ML20195B458

Rev 3 to CE NPSD-683, Development of RCS Pressure & Temp Limits Rept for Removal of P-T Limits & LTOP Requirements from Ts

ML20195B458
Person / Time
Site:	Fort Calhoun
Issue date:	05/31/1999
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML20195B434	List: LIC-99-0045, Forwards Application for Amend to License DPR-40,proposing to Relocate pressure-temp Curves,Predicated NDTT Shift Curve & LTOP Limits & Values from FCS TS to OPPD Controlled Document.Rev 3 to TR CE NPSD-683 & Supporting TSs Encl ML20195B437 ML20195B444 ML20195B452 ML20195B458
References
CE-NPSD-683, CE-NPSD-683-R03, CE-NPSD-683-R3, NUDOCS 9906010034
Download: ML20195B458 (150)

Text

{{#Wiki_filter:-___ _ _ _ _ -- C COMBUSTION ENGINEERING OWNERS GROUP CE NPSD-683 Rev.03 DEVELOPMENT OF A RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR THE REMOVAL OF P-T LIMITS AND LTOP REQUIREMENTS FROM THE TECHNICAL SPECIFICATIONS FINAL REPORT CEOG TASK 1120 prepared for the C-E OWNERS GROUP May 1999 dBB"bom$uY$I"EEhN$rfrig"NucIe"a~r" Power

       .48'1888A888888e3 P                   pon C          _

I LEGAL NOTICE This report was prepared as an account of work sponsored by the I Combustion Engineering Owners Group and ABB Combustion Engineering. Neither Combustion Engineering,Inc. nor any person acting on its behalf: l A. makes any warranty or representation, express or implied including the warranties of fitness for a particular. purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed in this report. Combustion Engineering,Inc.

h, , [ CE NPSD-683 Rev. 03 l THE DEVELOPMENT OF AN . ( RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR'THE REMOVAL i

  ;                                                       OF P-T LIMITS AND LTOP REQUIREMENTS FROM THE TECHNICAL SPECIFICATIC'NS

PREPARED FOR THE C-E OWNERS GROUP BY ABB COMBUSTION ENGINEERING NUCLEAR POWER COMBUSTION ENGINEERING, INC.

MAY 1999 VERIFICATION STATUS: COMPLETE The safety-related design infonnation in this report has been verified to be conect by means ofDesign Review using ChaMia 3.10-1 of QPM-101. Copyright 1999 CE,Inc. All Rights Reserved ABB CM= Engineenng NuclearPour

{ TABLE OF CONTENTS Section Page A INTRODUCTION............................................................ 6 ( A.1 B ACKGRC>U10') . . . . ., . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 l A.2 DESCRIPTION OF ACTIVITIES............................................. 7 A.2.1 PTLR Devel opmen t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 A.3 GENERIC PTLR.......................................................... 8 A.4 REACTOR COOLANT PRESSURE BOUNDARY OPERATIONAL DESCRIPTION............. 9 A.4.1 Genera 1............................................................ 9 A.4.2 Normal Opera ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 A.4.2.1 Reactor Vessel Boltup........................................... 9 A.4.2.2 Heatup......................................................... 10 A . 4 . 2 . 3 Co o1 down . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 A.4.3 Inservice Hydrostatic Pressure Test And Leak Tests . . . . . . . . . . . . . . . . 11 A.4.4 Rea ct or Core Opera ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 1.0 NEUTRON FLUENCE CALCUIATIONAL METRODS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 1.1 INPUT DATA........................................................... 13 1.1.1 Materials and Geometry ........................................... 13 1.1.2 Cross-Sections.................................................... 14 1.1.2.1 Multi-group Libraries......................................... 15 1.1.2.2 Constructing a Multi-group L1brary............................ 15 1.2 CORE NEUTRON SOURCE.................................................. 16 1.3 FLUENCE CALCULATION.................................................. 18 1.3.1 Transport cal cula ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 1.3.2 Synthesis of the 3 -D F1 uence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 1.3.3 Cavi ty Fluence Cal cula ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 1.4 METHODOLOGY QUALIFICATION AND UNCERTAINTY ESTIMATES.................. 22 1.4.1 Analytic Uncertainty Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 1.4.2 Corgparison wich Benchmark and Plant-Specific Measurements. . . . . . . . . 25

           ~1.4.2.1 Operating Reacter Measurements.................................                                                                    25 1.4.2.2 Pressure Vessel Simulator Measurements.........................                                                                    26 1.4.3 Overall Bias and Uncertain ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                              . 26 2.O    REACTOR VESSEL MATERIAL SURVIILIANCE PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 3.O    LOW TEMPERATURE OVERPRESSURE PROTECTION REQUIRENENTS...................                                                                     30 3.1 ImRODU CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 3.1.1 Scope.............................................................30 l        3.1.2 Background........................................................                                                                       31 1      3.2     GENERAL METHODOLOGY..................................................                                                                    34 3.2.1 Description.......................................................                                                                       34 3.2.2 L'1DP Evalua ti on Catqponen ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.3     TRANSIE!C ANALYSIS METHODOLOGY.......................................                                                                    35 3.3.1 L1MITING EVENT DETERM1 NATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.3.2 APPROACH AND MAJOR ASSUMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 3.3.3 LTOP RELIEF VALVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 0 3.3.3.1 General Description............................................ 40 3.3.3.2 Power-Operated Relief Valves...................................                                                                     40 3.3.3.3 SDC Relief       Valves..............................................                                                               42 3 . 3. 3 . 4 Pres suri zer Relief Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 3.3.4 ENERGY ADDITION EVENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 ABB Combustion Engineering Nuclear Power                                                                                                             2 CE NPSD-683 Rev. 03

I 3.3.5 MASS ADDITION EVENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4 3.4 LTOP EVALUATION METHODOLOGY.......................................... 46 3.4.1 CRITERIA FOR ADEQUATE LTOP SYSTM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6 3.4.2 APPLICAELE P-T LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6 3.4.3 LTOP MABLE TMPERATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 7 3.4.4 LTOP-RELATED LIMITING CONDITIONS FOR OPERATION. . . . . . . . . . . . . . . . . . . . 48 4.0 METHOD FOR CALCULATING BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURE

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(ART) ................................................................. 50 5.O APPLICATION OF FRACTURE MECHANICS IN CONSTRUCTING P-T CURVES........... 51 5.1 GENERAL.............................................................. 51 5.2 DETERMINATION OF THE MAXIMUM STRESS INTENSITY VALUES................. 54 5.2.1 General Method.................................................... 56 5.2.2 F1anges........................................................... 60 5.2.3 Nozz1es........................................................... 60 5.2.4 B el t l in e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 5.3 PRESSURE-TEMPERATURE LIMIT GENERATION METHODS........................ 63 5.3.1 General Description of PT Limits Generation . . . . . . . . . . . . . . . . . . . . . . . 63 5.3.1.1 Process Description............................................ 63 5.3.1.2 Regulatory Requirement......................................... 64 3 5.3.1.3 Reference Stress Intensity Factor.............................. 64 66 g 5.3.1.4 Calculation of Allowable Pressure.............................. 5.3.1.5 Analysis of HeatUp Transient................................... 66 ' 5.3.1.6 Analysis of Cooldown Transient................................. 67 5.3.1.7 Application of Output.......................................... 68 5.3.2 Thermal Analysis Methodol ogy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 5.3.3 ABB CMP PT Curve Me thod . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 5.3.3.1 Calculation of Thermal Stress Intensity Factors, Kn . . . . . . . . . . . 70 5.3.3.2 Calculation of Allow'able Pressure.............................. 71 5.3.4 Standard ASME PT Curve Me thod . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 5.3.5 199 6 ASRE Al t erna te Kn Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 5.4 TYPICAL PRESSURE-TEMPERATURE LIMITS.................................. 75 5.4.1 Beltline Limit Curves............................................. 76 5.4.2 Flange Limi c curves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 6 5.4.3 Composite Limit curves............................................ 77 5.4.4 operational Limit Curves.......................................... 78 5.4.5 Summarf . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 8 6.0 METHOD FOR ADDRESSING 10 CFR 50 MINIMUM TEMPERATURE REQUIREMENTS IN THE P-T CURVES ................................................................92 6.1 INSERVICE HYDROSTATIC PRESSURE TEST AND CORE CRITICAL LIMITS......... 92 6.2 MINIMUM BOLTUP TEMPERATURE........................................... 93 6.3 LOWEST SERVICE TEMPERATURE........................................... 93 7.0 APPLICATION OF SURVEILLANCE CAPSULE DATA TO THE CALCULATION OF ADJUSTED REFERENCE TEMPERA'IURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 5 8.0 SURREERY OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 7

9.0 REFERENCES

............................................................. 98 Appendices A Example of RCS Pressure and Temperature Limits Report B Example of Modified Technical Specifications ABD Combustion Engineering Nuclear Power 3 I CE NPSD-683 Rev. 03

f \ l l l LIST OF FIGURES l' , Figure No. Title Page l 5.1 Appendix G P-T Limits, Heatup............................. 81 5.2 Appendix G P-T Limits, Heatup............................. 82 l 5.3 Appendix G P-T Limits, Cooldown........................... 83 5.4 Appendix G P-T TAmits, Cooldown........................... 84 5.5 Appendix G Beltline P-T Limits, Hydrostatic............... 85 5.6- Appendix G Flange Limits, Heatup.......................... 86 i i

5. 7. Composite Appendix G P-T Limits, Heatup................... 87 5.8 -Composite Appendix G P-T Limits, Coo 1down................. 88 5.9 Composite Appendix G P-T Limits, Hydrostatic.............. 89 5.10 Typical Reactor Coolant System Pressure-Tamperature Idmits for Technical Specifications, Heatup...................... 90 5.11~ Typical Reactor Coolant System Pressure-Temperature Limits for Technical Specifications, Coo 1down.................... 91 i

l-t i l ABB Combustion Engineering Nuclear Power 4 CE NPSD-683 Rev. 03 A.

ABSTRACT An approach is presented in this report to relocate the Pressure-Temperature (P-T) limit curves, low temperature overpressure protection (LTOP) setpoint curves and values currently contained in the Technical Specifications (TS) to a licensee-controlled document. The approach is based upon criteria specified in NRC Generic Letter (GL) 96-03. As part of the relocation, additional considerations were the Reactor Vessel (RV) surveillance program, including the capsule withdrawal schedule, and the calculation of Adjusted Reference Temperature (ART), including the determination of the neutron fluence and analysis of post-irradiation surveillance capsule measurements. To substantiate relocation of the detailed information for affected Limiting Conditions for operation (LCOs), a new license controlled document called a RCS Pressure and Temperature Limits Report (PTLR) needs to be developed. This document is consistent with the requirements of Generic Letter 96-03 and contains the detailed information needed to support the pertinent LCOs which would remain in the Technical Specification. This topical report contains current methodology descriptions of RCS P-T limit development, LTOP criteria, ART calculation, RV Surveillance Program and Calculation of Neutron Fluence which supports the PTLR. An example of a PTLR is prepared along with the proposed changes to the subject Technical Specification. The enclosed sample PTLR is generic in nature and can be easily tailored to be suitable to any ABB CENP plant. ABB Combustion Engineering Nuclear Power 3 CE NPSD-683 Rev. 03

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I 4 A INTRODUCTION f 'In an effort to improve the maintenance of Technical Specifications, the Nuclear Regulatory Connaission (NRC) has issued Generic Letter (GL) 96-03, Reference 3, which allows the relocation of requirements from the Technical Specifications into another controlled document called an RCS Pressure and Temperature Limits Report (PTLR). This-relocation enhances the regulatory processing of frequently revised items such as Reactor

               Coolant System (RCS) Pressure-Temperature (P-T) limits, Low Temperature        -

Overpressure Protection (LTOP) setpoints, RV surveillance program post-irradiation test results, and neutron fluence calculation updates. Once f inccrporated into the plant's Technical specification, changes made in a PTLR would be controlled by the requirements of 10 CFR 50.59 and would no longer require a license amendment submittal to become effective. . This document is a product of a CE Owner's Group effort undertaken to create a generic PTLR document based on guidance presented in NRC GL 96-

03. This document is a complete revision and supercedes all previous revisions. This document has been reviewed and approved according to ABB CENP quality procedures for Quality Class 1 work. The verification status of this document is complete.

A.1 BACKGROUND

              -In 1972,'the Summer Addenda to the ASME Boiler and Pressure Vessel Code, Section III, incorporated Appendix G, " Protection Against Nonductile Failure". This Appendix, although nonmandatory, was issued to provide an acceptable' design procedure for' obtaining allowable loadings for ferritic pressure retaining materials in RCPB components.

Shortly after publication of ASME Code section III, Appendix G, a" new Appendix to 10 CFR 50 entitled " Appendix G - Fracture Toughness Requirements" was published in the Federal Register (July 17, 1973) and

             'became effective on August 16, 1973. This Appendix imposed fracture ABB Combustion Engineering Nuclear Power                                            6 CE NPSD-683 Rev. 03 i

1

    , toughness requirements on ferritic material of pressure-retaining components of the RCPB and mandated compliance with ASME Code Section III, Appendix G. Compliance with 10 CFR 50 Appendix G was applicable to all         E light water nuclear power reactors both currently operating and under                 j construction. 10 OFR 50 Appendix G, was further revised in 1979, 1983 and          .

1995. In addition to Appendix G, the RCPB must meet the requirements imposed by 10 CFR 50, Appendix A, General Design Criteria 14 and 31. These design criteria require that the reactor coolant pressure boundary be designed, fabricated, erected, and tested in order to have an extremely low ) l probability of abnormal leakage, of rapid failure, and of gross rupture. I The criteria also require that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, and testing loadings, the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized. Appropriate and conservative methods that protect the reactor coolant pressure boundary against nonductile failure have been developed by ABB Combustion Engineering Nuclear Power to comply with 10 CFR 50. A.2 DESCRIPTION OF ACTIVITIES The NRC issued GL 96-03 to advise licensees that they may request a license amendment to relocate cycle dependent information, such as the P-T limit curves and LTOP system limits from their plant Technical Specifications (TS) to a PTLR or similar controlled document. This task j addresses the development of the required information to be included in g the PTLR based on the generic letter. The guidance is divided into seven i provisions to be addressed in the PTLR. They are: l 1 Neutron Fluence Values i 2 Reactor Vessel Surveillance Program 3 LTOP System Limits ABB Combustion Engineering Nuclear Power 7 CE NPSD-683 Rev. 03

I l 4 Beltline Material Adjusted Reference Temperature (ART) 5 P-T Limits using limiting ART in the P-T Curve calculation 6 Minimum Temperature Requirements in the P-T curves 7 Application of Surveillance Data to ART calculations Each provision requires a methodology description be provided along with specific data about the operating plant. These provisions are specifically addressed in sections 1-7 of this document in conformance with the matrix of GL 96-03. The example PTLR shown in Appendix A is E organized to address each provision. Since this effort builds upon previous work (report CE NPSD-683 (Ref. 4)), the requirements of GL 96-03 are addressed by either creating new sections in the report or by drawing upon the work previously performed. In I addition, methodologies were updated to reflect the rule and code changes since the original issue of the topical, report. A.2.1 PTLR DEVELOPMENT The PTLR was developed on a generic basis such that it would apply to all I ABB CENP utilities. A review of typical LCOs for RCS P-T limits and LTOP requirements was performed and included in the generic PTLR.

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In order to support the PTLR, methodology descriptions were prepared and were incorporated as Sections 1-7 herein. The methodologies presented describe the development of RCS P-T limits, LTOP setpoints, RV Surveillance Programs and neutron fluence values. A.3 GENERIC PTLR l To facilitate development of a plant specific PTLR, an example PTLR is presented in Appendix A of this report. This PTLR is applicable to all ABB CENP utilities. l l ABB Combustion Engineering Nuclear Power 8 CE NPSD-683 Rev. 03

A.4 REACTOR COOLANT PRESSURE BOUNDARY OPERATIONAL DESCRIPTION i l l i A.4.1 GENERAL Currently 10 CFR 50, Appendix G imposes special fracture toughness requirements on the ferritic components of the Reactor Coolant Pressure Boundary (RCPB). These fracture toughness requirements result in pressure restrictions which vary with RCS temperature. Determinatien of these reutrictions requires that specific loading conditions be evaluated and the resulting pressure-tenperature limits not be exceeded. The specific I loading conditions, for which P-T limits are required, are as follows:

1. Normal operations which include reactor vessel boltup, heatup and cooldowm

2. Inservice hydrostatic pressure tests and leak tests

3. Reactor core operation A brief description of these conditions is provided to highlight the typical process followed to determine the physical loadings resulting from the particular operation.

A.4.2 NORMAL QPERATION A.4.2.1 Reactor Vessel Boltup ( Reactor vessel boltup loads are generated by stud tensioners when securing j the closure head against the reactor vessel. Prior to tensioning of the studs to the required preload, the reactor coolant temperature and the volumetric average temperature of the closure head region must be at or

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above the minimum boltup temperature. Once the studs have been tensioned, l the RCS is capable of being pressurized and heated. The heatup transient begins when a Reactor Coolant Pump (RCP) is started or when Residual Heat

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ABB Combustian Engineering Nuclear Ecuct 9 I CE NPSD-683 Rev. 03 l

Removal (RHR) system flow is altered to allow elevation of the RCS temperature, t A.4.2.2 Heatup - Heatup is the process of bringing the RCS from a COLD SHUTDOWN condition to'a HOT SHUTDOWN condition. The increase in temperature from COLD SHUTDOWN to NOT SHUTDOWN is achieved by RCP heat input and any residual

             -core heat.

During the heatup transient, the reactor coolant temperature is considered essentially the same throughout the RCS with the exception of the pressurizer. The pressurizer is used to maintain system pressure within the normal operating window which is between the minimum pressure associated with RCP net positive suction head (NPSH) or the RCP seal requirements, and the* maximum pressure meeting the RV material fracture toughness requirements. Also, the heatup rate should not exceed the rates specified by the pressure-temperature lind ts. A.4.2.3 Cooldown During'cooldown the RC$ is brought frem a HOT SHUTDOWN condition to a COLD SHUTDOWN conditio'n. Initially, coolant temperature reduction is achieved by removing heat through use of the steam generators by dumping the steam directly to the condenser or to the atmosphere through the Atmospheric

           . Dump Valve (ADV) .' The fluid temperature is decreased from approximately 550'T to 300*F using this method. To complete the cooldown the RHR System is utilized.

c Typically, cooldown is initiated by securing the RCP(s). Any remaining pug s provide coolant circulation through the RCS so that heat is transferred from the reactor coolant system to the secondary side of the steam generators. ~The RCS cooldown rate is controlled by the steam flow rate on the secondary side which is in turn controlled by the steam bypass control system or ADVs. The RCS pressure is controlled with the r-ABB Combustion Engineering Nuclear Power 10 CE NPSD-683 Jtev. 03 l'

pressurizer through use of heaters and spray. Once pressure and temperature have been reduced to within the design values of the RHR, the RHR can be utilized to control the cooldown rate and the remaining RCPs W can be stopped. It $ s advisable to initiate RHR flow prior to stopping all RCPs to provide sufficient mixing and minimize the thermal shock to RCPB components. The pressure during cooldown is maintained between the maximum pressure I needed to meet the fracture toughness requirements for this condition and the minimum pressure mandated by RCP NPSH requirements. The cooldown rate should not exceed the appropriate rates specified by the pressure-temperature limits. A.4.3 INSERVICE HYDROSTATIC PRESSURE TEST AND LEAK TESTS In order to perform a system leak test or hydrostatic pressure test, the system is brought to the HOT SHUTDOWN condition. The heatup or cooldown processes, described previously, would be followed to obtain a HOT SHUTDOWN condition. The pressure tests are performed in accordance with the requirements given I in ASME Code Section XI, Article IWA-5000. For the system leakage test, the test pressure should be at least the nominal operating pressure associated with 100% rated reactor power. In the case of the hydrostatic pressure test, the test pressure is determined by the requirements of ASME Code Section XI (Table IWB-5222-1). The minimum temperature for'the g required pressure is determined by the fracture toughness requirements and E guidance provided in 10 CFR 50, Appendix G. l A. 4. 4 REACTOR CORE OPERATION I The minimum temperature at which the core can be brought critical is controlled by core physics and safety analyses. This temperature is typically in excess of 500*F. The heatup process described previously is ABB Combustion Engineering Nuclear Power il CE NPSD-683 Rev. 03

F~ l l~ I used to attain the required temperature. Also, this minimum temperature is much higher than the requirements imposed by 10 CFR 50 Appendix G which ' ! address only brittle fracture concerns. t 1 i l l l l l l l 1 l l 1 I l l ABB Combustion Engineering Nuclear Power 12 CE NPSD-683 Rev. 03 i t

1.O NEUTRON FLUENCE CALCULATIONAL METHODS This section describes an outline of a general methodology for neutron fluence calculations. Due to the variety of dosimeter types which may be in use by any plant, and the plant specific nature of calculations for fluence, specific details of the methodology with regards to the dosimeter types used for the plant, methods qualification including analytical benchmark analyses to determine bias and uncertainty, and plant-specific methods and results (including uncertainties) shall be addressed'in~ detail by the plant-specific fluence analysis. The methods and assumptions described in this report apply to the calculation of vessel fluence for core and vessel geometrical and material configurations typical of ABB CDiP pressurized water reactors. This methodology meets the requirements of a proposed NRC Regulatory Guide (currently Draf t Regulatory Guide 1053: " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence"). The prediction of the vessel fluence is made by a calculation of the transport of neutrons from the core out to the vessel and cavity. The calculations consist of the following steps: (1) determination of the geometrical and material input data, (2) determination of the core neutron source, and (3) propagation of the neutron fluence from the core to the vessel and into the cavity. A qualification of the calculational procedure is described later. The discrete ordinate method should be used for the calculation of pressure vessel fluence. The DOT-4 code was commonly used in the United States and has been recently replaced by the DORT (2-D) and TORT (3-D) transport codes. 1.1 INPUT DATA 1.1.1 HATERIALS AND GEOMETRY Detailed material and geometrical input data are used to definer the physical characteristics that deternine the attenuation of the neutron ABB Combustion Engineering Nuclear Power 13 CE NPSD-683 Rev. 03

l l I flux from the core to the locations of interest on the pressure vessel. These data include material compositions, regional temperatures, and geometry of the pressure vessel, cere, vid internals. The geometrical input data includes the dimensions and locations of the fuel assemblies, reactor internals (shroud, core support barrel, and thermal shield), the pressure vessel (including identification and location of all welds and plates) and cladding, and surveillance capsules. For cavity dosimetry, I input data also includes the width of the reactor cavity and the material compositions of the support structure and concrete (biological) shielding, including water content, rebar and steel. The input data are based, to the extent possible, on documented and verified plant-specific as-built { dimensions and materials. The isotopic compositions of important constituent nuclides within each region are based on as-built materials data. In the absence of plant-specific information, nominal compositions and design dimensions can be used; however, in this case conservative estimates of the variations in the compositions and dimensions should be made and accounted for in the determination of the fluence uncertainty. The determination of the concentrations of the two major sources of isotopes responsible for the fluence attenuation (e.g., iron and water) are emphasized. The water density is based on plant full power operating g temperatures and pressures, as well as standard steam tables. The data E input includes an accounting of axial and radial variations in water { density caused by temperature differences in the core and inside the core barrel. 1.1.2 CROSS-SECTIONS I The calculational method to estimate vessel damage fluence uses neutron l 1 cross-sections over the energy range from ~0.1 MeV to ~15 MeV. The Draft j Regulatory Guide 1053 recommends the use of the latest version of the Evaluated Nuclear Data File (ENDF/B-VI). The ENDF/B-VI files were prepared under the direction of the Cross Section Evaluation Working Group l l (CSWEG) operated through the National Nuclear Data Center at Brookhaven National Laboratories (BNL). These data have been thoroughly reviewed, tested, and benchmarked. Cross-section sets based on earlier or equivalent nuclear data sets that have been thoroughly benchmarked for a specific application may be used for that application. l l ABB Combustion Engineering Nuclear Power 14 j CE NPSD-683 Rev. 03

