ML20141C119

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Non-proprietary NEDC-32720, Hatch Units 1 & 2 SAFER/GESTR- LOCA Loss-of-Coolant Accident Analysis
ML20141C119
Person / Time
Site: Hatch  
Issue date: 03/31/1997
From: Paradiso F, Stoll C, Yang A
GENERAL ELECTRIC CO.
To:
Shared Package
ML20141C095 List:
References
DRF-A13-00402, DRF-A13-402, NEDC-32720(NP), NUDOCS 9706240323
Download: ML20141C119 (46)
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Text

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I 9 GENuclear Energy NEDC-32720 DRF A13-00402 Class 1 i March 1997 Hatch Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis i A.I. Yang F. M. Paradiso g.. O i D DO K S 21 P PDR j

1' O GENuclearEnergy GeneralElectric Company 175 CurtnerAvenue San Jose, CA 95125 NEDC-32720 Class 1 March 1997 EDWIN I. HATCH NUCLEAR PLANT UNrrS 1 AND 2 SAFER /GESTR-LOCA LOSS-OF-COOLANT ACCIDENT ANALYSIS A. I. Yang F. M. Paradiso Approved: C. H. Stoll, Project Manager Hatch Extended Power Uprate Project 2

NEDC-32720 DISCLAIMER OF RESPONSIBILITY This document was prepared by the general Electric Company (GE). No other use, direct or indirect, of the document or the information it contains is authorized; and with respect to any unat.thorized use, neither GE nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy, or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information. Fmishing this document does not convey any license, expresss or implied, to use any patented im ention or any information of GE disclosed herein, or any rights to publish or make copies of the document without prior written permission ofGE. This document is the non-proprietary version of NEDC-32720P. It was generated under the Power Uprate contract between Georgia Power Company (GPC) and GE, as identified in Purchase Order Number 6012598, dated October 3,1996, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. ii ___ _j

i NEDC-32720 TABLE OF CONTENTS Page

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 3-2 3.3 Generic Analysis 3-3 3.4 Hatch Plant-Specific Analysis 3-4 4.0 INPUT 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 5-2 5.2 Compliance Evaluations 5-2 5.2.1 Licensing Basis PCT Evaluation 5-2 5.2.2 Upper Bound PCT Evaluation 5-2 ~ 5.3 Alternate Operating Mode Considerations 5-3 5.3.1 Power Uprate 5-3 5.3.2 Extended Load Line Limit Analysis (ELLLA) and ARTS 5-3 5.3.3 Increased Core Flow 5-3 5.3.4 Single-Loop Operation (SLO) 5-4 5.3.5 Feedwater Heater Out-of-Service and 5-4 Final Feedumta remperature Reduction \\PLHGR Lim': 5-4 6.0 ,. oIONS 6-1 i Iii

NEDC-32720 .l TABLE OF CONTENTS (Continued) Eage l

7.0 REFERENCES

7-1 APPENDIX - SYSTEM RESPONSE CURVES FOR THE NOMINAL A-1 AND APPENDIX K DBA RECIRCULATION LINE BREAK s e iV

NEDC-32720 LIST OF TABLES Iahlt lills Eage 3-l' Analysis Assumptions for Nominal Calculations 3-5 3-2 Analysis Assumptions for Appendix K Calculations 3-6 4-1 Plant Operational Parameters Used in the Plant Hatch 4-2 SAFER /GESTR-LOCA Analysis 4-2 Fuel Parameters Used in the Plant Hatch 4-3 SAFER /GESTR-LOCA Analysis 4-3 Plant Hatch SAFER /GESTR Analysis ECCS Parameters 4-4 4-4 Plant Hatch Single-Failure Evaluation 4-9 5-1 Comparison of DBA Recirculation Suction Line Break Results 5-5 for Plant Hatch Units 1 and 2 5-2 Summary of Recirculation Line Break Results for Plant Hatch 5-6 5-3 Plant Hatch Results for Non-Recirculation Line Breaks 5-7 5-4 Extended Load Line Limit Analysis Results 5-8 Comparison for Plant Hatch 5-5 Increased Core Flow Results Comparison For Plant Hatch 5-9 5-6 Single-Loop Operation Results Comparison for Plant Hatch 5-10 5-7 Feedwater Temperature Reduction Results 5-11 Comparison for Plant Hatch 6-1 SAFER /GESTR-LOCA Licensing Results for Plant Hatch 6-2 A-1 (Proprietary Information Deleted) A-1 O i v

