ML20199L918

Rev 0 to Final Rept CE NPSD-683, Development of RCS Pressure & Temp Limits Rept for Removal of P-T Limits & LTOP Requirements from Tech Specs

ML20199L918
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Site:	Fort Calhoun
Issue date:	12/31/1997
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML20199L879	List: LIC-98-0013, Forwards Application for Amend to License DPR-40,relocating pressure-temperature Curves,Predicted Radiation Induced NDTT Shift Curve & LTOP Limits,Per GL 96-03.Document Titled RCS Unit 1 RCS PT Limits Rept & CE NPSD-683,Rev 2 Encl ML20199L891 ML20199L895 ML20199L909 ML20199L918
References
CE-NPSD-683, CE-NPSD-683-R02, CE-NPSD-683-R2, NUDOCS 9802100067
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{{#Wiki_filter:________________ _ _ b t-dOMBUSTION ENOINEERING OWNERS GROUP CE NPSD-683 Rev.02 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 S42 prepared for the C-E OWNERS GROUP Decemoer 1997 j!gaigggg;oggggg,3

• Copyreht 1997 Combustion Engineering, Inc. All rights reserved

       , ABB Combustion Engineering Nuclear Operations                                           7%BRID J

                                                                                                                    ~

J I i I LEGAL NOTICE , This report was prepared as an accot.nt of work sponsored by the Combustion Engineering Owners Group and ABR Combustion Engineering. Neither Combustion Engineering, Inc nor any person acting on its behalf: gl A. makes any warranty or representation, expr<:st, or implie(f including the warranties of fitness for a particular purpose or merchantability, with l respect to the accuracy, completeness, or usefulness of the information L :ontained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately 2 owned rights; or B. assumes any liabilities with respect to the use of, or for -*amages , resulting from the use of, any information, apparatus, method or proces, disclosed in this report. 1 I E I: Combustion Engineering, Inc.

 .                                                                                                              g g.

i l l I THE DEVELOPMENT OF AN RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR THE LEMOVAL OF P-T LIMITS AND LTOP REQUIREMENTS FROM THE TECHNICAL SPECIFICATIONS I ' I PREPARED FOR THE C-E OWNERS GROUP I BY ABB COMBUSTION ENGINEERING NUCLEAR OPERATIONS COMBUSTION ENGINEERING, INC. I I ~ I I I z

TABLE OF CONTENTS Section Title Page ABSTRACT 3

1.0 INTRODUCTION

4

2.0 DESCRIPTION

OF ACTIVITIES 4 2.1 PTLR Development 5 3.0 GENERIC PTLR S 4.0 RESULTS 5

5.0 REFERENCES

7 APPENDICES A Combustion Engineering's Methodologies for the A-1 Development of Reactor Coolant System Pressure-Temperature Limits B Low Temperature Overpressure Protection Methodology B-1 and Relationship to Appendix G Pressure-Temperature Limits C Example of RCS Pressure and Temperature Limits C-1 Report D Example of Modified Technical Specifications D-1 I I I I

I 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. I.s 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 Limitino Conditions for Operation (LCO's), a new controlled document was developed v '. led an RCS Pressure and Temperature Limits Report (PTLR). This document is consistent with the requirements of Generic Letter 96-03 and contains the detailed information needed to support the pertinent LCO's which would remain in l the Tect.nical Specification. This generic PTLR contains current methodology l descriptions of RCS P-T limit development, LTOP criteria, ART calculation, RV Surveillance Program and Calculation of Neutron Fluence. An example of a PTLR I is prepared along with the proposed changts to the subject Technical Specification. I > The enclosed PTLR is generic in nature and can be easily tailored to be suitable to any C-E plant. I I

I

1.0 INTRODUCTION

In an effort to improve the maintenance of Technical Specifications, the Nuclear Regulatory Commission (NRC) has issued Generic Letter (GL) 96-03"'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 I (RCS) Pressure-Temperature (P-T) limits, Low Temperature Overpressure Protection (LTOP) setpoints, RV Surveillance Program and Fluence Calculation updates. Once incorporated 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 ef fort undertaken to ~ create a generic PTLR document based on guidance presented in NRC GL 96-I 03,

3.0 DESCRIPTION

OF ACTIVITIES The haC issued GL 96-03 to advise licensees that they may request a I license amendment to relocate cycle dependent information, such as, the pressure teu erature (P/T) limit curves and low temperature overpressure protection ( .. TOP) system limits from their plant Technical Specifications (TS) to a PTLR or similar controlled document. This task addresses the development of the required information to be included in the PTLR based I on the generic letter. The guidance is divided into seven provisions to be addressed in the P1LR. They are: 1 Neutron Fluence Values 2 Reactor Vessel Surveillance Program 3 LTOP System Limits 4 Beltline Material Adjusted Reference Temperature (ART) 5 Pressure-Temperature Limits using limiting ART in the P-T Curve calculation 6 Minimum Temperature Requirements in the P-T curves I

                                 ~1     Application of Surveillance Data to ART calculations Each provision requires a methodology description be provided along with g

specific data about the operating plant. These provisions are g specifically addressed in Appendices A and B of this document. The example PTLR shown in Appendix C is organized to address each provision. Since this ef fort builds upon previous work (report CE NPSD-683'"), 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 addition, methodologies were updated to reflect the rule and code changes 'since the original issue of the topical report. 2.1 PTLR Development The PTLR was developed on a generic basis such that it would apply to all C-E utilities. A review of typical LCO's for RCS P-T limits and LTOP requirements was performed and included in the generic PTLR. In order to support the PTLR, methodology descriptions were prepared I and were incorporated as Appendices A and B. The methodologies presented describe the development of RCS P-T limits, LTOP setpoints, RV Surveillance Programs and fluence values. 3.0 GENERIC PTLR To facilitate development of a plant specific PTLR, an example PTLR is presented in Appendix C of this report. This PTLR is applicable to all C-E utilities. 4.0 RESULTS I The results 9f this task provide a basis for the relocation of RCS Pressure-Temperature (P-T) limits, Low Temperature Overpressure Protection (LTOP) setpoints, RV Surveillance and Neutron Fluence reporting I

requirements from the Technical Specifications to another controlled document. I 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 Temperature Limits Report ( PTLR) , which contains the detailed information needed to comply with relocating the Limiting conditions for Operation from the Technical Specifications was developed. Methodology descriptions for developing RCS P-T limits, establishing LTOP setpoints, calculating ART, developing a RV Surveillance Program, and calculating Neutron Fluence to support the PTLR are provided in Appendices A and B. I An example PTLR and a sample Technical Specifications " mark-up" are provided in Appendices C & D, respectively. The example PTLR contains typical LCO's for RCS P-T limits and LTOP requirements for CE plants and can be tailored for plant specific submittals. The sample 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. I In conclusion, this report provides an acceptable and referenceable generic basis for the creation of plant specific PTLR reports. ( .

5.0 REFERENCES

1. Title 10 of the Code of Federal Regulations, Part 50, Appendix G, i

Fracture Toughness Requirements, 1995 Edition. l l g

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

3. NRC GL 96-03, Relocation of Pressure-Temperature Limit Curves and I

Low Temperature Overpressure Protection System Limits, January 31, 1996.

4. CE NPSD-683, Rev 00, "Duvelopment of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP requirements from the Technical Specifications, CEOG Task 661", March 1992. -

I I I I I I I I

I APPENDIX A Development of Reactor Coolant System Pressure-Temperature Limits Prepared for: Combustion Engineering Owners Group l l by: ABB Combustion Engineering Nuclear Operations 2000 Day Hill Road Windsor, CT 06095-0500 I I I

I TABLE ur CONTENTS .s Page

1.0 INTRODUCTION

A-6 1.1 SCOPE A-6 2.0 REACTOR COOLANT PRESSURE EOUNDARY OPERATIOhAL DESCRIPTION A-6 2.1 GENERAL A-6 2.2 NORMAL OPERATION A-7 I 2.2.1 2.2.2 REACTOR VESSEL BOLTUP HEATUP A-7 A-7 2.2.3 COOLDOWN A-8 2.3 INSERVICE HYDROSTATIC PRESSURE TEST AND LEAK TESTS A-9 2.4 REACTOR CORE OPERATION A-9 I 3.0 METHODOLOGY AND ANALYTICAL PROCEDURES A-9

3.1 BACKGROUND

A-9 3.2 NETHODOLOGIES A-10 I 3.2.1 NEUTRON FLUENCE CALCULATIONAL METHODS 3.2.1.1 Input Data A-10 A-11 3.2.1.1.1 Materials and Geometry A-11 I- 3.2.1.1.2 Cross-Sections A-12

                                                             -3.2.1.1.2.1 Multi-group Libraries.              A-12 3.2.1.1.2.2 Constructing a Multi-group Library A-13 3.2.1.2           Core Neutron Source                         A-13 3.2.1.3            Fluence Calculation                         A-15 3.2.1.3.1   Transport Calculation                  A-15 3.2.1.3.2    Synthesis of the 3-D Fluence           A-17 3.2.1.3.3    Cavity Fluence Calculations            A-18 I                                             3.2.1.4            Methodology Qualification and Uncertainty 'Jstimates                      A-19 3.2.1.4.1    Analytic Uncertainty Analysis          A-20 3.2.1.4.2   Comparison with Benchmark and Plant-Specific Heasurements            A-21 3.2.1.4.2.1 Operating Reactor Measurements      A-22 I

l A-2

I TABLE OF CONTENTS ( cont' d) Page 3.2.1.4.2.2 Pressure Vessel simulator Measurements A-22 3.2.1.4.2.3 Calculational Benchmarks A-22 3.2.1.4.3 overall Bias and Uncertainty A-22 3.2.2 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM A-23 3.2.3 METHOD FOR CALCULATING BELTLINE MATERIAL M ADJUSTED REFERENCE TEMPERATURE (ART) A-25 3.2.4 APPLICATION OF SURVEILLANCE DATA TO ART CALCULATIONS A-26 3.2.5 METHOD FOR DEVELOPING LTOP SYSTEM LIMITS A-28 3.2.6 APPLICATION OF THE LIMITING ART IN THE P-T CURVE CALCULATION A-28 3.2.6.1 General Method A-30 3.2.6.2 Flanges A-32 3.2.6.3 Nozzles A-32 5 3.2.6.4 Beltline A-33 3.2.7 METHOD FOR ADDRESSING 10 CFR 50 MINIMUM TEMPERATURE REQUIREMENTS IN THE P-T CURVES A-34 3.2.7.1 Inservice Hydrostatic Pressure Test and Core critical Limits A-34 3.2.7.2 Minimum Boltup and Lowest Service Temperature A-35 3.3 TECHNICAL DESCRIPTION FOR CALCULATING BELTLINE REGION PRESSURE-TEMPERATURE LIMITS A-36 M 3.3.1 GENERAL A-36 3.3.2 THERMAL ANALYSIS METHODOLOGY AND INFLUENCE COEFFICIENTS A-38 3.3.3 NORMAL OPERATION A-42 3.3.3.1 Cooldown Limit Analysis A-42 3.3.3.2 Heatup Limit Analysis A-43 4.0 TYPICAL PRESSURE-TEMPERATURE LIMITS A-45 4.1 BELTLINE LIMIT CURVES A-46 4.2 FLANGE LIMIT CURVES A-47 A-3 I Il

[ TABLE OF CONTENTS ( cont' d) 4.3 COMPOSITE LIMIT CURVES A-47 4.4 OPERATIONAL LIMIT CURVES A-48 5.0 stmeGRY AND CONCLUSIONS A-49

6.0 REFERENCES

A-50 I I I I I I I I I A-4 a

 .I m                            LIST or FIGURES

rigure Title Page 4.1-1 Appendix G Beltline P-T Limits, Heatup A-52 4.1-2 Appendix G Beltline P-T Limits, Heatup A-53 4.1-3 Appendix G Beltline P-T Limits, ccoldown A-54

4.1-4 Appendix G Beltline P-T Limits, Cooldown A-55 4.1-5 Appendix G Beltline P-T Limits, Hydrostatic A-56

'I 4.2-1 Appendix G Flange Limits, Heatup A-57 I 4.3-1 Composite Appendix G P-T Limits, Heatup A-50 I 4.3-2 Composite Appendix G P-T Limits, Cooldown A-59 4.3-3 Composite Appendix G P-T Limits, Hydrostatic A-60 4.4-1 Typical RCS P-T Limits for Technical Specifications, Heatup A-61 I 4.4-2 Typical ,RCS P-T Limits for Technical Specifications, Cooldown A-62 I A-5

I

1.0 INTRODUCTION

1.1 SCOPE This appendix describes the development of Reactor Coolant System (RCS) pressure-temperature (P-T) limits presented in the RCS Pressure-I Temperature Limits Report (PTLR). The methodology utilized to develop the RCS P-T limits is applicable to all CE Nuclear Steam Supply Systems (NSSS). This methodology addresses the analytical techniques and practices used to develop RCS P-T limits which are required to protect against nonductile (brittle) failure for the following loading conditions normal operations, inservice hydrostatic pressure tests and leak tests, and reactor core operation. Normal operation includes reactor vessel boltup, RCS heatup and RCS cooldown. l The specific analytical methods, which include Linear Elastic Fracture 3 Mechanics (LEFM) techniques, have been developed to meet the requirements of 10 CFR 50, Appendix G III as supplemented by ASME Boiler and Pressure Vessel Code (referred to hereafter as ASME Code) Section XI, Appendix G(6) . The definitions and terminology of 10 CFR 50 and ASME Code are used whenever appropriate. I 2.0 REACTOR COOLANT PRESSURE BOUNDARY OPERATIONAL DESCRIPTION I 2.1 GENERAT4 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. Determination of these restrictions require that specific loading conditions be evaluated and the resulting pressure-temperature limits not be exceeded. The specific loading conditions, for which P-T limits are required, are as follows: A-6

                                                                                                               ._. _J

l I

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

cooldown

2. Inservice hydrostatic pressure tests and leak tests

3. Reactor Core Operation A brief description of these conditions is provided to highlight the g

typical process followed to determine the physical loadings resulting from W the particular operation. 2.2 NORMAL OPERATION I 2.2.1 REACTOR VESSEL BOLTUP I Reactor vessel boltup loads are generated by stud tensioners when securing the closure head against the reactor vessel. Prior to tensioning of the g studs to the required preload, th< eactor coolant temperature and the W volumetric average temperature of the closure head region must be at or above the minimum boltup temperature. Once the studs have been tensioned, the RCS is capable of being pressurized and heated up. The heatup transient begins when a Reactor Coolant Pump (RCP) is started or when Residual Heat Removal (RHR) system flow is altered to allow elevation of the RCS temperature. 2.2.2 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 HOT 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 A-7 I

I

  =

requirements, and tha maximum pressure meeting the RV material fracture toughness requirements. Also, the heatup rate should not exceed the rates specifieu by the pressure-temperature limits. 3.2.3 COOLDOWN During cooldown the RCS is brought from a HOT SHUTDOWN condition to a COLD SHUTDOWN condition. Initially, coolant temperature reduction is achieved by removing heat through use uf 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

            $50'F to 300*F using this method. To complete the cooldown the RHR System is utilized.

