ML053200206

Attachment 12, EDO-33075, Class 1, Safety Analysis Report for Hope Creek Constant Pressure Power Uprate.

ML053200206
Person / Time
Site:	Hope Creek
Issue date:	11/30/2005
From:	Curran L, Hayes R, Schrull E General Electric Co
To:	Office of Nuclear Reactor Regulation
References
DRF 0000-0006-0455, FOIA/PA-2010-0209, LCR H05-01, LR-N05-0258 NEDO-33076
Download: ML053200206 (251)

Text

Attachment 12 LR-N05-0258 LCR H05-01 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REQUEST FOR LICENSE AMENDMENT EXTENDED POWER UPRATE NEDO-33076 "Safety Analysis Report for Hope Creek Constant Pressure Power Uprate" November 2005

GE Energy, Nuclear 3901 CastleHayne Road NEDO-33076 Wilmington, NC 28401 Class I DRF 0000-0006-0455 November 2005 SAFETY ANALYSIS REPORT FOR HOPE CREEK CONSTANT PRESSURE POWER UPRATE Prepared by: R. L. Hayes Approved by: E. D. Schrull, Project Manager General Electric Company Approved by: L. Curran, Project Manager PSEG Nuclear LLC

NEDO-33076 NON-PROPRIETARY INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33076P, Revision 1, and as such, has the proprietary information removed. An open and closed bracket as shown indicates the portions of the document that have been removed (( )).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between PSEG Nuclear LLC (PSEG) and GE, Hope Creek MELLLA/EPU Engineering Services Contract, dated May 10, 2002, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than PSEG, or for any purpose other than that for which it is intended, is not authorized; and, with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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NEDO-33076 Table of Contents EXECUTIVE

SUMMARY

.................. xi

1. INTRODUCTION . . .1-1 1.1. REPORT APPROACH .. 1-1 1.1.1 Generic Assessments .1-1 1.1.2 Plant Specific Evaluation .1-2 1.1.3 Report Generation and Review Process .1-2 1.2. PURPOSE AND APPROACH .. 1-4 1.2.1 Uprate Analysis Basis. 1-4 1.2.2 Computer Codes .1-4 1.2.3 Approach .1-5 1.2.4 Concurrent Changes Unrelated to CPPU .1-7 1.3 CPPU PLANT OPERATING CONDITIONS . .1-7 1.3.1 Reactor Heat Balance .1-7 1.3.2 Reactor Performance Improvement Features .1-8 1.4

SUMMARY

AND CONCLUSIONS .. 1-8

2. REACTOR CORE AND FUEL PERFORMANCE .2-1 2.1 FUEL DESIGN AND OPERATION .2-1 2.2 THERMAL LIMITS ASSESSMENT . .2-2 2.2.1 Minimum Critical Power Ratio (MCPR) Operating Limit .2-2 2.2.2 MAPLHGR and Maximum LHGR Operating Limits .2-2 2.3 REACTIVITY CHARACTERISTICS . .2-2 2.4 STABILITY .. 2-3 2.4.1 BSP Evaluation .2-3 2.4.2 Option III Evaluation .2-3 2.5 REACTIVITY CONTROL .. 2-4 2.5.1 Control Rod Scram .2-5 2.5.2 Control Rod Drive Positioning and Cooling .2-5 2.5.3 Control Rod Drive Integrity Assessment .2-7

3. REACTOR COOLANT AND CONNECTED SYSTEMS .3-1 3.1 NUCLEAR SYSTEM PRESSURE RELIEF/OVERPRESSURE PROTECTION ... 3-1 iii

NEDO-33076 3.2 REACTOR VESSEL .. ........................................ 3-3 3.2.1 Fracture Toughness .......................................... 3-3 3.2.2 Reactor Vessel Structural Evaluation .......................................... 3-4 3.3 REACTOR INTERNALS ........................................... 3-6 3.3.1 Reactor Internal Pressure Differences .......................................... 3-6 3.3.2 Reactor Internals Structural Evaluation .......................................... 3-6 3.3.3 Steam Dryer/Separator Performance .......................................... 3-10 3.4 FLOW INDUCED VIBRATION .......................................... 3-11 3.4.1 FIV Influence on Piping .......................................... 3-12 3.4.2 FIV Influence on Reactor Internal Components .......................................... 3-12 3.5 PIPING EVALUATION ........................................... 3-14 3.5.1 Reactor Coolant Pressure Boundary Piping.......................................... 3-14 3.5.2 Balance-Of-Plant Piping .......................................... 3-20 3.6 REACTOR RECIRCULATION SYSTEM ...................................... 3-24 3.7 MAIN STEAM LINE FLOW RESTRICTORS . ..................................3-26 3.8 MAIN STEAM ISOLATION VALVES ........................................ 3-27 3.9 REACTOR CORE ISOLATION COOLING/ISOLATION CONDENSER .. 3-28 3.10 RESIDUAL HEAT REMOVAL SYSTEM . .3-31 3.11 REACTOR WATER CLEANUP SYSTEM . .3-32

4. ENGINEERED SAFETY FEATURES . . .4-1 4.1 CONTAINMENT SYSTEM PERFORMANCE . .4-1 4.1.1 Containment Pressure and Temperature Response .4-2 4.1.2 Containment Dynamic Loads .4-4 4.1.3 Containment Isolation .4-5 4.1.4 Generic Letter 89-10 Program .4-5 4.1.5 Generic Letter 89-16 .4-6 4.1.6 Generic Letter 96-06 ..................................... 4-6 4.2 EMERGENCY CORE COOLING SYSTEMS .................................... 4-7 4.2.1 High Pressure Coolant Injection ..................................... 4-7 4.2.2 High Pressure Core Spray ..................................... 4-8 4.2.3 Core Spray or Low Pressure Core Spray ..................................... 4-8 4.2.4 Low Pressure Coolant Injection ..................................... 4-9 iv

NEDO-33076 4.2.5 Automatic Depressurization System ......................................................... 4-9 4.2.6 ECCS Net Positive Suction Head ......................................................... 4-10 4.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE . .................... 4-11 4.4 MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM ................... 4-12 4.5 FILTRATION, RECIRCULATION, AND VENTILATION SYSTEM ................ 4-13 4.6 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM ........... 4-15 4.7 POST-LOCA COMBUSTIBLE GAS CONTROL SYSTEM . ....................... 4-15

5. INSTRUMENTATION AND CONTROL ......................................................... 5-1 5.1 NSSS MONITORING AND CONTROL ......................................................... 5-1 5.1.1 Neutron Monitoring System ........................ ................................. 5-1 5.1.2 Rod Worth Minimizer/Rod Control and Information System ........... ............ 5-3 5.2 BOP MONITORING AND CONTROL ......................................................... 5-4 5.2.1 Pressure Control System ......................................................... 5-4 5.2.2 Turbine Steam Bypass System ......................................................... 5-5 5.2.3 Feedwater Control System ......................................................... 5-6 5.2.4 Leak Detection System ................. ........................................ 5-6 5.3 TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS .. 5-8 5.3.1 Main Steam Line High Flow Isolation ........................................................ 5-10 5.3.2 Turbine First-Stage Pressure Scram and Recirculation Pump Trip Bypass 5-11 5.3.3 APRM Flow Bi ased Scram ......................................................... 5-11 5.3.4 Rod Worth Minimizer/RCIS Rod Pattern Controller Low Power Setpoint 5-12 5.3.5 Rod Block Monitor ......................................................... 5-12 5.3.6 RCIS Rod Withdrawal Limiter High Power Setpoint .................................. 5-12 5.3.7 APRM Setdown in Startup Mode ......................................................... 5-13 5.4 CHANGES TO INSTRUMENTATION AND CONTROLS .......... ....................... 5-13

6. ELECTRICAL POWER AND AUXILIARY SYSTEMS ................................................. 6-1 6.1 AC POWER ......................................................... 6-1 6.1.1 AC Power (degraded voltage)........................................................................ 6-1 6.1.2 AC Power (normal operation).......................................................................

.6-2 6.2 DC POWER ......................................................... 6-2 6.3 FUEL POOL ......................................................... 6-3 6.3.1 Fuel Pool Cooling ......................................................... 6-3 V

NEDO-33076 6.3.2 Crud Activity and Corrosion Products ......................................................... 6-4 6.3.3 Radiation Levels ......................................................... 6-4 6.3.4 Fuel Racks ......................................................... 6-5 6.4 WATER SYSTEMS ......................................................... 6-6 6.4.1 Cooling Water Systems ......................................................... 6-6 6.4.2 Main Condenser/Circulating Water/Nonnal Heat Sink Performance ........... 6-8 6.4.3 Reactor Auxiliaries Cooling System ......................................................... 6-9 6.4.4 Turbine Auxiliaries Cooling System ......................................................... 6-10 6.4.5 Ultimate Heat Sink ......................................................... 6-10 6.5 STANDBY LIQUID CONTROL SYSTEM ..................................... 6-10 6.6 POWER DEPENDENT HVAC ......................................................... 6-11 6.7 FIRE PROTECTION ......................................................... 6-12 6.7.1 10 CFR 50 Appendix R Fire Event ......................................................... 6-13 6.8 OTHER SYSTEMS AFFECTED BY POWER UPRATE ............. ......................... 6-14

7. POWER CONVERSION SYSTEMS .......................................................... 7-1 7.1 TURBINE-GENERATOR ......................................................... 7-1 7.2 CONDENSER AND STEAM JET AIR EJECTORS ................................................ 7-2 7.3 TURBINE STEAM BYPASS ................................................. 7-3 7.4 FEEDWATER AND CONDENSATE SYSTEMS ................................................... 7-4 7.4.1 Normal Operation ......................................................... 7-4 7.4.2 Transient Operation ......................................................... 7-4 7.4.3 Condensate Demineralizers ...................... ................................... 7-6

8. RADWASTE AND RADIATION SOURCES ......................................................... 8-1 8.1 LIQUID AND SOLID WASTE MANAGEMENT . ................................8-1 8.2 GASEOUS WASTE MANAGEMENT .......................................... 8-2 8.3 RADIATION SOURCES IN THE REACTOR CORE ............................................. 8-4 8.4 RADIATION SOURCES IN REACTOR COOLANT ............................................. 8-5 8.4.1 Coolant Activation Products ......................................................... 8-5 8.4.2 Activated Corrosion Products and Fission Products ...................................... 8-6 8.5 RADIATION LEVELS ......................................................... 8-6 8.6 NORMAL OPERATION OFF-SITE DOSES ..................................... 8-8

9. REACTOR SAFETY PERFORMANCE EVALUATIONS .............................................. 9-1 Vi

NEDO-33076 9.1 ANTICIPATED OPERATIONAL OCCURRENCES . ..............................9-1 9.1.1 Transient Events ........................................................ 9-1 9.1.2 Alternate Shutdown Cooling Evaluation ....................................................... 9-2 9.2 DESIGN BASIS ACCIDENTS ........................................................ 9-3 9.3 SPECIAL EVENTS ........................................................ 9-5 9.3.1 Anticipated Transient Without Scram ........................................................ 9-5 9.3.2 Station Blackout ........................................................ 9-6 9.3.3 ATWS with Core Instability ........................................................ 9-7

10. OTHER EVALUATIONS . . . . .................................................... 10-1 10.1 HIGH ENERGY LINE BREAK ........................................................ 10-1 10.1.1 Steam Line Breaks ........................................................ 10-1 10.1.2 Liquid Line Breaks ........................................................ 10-2 10.2 MODERATE ENERGY LINE BREAK ........................................ 10-3 10.3 ENVIRONMENTAL QUALIFICATION ....................................... 10-4 10.3.1 Electrical Equipment ........................................................ 10-4 10.3.2 Mechanical Equipment With Non-Metallic Components ........................... 10-5 10.3.3 Mechanical Component Design Qualification ............................................. 10-5 10.4 TESTING ........................................................ 10-6 10.5 INDIVIDUAL PLANT EVALUATION ........................................ 10-8 10.5.1 Initiating Event Frequency ........................................................ 10-13 10.5.2 Component and System Reliability ........................................................ 10-17 10.5.3 Operator Response ........................................................ 10-19 10.5.4 Success Criteria ........................................................ 10-20 10.5.5 External Events ........................................................ 10-23 10.5.6 Shutdown Risk ........................................................ 10-26 10.5.7 PRA Quality ........................................................ 10-29 10.6 OPERATOR TRAINING AND HUMAN FACTORS . ...........................10-32 10.7 PLANTLIFE ........................................................ 10-33 10.8 NRC AND INDUSTRY COMMUNICATIONS . ................................10-35 10.9 EMERGENCY AND ABNORMAL OPERATING PROCEDURES . ................10-36

11. REFERENCES ........................................................ 11-1 vii

NEDO-33076 List of Tables No. Title 1-1 Glossary of Terms 1-2 Hope Creek Computer Codes Used for CPPU 1-3 Hope Creek Current and CPPU Plant Operating Conditions 3-1 Hope Creek Adjusted Reference Temperatures 3-2 Hope Creek Upper Shelf Energy - 40 Year Life (32 EFPY) 3-3 Hope Creek CUFs of Limiting Components 3-4 Hope Creek RIPDs for Normal Conditions (psid) 3-5 Hope Creek RIPDs for Upset Conditions (psid) 3-6 Hope Creek RIPDs for Faulted Conditions (psid) 3-7 Hope Creek RIPDs for Emergency Conditions (psid) 3-8 Hope Creek Reactor Internal Components - Summary of Stresses 3-9 Hope Creek BOP Piping MS and FW ASME Class 1 Piping 3-10 Hope Creek BOP Piping MS and FW ASME Class 2 and 3 Piping 4-1 Hope Creek Containment Performance Results 4-2 Hope Creek ECCS Performance Results 4-3 Hope Creek FRVS Iodine Removal Capacity Parameters 5-1 Hope Creek Analytical Limits for Setpoints 5-2 Hope Creek Instrument Scaling Changes for CPPU 5-3 Hope Creek Instrument Setpoint Changes for CPPU 5-4 Hope Creek Instrument Replacements for CPPU 6-1 Hope Creek CPPU Plant Electrical Characteristics 6-2 Hope Creek Offsite Electric Power System 6-3 Hope Creek Spent Fuel Pool Parameters for CPPU 6-4 Hope Creek Appendix R Fire Event Evaluation Results 6-5 Basis for Classification of No Significant Effect 8-1 Post-LOCA Vital Access Area Dose Rates and Occupancies 8-2 Post-LOCA TSC Dose 8-3 Post-LOCA OSC Dose 8-4 Post-LOCA Security Center (Guard House) Dose 9-1 Hope Creek Parameters Used for Transient Analysis viii

NEDO-33076 9-2 Hope Creek Transient Analysis Results 9-3 Hope Creek MSLBA Radiological Consequences 9-4 Hope Creek ILPBA Radiological Consequences 9-5 Hope Creek LOCA Radiological Consequences 9-6 Hope Creek FHA Radiological Consequences 9-7 Hope Creek CRDA Radiological Consequences 9-8 Hope Creek Key Inputs for ATWS Analysis 9-9 Hope Creek Results of ATWS Analysis 10-1 Hope Creek High Energy Line Break 10-2 Hope Creek Equipment Qualification for CPPU 10-3 Summary Comparison of Baseline and Updated CDF for Hope Creek 104 Summary of the Basis for Models Used in CPPU Delta Risk Calculations 10-5 Model Changes to Reflect Plant Physical Changes or Effects of CPPU Implementation 10-6 Sensitivity Calculation to Reflect Postulated Effects on Risk Metrics of CPPU Implementation 10-7 Summary of Initiating Event Frequency Effects of CPPU 10-8 Comparison of Initiator Contributors Between the CLTP And CPPU PRA Models 10-9 Comparison of SORV Probabilities (CPPU and CLTP) 10-10 Disposition of Key Actions for Potential HEP Re-Calculation 10-11 Changes in Success Criteria Included in the Risk Assessment 10-12 Hope Creek FAC Parameter Comparison for CPPU ix

NEDO-33076 List of Figures No. Title 1-1 Power/Flow Operating Map for CPPU 1-2 Hope Creek CPPU Heat Balance -Nominal 1-3 Hope Creek CPPU Heat Balance - 102% Power 2-1 Hope Creek Validation of Base BSP Regions for CPPU 2-2 Hope Creek ICA OPRM Trip Enabled Region for CPPU 3-1 MSIV Closure with Flux Scram 9-1 Turbine Trip with Bypass Failure 9-2 Generator Load Rejection with Bypass Failure 9-3 Feedwater Controller Failure - Maximum Demand 9-4 Feedwater Controller Failure - Maximum Demand with Bypass Out of Service 10-1 Generalized "Bathtub" Reliability Curve for a Component 10-2 PRA Self-assessment Process Applied to Hope Creek x

NEDO-33076 EXECUTIVE

SUMMARY

This report summarizes the results of all significant safety evaluations performed that justify uprating the licensed thermal power at Hope Creek. The requested license power level is an increase to 3840 MWt from the current licensed reactor thermal power of 3339 MWt.

GE has previously developed and implemented Extended Power Uprate using Licensing Topical Reports (LTRs), "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate" NEDC-32424P-A, February 1999 (ELTRI) and" Generic Evaluations for General Electric Boiling Water Reactor Extended Power Uprate" NEDC-32523P-A, February 2000 (ELTR2). Based on the Extended Power Uprate experience, GE has developed an approach to uprate reactor power that maintains the current plant maximum normal operating reactor dome pressure. This approach is referred to as Constant Pressure Power Uprate (CPPU) and is contained in the Licensing Topical Report (LTR) NEDC-33004P, Revision 4, "Constant Pressure Power Uprate," hereafter referred to as the CLTR.

This report provides a systematic application of the CLTR approach to Hope Creek, including the performance of plant specific engineering assessments and confirmation of the applicability of the CLTR generic assessments required to support a CPPU. Hope Creek has implemented a fuel vendor change by introducing GE14 fuel. Some topics in this CPPU report are directly fuel dependent because the fuel type affects the resulting evaluation or the consequences of transients or accidents. Because the first cycle CPPU core will contain some non-GE (legacy) fuel, this CPPU report does not reference the CLTR as the basis for areas involving reactor systems and fuel issues, consistent with the NRC's Conditions and Limitations on the use of the CLTR. For those topics, the evaluation methods from ELTRI and ELTR2 are applied. Topics that are evaluated using input from the fuel dependent evaluation topics are not considered directly fuel dependent and are dispositioned per the CLTR. For example, the Anticipated Transient Without Scram (ATWS) event evaluation is directly fuel dependent. The results from the ATWS evaluation are then used as input to the Standby Liquid Control System (SLCS) performance evaluation. However, the use of fuel dependent input from the ATWS evaluation does not make the SLCS evaluation fuel dependent, even though the SLCS evaluation must satisfy the input requirement. Instead, the SLCS evaluation is independent of the analysis bases used (e.g., fuel type) to provide the input performance requirements for SLCS.

It is not the intent of this report to explicitly address all the details of the analyses and evaluations described herein. For example, only previously NRC-approved or industry accepted methods were used for the analyses of accidents and transients, as referred to in the LTRs.

Therefore, the safety analysis methods have been previously addressed, and thus, are not explicitly addressed in this report. Also, event and analysis descriptions that are already provided in other licensing reports or the Updated Final Safety -Analysis Report (UFSAR) are not repeated within this report. This report summarizes the significant evaluations needed to support a licensing amendment to allow for uprated power operation.

Uprating the power level of nuclear power plants can be done safely within plant-specific limits and is a cost-effective way to increase installed electrical generating capacity. Many light water reactors have already been uprated worldwide, including many BWR plants.

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NEDO-33076 An increase in the electrical output of a BWR plant is accomplished primarily by generating and supplying higher steam flow to the turbine-generator. Hope Creek, as originally licensed, has an as-designed equipment and system capability to accommodate steam flow rates above the current rating. Also, the plant has sufficient design margins to allow the plant to be safely uprated significantly beyond its originally licensed power level.

A higher steam flow is achieved by increasing the reactor power along specified control rod and core flow lines. A limited number of operating parameters are changed, some setpoints are adjusted and instruments are recalibrated. Plant procedures are revised and power ascension testing is performed. Modifications to some non-safety power generation equipment will be implemented over time, as needed.

Detailed evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, environmental issues, and design basis accidents were performed. This report demonstrates that Hope Creek can safely operate at the requested CPPU level. However, non-safety power generation modifications must be implemented in order to obtain the electrical power output associated with the uprate power. Until these modifications are completed, the non-safety balance of plant equipment may limit the electrical power output, which in turn may limit the operating thermal power level to less than the rated thermal power (RTP) level. These modifications have been evaluated and they do not constitute a material alteration to the plant, as discussed in 10 CFR 50.92.

The evaluations and reviews were conducted in accordance with the CLTR or the ELTRs as approved by the NRC. The results of these evaluations and reviews are presented in the succeeding sections of this report:

• All safety aspects of Hope Creek that are affected by the increase in thermal power were evaluated;

• Evaluations were performed using NRC-approved or industry accepted analysis methods;

• No changes, which require compliance with more recent industry codes and standards, are being requested;

• The UFSAR will be updated for the CPPU related changes, after CPPU is implemented, per the requirements in 10 CFR 50.71(e);

• No new design functions that require modifications are necessary for safety related systems for the CPPU and any modification to power generation equipment will be implemented per 10 CFR 50.59;

• Systems and components affected by CPPU were reviewed to ensure there is no significant challenge to any safety system;

• Compliance with current Hope Creek environmental regulations were reviewed;

• Potentially affected commitments to the NRC have been reviewed; and xii

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• Planned changes not yet implemented have also been reviewed for the effects of CPPU.

The Hope Creek licensing requirements have been reviewed, and it is concluded that this CPPU can be accommodated (1) without a significant increase in the probability or consequences of an accident previously evaluated, (2) without creating the possibility of a new or different kind of accident from any accident previously evaluated, and (3) without exceeding any existing regulatory limits applicable to the plant, which might cause a significant reduction in a margin of safety. Therefore, the requested CPPU does not involve a significant hazards consideration.

The environmental assessment accomplished for Hope Creek demonstrates that while the proposed increase in capacity results in minor increases in the environmental effects, the CPPU can be accommodated with changes only to existing air permit limitations. The environmental assessment considered plant effects such as increased temperature of the circulating water leaving the plant, cooling tower air emissions, alternative power sources, plant modifications required to implement CPPU, and low level radioactive waste. The hourly maximum emissions of cooling tower particulates will increase above the current limitation and a variance request has been submitted. For all other factors, even with the increased output, the effects remain within the bounds of the original environmental effects identified in the Final Environmental Statement for Hope Creek.

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NEDO-33076

1. INTRODUCTION 1.1. REPORT APPROACH Uprating the power level of nuclear power plants can be done safely within certain plant-specific limits. Most GE BWR plants have the capability and margins for an uprate of 5 to 20% without major nuclear steam supply system (NSSS) hardware modifications. Many light water reactors have already been uprated worldwide. Over a thousand MWe have already been added by uprate in the United States. Several BWR plants are among those that have already been uprated. This evaluation justifies a CPPU to 3840 MWt, which corresponds to 115% of the current licensed thermal power (CLTP) for Hope Creek.

This report follows the generic format and content for CPPU licensing reports, as described in the CLTR (Reference 1). The NRC approved or industry accepted versions of the computer codes and analytical methodologies used for ECCS-LOCA evaluations; anticipated operational occurrences (AOOs) other accident, and special event evaluations; and piping evaluations are documented in Reference 2. The limitations on use of these codes and methods as defined in the NRC staff position letter reprinted in Reference 2 were followed for this CPPU analysis.

A glossary of terms is provided in Table 1-1.

1.1.1 Generic Assessments Many of the component, system and performance evaluations contained within this report have been generically evaluated in the CLTR, and found to be acceptable. The plant specific applicability of these generic assessments is identified and confirmed in the applicable sections of this report. Generic assessments are those safety evaluations that can be dispositioned for a group or all BWR plants by:

• A bounding analysis for the limiting conditions,

• Demonstrating that there is a negligible effect due to CPPU, or

• Demonstrating that the required plant cycle specific reload analyses are sufficient and appropriate for establishing the CPPU licensing basis.

Bounding analyses may be based on either a demonstration that previous pressure increase power uprate assessments provided in Reference 2 or 3 are bounding or on specific generic studies provided in the CLTR. For these bounding analyses, the current CPPU experience is provided in the CLTR along with the basis and results of the assessment. For those CPPU assessments having a negligible effect, the current CPPU experience plus a phenomenological discussion of the basis for the assessment is provided in the CLTR. For generic assessments that are fuel design dependent, the assessments are applicable to GE / Global Nuclear Fuel LLC (GNF) fuel designs up through GE14, analyzed with GE methodology and are evaluated in this report in accordance with ELTRI and ELTR2 (References 2 and 3).

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NEDO-33076 1.1.2 Plant Specific Evaluation Plant specific evaluations are assessments of the principle evaluations that are not addressed by the generic assessments described in Section 1.1.1. The relative effect of CPPU on the plant specific evaluations and the methods used for their performance are provided in this report.

Where applicable, the assessment methodology is referenced. If a specific computer code is used, the name of this computer code is provided in the subsection. If the computer code is identified in Reference 1, 2, 3, or 4, these documents may be referenced rather than the original report. Table 1-2 provides a summary of the computer codes used.

The plant-specific evaluations performed and reported in this document use plant-specific values to model the actual plant systems, transient response, and operating conditions. These plant-specific analyses are performed using a GE14 equilibrium core design for operation at 115% of CLTP for a cycle length of 18 months.

1.1.3 Report Generation and Review Process This PUSAR represents several years of project planning activities, engineering analysis, technical verification, and technical customer review. The final stages of the PUSAR preparation include PUSAR integration, additional customer review, on-site and offsite review committee review, and submittal to NRC. The Hope Creek CPPU project relied on the generic power uprate licensing topical reports (References 1, 2, and 3) submitted to and approved by NRC.

The project begins with the respective GE and PSEG Nuclear LLC (PSEG) Project Managers creating a Project Work Plan (PWP). This PWP, developed in accordance with GE engineering procedures, was used to define the plant-specific work scope, inputs and outputs required for Project activities. A Division of Responsibility (DOR) between PSEG and GE was used to further develop the work scope and assign responsible engineers (REs) from each organization.

A Task Scoping Document (TSD) applicable for each GE task was created prior to any technical work being performed. Each GE task RE submitted a Design Input Request (DIR) to the PSEG task RE interface to define the correct plant information for use in the GE task analysis and evaluation. Additional DIRs were submitted as the project continued. A plant-specific PUSAR "shell" was created that contains the appropriate depth of information (but not the specifics) expected in the final PUSAR.

All pertinent information is captured in an individual task Design Record File (DRF) maintained by the GE RE with oversight by the respective engineering manager. Each DRF contains the Quality Assurance records applicable to the task, including evidence of design verification.

A Draft Task Report (DTR) was created for every GE task; the DTR includes a description of the analysis performed, inputs, methods, and results obtained, and includes input to the applicable PUSAR section(s). The DTR was design verified, in accordance with the GE Quality Assurance Program, by a GE technical verifier and a GE Regulatory Services verifier, with oversight by the responsible GE technical manager and GE Project Manager. The DTR was transmitted by the GE Project Manager to PSEG and reviewed by the PSEG RE and other PSEG engineer, as appropriate. Subsequent commnents were resolved between the GE and the PSEG REs and a 1-2

NEDO-33076 Final Task Report (FTR) was developed. The FTR was again design verified (whether or not there were changes to the document), in accordance with the GE Quality Assurance Program, by a GE technical verifier and a GE Regulatory Services verifier, with oversight by the responsible GE technical manager and GE Project Manager. The GE Project Manager transmitted the FTR to the customer.

For the Hope Creek CPPU, PSEG personnel:

1. Conducted multidisciplinary technical reviews of GE evaluation reports (DTRs and FTRs) to ensure:

i. Appropriate use of design inputs; ii. Consistency with the applicable LTRs; and iii. Design basis and licensing basis requirements were addressed.

2. Provided technical review results, in the form of detailed comments, to GE performers;

3. Participated in discussions with GE REs to address and resolve comments; and

4. Applied the process for quality assurance of off-site services to GE.

The Regulatory Services RE integrated the individual PUSAR sections creating a Draft PUSAR that was design verified, in accordance with the GE Quality Assurance Program, by another GE Regulatory Services engineer, with oversight by the GE Regulatory Services Manager and the GE Project Manager. The GE Project Manager transmitted the verified Draft PUSAR to PSEG where it received another complete review by PSEG's technical personnel, project staff, and Licensing staff.

PSEG personnel generated questions and comments, which were responded to by GE's technical and Regulatory Services personnel.

PSEG performed technical assessments and reviews of analyses performed by GE Nuclear Energy, Global Nuclear Fuels (GNF), and GE Energy Services (GEES) in support of the Hope Creek CPPU project. PSEG reviewed design inputs, analysis methodologies, and results in the GE Design Record Files. Specific PSEG review activities included:

• Assessment of design/configuration controls at GE offices in July 2003.

• Design reviews and design review team meetings for technical rigor and acceptance of GNF Methods for mixed-core applications in Thermal Hydraulic Modeling; Nuclear Design Modeling; SAFER-GESTR LOCA Modeling; Safety Limit MCPR; Transient Selection Review; Fuel Rod Thermal Mechanical Performance Limits; and, Core Stability.

• Technical review of GE work in January 2004 at the GENE offices.

An additional technical assessment will be performed at the GE offices during 2005.

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NEDO-33076 1.2. PURPOSE AND APPROACH An increase in electrical output of a BWR is accomplished primarily by generation and supply of higher steam flow to the turbine generator. Most BWRs, as originally licensed, have an as-designed equipment and system capability to accommodate steam flow rates at least 5% above the original rating. In addition, continuing improvements in the analytical techniques (computer codes) based on several decades of BWR safety technology, plant performance feedback, operating experience, and improved fuel and core designs have resulted in a significant increase in the design and operating margin between the calculated safety analyses results and the current plant licensing limits. The available margins in calculated results, combined with the as-designed excess equipment, system, and component capabilities (1) have allowed many BWRs to increase their thermal power ratings by 5% without any NSSS hardware modification, and (2) provide for power increases up to 20% with some non-safety hardware modifications. These power increases involve no significant increase in the hazards presented by the plants as approved by the NRC at the original license stage.

The method for achieving higher power is to extend the power/flow map (Figure 1-1) along the Maximum Extended Load Line Limit Analysis (MELLLA). However, there is no increase in the maximum normal operating reactor vessel dome pressure or the maximum licensed core flow over their CLTP values. CPPU operation does not involve increasing the maximum normal operating reactor vessel dome pressure, because the plant, after modifications to non-safety power generation equipment, has sufficient pressure control and turbine flow capabilities to control the inlet pressure conditions at the turbine.

1.2.1 Uprate Analysis Basis Hope Creek is currently licensed at 3339 MWt. The CPPU RTP level included in this evaluation is 115% of the CLTP. Plant specific CPPU parameters are listed in Table 1-3. The CPPU safety analyses are based on a power level of 1.02 times the CPPU power level unless the Regulatory Guide 1.49 two percent power factor is already accounted for in the analysis methods consistent with the methodology described in, "General Electric Standard Application for Reactor Fuel,"

NEDE-2401 1-P-A and NEDE-2401 1-P-A-US, (latest approved revision).

1.2.2 Computer Codes NRC-approved or industry-accepted computer codes and calculational techniques are used to demonstrate compliance with the applicable regulatory acceptance criteria. The application of these codes to the CPPU analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER where applicable for each code. The limitations on use of these codes and methods as defined in the NRC staff position letter reprinted in ELTRI (Reference 2) were followed for this CPPU analysis. Any exceptions to the use of the code or conditions of the applicable SER are noted in Table 1-2. The application of the computer codes in Table 1-2 is consistent with the current Hope Creek licensing basis except where noted in this report.

1-4

NEDO-33076 1.2.3 Approach The planned approach to achieving the higher power level consists of the change to the Hope Creek licensing and design basis to increase the licensed power level to 3840 MWt, consistent with the approach outlined in the CLTR, except as specifically noted in this report. Consistent with the CLTR, the following plant specific exclusions are exercised:

• No increase in maximum normal operating reactor dome pressure

• No increase in the maximum licensed core flow

• No increase to currently licensed MELLLA upper boundary This report provides a systematic application of the CLTR approach to Hope Creek, including the performance of plant specific engineering assessments and confirmation of the applicability of the CLTR generic assessments required to support a CPPU. Hope Creek has implemented a fuel vendor change by introducing GE14 fuel. Some topics in this CPPU report are directly fuel dependent because the fuel type affects the resulting evaluation or the consequences of transients or accidents. Because the first cycle CPPU core will contain some non-GE (legacy) fuel, this CPPU report does not reference the CLTR as the basis for areas involving reactor systems and fuel issues, consistent with the NRC's Conditions and Limitations on the use of the CLTR. For those topics, the evaluation methods from ELTRI and ELTR2 (References 2 and 3) are applied (i.e., Sections 2.1, 2.2, 2.3, 2.4, 4.3, 9.1.1, and 9.3.1 in this report). Topics that are evaluated using input from the fuel dependent evaluation topics are not considered directly fuel dependent and are dispositioned per the CLTR. For example, the Anticipated Transient Without Scram (ATWS) event evaluation is directly fuel dependent. The results from the ATWS evaluation are then used as input to the Standby Liquid Control System (SLCS) performance evaluation.

However, the use of fuel dependent input from the ATWS evaluation does not make the SLCS evaluation fuel dependent, even though the SLCS evaluation must satisfy the input requirement.

Instead, the SLCS evaluation is independent of the analysis bases used (e.g., fuel type) to provide the input performance requirements for SLCS.

The plant-specific evaluations are based on a review of plant design and operating data, as applicable, to confirm excess design capabilities; and, if necessary, identify required modifications associated with CPPU. For specified topics, generic analyses and evaluations in the CLTR or the ELTRs as applicable, demonstrate plant operability and safety. Most of the dispositions are based on a 120% of OLTP increase and are bounding for the requested 115% of CLTP uprate. For this increase in power, the conclusions of system/component acceptability stated in the CLTR or the ELTRs are bounding and have been confirmed for Hope Creek. The scope and depth of the evaluation results provided herein are established based on the approach 1-5

NEDO-33076 in the CLTR or ELTRs and the unique features of the plant. The results of these evaluations are presented in the following sections:

a) Reactor Core and Fuel Performance: Specific analyses required for CPPU have been performed for a GE14 equilibrium cycle with the reactor core operating at CPPU conditions. Specific core and fuel performance is evaluated for each operating cycle, and will continue to be evaluated and documented for the operating cycles that implement CPPU.

b) Reactor Coolant System and Connected Systems: Evaluations of the NSSS components and systems have been performed at CPPU conditions. These evaluations confirm the acceptability of the effects of the higher power and the associated change in process variables (i.e., increased steam and feedwater flows). Safety-related equipment performance is the primary focus in this report, but aspects of reactor operational capability are also included.

c) Engineered Safety Feature Systems: The effects of CPPU power operation on the Containment, ECCS, Standby Gas Treatment system and other Engineered Safety Features have been evaluated for key events. The evaluations include the containment responses during limiting AQOs and special events, ECCS-LOCA, and safety relief valve (SRV) containment dynamic loads.

d) Control and Instrumentation: The control and instrumentation signal ranges and analytical limits for setpoints have been evaluated to establish the effects of the changes in various process parameters such as power, neutron flux, steam flow and feedwater flow.

As required, setpoint evaluations have been performed to determine the need for any Technical Specification setpoint changes for various functions (e.g., main steam line high flow isolation setpoints).

e) Electrical Power and Auxiliary Systems: Evaluations have been performed to establish the operational capability of the plant electrical power and distribution systems and auxiliary systems to ensure that they are capable of supporting safe plant operation at the CPPU power level.

f) Power Conversion Systems: Evaluations have been performed to establish the operational capability of various non-safety balance-of-plant (BOP) systems and components to ensure that they are capable of delivering the increased power output, and/or the modifications necessary to obtain full CPPU power.

g) Radwaste Systems and Radiation Sources: The liquid and gaseous waste management systems have been evaluated at limiting conditions for CPPU to show that applicable release limits continue to be met during operation at higher power. The radiological consequences have been evaluated for CPPU to show that applicable regulations have been met for the CPPU power conditions. This evaluation includes the effect of higher power level on source terms, on-site doses and off-site doses, during normal operation.

1-6

NEDO-33076 h) Reactor Safety Performance Evaluations: The limiting Updated Final Safety Analysis Report (UFSAR) analyses for design basis events have been addressed as part of the CPPU evaluation. All limiting accidents, AO0s, and special events have been analyzed or generically dispositioned consistent with the ELTRs and show continued compliance with regulatory requirements. ((

1))

i) Additional Aspects of CPPU: High-energy line break and environmental qualification evaluations have been performed at bounding conditions for CPPU to show the continued operability of plant equipment under CPPU conditions. The effects of CPPU on the Hope Creek Individual Plant Evaluation (IPE) have been analyzed to demonstrate that there are no new vulnerabilities to severe accidents.

1.2.4 Concurrent Changes Unrelated to CPPU Consistent with the NRC conditions and limitations on the use of the CLTR, Hope Creek is not requesting concurrent review of any changes listed among the restrictions applicable to the CLTR.

1.3 CPPU PLANT OPERATING CONDITIONS 1.3.1 Reactor Heat Balance The operating pressure, the total core flow, and the coolant thermodynamic state characterize the thermal hydraulic performance of a BWR reactor core. The CPPU values of these parameters are used to establish the steady state operating conditions and as initial and boundary conditions for the required safety analyses. The CPPU values for these parameters are determined by performing heat (energy) balance calculations for the reactor system at CPPU conditions.

The reactor heat balance relates the thermal-hydraulic parameters to the plant steam and feedwater flow conditions for the selected core thermal power level and operating pressure.

Operational parameters from actual plant operation are considered (e.g., steam line pressure drop) when determining the expected CPPU conditions. The thermal-hydraulic parameters define the conditions for evaluating the operation of the plant at CPPU conditions. The thermal-hydraulic parameters obtained for the CPPU conditions also define the steady state operating conditions for equipment evaluations. Heat balances at appropriately selected conditions define the initial and boundary conditions for plant safety analyses.

Figure 1-2 shows the CPPU heat balance at 100% of CPPU and 100% rated core flow. Figure 1-3 shows the CPPU heat balance at 102% of CPPU and 100% core flow.

Table 1-3 provides a summary of the reactor thermal-hydraulic parameters for the current rated and CPPU conditions. At CPPU conditions, the maximum nominal operating reactor vessel dome pressure is maintained at the current value, which minimizes the need for plant and licensing changes. With the increased steam flow and associated non-safety BOP modifications, the current dome pressure provides sufficient operating turbine inlet pressure to assure good pressure control characteristics.

1-7

NEDO-33076 1.3.2 Reactor Performance Improvement Features The UFSAR, core and fuel reload evaluations, and the Technical Specifications currently include allowances for plant operation with the performance improvement features and the equipment out-of-service (OOS) listed in Table 1-3. When limiting, the input parameters related to the performance improvement features or the equipment OOS have been included in the safety analyses for CPPU. The use of these performance improvement features and allowing for equipment OOS is continued during CPPU operation. The evaluations that are dependent upon cycle length are performed for CPPU assuming an 18-month cycle.

1.4

SUMMARY

AND CONCLUSIONS This evaluation has covered a CPPU to 11 5% of CLTP. The strategy for achieving higher power is to extend the MELLLA power/flow map region along the upper boundary extension.

The Hope Creek licensing requirements have been reviewed to demonstrate how this uprate can be accommodated without a significant increase in the probability or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from any accident previously evaluated, and without exceeding any existing regulatory limits or design allowable limits applicable to the plant which might cause a reduction in a margin of safety. The CPPU described herein involves no significant hazard consideration.

1-8

NEDO-33076 Table 1-1 Glossary of Terms Term Definition AC Alternating current ADS Automatic Depressurization System ADHR Alternate Decay Heat Removal AL Analytical Limit ALARA As Low As Reasonably Achievable ANS American Nuclear Society ANSI American National Standards Institute AOO Anticipated operational occurrences (moderate frequency transient events)

APRM Average Power Range Monitor ASME American Society of Mechanical Engineers AST Alternate Source Term ATWS Anticipated Transient Without Scram AV Allowable Value BHP Brake horse power BOP Balance-of-plant BWR Boiling Water Reactor BWROG BWR Owners Group CD Condensate demineralizer CDF Core damage frequency CFR Code of Federal Regulations CLTP Current Licensed Thermal Power CLTR Constant Pressure Power Uprate Licensing Topical Report CO Condensation oscillation CPF Condensate pre-filter 1-9

NEDO-33076 Term Definition CPPU Constant Pressure Power Uprate CR Control room CRD Control Rod Drive CRDA Control Rod Drop Accident CSC Containment Spray Cooling CS Core Spray CUF Cumulative usage factors DBA Design basis accident DC Direct current DFCS Digital Feedwater Control System DIR Design Input Request DOR Division of Responsibility DRF Design Record File DTR Draft Task Report EAB Exclusion Area Boundary ECCS Emergency Core Cooling System EFPY Effective full power years ELTR1 Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate" NEDC-32424P-A, February 1999 ELTR2 Generic Evaluations for General Electric Boiling Water Reactor Extended Power Uprate" NEDC-32523P-A, February 2000 EOC End of cycle EOP Emergency Operating Procedure(s)

EQ Environmental qualification FAC Flow Accelerated Corrosion F/D Filter-Demineralizer FFWTR Final Feedwater Temperature Reduction FHA Fuel Handling Accident FIV Flow induced vibration 1-10

NEDO-33076 Term Definition FLIM Failure likelihood index methodology FPCC Fuel Pool Cooling and Cleanup FRVS Filtration, Recirculation, and Ventilation System FTR Final Task Report FW Feedwater GE General Electric Company GH Guard house gpm Gallons per minute HX Heat exchanger HELB High Energy Line Break HCR Human cognitive reliability HEP Human error probability HgA Inches of mercury absolute HPCI High Pressure Coolant Injection HVAC Heating Ventilating and Air Conditioning HWC Hydrogen water chemistry ICS Integrated computer system IEEE Institute of Electrical and Electronics Engineers ILBA Instrument Line Break Accident IRM Intermediate Range Monitor KV Kilo Volt LCS Leakage Control System LDS Leak Detection System LERF Large early release frequency LHGR Linear Heat Generation Rate LLW Low Level Waste LOCA Loss-Of-Coolant Accident LOFW Loss of feedwater 1-11

NEDO-33076 Term Definition LOOP Loss of offsite power LPCI Low Pressure Coolant Injection LPRM Local Power Range Monitor LPSP Low Power Setpoint LPZ Low Population Zone MAAP Modular accident analysis program MAPLHGR Maximum Average Planar Linear Heat Generation Rate MBTU Millions of BTUs MCPR Minimum Critical Power Ratio MCR Main control room MELB Moderate Energy Line Break MELLLA Maximum Extended Load Line Limit Analysis MeV Million Electron Volts Mlb Millions of pounds MS Main steam MSIV Main Steam Isolation Valve MSL Main steam line MSLBA Main Steamline Break Accident MSRV(s) Main steam relief valve(s)

MSVV Main steam valve vault Mvar Megavar MWe Megawatt(s)-electric MWt Megawatt-thermal MSL Main steam line MVA Million Volt Amps NA Not Applicable NMCA Noble metal chemical addition NPSH Net positive suction head 1-12

NEDO-33076 Term Definition NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system NSSSS Nuclear steam supply shutoff system NUREG Nuclear Regulations OLTP Original Licensed Thermal Power OOS Out-of-service ORAM Outage Risk Assessment and Management AP Differential pressure - psi P25 25% of CPPU Rated Thermal Power PCP Primary Condensate Pump PCS Pressure Control System PCT Peak cladding temperature PJM Pennsylvania-New Jersey-Maryland PRA Probabilistic Risk Assessment PSA Probabilistic Safety Analysis PSEG Public Service Electric and Gas, PSEG Nuclear LLV PSF Performance-shaping factor psi Pounds per square inch psia Pounds per square inch - absolute psid Pounds per square inch - differential psig Pounds per square inch - gauge RBCCW Reactor Building Closed Cooling Water RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary RE Responsible Engineer RFP Reactor Feedwater Pump RHR Residual Heat Removal 1-13

NEDO-33076 Term Definition RHRSW Residual Heat Removal Service Water RIPD Reactor internal pressure difference(s)

RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RSLB Recirculation system line break RTP Rated Thermal Power RTNDT Reference temperature of nil-ductility transition RWCU Reactor Water Cleanup RWM Rod Worth Minimizer RP Radiation Protection Saldt CPPU alternating stress intensity Sm Code allowable stress limit SAMG Severe Accident Mitigation Guideline(s)

SAR Safety Analysis Report SBO Station blackout scfh Standard cubic feet per hour scfm Standard cubic feet per minute SCP Secondary Condensate Pump SDC Shutdown Cooling SER Safety Evaluation Report SJAE Steam Jet Air Ejectors SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single-loop operation SRM Source Range Monitor SRV Safety relief valve(s)

SRVDL Safety relief valve discharge line TACS Turbine Auxiliaries Cooling System 1-14

NEDO-33076 Term Definition TAF Top of active fuel TEDE Total Effective Dose Equivalent TLO Two (recirculation) loop operation TSC Technical Support Center TSD Task Scoping Document TSV Turbine Stop Valve UFSAR Updated Final Safety Analysis Report UHS Ultimate heat sink VWO Valves wide open Yr Year 1-15

NEDO-33076 Table 1-2 Hope Creek Computer Codes Used For CPPU *

'Task Computer '. Version or 'NRC Comments

'Code ' .:Revision Approved ____..--_.-.-..-:___

Nominal Reactor Heat ISCOR 09 Y(2) NEDE-2401 IP Rev. 0 SER Balance Reactor Core and Fuel TGBLA 06 Y NEDE-30130-P-A Performance PANACEA 11 Y (5) NEDE-30130-P-A GESAM 01 Y(3) NEDO-10958-A Reactor Powver/Flow Map BILBO 04V NA (1); NEDE-23504, February 1977 Thermal Hydraulic ODYSY 05 Y NEDC-32992P-A Stability Reactor Vessel Fluence DORTGOI I N (14)

TGBLA 6 Y (15)

Reactor Internal Pressure ISCOR 09 Y(2) NEDE-2401 IP Rev. 0 SER Differences LAMB 07 (4) NEDE-20566-P-A TRACG 02 Y NEDE-32176P, Rev 2, Dec 1999 NEDC-32177P, Rev 2, Jan 2000 NRC TAC No M90270, Sep 1994 (13)

Containment System SHEX 05 Y (8)

Response M3CPT 05 Y NUREG-0661 LAMB 08 (4) NEDE-20566-P-A Transient Analysis PANACEA 11 Y NEDE-30130-P-A (5)

ISCOR 09 Y(2) NEDE-2401 IP Rev. 0 SER ODYN 10 Y NEDO-24154-A SAFER 04 Y(6) NEDC-32424P-A, NEDC-32523P-A, (9), (10), (11)

TASC 03A Y NEDC-32084P-A, Rev. 2, July 2002 Anticipated Transient ODYN 10 Y NEDE-24154P-A Supp. 1, Vol.4 Without Scram STEMP 04 (7)

PANACEA 11 Y NEDE-30130-P-A ISCOR 9 Y (2) NEDE-2401 IP Rev. 0 SER TASC 03A Y NEDC-32084P-A, Rev. 2, July 2002 SHEX 05 Y (8)

Station Blackout SHEX 05 Y (8) 1-16

NEDO-33076 Task Computer e or NRC-Comments

..Code,,,.~

de`1 Re'viIsin

. o. .. Approved _ _ _ _ _ _ _ _ _ _ _

Appendix R Fire GESTR 08 (6) NEDE-23785-1 -PA, Rev. I Protection SAFER 04 (6) (9) (10) (11)

SHEX 05 (8)

Reactor Recirculation BILBO 04V NA (1)NEDE-23504, February System 1977 ECCS-LOCA LAMB 08 Y NEDO-20566A GESTR 08 Y NEDE-23785-1-PA, Rev. I SAFER 04 Y (9) (10) (11)

ISCOR 09 Y(2) NEDE-2401 IP Rev. 0 SER TASC 03A Y NEDC-32084P-A, Rev. 2, July 2002 Fission Product Inventory ORIGEN2 2.1 N Isotope Generation and Depletion Code High Energy Line Break COMPARE-MOD I A N(12) LA-7199-MS Probabilistic Risk MAAP 4.0.4 N (16)

Assessment

• The application of these codes to the CPPU analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER where applicable for each code. The application of the codes also complies with the SERs for the extended power uprate programs.

