Text

NEDO-33153, SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for Hope Creek Generating Station.
	ML042590419
Person / Time
Site:	Hope Creek
Issue date:	08/31/2004
From:	; General Electric Co
To:	; Office of Nuclear Reactor Regulation
References
	LR-N04-0378 DRF 0000-0026-3882, NEDO-33153
	Download: ML042590419 (136)
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GE NuclearEnergy NEDO-33153 DRF 0000-0026-3882 Class 1 August 2004 SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for Hope Creek Generating Station

GE Nuclear Energy 175 CurtnerAvenue San Jose, CA 95125 NEDO-33 153 DRF 0000-0026-3882 Class I August 2004 SAFER/GESTR-LOCA Loss of Coolant Accident Analysis fo r Hope Creek Generating Station

NEDO-33 153 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 the PSEG Nuclear LLC and GE for Fuel Bundle Fabrication and Related Services, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than PSEG Nuclear LLC, 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.

Copyright, General Electric Company, 2004.

NEDO-33 153 TABLE OF CONTENTS Pave

SUMMARY

S-1

1.0 INTRODUCTION

1-1

2.0 DESCRIPTION

OF MODELS 2-1 2.1 LAMB 2-1 2.2 SCAT/TASC 2-1 2.3 GESTR-LOCA 2-1 2.4 SAFER 2-2 3.0 ANALYSIS PROCEDURE 3-1 3.1 Licensing Criteria 3-1 3.2 SAFER/GESTR-LOCA Licensing Methodology 3-1 3.3 Generic Analysis 3-3 3.4 Hope Creek Plant-Specific Analysis 3-3 3.5 Analysis of Mixed Cores 3-5 4.0 INPUTS TO ANALYSIS 4-1 4.1 Plant Inputs 4-1 4.2 Fuel Parameters 4-1 4.3 ECCS Parameters 4-1 5.0 RESULTS 5-1 5.1 Break Spectrum Calculations 5-1 5.1.1 Recirculation Line Breaks 5-1 5.1.2 Non-Recirculation Line Breaks 5-3 5.2 Compliance Evaluations 5-3 5.2.1 Licensing Basis PCT Evaluation 5-3 5.2.2 Removal of the Current Requirement for Evaluation 5-3 of Upper Bound PCT 5.3 Expanded Operating Domain and Alternate Operating Modes 5-6 5.3.1 Increased Core Flow (ICF) 5-6 5.3.2 Reduced Core Flow (MELLLA / ELLLA) 5-6 5.3.3 Single-Loop Operation (SLO) 5-7 5.4 MAPLHGR Limits 5-8

6.0 CONCLUSION

S 6-1

7.0 REFERENCES

7-1 LAST PAGE B-39 iii

NEDO-33 153 APPENDICES A SYSTEM RESPONSE CURVES FOR NOMINAL A-1 RECIRCULATION LINE BREAKS B SYSTEM RESPONSE CURVES B-I FOR APPENDIX K RECIRCULATION LINE BREAKS iv

NEDO-33 153 LIST OF TABLES Table Title Pane 3-1 Analysis Assumptions for Nominal Calculations 3-7 3-2 Analysis Assumptions for Appendix K Calculations 3-8 4-1 Plant Parameters Used in Hope Creek SAFER/GESTR-LOCA Analysis 4-2 4-2 Fuel Parameters Used in Hope Creek SAFER/GESTR-LOCA Analysis 4-3 4-3 Hope Creek SAFER/GESTR-LOCA Analysis ECCS Parameters 4-4 4-4 Hope Creek Single Failure Evaluation 4-8 5-1 Summary of Hope Creek SAFER/GESTR-LOCA Results for Recirculation 5-10 Line Breaks 5-2 Summary of Hope Creek SAFER/GESTR-LOCA Results for Non- 5-12 Recirculation Line Breaks 5-3 Maximum Extended Load Line Limit Analysis Results Comparison for 5-13 Hope Creek 5-4 Single-Loop Operation Results Comparison for Hope Creek 5-14 6-1 SAFER/GESTR-LOCA Licensing Results for Hope Creek 6-2 6-2 Thermal Limits 6-3 A-l Nominal Recirculation Line Break Figure Summary A-2 B-I Appendix K Recirculation Line Break Figure Summary B-2 V

NEDO-33 153 vi

NEDO-33153 LIST OF FIGURES Fisaure Title Paae 2-1 Flow Diagram of LOCA Analysis Using SAFER/GESTR 2-3 3-1 Hope Creek Decay Heat Used for Nominal and Appendix K Calculations 3-9 4-1 Hope Creek ECCS Configuration 4-9 5-1 Nominal and Appendix K LOCA Break Spectrum Results for GE14 Fuel 5-15 5-2 Nominal and Appendix K LOCA Break Spectrum Results for SVEA-96+ 5-16 Fuel A-1 DBA Suction - Battery Failure (Nominal) -

3LPCI+ILPCS+ADS Available

a. Water Level in Hot and Average Channels A-3

b. Reactor Vessel Pressure A-4

c. Peak Cladding Temperature (GE14) A-5

d. Heat Transfer Coefficient (GE 14) A-6

e. ECCS Flow A-7

f. Peak Cladding Temperature (SVEA-96+) A-8

g. Heat Transfer Coefficient (SVEA-96+) A-9 A-2 80% DBA Suction - Battery Failure (Nominal) -

3LPCI+LPCS+ADS Available

a. Water Level in Hot and Average Channels A-10

b. Reactor Vessel Pressure A-1I

c. Peak Cladding Temperature (GE14) A-12

d. Heat Transfer Coefficient (GE 14) A-13

e. ECCS Flow A-14

f. Peak Cladding Temperature (SVEA-96+) A-15

g. Heat Transfer Coefficient (SVEA-96+) A-16 A-3 60% DBA Suction - Battery Failure (Nominal) -

3LPCI+LPCS+ADS Available

a. Water Level in Hot and Average Channels A-17

b. Reactor Vessel Pressure A-18

c. Peak Cladding Temperature (GE14) A-19

d. Heat Transfer Coefficient (GE14) A-20

e. ECCS Flow A-21

f. Peak Cladding Temperature (SVEA-96+) A-22

g. Heat Transfer Coefficient (SVEA-96+) A-23 A-4 1 ft2 Suction - Battery Failure (Nominal) -

3LPCI+LPCS+ADS Available

a. Water Level in Hot and Average Channels A-24

b. Reactor Vessel Pressure A-25

c. Peak Cladding Temperature (GE14) A-26

d. Heat Transfer Coefficient (GE]4) A-27

e. ECCS Flow A-28

f. Peak Cladding Temperature (SVEA-96+) A-29

g. Heat Transfer Coefficient (SVEA-96+) A-30 vii

NEDO-33153 LIST OF FIGURES (Continued)

Figure Title Page A-5 ((

a. Water Level in Hot and Average Channels A-31

b. Reactor Vessel Pressure A-32

c. Peak Cladding Temperature (GE14) A-33

d. Heat Transfer Coefficient (GE 14) A-34

e. ECCS Flow A-35

f. Peak Cladding Temperature (SVEA-96+) A-36

g. Heat Transfer Coefficient (SVEA-96+) A-37 B-I DBA Suction - Rated - Battery Failure (App. K)-

3LPCI+LPCS+ADS Available

a. Water Level in Hot and Average Channels B-3

b. Reactor Vessel Pressure B-4

c. Peak Cladding Temperature (GE14) B-5

d. Heat Transfer Coefficient (GE14) B-6

e. ECCS Flow B-7

f. Core Inlet Flow B-8

g. Minimum Critical Power Ratio (GE14 & SVEA-96+) B-9

h. Peak Cladding Temperature (SVEA-96+) B-10

i. Heat Transfer Coefficient (SVEA-96+) B-i 1 B-2 DBA Suction - MELLLA - Battery Failure (App. K)-

3LPCI+LPCS+ADS Available

a. W'ater Level in Hot and Average Channels B- 12

b. Reactor Vessel Pressure B-13

c. Peak Cladding Temperature (GE14) B-14

d. Heat Transfer Coefficient (GE14) B-15

e. ECCS Flow B-16

f. Peak Cladding Temperature (SVEA-96+) B-17

g. Heat Transfer Coefficient (SVEA-96+) B-18 B-3 80% DBA Suction -Battery Failure (App. K)-

3LPCI+LPCS+ADS Available

a. Water Level in Hot and Average Channels B- 19

b. Reactor Vessel Pressure B-20

c. Peak Cladding Temperature (GE14) B-21

d. Heat Transfer Coefficient (GE14) B-22

e. ECCS Flow B-23

f. Peak Cladding Temperature (SVEA-96+) B-24

g. Heat Transfer Coefficient (SVEA-96+) B-25 Viii

NEDO-33153 LIST OF FIGURES (Continued)

Figure Title PaMe B-4 60% DBA Suction -Battery Failure (App. K)-

3LPCI+LPCS+ADS Available

a. Water Level in Hot and Average Channels B-26

b. Reactor Vessel Pressure B-27

c. Peak Cladding Temperature (GE14) B-28

d. Heat Transfer Coefficient (GE14) B-29

e. ECCS Flow B-30

f. Peak Cladding Temperature (SVEA-96+) B-31

g. Heat Transfer Coefficient (SVEA-96+) B-32 B-5

))

a. Water Level in Hot and Average Channels B-33

b. Reactor Vessel Pressure B-34

c. Peak Cladding Temperature (GE14) B-35

d. Heat Transfer Coefficient (GE 14) B-36

e. ECCS Flow B-37

f. Peak Cladding Temperature (SVEA-96+) B-38

g. Heat Transfer Coefficient (SVEA-96+) B-39 ix

NEDO-33 153 ACKNOWLEDGEMENTS The following individuals contributed significantly toward completion of this report:

D. Abdollahian C. Bott C. Johnson A. Khan K. Knippel A. J. Lipps F. M. Paradiso J. Stott G. Thomas W-M. Wong x

NEDO-33 153 SUMM1ARY A design requirement for nuclear power plants is the capability to withstand Design Basis Accidents. One of the postulated accidents is a guillotine break in the largest size pipe connected to the reactor vessel. Historically, the analysis of the large break loss-of-coolant accident (LOCA) had been performed on a very conservative basis with margin added at every step of the calculation. This was done partly as a result of the restrictions imposed by the requirements of 10CFR50.46 and Appendix K, and partly to compensate for uncertainties inherent in the simplified models. However, after years of research with large-scale experiments and the development of the best-estimate codes, improved and more realistic boiling water reactor (BWNR) licensing models (i.e., SAFER/GESTR-LOCA) have been approved by the U.S. Nuclear Regulatory Commission (NRC). These new models calculate more realistic (yet conservative) peak cladding temperatures (PCT) to relieve unnecessary plant operating and licensing restrictions. More realistic analyses also predict actual plant response during postulated accidents and can be used as a basis for more appropriate operator actions. The LOCA analysis for Hope Creek uses these models and this licensing methodology.

The SAFER and GESTR-LOCA models are coupled, mechanistic, reactor system thermal hydraulic and fuel rod thermal-mechanical evaluation models. These models are based on realistic correlations and inputs. The SAFER/GESTR-LOCA methodology approved by the NRC allows the plant-specific break spectrum to be defined using nominal input assumptions.

However, the calculation of the limiting PCT to demonstrate conformance with the requirements of IOCFR50.46 must include specific inputs documented in Appendix K. The SAFER/GESTR-LOCA Application Methodology requires:

(1) The Licensing Basis PCT must be less than 2200'F. This Licensing Basis PCT is derived by adding appropriate margin for specific conservatism required by Appendix K of 10CFR50 to the limiting PCT value calculated using nominal values.

(2) The Licensing Basis PCT is required to be greater than the Upper Bound PCT.

(3) The NRC placed a restriction of 1600'F on the Upper Bound PCT in the Safety Evaluation Report (SER) approving the SAFER/GESTR-LOCA application methodology. This restriction is based on the range of test data and analyses used to generically qualify the SAFER code and application methodology. Therefore, it is required that the Upper Bound PCT be below 1600'F, otherwise additional plant S-1

NEDO-33 153 specific analyses must be performed.

The Upper Bound PCT limit of 1600°F was removed in a Supplemental Licensing Topical Report, Reference 8. Reference 8 shows that GE has performed the plant specific Upper Bound PCT calculations for its entire product line and unless there are significant changes to the plant's configuration, plant specific evaluation of Upper Bound PCT is not required.

The SAFER/GESTR-LOCA analysis for Hope Creek was performed in accordance with NRC requirements and demonstrates conformance with the Emergency Core Cooling System (ECCS) acceptance criteria of IOCFR50.46 Appendix K. A sufficient number of plant-specific break sizes were evaluated to establish the behavior of both the nominal and Appendix K PCTs as a function of break size. Different single failures were also investigated in order to clearly identify the worst cases. The Hope Creek specific ECCS analysis was performed with conservative values for the Peak Linear Heat Generation Rate (PLHGR) and initial Minimum Critical Power Ratio. This analysis is applicable to the rated thermal power of 3339 MWt (nominal assumptions) and the following operating conditions: Maximum Extended Operating Domain (MEOD) [includes Maximum Extended Load Line Limit (MELLL) and Increased Core Flow (ICF)], and Single Loop Operation (SLO). The analysis results demonstrated that the five acceptance criteria specified in IOCFR50.46 for ECCS performance analyses are satisfied. The Licensing Basis PCTs for Hope Creek are 1370'F for GE14 and 1540WF for SVEA-96+, which are below the 2200WF limit. Therefore, the Hope Creek specific analysis meets the NRC SAFER/GESTR-LOCA licensing analysis requirements.

S-2

NEDO-33 153

1.0 INTRODUCTION

This document provides the results of the Loss-of-Coolant Accident (LOCA) analysis performed by GE Nuclear Energy (GE-NE) for Hope Creek Generating Station. The analysis was performed using the SAFER/GESTR-LOCA Application Methodology approved by the Nuclear Regulatory Commission (NRC) (Reference 1). This analysis was performed assuming a rated thermal power level of 3339 MWt. The analysis addresses a core flow range from 76.6% to 105% of rated core flow and a single loop operation assuming a nominal power level of 2337 MWt at 60% of the rated core flow. This is the first application of the SAFER/GESTR-LOCA methodology for Hope Creek.

The LOCA analysis was performed in accordance with NRC requirements to demonstrate conformance with the ECCS acceptance criteria of IOCFR50.46. A key objective of the LOCA analysis is to provide assurance that the most limiting break size, break location, and single failure combination has been considered. Reference 2 documents the requirements and the approved methodology to satisfy these requirements.

The SAFER/GESTR-LOCA application methodology is based on the generic studies presented in the Reference 2 documentation. The approved application methodology consists of three essential parts. First, potentially limiting LOCA cases are determined by applying realistic (nominal) analytical models across the entire break spectrum. Second, limiting LOCA cases are analyzed with an Appendix K model (inputs and assumptions), which incorporates all the required features of IOCFR50 Appendix K. For the most limiting cases, a Licensing Basis Peak Cladding Temperature (PCT) is calculated based on the nominal PCT with an adder to account statistically for the differences between the nominal and Appendix K assumptions. The application methodology required a statistically derived Upper Bound PCT to be calculated to demonstrate the conservatism of the Licensing Basis PCT. The resulting Licensing Basis PCT would then conform to all the requirements of IOCFR50.46 and Appendix K.

As discussed in Section 3.2, further plant specific evaluation of Upper Bound PCT is no longer required to meet the SAFER/GESTR-LOCA application methodology requirements, unless there are significant changes in the plant's configuration.

1-1

NEDO-33 153 1-2

NEDO-33 153

2.0 DESCRIPTION

OF MODELS Four GE-NE computer models were used in the LOCA analysis to determine the LOCA response for Hope Creek. These models are LAMB, SCAT/TASC, GESTR-LOCA, and SAFER.

Together, these models evaluate the short-terrn and long-term reactor vessel blowdown response to a pipe rupture, the subsequent core flooding by ECCS, and the final rod heatup. Figure 2-1 is a flow diagram of these computer models, including the major code functions and the transfer of major parameters. The purpose of each model is described in the following subsections.

2.1 LANIB This model (Reference 3) analyzes the short-term blowdown phenomena for postulated large pipe breaks in which nucleate boiling is lost before the water level drops sufficiently to uncover the active fuel. The LAMB output (primarily core flow as a function of time) is used in the SCAT model for calculating blowdown heat transfer and fuel dryout time.

2.2 SCAT/TASC This model (Reference 3) completes the transient short-term thermal-hydraulic calculation for large recirculation line breaks. Developed for GEl I and later fuels with partial-length rods, an improved SCAT model (designated "TASC") is used to predict the time and location of boiling transition and dryout. The time and location of boiling transition is predicted during the period of recirculation pump coastdown. When the core inlet flow is low, TASC also predicts the resulting bundle dryout time and location. The calculated fuel dryout time is an input to the long-term thermal-hydraulic transient model, SAFER.

2.3 GESTR-LOCA This model (Reference 4) provides the parameters to initialize the fuel stored energy and fuel rod fission gas inventory at the onset of a postulated LOCA for input to SAFER. GESTR-LOCA also establishes the transient pellet-cladding gap conductance for input to both SAFER and SCAT/TASC.

2-1

NEDO-33 153 2.4 SAFER This model (References 5 and 6) calculates the long-term system response of the reactor over a complete spectrum of hypothetical break sizes and locations. SAFER is compatible with the GESTR-LOCA fuel rod model for gap conductance and fission gas release. SAFER calculates the core and vessel water levels, system pressure response, ECCS performance, and other primary thermal-hydraulic phenomena occurring in the reactor as a function of time. SAFER realistically models all regimes of heat transfer that occur inside the core, and provides the heat transfer coefficients (which determine the severity of the temperature change) and the resulting PCT as functions of time. For GEl I and later fuel analysis with the SAFER code, the part length fuel rods are treated as full-length rods, which conservatively overestimate the hot bundle power.