I Typically, cooldown is initiated by securing the RCP(s) . Any remaining pumps 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 I 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 utilited to control the cooldown rate and the remaining RCP's can be stopped. It is advisable to initiate RHR flow prior to stopping all RCP's to provide sufficient mixing and minimize the thermal shock to Reactor Coolant Pressure Boundary (RCPB) components. I The pressure during cooldown is maintained between the maximum pressure 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.

I I A-8 I

2.3 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 in ASME Code Section XI, Article IWA-5000. For th'e 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 required pressure is determined by the fracture toughness requirements and guidarce provided in 10 CFR 50, Appendix G. 2.4 REACTOR CORE OPERATION 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 used to attain the required temperature. Also, this minimum temperature is much higher than the requirerrents imposed by 10 CFR 50 Appendix G which address only brittle fracture concerns. 3.0 METHODOLOGY AND ANALYTICAL PROCEDURES I

3.1 BACKGROUND

I In 19"l2, the Section III Summer Addenda of the ASME Boiler and Pressure I Vessel Code 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. A-9 I

I Shortly after publication of ASME Code Section III Appendix G, a new Appendix to 10 CFR 50 entitled " Appendix G - Fracture Toughness I Requirements" was published in the Federal Register (July 17, 1973) and became effective on Augurt 16, 1973. This Appendix imposed fracture 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 light water nuclear power reactors both currently operating and under construction. 10 CFR 50 Appendix G, was further revised in July 1983 and 1995. I In addition to Appendix Gr the RCPB muet meet the requirements imposed by 10 CFR 50, Appendix A general Design Cr aria 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 probability of abnormal leakage, of rapid failure, and of gross rupture. 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, the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized. Appropriate and conservative methods, which protect the reactor coolant pressure boundary against nonductile failure, have been developed by combustion Engineering to comply with 10 CFR 50. I 3.2 HETHODOLOGIES 3.2.1 NEUTRON FLUENCE CALCULATIONAL METilODS The methods and assumptions desenbed in this report apply to the calculation of vessel fluence for core and vessel geometrical and material configurations typical of ABB-CE pressurized water reactors. This methodology meets the requirements of a proposed NRC Regulatory Guide (currently Draft Regulatory Guide 1053:

• Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence").

A-10

I The prediction of the vessel fluence is made by a calculation of the transport of neutrons frem the core out ta the vessel and cavity. The calculations consist of the following steps (1) determinatien of the geometrical and material input data, (2) determination of the core neutran I m source, and (3) propagation ef the neutron fluence from the core to tha vessel and into the cavir-y. A qualification of the calculational procedure is descrih d later. The discrete ordinate method is 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. Adherence to this methodology will accelerate NRC review of future fluence calculation submittals. 3.2.1.1 Input Data I 3.2.1.1.1 Materials and Geometry Detailed material and geometrical input data are used to define the physical characteristics that determine the attenuation of the neutron 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, core, and internals. The geometrical input data includes the dimensions and locations of the fuel assemblies, g 3 reactor internals (shroud, core support barrel, thermal shield, and neutron pads),' the pressure vessel (including identification and location of all welds and plates) and cladding, and survaillance capsules. For cavity dosimetry, input data also includes the width of the reactor cavity and the material compositions of the support structure and concrete I 3 (biological) shielding, including water content, rebar and steel. The data is based, to the extent possible, on documented and verified as-built dimensions and plant-specific materials, The isotepic compositions of important constituent nuclides within each region are based on as built materials data. In the absence of plant-specific information, nominal E E 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 A-ll I

I flucnce uncertainty. The determination of the concentrations of the 2 major isotopes responsible for the fluence attenuation (e.g., iren "nd water) are emphasized. The water density is based on plant full power operating temperatures and pressures, as well as standard steam tables. The data accounts for axial and radial variations in water density caused by temperature differences in the co:e and inside the core barrel. 3.2.1.1.2 Cross-sections The calculational method to estimate vessel damage fluence uses neutron I cross-sections over the energy range from'-0.1 MeV to ~15 MeV. Regulatory Guide 1053 recommends the use of the latest version of the The Draft Ovaluated Nuclear Data File (ENDF/B-VI). The ENDF/B-VI files were prepared under the direction of the Cross Sectjen Evaluation Working Group (CSWEG) operated through the National Nuclear Data Center at Brookhaven I National Laboratories (BNL). These data have been thoroughly reviewed, tested, and benchmarked. I 3.2.1.1.2.1 Multi-group Libraries. Since the discrete ordinates transport code used to determine the neutron l fluence uses a multi group approximatici., 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 Legendte

           . 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 eccuracy 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 utiderprediction 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 iron data are included in ENDF/B, version VI. t . I A-12

 .I A

I 3.2.1.1.2.2 Constructing a Multi-group Library. The ENDF/B-VI files were first processed into a problem-independent, fine-g multi-group, master library containing data for all required isotopes. 5 This master library (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. E The G library also contains 42 photon energy groups. The master library .as collapsed into a job (broad group) library over spectra that closely approximate the true spectra. The resulting library (BUGLE-95) contains 47 neutron an'f 20 photon groups. This reduction was accomplished with a one-dimensional calculation that included the discrete regions of the core, vessel internals, by-pass and downcomer water, pressure vessel, reactor cavity, shield, and support structures. This job library includes 27 anergy groups above -0.1 MeV. The collapsing was g performed over four dif ferent spectra typical of PWR's, i.e. the core, E downcomer, concrete and vessel. Both VITAMIN-B6 and BUGLE-96 are available from Oak Ridge National Laboratory. 3.2.1.2 Core Neutron Source The determination of the neutron source for the pressure vessel fluence calculations accounts for the temporal, seatial, and energy dependence together with the absolute source normalization. The spatial dependence of the source is based on ROCS /MC depletion calculations that incorporate core operation. The accuracy of the power distributions has been demonstrated in NRC approved Topical Reports (References 9, 10, 11). The depletion calculations are 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 scheme. The detailed state-point dependence must be accounted ,4 for, but a cycle average pswer distribution inferred from the cycle A-13 l I

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 nanagement 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 te avoid overprediction of the fluence. The pin-wise source distribution generated by MC is used for best-estimate, and represents the absolute source distribution in the assembly. The MC pin power distribution in (x-y) geometry is converted into a (r-0) distribution as required by the (r-0) DORT geometry with the RTHETA code. The local source is determined as the product of the fission rate and the 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 l 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 nominsi plane. A planar-octant representation is used for the octant-symmetric fuel-loading patterns typically used in ABB-CE plants. For evaluating dosimetry, the octant closest to the dosimeter capsules may be used. To accurately represent the important peripheral assembly geometry, a 0-mesh of about 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 A-14

I 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. 3.2.1.3 Fluence calculation 3.2.1.3.1 Transport calculation The transport of neutrons from the core to locations of interest in the pressure vessel is determined with the two-dimensional discrete ordinates transport proaram DORT in (r,0) geometries. An azimuthal (0) mesh using about 80 intervals over an octant in (r,0) geometry in the horizontal plane provides an accurate representation of g the spatial distribution of the material compositions and source 5 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 recommends 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 except g near material and source interfaces, where flux gradients can be large. E An S-8, fully symmetric angular quadrature is used for deterruining the fluence at the vessel. Past and future DOT /DORT models applied to ABB-cE plants meet these requirements. 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 I M i A-15 II I

I cylinder defined by 0< r< R' with a fictitious vacuum-bonndary condition applied at R'. From this initial calculation a boundary source is 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). I region had to be tested (e.g., by decreasing tlae inner radius of the outer The adequacy of the overlap region) to ensure that the use of the fictitious boundary condition at R' had not unduly af fected the boundary soures at R" or the results at the vessel. Current workstations do not present this limitation, and the entire problem can now be solved as one fixed source problem. A point-wise flux convergence criterion of < 0.001 should be used, and a sufficient number of iterations should be allowed within each group to ensure convergence. To avoid negative fluxes and improve convergence, a weighted difference model should be used. The adequacy of the spatial 1 HeV fluence through vessel and the resulting increased sensitivity to the iron inelastic-scattering cross-section and (b) the possibiltty of neutron streaming (i.e., strong directionally dependent) effects in the low-density materials (air and vessel insulation) in the cavity. Because I of the increased sensitivity to the iron cross-sections, ENDF/B-VI A-18 I

l cross-section data should be used for cavity fluence calculations. Typically, the width of the cavity together wi .. the close-to-beltline locations of the dosimetry capsules result in minimal cavity streaming effects, and an S8, angular quadrature is acceptable. However, when off-beltline locations are ..na l yz e d, the adequacy of the S8 quadrature to determine the streaming component must be demonstrated with higher-orden g g 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 bc represented as accurately as possible. Benchmark g g measurements involving simulated reactor cavities are recommended for methods evaluation. When both in vessel and cavity dosimetry measurements are available, an additional verification of the reacurements and calculations may be made by comparing the vessel inner-wall fluence determined from (1) the absolute fluence calculation, (2) the q extrapolation of the in-vessel measurements, and (3) the extrapolation of the cavity measurements. 3.2.1,4 Methodology Qualsfication and Uncertainty Estimates Regulatory Guide 1053 :equires that the neutron transport calculational methodclogy be qualified and that flux uncertainty estimates be determined. The neutron flux undergoes several decades of attenut. tion before reaching the vensal, and the calculation is sensitive to the material anu geometricil representation of the core and vessel internals, the neutron source, an d the numerical schemes used in its determination. The uncertainty estimates are used to determine the appropriate uncertainty allowance to be included in the application of the fluence estimate. While adhelence to the guidelines described in the Fegulatory Guide will generally result in accurate flue. ace estimates, the overall methodology must be gialified 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 da ta libraries used in the calculations, may have been found to be acceptable in previous applications, the qualification ensurro that the licent ee's f mplementation of the metbodology is valid. The methods qualification consists of three parts: (1) the analytic uncertainty analysis. (2) the comparison with benchmarks and A-19

[ plant-specific data, and (3) the estimate of uncertainty in calculated fluence. 3.2.1.4.1 Analytie Unoortainty Analysis 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 ditticulty of these calculations, the method's qualification by comparison to measurements must be made to 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 rei,iable estimate of the calculational bias (i.e., they_ represent a.statis*.ically significant measurement data base), the comparisons to measurement may be used to (1) determine the offeet 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.

           =As an additional qualification, the sensitivity of tha calculation to the important input and modeling parametors must be estermined and combined wish the uncertainties of the input

(' and modeling parameters to provide an independent estimate of the o1trall

             . calculational uncertainty.

[ -To demonstrate thre accuracy of the methodology, an analytic uncertainty analysis must be performed. 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),

h. Geometry (locations of components and deviations from the nominal dimensions),

Isotopic' composition of material (density and composition of coolant water, core barrel, thermal 1 shield, pressure vessel with cladding, and concre_te shield), _. A-20 L , -

I lieutton 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 J1st, is not neressarily exhaustive, and other uncertainties that are

                      ,                                                                         g specific to a particular reactor or a particular calculational method                    3 sliould be considered. In typical applications, the fluence uncertainty is rJominated by a few uncertainty components, such as the geometry, which are usually easily identified and substantially simplify the uncertainty analyais.

The pensitivity of the flux to the significant component uncertainties should be determined by a series of transport sensitivity calculations in which the calculativnal model input data and me,deling assumptions are varied and the effect on the calculated flux is determined. (A typical E sensitivtty would be 15% decrease in vessel >1 MeV fluence per E centimeter increann in vessel inside radius.) Estimates of the expected uncertainties in these input parameters must be made and combined with the corresponding f.'luence sensitivities to determine the total calculated. In performing calculations of surveillance capsule, it should be noted that the capsule fluence is extremely sensitive to the representation of the cap-ule geometry and intes;nal water region (if present), and the adequacy of the capsule rep.tesentation and mesh must be demonstrated using sensitivity. The capsule fluence and sipectra should accurately represent their sensitivity to the radial location of the apsule and its proximity to material interfaces (e.g., at the vessel, thermal shield, and concrete shield in the cavity). 3.2.1.4.2 comparison with Ber chmark and 71snt-specific Heasurenwnts calculational methods must be validated by comparison with measurements and calculational benchmarks. Three types of comparir.ons are required e operating reactor in-vessel or ex-vessel dosimetry measurements, A-21 I

e pressure vessel simulator e calculational benchmarks I The methods used to calculate the bc.3chmarks must be consistent with those I used to calculate fluence in the vessel. Calculated reaction rates at the dosimeter locations must agree with the meksurements to within about 20% for in-vessel capsules and 30% for cavity dosimetry. If the observed deviations are larger, the methodology must be exae,.ined and refined to 1:nprove the agreement. I 3.2.1.4.2.1 Operating Reactor Measurements comparisons of measurements and calculations should be performed for the specific reactor being analyzed or for reactors of similar physical and fuel management design. Plant-specific measurements provida the combined analytical and measurement uncertainties, including the ef* nt of as-built variations. A good estimate of the vessel attenuation can be obtained 1 I when both in-vessel and cavity dosimetry are available. 3.2.1.4.2.2 Pressure Vessel Simulator Measurements l A number of experimental benchmarks providing detector reaction rates in the peripherol fuel assemblies, within the vessel wall, and in the cavity are available for the purpose of methods calibration. i 3.2.1.4.2.3 calculational Benchmarks A calculational benchmark commissioned by the NRC and prepared by Brookhaven National 1,aboratory (Reference 12) provides very detail input description as well as the flux solution at several mesh points. An analysis of this benchmark, whf eh aditwsses both standard out-in and low-I leakage fuel managoraent, provides a detailed test of the cross sections and various calculational options for transport calculation. I I A-22 l 1

                                                                                                                                  )

