APPLICABLE FOR COOLDOWN RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 12 EFPY. CONTAIN MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. 2500 2250 2000 1750 w 1250 1000 O I 750 Z 500 250 Cooldown Rates 'F/Hr 0 20 40 60 100 Unaccaptabla Oparat1on Acceptable Operation I I I I s i I s v' 5 I 0 5P 100 150 200 250 300 350 400 450 50P INDICATED TEMPERATURE (OEG.F) 0 Figure 7. D. C, Cook Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 12 EFPY (With Margins) 371 di/0410aa:10 18

<<1 ~4 4r 4 ~ Il I1 .1 t%%

I ~ I MATERIAL PROPERTY" BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE C5556-2 RTNDT AFTER 32 EFPY: - 1/4T, 207'F 3/4T, 177'F CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. CONTAIN MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. 2500 8 Leak Test Limit 2250 I I 2000 I I I s ~ I !750 1500 Vl ! 250 a. ! 000 O I 750 OZ Unacceotaol e Operation 50~ F/Hr 500 250 Acceptable Operation Criticality Licit Iaaee on lnaervice Hy4roatatic Teat TWwreture (35)~) ~ for the Service Zrioe up to 3Z ER'T 0 50 100 150 200 250 300 350 400 450 500 INOICATEO TEMPERATURE (OEG.F) Figure 8. D. C, Cook Unit 2 Reactor Coolant System Heatup Limitations App1icab1e for the First 32 EFPY (Mith Margins) 371as/043040:10 19

i) Ol P 1J

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:. INTERMEDIATE SHELL PLATE C5556-2 , RTNDT AFTER 32 'FPY ' 1/4T 207 F 3/4T, 177'F CURVES APPLICABLE FOR COOLDOWN RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. CONTAIN MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. 2500 2250 2000 1750 1250 L000 Unacceptable OperatIon Acceptable Operation 500 I 2= C'," ~ 0 Cooldown Rates 'F/Hr 0 20 40 60 100 I I I 50 100 150 200 250 300 3 INDICATED TEMPERATURE (OEG ~ F) Figure 9. D. C. Cook Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 32 EFPY (With Margins) 3716s/041949:10 20

I 0

APPENDIX A HEATUP AND COOLDOWN DATA WITHOUT MARGINS FOR INSTRUMENTATION ERROR 371St/0410bQ:10 A-1

I 1j 'Is II

AMP COOLOOWN CURVES REG. GUIDE 1.99,REV.2 FOR PLATE C5556-2 04/17/89 -:THE: VOL'i.OPING DATA WERE PLOTTED FOR COOLOOWN PROFILE 1 ( STEADY-STATE 'COOLOOWN 'r) -.-

.1RRAOIAT1OH PERIOD'=-'10.'000 EFP YEARS FLAW DEPTH

~ AO'WIN T. r IHDICATEO;..'HDICATEO>> " - INDICATED ' INOXCATED,, IHDICATEOHDICATEO TEMPERATURE PRESSURE

,TEMPERATURE PRESSURE TEMPERATURE PRESSURE, (DEG.F) ~ (PS1)'.< ':" ' (OEG.F) (PSI) '. (OEG.F) (PSI) 85'.000 '""";<<":5(N;21".:-' 18 170.000- - 642. 13 ='.-,.--'5 - 255.000' ~ 1095.62. 2 90.000 ,,513.34, .19 175.000 656.27 36 260.000 1143.47 '3 "-<<'>> -.85'.OOO-, ".".,", "'517;7$ '""-':-""20 ~ '". 180.000 . 671.48, -- 37'--'" 265.000'~'1194.84 ~ 4 100. 000 522. 55 2 1 185. 000 687. 65, 38, 270.000, 1249. 75 5'.='rr'.".r,105;000'.>>"- .'<. "527'6$ '" 22 190)000 705.22 - '39, - 275.000 1309.08='-' ".'-'+>> >>>'~~4'Z"~"'>>~86 6 . 110,000 533. 19 23 195.000 724.07 40 280.000 1372.63 7 . 115;000:";" '39.02," "":=' 24. '00.000 744.22= 41 285.000-; 1440:59 8 120. 000,, 545.39,. 25, 205. 000, 766. 03 Q2 290.000, 1513. 87 9 -: '125."000r':"',~"..'52:24'~'-:4"'". ~ 26 '"'-" 210.000 789.30 ""'-P - 43 -'.295;000 " 159'2.21 ~ ', (-'.,"'r-"-'>>'"'"',10, 130.000.,, 559.61, 27, 215.000 814.47,, 44 300.000 1675.99 .:'" 11 ' i35!000 """--567;54>>:-'"'---'- -28 =-' 220.000 841'.39 ...- '5 305.000.~ 176$.86 ' -" << ~.--4'".'."-" 12 140.000 576.05, 29 225.000 870.44 46 310,000 1862.27 "". '13 '- 145.000 " 585.0$ '-'0 = 230.000 " 901.57, 47 .'15.000 ".; -196$.54 "=;, '.:: &~"" 14 150.000, 594.93 31, 235.000 '34.98 48, 320.000 2075.96 15 " 'i55;000 ":"":60$.51':"'. ': 32"---r 240.000 "'70.84 r:.r-4$ ",-.'25.000 - ".. 2193;76. -- 16 160,000 616.89 33 245.000 1009.61, 50 330.000 2319.71 '17r '. - 165.000.-". 629.12' ':-'34 250.000 1051.09; 51:' 335.000 2454.07 '.Ar r ~ 'r'.Vr.r. r r, 'prrrrr r *A '-:,'ri,r';,+h.."<<r;,') 'Y

l'