NEDC-32720 LIST OF FIGURES Figure Title P_ags 2-1 Flow Diagram ofLOCA Analysis Using SAFER /GESTR 2-3 3-1 Normalized Decay Power for Nominal and Appendix K Cases 3-7 4-1 Plant Hatch ECCS Configuration with Normal Diesel Alignment 4-11 5-1 Nominal and Appendix K LOCA Break Spectrum Results 5-12 A-1 (Proprietary Information Deleted) 4 9 r

NEDC-32720

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 bretk 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 10CFR50 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 still 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. Plant Hatch has utilized these models and this licensing methodology since 1986, and continues to use SAFER /GESTR-LOCA for this updated analysis. 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 spectmm 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 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 inputs. (2) The Upper Bound PCT is required to be less than the Licensing Basis PCT. Summary S-1

NEDC-32720 (3) The Upper Bound PCT has generally been demonstrated to be less than the Licensing Basis PCT when the limiting nominal PCT is 1600'F and lower. Therefore, it is desirable that the Upper Bound PCT be below 1600 F; otherwise, additional plant-specific analyses must be performed. The SAFER /GESTR-LOCA analysis for Plant Hatch was perfonned in accordance with NRC requirements and demonstrates conformance with the Emergency Core Cooling System (ECCS) acceptance enteria of 10CFR50.46. A sufficient number of plant-specific break sizes were evaluated to establish the behavior of both the nominal and Appendix K PCT as a function of break size. (Proprietary Information Deleted) In addition, many of the ECCS parameters were conservatively established relative to actual measured ECCS performance. SAFER /GESTR-LOCA was originally applied to Plant Hatch M 1986 (Reference 7). In anticipation of applying for a licensing amendment for both Hatch Units 1 and 2 to operate at an increased power level, the LOCA analyses have been performed at 108% of the current core thermal power using the latest approved version of SAFER /GESTR-LOCA. The results of this analysis may not be directly comparable to the 1986 analysis because the version of SAFER /GESTR-LOCA used in this analysis incorporates some coding modifications (Reference 9). This analysis is applicable from the current rated core thermal power of 2558 MWt up to a bounding analysis power of 2763 MWt and multiple operating conditions, including Extended Load Line Limit Analysis (ELLLA), Increased Core Flow (ICF), Single-Loop Operation (SLO), Feedwater Heater Out Of Service (FWHOOS) and Final Feedwater Temperature Reduction (FFWTR). This analysis is also applicable if GPC changes the safety-related power supply to the low pressure coolant injection (LPCI) motor-operated valves as described in Section 4.1. - (Proprietary Information Deleted) Therefore, Plant Hatch meets the NRC SAFER /GESTR-LOCA licensing analysis requirements. Summary S-2

1 NEDC-32720

1.0 INTRODUCTION

This document provides the results of the loss-of-coolant accident (LOCA) analysis performed by GE Nuclear Energy (GE-NE) for Plant Hatch. The analysis was performed using the SAFER /GESTR-LOCA Application Methodology approved by the Nuclear Rcgulatory Commission (Reference 1). The SAFER /GESTR-LOCA methodology was first applied to Plant Hatch in Reference 7. This analysis updates the Reference 7 analysis for plant operation at power uprate conditions. This analysis was performed assuming a bounding thermal power level of 2763 MWt, corresponding to 108% of the current licensed value of 2558 MWt and ECCS parameter relaxations that bound established parameters. The analysis also considered a core flow operating range of 71.4 Mlb/hr to 82.4 Mlb/hr for Unit 1 and 70.0 Mlb/hr to 80.9 Mlb/hr for Unit 2. The results for the analysis at the 2763 MWt power level conservatively bound current plant operation. at 2558 MWt. In addition, some of the Emergency Core Cooling System (ECCS) and related equipment performance parameters were conservatively established relative to actual ECCS performance. This LOCA analysis was performed in accordance with NRC requirements to demonstrate conformance with the ECCS acceptance criteria of 10CFR50.46. 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 for Plant Hatch. The SAFER /GESTR method described in NEDC-23785-PA (Reference 2) documents the requirements and approved methodology to satisfy these requirements. The SAFER /GESTR-LOCA application methodology is based on the generic studies presented in Reference 2. The approved application methodology consists of three main parts. First, potentially limiting LOCA cases are determined by applying realistic (nominal) analytical models across the entire break spectmm. Second, the limiting LOCA cases are analyzed with an Appendix K model (inputs and assumptions) which incorporates the required features of i 10CFR50 Appendix K. For the most limiting case, a Licensing Basis peak cladding temperature (PCT) is calculated bued on the nominal PCT with an adder to account statistically for the differences between the nominal and AppendixK assumptions. Finally, a statistically derived Upper Bound PCT is calculated to demonstrate the conservatism of the Licensing Basis PCT. The Licensing Basis PCT conforms to all the requirements of 10CFR50.46 and Appendix K. I Introduction 1-1