3.2.1.4.3 overall Bias and Uncertainty An appropriate conbination of the benchmark results provides the bias and g uncertainty to be applied to the predicted fluence. E i l 3.2.2 REACTOR VESSEL MATERIAL SURVEILIANCE PROGRAM ASTM E 185, " Standard Recom-ended Prr.ctice for Surveillance Tests for Nuclear 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 spe cimens , the design and placement in the reactor vessel of the test specimen compartments, and the means for measuring neutron fluence a;,d irradiation temperature. These are aspects pertaining to the design of the program. ASTM E 185 also provides a schedule 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 g Engineering owners Group plants was designed to meet or exceed the 3 requirements of the version of E 185 in effect at the time. For ea:h vessel, base metal was selected from one of the beltline plates and used to fabricate test specimens for pre-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 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 set g aside as archive material for subsequent use. Neutron flux and 5 temperature monitors, and test specimens from the surveillance plate, weld and HAZ together with specimens from a correlation monitor material were loaded into compartments and assembled into surveillance capsules. A mini num of six surveillance capsules were originally provided for each CEOG plant. Records were compiled which 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 A-23 ,

t I compartments, and the location of individus1 specimens in the compartments for each capsule assenbly. I The six surveillance vall capsules were installed in holders on the inside surface of the reactor vessel and within the region surrounded by the effective height of the active reactor core. The vessel wall location 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 which is being 'g monitored. (Note See plant-specific details for azimuthal location of 3 the wall capsules and, if applicable, for additional 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.) The surveillance capsule withdrawal schedule was originally established following the requirements of the version of E 105 in effect at the times the schedule may have been originally established based on the requirements of 10CFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements. The schedule called for at least three capsules to I be removed and tested during the design life of the reactor vessels remaining capsules were available to provide a higher frequency of testing the or to provide supplemental monitoring. The surveillance capsule withdrawal schedule can be modified as necessary for compliance with changes to Appendix H and for meeting plant-specific needs in keeping with the original purpose of monitoring changes in the fracture toughness properties of the vessel beltline materials. Post-irradiation testing is presently performed on the specimens from the withdrawn capsule in accordance with the requirements of ASTM E 185-82 (or I later versions, as specified in Appendix H) and 10CFR50, Appendix H. test data and evaluation results are compiled and presented in a report. The Application of the data for the PTLR are discussed in Section 3.2.4. The initial properties of the reactor vessel beltline plates and welds were establishci in parallel to the establishment of the reactor vessel surveillance program. For each of the beltline plates, Charpy impact tests and/or drop weight tests were performed to demonstrate compliance I with applicable ASME Code and vessel specification requirenients. The welding procedures used for beltline welds were qualified and the welding materials certified to applicable AWS, ASME Code and vessel specification A-24 I

requirements. Chemical analyses of the plates and weld deposits were obtained in accordance with the vessel specification. [Notet The data available for a specific vessel will vary because of differences in the g requirements for testing and certification.) The data were processed to 5 obtain a value of the initial reference temperature, RT.n, and of the copper and nickel content. For beltline plates and welds, the initial RT n was determined in accordance with the ASME Code, Section III, NB-2331, for which drop weight tests and Charpy impact tese* (complete transition curve) were performed. For the earlier CEOG vessels for which test requirements were dif f erer.t, the initial RT,n was determined using Branch Technical Position MTEB $-2, Fracture Toughness Requirements (for 4 Older Plants), or a generic value of initial RT pt was determined based on measurements for a pecific set of materials. Some CEOG sponsored efforts g which are pertinent are report CEN-109, December 1981, " Evaluation of E Pressurized Thermal Shock Effects due to small Break LOCAs with Loss of reedwater for the Combustion Engineering NSSS", and CENPSD-1039, December 1996 (Task 902), "Best Estimate Copper and Nickel Values in CE rabricated Reactor Vessel Welds". 3.2.3 METHOD TOR CALCULATING BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURE (ART) The ART is deterndned in accordance with Regulatory Guide 1.99, Revision 2 (May 1988) *, " Radiation Embrittlement of Reactor Vessel Materials". ART is determined as follows: ART = Initial RTun + 4 RTun + Margin Initial RTun is the reference temperature for the beltline plate or weld material as described in Section 3.2.2. A RTun is the shif t 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 g neutron fluence factor (using the neutron fluence at the vessel depth of M interest). The margin is the root mean squared value using the uncertainty in the initial RTan and the uncertainty in the shift in reference temperature. The uncertainty in the initial RTun for a measured value of RTan is based on the precision of the test methods the uncertainty for a generic value is the standard deviation of the data used I E A-25

l l l to obtain the generic value. The shift uncertainty for base material (plates) is 1*/ 'T and for welds is 20 'F. When credible surveillance data, as defined by Regulatory Guide 1.99, Revision 2, c a available, the chemistry factor may be modified and the uncertainty in the shift in reference temperature may be reduced. The I process is as described in the Guide and is discussed further in Section 3.2.4. [ Note: Upon issuance of a new revision of Regulatory Guide 1.99, the ART calculation methodology will be evaluated and, if applicable, the new methodelegy will be cited in subsequent revisions of the PTLR.) 3.2.4 APPLICATION OF SURVEILLANCE DATA TO ART CALCULATIONS Data from the reactor vessel surveillance program is 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 cn anomalous I 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 I for the vessel. The post-irradiation analysis of the surveillance capsule contents provides measurements of the neutron fluence and of the changes 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. For reactor vessels from which two or more surveillance capsules have be",n removed, an evaluation is performed to compare the measured RTut shif t to the predicted RTnn shift. Using Position 1.1 of Regulatory Guide 1.99, Revision 2, the mean predicted shift is calculated using the applicable chemistry facter from Table 1 or 2 of the Guide and the measured neutron fluences for the irradiated material. The predicted RTun shift plus two standard deviations for shift (base or weld metal, as applicable) should be greater than the measured shift for each of the surveillance capsule measurements. If this is not the case (i.e., one or more of the measured shifts exceeds the predicted RTan shift plus two standard deviations for A-26

                                                                   -       _ _ _ - _   i

shift), a supplement to the PTLR will be provided which demonstrates how those results will affect the approved methodology. When data are available from two or more capsules, an evaluation may be performed to determine 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 ) J are within 06 of the shift predicted using Position 2.1 (12 e4 if the  : fluence range is large), (4) the capsule irradiation temperature is , comparable to that of the vessel, and (5) the correlatinn 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 chendstry f actor 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 1.1, the revised values should be incorporated into the PTLR methodology. l If the Position 2.1 values result in a lower value of ART, either the 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 to the controlling vessel material and there are data for the controlling material available from another plant, the plant-specific PTLR may utili'te l another plant's surveillance data as the basis for the ART prediction methodology. If such data are employed, the source of the data will be identified and the basis for using it 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 Guide 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 A-27

( Task 904 which addresses methodology for the application of both plant-specific and sister plant data to ret h ART calculations. 3.2.5 METHOD FOR DEVELOPING LTOP SYSTEM LIMITS Appendix B describes the general criteria considered to ensure Low Temperature overpressure Protection (LTOP) for pressurized water reactors designed by Combustion Engineering. In addition, the re)ationship between LTOP setpoints and Recctor Coolant System (RCS) pressure-temperature (P-T) limits is discussed. LTOP is achieved in accordance with current criteria defined in the ASME Boiler and Pressure Vessel Code and 10 CFR 50, [ Appendix A, Design Criterion 15 and Design Criterion 31. The NRC guidance which ensures overprossure protection was published in NUREG-75/087 (currently NUREG 0000), and contains two documents, Standard Review Plan 5.2.2, " Overpressure Protection"I7) and Branch Technical Position RSB 5-2, "overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures"(8I. BTP RSB 5-2 requires protection of the Appendix G, paragraph. G-2215 of ASME Boiler and Pressure Vessel Code Section XI, Division 1, limits to meet the criteria established in 10 CFR 50, Appendix A. Specifically, the LTOP system is required to prevent pressure from exceeding a limit defined by 110% of the pressure which satisfies Appendix G. 3.2.6 APPLICATION or THE LIMITING ART IN THE P-T CURVE CALCULATION Practices, methodologies and techniques which are utilized in the development of the pressure-temperature limits, along with justification k of the aforementioned, are described briefly herein. Detailed technical descriptions of the pertinent items are given in Section 3.3. These { limits have been developed to meet the requirements of 10 CFR 50 Appendix o. A brief technical description of the procedures practiced by combustion Engineering to develop brittle fracture limits is given for the required components of the reactor coolant pressure boundary. These techniques are applicable to all Combustion Engineering NSSS's. These techniques have been applied to nuclear power plants designed to ASME Code editions later f A-28 I -

                                                                                       )

than the summer 1972 Addenda since the incorporation of Appendix G to 10 CTR $0 in 1973. These analytical techniques are based partially on Linear Elastic Tracture Mechanics (LErM) and prov'de appropriately conservative design loadings for the ferritic components of the reactor coolant pressure boundary to preclude brittle fracture. Currently, the ferritic components of the reactor coolant pressure boundary specifically addressed by Appendix G to section III of the ASME Code'h are delineated as follows:

1. Vessels

2. Piping, Pumps and Valves

3. Bolting The vessel is the only location for which a LEPH 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 1g. outlet Norzle The " beltline" refers to the region of the reactor vessel which is immediately adjacent to the fuel core and is exposed to highest levels of fluence. Typically, this would only be the large cylindrical shell section of the RV. However, for some plant designs this " beltline" region may include not only the shell sections but also the Vessel Wall A-29

l Transition. However, the material with the highest ART usually falls I within the cylindrical shell region. These locations have been analyzed utilizing the principles of LEIH described by Appendix G to Section III of the ASME Code. These analyses I considered plant heatup, plant cooldown and an isothermal leak test. A brief description of the general criteria follows. 3.2.6.1 oeneral Method I In accordance with Appendix G of the ASME Code, the mode I (opening model stress intensity factor, K , gis utilized and calculated at numerous intervals throughout the transient. The Kg was calculated at the crack tip of a postulated flaw. The postulated flaw site for the considered locations, except the flange and nortles, 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 and outside surfaces. The determination of the applied K g is based on the results of a two I dimensional heat transfer analysis and consideration of the primary membrane stress, om , primary bending stress, ob, secondary membrane stress, o,,, and secondary bending stress, o b. The resulting K for each m g component of stress can be calculated as follows: I Kg , = M, (membrane stress) KIb " Mb (bending stress) where M m and Hb are defined graphically in Appendix G of the ASME Code (2) , l'or computational simplicity, equations A3-4 and A3-6 of WRC Bulletin 175(33 were utilized, and are: 1.1M gE and Mt= 245 A-30

I Mb "I ) where M,MB= g correction 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 rene size T = the section thickness (in). For each point in the transient analyzed, the allowable pressure is determined by comparing the allowable stress intensity, Kgg, to the applied stress intensity and includes a conservative factor of safety. The value 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: Kgg = 26.78 + 1.231 exp!0.0145(T-RTNDT + 160)) kai Vin where, I T = crack tip temperature ('F) RTNDT = reference nil ductility temperature at the cracktip location i For plant heatup and plant cooldown, the following expression is used to 1 determine the allowable pressure Kyg > (2.0 Kg ,,+ 2.0 KIbI + IKIm + KIbI I PRIMARY SECONDARY For loak tests, the expression utilized to calculate the Kg due to test pressure is A-31 l

I Kg " (1.$ MIm + 1'b EIbI + IE!m + E!bI<E IR I I'RIMARY SECONDARY 3.2.6.2 rianges 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 aire 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 Kg is determined utilizing equations A3-1 and A3-2 along with rigures A3-1 and A3-2 f rom WRC Bulletin 175 ". The remainder of the 8 procedures, .s described previously are also applicable to the flange region. The flange region is also considered by either the design method I previously described or by applying the minimum requirements of 10 CPR 50 Appendix G. This minimum requirement is applicable when pressure exceeds 20 percent of the preservice system hydrostatic test pressure, the I temperature of the closure flange regions that are highly stressed by the bolt preload enust exceed the reference temperature of the material in those regions by at least 120'F for normal operation and by 90'r for hydrostatic pressure tests and leak tests. I 3.2.6.3 Nossles I In the case of the primary inlet and outlet nottles the method described in Appendix 5 to WRC Bulletin 175(3) Kg calculation Method for Nozzle, was utilized. In this analysis the postulated flaw site was equal to one-tenth of the vessel wall thickness and located on the inside corner of the nortle adjacent the vessel wall.- Again, the flaw site is confidently detectable A-32 I

l with the in-shop HDE techniques and consistent with the acceptance l'l standards of Sub-Article NB-5320 of Section III to the ASME Code. The appixed Kg due to membrane stress are determined utilizing Equation A5-9 in conjunction with figure A5-1, both from Reference 3. The bending stress intensity factor is calculated in the same manner as the other locations. The solution for the allowable pressure is still based on Kyp 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 unitradiated condition, show that for heatup, cooldown and isothermal leak test, the limiting locations are the vessel shell at the vessel flange, the inlet nozzle and the upper shell at the vessel wall transition, respectively. 3.2.6.4 Beltline In the developn.ent of operational limits, Combustion Engineering analyzes the reactor vessel beltline region considering the predicted et hets of neutron fluence over a specific time period. The beltline region is the g only location which receives sufficient neutron fluence to substantially 3 alter the toughness properties of the material. Therefore, the beltline region will become the controlling location when compared to the other reactor coolant system locations analyzed. Combustion Engineering considers the beltline region to be controlling, that is, the most limiting with respect to allowable pressure at any specific temperature, when the shift in RTNDT due to neutron radiation in the beltline causes the ART to be greater than the unitradiated RTHDT of the surrounding locations. This philosophy is consistent with the guidance given in Standard Review Plan 5.3.2, Pressure-Temperature Limits W . The beltline region is analyzed utilizing LEPH procedures described in Section 3.3 in con

• junction with the shift prediction methods of Regulatory A-33 1

[- Guide 1.99 Revision 2(6I to account for the reduction in fracture toughness due to neutron irradiation. 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 pressuriter are included. Pressucirer pressute indicator loop { uncertainties are also included and consequently, the limits are provided on coordinates of indicated pressurizer pressure versus indicated RCS (cold leg) temperature. 3.2.7 METHOD TOR ADDRESSING 10 CTR 50 MINIMUM TEMPERATURE REQUIPEMENTS IN THE P-T CURVES [ 3.2.7.1 Inservice mydreatatic Pressure Test and core Critioal Limite Both-10 CTR Part 50 Appendix G and the ASME Code, Section XI, Appendix G require the development of pressure-temperature limits which are applicable to inservice hydrostatic tests. For hydrostatic tests

                           . performed subsequent to loading fuel into the reactor vessel, prior to

[. cora criticality, the minimum test temperature is determined by evaluating K,g the mode I stress intensity factors. The evaluation of K is g h.- 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 which 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 CPR Part 50, specifies pressure-temperature limits for

{ core critical operation to provide additional margin during actual power operation. . j A-34

The pressure-temperature limit for core critical operation is based upon I 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 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 'r, and be at least 40

      'T higher than the minimum pressure-temperature curve for normal operation E

heatup or cooldown. For vessel pressure greater than 20 percent of 5 preservice system hydrostatic test pressure, the criteria are that the reactor vessel must be the larger of the minimum permissible temperature for the inservice system hyrdostatic 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 'r, and be at least 40 'r higher than the minimum pressure-temperature curve for normal operation heatup or cooldown. Note, that the core critical limits established utilizing this criteris are solely based upon fracture mechanics considerations. These limits u. not consider core reactivity safety analyses which can control the temperature at which the core can be brought critical. 3.2.7.2 Minimum Boltup and 1,owest Service Temperature I When developing operating limits for the RCS, two additional requirements are considered. These requirements are for the minimum boltup temperature and the lowest service temperature, m The minimum boltup temperature is established based on ASME Code Section . III, Subparagraph G-2222.c. It requires that the minimum metal temperature in the stressed region of the flange and adjacent shell region be at least the initial RTNDT of the material in the highly stressed region when stressed by the full intended bolt preload and by pressure not exceeding 20% of the pre-operational system hydrostatic test pressure. A-35 I

l 2 The lowest service temperature is defined by ASME Code as "the minimum I temperature of the fluid retained by the component or, alternatively, the calculated volumetric average inetal temperature expected during normal operation, whenever pressure exceed: 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. I I The lowest service temperature is established based on the limiting material RT yp7 for piping, pumps, and valves of the primary reactor coolant pressure boundary plus 100'F. I 3.3 TECHNICAL DESCRIPTION FDR CALCVLATING BELTLINE REGION PRESSURE

• TEMPERATURE LIMITS This section presents the analytical techniques and methodology for developing beltline pressure-temperature limits which are utilized in the composite RCS operating limits. The method is directly applicable to heatup, cooldown and inservice hydrostatic tests.