1.1.2.1 Multi-group Libraries since the discrete ordinates transport code used to determine the neutron fluence uses a multi-group approximation, the basic data contained within the ENDF files must be pre-processed into a multi-group structure. The development of a multi-group library considers the adequacy of the group structure, the energy dependence of the flux used to average the cross-sections over the individual groups, and the order of the Legendre expansion of the scattering cross-section. Sufficient details of the energy- and angular-dependence of the differential cross-sections (e.g., the minima in the iron total cross-section) should be included to preserve the accuracy in attenuation characteristics. It should be noted that in many applications the earlier ENDF/B-IV version and the first three Mods of the ENDF/B-V iron cross-sections result in substantial underprediction of the vessel inner-wall and of the cavity fluence. Updated ENDF/B-V iron cross-section data have been demonstrated to provide a more accurate determination of the flux attenuation through iron and are strongly recommended. These new iren data are included in ENDF/B, version VI. 1.1.2.2 Constructing a Multi-group Library The ENDF files (including ENDF/B-VI) were first processed into problem-independent, fine- multi-group, master library containing data for I all required isotopes. This master library (e.g. , VITAMIN-B6) was developed at Oak Ridge National Laboratory and includes a sufficiently large number of groups (199) such that differences between the shape of the assumed flux spectrum and the true flux have a negligible effect on the multi-group data. This library includes 62 energy groups above 1 MeV and 105 groups above 0.1 Mev. The library also contains 42 photon energy groups. l The master library is collapsed into a job (broad group) library over spectra that closely approximate the true spectra. The resulting library should contain ~47 neutron and ~20 photon groups. This reduction is L accomplished with a one-dimensional calculation that includes the discrete regions of the core, vessel internals, by-pass and downcomer water, pressure vessel, reactor cavity, shield, and support structures. This job

ABB Combustion Engineering Nuclear Power 1$

L CE NPSD-683 Rev. 03 * ~

library should include approximately 20 energy groups above ~0.1 MeV. The collapsing was performed over four different spectra typical of pWRs, i.e. the core, downcomer, concrete and vessel. Both master (VITAMIN-B6) and job libraries are available from oak Ridge National Laboratory. 1.2 CORE NEUTRON SOURCE The determination of the neutron source for the pressure vessel fluence calculations accounts for the temporal, spatial, and energy dependence together with the absolute source normalization. The spatial dependence of the source is based on two dimensional or three dimensional depletion calculations that incorporate actual core operation or from measured data. The accuracy of the power distributions shall be demonstrated. The depletion calculations may be performed in three dimensions, so as to provide the source in both the radial and axial directions. The core neutron source is determined by the power distribution (which varies significantly with fuel burnup), the power level, and the fuel management schema. The detailed state-point dependence must be accounted for, but a cycle average power distribution inferred from the cycle incremental burnup distribution can also be used. The cycle average power distribution is updated each cycle to reflect changes in fuel management. For the extrapolation to the end of life fluence, a best estimate power distribution is used, which is consistent with the anticipated fuel n.anagement of future cycles. The peripheral assemblies, which contribute the most to the vessel fluence, have strong radial power gradients, and these gradients are accounted for to avoid overprediction of the fluence. The pin-wise source distribution generated by the depletion calculation is used for best-estimate, and represents the absolute source distribution in the assembly. When the actual planar core rectangular geometry can not be modelled (e.g., in the case of (r-0) discrete ordinates calculations), the pin power distribution in (x-y) geometry is converted into a (r-0) distribution as required by the (r-0) transport code geometry. ABB Combustion D2gineering Nuclear Power 16 CE NPSD-683 Rev. 03

4 [ The local source is determined as the product of the fission rate and the f neutron yield. The energy dependence of the source (i.e., the spectrum) and the normalization of the source to the number of neutrons per megawatt account for the fact that changes in the isotopic fission fractions with fuel exposure (caused by Pu build-up) result in variations in the fission spectra, the number of neutrons produced per fission, and the energy [ released per fission. These effects increase the fast neutron source per megawatt of power for high-burnup assemblies. The variations in these physics parameters with fuel exposure may be obtained from standard lattice physics depletion calculations. This effect is particularly (. important for cycles that have adopted low-leakage refueling schemes in [ which once , twice , or thrice-burned fuel is located in peripheral locations. The horizontal core geometry is described using an (r,0) representation of the nominal plane. A planar-octant representation is used for the octant-symmetric fuel-loading patterns typically used in ABB CENP plants. For evaluating dosimetry, the octant closest to the dosimeter capsules may ( be used. For determining the peak fluence, fuel-loading patterns that are not octant synnetric may be represented in octant geometry using the octant having the highest fluence. For evaluating dosimetry, the octant in which the dosimetry is located may be used. To accurately represent the important peripheral assembly geometry, a 0-mesh of at least 40 to 80 ( angular intervals is applied over the octant geometry. The (r,0) representation should reproduce the true physical assembly area to within

          -0.5% and the pin-wise source gradients to within ~10%. The assignment of

{ the (x,y) pin-wise powers to the individual (r,0) mesh intervals is made on a fractional area or equivalent basis. The overall source normalization is performed with respect to the (r,0) source so that differences between the core area in the (r,0) representation and the true core area do not bias the fluence predictions. 1 ABB Combustion Engineering Nuclear Power 17 CE NPSD-683 Rev. 03

r-1.3 FLUENCE CALCULATION 1.3.1 TRANSPORT ChloCULATION The transport of neutrons from the core to locations of interest in the pressure vessel is determined with the two-dimensional discrete ordinates transport program DORT in (r,0) geometries. An azimuthal (0) mesh using at least 40 to 80 intervals over an octant in (r,0) geometry in the horizontal plane provides an accurate representation of the spatial distribution of the material compositions and source described above. The radial mesh in the core region is about 1 interval per centimeter for peripheral assemblies, and coarser for assemblies more than two assembly pitches removed from the core-reflector interface. The Regulatory Guide 1053 recomends that in excore regions, a spatial mesh that ensures the flux in any group changes by less than a factor of ~2 between adjacent intervals should be applied, and a radial mesh of at least ~3 intervals per inch in water and ~1.5 intervals per inch in steel should be used. Because of the relatively weak axial variation of the fluence, a coarse axial mesh of about 2 inches per mesh may be used in the axial (Z) geometry except near material and source interfaces, where flux gradients can be large. For the discrete ordinates transport code, an Se l a fully symmetric angular quadrature is used as a minimum for determining the fluence at the vessel. Past calculations were limited by computer storage and had to be performed in two or more ' bootstrap" steps to avoid compromising the spatial mesh or quadrature (the number of groups used usually does not affect the storage limitations, only the execution time). In this approach; the problem volume was divided into overlapping regions. In a two-step bootstrap calculation, for example, a transport calculation was performed for the cylinder defined by 0< r< R' with a fictitious vacuum-boundary condition applied at R'. From this initial calculation a boundary source is I determined at the radius R' = R' - A and was subsequently applied as the internal-boundary condition for a second transport calculation from R" to R (the true outer boundary of the problem). The adequacy of the overlap region had to be tested (e.g., by 6ecreasing the inner radius of the outer ABB Combustion Engineering Nuclear Power 18 CE NPSD-683 Rev. 03 1

region) to ensure that the use of the fictitious boundary condition at R' had not unduly affected the bouadary source at R' or the results at the vessel. Current workstations nomally do not present this computer storage limitation, and the entire problem can now be solved as one fixed source problem. I A point-wise flux convergence criterion of < 0.001 should be used, and a sufficient number M iterations should be allowed within each group to ensure convergent v. To avoid negative fluxes and improve convergence, a weighted difference model should be used. The adequacy of the spatial I mesh and angular quadrature, as well as the convergence criterion, must be demorstre,ted by tightening the numerics until the resulting changes are negligible. In discrete ordinates codes, the spatial mesh and the angular quadrature should be refined simultaneously. In many cases, these e- aluations can be adequately performed with a one-dimensional model. I Although the term " fluence calculation' is commonly used, one must recognize that the calculated quantity is a multi-group flux distribution, and that the fluence is obtained by integrating the flux over energy and over the duration of full power operation (in seconds). I l The transport calculations may be performed in either the forward or adjoint modes. When several transport calculations are needed for a specific geometry, assembly importance factors may be pre-calculated by either performing calculations with a unit source (with the desired I pin-wise source distribution) specified in the assembly of interest or by performing adjoint calculations. The adjoint fluxes are used to determine the fluence contribution at a specific (field) location from each source region, while the forward fluxes from the unit-source calculations determine the fluence at all locations an the problem. Once calculated, I these factors contain the required information from the transport solution. By weighting the source distribution of interest by the assembly importance factors, the vessel (or capsule) relative fluence may be determined without additional transport calculations, assuming the in-vessel geometry, material, and in-assembly source distribution remain the same. The use of forward solution is made on the basis of the number of configurations to be solved for the end of life fluence determination. " ABB Combustion Engineering Nuclear Power 19 CE NPSD-633 Rev. 03

The computational speed achieved with modern workstations may justify the exclusive use of forward solutions. In performing calculations of surveillance capsule fluence (Regulatory Position 1.4), it should be noted that the capsule fluence is extremely sensitive to the representation of the capsule geometry and internal water region (if present), and the adequacy of the capsule representation and mesh must be demonstrated using sensitivity calculations (as described in Regulatory Position 1.4.1). The capsule fluence and spectra are, sensitive to the radial location of the capsule and its proximity to material g interfaces (e.g., at the vessel, thermal shield, and concrete shield in 5 the cavity), and these should be represented accurately. The core shroud former plates can result in a 5-10% underprediction of the accelerated l surveillance capsule dosimeter response and should be included in the { model. (No significant effect is generally observed on the dosimeters g located at the vessel inner-wall and in the cavity.) E , 1.3.2 SYNTHESIS OF THE 3-D FLDENCE Since 3-D calculations are not usually performed, the Regulatory Guide g) 1053 recommends that a 3-D fluence representation be constructed by 5 synthesizing calculations of lower dimensions using the expression

                   @ (r, 0, z) =@ (r, 0)

• L(r,z) (Equation 1) where @ (r, 0) is the groupwise transport solution in (r,0) geometry for a representative plane and L(r,z) is a group-dependent axial shape factor.

Two simple methods available for determining L(r,z) are defined by the expressions L(r, z) = P(z) (Equation 2) where P(z) is the peripheral-assembly axial power distribution, or l L(r, z) =@ (r, z) /@ (r) (Equation 3) where @ (r) and @ (r, z) are one- and two-dimensional flux solutions, respectively, for a cylindrical representation of the geometry that preserves the important axial source and attenuation characteristics. The ABB Combustion Eng ering Nuclear Power 20 CE NPSD-683 Rev. 03

(r,z) plane should correspond to the azimuthal location of interest (e.g., peak vessel fluence or dosimetry locations. The source per unit height for both the (r, 0)- and (r)- models should be identical, and the true axial source density should be used in the (r,z) model. Equation 2 is only applicable when (a) the axial source distribution for all important peripheral assemblies is approximately the same or is bounded by a conservative axial power shape and (b) the attenuation characteristics do not vary axially over the region of interest. Since the axial flux distribution tends to flatten as it propagates from the core to the pressure vessel, for typical axial power shapes, use of Equation 2 will tend to overpredict axial flux maxima and underpredict minima. This underprediction is nonconservative and can be large near the top and bottom reflectors, as well as when minima are strongly localized as occurs in some fluence-reduction schemes. Equation 3 is applicable when the axial source distribution and attenuation characteristics vary radially but do not vary significantly in the azimuthal (0) direction within a given annulus. For example, this approximation is not appropriate when strong axial fuel-enrichment variations are present only in selected peripheral assemblies. In summary, an (r,0)-geometry fluence calculation and a knowledge of the peripneral assembly axial power distribution are needed when using Equation 2. Use of this equation may result in fluence overpredictions near the midplane at relatively large distances from the core (e.g., in the cavity) and underpredictions at axial locations beyond the beltline that are at relatively large radial distances from the core. Conservatism may be included in the latter case by using the peak axial power for all elevations. Both radial and axial fluence calculations are needed when using Equation 3; thus, it is generally more accurate in preserving the integral properties of the three-dimensional fluence. Both Equation 2 and Equation 3' assume separability between the axial and azimuthal fluence calculations, which is only approximately true. ABB Combustion Engineering Nuclear Power 21 CE NPSD-683 Rev. 03

Il l 1.3.3 CAVITY FLUDICE CALCVIATIONS Accurate cavity fluence calculations are used to analyze dosimeters located in the reactor cavity. The calculation of the neutron transport in the cavity is made difficult by (a) thse strong attenuation of the E > 1 MeV fluence through vessel and the resulting increased sensitivity to the iron inelastic-scattering cross-section and (b) the possibility of neutron streaming (i.e., strong directionally dependent) effects in the low-density materials (air and vessel insulation) in the cavity. Because g' of the increased sensitivity to the iron cross-sections, ENDF/B-VI E cross-section data should be used for cavity fluence calculations. Properly benchmarked alternative cross-sections may also be used, however, l for cavity applications, the benchmarking must include comparisons for operating reactor cavities or simulated cavity environments. Typically, j the width of the cavity together with the close-to-beltline locations of f the dosimetry capsules result in minimal cavity streaming effects, and an S8, angular quadrature is acceptable. However, when off-beltline locations are analyzed, the adequacy of the S8 quadrature to datermine the streaming component must be demonstrated with higher-order Sn calculations. The cavity fluence is sensitive to both the material and the local geometry (e.g., the presence of detector wells) of the concrete shield, and these should be represented as accurately as possible. Benchmark g measurements involving simulated reactor cavities are recommended for W methods evaluation. When both in vessel and cavity dosimetry measurements are available, an additional verification of the measurements and calculations may be made by comparing the vessel inner-wall fluence determined from (1) the absolute fluence calculation, (2) the g extrapolation of the in-vessel measurements, and (3) the extrapolation of EB the cavity measurements. 1.4 METHODOLOGY QUALIFICATION AND UNCERTAINTY ESTIMATES i Regulatory Guide 1053 requires that the neutron transport calculational l ( methodology be qualified and that flux uncertainty estimates be i determined. The neutron flux undergoes several decades of attenuation before reaching the vessel, and the calculation is sensitive to the 4 ABB Combustion Engineering Nuclear Power 22 CE NPSD-683 Rev. 03 I!

i 1 material and geometrical representation of the core and vessel internals, i the neutron source, and the numerical schemes used in its determination. The uncertainty estimater are used to determine the appropriate . uncertainty allowance to be included in the application of the fluence estimate. While adherence to the guidelines described in the Regulatory I cuide will generally result in accurate fluence estimates, the overall methodology must be qualified in order to quantify uncertainties, identify any potential biases in the calculations, and provide confidence in the fluence calculations. In addition, while the methodology, including computer codes and data libraries used in the calculations, may have been I found to be acceptable in previous applications, the qualification ensures that the licensee's implementation of the methodology is valid. The methods qualification consists of three parts: (1) the analytic uncertainty analysis, (2) the comparison with benchmarks and plant-specific data, and (3) the estimate of uncertainty in calculated I fluence. I 1.4.1 ANALYTIC CINCERTAINTy ANALYSIS I The determination of the pressure vessel fluence is based on both calculations and measurements; the fluence prediction is made with a calculation, and the measurements are used to qualify the calculation. Because of the importance and the difficulty of these calculations, the method's qualification by comparison to measurements must be made to I ensure a reliable and accurate vessel fluence determination. In this qualification, calculation-to-measurement comparisons are used to identify biases in the calculations and to provide reliable estimates of the fluence uncertainties. When the measurement data are of sufficient quality and quantity that they allow a reliable estimate of the I calculational bias (i.e., they represent a statistically significant measurement data base), the comparisons to measurement may be used to (1) determine the effect of the various modeling approximations and any calculational bias and, if appropriate, (2) modify the calculations by applying a correction to account for bias or by model adjustment or both. I As an additional qualification, the sensitivity of the calculation to the important input and modeling parameters must be detemined and combined with the uncertainties of the input and modeling parameters to provide an independent estimate of the overall calculational uncertainty. ABB Combustion Engineering NJclear Power 23 CE NPSD-683 Rev. 03

To demonstrate the accuracy of the methodology, an analytic uncertainty analysis must be perfonmed. This analysis includes identification of the important sources of uncertainty. For typical fluence calculations, these sources include:

           . Nuclear data (cross-sections and fission energy spectrum),
           . Geometry (locations of components and deviations from the nominal dimensions),
           . Isotopic composition of material (density and composition of coolant water, core barrel, thermal shield, pressure vessel with cladding, and concrete shield),
           . Neutron sources (space and energy distribution, burnup dependence),
           . Methods error (mesh density, angular expansion, convergence criteria, macroscopic group cross-sections, fluence perturbation by surveillance capsules, spatial synthesis, and cavity streaming).

This list is not necessarily exhaustive and other uncertainties that are specific to a particular reactor or a particular calculational method g should be considered. In typical applications, the fluence uncertainty is dominated by a few uncertainty components, such as the geometry, which are E usually easily identified and substantially simplify the uncertainty analysis. The sensitivity of the flux to the significant component uncertainties g should be determined by a series of transport sensitivity calculations in W which the calculational model input data and modeling assumptions are varied and the effect on the calculated flux is determined. (A typical sensitivity would be ~10-15% decrease in vessel #1 MeV fluence per centimeter increase in vessel inside radius.) Estimates of the expec*.ed g uncertainties in these input parameters must be made and combined with the W corresponding fluence sensitivities to determine the total calculated. Il I l 4 ABB Combustion Engineering Nuclear Power 24 CE NPSD-683 Rev. 03

1.4.2 COMPARISON WITH BENCIMARK AtO PLANT-SPECIFIC NEASURDENTS Calculational methods must be validated by comparison with measurements and calculational benchmarks. Three types of comparisons are required: I

• operating reactor in-vessel or ex-vessel dosimetry measurements, I e pressure vessel simulator e calculational benchmarks The methods used to calculate the benchmarks must be consistent with those used to calculate fluence in the vessel. Calculated reaction rates at the I dosimeter locations must agree with the measurements to within about 20%

for in-vessel capsules and 30% for cavity dosimetry. If the observed deviations are larger, the methodology must be examined and refined to improve the agreement. I 1.4.2.1 Operating Reactor Measurements l Comparisons of measurements and calculations should be performed for the specific reactor being analyzed or for reactors of similar physical and I fuel mane.gement design. This plant-specific data can be compared to the benchmark analyses to validate that plant-specific calculations are within the tolerances expected by the benchmark uncertainty. A good estimate of the vessel attenuation can be obtained when both in-vessel and cavity dosimetry are available. These measurements should not be used to bias or I adjust the fluence calculations unless a statistically significant number of measurements is available, the various dosimeter measurements are self-consistent and a reliable estimate of the calculational bias can be determined. Similarly, plant-specific biases should not be used unless sufficient reliable measurement data are available. As capsule and cavity I ineasurements become available, they should be incorporated into the operating reactor measurements data base and the calculational biases and uncertainties should be updated as necessary. l ABB Combustion Engineering Nuclear Power 25 CE NPSD-683 Rev. 03

I 1.4.2.2 Pressure Vessel Simulator Measurements A number of experimental benchmarks providing detector reaction rates in the peripheral fuel assemblies, within the vessel wall, and in the cavity are available for the purpose of methods calibration. These benchmark experiment were carried out by several laboratories, and dosimetry measurements using different techniques were compared to provide experimental results with well known and documented uncertainties. Examples include the Pool Critical Assembly (PCA), VENUS, and H.B. Robinson Unit 2 benchmarks. 1.4.2.3 Calculational Benchmarks A calculational benchmark commissioned by the NRC and prepared by Brookhaven National Laboratory (Reference 8) provides very detail input description as well as the flux solution at several mesh points. An analysis of this benchmark, which addresses both standard out-in and low-leakage fuel management, provides a detailed test of the cross sections and various calculational options for transport calculation. The benchmark calculation results may be used for methods qualification. The calculation being used as the benchmark must be the actual original referenced benchmark calculation, and not just a second independent calculation of the benchmark. 1.4.3 OVERALL BIAS AND UNCERTAINTY An appropriate combination of the analytical uncertainty analysis and the results of the uncertainty analysis based on the comparisons to the benchmark results provide the bias and uncertainty to be applied to the predicted fluence. n ABB Combustion Engineering Nuclear Power 26 I CE NPSD-683 Rev. 03

i 2.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM r This section addresses Provision 2 of Attachment 1 to Generic Letter 96-03 (Reference 3) on compliance with 10CFR50, Appendix H (Reference 17) . Appendix H presents the requirements for reacter vessel material surveillance programs. The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of AS W E185 (Reference 18) that is current on the issue date of the ASME Code, to which the reactor vessel was purchased. For each capsule withdrawal, the test procedures and reporting must meet the requirements of AS M E185-82 to the extent practicable for the configuration of specimens in the capsule. 1 ASTM E 185, " Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels,* provides for the monitoring and periodic evaluation of neutron radiation-induced changes in the mechanical properties of the vessel beltline materials. .The ASTM standard provides procedures for the selection of materials, the design and quantity of test specimens, the design and placement in the reactor vessel of the test specimen compartments, and the means for measuring neutron fluence and irradiation temperature. These are aspects pertaining to the design of the program. AS7H E 185 also provides the guidelines for a schedule for the withdrawal of capsules for testing and a procedure for the pre- and post-irradiation testing of the surveillance program materials, neutron fluence monitors and temperature monitors. The reactor vessel material surveillance program for Combustion Engineering owners Group plants was designed to meet or exceed the requirements of the version of ASTM E 185 in effect at the time that the t vessel was purchased. For each vessel, base metal was selected from one of the beltline plates and used to fabricate test specimens for pre-

• f irradiation testing and for inclusion in the surveillance capsule compartments. Similarly, a weldment was fabricated using portions of the beltline plates and the same welding process as used for one or more of the beltline welds; both weld metal and heat-affected-zone (HAZ) specimens I were fabricated from the weldment for pre-irradiation testing and for inclusion in the surveillance capsule compartments. A section from the surveillance plate and weld was retained as archive material. Neutron j flux and temperature monitors, and test specimens from the surveillance plate, weld and HAZ together with specimens from a correlation monitor ABB Combustion Engineering Nuclear Power 27 CE NPSD-683 Rev. 03

material were loaded into compartments and assembled into surveillance capsules. A minimum of six surveillance capsules were originally provided for each CEOG plant. Records were compiled that documented the source of the materials, including fabrication history, the location and orientation of test specimens in the original material, the design of the specimen g campartments, and the location of individual specimens in the compartments a for each capsule assenbly. The six surveillance wall capsules were installed in holders on the inside surface of the reactor vessel and within the region surrounded by the g effective height of the active reactor core. The vessel wall location 5 provides for irradiation of the surveillance materials under conditions closely approximating the neutron fluence rate, temperature, and variations thereof, over time of the reactor vessel that is being monitored. [ Note: See plant-specific details for azimuthal location of g the wall capsules and, if applicable, for additional surveillance or 5 dosimetry capsule locations. In some cases, additional capsules were installed in holders attached to the core barrel for accelerated irradiation or in the upper plenum region away from the beltline where the fast neutron fluence is negligible. In other cases, replacement g surveillance capsules have been installed in empty capsule holders to u l obtain additional vessel material or neutron fluence data. Examples of the latter are dosimetry capsules installed inside the vessel and outside in the annulus between the vessel and the biological shield.) The surveillance capsule withdrawal schedule was originally established following the requirements of the version of E 185 in effect at the time of vessel design / fabrication; the schedule may have been originally s ! l established based on the requirements of 10CFR50, Appendix H, Reactor l Vessel Material Surveillance Program Requirements. The schedule' called  ! for at least three capsules to be removed and tested during the design life of the reactor vessel. The remaining capsules were available to provide a higher frequency of testing if required or retained to provide supplemental information in the future. The surveillance capsule withdrawal schedule may be modified. Such proposed modifications will either be submitted to the NRC with a technical justification for approval or evaluated under the auspices of 10CFR50.59 for those plants not having the surveillance capsule removal schedule in the Technical Specifications. ABB Combustion Engineering Nuclear Power 28 CE NPSD-683 Rev. 03