2-2

NEDO-33 153 LAMB GESTR-LOCA -* SAFER SHORT-TERM THERMAL FUEL ROD LONG-TERM THERMAL HYDRAULIC TRANSIENT MODEL THERMAUMECHANICAL DESIGN HYDRAULIC TRANSIENT MODEL

/ OUTPUT \i OUTPUT CORE AVERAGE PRESSURE GAP CONDUCTANCE CORE INLET FLOW CORE INLET ENTHALPY ROD INTERNAL PRESSURE TASC 1,

.I TRANSIENT CRITICAL POWER MODEL I' OUTPUT OUTPUT PCT WATER LEVEL RESPONSE LOCATION AND TIME OF PRESSURE BOILING TRANSITION HEAT TRANSFER COEFFICIENT LOCAL OXIDATION Figure 2-1. Flow Diagram of LOCA Analysis Using SAFER/GESTR 2-3

NEDO-33 153 2-4

NEDO-33 153 3.0 ANALYSIS PROCEDURE 3.1 LICENSING CRITERIA The Code of Federal Regulations (IOCFR50.46) outlines the acceptance criteria for ECCS performance analyses. A summary of the acceptance criteria is provided below.

Criterion I - Peak Cladding Temperature - The calculated maximum fuel element cladding temperature shall not exceed 2200'F.

Criterion 2 - Maximum Cladding Oxidation - The calculated local oxidation shall not exceed 0.17 times the cladding thickness before oxidation.

Criterion 3 - Maximum Hydrogen Generation - The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinder surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

Criterion 4 - Coolable Geometrv - Calculated changes in core geometry shall be such that the core remains amenable to cooling.

Criterion 5 - Long-Term Cooling - After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

Conformance with Criteria I through 3 for Hope Creek is presented in this report. As discussed in Reference 3, conformance with Criterion 4 is demonstrated by conformance to Criteria 1 and

2. The bases and demonstration of compliance with Criterion 5 are documented in References 3 and 6, and remain unchanged by application of SAFER/GESTR-LOCA.

3.2 SAFER/GESTR-LOCA LICENSING METIIODOLOGY The SAFER/GESTR-LOCA licensing methodology approved by the NRC in Reference 1 allows the plant-specific break spectrum to be defined using nominal input assumptions. However, the 3-1

NEDO-33 153 calculation of the limiting PCT to demonstrate conformance with the requirements of I OCFR50.46 must include specific inputs and models required by Appendix K.

The Licensing Basis PCT is based on the most limiting LOCA (highest PCT) and is defined as:

PCT Licensing

.. = PCTmia.

o na + ADDER The value of ADDER is calculated as follows:

ADDER 2 = [PCTAppP - PCTNominaI] 2 + X ( 5PCT) 2 where:

PCTApp. K = Peak cladding temperature from calculation using Appendix K specified models and inputs.

PCTNOminaI= Peak cladding temperature from nominal case.

L ( 5PCTj) 2 =Plant variable uncertainty term.

The plant variable uncertainty term accounts statistically for the uncertainty in parameters that are not specifically addressed by I OCFR50 Appendix K.

To conform to l OCFR50.46 and the SAFER/GESTR-LOCA licensing methodology, the Licensing Basis PCT must be less than 2200'F.

Demonstration that the Licensing Basis PCT calculated above is sufficiently conservative is also required through the use of a statistical Upper Bound PCT as defined in Reference 2. The Upper Bound PCT is required to be less than the Licensing Basis PCT. This ensures that the Licensing Basis PCT bounds the expected PCT for at least 95% of all postulated limiting break LOCAs, which occur from limiting initial conditions. As part of the development of SAFER/GESTR-LOCA licensing methodology, GE-NE demonstrated that this criterion was satisfied generically for the BWR-3 through BWR-6 classes of plants. As shown in Reference 8, further plant specific Upper Bound PCT calculations are no longer required. In Reference 2, the application methodology was accepted on a generic basis for an Upper Bound PCT up to 1600"F. This 3-2

NEDO-33 153 1600'F restriction was removed in Reference 8. Section 5.2.2 demonstrates that the Licensing Basis PCTs for the fuels and conditions analyzed bound the estimated Upper Bound PCTs based on a plant-specific Upper Bound PCT calculation previously performed.

3.3 Generic Analysis Hope Creek was designed as one of the GE BWR/4 product line plants; however, the ECCS includes features that are used in the BWR-5/6 plants. The LPCI injection is into the bypass region of the core and part of the HPCI flow is used for high pressure core spray. As such, the Hope Creek response to a LOCA event is closer to that of a BWR-5/6 than a BWR-4. GE-NE performed a generic conformance calculation on the limiting hypothetical LOCA (Reference 2) for GE BWR plants which have LPCI injection into the bypass region (BWAIR-5/6 and some BWR/4 such as Hope Creek). The SAFER analysis of a typical BWRI6 was performed for this purpose. The limiting LOCA was determined from the nominal break spectrum as the break size and single ECCS component failure combination yielding the highest nominal PCT. The Appendix K calculation wvas then performed for this limiting LOCA event to establish the basis for the licensing evaluation.

The DBA suction break with failure of the High Pressure Core Spray (HPCS) was generically found to be the limiting break in the nominal break spectrum for BWR/6 plants. In Hope Creek, there is no HPCS, but the battery failure results in the unavailability of the high-pressure make-up system. The Hope Creek High Pressure Coolant Injection (HPCI) system injects part of its makeup flow through the core spray. As a result, these cases were used to perform the Appendix K calculations. The Licensing Basis PCTs were then calculated by combining the nominal PCTs with the adders described in Section 3.2.

3.4 Hope Creek Plant-Specific Analysis As discussed in the SER (Reference 2) the determination of the limiting case LOCA is based on:

1. The generic Appendix K PCT versus break size curve exhibits the same trends as the generic Nominal PCT versus break size curve for a given class of plants;

2. The limiting LOCA determined from Nominal calculations is the same as that determined from Appendix K calculations for a given class of plants; and 3-3

NEDO-33153

3. Both generic and Nominal PCT versus break size curve and Appendix K PCT versus break size curve for a given class of plants are shown to be applicable on a plant specific basis. Necessary conditions for demonstrating applicability include:

a. Calculation of a sufficient number of plant specific PCT points to verify the shape of the curve;

b. Confirmation that plant specific Appendix K PCT calculations match the trend of the generic curve for that plant class;

c. Confirmation that plant specific operating parameters have been conservatively bounded by the models and inputs used in the generic calculations;

d. Confirmation that the plant specific ECCS is consistent with the referenced plant class ECCS configuration.

Conformance to conditions I and 2 has been demonstrated in Reference 2. In order to show that conditions 3a and 3b have been satisfied, plant specific analyses for break sizes ranging from 0.05 ft2 to the maximum DBA recirculation suction line break (4.08 ft2 ) were performed.

Compliance with conditions 3c and 3d are demonstrated with a plant specific Upper Bound PCT calculation.

Different single failures were also investigated to identify the worst cases. The break spectrum was first evaluated using nominal analysis assumptions (Table 3-1). The potentially limiting cases were then analyzed again with the analysis assumptions specified for the Appendix K calculations (Table 3-2). The normalized decay heat fractions used are shown in Figure 3-1. The Hope Creek nominal and Appendix K results were compared to assure that the PCT trends as a function of the break size were consistent with one another and with those of the generic BWR/6 break spectrum curve documented in Reference 2.

The Hope Creek SAFER/GESTR-LOCA analysis was performed using conservative values for Peak Linear Heat Generation Rate (PLHGR) and Initial Minimum Critical Power Ratio (MCPR) for the fuel types analyzed. Inputs used in the analysis are given in Section 4.

3.5 Analysis of Mixed Cores The SAFER/GESTR-LOCA analysis assumes an equilibrium core loading. This approach is acceptable because of the channeled configuration of BWR fuel assemblies. There is no channel-to-channel cross flow inside the core and the only issue of hydraulic compatibility of the 3-4

NEDO-33 153 various bundle types in a core is the bundle inlet flow rate variation. In order to provide an acceptable response during normal operation and transients, the overall bundle design is constrained such that the hydraulic response is similar between different fuel product lines. As a result, there is no significant difference in the hydraulic response for a mixed core as compared to an equilibrium core.

The SAFER analysis is insensitive to mixed cores. The PCT is determined by hot channel response. The hot bundle hydraulics are driven by the overall core pressure drop. This basic premise is valid because no channel-to-channel interaction occurs during a LOCA. In addition, the SAFER single channel modeling is conservative when compared to a multiple channel model (such as TRACG). TRACG models several core regions with multiple channels in each region.

The conservatism in the SAFER modeling is shown in the Upper Bound PCT evaluation in Appendix A of NEDC-23785-1-PA, Volume III (Reference 4). This conservatism is on the order of 175 0 F, which is much greater than the PCT variation resulting from mixed cores.

The first peak PCT is primarily influenced by the timing of boiling transition at the various elevations in the bundle. The boiling transition in the bundle is governed by the core flow coastdown characteristics and the bundle powver level. The core flow coastdowvn is a core-wide phenomenon determined by the initial core flow and the recirculation pump coastdown, neither of which are dependent on the fuel type. The bundle power also affects the boiling transition time; a higher power bundle will experience an earlier and potentially deeper boiling transition.

Because of the channeled configuration of BWR fuel assemblies, there is no channel-to-channel cross flow inside the core. The boiling transition in one bundle will not affect the other bundles in the core. The second peak PCT is primarily influenced by bundle flooding from the bottom.

This is a low flow rate process that is governed by the ECCS system capacity. There is no channel-to-channel interaction during this time. Therefore, the transition from a mixed core to an equilibrium core is not expected to affect the second peak PCT response.