AMP COOLDOWN CURVES REG. GUIDE 1.99,REV 2 FOR PLATE C5556-2 04/17/89

THE FOLiOWiNi >DATi WERE PLOTTED FOR COOLOOWN OROFILE 2

--.-{20>DEG-F / ttR 'COOLOOWN ) .-:.::;,:,='=. IRRADIATION PERi00 "w','tl:000 EFP YEARS FLAW DEPTH .~ AOWIN T INDICATEO '-INDICATEO - '-- INOICATED INDICATED': INDICATEO. " INDICATEO TEMPERATURE PRE$ SURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE .- (DEG.f) .,i"'(PttI)~~~4.: .>>-.'OEG.F) (PSI) .(DEG.f) (PSI) "..?".";-"":"';".'~~~a>'?.'0',":>>~:"".'- 85;000;."': ">>458~48 <:Z.', 13 '45.000 =,, 545;60 ~ '- ~ 25 205.000 '739.03 ,2 h 90.000,470.69,,, t4 150.000 555.98,,26, 210,000,,763. 11 .."":=3" ' 95";000 '",,"'".7-'75'.-25~' ';,"" 15 -'="-'55 000 --. 567-:-17 .i "'.;.!'. 27- '"2 t5;000 "'.'*'. 799:.94?.'>' ",."'=~V.<<':.."""<,'w~>>%>s:"4-;+'.>I'PA 4 100. 000 480. 17 16 160. 000 579. 20 28 220. 000 818. 93 105.000 -;,. 8;485.48 --";-,"v. '17'""."'-', 165,000',"'h,' 592.06 '," ' '2$ ;, '" 225.000";., '- 850.03 ~ " " ">h':>>'~"~':. p".,'-..(-"',4>>"'.": 6,. 110.000 491. 19,, 18 170.000 606.00 30 230.000 883.38 -.7:: 1'15.000 -" ".'4$7 29 '->'-".'.i'=; '19 75.000' '21.04 31- " 235.000...'919.27 8 120,000 503.91 20

180.000 637.05 32 240.000 958.08 9 -'f25.000 >-"'-"'5tt>08'":""'::.'21 "';'t85.000 '664.47 "" 33 245;000' 999.63 10 130.000 h, 518.79, 22 190.000 673. 17 34 250.000 1044.25 '35.000 '~;.>'27.'t "~.' 23 " . '.195>000, . '693: 19 35 . " 255:000 "'= 1092.25 ' 12 140,000, 536.05

24 200.000 , 714,87

.AMP COOLDOWN CURVES REG. GUIDE,1-99,REV.2 FOR PLATE C5556-2

04/17/89 THE FOLLOlONQ DATA QEAK PLOTTED FOR CODLOOWN PROFILE "3; (40 DKG"F / HR COOLOOWN -:,.}.". 'RRADIATION PERIOD ~"" {2.000-:EFP YEARS FLAW DEPTH ~ AOWIN T

. INDICATEO:", -'INDIClTEO'; " '"" INDICATE'D INDICATED

'NDICATED '..;.:-'NDICATEO'EMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE '""'<<"'(DEG>F)."-'(PSI)<<1 > ~'<< (OEG.F) ~ (PSI)

'- 1<<z

-(DEG.F);- - <. (PSI): <." -"""85.000 '422;80'>>"" I."'3 - - 145;000 505.83 25 '205;000:., ~ --~'711.06 ~ ."'." ("<Y:<~k@c~'<,."."-'..- '.'>><.-'~<<'~ 2 90.000, 427< 14,,,, 14 150.000 516.61, 26 210.000, 737.79 '3."" -'5.000 "'<,:.~<434'-".85" "" 15 "'55.000 '28.48 ".":"" .,'"27 215.000, '"-" "768.80 4,< 100.000 436.93 16 160.000 541. 15 28 220.000,797.81 5 "'=' 105.000 '--'>>"-442:38'<<V>>~-:.'-'7.. -"-. 185<000 554 94 ' '=-29 '25 000 831121: - '.-:-" " '".'- ""-'"'"" -~."."'=.,'","~>><- >>-=c. 6 , 110.000 448.32 ,18 170.000 569.78 30 , 230.000 867.30 7 -'. 115.000 '-.<454 <78- --"".'19'75.000 585.88 31 235.000:,"'08.02 8 120. 000 461. 70, 20,, 180. 000.602. 92 32, 240.000, 947,64, 9 " 125.000 '. "'.'489.22'."," ': '1' "185.000 - ~ 621.52 ~ 33 "'- 245.000'-': 882.43 10 130. 000, 477. 32 22 190.000 64 1. 39 34 250. 000, 1040. 58 11 - ':"-"". 135.000 ~ "' 486.08 -'"w ".'"."23 '95.000 -",862 97- ~ - <<35 255.000:.<~'~ 1092.38 ~ '- '.':>>'<< '<'<<-'<~":<.*-. ~'~~k<k 12, 140.,000 , 495,43 .24, 200<<000,,... 686.02 <>h

ll

AMP COOLOOWN CURVES REG,GUIDE 1.99,REV.2 FOR PLATE C5556-2 04/17/89 "THE-'FOLLOWING DATA'ERE PLOTTED FOR COOLOOWN PROFILE 4

. (60 DEG"F / HR COOLOOWN."-.)

IRRADIATION'PERIOD < '2!000 EFP YEARS .'"~;-:,.':-,",,"',",, ~{'-;;.;~..~,",.;,:1~;-w;-sl:-. FLAW DEPTH %, AOWIN T "'--':=',"'-,1NDICATEO'-='::INDiiATEO':-" -: iNOICATED -INDI'CATED. i"-'- ' = "'INDICATEO'.-"ImICATEO -': '='-:- -'".".- ="~~~"="="P..";: '-~~{" N'EMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE i(DEC F')':-.".1{((PSI).~" =:

~ (OEG.F), (PSI) "'.':-=: (DEQ;F) ~ ~ 'PSI)-i"-" 'r~2'~.:~~i~X~'-"-~'88'.000 '>i'348;""t0;c "'<<,13 '48.000 465.16;, 24- '"'. 200;000'" '688.,17 -.."-'.-'-; ";.'..". ~;"";""Y"~N"~"'"'+." '~ . 2,, 90.000,382.60,... 14,, 150.000, 476.82 25 205.000 684,90 3 '"98:000"*,""-r. 387'<49 " ~ '8 r 185.000 '89.48 ~ 26 2'l0;000.;: l3.87 4, ,100.000, 392.77... 16 160.000 502.96 27 215.000, 744.91 {~"'8" "-'$08;000' ~~""'398.52 "~""" ': '17. 165.000 '17.67 '.';. 28 220.'000 " -'7$.32" 6 110.000 404.73 18 170.000 533.51 29 225.000 814.52 "'7: -" " 118.000 '""'l~- 411.43-"'", -"- 19 - 175{000 ~ .85083 ' =-- 30-:. 230000>>.'83.-34: " '--, ~ . ~';" .=-{'$~6< 8 120.000 418.72, 20 180.000 568.97 31 235.000 895. 11 125.000 '- 426;63. = 2l 185.000 ~ 588.77 32 240.000."".: 940.06 10 130.000 435. 17, 1 22 190.000 610. 22 33 245.000 988. 47 '11" '* 135;000 '-'44.36.=' " 23 195.000 '- 633.23 ' "'4 250.000 " 1040.52' ';~'.""'~; )'.;.?"'. 12 140,000 454.35