Post-irradiation testing is presently performed on the specimens from the withdrawn capsule in accordance with the requirements of ASTM E 185-82 (or later versions, as specified in Appendix H) and 10CFR50, Appendix H. The test data and evaluation results are compiled and presented in a report to the NRC within one year of the date of capsule withdrawal (unless an extension is requested and-then granted by the Director, Office of Nuclear Regulation). Application of the data for the PTLR are discussed in Sections 4.0 and 7.0. 1 The initial properties of the reactor vessel beltline plates and welds were established in parallel to the establishment of the reactor vessel i surveillance program. For each of the beltline plates, Charpy impact tests and/or drop weight tests were performed to demonstrate compliance with applicable ASME Code and vessel specification requirements. The welding procedures used for beltline welds were qualified and the welding materials certified to applicable AWS, ASME Code and vessel specification requirements. Chemical analyses of the plates and weld deposits were obtained in accordance with the vessel specification. The data were processed to obtain a value of the initial reference temperature, RTan, and of the copper and nickel content. [N6te: The data that are available for a specific vessel will vary because of differences in the requirements for testing and certification.) For beltline plates and welds, the initial RTan was determined in accordance with the ASME Code, Section III, l l

        -NB-2331, for which drop weight tests and Charpy impact tests (complete l         transition curve) were performed. For the earlier CEOG vessels for which test requirements were different, the initial RTme was determined using Branch Technical Position MTEB 5-2, Fracture Toughness Requirements (for older Plants), or a generic value of initial RTme was determined based on measurements for a specific set of materials. Some CEOG sponsored efforts which are pertinent are report CEN-189, December 1981, " Evaluation of l         Pressurized Thermal Shock Effects due to Small Break LOCAs with Loss of Foodwater for the Combustion Engineering NSSS*, and CE NPSD-1039, Revision 02, June 1997, CEOG Task 902, 'Best Estimate Copper and Nickel Values in      l CE Fabricated Reactor Vessel Welds'.

l I ABB Combustion Engineering Nuclear Power 29 CE NPSD-683 Rev. 03

n 3.0 LOW TEMPERATURE OVERPRESSURE PROTECTION REQUIREMENTS

3.1 INTRODUCTION

l. 3.1.1 SCOPE This section addresses Provision 3 of Attachment 1 to Generic Letter 96-03 (Reference 3) that allows low temperature overpressure protection (LTOP) limits developed using NRC-approved methodologies and contained in Technical Specifications -(TS) to be relocated to a plant-specific PTLR. The methods are described that ABB CENP utilizes l in the analyses supporting LTOP to ensure adequate protection of the reactor coolant pressure boundary (RCPB) and, especially, of the l reactor vessel, against brittle fracture during heatup, cooldown, and shutdown operations. These methods should be followed by the CEOG plants in the calculations of the plant-specific LTOP limits in their original PTLRs and revisions thereof. The relationship between LTOP setpoints and-limitations and Reactor Coolant System (RCS) pressure-temperature (P-T) limits is also discussed. l The P-T limits to be protected by LTOP tetpoints and limitations can be j 1 developed using the following methodologies:

1. ASE Code Section III, Appendix G (Reference 9), l

2. ASME Code Section XI, Appendix G (Reference 10),

3. ASME Code Case N-514 (Reference 11), and

4. ASME Code Case N-640 (Reference 21).

l

l. Presently, only Reference 9 has been endorsed in 10 CFR 50.55a and approved by the NRC for the use in establishing P-T limits. Appendix G P-T limits based on Reference 9 provides a basis for LTOP setpoints and limitations'at most of CEOG plants. The 1996 Edition of the Code (Reference 10) has not been endorsed in 10 CFR 50.55a and the use of it, as well as the use of ASME Code Case N-514 (Reference 11) and ASME Code i

Case N-640, has not been generically approved. Therefore, an exemption must be obtained from the NRC prior to the use of these methodologies. el ABB Combustion Engineering Nuclear Power 30 CE NPSD-683 Jtev. 03 t

In this section, P-T limits developed using the methodologies of { References 9 and 21 are identified as Appendix G P-T limits. Similarly, P-T limits developed using the methodologies of References 10 and 11 are identified as LTOP P-T limits, for they represent Appendix G P-T limits adjusted up by 10%, as prescribed by these documents. The LTOP P-T limits are used exclusively as a basis for LTOP setpoints and limitations. Additionally, two methods of calculating the LTOP enable temperatures are addressed: one is per Branch Technical Position RSP 5-2 (Reference

12) and another, as prescribed by ASME Code Case N-514 (Reference 11) and ASME Code Section XI, Appendix G (Reference.10). An exemption must be obtained for the use of References 10 or 11 similar to the P-T limits.

3.

1.2 BACKGROUND

l Current requirements defined in Section III, Article NB-7000 of the , ASME Boiler and Pressure Vessel Code provide for overpressure protection of the RCPB during power operation. Additional rcquirements are also given by 10 CFR 50, Appendix A, Design criterion 15 and Design Criterion 31. These criteria require that the RCS be designed with sufficient margin to ensure that the design conditions of the RCPB are not exceeded during normal operation including anticipated operational occurrences, and the RCPB be designed with sufficient margin to ensure that when stressed under operating, maintenance, testing, and postulated accident conditions, it behaves in a nonbrittle manner and the probability of rapid propagating fracture is minimized and very low. Consequently, the U.S. Nuclear Regulatory Commission (NRC) has provided guidance to ensure overpressure protection for normal operation and anticipated operational occurrences at conditions other than power operation. This guidance, published in NUREG-75/087 (currently NUREG-0800), is provided in Standard Review Plan 5.2.2,

             .._s -

ABB Combustion Engineering Nuclear Power 31 CE NPSD-683 Rev. 03 ,

I ' Overpressure Protection" (Reference 2), which includes BTP RSB 5-2 (Reference 12) I The primary concern of BTP RSB 5-2 pertains to operation at low temperatures, especially in a water-solid condition. The applicable operating limits in the low temperature region are based on an Appendix G evaluation which provides much lower allowable pressures than the design limit considered at normal operation (power operation) pressure and temperature. The consequences resulting from an overpressurization event at low temperatures are conspicuously threatening to the integrity of the RCPB. Therefore, BTP RSB 5-2 I requires protection of the Appendix G limits to meet the criteria established in 10 CFR 50, Appendix A. The LTOP system is required to protect the Appendix G limits, a standard requirement, or to limit the maximum pressure in the vessel to 110% of the pressure which satisfies the stress intensity factor equation defined in Reference 10, paragraph G-2215. The latter requirement, which was first introduced by Reference 11, effectively increases Appendix G P-T limits by 10% to arrive at LTOP P-T limits. As indicated in Section 3.1.1, an exemption must be obtained from the NRC for the use of the LTOP P-T limits as a basis for the LTOP setpoints and limitations. I LTOP is a combination of measures that ensure that the applicable P-T limits will not be exceeded during heatup, cooldown, and shutdown operations. The LTOP range is the operating condition when any RCS cold leg temperature is less than the applicable LTOP enable temperature and the RCS has pressure boundary integrity. The RCS does not have pressure boundary integrity when it is open to containment with a minimum area of the opening greater than, or equal to, a value specified in TS for a vent. The vent shall be capable of mitigating I the limiting LTOP events and the vent area shall be justified by analysis. The LTOP enable temperature is a temperature under which the LTOP relief valves must be aligned to the RCS for automatic protection. ABB Combustion Engineering Nuclear Power 32 CE NPSD-683 Rev. 03

I As a minimum, an LTOP system may include relief valves with a single setpoint that must be aligned below the enable temperature, and restrictions on RCS heatup and cooldown rates. Such a system would result when the P-T limits are not overly restrictive, the LTOP relief valves are of high capacity, and the relief valve setpoint allows for an acceptable operating window. Conversely, if the P-T limits are restrictive, the LTOP relief valves are small, and/or the operating window is challenged, the LTOP system may include a combination of valves, power-operated relief valves (PORV) with multiple setpoints, or with a variable setpoint controlled as a function of reactor coolant temperature. Other restrictions may be added to make the LTOP system adequate. The LTOP-related limitations are currently contained in TS, along with the applicable Appendix G P-T limit figures. P-T limits, except those for the reactor vessel, do not typically change, as these apply to the RCPB components that are not subject to irradiation. P-T limits based upon the reactor vessel beltline do change with time due to irradiation. As a result, every time P-T limits change, TS are E affected to incorporate new P-T limits and LTOP requirements. The TS E LTOP requirements may also be affected by plant modifications if these adversely impact LTOP analyses. Generic Letter 96-03 gives utilities the opportunity to avoid TS revisions due to changes in P-T limits by relocating the appropriate limits to plant-specific PTLRs. GL 96-03 establishes the conditions under which Lv0P system limits can be relocated from TS to a plant-specific PTLR. Attachment 1 to GL 96-03 specifies the requirements for the methodology that must be provided in the methodology report, which is the prerequisite to the PTLRs. According to Provision 3 of Attachment 1, the L'ICP methodology should include a description of how the LTOP system limits are calculated applying system / thermal hydraulics and fracture mechanics and should reference SRP 5.2.2, ASME Code Case N-514, ASME Coda Appendix G, Section XI, as applied in accordance with 10 CFR 50.55. ABB Cw.bustion E:3gineering Nuclear Power 33 CE NPSD-683 Rev. 03 1 _-_--___-_-__0

i

    ;                  The following sections describe the ABB CENP LTOP methodology that has been used to develop and analyze the LTOP systems for several of the 3                  CEOG plants and that should be adhered to in the plant-specific PTLRs and revisions thereof. Based on GL 96-03, following the initial NRC approval of this topical report and any plant-specific PTLR report that has this topical report as its basis, changes to LTOP setpoints and limitations due to regulatory changes, modifications to LTOP analysis methods and major assumptions, and/or LTOP system redesigns will require the NRC approval prior to implementation.

I 3.2 GENERAL METHODOLOGY 3.

2.1 DESCRIPTION

The ABB CENP LTOP methodology makes use of an iterative process in the I determination of LTOP limitations, which balances the adequacy of the LTOP system with acceptable operating restrictions. The methodology is based upon a presumption that an adequate LTOP system can be designed in more than one way by varying assumptions and targets such that the resulting operational restrictions are the most optimal. As an example, keeping the existing relief valve set point but further restricting the RCS heatup and cooldown rates may be more beneficial than keeping the rates but reducing the setpoint, which, in turn, will reduce the operating window. The utility will decide on the optimal I approach. LTOP is a safety-related function, as it protects the RCPB integrity. As a result, any analysis in support of LTOP shall be quality assured. Also, all LTOP-related revisions shall meet the requirements of BTP RSB 5-2 (Reference 12). It is imperative to refer to BTP RSB 5-2 while doing LTOP-related analyses. I I l l ABB Combustion Engineering Nuclear Power 34 CE NPSD-683 Rev. 03

I 3.2.2 LTOP EVALUATION COMPONENTS The analyses that support the determination of the LTOP requirements generally fall into three major analytical areas:

1) Analysis of P-T limits for use as operating guidelines and as a basis of LTOP requirements. The methodology for P-T limits is detailed in section'5.0.

2) Analyses of postulated overpressure events in the RCS, including energy addition (RCP start) and mass addition events. There analyses yield peak transient pressures which are compared with the P-T limits to identify LTOP-related limitations. The sources for the transients most often remain unchanged. However, changes in operational practices and plant configuration may cause changes in the applicable transients and/or temperatures.

3) L'IOP evaluation, which compares the applicable P-T limits and peak transient pressures to identify the LTOP-related limitations. LTOP evaluations may have different 3 objectives, depending on (1) whether a new LTOP system is designed, or (2) the current LTOP limitations need to be verified due to new P-T limits and/or revised pressure transient analysis, or (3) the existing limitations need to be relaxed.

The following sections describe the methods to be used in the various analyses that comprise the LTOP evaluations. 3.3 TRANSIENT ANALYSIS METHODOLOGY 3.3.1 LIMITI!G EVENT DETERMINATION The determination of LTOP-driven restrictions is basod upon the consideration of multiple requirements. Currently, 10 CFR 50 Appendix A requires that the initiating event must be established considering any condition of normal operation including anticipated ABB Combustion Engineering th2 clear Power 35 I CE NPSD-683 Fev. 03

i operational occurrences. Anticipated operational occurrences are defined therein as those conditions of nonnal operation which are I expected to occur one or more times in the life of the nuclear power unit. According to BTP RSB 5-2, 'All potential overpressurization events should be considered when establishing the worst-case event'. All potential causes of RCS overpressurization at low temperatures in the CEOG plants have been considered while the LTOP systems were being designed. Out of those, two types of events were then determined to challenge LTOP systems the most. They ares I (1) inadvertent energy addition to the RCS during an RCP start with the secondary steam generator inventory at a higher temperature than reactor coolant, and I (2) mass addition to the RCS during operation of HPSI pumps and/or charging pumps that results from an inadvertent Safety Injection Actuation Signal (SIAS). Presently, the RCP start continues to be the most limiting energy addition event. With respect to mass addition, the most limiting event is simultaneous operation of two HPSI and three charging pumps, or a combination with a highest flow rate, as allowed by TS. The applicability of the most limiting mass addition event may extend over the entire LTOP range, or may be restricted to a certain temperature range, in accordance to TS. If the applicability of the most limiting event is restricted, then other mass addition events, with a smaller number of operating pumps and/or flow rate restrictions (as allowed by I TS), are the limiting events at other temperatures. An additional qualifier for the limiting events is pressurizer water level. This is one of the design bases for LTOP limitations. Each energy addition and mass addition event's definition shall be supplemented by this parameter as 'under water-solid conditions" or *with a pressurizer steam volume of...t (or cu ft) . " If LTOP setpoints and limitations are i ABB Combustion Engineering Nuclear Power 36 L CE NPSD-683 Rev. 03 e

                                                                                     - . . m

l l l based on a transient analysis that assumes a steam volume in the IlI pressurizer, this limitation shall be included into TS, although the numerical value may be placed it the plant-specific PTLR. The analysis shall account for pressurizer level (volume) instrumentation uncertainty. I' 3.3.2 APPROACH AND MAJOR ASEUMPTIONS The limiting events shall be analyzed for each pump combination (mass addition), with each applicable means for transient mitigation, and for the most limiting fluid conditions in the RCS and pressurizer. If an LTOP system comprises of two or more PORV setpoints or a variable E PORV setpoint, and water-solid conditions in the pressurizer may exist during PORV discharge for transient mitigation, the transient analyses shall assume water-solid conditions and shall be performed for two or several setpoints to obtain peak pressure as a function of setpoint. . I l Similarly, if several HPSI and/or charging pump combinations may be l operable within the LTOP temperature range, each shall be analyzed with each applicable setpoint and water-solid conditions. The transient analyses should assume the most limiting allowable 3 operating conditions and systems configuration at the time of the postulated cause of the overpressure event, as required by BTP RSB 5-

2. Consequently, unacceptable peak pressures may result if only bounding analyses are performed, as these analyses typically assume the most limiting fluid conditions and plant configurations over the entire LTOP range. As an alternative, a transient analysis can be performed in a parametric manner for two or more initial reactor coolant temperatures, pressurizer water levels, RCS pressures, decay i heat rates, etc. Such an approach will yield lower peak pressures at the less limiting fluid conditions (where these apply), while l l

producing the bounding peak pressure values at the most limiting conditions that would only be applicable in a narrower temperature range. This approach will also benefit the LTOP evaluation, as a peak pressure database will be generated that can facilitate meeting the ABB Combustion Engineering Nuclear Power 37 I CE NPSD-683 Rev. 03

I ultimate goal - protection of the P-T limits with minimum operational limitations, a sufficient operating window, and best possible heatup and cooldown rates. If an operating restriction is introduced that either reduces the severity of the transient or eliminates it altogether (such as a limitation on pressurizer water level or racking out power to a HPSI pump), that restriction shall be placed in TS according to BTP RSB 5-2. Both energy and mass transient analyses will use the following major assumptions, unless a less restrictive approach is justified:

                 =

When the transient is mitigated by relief valves, only one valve shall be used in the transient analysis. This assumption meets the single failure criterion of BTP RSB 5-2. Past studies demonstrated that unavailability of one relief valve is the most limiting single failure with respect to the peak transient pressures. Relief valve I discharge characteristics shall be selected as indicated in Section 3.3.3.

                =

A pressure transient can be mitigated by pressurizer steam volume prior to reaching the relief valve setpoint. This assumption is optional, as it allows identification of the initial fluid conditions (e. g., pressurizer water level) under which a transient can be mitigated without challenging the relief valve. The analyses using this assumption will alsp assume a minimum time for an operator to initiate action to mitigate the event (see the assumption on operator inaction time below). -

               =

No credit may be taken for letdown, RCPB expansion, and heat absorption by the RCFB for transient mitigation. This assumption places the entire burden for transient mitigation on a single relief valve or pressurizer steam volume. l l ABB Combustion Engineering Nuclear Power 38 L CE NPSD-683 Rev. 03 r

l

• A water-solid pressurizer shall be assumed, with water at saturation at the initial pressure.

This assumption shall be used for bounding analyses. The assumption expedites the transient response and reduces g discharge flow rates for the PORVs and relief valves on the E pressurizer, as it reduces water subcooling at the inlet. If analyses are performed for other conditions as well, less limiting fluid conditions in the pressurizer may be justified, such as subcooled water or steam.

• Heat input from pressurizer heaters' full capacity shall be assumed.

This input increases transient pressure. e Decay heat shall be assumed as an additional input to maximize reactor coolant emansion. This assumption increases the peak pressure and is the result of an assumed loss of shutdown cooling heat removal capability. The method for calculating decay heat rate shall account for a reasonably fast cooldown to reach a temperature in the LTOP range aftcr reactor shutdown and reasonable time span before heatup is initiated. Decay heat input may not have to be included, if shutdown cooling can be relied upon. A justification will M re to be provided.

• Operator inaction time is 10 minutes.

This assumption requires analyzing the transients for 10 minutes from the start, except for a water-solid condition when peak pressure is reached within the first 20 seconds. Less time can be used if justified. If the plant can demonstrate that it g would take less than 10 minutes for the operator to recognize 5 and mitigate (terminate) the transient, less time can be usec. The justification shall be approved by the lac. e PORV setpoint for the analyses shall be greater than the nominal setpoint to account for the actuation loop uncertainty and pressure accumulation due to finite PORV opening time. ADB Combustion Engineering Nuclear Power 39 I CE NPSD-683 Rev. 03

[- ) This assumption recognizes that due to loop instrumentation uncertainties, the PORV may start its opening at a higher pressurizer pressure than the nominal setpoint (if the loop

                      " reads" low). Additionally, it accounts for pressure accumulation above the opening pressure during the time delay

[ between the signal initiation and the valve plug reaching the full flow position. See Section 3.3.3 for further discussion. 3.3.3 LTOP RELIEF VALVES 3.3.3.1 General Description Current ABB CENP system designs incorporate LTOP relief capability during low temperature operation of the RCS. This is done in one of several ways. LTOP is provided by either two PORVs on top of the ( pressurizer, two dedicated relief valves on top of the pressurizer, relief valves in the shutdown cooling (SDO) suction line, or a combination of the PORVs and SDC relief valves. The PORVs and the relief valves on the pressurizer are the only LTOP f relief valves with a setpoint that can be adjusted with relative ease. A change in the PORV or relief valve setpoint can be factored into the { LTOP transient analyses if needed, as these setpoints are for LTOP only. The SDC relief valves, on the other hand, are spring loaded relief valves with a fixed setpoint, whose main function to protect the SDC system. A setpoint change is not typically an option in the LTOP transient analyses involving these valves. The specifics of each type with respect to transient analyses are discussed below. [ 3.3.3.2 Power-Operated Relief Valves The PORVs at the CEOG plants are fast acting pilot operated valves, with stroke times of the order of milliseconds. l ~ ABB Combustion Engineering Nuclear Power 40 CE NPSD-683 Rev. 03

Note: A greater opening time is typically assumed in the analyses for consistency with the acceptance criteria during PORV testing. The PORVs may pass subcooled water, saturated water, and/or, steam, depending on the pressurizer conditions during transient mitigation. PORV discharge characteristics. for these fluids shall be developed, using appropriate correlations and a conservative back pressure, as applicable. Especially important is to account for discharge flow reduction due to flashing at the valve outlet when the discharged water has a low degree of subcooling. The characteristics, in a form of curves, should relate valve discharge flow rate with either PORV inlet pressure or pressurizer pressure, cover the anticipaced pressure range, and not be related to a setpoint. PORV inlet piping pressure drop should be taken into account in the curve in terms of pressurizer pressure. The curves shall be used in the transient analyses. PORV actuation loop instrumentation uncertainty and PORV opening time shall be accounted for in the determination of a conservative value j for the PORV opening pressure at the rated flow position. The addition of the uncertainty to the nominal setpoint will determine pressure at g the beginning of the opening, whereas addition of pressure B accumulation during the opening time will determine the highest pressure at the opening. The actuation loop instrumentation uncertainty shall be determined using guidance contained in Regulatory Guide 1.105 (Reference 20) and ISA Standard S67.04-1994 (Reference 13). Note that the 1994 Edition g of the Standard has not yet been approved by the NRC. The 1982 Edition E should be used, as it has been approved by the NRC. For development of a PORV setpoint curve for a continuously variable setpoint program, a conservative adjustment for uncertainty shall be applied to the entire curve. Alternatively, the curve can be divided into segments and an uncertainty for each segment shall be determined, based on the segment's slope. The PORV opening time shall be consistent with the acceptance criteria during in-service testing of the subject PORV (see the Note above). ABB Combustion Engineering Nuclear Power di CE NPSD-683 Rev. 03 1

[ The transient analyses shall assume a conservative PORV opening characteristic, which can be simplified by an assumption that during the opening time period, the PCRV remains closed and then opens instantaneously. Pressure accumulation during this time shall be added to the opening pressure (which is the nominal setpoint corrected for [ uncertainty) to obtain the maximum pressure at the opening which shall be used in the transient analyses. [ Should a setpoint change be contemplated, one or more new setpoints can be assumed and analyzed to provide a peak pressure vs. setpoint function for the LTOP evaluation. [ 3.3.3.3 SDC Relief Valves The SDC relief valves pass subcooled water, due to their location in ( the SDC system piping inside containment. The opening and discharge characteristic for these valves shall be consistent with the ASME s tandarn', 'or spring loaded safety relief valves and/or manufacturer's recommendat.ons. Typically, these valves start opening at 3% accumulation above the set pressure and reach rated flow position at lot accumulation. A setpoint change is not typically contemplated in UIOP transient analyses involving these valves, because of their function to also support SDC operation. [ 3.3.3.4 Pressurizer Relief Valves [ The safety relief valves on top of the pressurizer are the sole means for LTOP in one CEOG plant. These valves may pass subcooled water, saturated water, and/or steam, depending on the pressurizer conditions during transient mitigation. The opening and discharge characteristic ( for these valves shall be consistent with the ASME standards for spring leaded safety relief valves and/or manufacturer's recommendations, whichever is more conservative. Similar to the SDC { relief valves, these valves are spring loaded safety relief valves ABB Combustion Engineering Nuclear Power 42 CE HPSD-683 Rev. 03