1. Not a safety analysis code that requires NRC approval. The code application is reviewed and approved by GENE for "Level-2" application and is part of GENE's standard design process. Also, the application of this code has been used in previous power uprate submittals.

2. The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-2401 IP Rev. 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods.

3. The GESAM code is not approved by name. The methodology has been approved in NEDO-10958-A

4. The LAMB code is approved for use in ECCS-LOCA applications (NEDE-20566P-A and NEDO-20566A), but no approving SER exists for the use of LAMB in the evaluation of reactor internal pressure differences or containment system response. The use of LAMB for these applications is consistent with the model description of NEDE-20566P-A.

5. The physics code PANACEA provides inputs to the transient code ODYN. The improvements to PANACEA that were documented in NEDE-30130-P-A were incorporated into ODYN by way of Amendment 11 of GESTAR II (NEDE-24011-P-A). The use of TGBLA Version 06 and PANACEA Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A.

Watford (GE)

Subject:

"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, 1-17

NEDO-33076 GESTAR II Implementing Improved GE Steady-State Methods," (TAC NO. MA6481),

November 10, 1999.

6. The ECCS-LOCA codes are not explicitly approved for Transient or Appendix R usage. The staff concluded that SAFER is qualified as a code for best estimate modeling of loss-of-coolant accidents and loss of inventory events via the approval letter and evaluation for NEDE-23785P, Revision 1, Volume II. (Letter, C.O. Thomas (See NRC) to J.F. Quirk (GE),

"Review of NEDE-23785-1 (P), "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volumes I and II", August 29, 1983.) In addition, the use of SAFER in the analysis of long term Loss-of-Feedwater events is specified in the approved LTRs for power uprate: "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32424P-A, February 1999 and "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32523P-A, February 2000. The Appendix R events are similar to the loss of feedwater and small break LOCA events.

7. The STEMP code uses fundamental mass and energy conservation laws to calculate the suppression pool heatup. The use of STEMP was noted in NEDE-24222, "Assessment of BWR Mitigation of ATWS, Volume I & II (NUREG-0460 Alternate No. 3) December 1, 1979." The code has been used in ATWS applications since that time. There is no formal NRC review and approval of STEMP or the ATWS topical report.

8. The application of the methodology in the SHEX code to the containment response is approved by NRC in the letter to G. L. Sozzi (GE) from A. Thadani (NRC), "Use of the SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.

9. Letter, J.F. Klapproth (GE) to USNRC, Transmittal of GE Proprietary Report NEDC-32950P "Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model," dated January 2000 by letter dated January 27,2000

10. Letter, S.A. Richards (NRC) to J.F. Klapproth, "General Electric Nuclear Energy (GENE)

Topical Reports GENE (NEDC)-32950P and GENE (NEDC)-32084P Acceptability Review,"

May 24, 2000.

11. "SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-Jet Pump Plants," NEDE-30996P-A, General Electric Company, October 1987.

12. "A Computer Code for Transient Analysis of Volumes with Heat Sinks, Flowing Vents and Doors", LA-7199-MS

13. NRC has reviewed and accepted the TRACG application for the flow-induced loads on the core shroud as stated in NRC SER TAC No. M90270.

14. CCC-543 "TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport Version 2.8.14," Radiation Shielding Information Center (RSIC), January 1994.

15. The use of TGBLA Version 06 and PANACEA Version 11 was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE)

Subject:

"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC NO. MA6481), November 10, 1999.

16. MAAP is an industry-accepted code used for thermal-hydraulic analysis for many IPE submittals to the NRC.

1-18

NEDO-33076 Table 1-3 Hope Creek Current and CPPU Plant Operating Conditions Current*

Licensed CPPU Parameter Value Value Thermal Power (MWt) 3339 3840 Vessel Steam Flow (Mlb/hr)** 14.404 16.773 Full Power Core Flow Range Mlb/hr 76.6 to 105 94.8 to 105

% Rated 76.6 to 105 94.8 to 105 Maximum Nominal Dome Pressure (psia) 1020 No Change Maximum Nominal Dome Temperature (0F) 547.0 No Change Pressure at upstream side of turbine stop valve (TSV) (psia) 963.0 943.0 Full Power Feedwater Flow (Mlb/hr) 14.372 16.741 Temperature (IF) 422.6 431.6 Core Inlet Enthalpy (Btu/lb) *** 526.2 525.1

• Based on CLTP reactor heat balance.

• • At normal feedwater heating.

• At 100% core flow condition.

Currently licensed performance improvement features and/or equipment OOS that are included in CPPU evaluations:

1. Maximum Extended Load Line Limit Analysis (MELLLA)

2. End-of-Cycle (EOC) Coastdown (GESTAR Generic Analysis)

3. Single Loop Operation (SLO)

4. Feedwater Temperature Reduction (FWTR)

5. One SRV OOS

6. 3% SRV Setpoint tolerance

7. Increased core flow (ICF)

8. End-of-Cycle Recirculation Pump Trip (EOC RPT) OOS

9. Improved Scram Time

10. Thermal Power Monitoring

11. ARTS-MELLLA 1-19

NEDO-33076 Core Flow (Nllbthr) 0 10 20 30 40 50 60 70 80 90 too 110 120 120 . -, 3%0 100% EPU - 3840 Mwt 4500 100% CLTP - 3339 MWt 100% OLTP - 3293 MWt_

110 100% Core Flow - 100.0 Mlb/hr

- 4000 A: Natural Circulation MfELLLA Doundary Line D E F 100 B: Minimum Pump Speed 3840 MWt C: 57.6% Power/ 39.2% Flow D: 100.0% Power/ 94.8% Flow DI: 87.0% Power/ 76.6% Flow 86.95% PU Rod Line 3500

- 0 _ 4 F i339 MWt E: 100.0% Power/ 100.0% Flow (101.4% OLTP Rod Line)

E': 87.0% Power/ 100.0% Flow 3293 MWt e- 80 F: 100.0% Power/ 105.0% Flow F': 87.0% Power/ 105.0% Flow 3000 _

G: 20.6% Power/ 105.0% Flow

' 70 H: 20.6% Power/ 100.0% Flow 3 10 I: 20.6% Power/ 41.3% Flow 85.76% PU Rod ILine J: 52.9% Power/ 33.7% Flow J (I 00% OLTP Rod Line) 2500 '

r.

U 60

- - - - --- C __ _ . . _ _,_ _I U a

Increased Co" J 50 I* Flow Region - 2000 e E; . , I I-U UL I- . I. Mfinimnum Pump Speed

-_ 40 I 1500 U.

Natural Circulation, '

a UJ 30 1000 Cavitation Interlock 20 [

I If G 500 10 I

. . . . . . . . * . ..  ! . i . ... ... .. . i 0

0 10 20 30 40 50 60 70 80 90 100 ll0 120 Core Flow (%)

Figure 1-1 Power/Flow Operating Map for CPPU 1-20

NEDO-33076 Figure 1-2 Hope Creek CPPU Heat Balance - Nominal Legend

1. Flow, lbmrhr H ='Enthalpy, Btu/lbmr.';

F- Temperature, °F M = Moisture, % Main SteamFlowi, 116.773E+06#

• P=Prcssure, psia 0.52 M

• Carryunder= 0.40°/O 943 P

• j Min. I z.e

... Feed Main =-.- -'

e.d Flow 16.889E+06 # 16.7412+06 Y 410.3 H 410.3 H 431.6 °F 431.6 °F Ah= 1.1 H 1.480E+05 #

412.8 H 433.9 OF 3.200E+04 # 'Contro1 Rod Drive 1.480E+05 #

48.0 H t'-... "Feed Flow

. .. 525.0 H 77.0 OF 530.8 OF

• Conditions at upstream side ofTSV Core Thernnl Power 3840.0 Pump Heating 10.7 Cleanup Losses -4.9 Other System Losses -1.9 Turbine Cycle Use 3843.9 MWt 1-21

NEDO-33076 Figure 1-3 Hope Creek CPPU Heat Balance -102% Power

• Conditions at upstream side of TSV Core Thermal Power 3916.8 Pump Heating 10.7 Cleanup Losses 4.9 Other System Losses -1.9 Turbine Cycle Use 3920.7 MWt 1-22

NEDO-33076

2. REACTOR CORE AND FUEL PERFORMANCE This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapter 4, applicable to CPPU.

2.1 FUEL DESIGN AND OPERATION CPPU increases the power density proportional to the power increase. However, this power density remains within the current operating power density range of other BWRs. CPPU has some effects on operating flexibility, reactivity characteristics and energy requirements. The power distribution in the core is changed to achieve increased core power, while limiting the absolute power in any individual fuel bundle to within its allowable values.

At current or uprated conditions, all fuel and core design limits continue to be met by planned deployment of fuel enrichment and burnable poison. This is supplemented by core management control rod pattern and/or core flow adjustments. New fuel designs are not needed for CPPU to ensure safety. However, revised loading patterns, larger batch sizes and potentially new fuel designs may be used to provide additional operating flexibility and maintain fuel cycle length.

The reactor core design power distribution usually represents the most limiting thermal operating state at design conditions. It includes allowances for the combined effects on the fuel heat flux and temperature of the gross and local power density distributions, control rod pattern, and reactor power level adjustments during plant operation. NRC approved core design methods were used to analyze core performance at CPPU RTP. Detailed fuel cycle calculations, based on equilibrium GE14 core design for Hope Creek, demonstrate the feasibility of CPPU operation while maintaining fuel design limits. Thermal-hydraulic design and operating limits ensure an acceptably low probability of boiling transition-induced fuel cladding failure occurring in the core, even for the most severe postulated operational transients. Limits are also placed on fuel average planar linear heat generation rates and linear heat generation rates in order to meet peak cladding temperatures limits for the limiting Loss-of-Coolant Accident (LOCA) and fuel thermal mechanical design bases.

The subsequent reload core designs for operation at the CPPU RTP will take into account the above limits, to ensure acceptable differences between the licensing limits and their corresponding operating values.

CPPU may result in a change in fuel burnup, the amount of fuel to be used, and isotopic concentrations of the radionuclides in the irradiated fuel relative to the current level of burnup.

NRC-approved limits for burnup on the fuel designs are not exceeded. Also, due to the higher steady-state operating power associated with power uprate, the short-term curie content of the reactor fuel increases. The effects of higher power operation on radiation sources and design basis accident doses are discussed in Sections 8 and 9.2, respectively. CPPU has some effects on operating flexibility, reactivity characteristics, and energy requirements. These issues are discussed in the following sections based on GE experience and fuel characteristics.

The CPPU evaluations assumed the GE14 fuel type.

2-1

NEDO-33076 2.2 THERMAL LIMITS ASSESSMENT Operating limits ensure that regulatory and/or safety limits are not exceeded for a range of postulated events (e.g., transients, LOCA). This section addresses the effects of CPPU on thermal limits. An equilibrium GE14 core is used for the CPPU evaluation. Cycle-specific core configurations, evaluated for each reload, confirm CPPU capability, and establish or confirm cycle-specific limits, as is currently the practice.

For the subjects to be addressed in Sections 2.2.1 and 2.2.2, Hope Creek is consistent with the

[)) description provided in Sections 5.3.2 and 5.7.2 of ELTRI (Reference 2).

2.2.1 Minimum Critical Power Ratio (MCPR) Operating Limit The operating limit MCPR is determined on a cycle-specific basis from the results of the reload transient analysis, as described in Sections 5.3.2 and 5.7.2.1 of ELTRI and Section 3.4 of ELTR2 (Reference 3). This approach does not change for CPPU. ((

))A representative Operating Limit MCPR for CPPU operation is shown in Table 9-2.

2.2.2 MAPLHGR and Maximum LHGR Operating Limits The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and maximum Linear Heat Generation Rate (LHGR) limits will be maintained as described in GESTAR (Reference 4).

No significant change in operation is anticipated due to CPPU based on experience from other BWR uprates. The ECCS performance (ECCS-LOCA analysis) is addressed in Section 4.3, and also uses an equilibrium GE14 core for CPPU. ((

1))

2.3 REACTIVITY CHARACTERISTICS All minimum shutdown margin requirements apply to cold (68TF) conditions, and are maintained without change.

Operation at higher power could reduce the hot excess reactivity (typically by about 0.2 - 0.3%

Ak for each 5% power increase) during the cycle. This loss of reactivity does not affect safety, and is not expected to significantly affect the ability to manage the power distribution through the cycle to achieve the target power level. However, the lower hot excess reactivity can result in achieving an earlier all-rods-out condition. Through fuel cycle redesign, sufficient excess reactivity can be obtained to match the desired cycle length. Increasing hot reactivity may result in less hot-to-cold reactivity differences, and therefore, smaller cold shutdown margins.

However, this potential loss in margin can be accommodated through core design, and current design and Technical Specifications cold shut margin requirements are not affected. If needed, a bundle design with improved shutdown margin characteristics can be used to preserve the flexibility between hot and cold reactivity requirements for future cycles.

2-2

NEDO-33076 2.4 STABILITY Hope Creek Generating Station is currently operating under the requirements of reactor stability Long-Term Solution Option IIl. The Backup Stability Protection (BSP) solution is used when the Oscillation Power Range Monitor (OPRM) system fails, the effect of 115% thermal power uprate is addressed on the stability Option III solution (Reference 5).

2.4.1 BSP Evaluation The Base BSP Regions are shown in Figure 2-1. The Base BSP region boundary intercepts are, by default, defined using the ICA region boundary intercepts on the MELLLA boundary line and on the natural circulation line (NCL). The BSP regions are then established using the Generic Shape Function (GSF) to connect the region boundary intercepts. An evaluation is performed to determine the bounding region intercepts along the NCL and the MELLLA line. If these points fall within the Base BSP Region boundaries, then the Base BSP region boundaries are proposed for the cycle-specific backup stability solution. If these points fall outside the Base BSP region boundaries, then the regions should be expanded. For the equilibrium cycle evaluation, the BSP Scram Region and BSP Controlled Entry Region are the same as the Base BSP Regions as shown in Figure 2-1.

2.4.2 Option III Evaluation Option III is a detect-and-suppress solution, which combines closely spaced LPRM detectors into "cells" to effectively detect either core-wide or regional (local) modes of reactor instability.

Option III provides SLMCPR protection by generating a reactor scram if a reactor instability, which exceeds the specified trip setpoint, is detected. The acceptable setpoint is determined for each operating cycle per the NRC approved methodology discussed in Reference 5. The Option III stability reload licensing basis calculates the limiting OLMCPR required to protect the SLMCPR for both steady-state and transient stability events as specified in the Option III methodology. These OLMCPRs are calculated for a range of OPRM setpoints for MELLLA operation. Selection of an appropriate instrument setpoint is then made based upon the OLMCPR required to provide adequate SLMCPR protection. This determination relies on the DIVOM curve (Delta CPR Over Initial CPR Versus Oscillation Magnitude) to determine an OPRM setpoint that protects the SLMCPR during an anticipated instability event. Due to the concern regarding the generic regional mode DIVOM curve, a plant- and cycle-specific DIVOM slope is developed based on a cycle-specific TRACG evaluation.

The Oscillation Power Range Monitor (OPRM) is designed to provide the Option III automatic scram. The generic analyses for the Option III hot channel oscillation magnitude and the OPRM hardware were designed to be independent of core power. Because the OPRM hardware does not change, the hot channel oscillation magnitude portion of the Option III calculation is not affected by CPPU and does not need to be recalculated.

The Option III trip is armed only when plant operation is within the Option III trip-enabled region. The Option III trip-enabled region is defined as the region on the power/flow map with power 230% CLTP and core flow <60% rated core flow. For CPPU, the Option III trip-enabled region is rescaled to maintain the same absolute power/flow region boundaries. Because the 2-3

NEDO-33076 rated core flow is not changed, the 60% core flow boundary is not rescaled. The 30% CLTP region boundary changes by the following equation:

CPPU Region Boundary = 30% CLTP * (100% CPPU (% CLTP))

Thus for a CPPU of 115% of CLTP:

CPPU Region boundary = 30% CLTP * (100% . 115%) = 26.1% CPPU.

The CPPU stability OPRM trip enable region boundaries for Option III are shown in Figure 2-2 for Hope Creek Generating Station. The BSP analysis validates the adequacy of the OPRM Armed Region boundaries for the equilibrium GE14 core. The OPRM setpoint is evaluated for each reload core.

The OPRM system can only cause a scram when the plant is operating in the Option III Trip Enabled Region. The Trip Enabled Region will be defined in plant procedures and will be incorporated on the Hope Creek power/flow operating map. The Trip Enabled Region was modified for power uprate operation to maintain the pre-uprate absolute power and flow coordinates. The stability based Operating Limit Minimum Critical Power Ratio (OLMCPR) associated with the OPRM setpoint assures that the Critical Power Ratio (CPR) safety limit is not violated following an instability event. This is to be validated for every reload cycle.

2.5 REACTIVITY CONTROL The Control Rod Drive (CRD) system is used to control core reactivity by positioning neutron absorbing control rods within the reactor and to scram the reactor by rapidly inserting withdrawn control rods into the core. No change is made to the control rods due to the CPPU. The effect on the nuclear characteristics of the fuel is discussed in Section 2.3. The topics addressed in this evaluation for Hope Creek are:

Topic CLTR Disposition Hope Creek Result 2.5.1 Scram Time Response [

2.5.2 CRD Positioning 2.5.2 CRD Cooling 2.5.3 CRD Integrity ))

24

NEDO-33076 2.5.1 Control Rod Scram For pre-BWR/6 plants, the scram times are decreased by the transient pressure response, ((

)) At normal operating conditions, the accumulator supplies the initial scram pressure and, as the scram continues, the reactor becomes the primary source of pressure to complete the scram. ((

))

2.5.2 Control Rod Drive Positioning and Cooling

((

N' - '

2-5

NEDO-33076

)), and the automatic operation of the system flow control valve maintains the required drive water pressure and cooling water flow rate. Therefore, the CRD positioning and cooling functions are not affected. The CRD cooling and normal CRD positioning functions are operational considerations, not safety-related functions, and are not affected by CPPU operating conditions.

2-6

NEDO-33076 Plant operating data has confirmed that the CRD system flow control valve operating position has sufficient operating margin. ((

53 2.5.3 Control Rod Drive Integrity Assessment The postulated abnormal operating condition for the CRD design assumes a failure of the CRD system pressure-regulating valve that applies the maximum pump discharge pressure to the CRD mechanism internal components. This postulated abnormal pressure bounds the ASME reactor overpressure limit. ((

))

Other mechanical loadings are addressed in Section 3.3.2 of this report.

))

2-7

NEDO-33076 Core Flow (NMlb/hr) 0 10 20 30 40 50 60 70 80 90 100 110 120 120

• 4500 A: Natural Circulation 110 B: Minimum Pump Speed C: 57.6% Power/ 39.24 Flow MELLLA Boundary Line D: 100.0% Power/ 94.8% Flow D E F 4000 10O D': 87.0% Power/ 76.6% Flow 3840 MWt El 100.0% Power/ 100.0% Flow El: 87.0% Power/ 100.0% Flow 90 F: 100.0% Power/ 105.0% Flow 3500 F': 87.0% Power/ 105.0% Flow F' G: 20.6% Power/ 105.0% Flow 3339 MWIt I- H: 20.6% Power/ 100.0% Flow g 80 I: 20.6% Power/ 41.3% Flow 3000 >

J: 52.9% Power/ 33.7% Flow

. 70 ODYSY Based Scram Regi( onI i Ai-ICA' -2500 W C4 60 Boundary Line . C 0 Alt Region 11Boundaty Line Cs E 50 Base nSP Scram Region I 2000 E

.0: Boundary Line ODYSY Based Controlled Entry Bl-iCA Aj Region 11Boundary Line

[ 40 1500

• B2  ! E-U 30 n2.ICA f  : Increased Core Flow Region 1000 20 A .' I Cavitation Interlock . I G

, , 500 100% EPU - 3840 M~t 10 A. I 100% CUTP - 3339 M~t 100% Core Flow - 100.0 Mlb/hr 0 0 .- ,X .

, . _.--, . .. ~.I. .

--. . . . I .. ji.. . . . . . . . . ,,,.

0 0 10 20 30 40 50 60 70 80 90 100 110 120 Rated Core Flow (%)

Figure 2-1 Hope Creek Validation of Base BSP Regions for CPPU 2-8

NEDO-33076 Core Flow (Mi1b/lr) 0 10 20 30 40 50 60 70 80 90 100 110 120 12.^

500 A: Natural Circulation 0 B: Minimum Pump Speed C: 57.6' Power/ 39.29 Flow D: 100.01 Power/ 94.8eF low MELLLA3undaryLine D E 000 ICD D': 87.09 Power/ 76.69 Flow - ,

_ 3840MWA E: 100.01 Power/ 100.01 Flow E': 87.0% Power/ 100.0 Flow F. 100.0% Power/ 105.0% Flow .. 86.95%EPUJRodUne. - E- p - 500 I- F': 87.0t Power/ 105.0t Flow (t00)O%CLTPRL) 3339MW-Gt 20.69 Power/ 105.09 Flow- - - - - - 33 MW

-09 Hi 20.69 Power/ 100.09 Flow S0 I: 20.69 Power/ 41.39 Flow

o. . __ .. - . . . 3 J: 52.9% Power/ 33.79 Flow :3 000 3 07 0

-2 5v00 c. L.*

WC6 0 er 50 - --- -- - - -- 014 0 j 2000 C1

'e. 5 OPRM Tuip Enabled Incrcased Core Flow Region I Repon (Below 60%

2 40 - -  : Rated Core Flow, .

500 '

W eb Above 26.1 % Rated

= 330  :

0 power) t I ej 1000 2 0 . . Cavitation Interlock A.- . E -3840 MWt - 00

. . Bo 0 0o CLTP - 3339 MWt 1009 Core Flow - 100.0 M1b/hr

,0 j....

..9".,

a*-'-,

43,_,,,,

.l-O

.;I 0 10 20 30 40 50 60 70 80 90 100 110 120 Rated Core Flow (%)

Figure 2-2H1opc Creek OPRM Trip Enabled Region for CPPU 2-9

NEDO-33076

3. REACTOR COOLANT AND CONNECTED SYSTEMS This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapter 5, and to a very limited extent Chapter 3, that applies to CPPU.

3.1 NUCLEAR SYSTEM PRESSURE RELIEF/OVERPRESSURE PROTECTION The nuclear system pressure relief system topics addressed in this evaluation are as follows:

Topic CLTRDisposition Hope CreekResult Overpressure capacity Flow-induced vibration The nuclear system pressure relief system prevents overpressurization of the nuclear system during AOOs, the plant ASME Upset overpressure protection event, and postulated anticipated transient without scram (ATWS) events. The plant safety relief valves (SRV) along with other functions provide this protection. An evaluation was performed in order to confirm the adequacy of the pressure relief system for CPPU conditions. The adequacy of the pressure relief system is also demonstrated by the overpressure protection evaluation performed for each reload core and by the ATWS evaluation performed for CPPU (Section 9.3.1).

For Hope Creek, no SRV setpoint increase is needed because there is no change in the dome pressure or simmer margin. Therefore, there is no effect on valve functionality (opening/closing).

Two potentially limiting overpressure protection events are analyzed for CPPU: (1) Main Steam Isolation Valve Closure with Scram on High Flux (MSIVF) and (2) Turbine Trip with Bypass Failure and Scram on High Flux (Reference 2, Section 5.5.1.4). However, based on both plant initial core analyses and subsequent power uprate evaluations, the MSIVF is more limiting than the turbine trip event with respect to reactor overpressure. Recent extended power uprate evaluations show a 24 to 40 psi difference between these two events. Only the MSIVF event was performed because it is limiting. In addition, an evaluation of the MSIVF event is performed with each reload analysis.

3-1

NEDO-33076 The design pressure of the reactor vessel and reactor coolant pressure boundary (RCPB) remains at 1250 psig. The acceptance limit for pressurization events is the ASME code allowable peak pressure of 1375 psig (110% of design value). The overpressure protection analysis description and analysis method are provided in Reference 2. The MSIVF event is conservatively analyzed assuming a failure of the valve position scram. The analyses also assume that the event initiates at a reactor dome pressure of 1035 psia (which is higher than the nominal CPPU dome pressure), and one SRV out-of-service (OOS). Starting from 102% of CPPU RTP, the calculated peak reactor pressure vessel (RPV) pressure, located at the bottom of the vessel, is 1285 psig. The corresponding calculated maximum reactor dome pressure is 1265 psig. The peak calculated RPV pressure remains below the 1375 psig ASME limit, and the maximum calculated dome pressure remains below the Technical Specification 1325 psig Safety Limit. Therefore, the results are acceptable and within allowable values. The results of the CPPU overpressure protection analysis for the Hope Creek MSIVF event are consistent with the generic analysis in Reference 3. The Hope Creek response to the MSIVF event is provided as Figure 3-1.

SRV setpoint tolerance is independent of CPPU. CPPU evaluations are performed using the existing SRV setpoint tolerance analytical limit of 3% as a basis. Actual historical in-service surveillance of SRV setpoint performance test results are monitored separately for compliance to the Technical Specification requirements. Hope Creek has an ongoing evaluation program to resolve problems resulting in SRV surveillance testing exceeding the 3% tolerance.

FIV may increase incidents of valve leakage. Hope Creek has established administrative limits and actions for addressing a leaking SRV. In addition, operating procedures provide operators with immediate response actions for inadvertent SRV opening and stuck open SRV conditions.

The consequences of a stuck open SRV have been previously considered in the plant specific safety analyses and have been demonstrated to be non-limiting.

Increased main steam line flow may affect FIV of the piping and safety/relief valves during normal operation. The vibration frequency, extent, and magnitude depend upon plant specific parameters, valve locations, the valve design, and piping support arrangements. The FIV of the piping will be addressed by vibration testing during initial plant operation at the higher steam flow rates (see Sections 3.4.1 and 10.4).

))

3-2

NEDO-33076 3.2 REACTOR VESSEL The reactor pressure vessel (RPV) structure and support components form a pressure boundary to contain the reactor coolant and moderator, and form a boundary against leakage of radioactive materials into the drywell. The RPV also provides structural support for the reactor core and internals. The topics addressed in this evaluation are:

-Topic -lCLTR Disposition Hop reekResult 3.2.1 Fracture Toughness 3.2.2 Reactor Vessel Structural Evaluation (Components not significantly affected) 3.2.2 Reactor Vessel Structural Evaluation (Affected components) 3.2.1 Fracture Touahness The CLTR, Section 3.2.1 describes the reactor pressure vessel (RPV) fracture toughness evaluation process. RPV embrittlement is caused by neutron exposure of the wall adjacent to the core (the "beltline" region). Operation at the CPPU conditions results in a higher neutron flux, which increases the integrated fluence over the period of plant life.

The neutron fluence is reanalyzed for CPPU, based on flux calculated using 2-dimensional neutron transport theory (Reference 8); the neutron transport methodology is consistent with Regulatory Guide 1.190. The revised fluence is used to evaluate the vessel against the requirements of 10CFR50, Appendix G. The results of these evaluations indicate that:

a) The 32 effective full power year (EFPY) Reference Temperature shift is increased, and consequently, requires a change in the adjusted reference temperature, which is the initial RTNDT plus the shift. These values are provided in Table 3-1.

b) The upper shelf energy remains greater than 50 ft-lb for the design life of the vessel and maintains the requirements of Appendix G. The minimum upper shelf energy for beltline materials is 60 ft-lb. Detailed results for the upper shelf energy evaluation are provided in Table 3-2.

c) The beltline material reference temperature of the nil-ductility transition (RTNmT) remains below 2000 F.

d) The Pressure-Temperature (P-T) curves have been revised for CPPU conditions including the Adjusted Reference Temperature of Table 3-1. Amendment No. 157 to the Hope Creek Facility Operating License was issued on November 1, 2004. This amendment changed the Technical Specification P-T curves and included the CPPU conditions.

3-3

NEDO-33076 e) The surveillance program consists of three capsules. One capsule containing Charpy specimens was removed from the vessel after 6.01 EFPY of operation (end of Fuel Cycle 5) and tested. The remaining two capsules have been in the reactor vessel since plant startup. Hope Creek is participating in the BWRVIP Integrated Surveillance Program (ISP) and will comply with the withdrawal schedule specified by this program. The ISP specifies that Hope Creek shall remove the next capsule when it has fluence equal to the RPV 1/4T end of license fluence. This schedule is not changed by CPPU. With CLTP conditions, the ISP estimated this fluence to occur at 22 EFPY. With CPPU conditions, this is estimated to occur at approximately 23 EFPY.

The maximum normal operating dome pressure for CPPU is unchanged. Therefore, the hydrostatic and leakage test pressures are acceptable for the CPPU. Because the vessel remains in compliance with the regulatory requirements, operation with CPPU does not have an adverse effect on the reactor vessel fracture toughness.

3.2.2 Reactor Vessel Structural Evaluation 34

NEDO-33076 1))

The effect of CPPU was evaluated to ensure that the reactor vessel components continue to comply with the existing structural requirements of the ASME Boiler and Pressure Vessel Code. For the components under consideration, the 1968 code with addenda to and including winter 1969, which is the code of construction, is used as the governing code. However, if a component's design has been modified, the governing code for that component is the code used in the stress analysis of the modified component. The Hope Creek CPPU utilizes the original code of construction as the governing code for all components for CPPU conditions. New stresses are determined by scaling the "original" stresses based on the CPPU conditions (temperature and flow). The analyses were performed for the design, the normal and upset, and the emergency and faulted conditions. If there is an increase in annulus pressurization, jet reaction, pipe restraint or fuel lift loads, the changes are considered in the analysis of the components affected for normal, upset, emergency and faulted conditions.

3.2.2.1 Design Conditions Because there are no changes in the design conditions (i.e. temperature, pressure or flow) due to CPPU, the design stresses are unchanged and the Code requirements are met.

3.2.2.2 Normal and Upset Conditions The reactor coolant temperature and flows (except core flow) at CPPU conditions are only slightly changed from those at current rated conditions. A primary plus secondary stress analysis was performed to demonstrate that the stresses at the CPPU conditions continue to meet the requirements of the ASME Code. The evaluations were performed at conditions that bound the slight change in operating conditions. The fatigue usage was evaluated for the limiting location of components with a usage factor greater than 0.5. The analysis results for CPPU show that all components meet their ASME Code requirements. The stress and fatigue analysis results are provided in Table 3-3.

3-5

NEDO-33076 3.2.2.3 Emergency and Faulted Conditions The stresses due to Emergency and Faulted conditions are based on loads such as peak dome pressure, which are unchanged. Therefore, the Code requirements are met for all RPV components.

3.3 REACTOR INTERNALS The reactor internals include core support structure (CSS) and non-core support structure (non-CSS) components. The topics considered in this section are:

'Topic - CLTR Disposition Hope Creek'Resultl 3.3.1 Reactor Internals Pressure Differences 3.3.2 Reactor Internals Structural Evaluation 3.3.3 Steam Dryer Separator Performance 3.3.1 Reactor Internal Pressure Differences The increase in core average power alone would result in higher core loads and RIPDs due to the higher core exit steam quality. The maximum acoustic and flow-induced loads, following a postulated recirculation line break, were shown to be unaffected by the CPPU.

The RIPDs are calculated for Normal (steady-state operation), Upset, Faulted, and Emergency conditions for all major reactor internal components. ((

))

Tables 3-4 through 3-7 compare results for the various loading conditions between CLTP and CPPU for the vessel internals that are affected by the changed RIPDs.

3.3.2 Reactor Internals Structural Evaluation The reactor internals consist of the CSS components and non-CSS components. The reactor internals (excluding CRD) are not certified to the ASME code; however, the requirements of the code are used as guidelines in their design basis analysis. The evaluations and stress reconciliation in support of the thermal power increase are performed consistent with the design basis analysis of the components. The reactor internal components evaluated are:

Core Support Structure Components

• Shroud

• Shroud Support 3-6

NEDO-33076

• Core Plate

• Top Guide

• CRD and CRD Housing

• Control Rod Guide Tube

• Orificed Fuel Support

• Fuel Channel Non-Core Support Structure Components

• Steam Dryer

• Feedwater Sparger

• Jet Pumps

• Core Spray Line and Sparger

• Access Hole Cover

• Shroud Head and Steam Separator Assembly

• LPCI Coupling

• Core Differential Pressure and Standby Liquid Control Line The original configurations of the internal components are considered in the CPPU evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation.

The effects on the loads as a result of the thermal-hydraulic changes due to CPPU are evaluated for the reactor internals. All applicable loads and load combinations are considered consistent with the existing design basis analysis. These loads include the RIPDs, seismic loads, annulus pressurization (AP) and jet loads, flow induced and acoustic loads due to Recirculation Suction Line Break Loss-of-Coolant Accident (RSLB LOCA), and thermal loads. The RIPDs increase for some components/loading conditions as a result of CPPU. The flow conditions and thermal effects were considered in the evaluation, as applicable. The seismic response is unaffected by CPPU. As part of the GE14 Fuel transition, it was determined that the change to GE14 fuel had insignificant effects on the seismic and dynamic loads, and in turn, the structural integrity of the RPV internals. The acoustic and flow induced loads in the annulus as a result of the RSLB-LOCA are included in the evaluation and are bounded by CLTP values. The original plant design basis AP and jet loads remain bounded for CPPU.

3-7

NEDO-33076 The CPPU loads are compared to those in the existing design basis analysis. If the loads do not increase due to CPPU, then the existing analysis results bound the CPPU conditions, and no further evaluation is required or performed. If the loads increase due to the CPPU, then the effect of the load increase is evaluated further. ((

1))

Tables 3-8 presents the governing stresses for the reactor internals, which were quantitatively assessed. All stresses are within allowable limits and the reactor internal components are demonstrated to be structurally adequate for CPPU. The results of the qualitative assessment of the remaining internals are also presented.

The following reactor vessel internals are evaluated for the effects of changes in loads due to CPPU.

3-8

NEDO-33076 3-9

NEDO-33076 3.3.3 Steam Dryer/Separator Performance At Hope Creek, the performance of the steam separators and dryer has been evaluated to ensure that the quality of the steam leaving the RPV remains acceptable at CPPU conditions. CPPU results in an increase in saturated steam generated in the reactor core. For constant core flow, this in turn results in an increase in the separator inlet quality and dryer face velocity and a decrease in the water level inside the dryer skirt. These factors, in addition to the core radial power distribution affect the steam separator-dryer performance. Steam separator-dryer performance was evaluated at CPPU equilibrium cycle limiting conditions of high radial power peaking and the applicable core flow range shown on the power-flow map (Figure 1-1). The predicted steam moisture content is acceptable based on a revised moisture content performance specification of < 0.3 weight %.

The operation and performance of the NSSS components with moisture content up to the revised performance specification was determined to be acceptable. The effect of increasing steam 3-10

NEDO-33076 moisture content on the radiation source terms is addressed in Section 8.0. Steam dryer/separator performance operational testing is included in the CPPU implementation plan as described in Section 10.4 to ensure adequate operating limitations are implemented as required.

3.4 FLOW INDUCED VIBRATION The flow-induced vibration (FIV) evaluation addresses the influence of an increase in flow during CPPU on reactor coolant pressure boundary (RCPB) piping, RCPB piping components and RPV internals. The topics addressed in this evaluation are:

--Topic - CLTR Disposition -HopeCreekResult 3.4.1 Structural Evaluation of Recirculation Piping 3.4.1 Structural Evaluation of Main Steam and Feedwater Piping 3.4.1 Safety-Related Thermowells and Probes 3.4.2 Structural Evaluation of core flow dependent RPV Internals 3.4.2 Structural Evaluation of other RPV Internals

((i 3-11

NEDO-33076 3.4.1 FIV Influence on Piping Key applicable structures include the main steam (MS) system piping and suspension, the feedwater (FW) system piping and suspension, and the Reactor Recirculation System (RRS) piping and suspension. In addition, branch lines attached to the MS system piping or FW system piping are considered.

RRS drive flow is not significantly increased (< 5%) during CPPU operation. ((

The MS and FW piping have increased flow rates and flow velocities in order to accommodate CPPU. As a result, the MS and FW piping experience increased vibration levels, approximately proportional to the square of the flow velocities. The ASME Code and nuclear regulatory guidelines require some vibration test data be taken and evaluated for these high-energy piping systems during initial operation at CPPU conditions. Vibration data for the MS and FW piping inside containment will be acquired using remote sensors, such as displacement probes, velocity sensors, and accelerometers. A piping vibration startup test program, consistent with the ASME code and regulatory requirements, will be performed.

)) and FIV testing of the MS and FW piping system will be performed during CPPU power ascension.

The safety-related thermowells in the MS piping systems were evaluated and found to be adequate for the increased main steam flow as a result of CPPU. There are no safety-related thermowells or sample probes in the FW piping systems at Hope Creek. Non-safety related sample probes in the FW piping will be evaluated for vulnerability to failures that could result in loose parts at CPPU conditions.

3.4.2 FIV Influence on Reactor Internal Components 3-12

NEDO-33076 1]

The required reactor vessel internals vibration assessment of other RPV internals is described in the CLTR. CPPU operation increases the steam production in the core, resulting in an increase in the core pressure drop. There is a slight increase (3.4%) in maximum drive flow at CPPU conditions as compared to CLTP. The increase in power may increase the level of reactor internals vibration.

Analyses were performed to evaluate the effects of flow-induced vibration on the reactor internals at CPPU conditions. This evaluation used a reactor power of 3952 MWt and 105% of rated core flow.

This assessment was based on vibration data obtained during startup testing of the prototype plant (Browns Ferry Unit 1). For components requiring an evaluation but not instrumented in the prototype plant, vibration data acquired during the startup testing from similar plants or acquired outside the RPV is used. The expected vibration levels for CPPU region were estimated by extrapolating the vibration data recorded in the prototype plant or similar plants and on GE Nuclear Energy BWR operating experience. These expected vibration levels were then compared with the established vibration acceptance limits. The following components were evaluated:

a) Shroud b) Shroud head and moisture separator c) Jet pumps d) Feedwater sparger e) Steam dryer f) Jet pump sensing lines The results of the vibration evaluation show that continuous operation at a reactor power of up to 3952 MWt and 105% of rated core flow will not result in any detrimental effects on the evaluated reactor internal components (except the steam dryer).

))

3-13

NEDO-33076 During CPPU, the components in the upper zone of the reactor, such as the moisture separators and dryer, are mostly affected by the increased steam flow. Components in the core region and components such as the core spray line are primarily affected by the core flow. Components in the annulus region such as the jet pump are primarily affected by the recirculation pump drive flow and core flow. Because there is a slight increase (3.4%) in maximum drive flow with core flow remaining the same as compared to the CLTP condition, a small increase in FIV on the components in the annular and core regions are expected. However, the moisture separator and dryer are significantly affected by CPPU conditions.

The steam dryer and moisture separators are non-safety-related components. Recent uprate experience indicates that FIV at CPPU conditions may lead to high cycle fatigue failure of some dryer components.

A detailed evaluation will be performed to examine dryer components susceptible to failure at CPPU conditions. The results of the quantitative evaluation will be used to identify any additional modifications needed to maintain steam dryer structural integrity at CPPU conditions.

Any identified dryer modifications will be performed prior to CPPU implementation.

The calculations for CPPU conditions indicate that vibrations of all safety related reactor internal components are within the GE acceptance criteria. The analysis is conservative for the following reasons:

• The GE criteria of 10,000 psi peak stress intensity is less than the ASME Code criteria of 13,600 psi;

• The modes are absolute summed; and

• The maximum vibration amplitude in each mode is used in the absolute sum process, whereas in reality the peak vibration amplitudes are unlikely to occur at the same time.

Based on the above, it is concluded that FIV effects are expected to remain within acceptable limits at CPPU conditions.

3.5 PIPING EVALUATION 3.5.1 Reactor Coolant Pressure Boundary Piping The RCPB piping systems evaluation consists of a number of safety related piping subsystems that move fluid through the reactor and other safety systems. The topics addressed in this' evaluation are:

--I -- Topic  ;-. .-  ;- - C-LTRDisposition Hope sult Structural evaluation for unaffected safety [

related piping Structural evaluation for affected safety ]

related piping 3-14

NEDO-33076 As described in the CLTR, for most of the piping systems, the flow, pressure, temperature and mechanical loading do not increase for CPPU. ((

))

3-15

NEDO-33076 3-16

NEDO-33076 3-17

NEDO-33076 Section 3.1 demonstrates that the RCPB piping remains below the ASME pressure limit during the most severe pressurization transient.

The safety related thermowells and probes in the main steam line (MSL) and FW piping systems were evaluated for Hope Creek (see Section 3.4.1).

Main Steam and Associated Piping System Evaluation The MS piping system and associated branch piping (inside containment) was evaluated for compliance with currently licensed Section III of the ASME Boiler & Pressure Vessel Code, (as applicable). Included in the evaluation were the effects of CPPU on piping stresses, cumulative fatigue usage, piping supports including their components, RPV nozzles, penetrations, flanges and valves and for the effects of thermal expansion displacements on the piping snubbers, hangers and struts.

((I

))The SRV discharge loads are not affected because the SRV set pressures are not changed, therefore, there are no effects on the main steam and SRV piping design for this load. However, the increase in MS flow results in increased forces and moments from the turbine stop valve (TSV) closure transient. TSV load was used in the design of MS piping system. The MSIV closure 3-18

NEDO-33076 loads remains bounded by TSV loads, as the MSIV closure time is significantly longer than the turbine stop valve closure time.

Pipe Stresses A review of the changes in flow, pressure and temperature associated with CPPU indicates that piping load changes do not result in load limits being exceeded for the MS piping system and attached branch piping or for RPV nozzles. The original design analyses have sufficient margin between calculated stresses and ASME Boiler & Pressure Vessel Code allowable limits to justify operation at CPPU conditions. The pressure and temperature of the MS piping are unchanged for the CPPU. No new postulated pipe break locations were identified.

Similarly, the branch pipelines (SRVDL, RCIC, HPCI, RPV Vent connected to the MS pipe, and MSIV Drain) were evaluated to determine the effect of the increased MS flow on the lines. This evaluation concluded that the original design analyses have sufficient margin between calculated stresses and the allowable limits of Section III ASME Boiler & Pressure Vessel Code to justify operation at CPPU conditions. As with the MS piping, the pressures and temperatures for the associated branch piping systems do not change due to CPPU.

Pipe Supports The pipe supports for the MS piping system were evaluated for the increased loading associated with the turbine stop valve closure transient at CPPU conditions. The evaluations showed that the supports have adequate design margin to accommodate the increased loads.

The MS piping system was evaluated for the effects of thermal expansion displacements and concluded that they are within allowable values.

Feedwater Evaluation The FW system (inside containment) was evaluated for compliance with currently licensed Section III of the ASME Boiler & Pressure Vessel Code, and for the effect of thermal expansion displacements on the piping snubbers, hangers and struts. Piping interfaces with RPV nozzles, penetrations, flanges and valves were also evaluated.

Pipe Stresses A review of the changes in pressure, temperature and flow associated with CPPU indicates that piping load changes do not result in load limits being exceeded for the FW piping system or for RPV nozzles. The original design analyses have sufficient design margin between calculated stresses and the ASME Boiler and Pressure Vessel Code allowable limits to justify operation at CPPU conditions.

The design adequacy evaluation shows that the requirements of the ASME Boiler and Pressure Vessel Code requirements remain satisfied. Therefore, CPPU does not have an adverse effect on the FW piping design. A review of postulated pipe break criteria concluded that at three locations, cumulative fatigue usage exceeds postulated pipe break criteria limit. Reanalysis will be performed for these locations and structural modifications will be completed where required to ensure that 3-19

NEDO-33076 ASME Code stresses and fatigue usage factors will not exceed the criteria limit, prior to the implementation of CPPU.

Pipe Supports The pipe supports for the FW piping system were evaluated for the increased loading associated with the changes in pressure, temperature and flow at CPPU conditions. The evaluations showed that the supports have adequate design margin to accommodate the increased supports loads.

