ML053250466
| ML053250466 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/31/2005 |
| From: | Tuttle J General Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| DRF 0000-0026-3882, LCR H05-01, LR-N05-0328 NEDO-33172 | |
| Download: ML053250466 (172) | |
Text
Attachment 2 LR-N05-0328 LCR H05-01
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'IA.
NEDO-33172 SAFERJGESTR-LOCA Loss of Coolant Analysis for Hope Creek Generating Station at Power Uprate
GE Energy, Nuclear 175 CurtnerAvenue San Jose, CA 95125 NEDO-33 172 DRF 0000-0026-3882 Class I March 2005 SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for 1Hope Creek Generating Station at Power Uprate Approved by:
Jeff Tuttle Project Manager
NEDO-33 172 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, Gcncral Elcctric Company, 2005.
NEDO-33 172 TABLE OF CONTENTS Paie
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 2-1 3.0 ANALYSIS PROCEDURE 3-1 3.1 Licensing Criteria 3-1 3.2 SAFER/GESTR-LOCA Licensing Methodology -1 3.3 Generic Analysis 3-3 3.4 Hope Creek Plant-Specific Analysis 3-3 3.5 Analysis of Mixed Cores 3-4 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 Break-s 5-1 5.1.2 Non-Recirculation Line Breaks 5-3 5.2 Compliance Evaluations 5-3 5.2.1 LicensingBasisPCTEvaluation 5-3 5.2.2 Removal ofthe Current Requirement forEvaluation 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 5.3.4 Additional Power Conditions 5-7 5.4 MAPIHGR Limits 5-7
6.0 CONCLUSION
S 6-1
7.0 REFERENCES
7-1 LAST PAGE C-37 iii
NEDO-33 172 APPENDICES A SYSTEM RESPONSE CURVES A-i FOR NOMINAL RECIRCULATION LINE BREAKS B SYSTEM RESPONSE CURVES B-i FOR APPENDIX K RECIRCULATION LINE BREAKS C SYSTEM RESPONSE CURVES C-1 FOR NOMINAL NON-RECIRCULATION LINE BREAKS iv
NEDO-33 172 LIST OF TABLES Table Title Paste 3-1 Analysis Assumptions for Nominal Calculations 3-7 3-2 Analysis Assumptions for Appendix K Calculations 3-S 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 Creekl SAFER/GESTR-LOCA Analysis ECCS Parameters 44 44 Hope Creek Single Failure Evaluation 4-8 5-1 Summary of Hope Creek SAFER/GESTR-LOCA Results for Recirculation 5-9 Line Breaks 5-2 Summary of Hope Creek SAFER/GESTR-LOCA Results for Non- 5-11 Recirculation Line Breaks 5-3 Maximum Extended Load Line Limit Analysis Results Comparison for 5-12 Hope Creek 5-4 Single-Loop Operation Results Comparison for Hope Creek 5-13 6-1 SAFER/GESTR-LOCA Licensing Results for Hope Creek 6-2 6-2 Thermal Limits 6-3 A-1 Nominal Recirculation Line Break Figure Summary A-2 B-1 Appendix K Recirculation Line Break Figure Summary B-2 C-1 Nominal Non-Recirculation Line Break Figure Summary C-2 v
NEDO-33 172 LIST OF FIGURES Figure Title Paue 2-] Flow Diagram of LOCA Analysis Using SAFERIGESTR 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 GEM4 Fuel 5-14 5-2 Nominal and Appendix K LOCA Break Spectrum Results for SVEA-96+ 5-15 Fuel A-I DBA Suction - Battery Failure (Nominal) -
3LPCI+ILPCS+ADS Available
- a. Water Level in Hot and Average Channels A-3
- b. Reactor Vessel Pressure A4
- c. Peak Cladding Temperature (GE14) A-5
- d. Heat Transfer Coefficient (GE14) A-6
- e. ECCS Flow A-7
- f. Peak Cladding Temperature (SVEA-96+) A-S
- 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-I I
- c. Peak Cladding Temperature (GE14) A-12
- d. Heat Transfer Coefficient (GE14) A-13
- e. ECCS Flow A-14
- f. Peak Cladding Temperature (SVEA-96+) A-I 5
- 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. ReactorVessel Pressure A-I8
- 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 (GEI4) A-27
- e. ECCS Flow A-28
- f. Peak Cladding Temperature (SVEA-96+) A-29
- g. Heat Transfer Coefficient (SVEA-96+) A-30 vi
NEDO-33 172 LIST OF FIGURES (Continued)
Fieure Title Pace A-5 11
- a. Water Level in Hot and Average Channels A-3 1
- b. Reactor Vessel Pressure A-32
- c. Peak Cladding Temperature (GE14) A-33
- d. Heat Transfer Coefficient (GE14) A-34
- e. ECCS Flow A-35 A-6 1))
- a. Water Level in Hot and Average Channels A-36
- b. Reactor Vessel Pressure A-37
- c. Peak Cladding Temperature (SVEA-96+) A-38
- d. Heat Transfer Coefficient (SVEA-96+) A-39
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
3LPCI+LPCS+ADS Available
- a. Water 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-1 S vii
NEDO-33 172 LIST OF FIGURES (Continued)
Figure Title Paae 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 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 (GEl4) 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 (GE14) B-36
- e. ECCS Flow B-37
- f. Peak Cladding Temperature (SVEA-96+) B-38
- g. Heat Transfer Coefficient (SVEA-96+) B-39 C-1 Core Spray Line Break - Battery Failure (Nominal) -
3LPCJ+ADS Available
- a. Water Level in Hot and Average Channels C-3
- b. Reactor Vessel Pressure C-4
- c. Peak Cladding Temperature (GE14) C-5
- d. Heat Transfer Coefficient (GE14) C-6
- e. ECCS Flow C-7
- f. Peak Cladding Temperature (SVEA-96+) C-8
- g. Heat Transfer Coefficient (SVEA-96+) C-9 viii
NEDO-33 172 LIST OF FIGURES (Continued)
Figure Title Page C-2 Steamline Break Inside Containment -Battery Failure (Nominal)-
3LPCI+LPCS+ADS Available
- a. Water Level in Hot and Average Channels C-10
- b. Reactor Vessel Pressure C-lI
- c. Peak Cladding Temperature (GE14) C-12
- d. Heat Transfer Coefficient (GE14) C-13
- f. Peak Cladding Temperature (SVEA-96+) C-15
- g. Heat Transfer Coefficient (SVEA-96+) C-16 C-3 Steamline Break Outside Containment -Battery Failure (Nominal)-
3LPCI+LPCS+ADS Available
- a. Water Level in Hot and Average Channels C-17
- b. Reactor Vessel Pressure C-I 8
- c. Peak Cladding Temperature (GE14) C-19
- d. Heat Transfer Coefficient (GE14) C-20
- e. ECCS Flow C-21
- f. Peak Cladding Temperature (SVEA-96+) C-22
- g. Heat Transfer Coefficient (SVEA-96+) C-23 C-4 Feedwater Line Break -Battery Failure (Nominal)-
3LPCI+LPCS+ADS Available
- a. Water Level in Hot and Average Channels C-24
- b. Reactor Vessel Pressure C-25
- c. Peak Cladding Temperature (GE14) C-26
- d. Heat Transfer Coefficient (GE]4) C-27
- e. ECCS Flow C-28
- f. Peak Cladding Temperature (SVEA-96+) C-29
- g. Heat Transfer Coefficient (SVEA-96+) C-30 C-5 LPCI Line Break -Battery Failure (Nominal)-
2LPCI+LPCS+ADS Available