I with a fixed setpoint. A setpoint change can be considered in the LTOP analyses involving these valves. 3.3.4 ENERGY ADDITION EVENT An energy addition event can take place when the RCS is cooled via shutdown cooling, while the steam generators (SG) remain at a higher temperature. A temperature difference between tl.e secondary side'of the SG and reactor coolant will transfer heat in the SG tubes to the reactor coolant, thus raising coolant temperature and pressure. With a water-solid pressurizer, pressure quickly reaches the relief valve opening pressure, the valve then opens and starts to discharge. If the relief valve is a PORV and its capacity at the opening exceeds the flow rate equivalent to the resulting coolant expansion, the transient will be mitigated at the opening pressure and the valve may reclose at the reseat pressure only to open again as pressure rises to repeat the cycle. This valve cycling will continue until the cause of the transient is eliminated. The peak pressure in this case will be the maximum opening pressure. If PORV capacity at the opening is less than the transient input, I pressure will rise until equilibrium is reached, at which point discharge matches input. That equilibrium pressure will be the peak pressure in the transient. In the case of a SDC relief valve or a pressurizer relief valve, the { l peak pressure at the inlet, which will also be the equilibrium i t pressure, will either be maintained below 10% accumulation, if valve I capacity exceeds the input, or above 10% accumulation if a higher inlet pressure is needed to mitigate the transient. In case of the pressurizer relief valve, the obtained pressure at the valve inlet needs to be adjusted to the pressurizer by adding the inlet piping pressure drop. l l In the case with a steam volume in the pressuriter, the ==v4== pressure can be reached either prior to the valve opening, or after ABB Combustion Engineering Nuclear Power 43 l CE NPSD-683 Rev. 03 l l l 1

the opening during steam discharge, or after the opening but during water discharge, if the pressurizer becomes water solid within 10 minutes (or the justified operator inaction time) from the transient initiation. The analytical model for analysis of this event under water-solid conditions that ABB CENP uses includes equations for calculating heat transfer in the heated portions of the SG tubes from the secondary SG inventory to the reactor coolant. The model calculates fluid temperatures, specific volumes, relief valve discharge flow rates (after the valve opens), and other transient parameters every time increment. For a better accuracy, a 0.1 - 0.5 seconds increment is typically assumed in the analyses. Computer codes or hand calculations can be utilized for analyses of this event under other initial conditions, provided that the initial conditions are controlled as LTOP limitations in the TS. If analysis methods change, they shall be approved by the NRC prior to the use. A number of conservative assumptions are typically used in the analyses of this event to maximize peak pressures, in addition to those described in Section 3.3.2. These include: additional heat input from the RCP, fluid properties and heat transfer coefficients determined at the highest reactor coolant temperature, and instantaneous RCP start. 3.3.5 MASS ADDITION EVEt?? A mass addition event can take place whenever a HPSI and/or charging pump is aligned to the RCS. An inadvertent SIAS is assumed to initiate mass injection to the RCS from all the aligned pumps. The relief valve behavior in a mass addition event is similar to that described for an energy addition event (Section 3.3.4). As a different number of HPSI pumps and/or charging pumps may be operable in a particular temperature region, each pump combination represents an analytical I case and should be analyzed, rather than postulating the worst ABB Combustion Engineering Nuclear Power 44 CE NPSD-683 Rev. 03

possible combination over the entire LTOP temperature range. Mass addition is assumed to take place at the cold leg centerline and adjustments can be made to the pressurizer. The HPSI pump and charging pump inputs shall be maximized by addition of a conservative margin. The combined delivery of all operating pumps for a case is developed I in the form of a delivery curve representing flow to the cold legs as a function of pressurizer pressure. For analysis of this event, ABB CENP uses a method of equilibrium ,u pressures. The method consists of a superimposition of the relief valve discharge curve on the mass addition curve, both in terms of flow rate as a function of pressurizer pressure. The mass addition curve includes not only pump flow rates, but also energy inputs from decay heat, pressurizer heaters, and RCP (if operating) convertad into m equivalent flow rate. These additional flow rates are detennined by calculating reactor coolant temperature rise over an assumed period of time (10 minutes or as justified) resulting from these energy additions, which, in turn, determines reactor coolant expansion. The expansion is converted into the equivalent flow rate. That flow rate will have to be discharged by the relief valve. The pump delivery curve is shifted to the right by this additional flow rate value, which effectively increases the equilibrium pressure. The equilibrium pressure is determined at the intersection of the two curves. It signifies the pressure at which the mass input matches the relief a valve discharge flow rate. The equilibrium pressure is then compared with the maximum pressure at the valve opening (see Section 3.3.3) to identify the peak transient pressure. I  ! ABB Combustion Engineering Nuclear Power 45 CE NPSD-683 Rev. 03

                                                                                  }

3.4 LTOP EVALUATION METHODOLOGY 3.4.1 CRITERIA FOR ADEQUATE LTOP SYSTEM An adequate LTOP system will ensure that the applicable P-T limits are protected from being exceeded during postulated overpressure events with a minimal impact on plant operating flexibility. After the most limiting peak pressures from both the energy addition and mass additien transient analyses have been identified and linked to specific reactor coolant temperature range, these pressures are compared with the applicable P-T limits. As each LTOP limitation is terperature related, for it to be valid, the applicable P-T limit g pressure value shall be demonstrated to be above the applicable contra 111ng pressure at a given tecuperature. A controlling pressure is the cost limiting (greatest) transient pressure of all events postulated for the subject temperature range. 3.4.2 APPLICABLE P-T L1 HITS Presently, either Appendix G P-T 13mits or Appendix G P-T limits relaxed by a factor of 1.1 can be the basis for LTOP setpoints and limits. The relaxed Appendix G P-T limits, herein called LTOP P-T limits (see Section 3.1.1 for definitions), are developed using guidance provided in AIME Code Case N-514 iReference 11). Specifically, these limits represent 1100s of the pressure determined to satisfy Appendix G, para. G-2215 of ASME Code Section XI (Reference I 10) and are used exclusively in LTOP evaluations. As the use of the Code Case and the 1996 edition of the Code have not been formally approved by the NRC, an exemption must be obtained before it can be used to determine LTOP setpoints and limits. The Code Case's guidance has been incorporated into para. G-2215 of ASE Code Section XI, { Appendix G (Reference 10). Recently published ASME Code N-640 (Reference 21) can also be used for development of Aprendix G P-T limits, as indicated in Section 5.0. An exemption must be obtained to use the Code Case, as it has not been approved by the NRC. The methodology in Code Case N-640 yields less ? ABB Combustion Engineering Nuclear Power 46 L* CE NPSD-683 Rev. 03 P ~

limiting Appendix G P-T limits as compared to Reference 9. The P-T limits based on Reference 21 cannot be adjusted up by 10% as those resulting from Code Case N-514 or the 1996 Edition of the ASME Code. The P-T limits that are protected by LTOP are mostly those for the reactor vessel beltline (and flange, as applicable) and apply to RCS heatup, cooldown, and isothermal conditions. The P-T limits at the beltline are adjusted to the pressurizer using pressure correction factors. For TS, the P-T limits in terms of pressurizer pressure may or may not include pressure and temperature indication instrumentation uncertainties. As a basis for the LTOP evaluation, these adjusted P-T limits should not include pressure indication uncertaintias, but may include temperature indication uncertainty. If temperature indication uncertainty is not part of the P-T limits, it needs to be considered in the LTOP evaluation that detarmines LTOP-driven limitations. The plant-specific PTLR shall address , this issue. Pressure instrumentation uncertainty is accounted for in the determination of the PORV opening pressure, as described in Section 3.3.3.2. For the LTOP systems that use large capacity (over 1500 gpm) relief valves connected to the pressurizer, an adjustment must be made to account for the pressure differential between the reactor vessel and the pressurizer due to flow induced losses in the surge line. That pressure differential can either be included in the pressure correction factors for the P-T limits (see Section 5.3.1.7), or be added to the peak transient pressures. As this pressure differential is not present when the relief valve is closed, i. e., most of the time, using it for the adjustment of the P-T limits would unnecessarily restrict them at other times. Independent of which P-T limits are used as a basis for LTOP setpoints, the criterion for not exceeding these limits during postulated pressure transients remains valid.

. 4.3 L7CP ENABLE TDiPERATURES The LTOP system must be aligned and capable of mitigating any postulated overpressure event between the reactor vessel minimum ABB Combustion Engineering Nuclear Power                                          47 CE NPSD-kB3 Rev. 03

I I boltup temperature and the LTOP enable temperature. Exceptions to this requirement would be if the RCS were incapable of being I pressurized by establishing a sufficient vent area. The enable temperatures shall be determined by the guidance provided in BTP RSB 5-2, unless an exemption is obtained by the licensee that allows using the ASME Code Case N-514 methodology, which has been incorporated into Reference 10. Per BTP RSB 5-2, the LTOP enable. I temperature is defined as the water temperature corresponding to a metal temperature of at least RTg + 90*F at the beltline location (1/4t or 3/4t) which is controlling in the Appendix G limit calculation. The LTOP enable temperature shall account for the temperature gradient between the reactor coolant and metal at the contre,111ng location. Per ASME Code Case N-514 guidance, the LTOP enable temperature is at the greater of 200*F or the reactor coolant inlet temperature corresponding to a reactor vessel metal temperature less than RTg+ 50*F. The vessel metal temperature is the temperature at 1/4t at the beltline location. A single LTCP enable temperature value is typically determined for cooldown. With respect to heatup, however, LTOP enable temperature is a function of heatup rate. The finally selected LTOP enable temperature for heatup shall be that for the highest applicable heatup rate. The resulting enable temperatures are then corrected for instrumentation uncertainty, as applicable. A single value, equal to the greater of the two, may be used, if desired. Use of two values, one for heatup and another for cooldown, is also acceptable. 3.4.4 L7DP-RELATED L1MIT.DC CONDITIONS FOR OPERATION I As the reactor vessel gets irradiated with time, the Appendix G limits become more restrictive and additional limitations may be placed on ABB Combustion Engineering Nuclear Power 48 CE NPSD-683 Fev. 03

I operation of the plant. These operational restrictions shall be placed into TS, in accordance with BTP RSB 5-2. Typical restrictions that are placed on plant operations are listed , below. These restrictions are in addition to P-T limits and relief valve setpoints that are always included in TS. This list is not intended to be complete or be applicable to every plant but is provided as an overview of possible restrictions.

1. RCS heatup and cooldown rates are restricted to rates lower than the RCS design rates.

2. HPSI flow is restricted by locking out power to the pumps or closing header isolation valves and locking out power to the valves while in the LTOP region.

3. Charging pump operation is limited by locking out power to the pumps and either closing an appropriate valve, or using "

another means that will result in at least two actions / failures that would be required to start a pump.

4. The number of operating RCPs is limited.

5. Water solid operation is restricted to a temperature region.

6. Limitations on start of the first RCP are specified that may include the secondary-to-primary temperature differential, pressurizer level, and/or initial pressure.

Per Generic Letter 96-03, the restrictions shall remain in TS, but curves and numerical values may be relocated into the plant-specific PTLR. Ii I ABB Combustion Engineering Nuclear Power 49 CE NPSD-683 Rev. 03 m i l I i

4.O METHOD FOR CALCULATING BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURE (ART) This section addresses provision 4 of Attachment 1 to Generic Letter 96-03 (Reference 3) for the calculation of the adjusted reference temperature (ART). The ART is determined in accordance with Regulatory Guide 1.99, I Revision 2 (May 1988) (Reference 16), " Radiation Embrittlement of Reactor vessel Materials". ART is determined as follows: 1 ART = Initial RTm + A RTm + Margin i Initial RTm is the reference temperature for the beltline plate or weld material as described in Section 2.0. A RTm is the shift in reference temperature calculated using a chemistry factor (from Table 1 or 2, as applicable, of the Guide based on the copper and nickel content) and a neutron fluence factor (using the neutron fluence at the vessel depth of i interest). The margin is the root mean squared value using the uncertainty in the initial RTm, ci, and the uncertainty in the reference temperature shift, c.a The uncertainty in the initial RTm , a s , for a measured value of RTm is based on the precision of the test method; the uncertainty for a generic value is the standard deviation of the data used to obtain the generic value' The reference temperature shift uncertainty, ca , for baso material (e.g. plates) is 17 'F and for welds is 28 *F. When credible surveillance data, as defined by Regulatory Guide 1.99, Revision 2, are available, the chemistry factor may be modified and the uncertainty in the shift in reference temperature may be reduced. The process is as described in the Guide and is discussed further in Section 7.0. [ Note: Upon issuance of a new revision of Regulatory Guide 1.99, the I ART calculation methodology will be evaluated and, if applicable, the new i methodology will be cited in subsequent revisions of the PTLR.)

• When using the generic value for welds made using Linde 0091, 1092 and 124 and

( ARCOS B-5 weld fluxes, RTm = -560F, and of =17'F. ABB Combuscion Engineering Nuclear Powar 30 CE NPSD-683 Rev. 03

• 5.0 APPLICATION OF FRACTURE MECHANICS IN CONSTRUCTING P-T CURVES

       -This section addresses Provision 5 of Attachment 1 to Generic Letter 96-03, Reference 3, on calculation of pressure and temperature limit curves.

It presents the analytical techniques and methodology for developing beltline pressure-temperature limits that are utilized in the composite RCS operating limits. The method is directly applicable to heatup, cooldown and inservice hydrostatic tests. 5.1 GENERAL The analytical procedure for developing operational pressure-temperature limits for the reactor vessel beltline utilizes the methods of Linear Elastic Fracture Mechanics (LEPM) found in the ASME Boiler and Pressure vessel Code Section XI, Appendix G, Reference 10, in accordance with the requirements of 10 CFR Part 50 Appendix G, Reference 1. For these analyses, the Mode I (opening mode) stress intensity factors are used for-the solution basis. The general method utilizes Linear Elastic Fracture Mechanics procedures

      'which relates the size of a flaw with the allowable-loading which precludes crack initiation. This relation is based upon a mathematical stress analysis of the beltline material fracture toughness properties as prescribed in Appendix G to Section XI of the ASME Code.

The reactor vessel beltline is analyzed assuming a semi-elliptical surface flaw oriented in the axial direction with a depth of one quarter of the reactor vessel beltline thickness and an aspect ratio of one to six. ' Itis postulated flaw is analyzed at both the inside diameter location (referred to as the 1/4t location) and the outside diameter location (referred to as the 3/4t location) to assure the most limiting condition is recognized. The above flaw geometry and orientation is the postulated defect size (reference flaw) described in Appendix G to Section XI of the ASME Code. I ABB Ccabustion Engineering Nuclear Power $1 CE NPSD-683 Rev. 03

At each of the postulated flaw locations, the Mode I stress intensity factor, Ky , produced by each of the specified loadings is calculated and the sunmation of the Ky values is compared to a reference stress 1 I intensity, KIR. K IR is the critical value of Ky for the material and temperature involved. The result of this method is a relat' ion of pressure versus temperature for each reactor vessel operating period which precludes brittle fracture. K IR is obtained from a reference fracture toughness curve for low alloy reactor pressure vessel steels as defined in Appendix G to Section XI of the ASME Code. This governing curve is defined by the following expression: KIR = 26.78 + 1.223 exp[0.0145 (T-ART + 160)) ksiYin where, K IR = reference stress intensity factor, KsiYin T = temperature at the postulated crack tip, 'F ART = adjusted reference temperature at the postulated crack tip, *F (determined in accordance with Section 4.0) For any instant during the postulated heatup or cooldown, KIR is calculated at the metal temperature at the tip of the flaw, and the value of adjusted reference temperature at that flaw location. Also, for any g instant during the heatup or cooldown the temperature gradients across the m reactor vessel wall are calculated (see Section 5.3) and the corresponding thernal stress intensity factor, KIT, is determined. Through the use of j l superposition, the thermal stress intensity is subtracted from the 1 available K IR to determine the allowable pressure stress intensity factor and consequently the allowable pressure. l In accordance with the ASME Code Section XI Appendix G requirements, the I l general equations for determining the allowable pressure for any assumed rate of temperature change during Service Level A and B operation are: 2Kyg + KIT < EIR (1) ABB Combustion Engineering Nuclear Power 52 CE NPSD-683 Rev. 03

   .i 1.5KIM + EIT < EIR (Inservice. Hydrostatic Test) where, K

IM = All wable pressure stress intensity factor, Ksi Yin K IT = Thermal stress intensity factor, Ksi Yin K IR = Reference stress intensity factor, Ksi Yin In addition, the 1995 ASME Code, Section XI, Appendix G has introduced relief for pressurized water reactors (PWRs) with low temperature ove.rpressure protection (LTOP) systems. Section XI specifies load and ( temperature conditions to provide protection against failure during . reactor start-up and shutdown operation due to low temperature overpressure events that have been classified as Service Level A or B events. When using the Section XI Appendix G requirements, the LTOP systems are effective at coolant temperatures less than 200 'F or at coolant temperatures corresponding to a reactor vessel metal temperature less than RTm + 50 'r, whichever is greater. The LTOP systems will limit f the maximum pressure in the vessel to 110% of the pressure determined to satisfy equation (1) defined above. (. Recently published ASME Code N-_640 can also be used for development of Appendix G P-T limits. Code Case N-640 outlines the criteria to be followed in order to use Kre as the basis for establishing the reference fracture toughness limit, K im, value for the vessel. Use of the Kc fracture toughness limit will yield less limiting Appendix G P-T limits as compared to the use of Kra, the current fracture toughness limit. However, the use of this Code Case is restricted as follows: If a licensees wishes to use K e as the basis for establishing 2 the K:n value for the vessel, then the licensee may not use Code case N-514 as the basis for establishing the setpoints for the Low Temperature Overpressure Protection (LTOP) system. ABB Combustion Engineering Nuclear Power 53 CE NPSD-683 Rev. 03

I Presently, P-T and LTOP limits can be developed using methodologies outlined in one or more of the following references: I

1. 1989 ASME Code Section III, Appendix G (Reference 9),

2. 1996 ASME Code Section XI, Appendix G (Reference 10),

3. ASME Code case N-514 (Reference 11), and

4. ASME Code Case N-640 (Reference 21).

At this time, only ASME Code Section III, Appendix G has been endorsed in 10 CFR 50.55a and approved by the NRC for the use in the establishing P-T limits. Consequently, P-T limits using ASME Code Section III, Appendix G is the basis for P-T/LTOP setpoints and limits at most of the CEOG plants. However, recently introduced refinements to these methods, Items 2-4 above, will safely relieve restrictive operating limits. Unfortunately, at this point in time, the 1996 Edition of the ASME Code has not been endorsed in 10 CFR 50.55a end the use of it, as well as the use of ASME Code Case N-514 and ASME Code Case N-640, has not been generically approved. Therefore, a plant specific exemption must be obtained from the NRC prior to the use of these refined methodologies. 5.2 DETERMINATION OF THE MAXIMUM STRESS INTENSITY VALUES  ; 1 Practices, methodologies and techniques that are utilized in the development of the pressure-temperature limits, along with justification of the aforementioned, are described briefly herein. Detailed technical m descriptions of the pertinent items are given in Sections 5.3 and 5.4. , l These limits have been developed to meet the requirements of 10 CFR 50 Appendix G. A brief technical description of the procedures practiced by ABB Combustion Engineering Nuclear Power to develop brittle fracture limits is given for the required components of the reactor coolant pressure boundary. These techniques are applicable to all ABB Combustion g g ABB Combustion Engineering Nuclear Power 54 CE NPSD-683 Rev. 03

Engineering Nuclear Power NSSS's. These techniques have been applied to nuclear power plants designed to ASME Code editions later than the Summer 1972 Addenda since the incorporation of Appendix G to 10 CFR 50 in 1973. These analytical techniques are based partially on Linear Elastic Fracture Mechanics (LEFM) and provide appropriately conservative design loadings for the ferritic components of the reactor coolant pressure boundary to preclude brittle fracture. I currently, the ferritic components of the reactor coolant pressure boundary specifically addressed by Appendix G to Section III of the ASME Code (Reference 9) are delineated as follows:

1. Vessels

2. Piping, Pumps and Valves

3. Bolting I The vessel is the only location for which a LEFM analysis is specifically I required by 10 CFR 50 Appendix G. The test and acceptance standards to which the other components are designed are considered to be adequate to protect against nonductile failure.