• r

1

, AMP COOLOOWN CURVES REG. GUIDE 1,99,REV.2 FOR PLATE,C5556-2 04/17/89 C ". 'THK FaiLOitING DATA,WERE PLDTTED FOR COOLOOWN PROFILE 5" (;000 DEG-F/HR CaOLOOWN. ) -':'. "-.'..-"=-. '":~-:, e":~~-; "',.". IRRADIATION*'PERIOD <'2;000 EFP YEARS ~ '""x"'.'.",",, FLAW DEPTH,M AOWIN T C

i'INDICATEO'"-"INDICATEO""

'" INDICATED INDICATED,"'-",,'" INDICATEO,'NDICATEO',,-, -" >> '";""!N'"'.'~"'.-'EMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (BEG.F) w~~si-'"(PSI}~&'"~'=: ~ "~ (OEG.F) (PSI) , .- -"~' .="'- (DEG.F):- (PSI). ':. 85:000-"Z~.',::285-N" -"." - ~ 12 '40;000 370.S3 " ',i "23 '. 185.000 = "- - 578.78 ,, 2 90.000 .290.35. 13 145,000 382.83 24 200.000

605.64 3' -8B!000;K~"-'-.'295.67'.." 14, -150.000,. 396. 12 ~ 25 205.000, '638.83 4, 100. 000 301.46, 15,, 155. 000 4 10.49,, 26, 210. 000 670. 59 ', "'5'"">>'05':000 "'.,:;>'"307.80: ';..:!"..". 16..'60.000" 426.07 27 '15.000. 706.88 6, 110.000 314.67,, 17, 165.000,, 442.89 28 220.000 745.98 7 '=' ='"115'000 ",~~'"-'322'8'~-'~~"r-'8 '",'-'-170 000 461 12 "': '9 '- '225.000"'788 17 8 120.000 ,330.30 19,, 175.000 480.87

30 230.000 833.60 9 -'--,'125.000.'--'39.48 ' 20 ":,-.180.000 S02.08

. 3'1-'-. 235.000-.

".. 882.61-10 130.000 348.80 21 185.000 525. 14 32 240.000 935.38

".'$35:000:=>>';=' 3B9.27, '-".-': 22 " = "-190.000 '49-;89 -.-=' 245.000 =- '-'892.23' ~ h C C- ~ C

,....AMP 60F/HR HEATUP CURVE REG. GUIOE 1.99,REV.2 FOR PLATE C5556-2 04/ 17/89 THE FOLLOWING DATA WERE CALCULATEOFOR THE INSERVICE HYDROSTATIC LEAK TEST. ' '. "'MINIMUM INSERVICE LEAK TEST TEMPERATURE ( 12.000 EFPY) h '~ rh c-'-"" '- x": j~-"P~J<f'i-~~" "=" PRESSURE'(PSI) TEMPERATURE: (OEQ;F)

2000, 290 icing',,;rh hcV;,"cV-aha>>:.'..,~VF'c;,h.;W @Pc.>

h;." c-- 'h, = -, c, , '311 PRESSURE . (PSI) 2000 h rr PRESSURE STRESS, 1.5 K1M

-.'PSI)

(PSI'G.RT.IN. 21444,, 89745 '*'":r "',"". ~ '" '"2485': = '. -..26645 '=" ", '"."':..112505 C f C 7 'C C

$'/

AMP, 60F/HR HEATUP, CURVE.REG. GUIDE 1,99,REV.2 FOR PLATE C5556-2 04/17/89 COMPOSITE CURVE-PLOTTED.FOR'EATUP PROFILE 2 '" 'RRAOKATION PERIOD(,('.,'12;000 EFP YEARS ,FLAW DEPTH .% (1-AOWIN)T HEATUP RATE(S) (DEQ.F/HR) 60.0 " ('""..'.i<<h,":.': ' "'<< 'R.'"'='<<':e >INDiCATEO ""'.INDICATEO" '- INDICATED '" INDICATED "" TEMPERATURE

PRESSURE, TEMPERATURE PRESSURE

,". ':>: (DES.F)-"--'-'-'<<(" (PSI)'.A:<<~<<>>"'-'- "'OEG.F) " '- (PSI):.. .":"- -,":; - INDICATEO 'NDICi'TEO - '--.-'.--':. " ---,,":"'-'"-.".- -';:-".='""'"-- TEMPERATURE PRESSURE (DEQ F ) (>><<> (PSI ) P j <<('>>>P +& 1(+'~ z h ( 1 "='-'-:-." '5.000'.-...'.>"",".l;4~8" 2, 90.000 49+~ = - -3. ( '.."SS;OOO':..-:

.¹SS-.@C'.

4, 100.000 ,469<<48 'PS -'-'- 105.000-'>>'"-'". ¹IPAO 19 175.000 '-, 545.24 37 26$.000 ~1044.5$ '" ":> -.".h- '"-"";-~:.o~<<.'; -<<'q",>>'y'<<'"vl'<<<<>> 20 180.000 558.93 38 , 270.000 ,1095.79 457~$ 21 185.000 573.79 "

'-. 39' 2VS.OOO.'150:VS 22 190.000 589.71 40 280.000 1209,57

'3 '" ~ 195.000 '-807.06 ~,.'" ".'41>> '85.000 -'1272 94 =" '" 110.000,, 457. 04,24 200. 000 -'.=.'" 7 <<=:.-'-' 1$.000-:.:<'¹57;64:.'>"-:-:-"4'-<<2S-;".-:: 205.000 e., 120.000,, 459.42, . 26, 210.000 S:'-"'5.000 " '"482;44 -":..-<.>>.-'7- .-'i"'-215.000'0, 130.000 466.44 28 220.000 . 135.000; -:"4'l1>51'."( '-.: '-29.. -'25.000 12 140,000,477.47, ,, 30 230.000 >'"'13 "'- '-:1¹5;000"'F'--",.'484;¹0->>>>'~P "".31"'."> 235'.000 ...14,150.000, 492. 1932 240.000 -.-'-1$ ". <<'-.':-".~155:000<<.~:"'::-,('00.85.<<<< ~33'>'45'000.'6, 160.000, 510,50 34.,250.000