The FW piping system was evaluated for the effects of thermal expansion displacements and concluded that they are within allowable.

Other RCPB Piping Evaluation This section addresses the adequacy of the other RCPB piping designs, for operation at the CPPU conditions. The nominal operating pressure and temperature of the reactor are not changed by CPPU. Aside from MS and FW, no other system connected to the RCPB experiences a significant increased flow rate at CPPU conditions. Only minor changes to fluid conditions are experienced by these systems due to higher steam and feedwater flow from the reactor and the subsequent change in fluid conditions within the reactor. Additionally, dynamic piping loads for safety relief valves (SRV) lines at CPPU conditions are bounded by those used in the existing analyses. These effects have been evaluated for the RCPB portion of the RPV head vent line, SRV discharge piping, and MSIV drain piping as required.

These systems were evaluated for compliance with the ANSI B31.1 or ASME Code stress criteria (as applicable). Because none of these piping systems experience any significant change in operating conditions, they are all acceptable as currently designed.

3.5.2 Balance-Of-Plant Piping The Balance-of-Plant Piping (BOP) systems evaluation consists of a number of piping subsystems that move fluid through systems outside the RCPB piping. The topics addressed in this evaluation are:

Topic - CLTR Dispositio ope reekResult Structural evaluation for unaffected non-safety related piping Structural evaluation for affected non-safety related piping For some BOP piping systems, the flow, pressure, temperature, and mechanical loads do not increase. ((

3-20

NEDO-33076 3-21

NEDO-33076 1]

Large bore and small bore ASME Class 1, 2, and 3 piping and supports not addressed in Section 3.5.1 were evaluated for acceptability at CPPU conditions, and shown to be adequate as currently designed. The evaluation of the BOP piping and supports was performed in a manner similar to the evaluation of RCPB piping systems and supports (Section 3.5.1), using applicable ASME Section III, Subsections NB/NC/ND or B31.1 Power Piping Code equations. The original Codes of record (as referenced in the appropriate calculations), Code allowables, and analytical techniques were used and no new assumptions were introduced.

The Design Basis Accident (DBA)-LOCA dynamic loads, including the pool swell loads vent thrust loads, condensation oscillation (CO) loads and chugging loads were originally defined and evaluated for Hope Creek. The evaluation of the structures attached to the torus shell, such as piping system, vent penetrations, and valves are based on these DBA-LOCA hydrodynamic loads. For CPPU conditions, the DBA-LOCA torus shell response loads were re-evaluated (Section 4.1) and found acceptable and there are no resulting effects on the torus shell attached structures.

The effects of the CPPU conditions have been evaluated for the following piping systems:

• MS (outside containment) including Turbine Bypass Piping

• MSIV Drain Lines

• Main Steam Relief Valves

• Extraction Steam, Heater Vents and Drains

• FWF and Condensate

• RWCU - Outside Containment

• RHR - Outside Containment

• CS - Outside Containment

• HPCI - Outside Containment

• RCIC - Outside Containment

• SLC - Outside Containment

• RACS

• SACS

• TACS 3-22

NEDO-33076

• Fuel Pool Cooling and Clean-Up

• SRV Quenchers and Supports

• Filtration, Recirculation, and Ventilation System (FRVS)

• Off Gas

• Torus Attached Piping including ECCS Suction Strainers

• Control Rod Drive Pipe Stresses Operation at the CPPU conditions increases stresses on piping and piping system components due to slightly higher operating temperatures and flow rates internal to the pipes. For all systems, the maximum stress levels and fatigue analysis results were reviewed based on specific increases in temperature, pressure, and flow rate (see Tables 3-9 and 3-10). These piping systems have been evaluated and found to meet the appropriate code criteria for the CPPU conditions, based on the design margins between actual stresses and code limits in the original design. All piping is below the code allowables of the plant code of record, ASME B&PV Code, Div. 1,Section III, 1977 Edition through Summer 1979 Addenda for Class 1 piping and ASME BP&8V Code - Section III, Division I, 1974 Edition, through winter 1974 Addenda for Class 2 and 3 piping. No new postulated pipe break locations were identified.

Pipe Supports Operation at the CPPU conditions slightly increases the pipe support loadings due to increases in the temperature of the affected piping systems (see Tables 3-9 and 3.10).

The pipe supports of the systems affected by CPPU loading increases (MS, FW, Extraction Steam, Drains, Vent systems) were reviewed to determine if there is sufficient margin to code acceptance criteria to accommodate the increased loadings. This review shows that there is adequate design margin between the original design stresses and code limits of most of the supports (with the exception of Main Steam Outside of Containment - see below) to accommodate the load increase. The original design analyses have sufficient design margin to justify operation at the CPPU conditions.

Main Steam and Associated Piping System Evaluation (Outside containment)

The MS piping system (outside containment) was evaluated for compliance with Hope Creek criteria. Included in the evaluation were the affects of CPPU on piping stresses, piping supports and the associated building structure, turbine nozzles, and valves.

Because the MS piping pressures and temperatures outside containment are not affected by CPPU, there was no effect on the analyses for these parameters. The increase in MS flow results in increased forces from the turbine stop valve closure transient. The turbine stop valve closure loads bound the MSIV valve loads because the MSIV closure time is significantly longer than the turbine stop valve closure time. The MS analysis results are provided in Tables 3-9 and 3-10.

3-23

NEDO-33076 Pipe Stresses A review of the increase in flow associated with CPPU indicates that piping load changes do not result in load limits being exceeded for the main steam piping system outside containment. The original design analyses have sufficient design margin to justify operation at the CPPU conditions. The pressure and temperature of the MS piping is unchanged for CPPU.

Pipe Supports The pipe supports (primarily spring type supports) and turbine nozzles for the MS piping system outside containment were evaluated for the increased loading and movements associated with the turbine stop valve closure transient at CPPU conditions. This review shows that in most cases there is adequate design margin between the original design stresses and code limits of the supports and nozzles to accommodate the load increase. However, six pipe supports on the Main Steam System (Outside Containment) require modification, prior to CPPU implementation, in order to meet original code limits.

3.6 REACTOR RECIRCULATION SYSTEM The Reactor Recirculation System evaluation for CPPU addressed the following topics:

Topic -- °CLTR Disposition Hope Creek Result System evaluation NPSH Flow mismatch Single loop operation 3-24

NEDO-33076 The CPPU power condition is accomplished by operating along extensions of current rod lines on the power/flow map with no increase in the maximum core flow. The core reload analyses are performed with the most conservative allowable core flow. The evaluation of the reactor recirculation system performance at CPPU power demonstrates that adequate core flow can be maintained. The evaluation is based on as-built/clean design conditions and not on actual plant operating capability (aging, degradation, vibration issues, etc.) associated with actual plant capability.

The cavitation protection interlock remains the same in terms of absolute feedwater flow rates.

This interlock is based on subcooling in the external recirculation loop and thus is a function of absolute feedwater flow rate and feedwater temperature at less than full thermal power operating conditions. Therefore, the interlock is not changed by CPPU.

3-25

NEDO-33076 1]

SLO operation is limited to off rated condition and is not changed as a result of the CPPU.

3.7 MAIN STEAM LINE FLOW RESTRICTORS The main steam line flow restrictor evaluation for CPPU at Hope Creek addressed the following topics:

Topic CLTR Disposition Hope Creek Result Structural integrity

((

-:_ _ _ _ __ _ _ _ _ _ : _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1:1 v;sn-,i

- i.i

i!:......

3-26

NEDO-33076 Because the main steam line flow restrictors are stainless steel, the increase in steam flow rate has no significant effect on flow restrictor erosion. There is no effect on the structural integrity of the main steam flow element (restrictor) due to the increased differential pressure because the restrictors were designed and analyzed for the choked flow condition.

After a postulated steam line break outside containment, the fluid flow in the broken steam line increases until it is limited by the main steam line flow restrictor. ((

)) The Hope Creek restrictors were originally analyzed for these flow conditions and therefore the restrictors remain within the acceptable calculated differential pressure drop and choke flow limits under CPPU conditions.

3.8 MAIN STEAM ISOLATION VALVES The MSIVs evaluation for CPPU at Hope Creek addressed the following topics:

- Topic -CLTR Disposition Hope CreekResult Isolation performance Valve pressure drop ))

The MSIVs are part of the RCPB, and perform the safety function of steam line isolation during certain abnormal events. The MSIVs must be able to close within a specified time range at all design and operating conditions. They are designed to satisfy leakage limits set forth in the plant Technical Specifications. Hope Creek does not have an MSIV leakage control system.

The MSIVs have been evaluated, as discussed in Section 4.7 of Reference 3. The evaluation covers both the effects of the changes to the structural capability of the MSIV to meet pressure boundary requirements, and the potential effects of CPPU-related changes to the safety functions of the MSIVs. The generic evaluation from Reference 3 is based on (1) a 20% thermal power increase, (2) an increased operating dome pressure to 1095 psia, (3) a reactor temperature increase to 5560 F, and (4) steam and feedwater flow increases of about 24%. The evaluation from Reference 3 is confirmed applicable to Hope Creek. An increase in flow rate assists MSIV closure. The Hope Creek MSIVs have design features that ensure the MSIV closure times are not reduced below the stroke time limit. The closing time of the MSIVs is controlled by the design of the hydraulic control valves and the function of the damper.

The hydraulic damper senses the combined driving force of the pneumatic cylinder, the external closing springs, the steam drag force, the dead weight of the moving components, and the friction force. The steam drag force applied on the main disc increases due to an increase in steam flowrate. This force change is transmitted from the main disc to the valve stem, and then to the connecting hydraulic damper rod. It is then transmitted to the hydraulic damper and the 3-27

NEDO-33076 hydraulic control circuit. As the driving force increases due to the higher steam flowrate, a spring inside the hydraulic control valve reduces the opening of an internal variable orifice in order to compensate for the higher closing force. The net driving force stays unchanged due to this compensating mechanism. The self-compensating feature of the hydraulic control valve will maintain the closing time with little deviation despite the flow rate change.

Therefore, CPPU described herein is bounded by conclusions of the evaluation in Section 4.7 of Reference 3, and the MSIVs are acceptable for CPPU operation.

3.9 REACTOR CORE ISOLATION COOLING/ISOLATION CONDENSER The Isolation Condenser is not applicable to Hope Creek.

The Reactor Core Isolation Cooling (RCIC) system evaluation for CPPU at Hope Creek addressed the following topics:

Topic -CLTR Disposition Mope CreekResult System performance and hardware Net positive suction head Adequate core cooling for limiting LOFW events Inventory makeup - Operational ]

Level 1 avoidance 3-28

NEDO-33076 The RCIC system is required to maintain sufficient water inventory in the reactor to permit adequate core cooling following a reactor vessel isolation event accompanied by loss of flow from the FW system. The system design injection rate must be sufficient for compliance with the system limiting criteria to maintain the reactor water level above top of active fuel (TAF) at the CPPU conditions. The RCIC system is designed to pump water into the reactor vessel over a wide range of operating pressures. As described in Section 9.1.1, this event is addressed on a plant specific basis. The results of the Hope Creek plant specific evaluation indicate adequate water level margin above TAF at the CPPU conditions. Therefore, the RCIC injection rate is adequate to meet this design basis event.

An operational requirement is that the RCIC system can restore the reactor water level while avoiding Automatic Depressurization System (ADS) timer initiation and MSIV closure activation functions associated with the low-low-low reactor water level setpoint (Level 1). This requirement is intended to avoid unnecessary initiations of safety systems. The results of the Hope Creek plant specific evaluation indicates that the RCIC system is capable of maintaining the water level outside the shroud above nominal Level 1 setpoint through a limiting LOFW 3-29

NEDO-33076 event at the CPPU conditions. Thus, the RCIC injection rate is adequate to meet the requirements for inventory makeup. (see Section 9.1.1)

For the CPPU, there is no change to the normal reactor operating pressure and the SRV setpoints remain the same. There is no change to the maximum specified reactor pressure for RCIC system operation, ((

)) there are no physical changes to the pump suction configuration, and no changes to the system flow rate or minimum atmospheric pressure in the suppression chamber or condensate storage tank (CST). CPPU does not affect the capability to transfer the RCIC pump suction on low CST level from its normal alignment, the CST, to the suppression pool, and does not change the existing requirements for the transfer. For ATWS (Section 9.3.1) and fire protection (Section 6.7), operation of the RCIC system at suppression pool temperatures greater than the operational limit (170 'F) may be accomplished by using the dedicated CST volume as the source of water. The CST provides the dedicated water source for greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at Hot Shutdown conditions with the RPV isolated. Therefore, the specified operational temperature limit for the process water does not change with the CPPU. ((

)) The effect of CPPU on the operation of the RCIC system during Station Blackout events is discussed in Section 9.3.2.

The reactor system response to a loss of feedwater transient with RCIC is discussed in Section 9.1.1.

3-30

NEDO-33076 3.10 RESIDUAL HEAT REMOVAL SYSTEM The Reactor Heat Removal (RHR) system evaluation for CPPU at Hope Creek addressed the following topics:

Topic CLTR Disposition Hope Creek Result LPCI mode Suppression pool and containment spray cooling modes Shutdown cooling mode Steam condensing mode Fuel pool cooling assist The RHR system is designed to restore and maintain the reactor coolant inventory following a LOCA and remove reactor decay heat following reactor shutdown for normal, transient, and accident conditions. The CPPU effect on the RHR system is a result of the higher decay heat in the core corresponding to the uprated power and the increased amount of reactor heat discharged into the containment during a LOCA. For Hope Creek, the RHR system is designed to operate in one of the Low Pressure Coolant Injection (LPCI), Shutdown Cooling (SDC), Suppression Pool Cooling (SPC), Containment Spray Cooling (CSC) and Fuel Pool Cooling Assist modes. The Steam Condensing Mode (SCM) of RHR is not installed at Hope Creek.

The LPCI mode, as it relates to the LOCA response, is discussed in Section 4.2.4.

The SPC mode is manually initiated following isolation transients and a postulated LOCA to maintain the containment pressure and suppression pool temperature within designi limits. The CSC mode reduces drywell pressure, drywell temperature, and suppression chamber pressure following an accident. The adequacy of these operating modes is demonstrated by the containment analysis (Section 4.1).

The higher suppression pool temperature and containment pressure during a postulated LOCA (Section 4.1) do not affect hardware capabilities of RHR equipment to perform the LPCI, SPC, and CSC functions.

The Fuel Pool Cooling Assist mode, using existing RHR heat removal capacity, provides supplemental fuel pool cooling capability in the event that the fuel pool heat load exceeds the heat removal capability of the Fuel Pool Cooling and Cleanup (FPCC) system. The adequacy of fuel pool cooling, including use of the Fuel Pool Cooling Assist mode, is addressed in Section 6.3.1.

3-31

NEDO-33076 Shutdown Cooling Mode Steam Condensing Mode This mode is not installed at Hope Creek Therefore, the RHR system at Hope Creek is acceptable for CPPU operation.

3.11 REACTOR WATER CLEANUP SYSTEM The Reactor Water Cleanup (RWCU) system evaluation for CPPU at Hope Creek addressed the following topics:

Topic -- CLTRDisposition - Hope Creek Result System performance Containment isolation RWCU system operation at the CPPU RTP level slightly decreases the temperature within the RWCU system. This system is designed to remove solid and dissolved impurities from recirculated reactor coolant, thereby reducing the concentration of radioactive and corrosive species in the reactor coolant. The system is capable of performing this function at the CPPU RTP level.

The CPPU review included evaluations of water chemistry, heat exchanger performance, pump performance, flow control valve capability and filter / demineralizer performance. All aspects of performance were found to be within the design of RWCU at the analyzed flows. The RWCU analysis concludes that:

• There is negligible heat load effect;

• A small increase in filter / demineralizer backwash frequency occurs, but this is within the capacity of the Radwaste system;

• The slight changes in operating system conditions result from a decrease in inlet temperature and increase in FW system operating pressure;

• The RWCU filter / demineralizer control valve operates in a slightly more open position to compensate for the increased FW pressure; 3-32

NEDO-33076

• No changes to instrumentation are required; setpoint changes are not expected due to the negligible system process parameter changes; and Based on operating experience, the FW iron input to the reactor increases as a result of the increased feedwater flow. This input increases the calculated reactor water iron concentration from 16.02 ppb to 19.32 ppb for the designed normal RWCU flow rate of 133,000 Ibm/hr.

However, this change is well within design chemistry limits, and does not affect RWCU performance.

The effects of CPPU on the RWCU system functional capability have been reviewed, and the system can perform adequately during CPPU at the design RWCU system flow rate of 133,000 ibm/hr and the maximum flow rate of 148,000 Ibm/hr. This RWCU system flow results in a slight increase in the calculated reactor water conductivity (from 0.068 gmho/cm to 0.071 pmho/cm) because of the increased FW flow. The design basis conductivity bounds the increased CPPU conductivity.

The increase in FW line pressure has a slight effect on the system operating conditions regarding containment isolation. The effect of this increase was included the containment isolation assessment in Section 4.1.3.

3-33

NEDO-33076 Table 3-1 Hope Creek Adjusted Reference Temperatures - 40 Year Life (32 EFPY) k6....di.d. (Sh.I13) MM.1.fd V.M.tI. W#.kd end Lauw'.%o.d. tIoI.4..nl.Obth 1141.1d (Sh.03to 9.44)

ThckbsWmdn.65 10 32EFFPYP.kI0D n.M. 132-17 n.V2 32EFPP..P10T4...c.. 37E.17 nMr2 32 EFPYPsek1A4.T'-.. 17E.17 nia,?2 1hck ,- %nred 8.10 321EFPPeYl..K D b.uc I AE18 NcW2 32EFtfYPwv WS4T J. 70E.17 reknm2 32 EFPYPdksITWu.1e.- 7EE-17 nknV2 LPCINa1w Fo.1wg ,d We1d T>CkWeU Mashthw 8.10 3200PYNPl 1.0 umee. 47E017 n.r2 32EFPYP~xtleTfkjhxoe- 330.17 rdeff2 32 EFPYP F a 14T hjanAd 32E-17 r~en.2 14 W11.

T 2 EFY 22 EFPY 32 EFPY CCUPCfNT1 HEAT OR HEATILOT %C. 4" CF Rnd Fkl A KneI U..M Soft ART F rdv,7 7 S F *F PLATES Ur Sl N. S 5323011 00 7 018 44 .10 7.2E-17 1e O a 18 32 22 614 0c.s5 8 1 009 014 -1I 7.E-7 21 a It 21 42 3t N0. 824SI1 8 00S 57 1 1 T.E-17 10 0 9 10 37 39 o'.4 Spam; 1 007 0ss 44 .10 TJE- 17 18 a a 8 22 22 1404 EK253011 009 0ss 61 10 7.6E.17 10 e 0 19 27 50 h .4 EK32221I 00 064 SU 7 T7.E6.7 21 a 1 21 42 49 ha. 3 OK3 Il 015 071 113 1o 1.7E-17 28 a 14 2t 65 75 Pi. 3 l&0Bi 009 00 Us 19 37E.17 14 C 7 14 29 42 No. 8 2l I Oto 0 58 8 1o 3.72-17 10 e a 1 32 51 WELDS:

Ca"l

.3 SUAWJWI3 8SIMMs 000 054 100 -40 317Et17 27 e 13 27 f4 14

1. SAW/WIs O0M0O II125-=02S 0081 061 107 .30 17r.t7 20 a 13 20 a2 23 L425.SltAWI9W14WIS 19i.1205 000 0S4 10t -40 721 07 40 a 20 43 so 4a 61d415 8A&MW14,WIS D534tMOtt25C22D5 0821 0oa1 107 -30 7.42-17 3n a 20 39 78 45 IS3M. SAww14 319.0T2"Ms 001 a.3 20 -49 37E.17 5 0 2 5 10 .3 i314 aMAWIwe 50401235 0.01 051 20 .31 3 7E17 5 a 2 5 10 .21 3M. OMAWIWO 510"01 23s 00o 0s4 109 -40 3E7017 27 0 13 27 54 14

.61314 BAW IWu DrO t -

ID304 000 061 o 107 -49 3 70E17 23 0 13 25 03 I sOM. SAWuIWS 07c33t 101D-C= 010 00e 120 -40 3 7E.17 31 0 1e 31 Q2 22 580 45. wMYW/W7 5100220s 0." Q054 1og -40 z E-17 42 0 20 40 60 40 2.4.415. SAWIW7 C053040M12C225 008I 0 11i 107 -30 7.0-017 30 0 20 39 78 40 NEZUS:

u wozzu h7 1640811 012 OSD so -20 13E-I7 20 0 10 20 40 20 N7 1002411 0.14 082 105 -20 3-E-17 24 a 12 24 40 29 3IWel sb"W1 W179 01-01M25 002 0D1 27 -40 3.32.17 6 0 3 6 13 -27 SU4W4W170 51041205 001 003 20 .9 33E-17 6 a 2 5 6 .0 SwwJWI8 004-0t20$ 0.01 o01 20 -S 3. 5 aa.17 2 5 6 .2 3-34

NEDO-33076 Table 3-2 Hope Creek Upper Shelf Energy - 40 Year Life (32 EFPY)

,,;-,-Ii,* -iT ue. -;a

Locstion Het I USE . rItcm2) %cu -;USE -' .32EFPY USE PLATES

_Ower 5tK3230t l1 121 7.60E+17 0.07 185 111 L.rer 6C3511 107 l7.OE+17 0-09 10 9' Lower 6C4511 97 7lOE+17 008 9.5 89 Lower-Intetmediate 6t26631 102 7.605+17 0 07 85 93 Lower-Inteirmediate 5K2530 I 1 QO 70E517 008 95 T8 Lower-intermediate 5t3238 11 7G 7.60E+17 009 10 68 hrurrdiatedSLxefilanca 5K323811 91 7.60E+17 009 10 82 Interiediale 5Kt3025t1 75 3705.17 015 11.5 66 Intermediate bK20 I 1 75 370.E17 0 9 85 69 Intermediate 5K(2698 75 3 705.17 o 0 9 6B WELDS VerM1T Shell 3 SMAW I W13 6510-01205 92.5 3.70E417 0.09 11 82 Shell 3. SAW I W13 0530401t125402205 135 3.705E17 0081 1D.5 121 Shells465. SMAWwi4.w1S 510-01205 92-5 7.60E+17 0.09 13 80 Shells4&5: SAWIWI4.WI5 D5304011125.02205 135 7.60E+17 0.081 12.5 118 U'rarvndiated Survelflance D5304D 164 7.60E+17 0.09 125 144 Glrth Shells 314: SMAWIW6 519M205 109 3.70E+17 0.01 5.5 103 Stells 314: SMAWIWOG 6M041205 125 3.70E+17 0.01 5.5 118 Shells 3/4: SAWIM D53040t1810-02205 e5 3.70E+17 0.081 10.5 85 Shells 3O4: SAWIM D55733 r 1810-02205 8S 3.705417 0.10 11.5 s0 Shells 4r5 SA tW7 D?5304011125-02205 95 7.COE+17 0081 12.b 83 NOZZLES LPCI Forcino N7 1946811 79 l 30E+.17 0.12 10 71 N7 1002411 70 3.30E+17 0.14 10.5 63 LPCI Weod SMAWJW179 001o01205 1D9 330r.17 om 02 l 102 3-35

NEDO-33076 Table 3-3 Hope Creek CUFs of Limiting Components(')

--- P+OStress (ksi) -1CUF Component -Current CP j .Allowable - -Current -CPPU Allowable (ASME Code Core Spray Nozzle 13.76/ 13.79/ 43.1 (3 Sm) 0.796 0.796 1.0 84.57 (3) 84.74 (3)

Main Closure Stud 108.90 109.11 110 (3 Sm) 0.755 0.755 1.0 Shroud Support 22.03/ 24.41/ 69.9 (3 Sm) 0.672 0.672 1.0

_78.52(2.3) 80.9(2,3) 1 _

Notes:

(1) Only components with usage factors greater than 0.5 are included in this table all components have been qualified for a total plant life of 40 years.

(2) Thermal bending stress has been removed. The reactor vessels section evaluates the fatigue and P+Q stresses for the weld at the pressure boundary location of the shroud support. For this evaluation, conservatively, the highest stress in the whole shroud support is used. The reactor internals section considers the remainder of the structural evaluation of the shroud support.

(3) ASME Code Case interpretation 1441 is satisfied, i.e., P+Q stress range can exceed 3Sm when the specified modified fatigue evaluation requirements are satisfied. The numbers in the numerator are for the P+Q stresses without thermal bending and the numbers in the denominator are for P+Q stresses with thermal bending.

3-36

NEDO-33076 Table 34 Hope Creek RIPDs for Normal Conditions (psid)

Parameter CLTP* CPPU*

Core Plate and Guide Tube 18.40 19.70 Shroud Support Ring and Lower Shroud 24.44 27.09 Upper Shroud 6.25 7.59 Shroud Head 7.13 8.52 Shroud Head to Water Level (Irreversible**) 9.68 11.22 Shroud Head to Water Level (Elevation**) 0.86 0.75 Top Guide 0.54 0.54 Steam Dryer 0.34 0.47 Fuel Channel Wall 10.62 12.11

• CLTP equals to 3339 MWt and 105% core flow is assumed for CLTP and CPPU.

• • Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud to the exit of the separators.

. 3-37

NEDO-33076 Table 3-5 Hope Creek RIPDs for Upset Conditions (psid)

Parameter CLTP CPPU*

Core Plate and Guide Tube 20.80 22.10 Shroud Support Ring and Lower Shroud 26.84 29.49 Upper Shroud 9.38 11.39 Shroud Head 10.70 12.78 Shroud Head to Water Level (Irreversible**) 14.51 16.83 Shroud Head to Water Level (Elevation**) 1.29 1.12 Top Guide 0.58 0.58 Steam Dryer 0.44 0.61 Fuel Channel Wall 13.52 15.01

• CLTP is based on 104.2% of OLTP (= 3430 MWT) consistent with current RIPD licensing basis andlO5% core flow is assumed for CLTP and CPPU.

• • Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud to the exit of the separators.

3-38

NEDO-33076 Table 3-6 Hope Creek RIPDs for Emergency Conditions (psid)

Parameter CLTP* CPPU*

Core Plate and Guide Tube 21.00 23.00 Shroud Support Ring and Lower Shroud 29.50 33.00 Upper Shroud 11.80 13.60 Shroud Head 12.00 14.00 Shroud Head to Water Level (Irreversible**) 13.80 15.80 Shroud Head to Water Level (Elevation**) 1.60 1.50 Top Guide 0.30 0.40 Steam Dryer N/A N/A Fuel Channel Wall 11.60 13.50

• CLTP is based on 104.2% of OLTP (3430 MWt) consistent with current RIPD licensing basis and 105% core flow is assumed for CLTP and CPPU.

• • Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud to the exit of the separators.

3-39

NEDO-33076 Table 3-7 Hope Creek RIPDs for Faulted Conditions (psid)

Parameter CLTP* CPPU*

Core Plate and Guide Tube 22.50 24.00 Shroud Support Ring and Lower Shroud 44.00 44.00 Upper Shroud 25.50 27.00 Shroud Head 26.50 27.50 Shroud Head to Water Level (Irreversible**) 28.50 28.50 Shroud Head to Water Level (Elevation**) 1.40 2.50 Top Guide 0.70 1.00 Steam Dryery 11.00 11.00 Fuel Channel Wall 13.30 15.30

• CLTP is based on 104.2% of OLTP consistent with current RIPD licensing basis and 105%

core flow is assumed for CLTP and CPPU.

• • Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud to the exit of the separators.

• • • " The steam dryer pressure drop is for an MSLB outside primary containment and is based on the limiting hot standby condition; this condition has not changed with the CPPU.

340

NEDO-33076 Table 3-8 Hope Creek Reactor Internal Components - Summary of Stresses Item Component Category/  !--Stress/Load CLTP

- -Alloable CPPU

-Location Service Category Value V

_ _ - ~- -- Condition - - _ _ _

1 Shroud Normal/Upset Pm (psi) 10,030 10,870 21,450 2 Shroud Emergency and Qualitative Assessment (See Section 3.3.2 (a))

Faulted 3 Shroud Support Normal/Upset Pm (psi) 20,560 21,958 23,300 4 Shroud Support Normal/Upset Pm+Pb (psi) 21,220 22,663 35,000 5 Shroud Support Emergency and Pm (psi) 34,758 39,485 46,600 Faulted _

6 Shroud Support Emergency and Pm+Pb (psi) 36,635 41,617 69,900 Faulted _

7 Core Plate Qualitative Assessment (See Section 3.3.2(c))

8 Top Guide Qualitative Assessment (See Section 3.3.2(d))

9 CRD Housing Qualitative Assessment (See Section 3.3.2(e))

10 CRD Qualitative Assessment (See Section 3.3.2(f))

11 Control Rod Guide Qualitative Assessment (See Section 3.3.2(g))

Tube 12 Orificed Fuel Qualitative Assessment (See Section 3.3.2(h))

Support 13 Fuel Channel Qualified per Proprietary Fuel Design Basis 14 Steam Dryer Qualitative Assessment (See Section 3.3.2(j))

15 FW Sparger Qualitative Assessment (See Section 3.3.2(k))

16 Jet Pump Qualitative Assessment (See Section 3.3.2(1))

17 Core Spray Line Qualitative Assessment (See Section 3.3.2(m))

and Sparger 18 Access Hole Cover Normal/ Upset Qualitative Assessment (See Section 3.3.2 (n))

19 Access Hole Cover Emergency and Pm+Pb (psi) 32,111 32,466 47,300 Faulted 20 Shroud Head & Qualitative Assessment (See Section 3.3.2(o))

Steam Separator Assembly 21 LPCI Coupling Qualitative Assessment (See Section 3.3.2(p))

22 Core Delta P and Qualitative Assessment (See Section 3.3.2(q))

Liquid Control Line 3-41

NEDO-33076 Table 3-9 Hope Creek ASME Class 1 Piping Maximum percent increase in ASME Class 1 pipe stresses, usage factor, interface loads and displacements.

ASME CODE EQUATION MS (3)_ F"' (4)

Flow Pr&T Total Temp Press Total 9A N/A N/A N/A N/A N/A N/A 9B 4.7 0.0 4.7 N/A N/A N/A 9C 4.7 0.0 4.7 N/A N/A N/A 9D 4.7 0.0 4.7 N/A N/A N/A 10 2.4 0.0 2.4 (7) 0.0 (7) 12 N/A 0.0 0.0 (7) 0.0 (7) 13 N/A 0.0 0.0 (7) 0.0 (7) 14 (2) 1.2 0.0 1.2 (7) 0.0 (7)

Interface Loads (5) 23.9 0.0 23.9 (7) 0.0 (7)

Interface (Mr) Loads (6) 37.0 0.0 37.0 - - -

Thermal Displacement N/A 0.0 0.0 (7) 0.0 (7)

NOTES:

1. N/A - Not affected due to no pressure increase

2. Fatigue - Cumulative Usage Factor

3. MS, and MSIV Drain Lines

4. FW (Outside Containment)

5. For MS system, percent increase of total support load to account for Turbine Stop Valve Transient Load.

6. For MS system, percent increase of Turbine Stop Valve Closure transient load only, then recombined with other support loads.

7. The major portion of this analysis is located inside containment. The analysis nodes for outside containment are enveloped by nodal points inside containment and are considered acceptable.

3-42

NEDO-33076 Table 3-10 Hope Creek BOP Piping MS and FW - ASME Class 2 and 3 Piping Maximum percent increase in ASME Class 2 and 3 pipe stresses, interface loads and displacements ASME CODE EQUATION Main Steam (2) Feed Water (3)

Flow Pr&T Total Temp Press Total 8 N/A N/A N/A N/A N/A N/A 9B 4.7 0.0 4.7 N/A N/A N/A 9C 4.7 0.0 4.7 N/A N/A N/A 9D 4.7 0.0 4.7 N/A N/A N/A 10 N/A 0.0 0.0 (6) 0.0 (6) 11 N/A 0.0 0.0 (6) 0.0 (6)

Interface Loads (4) 23.9 0.0 23.9 (6) 0.0 (6)

Interface (Mr) Loads (5) 37.0 0.0 37.0 - - -

Thermal Displacement N/A 0.0 0.0 (6) 0.0 (6)

NOTES:

1 N/A - Not affected due to no pressure increase

2. MS, including Turbine Bypass (Outside Containment), MSIV Drain Lines, and Extraction Steam

3. FW (Outside Containment), Condensate, FW Heater Drains & Vents 4 For MS system, percent increase of total support load to account for Turbine Stop Valve Transient Load.

5. For MS system, percent increase of Turbine Stop Valve Closure transient load only, then recombined with other support loads.

6. The effects of a higher temperature was evaluated and found to be acceptable. The minimal increases in temperature, support loads and displacements were reviewed against established parameters and compared to margins within the calculations for these systems, and was found to be acceptable.

3-43

NEDO-33076 I

00 Tlm* ("c)

• 1

% 75 0 9

tO 1.0 20 30 40 s0 to 7.0 s0 0 Du 1.0 . a0 .0 so Tom ("c) Time (sec)

Figure 3-1 Response to MSIV Closure with Flux Scram (102% CPPU power, 105% core flow, and 1035 psia initial dome pressure) 3-44

NEDO-33076

4. ENGINEERED SAFETY FEATURES This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapter 6, that applies to CPPU. Regulatory Guide 1.70, Chapter 6 states, "engineered safety features are provided to mitigate the consequence of postulated accidents," and "are those (features) that are commonly used to limit the consequences of postulated accidents." NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 6.1.1, subsection I states, "Engineered safety features (ESF) are provided in nuclear plants to mitigate the consequences of design basis or loss-of-coolant accidents." The Hope Creek features evaluated within this section are designed to (directly) mitigate the consequences of postulated accidents, and thus, are classified in the plant Updated Final Safety Analysis Report (UFSAR) as engineered safety features, consistent with Regulatory Guide 1.70 and NUREG-0800.

4.1 CONTAINMENT SYSTEM PERFORMANCE This section addresses the effect of the CPPU on various aspects of the Hope Creek containment system performance. The topics addressed in this evaluation are:

Topic CLTRDisposition - Hope Creek Result 4.1.1 Pool temperature response 4.1.1 Wetwell pressure 4.1.1 Drywell temperature 4.1.1 Drywell pressure 4.1.2 Containment dynamic loads 4.1.3 Containment isolation 4.1.4 Motor-operated valves 4.1.5 Hardened vent system 4.1.6 Equipment operability ))

The UFSAR provides the containment responses to various postulated accidents that validate the design basis for the containment. Operation at CPPU causes changes in some of the conditions for the containment analyses. For example, the short-term DBA LOCA containment response during the reactor blowdown is governed by the blowdown flow rate. This blowdown flow rate is dependent on the reactor initial thermal-hydraulic conditions, such as vessel dome pressure and the mass and energy of the vessel fluid inventory, which change slightly with CPPU. Also, the long-term heatup of the suppression pool following a LOCA or a transient is governed by the ability of the RHR system to remove decay heat. Because the decay heat depends on the initial reactor power level, the long-term containment response is affected by CPPU. The containment 4-1

NEDO-33076 pressure and temperature responses have been reanalyzed, as described in Section 4.1.1, to demonstrate the plant capability to operate at CPPU RTP.

The analyses were performed in accordance with RG 1.49 and References 1 and 2, using GE codes and models (References 9 through 12). The application of the GE methods to power uprate evaluations has been reviewed and approved by the NRC (References 2, 17, 18 and 19).

The M3CPT code is used to analyze the short-term containment pressure and temperature response to the DBA-LOCA at CPPU conditions. This code was also used to analyze the short-term DBA-LOCA containment response for the UFSAR. The CPPU analysis used LAMB (Reference 10) to calculate the blowdown flow rates, which are then used as inputs to M3CPT.

This approach differs from that for the current UFSAR analysis, which uses a blowdown model built into M3CPT. Application of the LAMB blowdown model for the extended power uprate analysis is identified in Reference 2. The SHEX code is used for long-term containment response evaluations at CPPU conditions. SHEX was previously used for the Hope Creek UFSAR analysis.

The effect of CPPU on the containment dynamic loads due to a LOCA or SRV discharge has also been evaluated as described in Section 4.1.2. These loads were previously defined generically during the Mark I Containment Long Term Program (LTP) as described in Reference 13 and accepted by the NRC per References 13 and 14. Plant-specific dynamic loads were also defined (References 19 and 20), which were accepted by the NRC in Reference 13.

The evaluation of the LOCA containment dynamic loads is based primarily on the results of the short-term analysis described in Section 4.1.1.3. The SRV discharge load evaluation is based on no changes in the SRV opening setpoints at CPPU conditions.

The consequences of a LOCA occurring within the wetwell and the capability of the containment to withstand the effects of steam bypassing the suppression pool were evaluated. The Hope Creek containment analysis results demonstrate that CPPU does not significantly affect containment pressure and temperature response; therefore these other capabilities would not be significantly affected.

4.1.1 Containment Pressure and Temperature Response Short-term and long-term containment analyses results are reported in the UFSAR. The short-term analysis is directed primarily at determining the drywell pressure response during the initial blowdown of the reactor vessel inventory to the containment following a large break inside the drywell. The long-term analysis is directed primarily at the suppression pool temperature response, considering the decay heat addition to the suppression pool. The effect of CPPU on the events yielding the limiting containment pressure and temperature responses are provided below.

4.1.1.1 Long-Term Suppression Pool Temperature Response Short-term and long-term containment analysis results are reported in the UFSAR. The long-term analysis is directed primarily at the pool temperature response, considering the decay heat addition to the pool.

4-2

NEDO-33076 (a) Bulk Pool Temperature The long-term bulk pool temperature response with CPPU was evaluated for the DBA LOCA.

The analysis was performed at 102% of CPPU RTP. Table 4-1 compares the calculated peak values for LOCA bulk pool temperature. The current analyses have been performed using an RHR heat exchanger K-value of 307 BTU/sec- 0 F/HX and safety auxiliary cooling system (SACS) temperature of 100 0 F. The CPPU analysis was performed using a realistic decay heat model (ANS/ANSI 5.1, 1979 with 2a uncertainty and the recommendations of SIL 636, Rev. 1),

as compared to May-Witt decay heat model used in the current UFSAR analysis. The Hope Creek calculated peak bulk suppression pool temperatures are provided in Table 4-1 for both 102% of CLTP and 102% of CPPU RTP. This comparison shows that CPPU results in an increase of 11.30 F in peak pool temperature, based on the current method. The peak pool temperatures are below the suppression chamber structural design temperature of 3101F.

Therefore, the peak bulk pool temperature with CPPU is acceptable from a structural design standpoint.

Because the peak suppression pool temperature exceeds the current UFSAR value of 2120 F for NPSH, the evaluation of available NPSH for Core Spray (CS) and RHR pumps is needed. The results of that evaluation are documented in Section 4.2.6.

(b) Local Pool Temperature with SRV Discharge The local pool temperature limit for SRV discharge is specified in NUREG-0783 (Reference 14),

because of concerns resulting from unstable condensation observed at high pool temperatures in plants without quenchers. The local pool temperature has been evaluated for CPPU. This analysis was performed at 102% of 3952 MWt, which exceeds the 104.3% of RTP requirement in Reference 15. The local peak suppression pool temperature is 202.1degrees F, which is below the limit of 204.1 degrees F (Reference 15) and meets the NUREG-0783 criteria. Therefore, the peak local suppression pool temperature at Hope Creek is acceptable for CPPU conditions.

4.1.1.2 Short-Term Gas Temperature Response The short-term DBA-LOCA analysis covers the blowdown period when the maximum drywell airspace temperature occurs. These analyses were performed at 102% of 3952 MWt, using methods reviewed and accepted by the NRC during the Mark I Containment LTP (Reference 13) with the break flow calculated using a more detailed RPV model (Reference 10) previously approved by the NRC. The calculated peak drywell airspace temperatures are provided in Table 4-1. Table 4-1 also shows the values from calculations with CLTP RTP using the same method as CPPU. CPPU increases the calculated peak drywell airspace gas temperature by 37F. The peak drywell air space temperature remains below the containment structural design basis temperature of 340TF. Therefore, the short-term drywell airspace gas temperature responses with CPPU are acceptable.

Table 4-1 shows the peak calculated suppression chamber temperature. The short-term suppression chamber temperatures with CPPU are well below the suppression chamber design temperatures. Therefore, the short-term suppression chamber gas temperature responses with CPPU are acceptable.

4-3

NEDO-33076 4.1.1.3 Short-Term Containment Pressure Response Short-term containment response analyses were performed for the limiting DBA LOCA, assuming a double-ended guillotine break of a recirculation suction line to demonstrate that CPPU does not result in exceeding the containment design limits. The short-term analysis covers the blowdown period when the maximum drywell pressures and differential pressures between the drywell and wetwell occur. These analyses were performed at 102% of 3952 MWt, using methods reviewed and accepted by the NRC during the Mark I Containment LTP (Reference 13) with the break flow calculated using a more detailed RPV model (Reference 12) previously approved by the NRC. The results of these short-term analyses are summarized in Table 4-1 for comparison against the drywell design pressure. Also included in Table 4-1 is a comparison of the peak drywell pressure calculated with the CPPU method at CLTP against the UFSAR values. The drywell pressure increases by 3 psi as a result of CPPU. The design pressure bounds the maximum drywell pressure value for CPPU.

4.1.2 Containment Dynamic Loads 4.1.2.1 Loss-of-Coolant Accident Loads The LOCA containment dynamic loads analysis for CPPU is based primarily on the short-term LOCA analyses. These analyses were performed at 102% of 3952 MWt, using methods reviewed and accepted by the NRC during the Mark I Containment LTP (Reference 13) with the break flow calculated using a more detailed RPV model (Reference 10) previously approved by the NRC. The NRC approved the use of this model for CPPU containment evaluations in Reference 2. These analyses provide calculated values for the controlling parameters for the dynamic loads throughout the blowdown. The key parameters are drywell and suppression chamber pressure, vent flow rates, and suppression pool temperature. The LOCA dynamic loads for CPPU include pool swell, condensation oscillation (CO), and chugging. For Mark I plants like Hope Creek, the vent thrust loads are also evaluated.

The short-term containment response conditions at 102% of 3952 MWt are within the range of test conditions used to define the pool swell and condensation oscillation loads for the plant.

The peak drywell pressure from these analyses is given in Table 4-1. The long-term response conditions at 102% of 3952 MWt in which chugging would occur are within the conditions used to define the chugging loads. The vent thrust loads at 102% of 3952 MWt are calculated to be less than plant-specific values calculated during the Mark I Containment LTP because the LAMB computer code was used to calculate the blowdown flows and enthalpies. Therefore, the LOCA dynamic loads definition is not affected by CPPU.

4.1.2.2 Safety Relief Valve Loads The Safety Relief Valve (SRV) air-clearing loads include SRV discharge line (SRVDL) loads, suppression pool boundary pressure loads and drag loads on submerged structures. These loads are influenced by SRV opening setpoint pressure, the initial water leg in the SRVDL, SRVDL geometry, and suppression pool geometry. For the first SRV actuations following an event involving RPV pressurization the only parameter change potentially introduced by CPPU, which can affect the SRV loads, is an increase in SRV opening setpoint pressure. However, this CPPU 4-4

NEDO-33076 does not include an increase in the SRV opening setpoint pressures. CPPU may reduce the interval between subsequent SRV actuations, which may affect the load definition for subsequent actuations.

The maximum column of water in the SRVDL after SRV opening and reclosing (i.e., cycling) is determined by the SRVDL geometry, pool water temperature, and SRVDL vacuum breaker characteristics. If a subsequent SRV actuation occurs immediately after the SRV closure, the actuation would occur with an elevated water leg in the SRVDL, which can increase the SRVDL loads (i.e., a larger mass of water would be expelled upon subsequent SRV actuation). To mitigate the effects of a subsequent SRV actuation, Low-Low Set (LLS) logic was previously implemented at Hope Creek to extend the time between SRV closure and subsequent actuations (Reference 21). Analysis at 102% of 3952 MWt shows that there is at least 18 seconds to re-open a LLS valve, which is greater than the lower limit of 4 seconds set in Reference 21. This time period is sufficient to allow the water leg of the SRV piping to clear as well as to mitigate the thrust loads on SRV discharge piping. Therefore, the LLS system performs as required for CPPU RTP conditions.

4.1.2.3 Subcompartment Pressurization The mass and energy releases that affect the annulus pressurization loads on the biological shield wall caused by a postulated recirculation suction line break (RSLB) or FW line break (FWLB) were evaluated at CPPU conditions. Using methods consistent with the existing Hope Creek design and licensing basis, the mass and energy releases at CPPU conditions exceed the CLTP values, but are less than the releases from the MELLLA condition for comparable points considered. The minimum recirculation pump speed with FWTR case provides the limiting mass and energy releases for both RSLB and FWLB loads evaluation.

4.1.3 Containment Isolation The system designs for containment isolation are not affected by CPPU. Isolation actuation devices located inside and outside of containment were designed to operate at a containment pressure that bounds the post-accident pressures at CPPU conditions. The capabilities of isolation actuation devices to perform during normal operations and post-accident conditions have been determined to be acceptable. Some of the containment isolation MOVs discussed in Section 4.1.4, have reduced operating margins, but remain capable of performing their isolation function. Therefore, the Hope Creek containment isolation capabilities are not adversely affected by the CPPU.

4.1.4 Generic Letter 89-10 Program The process parameters of temperature, pressure, and flow for GL 89-10 MOVs were reviewed, and minor changes in the process parameters were identified as a result of operation at CPPU conditions. Peak drywell pressure increases from 48.1 psig to 50.6 psig, peak suppression pool pressure increases from 27.5 psig to 27.7 psig, and fluid velocities increase in some systems, causing an increase in dynamic differential pressure (dp). Based on the margins in the existing capabilities for these valves, none of these changes are expected to affect component or system operability. Therefore, the Hope Creek GL 89-10 MOVs remain capable of performing their 4-5

NEDO-33076 intended design basis functions at CPPU conditions. MOV calculations will be revised as necessary to include the effect of CPPU conditions on calculated margins.

Any increase in room temperatures at the CPPU conditions does not affect any GL 89-10 MOVs because the existing analyses use the design maximum room temperature. Therefore, the Hope Creek GL 89-10 MOVs remain capable of performing their intended design basis functions at the CPPU conditions.