NEDC-32720

2.0 DESCRIPTION

OF MODELS Four GE-NE computer models were used to determine the LOCA response for the Plant Hatch LOCA analysis. These models are LAMB, SCAT / FASC, SAFER, and GESTR-LOCA (References 2 and 3). 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 fuel rod heatup. Figure 2-1 is a flow diagram of these computer models, indicating the major code functions and the transfer of major pc.rameters. The purpose of each model is described in the following subsections. 2.1 LAMB This model 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 (most importantly, 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 / FASC This model completes the transient short-term thermal-hydraulic calculation for large recirculation line breaks. The time and location of boiling transition are predicted during the period of recirculation pump coastdown. When the core inlet flow is low, SCAT 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. For Gell (and later fuel), an improved SCAT model (designated "TASC") is used to predict the time and location of boiling transition and dryout. This model explicitly models the axially varying flow areas and heat transfer surface resulting from the gel 1 (and later fuel) part length fuel rods, and incorporates the critical power correlation for GE13 (Reference 6). 2.3 GESTR-LOCA This model 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 estabhshes the transient pellet-cladding gap conductance for input to both SAFER and SCAT /TASC. Description ofModels 2-1

NEDC-32720 2.4 SAFER This model 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 PCT as functions of time. For fuels with part length fuel rods, the fuel rods are treated as full-length rods, which conservatively overestimates the hot bundle power. Description ofModels 2-2

sUo \\ L E D LO E NE AM S I MT N IC RN O F P F EE E HI SE R T T ERO S E N U R UC F MA P T A RR T CELSR SE S ET U PVEF R ERS TC O GL PN T L I R A S NU E R F OA T T LR A G T D W A / Y E R [ H E FASgn i N s G U I E S R s i ED EU CS L NS C DIC AE a A A T R n T CP M O ON U UL L RA P T DA R LH 1 T EC U NN C R S UE O O O E E FM CT G / L PN lA AI f M GD o R O E R m H [ m T g a M w o \\ N H L ED LO F Y M AM P L ON MT WL A EO R N A CL MT I E H TE II m E L I TS T FT C ID TS HI N S R T DN B E A CO U NA g N U M MA 1 T LT P T TM P / T AR E A RR T L ET U E T NR L A EE U NG T O N C SW O ON C TU RI S NO Tl ^OE I R l OU R AP Ai R O R CO E HA SR C T OB D L Y [ / H N, .a 4> vw ili!

NEDC-32720 l 3.0 ANALYSIS PROCEDURE 3.1 LICENSING CRITERIA The Code of Federal Regulations (10CFR50.46) outlines the acceptance criteria for ECCS analyses. A summary of the acceptance criteria is provided below: Criterion 1 - Peak Cladding Temperature: The calculated maximum fue, element cladding temperature shall not exceed 2200 F. Criterion 2 - Maximum Claddinn Oxidation: The calculated total local oxidation shall not exceed 0.17 times the total cladding thickness before oxidation. Criterion 3 - Maximum Hydrogen Ecueration: 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 - Lonn-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. The conformance with Criteria 1 through 3 for Plant Hatch 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 Reference 3 and remain unchanged by application of SAFEPJGESTR-LOCA. The licensing methodology using SAFER /GESTR-LOCA is discussed in Section 3.2. 4 Analysis Procedure 3-1