Fuels from other vendors are analyzed in GE's thermal-hydraulic methodologies, including SAFER/GESTR-LOCA, as if they were GE fuel. The inputs to the thermal-hydraulic codes are flexible and can be adapted to a large variety of bundle designs. Sufficient information is obtained from the other vendor to allow modeling the thermal-hydraulic behavior of the other vendor's fuel using GE's codes. Most inputs can be used directly (e.g., dimensions, weights, material properties). A controlled benchmarking approach is used to model critical fuel 3-5

NEDO-33 153 performance correlations (e.g., boiling transition, bundle pressure drop) in a format compatible with GE's methods.

3-6

NEDO-33 153 Table 3-1 ANALYSIS ASSUMPTIONS FOR NOMINAL CALCULATIONS (Reference 2)

1. Decay Heat 1979 American Nuclear Society (ANS)

(Figure 3-1 )

2. Transition Boiling Temperature Iloeje correlation

3. Break Flow 1.25 HEM(t) (subcooled) 1.0 HEM"'l (saturated)

4. Metal-Water Reaction EPRI coefficients

5. Core Power 3339 MWt

6. Peak Linear Heat Generation Rate See Table 4-2

7. Bypass Leakage Coefficients Nominal values

8. Initial Operating Minimum Critical Power See Table 4-2 Ratio (MCPR)

9. ECCS Water Enthalpy (Temperature) 88 Btu/lbm (120 F)

10. ECCS Initiation Signals (See Table 4-3)

11. Automatic Depressurization System 120-second delay time (Table 4-3)

12. ECCS Available Systems remaining after worst case single failure.

13. Stored Energy Best Estimate GESTR-LOCA

14. Fuel Rod Internal Pressure Best Estimate GESTR-LOCA

15. Fuel Exposure Limiting fuel exposure which maximizes PCT (1) HEM: Homogeneous Equilibrium Model.

3-7

NEDO-33 153 Table 3-2 ANALYSIS ASSUMPTIONS FOR APPENDIX K CALCULATIONS (Reference 2)

1. Decay Heat 1971 ANS + 20% Decay Heat (Figure 3-1)

2. Transition Boiling Temperature Transition boiling allowed during blowdow.n only until cladding superheat exceeds 300'F.

3. Break Flow Moody Slip Flow Model with discharge coefficients of 1.0, 0.8, and 0.6.

4. Metal-Water Reaction Baker-Just

5. Core Power 3430 MWt

6. Peak Linear Heat Generation Rate See Table 4-2

7. Bypass Leakage Coefficients Same as Table 3-1

8. Initial Operating Minimum Critical Power See Table 4-2 Ratio (MCPR)

9. ECCS Water Enthalpy (Temperature) Same as Table 3-1

10. ECCS Initiation Signals Same as Table 3-1

]1. Automatic Depressurization System Same as Table 3-1

12. ECCS Available Same as Table 3-1

13. Stored Energy Same as Table 3-1

14. Fuel Rod Internal Pressure Same as Table 3-1

15. Fuel Exposure Same as Table 3-1 (1) 102.7% of nominal core power of 3339 MWt, per Hope Creek request.

3-8

NEDO-33 153 1.2 1

g 0.8-

-1971 ANS + 20% (Appendix K)

= 0.6 - - 1979 ANS (Nominal)

ZO0.4-0.2 -

0 0.01 0.1 1 10 100 1000 10000 Time After Break (seconds)

Figure 3-1. Hope Creek Decay Heat Used for Nominal and Appendix K Calculations 3-9

NEDO-33 153 3-10

NEDO-33 153 4.0 INPUT TO ANALYSIS 4.1 PLANT INPUTS The plant input parameters to Hope Creek LOCA analysis are presented in Tables 4-1, 4-2 and 4-

3. Table 4-1 shows the plant operating conditions, Table 4-2 shows the fuel parameters, and Table 4-3 identifies the key ECCS parameters used in the analysis. Table 4-4 identifies the combinations of single failures and available systems specifically analyzed for the Hope Creek ECCS configuration, illustrated in Figure 4-1.

4.2 FUEL PARAMETERS All SAFER/GESTR-LOCA analyses were performed with a conservative Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) at the most limiting combination of power and exposure (Table 4-2). These values were carefully selected in order to meet the IOCFR50.46 acceptance criteria. The values shown in Table 4-2 for MAPLHGR and PLHGR are used for the Nominal and Appendix K analyses. The axial power shape is varied for each analyzed power /

flow condition to place the hot bundle on the PLHGR limit while the bundle power is on the MCPR limit.

4.3 ECCS PARAMETERS The Hope Creek SAFER/GESTR-LOCA analysis incorporates values for the ECCS performance parameters that are consistent with the current Technical Specifications. Table 4-3 shows a summary of specific performance input parameters used in the analysis. Table 4-3 is applied to all fuel types and initial conditions.

4-1

NEDO-33153 Table 4-1 PLANT PARAMETERS USED IN HOPE CREEK SAFER/GESTR-LOCA ANALYSIS Plant Parameters Nominal Appendix K Core Thermal Power (MWt) 3339 3430 Corresponding Power (% of 3339 MWt) 100 102 Vessel Steam Output (Ibm/hr) 14.404 x 106 14.887 x 106 Rated Core Flow (Ibm/hr)(') 100 x 106 100 x 106 Vessel Steam Dome Pressure (psia) 1020 1055 Maximum Recirculation Suction Line 4.08(2) 4.08(2)

Break Area (ft2 )

(I) The break spectrum determination of worst single failure was performed at rated core flow of 100 Mlb/hr. The limiting LOCA cases were analyzed for a core flow range of 76.6 Mlbfhr to 105 Mlb/hr (76.6% to 105%) of rated core flow at 3339 and 3430 MWt core power.

(2) Includes area of bottom head drain.

4-2

NEDO-33 153 Table 4-2 FUEL PARAMETERS USED IN HOPE CREEK SAFER/GESTR-LOCA ANALYSIS Anal)sis Value Fuel Parameter GE14 SVEA-96+

PLHGR (kW/ft)

- Appendix K (( See Note 3

- Nominal ))

MAPLHGR (kW/ft)

- Appendix K 12.82x1 .02 See Note 3

- Nominal 12.24 Worst Case Pellet Exposure (l) See Note 3 (MWd/MTU) See__Note_3 Initial Operating MCPR

- Analysis Limit 1.25 1.25

- Appendix K 1.25 + 1.02 1.25 *.1.02

- Nominal 1.25 + 0.02 1.25 + 0.02 Number of Fuel Rods per Bundle (2) 92 96 Notcs: (1) This is the exposure at the knee in the PLHGR curve for each fuel. It represents the limiting operating condition resulting in the maximum calculated PCT at anytime during the fuel bundle life.

(2) GEI4 (lOxlO) has 2 water rods occupying a 8-rod space. SVEA-96+ (loxlO) has no water rods, but has a water channel occupying a 4-rod space.

(3) The PLHGR curve for SVEA-96+ fuel is Westinghouse Proprietary and cannot be included. The SVEA-96+ analysis was performed at the exposure with the highest PLHGR.

4-3

NEDO-33 153 Table 4-3 HOPE CREEK SAFERIGESTR-LOCA ANALYSIS ECCS PARAMETERS 1 Lowv Pressure Coolant Injection (LPCI) System Analysis Variable Units Value

a. Maximum vessel pressure at which pumps can inject flow psid (vessel to 286 drywvell)

b. Minimum flowv to reactor vessel with minimum flow bypass valve open

Vessel pressure at which below listed flow rates are psid (vessel to 20 quoted drywell)

I LPCI pump gpm 900ooo"

2 LPCI pumps gPm I80000"'

3 LPCI pumps gpm 270007l)

4 LPCI pumps gpm 3600001)

c. Minimum flow to reactor vessel at 0 psid with minimum flow valve open

1 LPCI pump gpm 10600(l)

2 LPCI pumps gpm 220012)

3 LPCI pumps gpm 31800(')

4 LPCI pumps gpm 424000)

d. Initiating Signals

Low water level (LI) inches (above 378.5 vessel zero) . -

Or

High drywell pressure psig 2.0

e. Vessel pressure at which injection valve may open psig 360

f. Time from initiating signal (Item Ld) to system capable of sec 40(2) delivering full flow (power available, pump at rated speed, and injection valve fully open)

g. Injection valve stroke time-opening sec 242)

(°) These flow rates assume the minimum flow bypass valve does not close. Flow rates are increased by 1000 gpm per pump when the bypass valve closes. These flow rates to the vessel are reduced by 80 gpm to account for leak-age.

(2) This does not include signal processing delay time (I sec).

4-4

NEDO-33 153 Table 4-3 (cont,)

HOPE CREEK SAFER/GESTR-LOCA ANALYSIS ECCS PARAMETERS

2. Core Spray (CS) System Analysis Variable Units Value

a. Maximum vessel pressure at which pumps can inject flow psid (vessel to 289 drywell)

b. Minimum flow to reactor vessel

. Vessel pressure at which below listed flow rate is psid (vessel to 105 quoted drywvell) _ __

. Minimum flow of one core spray loop gpm 5650"'

c. Minimum flow of one core spray loop at 0 psid gpm 7000(2)

d. Initiating Signals . -

. Low water level (LI) inches (above 378.5 vessel zero) or

. High dry well pressure psig 2.0

e. Vessel pressure at which injection valve may open psig 425

f. Injection valve stroke time-opening sec 12(l_

g. Time from initiating signal (Item 2.d) to system capable of sec27 delivering full flow (power available, pump at rated speed and injection valve fully open)

° This does not include signal processing delay time (I sec).