17. >>>~.'s".:-j65'000

-:,===..."-; -521 '. 12 <<--'(".; 35 -'*,. 255,000 ., 18 , 170.000 532. 71 36 260.000 625.76 ,.645;78 667.48

.690;70 715.85 742.72 771. 78 802.89

,836.27 872. 14 910. 88 " 952.33 996.79 , 42 (43 44 -"-45 46 47 48 49 50 '(>>"':-"'1 52 >. 53 290.000 1340.67 -.'29$.000>>" 1413.27'= " (-<<:'. ":< '>>>>'>""-:,':>><<,".>> 300. 000 1490. 95 305.'000 . ':-15'l4. 43 310.000 1663.52 ." 31$.000 1759'.00 320.000 1860.76 . 325.000. ~~ ','970.00 "",:hei',. >h, >(~>, 4':, 'Y', 'v'.:"",'<:...i><<<<'4" 330.000 2086.46 - 335.000 -.- 2210.89; ~ =

:

'.'"."-::-:;;-"-."-':"-=."-.7 .. <<>'-'.:K;>> 340.000 . 2329.54 345.000 . '-.. '<<24521 VB ( y p '( V Y,( h<< ( V >> Y

'L 1 "f I'

AMP COOLOOWN CURVES REG. GUIDE 1,99,REV.2,FOR PLATE C5556-2,

04/17/89 ""'-THE'FOLL'OQINO DATA MERE'PLOTTED'FOR COOLOOWN PROFILF. t '.( STEADY"'STATE COOLOOWN ','-;.;, ~": '". ' 'w:,-', " *",..'". ":";<"~"~ IRRADIATION PERIOb ~>.'32.000'FP>>YEARS FLAW DEPTH AOWIN T, INDICATEO,' INDICATEO = - ,> INO1CATED INDICATED-,: "'.INDICATEO. INDICATEO TEMPERATURE PRESSURE

TEMPERATURE PRESSURE

TEMPERATURE PRESSURE ~'";'.>--<<: - (DEG.F) -::::.>(PSI)." 1 (OEG.F) (PSI). ' (DEG.F). ~ ',(PSI) 'f , '"'85.000 ".: ";."480.'"l8'."':',,i'., 20 '180.000 '-.-..'95'.92 ".'"'.,'. '8', 275.000' '013".'48 '" '. " 2 90.000 , 492,78 21, , 185.000 606.58 40 280.000 1055.25 -'3 '-" ~ 85.000 ,.:-'= 485;68 -" 22 " 190.000

618.04:. 41 - '285.000 tt00.'f0 4 100.000 498.79,

23, '195.000 630.35 42 290.000

1148,28 '-";-"-5-'.'0$.000'- =,- 502. i3 =t-"'- ""-'= 24. -

200.OOO 643:46

= - 43 '95.000',; .tf89.9$ -:,='<< '- -.=,-".-'-'. -.":""'-'-"--" "-'.'":V'"'-" -' 110.000 505.73 25 205.000 657.70 44 300.000 1255.36 ""'"""7-" -:- ti5.000 '"'. ='-= 509.60 N'"'. ""26 " " '10 000 '>.-'72'.99 " "-'"*45 ".0$.000 '" - 1315. tt 8,, 120,. 000, 513. 75,,, 27 215.000, 689. 29 46, 310. 000 ...1379. 04 i; 8 '. ", >>'t25.'000':-;":::.~'-.'518<<22 -,Yi'" '-"-:".28'.:,220.000 ~ 706.99 " '47,', 315:000. ',','t447.'54. =.;.:;", -":.> >>,'Kc':,"'"".>'>.':.'"-.'A~",.','~.'0 130.000 523.03 ,, 29 225.000 725.97 , 48, 320.000 1521.18 -;f35;000.";":- "- 528.18 ,-~':-,'--- 30:-": 230.000 746.26 ".---'8 , 325.000 ~ -=1600.05 .':-~:;~'~~ ",:i::,;,:.-:<<<<'~":"-<<.;"."'";:~"5N ,12 140.000 ... 533.75 . 31 ,235.000, 768.20 , 50 330.000 ,, 1684.61 " '" t3->> -';.~ 14$.000:.-'> >>'"'539.6t"= -- 32

"- 240.000 " '91.66

'=: 51 = . 335;000 1775 'l2' -~~"'<<>' 14 1S0.000, 546.03, 33 245.000, 817.01 52 340.000, 1872.08

tS '>---'55'.00052.93.'34'>-':

2SO.OOO-." 844'.12 .. '3 345.000 . " 1975.92 -'..'-:.">> ='"'.-'..'". <.'.-'-~.'."8.'.., . 16 ,, 160.000 560.35

35 255.000 873.20 54 350.000 2086.92 -~ 17 165:000 -:.".~ 568.33 .> "i(:i""-36-" -'80.000'-.' 904.72 55 355:000 2205.68 ':..'-.-".7Y<<.'."=":;<<<<w> '"""';;"",'."<<<,"'a~)'4 18 170.000, 576. 91 37, . 265.000, 938. 36, 56 360. 000 2332. 27 t8 -" : 175.000 " "- 586 Of> ": "" 38" '270.000 .974.48 57." "365'.000 467.'28 '"".~" "::<':""~'":-'"'."-<"-'-"1'""'""""" ~..';

%1

.AMP, COOLDOWN CURVES REG. GUIDE 1.99,REV.2 FOR PLATE C5556.-2 04/17/89 'tA '."" THE'I aLiOiiiNG D'ATA'VERE >LO'TTED t OR COOLOOWN PROFIt.E 2 -'2O'DEG-r /. HR,COOiDOWN )

"7" '-'RRAOfATION PERIOD 4" 32.000 EFP VE4RS" FLAW DEPTH

~ AOWIN T y>>>>

INDICATEO '"-" INDICATED

. ": ~ -- .INDICATED INDICATED TEMPERATURE ,PRESSURE TEMPERATURE PRESSURE -'= - (DEG;F)" ~'.':>>(PSI):>";"'",'" (OEG.F) - (PSI)