- a. Water Level in Hot and Average Channels C-31
- b. Reactor Vessel Pressure C-32
- c. Peak Cladding Tcmperature (GE14) C-33
- d. Heat Transfer Coefficient (GE 14) C-34
- e. ECCS Flow C-35
- f. Peak Cladding Temperature (SVrEA-96+) C-36
- g. Heat Transfer Coefficient (SVEA-96+) C-37 ix
NEDO-33 172 ACKNOWLEDGEMENTS The following individuals contributed significantly toward completion of this report:
D. Abdollabian A. Khan K. Knippel A.J. Lipps J. Stott W-M. Wong x
NEDO-33 172
SUMMARY
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 IOCFR50.46 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 (BWR) 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 10CFR50.46 must include specific inputs documented in Appendix K. The SAFERJGESTR-LOCA Application Methodology requires:
(1) Thle licensing Basis PCT must be less than 22001F. 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 16000 F on the Upper Bound PCT in the Safety Evaluation Report (SER) approving the SAFERIGESTR-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 16001F, otherwise additional plant S-1
NEDO-33172 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 10CFR50.46 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 3840 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 perfonnance analyses are satisfied. The Licensing Basis PCTs for Hope Creek are 1380 0 F for GE14 and 1540'F for SVEA-96+. which are below the 2200'F limit. Therefore, the Hope Creek specific analysis meets the NRC SAFERIGESTR-LOCA licensing analysis requirements.
S-2
NEDO-33 172
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 3840 MWt. The analysis addresses a core flow range from 94.8% 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. Additional analysis performed at core thermal power levels at 3506 MWt and 3673 MWt and rated core flow. Calculated results are also included in the report for comparison.
The LOCA analysis was performed in accordance with NRC requirements to demonstrate conformance with the ECCS acceptance criteria of I 0CFR50.46. 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 SAFERIGESTR-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 I OCFR50.46 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-33 172
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-term 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 LAMB 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 GEI 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 SCATITASC.
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 2-1
NEDO-33 172 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 GE] 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-33 172 LAMB GESTR-LOCA SAFER SHORT-TERM THERMAL FUEL ROD LONG-TERM THERMAL HYDRAULIC TRANSIENT MODEL THERMALIMECHANICAL DESIGN HYDRAULIC TRANSIENT MODEL OU'rTPUT I' I OUTPUT CORE AVERAGE PRESSURE CORE INLET FLOW GAP CONDUCTANCE CORE INLET ENTHALPY ROD INTERNAL PRESSURE
\I ,
TASC TRANSIENT CRlTICAL POWER MODEL OUTPUT OUTPUT POT WATER LEVEL RESPONSE I LOCATION AND TIME OF PRESSURE BOILING TRANSITION HEATTRANSFERCOEFFICENT LOCAL OXIDATION Figure 2-1. Flow Diagram of LOCA Analysis Using SAFER/GESTR 2-3
NEDO-33 172 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 1 - Peak Cladding Temperature - The calculated maximum fuel element cladding temperature shall not exceed 2200'F.
Criterion 2 - Maximum Claddina Oxidation - The calculated local oxidation shall not exceed 0.17 times the cladding thickness before oxidation.
Criterion 3 - Maximum Hvdroeen 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 Geometry - Calculated changes in core geometry shall be such that the core remains amenable to cooling.
Criterion 5 - Lona-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 I 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 SAFERIGESTR-LOCA LICENSING METHODOLOGY 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 calculation of the limiting PCT to demonstrate conformance with the requirements of IOCFR50.46 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:
3-i
NEDO-33 172 PCTI = PCTNomina + ADDER
.cnsing The value of ADDER is calculated as follows:
ADDER2 = [PCTAn,. K - PCTNonnial ] 2+ E ( 5PCTi 2 where:
PCTA,,. ; = Peak cladding temperature from calculation using Appendix K specified models and inputs.
PCTNominatl -Peak cladding temperature from nominal case.
E: ( 8PCT1 ) 2 = Plant variable uncertainty term.
The plant variable uncertainty term accounts statistically for the uncertainty in parameters that are not specifically addressed by IOCFR50 Appendix K.
To conform to IOCFR50.46 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 1600WF 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 3-2
NEDO-33 172 Hope Creek was designed as one of the GE BA'R/4 product line plants; however, the ECCS includes features that are used in the BWAR-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 BAR4. 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 (BWR-5/6 and some BWR/4 such as Hope Creek). The SAFER analysis of a typical BWR/6 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 was 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; the high-pressure make-up system is not available due to the limiting failure of Channel A DC failure. 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 Plaiit-Speciflc 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. 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; 3-3
NEDO-33 172
- 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 1 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 (4.085 f9) were performed.
Compliance wvith 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 performned 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.
3-4
NEDO-33 172 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 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 111 (Reference 4). This conservatism is on the order of 1750 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 power level. The core flow coastdown 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 3-5
NEDO-33 172 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 performance correlations (e.g., boiling transition, bundle pressure drop) in a format compatible with GE's methods.