(2) The flow rate to vessel is reduced by 100 gpm to account for leakage.

4-5

NEDO-33 153 Table 4-3 (cont,)

HOPE CREEK SAFER/GESTR-LOCA ANALYSIS ECCS PARAMETERS

3. High Pressure Coolant Injection (IIPCI) System Analysis Variable Units Value

a. Operating Vessel Pressure Range psid (vessel to 200 to drywell) 1141

b. Minimum flow required over the entire above pressure gpm 5600 range

c. Minimum rated HPCI flow injected through the core spray gpm 2000 sparger

d. Initiating Signals

. Low water level (L2) inches (above 469.5 vessel zero) or

High drywell pressure psig 2.0

c. Maximum allowable time delay from initiating signal sec 35(i)

(Item 3.d) to system capable of delivering full flow (pump at rated speed and injection valve fully open)

(11 This does not include signal processing delay time (I sec).

4-6

NEDO-33 153 Table 4-3 (cont,)

HOPE CREEK SAFER/GESTR-LOCA ANALYSIS ECCS PARAMETERS

4. Automatic Depressurization System (ADS)

Analysis Variable Units Value

a. Number of ADS valves

Total number of relief valves with ADS function 5

Total number of relief valves with ADS function assumed 5-.

available in the analysis

b. Pressure at which below listed capacity is quoted psig 1125

c. Minimum flow capacity for one ADS valve Ibm/hr 800000

d. Initiating Signals

Low water level (LI) inches (above 378.5 vessel zero).

and

High drywell pressure psig 2.0 or High drywell pressure bypass timer timed out sec 360 and

Confirming signal that I LPCI or LPCS is running

e. Delay time from initiating signal completed to time valves sec 120('

start to open

(°) The small break analyses assume five ADS valves to be functioning, but the ADS sensitivity studies were analyzed assuming four ADS valves are functioning.

(2) This does not include signal processing delay time (I sec).

4-7

NEDO-33 153 Table 4-4 HOPE CREEK SINGLE FAILURE EVALUATION Assumed Failured1 Systems Remaining (2l Channel A DC Source (Battery) I LPCS, 3 LPCI, ADS(3 )

LPCI Injection Valve (LPCI IV) 2 LPCS, 3 LPCI, HPCI, ADS"3 )

Diesel Generator (D/G) I LPCS, 3 LPCI, HPCI, ADS' 3 I HPCI 2 LPCS, 4 LPCI, ADS(3_

Other postulated failures are not specifically considered because they all result in at least as much ECCS capacity as one of the assumed failures.

(2) Systems remaining, as identified in this table, are applicable to all non-ECCS line breaks. For a LOCA from an ECCS line break, the systems remaining are those listed, less the ECCS system in which the break is assumed.

(3) Five ADS valves are assumed for the small break analyses. Four operable ADS valves (one non-functioning ADS in addition to the single failure) are conservatively assumed for large break analyses and a separate small break sensitivity study to determine the impact of an ADS valve out-of-service. Analysis of four ADS valves shows that the ADS failure is bounded by the battery failure.

4-8

NEDO-33 153 DIG

~DIG -- ---- I --

i D/cqLQ I I I DI I I - - ' -l I I

'NOTE: BOTH CORE SPRAY PUMPS IN A SYSTEM MUST OPERATE TO ASSURE ADEQUATE SPRAY DISTRIBUTION Figure 4-1. Hope Creek ECCS Configuration 4-9

NEDO-33 153 4-10

NEDO-33 153 5.0 RESULTS 5.1 BREAK SPECTRUM CALCULATIONS 5.1.1 Recirculation Line Breaks The recirculation line break spectrum was analyzed for the GE14 and SVEA-96+ fuel types using the nominal and Appendix K assumptions and inputs discussed in Section 4.0. The bottom head drain flow path was included in the recirculation line break cases. The results are listed in Table 5-1 and it can be seen that battery failure is the limiting single failure for both large and small breaks. A sufficient number of breaks were analyzed to establish the shape of the PCT versus break area curve (break spectra shown in Figure 5-1 for GE]4 and Figure 5-2 for SVEA-96+). This ensures that the limiting combination of the break size, location, and single failure has been identified and is consistent with that determined in the generic evaluation.

5.1.1.1 Nominal Calculations The nominal assumptions used in the analysis are listed in Table 3-1. Table 5-1 is a summary of the results. The resulting PCTs, plotted for the break spectra in Figures 5-1 and 5-2, show that nominal PCT decreases with decreasing break size to the 0.5 ft2 range, which is consistent with the trends observed in the generic break spectra, Reference 2. In the large break range, the cladding temperature histories show two peaks during the heatup period. The first peak is due to early transition to film boiling (dryout) and is not sensitive to differences in break sizes. The second peak temperature is caused by core uncovery. Except for DBA break with SVEA-96+

fuel, which has a low 15' peak PCT, the nominal PCTs for the large breaks (>I ft2 ) are 1 'tpeak limited; the 2nd peak PCT is strongly dependent on the ECCS performance. The dryout times were calculated for DBA suction break for both GE14 and SVEA-96+ fuels. The dryout times for other large break sizes were estimated based on the DBA dryout times adjusted for the smaller break sizes. No adjustment for penetration of the early boiling transition is made. This approach results in conservative estimation of the dryout times for non-DBA large breaks. The PCTs for the recirculation suction line breaks with nominal conditions are shown in Table 5-1.

Most of the break sizes in Table 5-1 are analyzed with a LPCI flow rate that assumes the minimum flow bypass valve closes. The limiting breaks are also analyzed with reduced LPCI flow that assumes the minimum flow bypass valve remains open. f[

5-1

NEDO-33 153 For small breaks (< 1.0 ft2 ), ECCS injection depends on reactor depressurization due to initiation of the Automatic Depressurization System (ADS). The highest calculated PCT in the small break range occurs near 0.1 ft2. The calculated PCT decreases as the break size increases above the limiting small break and decreases as the break size decreases below the limiting small break size. For small breaks that do not experience early film boiling, the cladding heatup occurs due to core uncovery.

)) The system response time histories for selected nominal cases are plotted in Appendix A.

5.1.1.2 Appendix K Calculations Appendix K assumptions used in the analysis are listed in Table 3-2. Using the Appendix K input assumptions; DBA analyses with battery failure are performed for GE14 and SVEA-96+

fuels. Three large break sizes (100%, 80% and 60% DBA) and the limiting small break wvere analyzed using the Appendix K assumptions. This is intended to examine the sensitivity of Appendix K PCT to break size and to assure that the limiting break is consistent with the generic Appendix K results. The analysis of these three large break cases satisfies the Appendix K requirement for use of the Moody Slip Flow model with three discharge coefficients of 1.0, 0.8 and 0.6 (Table 3-2).

((

)) The results of the Appendix K analyses are 5-2

NEDO-33 153 also shown in Table 5-1, and the plotted system response time histories for selected cases are plotted in Appendix B.

5.1.2 Non-Recirculation Line Breaks Non-recirculation line breaks were analyzed for both GE14 and SVEA-96+ fuels using nominal assumptions with battery failure. The results of these analyses (Table 5-2) show that these postulated breaks are significantly less severe than the postulated recirculation line breaks (Table 5-1).

5.2 COMPLIANCE EVALUATIONS 5.2.1 Licensing Basis PCT Evaluation The Hope Creek Appendix K results confirm that the limiting DBA break is the recirculation suction line, which is consistent faith the BWR-5/6 generic conclusions and demonstrate that the battery failure is limiting for all fuels. [

1]

The Licensing Basis PCTs for Hope Creek are calculated for SVEA-96+ and GE14 fuel types based on the above Appendix K PCTs using the methodology described in Section 3.2 at the MELLLA condition and assuming the LPCI bypass valve does not close. Hope Creek unique variable uncertainties, including backflow leakage, ECCS signal, stored energy, gap pressure, and ADS time delay, were evaluated for both fuel types to determine plant-specific adders. The calculated licensing Basis PCTs are 1370'F for GE]4 and 1540TF for SVEA-96+.

5.2.2 Removal of the Current Requirement for Evaluation of Upper Bound PCT The NRC SER approving the original SAFER/GESTR-LOCA application methodology (described in Reference 2) placed a restriction of 1600'F on the Upper Bound PCT calculation.

Additional supporting information was needed to support the use of the methodology for Upper Bound PCTs in excess of this limit. GENE provided this information on a generic basis in Reference 8. GENE received an SER from the NRC (Reference 7) eliminating the 16000 F restriction on the Upper Bound PCT. The elimination of the restriction on the Upper Bound PCT is applicable to all plants using the SAFER/GESTR-LOCA application methodology described 5-3

NEDO-33 153 in Reference 2, including Hope Creek. In addition, the 16001F restriction on the Upper Bound PCT is no longer applicable when evaluating the effect of changes and errors reported under the requirements of I OCFR50.46.