"8$ ;000" '>>','>>448".03 l'>>.'",'"'8'- 188.000 "i

-8t0.V2 >~".':. 2, 90.000,448.71 16 160.000 518.49 3""-~ 8$.000 ">> ': '451-62... '--'. '7 - 168.000 826.88 = , 100,000 454.74 18 170.000 535.90,, ":-S.'.>>-"'05.'000 'Y1458:t4 -." 19 '1V8.000 " 'S45.83 6 110.000,,, 46,1.79 ,,.,,, 20, t80.000 555.99 ",."7'i=-'-- -'::115.000; ~.-" ~"465;"t4'-';.~ -~,"=:~21 "*- >>~ '188;000 86V:28 8 120.000, 469.99 22 190.000 579.41 = 8-"" ~ "125.000 "': -474'.60'"-"='~ -'- -23 - =188.000 892.37 10 130.000, 479. 55 .24 200.000 606. 44 .t35:000- '.; - ""~484.9t ' '8 .208.000*62t.60'.>>'2, '140.000, 490.66., 26 210.000 637.75 13 '- '45.000 '-'."-"" 488.80:: -~- '27 -. 218:000 685;32 14 150.000,, 503.49,, ,28 220.000 674. 18 INDICATED.. INDICATED TEMPERATURE ,PRESSURE (DEG.F) (PSI) .: ~:"~'~%'g'V>>s! ".-'-",<'.:-'"'8 30 '31 , 32 '33 34 35-- 36 37 38 >> 39 40 '..." 4t 'V 225.000- =684;37 .-:""'F'~"'-'",".-;<" 5=. ',"'v~"PA.>> 230.000 716.23 238.000, 739-;59 240.000

764.88 24$. 000 - 791.. 94-250. 000,, 821. 18 255.000 "-'"" 852.55 260.000 886. 18 265.000 922.38 " 270.000 961. 48 27$.000= '. 1003.38 280.000 1048.39 285.000'. 1098.79 2 S 1- '>>),

4 J r

AMP,. COOLOOWN CURVES REG, GUIDE 1.99,REV.2 FOR PLATE C5556;2 04/17/89 THE FOLLOltINQ'DATA-WERE'LOTTED FOR CdOLOOWN PROFILE '3 ->> (40 DEG"F, /'R COOLOOWN ""IRRADIATIOMPERIOO"'32:000 EFP "YEARS -'.'- FLA'W DEPTH ~ AOWIN T - - INDICATEO' 'INDICATEO'-.- INDICATED INOICATED-'EMPERATURE PRESSURE TEMPERATURE PRESSURE (DEC'.F)- '-: 'L(PSX)'P>i'" -'-'"." '";. (OEG.F) (PSI) 1 -"-.-85:000 - "-"-400!'05~@ 'g, 15 . 155.000'-,,: 467;79 -, 29 2 90.000

,,403,64 16 160.000 475.96 30 '- "-.: 3 - -".--.-95;000>>='M'-::406.'52:;,'.'~ 17: , 165.000 ; "-: 484.82 '" '; 31 ~ ,, 4 100.000 409.68 18, 170.000 494.25

32 '<<'""'. 5--';,'".'.-105 000-'":."-"-',""'413;.'l3"': '" 19 " '175.000'- =" ".504.58 ~ '=:;.-'," 33 6,, 110,000, 416.86 20, 180.000, 515.64 34 --.'" 7": -"-'."115.'-000 -",.::-"."--'420;92.."'."." '-:.'21* l -185;000.: '27.63 - =."." 35 8, 120,000 425.29

22 190.000 540.42, 36 ""'~""'-9-';:'">>.".125;000 "-.":i'30.05 '"'i""'3 "195;000 '. ".'54.35 . "'7 .10 130.000

.435. 18, 24 200.000 569.34 38 ~""ff-'~"".,""-'f35000 '>>.;." ".'"4'40t68 -"25'; " 205.000 ',".585.40.-'39 12

140,000,,446.68 , 26 210.000 602.81... ,40 f3 '." 145-.000,. '-'.'453 1$ ~ '" 27 .'215;000 = 621.60 '41 14, ,, 150.,000.,460.20 , 28 220.000 641.66, 'NDICATEO. INDICATEO TEMPERATURE PRESSURE (DEG.F) '-(PSI).

- "."<-~ - ='=~-."'-."."-'..>>:" "'~'>.-' 225-.000 -,.'63.4'f'~=-.-'-';Y:j'-,

~'7<'AY~". ',< ~~j"'V-"

, 230.000 686.75 235,000 w '" =712 03 -.=~,. -' ' ~'40,000 739.04. '"'-', 245.000'" "" '788 31 " ') 250.,000 799.66 255.000 "33.40 " ~;.""." '.>>,~""'~.",Vi";.i~-.i'":. 260.000 , 869.82 270.000 950.97 280.000... 1044.85 285.000 1097. 17

AMP COOLDOWN CURVES REG. GUIDE 1.99,REV.2 FOR PLATE C5556-2 04/17/89 '"TKE FOL'LOWINO 'OATS WERE PLOTTED FOR COOLQOWN PROFILE 4 (60 DEG F',/ -'RRA01ATION PERIOD > "32.000 EFP YEARS FLAW DEPTH % AOWIN T KR'COOLOOWN )'INDICATEO" '~ INDICATED.>>' NDICATED TEMPERATURE PRESSURE TEMPERATURE (DEG.F), ~: "'~ '(PSI):~.:i ~.-- = 'DEG.F) 85.000';";:.".- "g'354>69>>>>"a;~~'.-'i '5 155.000'" 2, 90.000 357.38 16 160.000 3 "-,: -'-"; BS;000" '~+ '380;35- .; "-':.-: ~ --17. ': 165.000 ~ . 4, 100.000,, 363.52, 18 170.000 5'. '.'" 10$ :000",<~N'387";05".<<"'y..",'.'-.19 " 175.000 6 1 10. 000 370. 87 20 180. 000 "-7 ->>~~-.>> 115:000>>',~ >>:-375'06~>>~'P.- <<'2t = -'=. '185.000 8 120. 000 379. 58, 22 190. 000 " '- 9-="" 125.000-"--~~84.52".I"'-">>'-"'"""23'" ""'"195a000 ,10 130.000 389.86 24 200.000 >>-'-'-'-11 " '35:000 " -:-'='<395.:68 -'=--.'~ 25 " -:-,.'205.-000 - 12, t40.000, 401.96,, 26 210.000 13 ">>'> 145;000' -'.::"408.73-': 27-'. 215,000 '4 150.000 416. 10 424.t 1 432.74 ,442.04 452. 14 463!'08 474.87 487.64 501. 31 516'; 19 532.21 549.43 568.09 588.11 - 28 29 -'- 30 31 >>'32 , 33 -'4 35 38 37 ."">>'- '"'<<'"38 39 40 - INDICATED'."" PRESSURE (PSI) = . 'NDICATEO '..INDICATEO TEMPERATURE PRESSURE (DEG.F) 'PS?), '20.000 '.609.'Bt 225.000 . 633.21 230.000 . "-'659.32 '35.000 685.37 240.000. =-Vt4".67 "'-'= -.~ "-"'.'-"="".a!'...-',". ~-. ~". 245.000, 746.08