NEDC-32720 3.2 SAFER /GESTR-LOCA LICENSING METHODOLOGY The SAFER /GESTR-LOCA licensing methodology approved by the NRC in Reference 1 l allows the plant-specific break spectrum to be defined using the nominal input assumptions. However, the calculation of the limiting PCT to demonstrate conformance with the requirements of 10CFR50.46 must include specific inputs and models documented in Appendix K. The Licensing Basis PCT is based on the most limiting LOCA (highest PCT) and is defined as: ) PCTLicensing = PCT ominal + ADDER. N I i The adder is calculated as follows: j 2 App. K - PCT ominal 32 + I(S PCT )2, ADDER = [ PCT N i where: PCTApp.K Peak cladding temperature from calculation using Appendix K specified = models and inputs. PCT ominal N Peak cladding temperature from nominal case. j = I(8 PCT)i Plant variable uncertainty term. = ) The plant variable uncertainty term accounts statistically for the uncertainty in parameters which are not specifically addressed by 10CFR50 Appendix K. To conform with 10CFR50.46 and the SAFER /GESTR-LOCA licensing methodology, the t Licensing Basis PCT must be less than 2200'F. I i l 1 Analysis Procedure 3-2 I

_.m.___ I NEDC-32720 i l Conformance evaluation of the nominal PCT is also required through the use of a statistical Upper Bound PCT as defined in Reference 1. The Upper Bound PCT is a function of the limiting break Nonunal PCT, modeling bias, and plant variable uncertainty. The Upper Bound PCT is defined as: l PCTUpper Bound = PCT ominal + A4-maxgeneric + (A 3 + 2sA3) N ) where: i l A4-maxgeneric = Modeling Bias. This term accounts for errors in modeling processes for which experimental data is available for comparison. These are primarily the LOCA thermal-hydraulic processes. i 8 1 { (3 + 2sA3) Plant Variable Uncertainties. This term accounts for the uncertainties due = l to inputs to the model. These are typical plant parameters with associated uncertainties in their measured values. l The Upper Bound PCT is required to be less than the Licensing Basis PCT. This ensures that the Licensing Basis PCT is in all cases, greater than the 95th percentile of the PCT l distribution for the limiting case LOCA, and for all LOCAs within the design basis. As part of the j development of SAFER /GESTR-LOCA licensing methodology, GE-NE demonstrated that this j criterion was satisfied for the BWR/3-4 class of plants. The application methodology was also accepted on a generic basis for the Upper Bound PCT up to 1600 F. For Plant Hatch, fuel and plant-specific evaluations were performed to demonstrate conformance to these licensing criteria. 7 3.3 GENERIC ANALYSIS For the GE Boiling Water Reactor (BWR)-3/4 product lines, a generic Appendix K j conformance calculation was performed for the limiting hypothetical LOCA (Reference 1). The i limiting LOCA was determined from the nominal break spectmm at that break size and ECCS i component. failure combination that yielded the highest Nominal PCT. The Appendix K calculation was established as the basis for the licensing evaluation and determining factor operating limits. 1 The PCT calculated as described above maintains margins for licensing evaluations (i.e., the Licensing Basis PCT is at least the upper 95th percentile PCT). This was verified by separate l Analysis Procedure 3-3 _ ~, _

NEDC-32720 calculations to determine the upper 95th probability values of PCT at the most limiting conditions determined from the nominal break spectrum calculations. These calculations were performed to qualify the " Appendix K Procedure" as being sufficiently conservative. (Proprietary Information Deleted) As a result, this case was used to perform the Appendix K calculation. The results of the Appendix K calculation demonstrated that a discharge coefficient of 1.0 in the Moody Slip Flow Model yields the highest calculated PCT. The BWR-3/4 Licensing Basis PCT was established 4 using an adder to account for the required Appendix K models. The Upper Bound PCT (95% probability PCT) was also established generically to demonstrate that the Licensing Basis PCT was above the Upper Bound PCT. This generic evaluation demonstrated that a PCT margin in excess of 140 F existed between the Upper Bound PCT and the Licensing Basis PCT (Reference 2). 3.4 HATCH PLANT-SPECIFIC ANALYSIS The specific analysis performed for Plant Hatch consisted of break sizes ratgir.g from 0.05 ft2 to the maximum DBA recirculation suction line break (4.16 ft2). The break spectmm was first evaluated using the analysis assumptions for nominal calculations (Table 3-1). Limiting LOCA cases were then analyzed again with the analysis assumptions specified for the Appendix K calculation (Table 3-2). The Plant Hatch nominal and Appendix K PCT results were compared to assure that the PCT trends as a function of break size were consistent with one another and with those of the generic BWR-3/4 break spectrum curves (Section 3.1). j Analysis Procedure 3-4