3.3.1 GENERAL The ana11 .ical procedure for developing operational pressure-temperature limits for the reactor vessel beltline utilizes the methods of Linear Elastic Fracture Mechanics (LE!H) found in the ASME Boiler and Pressure vessel Code Section XI, Appendix G(6) in accordance with the requirements of 10 CFR Part 50 Appendix GIII. For these analyses, the Mode I (opening mode) stress intensity factors are used for the solution basis. I 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. A-36 __J

I 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 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.

• t each of the postt' lated flaw locations, the Mode I stress intensity factor, Kg , produced by each of the specified lo.idings is calculated and the summation of the Kg values is compared to a reference stress intensity, Kgg. K gg is the critical value of K for y the material and temperature involved. The result of this method is a relation of pressure versus temperature for each reactor vessel operating period which precludes brittle fracture. K yp 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 e defined by the following expression Kg p = 2 6. 7 8 + 1. 2 3 3 exp ( 0. 014 5 (T- ART + 160 ) ) kaiVin

where, I

Egg = reference a ress intensity factor, KsiYin T = temperature at the postulated crack tip, 'T g ART = adjusted reference temperature at the postulated crack u tip, 'T (predicted in accordance with Section 3.2.3) For any instant during the postulated heatup or cooldown, K gg 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 instant during the heatup or cooldown the temperature gradients across the g reactor vessel wall are calculated (see Section 3.3.2) and the g corresponding thermal stress intensity factor, Kg7, is determined. A-37

L F b Through the use of superposition, the thermal stress intensity is subt:aeted from the available M gg to determine the allowable pressure stress intensity factor and consequently the allowsbie pressure. F L In accordance with the ASME Code section XI Appendix G requirements, the general equations for determining the allowable pressure for any assumed rate of temperature change during Service Level A and B operation ares 2Kgg + KIT < Kgg 433 1.5Kyg + Kg7 < KIR (Inservice Hydrostatic Test) where, { K gg = Allowable pressure stress intentity factor, KJi Vin K IT

                                                                            =

Thermal stress intensity factor, Kai Vin K yp = Reference stress intensity factor, Kai Vin In addition, the 1995 ASME Code, Section XI. Appendix G has introduced relief for pressurized water reactors (PWRs) with low temperature overpressure 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 .9ervice Level A or B events. When using the Section XI Appendix G requirements, the LTOP systems are effective at coolant temperatures less than 200 *r or at coolant temperatures corresponding to a reactor vessel metal temperature 1ess than RTm + $0 'r, whichever is greater. The LTOP systems will limit [ the maximum pressure in the vessel to 110% of the pressure determined to satisfy equation (1) defined above. [ . [ A-38 c ._ .. . . .

I 3.3.2 THEPyAL NJALYSIS METHODOLOGY NJD INTLUENCE COEFFICIENTS The Mode I thermal stress intensity factor is obtained through a detailed thermal analysis of the reactor vessel beltline wall using a computer code. In this code a one dimensional, tnree noded, isoparametric finite element suitable for one dimensional axisymmetric radial condt. : tion-convection heat transfer is used. The vessel wall is divAded into elements and an accurate distribution of temperature as a function of radial location and transient time is calculated. The code utilises g convective boundary on the inside wall of the vessel and an insulated g boundary on the outside wall of the vessel. Variation of traterial properties through the vessel wall are permitted 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 I is governed by a partial differential equation, 2 Or 0T 1 Pr pC bt =K +-- pg 2 r Or subject to the following boundary conditions at the inside and outside I wall surface locations: Or At r = rg -K p = h(T-Ty , er At r = r -

                                                      = 0 o             or where, l

p = density, Ib/ft 3 C = specific heat, btu /lb 'r K = thermal conductivity, btu /hr-ft 'r T = vessel wall temperature, 'T "" r = radius, ft A-39 I

I I t a time, hr h a convective heat transfer coefficient, btu /hr-ft 2.,r Te = RCS coolant temperature, 'r

                                                  =

rg, ro inside and outside radii of vessel wall, it I The above is solved numerically using a finite element model to determine wall temperature as a function of radius, time, and thetmal rate. The thermal stress intensity factors are determined from the temperature profile through the beltline wall using thermal influence coefficients and superposition. ASME Code Section XI Appendix G recognizes the limitations of the method which it provides for calculating K IT because of the assumed temperature profile. Since a detailed heat transfer analysis results in varying temperature profiles (and consequently varying thermal stresses), I this alternate enethod for calculating M IT was employed as suggested by Subparagraph G-2214.3 of Reference 2.

I The superposition technique employed is temperature profile based rather than the stress nrofile based which is typically used. A third order polynomial fit to the temperature distributions in the wall was used and is given by:

T(n) =Co+C1 (1 */h) +C 2 II' / h) +C3 (1 X/h) where, T(x) I

                                                                                =  Temperature at radial location x from inside wall surface                                 <

C,C,C,C3 O g 2 = coefficients in polynomial fit x = Distance throug). beltline w.sil, in h = Beltline wall thickness, in I These coefficients are utilized in determination of the applied stress intensity. I A-40 I

I Temperature based influence coefficients for determination of the thermal stress intensity factor, M IT, are used. The influence coefficient are dependent upon the geometrical parameters associated with the maximum g postulated defect, and the geometry of the reactor vessel beltline region 8" (i.e., ro/rg, a/c, a/t), along with the unit loading. These influence coefficients were computed for various reactor vessel geometries using a two dimensional reactor vessel finite element model. The postulated flaws were modeled with special quarter point crack tip elements to simulate the singularity present in the stress field due to g the presence of a crack. The influence coefficients were calculated for m unit loadings due to linear, quadratic and cubic thermal gradients, and pressure. These influer.ce coef ficients were then corrected to account for three dimensional effects. For pressure loadings, unit values of the load distributions were used to compute the influence coef ficients. The unit value chosen for internal pressure was 1000 psi. The general equation to compute the Mode I stress intensity factors is for thermal and pressure loading conditions is as follows: 3 K (a) = E C K b i=0 where, My(a) = Total applied stress intensity factor due to loading condition at crack depth, a Cg = Polynomial coefficients f rom the curve fit to the temperature or stress distribution through the vessel wall Kg* = Fracture mechanics influence coefficients for a specified loading condition for each term of the l polynomial expression for the temperature or stress distribution through the vessel wall , a a crack depth, in A-41 l I

[ The Kg for each loading con:iition is then summed and compared to the allowable K IR to determine the allowable pressure. r [ The above described methodology utilizes a finite element based approach to generate influence coefficients for the calculation of stress intensity ( factors at the crack tip. Recently, the ASME Code has proposed a similar influence coefficient based approach in Section XI, Appendix G based on work performed by Raju-Newman on cylindrical vessels with 3-D semi-elliptical flaws. Both methods are fundamentally identical and either method is capable of providing adequate influence coefficients to be used in calculating stress intensity factors. 3.3.3 NOPJ4AL OPERATION ( 3.3.3.1 cooldewn Limit _ Analysts During cooldown, membrane and thermal bending stresses act together in tension at the reactor vessel inside wall. This results in the pressure stress intenrity factor, K yg, and the thermal stress intensity factor, f K IT, acting in unison creating a high stress intensity. At the reactot vessel outside wall the tensile pressure stress and the compressiva ( thermal stress act in opposition ret.ulting in a lower total stress than at the inside wall location. Also, the shift in RTNDT and the associated reduction in fracture toughness are less severe at the outside wall compared to the inside wall location. Consequently, the inside flaw location is more limiting and is analyzed for the cooldown event. Utilizing the metal temperature and adjusted reference temperature at the { 1/4t location, the reference stress intensity is determined. From the method provided in Section 3.3.2, the through wall temperature gradient is calculated for the assumed cooldown rate to determine the thermal stress intensity factor. In general, the thermal stress intensity factors are found using the temperature profile through the wall as a function of transient time as described in Seccion 3.3.2. They are then subtracted from the available Kgg value to find the allowable pressure stress intensity factor and consequently the allowable pressure. A-42

The cooldown pressure-temperature curves are thus generated by calculating the allowable pressure on the reference flaw at the 1/4t location based upons Km i = Ka - Mit 2 where, K gg = Allowable pressure stress intensity factor as a function of coolant temperature, Ksi Vin K yp = Reference stress intensity factor as a function of coolant temperature, Ksi Vin K IT = Thermal stress intensity factor as a function of I coolant temperature, Kai Vin To develop a composite pressure-temperature limit for the cooldown event, the isothermal pressure-ternperature limit must be calculated. The isothermal pressure-temperature limit is then compared to the pressure-temperature limit associated with a cooling rate and the more restrictive == allowable pressure-temperature limit is chosen resulting in a composite rate limit curve for the reactor vessel beltline. 3 3.3.3.2 Reatup Limit Analysis During a heatup transient, the thermal bending stress is compressive at g the reactor vessel inside wall and is tensile at the reactor vessel E outside wall. Internal pressure creates a tensile stress at the inside wall as well as the outside wall locations. Consequently, the outside wall location has the larger tot.il stress when compared to the inside wall. However, neutron embrittlement (the shif t in material RTNDT and the associated reduction in fracture toughness) is greater at the inside location than the outarde. Therefore, both the inside and outside flaw locations must be analyzed to assure that. the most limiting condition is achieved. A-43

L [ As described in the cooldown case, the reference stress intensity factor is calculated based on the metal temperature at the tip of the flaw and the adjusted reference temperature at the flaw location. For heatup the reference stress intensity is calculated for both the 1/4t and 3/4t locations. Using the finite element method described in section 3.3.2, the temperature profile through the wall and the metal temperatures at the tip of the flaw are calculated for the transient history. This information is used to calculate the thermal stress intensity factor at the 1/4t and 3/4t locations using the calculated wall gradient and thermal influence coefficients. The allowable pressure stress intensity is then determined by superposition of the thermal stress intensity factor with the available reference stress intensity at the flaw tip. The allowable pressure is then derived from the calculated allowable pressurs stress ( intensity factor. 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 the isothermal steady state condition would represent the limiting P-T limit. However, the isothermal condit'.on may not always pnovide the limiting pressure-temperature limit for the 1/4t location during a heatup transient. This it 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 Lepresentation 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 recogni:ed. These limits account for clad and film differential temperatures and for the gradual buildup of wall differential temperatures with time. At the 3/4t location the pressure stress and thermal stresses are tensile

 -resulting in-the maximum stress at that location. Pressu'.e-temperature A-44
                                                  .                            I

Il limits are calculated for the 3/4t location accounting for clad and film differential temperature and the buildup of wall temperature gradients with time using the method described in Section 3.3.2. The allowable pressure is derived based upon a flaw at the 3/4t location by superposition of the thermal stress intensity with the available reference stress intensity for the metal temperature and adjusted reference terrperature at that position. To develop composite pressure-terr.perature limits for the heatup transient, the isothermal, 1/4t heatup, and 3/4t heatup pressure-temperature limits are compered for a given thermal rate. Then the most restrictive pressure-temperature limits are combined over the complete teroperature interval resulting in a composite race limit curve for the reactor vessel beltline for the heatup event. 4.0 TYPICAL PRESSURE-TIMPERATURE 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 3.0 and in 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.] Reactor Vessel Data Design Pressure = [2500) psia I operating Pressure = [2250) psia Design Temperature = (650]'r Vessel I.R. to Wetted Surface = [87.227) in. Cladding Thickness = (5/16) in. l Beltline Thickness = [8.625) in. s-4, I I

l l I Haterial Cladding - (Type 304 Stainless steel) Beltline - (3A-533 Grade B Class 1) l Beltline Adjusted Reference Temperature l Flaw Location Adjusted RTilDT ('r) l l 1/4 T (191,0) 3/4 7 [13;,o) Initial RT, joy riange Region = (+30)'r Piping, Pumps and Valves = (+90)'r j Pressure and Temperature Correction Frctors 4 AT = (+6)'r I (For Te < 200'rs AP = -77 psi (2 RCP's operating)) (For Te > 200'r; AP = -89 psi (3 RCP's operating)) 4.1 BELTLINE LIMIT CURVES I The beltline pressure-temperature limits calculated for heatup and cooldown are depicted in Figures 4.1-1 through 4.1-4 and have beer, I developed utilizing the methodology described in Section 3.3. These figures provide the operating limitations for the beltline region in terms of an allowable pressure over the operating temperature range for various linea' rates of temperature change. Also, these figures have been corrected to indicated pressurizer pressure and cold leg temperature (Te l. I Depicted in rigure 4.1-5 is the beltline pressure-temperature curve for inservice hydrostatic test. This limit curve is typically developed for an isothermal 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. Again, Combustion Engineering's A-46

I practice is to recommend a minimum tempera'- . for inservice hydrostette. tM based on a test pressure corresponding to * .1 times the design pressure. 4.2 rLANGE LIH1T CURVES The ves.iel flange limits, resulting from the detailted analysis described in section 3.2, are shown in Figure 4.2-1. This figure has been corrected to indicated pressurizer pressure and cold leg temperature. 4.3 CoKP051TE LIMIT CURVES I. 1 The beltline pressure-temperature limits and flange pressure-temperaturo limits discussed in the previous sections form the basis for the cetnp': site g, limit curves. In addition, the requirements described in Section 3.4 are Ei also considered when developing the composite RCS P-T limits. During the development of the composite limits, the heatup and cooldown I rates are chosen based on numerous considerations. The issues involved 4r enabitshing these maximum rates include the impact on the operating window, the selection of the Low Tomperature Overpressure Protection setpoint(s), the plant's physiert limitations, and the economical impact ass miated with loss of electrical power generation. The relative importance of these items are different for each utility and therefore uro not addressed directly in this report. For the purpose of illustration, composite limits were developed t'or heatup and cooldown and are presented in Figures 4.3-1 and 4.3-2, j respectively. These figures show arbitrary rates selected for hentup and cooldown which 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 with respect to operating temperature. The minimum boltup temperature was conservatively estLblished tu be (80l'r and the lowest service temperature was established to be [196)*F. Dath A-47 I

bli g requirements are depicted as part of the composite heatup and cooldown j limits. The composite limit curve for inservice hydrostatic test is shown in i- Figure 4.3-3. The minimui .emperature for inservice hydrostatic pressure test, (322)'r was established based on a test pressure of (2427] psia (1.1 times normal operating pressure). l The limitations associated with core critical operation are developed along with the PTLR figures. These are presented in Section 4.4. 4.4 OPERATIONAL LIMIT CURVF.S r I The operational limits developed for utilities are based on the composite limits presented in the previous section. Typical representations of figures developed for inclusion in the PTLR are presented in Figures 4.4-1 and 4.4-2.