3-6
NEDO-33 172 Table 3-1 ANALYSIS ASSUMPTIONS FOR NOMINAL CALCULATIONS (Reference 2)
- 1. Decay Beat 1979 American Nuclear Society (ANS)
(Figure 3-1)
- 2. Transition Boiling Temperature Iloeje correlation Break Flow 1.25 HEMWl) (subcooled) 1.0 HEM(]) (saturated)
- 4. Metal-Water Reaction EPRI coefficients
- 5. Core Power 3840 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: Honiogczicous Equilibrium Modcd.
3-7
NEDO-33 172 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 blowdown 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 3917 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
- 11. 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-I
- 15. Fuel Exposure Same as Table 3-1 (1) 102%ornonuinal corc po%%cr(.02 x 3840 MWZ) and 102%orbtiidlc powcrwcrc uscd in 1lhc Appendix K analysis.
3-8
NEDO-33 172 1.2 4 0.8*
3:
0 a-a, - 1971 ANS + 20% (Appendix K) o 0.6 1979 ANS (Nominal) ......
.0 z 04-
.\.
0.2 -
n I U _-
0.01 0.1 1 10 100 1000 10000 Time After Break (seconds)
Figure 3-1. Hope Creek Decay Heat Used forNominal and Appendix K Calculations 3-9
NEDO-33 172 4.0 INPUT TO ANA'LYSIS 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 10CFR50.46 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 those documented in Reference 11 and the current Technical Specifications. Table 4-3 shows a summary of specific performance input parameters used in the analysis (and in the Reference 11 analysis). Table 4-3 is applied to all fuel types and initial conditions. Note that the analysis is performed with a LPCI flow rate that assumes the minimum flow bypass valve remains open.
4-1
NEDO-33 172 Table 4-1 PLANT PARAMETERS USED IN HOPE CREEK SAFER/GESTR-LOCA ANALYSIS Plant Parameters Nominal Appendix K Core Thermnal Power (MWt) 3840 3917 Corresponding Power (% of 840 MWIt) 100 102 Vessel Steam Output (Qbm/hr) 16.773 x 106 17.198 x 106 Rated Core Flow (lbm/hr)(1) 1O x 106 10 x 106 Vessel Steam Dome Pressure (psia) 1020 1055 Maximum Recirculation Suction Line 4.085(2) 4.085(2)
Break Area (ft2 )
(l) 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 94.8 Mlb/hr to 105 Mlb/hr (94.8% to 105%) of rated core flow at 3840 and 3917 MWt core power.
(2) Includes area of bottom head drain.
4-2
NEDO-33 172 Table 4-2 FUEL PARAMETERS USED IN HOPE CREEK SAFER/GESTR-LOCA ANALYSIS Analvsis Value Fuel Parameter GE14 SNTEA-96+
PLHGR (IkW/ft)
- Appendix K See Note 3
- Nominal ))
MAPLHGR (k}'F/ft)
- Appendix K 12.82x .02 See Note
- Nominal 12.24 Worst Case Pellet Exposure '] See Note 3 (MWd/MTU) SeeNote_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 xokes (I) This is tlhecxposurc at tlh knec i tlie PLHGR cunre for cacl fucl. Ii represents the limitig,opCrotiig condition resulting in the ninxiurn calculatcd PCT at anytime during the fuel bundle life.
(2) GE 14 (I OxIO) has 2 wvaler rods occupying a 8-rodspace. SVEA-96+ (IOxlO) has no water rods, but hias a watcr channel occupying a 4-rod space.
(3) Thc PLHGR curve for SVEA-96+ fuel is Westinghouse Proprictanu and cannot be included. Thc SVEA-96+ analysis was perfonied at the exposure wilh the highest PLHGR.
4-3
NEDO-33172 Table 4-3 HOPE CREEK SAFER/GESTR-LOCA ANALYSIS ECCS PARAMETERS
- 1. Low Pressure Coolant Injection (LPCI) System Analysis Variable Units 'aliue
- a. Maximum vessel pressure at which pumps can inject flow psid (vessel to 286 drywell)
- b. Minimum flow to reactor vessel with minimum flow bypass valve open
- Vessel pressure at which below listed flow rates are psid (vessel to 20 quoted drywell) 9______
- ILPCI pump gpm (i
- 2 LPCI pumps gpm 180001i)
- 3 LPCI pumpsm gp 27000Q"
- 4 LPCI pumps gpm 36000("
- c. Minimum flow to reactor vessel at 0 psid with minimum flow bypass valve open
- 1 LPCI pump gPm 10600"'
- 2 LPCI pumps gpm 21200("
- 3 LPCI pumps gpm 318001'"
- 4 LPCI pumps gpm 42400("
- 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 I.d) to system capable of sec ot delivering full flow (power available, pump at rated speed, and injection valve fully open)
- g. Injection valve stroke time-opening sec 24(2)
(I) Thcsclow mrtcsassumtctihemininumnflo- bypass valvcdocsnotclosc. Flow ratcsarcincreascdby 1000_gpm per pump i^1icii thc bypass valve closes. Tlicsc low rtes to tllc Vcsscl are reduccd by 8 gpin to account for Icakagc.
(2 This does not include signal proccssing dclay timc (I scc).
4-4
NEDO-33 172 Table 4-3 (cont,)
HOPE CREEK SAFER/GESTR-LOCA ANALYSIS ECCS PARAMETERS
- 2. Core Spray (CS) System Analysis Variable Units Valtle
- 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 drywell) _ _
. Minimum flow of one core spray loop gpm 56507)
- c. Minimum flow of one core spray loop at 0 psid gpm 7000(2)
- d. Initiating Signals
- Low water Ievel (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_ 1_
- 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)
(1} This does not include signal processing dclay time (I scc).
2' Tnc flow rate to vcsscl is reduced by 100 gpm to account for leakagc.
4-5
NEDO-33 172 Table 4-3 (cont,)
HOPE CREEK SAFER/GESTR-LOCA ANALYSIS ECCS PARAMETERS
- 3. High Pressure Coolant Injection (HPCI) 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 IHPCI 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
- e. Maximum allowable time delay from initiating signal sec35"'
(Item 3.d) to system capable of delivering full flow (pump at rated speed and injection valve fully open)
(I) This does not include sigatl processing dclay time (I scc).