d NEDC-32720 Table 3-1 ANALYSIS ASSUMPTIONS FOR NOMINAL CALCULATIONS (Reference 2) 4 1. Decay Heat 1979 American Nuclear Society (ANS) (Figure 3-1) 2. Transition Boiling Temperature Iloeje Correlation 3. Break Flow 1.25 HEM (1)(Subcooled) 1.0 HEM (1)(Saturated) 4. Metal Water Reaction EPRI Coefficients 5. Core Power 2763 MWt i 6. Peak Linear Heat Generation Rate (Proprietary Information Deleted)(2) 7. Bypass Leakage Coefficients Nommal Values i 8. Initial Operating Minimum Critical (Proprietary Information Deleted) Power Ratio (MCPR)(3) (Proprietary Information Deleted) 9. ECCS Water Enthalpy(Temperature) 88 Btu /lbm (120 F)

10. ECCS Initiation Signals See Table 4-3
11. Automatic Depressurization System 130-Second Delay Time (Table 4-3)
12. ECCS Available -

Systems remaining after worst 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 maxunizes PCT (1) HEM: Homogeneous Equilibrium Model (2) (Proprietary Information Deleted)

(3} (Proprietary Information Deleted) Analysis Procedure 3-5

NEDC-32720 Table 3-2 ANALYSIS ASSUMPTIONS FOR APPENDIX K CALCULATIONS (Reference 2) 1. Decay Heat 1971 + 20% Decay Heat (Figure 3-1) 2. Transition Boiling Temperature Transition boiling allowed during blowdown until cladding superheat exceeds 300 F 3. Break Flow Moody Slip Flow Break Flow Model with discharge coefficients of 1.0, 0.8, and 0.6 q 4. Metal-Water Reaction Baker-Just 5. Core Power 2818 MWt 6. Peak Linear Heat Generation Rate (l) (Proprietary Information Deleted) 7. Bypass Leakage Coefficients Same as Table 3-1 8. Initial Operating Minimum Critical (Proprietary Information Deleted) Power Ratio (MCPR)(2) (Proprietary Information Deleted) 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 Internu Pressure Same as Table 3-1
15. Fuel Exposure Same as Table 3-1 e

(1) (Proprietag Information Deleted) (2) (Proprietary Information Deleted) Analysis Procedure 3-6

.... _. =... - -... NEDC-32720 ) 1 . (Proprietary Information Deleted) _ i i i l t 1 l l l* I Figure 3-1. Normalized Decay Power for Nominal and Appendix K Cases t AnalysisProcedure 37 9 4

NEDC-32720 4.0 INPUT TO ANALYSIS 4.1 PLANT INPUTS The significant plant input parameters to the Plant Hatch 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 ECCS parameters. Plant Hatch will be making some changes in the ECCS power supply and eliminating the LPCI inverters. (The LPCI inverters currently power the LPCI injection valves and the recirculation loop discharge valves.) These changes will affect the ECCS available following a single failure. Table 4-4 identifies the combinations of break locations, single failures and available systems specifically analyzed for the Plant Hatch ECCS configuration (Figure 4-1). The systems remaining in this table represent the most limiting set of systems available for each failure for all possible emergency diesel alignments and for the ECCS power supply arrangement before and after the planned modifications. 4.2 FUEL PARAMETERS The SAFER /GESTR-LOCA analyses were performed with a bounding Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) at the most limiting combination of power and exposure (Table 4-2). (Proprietary Information Deleted) 4.3 ECCS PARAMETERS The Plant Hatch SAFER /GESTR-LOCA analysis incorporates values for some ECCS performance parameters that are more conservative than the expected equipment performance. The intent was to perform the analysis in a very conservative manner relative to the expected equipment performance to support relaxation of equipment requirements. Table 4-3 contains the speciSc EC CS performance input parameters used in the evaluation. l Input to Analysis 4-1