The reactor vessel regions considered in the analysis to establish brittle fracture limits are as follows: la. Beltline Ib. Vessel Wall Transition Ic. Bottom Head Juncture ld. Core stabilizer Lugs le. Flange Region If. Inlet Nozzle lg. Outlet Nozzle The " beltline

• refers to the region of the reactor vessel that immediately surrounds the reactor core and is exposed to the highest levels of fast neutron fluence. Typically the beltline is restricted to the large I cylindrical shell section o.' the RV below the vessel wall transition. For ABB Combustion Engineering Nuclear Power $$

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some plant designs, the beltline region may also include the vessel wall transition. Typically, in either case, the material with the highest ART value falls within the cylindrical shell region below the vessel wall transition. These locations have been analyzed utilizing the principles of LEFM described by Appendix G to Section III of the ASME Code. These analyses considered plant heatup, plant cooldown and an isothermal leak test.. A brief description of the general criteria follows. 5.2.1 GDIERAIa NETHOD In accordance with Appendix G, Section III of the ASME Code, the mode I (opening mode) stress intensity factor, Kr, is utilized and calculated at numerous intervals throughout the transient. The Kr is calculated at the crack tip of a postulated flaw. The postulated flaw size for the = considered locations, except the flange and nozzles, are assumed to have a depth equal to one-fourth of the section thickness and a length equal to 1-1/2 times the depth. At each of these structural locations, flaws are analyzed on the inside surface for cooldown transient events and at both the inside and outside surfaces for heatup transient events. The detertaination of the applied Ky is based on the results of a two dimensional heat transfer analysis and consideration of the primary 4 membrane stress, opm, primary bending stress, o 3, p secondary membrane stress, osm, and secondary bending stress, os b. The resulting Ky for each component of stress can be calculated as follows: K Im3

                               *M m x (membrane stress) 10 ,where i = p or s Kyg, = Mb x (bending stress) cid where i = p or s where M ,and Mb are defined in Appendix G, Section III of the ASME Code (Reference 9). For computational simplicity, equations A3-4 and A3-6 of WRC Bulletin 175 (Reference 14) were utilized, where gM and Mb is ABB Combuscion Engineering Nuclear Power                                             56 I

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equivalent to Mm and M bin Appendix G of the ASME Code (Reference 9), as follows: 1 1.1M yE and Mt= i 20 i M M .< g a I where: M,MB= g c rrection factors dependent on the ratios of crack depth to section thickness and crack depth to crack length Q = the flaw shape factor modified for plastic zone size T = the section thickness (in). 1.1MgE here f = p or s KIm "MtX 6fm " f im I KIb " Mb X#1b ~ ib here i = p or s For each point in the transient analyzed, the allowable pressure is i determined by comparing the reference stress intensity, KIR, to the applied stress intensity with a conservative factor of safety. The value g of K IR is obtained at the crack tip location based on the crack tip temperature for the specific time point in the transient and determined based on the following equation: KIR = 26.78 + 1.223 exp[0.0145(T-RTg + 160)) ksi Yin where, ABB Combustion D1gineering Nuclear Power CE NPSD-683 Rev. 03 b

I, T = crack tip temperature (*F) at 1/4T and 3/4T locations RTNDT = reference nil ductility temperature at each cracktip location ' For plant heatup and plant cooldown, the following expression is used to determine the allowable pressure: KIR > 2.0 Kyp + KIT Substituting, i KIR > 2.0 {KImp + Kyg,) + Gym, + KIbs PRIMARY SECONDARY l For leak tests, the expression utilized to calculate the Ky due to test pressure is: i KIR > 1.5 Kyp + KIT Substituting, KIR > 1.5 {KImp + KIbp) + {K Ims + Ibs} PRIMARY SECONDARY Table 1 of 10 CFR Part 50, Appendix G outlines the pressure and temperature requirements for the reaccor pressure vessel for the normal and hydrostatic pressure and leak tests operating conditions. The table provides specific guidance on P-T requirements for critical and non-critical core conditions. The guidance is centered en P-T limits developed using the fracture toughness methods of ASME Section XI, Appendix G. Table 1 of 10 CFR Part 50, Appendix G, also sets criteria to establish the minimum temperature requirements for the reactor vessel. Composite P-T limit curves are normally generated by calculating the most conservative P-T limit points established by using the methods of ASME ABB Combustion Engineering Nuclear Power $8 I CE NPSD-683 Rev. 03 I

( Section XI,' Appendix G, and the methods for the minimum temperature requirements. 4 The minimum temperature requirements for the reactor vessel, as required by Table 1 to 10 CFR Part 50, Appendix G, are as follows: For pressure testing conditions of the reactor coolant system (RCS), when the'RCS pressure is less than or equal to 20% of the preservice hydrostatic test pressure (PHTP), and the reactor core is not critical, the minimum temperature requirement for the reactor vessel must be at least as high as the adjusted reference temperature limiting material in the closure flange region stressed by bolt preload. For pressure testing conditions of the RCS, when the RCS pressure is greater than 20% of the PHTP and the reactor core is not critical, the minimum temperature requirement for the reactor vessel must be at least as high as the adjusted reference temperature for the limiting material in the closure flange region plus 90 'F. For normal operations, when the RCS pressure is less or equal to 20% of the PHTP and the reactor core is not critical, the minimum temperature requirement for the reactor vessel must be at least as high as the adjusted reference temperature for the limiting material in the closure flange region stressed by bolt preload. For normal operations, when the RCS pressure is greater than 20% of the PHTP and the reactor core is not critical, the minimum temperature requirement for the reactor vessel must be at least as high as the adjusted reference temperature for the limiting material in the closure flange region stressed by bolt preload plus 120 'F. For normal operations, when the RCS pressure is less than or equal to 20% of the PHTP and the reactor core _i_ss critical, the minimum ABB Combustion Engineering Nuclear Power $9 CE NPSD-683 Rev. 03

temperature requirement for the reactor vessel must be at least as high as the adjusted reference temperature for the limiting material in the closure flange region stressed by bolt preload plus 40 *F, or the 3 minimum permissible temperature for the inservice hydrostatic pressure test, whichever is larger. For normal operations, when the RCS pressure is greater than 20% of the Ii l PHTP and the reactor core ,i_s critical, the minimum temperature requirement for the reactor vessel must be at least as high as the adjusted reference temperature for the limiting material in the closure flange region stressed by bolt preload plus 160 *F, or the minimum permissible temperature for the inservice hydrostatic pressure test, whichever is larger. 5.2.2 FLAtCES The flange is analyzed assuming a flaw size of 0.75 inch and is smaller than a one-quarter depth flaw. This smaller flaw size is permitted by Article G-2120 of Appendix G to Section III of the ASME Code and is based on the ability to confidently detect this flaw size utilizing in-shop non-destructive examination (NDE) techniques (e.g., radiography, ultrasonic testing, etc.) and is consistent with the acceptance standards of sub-Article NB-5320 of Section III to the ASME Code. The applied Ky is determined utilizing equations A3-1 and A3-2 along with Figures A3-1 and A3-2 from NRC approved WRC Bulletin 175, Reference 14. The remainder of the procedures, as described previously are also = applicable to the flange region. S.2.3 N0ZZLES . In the case of the primary inlet and outlet nozzles the method described in Appendix 5 to NRC approved WRC Bulletin 175, Reference 14, Ky g Calculation Method for Nozzle, was utilized. ABB Combustion Engineering Nuclear Power 60 I CE NPSD-683 Rev. 03

In this analysis the postulated flaw size was equal to one-tenth of the vessel wall thickness and located on the inside corner of the nozzle adjacent the vessel wall. Again, the flaw size is confidently detectable with the in-shop NDE techniques and consistent with the acceptance standards of Sub-Article NB-5320 of Section III to the ASME Code. I The applied K y due to membrane stress are determined utilizing Equation A5-1 in conjunction with Figure A5-1, both from Reference 14. The bending I stress intensity factor is calculated in the same manner as the other locations. The solution for the allowable pressure is still based on K IR as the maximum allowable stress intensity factor for the particular crack tip temperature. The relations previously cited for heatup and cooldown, and leak test were applied in determining the applicable limits. The results of these analyses, in the unirradiated condition, show that I for heatup, cooldown and isothermal leak test, the limiting locations are the vessel shell at the vessel flange, the inlet nozzle and the upper I shell at the vessel wall transition, respectively. I I I l 1 1 ABB Combustion Engineering Nuclear Power 61 CE NPSD-683 Rev. 03

       . i-.            -
                                                                        . i       _-

S.2.4 BELTLDG I i i j In the development of operational limits, ABB analyzes the reactor vessel beltline region considering the predicted effects of neutron fluence over g a specific time period. The beltline region is the only location that E receives sufficient neutron fluence to substantially alter the toughness properties of the material. Therefore, the beltline region will likely become the controlling location when compared to the other reactor coolant system locations analyzed. ABB considers the beltline region to be g controlling, that is, the most limiting with respect to allowable pressure W at any specific temperature, when the shift in RTgg due to neutron radiation in the beltline causes the ART to be greater than the unirradiated RTy g of the surrounding locations. This philosophy is consistent with the guidance given in Standard Review Plan 5.3.2, Pressure-Temperature. Limits, Reference 15. Pressure-Temperature limits for the beltline are generated based on procedures described in Sections 5.3 through 5.4 in conjunction with the shift prediction methods of Regulatory Guide 1.99 Revision 2, Reference 16, to account for the reduction in fracture toughness due to neutron irradiation, a The operational limits as indicated in the control room account for the temperature differential between the reactor vessel base metal and the reactor coolant bulk fluid temperature. Corrections for elevation and flow induced pressure differences between the reactor vessel beltline and prossurizer are included. Pressurizer pressure indicator loop uncertainties are also included and consequently, the limits are provided on coordinates of indicated pressurizer pressure versus indicated RCS (cold leg) temperature. E l I ABB combustion Engineering Nuclear Power 62 I CE NPSD-683 Rev. 03

5.3 PRESSURE-TEMPERATURE LIMIT GENERATION METHODS 5.3.1 GENERAL DESCRIPTION OF PT LIMITS GENERATION 5.3.1.1 Process Description PT limits are generated via any one of the following three approaches to calculate P-Allowable. These are based on the same general method utilizing Linear Elastic Fracture Mechanics procedures. The primary difference between the approaches is in the calculation of the thermal stress . intensity factor, Kn, at the W T and M T crack tip locations. Once En is determined, the Appendix G, ASME Section XI requirement can be used to relate the size of a flaw with the allowable loading that precludes crack initiation, thus generating an allowable pressure. This relation is based upon a stress analysis of the reactor vessel beltline and upon experimental measurements of the beltline material fracture toughness properties, as prescribed in Appendix G to Section XI of the ASME Code, Reference 10. The general process to generate PT Limits is as follows: a) Determine the limiting adjusted reference temperature for the postulated 1/4T and 3/4T crack tip locations of the reactor vessel. b) Perform a thermal analysis of a set of constant rate heatup and ' cooldown transients on a particular vessel geometry to obtain through-wall temperatures. c) Calculata thermal stress intensity factor, Kn, at the postulated crack tips for each time point in each transient. d) Calculate material reference stress intensity factor, Kn, at the postulated crack tips for each time point in each transient. e) Calculate the transient P-Allowable by subtracting the thermal stress intensity factor, Kn, from the material reference stress intensity factor, Km, via the Appendix G requirement and so,1ving for the allowed pressure loading for each point in the transient which does not exceed this requirement, f) Calculate the Isothermal P-Allowable from the material reference stress intensity factor, Ka, via the Appendix G requirement and ABB Combustion Engineering Nuclear Power 63 CE NPCD-683 Rev. 03 l

solving for the allowed pressure loading which does not exceed this requirement (For the Isothermal condition, the thermal stress intensity factor, Kn, is assumed to be zero). g) Determine minimum P-Allowable as the minimum of the Heatup/Cooldown transient P-Allowable and the Isothermal P-Allowable at the postulated crack tips. (These results are tabularized and plotted as the Heatup/Cooldown PT Limits for a particular vessel) . The following sections provide additional detail as to some of the specifics outlined in the general procedure above. In addition, the analysis of heatup and cooldown transients are described and discussed. 5.3.1.2 Regulatory Requirement In accordance with the ASME Code Section XI, Appendix G, Reference 10, requirements, the general equation to be satisfied for any assumed rate of temperature change during Service Level A and B (Normal and Upset Loads, respectively) operation is s = 2Km + Kn < Kn Reference 10 where, Km = Allowable pressure stress intensity factor, Ksi

                                                                                       )

{ l Krr = Thermal stress intensity factor, Ksi b I Kra = Reference stress intensity factor, Ksi b 5.3.1.3 Reference Stress Intensity Factor ~ l At each of the postulated flaw locations, the Mode I stress intensity factor, Kr, produced by each of the specified loads, is calculated and the summation of the Kr values is compared to a reference stress intensity factor, Krn.. The result is a relationship of pressure versus temperature for reactor vessel operating limits that preclude brittle fracture. Kra is currently defined as Ka which is the lower bound of crack arrest critical K values measured as a function of temperature. Another material stress intensity factor, Krc, is based on the lower bound of static initiation critical K values measured as a function of temperature. Both Ku and Kre ABB Combustion Engineering Nuclear Power 64 CE NPSD-683 Rev. 03

are obtained from a reference fracture toughness curve for reactor pressure vessel low alloy steels as defined in Appendix G and Appendix A to Section XI of the ASME Code. These governing curves are defined by the following expressions: Ki a = 26.78 + 1.223e I ' ' '""T* **,n *2"" Reference 10 Krc = 33.20 + 2. 806e to.czour-m + soon

                                                             ,m        Reference 10 where, Krx    = Crack arrest reference stress intensity factor, Ksi 5

Kre = Crack initiation reference stress intensity factor, Ksi b I T = temperature at the postulated crack tip, *F RTun = adjusted reference nil ductility temperature at postulated crack tip, *F I For any instant during the postulated heatup or cooldown, K: A or Kre is calculated using the metal temperature at the tip of the flaw, as well as I the value of adjusted RTun at that flaw location. Note: The u e of Krc as the basis for establishing the reference fracture toughness limit, Kra, value for the vessel is currently outlined in ASME code N-640. Use of the Kre fracture toughness limit will yield less limiting AppendiY G P-T limits as compared to the use of Krx, the current I fracture toughness limit. restricted as follows: However, the use of this Code Case is I

                       - If a licensees wishes to use Kze as the basis for establishing the Kra value t'or the vessel, then the licensee may not use ASME Code case N-514 as the basis for establishing the setpoints for the Low Temperature Overpressure Protection (LTOP) system.

I Presently, on',y ASME Code Section III, Appendix G has been endorsed in 10 CFR 50.55a and approved by the NRC for the use in the establishing P-T ABB Combustion Engineering Nuclear Power 65 CE NPSD-683 Rev. 03

limits. Neither, ASME Code 514 or 640 has been generically approved by the NRC at this time. Consequently, a plant specific exemption must be obtained from the NRC prior to the use of either of these code cases. 5.3.1.4 Calculation of Allowable Pressure The Appendix G equation re3ating Krx, Kit, and Kra is rearranged as shown I below to solve for the allowable pressure stress intensity factor, Krx, as a function of time with the calculated Kia and KIT values. As shown in the following equation, che thermal stress intensity is subtracted from the E available Kr to determine the allowable pressure stress intensity factor E and consequently the allowable pressure: Kyg= IR - IT 2 where, K:n = Allowable pressure stress intensity factor as a function of coolant temperature, Ksi b Kra = Reference stress intensity factor as a function of coolant temperature, Ksi b Kzt = Thermal stress intensity factor as a function of coolant temperature, Ksi b I The allowable pressure is derived from the calculated allowable pressure g stress intensity factor, Krx, shown above. The value of Krx will depend on 5 the approaches discussed in Sections 5.3.3 through 5.3.5. 5.3.1.5 Analysis of HeatUp Transient I During a heatup transient, the thermal bending stress is compressive at the reactor vessel inside wall and is tensile at the reactor vessel outside wall. Internal pressure creates a tensile stress at the inside wall as well as the outside wall locations. Consequently, the outside ABB Combustion Engineering Nuclear Power 66 I CE NPSD-683 Rev. 03 I

wall location has the larger total stress when compared to the inside wall. However, neutron embrittlement, shift in material RTme, and reduction in fracture toughness are greater at the inside location than at the outside. Therefore, results from both the inside and outside flaw locations must be compared to assure that the most limiting condition is recognized. It is interesting to note that a sign change occurs in the thermal stress through the reactor vessel beltline wall. Assuming a reference flaw at the 1/4t location, the thermal stress tends to alleviate the pressure stress indicating that the isothermal steady state condition would represent the limiting P-T limit. However, the isothermal condition may not always provide the limiting pressure-temperature limit for the 1/4t location during a heatup transient. This is due to the difference between the base metal temperature and the Reactor Coolant System (RCS) fluid temperature at the inside wall. For a given heatup rate (non-isothermal), the differential temperature through the clad and film increases as a function of thermal rate, resulting in a crack tip temperature which is lower than the RCS fluid temperature. Therefore, to ensure the accurate representation of the 1/4t pressure-temperature limit during heatup, both the isothermal and heatup rate dependent pressure-temperature limits are calculated to ensure the limiting condition is recognized. These limits account for clad and film differential temperatures and for the gradual buildup of wall differential temperatures with time. To develop minimum pressure-temperature limits for the heatup transient, the isothernal conditions at 1/4t and 3/4t, 1/4t heatup, and 3/4t heatup pressure temperature limits are compared for a given thermal transient. The most restrictive pressure-temperature limits are then combined over the complete temperature interval resulting in a minimum PT curve for the reactor vessel beltline for the heatup event. 5.3.1.6 Analysis of Cooldown Transient f During cooldown, membrane and thermal bending stresses act together in tension at the reactor vessel inside wall. This results in the pressure stress intensity factor, Em, and the thermal stress intensity factor, Kn., acting in unison creating a high stress intensity. At the reactor vessel outside wall, the tensile pressure stress and the compressive thermal stress act in opposition, resulting in a lower total stress than at the I ABB Combustion Engineering Nuclear Power 67 CE NPSD-683 Rev. 03 u

I inside wall location. Also, neutron embrittlement, the shif t in hTn:rr, and the reduction in fracture toughness are less severe at the outside wall compared to the inside wall location. Consequently, the inside flaw location is limiting for the cooldown event. =- To develop a minimum pressure-temperature limit for the cooldown event, the isothermal pressure-temperature limit must be calculated. The isothermal pressure-temperature limit is then compared to the pressure- g temperature limit associated with cooling rate, and the more restrictive B allowable pressure-temperature limit is chosen, resulting in a minimum PT limit curve for the reactor vessel beltline. j 5.3.1.7 Application of output The pressure-temperature limits developed using the method described above

                                                                                        )

account for the temperature differential between the reactor vessel base metal and the reactor coolant bulk fluid temperature. However, uncertainties for instrumentation error, elevation, and flow induced differential pressure corrections are not accounted for and should be included by the plant when final P-T limits are developed. 5.3.2 THERMAL ANALYSIS METHODOLOGY The first step in P-T limits generation is a detailed thermal analysis of I E the reactor vessel beltline wall to calculate the Mode I thermal stress intensity factor, Kn. One dimensional, three noded, isoparametric finite elements suitable for one-dimensional axisymmetric radial conduction-i convection heat transfer are used. The vessel wall is divided into elements and an accurate distribution of teuperature as a function of radial location and transient time is calculated. Convective boundary , conditions on the inside wall of the vessel and an insulation boundary on the outside wall of the vessel are used in the analysis. Variation of material properties through the vessel wall is permitted thus allowing for the change in material thermal properties between the cladding and the base metal. = In general, the temperature distribution through the reactor vessel wall is governed by the partial differential equation, I ABB Combustion Engineering Nuclear Power 68 I CE NPSD-683 Rev. 03 I

pC =K + (Reference 19)

• O*

_br , subject to the following boundary conditions at the inside and outside wall surface locations (Reference 19, p. 109):

                                              &r At r=r i          -KBr = h (T-Te)
                                           &r i                where, At r=r o
                                          -=0 Br p      =         density, lb/ft s C      =         specific heat, Btu /lb *F K      =         thermal conductivity, Btu /hr-ft *F T     =         vessel wall temperature, 'F I                       r t
                               =
                               =

radius, ft time, hr h = convective heat transfer coefficient, Btu /hr-f p'-

                                         'F Te     a        RCS coolant temperature, 'F rg, r. =        inside and outside radii of vessel wall, ft The above expression is solved numerically using a finite element model to determine wall temperature as a function of radius, time, and thermal rate. The results are applicable to all methods presented below.

I 5.3.3 ABB CMP PT CURVE FETHOD I For the ABB CENP PT Curve Methodology, the reactor vessel beltline region is analyzed assuming a semi-elliptical surface flaw oriented in the axial direction with a depth of one quarter of the reactor vessel beltline thickness and an aspect ratio of one to six. This postulated flaw is analyzed at botih the inside diameter location (referred to as the 1/4t { location) and the outside diameter location (referred to as the 3/4t location) to assure the most limiting condition is achieved. The above flaw geometry and orientation is the maximum postulated defect size (reference flaw) described in Appendix G to Section III of the ASME Code, ABB Combuscion Engineering Nuclear Power 69 - CE NPSD-683 Rev. 03

i Reference 9. This methodology generates results at the crack tips based on unit loads of pressure and temperature as described in the following sections. 5.3.3.1 Calculation of Thermal Stress Intensity Factors, Kn-ASME Section III Appendix G, Reference 9, recognizes the limitations of the original method provided for calculating Kn because of the assumed temperature profile. Since a detailed heat transfer analysis results in time varying temperature profiles (and consequently varying thermal stresses), an alternate method for calculating Kn is employed as suggested by Article G-2214.3 of Reference 9. The alternate method employed uses a polynomial fit of the temperature profile and superposition using influence coefficients to calculate Kn.. The influence coefficients are calculated using a 2-dimensional finite element model of the reactor vessel. The superposition technique employed is temperature profile based rather I than the stress profile based which is typically used. A third order polynomial fit to the temperature distributions in the wall was used and is given by: T(x) =Co+C1 (1 */3) +C2 Il~ l h +C 3 (l>/ h) where, T(x) = Temperature at radial location x from inside wall surface g Co,C3 ,c2 ,C3 = coefficients in polynomial fit 3 x = Distance through beltline wall, in h = Beltline wall thickness, in These polynomial fit coefficients are utilized in determination of the g app 3ted stress intensity. = In the following section, temperature based influence coefficients, K *, for detentination of the thermal stress intensity factor, KIT, are ABB Combustion Engineering Nuclear Power 70 I CE KPSD-683 Rev. 03

discussed. The influence coefficients are dependent upon the geometrical parameters associated with the maximum postulated defect, and the geometry of the reactor vessel beltline region (i.e., ro /rg, a/c, a/t), along with the unit loading. { 1 5.3.3.2 Calculation of Allowable Pressure As presented above, the Appendix G equation relating Km, Kn, and Kra is rearranged to solve for the allowable pressure stress intensity factor, Y.m , as a function of time with the calculated Kra and Kn values. As shown in the following equation, the thermal stress intensity is subtracted from the available Krn to determine the allowable pressure stress intensity factor and consequently the allowable pressure: I 'IR IT l Kyg= 2 f where, . Km = Allowable pressure stress intensity factor as a function of coolant temperature, Ksi b Kra = Reference stress intensity factor as a function of coolant temperature, Ksi b Kn = Thermal stress intensity factor as a function of f coolant temperature, Ksi b For pressure loadings, unit values of the load distributions were used to ( compute the influence coefficients. The unit value chosen for internal pressure was 1000 psi. The general equation to compute the Mode I stress intensity factors for thermal and pressure loading conditions is as follows: 3 ( K (a) I

                         =     I C K$ b i=0     i 1 where, ABB Combustion Engineering Mtclear Power                                                71 CE NPSD-683 Rev. 03

I K7(a) = Total applied stress intensity factor due to loading condition at crack depth, a Ci = Polynomial coefficients from the curve fit to the temperature or stress distribution through the vessel wall E*g = Fracture mechanics influence coefficients for a specified loading condition for each term of the g polynomial expression for the temperature or stress E distribution through the vessel wall a = crack depth, in The Ky for each loading condition is then summed and compared to the allowable K IR to determine the allowable pressure. The allowable pressure is derived from the calculated allowable pressure , stress intensity factor, Km, shown above. For calculation purposes, the-allowable pressure can be represented by the following expression once the allowable pressure stress intensity factor is determined. K I P- Allowable = A IM where, P-Allowable = allowable pressure as a function of time or coolant temperature, Ksi Km = allowable pressure stress intensity factor, Ksi 5 Km' = pressure stress intensity factor for 1000 psia internal pressure as determined from a finite element model, Ksi b 5.3.4 STANDARD ASME PT CURVE NETHOD As intended, ASME Section III, Appendix G, Reference 9, provides sufficient guidance and direction through figures and text to perform Pressure-Temperature calculations in a straight-forward ABB Combustion Engineering Nuclear Power 72 j CE NPSD-683 Rev. 03

r fashion. The following outlines the ASME Appendix G calculational procedure used in this report to generate the allowable pressure. Beginning with Equation (1) of G-2215, the general equation for

       -determining the allowable pressure for any assumed rate of temperature change during Service Level A and B operation is:

2Km + Krr < Kra then, solving for Em, we have Km < (Kra - E n) /2 where Km = 4

• c. = %
• Pr/t where c. = Pr/t (Membrane hoop stress) substituting and solving for P-Allowable (ksi), we have P-Allowable < (Kra - Kn) *t / (2
• r *&)

where, P-Allowable = Allowable pressure, Ksi from Kn = Me

• AT.

where Mc from Figure G-2214-1, 1996 ASME Code or Figure G-2214-2, Pre-1996 ASME Code, and AT, = T(oD) - T(ID) from Heat Transfer Analysis at each time point (Section 5.3.3.2) - Kr, = Reference stress intensity factor, Ksi b , per Figure G-2210-1

              &            = From either Figure G-2214-1 Pre-1996 ASME Code, or via formula specified in 1996 ASME Code.

t = Base Metal Wall Thickness, in r = Base Metal Inner Radius, in This formulation is used in conjunction with the basic data identified above, along with a common through-wall temperature ABB Combustion Engineering Nuclear Power 73 CE NPSD-683 Rev. 03

• I analysis of the heatup and cooldown transients to generato P-Allowable.