4-6
NEDO-33 172 Table 4-3 (cont,)
HOPE CREEK SAFER/GESTR-LOCA ANALYSIS ECCS PARAMETERS
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 lbm/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
- e. Delay time from initiating signal completed to time valves sec 1 20t4) start to open r3' TiC smali break anal!ses assumc fivc ADS valves to be functioning,. but tlic ADS sensitivity studies toere anmly-zd assuming four ADS valves are futnctioning.
(i' This does not include signal processing dlaby time (I sec).
4-7
NEDO-33 172 Table 4-4 HOPE CREEK SINGLE FAILURE EVALUATION Assumed Failure°) I Systems Remaining( 2 )
Channel A DC Source (Battery) I LPCS. 3 LPC1. ADS'-"
LPCI Injection Valve (LPCI IV) 2 LPCS. 3 LPCL HPCI. ADS 3 3 Diesel Generator (D/G) I LPCS, 3 LPC1. HPCI, ADS(3)
Oticr postulated failures are not specirlcally considered because they all result in at least as much ECCS capacity as onc of thc assumed failures.
(2 Systems remaining, as identified in this table, arc applicabic to all non-ECCS line breaks. For a LOCA from an ECCS linc break. thc systems remaining are those listed. less thc ECCS svstcm in which the brcak is assumed.
1 Fivc ADS valves are assumcd for thc snall brcak analr scs. Four operabI ADS valves (onc non-functioning ADS in addition to the singIc failurc) are conscrvatively assumcd for largc brak analyscs and a scparatc smalI break sensitivitv studs to determine the impact oran ADS valvc out-or-scnricc.
4-8
NEDO-33 172 DD/G D/G(
(~ A D y I I QOG I I B I II I! I I
, - l- I 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-33 172 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 from DBA 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. Both I' peak and 2 nd peak PCTs are provided for large breaks in Table 5-1 to show the trend of PCTs with break sizes.
Except for DBA break with SVEA-96+ fuel, which has a low I" peak PCT, the nominal PCTs for the large breaks (Cl ft2 ) arc 1" peak limited; the 2nd peak PCT is strongly dependent on the ECCS performance. The dryout times were calculated for DBA suction break for both GEI4 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. All break sizes in Table 5-1 are analyzed with a LPCI flow rate that assumes the minimum flow bypass valve remains open. ((
5-1
NEDO-33 172 1]
For small breaks (c 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 were 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).
1))
The results of the Appendix K analyses are also shown in Table 5-1, and the plotted system response time histories for selected cases are plotted in Appendix B.
5-2
NEDO-33 172 5.1.2 Non-Recirculation Line Breaks Non-recirculation line breaks were analyzed for both GEI4 and SVEA-96+ fuels using nominal assumptions with battery failure. All breaks are consistently analyzed with a LPCI flow rate that assumes the minimum flow bypass valve remains open. 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 with the BWR-5/6 generic conclusions and demonstrate that the battery failure is limiting for all fuels. ((
))
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. Hopc 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 PCT is 1380'F for GE14. The Licensing Basis PCT of 1540'F for SVEA-96+ documented in Reference I I is unchanged at power uprate conditions.
5.2.2 Removal or the Current Requirement for Evaluation of Upper Bound PCIT The NRC SER approving the original SAFERIGESTR-LOCA application methodology (described in Reference 2) placed a restriction of 16000 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 S. GENE received an SER from the NRC (Reference 7) eliminating the 16001F 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 in Reference 2, including Hope Creek. In addition, the 1600'F restriction on the Upper Bound 5-3
NEDO-33 172 PCT is no longer applicable when evaluating the effect of changes and errors reported under the requirements of IOCFR50.46.
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 BW'R 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.
((
5A
NEDO-33 172 1]
As demonstrated in the discussions above, the Upper Bound is no longer restricted by the 16000 F limit. Therefore, when evaluating the effect of changes and errors reported under the requirements of 10CFR50.46, the effect on the Upper Bound PCT no longer needs to be evaluated.
5.3 EXPANDED OPERATING DOMAIN AND ALTERNATE OPERATING MODES Extended operating domains and alternate operating modes are presented as sensitivity studies to 5-5
NEDO-33 172 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)
((I 1))
5.3.2 Reduced Core Flowv (MELLLA / ELLLA)
The higher rod line in the MELLL region permits reactor operation at 94.8% of rated core flow for the rated power of 3840 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 flow 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 flow effects on the ECCS analyses were generically addressed in Reference 9, which was approved by the NRC in Reference 10. These studies demonstrated that no MAPLHGR multiplier was required for low core flow operation for the BWR-5/6 plant class, which has ECCS similar to Hope Creek. The SAFERIGESTR-LOCA analysis for low core flow conditions in the MELLL region was evaluated for Hope Creelk using the same ECCS inputs as used for the rated core flow conditions.
I((
)) The analysis was performed with both nominal and Appendix K assumptions. The results are shown in Table 5-3 with rated core flow results presented for comparison. The MELLLA results at 3339 MWt and 76.6% rated core flow documented in Reference 11 are also included for completeness.
5-6
NEDO-33 172
]1 5.3.3 Single-Loop Operation (SLO)
The ECCS performance for Hope Creek under SLO was evaluated using SAFER/GESTR-LOCA for the DBA break with battery failure. The analysis approach in determining the SLO multiplier on MAPLHGR and LHGR, and the calculated SLO multipliers for both GEI4 and SVEA-96+
fuels are documented in Reference 11. ((
The operating conditions for SLO are not changed with uprated power conditions. Therefore, the calculated SLO multiplier of 0.8 documented in Reference II for both GE14 and SVEA-96+
remain valid. The calculated SLO multipliers are conservative and assure that the SLO results satisfy the acceptance criteria of I OCFR50.46 and the NRC SER requirements for the SAFER application methodology. (3 1f 5.3.4 Additional Power Conditions Additional SAFER/GESTR-LOCA analyses were performed at 3506 MWt and 3673 MWt and at 100% rated core flow conditions for Hope Creek with consistent input assumptions. Both nominal and Appendix K assumptions were considered and the analyses were performed for both GE14 and SVEA-96+ fuels. The power levels were selected by PSEG for future interested plant operating power conditions at Hope Creek. The Licensing Basis PCT determined in Section 5.2.1 was calculated at 3840 MWt based on MELLLA conditions. If Hope Creek is licensed at a power level of 3506 MWt or 3673 MWt, the plant Licensing Basis PCT needs to be addressed.