NEDC-32720 Table 4-1 PLANT OPERATIONAL PARAMETERS USED IN PLANT HATCH SAFER /GESTR-LOCA ANALYSIS Plant Parameters Nominal Appendix K Udt l Udt2 Udt 1 Udt2 Core Thermal Power (MWt) 2763(1) 2763 2818 2818 Corresponding Power (% of 2763 MWt) 100 100 102 102 Vessel Steam Output (M1b/hr) 11.54 11.98 11.81 12.26 Core Flow (Mlb/hr)(2) 78.5 77.0 78.5 77.0 Feedwater Temperature ( F) 397.5 425.1 399.5 427.3 Vessel Steam Dome Pressure-(psia) 1050 1050 1053 1053 Maximum Recirculation Suction Line 4.14 3.99 4.14 3.99 Break Area (g2)(3) (1) The core thermal power correspcnds to a thermal power level of 108% of the current licensed value of 2558 MWt. (2) The break spectmm determination of worst single failure was performed at a rated core flow of 78.5 Mlb/hr for Unit I and 77.0 NCb/hr for Unit 2. Other analyses were performed as described in Section 5 to cover operation from 71.4 h0b/hr to 82.4 Mlb/hr for Unit I and 70.0 Mlb/hr to 80.9 Mlb/hr for Unit 2. 2 (3) Includes 0.02 R drainline. Input to Analysis 4-2

NEDC-32720 Table 4-2 FUEL PARAMETERS USED IN THE PLANT HATCH SAFER /GESTR-LOCA ANALYSIS Analysis Value Fuel Parameter GE9 GE13 PLHGR (kW/ft) - Appendix K - Nominal MAPLHGR (kW/ft) - Appendix K - Nominal Worst Case Pellet Exposure for ECCS Evaluation (l)(mwd /MTU) Initial Operating MCPR(2) - Appendix K - Nominal Axial Peaking Factor 1.4 1.4 Number ofFuel Rods per Bundle 60 74 (Proprietary Information Deleted) (1) Represents the limiting operating condition resulting in the maximum calculated PCT at anytime during the fuel bundle life. (2) (Proprietary Information Deleted) Input to Analysis 4-3

NEDC-32720 Table 4-3 PLANT HATCH SAFER /GESTR ANALYSIS ECCS PARAMETERS 1. Emergency Diesel Generators Variable Units Analysis Value Unit 1 Unit 2

a. Emergency Diesel Generator Startup Time sec 21 21 (from initiation signal to emergency bus being powered) 2.

Low Pressure Coolant Injection (LPCI) System Variable Units Analysis Value Unit 1 Unit 2

a. Vessel Pressure at which Flow May Commence psid 207 223
b. Minimum Rated Flow Vessel pressure at which LPCI flow rates psid (vessel 20 20 are quoted to drywell) 2 LPCI pumps injecting into 1 loop gpm 15,660 15,570 2 LPCI pumps injecting into 2 loops 8Pm 17,280 18,180 Initiating Signals and Setpoints c.

Low-Low-Low Water Level inches above TAFm TAFW vessel "zero" or High Drywell Pressure Psig 2.0* 2.0*

d. Maximum Vessel Pressure at which LPCI psia 380 380 Injection Valve Can Open Maximum Allowable Time Delay from Initiating sec 64 64 c.

Signal to Pump at Rated Speed and Capable of Rated Flow f. Injection Valve Stroke Time @ sec 63

  • 63
  • l Input to Analysis 4-4

NEDC-32720 Table 4-3 PLANT HATCH SAFER /GESTR ANALYSIS ECCS PAPAMETERS (Continued) 2. Low Pressure Coolant Injection (LPCI) System (Continued) Variable Units Analysis Value Unit 1 Unit 2

g. Recirculation Discharge (DSCG) Vulve@)

Pressure permissive for closure Psia 315 315

  • DSCG valve stroke time sec 43 43 3.