The effect of the CPPU on the potential for pressure locking/thermal binding effects (GL 95-07) was also reviewed for CPPU. Operation at the CPPU conditions increases post-accident drywell and torus temperatures. However, the increased CPPU conditions are bounded by the existing analyses, which use design temperatures. Therefore, the GL 95-07 conclusions remain valid for the CPPU.

Air Operated Valves The air operated valve (AOV) process parameters of temperature, pressure, and flow were reviewed and no changes to the functional requirements of any AOV were identified as a result of operating at the CPPU conditions.

Operation at the CPPU conditions is within the pressure and temperature capability of the AOVs.

Therefore, the Hope Creek AOVs remain capable of performing their design basis function.

4.1.5 Generic Letter 89-16 The quantitative results below conservatively assumed power operation at 4031 MWT, which corresponds to 120% of OLTP plus a 2% uncertainty. This bounds the CPPU conditions.

The hardened vent is designed to mitigate loss-of-decay-heat removal by providing sufficient wetwvell venting capability to prevent further containment pressurization with the containment at its pressure limit. The vent is designed with sufficient capacity to accommodate decay heat input equivalent to 1% of CLTP. This corresponds to 0.83% of 4031 MWT. Decay heat decreases to 0.83% 7.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after shutdown. The drywell pressure limit is not reached until 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the wetwell pressure limit is not reached until 7.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after loss-of-decay-heat removal capability following operation at 4031 MWT. Thus, under CPPU conditions, the existing hardened vent capacity is adequate to relieve decay heat when the containment pressure limit is reached.

4.1.6 Generic Letter 96-06 The Hope Creek response to Generic Letter 96-06, "Assurance of Equipment Operability and ContainmentIntegrity DuringDesign-BasisAccident Conditions, " was reviewed for CPPU post accident conditions. The issues identified in the GL 96-06 review were addressed through procedural changes and piping modifications (i.e., installation of pressure relief valves). These modifications and procedural changes are unaffected by CPPU. Therefore, the existing Hope Creek response to Generic Letter 96-06 remains valid for CPPU.

4-6

NEDO-33076 4.2 EMERGENCY CORE COOLING SYSTEMS Each Hope Creek emergency core cooling system (ECCS) is discussed in the following subsections. The effect on the functional capability of each system, due to CPPU is addressed.

The assumption of constant pressure minimizes the effect of CPPU for ECCS evaluation. The topics addressed in this evaluation are:

-Topic CLTR Disposition Hope Creek Result 4.2.1 High Pressure Coolant Injection 4.2.2 High Pressure Core Spray 4.2.3 Core Spray 4.2.4 Low Pressure Coolant Injection 4.2.5 Automatic Depressurization 4.2.6 ECCS Net Positive Suction Head 4.2.1 High Pressure Coolant Injection The HPCI system is designed to pump water into the reactor vessel over a wide range of operating pressures. The primary purpose of the HPCI is to maintain reactor vessel coolant inventory in the event of a small break LOCA that does not immediately depressurize the reactor 4-7

NEDO-33076 vessel. In this event, the HPCI system maintains reactor water level and helps depressurize the reactor vessel. The adequacy of the HPCI system is demonstrated in Section 4.3.

For CPPU, there is no change to the maximum nominal reactor operating pressure and the SRV setpoints remain the same. ff 4.2.2 High Pressure Core Spray The High Pressure Core Spray system is not applicable to Hope Creek.

4.2.3 Core Spray or Low Pressure Core Spray The Low Pressure Core Spray (LPCS) system is not applicable to Hope Creek.

The CS system is automatically initiated in the event of a LOCA. When operating in conjunction with other ECCS, the CS system is required to provide adequate core cooling for all LOCA events. There is no change in the reactor pressures at which the CS is required.

The CS system sprays water into the reactor vessel after it is depressurized. The primary purpose of the CS system is to provide reactor vessel coolant inventory makeup for a large break LOCA and for any small break LOCA after the reactor vessel has depressurized. It also provides spray cooling for long-term core cooling in the event of a LOCA. The CS system meets all 4-8

NEDO-33076 applicable safety criteria for the CPPU. The adequacy of the CS system performance is demonstrated by the margins discussed in Section 4.3.

IT The peak suppression pool temperature (212.30 F) during a limiting event exceeds the current maximum operating temperature of 2120 F for the CS pump seals. The CS pump seals have been qualified to a temperature of 218 IF Therefore, the CS system at Hope Creek is acceptable for CPPU operation.

4.2.4 Low Pressure Coolant Injection The Low Pressure Coolant Injection (LPCI) mode of the RHR system is automatically initiated in the event of a LOCA. The primary purpose of the LPCI mode is to help maintain reactor vessel coolant inventory for a large break LOCA and for any small break LOCA after the reactor vessel has depressurized. The LPCI operating requirements are not affected by CPPU. The adequacy of this system is demonstrated by the margins discussed in Section 4.3.

))

4.2.5 Automatic Depressurization System 4-9

NEDO-33076 The ADS uses safety/relief valves to reduce the reactor pressure following a small break LOCA when it is assumed that the high-pressure systems have failed. This allows the CS and LPCI to inject coolant into the reactor vessel. The adequacy of this system is demonstrated by the margins discussed in Section 4.3. The ADS initiation logic and valve control is not affected by CPPU conditions. The CPPU does not change the conditions at which the ADS must function.

o((

4.2.6 ECCS Net Positive Suction Head CPPU increases the reactor decay heat, which increases the heat input to the suppression pool.

This increased heat input increases the peak suppression pool water temperature, which may affect RHR, CS and HPCI pump operation. As discussed in Section 4.1.1 and shown in Table 4-1, the calculated peak suppression pool temperature for the most limiting case, the LOCA, is 212.30 F, which exceeds the current peak pool limit of 2121F. Peak suppression pool temperatures for an ATWS, Appendix R, and SBO event are bounded by the most limiting LOCA case.

The net positive suction head (NPSH) requirements for the RHR and CS pumps were analyzed using 0 psig containment pressure, as required by RG 1.1, a peak suppression pool water temperature of 218 0F, and maximum calculated pump flows, which exceed the design basis pump flows. Using these conditions, the NPSH available is greater than the NPSH required for the RHR and CS pumps.

4-10

NEDO-33076 The NPSH requirements for the HPCI pump are based on design basis peak suppression pool temperatures for transient operation of 170'F, 0 psig containment pressure, as required by RG 1.1, and design basis pump flow. Under CPPU conditions, the peak suppression pool temperatures for transient operation of the HPCI pump does not exceed 170'F (including ATWS, Appendix R and SBO events). The requirements of the Condensate Storage Tank (CST) are not affected by CPPU; therefore, the NPSH requirements for the HPCI pump from the CST are not affected by CPPU.

The general methods used to calculate the ECCS suction strainer debris loading and head losses following a design basis LOCA at Hope Creek are based on NEDO-32686 Rev. 0, Utility Resolution Guidance for ECCS Suction Strainer Blockage, and NUREG/CR-6224, Parametric Study of the Potential for BWR ECCS Strainer Blockage due to LOCA Generated Debris. Hope Creek calculations used conservative estimates for fibrous debris (425 cu. ft.), suppression pool sludge (150 lblyr), dirt/dust (150 lb), rust flakes (50 lb), paint chips (85 lb), and unqualified coatings (270 lb). CPPU conditions do not affect the methods for calculating these values.

Strainer approach velocities and suppression pool turbulence are inputs to the calculation of suction strainer debris loading. Because RHR and CS pump flow rates do not change for CPPU, strainer approach velocities are not affected. Suppression pool turbulence at CPPU conditions remains within the conditions used to define the chugging loads (Section 4.1.2. of this document). Because CPPU conditions do not result in new HELB locations, the existing Hope Creek calculation for zones of influence remains valid. Therefore, the suction strainer debris loading analysis remains bounding for CPPU.

4.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE The Emergency Core Cooling Systems (ECCS) are designed to provide protection against hypothetical loss-of-coolant accidents (LOCA) caused by ruptures in the primary system piping.

The ECCS performance under all LOCA conditions, and their analysis models, satisfy the requirements of 10CFR50.46 and 10CFR50 Appendix K. The results of the ECCS-LOCA analysis using NRC-approved methods are summarized in Table 4-2.

The Licensing Basis PCT is determined based on the calculated nominal PCT with an adder to account for uncertainties. The adder is derived from calculations that are in conformance with the requirements of 10CFR50 Appendix K.

For CPPU (115% of CLTP), the Licensing Basis PCT for GE14 fuel is 1380'F at MELLLA conditions (94.8% of rated core flow). The comparable GE14 Licensing Basis PCT for the CLTP conditions is 1370'F at MELLLA condition (76.6% of rated core flow). The calculated results show significant margin to the licensing limit of 22000 F.

For Single Recirculation Loop Operation (SLO) a multiplier is applied to the Two-Loop Operation LHGR and MAPLHGR limits. Application of the appropriate LHGR/MAPLHGR multiplier for SLO operation assures the expected SLO PCT is less than the calculated PCT for Two-Loop Operation.

The increase in reactor power due to CPPU has a small effect on the Licensing Basis PCT, and has a negligible effect on the local clad oxidation, the hydrogen generation, the coolable 4-11

NEDO-33076 geometry, and the long-term cooling. The LOCA evaluations with the equilibrium core of GE14 fuel demonstrate compliance with the ECCS acceptance criterion.

4.4 MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM The Hope Creek topics addressed in this evaluation are:

Topic CLTR Dspoition -Hope CreekResult Iodine intake U With the exception of resealing of some instrumentation, CPPU does not require any changes to the MCR. Heat sources in the main control room (MCR) are due to equipment, ambient outside air temperature, and personnel, and do not change with CPPU. There are no changes to the MCR envelope and there are no significant changes to the temperatures in the adjacent walls and ceilings. Accordingly, there is no change in the heating and cooling loads, required ventilation flow, or the MCR capability to establish isolation and maintain positive pressure with respect to outside boundaries.

The radiological effect of CPPU on control room (CR) habitability is due to an increase in the core iodine activity released during the DBAs. Performance of the CR emergency filtration (CREF) system is not affected as a result of the CPPU. The CR habitability was re-analyzed for various DBAs using the CPPU core inventory and the previously approved Alternative Source Term (AST) methodology. The post-CPPU CR doses resulting from the various DBAs are within the applicable regulatory limits and in compliance with 10 CFR 50.67 requirements. The CREF system is not credited for the first 30 minutes after the onset of a LOCA. The non-LOCA CR doses are analyzed with the CR in normal mode operation without CREF initiation. The appropriate CR air intake x/Qs are developed based on the 7-years site-specific meteorology data and release point/receptor geometries. Post-LOCA CR charcoal filter shine dose is calculated to be insignificant. Four-inch deep CREF charcoal beds are tested using the Generic Letter 99-02 requirements, and charcoal filter efficiencies calculated based on the test results are used in the DBA analyses. Per Regulatory Guide 1.183, Table 1, a total of 30% of the core iodine is released to the drywell. Of the released iodine, 4.85% of the elemental iodine and 0.15% of the organic iodide are conservatively assumed to be collected on the CREF charcoal beds without crediting radiological decay, atmospheric transport dilution, removal by the Filtered Recirculation Ventilation System (FRVS) charcoal filtration, and holdup and plateout of iodine on the main steam piping surface. Despite the increase in iodine core inventory as a result of CPPU, the iodine loading on the CREF charcoal filters remains a small fraction of the allowable limit of 2.5 mg of total iodine per gram of activated carbon, as required by Regulatory Guide 1.52. The results of the CR habitability analyses indicate that the CREF charcoal beds provide adequate radiation protection to the CR operators during design basis accident conditions including a LOCA with the assumed CR unfiltered inleakage of 350 cfm. The actual measured CR unfiltered inleakage is less than 210 cfm.

4-12

NEDO-33076 Table 9.3 addresses the LOCA accident doses for the Main Control Room.

4.5 FILTRATION, RECIRCULATION, AND VENTILATION SYSTEM The Filtration, Recirculation, and Ventilation System (FRVS), referred to as the Standby Gas Treatment System in the CLTR, is designed to maintain secondary containment at a negative pressure and to filter the exhaust air for removal of fission products potentially present during abnormal conditions. By limiting the release of airborne particulates and halogens, the FRVS limits off-site dose following a postulated design basis accident. The topics addressed in this evaluation are:

Topic -CLTRDisposition Hope Creek Result Flow capacity ((

Iodine removal capability ))

[4[__

4-13

NEDO-33076 1]

The design flow capacity of the FRVS was selected to maintain the secondary containment at the required negative pressure to minimize the potential for exfiltration of air from the reactor building. ((

))The total (radioactive plus stable) post-LOCA iodine loading on the charcoal adsorbers increases proportionally with the increase in core iodine inventory, which is proportional to core thermal power (Section 9.2). Sufficient charcoal mass is present at Hope Creek so that the post-LOCA iodine loading on the charcoal remains below the guidance provided by Regulatory Guide 1.52.

While decay heat from fission products accumulated within the system filters and charcoal adsorbers increases in proportion to the increase in thermal power, the cooling air flow required to maintain components below operating temperature limits is well below the cooling flow capability of the system. A water deluge system is provided to maintain the charcoal adsorbers below the ignition temperature in the event of loss of cooling flow.

In support of the above conclusions, (( )) has been performed in the CLTR to evaluate systems that implement Alternate Source Term (AST) in accordance with Regulatory Guide 1.183. ((

))

Results of the AST evaluation, applicable to Hope Creek, show that the maximum charcoal loading, based on only 50 pounds of charcoal per adsorber train, is approximately 0.26 mg of total iodine per gram of charcoal, well below the 2.5 mg/gm maximum value in Regulatory Guide 1.52. The maximum component temperature is approximately 168 0F with normal flow conditions and, under conditions of a failed fan, charcoal temperature is maintained below the 625 0F charcoal ignition temperature by water deluge.

4-14

NEDO-33076

((

1]

4.6 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM Hope Creek does not use a Main Steam Isolation Valve Leakage Control System (MSIV-LCS).

4.7 POST-LOCA COMBUSTIBLE GAS CONTROL SYSTEM The revised 10 CFR 50.44 (68FR54123, dated September 16, 2003) does not define a design basis LOCA hydrogen release and eliminates the requirements for hydrogen control systems to mitigate such releases. Hope Creek License Amendment Number 160, issued on August 9, 2005, eliminated the requirements for hydrogen recombiners and hydrogen/oxygen monitors.

4-15

NEDO-33076 Table 4-1 Hope Creek DBA LOCA Containment Performance Results CLTP CPPU' Limit UFSAR CPPU Method"'

Parameter Peak Drywell Airspace 48.1 47.6 50.6 62 Pressure (psig)

Peak Drywell Airspace 291 295 298 340 Temperature (0F)

Peak Bulk Pool 210 201 212.33 212/3104 Temperature (0F)

Peak Wetwell 27.5 27.6 27.7 62 Airspace Pressure (psig)

Peak Wetwell N/R5 198.2 212.2 310 Airspace Temperature (OF)

1. The CPPU analysis was performed at 102% of 3952 MWt using a realistic decay heat model (ANS/ANSI 5.1, 1979 with 2a uncertainty), as compared to May-Witt decay heat model used in the current UFSAR analysis.

2. The values in the "CPPU Method" column are the CLTP values recalculated using the CPPU methodology for comparison to the calculated CPPU values.

3. Reported value is based on 102% of requested CPPU RTP (3840 MWt)

4. The peak bulk suppression pool temperature at CPPU RTP exceeds the current licensing limit, but stays below the design limit. The core spray and RHR pump NPSH margin is addressed in Section 4.2.6.

5. No long term UFSAR wetwell airspace temperature available.

4-16

NEDO-33076 Table 4-2 Hope Creek ECCS Performance Results 10 CFR 50.46 Parameter CLTP CPPU Limit Method SAFER/GESTR SAFER/GESTR Power 101.4 % OLTP 115% CLTP

1. Licensing Basis 1370 (GE14) 1380 (GE14) < 2200 Peak Clad Temperature, (PCT) 0F

2. Cladding <1.0 <1.0 <17 Oxidation, %

Original Clad Thickness

3. Hydrogen <0.1 <0.1 <1.0 Generation (Core wide Metal-Water Reaction) %

4. Coolable OK OK Meet 1 and 2, above Geometry

5. Core Long OK OK Core flooded to TAF Term Cooling or Core flooded to jet pump suction elevation and at least one core spray system is operating at rated flow.

4-17

NEDO-33076 Table 4-3 Hope Creek FRVS Iodine Removal Capacity Parameters ri:

4 I.

4 4 4 4 1]

(I) Hope Creek uses a water deluge system to limit maximum temperature.

(2) Hope Creek criteria is <0.5% first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, <0.25 day 2 through day 30 4-18

NEDO-33076

5. INSTRUMENTATION AND CONTROL This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapter 7, as it applies to CPPU. The principle instrumentation affected by CPPU is addressed in the following.

5.1 NSSS MONITORING AND CONTROL The instruments and controls used to monitor and directly interact with or control reactor parameters are usually within the NSSS. Changes in process variables and their effects on instrument performance and setpoints were evaluated for CPPU operation to determine any related changes. Process variable changes are implemented through changes in normal plant operating procedures. Technical Specifications address those instrument allowable values and/or setpoints for those NSSS sensed variables that initiate protective actions. The effect of CPPU on Technical Specifications is addressed in Section 5.3. The topics addressed in this evaluation are:

Topic -CLTRDisposition -lHope CreekResult 5.1.1.1 Average Power Range Monitors, Intermediate Range Monitors, and Source Range Monitors 5.1.1.2 Local Power Range Monitors 5.1.1.3 Rod Block Monitor 5.1.2 Rod Worth Minimizer/Rod Control Information System 5.1.1 Neutron Monitoring System CPPU affects the performance of the Neutron Monitoring System. These performance effects are associated with the Average Power Range Monitors (APRMs), Intermediate Range Monitors (IRMs), Local Power Range Monitors (LPRMs), Rod Block Monitor (RBM), and Rod Worth Minimizer (RWM).

5-1

NEDO-33076 5.1.1.1 Average Power Range Monitors, Intermediate Range Monitors and Source Range Monitors At rated power, the increase in power level increases the average flux in the core and at the in-core detectors. The APRM power signals are calibrated to read 100% at the new licensed power.

CPPU has little effect on the IRM overlap with the SRMs and the APRMs. Using normal plant surveillance procedures, the IRMs may be adjusted, as required, so that overlap with the SRMs and APRMs remains adequate.

The SRM, IRM, and APRM Systems installed at Hope Creek are in accordance with the requirements established by the GE design specifications. ((

5.1.1.2 Local Power Range Monitors

, ~ ~~~ " '" .

At CPPU RTP the average flux experienced by the detectors increases due to the average power increase in the core. The maximum flux experienced by an LPRM remains approximately the same because the peak bundle powers do not appreciably increase. Due to the increase in neutron flux experienced by the LPRMs and traversing incore probes (TIPs), the neutronic life of 5-2

NEDO-33076 the LPRM detectors may be reduced and radiation levels of the TIPs may be increased. LPRMs are designed as replaceable components. The LPRM accuracy at the increased flux is within specified limits, and LPRM lifetime is an operational consideration that is handled by routine replacement. TIPs are stored in shielded rooms. The normal plant operation radiation protection program can accommodate the small increase in radiation levels.

The LPRMs and TIPs installed at Hope Creek are in accordance with the requirements established by the GE design specifications. ((

5.1.1.3 Rod Block Monitor

((

The increase in power level at the same APRM reference level results in increased flux at the LPRMs that are used as inputs to the RBM. The RBM instrumentation is referenced to an APRM channel. Because the APRM has been rescaled, there is only a small effect on the RBM performance due to the LPRM performance at the higher average local flux. The change in performance does not have a significant effect on the overall RBM performance.

The RBMs installed at Hope Creek are in accordance with the requirements established by the GE design specifications. ((

1))

5.1.2 Rod Worth Minimizer/Rod Control and Information System The Rod Control and Information System (RCIS) is not applicable to Hope Creek.

The RWM is a normal operating system that does not perform a safety related function. The function of the RWM is to support the operator by enforcing rod patterns until reactor power has reached the appropriate level. ((

))The power-dependent instrument setpoints for the RWM are included in the plant Technical Specifications (see Section 5.3.4).

((I 5-3

NEDO-33076 11 5.2 BOP MONITORING AND CONTROL Operation of the plant at CPPU has minimal effect on the balance-of-plant (BOP) system instrumentation and control devices. Based on CPPU operating conditions for the power conversion and auxiliary systems, most process control valves and instrumentation have sufficient range/adjustment capability for use at the CPPU conditions. However, some (non-safety) modifications may be needed to the power conversion systems to obtain CPPU RTP. No safety-related BOP system setpoint change is required as a result of the CPPU, with the exception of main steam line high flow discussed in Section 5.3.1. The topics considered in this section are:

Topic ,'CLTR2Disposition 'Hope Creek Result 5.2.1 Pressure Control System 5.2.2 Turbine Steam Bypass System (Normal Operation) 5.2.2 Turbine Steam Bypass System (Safety Analysis) 5.2.3 Feedwater Control System (Normal Operation) 5.2.3 Feedwater Control System (Safety Analysis) 5.2.4 Leak Detection System ))

5.2.1 Pressure Control System

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-"i:'-

5-4

NEDO-33076 The PCS is a normal operating system to provide fast and stable responses to system disturbances related to steam pressure and flow changes to control reactor pressure within its normal operating range. This system does not perform a safety function. Pressure control operational testing is included in the CPPU implementation plan as described in Section 10.4 to ensure adequate turbine control valve pressure control and flow margin is available.

5.2.2 Turbine Steam Bypass System

[1 The Turbine Steam Bypass System is a normal operating system that is used to bypass excessive steam flow. The absolute flow capacity of the bypass system is unchanged. The bypass flow capacity is included in some anticipated operational occurrence (AOO) evaluations (Section 9.1).

((

))

5-5

NEDO-33076 5.2.3 Feedwater Control System

((

The Feedwater Control System is a normally operating system to control and maintain the reactor vessel water level. CPPU results in an increase in feedwater flow. Feedwater control operational testing is included in the CPPU implementation plan as described in Section 10.4 to ensure that the feedwater response is acceptable. Failure of this system is evaluated in the reload analysis for each reload core with the feedwater controller failure maximum demand event. A LOFW event can be caused by downscale failure of the controls. The LOFW is discussed in Section 9.1.

((

1]

5.2.4 Leak Detection System

/:,;~ I-.'-:-'.'-,,,':.',

5-6

NEDO-33076 1]

The only effect on the LDS due to CPPU is a slight increase in the feedwater temperature and steam flow. ((

)) The increased feedwater temperature results in a small increase in the main steam tunnel temperature (< 0.50 F).

Mi

)) Main steam line high flow is discussed in Section 5.3. 1.

5-7

NEDO-33076

))

5.3 TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS Technical Specifications instrument allowable values and/or setpoints are those sensed variables, which initiate protective actions and are generally associated with the safety analysis. Technical Specification allowable values are highly dependent on the results of the safety analysis. The safety analysis generally establishes the analytical limits. The determination of the Technical Specification allowable values and other instrument setpoints includes consideration of measurement uncertainties and is derived from the analytical limits. The settings are selected with sufficient margin to minimize inadvertent initiation of the protective action, while assuring that adequate operating margin is maintained between the system settings and the actual limits.

There is typically substantial margin in the safety analysis process that should be considered in establishing the setpoint process used to establish the Technical Specification allowable values and other setpoints.

Increases in the core thermal power and steam flow affect some instrument setpoints. These setpoints are adjusted to maintain comparable differences between system settings and actual limits, and reviewed to ensure that adequate operational flexibility and necessary safety functions are maintained at the CPPU RTP level. Where the power increase results in new instruments being employed, an appropriate setpoint calculation is performed and Technical Specification changes are implemented, as required. If there is no change in the instrument equipment, the simplified process outlined in the CLTR may be used to determine the instrument allowable value and setpoint.

Per the CLTR, [

)) The justification for implementing this simplified process for the individual Technical Specification setpoints is provided for each instrument below. ((

5-8

NEDO-33076 Table 5-1 summarizes the current and CPPU Analytical Limits for Hope Creek.

The topics addressed in this evaluation are:

Topic CLTR'Disposition Hope eek esult 5.3.1 Main Steam Line High Flow ((

Isolation - Setpoint Calculation Methodology 5.3.1 Main Steam Line High Flow Isolation - Setpoint Value 5.3.2 Turbine First-Stage Pressure Scram Bypass - Setpoint Calculation Methodology 5.3.2 Turbine First-Stage Pressure Scram Bypass - Setpoint Value 5.3.3 APRM Flow-Biased Scram -

Setpoint Calculation Methodology 5.3.3 APRM Flow-Biased Scram -

Setpoint Value 5.3.4 Rod Worth Minimizer/ RCIS Rod Pattern Controller Low Power Setpoint - Setpoint Calculation Methodology 5.3.4 Rod Worth Minimizer/ RCIS Rod Pattern Controller Low Power Setpoint - Setpoint Value 5.3.5 Rod Block Monitor 5.3.6 RCIS Rod Withdrawal Limiter High Power Setpoint 5.3.7 APRM Setdown in Startup Mode -

Setpoint Calculation Methodology 5.3.7 APRM Setdown in Startup Mode -

Setpoint Value 5-9

NEDO-33076 The instrument function analytical limit (AL) is the value used in the safety analyses to demonstrate acceptable nuclear safety system performance is maintained. The allowable value (AV) and nominal trip setpoints (NTSP) are then chosen/calculated such that the instrument will function before reaching the AL under the worst-case environmental/event conditions. The instrument setpoints account for the measurable instrument characteristics (e.g., drift, accuracy, and repeatability).

5.3.1 Main Steam Line High Flow Isolation The main steam line high flow isolation setpoint is used to initiate the isolation of the Group I primary containment isolation valves. The only safety analysis event that credits this trip is the main steam line break accident. For this accident, there are diverse trips from high area temperature and high area differential temperature. For Hope Creek, there is sufficient margin to choke flow, so the analytical limit (AL) for CPPU is maintained at the current percent of rated steam flow in each main steam line.

For Hope Creek, the AL of 140% of steam flow is not changed and no new instrumentation is required (the existing instrumentation has the required upper range limit to re-span the instrument loops to accommodate the new setpoint). Therefore, a new setpoint is calculated 5-10

NEDO-33076 using the methodology as noted in Section 5.3 and the appropriate Technical Specifications changes have been provided.

5.3.2 Turbine First-Stage Pressure Scram and Recirculation Pump Trip Bypass CPPU results in an increased power level and the high-pressure turbine (HPT) modifications result in a change to the relationship of turbine first-stage pressure to reactor power level. The turbine first-stage pressure setpoint is used to reduce scrams and recirculation pump trips at low power levels where the turbine steam bypass system is effective for turbine trips and generator load rejections. In the safety analysis, this trip bypass only applies to events at low power levels that result in a turbine trip or load rejection. ((

)) It is further reduced to be consistent with the power level at which the Technical Specification Thermal Limits must be monitored. This reduction is in the conservative direction.

((

)) Therefore, a new setpoint was calculated and the appropriate Technical Specifications AV change has been provided.

To assure that the new value is appropriate, CPPU plant ascension startup test or normal plant surveillance will be used to validate that the actual plant interlock is cleared consistent with the safety analysis.

5.3.3 APRM Flow Biased Scram

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.' .' ':-" .:'...,,:':#-"'".'.'. '; c . i' ' : , .: "'. "i;--." -'..  :-''-1 5-11

NEDO-33076 The NTSPs were adjusted by the same difference as the changes in the AVs. This allows the current license basis to be maintained through the application of the same uncertainties in the same manner as previous setpoint evaluation.

5.3.4 Rod Worth Minimizer/RCIS Rod Pattern Controller Low Power Setpoint The RCIS Rod Pattern Controller is not applicable to Hope Creek.

The Rod Worth Minimizer low power setpoint (LPSP) is used to bypass the rod pattern constraints established for the control rod drop accident at greater than a preestablished low power level. The measurement parameter is steam flow.

((

)) This approach does not affect the limitations on the sequence of control rod movement to the absolute core power level for the LPSP associated with the requirements of the control rod drop accident. The RWM main steam instrumentation is being replaced to provide adequate measurement range for CPPU and therefore, a new setpoint was calculated and the appropriate Technical Specifications change will be provided.

5.3.5 Rod Block Monitor The RBM rod block is no longer credited in the evaluation of the control rod withdrawal error as described in Section 4.0 of Reference 29.

5.3.6 RCIS Rod Withdrawal Limiter High Power Setpoint The RCIS Rod Withdrawal Limiter High Power Setpoint is not applicable to Hope Creek.

5-12

NEDO-33076 5.3.7 APRM SetdoNvn in Startup Mode

((

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The value for the Technical Specification safety limit for the reduced pressure or low core flow condition is established to satisfy the fuel thermal limits monitoring requirements described in Section 9.1.1. Because the thermal margin monitoring requirement is reduced from 25% to 24%,

the APRM Scram Setdown AV is reduced the same amount, i.e., from 20% to 19%. Similarly, the APRM Rod Block Setdown AV is reduced from 14% to 13% to maintain the same margin between the Rod Block and Scram AVs for operational flexibility. The Scram and Rod Block Setdown NTSPs are reduced by the same amount. This allows the current license basis to be maintained through the application of the same uncertainties in the same manner as the previous setpoint evaluation.

5.4 CHANGES TO INSTRUMENTATION AND CONTROLS In the CLTR SER, the staff requested that the plant -specific submittal address all CPPU-related changes to instrumentation and controls, such as scaling changes, changes to upgrade obsolescent instruments, and changes to the control philosophy. Tables 5-2 through 5-4 provide this information.

5-13

NEDO-33076 Table 5-1 Hope Creek Analytical Limits For Setpoints Analytical Limits Parameter Current CPPU APRM Calibration Basis 3339 MNWt 3840 MWt APRM Flow Biased Simulated Thermal Power Upscale Scram' TLO Clamped AV (%RTP) 115.5 No change 2 TLO Flow Biased (%RTP)3 AV 0.66 Wd +69 AV0.57Wd+61 SLO Flow Biased (%RTP) 3 AV 0.66 (Wd-AV) +69 AV 0.57 (Wd-AW) + 61 APRM Flow Biased Neutron Flux Upscale (Rod Block) '

Clamped AV (%RTP) 111 No change2 TLO Flow Biased (%RTP) 3 AV 0.66 Wd +60 AV0.57Wd+56 SLO Flow Biased (%RTP) 3 AV 0.66 (Wd-AW) +60 AV 0.57 (Wd-AW) +56 AW =9 % for SLO APRM Setdown (%RTP)

Scram - AV 20 19 Rod Block - AV 14 13 Rod Block Monitor Low Power Setpoint (%RTP) 30 No Change Rod Worth Minimizer (%RIT) 10 8.6 Main Steam Line High Flow (% rated steam flow) 140 140 Main Steam Line High Flow (differential pressure) 114.7 psid 176.2 psid Turbine First-Stage Pressure Scram 30 24 and Recirculation Pump Trip Bypass (%RTP) 1 Hope Creek does not have ALs for this setpoint function 2 (( 11 3 No credit is taken in any safety analysis for the flow referenced setpoints.

5-14

NEDO-33076 Table 5-2 Hope Creek Instrument Scaling Changes for CPPU Parameter /,Device CurrentRange CPPU Range MSL Flow to NSSS Isolation Logic (psid (Mlb/hr)) 0 - 150 0 - 267.73 (0-5.76) (0 - 6.709)

MSL Flow to Digital FW Control System (psid (Mlb/hr)) 0 - 76.96 0 - 116.07 (0 - 4.25) (0 - 5.00)

MSL Flow Recorders, Indicators, computer points (Total in 0- 17 0 -20 Mlb/hr)

FW Flow Transmitters, computer points (Individual in Mlb/hr) 0 - 8.5 0 - 10 FW Flow Recorder, computer points (Total in Mlbfhr) 0 - 17 0 - 20 Condensate Pre-Filter Flow (gpm) 0 - 11,000 0 - 14,000 Condensate Demineralizer Flow (gpm) 0 - 6,000 0 - 6,500 5-15

NEDO-33076 Table 5-3 Hope Creek Instrument Setpoint Changes for CPPU Parameter /IDevice . - - -  : - -. :Current - CPPU

'Nominal Nominal

Setpoint  :'Setpoint Hydrogen Water Chemistry Injectioni (Mlb/hr) 4.321 5.022 Primary Condensate Pump 75% Permissive 2 (% FW flow) 63.3 60.8 Secondary Condensate Pump 85% Permissive2 (% FW flow) 71.8 69.5 Condenser Variable Exhaust Pressure Alarm (inches HgA) 5.0 - 5.5 5.5 - 6.0 Condensate Pre-Filter System Differential Pressure (psid). 15 18 Off-gas Treatment System High Flow Alarm (scfh) 4,000 4,400 Stator Water High Bulk Water Temperature Alarm 3 (C) 77 74 Stator Water High Bulk Water Temperature Runback3 CC) 82 79 Stator Bar High Outlet Temperature Alarm3 (C) 82 79 Stator Between Bar High Outlet Temperature 3 (C) 77 74 Stator Water Cooling Low Flow Alarm 3 (gpm) 548.5 712 Stator Water Cooling Inlet Low Pressure Alarm 3 (psig) 15 712 gpm4 Stator Water Cooling Inlet Low Pressure Runback 3 (psig) 13 673 gpm 4

1. The Digital FW Control System processes the signals from both FW flow transmitters for the Hydrogen Water Chemistry Injection setpoint. The setpoint remains the same in % RTP (30% RTP) but is revised in terms of FW flow because of the increase in total rated FW flow.

2. The Digital FW Control System processes the signals from both FW flow transmitters for the permissives for the Primary and Secondary Condensate Pumps. The setpoints are revised because of the increase in total rated FW flow and full-scale range.

3. The Stator Water Cooling System components were replaced during RF12 as a part of the Low Pressure Turbine Replacement.

4. Field Installation Testing will determine the actual pressure (psig) corresponding to the specified flow (gpm) requirement.

5-16

NEDO-33076 Table 54 Hope Creek Instrument Replacements for CPPU Parameter /Device CPPU Change, MSL Flow Transmitter The current transmitters have a range limit of 0 - 100 psid and cannot accommodate the CPPU calibrated range of 0 - 116.7 psid.

They will be replaced with a model with 0 -

300 psid range.

Condensate Pre-filter Flow Element The current flow elements have a range limit of 0 - 11,000 gpm and cannot accommodate the CPPU calibrated range of 0 - 14,000 GPM. These will be replaced with a model capable of 0 - 14,000 GPM range.

Stator Water Cooling System' Stator Water pressure and flow increase to obtain the increased generator rating. The following stator water cooling system l&C components were replaced: Flow Orifice, Flow Meter, and Winding Inlet Pressure Gauge.

High Pressure Turbine Instrumentation HP Turbine Replacement DCP will identify the instruments to be replaced.

1. The Stator Water Cooling System components were replaced during R12 as a part of Low Pressure Turbine Replacement.

5-17

NEDO-33076

6. ELECTRICAL POWER AND AUXILIARY SYSTEMS This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapters 8 and 9, that applies to CPPU.

6.1 AC POWER The Hope Creek AC power supply includes both off-site and on-site power. The on-site power distribution system consists of transformers, buses, and switchgear. Alternating current (AC) power to the distribution system is provided from the transmission system or from onsite Diesel Generators.

Plant electrical characteristics are given in Table 6-1. The topics addressed in this evaluation are:

Topic -CLTR Disposition Hope Creek Result AC power (degraded voltage) ((

AC power (normal operation) 6.1.1 AC Power (degraded voltage)

The on-site power distribution system loads were reviewed under normal and emergency operating scenarios for CPPU conditions. Loads were computed based on equipment nameplate data or brake horsepower (BHP). These loads were used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the CPPU conditions is achieved for normal and emergency conditions by operating equipment within the nameplate rating running kW or BHP.

There is no significant change in electrical demand load associated with the power generation system; therefore, the existing load flow and short circuit calculations verify the adequacy of the on-site AC system for the proposed changes. The existing protective relay settings are adequate to accommodate the increased load on the 7.2 KV and 4.16 KV power systems. Selective coordination is maintained between the M-G Set and pump motor feeder breakers and the 7.2 KV/4.16 KV switchgear main feeder breakers.

Station loads under emergency operation/distribution conditions (Emergency Diesel Generators) are based on equipment nameplate data, except for the ECCS pumps where a conservatively high flow BHP is used. CPPU conditions are achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the required pump motors; therefore, under emergency conditions the electrical supply and distribution components are adequate.

((I

))The systems have sufficient capacity to support all required loads to achieve and maintain safe shutdown conditions and to operate the ECCS equipment following postulated accidents and transients.

6-1

NEDO-33076 A grid stability analysis has been performed, considering the increase in electrical output, to demonstrate conformance to General Design Criteria 17 (10 CFR 50, Appendix A). The analysis establishes grid voltage schedules, generator reactive power limits and reduced generation limits that are required under certain pre-event outages. At CPPU RTP, the reactive limits of the generator are +315 / - 428 MVAR. The minimum limit of 315 MVAR is required to maintain generator stability. A new 500 KV circuit breaker has been added to the Hope Creek switchyard to improve system stability. The voltage schedules and reactive power limits will be incorporated into operating guides prior to CPPU operation.

6.1.2 AC Power (normal operation)

The existing off-site and on-site electrical equipment was determined to be adequate for normal operation with the uprated electrical output as shown in Table 6-2. The review concluded the following:

• The Isolated Phase Bus Duct will be upgraded to 34,000 Amperes, to accommodate the higher Generator output at the CPPU conditions.

• The BHP of the recirculation MG set motors increases 6.0% for CPPU but remains within its nameplate capability.

• The electrical demand load associated with power generation system motors for the primary and secondary condensate pumps will increase for CPPU but will remain within their nameplate capability. These system pumps experience increased flow demand at CPPU conditions.

• The existing main power transformers have been replaced for operation with the CPPU-related electrical output of the upgraded generator.

• The existing 500 KV switchyard buses, breakers, and switches are adequate for CPPU operations.

• The protective relaying for the main generator, transformer, and switchyard is adequate for the CPPU generator output. Setpoint changes (e.g. out of step and overload) will be required.

• The main generator rating has been increased to 1373 MVA for CPPU. As a result stator cooling water system has been upgraded.

6.2 DC POWER The Hope Creek DC power distribution system provides control and motive power for various systems/components within the plant. The topics addressed in this evaluation are:

Topic -'CLTR Disposition Hope Creek Result-DC power requirements -

6-2

NEDO-33076 The direct current (DC) loading requirements in the UFSAR were reviewed, and no reactor power-dependent loads were identified.

The DC power distribution system provides control and motive power for various systems/components within the plant. In normal and emergency operating conditions, loads are computed based on equipment nameplate ratings. These loads are used as inputs for the computation of load, voltage drop, and short circuit current values.

Operation at the CPPU conditions does not increase any load beyond nameplate rating or revise any component operating duty cycle; therefore, the DC power distribution system remains adequate.

6.3 FUEL POOL The Hope Creek fuel pool systems consist of storage pools, fuel racks, the Fuel Pool Cooling and Cleanup system (FPCC). The objective of the fuel pool system is to provide specially-designed undenvater storage space for the spent fuel assemblies. The objective of the fuel pool systems is to remove the decay heat from the fuel assemblies and maintain the fuel pool water within specified temperature limits. The effects of CPPU on the Hope Creek fuel pool are addressed in the following evaluation:

Topic CLTR Disposition lHope Creek Res' 6.3.1 Fuel Pool Cooling (normal core offload) ((

6.3.1 Fuel Pool Cooling (full core offload) 6.3.2 Crud Activity and Corrosion Products 6.3.3 Radiation Levels 6.3.4 Fuel Racks 6.3.1 Fuel Pool Cooling The Hope Creek spent fuel pool (SFP) bulk water temperature must be maintained below the licensing limit of 150 ° F. The evaluation is based on a Safety Auxiliaries Cooling System (SACS) temperature of 95 IF, which is the maximum (non-accident) temperature based on the TS UHS temperature limit, minimum cooling flow rates, and minimum heat exchanger performance. The limiting condition is a full core discharge with all remaining storage locations filled with used fuel from prior discharges. A normal batch offload is assumed for outage planning. The batch offload was analyzed with both trains of FPCC operating, and assumes a single failure of RHR fuel pool cooling assist mode. The full core offload scenario was analyzed with both trains of the FPCC system in operation and one train of RHR in fuel pool cooling assist mode, without assuming a single failure. The full core offload scenario considers two plant configurations. One configuration maintains one train of RHR in the shutdown cooling mode for core cooling, and the other configuration takes advantage of natural circulation for core 6-3

NEDO-33076 cooling without using the RHR loop. The key results of these analyses are presented in Table 6-

3. The operational temperature requirement of less than 135 ° F assures operator comfort for batch offload, and provides ample margin against an inventory loss in the fuel pool due to evaporation or boiling.

The CPPU SFP heat load is higher than the CLTP heat load. The CPPU heat loads at the limiting full core offload condition and the normal batch offload are calculated and then the bulk pool temperature is determined to evaluate the FPCC system adequacy. The CPPU does not affect the heat removal capability of the FPCCS or the fuel pool cooling assist mode of the RHR system. The CPPU results in slightly higher core decay heat loads during refueling. Each reload affects the decay heat generation in the SFP after a batch discharge of fuel from the reactor. The full core offload heat load in the SFP reaches a maximum some time after the full core discharge.

Based on the heat load evaluations, the SFP bulk temperature remains less than 150 0 F and is acceptable for CPPU conditions.

The SFP normal makeup source is from the Seismic Category II Condensate Storage system, with a capacity of 75 gpm and is not affected by CPPU and remains adequate for CPPU conditions for non-accident cases.

In the unlikely event of a complete loss of SFP cooling capability, the SFP would reach the boiling temperature in 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in the worst case condition after the limiting full core offload.

The boil-off rate would be 130 gpm (see Table 6-3). Three Seismic Category I emergency makeup sources, the Low Pressure Coolant Injection (LPCI), Emergency fire makeup system, and the Service Water System, have a makeup capability greater than 130 gpm.

Existing plant instrumentation and procedures provide adequate indications and direction for monitoring and controlling SFP temperature and level during normal batch offloads and the case of the limiting full core offload. Symptom based operating procedures exist to provide mitigation strategies including placing additional cooling trains or systems in service, stopping fuel movement, and initiating make-up if necessary. The symptom based entry conditions and mitigation strategies for these procedures do not require changes for CPPU.

6.3.2 Crud Activity and Corrosion Products The total crud in the SFP increases slightly, assuming that all residual crud in the Reactor Coolant System (RCS) is transported to the SFP. However, the increase is negligible, and SFP water quality is maintained by the FPCC system.

6.3.3 Radiation Levels The normal radiation levels around the SFP are expected to increase slightly, primarily during fuel handling operation. Radiation levels in those areas of the plant directly affected by the reactor core and spent fuel will increase by the percentage increase in the average power density of the fuel bundles. Therefore, for a CPPU increase of 15%, the radiation dose rates increase by 15%.

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NEDO-33076 The shielding design of spent fuel pools is typically very conservative from the perspective of radiation exposure such that changes in the fuel inventory / bundle surface dose rate of 15%

result in inconsequential changes in occupational dose. The post-CPPU radiation exposures around the SFP including the accessible areas adjacent to the sides or bottom of the SFP are expected to be within the allowable dose rate limit of the existing radiation zone designation.

The expected increase in post-CPPU radiation dose rate 2 feet above the refueling platform is less than 15% of the existing dose rate during the transition of spent fuel assemblies from the RPV to the SFP. Such changes will have little effect on plant operations or ALARA exposure.

Hope Creek radiation protection procedures and the radiation monitoring program would detect any changes in radiation levels and initiate appropriate actions.

6.3.4 Fuel Racks The increased decay heat from the CPPU results in a higher heat load in the fuel pool during long-term storage. The fuel racks are designed for higher temperatures (212 0 F) than the licensing limit of 150 ° F. The FPCC system, in conjunction with the RHR fuel pool cooling assist mode, assures that the licensing limit is maintained.

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NEDO-33076 6.4 WATER SYSTEMS The Hope Creek water systems are designed to provide a reliable supply of cooling water for normal operation and design basis accident conditions. The topics addressed in this evaluation are:

Topic -;CLTR Disposition - Hope Creek Result Water systems performance (safety related)

Water systems performance (normal operation)

Suppression pool cooling (RHR service operation)

Ultimate heat sink 6.4.1 Cooling BWater Systems The Hope Creek cooling water systems include a non-safety related circulating water system to transfer the heat from the main condenser to a cooling tower and a once-though, safety related Station Service Water System (SSWS) to remove the heat from the closed loop Safety Auxiliaries Cooling System (SACS), the closed loop non-safety related Reactor Auxiliaries Cooling System (RACS), and the closed loop non-safety Turbine Auxiliaries Cooling System (TACS). The SSWS uses the Delaware River as the UHS.

6.4.1.1 Safety-Related Loads 6.4.1.1.1 Station Service Mater System The safety-related SSWS is designed to provide a reliable supply of cooling water to the SACS during normal operation, normal shutdown, loss-of-offsite power (LOOP) and following a loss-of-cooling accident (LOCA). The SSWS also provides a reliable supply of cooling water to the RACS during normal operation and during LOOP events without a LOCA.

The CPPU effect is bounded by the LOCA analysis. The SSWS contains sufficient redundancy in pumps and heat exchangers to assure that adequate heat removal capability is available during all modes of operation at CPPU.

The CPPU LOCA evaluation maintained the same assumptions on components cooled, flow rates, and methodology as the CLTP analysis. The LOCA evaluation conservatively assumes only one SACS loop in operation with two SACS pumps and heat exchangers; it then determines the maximum allowable SSWS temperature that removes the required heat and maintains the SACS supply temperature at or below its allowable maximum.

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NEDO-33076 As discussed in Section 6.4.1.1.2, the total CPPU LOCA heat loads from SACS were not increased over the conservative values assumed by the CLTP analysis.