The calculated results are shown in Table 5-4.
5-7
NEDO-33 172 5.4 MAPLIIGR 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) / LHGRFAC(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 discussed in the above paragraph. The SLO multiplier is applicable to all fuel rod exposures.
5-8
NEDO-33 172 Table 5-1
SUMMARY
OF HOPE CREEK SAFER/GESTR-LOCA RESULTS FOR RECIRCULATION LINE BREAKS"'
((
I IX I I I vI I I1 5-9
NEDO-33 172 Table 5-1 (continued)
SUMMARY
OF HOPE CREEK SAFER/GESTR-LOCA RESULTS FOR RECIRCULATION LINE BREAKS"'t
((
I I I II I I I I_
1]
5-10
NEDO-33 172 Table 5-2
SUMMARY
OF HOPE CREEK SAFER/GESTR-LOCA RESULTS FOR NON-RECIRCULATION LINE BREAKS"l)
(Nominal Analysis Basis)
I IX 5-11
NEDO-33 172 Table 5-3 MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS RESULTS COMPARISON FOR HOPE CREEK"')
LIMITING LOCA: DBA - Recirculation Suction Line Break, Battery Failure
] Il 7I1 I I I I_ I 5-12
NEDO-33 172 Table 5-4 SAFER/GESTR-LOCA RESULTS COMPARISON AT VARIOUS POWER CONDITIONS FOR HOPE CREEK(')
DBA Suction Break, Battery Failure, at 100% Rated Core Flow
((
t t
))
5-13
NEDO-33 172
[I 1]
Figure 5-1. Nominal and Appendix K LOCA Break Spectrum Results for GE14 Fuel 5-14
NEDO-33 172 1]
Figure 5-2. Nominal and Appendix K LOCA Break Spectrum Results for SVEA-96+ Fuel 5-15
NEDO-33 172
6.0 CONCLUSION
S LOCA analyses have been performed for Hope Creek at thermal power of 3840 MWt (nominal assumptions) using the GE SAFERIGESTR-LOCA Application Methodology approved by the NRC. These analyses were performed to demonstrate conformance with 10CFR50.46 and Appendix K, and thus, support a revised licensing basis for Hope Creek with the GE SAFERIGESTR-LOCA methodology.
As the SAFERIGESTR-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 10CFR50.46 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 GE14 and SVEA-96+ fuel types in the ECCS-LOCA evaluation are summarized in Table 6-2.
6-1
NEDO-33 172 Table 6-I SAFERIGESTR-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 1380 1540 < 22000 F Temperature (Licensing Basis)
- 5. Maximum Local <I% <1% <17%
Oxidation
- 6. Core-Wide Metal-Water <0.1% <0.1% <1%
Reaction
- 7. Coolable Geometry Items 4 & 5 PCT < 2200'F and Local Oxidation <
17%
- 8. 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 thejet 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-33 172 Table 6-2 Thermal Limits Analysis Limit PARAMETER GE14 SV'EA-96+
PLHGR - Exposure Limit Curve GWD/MT k\\I/ft GWDtMT kWI/ft Note I Note I Note I Note 1 Note I Note 1 Note I Note I MAPLHGR - Exposure Limit Curve GWD/MT k .Vft GWD/MT kW/ft 0 12.82 Note I Note I 21.09 12.82 Note 1 Notc I 63.50 8 Note I Note 1 70.00 5 Note 1 Note I Initial Operating MCPR 1.25 1.25 Minimum R-Factor [ ]
SLO Multiplier on PLHGR & MAPLHGR 0.80 0.80 Nok% (I) ThIc PLIGR curvc for SVEA-96+ fucl is Wcstingbousc Pitprictar and cannot bc included. ThC SVEA-96+ amalysis AAas perfonncd at the cxposure with tleC highiest PLHGR. The MAPLHGR curve for SVEA-96+ is based on the LHGR cun-c, so it is omintcd.
6-3
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 Ill, SAFER/GESTR Application Methodology," NEDE-23785-1-PA, General Electric Company, Revision 1, October 1984.
- 3) "General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with I 0CFR50 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 1, 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. Ill, Supplement 1, Revision 1, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume 111, 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 111, 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.
- 11) "SAFER/GESTR-LOCA Loss of Coolant Accident Analysis for Hope Creek Generating Station," NEDC-3)3153P, Revision 1, September 2004.
7-1
NEDO-33 172 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-I
NEDO-33 172 Table A-1 NOMINAL RECIRCULATION LINE BREAK FIGURE
SUMMARY
Notes: All Plots arc for GE 14 fucl. cxccpt wlhcn notcd.
- Rccirc. Break DBA 80% DBA 60% DBA 1.0 ft[
- Singlc Failure Battcry Battety Battenr Battcry Water Lcvel in A-la A-2a A-3a A4a A-5a A-Oa*
Hot & Av-crac Cltanncls Reactor Vessel A-lb A-2b A-3b A-4b A-5b A-6b*
Pressurc Peak Cladding A-1c.f* A-2c.f* A-3c.f1 Alc.f* A-5c A-6c*
Tempcralure Heat Tmnsrcr A-ld g* A-2d g* A-3dg* A4d..g* A-5d A.6d*
Coefficient ECCS Flow A-lc A-2c A-3c A4c A-5c A.6c*
- Plots for SVEA-96+ are included.
A-2
0 ta Figure A- I a. Water Level in Hot and Average Channels - DBA Suction - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
z 0
0 Is J-3 11.)
Figure A- lb. Reactor Vessel Pressure - DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
0
%--J Figure A-Ic. Peak Cladding Temperature (GE14) - DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
m aa c&
"3-Figure A-I d. Heat Transfer Coefficient (GE14) - DBA Suction - Battery Fnilure (Nominal) - 3LPCI-LPCS+ADS Available
r-C Cp aJ Figure A-le. ECCS Flow - DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
0
>0 (4
> ~to Figure A-If. Peak Cladding Temperature (SVEA-96+) - DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
oz
'a L--4 Figure A-lg. Heat Transfer Coefficient (SVEA-96+) - DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
oM
> 0d 0 wthii Figure A-2a. Water Level in Hot and Average 80% DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
a I
-J t'3 Figure A-2b. Reactor Vessel Pressure - 80% DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
P-"
> To I') tA I'-J Figure A-2c. Peak Cladding Temperature (GE 14) - 80% DI3A Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
z I 0 t.