Plant-specific Upper Bound PCT Calculation The primary purpose of the Upper Bound PCT calculation is to demonstrate that the Licensing Basis PCT is sufficiently conservative by showing that the Licensing Basis PCT is higher than the Upper Bound PCT. The NRC SER approving the SAFER/GESTR-LOCA application methodology also required confirmation that the plant-specific operating parameters have been conservatively bounded by the models and inputs used in the generic calculations. The SER also required confirmation that the plant-specific ECCS configuration is consistent with the referenced plant class ECCS configuration for the purpose of applying the generic LTR Upper Bound PCT calculations to the plant-specific analysis. Because of the wide variation in plant specific operating parameters and ECCS performance parameters within the BWR product lines, it is difficult to judge whether an individual plant is bounded by the generic calculations.

Therefore, the practice has been to calculate the Upper Bound PCT on a plant-specific basis rather than rely on the generic Upper Bound PCT calculations in order to demonstrate that the Licensing Basis PCT is sufficiently conservative.

Reference 8 provided generic justification that the Licensing Basis PCT will be conservative with respect to the Upper Bound PCT and that the plant-specific Upper Bound PCT calculation was no longer necessary. The NRC SER in Reference 7 accepted this position by noting that because plant-specific Upper Bound PCT calculations have been performed for all plants, other means may be used to demonstrate compliance with the original SER limitations. These other means are acceptable provided there are no significant changes to the plant configuration that would invalidate the existing Upper Bound PCT calculations. For the purposes of the Upper Bound PCT calculation, the plant configuration includes the plant equipment and equipment performance (e.g., ECCS pumps and flow rates), fuel type, and the plant operating conditions (e.g., core power and flow) that may affect the PCT calculation. In order to demonstrate continued compliance with the original SER limitations, the PCT effect due to the changes in the plant configuration must be reviewed in order to confirm that the conclusions based on the original Upper Bound PCT calculation have not been invalidated by the changes.

5-4

NEDO-33 153

))

As demonstrated in the discussions above, the Upper Bound is no longer restricted by the 1600'F limit. Therefore, when evaluating the effect of changes and errors reported under the requirements of IOCFR50.46, the effect on the Upper Bound PCT no longer needs to be evaluated.

5-5

NEDO-33 153 5.3 EXPANDED OPERATING DOMAIN AND ALTERNATE OPERATING MODES Extended operating domains and alternate operating modes are presented as sensitivity studies to the break spectrum analyses performed at rated conditions. Only the limiting DBA recirculation line break/failure combination is analyzed using nominal and Appendix K assumptions. The limiting break/failure combination is usually not affected by changes in the power / flow conditions.

5.3.1 Increased Core Flow (ICF) 5.3.2 Reduced Core Flow (MELLLA / ELLLA)

Although the plant is currently operating in the Extended Load Line Limit (ELLL) region, it is expected to transition to the Maximum Extended Load Line Limit (MELLL) region. The rod line in the ELLL region permits reactor operation at 90.9% of rated core flow for the rated power of 3339 MWt. The higher rod line in the MELLL region permits reactor operation at 76.6% of rated core flow for the rated power of 3339 MWt. For the low core flow portion of the MELLL region, boiling transition at the high power fuel nodes can occur sooner than at the rated core flowa conditions. This phenomenon is referred to as early boiling transition (EBT). If EBT occurs for the higher power node as a result of the reduced initial core flow, the resulting PCT can exceed the corresponding results for the rated core flow. Low core floxv effects on the ECCS analyses were generically addressed in Reference 9, which wvas approved by the NRC in Reference 10. These studies demonstrated that no MAPLHGR multiplier wvas required for low core flow operation for the BWR-5/6 plant class, which has ECCS similar to Hope Creek. The SAFER/GESTR-LOCA analysis for low core flow conditions in the MELLL region was evaluated for Hope Creek using the same ECCS inputs as used for the rated core flow conditions.

)) The analysis was 5-6

NEDO-33 153 performed with both nominal and Appendix K assumptions. The results are shown in Table 5-3 with rated core flow results presented for comparison.

1[

))

5.3.3 Single-Loop Operation (SLO)

))

This evaluation is intended to address SLO analysis as it relates to LOCA; other SLO analyses are not covered here.

The ECCS performance for Hope Creek under SLO was evaluated using SAFER/GESTR-LOCA for the DBA break with battery failure. ((

5-7

NEDO-33 153

)) The calculated results for Hope Creek under SLO are shown in Table 5-4.

5.4 M1APL11GR LIMITS The SAFER/GESTR-LOCA analysis was performed with a bounding Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) at the most limiting combination of power and exposure for each analyzed fuel type (SVEA-96+ and GE14). The ECCS-based exposure dependent MAPLHGR limits are determined on a fuel type bases.

Although the analyses does not credit any reductions in LHGR or MAPLHGR during two-loop operation, application of either the APRM setpoint requirements or the ARTS based fuel thermal-mechanical design analysis limits [LHGRFAC(p) I 1,HGRFAC(f) or MAPFAC(p) /

MAPFAC(f)] are required to ensure that off-rated conditions not specifically analyzed will not be limiting.

In Single Loop Operation, specific multipliers on PLHGR and MAPLHGR are required. The SLO multiplier is independent of the two-loop limits. The SLO multiplier is applicable to all fuel rod exposures.

5-8

NEDO-33 153 Table 5-1

SUMMARY

OF HOPE CREEK SAFER/GESTR-LOCA RESULTS FOR RECIRCULATION LINE BREAKS"'l

((

__ I __ I __

I Il

__ I_ _I _ __

))

5-9

NEDO-33 153 Table 5-1 (continued)

SUMMARY

OF HOPE CREEK SAFER/GESTR-LOCA RESULTS FOR RECIRCULATION LINE BREAKS("'

II -

__I_

_ 1 _ _

))

5-10

NEDO-33153 Table 5-2

SUMMARY

OF HOPE CREEK SAFER/GESTR-LOCA RESULTS FOR NON-RECIRCULATION LINE BREAKS(')

(Nominal Analysis Basis)

((

I.

5-11

NEDO-33 153 Table 5-3 MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS RESULTS COMPARISON FOR HOPE CREEK"l)

LIMITING LOCA: DBA - Recirculation Suction Line Break ll I _ _ __ __ II__

I]

5-12

NEDO-33 153 Table 5-4 SINGLE LOOP OPERATION RESULTS COMPARISON FOR HOPE CREEK DBA Suction Break Analysis Basis Parameter GE14 SN'EA-96+

Nominal Two-Loop Operation PLHGR & ((

Battery Failure MIAPLHGR Alultplier PCT (0 F)

Single Loop Operation PLH1GR &

MAPL1IGR Battery Failure Multplier PCT (0 F)

Appendix K Two-Loop Operation PLIIGR &

Battery Failure MIAPLHGR Multplier PCT (0 F)

Single Loop Operation PL-IGR &

MIAPLIiGR Battery Failure 'Multplier PCT (0 F) 5-13

NEDO-33 153 1]

Figure 5-1. Nominal and Appendix K LOCA Break Spectrum Results for GE14 Fuel 5-14

NEDO-33 153

))

Figure 5-2. Nominal and Appendix K LOCA Break Spectrum Results for SVEA-96+ Fuel 5-15

NEDO-33 153

6.0 CONCLUSION

S LOCA analyses have been performed for Hope Creek using the GE SAFER/GESTR-LOCA Application Methodology approved by the NRC. These analyses were performed to demonstrate conformance with I0CFR50.46 and Appendix K, and thus, support a revised licensing basis for Hope Creek with the GE SAFER/GESTR-LOCA methodology.

As the SAFER/GESTR-LOCA results presented in Section 5 indicate, a sufficient number of plant-specific PCT points have been evaluated to establish the shape of both the nominal and Appendix K PCT versus break size curves. The analyses demonstrate that the limiting Licensing Basis PCT occurs for the recirculation suction line break DBA with Battery failure at MELLLA conditions.

Table 6-1 summarizes the key SAFER/GESTR licensing results for Hope Creek. The analyses presented are performed in accordance with NRC requirements and demonstrate conformance with the ECCS acceptance criteria of IOCFR50.46 as shown in Table 6-1. Therefore, the results documented in this report may be used to provide a new LOCA Licensing Basis for Hope Creek.

The thermal limits applied to the GEI4 and SVEA-96+ fuel types in the ECCS-LOCA evaluation are summarized in Table 6-2.

6-1

NEDO-33 153 Table 6-1 SAFER/GESTR-LOCA LICENSING RESULTS FOR HOPE CREEK SAFER/GESTR-LOCA LICENSING RESULTS ACCEPTANCE Parameter CRITERIA

1. Limiting Break DBA (Recirculation Suction Line)

2. Limiting ECCS Failure Battery

3. Fuel Type GE14 SVEA-96+

4. Peak Cladding 1370 1540 < 22000 F Temperature (Licensing Basis)

6. Maximum Local <1% <1% <17%

Oxidation

7. Core-Wide Metal-Water <0.1% <0.1% <1%

Reaction

8. Coolable Geometry Items 4 & 6 PCT < 22007F and Local Oxidation <

_ _ _ __ __ __ __ _ _ _ _ _ _ _17 %

9. Long-Term Cooling Core reflooded above Top Core temperature of Active Fuel (TAF) acceptably low and or long-term decay heat Core reflooded to the top of removed; met by Core the jet pump, suction and reflooded above Top one Core Spray system in of Active Fuel (TAF) operation or Core reflooded to the top of the jet pump suction and one Core Spray system in operation 6-2

NEDO-33 153 Table 6-2 Thermal Limits Analysis Limit Parameter GE14 SV'EA-96+

PLHGR - Exposure Limit Curve GWD!MT kW/ft GWD/MT l W/ft R Note I Note I Note I Note I Note I Note I j] Note I Note I MAPLHGR - Exposure Limit Curve GWD!MT kW/ft GWD/MT kWMft 0 12.82 Note I Note I 21.09 12.82 Note I Note I 63.50 8 Note I Note 1

_ 70.00 5 Note I Note I Initial Operating MCPR 1.25 1.25 Minimum R-Factor [](( ))

SLO Multiplier on PLHGR & MAPLHGR 0.80 0.80 Notes: (I) The PLHGR curve for SVEA-96+ fuel is Westinghouse Proprietary and cannot be included. The SVEA-96+ analysis was performed at the exposure with the highest PLHGR. The MAPLHGR curve for SVEA-96+ is based on the LHGR curve, so it is omitted.