6.4.1.1.2 Safety'Auxiliaries CoolingSystem The safety-related Safety Auxiliaries Cooling System (SACS) provides cooling for the following equipment and systems during and following the most demanding design basis event, the LOCA:

RHR Heat Exchangers RHR pump seal and motor bearing coolers Diesel Generator Coolers Diesel Generator Room Coolers RHR Pump room coolers HPCI pump room coolers RCIC pump room coolers CS pump room coolers Filtration Recirculation and Ventilation System coolers Class IE equipment chillers Control room chillers Containment instrument gas compressor Post accident sampling station Spent Fuel Pool Heat Exchangers (optional load)

The diesel generator loads, gas compressor loads, RHR Pump Seal loads, Chillers, and FRVS system loads remain unchanged for LOCA conditions following CPPU operation. The calculated SACS cooling .loads for the RHR Heat Exchanger and the ECCS room coolers increase only minimally because of the small change from the assumed suppression pool temperature of 212.0'F at CLTP to the calculated temperature of 212.30 F at CPPU.

The SACS LOCA heat load calculation conservatively assumes that SFP cooling is not shed; however, an over conservatism was removed from this assumption. The CLTP LOCA calculation assumed the maximum SPF heat load immediately following a full fuel offload. The CPPU calculation credits the delay between offload and returning to power operation. This change results in a lower CPPU SFP heat load as well as no net increase in the total SACS LOCA heat load assumed between CLTP and CPPU. The containment cooling analysis in Section 4.1.1 shows that 6-7

NEDO-33076 the post-LOCA RHR heat load increases due to an increase in the maximum suppression pool temperature that occurs following a LOCA. The post-LOCA containment and suppression pool responses have been calculated based on an energy balance between the post-LOCA heat loads and the existing heat removal capacity of the RHR system and SACS. As discussed in Sections 3.5.2 and 4.1.1, the existing suppression pool structure and associated equipment have been reviewed for acceptability based on this increased post-LOCA suppression pool temperature.

Therefore, the containment cooling analysis and equipment review demonstrate that the suppression pool temperature can be maintained within acceptable limits in the post-accident condition at CPPU based on the existing capability of the SACS and SSWS. The SACS system has sufficient capacity at CPPU to supply adequate cooling to the spent fuel pool heat exchangers. In addition, the SACS system has sufficient capacity to serve as a standby coolant supply for long term core and containment cooling as required for CPPU conditions.

The SACS is used to supply flow to the Turbine Auxiliary Cooling System (TACS) during normal operating conditions. The TACS flow rate is slightly increased for CPPU, but normal operation is not the limiting SACS condition. The SACS to TACS flow path is isolated under LOOP or LOCA conditions.

6.4.2 Main Condenser/Circulating Water/Normal Heat Sink Performance The main condenser, circulating water, and heat sink systems are designed to remove the heat rejected to the condenser and thereby maintain adequately low condenser pressure as recommended by the turbine vendor. Maintaining adequately low condenser pressure assures the efficient operation of the turbine-generator and minimizes wear on the turbine last stage buckets.

CPPU operation also increases the heat rejected to the condenser and, therefore, reduces the difference betveen the operating pressure and the recommended maximum condenser pressure. If condenser pressures approach the main turbine backpressure limitation, then reactor thermal power reduction would be required to reduce the heat rejected to the condenser and maintain condenser pressure within the main turbine requirements.

The main condenser and circulating water system are not being modified for CPPU operation. In anticipation of CPPU, the cooling tower (heat sink) was modified in the spring of 2003 to improve the water spray distribution across the cooling tower cross-area in order to improve the tower's performance. The performance of these systems was evaluated for CPPU. This evaluation was based on a design duty over the actual range of circulating water inlet temperatures, and confirms that the condenser, circulating water system, and heat sink are adequate for CPPU operation. These systems maintain adequate condenser backpressure on all but the most severe summertime conditions. Under the most severe summer conditions, a slight power reduction may be required to avoid exceeding the allowable condenser backpressure. The evaluation of these systems at CPPU conditions indicates that the plant continues to meet all environmental permit conditions related to the UHS (Delaware River) and the plant cooling towers. The effect of CPPU on the flooding analyses is addressed in Section 10.1.2.

6-8

NEDO-33076 6.4.2.1 Circulating Water System Transient Operation The loss of one or more circulating water (CW) pumps results in high condenser pressure potentially resulting in a turbine trip. A Reactor Recirculation System (RRS) runback is provided to reduce the potential for a turbine trip by rapidly reducing reactor power and core flow to compensate for the loss of one or more of the CW pumps. The RRS runback has two speed limiters, one set at 45% speed (Intermediate runback) and the other is set at 30% speed (Full runback). The RRS Intermediate runback results in approximately 70% power and 60%

core flow. The RRS Full runback results in approximately 60% power and 50% core flow.

The current RRS runback logic related to the CW System and the planned change for CPPU conditions are as follows:

Event CLTP CPPU Loss of I CW pump with 4 CW pumps initially Intermediate No Change operating concurrent with condenser backpressure Runback

> 4.5 In HgA Loss of I or more CW pumps resulting in 2 or less Full Runback No Change CW pumps operating concurrent with condenser backpressure >5.8 In HgA Condenser backpressure > 6.0 In HgA with 3 or 4 None Intermediate CW pumps initially running Runback For CPPU operation, a RRS Intermediate runback is added to reduce the potential of a low vacuum turbine trip. Currently, operators have the option of initiating a manual turbine trip if condenser backpressure exceeds 6.5 In HgA. The RRS runback on increased condenser pressure

(> 6.0 In HgA) with 3 or 4 CW pumps running is initiated in anticipation of further degradation potentially causing a manual or automatic turbine trip.

Hope Creek will perform a comprehensive Dynamic Analysis to verify the RR runback capability at CPPU conditions for all line-up configurations. The Dynamic Analysis will use the latest best-estimate Thermal-Hydraulic model (THOR-BOP) recently installed at the HC Simulator. The changes to the RRS runback logic will be performed as part of the CPPU implementation. Proper operation of the RRS runback logic will be verified in a functional test as part of the implementation of the design change.

6.4.3 Reactor Auxiliaries Cooling System The heat loads on the Reactor Auxiliaries Cooling System (RACS) slightly increase (<0.2% for normal operations and for a Loss of Offsite Power without a LOCA condition) as a result of CPPU. The increase in the RACS heat loads results from an increase in the pumping power input to the Reactor Recirculation system pumps and a minimal increase in the drywell cooling loads. However, the total RACS heat load at CPPU remains bounded by the design heat load.

The flow rates in the systems cooled by the RACS do not change significantly due to CPPU (e.g., Recirculation < 5% and RWCU pumps cooling 0%) and, therefore, are minimally affected by CPPU. The operation of the remaining equipment cooled by the RACS (e.g., sample coolers 6-9

NEDO-33076 and drain sump coolers) is not power-dependent and is not affected by CPPU. The RACS contains sufficient redundancy in pumps and heat exchangers to assure that adequate heat removal capability is available during normal operation. Sufficient heat removal capacity is available to accommodate the RACS heat load at CPPU conditions.

6.4.4 Turbine Auxiliaries Cooling System The heat loads on the TACS which are power-dependent and are increased by CPPU, include those related to the operation of the generator stator coolers, iso-phase bus heat exchanger, the Condenser Compartment Unit Coolers and Fans, and the Turbine Building Chiller Condensers and Pump Out Unit Coolers. Because the TACS flow to these components can be increased to compensate for the increased heat load, there is no increase in TACS operating temperature at CPPU conditions.

6.4.5 Ultimate Heat Sink The ultimate heat sink (UHS) is the Delaware River. The UHS intake temperature is unaffected by operations at CPPU conditions.

The existing UHS system provides a sufficient quantity of water at a temperature within Technical Specification limits-to perform its safety related functions at CPPU.

6.5 STANDBY LIQUID CONTROL SYSTEM The Standby Liquid Control System (SLCS) is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that all or some of the control rods cannot be inserted. This system pumps a highly enriched sodium pentaborate solution into the vessel, to provide neutron absorption and achieve a subcritical reactor condition. SLCS is designed to inject over a wide range of reactor operating pressures. The following topics are addressed in this evaluation:

Topic --ICLTR Disposition 'Hope Creek'Result Core shutdown margin System performance and hardware Suppression pool temperature following limiting ATWS events 6-10

NEDO-33076 The boron injection rate requirement for maintaining the peak suppression pool water temperature limits, following the limiting ATWS event with SLCS injection, is not increased for CPPU.

Based on the results of the plant specific ATWS analysis, the maximum reactor upper plenum pressure following the limiting ATWS event reaches 1179 psia during the time the SLCS is analyzed to be in operation. Consequently, there is a corresponding increase in the maximum pump discharge pressure and a decrease in the operating pressure margin for the pump discharge relief valves. The pressure margin for the pump discharge relief valves remains above the minimum value needed to assure that the relief valves remain closed during system injection. In the event that the SLCS is initiated before the time that the reactor pressure recovers from the first transient peak, resulting in opening of the SLC relief valves, the reactor pressure must reduce sufficiently to ensure SLC relief valve closure. Analysis results indicate that the reactor pressure reduces sufficiently from the first transient peak to allow the SLC relief valves to close.

The SLCS ATWMS performance is evaluated in Section 9.3.1 ((

)) for CPPU. The evaluation shows that CPPU has no adverse effect on the ability of the SLCS to mitigate an ATWS.

6.6 POWER DEPENDENT HVAC The heating ventilation and air conditioning (HVAC) systems consist mainly of heating, cooling supply, exhaust, and recirculation units in the turbine building, reactor building, and the drywell.

CPPU results in slightly higher process temperatures and small increases in the heat load due to higher electrical currents in some motors and cables. The topics addressed in this evaluation are:

I -Topic. l CLTR Disposition Hope Creek Result Power dependent HVAC performance The affected areas are the drywell, the steam tunnel, and the ECCS rooms in the reactor building; and the moisture separator areas, the feedwater heater rooms, condenser area, condensate pump areas, and the steam driven feedwater pump rooms in the turbine building. Other areas in the reactor building and the turbine building are unaffected by the CPPU because the process temperatures remain relatively constant.

The increased heat loads during normal plant operation result in an insignificant (0.21F) increase in the drywell and < 0.5 0F increase the main steam tunnel. In the turbine building, the maximum temperature increase in the moisture separator areas, feedwater heater rooms, condenser area, condensate pump areas, and the steam driven feedwater pump rooms is 3.5 0 F.

The small increase in the post LOCA suppression pool temperature, from the assumed peak of 212 0 F to the calculated CPPU temperature of 212.30 F, results in a negligible (-0.01 0F) increase in the ECCS room temperatures during a LOCA.

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NEDO-33076 Other HVAC systems were reviewed and are unaffected. Because CPPU does not result in any significant building temperature changes, changes to the heat transfer between buildings are insignificant. The Diesel Generator remains below rated capacity and there are essentially no electrical load or process temperature changes in the design basis heat load for this area; therefore, there is no increase in the design basis heat load in this area. The SFP area is within the Reactor Building and is serviced by the Reactor Building Ventilation. CPPU does not add any equipment in the SFP area nor change the SFP maximum process temperature. Therefore, CPPU does not adversely affect the normal or accident SFP heat loads to the Reactor Building Ventilation System. CPPU does result in a small increase in the volume of liquid and solid radwaste, but these do not affect the process temperature or electrical load changes. CPPU does increase the amount of hydrogen gas production by radiolysis, but the amount remains within the original design for the Off-Gas System recombiners. Therefore, the ventilation in the radwaste handling areas is not adversely affected by CPPU.

Based on a review of design basis calculations and design temperatures, the design of the HVAC is adequate for the CPPU.

6.7 FIRE PROTECTION This section addresses the effect of CPPU on the fire protection program, fire suppression and detection systems, and reactor and containment system responses to postulated 10 CFR 50 Appendix R fire events. The topics addressed in this evaluation are:

-Topic - CLTR'Disposition Hope Creek Result Fire suppression and detection systems Operator response time Peak cladding temperature Vessel water level Suppression pool temperature

((

)) Any changes in physical plant configuration or combustible loading as a result of modifications to implement the CPPU will be evaluated in accordance with the administrative controls in the plant modification and fire protection programs. These administrative control programs are not affected by CPPU. The safe shutdown systems and equipment used to achieve and maintain cold shutdown conditions do not change, and are adequate for the CPPU conditions. The operator actions required to mitigate the consequences of a fire are not affected.

Therefore, the fire protection systems and analyses are not affected by CPPU.

The reactor and containment response to the postulated 10 CFR 50 Appendix R fire event at CPPU conditions is evaluated in Section 6.7.1. The results show that the peak fuel cladding temperature, reactor pressure, and containment pressures and temperatures are below the acceptance limits and demonstrate that there is sufficient time available for the operators to 6-12

NEDO-33076 perform the necessary actions to achieve and maintain cold shutdown conditions. Therefore, the fire protection systems and analyses are not adversely affected by CPPU.

6.7.1 10 CFR 50 Appendix R Fire Event A plant-specific evaluation was performed to demonstrate safe shutdown capability in compliance with the requirements of 10 CFR 50 Appendix R assuming CPPU conditions. The limiting Appendix R fire event was analyzed assuming CLTP and CPPU. The fuel heatup analysis was performed using the SAFER/GESTR-LOCA analysis model. The containment analysis was performed using the SHEX model. This evaluation determined the effect of CPPU on fuel cladding integrity, reactor vessel integrity, and containment integrity as a result of the fire event.

The major operator actions for a limiting postulated Appendix R fire event using Remote Shutdown System (RSS) are described below:

1) One stuck SRV opens at time zero and remains open throughout the event;

2) RCIC injection occurs at 10 minutes by operator action from remote shutdown panel;

3) Suppression pool cooling is initiated at 20 minutes by operator action;

4) At 60 minutes, when the vessel pressure reaches 80 psig, the vessel water level is raised to the main steam line elevation using one RHR pump in LPCI mode.

Alternate shutdown cooling is initiated when an SRV(s) is held open to allow water to flow to the suppression pool.

The above scenarios were developed based on the descriptions in the Hope Creek UFSAR and plant shutdown procedures.

The results of the Appendix R evaluation for CLTP and CPPU provided in Table 6-4 demonstrate that the fuel cladding integrity, reactor vessel integrity and containment integrity are maintained and that sufficient time is available for the operator to perform the necessary actions.

No changes are necessary to the equipment required for safe shutdown for the Appendix R event.

One train of systems remains available to achieve and maintain safe shutdown conditions from either the main control room or the remote shutdown panel. Therefore, CPPU has no adverse effect on the ability of the systems and personnel to mitigate the effects of an Appendix R fire event, and satisfies the requirements of Appendix R with respect to achieving and maintaining safe shutdown in the event of a fire. These results demonstrate that CPPU does not increase the potential for a radiological release as the result of a fire.

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NEDO-33076 6.8 OTHER SYSTEMS AFFECTED BY POWER UPRATE This section addresses the effect of CPPU on systems not addressed in other sections of this report. The topics addressed in this evaluation are:

I - -I

, -, Topic

-: ..;-: CLTR Disposition ; 'n Hope Creek Result IOther systems

. T'- . .. .. , . . .

1 Based on experience and previous NRC reviews, all systems that are significantly affected by CPPU are addressed in this report. Systems not addressed by this report are not significantly affected by CPPU.

((

1]

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NEDO-33076 Table 6-1 Hope Creek CPPU Plant Electrical Characteristics Parameter -,.. - - . CPPU Gross Generator Output (MWe) 1287 Rated Voltage (KV) 25 Power Factor 0.9375 Current Output: 31,710 Isolated Phase Bus Duct Rating (A):

Main Section (A) 34,000 Delta Section (A) 19,500 Main Transformers Rating (MVA) Per 466.7 Phase Table 6-2 Hope Creek Offsite Electric Power System

-Component Rating CPPU Output Generator (MVA) 1373 1373 Isolated Phase Bus Duct (kA) 34 34 Main Transformers (MVA) 1400.1 1373 Switchyard (limiting) (MVA) 2600 2600 6-15

NEDO-33076 Table 6-3 Hope Creek Spent Fuel Pool Parameters for CPPU Conditions /Parameter- Results iLirfit Configuration I -Batch Offload Both trains ofFPCCin service, and the heat load in the RP V cooled by the other RHR Time to initiate fuel transfer to SFP (hr) 59 > 24 Peak SFP Temperature (0 F) 134.9 < 135 Time to Peak SFP Temperature (hr) 115 NA Time to boil (l) (hr) 30 NA Boil off rate (gpm) 37 Configuration2 - Full Core Offload Both trainsof FPCCandNormal RHR FPCAssistmode in service, and the heat load in the RPV cooled by the other RHR.

Time to initiate fuel transfer to SFP (hr) 24 > 24 Peak SFP Temperature (0F) 139.5 < 150 Time to Peak SFP Temperature (hr) 154 NA Time to boil (l) (hr) 6.4 Boil off rate (gpm) 100 Configuration3 - Full Core Offload with Alternate RHR FPCAssist Mode Both trainsofFPCCand4AlternateRHR FPCAssist mode in service, and the heat load in the RPVcooled by.4lternateRHR FPCAssist Mode Time to initiate fuel transfer to SFP (hr) 74 > 24 Peak SFP Temperature (0 F) 149.9 < 150 Time to Peak SFP Temperature (hr) 90 NA Time to boil (l) (hr) 5 Boil off rate (gpm) 130 (1) "Time to boil" represents the time to reach 210.5 °F for Configuration I or 212 ° F for Configurations 2 and 3, after loss of all cooling at the peak temperature. Configuration I never reaches boiling point (212e F), but reaches the maximum temperature of 210.5 e F.

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NEDO-33076 Table 64 Hope Creek Appendix R Fire Event Evaluation Results PM :pC U( -'AP. 'R Criteria Cladding Heatup (PCT) (oF)(2) 589 591

• 1500 Operator Action Time to Start RCIC 10 10 (3)

(minute)

Primary System Pressure (psig)(4 ) 1112.1 1119.9

• 1375 Primary Containment Pressure (psig) 9.3 11.0 < 62 Drywell Airspace Temperature (IF) 300.3 300.2 < 340 Suppression Pool Bulk Temperature (OF) 195.2 205.9

• 310 Net Positive Suction Head(5) Yes Yes Adequate for system using designated water source (1) (( I' (2) Initial steady-state fuel temperature (3) To maintain the core covered.

(4) (( 1]

(5) NPSH demonstrated adequate, see Section 4.2.6.

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NEDO-33076 Table 6-5 Basis for Classification of No Significant Effect 6-18

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___ __________ I __________ I ________ I __________

• + + 4-t 4- 4- 1-I .1.

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4 4 A I I 4 4 6-22

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7. POWER CONVERSION SYSTEMS This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapter 10, that applies to CPPU.

7.1 TURBINE-GENERATOR The Hope Creek turbine-generator converts the thermal energy in the steam into electrical energy. The topics addressed in this evaluation are:

Topic -CLTR Disposition - Hope Creek Result Turbine-generator performance Turbine-generator missile avoidance The turbine and generator was originally designed with a maximum flow-passing capability and generator output in excess of rated conditions to ensure that the original rated steam-passing capability and generator output is achieved. This excess design capacity ensures that the turbine and generator meet rated conditions for continuous operating capability with allowances for variations in flow coefficients from expected values, manufacturing tolerances, and other variables that may adversely affect the flow-passing capability of the unit. The difference in the steam-passing capability between the design condition and the rated condition is called the flow margin.

The turbine-generator was originally designed with a flow margin of 5%. The current valves wide open (VWO) throttle steam flow is 14.81 Mlb/hr at a throttle pressure of 963 psia. The generator is rated at 1,373 MVA, which results in a rated electrical output (gross) of 1,287 MWe at a power factor of 0.9375.

The main generator stator cooling water system was upgraded by changing the pump impeller.

This allowed the generator to be upgraded from 1300 MVA to 1373.1 MVA. No other modifications are planned for the main generator or the generator hydrogen coolers.

With the generator upgraded to 1373.1 MVA the isolated phase bus rating needs to be increased from 32000 Amps to 34000 Amps for the forced cooled section of the bus. This will be accomplished by changes to the cooling unit. The self-cooled section of the bus is adequate for an increase from 18500 Amps to 19500 Amps.

The evaluation of the turbine gland seal system, taking into account the modification of the Hope Creek main turbine to accept the increased steam flow at CPPU operating conditions, demonstrated that the system is capable of adequately performing its design function without modification with the exception that four of the six gland seal system relief valves are planned to be replaced with valves with a higher relief setpoint pressure. No increase in capacity or changes in any control settings are required for the Hope Creek CPPU.

7-1

NEDO-33076 The high-pressure and low-pressure turbine rotors at Hope Creek (for both CLTP and CPPU RTP) have integral, non-shrunk on wheels. Per CLTR Section 7.1, a separate rotor missile analysis is not required for plants with integral wheels.

The overspeed calculation compares the entrapped steam energy contained within the turbine and the associated piping, after the stop valves trip, and the sensitivity of the rotor train for the capability of overspeeding. The entrapped energy increases for the CPPU conditions. The hardware modification design and implementation process establishes the overspeed trip settings to provide protection for a turbine trip.

The Hope Creek CPPU does not result in increases in system pressures, configurations, or equipment overspeed that would affect the evaluation of internally generated missiles on safety-related or nonsafety-related equipment.

7.2 CONDENSER AND STEAM JET AIR EJECTORS The Hope Creek condenser converts the steam discharged from the turbine to water to provide a source for the condensate and feedwater systems. The steam jet air ejectors (SJAE) remove noncondensable gases from the condenser to improve thermal performance. The topics addressed in this evaluation are:

Topic - CLRDisposition Hope Creek Result Condenser and SJAE ]

The condenser and SJAE functions are required for normal plant operation and are not safety related.

The main condenser can support CPPU operation. The condenser was evaluated for performance at CPPU conditions assuming cleanliness levels as low as 70%, condenser tube pluggage of 0.5%, nominal Circulating Water flow rates, and a range of Circulating Water temperatures. At CPPU conditions, the current main turbine backpressure limitation may require load reductions at the upper end of anticipated circulating water inlet temperatures.

Condenser hotwell capacities and level instrumentation are adequate for CPPU conditions. Periodic eddy current testing and water chemistry monitoring are performed which will provide monitoring of the effect of CPPU RTP operation on the condenser tubes.

The design of the condenser air removal system is not adversely affected by CPPU and no modification to the system is required specifically for CPPU. However, one of the two SJAEs cannot operate satisfactorily at condensate temperatures greater than 1300 F. The SJAE will be restored to its as-intended design configuration as part of the CPPU effort. The following aspects of the condenser air removal system were evaluated for this determination:

• Non-condensable gas flow capacity of the SJAE system;

• Capability of the SJAEs to operate satisfactorily with available dilution / motive steam flow; 7-2

NEDO-33076

• SJAEs and inter-condensers' performance at the higher expected non-condensable flow and condenser pressure conditions for CPPU, considering water vapor carryover and the maximum expected condensate temperature and flow rate; and

• Mechanical vacuum (hogging) pump capability to remove required non-condensable gases from the condenser at start-up conditions.

The physical size of the primary condenser and evacuation time are the main factors in establishing the capabilities of the vacuum pumps. These parameters do not change. Because flow rates do not change, there is no change to the holdup time in the pump discharge line routed to the reactor building vent stack. The capacity of the SJAEs is adequate because they were originally designed for operation at flows greater than those required at CPPU conditions.

7.3 TURBINE STEAM BYPASS The Turbine Steam Bypass System provides a means of accommodating excess steam generated during normal plant maneuvers and transients. The topics addressed in this evaluation are:

Topic -:CLTRDisposition Hope Creek Result Turbine steam bypass (normal operation)

Turbine steam bypass ]

(safety analysis)

At CPPU conditions, rated reactor steam flow is 16.773 Mlb/hr, resulting in a bypass capacity of 22.18% of CPPU rated steam flow. The bypass capacity at Hope Creek remains adequate for normal operational flexibility at CPPU RTP.

The bypass capacity is used as an input to the reload analysis process for the evaluation of events that credit the Turbine Steam Bypass System (see Table 9-1). [f 7-3

NEDO-33076 7.4 FEEDWATER AND CONDENSATE SYSTEMS The Feedwater and Condensate Systems provide the source of makeup water to the reactor to support normal plant operation. The topics addressed in this evaluation are:

Topic CLTR Disposition Hope Creek Result Feedwater and condensate systems J]

The FW and condensate systems do not perform a system level safety-related function, and are designed to provide a reliable supply of FW at the temperature, pressure, quality, and flow rate as required by the reactor. However, their performance has a major effect on plant availability and capability to operate at the CPPU conditions. The FW and condensate systems meet both the normal and transient operational requirements when operating at CPPU conditions with the following complement of condensate and feedwater pumps:

1. Three motor driven Primary Condensate Pumps (PCP)

2. Three motor driven Secondary Condensate Pumps (SCP)

3. Three turbine driven Reactor FW Pumps (RFP)

When operating with less than all pumps, procedural guidance will limit the maximum allowable power so that these objectives are met.

7.4.1 Normal Operation All condensate pumps and RFPs remain within their original nameplate ratings at CPPU conditions. The existing RFP turbine speed limit setpoints are not increased for CPPU. The FW heater drain lines, level control valves, and high water level dump valves were analyzed for the increased flows and conditions at CPPU. The usage factor for that part of the FW piping that constitutes a portion of the reactor coolant pressure boundary (RCPB) is addressed in Section 3.5. As the FW heaters are non-safety-related ASME Section VIII pressure vessels and their associated non-RCPB piping is non-safety-related, usage factor analysis does not apply and was not performed.

The FW heaters have been evaluated for CPPU conditions. All FW heaters are adequate for the CPPU conditions except for the 5th point FW heaters. Because this FW heater is the Moisture Separator normal drain path, its design temperature and pressure are being re-rated to envelope the higher Moisture Separator relief valve setpoints. The re-rating is being done through the original equipment manufacturer (OEM). In addition, the OEM is providing revised data sheets and nameplates for all of the FW heaters and drain coolers.

7.4.2 Transient Operation To account for FW demand transients, Hope Creek performed two independent hydraulic analyses for operation with a 3PCP/3SCP/3RFP pump line-up to ensure adequate margin above CPPU FW flow is available. One analysis assumed 105% of CPPU rated FW flow. The second analysis assumed 108% CPPU rated FW flow. Assuming anticipated pump wear, these analyses 7-4

NEDO-33076 showed that the predicted operating parameters were acceptable and within the component capabilities. The RFP speed limiter setting corresponds to approximately 112% of CPPU rated FW flow.

None of the condensate pumps or RFPs is being modified or upgraded to increase the capacity, and the CLTP RFP speed limiter and overspeed trip setpoints are maintained. Thus, the maximum postulated feedwater runout flow, measured in Ibm/hr or percent of CLTP, is unchanged.

The, historical requirement for 10% excess capacity is a nominal value; recent transient analyses have shown that a 5% flow margin is adequate to allow the feedwater and condensate control systems to successfully mitigate reactor level transients. The actual value of transient flow demanded and obtained depends on controller settings, pump inertia, initial power, level error and steam flow/feed flow mismatch gain. The current Digital FW Control System (DFCS) provides improved response to FW flow transients. If DFCS fails to provide appropriate transient response, the plant will scram on low RPV level (L3) or hi-hi level (L8). While unlikely, these are plant availability concerns and not safety concerns because these conditions are bounded by the transient analyses.

The DFCS response will be tested as part of CPPU power ascension test program. Each RFP controller will be placed in manual and then the 5% flow step changes will be inserted. Upon satisfactory control system response to the 5% flow step changes, the 10% flow step changes will be inserted. To verify the maximum FW runout capability, the pressure, flow and controller data will be measured during power ascension testing and compared against acceptance criteria.

Currently, the Reactor Recirculation System (RRS) runback is used to reduce the potential for a reactor low level scram (L3) on the loss of a RFP concurrent with RPV level less than 30-inches or a loss of a condensate pump. The RRS runback is designed to rapidly reduce power and core flow to a level within the capability of the operating condensate and feedwater pumps. The RRS runback has two speed limiters, one set at 45% speed (Intermediate runback) and the other set at 30% speed (Full runback). Currently, a RRS Intermediate runback results in approximately 70%

power and 60% core flow. A RRS Full runback results in approximately 60% power and 50%

core flow.

Upon receipt of a tripping signal of a RFP concurrent with RPV level less than 30-inch or a tripping signal of a condensate pump, RRS will initiate a fast runback resulting in lower thermal power and core flow. Concurrently, the DFCS reduces the FW flow demand to match the lower power and core flow. The RRS runback logic for PCP and SCP is only armed when FW flow is greater than 75% and 85%, respectively. The current RRS runback logic and planned changes for CPPU conditions are as follows:

7-5

NEDO-33076 Event: CLTP CPPU Loss of 1 PCP resulting in a Full Runback Intermediate 2PCP/3SCP/3RFP line-up Runback Loss of I SCP resulting in a Intermediate Same 3PCP/2SCP/3RFP line-up Runback Loss of I RFP concurrent w/ level less than Intermediate Same 30-inches resulting in a 3PCP/3SCP/2RFP Runback line-up Hope Creek has performed a steady-state analysis to verify capability of the various pump line-up configurations at CPPU conditions. The analysis results indicate that a RRS Full runback for a loss of PCP is not necessary and can be changed to a RRS Intermediate runback. Recent plant data indicates scram avoidance is successful for Intermediate runback events.

Hope Creek will perform a comprehensive Dynamic Analysis to further verify the RRS runback capability at CPPU conditions for all line-up configurations. In addition, the Dynamic Analysis will include verification that a PCP or SCP trip will not result in unacceptable NPSH conditions.

The Dynamic Analysis will use the latest best-estimate Thermal-Hydraulic model (THOR-BOP) recently installed at the Hope Creek Simulator. The changes to the RRS runback logic will be performed as part of the CPPU implementation. Proper operation of the RRS runback logic will be verified in a functional test as part of implementation design change. DFCS response will also be verified as part of CPPU power ascension testing.

7.4.3 Condensate Demineralizers The effect of CPPU on the condensate demineralizers (CDs) and pre-filters (CPFs) was reviewed.

The system supports CD full flow operation during resin replacement without requiring a plant power reduction. The system experiences slightly higher loadings resulting in slightly reduced CD run times and increased CPF backwashing frequency. However, the reduced run times and increased backwashing frequencies are acceptable (refer to Section 8 for the effect on the radwaste systems).

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NEDO-33076

8. RADWASTE AND RADIATION SOURCES This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapters 11 and 12, that applies to CPPU.

8.1 LIQUID AND SOLID WASTE MANAGEMENT The Hope Creek Liquid and Solid Radwaste System collects, monitors, processes, stores, and returns processed radioactive waste to the plant for reuse or for discharge. The topics addressed in this evaluation are:

-Topic CLTRDisposition 'Hope Creek Result Coolant fission and corrosion product levels Waste Volumes ]

The average annual volume of liquid waste release prior to dilution at Hope Creek is 1.898E+8 liters. The average annual volume of solid waste is 732 cubic meters, which includes spent resin, filter sludge, evaporator bottoms, dry compressible waste, irradiated components, and contaminated oil. These average values are based on the years 2000 through 2004. For the most recent calendar year, 2004, the estimated volume of liquid Low Level Waste (LLW) generated is 1.823E+08 liters; the estimated volume of solid LLW generated is 1316 cubic meters.

The single largest CPPU effect is the increased liquid and wet solid waste from the backwash of the Condensate Pre-Filters (CPF). CPPU results in an increased flow rate through the Condensate Pre-Filters, resulting in a reduction in the average time between backwashes and reduced CPF filter life expectancy. This reduction does not affect plant safety. The increased FW flow causes the soluble and insoluble reactor water iron concentration and conductivity to increase. Therefore, the RWCU filter-demineralizer (F/D) requires more frequent backwashes to maintain the reactor water chemistry. More frequent condensate demineralizer and RWCU demineralizer resin replacements result in solid waste increases.

The CPF and RWCU F/D backwashes are routed to the Waste Sludge Phase Separator (WSPS) and Clean-Up Phase Separator (CUPS), respectively. These separators decant the liquid from the slurry waste to allow separate processing of the liquid and solid waste. Hope Creek presently backwashes CPFs more frequently than required by CPPU conditions, in order to improve CPF performance and extend the filter life expectancy. Although the existing backwash frequency bounds the CPPU required frequency; as part of the implementation effort Hope Creek will evaluate more frequent backwashes to retain optimal CPF performance. The CUPS have been evaluated and are not limiting.

The floor drain collector subsystem and the waste collector subsystem both receive periodic inputs from a variety of sources including the liquids from the WSPS and CUPS. Neither subsystem is expected to experience a large increase in the total volume of liquid and solid waste due to operation at the CPPU condition. The design of the Hope Creek equipment and floor 8-1

NEDO-33076 drains inside and outside of containment has been evaluated to ensure any CPPU-related liquid radwaste increases can be processed. Hope Creek has sufficient capacity to handle added liquid increases required,, i.e., it can collect and process the drain fluids. The drainage systems backflow at maximum flood levels and infiltration of radioactive water into non-radioactive water drains do not change as a result of CPPU. The drainage systems design capability to withstand the effects of earthquakes and to be compatible with environmental conditions does not change as a result of CPPU. Therefore, CPPU does not affect system operation or equipment performance.

The increased loading of soluble and insoluble species increases the volume of the liquid processed wastes by 2.2% and the volume of the solid processed wastes by 14.7%. The total volume of liquid and solid processed waste does not increase appreciably (as compared to the Radwaste System capacity) because the only increase in processed waste is due to more frequent backwashes of the CPF and RWCU filter demineralizers and more frequent replacement of resin and filter elements. The total liquid and solid increases are within the Radwaste System capacity. Therefore, CPPU does not have an adverse effect on the processing of liquid and solid radwaste, and there are no significant environmental effects.

8.2 GASEOUS WASTE MANAGEMENT The topics addressed in this evaluation are:

Topic - c

-LTDisposition -Hope Creek Result Offsite release rate Recombiner performance

((:

- . ad ' ;A ! ;- , .t - l > *^: ' . ; -.-  ? -: : . ,- , - -

8-2

NEDO-33076 1]

The primary function of the Gaseous Waste Management (Offgas) System is to process and control the release of gaseous radioactive effluents to the site environs so that the total radiation exposure of persons in offsite areas is within the guideline values of 10 CFR 50, Appendix I. The Offgas System involves the management of condenser air removal system; gland seal exhaust and mechanical vacuum pump operation exhaust; and building ventilation system exhausts. Plant procedures exist to test for air infiltration (e.g., condenser) and repair as needed to maintain the Offgas System functional.

The radiological release rate is administratively controlled to remain within existing site release rate limits, and is a function of fuel cladding performance, main condenser air inleakage, charcoal adsorber inlet dew point, and charcoal adsorber temperature. ((

))

The administrative controls mentioned above to maintain the offgas radiological release rate below limits include power reduction or shutdown, reducing main condenser air in-leakage (increasing charcoal adsorber holdup time), and local power suppression (inserting control rods near a leaking fuel bundle). In addition, decreasing adsorber temperature (increasing dynamic adsorption coefficients and holdup times) can be effective in dealing with slow increases in offgas release rate.

Hope Creek has TS requirements and administrative controls to limit fission gas releases to the environment. Plant procedures or programs exist for reducing core power, suppressing power near leaking fuel, and repairing condenser air inleakage if necessary to maintain the offgas limits. These procedures are not affected by CPPU.

((:

)) Thus, the recombiner and condenser, as well as downstream system components, are designed to handle an average increase in thermal power of as much as 57% relative to OLTP, without exceeding the design basis temperatures, flow rates, or heat loads. The evaluation of the Offgas System and those connected to it for CPPU concludes that sufficient capacity exists without 8-3

NEDO-33076 modification to process expected offgas. Therefore, the gaseous radwaste system at Hope Creek is confirmed to be consistent with GE design specifications for radiolytic flow rate ((

1]

8.3 RADIATION SOURCES IN THE REACTOR CORE During power operation, the radiation sources in the core are directly related to the fission rate.

These sources include radiation from the fission process, accumulated fission products and neutron reactions as a secondary result of fission. Historically, these sources have been defined in terms of energy or activity released per unit of reactor power. Therefore, for a CPPU, the percent increase in the operating source terms is no greater than the percent increase in power.

The topics addressed in this evaluation are:

Topi Post operational radiation sources for radiological and shielding analysis

. CLTR Disposition. .C 1 ]

The post-operation radiation sources in the core are primarily the result of accumulated fission products. Two separate forms of post-operation source data are normally applied. The first of these is the core gamma-ray source, which is used in shielding calculations for the core and for individual fuel bundles. This source term is defined in terms of MeV/sec per Watt of reactor thermal power (or equivalent) at various times after shutdown. The total gamma energy source, therefore, increases in proportion to reactor power.

The second set of post-operation source data consists primarily of nuclide activity inventories for fission products in the fuel. These are needed for post-accident and spent fuel pool evaluations, which are performed in compliance with regulatory guidance that applies different release and transport assumptions to different fission products. The core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which develops "equilibrium" activities in the fuel (typically 3 years). Most radiologically significant fission products reach equilibrium within a 60-day period. ((

84

NEDO-33076

))

The results of this assessment are accounted for in the plant radiation protection program.

8.4 RADIATION SOURCES IN REACTOR COOLANT Radiation sources in the reactor coolant at Hope Creek include activation products, activated corrosion products and fission products. The topics addressed in this evaluation are:

T opic -CPPU Disposition Hope Cree IRs 8.4.1 Coolant Activation Products 8.4.2 Activated Corrosion Products and Fission Products The CLTR, Section 8.4, requires a plant specific evaluation for radiation sources in the coolant.

8.4.1 Coolant Activation Products During reactor operation, the coolant passing through the core region becomes radioactive as a result of nuclear reactions. Coolant activation products, primarily N16, are the dominant source of gamma radiation fields in the turbine building. Because these sources are produced by activation of coolant in the core region, their rates of production are proportional to power. The activation of the water is in approximate proportion to the increase in thermal power. As a result, the activation products, observed in the reactor water, increase in approximate proportion to the increase in thermal power. ((

)) Nevertheless, the radiation field resulting from activation products will increase with CPPU primarily due to the increased steam flow and the resultant decrease in transit time for the activation products from the reactor pressure vessel to the turbine complex. Because these activation products typically have extremely short half-lives, on the order of seconds, the decrease in transit time will result in a measurable increase in downstream activity. The activation products in the steam are not bounded by the original design basis concentration, but current operations with Hydrogen Water Chemistry (HWC) demonstrates sufficient margin in design to allow operations with enhanced activation products. The increase in N16 in the turbine components due to CPPU is approximately 16% for a 20% increase in steam flow. This can be compared to the increase due to HWC (factor of 4.3).

8-5

NEDO-33076 8.4.2 Activated Corrosion Products and Fission Products The reactor coolant contains activated corrosion products, which are the result of metallic materials entering the water and being activated in the reactor region. Under the CPPU conditions, the feedwater flow increases with power, the activation rate in the reactor region increases with power, and the filter efficiency of the condensate demineralizers may decrease as a result of the feedwater flow increase. The net result is an increase in the activated corrosion product production. However, the corrosion product concentrations do not exceed the design basis concentrations as a consequence of the CPPU. Therefore, no change is required in the Hope Creek design basis activated corrosion product concentrations for the CPPU.

Fission products in the reactor coolant are separable into the products in the steam and the products in the reactor water. The principle activity in the steam exclusive of activation products consists of noble gases released from the core plus carryover activity (moisture) from the reactor water. The design basis for noble gases is 0.1 curies/second after thirty minutes decay for normal operations. An evaluation of steam fission and corrosion products based upon current standards at CPPU conditions with the revised moisture content limits (see Section 3.3.3), show the plant design basis to be bounding on CPPU predicted concentrations. Therefore, the design basis activity is not exceeded and designs based upon those levels of activity in the steam are conservative for the Hope Creek CPPU. The Technical Specification limit for offgas activity does not change for the CPPU.

The fission product activity in the reactor water, like the activity in the steam, is the result of minute releases from the fuel rods. The evaluation of activity levels for fission products at CPPU conditions remain bounded by the design basis. The Technical Specification limit for reactor water concentrations does not change for the CPPU.

8.5 RADIATION LEVELS For CPPU at Hope Creek, normal operation radiation levels increase slightly. The post-CPPU radiation exposure assessment in the turbine building complex was performed based the operational data obtained by radiological surveys during the implementation of the hydrogen water chemistry (HWC) with a hydrogen injection rate of 35 scfm. Due to the conservative higher-than-expected radiation source terms and analysis techniques used for the original plant shielding design to maintain the plant exposure within the allowable radiation zone limit and ALARA (As Low as Reasonably Achievable), the increase in post-CPPU radiation levels does not affect the existing radiation zoning or shielding in the various areas of the plant. The Hope Creek topics addressed in this evaluation are:

8-6

NEDO-33076 Topic CLTR Disposition ; Hope Creek Result Normal operational radiation levels Post-operation radiation levels Post-accident radiation levels ))

The post-CPPU operational radiation levels in most of the affected plant areas including the Turbine Building complex are expected to increase by less than 20 percent. The increased normal radiation doses were evaluated and determined to have no adverse effect on safety-related plant equipment as indicated in sections 10.3.1 and 10.3.2. Individual worker exposures can be maintained within acceptable limits by controlling access to radiation areas using the site ALARA program. The Person-Rem exposure mainly consists of the radiation exposures from the refueling outage, the major contributor, and from the normal plant maintenance. The N-16 related increase in the normal plant maintenance exposure adds negligible dose to the total annual exposure. The use of Radiation Protection (RP) Procedural controls and effective ALARA program can reduce the refueling outage related exposure, which compensates the increased post-CPPU radiation exposure for the total annual exposure.

The post-CPPU radiation dose and allowable occupancy in the vital areas requiring post-accident access (NUREG-0737, Item II.B.2) for accident mitigation were evaluated using the AST and CPPU core inventory. The allowable occupancies were calculated based on maximum Total Effective Dose Equivalent (TEDE) dose rates and allowable TEDE dose. RP personnel will determine the required duration of occupancy based on the actual radiation surveys prior to accessing the vital areas if the need arises. The list of vital areas, post-CPPU TEDE dose rates, and resulting occupancies are shown in Table 8-1. The post-CPPU TSC doses from the various release paths following a LOCA are shown in Table 8-2. The post-CPPU Operational Support Center (OSC) and Security Center doses from the various release paths following a LOCA are shown in Tables 8-3 and 8-4 respectively. The Technical Support Center (TSC) is habitable for the post-accident emergency preparedness activities. The OSC is monitored, and procedures provide for relocation of the OSC functions. In the event the Security Center becomes uninhabitable an alternate location is available.

The post-CPPU drywell 40-year normal integrated doses are expected to increase because of increased core fission neutron and gamma doses, and associated increased reactor coolant and main steam doses. The existing normal core neutron and gamma integrated doses in the drywell remain bounding due to conservatism in the ANISN shielding model. The 40-year normal integrated doses in the turbine building and in some auxiliary building rooms are expected to increase due to the CPPU. The 40-year normal integrated doses in the RWCU equipment rooms remained bounding due to the protracted N-16 transit times for the various RWCU components, which allow substantial radioactive decay to reduce the post-CPPU N-16 related radiation exposure increase to a negligible level. The existing 40-year normal integrated doses in the 8-7

NEDO-33076 reactor building and balance of the plant remain bounding for CPPU. The increased post-accident radiation doses have no adverse effect on safety-related plant equipment as indicated in sections 10.3.1 and 10.3.2.

Section 9.2 addresses the accident doses for the Main Control Room.

8.6 NORMAL OPERATION OFF-SITE DOSES The primary sources of normal operation offsite doses at Hope Creek are (1) airborne releases from the Offgas System and, (2) gamma shine from the plant turbines, and (3) liquid effluent releases from the radwaste system. The topics addressed are:

Topic CLTR Disposition - Hope Creek Result Plant gaseous emissions Plant skyshine from the turbine The Hope Creek Radioactive Effluent Controls Program provides for the control of radioactive effluents and for maintaining the doses to member(s) of the public from radioactive sources as low as reasonably achievable. The post-CPPU liquid effluent releases are expected to increase by 2.2 percent. The maximum average annual offsite dose from liquid effluent release is less than 0.25% of the allowable limit. The gaseous effluent releases are not expected to change due to the CPPU. The measured N-16 skyshine dose during implementation of the hydrogen water chemistry (HWC) at the offsite location was negligibly small (less than 0.01 mr/hr). The post-CPPU N-16 related increase adds negligibly to the offsite skyshine dose. Therefore, the offsite doses from noble gases, airborne particulates, iodine, tritium, and liquid effluents are insignificantly affected and considered to be bounded by the current offsite dose analysis. The existing offsite doses due to the effluent releases are a small fraction of the regulatory limits of 10 CFR 50, Appendix I, and expected to remain bounding for the resulting post-CPPU doses.

The post-CPPU offsite doses to a member of public are estimated to be a fraction of the regulatory limits set forth in 10 CFR 20.1301 and 1302 and 40 CFR 190, Subpart B.

8-8

NEDO-33076 Table 8-1 Post-LOCA Vital Access Area Dose Rates and Occupancies Vital Access Post-LOCA -Allowable Area .Dose Rate (rem/hr) Occupancy

-Locations `Whole Bodyi TEDE (hr)

Operational Support Center 8.30E-03 1.50E+00 Guardhouse 5.90E-03 3.1OE-01

• PASS Sample Station 4.10E-02 3.70E+00 1.3 PASS Analysis Room 4.10E-02 3.70E+00 1.3 Diesel Generator and 3.60E-02 1.30E+00 3.8 Accessories FRVS RMS Skid 5.80E-02 5.10E+00 0.9 Remote Shutdown Panel Area 5.80E-02 5.1 OE+00 0.9 HP/Access Control Point 5.80E-02 5.10E+00 0.9 Cafeteria (Room 109) 1.50E-02 1.40E+00 3.5 Training Rooms (103 & 104) 1.50E-02 1.40E+00 3.5 Maintenance Shop 1.50E-02 1.40E+00 3.5

• The dose rates are provided for information only.

Note: The maximum dose rates between 0.50 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are listed in the above table. The allowable occupancy is calculated based on the allowable TEDE dose limit of 5 rem and the TEDE dose rate (i.e., for the Pass Sample Station: (5 rem TEDE)/(3.7 rem TEDE/hr) = 1.3 hr).

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NEDO-33076 Table 8-2 Post-LOCA TSC Dose Post-LOCA .