Core Spray (CS) System Variable Units Analysis Value Unit 1 Uniti a. Vessel Pressure at which Flow May Commence psid 284 375

b. Minimum Rated Flow Vessel pressure at which CS flow rate is psid (vessel 113 113 e

quoted to drywell) 1 CS loop gpm 4000 4000

c. Initiating Signals and Setpoints U

U Low-Low-Low Water Level inches above TAF) TAF) e vessel "zero" or High Drywell Pressure psig 2.0(2) 2.0(2) e

d. Run-out Flow at 0 psid (vessel to drywell) for gpm 5470 5310 One CS Loop
e. Maximum Vessel Pressure at which CS psia 380 380 Injection Valve Can Open Input to Analysis 4-5

NEDC-32720 Table 4-3 PLANT HATCH SAFER /GESTR ANALYSIS ECCS PARAMETERS (Continued) 3. Core Spray (CS) System (Continued) Variable Units Analysis Value Unit 1 Unit 2 f. Injection Valve Stroke Time sec 20* 20*

g. Maximum Allowable Delay Time from Initiating sec 34 34 Signal to Pump at Rated Speed with Emergency Diesel Power 4.

High Pressure Coolant Injection (HPCI) System @ Variable Units Analysis Value Unit 1 Unit 2

a. Operating Vessel Pressure Range psia 165 to 165 to 1210 1210
b. Minimum Flow Required Over the Entire gpm 4250 4250 Operating Vessel Pressure Range
c. Maximum Vessel Pressure at which Pump Can psia 1210 1210 Inject Flow
d. Initiating Signals and Setpoints Low-Low Water Level inches above 89 89 TAF*

or High Drywell Prcssure Psig 2.0* 2.0A

e. Maximum Allowable Time Delay from Initiating (sec) 50 50 Signal to Rated Flow Available and Injection Valve Wide Open input to Analysis

~ 4-6

NEDC-32720 Table 4-3 PLANT HATCH SAFER /GESTR ANALYSIS ECCS PARAMETERS (Continued) 5. Automatic Depressurization System (ADS) Variable Units Analysis Value Unit 1 Unit 2 a.

1. Total Number of Relief Valves with ADS 7

7 Function

2. Total Number ofRelief Valves with ADS 5

5 Function Assumedin Analysis *

b. Pressure at which ADS Capacity Is Quoted psig 1080 1090
c. Minimum Flow Capacity of Any 5 Valves Ibm /hr 3.94 4.345 x10W x10*)
d. Initiating Signals and Setpoints Low-Low-Low Water Level inches above TAFm TAFW e

vessel "zero" High Drywell Pressure psig 2.0* 2.0W e DE Low-Low-Low Water Level inches above TAFW TAFW Vessel"zero" High Drywell Bypass Timer minutes 13 13 e Additional Delay e. ADS Timer Delay sec 130 130 1 l l Input to Analysis 4-7 l

Table 4-3 PLANT HATCH SAFER /GESTR ANALYSIS ECCS PARAhETERS (Continued) NOTES: (I) Top of Active Fuel (358 inches above vessel zero). (2) No credit is taken for the high drywell pressure initiation signal in the Appendix K analysis. (3) Flow is assumed to be achieved prior to valve being full open. (4) No credit is taken for HPCI in these evaluations. Therefore, the parameter values in this table have no bearing on the ECCS results presented in this report. (5) Two ADS valves are assumed unavailable to conservatively bound the scenario of an ADS valve out of service and a single failure of another ADS valve for all LOCA cases. (6) Power to these valves is assumed to coincide with the 21-second emergency diesel generator startup time. haput to Analysis 4-8

NEDC-32720 Table 4-4 PLANT HATCH SINGLE-FAILURE EVALUATIONm The table below shows the various combinations of Automatic Depressurization System (ADS), High Pressure Coolant Injection (HPCI) System, Low Pressure Coolant Injection (LPCI) System and Core Spray (CS) System which might be operable in an assumed Design Basis Accident situation. In performing the ECCS performance analysis with SAFER /GESTR, it is assumed that no postulated single active component will result in less than certain minimum combinations of systems remaining operable. The following single active failures will be considered in the ECCS performance evaluation: Assumed Failure (2) Recirculation Suction Break Recirculation Discharge Break Systems Remaining" Systems Remaining Station Service Battery 2CS+2LPCI+ ADS 2CS+ ADS Diesel Battery: Swing Diesel B 2CS+1LPCI+HPCI+ ADS 2CS+HPCI+ ADS Dedicated Diesel (A or C) ICS+2LPI2+HPCI+ ADS ICS+1LPCI+HPCI+ ADS LPCIInjection Valve 2CS+2LPCI+HPCI+ ADS 2CS+HPCI+ ADS Diesel Generator: Swing Diesel B 2CS+1LPCI+HPCI+ ADS 2CS+HPCI+ ADS Dedicated Diesel (A or C) ICS+3LPI2+HPCI+ ADS 1CS+1LPCI+HPCI+ ADS NOTES: U} The single failures shown in this table reflect the most limiting set of single failures based on the configuration of both units. The key to the abbreviations in this table are as follows: 1LPCI - one LPCI pump injects into one recirculation loop 2LPCI - two LPCI pumps inject into one recirculation loop 2LPI2 - two LPCI pumps inject into two recirculation loops 3LPI2 - three LPCI pumps inject into two recirculation loops input to Analysis 4-9