6-3

NEDO-33 153 6-4

NEDO-33153

7.0 REFERENCES

1) Letter, C.O. Thomas (NRC) to J.F. Quirk (GE), "Acceptance for Referencing of Licensing Topical Report NEDE-23785, Revision 1, Volume III (P), 'The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident," June 1, 1984.

2) "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume III, SAFER/GESTR Application Methodology," NEDE-23785- I-PA, General Electric Company, Revision 1, October 1984.

3) "General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K," NEDO-20566A, General Electric Company, September 1986.

4) "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident; Volume I, GESTR-LOCA - A Model for the Prediction of Fuel Rod Thermal Performance," NEDC-23785-2-P, General Electric Company, May 1984.

5) "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.

6) "Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model,"

NEDC-32950P, January 2000.

7) Stuart A. Richards (NRC) to James F. Klapproth (GENE), Review of NEDE-23785P, Vol. III, Supplement 1, Revision 1, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1, Additional Information for Upper Bound PCT Calculation," (TAC No. MB2774), February 1, 2002.

8) "GESTR-LOCA and SAFER Models for Evaluation of Loss-of Coolant Accident, Volume III, Supplement I - Additional Information for Upper Bound PCT Calculation, NEDE-23785P-A, Supplement 1, Revision 1, March 2002.

9) Letter, R. L. Gridley (GE) to D. G. Eisenhut (NRC), "Review of Low-Core Flow Effects on LOCA Analysis for Operating BWRs - Revision 2," May 8, 1978.

10) Letter, D. G. Eisenhut (NRC) to R. L. Gridley (GE), "Safety Evaluation Report on Revision of Previously Imposed MAPLHGR (ECCS-LOCA) Restriction for BWRs at Less Than Rated Core Flow," May 19, 1978.

7-1

NEDO-33 153 7-2

NEDO-33 153 APPENDIX A SYSTEM RESPONSE CURVES FOR NOMINAL RECIRCULATION LINE BREAKS Included in this Appendix are the system response curves for Hope Creek. Table A-I shows the figure numbering sequence for the nominal recirculation breaks.

A-l

NEDO-33 153 Table A-I NOMINAL RECIRCULATION LINE BREAK FIGURE

SUMMARY

Notes: All Plots are for GEI4 fuel, except when noted.

All the figures are for the analyses performed with the LPCI bypass valve closed except where noted as "Low LPCI."

- Recirc. Break DBA 80% DBA 60% DBA 1T0 fje [

- Single Failure Battery Battery Battery Battery Low LPCI Water Level in A-la A-2a A-3a A-4a A-5a Hot & Average Channels Reactor Vessel A-lb A-2b A-3b A-4b A-5b Pressure Peak Cladding A- Ic,f* A-2c,f* A-3c,f* A-4c,f* A-5c,f*

Temperature Heat Transfer A-ldg* A-2d,g* A-3d,g* A-4d,g* A-5d,g*

Coefficient ECCS Flow A-le A-2e A-3e A-4e A-5e

Plots for GE14 and SVEA-96+ are included.

A-2

z 0

ow (j,'

Figure A-la. Water Level in Hot and Average Channels - DBA Suction - Battery Failure (Nominal) -

U-~ 3LPCI+LPCS+ADS Available

z (I

to 6 Figure A-lIb. Reactor Vessel Pressure - DI3A Suction - Battery Failure (Nominal) - 3LPC1+LPCS+ADS Available

z Di 0

'-I LtA 3LPCI+LPCS+ADS Figure A-Ic. Peak Cladding Temperature (GE1 4)- DBA Suction - Battery Failure (Nominal)

Available

z M

t, Failure (Nominal) - 3LPCI+LPCS+ADS Figure A-Id. FlTeat Transfer Coefficient (GEI4) - DBA Suction - Battery Available

zMr 0

LA3

.-. j Figure A-lc. ECCS Flow - DI3A Suction - B~attcry Pailurc (Nominal) - 3LPCI+LPCS+ADS Available

z 00 U' Figure A-If. Peak Cladding Temperature (SVEA-96+) -DBA Suction -Battery Failure (Nominal) -3LPCI-FLPCS+ADS

6. Available

z a:

0 Ij Lh Failure (Nominal) - 3LPCI+LPCS+ADS Figure A- 1g. Heat Transfer Coefficient (SVEA-96+) - D1BA Suction - Battery 6- Available

-9

-9 ZI z

>o -F0

.LA IjJ Battery Failure (Nominal) - 3LPCT+LPCS+ADS Figure A-2a. Water Level in Hot and Average 80% DI3A Suction -

6-dAvailable

zm 9

Lw

.A) w Figure A-2b. Reactor Vessel Pressure - 80% DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available

z 80% DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Figure A-2c. Peak Cladding Temperature (GE14) -

Available

3z 0

I w

(Nominal) - 3LPCI+LPCS+ADS Figure A-2d. Heat Transfer Coefficient (GE14) - 80% DBA Suction - Battery Failure Available

z 9

O 0e 6 Figure A-2c. ECCS Flow - 80% DBA Suction - Battcry Failure (Nominal) - 3LPCI+LPCS+ADS Available

z 0

I _

h w

(-A Figure A-2f. Peak Cladding Temperature (SVEA-96+) - 80% DBA Suction - Battery Failure (Nominal) -

3LPCI+LPCS+ADS Available

z m0 c ~I1 w

- 80% DBA Suction - Battery Failure (Nominal)

L Figure A-2g. Heat Transfer Coefficient (SVEA-96+)

3LPCI+LPCS+ADS Available

-9 w

Channels - 60% DBA Suction - Battery Failure (Nominal) and Average Figure A-3a. Water Lcvel in I-lot 3LPCI+LPCS+ADS Available

z

> p o 0 Figure A-3b. Reactor Vessel Pressure - 60% DBA Suction - Battery Failure (Nominal) -

. ILPCS + 2LPCI + ADS Available

z97-r Battery Failure (Nominal) -

Figure A-3c. Peak Cladding Temperature (GE 14) - 60% DBA Suction -

I1LPCS + 2LPCI + ADS Available

z 0

t t

_ Figure A-3d. Heat Transfer Coefficient (GE14)- 60% DBA Suction - Battery Failure (Nominal) -

I LPCS + 2LPCI + ADS Available

rri tz Fta 3LPCI+LPCS+ADS Available 60% DBA Suction - B~attery Failure (Nominal)

Figur -

Figure A-3e. ECCS Flow -

z Tri w

bqe.j (Nominal) -

- Battery Failure (SVEA-96+) - 60% DBA Suction A-3f. Peak Cladding Temperature Available Figure +ADS ILPCS + 2LPCI

z Ml w

~w w

(Nominal) -

Coefficient (SVEA-96+) - 60% DBA Suction - Battery Failure Figure A-3g. Heat Transfer ILPCS + 2LPCI + ADS Available

z qjj (Nominal) - 3LPCI+LPCS+ADS

- I ft' Suction - Battery Failure in Hot and Average Channels Figure A-4a. Water Level 6 Available

z rI 0

(A w

w

_ Figure A-4b. Reactor Vessel Pressure - 1 ft2 Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available

z M

>~00 (ON LAP QSuction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Figure A-4c. Peak Cladding Temperature (GE 14) -

Available

z M

0 ien

Figure A-4d. H-eat Transfer Coefficient (GE 14) -1f 2 Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available

z 0

LA kJJ t'j W 00 Available

- Figure A-4c. ECCS Flow - I £12 Suction - Battery Failure (Nominal) - 31PCI+LPCS+ADS

zM C:

> L~0

'0 LA Figure A-4f. Peak Cladding Temperature (SVEA-96+) - 1 ft2 Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available

1-9 9

S>

00 LA~

__-I Figure A-4g. Ileat Transfcr Coefficient (SVEA-96+) - I ft2 Suction - Battery Failure (Nominal) - 3LPCI+IPCS+ADS Available

m 9

0 t7

!o tA Uj Figure A-Sa. Water Level in Hot and Average Channels -

z 9Ln t'j

'-I Figure A-5b. Reactor Vessel Pressure -

z M

U I

'jJ (GE14) -

Figure A-5c. Peak Cladding Temperature

T-9

~-

z

> C" p

I Figure A-5d. Heat Transfer Coefficient (GE 14) -

-9 zm 0

p

!0 Wn LA (AI

_ Figurc A-5e. ECCS Flow -

-- 9 r-9 Z~

9 Figure A-5f. Peak Cladding Temperature (SVEA-96+) -

j

z LA.