Activity Release TSC Dose (rem)

Path TEDE Containment Leakage 7.76E-01 ESF Leakage 1.48E-01 MSIV Leakage 2.63E+00 Containment Shine 0.OOE+00 External Cloud 4.08E-02 FRVS Filter Shine 3.01E-01 Total 3.90E+00 Allowable Dose Limit 5.OOE+00 Note: Results are from the existing design analysis, which is based on an assumed core thermal power level of 1.02 times CPPU RTP..

8-10

NEDO-33076 Table 8-3 Post-LOCA OSC Dose Post-LOCA Activity Release , ^OSC Dose (rem)

-Path *TEDE-Containment Leakage 1.02E+O1 ESF Leakage 2.15E+01 MSIV Leakage 5.73E+01 Total 8.90E+01 Allowable Dose Limit 5.OOE+00 OSC dose rate calculation:

OSC Total Dose = 8.90E+01 rem TEDE assuming 100 percent occupancy from 0 to 30 days OSC Dose Rate = (8.90E+01 rem TEDE)(1000 mrem/rem)/(720 hr) = 123.6 mrem/hr TEDE > 15 mrem/hr guideline value Notes:

Results are from the existing design analysis, which is based on an assumed core thermal power level of 1.02 times CPPU RTP.

Dose rates may exceed the criteria for areas requiring continuous occupancy during the most limiting design basis event. However, plant procedures provide for continuous monitoring of the OSC and relocation to the TSC or other location if loss of habitability occurs.

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NEDO-33076 Table 84 Post-LOCA Security Center (Guard House) Dose Post-LOCA Activity Release., GiH Dose (rem)

'Path' TEDE Containment Leakage 1.09E+00 ESF Leakage 2.27E+00 MSIV Leakage 1.22E+01 Total 1.56E+01 Allowable Dose Limit 5.OOE+00 Guard House (GH) dose rate calculation:

GH Total Dose = 1.56E+01 rem TEDE assuming 100 percent occupancy from 0 to 30 days.

GH Dose Rate = (1.56E+01 rem TEDE)(1000 mremlrem)/(720 hr)= 21.67 mrem/hr TEDE > 15 mrem/hr guideline Notes:

Results are from the existing design analysis, which is based on an assumed core thermal power level of 1.02 times CPPU RTP.

Dose rates may exceed the criteria for areas requiring continuous occupancy during the most limiting design basis event. However, an alternate location is available if loss of habitability occurs.

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NEDO-33076

9. REACTOR SAFETY PERFORMANCE EVALUATIONS This section primarily focuses on the information requested in Regulatory Guide 1.70, Chapter 15, which applies to CPPU.

9.1 ANTICIPATED OPERATIONAL OCCURRENCES 9.1.1 Transient Events The UFSAR evaluates the effects of a wide range of potential plant Anticipated Operational Occurrences (commonly referred to as transients). Disturbances to the plant caused by a malfunction, a single equipment failure or an operator error are investigated according to the type of initiating event per Regulatory Guide 1.70, Chapter 15. Appendix E of ELTRI (Reference 2) identifies the limiting events to be considered in each category of events. The generic guidelines also identify the analytical methods, the operating conditions that are to be assumed, and the criteria that are to be applied.

The following paragraphs address each of the limiting events and provide a summary of the resulting transient safety analysis. The results given here are for a GE14 equilibrium core, and show the overall capability of the design to meet all transient safety criteria for CPPU operation.

Table E-1 of ELTR1 (Reference 2) provides the specific events to be analyzed for the CPPU, the power level to be assumed, and the computer models to be used. The transients that are not listed in Table E-1 are generally milder versions of the analyzed events. The CPPU analysis uses the GEMINI transient analysis methods listed in Appendix E of ELTRI.

The reactor operating conditions that apply most directly to the transient analysis are summarized in Table 9-1. They are compared to the conditions used for the UFSAR and the most recent GE reload fuel cycle (Reload 12) analyses. Most of the transient events are analyzed at the full power and maximum allowed core flow operating point on the power/flow map, shown in Figure 2-1. Direct or statistical allowance for 2% power uncertainty is included in the analysis. ((

)) The Safety Limit MCPR (SLMCPR) in Table 9-1 was used to calculate the MCPR Operating Limits required for the analyzed events. For all pertinent events, one SRV is assumed to be out-of-service. The MSIV closure (with flux scram) overpressurization analysis is provided in Section 3.1. A discussion of other equipment out-of-service options is provided in Section 1.3.2.

The limiting events for each limiting transient category from Table E-I of ELTR1 were analyzed. Their inputs and results revise the licensing basis for the transient analysis to the CPPU RTP. The limiting transient analysis results for the full CPPU RTP condition are provided in Table 9-2, and Figures 9-1 through 9-4. As shown in the table and figures, no change to the basic characteristics of any of the limiting events is caused by the CPPU.

9-1

NEDO-33076 The severity of transients at less than rated power are not significantly affected by the CPPU, because of the protection provided by the power and flow dependent limits.

The historical 25% of RTP value for the Technical Specification (TS) Safety Limit, some thermal limits monitoring Limiting Conditions for Operation (LCOs) thresholds, and some Surveillance Requirements (SRs) thresholds is based on generic analyses (evaluated up to -50%

of original RTP) applicable to the plant design with highest average bundle power (the BWR6) for all of the BWR product lines. As originally licensed, the highest average bundle power (at 100% RTP) for any BWR6 is 4.8 MWt/bundle. The 25% RTP value is a conservative basis, as described in the plant Technical Specifications, ((

1))

The Loss of Feedwater Flow (LOFW) transient was analyzed for CPPU. During a LOFW transient and assuming an additional single failure of HPCI, reactor water level is automatically maintained above the top of the active fuel (TAF) by the RCIC system, without any operator action. Because of the increased decay heat from the CPPU, slightly more time is required for the automatic systems to restore water level. Operator action is only needed for long-term plant shutdown. After water level is restored, the operator manually controls water level, reduces reactor pressure, and initiates RHR shutdown cooling. These sequences of events do not require any new operator actions or shorter operator response times. Therefore, the operator actions for a LOFW transient do not significantly change for the CPPU.

9.1.2 Alternate Shutdown Cooling Evaluation The Hope Creek UFSAR Section 15.2.9.3 provides a qualitative evaluation of the Alternate Shutdown Cooling (ASDC) mode of decay heat removal using only safety grade equipment.

CPPU conditions have no effect on this qualitative evaluation because none of the equipment is modified for or affected by CPPU operation.

9-2

NEDO-33076 9.2 DESIGN BASIS ACCIDENTS This section addresses the radiological consequences of Design Basis Accidents (DBAs) for Hope Creek. The topics addressed in this evaluation are:

T'opic -CLTRDisposition Hope'Creek Result Main Steam Line Break Outside Containment ((

Instrument Line Break LOCA Inside Containment Fuel Handling Accident Control Rod Drop Accident

• The CLTR allows for the use of a ((

)) evaluation for all of the DBA radiological consequence topics.

The DBA analyses described in this section are based on a core inventory for a core thermal power level of 4,031 MWt. This bounds the CPPU of 3840 MWt.

Main Steam Line Break Accident (MSLBA) Outside Containment The coolant and main steam source terms are affected by the CPPU. Therefore, the MSLBA is analyzed using the uprated coolant and MS source terms, guidance in Appendix D of Regulatory Guide 1.183, and the TEDE dose criteria in Table 6 of RG 1.183. Because no fuel damage occurs during a MSLBA at Hope Creek, the released activity is the maximum coolant activity allowed by the technical specifications. The iodine concentrations in the primary coolant are assumed to correspond to the maximum value of 4.0 [pCi/gm Dose Equivalent (DE) I-131 ( pre-accident iodine spike) and a value of 0.2 pCi/gm DE 1-131 equilibrium iodine activity for continued full power operation. The CREF system is not credited in the analysis. The post-MSLBA EAB, LPZ, and CR doses are summarized in Table 9-3, and shows that all doses are within their applicable regulatory limits.

Instrument Line Pipe Break Accident (ILPBA)

The ILPBA is analyzed assuming the iodine concentration in the primary coolant at 4 pCi/g Dose Equivalent (DE) 1-131. Of the 25,000 pounds of coolant released from the instrument line break, 6,000 pounds flashes to steam. All of the iodine in the coolant, which flashes to steam is assumed to enter the steam phase with the coolant, and that 10 % of the iodine remaining in solution in the coolant becomes airborne. The activity released from the break is assumed to mix with 50% of the reactor building volume prior to being released to the environment via the Reactor Building Ventilation System (RBVS) through the South Plant Vent (SPV). The post-ILPBA activity is assumed released instantaneously as a single puff and the CREF charcoal filtration systems are not credited in the analysis. The post-ILPBA EAB, LPZ, and CR doses are summarized in Table 9-4, which shows that all doses are within their applicable regulatory limits.

9-3

NEDO-33076 Loss of Coolant Accident The post-LOCA EAB, Low Population Zone (LPZ), and CR doses are analyzed using the CPPU core inventory, AST, guidance in Appendix A of Regulatory Guide 1.183, and an increased CR unfiltered inleakage of 350 cfm. Adherence to the guidance in RG 1.183, and the use of the specific values/limits contained in the TS with as-tested post-accident performance of the safety grade engineered safety functions (ESF), provide the assurance for sufficient safety margin, including a margin to account for analysis uncertainties. The concrete shielding associated with the CR and reactor building provides adequate protection to reduce the external cloud and containment shine doses to a CR operator to a negligible amount. The post-LOCA CR filter shine dose is calculated and added to the doses from other sources. The post-LOCA EAB, LPZ, and CR doses are summarized in Table 9-5, and shows that all doses are within their applicable regulatory limits.

Fuel Handling Accident (FHA)

The post-FHA EAB, LPZ, and CR doses are analyzed using the CPPU core inventory, guidance in Appendix B of Regulatory Guide 1.183, iodine and noble gas activity released from 124 damaged fuel rods with a radial peaking factor of 1.75, and the TEDE dose criteria in Table 6 of RG 1.183. Because the containment is open during fuel handling operations (i.e., containment hatch C-9 and RB truck bay door are open), the radioactive material escaping from the reactor cavity pool to the containment would be released to the environment over a 2-hour time period as a ground level release through the reactor building truck bay door. The CREF system is not credited in the analysis. The post-FHA EAB, LPZ, and CR doses are summarized in Table 9-6, and shows that all doses are within their applicable regulatory limits.

Control Rod Drop Accident (CRDA)

The post-CRDA EAB, LPZ, and CR doses are analyzed using the CPPU core inventory, guidance in Appendix C of Regulatory Guide 1.183, iodine, noble gas, and alkali metals released from 850 damaged fuel rods and 0.77% melted fuel, a radial peaking factor of 1.75, and the TEDE dose criteria in Table 6 of RG 1.183. The activity released from the gap and fuel pellets is assumed instantaneously mixed in the reactor coolant within the pressure vessel and transported to the condenser where it is released to the atmosphere as a ground-level release at a rate of 1%

per day for one day. The CREF system is not credited in the analysis. The post-CRDA EAB, LPZ, and CR doses are summarized in Table 9-7, and shows that all doses are within their applicable regulatory limits.

9-4

NEDO-33076 9.3 SPECIAL EVENTS This section considers two special events: ATWS and SBO. The topics addressed in this evaluation are:

- Topic .CLTR Disposition H ope Cree Result 9.3.1 Anticipated Transient Without Scram ((

9.3.2 Station Blackout 9.3.3 ATWS with Core Instability 9.3.1 Anticipated Transient Without Scram Hope Creek meets the Anticipated Transients Without Scram (ATWS) mitigation requirements defined in 10 CFR 50.62:

1. Installation of an Alternate Rod Insertion (ARI) system.

2. Boron injection equivalent to 86 gpm.

3. Installation of automatic Recirculation Pump Trip (RPT) logic (i.e., ATWS-RPT).

In addition, a plant-specific ATWS analysis was performed to ensure that the following ATWS acceptance criteria are met:

1. Peak vessel bottom pressure less than ASME Service Level C limit of 1500 psig.

2. Peak cladding temperature within the 10 CFR 50.46 limit of 2200'F.

3. Peak cladding oxidation within the requirements of 10 CFR 50.46.

4. Peak suppression pool temperature shall not exceed 201F.

5. Peak containment pressure shall not exceed 62 psig.

The key inputs to the ATWS analysis are provided in Table 9-8.

The ATWS analysis was performed as discussed in Section L.3 of ELTRI, using the ODYN code (Reference 26). The analyzed events have been shown to be the limiting events for ATWS calculations.

The ATWS analysis was performed for current rated power and for the CPPU to demonstrate the effect of the CPPU on the ATWS acceptance criteria. The limiting results for each of the ATWS evaluation acceptance criterion are provided in Table 9-9.

9-5

NEDO-33076 These results of the ATWS analysis meet the above ATWS acceptance criteria. Therefore, the plant response to an ATWS event at the CPPU conditions is acceptable.

9.3.2 Station Blackout Station blackout (SBO) was re-evaluated using the NRC approved SHEX code (Reference 16) and the guidelines of NUMARC 87-00. The existing SBO evaluation was performed using the MAAP Code. For CPPU, a single bounding event was analyzed that assumes only the RCIC system is available to control RPV water level. HPCI was assumed to be unavailable. The results of this analysis bound a "HPCI system only" scenario.

The plant responses to and coping capabilities for an SBO event are affected slightly by operation at CPPU RTP, due to the increase in the initial power level and decay heat. Decay heat was conservatively evaluated assuming end-of-cycle and GE-14 fuel. There are no changes to the systems and equipment used to respond to an SBO, nor is the required coping time (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) changed.

Areas containing equipment necessary to cope with a station blackout event were evaluated for the effect of loss-of-ventilation due to an SBO. The evaluation shows that equipment operability is bounded due to conservatism in the existing design and qualification bases. The battery capacity remains adequate to support HPCI/RCIC operation after CPPU. Adequate compressed gas capacity exists to support MSRV actuations.

The current condensate storage tank inventory reserve (135,000 gal.) for HPCI/RCIC use ensures that adequate water volume is available to remove decay heat, depressurize the reactor, and maintain reactor vessel level between Level 2 and Level 8 (approximately 109,000 gal.

required). Peak containment pressures and temperatures remain within design bases. Consistent with the DBA-LOCA condition, the required NPSH margin for the RHR pumps has been evaluated (see Section 4.2.6) and a component acceptability review has been completed (see Section 3.9).

Based on the above evaluations, Hope Creek continues to meet the requirements of 10 CFR 50.63 after the CPPU.

9-6

NEDO-33076 9.3.3 ATWS with Core Instability

((

• ].

The ATWS with core instability event occurs at natural circulation following a recirculation pump trip. Therefore, it is initiated at approximately the same power level as a result of CPPU operation because the MELLLA upper boundary is not increased. The core design necessary to achieve CPPU operations may affect the susceptibility to coupled thermal-hydraulic/neutronic core oscillations at the natural circulation condition, but would not significantly affect the event progression.

Several factors affect the response of an ATWS instability event, including operating power and flow conditions and core design. The limiting ATWS core instability evaluation presented in References 29 and 30 was performed for an assumed plant initially operating at OLTP and the MELLLA minimum flow point. ((

)) CPPU allows plants to increase their operating thermal power but does not allow an increase in control rod line. ((

)) The conclusion of Reference 28 and the associated NRC SER that the analyzed operator actions effectively mitigate an ATWS instability event are applicable to the operating conditions expected for CPPU at Hope Creek.

Initial operating conditions of Feedwater Heater Out of Service (FWHOOS) and Final Feedwater Temperature Reduction (FFWTR) do not significantly affect the ATWS instability response reported in References 29 and 30. The limiting ATWS evaluation assumes that all feedwater heating is lost during the event and the injected feedwater temperature approaches the lowest achievable main condenser hot well temperature. The minimum condenser hot well temperature is not affected by FWHOOS or FFWTR. Thus, as compared to the event initiated from a normal feedwater temperature condition, the event initiated from either the FWHOOS or FFWTR condition would have less moderator reactivity insertion based on a smaller temperature difference between the initial and final feedwater temperatures. Therefore, the power oscillation for FWHOOS or FFWTR is expected to be no worse than for the normal temperature condition.

9-7

NEDO-33076 9-8

NEDO-33076 Table 9-1 Hope Creek Parameters Used for Transient Analysis Parameter Base UFSAR Cycle' Analysis - CPPU Rated Thermal Power (MWt) 3293 3339 3840 Analysis Power (% Rated) 100 100 / 102 100 / 102 2 Analysis Dome Pressure (psig) 1005 1005 1005 Analysis Turbine Pressure (psig) 965 961 3 9463 Rated Vessel Steam Flow (Mlb/hr) 14.159 14.404 16.773 Analysis Steam Flow (% Rated) 100 100 100 Rated Core Flow (Mlb/hr) 100 100 100 Rated Power Core Flow Range (% Rated) 100 76.6-105 94.8 - 105 Analysis Core Flow 4 100 105 105 Normal Feedwater Temperature (OF) 420.0 422.6 431.6 Steam Bypass Capacity (% Rated Steam flow) 25.0 25.0 21 No. of SRVs assumed in the analysis 13 13 5 13 5 No. of Safety Valves assumed in the analysis 13 13 13 No. of Relief Valves assumed in the analysis 13 13 13 MCPR Safety Limit 1.07 1.06 1.10 I Reload 12 (Cycle 13) results provided for comparison.

2 There are some analyses as indicated in Table 9-2 that were performed based on 3952 MWt.

3 Reload and CPPU analysis based on measured steam line pressure drop.

4 All analysis at maximum core flow unless explicitly noted otherwise.

5 One of the lowest pressure setpoint SRVs is assumed to be out of service for transient analysis.

9-9

NEDO-33076 Table 9-2 Hope Creek Transient Analysis Results

.OLMCPR:

Enent Peak Peak ACPR Op ption B Neutron Heat Flux:Flux

( of -Rated '(% of Rated:

CPPU)CPPU) _ _ _ _ _ _

Load Rejection with Bypass Failure 360 118 0.27 1.59 1.42 Turbine Trip With Bypass Failure 381 119 0.27 1.59 1.42 FW Controller Failure Max Demand 313 119 0.23 1.56 1.39 Loss of Feedwater Heating (1) (1) 0.17 1.27 Rod Withdrawal Error (1) (1) 0.17 1.27 Slow Recirculation Increase (2) (2) (2) MCPRf Fast Recirculation Increase 5 74 46 0.22 (3) (3)

Load Rejection with Bypass 243 110 0.19 1.51 1.34 MSIV Closure - All Valves 233 104 0.18 (4) (4)

MSIV Closure - One Valve 120 109 0.11 (4) (4)

Loss of Feedwvater Flow5 102 102 (2) (2) (2)

Loss of One Feedwater Pump 5 100 100 (2) (2) (2)

1. Peak neutron flux and peak heat flux are not reported for the slow transients.

2. Not applicable.

3. Fast recirculation increase is bounded by offrated limits.

4. Bounded by the Load Rejection with Bypass Failure.

5. The percent of power is based on 3952 MWt. The analysis was not performed based on 3840 MWt because the analysis based on 3952 MWt is more limiting.

9-10

NEDO-33076 Table 9-3 Hope Creek MSLBA Radiological Consequences MSLBA

..-:.:..,, ' . 'Pre-accident Iodine Spike,'-. .. ,..

TEDE Dose (rem)

' Receptor Location

' CR [ EAB 'LPZ Calculated Dose 3.60E+00 9.42E-01 9.45E-02 Allowable TEDE Limit 5.OE+00 2.5E+01 2.5E+01

'MSLBA

.Maximumn<Equilibrium lodine Concentration for.

-.Continued Full Power.Operation

'TEDE Dose (rem)

'Receptor.Location' CR EAB ' ' ' ;Z Calculated Dose 1.81E-01 5.61E-02 5.63E-03 Allowable TEDE Limit 5.0E+00 2.5E+00 2.5E+00 9-11

NEDO-33076 Table 94 Hope Creek ILPBA Radiological Consequences

..ILPBA TEDE Dose (rem);,

.Receptor Location"

-:;

-CR - EAB PZ Calculated Dose 2.22E-01 5.07E-02 5.07E-03 Allowable TEDE Limit 5.0E+00 2.5E+00 2.5E+00 9-12

NEDO-33076 Table 9-5 Hope Creek LOCA Radiological Consequences Post-LOCA-'" LOCAsm' Activity Release TEDE Dose (Rem)n Path ... .....-

I ..... on . ....... .. ,. . ,

' ,, ', CR - EAB ] .P 1 Containment Leakage 1.05E+00 3.73E-01 1.62E-01 ESF Leakage 1.25E+00 1.91E-01 9.79E-02 MSIV Leakage 2.13E+00 2.63E+00 4.56E-01 Containment Purge 0.00E+00 0.OOE+00 0.OOE+00 Containment Shine 0.OOE+00 0.OOE+00 0.OOE+00 External Cloud 0.OOE+00 0.00E+00 0.OOE+00 CR Filter Shine 2.46E-03 0.OOE+00 0.00E+00 Total Calculated Dose 4.43E+00 3.199E+00 7.16E-01 Allowable TEDE Limit 5.OE+00 2.5E+01 2.5E+01 9-13

NEDO-33076 Table 9-6 Hope Creek FHA Radiological Consequences In 'Reactor Building

'TEDE'Dose (remn)

.. .'Receptor Location

, -,X'CR ., , l: EAB :'-: -: : LPZ I I.I Calculated Dose 3.31 E+00 5.27E-01 5.27E-02 Allowable TEDE Limit 5.OE+00 6.3E+00 6.3E+00 Table 9-7 Hope Creek CRDA Radiological Consequences

CRDA-TEDE Dose (rem)

Recepior Location,

R EAB L-PZ Calculated Dose 1.37E-01 2.92E-02 6.23E-03 Allowable TEDE Limit 5.0E+00 6.3E+00 6.3E+00 9-14

NEDO-33076 Table 9-8 Hope Creek Key Inputs for ATWS Analysis Input Variable CLTP CPPU Reactor power (MWt) 3339 3952 Reactor dome pressure (psia) 1005 1005 SRV capacity (Mlbm/hr) 12.38 12.38 High pressure ATWS-RPT (psig) 1101 1101 Number of SRVs Out-of-service (OOS) 1 1 Table 9-9 Hope Creek Results of ATWS Analysis*

Acceptance Criteria CLTP** CPPU Peak vessel bottom pressure (psig) 1343 1437 Peak suppression pool temperature (0 F) <199 199 Peak containment pressure (psig) 8.0 9.1 Peak cladding temperature (0 F) 1589 1446 Local cladding oxidation <17 <17

• Cladding temperature and oxidation remain below their 10 CFR 50.46 limits.

• • To examine the effect of CPPU, a baseline is established at the CLTP level, assuming the current licensed equipment performance assumptions and plant parameters.

9-15

NEDO-33076 V700

-a-Vessel Press Rise(psi) 3050 -Safety Valve Flaw

-RafeValveFlo

- BypassValve Flow 250 I 'A 175.0

• 12 nD ISA J ,, J

.2o .

0D 10 2 T3 0 e 40 10 aa 72 Time(s ¢)

I 20 AD 40 so 40 70 ..a Ia A 'a Time(s"c) Time(sac)

Figure 9-1 Turbine Trip with Bypass Failure

(@ 100% CPPU RTP and 105% Core Flow) 9-16

NEDO-33076 at 0e0 1. 20 30 so to 00 7.0 3.0 40 G0 so 70 Time ("c) Tim. ("c)

Figure 9-2 Generator Load Rejection with Bypass Failure

(@ 100% CPPU RTP and 105% Core Flow) 9-17

NEDO-33076

-B-Vessel PressRise(psi)

- Safety Vahe Fiw

=a - Reletf Vbei Fbw

- Bypass Valve Flow Os "So

'I*S 1n00 r125 0 750 O a no 20 40 so to 10O 120 MO 1O.0 Isa 00 2.0 40 a0 *80 100 15 10 80 Io Time (see) Tire (see)

Ir1 20 o0 20 40a 0o I0 00 110 U10 0o hSO 00 10 4.60 0 so 18o 110 40 "Jo0 10 Time (s.ec) Time (see)

Figure 9-3 Feedwater Controller Failure - Maximum Demand

(@ 100% CPPU RTP, 105% Core Flow and 431.6°F Feedwater Temp.)

9-18

NEDO-33076 a'

00 20 40 s0 a0 10 10 140 IS0 10 *0 20 40 *0 *0 100 120 140 160 180 rim. (80c) Time (oec)

I II' Time("c) Tkne ("ec)

Figure 94 Feedwater Controller Failure - Maximum Demand with RPT Out of Service

(@ 100% CPPU RTP, 105% Core Flow & 431.6 0F Feedwater Temp.)

9-19

NEDO-33076

10. OTHER EVALUATIONS 10.1 HIGH ENERGY LINE BREAK High-energy line breaks (HELBs) are evaluated for their effects on equipment qualification. The topics addressed in this evaluation are:

Topic -CLTR Disposition  ; Hope Creek Result 10.1.1 Steam lines j[

10.1.2 Liquid lines ))

)) The result of the Hope Creek evaluation of HELBs is provided in Table 10-1.

10.1.1 Steam Line Breaks Main Steam Line Breaks CPPU has no effect on main steam line breaks because steam conditions at the postulated break locations are unchanged. CPPU has no effect on the steam pressure or enthalpy at the postulated break locations. Therefore, CPPU has no effect on the mass and energy releases from a HELB in a main steam line.

HPCI Steam Line Breaks CPPU has no effect on HPCI steam line breaks because steam conditions at the postulated break locations are unchanged. CPPU has no effect on the steam pressure or enthalpy at the postulated break location. Therefore, CPPU has no effect on the mass and energy releases from a HPCI line break.

10-1

NEDO-33076 RCIC Steam Line Breaks CPPU has no effect on RCIC line breaks because steam conditions at the postulated break locations are unchanged. CPPU has no effect on the steam pressure or enthalpy at the postulated break location. Therefore, CPPU has no effect on the mass and energy releases from a RCIC line break.

10.1.2 Liquid Line Breaks Operation at CPPU conditions requires an increase in the steam and feedwater flows, which results in a slight increase in downcomer subcooling. This increase in subcooling may lead to increased break flow rates for liquid line breaks. Only the mass and energy releases for HELBs in the RWCU and FW systems may be affected by CPPU and were re-evaluated at CPPU conditions.

RWCU Line Breaks An evaluation of the mass and energy releases for RWCU line breaks at CLTP and CPPU conditions indicated that the CPPU mass releases for RWCU line breaks increases by a maximum of 35% from CLTP (100% RTP, 100% rated core flow) to the MELLLA minimum recirculation pump speed region with reduced feedwater temperature condition. The enthalpy of the fluid released decreases by less than 1% from CLTP to rated CPPU conditions due to increased subcooling in the reactor recirculation fluid. Based on these results, the effects of increased mass/energy release on reactor building pressure, temperature and relative humidity profiles at CPPU conditions were evaluated.

Reactor Building (RB) subcompartment pressures and temperatures post RWCU line break were determined at each break location at CPPU conditions. The resulting pressures and temperatures were found to be within the current licensing values.

Feedwater System Line Break The CLTP mass and energy releases for FW line breaks are affected by changes in the FW system including increased FW flow rates. The mass and energy releases for the FW line breaks were re-analyzed at CPPU conditions. Energy release from the FW line break at CPPU conditions is bounded by the energy release from a MS line break at current licensed conditions.

Pipe W'hip and Jet Impingement Pipe whip and jet impingement loads resulting from high energy pipe breaks are directly proportional to system pressure. Because CPPU conditions either do not result in an increase of pressure in high-energy piping or the increase in pressure is bounded by the original analysis basis pressure, there is no change on existing pipe whip or jet impingement loads on HELB targets or pipe whip restraints. Additionally, a review of pipe stress calculations determined that 10-2

NEDO-33076 the feedwater temperature increases associated with CPPU conditions will not result in pipe stress levels above the thresholds required for postulating HELBs, except at locations already evaluated for breaks. As a result, CPPU conditions do not result in new HELB locations, nor affect existing HELB evaluations of pipe whip restraints and jet targets.

The review of the postulated pipe break criteria determined that for the FW piping at three locations, the cumulative fatigue usage exceeds the postulated pipe break criteria limit. The existing calculations for these locations will be reviewed to reconcile the cumulative fatigue usage prior to implementation of the CPPU.

Internal Flooding The fluid volumes in tanks and vessels with potential for flooding do not change for CPPU.

Internal flooding for postulated RWCU line breaks increases due to the increased mass release at CPPU conditions. The evaluation shows that for the rooms affected by postulated RWCU line breaks; the CPPU mass release will result in a maximum increase in flooding levels of 36%.

Flooding due to a high energy line break in rooms previously evaluated for flooding, will not preclude a safe shutdown of the plant because there is no essential safety equipment in the rooms, equipment has already been evaluated for wetting/flooding, and/or the equipment is located above the flood level. The increased flooding levels will not affect plant safety.

10.2 MODERATE ENERGY LINE BREAK Moderate energy line breaks (MELBs) are evaluated for their effects on equipment qualification.

The topics addressed in this evaluation are:

"Topic  : CLTRDisposition Hope Creek&Result Flooding ((

Environmental Qualification ]

((

System design limits (design pressure) used as input to the Moderate Energy Line Break (MELB) flooding analyses are not changed by CPPU. Therefore, the Hope Creek MELB internal flooding evaluations are not affected by the CPPU and the design change process ensures continued evaluation of all changes for effect on MELB flooding.

10-3

NEDO-33076 10.3 ENVIRONMENTAL QUALIFICATION Safety related components are required to be qualified for the environment in which they are required to operate. The topics addressed in this evaluation are:

vTopic -CLTRDisposition Hope Creekl Resultl 10.3.1 Electrical Equipment 10.3.2 Mechanical Equipment With Non-Metallic Components 10.3.3 Mechanical Component Design Qualification 10.3.1 Electrical Equipment The safety-related electrical equipment was reviewed consistent with the requirements of 10 CFR 50.49 to assure the existing qualification for the normal and accident conditions expected in the area where the devices are located remain adequate. The 10 CFR 50.49 acceptance criteria including pressure, temperature, and radiation were used in making this determination. Table 10-2 provides a listing of the EQ effects and parameter changes associated with CPPU.

Inside Containment Environmental qualification (EQ) for safety-related electrical equipment located inside the containment is based on main steam line break and/or DBA-LOCA conditions and their resultant temperature, pressure, humidity and radiation consequences, and includes the environments expected to exist during normal plant operation. Normal temperatures increase slightly as a result of CPPU conditions, but remain bounded by the normal temperatures used in the EQ analyses. The current accident conditions for temperature and pressure are modified for CPPU conditions as provided in Table 4-1. The post-accident peak temperature and pressure for CPPU conditions increase slightly but remain bounded by the peak temperature and pressure conditions used in the EQ analyses.

Radiation EQ for safety-related electrical equipment located inside the containment is based on the radiation environment expected to exist during normal plant operation, post-LOCA conditions, and the resultant cumulative radiation dose consequences. The analyzed maximum core CPPU gamma and neutron radiation levels under 40-year normal plant operating conditions remain bounding due to conservatism in the existing plant ANISN shielding model in the drywvell areas that are populated with the most safety related electrical equipment (i.e., primary containment zones 1 through 5). The 100-days post-LOCA gamma radiation levels increase by less than 20% in these areas. The total integrated doses (normal plus accident) for CPPU conditions do not adversely affect qualification of the equipment located inside containment due to the compensating margin in the qualified dose. Using the worst-case dose, the increased radiation doses result in a reduction of the radiation life of Target Rock Solenoid Valves located inside containment. However, a case-specific analysis was performed, which determined that the 10-4

NEDO-33076 radiation life of Target Rock Solenoid Valves could be extended up to design life of the plant.

Therefore, the qualified life of these solenoids was extended to the remaining plant life.

Outside Containment Accident temperature, pressure, and humidity environments used for qualification of equipment outside containment result from an MSLB, or other HELBs, whichever is limiting for each plant area. The HELB temperature and pressure profiles for CLTP conditions were determined to be bounding for CPPU conditions. The accident temperatures outside containment resulting from a LOCA/MSLB inside containment remain unchanged. The normal temperature, pressure, and humidity conditions slightly increased in some rooms containing EQ equipment as a result of CPPU. However, the design limits used for EQ evaluations bound the increased levels.

The post-accident radiation exposure in the reactor building remains bounding for the CPPU condition. The post-CPPU 40-year normal gamma radiation levels in the main steam tunnel (Room No. 4316) and some areas of the turbine and auxiliary buildings are estimated to increase by less than 20% due to the CPPU. Despite the increase in the post-CPPU 40-year normal integrated dose, the qualified life of equipment located in the affected areas remains bounding due to the compensating margin in the qualified dose except the Barksdale Pressure Switches located in the turbine building, room 1313. The increased radiation doses result in a reduction of the radiation life of Barksdale Pressure Switches. A case-specific analysis was performed and determined that the radiation life of Barksdale Pressure Switches could be extended up to the design life of the plant. Therefore, the qualified life of these switches was revised to be acceptable for the remaining life of the plant.

10.3.2 Mechanical Equipment With Non-Metallic Components The temperatures in the areas containing safety related mechanical equipment do not increase from the CLTP levels. The accident radiation level and the normal radiation level increase due to CPPU as discussed in Section 10.3.1.

Reevaluation of the safety related mechanical equipment with non-metallic components identified some equipment potentially affected by the CPPU conditions. The qualification of this equipment (resilient seat check valves and LISEGA Type Hydraulic Snubbers) was resolved by reanalysis.

10.3.3 Mechanical Component Design Qualification The mechanical design of equipment/components (pumps, heat exchangers, etc.) in certain systems is affected by operation at CPPU due to slightly increased temperatures, and in some cases, flow. The revised operating conditions do not significantly affect the cumulative usage fatigue factors of mechanical components.

The effects of increased fluid induced loads on safety-related components are described in Sections 3 and 4.1. Increased nozzle loads and component support loads due to the revised operating conditions were evaluated within the piping assessments in Section 3. These increased loads are insignificant, and become negligible (i.e., remain bounded) when combined with the 10-5

NEDO-33076 governing dynamic loads. Therefore, the mechanical components and component supports are adequately designed for CPPU conditions.

10.4 TESTING Testing is required for the initial power ascension following the implementation of CPPU. The topics addressed in this section are:

...ic .. ,HpC'k sut . .. -...

Topic CLTR Disposition Hope Creek Result Plant/Component Testing Large Transient Testing ]

((]

Based on the analyses and GE BWR experience with uprated plants, a standard set of tests has been established for the initial power ascension steps of CPPU. These tests, which supplement the normal Technical Specification testing requirements, are as follows:

• Testing will be performed in accordance with the Technical Specifications Surveillance Requirements on instrumentation that is re-calibrated for CPPU conditions. Overlap between the IRM and APRM will be assured.

• Data will be taken at points from 90% up to 100% of the CLTP, so that system performance parameters can be projected for CPPU power before the CLTP RTP is exceeded.

• CPPU power increases will be made in predetermined increments of power. Operating data, including fuel thermal margin, will be taken and evaluated at each step. Routine 10-6

NEDO-33076 measurements of reactor and system pressures, flows, and vibration will be evaluated from each measurement point, prior to the next power increment. Radiation measurements will be made at selected power levels to ensure the protection of personnel.

• Control system tests will be performed for the reactor feedwater/reactor water level controls, pressure controls, and recirculation flow controls, as applicable. These operational tests will be made at the appropriate plant conditions for that test at each of the power increments, to show acceptable adjustments and operational capability.

• Steam dryer/separator performance will be confirmed within limits by determination of steam moisture content as required during power ascension testing..

• Testing will be done to confirm the power level near the turbine first-stage scram bypass setpoint.

The same performance criteria will be used as in the original power ascension tests, unless they have been replaced by updated criteria since the initial test program. ((

h]

Hope Creek does not intend to perform large transient testing involving an automatic scram from a high power. Transient experience at high powers and for a wide range of operating power levels at operating BWR plants has shown an acceptable correlation of the plant transient data to the predicted response. The operating history of Hope Creek demonstrates that previous transient events from full power are within expected peak limiting values. The transient analysis performed for the Hope Creek CPPU demonstrates that all safety criteria are met and that this uprate does not cause any previous non-limiting events to become limiting. Based on the similarity of plants, past transient testing, past analyses, and the evaluation of test results, the effects of the CPPU RTP level can be analytically determined on a plant specific basis. No new design functions that would necessitate modifications and large transient testing validation were required of safety related systems for the CPPU. The instrument setpoints that were changed do not contribute to the response to large transient events. No physical modification or setpoint changes were made to the SRVs. No new systems or features were installed for mitigation of rapid pressurization anticipated operational occurrences for this CPPU. A scram from high power level results in an unnecessary and undesirable transient cycle on the primary system. Therefore, additional transient testing involving a scram from high power levels is not justifiable. Should any future large transients occur, Hope Creek procedures require identification of any anomalous plant response and verification that all key safety-related equipment, required to function during the event, operated as anticipated or expected. Existing plant event data recorders are capable of acquiring the necessary data to confirm the actual versus expected response.

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NEDO-33076 Transient mitigation capability is demonstrated by other tests required by the Technical Specifications. In addition, the limiting transient analyses are included as part of the reload licensing analysis.

10.5 INDIVIDUAL PLANT EVALUATION Probabilistic risk assessments (PRAs) are performed to evaluate the risk of plant operation. The topics considered in this section are:

-Topic CLTRDisposition, Hope Creek Result 10.5.1 Initiating Event Frequency 10.5.2 Component Reliability 10.5.3 Operator Response 10.5.4 Success Criteria 10.5.5 External Events 10.5.6 Shutdown Risk 10.5.7 PRA Quality Analysis Framework and Results Summary Regulatory Guide 1.174 (Reference 30) provides the guidance framework for using PRA in risk-informed decisions for plant-specific changes to the licensing basis. The quantitative risk metrics chosen by the NRC in Regulatory Guide 1.174 are the changes to Core Damage Frequency (CDF) and Large Early Release Frequency (LERF).

The RG 1.174 acceptance guidelines consider both the initial values and the magnitudes of the changes in CDF and LERF as a result of the proposed change to the licensing basis.

Hope Creek PRA models for internal events, including internal floods, that represent the Current Licensing Thermal Power (CLTP) and the Constant Pressure Power Uprate (CPPU) model are developed. External event initiators and shutdown conditions are each addressed in the evaluation, but are not explicitly included in the quantitative discussion. External event initiators and shutdown events are each determined to be very small contributors to the change in risk associated with CPPU implementation.

The resulting quantitative changes in risk metrics associated with CPPU implementation are summarized in Table 10-3 and described below.

• The CLTP and CPPU CDFs are both well below IE-4 events per year for the internal events. The change in CDF associated with CPPU implementation is 6.8E-7/yr.

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• The CLTP and CPPU LERFs are both well below IE-5 events per year for the internal events. The change in LERF associated with CPPU implementation is 6.1E-8/yr.

These two risk metric changes place the change in risk for Hope Creek in Region III (very small risk change) of the RG 1.174 acceptance guideline.

In Region III, the calculated increase in CDF is less than IE-6 and the calculated increase in LERF is less than IE-7 per year. As stated in Regulatory Guide 1.174,

• When the calculated increase in CDF is very small, which is taken as being less than 10-per reactor year, the change will be considered regardless of whether there is a calculation of the total CDF (Region III). While there is no requirement to calculate the total CDF, if there is an indication that the CDF may be considerably higher than 104 per reactor year, the focus should be on finding ways to decrease rather than increase it.

• When the calculated increase in LERF is very small, which is taken as being less than 10-7per reactor year, the change will be considered regardless of whether there is a calculation of the total LERF (Region III). While there is no requirement to calculate the total LERF, if there is an indication that the LERF may be considerably higher than 10-5 per reactor year, the focus should be on finding ways to decrease rather than increase it.

The CDF changes are very small because the effect of CPPU is limited to a few success criteria differences, small changes in initiating event frequency, and small changes in consequential effects including available timing for crew actions.

The change in LERF is very small due to the fact that mitigation capability to prevent radionuclide release is not significantly affected by the CPPU. This mitigation capability includes:

• Containment flooding for cooling the core

• Containment spray system for scrubbing fission products and cooling debris

• The containment capability itself

• The reactor building as a fission product retention location (Not credited in the PRA)

• RHR system for containment cooling There are slight changes in accident progression timing resulting from the increased decay heat.

However, the slight changes are negligible compared with the overall timing of the core melt accident progression.

Therefore, LERF changes are primarily due to the changes in Level 1 results, i.e., CDF accident frequency increases. The Level 1 CDF results are dominated by accident sequences that do not contribute to LERF. Specifically, the dominant sequences in Level I include late SBO releases 10-9

NEDO-33076 and loss of decay heat removal. Other effects associated with core melt progression due to CPPU have a negligible effect on LERF.

Given the minor change in Level 1 CDF results, minor changes in the Level 2 release frequencies can be anticipated. Such changes are directly attributable to the minor changes in initiating event frequencies, short-term accident sequence timing, and the effect on human error probabilities (HEPs).

Fission product inventory in the reactor core is higher as a result of the increase in power due to the CPPU. The increase in fission product inventory results in an increase in the total radioactivity available for release given a severe accident. The total activity available for release is approximately 20% higher based on the assumption that the power uprate is 20%. However, this does not affect the definition or quantification of the LERF risk measure used in Regulatory Guide 1.174.

PRA Models Used for Risk Calculations The Hope Creek PRA has been updated (designated Revision 2005B) to ensure that the level of detail, fidelity with the as-operated, as-built plant (including CPPU implementation), and quality of the PRA all are acceptable to support the use of the PRA for applications.

To ensure the PRA addresses fidelity and quality, the Revision 2005B model includes recent operating experience for incorporation of plant trips, outages and failure data, plant modifications (including CPPU related changes), changes to plant procedures, changes in operator training, and success criteria based on enhanced MAAP calculations which are all representative of the CPPU condition. The CPPU model incorporated all of the changes, plant modifications, and power increase effects into this new model (Model 2005B). In addition, the 2005B model included significant improvements to resolve PRA peer review comments [See 4 to the CPPU License Change Request] and upgrade the model to Capability Category II of the ASME PRA Standard. (Reference 31)

Consistent with previous CPPU risk assessments (References 36, 37, 38), two PRA models are required for the evaluation of the change in risk metrics: (1) the CPPU condition, and (2) the CLTP configuration. Typically, the CLTP model forms the base model because it is the model of record, and the CPPU model is developed from the base model to represent the new configuration. However, the Hope Creek CPPU model was the one recently developed and upgraded from a previous peer reviewed Hope Creek model. Therefore, the CLTP model was derived from the recently updated CPPU model.

As noted, two categories of changes and improvements were identified for inclusion in the model update: CPPU related changes and Non-CPPU related changes. The CPPU related changes included CPPU related hardware and configuration changes, updated thermal hydraulic analysis, procedure changes, success criteria changes, initiating event frequencies, and timing changes all associated with the power uprate. The Non-CPPU changes which are included in both models are generic model improvements related to the ASME PRA Standard (Reference 31), as well as routine updates for data, human reliability analysis, non-CPPU related configuration changes and correction of identified errors and omissions. These non-CPPU changes are necessary to ensure 10-10

NEDO-33076 a detailed, robust PRA model to resolve PRA Peer Review comments and to upgrade the model consistent with the ASME PRA Standard. (Reference 31) [See Attachment 14 to the CPPU License Change Request.]

The CLTP model, then, also requires the same pedigree as the CPPU model. Therefore, the following approach is used to:

• Establish the CPPU Model as described above (2005B).

• Establish the CLTP model by incorporating only the non-CPPU changes referred to above

• Determine the delta risk resulting from the power uprate by comparing the CPPU model to the CLTP model.

The change in risk metrics associated with CPPU implementation is calculated with the two models as described above. The two models are compared in tabular form in Table 10-4. Table 104 provides a comparison of the potential differences in the two models, CLTP and CPPU, tabulated by PRA element. In Table 10-4, Column 2 represents the CLTP configuration and is derived from the Revision 2005B as noted. Column 3 represents the CPPU condition.

Table 10-5 describes the most significant differences between the CLTP and CPPU PRA models.

The changes are evaluated to provide input to decision makers regarding the licensing change to the Hope Creek licensed power level. Table 10-5 summarizes those differences that are judged appropriate to calculate a realistically conservative change in the risk metrics of CDF and LERF.

The approximate contribution to the change in CDF is also reported.

In addition to the differences identified in Table 10-5, there are other changes that are perceived as conceivable effects due to CPPU implementation. These changes are the subject of sensitivity evaluations that point out the magnitude of variation in the risk metrics of ACDF and ALERF for these potential effects. Table 10-6 provides the results of these sensitivity cases. The sensitivity cases documented in Table 10-6 demonstrate that even using extreme bounding assumptions on parameters still result in the ACDF risk metric remaining in Region III or just barely in Region II.

Plant Modifications The plant modifications associated with the CPPU have been identified by PSEG as input to this assessment. The modifications to be implemented as part of the power uprate are discussed below.

Plant changes that may affect the risk profile:

1. Turbine First Stage Pressure (TFSP) Scram Bypass Permissive setpoint will be changed from 30% RTP to 24% RTP. As a result, many inputs to scram will be "armed" at a lower power level than before CPPU. This will reduce the margin between the nominal operating band and the scram setpoint. There is a slight possibility that operational transients could induce a scram with a higher frequency than observed in past operating 10-11

NEDO-33076 experience. This, however, has not been observed among the BWR plants that have implemented extended power uprates. This could result in a turbine trip frequency increase.

2. All three reactor feed pump turbines (RFPT) will operate at a higher speed than before CPPU. Given a Feedwater Controller Failure, the time available for operator action may decrease. There is a small probability that this operational transient could induce a scram with a higher frequency than previously observed.

3. The condenser backpressure operating value will be increased from 4.0 in. HgA to 4.5 -

4.8 in. HgA, nominally during summer months. This will reduce the operating margin to the turbine trip signal due to high backpressure. This means that the scram signal will be reached faster at CPPU. This reduces the operational margin between the operating condenser vacuum and the turbine trip (7.5 in. HgA), trip of all feedwater (10.0" HgA),

and MSIV Closure setpoints (21.5 in. HgA). There is a slight possibility that operational transients could induce a scram with a higher frequency than observed in past operating experience. This, however, has not been observed among the BWR plants that have implemented extended power uprates.