5.3.5 1996 ASE ALTERNATE Kn. METHOD In 1996, ASE Section XI, Appendix G, Reference 10, an alternative approach to calculate Kn is offered as a substitute to the approach discussed in Section 5.3.4. This alternative is based on an influence coefficient approach which generates less conservative results than the standard ASME Section XI, Appendix G approach. As specified in G-2214.3 (b), the alternative approach is valid to calculate Kn for radial thermal gradients for any thermal stress distribution at any specified time for a M-thickness surface defect. The t umulation is very concise as to its application to plant heatup and cooldown cycles, and is specified as follows: For an inside surface defect during cooldown, En = (1.0359Co + 0.6322C2 + 0.4753C + 0.3855C 3 )h For an outside surface defect during heatup, Kn = (1.043Ce + 0.630C1 + 0.481C2 + 0. 4 01C 3 )h The coefficients Co, C 2 , C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using: a (x) = co + Cs (x/a) + c2 (x/a) * + c, (x/a) ' where x is a variable that represents the radial distance, in., from the appropriate (i.e., inside or outside) surface and a is the maximum crack depth, in. This formulation is used in conjunction with the basic data identified in g Section 5.3.4 along with a common through-wall temperature analysis, as 5 discussed in Section 5.3.2, of the heatup and cooldown transients to generate P-Allowable. P-Allowable is calculated as shown in the Standard ASME PT Curve Method, Section 5.3.4, with the appropriate substitution of Kn from this method. ABB Combustion Engineering Nuclear Power 74 CE NPSD-683 Rev. 03

i 5.4 TYPICAL PRESSURE-TEMPERATURE LIMITS This section presents example pressure-temperature limits for the reactor vessel beltline region and the reactor flange region. These limits were developed using the methods described in Section 5.1 through 5.3 and in I conjunction with the following information. [ Note: This information is not intended to be representative of all reactor vessels and is provided for illustration purposes only.] I Reactor vessel Data I Design Pressure = [2500) psia Operating Pressure = [2250) psia l Design Temperature = [650]*F Vessel I.R. to Wetted Surface = [87.227) in. Cladding Thickness = [5/16) in. Beltline Thickness = [8.625) in. 1 Material Cladding - [ Type 304 Stainless Steel) I Beltline - [SA-533 Grade B Class 1) 1 Beltline Adjusted Reference Temperature ., Flaw Location Adjusted RTyg (*F) [ 1/4 T [191.0) 3/4 T [137.0) Initial RTrg Flange Region = [+80]*F Piping, Pumps and Valves = [+90)*F ABB Combustion Engineering Ruclear Power 75 L CE NPSD-683 Rev. 03 r

l I Pressure and Ternperature Correction Factors AT = [46)'F [For Te < 200'F; AP = -77 psi (2 RCP's operating))

                                                                   ~

(For Te > 200*F; AP = -69 psi (3 RCP's operating)) S.4.1 BELTLINE LIMIT CtTRVES ' The beltline pressure-temperature limits calculated for heatup and  ! I cooldown are depicted in Figures 5.1 through 5.4 and have been developed i utilizing the ABB CDIP methodology described in Section 5.1 through 5.3. These figures provide the operating limitations for the beltline region in terms of an allowable pressure over the operating temperature range for , various linear rates of temperature change. Also, these figures have been corrected to indicated pressurizer pressure and cold leg temperature (T c). Depicted in Figure 5.5 is the beltline pressure-temperature curve for inservice hydrostatic test. This limit curve is typically developed for an isothernal condition. Again, this figure has been corrected to indicated pressurizer pressure and cold leg temperature. The purpose of this figure is to establish the minimum temperature corresponding to the required hydrostatic test pressure. Note that ABB Combustion Engineering Nuclear Power's practice is to recommend a minimum temperature for inservice hydrostatic test based on a test pressure corresponding to 1.1 times the design pressure. S.4.2 FLANGE LIMIT CCTRVES Tae vessel flange limits, resulting from the detailed analysis described in section 5.2.2, are shown in Figure 5.6. This figure has been corrected to indicated pressurizer pressure and cold leg temperature. I ABB Combustion Engineering Nuclear Power 76 CE NPSD-683 Rev. 03

5.4.3 COMPOSITE LIMIT CURVES The beltline pressure-temperature limits and flange pressure-temperature limits discussed in previous sections form the basis for the ecmposite limit curves. In addition, the requirements described in Section 5.2.3 are also considered when developing the co=posite RCS P-T limits. During the development of the composite limits, the heatup and cooldown rates are chosen based en numerous considerations. The issues involved in establishing the-a aaximum rates include the impact on the operating window, the selection of the 1.ow Temperature cierpressure Protection setpoint (s) , the plant's physical limitations, and the economical impact associated with loss of electrical power generation. The relative importance Of these items is different for each utility and therefore is not addreseed directly in this document. For the purpose of illustration, composite limits were developed for heatup and cooldown and are presented in Figures 5.7 and 5.8, respectively. These figures show arbitrary rates selected for heatup and cooldown that will be used to develop the PTLR figures. Included in the figures are all of the analyzed locations and additional requirements necessary to determine which specific location is controlling witn respect to operating temperature. Again, for the purpose of illustration, the minimum boltup temperature was conservatively established to be [80]*F and the lowest service temperature was established to be [ 19 6 ) *F . Both requirements are depicted as part of the composite heatup and cooldown limits. The composite limit curve for inservice hydrostatic test is shown in Figure 5.9. The minimum temperature for inservice hydrostatic pressure test, [322)*F was established based on a test pressure 'of [2427) psia (1.1 times normal operating pressure). The limitations associated with core critical operation are developed along with the PTLR figures. These are presented in Section 5.4.4. ABB Combustion Engineering Nuclear Power 77 CE NPSD-683 Rev. 03 t

S.4.4 OPERATIONAL LIMIT CURVES The operational limits developed for utilities are based on the composite limits presented in the previous section. 'Iypical representations of figures developed for inclusion in the PTLR are presented in Figures 5.10 and 5.11. Figure 5.10 presents typical heatup limits developed to protect the RCS from brittle fracture. Included with the actual heatup limits are the limits representing inservice hydrostatic test and limits pertaining to core critical operation. The core critical limits were established based on the requirements given in Section 6.1. In addition, the allowable rates utilized in development of the heatup limits are also given as maximum heatup rates for the appropriate temperature range. Figure 5.11 presents typical cooldown limits established to protect the RCS from brittle fracture. Again, limits representing inservice hydrostatic test are also present with the cenposite cooldown limits. The allowable rates, utilized to develop the cooldown limit curve, are also listed as maximum cooldown rates for the appropriate temperature range, f The limitations for critical operation of the core are usually not i presented as part of the cooldown PTLR figure. i 5.4.5

SUMMARY

This section describes methodologies and practices utilized in the development of reactor coolant system pressure-temperature limits. The methodology was developed to meet the specific criteria of 10 CFR 50, Appendix G, Fracture Tooghness Reg'tirements and 10 CFR 50, Appendix A, Design criterion 14 and Design C.riterion 31. l The current requirements imposed by 10 CFR 50, Appendix G, apply to I pressure-retaining components of the reactor coolant pressure boundary ABB Combustion Engineering Nuclear Power 78 CE NPSD-683 Rev. 03 l

q which are fabricated from ferritic material and apply to any condition of normal operation, including anticipated operational occurrences and system hydrostatic pressure tests. Section A.4 provides a list and an operational description of the conditions that require pressure-temperature limits. The method and analytical procedures used in the development of the reactor coolant system pressure-temperature limits are based on linear I elastic fracture mechanics techniques described in ASME Boiler and Pressure Vessel Code, Section XI, Appendix G, Fracture Toughness criteria for Protection Against Failure. As noted previously, the required loading conditions are described in Section A.4. As discussed in Section 5.2, the only component specifically requiring a LEFM analysis is the reactor vessel. Additional details on the reactor vessel locations that were analyzed and the technical methodology are also provided. I The results of the LEFM analysis performed for the reactor vessel provided the limiting locations in the unirradiated condition for heatup, cooldown I and isothermal leak test. The limiting locations considered are the vessel shell at the vessel flange, the inlet nozzle and the vessel wall transition region. These results are considered in the development of composite RCS operating limits. Typically, when the RCS operating limitt are developed for a specific time period, the be'etline becomes the most limiting location in the reactor vessel because of the effects of neutron irradiation. Therefore, when RCS operating limits are developed, the beltline is analyzed considering the effect of neutron irradiation in accordance with Regulatory Guide 1.99 Revision 2 (see Section 4.0 and I 7.0), and the vessel flange region is considered, as a minimum, per the requirements of 10 CFR 50 Appendix G (see section 6.0) . To illustrate the application of these methodologies and practices, RCS pressure-temperature limits are discussed in Sections 5.1 through 5.4 for a typical plant. Included is a description of the process utilized to develop composite limits which protect the reactor coolant pressure boundary from brittle fracture and typical technical specification figures which specifically address the requirements of 10 CFR 50 Appendix G ABB Combustion Engineering Nuclear Power 79 CE NPSD-683 Rev. 03 L

providing limits for normal operation, inservice hydrostatic test, and core critical operation. l i ABB Cottbustion Engineering Nuclear Power 80 CE NPSD-683 Rev. 03

                                                                               ~

l l 1

( ( FIGURE 5.1 APPENDIX G P-T UMITS HEATUP 2,500 ABB CENP PT CURVs METHOD 30 FM _ SO FM MFM 90 FM 2,000

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50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE, Tc, DEG. F Te < 200*F, AP = -77 psi ART Te 2 200*F. AP = -69 psi 1/4t = 191.0*F AT = +6*F 3/4t = 137.0*F ( MB Combustion Engineering Nuclear Power 81 L CE NPSD-683 Rev. 03

I FIGURE 5.2 APPENDIX G P-T LIMITS i HEATUP l 2,500 , , , , ABB CENP PT CURVE METHOD ,pm 40 FMR 60FMR

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e a, I I o 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE, Tc, DEG. F Te < 200*F, AP = -77 psi ART T,2 200*F, AP = -69 psi 1/4t = 191.0*F AT = WF 3/4t = 137.0*F ABB Co:rbastion Engineering Nuclear Power 82 l CE NPSD-683 Rev. 03

i FIGURE 5.3 APPENDIX G P-T LIMITS COOLDOWN I 2,500 ABB CENP PT CURVE METHOD I  : 2,000 a

a. _

I uJ g 1,500 l E E N

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0 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE, Tc, DEG. F I Te < 200*F, AP = -77 psi ART Te 2 200*F, AP = -69 psi 1/4t = 191.0*F I AT = +6*F 3/4t = 137.0*F I l ABB Cotr.bustion Engineering Nuclear Power 83 CE NPSD-683 Rev. 03 r-

FIGURE 5.4 APPENDIX G P-T LIMITS COOLDOWN 2,500 , , , ABB CENP PT CURVE METHOD

I 2,000 e

i CL -

ut 1,500 E cc - 5 _ E ~ 3 /

     $     1,000 E                        "

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o 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE, Tc, DEG. F Te < 200*F, AP = -77 psi ART Te 2 200*F, AP = -69 psi 1/4t = 191.0*F AT = +6*F 3/4t = 137.0*F ABB Combustion Engineering Nuclear Power 84 CE NPSD-683 Rev. 03

[. i ( FIGURE 5.5 APPENDIX G BELTLINE P-T LIMITS [ 2,500' , , HYDROSTATIC ABB CENP PT CURVE METHOD

                                  ~

2,000

c. -

uf g 1,500 W L - sc [ d - a ( @ 1,000 8 Y - {. _W

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500 0 ' 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE, Tc, 'F Te '< 2001, AP = -77 psi ART Te 2 200*F, AP = -69 psi 1/4t = 191.0T AT = +6*F 3/4t = 137.07 ABB Combustion Engineering Nuclear Power 85 L CE NPSD-683 Rev, 03

1 I FIGURE 5.6 APPENDIX G FLANGE LIMITS 2,500 r 2,000 10C FMR a

n. _

ur C 1,500 h _ m Fma Y - 5 N iy 1,000

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5 g 5 _ 500 I o 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE, Tc, DEG. F Te < 200*F, AP = -77 psi ART Te 2 200*F, AP = -69 psi 1/4t = 191.0*F j AT = +6*F 3/4t = 137.0*F ABB Combustion Engineering Nuclear Power 86 I CE NPSD-683 Fev. 03

FIGURE 5.7 COMPOSITE APPENDIX G P-T LIMITS HEATUP 2,500 l LCMEST SERVDE pg

                       .       TBPERATURE (19i'F)

ABB CENP PT CURVE METHOD g 2,000

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cc d us y E 1.000 E 8 f 4 / 9 - now,PRssERveE

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                      . _     _ MHNUM BOLTUP mPENTURE o

50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE, Tc, DEG. F Te < 200*F, AP = -77 psi ART Te 2 200*F, AP = -69 psi 1/4t = 191.0*F AT = +6*F 3/4t = 137.0*F ABB Combustion D3gineering Nuclear Power 87 CE NPSD-683 Rev, 03

                                                                                                                             )

I FIGURE 5.8 COMPOSITE APPENDIX G P-T LIMITS COOLDOWN 2.500 , , i i ABB CENP PT CURVE METHOD _ meEnATuRE - (19i F) 2,000

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g 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE, Tc, DEG. F Te < 200*F, AP = -77 psl ART T 2 200*F, AP = -69 psi 1/4t = 191.0*F AT = +6*F 3/4t = 137.0*F ABB Combustion Engineering Nuclear Power 88 CE NPSD-683 Rev. 03

FIGURE 5.9 COMPOSITE APPENDIX G P-T UMITS gg HYDROSTATIC RA ~ (1MF) ABB CENP PT CURVE METHOD

           !               :                                                          %       IELTLNE 1 500 i              :

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I , _ _ MINNUMSOLTUP TEMPERJTURE l _ o 50 100 150 200 250 300 350 400 450 500 l INDICATED RCS TEMPERATURE, Tc,'F Te < 200*F, AP = -77 psi ART l Te 2 200*F, AP = -69 psi 1/4t = 191.0*F AT = 46*F 3/4t = 137.0*F ABB Combustion D2gineering Nuclear Power 89 " CE NPSD-683 Rev. 03

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9ghO @ h 3e*2r=(Eg e e - km "d9$yOa gtS*{n,t g g Oy g "0 h N O RO'$y MG*, o,

6.0 METHOD FOR ADDRESSING 10 CFR 50 MINIMUM TEMPERATURE REQUIREMENTS IN THE P-T CURVES 6.1 INSERVICE HYDROSTATIC PRESSURE TEST AND CORE CRITICAL LIMITS Both 10 CFR Part 50 Appendix G and the ASIG Code, Section II, Appendix G require the development of pressure-temperature limits which are applicable to inservice hydrostatic tests. For hydrostatic tests perfonned subsequent to loading fuel into the reactor vessel, prior to core criticality, the minimum test temperature is determined by evaluating K,y the mode I stress intensity factors. The evaluation of K y is performed in the same manner as that for normal operation heatup and cooldown conditions except the factor of safety applied to the pressure stress intensity factor is 1.5 versus 2.0. From this evaluation, a pressure-temperature limit that is applicable to inservice hydrostatic tests is established. The minimum temperature for the inservice hydrostatic test pressure can be established conservatively by determining that the test pressure corresponding to 1.1 times normal operating pressure and locating the corresponding temperature. Hydrostatic testing of the reactor vessel after achieving core criticality is not allowed. Appendix G to 10 CFR Part 50, specifies pressure-temperature limits for core critical operation to provide additional margin during actual power operation. The pressure-temperature limit for core critical operation is based upon the following criteria. For vessel pressure less than or equal to 20 percent of preservice system hydrostatic test pressure, the criteria are that the reactor vessel temperature must be the larger of the minimum permissible temperature for the inservice system hydrostatic pressure test or the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload plus 40 'F, and be at least 40 'F higher than the minimum pressure-temperature curve for normal operation heatup or cooldown. For vessel pressure greater than 20 percent of preservice system hydrostatic test pressure, the criteria are that the reactor vessel temperature must be the larger of the minimum permissible ABB Combustion Engineering Ruclear Power 92 CE NPSD-683 Rev. 03

l temperature for the inservice system hydrostatic pressure test or the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload plus 160 *F, and be at least 40 *F higher than the minimum pressure-temperature curve for normal operation heatup or cooldown. Note that the core critical limits established utilizing this criterion are based solely upon fracture mechanics considerations. These limits do not consider core reactivity safety analyses that can control the temperature at which the core can be brought critical. 6.2 MINIMUM BOLTUP TEMPERATURE The minimum boltup temperature is established based on ASME Code Section III, Subparagraph G-2222.c. The recommendation is as follows from Reference 9 :

     " ... when the flange and adjacent shell region are stressed by the full intended bolt preload and by pressure not exceeding 20% of the pre-operational system hydrostatic test pressure, minimum metal temperature in the stressed region should be at least the initial RT    temperature for the material in the stressed region plus any effects of irradiation at the stressed regions.'

6.3 LOWEST SERVICE TEMPERATURE The lowest service temperature is defined by the ASME Code as "the minimum temperature of the fluid retained by the component or, alternatively, the calculated volumetric average metal temperature expected during normal operation, whenever pressure exceeds 20% of the pre-operational system hydrostatic test pressure *. The requirement is applicable to piping, pumps, and valves and is intended to protect these components from brittle fracture. ABB Combustion Engineering Nuclear Power 93 , CE NPSD-683 Rev. 03 Il

The lowest service temperature is established based on the limiting RT g for ferritic low alloy steel piping, pump, and valve materials in the primary coolant reactor pressure boundary. The lowest service temperature is the highest RTg for those materials plus 100*F. ABS Combustion Engineering Nuclear Power 94 CE NPSD-683 Rev. 03

r i ) 7.0 APPLICATION OF SURVEILLANCE CAPSULE DATA TO THE CALCULATION OF ADJUSTED REFERENCE TEMPERATURE This section addresses Provision 7 of Attachment 1 to Generic Letter 96-03 (Reference 3) on application of surveillance capsule data. l Data from the reactor vessel surveillance program are used for two related purposes. The original purpose was to provide a system to monitor the radiation-induced changes to the toughness properties and provide assurance that the vessel materials are not behaving in an anomalous l

manner. The second purpose is to provide plant specific data for reactor vessel integrity analysis. Irradiation of materials in the surveillance capsules exposes specimens which are representative of the reactor vessel beltline in an irradiation environment nearly identical to the environment for the vessel. The post-irradiation analysis of the surveillance capsule contents provides measurements of the neutron fluence and of the changes j in toughness properties of the surveillance plate and weld materials. ,

These data can be used to refine both calculations of the vessel fluence and predictions of the adjusted reference temperature for the beltline materials. . I When data are available from two or more capsules, an evaluation may be i performed to deterinine whether the data are credible as defined in

     . Regulatory Guide 1.99, Revision 2. The data are deemed credible if (1) one or more of the surveillance materials is controlling for that vessel with respect to the ART, (2) the Charpy data scatter does not cause ambiguity in the determination of 30 ft-lb shift, (3) the measured shifts are within ca of the shift predicted using Position 2.1 (2 04 if the l       fluence range is large), (4) the capsule irradiation temperature is comparable to that of the vessel, and (5) the correlation monitor material data, if available, is within the scatter band of the known data for that material. The credible data can then be applied following Position 2.1 of the Guide to calculate a new chemistry factor for that material and to reduce the standard deviation for shift by half. If the revised chemistry factor and reduced standard deviation from application of Position 2.1 result in a higher value of ART than from that calculated using Position 1.1, the revised values should be incorporated into the PTLR methodology.

If the Position 2.1 values result in a lower value of ART, either the ABB combustion D3gineering Nuclear Power 95 CE NPSD-683 Rev. 03 L

Position 2.1 values will be incorporated or the original PTLR methodology will be retained. When the plant-specific surveillance capsule data are credible in all respects except for the match of the surveillance material heat number to the controlling vessel material heat number and there are data for the controlling material heat number available from another plant, the plant-specific PTLR may utilize surveillance data from that other plant as the basis for the ART prediction methodology. If such data are employed, the source of the data will be identified, the correspondence of the material heat numbers will be confirmed, and the basis for the manner in which the E data are applied will be provided. The basis could be a previously generated safety evaluation report Which would be referenced or a newly generated evaluation in which the licensee's surveillance data and the sister plant surveillance data are assessed with respect to the credibility criteria of Regulatory cuide 1.99, Revision 2 and, in addition, with respect to irradiation environment factors (e.g., neutron spectrum and irradiation temperature). Some recent CEOG sponsored efforts which are applicable to this discussion are CEOG Task 621 which addresses methodology for the application of sister plant data and CEOG Task 904 which addresses methodology for the application of both plant-specific and sister plant data to refine ART calculations. I I I ABB Combustion Engu ring Nuclear Power 96 CE NPSD-683 Rev. 0: I

g r I 8.0

SUMMARY

OF RESULTS The results of this task provide a basis for the relocation of RCS P-T limits, LTOP setpoints, RV Surveillance and Neutron Fluence reporting requirements from the Technical Specifications to another controlled

document called a PTLR.

Methodology descriptions for developing RCS P-T limits, establishing LTOP setpoints,; calculating ARTi developing a RV Surveillance Program, and calculating Neutron Fluence to support the PTLR are provided in Sections 1-7 and is considered the topical report. A generic approach for the relocation of the detailed information for the affected Limiting Conditions for operation from the Technical Specifications based on GL 96-03 was used. A generic document, called an RCS Pressure and Temqperature Limits Report (PTLR), which contains the detailed information needed to comply with relocating the Limiting conditions for operation from the Technical Specifications can be developed based on information in the topical report. l An example PTLR and a sample Technical Specifications

• mark-up* are I provided in Appendices A & B, respectively. The exanple PTLR contains-typical LCO's for RCS P-T limits-and L'IOP requirements for ABB CENP plants and can be tailored for plant specific submittals. The sanqple Technical Specifications
• mark-up* is provided for illustrative purposes only. CEOG utilities should prepare specific
• mark-ups* of their current Technical Specifications for their individual submittals.

1' In conclusion,-this report provides an acceptable and referenceable generic basis for the creation of plant specific PTLR reports. 1 1 l ABB Combustion Engineering Nuclear Power 97 l CE NPSD-683 - Rev. 03 -

( [-

9.0 REFERENCES

f

1. Title 10, code of Federal Regulations, Part 50, Appendix G, Fracture Toughness Requirements, Federal Register, Vol. 60, No. 243, Decerber 19, 1995, page 65474.

2. U.S. Nuclear Regulatory Commission, Standard Review Plan 5.2.2, overpressure Protection, Revision 2, November 1988.

f 3. NRC GL 96-03, Relocation of Pressure-Temperature Limit curves and Low Temperature overpressure Protection System Limits, January 31, 1996. ,

4. CE NPSD-683, Rev 02, " Development of a RCS Pressure and Temperature 5 Limits Report for the Removal of P-T Limits and LTOP requirements from the Technical Specifications, CEOG Task 942", December 1997.

5. "The ROCS and DIT Computer Codes for Nuclear Design,' CENPD-266-P-A,

[ April 1983.

6. *C-E Methodology for Core Design Containing Gadolinia-Urania

( Burnable Absorbers," CENPD-275, Rev. 1-P-A, May 1988.

7. " Methodology for Core Designs Containing Erbium Burnable Absorbers,"

CENPD-382-P-A, August 1993.

8. J.F. Carew, et. al, " Pressure Vessel Fluence Calculation Benchmark

[ Problems and Solutions," !MREG/CR-6115 (Draft). I 9. ASME Boiler and Pressure Vessel Code Section III, Appendix G,

• Protection Against Nonductile Failure", 1986 Edition.

I

10. ASME Boiler and Pressure Vessel Code, 1996 Edition f - Section XI, Appendix A, " Analysis of Flaws".

ABB Combustion Engineering Nuclear Power 98 CE NPSD-683 Rev. 03

            - Section XI, Appendix G, " Fracture Toughness criteria for Protection against Failure".

11. Cases of ASME Boiler and Pressure Vessel Code, Case N-514, " Low Temperature overpressure Protection," section XI, Division 1, dated February 12, 1992,

12. U. S. Nuclear Regulatory Commission, Branch Technical Position RSB 5-2, 'Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures,* Revision 1, November 1988.