Fw Figure A-5g. Heat Transfer Coefficient (SVEA-96+) -

NEDO-33 153 APPENDIX B SYSTEM RESPONSE CURVES FOR APPENDIX K RECIRCULATION LINE BREAKS Included in this Appendix arc the system response curves for Hope Crcek. Table B-I shows the figure numbering sequence for the Appendix K recirculation breaks.

B-I

NEDO-33 153 Table B-I APPENDIX K RECIRCULATION LINE BREAK FIGURE

SUMMARY

Note: All Plots are for GE14 fuel, except when noted.

All the figures are for the analyses performed with the LPCI bypass valve closed except where noted as "Low LPCI."

- Recirc. Break DBA Suction - DBA Suction - 80% DBA 60% DBA

- Single Failure Rated - Battery MELLLA - Suction - Suction -

Low LPCI Battery Battery Failure Battery Failure Low LPCI Water Level in B-la B-2a B-3a B-4a B-5a Hot & Average Channels Reactor Vessel B-lb B-2b B-3b B-4b B-5b Pressure Peak Cladding BI1c,h* B-2c,f* B-3c,f* B-4c,f* B-5c,f*

Temperature Heat Transfer B-Idi* B-2d,g* B-3d,g* B-4d,g* B-5d,g*

Coefficient ECCS Flow B-le B-2e B-3e B-4e B-5e Core Inlet Flow B-If MCPR B-1g

Plots for GE14 and SVEA-96+ are included.

B-2

z cr1 tz 0 t~a t9 (I

(.3

.- j Figure B-la. Water Level in F-lat and Average Channels (01314) - DBA Suction - Rated - Battery Failure (App. K) -

3LPCI+/-LPICS+ADS Available

ztz 9

w

(.bJ Figure B- Ib. Reactor Vessel Pressure (GE 14) - DBA Suction - DBA Suction - Rated - Battery Failure (App. K) -

3LPCI+LPCS+ADS Available

rn oz uJ Rated Battery Failure (App. K) -

Figurc B-Ic. Peak Cladding Temperature (GE14) - DBA Suction 3LPCI+LPCS+ADS Available

z Figure B-Ild. Hecat Transfer Coefficient (GE 14) - DI3A Suction - Rated - Battery Failure (App. K)- 3LPCJ+LPCS+ADS Available

z L~Lh Figure B- I . ECCS Flow (GEI14) - DBA Suction - Rated - Battery Failure (App. K) - 3LPCI+LPCS+ADS Available

z rnI 03 tz Figure B-If. Core Average Inlet Flow (GE14) - DBA Suction - Rated - Battery Failure (App. K)-

3LPCI+LPCS+ADS Available

z p

to

'0 w

ILA

- Rated Battery Failure (App. K)-

Figure B- 1g. Minimum Critical Power Ratio (GE14 & SVEA-96+) - DBA Suction 3LPCI+LPCS+ADS Available

zm 9

I o

U-C, 6 Figure 13-lh. Peak Cladding Temperature (SVEA-96+) - DBA Suction - Rated - Battery Failure (App. K)-

3LPCI+LPCS+ADS Available

zm m

.. 9 t~o Itj 6--i

- Battery Failure (App. K) -

Figure B- l i. Heat Transfer Coefficient (SVEA-96+) - DBA Suction - Rated 3LPCI+LPCS+ADS Available

z 0

0 mS 1

- MELLLA - Battcry Failure (App. K) -

_ Figure B-2a. Water Level in I-lot and Average Channels - DBA Suction 3ILPCI+LPCS+ADS Available

z 9

tz kAJ Battery Failure (App. K) - 3LPCI+LPCS+ADS 6 ~Figure B-2b. Reactor Vessel Pressure- DBA Suction - MIELLLA -

Available

-9 z

wm L-LA Figure B-2c. Peak Cladding Temperature (GE 14) - DBA Suction - MELLLA - Battery Failure (App. K)-

3LPCI+LPCS+ADS Available

m oz tz 0 c-f t.w Figure B-2d. [Teat TransferCoefficient (GE14) - DBA Suction - MELLLA - Battery Failure (App. K) -

3LPCI+LPCS+ADS Available

Z 9

C) 0 kw

.fI ti

- Battery Failure (App. K) - 3LPCI+LPCS+ADS Available Figure B-2e. ECCS Flow - DBA Suction - MELLLA

z 9

W

-J-(App. K)-

Figure B-2f. Peak Cladding Temperature (SVEA-96+) - DBA Suction - MELLLA - Battery Failure 3LPCI+LPCS+ADS Available

z 0

O 0

tfA 6-LtJ

-j Figure B-2g. H-eat Transfer Coeff icient (SVEA-96+) - DBA Suction - MELLLA - Battery Failure (App. K) -

3LPCI+LPCS+ADS Available

ozrnI I-to 0t:

tA Figure B-3a. Water Level in Hot and Average Channels - 80% DBA Suction -Battery Failure (App. K) -

3LPCI+LPCS+ADS Available

z7 m

p0 u3 co tN)

C) (If tA Figure B-3b. Reactor Vessel Pressure - 80% DBA Suction -Battery Failure (App. K) - 3LPCI+LPCS+ADS Available

z m

Figure B-3c. Peak Cladding Temperature (GE14) - 80% DBA Suction -Battery Failure (App. K) - 31,PCI+LPCS+ADS Available

-9 z

Ml 0

6--tJ (App. K) - 3LPCI+LPCS+ADS Figure B-3d. Heat Transfer Coefficicnt (GE14) - 80% DBA Suction -Battery Failure Available

z tk) 9 r"ti

(~3 Figure B-3e. ECCS Flow - 80% DBA Suction -Battery Failure (App. K) - 3LPCI+LPCS+ADS Available

~-

z LA Figure B-3f. Peak Cladding Temperature (SVEA-96+) - 80% DBA Suction -Battery Failure (App. K) -

3LPCI+LPCS+ADS Available

zm 9

rT w %

tA U' 6 Figure B-3g. Heat Transfer Coefficient (SVEA-96+) - 80% DBA Suction -Battery Failure (App. K) -

3LPCI+LPCS+ADS Available

CI 0

w t!,) w (App. K) -

Figure B-4a. Water Level in I-lot and Average Channels - 60% DBA Suction -Battery Failure 3LPCI+LPCS+ADS Available J0

-9

-9 z

rn 0M

-J

'I VIJ 6 Figure B-4b. Reactor Vessel Pressure - 60% DBA Suction -Battery Failure (App. K) - 3LPCI+LPCS+ADS Available

zrr C9 W

t~U.)

60% DBA Suction -Battery Failure (App. K) - 3LPCl+LPCS+ADS 6 Figurc B-4c. Peak Cladding Temperature (GE 14) -

Available

6-9 ~w Figure B-4d. fHeat Transfer Coefficient (GE14) - 60% DBA Suction -Battery Failurc (App. K)-3LPCI+LPCS+ADS Available

z 9

an L(A 6.J Figure B-4e. ECCS Flowv - 60% DI3A Suction -Battery Failure (App. K) - 3LPCI+LPCS+ADS Available

z rrl up Figure B-4f. Peak Cladding Temperature (SVEA-96+) - 60% DI3A Suction -Battery Failure (App. K) -

3LPCI+LPCS+ADS Available

wz 0

t~j

~we LA IFigure B-4g. Heat Transfer Coefficient (SVEA-96+) - 60% DBA Suction -Battery Failure (App. K) -

6-J 6_J 3LPCI+LPCS+ADS Available

zm CZ to w

to to w

('

6Figure B-5a. Water Level in Hot and Average Channels -

143 r14 Fiue135.Reco ese resra

-4

z rC) tl

'jJ 0

LA w LA 6 Figure B-5c. Peak Cladding Temperature (GE14) -

zrri 0

w p tz (ah

_ Figure B-5d. Heat Transfer Coefficient (GE 14) -

z CZ 0 W W LFF Figure 13-5c. I3CCS Flow -

A

I-"

z wz 00 t~3 9 0

L PiA 6-j Figure 13-5f. Peak Cladding Temperature (SVEA-96+) -

I

z rrI tz w

F r - H0 U)

Figure B-5g. Heat Transfer Coefficient (SVEA-96+) -

A

LR-N04-0378, NEDO-33153, SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for Hope Creek Generating Station.

Contents

Text

SUMMARY

1.0 INTRODUCTION

2.0 DESCRIPTION

6.0 CONCLUSION

7.0 REFERENCES

1.0 INTRODUCTION

2.0 DESCRIPTION

SUMMARY

SUMMARY

SUMMARY

6.0 CONCLUSION

7.0 REFERENCES

SUMMARY

SUMMARY

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LR-N04-0378, NEDO-33153, SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for Hope Creek Generating Station.

Text

SUMMARY

1.0 INTRODUCTION

2.0 DESCRIPTION

6.0 CONCLUSION

7.0 REFERENCES

1.0 INTRODUCTION

2.0 DESCRIPTION

SUMMARY

SUMMARY

SUMMARY

6.0 CONCLUSION

7.0 REFERENCES

SUMMARY

SUMMARY

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