"-" 250.000

• >>--y 779.88'-'55.000816.51 260.000 '"

BSS;.75 265.000 898.05 - 270.000 ~ '."'43 52 275.000, ,992.47

• -280.000

'104S>>12 >>, >>y).9 '; i>>S. yy -"y. yr C y C C. .*-y y ~ y

,T

AMP COOLOOWN CURVES REG. GUIDE 1.99,REV.2 FOR PLATE C5556-2

04/17/89 -7'NE-taLLOWiNG'DAT'a'ERE::i LritiEO FOR COOLOOWN PROFILE-.S'.,( 100 OEG-F/NR COOLOOWN: ) '=:"'..:,.--: -'=., ":- " - --'~-.'.'-'.-'-:.-=-.".," - -'::=e<..'-='. 'IRRAOiiTI'ON:PERiOO. " -:32-.'OOO'EFP YEARS FLAW,DEPTH, ~ AOWIN T, INDItATEO:.>> -IHDICATEO:;.'---', .- INDICATED INDICATED,'.. 'INDICATEO '-.-IHDICATEO TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE -'-~'.""'"""'V(DEG.F),'~'."'>>,:, (PSI)~>>,",&>>'-" ' ('OEG.F) '-(PSI) "'."""";"' ' {DEG.F). ":- --{PSI).' -. =.-'>>>> i.".V':."-"->>-" "" *:"'.-".".i:;~-'i.-"~'P' 85.000-.". 258.;40~6>>i '-". 14 .. 1SO.OOO 325i47 -'-..'7 '.:, 215;000 '.,': ~ '523".34"'-.,",5">-,."-;.'"<>>~-~'< <<;">>F" >> " 2 90,000, 26t,17,, 15 155.000 ,334.49 , 28 220.000 548.47 5 ~ 000 "" r V;.'28$26%~>>'.+ 7,':...'.18 180 000 '",344 25.'.;4 29 '- H'-.22$ OOO,.575 "l8'>>>> '< "-'.-.i'" Pg""+""~R:. +W""~ 4, , 100.000 . 267.64 17 165.000 354.88 30 230.000 605.08 5': 105.000 -<<."271 40-"'K'.-~,'-" 18 170;000 " 386.32 '- 31 -.'.-"'23$ :000 .~ 638.74 '. -.-: --';:.,.'~%=.;,.';;,';.o'- - ':-~ >>"..9 6 1 10.000 275. 49 19, 175. 000 378. 82 32 240. 000, 671. 01 ,7 "-.. f tS,OOO '"'-"'.:-. 280:Ot:,'-5.>>'..20 . '.-180.000..',.392.31 . : ', 33, 2 $.000 '.." 707;85.' 120.000, , 284.90, . 21, 185.000, 406.91, 34 250.000 747.53 -'..": 9'-'-'-- 12$.000 '" i"'.".':290.31.".".'. " '2 ' 190.000 '22.73 3$ 255.000 790;35 = 10, . 130,000..., 296. 18, 23,195.000 439.81, 36 260.000 836.45 tt '"'*'35.'000 '<:302.62 '"'"""- 24 ":: '200;000 458.32 ".=-.';. 37 -. 26$.000 - '888.17 12 140.000 ,, 309.60 25 205.000 ,478.38 38 , 270.000 939.69 13 -' 145.000 ",';"":""3t7.24:.":".. ";28 ;"-....'"210.000 -"'.,499.92. >>'..: 39 275.000 ,"-'997.32-0 F+ XA:

AMP, 60F/HR HEATUP CURVE REG. GUIDE 1,99,REV.2 FOR PLATE C5556.-2 04/17/89 THE FOLLOWING DATA WERE CALCULATEDFOR THE INSERVICE HYDROSTATIC LEAK TEST, t "i '""MINIMUMINSERVICE LEAK-TEST TEMPE'RAZURE. ( 32;000 EFPY) .! '-""'. -=;.'-"--" '="~.='.~'--" "'.~.~ -"'."-',"!2"'"1 " '> """'-",o-'gg';"ga..'"2"'~'-":PRESSURE" (PSI) TEMPERATURE.. (OEG. F) 2000 ,3 19 fi '*"'~'."- ",.'-'=-'- 2485 A ? PRESSURE -"'- 'PSI) ,PRESSURE, STRESS , 1.5 K1M 'PSI) "(PSI SOiRT.IN.)- -'"'.'" ---' ';,--<<)~i'"'000 2485 21444 - 2664$ 897,45 "5'5 y+> )8'> j, ~ 8f 8+ 'jg-xc'y$ xv+'(x 4'

. AMP 60F/HR,HEATUP CURVE REG. GUIDE 1.99,REV.2 FOR PLATE C5556-2 COMPaSrTE 'CU'RVE PL>>DlTED. FOR 'HEA'TUO'P'ROFILE 2", 'EATUt RATE(S') ':(DEG.'F/HR) 'O'<<0 IRRAOiATION.PERIOO i 32..000 EFP YEARS FLAW DEPTH > (1-AOWIN}T 04/17/89 '""<:-'.INDICATEO '~ INDICATEO:-.'-='- TEMPERATURE, PRESSURE (DEa. F )-:" "';.%(V%I<<)~.':;-'-: '.. INDICATED INDICATED"': ~ INDICATEO ". INDICATEO TEMPERATURE PRESSURE TEMPERATURE PRESSURE (OEG.F).: '-'(PSI) '--:=-"' (DEQ.F)= " ';=(PSI) '"--;;dS; 000 ~~"V."-.':hatt~ 2, 90,000, ~IB """3 """--9$ -000-: ' 1 10. 000 ~5 1<<':.>> fi7"'fi' '"1 t5';000- '~:-'- '428."65.'I '=".- 21 "'15.000 8. 120,000 . 428. 66 28 220. 000 9 '". 125:000" '"'29".72.:=>~~"-'~- -",'29 -225;000 10...130.,000 .431.59,, 30. 230.000 1 t' -'.:f35.000