 ,                         Figure 4.4-1 presents typical heatt.p limits developed to protect the RCS from brittle fracture.

I Included with the actual heatup limits are the limits representing inservice hydrostitic test and limits pertaining to core aritical operatt.on. Tie core critical limits were established based on the requirements gn.r in uction 3.2.*l. 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 4.4-2 presents typical cooldown limits established to protect the. RCS from brittle fractur . Again, limits representing inservice I hydrostatic test are also present with the composite cooldown limits. The allowable rates, utilized to develop the cooldown limit curve, are also listed as maximum cooldown rates for the appropriate temperature range. The limitations for critical operation of the core are usually not presented as part of the cooldown PTLR figure. I I A-48 I A

I 5.0 stMMARY

                           /

This report described methodologies and practices utilized in tt.e development of reactor coolant system pressure-temperature limits. The methodology was developed to meet the spet.ific c: :,teria of 10 CFR 50, Appendix G, Fracture Toughness Requit;er.ents and 10 CFR 50, Appendix A, Design Criterion 14 and Design Criterion 31. The current requirements imposed oy 10 CFR 50, App >endix G, atply to pressure-retaining components of the reactor coolant pressure .mundary which are fabricated from ferritic material and apply to any c< iltion of normal operation, including anticipated operational occurrenc 3 and system hydrostattic pressure tears. Section 2.0 providas a list and an operational description of the conditions which require pres.3ure-temperature limits. The method and analytical procedures used in the development of the reactor coolant system pressure-temperature limits are based on line n elastic fracture mechanics techniques described in ASME Boiler and r Pressure Vescal Code, Section XI, Appendix G, Fracture Toughness Criteria for Protectica Against Failure. As noted previously, the required loading conditions are described in Section 2.0. As discussed in Section 3.0, the only component apecifically requiring a LEEN analysis is the reactor vessel. Additional details on the reactor vessel locations which weru analyzed and the technical methodology are also provided. The results of the LEth analysis performed for the reactor vessel provided the limiting locations in the unirradiated condition, for heatup, cooldown and isothermal loak tests the vessel shell at the vest.el flange, the inlet nozzle and the vessel wall transition region, respectively. These results are considered in the development of composite RCS operating li:cits. Typically, when the RCS operating limits are developed for a specific time period, the effect of neutron irrad1Ation has caused the beltline region to beccme the most limiting location in the reactor vessel. Therefore, g when RCS optrating limits .re developed, the beltline region is analyze'd 5 considering the ef fect of neutron irradiation in accordance with 10 CFR 50 A-49 I

I . I Appendix G (i.e., U.S. Nuclear Regulatory Guide 1.99 Revision 2), and the vessel flange region is considered, as a minimum, per the requirements of I 10 CFR 50 Appendix G. To illustrate the application of these methodologies and practices, RCS pressure-temperature limits are presented in Section 4.0 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 providing I limits for normal operation, inservice hydrostatic test, and core critical operation. I

6.0 REFERENCES

1. Title 10 of the Code of Federal Regulations, Part 50, Appendix G, Fracture Toughness Requirements, 1995 Edition.

2. ASME Boiler and Pressure Vessel Code Section III, Appendix G, I Protection Against Nonductile Failure, 1986 Edition.

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

August 1972.

4. U.S. Nuclear Regulatory Commission, Standard Review Plan 5.3.2,

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

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

6. ASME Boiler and Pressure Vessel Code Section XI, Appendix G, " Fracture Toughness Criteria for Protection Against Failure," 1995 Edition.

A-50 I

                                                                                       . /N D Ib

I' I

7. U.S. Nuclear Regulatory Commission, Standard Review Plan 5.2.2,

           " Overpressure Protection," Revision 2, November 1988.

8. Branch Technical Position RSB 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures,"

Revision 1, November 1988.

9. "The ROCS and DIT Computer Codes "or Nuclear Design," CENPD-266-P-A, April 1983,

10. "C-E Methodology for Core Design Containing Gadolinia-Urania Burnable Absorbers," CENPD-275, Rev. 1-P-A, May 1988.

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

CENPD-382-P-A, August 1993.

12. J. F. Carew, et. al, " Pressure Vessel Fluence Calculation Benchmark Problems and Solutions," NUREG/CR-6115 (Draft).

I I I I A-51 1

FIGURE 4.1-1 APPENDIX G P T LIMITS ~ HEATUP 2.500 ISO 10 FMR -

                               ~

30 FMR - SCFMR

                               ,                                                       70 FMR
                                                                    ,                  90 FMR

) 2,000 l -

                               ~

g / ( ' l 0 - d g 1,500 l I E

o. -

E

                               ~

I 5 m

                $       1,000
                                                                                     /

I @ O

                               ~

90 FMR 70 FMR soFMR I g '30 FMR E _ ISO InFMR - 500 m o s I  ; I " ' O 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE -Tc, DEG. F I Tc < 200*F, AP = -77 psia Tc 2 200*F, AP = -69 psia AT = + 6' F ART 1/4t = 191.0'F 3/4t = 137.0"F I I ^-'

I FIGURE 4.1-2 I APPENDIX G P-T LIMITS HEATUP 2.500 , I ---* ISO 20 FMR 40VMR

                   -                                                                60 FMR 80FMR 2,000 I
                  ~

5 - E - a g 1,500 , m l E - m ~ m 1,000 l g _ 1 )0 FMR - 80FMR

                 ~

60 FMR

                 -           40 FMR      ~

Z 20 FMR - 500 - I 7 /

                             % [r$                                                                                    l w o
                                                     /

s / O I 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE -Tc, DEG. F Te < 2OO'F, AP = -77 psia ART Tc 2 2OO'F, AP = -69 psia 1/4t = 191 O'F I AT = + 6'F 3/4t = 137.O'F I! A-53

l l FIGURE 4.1-3 APPENDIX G P-T LIMITS COOLDOWN 2.500 I  : 2.000 5 - E I uf g 1.500 E - a:-

                       ~

l @ 1.000

                                                                                /

E '= o . 10 FMRl

         &             - 30 FMR
                       . 50FMR
         $             _ 70 FMR 500                                             '

I ~

%97//

JV / l/ _._ ,p , l . as/

                                     -J-//

0 ' ' ' 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE -Tc, DEG. F Te < 200'F, AP = -77 psia ART Tc 2 200*F, AP - -69 psia 1/4t - 191.0'F AT = + 6

• F 3/4t = 137.0'F A-54

__ _ _ _ _ _ _ _ _ _ _ ]

                   .-                          .      . _ - ~                                   ..

FIGURE 4.14 APPENDIX G P T LIMITS COOLDOWN 2,500

                 -                                            I 2,000
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E - uf g 1,500 s _

                                                                                                   /

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                                                                                       /

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                               '83 E             _

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                 - 6oFMn O                     eorms          -

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                    '           '                            '     '     '                  'l 0

50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE Tc, DEG. F Tc < 200'F, AP = 77 psia ART I Te 2 200'F, AP = -69 psia 1/4t = 191.00F AT = + 6' F l 3/4t = 137.O'F A-55 I

[. - [l-- FIGURE 4.15 APPENDIX G BELTLINE P-T LIMITS HYDROSTATIC 2.500 ... - 2,000-f- < -

                  .[               _

\ k** _ r a: d - r-g ; - 1,000 g .

                                                                                     /
                  ?
                  -h-y             /

500-

                                    ~

0 ' 50 100- 150 200 -250 300 350 400 450 500 INDICATED RCS TEMPERATURE - Tc, sF Tc < 200*F, AP = -77 psia ART-Te it 200'F, AP = -69 psia - 1/4t = 191,0*F AT = + 6'F 3/4t = 137.O'F A-56

                                                                                           . . , ,                                      N

1 I FIGURE 4.21 APPENDlX G FLANGE LIMITS I 2.500 HEATUP i I  : I I 2,000 1M FMR I $ l E - 1,500 I _ T5FMR

o. -

m ^ I sv>

          $     1,000
                       ~

sor m R Y - 5 - I  ! [ 500 E i 0 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE -Tc, DEG. F Te < 200'F, AP = -77 psia ART i Te ;t 200*F, AP = 69 psia 1/4t = 191.O'F AT = + 6'F 3/4t = 137.0*F I g .5>

I I FIGURE 4.3-1 COMPOSITE APPENDIX G P-T LIMITS I 2,500 , I HEATUP _ @'EL*My"'_ S *- EEE (IIH; F) I 100 FAR 2.000

                       ~

I $ ! U) SO FA4R A ~ BELTLINE

                                                                                                            /

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                                                                                                                   \
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BGTLINE I l 5 v> 1,000 I l @ 'I !N -

                        ~
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5

                        - 20% PRESERVICE lI                       _

x 500 7

w. #

I . _ . MINIMUM lCLTUP TEMPERAFJRE I L ' '' ' ' ' ' ' ' ' ' O 50 100 150 200 250 300 350 400 450 500 INDICATED RCS TEMPERATURE Tc, DEG. F g Tc < 200'F, AP = 77 psia ART M Tc 2 200'F, AP = -69 psia 1/4t = 191.0'F AT = + 6 ' F 3/4t = 137.0'F g &Se o

FIGURE 4.3-2 COMPOSITE APPENDIX G P-T LIMITS COOLDOWN 2.500 LOWE ST SERVICE TEnJPERATURE ~ (19h F) 2.000 E a E g 1,500 5 0 - E

o. -

a:

                 ~

5 E 1,000 ) E _ 8 r -

                  - 20% PRESERVICE                                         /

l

    @                 HYDRO        s, s

E 500

                                            #                      s y                      s N

20 rmR eO.TLINE

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I FIGURE 4.5-3  : COMPOSITE APPENDIX G P-T LIMITS HYDROSTATIC 2.500 ' 6

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APPENDIX B Low Temperature Overpressure Protection Requirements and Relationship to Pressure-Temperature Limits I I Prepared fort II Combustion Engineering Owners Group I by: ABB Combustion Engineering Nuclear Operations I 2000 Day Hill Road Windsor, CT 06C95-0500 I I I I I I I

I I TABLE OF CONTENTS Page

1.0 INTRODUCTION

B-3 1.1 Scope B-3 1.2 Background B-3 I 2.0 GENERIC LTOP REQUIREMENTS 2.1 Limiting Event Determination B-4 B-4 2.2 Analytical Considerations B-5 2.3 LTOP Enable Temperatures B-6 2.4 Typical Operational Restrictions B-7

3.0 REFERENCES

B-8 I I I I I I I "-

I

1.0 INTRODUCTION

1.1 ocops

 .I This report describes the general criteria considered by Combustion Engineering to ensure adequate low temperature overpressure protection (LTOP) for pressurized water reactors. In addition, the relationship between LTOP setpoints and Reactor Coolant System (RCS) pressure-tempe ature (P-T) limits is discussed.

I 1.2 Background Current requirements defined in Section III, Article NB-7000 of the ASME Boiler and Pressure Vessel Code provide for overpressure protection of the reactor coolant pressure boundary during power operation of the reactor. These requirements address protection of the design limits of the Reactor Coolant System. Additional requirements 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 reactor coolant pressure boundary are not exceeded during normal operation including anticipated operational occurrences, and the reactor coolant pressure boundary 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.

':I            Consequently, *he U.S. Nuclear Regulatory Commission (NRC) has provided guidance to ensure overpressure protection for normal I          operation and aiticipated operational occurrences at conditions other than power operation. This guidance, published in NUREG-75/087 (currently NUREG-0800) contains two documents, Standard Review Plan 5.2.2,  "overpressare Protection"(1) and Branch Technical Position RSB 5-2,  "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures"(2)  .

B-3

} I l The primary concern of Branch Technical Position (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 ceasequences resulting from an overpressurization event at low temperatures are conspicuously threatening to the integrity of the reactor coolant pressure boundary. Therefore, BTP RSB 5-2 requires I M protection of the Appendix G limits to meet the criteria established in 10 CFR 50, Appendix A. The LTOP system is required to protect a limit defined by 110% of the pressure which satisfies the stress intensity factor equation defined in Appendix G, paragraph G-2215 of ASME Boiler and Pressure Vessel Code Section XI, Division 1. I 2.0 GENERIC LTOP REQUIREMENTS 2.1 Limiting Event Determination The determination of LTOP-driven restrictions is based upon the I 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 operational occurrences. Anticipated operational occurrences are defined as those conditions of normal operation t.hich are expected to occur one or more times in the life of the nuclear power unit. In addition to the initiating event, the most limiting concurrent single g active component failure must be assumed. 3 Combustion Engineering classifies and separates the initiating events in the LTOP region into two categories, mass addition events and energy addition events. The mass addition events are the result of simultaneous operation of charging and HPSI pumps that inject mass into the RCS. Depending on temperature, the combinations vary, with the most limiting being that resulting from inadvertent safety injection actuaticn. The energy addition event 4:

• reactor coolant pump start with secondary-to-prinury temperature dif ferential. In both 3 B-4

I I instances, the limiting concurrent single active component failure was determined to be loss of pressure relief capability (i.e., a single LTOP relief valve or relief valve train). Combustion Engineering then I analyzes, within each category, a matrix of cases which considers the initiating events combined with the plausible single active component failures to determine their respective peak transient pressures. Upon determination of the peak transient pressures using appropriate analytical methods, a comparison is made 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. If the peak transient pressure exceeds 110% of the applicable P-T limit curve, operational restrictions are imposed to preclude violation of 110% of the applicable Appendix G pressure limit. These operational restrictions are then identified for inclusion in the plant technical specifications as required by BTP RSB 5-2 or in tt. PTLR. I 2.2 Analytical Considerations Current CE system designs incorporate LTOP relief capability during low temperature operation of the Reactor Coolant System. This ir done in one of several ways. LTOP is provided by either two power-operated relief valves (PORV's) on top of the pressurizer, two spring loaded relief valves (generically called relief valves) on top of the pressurizer, relief valves in the Shutdown Cooling System suction, or a combination of the PORVs and shutdown cooling suction relief valves. The methodologies described below have been developed to address LTOP I and are appropriate for the CE designed systems. As described previously, CE's approach to address LTOP is typicsily twofold. Consideration is given first to the identification of the limiting overpressurization events in the RCS which are then analyzed to determine peak transient pressures. Secondly, an LTOP evaluation is performed to compare the peak transient pressures to the applicable P-T limits and to assess required operational restrictions.