Hope Creek is currently operated with all 3 Reactor Feedwater Pumps (RFPs), 3 Primary Condensate Pumps (PCPs) and 3 Secondary Condensate Pumps (SCPs) at full power and has automatic Reactor Recirculation (RR) Runback on loss of one RFP or PCP or SCP. This logic remains the same for CPPU with only minor adjustment expected. The existing RR runback circuit has actuated several times during the past 8 years and has not resulted in a scram.

Other plant changes are expected to result in equivalent operational and reliability conditions and are not expected to affect the risk profile (i.e., not result in an increased risk after CPPU implementation). These changes include the following:

• Operating range flexibility analysis (MELLLA)

• Addition of a 500 kV breaker in the Hope Creek substation for grid stability

• Replacement of the A and B phase generator step up (GSU) transformers

• Main generator stator water cooling pump upgrades

• Replacement of the HP and LP turbines

• Cooling tower improvements

• Feedwater heater dump valve replacement

• Moisture separator upgrades

• Replacement of analog EHC with digital EHC

• Modifications to the isolated phase bus to increase rating Note that model changes and sensitivity cases are presented in tables 10-4 and 10-5 to indicate the potential effect of these changes on the risk metrics.

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NEDO-33076 Procedural Changes Adjustments to the Hope Creek Emergency Operating Procedures (EOPs) and Severe Accident Mitigation Guidelines (SAMGs) will be made to be consistent with CPPU operating conditions.

In almost all respects, the EOPs and SAMGs are expected to remain unchanged because they are symptom-based; however, certain parameter thresholds and graphs are dependent upon power and decay heat levels and will require revision.

Based on the CPPU evaluations, EOP variables that play a role in the PRA and may require adjustment for the CPPU include:

• Boron Injection Initiation Temperature (BIIT)

• Heat Capacity Temperature Limit (HCTL)

• Primary Containment Pressure Limit (PCPL)

These variables may require adjustment to reflect the change in power level, but will not be adjusted in a manner that involves a change in accident mitigation philosophy. Because the HCTL and PCPL relate to long-term scenarios, any changes in the scenario timings associated with CPPU changes to these curves will be minor (e.g., changes on the order of 10-15 minutes over accident times greater than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) and would not significantly affect the human error probabilities in the PRA. No major perturbations in procedures have been identified.

Plant Operating Conditions The key plant operational modifications to be made in support of the CPPU are:

• Increase in reactor thermal power from 3339 MWt to 3840 MWt; equal to a 15% increase in thermal power, however, in general, the PRA is based on an assumed 20% of thermal power increase.

• Feedwater/Condensate flow rates will increase by approximately 20% over the flow rates at OLTP.

• Operation at a higher condenser backpressure.

10.5.1 Initiating Event Frequency Thirty-five initiating event categories are considered in the baseline and updated PRAs:

seventeen transient initiator categories; eleven internal flooding initiator categories and seven LOCA initiator categories. The initiator categories for the CPPU and CLTP models are the same.

Hope Creek operating experience (CLTP) and generic data have been used to provide Bayesian updated initiating event frequencies for the transients. The CPPU does not result in plant equipment operation beyond the design ratings and conditions. In addition, initial CPPU 10-13

NEDO-33076 experience from eleven BWRs has been reviewed to ascertain whether there may be an increase in initiating event frequencies due to CPPU related causes. Based on the initial BWR CPPU experience (approximately 12 reactor years), there does not appear to be any trend toward increased scram frequency due to CPPU implementation. This analysis did identify issues related to steam dryer vibration or small bore pipe vibration that have led to manual shutdowns and outages. However, no increase in the total manual shutdown frequency was found, therefore, these do not measurably affect the risk metrics. Nevertheless, the comparative risk assessment has included an increase in the scram frequency to analytically account for possible higher scram frequencies associated with some reductions in operating margins. The transient categories with the most potential to be affected by the CPPU are those that can be influenced by reactor or turbine trip setpoints or margins to those setpoints, such as reactor scram, system isolations, and operating equipment trips. A review of these transient categories concluded that the operational margins remain adequate such that changes to the baseline PRA are not required to reflect CPPU conditions. Nevertheless, increases in the frequency of "turbine trip with bypass" due to possible turbine trip contributors identified in Table 10-7 are modeled in the CPPU model. The Turbine Trip frequency increase encompasses any risk effect of changes in manual shutdown frequency.

Table 10-7 summarizes the potential changes in the initiating event frequencies. This is modeled for the CPPU conditions by increasing the turbine trip frequency of the Revision 2005B model by 21% relative to the CLTP assessed turbine trip frequency. The 21% increase is based on engineering judgment. The ACDF sensitivity to this assumed CPPU effect is included in Table 10-5. The contributors are identified in Table 10-7 and discussed below.

• The TCV fast closure and TSV closure being enabled at a lower power level than before CPPU may result in additional scram challenges. This is considered a small effect; however, for the delta risk calculation associated with CPPU implementation, this is modeled as a 10% increase in the turbine trip frequency.

• The FW controller failure may result in a shorter time for operator action in the CPPU conditions compared with the CLTP conditions. The Boolean combination of FW controller failure and operator response failure is estimated to occur at less than IE-3/yr.

This is not considered numerically significant relative to the turbine trip frequency.

• The reduction in margin between the condenser backpressure setpoint and the nominal operating conditions as a result of CPPU implementation may result in additional scram challenges. This is considered a small effect, however, for the delta risk calculation associated with CPPU implementation, this is modeled as a 10% increase in the turbine trip frequency.

• Changes to the reactor recirculation runback logic described in sections 6.4.2 and 7.4.2 are not expected to cause an increase in spurious trips or failures to actuate when required, however, for the delta risk calculation associated with CPPU implementation, these changes are modeled as a 1% increase in turbine trip frequency.

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NEDO-33076 Postulated effects related to flow accelerated corrosion (FAC) might result in an increase in pipe failure frequency in feedwater or main steam lines. In addition, flow induced vibration (FIV) may cause failures of small-bore piping leading to manual reactor scrams for repairs. Because the nominal RPV pressure does not change at the CPPU conditions, the increase in power requires the feedwater and steam flows to increase. The basic robust Hope Creek design and the continued regular pipe inspection and analysis programs minimize the postulated effects.

Therefore, no changes in LOCA or turbine trip frequencies are anticipated. However, sensitivity evaluations are performed to demonstrate the potential risk effect associated with increased LOCA and turbine trip frequencies.

• FIV and FAC Steam flow and feedwater flow will increase as a result of the CPPU implementation. It has been postulated that the increase in flow will result in the potential for increases in vibration (e.g., at different harmonics than under current conditions) and will cause an increase in FAC. Both phenomena are anticipated and steps are included in the CPPU implementation at Hope Creek to address both items. Because of these preparations, neither of the phenomena is anticipated to result in failures at Hope Creek.

• FIV Modeled in the PRA Some components may be subjected to an increase in FIV due to higher flows and potentially different frequency harmonics. This effect has been observed in the operating experience with CPPU implementation at some of the eleven BWRs that have implemented CPPU, e.g.:

- Adverse small-bore pipe vibration associated with different plant harmonics after CPPU implementation has caused increased leakage into the drywell

- Adverse flow induced vibration on the dryer assembly at Quad Cities Unit 2 (BWR/3) has caused forced plant shutdowns Sensitivity cases in Table 10-6 evaluated issues related to small-bore attached pipe vibration that has occurred in plants that have implemented CPPU.

This FIV quantification case is only considered a sensitivity case because the possibility of increased initiating event frequency is considered a pessimistic assumption and in addition, would only appear during the "break-in" period or the initial two years of plant operation at CPPU.

A sensitivity analysis by increasing the Inadvertently Open SRV (IORV) Initiating Event frequency was also performed and found that the effect on CDF is not significant.

• FAC Modeled in the PRA The FAC phenomena have been assessed for Hope Creek and do not result in component failures for the Hope Creek CPPU condition.

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NEDO-33076 The periodic inspections of the Hope Creek piping, vessels, and components are expected to identify any areas that are degraded due to FAC and to identify the appropriate steps to prevent failures. Therefore, the PRA does not include additional failures in the model to address these slowly developing failure modes that can be discovered by periodic inspections.

Nevertheless, a quantitative sensitivity case was performed that doubled the large LOCA initiating event frequency. This sensitivity case is summarized in Table 10-6.

No effects that result in changes to internal flood or special initiating events (support systems) have been identified associated with CPPU implementation. The Revision 2005B PRA model identifies nine initiators as a result of the loss or degradation of support systems. The duty on these systems is essentially unchanged as a result of the CPPU. Therefore, the frequency of these special initiators does not change at CPPU conditions.

There is the potential that the PSEG grid is lost following a trip of the Hope Creek unit. Rapid separation of a large generating unit from the grid has the potential to cause grid instability and loss of offsite power. This possibility is represented in the PRA. A grid stability analysis has been performed, considering the increase in electrical output, which demonstrates conformance to General Design Criterion 17. In addition, the PJM Interconnection grid analysis for the worst-case three-phase or single-phase fault identified the requirement of adding another 500kV breaker to the plant switchyard. This plant modification (part of CPPU) will make the postulated Hope Creek electrical fault effects on grid stability acceptable. The implementation of this plant change makes the CPPU and CLTP grid response the same. Therefore, the frequency of loss of offsite power events due to grid instabilities is not affected by the CPPU.

The LOOP frequency is based on PJM grid data because this is deemed more representative than the limited information available from the Hope Creek history of LOOP events. The LOOP initiator frequency of 3.04E-2/year is calculated using a Bayesian approach with the prior distribution based on 1980-1995 PJM operating experience.

The LOOP non-recovery probabilities entered into the PRA model are calculated from the modified Weibull equation that has been fit to the data. The time used in this equation is the time to recover off-site power and tie in to the plant buses. The non-recovery probability assessment is divided into three LOOP categories of plant-centered, grid related and severe weather following the approach of NUREG/CR-5496.

In addition, consequential LOOP induced by a Hope Creek scram has been incorporated into the 2005B model. The conditional LOOP probabilities are a function of whether the Hope Creek scram has a coincident LOCA signal generated or not. The table below shows the conditional LOOP probability and its contribution to CDF for the CPPU model.

Conditional LOOP CPPU 2005B Model Hope Creek Scram Event Probability  % Contribution to CDF Without LOCA Signal 3E-3 8.3%

With LOCA Signal IE-2 < 0.2%

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NEDO-33076 Reactor Protection System reliability is not changed as a result of the CPPU implementation.

The total frequency of all ATWS scenarios is estimated by multiplying the frequency of the transient events by the likelihood of failure to scram. As discussed above, the frequency of occurrence of the transient initiators does not change after the modifications are implemented except for the postulated change in the turbine trip initiating event frequency. No modifications are anticipated to affect the likelihood of scram. Therefore, the frequency of ATWS challenges may change slightly, but only due to the initiating event frequency changes.

Table 10-8 provides a breakdown of initiator contributions to CDF for the CLTP and CPPU conditions. Six (6) Hope Creek initiators are not included in the evaluation because those cutsets were below the truncation value.

For the majority of the initiator categories, the CPPU has no effect on the frequency of occurrence. The initiator category potentially most affected is the turbine trip with bypass.

10.5.2 Component and System Reliability No increase in component failure rates is anticipated. Under CPPU conditions, equipment operating limits, conditions, and/or ratings are not exceeded. Existing plant component monitoring programs detect degradation if it occurs and corrective action is taken in a timely manner. It is possible that CPPU conditions may result in selected components requiring refurbishment or replacement more frequently; however, the functionality and reliability of components and systems is maintained at the current standard.

Therefore, individual component failure probabilities are not expected to increase as a result of the CPPU implementation.

Some of the considerations included in the individual component evaluation are as follows:

• Break-in period Components being replaced are considered to be replacement "in kind" of the components. Therefore, no increase in failure probability is anticipated. The exception to this is that there may be some increase in failure probability associated with the initial "break-in" period for these components. This has been accounted for in the CPPU condition calculation by increasing the BOP component failures leading to initiating events for the initial two years of CPPU operation, i.e., projected "break-in" period.

Historical evidence suggests that during the implementation of major plant changes some design, installation or operating issues arise that result in slightly reduced reliability during the early stages of operation. This early stage or "break in" period is generally in the time frame of 1 to 2 year duration.

Figure 10-1 is a representation of the theoretical "bathtub" curve representing individual component failure rates as a function of time in component life.

Region A of Figure 10-1 represents the high initial failure rate of components generally referred to as the "break-in" portion of component's life. Wear-out phenomena (Region C) are generally considered to occur after the useful life of a component. The details of 10-17

NEDO-33076 this curve vary for every component, but these major characteristics are nearly always present. Because the CPPU implementation involves some substantial changes to the balance of plant (BOP) systems, the effects of the "break-in" period and possible higher failure rates are explicitly evaluated in the CPPU risk assessment. The effect on risk is small.

• Reactor Vessel Integditv The CPPU condition results in increased neutron flux on the reactor vessel. This increased fluence will result in increases in the nil ductility temperature. However, the effect on BWRs, and Hope Creek in particular, are so small as to not affect the assessed RPV failure probability.

No changes are being made to the limiting Pressure-Temperature curve in the Technical Specification for the CPPU implementation because the existing curves account for the increased fluence due to CPPU.

The change in risk metrics for Hope Creek due to increased risks of pressure vessel failure associated with the increased neutron flux after power uprate are found to be so small as to be negligible changes in risk.

• SORV Failure Probability The SRV setpoints have not been changed as a result of the Hope Creek CPPU. Given the power increase of the CPPU, one may postulate that the probability of a stuck open relief valve (SORV) during an accident scenario would increase due to an increase in the number of SRV cycles.

The SORV probability is modified assuming that the SORV probability is linearly related to the number of SRV cycles, but the number of cycles is not necessarily directly related to the reactor power increase. The increase in the number of SRV cycles during accident response is estimated here by comparing the results of the MAAP runs performed in support of this analysis.

Review of MAAP cases for CPPU versus CLTP indicates that the number of SRV cycles associated with CPPU implementation in the first few hours of the accident progression increases by 13% for the CPPU power level versus the CLTP power level for non-ATWS conditions and by 9% for ATWS conditions.

Using this information, the Hope Creek PRA CPPU and CLTP SORV probabilities are summarized in Table 10-9.

The CPPU conditions are calculated to have an increased probability of an SORV in comparison with the CLTP configuration. This results from an increase in the number of SRV challenges assigned to transients and ATWS.

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• Structural Evaluations This assessment did not identify issues associated with postulated effects from the CPPU on the PRA modeling of structural (e.g., piping, vessel, containment) capabilities.

10.5.3 Operator Response The Hope Creek risk profile, like other plants, is dependent on the operator actions for successful accident mitigation. The success of these actions is in turn dependent on a number of performance shaping factors (PSF). The PSF that is principally influenced by the CPPU is the time available for the operator to detect, diagnose, and perform the required actions. The higher power level results in reduced times available for some actions. To quantify the potential effect of this PSF, deterministic thermal hydraulic calculations using the MAAP computer code were used. Also changes in the response of the SACS system (the intermediate safety system cooling loops) were evaluated as they may influence crew actions. One of the SACS heat loads is the RHR heat exchangers, and therefore, suppression pool cooling is affected by the CPPU implementation due to the higher decay heat load that must be handled. The SACS calculations were performed using deterministic models developed by PSEG and confirmed to be consistent with MAAP.

The human reliability analysis (HRA) evaluates the effect of the CPPU implementation on operator response capabilities following an initiating event requiring safe plant shutdown. The operator response evaluation for Hope Creek CPPU risk assessment included a close examination of the 123 post-initiating event operator actions used in the Revision 2005B model.

No new actions are added to the model specifically in responding to the CPPU. The evaluation addresses all post-initiating event actions by considering two factors: 1) the reduction in the amount of time available for the operators to diagnose and execute an action; and, 2) whether the reduced times available are an influence on the error rates calculated for the human actions in the PRA.

Operator action dependencies are assessed by two techniques: (1) embedding the dependency in the model (ATWS actions); (2) post processing the cutsets to identify cutsets with multiple HEPs. The post processing quantification assessment captures cutsets containing more than one action and assesses their dependencies.

The CPPU evaluation used MAAP 4.0.4 detailed thermal hydraulic calculations to calculate the time available for the operator response to accidents. Similar (corresponding) calculations for the CLTP configuration are not available. However, the time available for operator response for the CLTP configuration can be estimated using available MAAP calculations.

The estimated decreases in times available for the CPPU conditions are as follows:

• Except where noted in Table 10-10, the CPPU power level is assumed in the model to be 20% higher than CLTP and results in approximately a 25% decrease in the CPPU available operator response time for postulated accidents such as loss of make-up capability.

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• Time to reach TAF is decreased by slightly more than 20% for the CPPU condition. This is related to such things as increased RPV inventory (decreased voids). It is reasonable to assume a 25% shorter time available for such short-term actions dependent on the loss of inventory and for LOCA event responses for the CPPU condition.

• ATWS response action time is expected to be approximately proportional to the power change for CPPU condition. Therefore, a 20% shorter time available for some ATWS response actions is assumed. In some other ATWS response actions, the core power level is not the controlling variable and no change in time results. In addition, Hope Creek has automatic SLCS initiation as a design feature. Therefore, the change in the operator HEP for SLCS manual backup to this automatic initiation has a negligible effect on the risk metrics for the implementation of CPPU. This design feature is different from previous BWR CPPU submittals.

Where noted in Table 10-10, CPPU timing for some operator actions was calculated based a decay heat 12.3% greater than OLTP. This is adequate, based on engineering judgment, to represent decay heat for CPPU for these operator actions.

The post-initiator HEPs in the CLTP model were evaluated using the same method as the CPPU PRA model, with the time available modified to reflect the above differences in available time for operator action between the CPPU and CLTP conditions. The HEPs used in CPPU Model were developed with analysis and operator interviews conducted with the latest EOPs.

Table 10-10 summarizes an evaluation of the top 20 Hope Creek operator actions (ranked by Fussell-Vesely) for CPPU, down to a F-V of 0.01.

The HEPs are, in general, calculated using the EPRI Cause-Based Methodology (Reference 34) for the cognitive portion of the analysis (as implemented in the EPRI HRA Calculator). The EPRI calculator methodology results in minimal effects on the calculated HEPs due to CPPU.

Some baseline Hope Creek HEPs were calculated using a combination of the Cause-Based Methodology the Accident Sequence Evaluation Program (ASEP) time reliability correlation (Reference 35). These HEPs were also recalculated using the estimated time available.

All modified HEPs with longer times available for diagnosis and action are then input directly into the CLTP PRA model.

10.5.4 Success Criteria The ability to safely achieve a safe stable state following a challenge (initiating event) depends on the success of systems used for core damage prevention. The success criteria for these systems could be influenced by the CPPU implementation.

The success criteria for the Hope Creek PRA are derived based on realistic evaluations of system capability over the 24-hour mission time of the PRA analysis. These success criteria, therefore, may be different than the design basis assumptions used for licensing Hope Creek. This report examines the risk profile changes, caused by CPPU, from this realistic perspective to identify changes in the risk profile that may result from severe accidents on a best estimate basis.

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NEDO-33076 For discussion purposes, the success criteria are divided into the following functions:

• Reactivity control

• RPV pressure control

• RPV inventory control

• Containment pressure and temperature control For each of these functions, individual systems provide the response. There are a number of potential effects of the CPPU that could alter success criteria of the systems used for accident prevention. They include the following:

• Time to reach TAF

• Heat load to the suppression pool

• Blowdown loads

• RPV overpressure margin (number of SRVs required)

• Depressurization (number of SRVs required)

• Flow rates or numbers of systems required The latest deterministic calculations representing the CPPU conditions are included in the Hope Creek PRA thermal hydraulics model (MAAP 4.0.4). Each of these functions is discussed along with the potential CPPU effects that may alter the success criteria. The following subsections discuss different aspects of the success criteria as used in the PRA.

10.5.4.1 Reactivity Control The scram function is not affected by CPPU because operation of the Control Rods, the Control Rod Hydraulic System and the Reactor Protection System (RPS) are not affected by the CPPU.

Due to the increased nominal initial power level, the time available to perform selected operator actions decreases. Examples include initiation of the SLCS in the ATWS scenarios and control of the RPV level following a transient. These effects are discussed in Section 10.5.3. However, the success criteria for the SLC System are not changed. Note that Hope Creek has automatic initiation logic for SLC and therefore is not dependent on operator action for successful injection.

10.5.4.2 RPV Pressure Control There are two aspects of RPV Pressure Control:

• Overpressure protection

• RPV depressurization 10-21

NEDO-33076 Overpressure Protection The RPV overpressure protection is important for all conditions. The two that are potentially affected are discussed here: (1) transient response; and, (2) ATWS response. The transient overpressure protection is accomplished by successful operation of 4 of the 14 available SRVs.

This applies equally to CLTP and CPPU conditions based on MAAP calculations and GE calculations.

The failure-to-scram conditions are believed to present more severe challenges to the pressure relief capability. Recirculation pump trip (both pumps) is found to be required for ATWS isolation events before and after CPPU implementation. The number of SRVs required for this condition increases from 11 of 14 to 12 of 14 SRVs with the CPPU implementation. This represents a change in success criteria to be assessed in the PRA based on GE calculations.

Other than these changes, for RPV integrity success criteria, no changes are identified. See further discussion under Section 10.5.2 regarding reactor vessel integrity.

RPV Depressurization The CPPU MAAP calculations indicate that 2 SRVs are required to provide adequate RPV depressurization when required to assure adequate low-pressure injection for the CPPU conditions. The success criteria were previously identified to be a single SRV for the CLTP conditions.

Therefore, the success criteria in the PRA for CLTP and CPPU are different.

The failure probability to depressurize is dominated by the following:

• Support system failures - not affected by CPPU

• Operator errors - see Operator Response (Section 10.5.3)

• Common cause failures of SRVs This latter item is influenced by the success criteria. However, with 14 SRVs, the inability to open I SRV or 2 SRVs is indistinguishable in CCF probabilities from available sources.

Therefore, the quantitative effect is conservatively assessed and it results in a negligible (1.OE-8/yr) influence on the risk metrics of ACDF and ALERF.

10.5.4.3 RPV Inventory Control The RPV inventory control success criteria are based on an integer number of system trains.

This treatment in PRAs usually results in substantial margins between the delivered flow and the flow required to mitigate accidents. This turns out to be true for Hope Creek for those systems able to satisfy the success criteria of preventing core damage.

Note that CRD injection is not adequate as the sole method of RPV inventory control for CPPU or CLTP configurations, but CRD is assumed effective at extended times into an event when decay heat is low.

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NEDO-33076 No change in the PRA models are required to model the integer number of injection systems required for RPV makeup for various accidents.

10.5.4.4 Containment Pressure and Temperature Control There are a number of different aspects to the containment success criteria including the following:

• Ultimate containment capability

• Vapor suppression system

• RHR heat removal

• Venting

• Drywell sprays Each of these was evaluated using Hope Creek MAAP models. The associated success criteria for each were unchanged for CPPU. This is because there is sufficient margin in the system capabilities to accommodate the CPPU effects. Therefore, no logic changes to the PRA were required. The timing of operator response actions for containment control actions are affected by the change in decay heat. These are addressed in Section 10.5.3 and found to be negligible for the relatively long allowable times for operator actions associated with controlling containment conditions.

Based on the above assessment of the individual functional success criteria, no changes to the Level I or the Level II event tree structure are required.

Table 10-11 summarizes the success criteria changes required by the CPPU implementation.

10.5.5 External Events The effect of the CPPU was reviewed to determine whether any new plant vulnerabilities exist from the occurrence of internal fires, seismic events, and other external events. Equipment changes associated with the CPPU are minor. The functionality and reliability of components and systems are maintained at the existing standards. Therefore, individual component failure probabilities are not expected to increase as a result of the CPPU.

The Hope Creek IPEEE for Fire, Seismic, high winds, external floods, and transportation was reviewed to determine whether there were any existing conditions where the CPPU could introduce new vulnerabilities.

10.5.5.1 Seismic Based on a review of the Hope Creek IPEEE seismic analyses, the CPPU implementation has little or no effect on the seismic qualifications of the systems, structures and components (SSCs).

The decrease in time available for operator actions, and the associated increases in calculated HEPs, is judged to have a non-significant effect on seismic-induced risk. Industry BWR seismic 10-23

NEDO-33076 PRAs have typically shown (e.g., Peach Bottom input to the NUREG-1 150 study (Reference 39); Limerick Generating Station Severe Accident Risk Assessment (Reference 40);

NUREG/CR-4448) that seismic risk is overwhelmingly dominated by seismic induced equipment and structural failures.

Note that the Hope Creek IPEEE seismic analyses and seismic walkdowns were not re-performed in support of this risk assessment. The effects of the CPPU on the different aspects of seismic risk modeling were assessed based on knowledge of the Hope Creek IPEEE and the Hope Creek CCPU. No significant changes to existing equipment mountings or building structures will be made as part of the CPPU that would affect the Hope Creek IPEEE. The CPPU equipment replacements will be installed using anchorages that are similar to existing equipment anchorages.

Based on the above discussion, it is judged that the percentage increase in the Hope Creek seismic risk due to the CPPU implementation is much less than that calculated for internal events.

The IPEEE concluded that seismic evaluations did not identify any unique or new vulnerability for Hope Creek. Because the CPPU modifications do not affect the seismic response, no new vulnerabilities are introduced as a result of a seismic event. The conclusion of the IPEEE remains applicable.

The Hope Creek seismic PRA model developed under the IPEEE program does not meet the ANS PRA Standard for external events and has not been quantified with the latest 2005B model.

Nevertheless, a quantitative seismic model is available to provide some additional risk information into the effects associated with this PRA application.

The seismic initiating events from the IPEEE and their consequential effects on SSCs have been incorporated into the previous Hope Creek model. The results indicate the seismic CDF is dominated by the following:

• Seismic Induced Loss of DC power 71.5%

• Seismic Induced Late SBO events 21.0%

92.5%

These contributors are judged not to be subject to effects from CPPU implementation that would modify:

• Equipment failures

• Success Criteria

• Operator response timing or actions

• LERF 10-24

NEDO-33076 The residual events of 7.5% of the seismic CDF may be minimally affected by crew response and equipment failures, but there is no significant change that would alter the ACDF conclusion from the internal events assessment.

The seismic LERF is not affected by the CPPU implementation for 92.5% of the seismic induced CDF. The same effects on containment, mitigation, and evacuation from seismic events exist for both CLTP conditions and CPPU conditions.

The LERF from seismic events may be a relatively large percentage of the seismic CDF due to increased difficulties with communication and evacuation. However, ALERF due to the CPPU implementation is considered to be minimal because the same sequences, frequencies, and relative source terms apply whether it is the CLTP or CPPU model.

10.5.5.2 Fire The Hope Creek plant risk due to internal fires was evaluated as part of the Individual Plant Examination of External Events (IPEEE) Submittal.

Note that the Hope Creek IPEEE fire analyses and fire walkdowns were not re-performed in support of this risk assessment. The effects of the CPPU on the different aspects of fire risk modeling were assessed based on knowledge of the Hope Creek Fire IPEEE and the Hope Creek CCPU. The results of the fire IPEEE analysis indicates that the risk is dominated by the fire induced failures of equipment or accessibility and not on the variables affected by the CPPU implementation such as success criteria or time available for crew actions.

The CPPU will not result in any significant changes to combustible loadings throughout the plant or changes to fire area boundaries. In addition, the Hope Creek CPPU does not involve any changes to the fire detection or fire protection systems in the plant.

Based on this assessment, it is concluded that no unique or significant effects on the fire risk profile result from the CPPU implementation.

The IPEEE review concluded that there are no new fire-induced vulnerabilities associated with the CPPU. The fire zones, fire loading, and safe shutdown paths for Hope Creek do not change for CPPU; therefore, there is no increase in the vulnerability to internal fires associated with CPPU implementation.

10.5.5.3 High Winds and Floods The Hope Creek Plant/Facilities design is robust in relation to the Standard Review Plan (SRP)

Criteria and walk-downs performed to support IPEEE did not reveal any potential significant vulnerability that was not included in the original design basis analysis. Because there are no external or other structural changes associated with the CPPU, there are no new vulnerabilities introduced from wind or flood events.

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NEDO-33076 10.5.5.4 Transportation and Nearby Facilities There are no changes in the CPPU that could be affected by transportation or nearby facility accidents. Thus, there are no new vulnerabilities introduced from transportation and nearby facility accidents.

10.5.6 Shutdomwn Risk The effect of the CPPU on shutdown risk is similar to the effect on the at-power Level 1 PRA.

Based on the insights from the at-power PRA, the areas of review appropriate to shutdown risk are the following:

• Initiating Events

• Success Criteria

• Human Reliability Analysis The following qualitative discussion applies to the shutdown conditions of Hot Shutdown (Mode 3), Cold Shutdown (Mode 4), and Refueling (Mode 5). The CPPU risk effect during the transitional periods such as at-power (Mode 1) to Hot Shutdown and Startup (Mode 2) to at-power are judged to be subsumed by the at-power Level I PRA. This is consistent with the U.S.

PRA industry, and with NRC Regulatory Guide 1.174, which states that, not all aspects of risk need to be addressed for every application. While higher conditional risk states may be postulated during these transition periods, the short time frames involved produce a insignificant effect on the long-term annualized plant risk profile.

At Hope Creek, shutdown risk is managed in accordance with outage management program and outage risk assessment procedures. These procedures provide a process for managing and assessing outage risk for both planned and forced outages. The process is based on NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management (Reference 32) and also satisfies the requirements of the Maintenance Rule (IOCFR50.65). The shutdown risk assessment process for Hope Creek monitors the following Key Safety Functions: shutdown cooling, electrical power, inventory control, reactivity control, spent fuel pool cooling and secondary containment. Color codes (GREEN, YELLOW, ORANGE and RED) are utilized to identify risk levels based upon defense in depth considerations. Contingency plans are required to manage the risk associated with plant configurations that are categorized as ORANGE (minimum allowed defense-in-depth). The CPPU has no effect on the process or procedures for managing shutdown risk.

A Probabilistic Risk Assessment (PRA) model to quantitatively evaluate shutdown risk in terms of CDF and LERF is not available for Hope Creek. Instead, PSEG Nuclear utilizes the Outage Risk Assessment and Management (ORAM) computer code to assist in the management of risk during shutdowns. The ORAM model contains the Safety Functional Assessment Trees (SFATs) used to assess the configuration risk associated with each Key Safety Function (KSF),

along with the Fault Trees (FT), User Variables (UV) and the Plant Configuration Database (PCDB) that support them. The CPPU has no direct effect on the defense in depth considerations associated with plant configuration and therefore no direct effect on any of the 10-26

NEDO-33076 logic contained in the Fault Trees or SFATs. However, the CPPU will increase the decay heat load following shutdown, which will affect the time interval before alternate decay heat removal (DHR) systems can be used. The decay heat level is accounted for in ORAM (User Variables based on time from shut down, such as DECAYH and FPHEAT). The ORAM model is sufficiently conservative such that these variables do not need to be revised as a result of the relatively small increase in decay heat at the extended times of plant shutdown for CPPU.

Decay heat loads and heat-up calculations (time to boil curves) are prepared sufficiently early in the planning stage of Refueling Outages (RFO) to allow consideration during the scheduling process. This involves consideration of the time available to implement mitigating actions if decay heat removal is unexpectedly lost. This information is used by the outage planning team to ensure that heat removal systems are available and that appropriate contingency plans are made for maintenance and testing of systems. Finalized decay heat load and heat-up curves are provided to the on-duty Operations Shift Superintendent immediately prior to the start of refueling outages.

The following is a more detailed discussion of the specific risk effects on shutdown operations associated with implementing a CPPU:

Increased decay heat - The additional decay heat load is not so large that new equipment is being added (for example, larger heat exchangers or an additional S/D cooling loop) or that success criteria are changed (for example, should two cooling loops be required to prevent core boiling instead of one). The existing plant equipment is sufficient to remove the additional decay heat. There is no direct effect on the ORAM FTs or SFATs, and the determination of risk based on defense in depth considerations will be unaffected by the CPPU.

Increased time to reach shutdown - With greater decay heat, it take longer in theory to cool down to the lower operational modes. However, this will not necessarily be realized in practice for normal shutdowns.

• While the calculated duration to reduce reactor coolant temperature to 200 'F after plant shutdown increases from 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, actual plant cooldowns are typically performed more slowly and are not expected to be affected by CPPU.

• Experience indicates that this evolution has at times taken up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> pre-CPPU. The first part of the cool down, to approximately 80 psig, involves drawing steam from the reactor and condensing it. The heat removal capability/timing associated with this method of cooling down is not challenged by CPPU. The next block of time is involved with lining up, flushing, pre-warming and placing RHR in service for SDC. This also is not affected by CPPU. The last block of time is involved with cooling down using RHR.

One RHR heat exchanger is presently used, and this is not expected to be different after CPPU. The maximum administrative cool down rate of 90 degrees/hour will still be within the heat removal capability of one RHR heat exchanger after implementing the CPPU. During this block of time operators will be closing the Bypass Valve Opening Jack, closing the MSIVs, closing steam drains and preparing to open the head vents, etc.

It is not expected that this sequence will be altered in any significant way due to implementation of the CPPU.

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NEDO-33076 Longer mission times - From a PRA perspective, increased times to reach shutdown would result in longer "mission times" for the primary Decay Heat Removal (DHR) components. In PRA space, the Test & Maintenance (T&M) and Fails to Start (FS) basic events would not be affected by CPPU. The Fails to Run (FR) basic events for the normal DHR systems would be slightly affected by these increased "mission times", however, the T&M and FS basic events drive the risk. The FR basic events have much lower probabilities (typically one to two orders of magnitude) and the additional mission time is not significant enough to cause the FR basic events to become as important as T&M or FS. Therefore, the increase in risk associated with longer mission times is insignificant when qualitatively analyzed from a PRA perspective.

Loneer times before alternative decay heat removal systems can be used - As discussed above, heat up curves are provided for outage scheduling purposes. One use of these curves is to ensure that alternate decay heat removal systems are not placed in service before they are known to have sufficient heat removal capability to meet all requirements and limitations. However, the possibility of using these systems on an emergency basis is considered here. A representative set of CPPU heat up curves was prepared by replacing the decay heat levels in the RF12 (Fall 2005) heat up curves with decay heat levels for a CPPU fuel load. These CPPU heat up curves were then superimposed over the actual RF12 heat up curves. If the various alternate decay heat removal systems were placed in service on Day 4, the superimposed heat up curves predict no more than an additional 10 degree F peak Fuel Pool temperature, and lower additional peak temperatures when placed in service on subsequent days. The highest projected peak temperature is 175 deg. F based on the alternate shut down cooling configuration of 2 RWCU pumps, I RWCU H/X and 1 FPC pump with initial SACS temperature at 75 deg. F and SFP at 100 deg. F. Use of the alternate system in this off-normal situation would still provide 36 degrees of margin to fuel pool boiling. It is judged that this presents no significant increase in risk.

Coincidentally, the existing treatment of DECAYH in the ORAM model is conservative and does not need to be revised for CPPU implementation. This variable addresses the relative amount of existing decay heat (high, medium-high, medium or low) in terms of time after shut down as follows:

DECAYH = H <3 days MH <6 days M <30 days L >30 days.

DECAYH is only used in the Shut Down Cooling (SDC) SFAT. When DECAYH becomes M (due to time after shut down > 6 days) then the SDC SFATs first begin to give credit for alternate decay heat removal (RWCU and/or FPC). More credit is given to alternate decay heat removal methods as time passes because the alternate DHR methods will be more capable of removing the decay heat loads as the loads decrease over time.

Shorter times to boiling - The outage management process requires the development of decay heat curves (heat up curves) to be used in accordance with outage management and risk assessment procedures. A review and comparison of CLTP and CPPU heat up curves mentioned above indicates that CPPU "times to boil" will generally be about 13% shorter. The risk significance for operator response times is discussed below.

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NEDO-33076 Shorter time for operator responses - This is a direct result of shorter times to boiling/boil-off.

The effect on operator response times from time zero until cold shutdown is achieved is addressed in the on-line, PRA based risk assessment. The following discussion covers the times and configurations subsequent to opening the Reactor Head Vents and entry into Operational Condition 4 (COLD SHUTDOWN). This discussion considers the reduction in postulated operator response times from the time of a postulated loss of normal DHR until the water inventory boils down to the top of active fuel (TAF). Fuel damage is conservatively assumed to occur when water level reaches TAF.

The most limiting time and configuration for this condition is immediately after venting the reactor vessel and entering Operational Condition 4. The decay heat and initial bulk water temperature are highest and the water volume is lowest compared to later in the outage. The initial bulk water temperature is assumed to be 199 'F. and water is assumed to be at normal level for entering Operational Condition 4. Calculations were performed to estimate the time to heat up the water inventory to saturated conditions at 212 'F, and then to boil down to the top of active fuel. The earliest that this could occur for CPPU decay heat levels is 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> from the time of shutdown. The calculation was performed using these initial conditions and was repeated for several later event times and conditions to assess the dynamics of the situation.

For the CLTP decay heat load, the calculated response time is 256 minutes (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 16 minutes). For the CPPU decay heat load, the calculated response time is 223 minutes (3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 43 minutes). This is a reduction of response time of 33 minutes, or approximately 13%.

This represents the most challenging time for a loss of normal DHR to occur. At all subsequent times, the calculated times for operator response are longer than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 43 minutes due to the natural reduction of the decay heat load and, for refueling outages, the larger water inventory as the refueling outage progresses. The reduction in calculated response times is consistently 13%. The cited time reduction is not considered significant with respect to operator response when there is more than three hours to make a diagnosis and carryout the actions.

Therefore, the Hope Creek CPPU has no effect on the process controls for shutdown risk management and a negligible effect on the overall shutdown risk profile.

10.5.7 PRA Quality 10.5.7.1 Overview of PRA The Hope Creek PRA has been in the process of update for several years. The latest revision is model 2005B used for the evaluation of risk associated with CPPU implementation.

PSEG procedures provide the details describing the use of the PRA at Hope Creek to support the Maintenance Rule. The PRA assists in establishing performance criteria, balancing unavailability and reliability for risk significant SSCs and setting goals, and provides input to the Expert Panel for the risk significance determination process when revisions to the PRA take place.

Because the PRA is actively used at Hope Creek, a formal process is in place to evaluate and resolve PRA model-related issues as they are identified.

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NEDO-33076 10.5.7.2 PRA Quality Summary The quality of the Hope Creek PRA models used in performing the risk assessment for the Hope Creek CPPU is established and maintained by:

• Sufficient scope and level of detail in PRA

• Active maintenance of the PRA models and inputs

• Comprehensive Critical Reviews Scope and Level of Detail The Hope Creek PRA is of sufficient quality and scope for this application. The Hope Creek PRA modeling is highly detailed; including a wide variety of initiating events (e.g., transients, internal floods, LOCAs inside and outside containment, support system failure initiators),

thorough modeling of systems, extensive level of detail, operator actions (including dependency treatment), and extensive common cause evaluation.

Maintenance of Model. Inputs, Documentation The Hope Creek PRA model and documentation were updated to reflect the current plant configuration and to reflect the accumulation of additional plant operating history and component failure data. The current Hope Creek PRA model of record used in the CPPU analysis is Revision 2005B. This has been developed in 2005 based on the analysis and data used to update the model in 2003.

Significant changes to the following PRA elements have been performed to respond to the PRA Peer Review and the expectations of the ASME PRA Standard. (Reference 31) These changes include the following:

• Completely new Human Reliability Analysis (HRA)

• Revised accident sequence models (Event Trees)

• New MAAP calculations to support the success criteria and accident sequence timing for the CPPU condition

• Updated data (initiating events, component failure data, and unavailability data)

• Modified system models

• Updated common cause failure probabilities incorporating the latest NRC data compiled by INEEL

• The addition of internal flood accident sequences Critical Reviews A BWROG pilot PRA Peer Review was performed for Hope Creek in 1996. In response to this review, the PRA model was updated in 1999. Significant comments (A and B) were addressed in the 1999 revision.

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NEDO-33076 Subsequently, PSEG participated in a full PRA Peer Review Certification of the Hope Creek 1999 PRA model administered under the auspices of the BWROG Peer Certification Committee, using the NEI 00-02 predecessor document. The purpose of the PRA peer review process is to establish a method of assessing the technical quality of the PRA for the spectrum of its potential applications.

The Peer Review evaluation process used a tiered approach with standardized checklists allowing for a detailed review of the elements and the sub-elements of the Hope Creek PRA to identify strengths and areas that needed improvement. The review system used allowed the Peer Review team to focus on technical issues and to issue their assessment results in the form of a "grade" of 1 through 4 on a PRA sub-element level. To reasonably span the spectrum of potential PRA applications, the four grades as defined by NEI 00-02 were employed. All of the significance level A and B Facts and Observations from the 1999 Peer Review of the 1999 PRA model have been addressed by PSEG in the current Hope Creek PRA Revision 2005B model. A summary of the facts and observations is provided separately.

In addition, PSEG performed a self-assessment relative to the ASME PRA Standard. (Reference

31) The self-assessment was carried out using the Supplementary Guidance on self-assessment provided by NEI. (Reference 33) The process also made use of the independent peer review of the 1999 Hope Creek PRA model.

The self-assessment demonstrated the Hope Creek PRA capability relative to ASME PRA Standard Capability Category II.

The flowchart in Figure 10-2 shows the context for the Hope Creek self-assessment with respect to the Hope Creek PRA and other related activities.

The Hope Creek PRA self-assessment included the steps identified in the NEI supplementary guidance (Reference 33). Based on the ASME PRA Standard, the following steps were performed relative to the development of the Hope Creek PRA 2005B model as part of the self-assessment:

a. Beginning with Initiating Events Analysis, the supporting requirements of the ASME Standard (corresponding to Capability Category II) that were indicated as "partially addressed" or "not addressed" by the peer review process were identified.

b. For each supporting requirement so identified, a determination was made as to whether the supporting requirement was addressed in the Hope Creek PRA. The basis for the determination that the supporting requirement is addressed, including the update of the PRA in 2003 to specifically address the supporting requirements, is documented.

c. For supporting requirements identified as "addressed" by the peer review process, the peer review report was reviewed to determine if these sub-elements were assigned a grade of 3 or higher. If a grade less than 3, or a conditional grade 3 was provided, or if significant facts and observations needed to be reconciled, then a determination was made of which capability level in the ASME PRA Standard was met, and this determination was 10-31

NEDO-33076 documented. This included consideration of the resolution of facts and observations that are incorporated in the development of the Hope Creek PRA 2005B model. The resolution of facts and observations from the PRA Peer Review included all of the significance level A and B Facts and Observations.

d. Following assessment of the supporting requirements, the information produced was reviewed and a determination was made as to whether the high level objective for the initiating events analysis section was met.

The above process was repeated for each PRA element as defined in the ASME PRA Standard.

10.5.7.3 PRA Quality Conclusion The Hope Creek Internal Events PRA has been peer reviewed using the industry guidance for the PRA Peer Reviews. A PRA self-assessment to compare the Hope Creek PRA to the ASME PRA Standard per the NEI supplementary guidance has been performed. These activities demonstrate that the Hope Creek PRA is suitable to support PRA applications that require ASME PRA Capability Category II.

10.6 OPERATOR TRAINING AND HUMAN FACTORS Some additional training is required to enable plant operation at the CPPU RTP level. The topics addressed in this section are:

l Topic - - lCLTR Disposition Hope Creek Result Operator training and human factors ]

The additional training required to operate the plant in CPPU conditions is minimal.

For CPPU conditions, operator responses to transient, accident and special events are not affected. Most abnormal events result in automatic plant shutdown (scram). Some abnormal events result in automatic RCPB pressure relief, ADS actuation and/or automatic ECCS actuation (for low water level events). All events result in safety-related systems, structures, and components (SSCs) remaining within their design allowables. CPPU does not change any of the automatic safety functions. After the applicable automatic responses have initiated, the subsequent operator actions (e.g., maintaining safe shutdown, core cooling, and containment cooling) for plant safety do not change for CPPU.

The analog and digital inputs for the Plant Computer Systems, including the Safety Parameter Display System (SPDS), will be reviewed to determine the effects from CPPU. This includes required changes to monitored points, calculations, alert and trip setpoints. Various changes in EOP curves and limits, if required, will also require an update of the SPDS. Any changes required to the Plant Computer Systems will be completed prior to operation at CPPU conditions.

Following a review of the CPPU modifications and identified key procedure changes, recommendations for operator training and simulator changes and a final determination of the 10-32

NEDO-33076 operator training needs will be made, consistent with the Hope Creek training program for selection of modifications for operator training. Any modifications required for CPPU will be evaluated for their effect on the Plant Computer Systems and any required changes (including any new monitoring points) will be addressed as a part of the modification. Any changes made will be discussed as a part of the operator-training program for CPPU.

Training required to operate the plant following CPPU will be conducted prior to operation of the unit at the CPPU conditions. Data obtained during CPPU testing will be incorporated into the training as needed. The classroom training will cover various aspects of CPPU including changes to parameters; setpoints, scales, procedures, systems and CPPU test procedures. The classroom training will be combined with simulator training. The simulator training will include, as a minimum, a demonstration of transients that show the greatest change in plant response at CPPU RTP compared to CLTP.

The simulator changes will include hardware changes for new or modified control room instrumentation and controls, software updates for modeling changes due to CPPU (e.g., HP turbine modifications), setpoint changes, and re-tuning of the core physics model for cycle specific data. The Plant Computer Systems included in the Hope Creek Simulator will also be updated for CPPU modifications prior to CPPU implementation.

Simulator performance will be verified by conducting a validation test and evaluating the results against predicted performance data based on design and engineering analysis data as required in ANSI/ANS 3.5 - 1993. This test will include a demonstration that the simulator represents the plant to the scope required by Section 3 of ANSI/ANS 3.5 - 1993, "General Requirements."

Section 4 of ANSI/ANS 3.5 - 1993, "Simulator Testing and Validation," provides the criteria for validating that these requirements are met. Operating data will be collected during CPPU implementation and testing. This data will be compared to simulator data as required by ANSI/ANS 3.5 - 1993, Section 4.4.1.