• "..'-:.". >.".434 324."--'>>'-.; 31: ~",'..235.000 12 140. 000,

,437. 74 ., 32, 240. 000 -><<:-- t3 -'?fi4$ :000'~ -'-'44'1:BS -': I.<<>><<. '3 245;.000 14 150.000 446.67 34

250.000 15 ";,155.000 "-',<<'"452:2t ~.'"-"<<' 35 ',-"'SS.OOO 16 160.,000 ., 458.39... 36, 260.000 -"" '17fi'""'165:000, ',-465.21--:-; 37 - '85.000 18 170.000 . 472.84, 38 270.000 "; ~.19 '" -17S 000?'"=-""" '481 ~ 14"""- 39 ""'" 275.000 20 180.000,, 490.18, 40 280.000 578.23.',.:-. " 47 594.99 48 813. 11 '.=:.- -'-" 49 632.71 50 853.64-..'. = 51 = 676.27 52 '700.41 ~ "' "., 53 726.59 54 '54.61 fi - ~'.~-'>>- SS 784.63 , 56 817.'10 "fi 57 851.82, 58 889 08 ' "'-'59 929.09, 31$.000' ';" 1304.48 " -<<:, "~" 320.000 1374.83 32$.000 '-'- 1450.01 330.000 1530.61 335.000 '618.85.. '- -,-'-.'.'". ~""'.' '.z;",.:-". ". 340.000 1709.26 345.000 ='801,94 "- -:>>.">>'.';.2, '.".?,;"; -'-.'".- '"'-" ~..","";";",3 350.000, 1913. 47 360.000, 2113.98 365 000 '-':. 2222'3 . :."""" '-v";i:':: 370.000 2337.74 375.000 - > 2461:04 = '-'.- ',:"", ~;.,"- ~-="-';:='~:-:-'=,"'fi=-"."".'9:.,-.-; " 21'"-'fi'185;000 "': 499.94 '"-;",'"" "='45~ ": 28$':000'-'.'" '.'972.05 22, 190.000, 510.60, 42, 290.000,,1018. 19 1jgg> A+3 '? " '195.000?> '-.'22 < 16,',.<<fi>: ~ 43-, "?fi '. "295.000'?<<"'!>'>. 1067.'73 '.';:>>,. ';.>>:-, <<g",PP'fi, '?*;-~~>>~?";~';>""'$P'fi'?'..."'< qj::. 24,, 200.000,, 534.64 . 44, 300.000 1120,91

"-'":-'s'4:-'5.'

~ 205;000 '-.'-" -548.02 ~-.> - -'='-45 '05.000""",:..1177.96'- 26 2 10. 000 562. 55 46 310. 000 1239. 07 .<<-?".,fi.',, >> fi ~ ? <<,, 'r'>>.fi>>> fi? yfi~fififi>""' fi ?>,- ? ? ?.>. p >fii"::,'fi <<.fi I

APPENDIX B HEATUP AND COOLDOMN DATA HITH MARGINS FOR INSTRUMENTATION ERROR 37 Ids/Oil 989:10 B-1

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Attachment 1 TO AEP:NRC:1077A DESCRIPTION AND 10 CFR 50.92 ANALYSES FOR CHANGES TO THE DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATIONS

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Attachment 1 to AEP:NRC:1077A Page 1 DESCRIPTION OF AMENDMENT REQUEST In this application to amend the Donald C. Cook Nuclear Plant Unit 2

license, we propose to revise the Technical Specifications by implementing a Core Operating Limits Report (COLR) for Unit 2 in accordance with Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications."

The proposed Technical Specification (T/S) changes are as follows: Definition 1.39, "CORE OPERATING LIMITS REPORT," was added in accordance with Generic Letter 88-16 2. T/S 3.1.1.4, "MODERATOR TEMPERATURE COEFFICIENT," was revised by removing the cycle-specific moderator temperature coefficient limit for end of life and placing it in the COLR in accordance with Generic Letter 88-16. T/S 4.1.1.4.2.b, a surveillance requirement for T/S 3.1.1.4, was revised by referencing the 300 ppm surveillance limit in the COLR. 4, T/S 3.1.3.1, "MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT," was revised by removing the reference to Figure 3.1-1, "ROD BANK STEP POSITION AS A FUNCTION OF POWER," and referencing the COLR. 5. T/S 3.1.3.4, "ROD DROP TIME," was revised by removing the value for the fully withdrawn position and placing it in the COLR in accordance with Generic Letter 88-16. T/S 3.1.3.5, "SHUTDOWN ROD INSERTION LIMIT," was revised by removing the value for the insertion limit and placing it in the COLR in accordance with Generic Letter 88-16. The T/S was also revised by changing the phrase "fullywithdrawn" to a phrase that references the insertion limit. 7. T/S 3.1.3.6 was revised by removing the reference to Figure 3.1-1, "ROD BANK STEP POSITION AS A FUNCTION OF POWER," and referencing the COLR, and by moving Figure 3.1-1, page 3/4 1-26, from the T/Ss to the COLR in accordance with Generic Letter 88-16. T/S 3.2.1, "AXIALFLUX DIFFERENCE," was revised by removing the cycle-specific target band limits from the T/S and referencing the COLR, and by removing Figure 3.2-1, "ALLOWABLE DEVIATION FROM TARGET FLUX DIFFERENCE," page 3/4 2-4, and placing it in the COLR in accordance with Generic Letter 88-16. Surveillance requirement 4.2.1.4 was also revised by removing the values for the axial flux difference target band and referencing the COLR.

Attachment 1 to AEP:NRC:1077A Page 2 9. T/S 3.2.2, "HEAT FLUX HOT CHANNEL FACTOR F (Z)," was revised by moving the cycle-specific F limit from th9 T/S to the