B-5

In performir.g pressure transient analyses, significant effort is placed upon determination or justification of plant specific assumptions and appropriate analytical modeling. The assumptions are made based upon review of plant operating procedures and practices. Per BTP RSB 5-2, the most limiting potential plant configurations are assumed in transient analyais. The pressure transient analyses of the limiting pressuriration events are typically performed in a parametric manner. The variables may include relief valve setpoint, initial Rcs pressure and temperature, pressurizer water level and fluid conditions, PORV opening time (where applicable), secondary-to-primary temperature differential (energy addition transient), core decay heat, etc. Such an approach generates a sufficient d4.ta base of the peak transient pressures that allows the identification of the most optimal operating limitations while providing adequate LTOP. Relief valve opening characteristics, PORV activation loop uncertainties (where applicable), PORV inlet' piping pressure losses (where applicable), RCS flow AP differences and elevation differences are all taken into account to determine peak transient pressures in the pressurizer. These peak transient pressur:es are then evaluated along with the applicable Appendix G P-T limits during an LTOP evaluation. By g selecting various operatic.nal limitations and applicable heatup and E cooldown rates, an optimal LTOP system is developed which assures adequate LTOP, i.e., protection of the 110% of the Appendix G pt; essure limit, in combination with an acceptable operating window and reasonable operational limitations. 2.3 LTOP Enable Temperatures The LTOP system must be aligned and capable of mitigating any overpressurization event between the vessel utinimum boltup temperature E and the LTOP enable t emperature. Exceptions to this requirement would i, B-6 s - _ I

be if the RCS was incapable of being pressurized by establishing a sufficient vent area or the vessel bolts are untensioned. h The enable temperatures are determined by the guidance provided in Branch Technical Position RSB 5-2 III' The LTOP_ enable. temperature is at the greater of 200'r or the reactor coolant inlet temperature corresponding to a reaccor-vessel metal temperature less than RTNDT +

                      $0'F. The vessel metal temperature is the temperature at 1/4t at the beltline location. . The resulting enable temperatures are then corrected for instrumentation uncertainty.

h 2.4'= Typical Operational Restrictions In order to maintain an acceptable operating window between t'ne RCP operating curves and relief valve setpcint, operational reattictions may have been imposed to lessen ths aewrity of the event or eliminate some events all together. As the reactor vessel ages, the Appendix G-limits become more restrictive and additional restrictions are placed on operation of the. plant. These operational rests:ictions are identified for inclusion into the Technical Specifications, h Typical restrictions which-are placed on plant operations are list.ed { below. This list is not intended to be comp bte or be applicable to

                    - every plant but is provided ~as an overview of .:ossible restrictions.

Typical LTOP Operational Restrictions:-

                     . l '. RCS heatup and cooldown rates:gre restricted to rates lower than the RCS design, rat.es.

{-: 2.- High pressure safety injection:(HPSI) flow is restricted by locking cut 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.

4. The numbe'r of operating RCPs is limited.

{-( - B-7 [ ,... J

5. Water solid operation is restricted.

I

6. Limitations on RCP starts are specified that may include the I

secondary to primary temperature differential, pressuriter level, and/or initial pressure.

3.0 REFERENCES

Is

1. U.S. Nuclear Regulatory Conunission, Standard Review Plan 5.2.2, I

              " Overpressure Protection," Revision 2, Novernber 1988.

2. Branch Tecnnical Position RSB 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temper;6tures,"

Revision 1, November 1988. I I I I

                                                                                               ~l B-8 I

                . . . - . - - . . - . - - . - ~ .         . - - . - - - - - . - - - - - - . . - . - - . .   - - - - . . . - - . - - - . . ~ . . . . .

i i 4 i 4 Y f j

l 1

I i 1 i 1' I APPENDIX C ) . 1 3 i i , 11 I a f i EXAMPLE OF i 6 h IE RCS PRESSURE AND TEMPERATURE LIMITS REPORT D t i at e k i I 4 lI i

RCS PRESSURE AND TEMPEPATURE L2MITS REPORT FOR (NAMEJ UNIT (KJ I I .able of Contents Page

1.0 INTRODUCTION

C-4 2.0 GL 96-03 PROVISION REQUIREMENTS C-4 2.1 Neutron Fluence Values I 2.2 Reactor Vessel Surveillance Program C-4

                                                                                 .4 2.3    LTOP System Limits                                           C-5

.g 2.3.1 LCO (3.2.2.2J Boration Systems 'W Flow Paths - Shutdown C-6 2.3.2 LCO (3.2.2.3J Reactivity control Systems Charging Pump - Shutdown C-6 2.3.3 LCO (3.3.2.2J Engineered Safety Features Actuation System Instrumentation C-7 2.3.4 LCO (3.4.9.2) Pressure / Temperature Limita - Reactor Coolant System C-7

2. 3. 5 LCO (3. 4.23J Reactor Coolant System Power Operated Relief Valve C-7 2.3.6 LCO (3.5.3J Emergency Core Cooling Systems, I ECCS Subsystems - Tavg < (325'FJ 2.3.7 LCO (3.4.14) Reactor Coolant System C-8 Reactor Coolant Pump - Starting I 2.4 Beltline Material Adjusted Reference Temperature (ART)

C-9 C-9 2.5 Pressure-Temperature Limits using limiting ART in the P-T Curve calculation C-9 I 2.6 Minimum Temperature Raquirements in the P-T curves C-10 2.7 Application of Surveillance Data to ART calculations C-11 C-2 1

RCS PRESSURE AND TEMPERATURE LIM 8TS REPORT FOR (NAMEJ UNIT (A)

3.0 REFERENCES

C-13 4.0 LIST OF FIGURES 4.1 (Naaw) Unit (AJ P/T Limits (2) EFPY g Heatup and Core Critical C-14 W 4.2 (Namne) Unit (AJ P/T Limits (*) EPPY > Cooldown and Inservice Test C-15 4.3 (Name] Unit (AJ P/T Limits (EJ EFPY Maximum Allo'<able Cooldown Rates C-16 4.4 Maximum Allowable Heatup and Cooldown Rates, Single HPSI Pump I Operation C-17 I I I I I I I I I I cs I t_ - I-

RCS PRESSURE AND TEMPERATURE LIMITS REPORT FOR (NAMEJ UNIT (KJ I

1.0 INTRODUCTION

This PTLR for (NAME) Unit (KJ contains Pressure-Temperature IP-T) limits j corresponding to (KJ Effective Full Power Years (EFPY) of operation. In i addition, this report contains Low Temperature Overpressure Protection (LTOF) 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 categorys 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 using the NRC-accepted methodologies as described in Appendix B, i Reference 3.2 (, except as snodified in Reference 3.z] . 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 (NAMEJ Unit (KJ corresponding to (locations on the vessel] for (2) offective full power years (EFPY) is (3.6x20**] neutrons per square centimeter (n/cm*) with en associated uncertainty of i (....). 2.2 Reactor Vessel Surveillance Program The reactor vessel surveillance program and the surveillance capsule withdrawal cre described in Appendix B, Reference 3.2 and Reference 3. (g . . plant specific g details includimg withdrawal schedu.le reference). The reports describing the . E post-irradiation evaluation of the =urveillance capsules are contained in Reference 3.(s . . Post-irradiation era 1uation reference) I . I I C-4

RCS PRESSURE AND TEMPEPATURE -LIMIT 3 hZPORT FOR (NAMEJ UNIT fRJ 2.3 LTOP System Limits The LTOP requirements have been developed by making a comparison between the peak transient pressures and the appropriate Appendix G pt. essure-temperature lint.tt curven. The acceptability criterion regarding each particular transient is that the peak tranalent pressure does not exceed 110% of the applicable Appet. dix G pressure lindt'. These requirements for LTOP have been established . based tn NF.C-accepted methodrilogies and are described in. Appendix B, Reference ' 3.2 and speciflad in thee Bases Section for Technical Specification (A.B.C], Referance 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-apecific I representation of LCOs which are in place at some of the participant 's operating plants. Not all plancs currently have ew:b of these particular LCOs, aepending on the corrplexity of the plant 's current LTOP analysis. Dependirm upon the LTOP analysis at each unit (which is unique), some of these may not be applicable. ) LCO (3.1.2.1) Boration Systems Flow Paths - Shutdown LC) f3.1.2.3J Reactivity Control Systems Charging Pump - Shutdown E LCo (3.3.2.1) Engine 4 red Safety Features Actuation System Instro*ntation LCO f3.d.9.1J Pressure / Temperature Limits - Reactor Coolant Syst.m LCo (3.4.13] Reactor Coolant System Power Operated Relief Valve LCO (3.5.J] Emergency Core Cooling Systems, ECCS Subsystems - Tavg < [325'TJ LCO (3.4.1d] Reactor Coolant Systo.n Reactor Coolant Pump A Starting The LTOP spec.ific requirem nts for each LCO are presnnted in the following subsections. \m I C-5 I

{ RCS PRESSURE AND TEMPERATURE LIMITS REPORT roR (Maas) UNIT (KJ [. 3.3.1 Boration Systems riow Paths - Shutdown ((Eco 3.2.2.2)) 2.3.1.1 The flow path from the RWT to the RCS via a single HPSI pump shall { only be established its [ a. The RCS pressure boundary does not exist, or

b. (NeJ changing pumps are operable and the RCS heatup and

[, cooldown rates shall be limited to those in rigure (3.1-11 h At RCS temperatures below (115'F), 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. High Pressure Header Auxiliary Header (ECV-3626) p.cV-3617J (ECV-3626J (acV-3627J (ncV-3636) (ncY-3637J (EcV-3646) (McV-2647J 2.3.2 Reagivity control sys tems, charging Pumps - Shutdown (Leo ft.1.2.31) ' 2.3.2.1 The flow path from the RWT to the RCS via,a single HPSI pump shall be established only if (~ a. The RCS pressure boundary does not exist, or

b. (NeJ. charging pumps are operable and the RCS heatup and

{ cooldown rates shall be limited to those in rigure (3.2-1). At RCS temperatures below (128'FJ, 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 th O C-6

Acs PRESSURE AND TEMPERATURE LIMI?S RETORT TOR (NANEJ UN37 (K) power supply, or by other trieans to prevent the valves from cpening a c t or'ia t i c a ll y. High Prest;ure Header AuxiliarL Header (HW *3616) (HCV-3617) (HW~3626) (HCY-36.17) (HCV-3636) [HCY-3631) (;tCV-3646) (HcV-3647) 2.3.3 Er2girwered safety Features Actuation System Instrumentation (LCO g (3.3.2.1)> M 2.3.3.1 At (3(;5'r) and less, the required operable HPSI pump shall be in a pt.1-to-lock and will not start automatically 2.3.4 Pressure /Terrperature Limits - Reacter coolant System (LCO f3.4.9.IJ) 2.3.4.1 'the RCS (except the pressurizer) tempterature and pressure shall be limited in accordance with the lirait lines shown on rigures(3.4.S-2) g and (3.4.9-2) during heatup, coaldown, criticalit/, and inservico N jeak and hydrostativ testing 2.3.5 Reactm conlant System Power operated Relief italves ((Leo 3.4.23))_ I 2.3.5.1 The setpoints for the pcwer operated relief valveu shall be as folless. - l a. A s6troint of less than or equal to (350 ps.taJ shall be selected: 1

                            .1.         During cooldown when the temperature of any RCS cold leg is less than or equa4 to (225'rJ cnd l

I-l C-7 } , II

I E RCS PRESSURE AND TEMPERATURE LIMITS REPORT TOR (NDE) UNIT (KJ I 2. During heatup and isothermal conditions when the tenperature of any RCS cold le; is less than or equal to (193 *TJ . I b. A setpoint of less than or equal to (530 pata) shall be selected: I

1. During cooldown when the temperature of any P,CS cold leg is greater than (215'r) and less than or equal to the LTOP Enable Temperature for cooldown.

2. During heatup and isothermal conditions when the temperature of any RCS cold leg is greater than or equal 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'rJ.

b. The LTOP Enable Temperature f or er.cidown is (281 'r) .

3.3.6 Emergency core cooling systems, Ecc5 'tavo < (325'r] (ftco 3.5.JJi I 8 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. 3 3.6.2 Prior to decreasing the reactor coolant system temperature below (236*TJ, all high pressure safety injection pumps shall be disabled and their associated header stop valves closed except as allowed by Specifications (J. J.2. 2 and J. A.2. JJ . I I C-8 a

RCS PRESSURE AND TDiPERATURE LIMITS REPORT POR (XANCJ UNIT (KJ 2.3.7 Reactor toolant System, Reactor Coolant Pump - Starting ((LCO J.d.2d)) 2.3.7.1 If the steam generator temperature exceeds the primary temperature by more than (30'r), no idle ret.ctor coolent pump shall be started. 2.4 Beltline Material Adjusted Reference Temperature (ART) I The calculation of the adjusted reference temperature (ART) for the beltline region has been performed using the NRC-accepted methodologies as described in Appendix B, Reference 3.2. Provision 7 (section 2.7, Application of surveillance Data) (vas/was not) used to refine the chemistry factor and the margin term. The limiting ART values in the beltline region for the (NAMEJ Unit (K) corresponding to (z) Effective rull Power Years (EPPY) for the 1/4t and 3/4t locations are Location ART Material 1/4t (xxx TJ [ . . . Limiting Piste or Weld Haterial Identification ... J 3/4t (xxx TJ [ . . . Limiting Piste or Weld Material Identifscation ... ] The RTm value for (MANr) Unit (KJ which is calculated in accordance with 10 CFR 50.61 is (xxx TJ which corresponds to (Limiting Place or Weld IdentifierJ. Provision 7 (Section 2.7, Application of surveillance Data) (was/vas not) used to refine the chemistry factor and the margin term. 2.5 Pressure-Temperature Limits using limiting ART in the P-T Curve calculation The limits for LCO J.4.9.1 are presented in t5e subsection that follows. The analytical methods used to develop the RCS pressure-temperature limits are based on NRC-accepted methodologies and discussed in Appendix A. The methodology is also m.:umented in the Bases for Technical Specification (A.M.CJ. C-9

RCS PRES $URE AND TD4PEPATURE LIMITS REPORT TOR (MANEJ UNIT (K) l The RCS PRESSURE-TD4PEPATURE LIHits REPORT will be updated prior to exceeding I the RTHDT utilized to develop the current heatup and cooldown curves. The RCS PRESSURE-TEMPEPATURE LIMITS REPORT, including any revisions or supplements l thereto, shall be provided, upon issuance of new heatup and cooldown curves to the NRC Document Control Desk with copies to the kegional Administrator and 1 Resident Inspector. 2.5.1 RCs Pressure and Temperature (P/T) Eimits ((Leo 3. 4. 9. 21)_ 2.5.1.1 The RCS temperature rate-of-change limits are 4

a. A maximum heatup of (50)T in any 1-hour period. .