-4 t')
Figure A-2d. Heat Transfer Coefficient (GE 14) - 80% DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
9 I'.)
Figure A-2c. ECCS Flow - 80% DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
a> 0 1 a Figure A-2f. Peak Cladding Temperature (SVEA-96+) - 80% DBA Suction - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
> II 0a I'-
Figure A-2g. Heat Transfer Coefficient (SVEA-96+) - 80% DBA Suction - Bailery Failure (Nominal) -
3LPCl+LPCS+ADS Available
C 0
4 W4
-4 L--j Figure A-3a. Water Level in Hot and Average Channels - 60% DBA Suction - Battery Failure (Nominal) -
3LPCT+LPCS+ADS Available
0
> a 00 co -
L--j Figure A-3b. Reactor Vessel Pressure - 60% DBA Suction - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
> ba I 0-Figure A-3c. Peak Cladding Temperature (GE14) - 60% DBA Suction - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
oz
> 0 Lij
-3 L
Figure A-3d. Heat Transfer Coefficient (GE 14) - 60% DBA Suction - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
F-"
0 La t'
ijJ Figure A-3e. ECCS Flow - 60% DBA Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
> 0a^
L-.3 FigureA-3f. Peak Cladding Temperature (SVEA-96+) - 60% OBA Suction - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
ri
> a
-J.
t'3 Figure A-3g. I-Heat Transfer Coefficient (SVEA-96+) - 60% DBA Suction - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
0 D
Figure A-4a. Water Level in Hot and Average Channels - ft2 Suctionl - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
a t'
tA
-3 to t--J Figure A-4b. Reactor Vessel Pressure - I 2 Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
0
~L) t')
'-aj Figure A-4c. Peak Cladding Temperature (GE 14) - 1 ft Suction - Battery Failure (Nominal) 2 - 3LPCI+LPCS-IADS Available
09 Figure A-4d. Heat Transfer Coefficient (GE 14) - 1 ft2 Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
zrrs
> 0
'A.
00 6.-3 Figure A-4e. ECCS Flow - I ft Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
'9 a
t~j ko 0w
-3 IN)
Figure A-4f. Peak Cladding Temperature (SVEA-96+) - I ft2 Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
taa
>0 Figure A-4g. Heat Transfer Coefficient (SVEA-96+) - I ft Suction - Battery Failure (Nominal) - 3LPCI+LPCS+ADS 2
Available
0 Li
-J N')
L-j Figure A-5a. Water Level in Hot and Average Channels -
9 0
tU)
-j toJ Figure A-5b. Reactor Vessel Pressure -
0 tU, La
-a hi 6-i Figure A-5c. Peak Cladding Temperature (GE14) -
0 wA
-tD.
-J.
Figure A-Sd. Heat Transfer Coefficient (GE 14)
a
- > 0
-F N)
L-j Figure A-5c. EGGS Flow -
0 U'A 'A.)
CS "3e Figiure A-6a. Water Level in Hot and Average Channels (SVEA-96+) -
A
-4 t,.
t3-F-P Figure A-6b. Reactor Vessel Pressure (SVEA-96+) -
z mr 0
0 t'a LA.
00
-:2 I',
Figure A-6c. Peak Cladding Temperature (SVEA-96+)
00
- > 'a so -a w)
LHsj Figure A-6d. Heat Transfer Coefficient (SVEA-96+) -
zrr 0
~> 0 0 'a L--j Figure A-6e. ECCS Flow (SVEA-96+)
NEDO-33 172 APPENDIX B SYSTEM RESPONSE CURVES FOR APPENDIX K RECIRCULATION LINE BREAKS Included in this Appendix are the system response curves for Hope Creek. Table B-1 shows the figure numbering sequence for the Appendix K recirculation breaks.
B-I
NEDO-33 172 Table B-I APPENDIX K RECIRCULATION LINE BREAK FIGURE
SUMMARY
Notc: All Plots arc for GEI4 fucl. cxccpt whcn notcd.
- Rccirc. Brcak DBA Suction - DBA Suction- 80°/o DBA 6MA DBA
- Singic Failure Rated - Balumn MELLLA - Suction - Suction -
Battcry Battciy Failurc Battry Failure WatcrLcvcl in B-la B-2a B-3a B4a B-5a Hot &8Averagc Chauncls Rcactor Vcssel B-lb B-2b B-3b B-lb B-Sb Pressurc Pcak Cladding B-lc.h* B-2c.S* B-3c.f* B~c.f* B-5c.f*
Tcmupcrature Hcat Transfcr B-ld.i* B-2d.g* B-3d.g* B4d.g* B-5d.g*
Cocfficient ECCS Flow B-lc B-2c B-3c B-c B-5c Corc InIct Flow B-If MCPR B-Ig
- Plots for SVEA-96+ are included.
B-2
z To4 Li)j Figure B-la. Water Level in l-lot and Average Channels (GE4) - DBA Suction - Rated - Battery Failure (App.
K) - 3LPCI+LPCS+ADS Available
'9 e
0
-4 lo Figure B-lb. Reactor Vessel Pressure (GE014) - DBA Suction - DBA Suction - Rated - Battery Failure (App.