COLR, and moving the associated figures (Figures 3.2-2, page 3/4 2-8, and 3.2-2(a),

page 3/4, 2-8a) into the COLR in accordance with Generic Letter 88-16. 10. T/S 3.2.3, "NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F ~'," was revised by moving the cycle-specific F" and its power factor multiplier from the T/S to the COLR in accordance with Generic Letter 88-16. ll. T/S 3.2.6, "ALLOWABLE POWER LEVEL - APL", was revised by moving Figure 3.2-3, page 3/4 2-8b, V(Z) AS A FUNCTION OF CORE HEIGHT," from the T/S to the COLR, and referencing the cycle-specific F limit in the COLR in accordance with Generic Letter 88-16. SParting in Cycle 8, a cycle-specific V(Z) function will replace the current generic V(Z) calculated by Advanced Nuclear Fuels. 12. T/S 6.9.1.11, "CORE OPERATING LIMITS REPORT," was added to the T/Ss in accordance'ith Generic Letter 88-16. 13. In addition corresponding changes have been made to the appropriate Bases sections. B. 'BACKGROUND Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," provides guidance for relocating cycle-specific parameters from the T/Ss so that future reload designs can be implemented in accordance with 10 CFR 50.59. The generic letter requires that the cycle-specific parameters be maintained in a COLR and the NRC be informed of changes to the COLR. Any changes to the operating limits will made using approved NRC methodologies for Cook Nuclear Plant. This change will not affect the operation or safety of Cook Nuclear Plant Unit 2 since the actions required to be completed should a limit be exceeded will not be removed from the T/Ss. The proposed T/S changes include the removal of the moderator temperature coefficient, the shutdown rod insertion limit, the control rod insertion limits (Figure 3.1-1), the axial flux difference operational limits (Figure 3.2-1), the heat flux hot channel factor limit and associated factors (Figures 3.2-2 and 3.2-2(a)), the nuclear enthalpy hot channel factor limit and associated

factors, and the allowable power level factors F

, K(Z), and V(Z) from the T/Ss. The proposed T/Ss retain the Limiting Condition for Operation (LCO) wording and surveillance requirements by referring to the specific limits which are provided to the NRC in a COLR in accordance with the proposed T/S 6.9.1 ~ 11. Relocation of these cycle-specific limits is consistent with the guidance provided

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Attachment 1 to AEP:NRC:1077A Page 3 by the NRC in Generic Letter 88-16. Removal of the limits from the T/Ss will allow cycle-specific limits to be implemented in accordance with approved methodology in a timely and cost effective manner. C. JUSTIFICATION These T/S changes comply with the NRC guidance given in Generic Letter 88-16. Implementation of a COLR for each unit will allow for cycle-specific parameter limit changes in accordance with the referenced approved methodology of Section 6.9.1.11. The relocation of the cycle-specific parameter limits to the COLR will allow for greater flexibilityin optimizing designs to enhance the economics for each cycle. D. SAFETY EVALUATION The relocation of the cycle-specific parameters does not affect the operation of Cook Nuclear Plant Unit 2. The revised T/S will require the same actions to be taken when a limit is exceeded as required by the present T/S. Revisions to the COLR will be made in accordance with '10 CFR 50.59 and the NRC will be notified of all revisions in accordance with proposed T/S 6.9.1.11. All revisions to the COLR will be based on NRC-approved methodologies. The proposed revision to the T/Ss simply revises the method by which changes to the cycle-specific parameter limits can be requested and implemented. The parameters and methodologies associated with calculating them are presented below:

1) Revisions to the Moderator Temperature Coefficient, Rod Drop Time Rod Insertion, Shutdown Rod Insertion, and Control Rod Insertion Limits will be based on the Westinghouse methodology previously reviewed and approved by the NRC and described in WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology."

2) The Axial Flux Difference Limits will be, determined in accordance with Westinghouse methodology previously reviewed and approved by the NRC and described in WCAP-10216-P-A, Part B, "Relaxation of Constant Axial Offset Control/F Surveillance Technical Specification;" WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology;" and WCAP-8385, "Power Distribution Control and Load Following Procedures."

3) The Heat Flux Hot Channel Factor F (Z) will be determined in accordance with Westinghouse methodol9gy previously reviewed and approved by the NRC and described in WCAP-10266-P-A, Rev.

2, "The 1981 Version of Westinghouse Evaluation Model Using BASH Code;" WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology;" and WCAP-10216-P-A, Part B, "Relaxation of Constant Axial Offset Control/F Surveillance Technical Specification."

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4) The Nuclear Enthalpy Rise Hot Channel Factor F

will be determined in accordance with Westinghouse methodology previously reviewed and approved by the NRC and described in WCAP-10266-P-A, Rev. 2, "The 1981 Version of Westinghouse Evaluation Model Using BASH Code"; and WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology." Calculating the cycle-specific parameter limits in accordance with an NRC-approved methodology ensures the limits are consistent with the applicable safety analysis addressed in the Updated Cook Nuclear Plant FSAR. In addition, the proposed T/S 6.9.1.11 is written to require that the limits be maintained in accordance with the approved methodology. As discussed in this submittal, there is reasonable assurance that the T/S changes associated with relocating cycle-specific parameters out of the T/S will not adversely affect the health and safety of the public. E. NO SIGNIFICANT HAZARDS EVALUATION The changes presented in this amendment request are purely editorial in nature. Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (1) Involve a significant increase in the probability or consequence of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. The following is provided for the three categories of the significant hazards consideration standards. Criterion 1 Does the change involve a significant increase in the probability or consequences of an accident previously evaluated2 The moderator temperature coefficient limit, rod drop time rod insertion limit, shutdown rod insertion limit, control rod insertion limit, axial flux difference operational limits, heat flux.hot channel factor limit, and nuclear enthalpy rise hot channel factor limit are cycle-s'pecific parameters. The removal of the cycle-specific parameters from the T/Ss has no influence or impact on the probability or consequences of a previously evaluated accident. The cycle-specific parameter

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Attachment 1 to AEP:NRC:1077A Page 5 limits, although not in the T/Ss, will be maintained in the COLR and referenced in the Cook Nuclear Plant T/Ss. The proposed amendment still requires the same action be taken if limits are exceeded as is required by current T/Ss. Future reloads will be evaluated using NRC-approved methodologies and will be examined per the requirements of 10 CFR 50.59. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. Criterion 2 Does the change create the possibility of a new or different kind of accident from any accident previously evaluated2 There is no physical alteration to any plant system, nor is there a change in the method by which any safety related system performs its function. As stated

above, the proposed change is administrative in nature.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3 Does the change involve a significant reduction in a margin of safety? The margin of safety is not affected by the removal of cycle-specific parameter limits from the T/Ss. The proposed amendment still requires operation within the core limits as determined from the NRC-approved reload design methodologies. Appropriate actions will continue to be taken if limits are violated. The development of the limits for future reloads will continue to conform to those methods described in NRC-approved documentation. In addition, each future reload will involve a 10 CFR 50.59 review. Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

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. 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.

Attachment 1 to AEP:NRC:1077A Page 6 F. ENVIRONMENTAL EVALUATION I&M has evaluated the proposed changes and determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increases in the amounts of any effluents that may be released off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibilitycriterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed changes is not required.

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