I

b. A maximum cooldown rate consistent with rigure (2.1-3).

c. A maximum temperature change of 5 5'r 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 testing, and criticality are specified by rigures (2. 2-2J, (2. 2-2) and (2.1-3). 2.6 Minimum Temperature Requirements in the P-T curves The. minimum temperature requirements specified in Appendix G to 10 CTR 50 are cpplied to the P/T curves using the NRC-accepted methodologies as described in Appendix B, Reference 3.2. The minimum temperature values applied to the P/T curves for (MAMr) Unit (K) corresponding to (2) Effective Full Power Years (ETPY) are Location Min Temperature BoltUp . (acxx TJ Hydrotest (macr TJ 4 e C-10 1

RCS PRE 55UhE AND TEMPEPATURE LZHITS REPORT FOR (xnNrJ UNIT (KJ 2.7 Application of Surveillance Data to ART calculations Post-irradiation surveillance capsule test results for (NAMEJ Unit (KJ are given in (Reference 3.s) . The test results (do/do notJ meet the credibility criteria of Regulatory Guide 1.99 Revision 2. (no criteria were seet as follove t a) the surveillance program piste or veld dmplicates the contro11 Lag reactor vessel belt 1Lne material in texans of ART / b) Charpy data scatter doas g1 wase ambiguity in the aetezmination of the 30 ft-lb shift l

                   .,    t .    ...x.d e                 .r. ....L.t.nt                       ,r.d .t.d . - C.,
                                                     .                                      t g

d) the capsule irradiation temperature is coeparable to that of the vesself and e) correlation monitor data (are/are not) available and are consistent with the known data for that material.

                                          *                )
                                                     .\*

ff ne data sypporting the credihtlity analysts are presented in (reference).J (In the case where sister vessel auxveillance data are available for use, the preceding should be sqppleenented. .. s as Apdicated under Provision 7 (section 3.2.4) of Appendix A, Reference. 3.2.'s 'The appplenmental infozmation should address ditforences between the two sister plants in tezans of irradiation environment and estabitsk the app 1Loabk1Lty of the data.J The credible surveillance data (were/were notJ used to refine the chemistry factor and the margin term. (ne process for applying the credible surveillance data is described under Provision 7 (Section 3.2.4, &mA4= A, Reference 3.2) in the Methodology and follove that prescribed in Position 2.1 of Begulatory Guide 1.99, Bevision 2. De data used and the calculations perfozamed are given belovt C-11

RCS PRESSURE AND TEMPEPATURE LIMITS REPORT FOR (NANrJ UNIT (KJ ~ R~ port Capsule TD T1uence Chif t T1uence Tactor, f_ (f) E (f x shif t) L Rafined Cheekstzy Tactor e CT(R)

• E(f x shift)

E(f)# og n (17 or 28) T Astined as n (17 or 28) T /2 ~ l

~

Refined ART = Znitial Rtadt + CT(R) x i + 2( ) + 0,# ) W f h C-12

RCS PRESSURE AND TEMPERATURE LIH2TS REPORT FOR (70LNEJ UNIT (KJ I l

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 01, " Development of a acS Pressure and Temperature. Limits Report for the Removal of P-T Limits and LTOP requirements from the Technical Specifications, CEOG Task 942", January 1997. 3.3 Each spec A.B.C for (Itamme) tinkt (X) ... (3. q ZT not in Tech spea . . . Reference for Plant specific surretilance capsule withdravs1 Schedule ] (3.s Reference for post-triadiation erstuation of surveL11ance capsules ] (3.s Reference for Tiuence value saodification 1 I l I I I C-13

                                    . - - - _ - _ _ _ - _ _ - - - - - - - - - - = - - -                                                                                                                                           -                               -

FIGURE 41 lNAME] UNIT [A) P/T UMITS. [ j EFPY HEATUP AND CORE CRITICAL s00 ..-n.-. 3a-s 3._-:- -7.tv: : ~7 . .. m- [

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     'INAME) UNIT (Al                                                                                                                                                                                                                                           AMENDMENT NO. [Y1 C-14 I                                                                                                                                                                                                                                                m . .                                                                     . _

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E ) FIGURE 4 3 INAMEl UNIT lAl P/T LIMITS, I IEFPY MAXIMUM ALLOWABLE COOLDOWN RATES I  ! l l ' ion - - l l RATF. DtO FMR TiMD (IMIT, Of.O F i 20 airl ' _ 30 61 125 146  !  ! 145166 60 l/ 16f..ies 7 s .. , .. / too esos f l 1 { l / h" ' I _ / l I d 8 I

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I ao too iso i4o iso Tc INDICATED REACTOR COOLANT TEMPERATURE, DEG F no no I NOTE. A MAXIMUM COOLDOWN RATE OF 100 DEG FMR IS ALLOWED AT ANY TEMPERATURE ABOVE 195 DEG. F I k i _ - ,. _ _ ..,,, i t c.,.

I FIGURE 4 4 MAXIMUM ALLOWABLE HEATUP AND COOLDOWN RATES, SINGLE HPSI PUMP IN OPERATION I

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I I I I I I (NAME) UNIT IAl AMENDMEN T NO. [Y) I C-17 l

[ [ E E r L E APPENDIX D [ [ [ EXAMPLE OF MODIFIED { TECHNICAL SPECIFICATIONS [ [ Appendix D has not been modified from Rev 00 of this report [ [ [ o.1 c . . . . _ _ . . .. .. .. .

_ _ . _ _ . _ _ _ _ . _ _ - - - - - - - - - - ' - - ~ ~ ~ ~ l L F L INOf% F L CEP !N tit o.15 r' I $!? TION t l NOE 1.23 P ro c e s s C e n t ro l P ro g r a m ( P C ? ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . l e 13 [ 1. 2 f. Furge Purging............................................. 1-5 1.25 Ra d T tral ow r . .. . 10hD~

                                                                                                                     ....................               16 l J!                       ,CS hr.15etrC*E1" mfara /C L/S tY.! S ydth . . .
                          .?                                                                                              ...     .....

R ea ciof i r i p TyTT.efii Res pon s e i me .T. C. . . . . . . . . . . . . . . . . . . 1-6 ..

                                         '7 1.$'PR e po r t a b l e E v e n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6

1. (h Sh i e l d Su l l d i ng I n t e g ri ty . . . . . . . . . . . . . . . . . . . . . . . .'. . . . . . . .1.-6. .

9 ' I 29 Shutdewn .".argin............................................. 1-6 Jo

1. ite 60undary................................................ l*6

[- i Q S,o uce neck................................................ 1-6

1. @ S t.a g g e r e d T e s t Ba s i s . . . . . . . . . . . . . . . . . . '. . . . . . . . . 1.- .7 . . . . . . . . . . .

[- i .ew .3ther.aier................................................ rc i-7 { 1. h nidentified Leakage.............'........................... 1-7 1.hUnrestrictedArta.........'.................................1-7

                                       % l.

[ l. Unrodded Integrated Radial Feaking Factor - Fr . . . . . I-I I. UP. rodded Planar Radial Peaking Facter - F

                                      %                                                                              xy..................             I-7

[. , . [ - (NA*E.) UNIT ( A) la NET W. WI 0-2

CEFU4(TIONS 7

        ,    IDENTIF tED LEAKAGE 1.15      10!NTIFIE0 l.EAXAGE shall be:

a. I Leakage (except CONTROLLEO LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or 'E b. Leakage into the containment atmosphere from sources that are both 3 specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDAM LEAXAGE, or

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

LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE

         .cond
            .l .16.t.i,3a,,

The LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE is_tha , erating

          < 2 l'F du                      cold leg temperature is < 04 F ~during heatup ce                   l ce do      and (2) the Reactor Coo ent ys e.

7, in, gri y, . p ess re bocndtry I/ ace when the 3eactor or Coolant System does not have pressure boundary integrity Coolant System is open .to containment and the minimup area df J N4 A the Reactor Coolant System o enino is great th n (": I NEMSE . rIU3weiclhsn-r it 1%c rcnecJok m,rwbe o t.ok h1709 7 o e

                                                         &a 4 hf rr bou tMaMid M At- ACS
                                                   /epor,',

ch 1.17 MEMBER (S) 07 THE PUBLIC shall include all persons who are not occupation-ally associated wtth the plant. This category don not include employees of 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-t.ional, occupational or other purposes not associated with the plant. OFFSITE DOSE CALCULATION MANUAL {0DCN) 1.18 The OFFSITE DOSE CALCULATI0li MANUAL shall contain the current methodology and parameters used in the calculations of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent I, monitoring alarm / trip setpoints and shall include the Radiological Environ-mental Sarc.ple point locaticas. I - (tu+El UNIT ( A) 1.a A.c c e T to. (Y) - ( l 0-3 'I -

OEFINITIONS RATED THER."At POWER - L I 1.25 RATED THERML POWER shall te a total reactor core beat transfer rate to the reactor coolant of 2700 Midt. - REACTOR TRIP SYSTEM RESPONSE TIME l E 7 -

1. The REACTOR TRIP SYSTEN RESPONSE TDtE shall be the tirne interval from whe the rnonitored parameter exceeds its trip setpoint at the channel sensor a until electrical power is interrupted to the CEA drive mechanism. g REPORTABLE EVENT .

78 '

1. A REPORTABLE EVEKT shall be any of those conditions specif fed in Section 50 J to 10 CFR Part 50. '

SHIELO BUILDING IKTEGRITY g l u,

         .-              1.h SHIELD BUILDING !KTEGRITY shall exist when;

(

a. Each door is closed except when the access opening is being used for normal transit entry and exit;

b. The shield building ventilation system is in compliance with Specification 3.6.6.1, and -

f'

c. The sealing mechanism associated with each penetration (e.g.,

welds, bellows or 0-rings) is OPERABLE. EftrTOOWN MRGIN

                  ~

1 S DOWN MRGIN shaTl Se the instantaneous amount of reactivity by which the reactor is soberitical or would be subcritical from its present condition h assuming all full-length contrul element assemblies (shutdown and regulating) e 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 lan(is neither owned, leased, nor otherwise controlled by the licensee. -

( ' SOURCE CHECX ,,' 1.@nA SOURCE CHECK shall be the qualitative assessment of chan a when the channel sensor is exposed to a radioactive source. g l ' (NAME] UNIT (A) , AMDO!D(T(C. 0-4 i u

E INsrRT A L. PCs PPls SUPI-TIMPIPATUPI LIMITS PIPORT 1.26 The RCS PRE 55URI-TIKPEMTURE LIMITS P.EPORT (PTLR) is a fluence dependent re po rt providing Limiting condition of operativne for heatup, cooldown, inservice hydrostatic and leak testing, and cos e criticality 11mits in the for a of Pressure-Temperature (p-T) 11mits to ensure prevention of brittle fracture. In addition, this report establishes Limiting Conditions of Operations which provide Low Temperature overpressure Protection (LTOP) to assure the P-T limits are not exceeded during the most ilmiting LTOP event. The P-T 11mits 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. I I-I I . I I ( .

                                                                                    ~

DEFINITIONS STac;E?.E0 TEST BA5!5

1. A STAGGEREO TEST !A5!5 shall tensis: of:

a. A test schedule for n systems, subsystems, trains obother designated compcnents obtained by dividing the specified test interval into n eRual subintervals, and

b. The testing of one system, subsystem, train or other designated component at the beginning of each subintervel.

THE2.'ul POWER 1.@74THEPJ4AL POWER shall be the total reactor core heat the reactor coolant. l tr UNICENTIFIED LEAXAGE

1. Vi!DENTIFIED LEAKAGE shall be all leakage which is not 10!NTIFIED iLEA GE or CONTROLLED LEARAGE. l Ur,8.ESTRICTED AREA

1. hrioAn UNRESTRICTED AREA shall be any area at or beyond the SITE SO access to which is not controlled by the licensee for purposes of protaction l of individuals from exposur: to radiation and radioactive materials, or any area within the SITE B0UNDARY used for residential quarters or for inc'.strial, cer: .ercial, institutional, and/or recreational purposes.

UNR000E0 INTEGRATED RA0!AL PEAK!!!G FA* TOR - F,. , I 1.[The UNR000E0 INTEGRATIO RADIAL PEAKING FACTOR is the ratio of pin power to' the average pin power in an unrodded core, excluding tilt. l UNRODDED PLANAR RADIAL PEAK!NG FACTOR - F,y

    -        1.h7 T e UNR000E0 PLANAR RADIAL PEAKING FACTOR is the l maxi peak to average power density of the individual fuel rods in any of the unrodded horizontal planes, excluding tilt.

1 . 9 0 0.e L I

                                     . _ _ - _ _ - _ - - - - - - - - - - - - - - - - - - -                                        ~

{ L

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                              . A f e 5 r 3 7!t15 3 / a .1,2 BORAT10N SYSTEMS Ft,0W PATHS - SHUTOOWN                                                                                                                   ('

F L L(MITING CONDITION FOR OPERATION 3.1.2.1 [ As a minimum, one of the following boron OPEP.A3LE and capable of being powered from an OPERACLE emergency power s , s. p A flow path from the boric acid makeup tank via eith'r a L boric acid pump or a gravity feed connection and any charging pump to the Reattor 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 injection pum@gjg to the Reactor Coolant System if only the refu .lin water { tank in Specification 3.1.2.7b is OPEP.ABLE. Ab APPLICABILITY: MODES 5 and 6. b A CT I 0t),: { CORE ALTERATIONS or positive reactivity changes until at lea . path is res tored to OPERABLE status. g-

                                                                                                                                                    \;.