K) - 3LPCI+LPCS+ADS Available
z m
a I')
L--j Figure B-I c. Peak Cladding Temperature (GE14) - DBA Suction - Rated - Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
0 wu 0
'N)
Figure B-Id. Heat Transfer Coefficient (GE14) - DBA Suction - Rated - Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
w9 a
toj
'S Fvigure B-Ie. ECCS Flow (GE 14) - DBA Suction - Rated - Battery F~ailure (App. K) -3LPCI+LPCS+ADS Available
0 00
'JJ 6.-
Figure B-If. Core Average Inlet Flow (GE14) - DBA Suction - Rated - Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
w 0
\b U.)uj "3-Figure B-lg. Minimum Critical Power Ratio (GE14 & SVEA-96+) - DBA Suction - Rated - Battery Failure (App. K)-
3LPCI+LPCS+ADS Available
Cd a I 0 W iJW
-J Figure B-i h. Peak Cladding Temperature (SVEA-96+) - DBA Suction - Rated - Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
0o0
-.3 I')
L- j Figure B-li. pleat Transfer Coefficient (SVEA-96+) - DBA Suction - Rated - Battery Failure (App. K)-
3LPCI+LPCS+ADS Available
z w a I-.3
.. to Figure B-2a. Water Level in F-lot and Average Channels - DBA Suction - MELLLA - Battery Failure (App. K) 3LPCI+LPCS+ADS Available
z m
Ra "3
Figure B-2b. Reactor Vessel Pressure- OBA Suction - MELLLA - lBattery Failure (App. K) -
3 LPCI+LPCS+ADS Available
I 0
-I Figure B-2c. Peak Cladding Temperature (GE 14) - DBA Suction - MIELLLA - Battcry Failure (App. K) -
3LPCI+LPCS+ADS Available
NN NN UU WU LLLL LLLL yy yy NNN NN UU UU LL LL yy yy NNNN NN UU 13 LL LL yy yy NN NNNN UU U1 LL LL yyyy NN NNN UU UU LL L LL L yy NN NN UU 13 LL LL LL LL yy NN NN WUUUW LLLLLLL LLLLLLL yyyy USERID: NULLY DOCUMENT NUMBER: 71 PRINTED ON 3/29/05 10:46:28 AM
Yeager, Linda: L. .- :..
From: Thompson, Jack W.
Sent: Tuesday, March 29, 2005 6:22 AM To: Yeager, Linda L.
Subject:
RE: Credit Card Supervisor update Robert Kolo
-Original Message--
From: Yeager, Linda L Sent: Monday, March 28, 2005 3:55 PM To: Hassler, Matthew J.; Racer, Jacqueline J.; Thompson, Jack W.
Subject:
Credit Card Supervisor update Importance: High Please provide your respective Supervisor and employee number. We need to change Paymentnet with your new supervisor. Thank you.
268 HASSLER MATTHEW J OUTAGE 5130200 1350 0.00745.2.3.5 01, 3
293RACER JACQUELINE J REFUELING OUTAGE GROUP 5130200 1350 0.00745.7.1.1 01 7
837THOMPSON JOHN W OUTAGE MAINTENANCE 5130200 1350 0.01049.2.3.5 01; 0
Linda Yeager Financial Analysis & Controls, M/C N07 Wk:856/339-7850 Fax: 856/339-1369 Email: Linda.Yeaqer(ZPSEG.com
t0 t.-.3 Figure B-2d. Heat Transfer Coefficient (GE14) - DBA Suction - MELLLA - Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
we W
M L--
Figure B-2e. ECCS Flow - DBA Suction - MELLLA - Battery Failure (App. K) - 3LPCI+LPCS+ADS Available
z I MI I--
-J Figure B-2f. Peak Cladding Temperature (SVEA-96+) - DBA Suction - MELLLA - Battery Failure (App. K) -
3LPCT+LPCS+ADS Available
z 1C 00
-4 Figure B-2g. Heat Transfer Coefficient (SVEA-96+) - DBA Suction - MELLLA - Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
10w L'-P Figure B-3a. Water Level in Hot and Average Channels - 80% DBA Suction -Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
oz Cr zw I\)
Figure B-3b. Reactor Vessel Pressure - 80% DBA Suction -Battery Failure (App. K) - 3LPCI+LPCS+ADS Available
0 Figure B-3c. Peak Cladding Temperature (GE14) - 80% DBA Suction -Battery Failure (App. K) -
3LPCT+LPCS+ADS Available
CO C Figure B-3d. Heat Transfer Coefficient (GE14) - 80% DBA Suction -Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
-z td U
C t)W ('a W'
-J t-Figure B-3e. ECCS Flow - 80% DBA Suction -Battery Failure (App. K) - 3LPCI+LPCS+ADS Available
zI iW 0
Figure B-3f. Peak Cladding Temperature (SVEA-96+) - 80% DBA Suction -Battery Failure (App. K) -
3LPCI4LPCS+ADS Available
W~0O
-a Figure B-3g. Heat Transfer Coefficient (SVEA-96+) - 80% DBA Suction -Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
(ON w
-4 Figure B4a. Water Level in Hot and Average Channels - 60% DBA Suction -Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
z To4 ba 0 t'.
-3 Figure B-4b. Reactor Vessel Pressure - 60% DBA Suction -Battery Failure (App. K) - 3LPCI+LPCS+ADS Available
z e 0 on w 6_t.3 Figure B4c. Peak Cladding Temperature (GE14) - 60% DBA Suction -Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
td I
-9
'0 Z4 Figure B4d. Heat Transfer Coefficient (GE14) - 60% DBA Suction -Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
z 0
0 t'.)
0 w
--A W
Figure B-4e. ECCS Flow - 60% DBA Suction -Battery Failure (App. K) - 3LPCI+LPCS+ADS Available
5 0
- w4 Figure B3-4f. Peak Cladding Temperature (SVEA-96+) - 60% DBA Suction -Battery Failure (App. K) -
3L1PCI+LPCS+ADS Available
z tz 0 Figure B4g. Heat Transfer Coefficient (SVEA-96+) - 60%/o DBA Suction -Battery Failure (App. K) -
3LPCI+LPCS+ADS Available
z 0
-j t-J Figure B-5a. Water Level in Hot and Average Channels -
rn1 0
tw
'-3~
t.2 Figure B-5b. Reactor Vessel Pressure -
Ul t'S)
Figure B-Sc. Peak Cladding Temperature (GE14)
200 w
-3 t'3 Figure B-5d. Heat Transfer Coefficient (GE14) -
Available
z
-4
.x Figure B-Se. ECCS Flow -
z to'
-4 Figure B-5f. Peak Cladding Temperature (SVEA-96+) -
CO 0 6
%0 Figure B-5g. Heat Transfer Coefficient (SVEA-96+) -
NEDO-33172 APPENDIX C SYSTEM RESPONSE CURVES FOR NOMINAL NON-RECIRCULATION LINE BREAKS Included in this Appendix are the system response curves for Hope Creek. Table C-I shows the figure numbering sequence for the Nominal Non-recirculation breaks.
C-1
NEDO-33 172 Table C-I NOMINAL NON-RECIRCULATION LINE BREAK FIGURE
SUMMARY
Note: All Plots are for GE14 fuel, except when noted.