13. ANSI /ISA-S67.04, Part I-1994, 'setpoints for Nuclear Safety-Related Instrumentation, approved August 24, 1995.

14. WRCB 175 (Welding Research Council Bulletin 175), "PVRC Recommendations on Toughness Requirements for Ferritic Materials,*

August 1972.

15. U.S. Nuclear Regulatory Comission, Standard Review Plan 5.3.2,

           " Pressure-Temperature Limits", Rev. 1, July 1981.                      E l

16. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision l 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988.

17. Title 10, code of Federal Regulations, Part 50, Appendix H, Reactor I

Vessel Material Surveillance Program Requirements, Federal Register, Vol. 60, No. 243, December 19, 1995, page 65476.

18. AS'IM E185, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society .

for Testing and Materials, Annual Book of Standards, Volume 12.02.

19. " Heat Transfer, A Basic Approach," M. Cecati Ozisik, McGraw Hill '

Book Company, 1985. ABB Combustion Engineering Nuclear Power 99 CE NPSD-683 Rev. 03

20.  ?! . S . Nuclear Regulatory Connission, Regulatory Guide 1.105, Revision 2, February 1986, " Instrument Setpoint For Safety Related Systems"

21. Cases of ASME Boiler and Pressure Vessel Code, Case N-640,

            " Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", Approval Date-2/26/99, To be Published.

ABB Combustion Engineering Nuclear Power 100 CE NPSD-683 Rev. 03

p.,..w.--.....,.. k APPENDIX A

       .q EXAMPLE OF RCS PRESSURE AND TEMPERATURE LIMITS REPORT May 1999
                                                                                 /

ABB Combustion Engineering Nuclear Power CE NPSD-683 Rev. 03

                                                         .       3

[El06) UNIT [X] RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) taarrJ Not to be used for operation. For illustration only. [ Note: This example is fouratted so that the " plant specific information" or

  " optional" items are in Sold / Italic font and are enclosed in square brackets as shown on this page.)

ABB Combustion Engineering Nuclear Power CE NPSD-683 }tev 03

RCS. PRESSURE AND TEMPERATURE LIMITS REPORT FOR [MANEJ UNIT (21 Table of Contents Page

1.0 INTRODUCTION

A-4 2.0 GL 96-03 PROVISION REQUIREMENTS A-4 2.1 Neutron Fluence values A-4 2.1.1 Input Data 2.1.1.1 Materials And Geometry 2.1.1.2 Cross Sections 2.1.1.2.1 Multi-Group Libraries 2.1.1.2.2 Construction of the Mul.ti-Group Library-2.1.2 Core Neutron Source 2.1.3 Fluence Calculation 2.1.3.1 Transport Calculation 2.1.3.2 Synthesis of the 3-D Fluence

                 ~

2.1.3.3 Cavity Fluence Calculations 2.1.4 Methodology Qualification and Uncertainty Estimates 2.1.4.1 Analytic Uncertainty Analysis 2.1.4.2 Comparison with Benchmark and Plant-Specific Measurements 2.1.4.2.1 operating Reactor Measurements 2.1.4.2.2 Pressure Vessel Simulator Measurements 2.1.4.2.3 Calculational Benchmarks 2.1.4.3 overall Bias and Uncertainty 2.2 Reactor Vessel Surveillance Program A-4 ABB Combustion Engineering Nuclear Power CE NPSD-683 Rev. 03 A-2

RCS PRESSURE AND TF}ZPERATURE LIMITS REPORT FOR [NAMEJ UNIT (I] 2.3 LTOP System Limits A-5 2.3.1 LCO f3.2.2.2J Boration Systems Flow Paths - Shutdown A-6 m 2.3.2 LCO (3.2.2.3J Reactivity Control Systems Charging Pump - Shutdown A-6 2.3.3 LCO [3.3.2.2J Engineered Safety Features Actuation System Instrumentation A-7

2. 3.4 LCO f3.4.9.2J Pressure /'lemperature Limits - Reactor Coolant System A-7 2.3.5 LCO (3.4.23J Reactor Coolant System Power Operated Relief Valve A-7 2.3.6 LCO (3.5.3J Emergency Core Cooling Systems, ECCS Subsystems - Tavg < (325*FJ A-B 2.3.7 LCO [3.4.24J Reactor Coolant System Reactor Coolant Purp - Starting A-9 2.4 Beltline Material Adjusted Reference Temperature (ART) A-9 2.5 Pre sure-Temperature Limits using limiting ART in the P-T Curve calculation A-9 2.6 Minimum Temperature Requirements in the P-T curves A-10 2.7 Applicatien of Surveillance Data to ART calculations A-ll

3.0 REFERENCES

A-13 4.0 LIST OF FIGURES 4.1 [Name] Unit [A] P/T Limits [KJ EFPY I Heatup and Core Critical A-14 4.2 (NaaneJ Unit {AJ P/T Limits (KJ EFPY Cooldown and Inservice Test A-15 g 4.3 (Name] Unit [AJ P/T Limits (KJ EFPY E Maximum Allowable Cooldown Rates A-16 4.4 Maximum Allowable Heatup and Cooldown Rates, Single HPSI Pump I Operation A-17 ABB Ccmbustion Engineering Nuclear Power A-3 \ CE NPSD-683 Rev. 03 '

RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR [NAME] UNIT [X]

1.0 INTRODUCTION

This PTLR for [NAME] Unit [2] contains Pressure-Temperature (P-T) limits corresponding to [5J Effective Full Power Years (EFPY) of operation. In addition, this report contains Low Temperature Overpressure Protection (LTOP) specific requirements which have been developed to protect the P-T limits from being exceeded during the limiting LTOP event. The Technical Specifications affected by this report are listed below and are separated into the appropriate category; P-T limits or LTOP requirements. 2.0 GL 96-03 PROVISION REQUIREMENTS 2.1 Neutron Fluence Values The reactor vessel beltline neutron fluence has been calculated for the critical locations in accordance with the general methodologies as described in Section 1.0 of the main body of this document. The following discussion gives the results of the fluence calculation followed by the details of the calculational analysts for the [NAME] Unit [X]. The peak value (s) of neutron fluence (E > 1 MeV) at the vessel clad interface used as input to the Adjusted Reference Temperature (ART) calculations for [NAME] Unit [X] corresponding to [ locations on the vessel] for (Z] effective full power years (EFPY) is [3.6x20"] neutrons per square centimeter (n/cm*) with an associated uncertainty of i [....J. 2.1.1 Input Data 2.1.1.1 Materials and Geometry [ Details of materials and geometry in accordance with section 1.1.2 of Ref. 3.2] 2.1.1.2 Cross sections [Detatis of the cross sections used in accordance with section 1.1.2 of Jtef. l 3.2] l ABB Combustion Engineering Nuclear Power A-4 CE NPSD-683 Rev. 03

RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR [NAMEJ UNIT [XJ 2.1.1.2.1 Multi-group Libraries [ Details of the multi group cross section library in accordance with section 1.1.2.1 of Ref. 3.2) 2.1.1.2.2 Construction of the Multi-Group Library [Detakis of the construction of the multi group library in accordance with section 1.1.2.2 of Ref. 3.2] 2.1.2 Core Neutron Scurce [ Details of the core neutron source in accordance with section 1.2 of Ref. 3. 2]. 2.1.3 Fluence Calculation 2.1.3.1 Transport Calculation [Detatis of the transport calculation used in accordance with section 1.3.1 of Ref. 3.21 2.1.3.2 Synthesis of the 3-D Fluence [ Details of the 3-D fluence synthesis in accordanne with section 1.3.2 of Ref. 3.2) 2.1.3.3 Cavity Fluence Calculations [ Details of the cavity fluence calculation in accordance with section 1.3.3 of Ref. 3.2] 2.1.4 Methodology Qualification and Uncertainty Estimates I (Details of the methodology gn=14 fication and uncertainty estimates used in l accordance with section 1.4 of the main body] E 2.1.4.1 Analytic Uncertainty Analysis [ Details of the analytical uncertainty analysis in accordance with section 1. 4.1 of the main body) 2.1.4.2 Comparisen with Benchmark and Plant-specific Measurements [ Details of the comparisons with h=="""> and plant-specific measurements in accordance with section 1.4.2 of Ref. 3.2) 2.1.4.2.1 Operating Reactor Measurements [ Details of the reactor measurements comparisons with calculation in accor anne e with section 1.4.2.1 of Ref. 3.2) l 1 I ABB Combustion Engineering Nuclear Power A.S CE NPSD-683 Fev. 03 I

RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR (NAME7] UNIT (XJ i 2.1.4.2.2 Pressure Vessel Simulator Measurements (Details of the pressure vessel simn1= tor b=neh==* analyses performed in accordance with section 1. 4.2.2 of Ref. 3.2) 2.1.4.2.3 Calculational Benchmarks (Details of the calculational henchmark for methods gn=1sfication in accordance trith section 1.4.2.3 of Ref. 3.2] l 2.1.4.3 overall Bias and Uncertainty (Details of the overall bias and uncertainty an=1ysis in accordance seith section 1.3.1 of Ref. 3.2] I 2.2 Reactor Vessel Surveillance Program I The reactor vessel surveillance program and the surveillance capsule withdrawal are described in Section 2, Reference 3.2 and Reference 3. (g . . plant specific details including withdrawal schedule ret'erence]. The reports describing the post-irradiation evaluation of the surveillance capsules are contained in Reference 3. [s . . post-irradiation evaluation reference) 2.3 LTOP System Limits The LTOP require: tents have been developed by making a comparison between the peak transient pressures and the appropriate Appendix G pressure-temperature limit curves. The acceptability criterion regarding each particular transient is that the peak transient pressure does not exceed 110% of the applicable Appendix G pressure limit. The requirements for LTOP have been established based on NRC-accepted methodologies and are described in Section 3, Reference 3.2 and specAfied in the Bases Section for Technical Specification [A.B.c1, Reference 3.3. The affected Technical Specification Limiting Conditions for Operation (LCO's)

~     which ensure adequate LTOP are:

(NOTE: The following LCOs are presented as non-specific representation of LCOs which are in place at some of the e ABB Combustion Engineering Nuclear Power A-6 CE NPSD-663 Rev. 03 s i . .

RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR (NAMEJ UNIT (KJ I participant 's opera ting plants. Not all plants currently have each of these particular LCOs, depending on the complexity of the plant 's current LTOP analysis. Depending upon the LTOP analysis at each unit (which is unique), some of these may not be applicable.] LCO {3.2.2.2] Boration Systems Flow Paths - Shutdown LCO (3.2.2.3J Reactivity Control Systems Charging Pump - Shutdown LCO (3.3.2.2J Engineered Safety Features Actuation System Instrumentation LCO [3.4.9.21 Pressure / Temperature Limits - Reactor Coolant System LCO {3.4.23J Reactor Coolant System Power Operated Relief Valve LCO [3.5.3J Emergency Core Cooling Systems, ECCS Subsystems - Tavg < (325*f] LCO [3.4.24] Reactor Coolant System Reactor Coolant Pump - Starting The LTOP specific requirements for each LCO are presented in the following subsections. 2.3.1 Boration Systems Flow Paths - Shutdown ({LCD 3.2.2.2J) 2.3.1.1 The flow path from the RWT to the RCS via a single HPSI pump shall only be established if:

a. The RCS pressure boundary does not exist, or

b. (NoJ charging pumps are operable and the RCS heatup and cooldown rates shall be limited to those in Figure (3.2-21 At RCS temperatures below (125'TJ, any (two] of the following valves in the operable HPSI header shall be verified closed and have their power removed by removing their motor circuit breakers from the power supply, or by other means to prevent the valves from opening automatically.

ABB Combustion Engineering Nuclear Power CE NPSD-683 Rev. 03 A.7 I

n RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR [KAMEJ UNIT [XJ { High Pressure Header Auxiliary Header I [BCV-3616] [BCV-3617] [BCV-3626] [BCV-3627) [BCV-3636) [BCV-3637) ' [BCV-3646] [BCV-3647] 2.3.2 Reactivity Control Systems, Charging Pumps - Shutdown (LCo [3. 2. 2. 3J ) I 2.3.2.1 The flow path from the RWT to the RCS via a single HPSI pump shall I be established only if: I

a. The RCS pressure boundary does not exist, or

b. [No] charging pumps are operable and the RCS heatup and cocidown rates shall be limited to those in Figure [3.2-2].

At RCS temperatures below [21S*FJ, any [two] of the following valves in the cperable HPSI header shall be verified closed and have their power removed by removing their motor circuit breakers from the I power supply, or by other means to prevent the valves from opening automatically. High Pressure Header Auxiliary Header [BCV-3616) [BCV-3617] [BCV-3626) [BCV-3627] [BCV-3636] [BCV-3637] [BCV-3646] [BCV-3647] I 2.3.3 Engineered Safety Features Actuation System Instrumentation (LCO [3.3.2.1J) 2.3.3.1 At [365*FJ and less, the required Operable HPSI pump shall be in a l pull-to-lock and will not start automatically ABB Combustion Engineering Nuclear Power A-8 CE NPSD-683 Rev. 03 i l

RCS PRESSURE AMD TEMPERATURE LIMITS REPORT FOR (NAMEJ UNIT [2] 2.3.4 Pressure / Temperature Limits - Reactor Coolant System (LCO (3.4.9.2J) 2.3.4.1 The RCS (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures (3.4.9-2J and f3.4.9-2] dtring heatup, cooldown, criticality, and inservice leak and hydrostatic testing 2.3.5 Reactor Coolant System Power Operated Relief Valves ((ICO 3.4.237) I 2.3.5.1 The setpoints for the power operated relief valves shall be as I i follows: l

a. A setpoint of less than or equal to (350 psiaJ shall be selected: E

1. During cooldown when the temperature of any RCS cold leg is less than or equal to (215'FJ and l

2. During heatup and isothermal conditions when the 1

j temperature of any RCS cold leg is less than or equal to (193*FJ.

b. A setpoint of less than or equal to (530 psiaJ shall be selected: 3

1. During cooldown when the temperature of any RCS cold leg is greater than [225*FJ and less than or equal to the LTOP Enable Temperature for cooldown.

2. During heatup and isothermal conditions when the i

temperature of any RCS cold leg is greater than or equal I to (193*FJ and less than or equal to the LTOP Enable ' Temperature for heatup. 2.3.5.2 The LTOP Enable Temperatures are defined as follows:

a. The LTOP Enable Temperature for heatup is (304 *FJ .

I ABB Combustion Engineering Nuclear Power A.9 CE NPSD-683 Rev. 03 I

1 RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR (NAMEJ UNIT (2)

b. The LTOP Enable Temperature for cooldown is (282*FJ .

2.3.6 Emergency Core Cooling Systems, ECCS - Tavq < [325 *FJ ([LCO 3.5.3J) 2.3.6.1 Prior to decreasing the reactor coolant system temperature below (270*FJ, a maximum of only one high pressure safety injection pump shall be OPERABLE with its associated header stop valve open. I 2.3.6.2 Prior to decreasing the reactor coolant system temperature below (236*FJ, all high pressure safety injection pumps shall be disabled I and their associated header stop valves closed except as allowed by Specifications [3.1.2.1 and 3.1.2.31 2.3.7 Reactor Coolant System, Reactor Coolant Pump - Starting - ( {LCO J. 4. 24J) I 2.3.7.1 If the steam generator temperature exceeds the primary temperature by more than {30*FJ, no idle reactor coolant pump shall be started. 2.4 Beltline Material Adjusted Reference Temperature (ART) The calculation of the adjusted reference temperature (ART) for the beltline region has been performed using the NRC-accepted methodologies as described in Section 4, Reference 3.2. Application of Surveillance Data [was/was not] used to refine the chemistry factor and the margin term (see Section 2.7) . I The limiting ART values in the beltline region for the (MAMEJ Unit (21 corresponding to (EJ Ef fective Full Powa'- v- .rs (EFPY) for the 1/4t and 3/4t locations are: I Location ART Material 1/4% [xxx *FJ [... r.4miting Plate or Wald Material Identification . . . J 3/4t [xxx *FJ [... Lindting Plate or Weld Haterial Identification . . . J l ABB combustion Engineering Nuclear Power CE NPSD-683 Rev. 03 A-10 M

                                                                                                 -____a

RCS PRESSURE AND TEMPEFATURE LIMXTS REPORT FOR INAMEJ UNIT (IJ The RTns value for (NAMEJ Unit (21 which is calculated in accordance with 10 CFR 50.61 is [xxx 'r] uhich corresponds to [ Limit. tug Plate or Held Identifier]. Application of Surveillance Data fras/was notJ used to refine the chemistry W factor and the margin term (see Section 2.7) . 2.5 Pressure-Temperature Limits using limiting ART in the P-T I Curve calculation The limits for Ico 3.4.P.2 are presented in the subsection that follows. The analytical methods used to develop the RCS pressure-temperature limits are based , on NRC-accepted methodologies and discussed in Section 5 of Reference 3.2. The methodology is also documented in the Bases for Technical Specification (A.B.CJ. The RCS PRESSURE-TEMPERATURE LIMITS REPORT will be updated prior to exceeding the RTNDT utilized to develop the current heatup and cooldown curves. The RCS g PRESSURE-TEMPERATURE LIMITS REPORT, including any revisions or supplements thereto, shall be provided, upon issuance of new heatup and cooldown curves to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. 2.5.1 RCS Pressure and Temperature (P/T) Limits ((Ico 3. 4. 9. 2 J) 2.5.1.1 The RCS temperature rate-of-change limits are: I

a. A' maximum heatup of f50] *F in any 1-hour period.

b. A maximum cooldown rate consistent with Figure (2.2-3J.

c. A maximum temperature change of < 5*F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.5.1.2 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak g testing, and criticality are specified by Figures (2.2-2J, [2.2-2] and 5 [2. 2-3J . ABB Combustion Engineering Nuclear Power A-11 CE NPSD-683 Rev. 03 . 1

RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR (NAMEJ UNIT (XJ 2.6 Minimum Temperature Requirements in the P-T curves The minimum temperature requirements specified in Appendix G to 10 CFR 50 are applied to the P/T curves using the NRC-accepted methodologies as described in Section 6 of Reference 3.2. The minimwn temperature values applied to the P/T curves for (NAMEJ Unit {XJ corresponding to lEJ Effective Full Power Years (EFPY) are: Location Min Temperature BoltUp (xxx TJ Hydrotest {xxx TJ I ... ... I 2.7 Application of Surveillance Data to ART calculations I Post-irradiation surveillance capsule test results for (NAMEJ Unit (XJ are given in [ Reference 3.s]. The test results [do/do not] meet the credibility criteria of Regulatory Guide 1.99 Revision 2. [The criteria were met as follows: a) the surve111= nee program plate or veld dqplicates the contro11 Lag reactor u ,h1 hel-uine materisi in tezas of ART; I b) Charpy data scatter does not cause ambiguity in the det=m4 nation of the 30 ft-lb shift; l c) the measured shifts are consistent with the predicted shifts; d) the capsule irradiation tenperature is ccaparable to that of the vessel: and i e) correlation monitor data [are/are notJ available and are consistent with the known data for that material. l The data supporting the credibility analysis are presented in (reference].J l ABB Combustion Engineering Nuclear Power A-12 CE NPSD-683 Rev. 03

ROS PRESSURE AND TEMPEPATURE LIMITS REPORT FOR [NAME] UNIT (X) [In the case where sister vessel surveillan:e data are available for use, the preceding should be supplemented as indicated under Section 7 of Reference 3.2. The supp1="wntal info 1mation should address differences between the two sister plants in texms of irradiation environment .and establish the applicability of the data.] The credible surveillance data {were/were not] used to refine the chemistry factor and the nargin term. [The process for apply 1ng the credible surveillance data is described smder section 7 of Reference 3.2 in the Methodology and follows that prescribed in Position 2.1 of . Regulatory Guide 1.99, Revision 2. The data used and the calculations perfozmed are given below: Report Capsule ID Fluence _Shi Q Fluence Fnctor, f (f)E (f x shift) I Refined Chemistry Factor, CT(R)= E(f x thift) E(f)# 03 = (17 or 28) *F Refined as = (17 or 28) *F /2 0 2 Refined ART = Znitial Rtadt + CF(R) xf+ 2( )+ ] I l I ( I ABB Combustion Engineering Nuclear Power A-13 \ CE NPSD-683 Rev. 03 I 1

RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR [MAME] UNIT [X] 3.0-. REFERENCES 3.1 NRC GL 96-03, " Relocation of Pressure-Temperature Limit Curves and Low Temperature Overpressure Protection System Limits", January 31, 1996. 3.2 CE NPSD-683, Rev 03, " Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP requirements from the Technical Specifications,", May 1999. 3.3 Tech Spec A.B.C for [Name] Dhit [X] ... [3. q IT act La Tech Spec . . . Reference for Plant Specific Suzv=41'=~e Capsule Withdrawal Sek=A"Te ] [3.s Reference for post-irradiation evaluation of survei11 mar ~e capsules ] [3.s Reference for Fluence value modification ] f ABB Combustion Engineering Nuclear Power A-14 CE NPSD-683 Rev. 03 .