10.7 PLANT LIFE The plant life evaluation identifies degradation mechanisms influenced by increases in fluence and flow. The topics addressed in this evaluation are:

Topic - -CLTR Disposition: Hope Creek Result Irradiated Assisted Stress Corrosion Cracking Flow Accelerated Corrosion Hope Creek has a procedurally controlled program for the augmented nondestructive examination (NDE) of selected RPV internal components in order to ensure their continued structural integrity. The inspection techniques utilized are primarily for the detection and characterization of service-induced, surface-connected planar discontinuities, such as intergranular stress corrosion cracking (IGSCC) and irradiation-assisted stress corrosion cracking (IASCC), in welds and in the adjacent base material. Hope Creek belongs to the BWR Vessel and Internals Project (BWRVIP) organization and implementation of the procedurally controlled program is consistent with the BWRVIP issued documents. The inspection strategies 10-33

NEDO-33076 recommended by the BWRVIP consider the effects of fluence on applicable components and are based on component configuration and field experience.

Components selected for inspection include those that are identified as susceptible to in-service degradation and augmented examination is conducted for verification of structural integrity.

These components have been identified through the review of NRC Inspection and Enforcement Bulletins (IEBs), BWRVIP documents, and recommendations provided by General Electric Service Information Letters (GE SILs). The inspection program provides performance frequency for NDE. Components inspected include the following:

• Core spray piping

• Core spray spargers

• Core shroud and core shroud support

• Jet pumps and associated components

• Top guide

• Lower plenum

• Vessel ID attachment welds

• FW sparger Continued implementation of the current procedure program assures the prompt identification of any degradation of reactor vessel internal components experienced during CPPU operating conditions. Reactor vessel water chemistry conditions are also maintained consistent with the EPRI and established industry guidelines.

There are no known un-repaired flaws remaining in the reactor recirculation piping. Hydrogen is currently injected into the primary system for IGSCC mitigation in the recirculation piping.

Changes to the mitigation process are not required for CPPU.

The service life of most equipment is not affected by CPPU. ((

Two components are predicted to exceed the BWRVIP-26 threshold fluence level of 5 x 10A20 n/cm2. Peak fluence for the Top Guide is predicted to reach 1.76 x 10A22 n/cm2. Peak fluence for the shroud is predicted to reach 2.38 x 10^21 n/cm2. The current inspection strategy for the reactor internal components is expected to be adequate to manage any potential effects of CPPU.

The Hope Creek procedurally controlled FAC program uses selective component inspections to provide a measure of confidence in the condition of systems susceptible to FAC. These selective inspections are the basis for qualifying un-inspected components for further service. This approach is based upon program guidelines developed by the Electric Power Research Institute (EPRI), and the American Society of Mechanical Engineers (ASME). The criteria for selecting components for inspection after the CPPU will be the same as used under CLTP. In addition to this aggressive monitoring program, selected piping replacements have been performed to maintain suitable design margins. Where possible, FAC resistant replacement materials are used to mitigate future occurrences of FAC.

10-34

NEDO-33076 A CHECWORKSTMt FAC model (in accordance with the CHECWORKST mFAC users guide and EPRI modeling guidelines) has been developed for Hope Creek to predict the FAC wear rate (single and two-phase fluids) and the remaining service life for each piping component. As a minimum, the controlled CHECWORKS TM FAC model is updated after each refueling outage.

The FAC models are also used to identify FAC examination locations for the outage examination list and uses empirical data input to the model.

Process variables that influence FAC at Hope Creek include:

• Moisture content

• Water chemistry

• Temperature

• Oxygen

• Flow path geometry and velocity

• Material composition Hope Creek has predicted CPPU system operating conditions that will be used as inputs to the CHECWORKSTm FAC model. Implementation of CPPU will affect moisture content, temperature, oxygen, and flow velocity. For some systems the moisture content and oxygen will increase but remain within the CHECWORKSTM FAC model parameter bounds. Selected portions of some system piping are predicted to increase a maximum of 13TF. Depending on operating power levels, flow velocity increases of up to -20% will occur. The Hope Creek CHECWORKSTMt FAC model is capable of accepting these CPPU related parameter changes.

Based on experience at CLTP operating conditions and previous FAC modeling results, it is anticipated that the CPPU operating conditions may result in the need for additional FAC monitoring points. The CHECWORKSTM FAC modeling techniques allow for the identification of additional monitoring points required for CPPU. The CHECWORKSTm FAC program targets FAC susceptible piping and components and includes the installation of FAC resistant material.

Table 10-12 compares key parameter values (CLTP and CPPU) affecting FAC.

The increased main steam and FW flow rates at CPPU conditions do not significantly affect the potential for FAC. Therefore, the Hope Creek program for FAC is adequate to manage any potential effects of the CPPU on NSSS, turbine-generator (T-G), and BOP components. The reactor internals inspection and FAC programs do not significantly change for CPPU. In addition, the Maintenance Rule provides oversight for the other mechanical and electrical components, important to plant safety, to guard against age-related degradation.

10.8 NRC AND INDUSTRY COMMUNICATIONS NRC and industry communications could affect the plant design and safety analyses. However, as stated in Section 6.8, all of the systems significantly affected by CPPU are addressed in this 10-35

NEDO-33076 report. In addition, all of the plant safety analyses affected by CPPU are addressed in this report.

As a result, evaluations of plant design and safety analyses affected by the communications in place are inherently included in the plant specific CPPU assessments. Therefore, it is not necessary to review prior dispositions of NRC and industry communications and no additional information is required in this area.

10.9 EMERGENCY AND ABNORMAL OPERATING PROCEDURES Emergency and abnormal operating procedures can be affected by CPPU. Some of the Emergency Operating Procedures (EOPs) variables and limit curves depend upon the value of rated reactor power. Some Abnormal Operating Procedures (AOPs) may be affected by plant modifications to support the higher power level. The topics addressed in this section are:

Topic CLTR Disposition Hope Creek Result Emergency operating procedures Abnormal operating procedures EOPs include variables and limit curves, defining conditions where operator actions are indicated. Some of these variables and limit curves depend upon the RTP value. Changing some of the variables and limit curves requires modifying the values in the EOPs and updating the Hope Creek support documentation. EOP curves and limits may also be included in the SPDS and will be updated accordingly.

The charts and tables used by the operators to perform the EOP's are reviewed for any required changes prior to each core reload. The EOPs will be reviewed for any changes required to implementing CPPU. The operators will receive training on these procedures as described in Section 10.6.

AOPs include event based operator actions. Some of these operator actions may be influenced by plant modifications required to support the increase in rated reactor power. No significant AOP revisions are foreseen and the effect on operator actions is minimal. Most of the changes are directly tied to changes in the listing of MWt or MWe in AOPs. Others are tied to specific parameters such as the power-flow map, rated steam flow or pump speed. Some of the setpoints used in the AOPs will change due to CPPU. All AOPs will be reviewed for CPPU conditions and necessary revisions will be completed prior to CPPU implementation. The operators will receive training on these procedures as described in Section 10.6.

10-36

NEDO-33076 Table 10-1 Hope Creek High Energy Line Break

%'Increase (Chaige) Due to CPPU ,

=' - Break Location --

Break'Location '-- - Mass Release - Pressure

<--; emperature, Main Steam Line Break in Steam Tunnel No change No change No change Feedwater Line Breaks in Steam Tunnel (1) No change No change RCIC Steam Line Breaks in Reactor Building No change No change No change HPCI Steam Line Breaks in Reactor Building No change No change No change RWCU Breaks in Reactor Building 35%(2) No change No change Notes:

(1) - FW line blowdown mass release increase was not explicitly calculated. MS line break at current licensed conditions bounds FW line break at CPPU conditions. Maximum FW line energy release occurs at ICF at CPPU RTP and minimum FW temperature conditions.

(2) RWCU Pump Suction Break at minimum pump speed, CPPU conditions, and minimum FW temperature.

10-37

NEDO-33076 Table 10-2 Hope Creek Equipment Qualification for CPPU Parameter CPPU Effect

• Accident CPPU peak temperature increase assumption inside containment for EQ No Change (l)

• Accident CPPU peak pressure increase assumption inside containment for EQ No Change (l)

• Normal CPPU plant operation plus CPPU accident radiation increase assumption inside containment for EQ (Excludes Zone 2) 16% (2

• Solenoid Valves located within containment potentially affected by higher radiation levels Note 3

• Normal CPPU temperature increase assumption outside containment for EQ No Change (4)

• HELB CPPU flooding level increase assumption outside containment for EQ (Rooms 4403, 4405,4502, 4503 and 4506) <1.74 ft.

• Temperature elements, conax seals, process switches and cables located outside containment potentially affected by higher HELB flooding levels Note 5

• MELB CPPU flooding level increase assumption outside containment for EQ No Change

• Accident CPPU temperature increase assumption outside containment for EQ No Change

• Accident CPPU pressure increase assumption outside containment for EQ No Change

• Normal CPPU radiation level increase assumption outside containment for EQ (Select rooms in Reactor, Turbine and Auxiliary Buildings) 16%

• Accident CPPU radiation increase assumption outside containment for EQ No Change

• Pressure switches located outside containment potentially affected by higher radiation level Note 6 Notes:

(1) The peak accident temperature and pressure increase due to CPPU but remain bounded by the worst-case accident profile assumed for EQ purposes.

(2) The increase in the total integrated dose (normal plus accident) for CPPU conditions is based on a comparison of the CPPU total integrated worst-case dose of 8.4E7 rad gamma (i.e., normal 40-year CPPU plant operation gamma radiation dose in Zone I of5.2E7 rad plus the 2400-hour accident CPPU gamma radiation dose in Zone I of 3.2E7 rad) with the CLTP worst-case normal plus accident dose (7.24E7 rad gamma).

(3) The increase in the total integrated dose (normal plus accident) for CPPU conditions may result in a reduction of the radiation life for some Target Rock solenoid valves located inside primary containment.

(4) The only normal CPPU temperature increase above EQ design temperatures occurs in Room 1512; however, there is no EQ equipment installed in the affected room.

(5) No safety related equipment is affected by the HELB increased flooding levels.

(6) The increase in the total integrated dose (normal plus accident) for CPPU conditions may result in a reduction of the radiation life for Barksdale pressure switches located outside containment.

10-38

NEDO-33076 Table 10-3 Summary Comparison of Baseline and Updated CDF for Hope Creek

-Hope Creek - -CLTP( I -- CPPU( 2 )

Total CDF (yr7 1 )(3 ) 9.42E-06 L.O1E-05 lLERF (yr')(3)l 2.37E-07 2.98E-07 A CDF = 6.8E-7/yr(4)

A LERF = 6.1 E-8/yr(4 )

(1) The CLTP PRA model is developed by removing the CPPU specific model changes from the CPPU PRA model.

(2) The CPPU estimates are developed from the Revision 2005B model, which includes the CPPU power level and plant changes, and the possible changes postulated to affect the turbine trip initiating event frequency.

(3) Includes Internal Events (including internal flooding). The internal flooding contributors are relatively small contributors to the risk metrics.

(4) These risk metric changes place the change in risk for Hope Creek in Region III (very small risk change) of the RG 1.174 acceptance guideline 10-39

NEDO-33076 Table 104 Summary of the Basis for Models Used in CPPU Delta Risk Calculations Topic CLTP.Model CPPU Modell-

- -Highlights of

-_ _ - . :;Model(

- . ) .~:

2

. . X . :l .

Initiating Events Same as 2005B model except Hope Creek Data Bayesian that turbine trip Initiating updated(2)l Event Frequency (IEF) is not increased. Increased turbine trip Initiating Event Frequency (lEF) based on postulated changes in margins Success Criteria & Thermal Modified to be representative CPPU specific analysis with Hydraulic Analysis of CLTP. MAAP 4.0.4 RPV depressurization success RPV depressurization success is 1 SRV is 2 SRVs RPV overpressure protection RPV overpressure protection for ATWS response is 11 of for ATWS response is 12 of 14 SRVs 14 SRVs

[The CPPU licensing submittal does not credit containment overpressure to provide adequate NPSH to the ECCS pumps.]

Component Failure Data & Same as 2005B model except Hope Creek Data Bayesian Unavailability Data for reduced probability of updated SORV HRA Modified timing of HEPs to Hope Creek specific analysis be characteristic of CLTP for CPPU condition and configuration, but used same sequence timing HRA method as 2005B model Systems and Plant Modified to demonstrate Hope Creek specific CPPU Configuration differences with CPPU condition condition:

No spurious RR No TT on Loss of single FW l_ pump 1040

NEDO-33076 Table 104 Summary of the Basis for Models Used in CPPU Delta Risk Calculations l__

opic Topc-

-;__-_-__-______;-__-_I CLTP Mode(l)--

I J

CPPU Model Highlights of 2005B Model( 2 -

Structural Same as 2005B model Model 2005B:

No significant structural changes since IPE and 1999 model Common Cause Same as 2005B model Updated Common Cause Analysis with latest NRC data Accident Sequence Structure Same as 2005B model 2005B model Quantification Process Same as 2005B model 2005B model Level 2 Structure and phenomena are 2005B model the same as 2005B model Notes:

(1) The model used in the CLTP assessment is the 2005B model with a truncation at 5E-I 1/yr and appropriate changes to reflect the initiating frequencies, success criteria, and component failure rates associated with current plant operation.

(2) The Hope Creek Base Model (2005B) is representative of the CPPU condition. It has incorporated Hope Creek operating history (component data and initiating event data). Changes in component data as a result of CPPU would be speculative. Therefore no attempt to include hypothetical increases is made. Future PRA updates will capture any observed changes. It is noted that power uprates implemented at other BWRs have not caused any noteworthy increases in the scram failure rate, however, the turbine trip frequency is increased for the CPPU model. The 2005B model is quantified at a truncation of 5E-1 1/yr and the reported CDF is 1.0lE-05/yr.

10-41

NEDO-33076 Table 10-5 Model Changes to Reflect Plant Physical Changes or Effects of CPPU Implementation

-Risk Metric Technical Item Approach

.,Ppq 'ACDF . .

(per .. yr)(l)

.,,,.,...p ... .

Reduction in the margin available: For CPPU model increase the turbine 2.5E-7 (1) Turbine First Stage Pressure trip initiating event frequency by (TFSP) scram bypass 21%.

permissive (margin decreased)

(2) Time available for operator action given FW controller failure may decrease (3) Reduced margin to condenser backpressure setpoint (4) Spurious recirculation runback or failure to actuate HRA timing for implementation Implement modified HEPs( 3 ) for the 3.1E-7 reduced in CppU.(2 ) CLTP model to represent the increased time available under CLTP conditions.

SORV probability increases in Implement modified SORV failure 1.4E-8 CPPU model due to increased probability in the CLTP model to challenges represent the decrease in challenges for the CLTP configuration.

Success criteria for depressurization Implement modified depressurization 5.0E-9 modified common cause failure probability in the CLTP model to reflect previous success criteria.

Success criteria for ATWS Implement modified overpressure overpressure protection common cause failure probability for (elow ATWS in the CLTP model to reflect truncation) the previous success criteria.

10-42

NEDO-33076 Notes to Table 10-5:

(1) The ACDF is measured between the PRA for the current licensed thermal power (CLTP) condition and the proposed constant pressure power uprate (CPPU) configuration.

(2) The Hope Creek plant has automatic SLCS initiation as a design feature. The change in the HEP for SLC manual backup to this automatic initiation has a negligible effect on the risk metrics for the implementation of CPPU. This design feature is different from previous BWR CPPU submittals.

(3) Modified HEPs:

Designator l Description NR-UIX-DEP-SRV FAILURE TO DEPRESSURIZE WITH SRV W/O HIGH PRES. INJ.

RX-FW-ADS DEPENDENT HEP - FAILURE OF MANUAL FW CONTROL AND ADS SWS-XHE-FO-START FAILURE TO START SW PUMPS WHEN REQUIRED SAC-XHE-FO-HEAT SACS HEAT LOAD MANIPULATION SAC-XHE-FO-HEA5A DEPENDENT HEP FAILURE OF SACS HEAT LOAD MANIPULATIONAND OPEN 2355A LOCALLY SAC-XHE-FO-HEA5B DEPENDENT HEP - FAILURE OF SACS HEAT LOAD MANIPULATION AND OPEN SACS SW HX VALVE 2355B SAC-XHE-FO-HEAIA DEPENDENT HEP FAILURE OF SACS HEAT LOAD MANIPULATION AND OPEN 2371A LOCALLY SAC-XHE-FO-HEAIB DEPENDENT HEP FAILURE OF SACS HEAT LOAD MANIPULATION AND OPEN 237 1B LOCALLY 10-43

NEDO-33076 Table 10-6 Sensitivity Calculation to Reflect Postulated Effects on Risk Metrics of CPPU Implementation Risk Metric;

-Technical 1term Approach 'ACDF (per yr)

Flow assisted corrosion may increase main Double the Large LOCA 2.3E-7 steam or feedwater pipe failure probability. frequency in the CPPU model for a bounding estimate.

Flow induced vibration could result in Double the turbine trip frequency 1.6E-6 increased scram frequency due to effects on in the CPPU model for a bounding (a) reactor internals; or, (b) small bore estimate.

attached pipe.

There may be an initial "break-in" period in (Included in above sensitivities) which there is an increase in component failures or pipe challenges. This is expected to potentially manifest itself in an increase in scram frequency.

Inadvertently open SRV (IORV) initiating Double the inadvertently open 1.7E-7 event frequency SRV initiating event frequency in the CPPU model for a bounding estimate 10-44

NEDO-33076 Table 10-7 Summary of Initiating Event Frequency Effects of CPPU Estimated 'Magnitude of Potential -Initiating Event Frequency Item- Change ';

CPantEffect Anticipated Transients Turbine First Stage Pressure .r7 Frequency IE-2/yr 1 (TFSP) scram bypass increase (Use 10% increase) permissive (margin decreased)

Time available for operator rr Frequency IE-3/yr 2 action given FW controller increase (Negligible) failure may decrease 3 Reduced margin to condenser increase (Use 10E%2/ncrease) backpressure setpoint 4 Spurious runback or failure to rr Frequency Negligible actuate increase (Use 1% increase)

LOCAs 5 LOCA Frequency (2)

FAC and FIV increase (_)

6 TT Frequency (2)

Flow induced vibration increase Special Initiators 7 None None None Internal Flood 8 None None None Loss of Offsite Power 9 None None None ATWS 10 None None None (1) The changes implemented in the CPPU model relative to the 2003A baseline model are identified in parenthesis.

(2) No effect on scram or LOCA frequency is expected because of the robust Hope Creek pipe inspection and analysis programs plus adequate safety margin built into the plant.

1045

NEDO-33076 Table 10-8 COMPARISON OF INITIATOR CONTRIBUTORS BETWVEEN THE CLTP AND CPPU PRA MODELS

CLTP  ::_-J -_ - CPPU Frequency I- CDF Frequency . - CDFI Initiator Categor-y Identifier (per r F-V (per yr) - (per yr) F-V (per yr)

Transient Initiator FPS RUPTURE IN CONTROL DIESEL BUILDING %FLFPS-CD 8.20E-05 0 0 8.20E-05 6.73E-06 6.80E-1 I FPS RUPTURE OUTSIDE CONTROL ROOM %FLFPS-CR 1.1013-05 7.2413-04 6.82E-09 1.10E-05 6.7613-04 6.83E-09 SACS A RUPTURE %FLSACS-A 2.70E-04 1.0113-03 9.51 E-09 2.7013-04 9.4813-04 9.57E-09 SACS B RUPTURE %FLSACS-B 2.70E-04 1.1313-03 1.0613-08 2.70E-04 1.06E-03 1.07E-08 SW RUPTURE IN SACS A ROOM %FLSW-SACS-A 4.10E-06 0 0 4.10E-06 0 0.00E+00 SW RUPTURE IN SACS B ROOM %FLSW-SACS-B 4.10E-06 0 0 4.10E-06 0 0.00E+00 SW LOOP A RUPTURE IN RACS AREA %FLSWA-RACS OOE-05 1.2613-04 1.1913-09 .. l.OOE-05 1.181E-04 1.19E-09 SW LOOP B RUPTURE IN RACS AREA %FLSWB-RACS 1.00E-05 1.26E-04 1.1913-09 1.00E-05 1.18E-04 1.1913-09 TURBINE BUILDING FLOOD %FLTB-CW 1.3013-04 0 0 1.30E-04 0 0.0013+00 TORUS RUPTURE IN TORUS ROOM %FLTORUS 2.8013-06 4.73E-03 4.4613-08 2.80E-06 4.411E-03 4.4513-08 TORUS SUCTION LINE RUPTURE IN ECCS ROOM %FLTORUSRB 3.5013-05 5.2013-05 4.9013-10 3.5013-05 4.8513-05 4.90E-10 LOSS OF AC BUS A INITIATING EVENT %IE-ACA 1.53E-04 9.85E-06 9.28E-1 1 1.5313-04 3.2613-05 3.2913-10 LOSS OF AC BUS B INITIATING EVENT %IE-ACB 1.53E-04 0 0 1.53E-04 0 0.0013+00 LOSS OF AC BUS C INITIATING EVENT %IE-ACC 1.5313-04 0 0 1.5313-04 1.52E-05 1.5413-10 LOSS OF AC BUS D INITIATING EVENT %IE-ACD 1.5313-04 1.4713-03 1.3813-08 1.5313-04 1.4013-03 1.411E-08 LOSS OF DCA & DCB %IE-DCAB 9.0013-07 2.6213-04 2.4713-09 9.00E-07 2.5813-04 2.611E-09 LOSS OF HVAC %IE-HVAC 2.8013-03 0 0 2.8013-03 0 0.00E+00 LOSS OF INSTRUMENT AIR INITIATOR %IE-IAS 1.2013-02 3.2313-02 3.0413-07 1.2013-02 3.11 E-02 3.14E-07 MANUAL SHUTDOWN INITIATING EVENT %IE-MS 9.4413-01 9.9913-02 9.41 E-07 9.4413-01 9.84E-02 9.9413-07 10-46

NEDO-33076 Table 10-8 COMPARISON OF INITIATOR CONTRIBUTORS BETWEEN THE CLTP AND CPPU PRA MODELS CLTP ____J___ CPPU-Frequency I CDF -Frequency -: CDF Initiator Category,, -liitfier (per yr) j F-V r-V , (per yr)'

LOSS OF RACS %IE-RACS 1.56E-05 0 0 1.56E-05 0 0.00E+00 LOOP WITH AC RECOVERED %IE-RLOOP 1.74E-02 7.78E-03 7.33E-08 1.74E-02 7.94E-03 8.02E-08 LOSS OF SACS INITIATING EVENT %IE-SACS 1.12E-04 3.93E-03 3.70E-08 1.12E-04 3.68E-03 3.72E-08 LOSS OF SERVICE WATER INITIATING EVENT %lE-SWS 2.50E-04 3.75E-02 3.53E-07 2.50E-04 3.53E-02 3.57E-07 LOSS OF CONDENSER VACUUM %IE-TC 9.35E-02 6.02E-02 5.67E-07 9.35E-02 6.03E-02 6.09E-07 LOSS OF OFFSITE POWER INITIATING EVENT %IE-TE 3.04E-02 5.22E-01 4.92E-06 3.04E-02 4.98E-01 5.03E-06 LOSS OF FEEDWATER %IE-TF 5.05E-02 2.68E-02 2.52E-07 5.05E-02 2.71 E-02 2.74E-07 INADVERTENTLY OPEN SRV INITIATING EVENT %IE-TI 2.24E-02 1.63E-02 1.54E-07 2.24E-02 1.60E-02 1.62E-07 MSIV CLOSURE %IE-TM 2.69E-02 1.31 E-02 1.23E-07 2.69E-02 1.32E-02 1.33E-07 TURBINE TRIP WITH BYPASS %IE-TT 1.03E+00 1.25E-01 1.18E-06 1.25E+00 1.57E-01 1.59E-06 LOCA Initiator EXCESSIVE LOCA EVENT %IE-R 11.OOE-08 1.06E-04 9.99E-10 I.OOE-08 9.91 E-05 I1.OOE-09 LARGE LOCA INITIATOR %IE-A 3.00E-05 2.42E-02 2.28E-07 3.00E-05 2.26E-02 2.28E-07 MEDIUM LOCA (WATER) INITIATOR %IE-SI 3.00E-05 1.22E-02 I.15E-07 3.00E-05 1.14E-02 I.15E-07 SMALL BREAK LOCA %IE-S2 4.00E-04 5.80E-03 5.46E-08 4.00E-04 5.62E-03 5.68E-08 ISLOCA INITIATOR FOR ECCS DISCHARGE PATHS %IE-ISLOCAD 1.63E-05 9.19E-04 8.66E-09 1.63E-05 8.60E-04 8.69E-09 ISLOCA INITIATOR FOR SDC SUCTION PATH %IE-ISLOCAS 5.01 E-07 2.71 E-04 2.55E-09 5.01E-07 2.53E-04 2.56E-09 BREAK OUTSIDE OF CONTAINMENT INITIATING EVENT %IE-BOC 6.OOE-08 1.66E-03 1.56E-08 6.00E-08 I.55E-03 1.57E-08 10-47

NEDO-33076 Table 10-9 Comparison of SORV Probabilities (CPPU and CLTP)

CLTP SORV. .- CPPU SORV Accident Scenario .:-. w Probability Probability Turbine Trip 1.5E-2 1.8E-2 Isolation Event/SLOCA 4.8E-2 5.4E-2 ATWS 5.288E-2 5.4E-2 10-48

NEDO-33076 Table 10-10 Disposition of Key Actions for Potential HEP Re-Calculation Action Time A ailable

_ _ __ _ _ _ _ _ _H EP ~- _

l Action Basis of A Re-Calculation Dscrfiptioni- -rniportance LTP cpecessar Comment Basic Event ID -:: - -: .- . N -  :: - -- Cm. n NR-XTIE-EDG Failure to F-V = 0.399 4 hrs. 4 hrs. No This operator action is a place holder in the PRA, Crosstie modeled in the Hope Creek PRA with an HEP of Diesel 1.0. This action is not proceduralized and the crew Generator to indicated they would not perform it. As such, the opposite bus CPPU has no effect on the current modeling of this operator action.

ACP-XHE-RE- Failure to F-V 0.228 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> No This is an offsite power recovery term. The time SW04H Recover frame is based on nominal modeling time phases for Severe LOOP scenarios determined principally by battery Weather depletion time. The recovery failure probability is LOOP (4 based on statistical analysis of the duration of Hours) industry LOOP events and not directly on HEP calculations. The CPPU does not affect the appropriateness of this time frame nor the recovery failure probability.

NR-XTIE- Failure to F-V = 0.177 3 hrs. 3 hrs. No This action is to cross tie power to a battery charger CIIARG Crosstie before the battery discharges. The CPPU does not Energized Bus affect the battery discharge time.

to Battery Charger Breaker 10-49

NEDO-33076 Table 10-10 Disposition of Key Actions for Potential HEP Re-Calculation Action Time Available  :

l_ _ _ _ -_ _ _ - ,H EP Action  : Basis of- - Re- Calculation Description Impoilance CLTP cru ^

ecessiy Comment-

-Basic Evejit ID ACP-XHE-RE- Failure to F-V =0.154 4 hrs. 4 hrs. No This is an offsite power recovery term. The time PC04H Recover Plant frame is based on nominal modeling time phases for Centered and LOOP scenarios determined principally by battery Grid Related depletion time. The recovery failure probability is LOOP (4 based on statistical analysis of the duration of Hours) industry LOOP events and not directly on HEP calculations. The CPPU does not affect the appropriateness of this time frame nor the recovery failure probability.

SAC-XHE-FO- Operator F-V = 0.116 46 minutes 40 minutes No This action corresponds to the local manipulation of HEA5B Action SWS- (3) the SSW discharge valve 2355B on the SACS heat XHE-FO- exchanger. Based on lack of accessibility, no clear 2355B, Failure procedural guidance, and crew interviews this IIEP to Open has been assigned an HEP=1.0 for both CLTP and SACS-SW CPPU models.

Heat Exchanger Valve 2355B Locally 10-50

NEDO-33076 Table 10-10 Disposition of Key Actions for Potential HEP Re-Calculation Action Time Available 1_____ _ --- HEP -

Action Bask of Re- Calculation Description Importance  : CLTP - CP Necessary- - Comment Basic Event ID NR-VENT-5-03 Failure to F-V = 0.115 -20 hrs. 20 irs. No This operator action represents failure to align the Initiate (') containment vent. The time frame is 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> based Containment on the time to reach the containment vent pressure.

Venting The CPPU does not affect the appropriateness of this extremely long time frame nor the failure probability determined based on this long time frame.

ADS-XHE-OK- Automatic F-V = 0.075 -14 minutes 12 minutes No This is not a human error. This action is to INHIB ADS Inhibited successfully inhibit automatic ADS actuation. An (Non-ATWS)- override success probability of 1.0 is used. CPPU

-Success Of implementation is not judged to affect his The Action probability. Any decrease in the success probability associated with CPPU implementation would decrease the risk of CPPU implementation.

ACP-XIIE-RE- Failure to F-V = 0.066 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> No This is an offsitc power recovery term. The time SW20H Recover frame is based on nominal modeling time phases for Severe LOOP scenarios determined principally by battery Weather depletion time. The recovery failure probability is LOOP (20 based on statistical analysis of the duration of Hours) industry LOOP events and not directly on HEP calculations. The CPPU does not affect the appropriateness of this time frame nor the recovery failure probability.

10-51

NEDO-33076 Table 10-10 Disposition of Key Actions for Potential JIEP Re-Calculation Action Time Avilable-Action Basis of Re- Calculation .

Description Imporiance CLnomm et,

,Bdsic E-venti IDce CTcru Ncssj CAC-XHE-FO- Failure to F-V = 0.064 80 minutes 69 minutes Yes The operator action to vent the containment so that NPSH prevent steam (3) NPSH is not lost for pumps using the suppression binding of pool. No change in the HEP using the Cause Based ECCS pump Decision Tree Method, EPRI TR 100259.(2)

During Cont (Rcecrence 34A]

Vent NR-SPL-LVLL-4 Failure to F-V = 0.064 > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> No This operator action represents failure to align the Align Core (I) (1) Core Spray to the CST for injection post Spray to the containment failure. The time frame is > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> CST for Late based on the time to reach the ultimate containment Injection (Post failure pressure. The CPPU does not affect the Containment appropriateness of this time frame nor the failure Challenge) probability determined based on this long time frame.

SAC-XHE-FO- Operator F-V = 0.056 46 minutes 40 minutes Yes This is a dependent HEP combination. The HEA5A Action SWS- (3) manipulation of SACS heat loads is evaluated in the XHE-FO- PRA for the worst case conditions of high river 2355A, water temperature and high SACS temperatures.

Failure to For these conditions, the time frames for crew action Open SACS- result in a change in the calculated HEP. This action SW Heat is required for certain SACS configurations that may Exchanger occur following a LOOP event. The local opening Valve 2355A of the 2355A valve is set to 1.0.

Locally 10-52

NEDO-33076 Table 10-10 Disposition of Key Actions for Potential HEP Re-Calculation Action Time Available

_ __ _ _ _ __ _ _ _ _H EP -

Action - Basis of Re- Calculation BasDesiption Imporvance. CLT Necessary Coment UVI-XIIE-FO- Failure to F-V = 0.053 -. 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> No This operator action is modeled in the Hope Creek ALIGN Align FP for (I) PRA with an HEP of 0.99 due to procedural Late RPV limitations. The SAGs direct use of FP for RPV Injection injection, but FP injection is not referenced in the EOPs. As such, The CPPU has no effect on the current modeling of this operator action.

SWS-XHE- Failure to F-V = 0.053 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> No This operator action is modeled in the Hope Creek PROC Align SSW for )

(I PRA with an HEP of 1.0 due to procedural Late RPV limitations. The SAGs direct use of SSW for RPV Injection injection, but SSW injection is not referenced in the EOPs. As such, The CPPU has no effect on the current modeling of this operator action.

NR-UIX-DEP- Failure to F-V = 0.047 -33 minutes 27 minutes Yes The Hope Creek PRA uses a value of 27 minutes for SRV Depressurize the HEP calculations for depressurization based on with SRV w/o MAAP Cases IB-LI-3-SBO (IICOOIO) and ID-LI-High Pressure. 7B3 (HCO017). The MAAP cases indicate that the Injection. time allowable for the CPPU case is reduced approximately 6 minutes. This decrease in time is calculated to result in a change in the quantified HEP. This basic event change was included in the evaluation of the change in risk metrics (see Table 10-8) as one of the contributors to the risk increase.

10-53

NEDO-33076 Table 10-10 Disposition of Key Actions for Potential HEP Re-Calculation

'Action Time Available

_ ___ __ __ _ _ _ IEP~,

Acdion Basis of - Re- Calculation B v Description Importance CLTP - cpru Necessary Comment

.Basic Event ID :-*.-  :.-~:

NR-%IE-SWS Non-recovery F-V = 0.035 No Not quantified - it is judged that the probabilities are of %lE-SWS not significantly different based on plant response and calculations using the Cause Based Decision Tree Method, EPRI TR 100259.(2) (Reference 34)

RX-FW-ADS Dependent F-V = 0.02 30 min. 27 min. Yes The constituent events of this combination HEP are Operator NRQFWLVH4M-03 and NR-UIX-DEP-SRV. For Actions - event NRQFWLVH4M-03, the time frame for the Operator. Fails operator action is estimated to be 4 minutes based on FW Control operator available interviews.

for operatorsThis time is FW to reduce bascd on before flow the time and ADS potentially reaching the Level 8 high level trip following a scram. The 4 min. time frame is expected to be dependent on the response of the FW control system and not significantly affected by CPPU. For event NR-UIX-DEP-SRV, the CPPU effect on the HEP has been calculated. (See NR-UIX-DEP-SRV in this table.) The dependent HEP combination failure probability is reassessed to determine the effect of CPPU. The basic event was modified and included in Table 10.8 as one of the contributors to the risk increase.

10-54

NEDO-33076 Table 10-10 Disposition of Key Actions for Potential HEP Re-Calculation Action Time Available

- - - - - - - -HEP Action- Basis of Re- Calculation Description Importance CLTP Cppu Necessary- Coniment Basic Event ID.

SAC-XHE-FO- SACS Heat F-V = 0.019 46 minutes 40 minutes Yes The manipulation of SACS heat loads is evaluated in HEAT Load (3) the PRA for the worst case conditions of high river Manipulation water temperature and high SACS temperatures.

For these conditions, the time frames for crew action result in a change in the calculated HEP. This action is required for certain SACS configurations that may occur following a LOOP event.

RHS-REPAIR- Repair/Recove F-V = 0.019 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> No This is a recovery term for long term loss of DHR TR ry of RHR For (1) sequences. The time frame is 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> based on the Loss of DHR time to pressurize the containment and close the Events SRVs. The recovery failure probability is based on a mean time to repair of 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> for pumps and not directly on HEP calculations. The CPPU does not affect the appropriateness of this time frame nor the recovery failure probability determined based on their long time frame.

IGS-XHE-FO- Failure to F-V =0.011 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> No This action supports the containment vent action.

V5125 open cross The timing required is in excess of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. No connect valve measurable difference in the calculated HEP is found for CPPU.

10-55

NEDO-33076 Table 10-10 Disposition of Key Actions for Potential HEP Re-Calculation Action Time AVAiiable Acdion Basis of Re- Calculation, lDescription Importance LT Ucessary  :', , ent 13asi e Event ID , ,,,-i , ... - , .

NR-RHR-INIT-L Failure to F-V = 0.010 '-20 hours 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> No This is a system initiation action for long term loss initiate RIIR (I) of DHR sequences. The time frame is 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> (Late) based on the time to pressurize the containment and close the SRVs. The small relative change in the time available for diagnosis and action due to CPPU implementation does not affect the calculation fo the HEP due to the extremely long time available from the initial cue. The CPPU does not affect the appropriateness of this time frame nor the recovery failure probability determined based on their long time frame.

Note:

(1) The action time available for the CLTP case is expected to be approximately the same or slightly more; however, a formal assessment of the time available for the CLTP case is not necessary in determining whether a change in the HEP calculation is warranted. The actions for which this note applies have HEPs that are conservative in nature and would not be affected by the potential changes in available timings due to the CPPU.

(2) The HEPs are, in general, calculated using the EPRI Cause-Based Methodology for the cognitive portion of the analysis (as implemented in the EPRI HRA Calculator). The EPRI calculator methodology results in minimal effects on the calculated HEPs due to CPPU implementation.

(3) CPPU action time is calculated based on a decay heat 12.3% greater than OLTP.

10-56

NEDO-33076 Table 10-11 Changes in Success Criteria Included in the Risk Assessment Success Criteria CLTP CPPU Reactivity Control RPS and SLC No Change RPV Overpressure Protection

• ATWS RPT + ll ofl4 SRVs RPT + 12 of l4 SRVs

• Non-ATWS 4 of 14 SRVs No Change RPV Depressurization 1 SRV 2 SRVs RPV Inventory Makeup

• High Pressure* FW, HPCI, RCIC No Change

• Low Pressure* Condensate, LPCI, CS No Change

• Sequence dependent 10-57

NEDO-33076 Table 10-12 Hope Creek Piping FAC Parameter Comparison for CPPU Parameter - .lAiloiableInput l .sCLTPl .120% OLTP CPPU Values l_ ._._ --- ;_._;l_-_ Range , Values, Values l'Within Range l Condensate Mass Flow Rate I - 100,000,000 14,447,970 17,471,844 Yes (Ibm/hr)

Velocity (ft/sec) Calculated in 8 - 24 9 -28 Yes program Steam Quality %) 0 to 100 0 0 Yes Operating 0 to 750 154 - 365 153 - 384 Yes Temperature (0F)

Cross Around HP Exhaust to Steam MS Mass Flow Rate I - 100,000,000 13,010,700 15,156,040 Yes (Ibm/hr)

Velocity (fi/sec) Calculated in 156 145 Yes program Steam Quality %) 0 to 100 <1 <1 Yes Operating 0 to 750 373 392 Yes Temperature ( IF)

Extraction Steam (#5 and #6)

Mass Flow Rate 1 - 100,000,000 <1,058,668 <1,229,102 Yes (Ibm/hr)

Velocity (ft/sec) Calculated in < 150 < 146 Yes program Steam Quality %) 0 to 100 <1 <1 Yes Operating 0 to 750 374-427 392 -446 Yes Temperature ( IF) 10-58

NEDO-33076 Table 10-12 Hope Creek Piping FAC Parameter Comparison for CPPU continued

.Parameter l_-_._-_--_._'--l_

1 Allowable Input Range 1-CLTP [ 120% LTP

,alues V.:Va',es

.CPPUValues

'Within Rang'e i

Feedwater Mass Flow Rate 1 - 100,000,000 14,348,780 17,340,74 Yes (Ibm/hr)

Velocity (ft/sec) Calculated in 9 - 19 10 - 23 Yes program Steam Quality %) 0 to 100 0 0 Yes Operating 0 to 750 365 -421 384-436 Yes Temperature (fF)

Heater Drains Mass Flow Rate I - 100,000,000 <5,334,036 <6,620,495 Yes (Ibm/hr)

Velocity (ftlsec) Calculated in 2 - 21 3 -25 Yes program Steam Quality %) 0 to 100 <1 <1 Yes Operating 0 to 750 218 - 377 228 -396 Yes Temperature (TF)

Moisture Separator Drains Mass Flow Rate I - 100,000,000 1,633,064 1,799,507 Yes (Ibm/hr)

Velocity (ft/sec) Calculated in 2 -9 2 - 10 Yes program Steam Quality %) 0 to 100 0- 0.004 0 - 0.004 Yes Operating 0 to 750 370 - 373 388 - 391 Yes Temperature ( IF) 10-59

NEDO-33076 Table 10-12 Hope Creek Piping FAC Parameter Comparison for CPPU continued

'Parameter, Allowable Input

.Rangc CLTP

'120% OLTP I CPPU Values

.. l--Vaus5'I,l.Whin'Range.'l l.. Vlues

1. 4 Extraction Steam to Seal Steam Mass Flow Rate I - 100,000,000 28,007 28,195 Yes (Ibm/hr)

Velocity (ftsec) Calculated in 116 93 Yes program Steam Quality %) 0 to 100 <1 <1 Yes Operating 0 to 750 322 345 Yes Temperature ( IF) 10-60

NEDO-33076 Figure 10-1 Generalized "Bathtub" Reliability Curve for a Component 10-61

NEDO-33076 NEI Self Assessment Process Guideline Disposition of Revision 2005B

_ Changes . PRA Model Figure 10-2 PRA Self-assessment Process Applied to Hope Creek 10-62

NEDO-33076

11. REFERENCES

1. GE Nuclear Energy, "Constant Pressure Power Uprate," Licensing Topical Report NEDC-33004P-A, Revision 4, Class III (Proprietary), July 2003.

2. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Licensing Topical Reports NEDC-32424P-A, Class III (Proprietary), February 1999.

3. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," Licensing Topical Reports NEDC-32523P-A, Class III, February 2000; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999.

4. GE Nuclear Energy, "General Electric Standard Applications for Reactor Fuel", NEDE-2401 1-P-A-14, June 2000.

5. GE Nuclear Energy, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465-A, August 1996.

6. NEDO-32983-A "GE Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations," Rev 0, December 2001.

7. GE Nuclear Energy, "The GE Pressure Suppression Containment System Analytical Model," NEDM- 10320, March 1971.

8. GE Nuclear Energy, "The General Electric Mark III Pressure Suppression Containment System Analytical Model," NEDO-20533, June 1974.

9. GE Nuclear Energy, "Maximum Discharge of Liquid-Vapor Mixtures from Vessels,"

NEDO-21052, September 1975.

10. GE Nuclear Energy, "General Electric Model for LOCA Analysis In Accordance With 10 CFR 50 Appendix K," NEDE-20566-P-A, September 1986.

11. NUREG-0800, U.S. Nuclear Regulatory Commission, Standard Review Plan, Section 6.2.1.1.C, "Pressure - Suppression Type BWR Containments," Revision 6, August 1984.

12. NUREG-0661, "Mark I Containment Long-Term Program Safety Evaluation Report,"

July 1980.

13. GE Nuclear Energy, "Mark I Containment Program Load Definition Report," NEDO-21888, Revision 2, November 1981.

14. NUREG-0783, Suppression Pool Temperature Limits for BWR Containment, November 1981.

15. GE Nuclear Energy, Hope Creek Generating Station Suppression Pool Temperature Response, NEDC-30154, June 1983.

16. Letter to Gary L. Sozzi (GE) from Ashok Thadani (NRC) on the Use of the SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis, July 13, 1993.

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NEDO-33076

17. Letter to Patrick W. Marriott (GE) from William T. Russel (NRC) forwarding the Staff Position Paper on General Electric Boiling Water Reactor Power Uprate Program (TAC NO. M79384), September 30, 1991.

18. Letter to G. L. Sozzi (GE) from Dennis M. Crutchfield (NRC) forwarding the Staff Position Concerning General Electric Boiling Water Reactor Power Uprate Program (TAC No. M91680), February 8, 1996.

19. "Hope Creek Generating Station Plant Unique Analysis Report Volume I General Criteria and Loads Methodology", BPC-01-300-1, Revision 0, January 1984.

20. "HCGS Plant Unique Analysis Report (PUAR)", transmitted under cover from R.L.

Mittl (PSE&G) to A. Schwencer (NRC), February 10, 1984.

21. GE Nuclear Energy, "Evaluation of MARK I S/RV Load Cases C3.1, C3.2, C3.3 For the Hope Creek Nuclear Generating Station, NEDC-22200, August 1982.

22. Letter LR-N97792 to USNRC from Public Service Electric and Gas Company, 90 Day Response to Generic Letter 97-04, Assurance of Sufficient Net Positive Suction Head (NPSH) for Emergency Core Cooling (ECCS) and Containment Heat Removal Pumps, Hope Creek Generating Station, December 30, 1997.

23. NEDO-32686, the BWROG Utility Resolution Guidance for ECCS Suction Strainer Blockage, November 1996.

24. GE Nuclear Energy, "General Electric Instrument Setpoint Methodology," NEDC-31336P-A, Class III (Proprietary), September 1996.

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27. GE Nuclear Energy, "ATWS Rule Issues Relative to BWR Core Thermal-Hydraulic Stability," NEDO-32047-A, June 1995.

28. GE Nuclear Energy, "Mitigation of BWR Core Thermnal-Hydraulic Instabilities in ATWS," NEDO-32164, December 1992.

29. Hope Creek Generating Station APRM / RBM / Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS / MELLLA)," NEDC-33066P, Revision 2, February 2005.

30. US Nuclear Regulatory Commission, Regulatory Guide 1.174 "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, November 2002.

31. "ASME PRA Standard," ASME-RA-S-2002.

32. NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management,"

1991.

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33. NEI, "Industry Guidance for Addressing the ASME PRA Standard," DRAFT, August 15, 2002.

34. Parry, G.W., "An Approach to the Analysis of Operator Actions in Probabilistic Risk Assessment," EPRI TR-100259, June 1992.

35. Swain, A.D., Guttmann, H.E., "Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications," NUREG/CR-1278, August 1983.

36. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 222 to Facility Operating License No. DPR-71 and Amendment No. 247 to Facility Operating License No. DPR-62, Carolina Power & Light Company, Brunswick Steam Electric Plant, Units I and 2, Docket Nos. 50-325 and 50-324

37. Safety Evaluation by the Office of Nuclear Reactor Regulation to Facility Operating License No. DPR-19 and to Facility Operating License No. DPR-25 Exelon Generation Company, LLC Dresden Nuclear Power Station, Units 2 and 3 Docket Nos. 50-237- and 50-249.

38. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 202 to Facility Operating License No. DPR-29 and Amendment No. 198 to Facility Operating License No. DPR-30 Exelon Generation Company, LLC and Midamerican Energy Company Quad Cities Nuclear Power Station, Units I and 2 Docket Nos. 50-254 and 50-562.

39. "Evaluation of Severe Accident Risks: Peach Bottom, Unit 2," NUREG/CR-4551, Vol.

4, December 1990.

40. "Severe Accident Risk Assessment, Limerick Generating Station," Volume 1, Main Report, prepared for PECo by NUS Corporation, April 1983.

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