- Non-Recirc. Core Spray Line Stcamlinc Steamlinc FecdwatcrLinc LPCI Linc Break - Battery Inside Outside -Battery -Battery
- Single Failure Containment - Containunent -
Battery Battcery WVater Level in C-la C-2a C-3a C-a C-5a Hot & Averagc Chanmels Reactor Vesscl C-lb C-2b C-3b C4b C-5b Pressure Peak Cladding C-lch* C-2cf* C-3c,f'* C-4cp C-5cf*
Temperature Hcat Transfer C-ld,i* C-2d,g* C-3d,g* C4dg* C-5dg*
Coefficient ECCS Flo C-lc C-2c C-3c C-Ie C-5c
- Plots for SVEA-96+ are included.
C-2
C) 0 I S tij
-3 tQ Figure C- I a. Water Level in Hot and Average Channel s - Core Spray Line Break - Battery Failure (Nominal) -
3LPCI+ADS Available
zn 0
t-
-3 Figure C- lb. Reactor Vessel Pressure - Core Spray Line Break -lBattery Failure (Nominal) - 3LPCI+ADS Available
z m
a tj) 6LJ Figure C-1c. Peak Cladding Temperature (GE14) - Core Spray Line Break - Battery Failure (Nominal) -
3LPCI+ADS Available
0
- 0
-3 IQ Figure C-ld. Fleat Transfer Coefficient (GE 14) - Core Spray Line Break - Battery Failure (Nominal) -
3LPCI+ADS Available
0
-4
-4 I',)
Figure C-le. ECCS Flow - Core Spray Line Break- Battery Failure (Nominal) - 3LPCI+ADS Available
z 000 L.tjJ
,--4 Figure C-I f. Peak Cladding Temperature (SVEA-96+) - Core Spray Line Break - Battery Failure (Nominal) -
3LPCI+ADS Available
a o0
-. 3 Figure C-lg. Heat Transfer Coefficient (SVEA-96+) - Core Spray Line Break- Battery Failure (Nominal) -
3LPCT+ADS Available
I 0 0U.
I'.)
Figure C-2a. Water Level in Hot and Average Steamline Break Inside Containment - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
9 0 Figure C-2b. Reactor Vessel Pressure - Steamline Break Inside Contaimnent - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
£ to 0
tjj to t~j L--J Figure C-2c. Peak Cladding Temperature (GE14) - Steamline Break Inside Containment - Battery Failure (Nominal) -
3LPCT+LPCS+ADS Available
ta Figure C-2d. Heat Transfer Coefficient (GE14) - Steamline Break Inside Containment - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
£ 0
-13 Figure C-2e. ECCS Flow - Steamliine Break Inside Containnment -Battery Failure (Nominial) -3LPCI+/-LPCS+ADS Available
o9 I 0 1I)
L--4 Figure C-2f. Peak Cladding Temperature (SVEA-96+) - Steamline Break itside Containment - Battery Failure (Nominal) 3LPCI+LPCS+ADS Available
£ 0
,U)
-J Figure C-2g. Heat Transfer Coefficient (SVEA-96+) - Steamline Break Inside Containment - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
o 0
-a N)
Figure C-3a. Water Level in Hot and Average Channels - Steamline Break Outside Containment - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
n 0 00 U
-J L..
Figure C-3b. Reactor Vessel Pressure - Steamline Break Outside Containment - Battery Failure (Nominal) -
3LPCt+LPCS+ADS Available
z lo 10 LI Figure C-3c. Peak Cladding Temperature (GE14) - Steamline Break Outside Containment - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
a sO ?.
6
-j Figure C-3d. Heat Transfer Coefficient (GE14) - Steamline Break Outside Containment - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
C)a 0
-J I')
Figure C-3e. ECCS Flowv - Steamline Break Outside Containment - Battery Failure (Nominal) - 3LPCI+LPCS-ADS Available
(Nominal) -
3LPCT+LPCS+ADS Available
m 0
(&s a a A.,s a".O UO- k ... U 3Av 3LPCI+LP'CS+ADS Available
2 o9 a
-11 L-I')
Figure C-4a. Water Level in Hot and Average Channels - Feedwater Line Break - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
z4 m
00 to
(./1t, VDJ Figure C-4b. Reactor Vessel Pressure - Feedwater Line Break - Battery Failure (Nominal) - 3LPCI4-LPCS+ADS Available
zM na I 03 t.a
"--J Figure C-4c. Peak Cladding Temperature (GE14) - Feedwater Line Break - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
0 ow j
Figure C-4d. Hleat Transfer Coefficient (GEl4) - Feedwater Line Break - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
2 0'ah I-Figure C-4e. ECCS Flow - Feedwater Line Break - Battery Failure (Nominal) - 3LPCI+LPCS+ADS Available
zm 2 0 O .
bj N-3 Figure C-4f. Peak Cladding Temperature (SVEA-96+) - Feedwater Line Break - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
oA 0
O In
.4 Figure C-4g. [Teat Transfer Coefficient (SVEA-96+) - Feedwvater Line Break - Battery Failure (Nominal) -
3LPCI+LPCS+ADS Available
9 n 00
- ajJ L--J Figure C-5a. Water Level in Hot and Average Channels - LPCI Line Break - Battery Failure (Nominal) -
2LPCI+LPCS+ADS Available
m z4 U
O-Li) p t')
t~3 Figure C-5b. Reactor Vessel Pressure - LPCI Line Break - Battery Failure (Nominal) - 2LPCI+LPCS+ADS Available
9 "a
T-4 "3
Figure C-Sc. Peak Cladding Temperature (GE 14) - LPCI Line Break - Battery Failure (Nominal) - 2LPCI+LPCS+ADS Available
z I)
C ULA 1"3 Figure C-5d. [Teat Tranisfer Coefficienit(0214)- LPCI Line Break- Battery Failure (Nomiinal) - 2LCI+LPCS+ADS Available
09 n
Uj, CO VI, I
lo Figure C-Se. ECCS Flow - LPCI Line Break - Battery Failure (Nominal) - 2LPCI+LPCS+ADS Available
a 0 ta tt'3-Figure C-5f. Peak Cladding Temperature (SVEA-96+) - LPCI Line Break - Battery Failure (Nominal) -
2LPCI+LPCS+ADS Available
zm I0 L.J I'.)
Figure C-5g. Heat Transfer Coefficient (SVEA-96+) - LPCI Line Break - Battery Failure (Nominal) -
2LPCI+LPCS+ADS Available