                                                                                                                                                                                                                                                                   ]

] RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR (NAMEJ UNIT (XJ FIGURE 4-1 [NAME] UNIT [A] P/T LIMITS,[ ] EFPY HEATUP AND CORE CRITICAL 2500 _ . . . . . _ . _ msgRyncs 2:  : ~1 - - l ~~~~*0.C HYDROSTATIC TEST I~ _I, i l eerup -

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0 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE, T. *F [NAME} UNIT [A] AMENDMENT NO. [Y] ABB Co:::bustion Engineering Nuclear Power A-15 L CE NPSD-683 Rev. 03 e _ , . . _ . . _ = - . . . . . . - -

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RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR [NAMEJ UNIT [XJ I FIGURE 4-2 [NAME] UNIT [A] P/T UMITS,[ ] EFPY I COOLDOWN AND INSERVICE TEST 2500 .._ ,.__, . . . _ i NSERVCE -'~-1 ' HYDROSTATC

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g ggp p MMMUM PRESSURE FOR SDCOPERATION n 0 100 200 300 400 500 600 (NDICATED REACTOR COOLANT TEMPERATURE, T., 'F I [NAME] UNIT [A] AMENDMENT NO. [Y) I ABB Con:bustion Engineering Nuclear Power A-16 I CE NPSD-683 Rev. 03

[. RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR fMANEJ LTIT (ZJ FIGURE 4-3 [NAME] UNIT [A] P/T LIMITS, [ ]EFPY MAXIMUM ALLOWABLE COOLDOWN RATES f i '

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40 100 120 140 160 180 200 Tc-INDICATED REACTOR COOLANT TEMPERATURE, DEG. F NOTE: A MAXIMUM COOLDOWN RATE OF 100 DEG, F/HR IS ALLOWED AT ANY TEMPERATURE ABOVE 195 DEG. F [NAME] UNIT [A] AMENDMENT NO. [Y] \ ABB Combustion Engineering Nuclear Power A-17 CE NPSD-683 Rev. 03

RCS PRESSURE AND TEM?ERATURE LIMITS REPORT TOR [NAME] UNIT (ZJ FIGURE 4-4 MAXIMUM ALLOWABLE HEATUP AND COOLDOWN RATES, g SINGLE HPSI PUMP IN OPERATION 3 100 - 80 - g 80 -

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Tc-INDICATED REACTOR COOLANT TEMPERATURE. DEG. F Il I I, 1 l I I (NAMEJ UNIT [A] AMENDMENT NO. [Y] ABB Combustion Engineering Nuclear Power A-18 CE NPSD-683 Rev. 03

APPENDIX B EXAMPLE OF MODIFIED TECHNICAL SPECIFICATIONS Appendix B is not be modified from Rev 00, 01, or 02 of this report. ABB Combustion Engineering Nuclear Power B-1 CE NPSD-683 Rev. 03 l

I_____ INDEX DE.!NITIONS

• 1 li SECTION

_?i3E

1. 23 ' F rec es s Cen t ro l P = gram ( FC? ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.24 ? u rg e - P u rg i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

( 1-3 1.25 ;ta .d Th-r=al 8 ewer . . . Abh ' i

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[ 1.@w9sh iel d su i l di ng Integri ty. . . . . . . . . . . . . . . . . . . . .1.-5. . . . . . . . . . . . . 1 g shu tdewn .Ma rgi n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. -.6. . . . . . . . . . ( i.q.3ouesouneary................................................ i-6 t i .g*2 sourceChece................................................ L 1-6 1.h'-3 Staggered Test 3 asis..................'....................... i-7 [ 1. Q Ther=al .:cwer............................................... N 1-7 1.@unieentifiedLeekage.............'...........................1-7 [ 1.h .5 Un re s tri cted Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . N I., 1-7 1 1. Unr:dded Integrated Radial Feaking Factor - Fr - I-7

1. Unrodded Planar Radial Peaking Facter - F n 27................. 1-7

[ [ I' L [ W.:.) UNIT [A] !a

                                                                                                                           ' AMENDVE.NT NO. [Y)

ABB Combustion Engineering Nuclear Power B-2 CE NPSD-683 Rev. 03

__ ___ -- ~ g i CEFINITIONS f g IDENTIFIED LEAXAGE 1.15 IDENTIFIED LEAXAGE shall be: ( a. Leskage (except CONTROLLED LEAKAGE) into closed systems, such as pumo seal or ' valve packing leaks that are captured, and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAXAGE, or

c. Reactor Coolant System leakage through a steam generator to the

{ secondary system. j LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE

           .J.16 condi iThe    LOW w an TEMPERATURE I                 RCS OVERPRESSURE PROTECTION RANGE is that.o erating t = cold le temperature is < 304*F during heatup er f                                                                                                                        I 7 <   2 1*r during in.-gri    y.       cooldown and (2 the Reactor. Coo ant Syste. la~s.~ pressure boor}dary g                       ine     , ea .or Coolant System does not have .nressure boundary inte<fity
  @b/A A when the .9eactor Coolant System is open to containment and the minimum area 6f the Reactor Co'olant System coenino is gr' eat r than 7                  e   . Ench ~
                  /c.s.s -s%n h*Duft   - 1~rmp're %e res Aveacce'obe s

A 'ta!.ke 5/VasAed in Aa ACS Lfo/ I*na 4 % f.ssnN.s Arper,S, HEM 3ER ) ur i: Uout; " -- ^~ - -- 1.17 MEM3ER(S) 0F THE PUSLIC shall include all persons who are not occupation-ally associated with the plant. This category does not include employees of f the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recrea-tJonal, occupational or other ptrrposes not associated with the plant. OFFSITE DOSE CALCULATION MANUAL {0DCM) 1.18 The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and parameters used in the calculations of offsite doses due to radioactive l gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and shall include the Radiological Environ-mental Sample point locations. (NAME] UNIT [A] l-4 AMENDMENT No. [Y1 ABB Combustion Engineering Nuclear Power B~3 J CE NPSD-683 Rev. 03

DEFINITIONS I RATED THEP".AL POWER 1.25 RATED THER"At POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 ."W:. REACTOR TRIP SYSTEM RESPONSE TIME . 7 1.36 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from whe the monitored parameter exceeds its trip setpoint at the channel sensor untti electrical power is interrupted to the CEA drive mechanism. REPORTABLE EVENT 78

1. A REPORTABLE EVENT shall be any of those conditions .specified in Section a 50 to 10 CFR Part 50.

l SHIELD BUILDING INTEGRITY

1. h HIELO BUILDING INTEGRITY shall exist when:  !

a. Each door is closed except when the access opening is being used for nor:ral transit entry and exit;

b. The shield building ventilation ' system is in cocpliance with Specification 3.6.6.1, and

c. The sealing me'chanism associated with each penetration (e.g., g .'

welds, bellows or 0-rings) is OPERABLE. SHUTDOWN MRGIN lh5 DOWN MRGIN shat 1.6e the instantaneous'asount of reactivity by which the reactor is suberitical or would be subtritical from its present condition assuming all full-len.gth control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn. SITE BOUNDARY

1. The SITE BOUNDARY shall be that line beyond which the land 'is neither owned. leased, nor otherwise controlled by the licensee.

SOURCE CHECX 2 1.h<ASOURCECHECKshallbethequalitativeassessmentofchannelresponse when the channel sensor is exposed to a radioactive source. [tWiE) UNIT [Aj AMENDMENT No. [ ABB Combustion Engineering Nuclear Power B-4 g CE NPSD-683 Rev. 03 l

INSERT A RCS PRESSURE-TEMPERATURE LIMITS REPORT 1.26 The RCS PRESSURI-TEMPERATURE LIMIT REPORT (PTLR) is a fluence dependent report providing Limiting Condition of Operatiens for heatup, cooldown, inservice hydrostatic and leak testing, and core criticality limits in the form of Pressure-Temperature (P-T). limits to ensure prevention of brittle fracture. In addition, this report establishes Limiting conditions of Operation which provide Low Temperature Overpressure Protection (LTOP) to assure the P-T limits are not exceeded during the most limiting LTOP event. The P-T limits and LTOP criteria in the PTLR are applicable through the time period specified. NRC approved methodologies are used as the basis for the LCO's provided in the PTLR. L ABB Combustion Engineering Nuclear Power B-5 CE NPSD-683 Rev. 03

I OEFINITICNS ST.:.GGERED TEST 3 ASIS 3 i t.h A STAGGERED TEST BASIS snali of: c:nsis:

            ,            a. A test schedule for n sys: ems, subsystems, : rains cr-other i

desicnated components obtained by dividing tne specified

                                   ~

test interval into n equal sucintervals, and b. The testing of one system, subsystem, train or other desig ated co=ponent at the beginning of each subinterval.

            'THE?JML POWER 74 1.3)3 the    reactorTHERi4AL coolant. POWER shall be the total reactor core heat transfer UNIDENTIFIED LEAXAGE l           75 jl.$ UNIDENTIFIED LEAXAGE shall be all leakage which is not IDENTIFIED ILEAXAGE or COHTROLLED LEAXAGE.                                                                                   l UNRESTRICTED AREA

1. .n An UNRESTRICTED AREA shall be any area at or beyond the SITE EOU.*iDARY access to which is not controlled by the. licensee for purposes of protaction l of individuals from exposure to radiation and radioactive materials, er any area within the SITE SOUNDAPJ used for residential quarters or for ind:strial, cc=ercial, institutional, and/or recreational purposes.

UliRODDED INTEGRATED RADIAL PEAKING FACTOR r

                                                                                            -F 1.

T e UNRODDED INTEGRATED RADIAL PEAKING FACTOR is the ratio of the peak pin power to the average Din power in an unrodded core, excluding tilt. l UNRODDED PLN{AR RADIAL PEAXING FACTOR - F,y 1. T e UNRODDED PLANAR RADIAL PEAKING FACTOR is thei maximum peak to average power density of the individual fuel rods in any of :. e ' unrodded horizontal planes, excluding tilt. W) MT %) I-7 AMENDMENT NO. [Y) ABB Combustion Engineering Nuclear Power CE NPSD-683 Rev. 03 B-6

. . _ .. .piut 3:372-5 3/a.i.2 SORATION 5YSTEMS I l FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3 .1. 2 .1 As a minimum, one o f the following boren injection flow paths shall ce 0?ERABLE. and capable of being powered fecm an 0?ERACLE emergency power sc a.

A flow path from the boric acid makeup tank via either a boric acid pump or a gravity feed connection and any chargina pump to the Reactor Coolant System if only the boric acid makeup tank in Specification 3.1.2.7a is OPERABLE, or b. The flow path from the refueling water tank via either a charging pump or a high pressure safety in.jection pump 9~dekic to the Reactor Coolant System if only the refueline water l E tank in Specification 3.1.2.7b is OPERA 3LE. Q APPLICABILITY: N00ES 5 and 6. ACTION : With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status. SURVEILLANCE REOUIREMENTS 4.1.2.1 OPERABLE:At least one of the above required flow paths shall be demonstrated I a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not I locked, sealed, or otherwise secured in position, is in its g correct position. g

          '-*T.he         flow path be established        if: (afrom the RWT to the RCS via a single HPSI pump (b) no charging pumps a)re o~perable.the RCS pressure boundary does not e i                                                 In this case all charging pumps shall      be  disabled,
         ) accordance with Fig. 3.1-lb.and  heatup and  cooldown    rates shall be limited in f

At RCS temperatures below 115'F, any two of the following vanes in the s-operable HPSI header shall be verified closed and have their power remo<e::  ! E g Hich Pressure Header Auxiliary Header ' HCV-3616 HCV-3626 HCV-3617 .' fl6mOVI TO HCV-3627 ' HCV-3636 /YLA HCV-3646 HCV-3637 l HCV-3647 E [ E (NAMEl UNIT [A] 3/4 1-8 AMENDME:NT NO. IY) ABB Combustion E.ngineering Buclear Power B-7 CE NPSD-683 Rev. 03

I INSERT B The flow path from the RNT to the RCS via a single HPSI, pump shall only be established if the requirements in the PTLR are met. 9 ABB Combustion Engineering Nuclear hwer g.g CE MPSD-683 Jtev. 03

T a > too - E so - 5 - w 6c - Heatup 5 4o . 20 - O' '

                                                   '                                    C::' ; n          '

so 1oo 120 140 1sc iso 200 22o Tc . IND4CATED REACTOR COOLANT TElePERATURE.Y l s i F4GURE 3.1-10 edAXedlJM ALLOWASLE HEAT!JP AND COOLDowM RATES, SaNGLE Hest PuesP IN OPERATION

                                                       .A                     w       h

[tn50V'f hO WLk (NAME) UNIT [A] 3/4 1- 9a ~ m m NO. M ABB Combustion Engineering Nuclear Power CE NPSD-683 Rev. 03 B-9

5

                  ;' :.EACTIt'ITY CONTROL SYSTEMS lCHARGINGPUMD - SHUTDOWN l.

F LIMITING CONDITION FOR OPERATION I b ele)c. 3 .1. 2_. 3 5 At least one charging pump or one high pressure safety injection cation 3.1.2.1 shall be OPERABLE and capable of being I

                ! OPERABLE emergency bus.

APPLICABILITY: NODES 5 and. 6. ACTIO(1: h tiith no charging pump or high pressure safety injection pum

               ' all operations involving CORE ALTERATIONS or positive reactivity changes un at least one of the required pumps is restored to OPERASLE status.

SURVEILLANCE REOUIREMENTS g 4 .1 . 2 . 3 E At least one of the above required pumps shall be demonstrated OPEP.AELE by verifying the charging pur!p develops a flow rate of greater than or equal to 40 gpm or the high pressure sa fety injection pump develops a Specftotal fication head of4.0.5. grea,ter than or equal to 2571 f t. when tested pursuant to

               *The flow path from the RtiT to thh RCS via a single HPSI pump shall be established only if: (a) the RCS pressure boundary does not exist, or                 E (b) no charging puinps are ophrable. In this case, all charging pumps                 3 shall be, disabled and heatup and cooldown rates shall be limited in                   g accordance with Fig. 3.1-lb.

gfmovi g G Pr24 At RCS tempeiatures below ll5*F, any two of the following valves in the operable HPSI header shall be verified closed and have their power removed: Hich Pressure Header Auxiliary Header HCV-3616  ! HCV-3617 HCV-3626 HCV-3627 E HCV-3636 HCV-3637 5 HCV-3646 HCV-3647 [h*AME) UNIT {A) 3M 1-12 AMENDMENT NO. [Y) ' , I ABB Combustion Engineering Nuclear Power B-10 g CE NPSD-683 Rev. 03 l

INSERT B The flow path frcm the RWT to the RCS via a single HPSI pump shall only be established if the requirements in the PTLR are met. ABB Combustion Engineering Nuclear Power B-11 l CE NPSD-683 Rev. 03 l

l REACTOR CCCLANT SYSTE.M 3/a.4.9 PRESSURE /TE.MPERATURE L TM?TS

EACTOR COOLANT SYSTEM

                            ~

LIMITING CONDITION FOR OPERATTON 3.4.C.1 The Reactor Coolant System (except the pressurizerl temperature and

            <Mure shall be limited in accordance with the limit lines shown on Figures
            , 3.4-2a, 3.4-2b and 3.4-3 during heatup, cooldown, criticality, and inservice .

and hydrostatic testing. APPLICABILITY: Atalltimesh C8 # l ACTION: Qgge fr7SC U I With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an analysis to detemine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; detemine that the I Reactor Coolant System remains acceptable for continued operations or be in less _to at leastthe 2 0*FHOTwi' inSTANDBY 2.4-3 wi-thin

                                                 .the' following 30'  hours inthe  next 6 accordance    hours with'                               and reduce LC0 3. $. 9. I I                                                                ~

7When the flow path from th_) RyT to the RCS via a ,si.ngle HPSI pump is

             , established per 3.1.2.3, the heatup and cooldown rates shall be established
              '==-1==              m ng. 2

n. as indic4kd id He 97%f.

(During hydrostatic testing operations above system design pressure, e maximum temperature change in any one hour period shall be limited to 5'~. I I l I [NAME) UNIT [A] 3/4 4-21 AMENDMENT NO. [Y1 i ABB Combustica Engineering Nuclear Power B-12 CE NP5D-683 Rev. 03

INSERT C The cc:rJoination of RCS pressure, RCS temperature and RCS heatup and cooldowm rates sh.11 be maintained within the limits specified in the RCS PRESSURE-TEMPEPATURE LIMITS REPORT. ABB Combustion Engineering Nuclear Power B-13 CE NPSD-683 Rev. 03

              REACTOR COOLANT SYSTEM                           ,

SURVEILLANCE REQUIREMENTS 4.4.9.1

a. The Reactor Coolant System temperature and pressure shall be detemined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

b. The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minutes prior to achieving reactor criticality.

c. The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4- . T' e -=s i ts of es e

                           . -2b and   minations shall  be used to update Figures    3.4-2a, '

m :3.4-3. . 4ht $C.S Atc.ssdrd ~iist1/r/CkW l.tkrs'b AC,00rA I I I I h l l ' h (NAME] UNIT (A) 3/4 4-22 AtENDMENT NO. (Y1 ABB combustion Engineering Nuclear Power B-14 CE NPSD-683 Rev. 03

( E/7? C V E T O/ A V

                                                   ?!GUTC 3.4-2a l

[(Ch5J L' NIT [ A } ?/~ L*.}!ITS, { j T.T?Y i :.ATU? AND CO?2 C?lTICM., i 2500,. tSOTM.RWAL TO SCT/HR T_ 2000!

a. A W

m 5 e

.W I     1500 E
               -      LOWEST :

5 sEnycg E - TEnaPERATURE - __ g- *== N

                                                                              ~' CORE CRfTCAL -

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                                              =!

f- ALLOWASLE MATUP RATES 300 - R ATE.*F/tst TEMP. LandKT, *F ESOT/HR s ~

                                                                   .==               $0          AT ALL
                             - m.4GL 90LTUP TM 309                                                                   E g

0 u -- -  : 23 -~ - O 100 200 soo 4og Soo a TC INDICATED RCS TEMPERATURE. "f I (tetEl UNIT (Al 3/4 4-2.3a AMENDMENT NO. [Y) ABB Corr.bustion Engineering Nuclear Power B-25 g CE NPSD-683 Rev. 03 ll _J

TIG2,Z 3.4-2b [ M E} UNIT [Al :/T L MITo, i j ,,,.rpy l CCONN AN"O INSERVICI T.5T 2500 _ _

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0 100 200 300 400- 500 T GeoCATED RCS TEAdPERATURE,*F c A [A] gggygg3 g 4 2 AME:Noxert NO. [Y) f l ABB Combustion Engineering Nuclear Power B-16 CE NPSD-683 Rev. 03

I T:'G.T?2 3,.4-3 {NA:-2) UNIT [A], i } II'?Y- I MAXDE'M A*.l.CG3!.I CI'CvS~ ?.A~ES f x - I 1& . .

                                   , RATE. *F/HR                                                 -

TEM 9. LIMIT. *F 20 <1:5 '

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30 125 146 g B0 40 14166 3I

                             . . .             50                           16136 75                          185-196 3                           100 2         . . _ . . .                                         > 196 w
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g 3 C' ' 80 100 120 140

            \                                                                          160               180              200 Tc . IMOCATED REACTOR COOLANT TEMPERATURE. *F I
     /(/r70Y[ TO                            NOTE: A MAXlMUM COOLDOWN RATE OF 100*F/HR 13 ALLOWED AT ANY
       /7'2.8                                         TEMPERATURE ASOVE 195'F l        E g

I

 *t4AF.r".] UNIT ( A)

AMENDMENT NO. {Y3 I ABB Combustion Engineering Nuclear Power B-17 CE NPSD-683 Rev. 03 g l

            "E*CTOR C00LsNT SYSTE.M PC'4ER OPERATED RELIEF VALVES                                   *
                                                                         //) 4CddiEQt1R tus'Nt 'Hwse LIMITING CONDITION CR OPERATION                               @d/O8 /h Me /M

3. .13 Two power operated relief velves (PORVs) s' hall be OPERAS' e ith e.eir setpoints selected to the low temoera tur'e mode of operation

a. .A setpoint of less tnan or equal to 350 psia shall be selec:ed:

1 During cooldown when the temperature of any RCS cold leg is less than or eque.1 to 215'F and

2. During heatup and isother=al conditions when the tempera ture of any RCS cold ' leg is less than or equal to 193*F.

b. A setpoint of less than or equal to 530 psia shall be selected:

1. During cooldown when the temperature of any RCS cold leg I 2.

is greater than 215'F and less than or equal to 281*F. During heatup and isothermal conditions when the temperature of any PCS cold leg is greater than or equal to 193*F and I less than or equal to 304*F. w - Nmov[To APPLICABILITY: MODES 4' and 5*. fo 7cg ACTION:

a. With less than two PORVs OPERABLE and while at Hot Shutdown during a planned cooldown, both PORVs will be returned to OPERABLE status prior to entering the applicable MODE unless:

1 The repairs cannot be accomplished within 24 hours or the repairs cannot be performed under hot conditions, or I 2. Another action statement requires cooldown, or

3. Plant and personnel safety requires cooldown to Cold Shutdown with extreme caution.

b. With less than two PORVs OPERABLE while in COLD SHUT 00WN, both PORVs will be returned to OPERABLE status prior to startup.

c. The provisions of Specification 3.0.4 are not applicable.

       " SURVEILLANCE REOUTREMENTS 4.4.13 The FORVs shall be verified OPERABLE by:

a. Verifying the isolation valves are open when the PORVs are reset to the low temperature mode of operation.

b. L Performance of a CHANNEL FUNCTIONAL TEST of the Reactor Coolant System overpressurization protection system circuitry up to and including the relief valve solenoids once per refueling outage.

c. Performance of a CHANNEL CALIBRATION of the pressurizer pressure sensing channels once per 18 months.

                                                                                                       ,,     4 p,, (yop -

tr,ohte *Gp.m wee nacMc.d h r'

                                                                              $($ ~5Yf$$drf. Vier /ttdthff tReactor Coolant System cold leg temperature bel w 30 *F                LEJJ &/er/.

PORVs are not required below 140*F when RCS does not have pressure boundary i ntegrity. { [NAME] UNIT (A) 3/4 4-59 AMENDMENT NO. [Y) ABB Combustion Engineering Nuclear Power B-18 CE NPSD-683 Fev. 03

                                       ~

1 l REACTOR CCOLANT ? UMP - STARTING LIMIT *NG CONDITION FOR CPERATION j 3 .ic If t s teem generator tempera ture exceeds the ::rir.ary temperature by more than 30*F the first idle reactor coolent pumo shell not be started. i

                     +he mgnibe'e pe/Shd th He tc.s Grewrc- nmge:Nrc Le:S Enal(f7 APPLICABILITY: MODES a and 5.

ACTION: If a reactor cooient pump is started when the steam generator temperature exceeds primary temperature by more the 30 ~ evaluate the subsequent I transient to determine compliance with Specification 3.4.9.1. c ma fepor/hnllsk rt t). . spec l4cd b Mc 4cs Ac.surc-77mperad/m LAnW SURVEILLANCE REOUIREMENTS l l 4.4.14 Prior to starting a reactor coolant pump, verify tha *h steam generator temperature does not exceed primary temperature by more tha 30*F. A C $ /> /U (SfFC hf th hC $$$ fussare-6pera/""4&Ms deped (mt.). I I I I of cfa:rl 40 At L1'90ems le 7Empcrabre I

                                                  .SptctSed ih Me drC Pet.1.r5Lc -           g.
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Timpeexr'ure Lihu%s 4por/ (PrsE). E i fReactor Coolant System Cold Leg Temperature is less thanh OWE] UNIT (A] 3/4 4-60 yq:gp3gg7 30, ty) I ABB Combustion Engineering Nuclear Power B-19 , CE NPSD-683 Rev. 03

l, EMER3ENCY CORE COOLING' SYSTEMS ECCS SUESYSTEMS --Te y, < 325'? LIMITING CONDITION FOR OPERATION 3.5.3 As a_ minimum, one; ECCS subsystem comprised o f the following shall be OPERABLE: '

a. In MODES 3* and 4f, one ECCS subsystem com::osed of one OPERABLE high pressure safety. injection pump and one OPERABLE flow path capable

                          . of taking suction from the refueling water storage tank on a safety injection actuation signal and automatically transferring suction
             .'Txugr@ to the containment sump on a sump recirculation actuation signal,

b. Prior to decreasing the reactor coolant system temperature below I #0 270*F a maximum of only one high pressure safety injection pump N shall be OPERABLE with its associated header stop valve open.

c. Prior to decreasing the reactor coolant. system temperature below ~l 236*F.all high pressure safety injection pumps shall' be disabled  !

and their associated header stop valves closed except as allowed by Specifications 3.1.2.1 and 3.1.2.3. APPLICABILITY: MODES 3* and 4 . l ACTION:

a. With no ECCS' subsystems OPERA 8LE in MODES 3* and #

4 , imediately restore one ECCS subsystem to OPERA 8LE status or be in COLD SHUTDOWN i

      //#49cr w ,rg within 20 hours.

sum '

             &@g b. With RC3 temperature below 270*F and' with more than the allowed high 7       pressure safety infection pump OPERABLE or injection valves and header isolation valves open, immediately disable the high pressure safety injection pump (s) or close the header isolation valves.

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actBatioii cycles to date.

SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall- be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2. . l 4.5.3.2 The high pressure safety injection pumps shall be verified inoperable and th sso i ted header stop valves closed prior to decreasing below the M84' - bove tpeci ie Reactor Coolant System temperature nd once per month when the React 5r Io6 Tant System is at refueling temperatures. Ab

              *With pressurizer pressure < 1750 psia, J/kc     Q th kN iREACTOR COOLANT SYSTEM cold leg temperature above 250*F.

INAME) UNIT [A) 3/4 54 AMENDMENT NO. (Y) ABB Combustion Engineering Nuclear Power B-20 l CE NPSD-683 Rev. 03 1..

INSERT D

b. Additional operability requirements for high pressure safety injection pumps are provided in the ROS Pressure-Temperature Limits Report and shall be adhered to.

T h Y ABB Combustion Engineering Nuclear Power B-21

 ,      CE NPSD-683 Rev. 03

I INSERT E

b. If the requirements of the RCS Pressure-Temperature Limits report have not been satisfied, initiate action to provide immediate compliance with the requirements.

I l l 1 i l I I l l I I 1 ABB Combustion Engineering Nuclear Power B-22 CE NP5D-683 Rev. 03 1}}