SURVEILLANCE REQUIREMENTS I 4.1.2.1 L OPERADLE:At least one of the above required flow paths shall be demonstrated p a. L At least once per 31 days by verifying that each valve (manual . power operated or automatic) in the flow path that is not locked,position. correct sealed, or otherwise secured in position, is in its { ~

           ~ ' be T.he      flow path established        if:                                from the RVT to the RCS via a single HPSI pump s

{ (b) no charging pumps Are(a) the RCS pressure boundary does not exist, or operable. In this case all charging pumps accordance with Fig. 3.1-lb.shall be disabled, and heatup and cooldown rat b At RCS temperatures below 115'F, any two of the following valves i n the (; f operable HPSI header shall be verified closed and have their power re High Pressure Header Auxiliary Header F HCV-3616 .. HCV-3626 HCV-3617 L /&rrnov( 7D/ PTLE HCV-3636 HCV-3627 (~ HCV-3646 HCV-3637 HCV-3647 i (NAMEl UNIT (A) 3/4 I8 N N'). (Y) 07 r -

  - . . - - . . ~ - - - - - -                 . - -     .- -      .- - ----            ~.       .--               -

d 1 i i i INSERT 5 5 i The flow path from the RWT to the RC5 via a single HPs! pump shall only se established if the requirements in the PTLA are mett I I: e e

      'O                      gm 0t # -                         p         4               g e

O e I l l l o.e g t

r ' L. P 100 - f 80 - r * - I tk w 60 - Heatup ~ 40 - l O i , , C W own

• 80 ' '

I 100 120 140 th 18O 200 220 , Tc . If40+CATED RE ACTOR COOL ANT TEWR ATM y l I f . FM 3.1 1b

                                   "XM ALLOWAstg Hi!ATUP AM COOLDOWN RATES                                                                           *
                                                                                                ,                                                  n E MS PU8dP IN OPER ATION I                                                               ~
                                            ~

fC110/C hO me 4 e

'        ( M ) UNIT (A)                                   3/4 1 9a AMENDMENT NC. [Yl 0-9 l                                                                                                     . . . . .        .   .

i lREACTIV!TYCONTROLSYSTEMS I CHARGING PUMP - SHUT 00WN t.lMITING CONDITION FOR OpfRATION __ g be sc/c, ~ 3J.2_.3 At least one charging pump or one high pressure sa fety injection g

        -                        pum;P,)in the boron injection flow path required OPERABLE pursuant to Speciff-                     g cation 3.1.2.1 shall be OHRAB E and capable of being powered from an OPERABLE emergency bus.                                                                             3 Juse dr@

Appl ICA8 ILITY,: NODES S and 6. ACTION : l j With no charg619 pump or high pressure safety injection pump *0PERABLE, suspend f all operatbns involving CORE ALTERAi!ONS or positive reactivity changes until as iend one of the required pumps is restored to OPERABLE status. l SURVEILLAPCE REOUIRENENTS E g 4.1.2.3 At least oct of the above required pumps shall be demonstrated OPERAELE by verifying the charging pur!p develops a flow rate of greater than or equal to 40 gpm or the high pressure safety injection pump develops a total head of4.0.5 Specification gree,ter than or equal to 2571 f t, when tested pursuant to g g

  ...             - . . - p
                            'The flow path from the RWT to thb RCS via a single HpS! pump shall be 1 established oniy if: (a) the RCS pressure boundary does not exist, or (b) no charging pumps are ephrable, in this case, all charging pt.Lps shall be, disabled and heatup and cooldown rates shall be limitut in                       ,

accordance with Fig. 3.1-lb.

        , fpnovi At RCS temper.atures below ll5'F o,1y two of the follcwing valves in the                     ;

TD pr~td operable HPS! header shall be verified closed and have their power rernoved; High pressure Header Auxiliary Hea' der HCV-3616 HCV-3617 i HCV-3626 HCV-3627 , l  :'cV- 3636 HCV-3637 - HCY.3646 HCV'3647 (NAMI) UNIT (Aj 341-12 ,c, gy) l' 0-10 (

L L INSERT 3 The flow path from the RWT to the RCS via a sir.gle hP$I pump shall only be establianand if the requiremento in the PTLR are toet. .I I .. I I ~ I . I. g I - I e 6 0-11

lEACTOR CCOL AttT SYSTEM 3/a,a.9 PRESSURE /TEMPE?.ATURE L IMITS

               *! ACTOR COOLANT SYSTEM LIMIT!t1G C0t10!T!0tl FOR OPERATICf4 surizerl temperature and pressure 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 ,

leak and hydrostatic testing. APPLICABILITY: AtalltimesO C # ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perfom an analysis to detemi,e the effects of the out-of-limit condition on the fracture

oughness properties of the Reactor Coolant System; detemine that the Reactor Coolant System remains acceptable for continued operations or be '

I In lat to_ e sleast s th - HOT 2 0*F STANOBY within the wi in .the' following next in 30' hours 6 accordance hours and wfreduce th' the RCS T**9 3

   ..        Fleures 3.4- b and 3.4-3                                                                        g L e o .3. 4. 9. I I

fen 10 N lo ff2 Q ' then the flow path from thh RWT to the RCS via a . single HPSI pump is

               , established per 3.1.2.3, the heatup and cooldown rates shall be established in accordance with Fig. 3.1-16.

10uring hydrostatic testing operations above system design pressure, a maximum temperature change in any one hour period shall be limited to 5'F. 3 l

             -> - mi                                    3,g,1                                   .m           I 0-12 l

[ INSERT C [ 3.4.9.1

                                               ;he corn.bination of RCS pressure, Rc3 temperature and Res hestup and cooldown rates shall be maintained within the limits specified in the Res PRIS$URE-TIMPIAATURI LIMITS REPORT.

{ t [ i [  ! S [ [ - . . . _ . . [ . . O 4 0 em 6 0-13 I,

REACTOR COOLANT SYSTEM I SURVElllANCE REOUIREMENTS I 4.4.9.1

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

g

b. The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minuter 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 as its o

ns shall be used to update Figures 3.4-2a. l

                                                                                                 .. r 5%e /C.5 #aa.ure Theyr/cb&

l.dnll.s fcpoe/. I I

    ~

I g (I (tWE) UNIT (Al 3/4 4-22 AMDOENT to. (Y) 0-14 t

L_______--_-_-__--__-_--- [P710V[ 70 A& I i;Gh?: 3.4 22 IM) UNIT (AJ P/T LD!!T3. [ j E77y HTATUP AND COP 2 C?lTICM,, . 1 2$00. tSOTHE.RW A L -- 1

                                                              -TO sov/m r_
                                                                                 ~_

M

• 4 2000 '

i- - I. ' g 1500 ,. 5 LOWEST - 0 55 sr.nycz -- E & ': " EE TE6aPtR ATLJRE 33 1001 -

                         ~                                                                   -
                                                                                                 ' CORE Cpt!TK: ALE o              ; __

g 1000 ' j

                        - __ ISOTHrfted AL _

I f- - i

             @                                         _. ; . e-c.,,,,,,                            ._

_f ' ALL,0WA ELE HE ATW R AT ES 3,,

                                               - w --                             __

I, - E sevim :.=; _-- aATE. r/HR TEw. twarT, .r . 50 AT ALL . N DOLTUP TdiP. sov ,' o-.--= - m---- W 200 300 ' 400 300 T C

• INDIC ATED RCS TEldeta ATIJRg, p

                                                                                         ^_

(NAMEl UNIT ( A) 3/4 4 234 g,g g g , (7y 0-15

                                                                                        ..~.           __

i!G2.I 3. 4- 2 b (.M } LHIT ( A) P/T LD:I o, i l ETPY ' ' 'N CCOLDDWN AND INSIRVIC'I TEST , 2500-(

                                                                                                                                                                                                      - v ., _                         -- .:: -                 I

---- _; 1

                                                                                                #tSERVCE -                                                           l            .'

i HYDROSTATC _ u

                                                                                                                                                                    !           i                                =                                              '
                                                                                                                                                                                                             ..- c TEST                                                               -            .-                             -

:: .- .=

2000 -

                                                                                                                                                                                                                                          ._ m                  l 4                                       .                                                                                             .-                                       . -:.= -

B a.

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                                                  - =                                                                                                 l ; -l                                                             .

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                                                ..                              .                                                                            r-- :                                         .n ..a . --

m -- . . _. - , - 1oog/HR 70.=.

             ,           15001 _._J LOWEST                                                                                                    !              !                     ISOTHERM                                AL E-a:            -
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         ,          w                                                                                          .                           .              .-

N -----' TMMAT LAtf. T-

                                                                                                                                                                                                                               ~~

190T ' ' _ := - --

2. -

90 L-~.  :  !=

                                                         = - -                                                :

l' T.- _-- = _,__.-g .

                                                                                                                                                                                                                        .. ~ .. -

1000 -

                   >=
                                                                                                                                                                ._.-~4                                            ,,

ISOTMutMAL- = m - L :-, -

  .   ..        e .

G, ~ :.. - .::.,,,, 500',_ ~ - - --

                                                                              '        w                                                                                                                                                                  '
                                           .                                    9%--sk_ _ _ _ .
                                                                        # j y s s ,.-

[;,MY/M " _- f M T/>#f"' p # _-

449,y(W=.. m y,

ygeg-n m ..-_: e

                                                                                                                                                 -,,=

. t001/M E [-

M M NM Th. $@Y l I 0,~ e. 4 .

   .                             0                                         100                               200                                   300                                           400'                                       500                 ,

q T e m ito E s T m A M T

                                                                                                #'                   -"V'M

('_ (M) MT (A) l 3/4 4-23b AMDO ENTto. (?) nwww ro me . 0-16

t FIGl?2 3,.4-3

                                 /

[ (NM2] UNIT ( A l . ( l EJ'?Y- . MAXIMUM ALLOC32 CODLIO.N MG ' s

                                                                                                                     <                       w                                                                \.

too <

                                                   .        RATE. "F/HR                                                        TEMP. LIMIT. OF

( 20 <125 * ~ 30 125 146 - 80 40 145 146 (

                                               . . .                                           50                                         166 186 i

75 186 196 3 100 g ... > 196 O ~ ~ . . . 00 w'

                       ~                                       .
                                                                                                                                                                       .f=.
                                                                                                                                                                                                      ~

j / . . . . {

                                                                                                                                           . . .g *                    -

, m . _. . .. .. ( 20 #

   .                                                                                                                                                            . . ~ . . . - - .                                              .

0 ' ' So 100 120 l 140 160 180 200 Te . INOCATED REACTOR COOLANT TEMPERATURE. 'F ' NOTE: g(p10 Vd~ IO A MAXIMUM COOLDOWN RATE OF 100*F/HR 13 ALLOWED AT ANY PYL2 TEMPERATURE A80VE 196*F M] UNIT Sl f* ~ C AteDerr NO. .(Yi 5 D-17 e )

AEACT0*4 COOLANT SYSTEM 70' DER OPERATED RELIEF VALVES ' LIMITING CONDITION FOR OPERATION 3.a.13 Two power operated relief valves (PM'isi s6all be 0? wit setpoints selected to the low temperature r. ode of operation (@ERABLc D)h :heirL *n ceco

                                                                                        .~     ~             n qe.spe e          .:

a. A setpoint of less than or equal to 350 psia shall 7 'ected: ,me-1 During cooldown when the tempera ture of any RCS :cid icg is less than or eque.1 to 215'F and l

2. During heatup and isothermal conditions when the temperature nf any RCS cold leg is less than or equal to 193*P.

l

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

1 During coohfown when the temperature of any RCS cold leg is greater than 215'F and less than or equal to 281*F.

2. Duritig heatup and isothermal conditions when the temperature s of any RCS cold leg is greater chan or equal to if3*F and E!mOV[70 APPLICABILITY: MODES 4' and S*. pyc.g 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

2. Another action statement requires cooldown, or

3. Plant and personnel safety requires cooldown to Co.d 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 REQUIREMENTS 1.4.13 The PORVs sha" be verified OPERABLE by:

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

b. 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. Perfonnance of a CHANNEL CALIBRATI0r: of the pressurizer pressure

           >                  sensing channels once per 18 months.

kn N "or P /*#'dT##E Ench te Ge m wre .ssecMrd ah E. t c1 R essw e. % petarbre iReactor Coolant System cold leg temperature bel w 30 *F LMJ #</or /.

                   *PORVs are not required below 140*F when RCS does not have pressure boundary integrity.                                       -

[ [NAME] UNI (Al 3/4 4-59 NN *

• lYI 0-18 t

\ . . PjACTC2 C00LAt4T PUMP - STARTit4G ' P L IMITitiG CON 0lTI0t4 FOR OPERAT10t1 l3 j, E

        .14 I core than s team generator temperature exceeds the primary temperature by ted, the    first l-he     idle reactor mggniMe        .sp:cined      coolent   id nepump   shall 7?mwar la Awar-   no't be star,we imohL 0,w l (N W . g l APPLICABILITY:            MODES 4 and 5.          .

I ACTI0ti: l If a reactor coolant pump is started when the steam generator temperature exceeds primary temperature by more tha 30 evaluate the subsequent transient to determine compliance with Specification 3.4.9.1. l Ge fepor/ ach,r,2/).ib sc .sp a.tes ,;, a c < a ne w a- s ers k a i,w h SURVEILLANCE REQUIREMEtlTS 4.4.14 Prior to starting a reactor coolant pump, verify that the steam generator temperature does not exceed primary temperature by more than 30*F.

                                                                                                                                       ~

hi or egast lo Ik LfoPEml/s 7Emwnbre

                                                                            .speciSed ih Me der Pa.ssile -                    '

bperslure Luh,%s dep o d (P 72 2). (%

  # Reactor Coolant Syst;em Cold Leg Temperature is less than~

o M l UNIT (Al 3/4 4-60 - em go, [y] 0-19 I

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T < 325'? LIMITING CONDITION FOR OPERATION 3.S.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. In MODES 3* and 4 , one ECCS subsystem composed 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 nWM@ to the containment sump on a sump recirculation actuation signal.

b. Prior to decreasing the reactor ccolant system temperature below 270'F a maximum of only one high pressure safety injection pump W g shall be OPERABLE with its associated header stop valve open. E

c. Prior to decreasing the reactor coolant system temperature below 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.

g - APPLICABILITY: MODES 3* and 4 ACTION:

a. # '

                                                                                                     ~

With no ECCS subsystems OPERABLE in. MODES 3* and 4 , imediately restore one ECCS subsystem to OPERABLE status or be in COLD SHUT 00VN ff#WE wm/ within 20 hours. Ltd( && b . With RCS temperature below 270'F and with more than the allowed high pressure safety injection pump OPERABLE or injection valves and header isolation valves open, imediately 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 subs..itted to the Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

SURVEILLANCE REQUIREMENTS 4.S.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2. 4.S.3.2 The high pressure safety injection pumps shall be verified inoperable MM and th asso iated header stop valves closed prior to decreasing below the bv specirie Reactor Coolant System temperaturegnd once per month when the Reactor oo ant System is at refueling temperaturen bb

                                                                             .5)%cihk;d Nr fhe}0A
                'With pr'.ssurizer pressure < 1750 psia.                                                   1
                # REACTOR COOLANT SYSTEM cold leg temperature above 250*F.                                  '

(NAMEl UNIT (A) 3/4 S-7. - g), [yj 0 20 I

INSERT 3 b. Additional operability require.ments for high pressure safety injectien PCP8 are Provided in the RC5 Pressure-Tex.peraturo I,1.mits Report and shal; I be adhered to. I i 4 i l l I l 'I . I - . . . . _ . g . I I W 9 I . 0-21

                                                                      . . . . . . . . _ . . . . _ . _ _ . ,                   ,]

I

                                                                         *NSIR7 I

b. If the req.:irements cf the RCS Pre s sure-Te.mperature 7.i=it s Report have no:

been satisfied, initiate a:: ion to provide immediate compliance with th e requirement s . I I I I I I

   .                                                                                                   I e

0 I D-22 I1! I}}