NEDC-32720 Table 4-4 (Cont.) (2) Other postulated failures are not specifically considered because they all result in at least as much ECCS capacity as one of the above assumed failures. With no credit for HPCI operation (see Note 4), the systems remaining for the diesel battery failure bound all other possible single failures. U} Systems remaining for the recirculation suction break are applicable to all non-ECCS line breaks. For an ECCS line break, the systems remaining are those listed for the recirculation suction break, less the ECC train in which the break is assumed. (4) No credit is taken for HPCI operation. 0} All LOCA analyses are performed assuming two ADS valves are unavailable to conservatively bound the scenario of an ADS valve out of service and a single failure of another ADS valve for all LOCA cases. Input to Analysis 4-10

NEDC-32720 1 A r i---------- '----------- a--, s A A C p D B B N / V A ( O E s LJ Figure 4-1. Plant Hatch ECCS Configuration with Normal Diesel Alignment i Input to Analysis 4-11

NEDC-32720 5.0 RESULTS 5.1 BREAK SPECTRUM CALCULATIONS 5.1.1 Recirculation Line Breaks A sufficient number of breaks were analyzed for Plant Hatch with the potentially limiting single failures using nominal assumptions (Table 3-1) and the inputs discussed in Section 4.0. The potentially limiting single failures included a diesel battery for the swing diesel B and the dedicated diesels A and C, since they result in the muumum number of systems available. Both of these failures leave three low pressure ECC systems available (with no credit taken for the HPCI). The swing diesel battery failure leaves 2CS + ILPCI + ADS available and the dedicated diesel battery failure leaves ICS + 2LPI2 (2 LPCI injecting into two recirculation loops) + ADS. The total ECCS flow rate is about the same for both cases, but the injection is more into one vessel region than the other for each case. (Proprietary Infonnation Deleted) Key break spectrum points were analyzed again using the Appendix K input assumptions. The 100%, 80%, and 60% DBA cases also satisfy the Appendix K requirement for using the Moody Slip Flow Model with three discharge coefficients of 1.0,0.8, and 0.6, respectively. Table 5-2 summarizes the Appendix K PCT results for the break sizes analyzed for Plant Hatch. (Proprietary Information Deleted) Again, Table 5-2 summarizes the PCT results for all recirculation break sizes analyzed specifically for Plant Hatch. The Appendix to this report contains the plots of the system responses for the DBA for both the nominal calculation and Appendix K calculation. k Residts 5-1

NEDC-32720 5.1.2 Non-Recirculation Line Breaks Table 5-3 shows the results for non-recirculation line breaks. These analyses results clearly demonstrate that these postulated breaks are significantly less limiting than the postulated recirculation line breaks. 5.2 COMPLIANCE EVALUATIONS 5.2.1 Licensing Basis PCT Evaluation The results in Section 5.1.1 confirm that the limiting break is the recirculation suction line DBA, which is consistent with the BWR-3/4 generic conclusions. The Plant Hatch calculations demonstrate that GE13 is the limiting fuel type. (Proprietary Information Deleted) 5.2.2 Upper Bound PCT Evaluation For the BWR-3/4 plants, the generic Appendix K PCT versus break size curve exhibits the same trends as the generic nominal PCT versus break size curve, and the limiting LOCA determined from nominal PCT calculations is the same as that determined from the Appendix K PCT calculations. The Plant Hatch specific results presented in Section 5.1 demonstrate the applicability of the BWR-3/4 generic nominal PCT and Appendix K PCT versus break size curves to the Plant Hatch. As described in Section 4, the Plant Hatch analysis utilizes ECCS parameters conservatively established with respect to both the actual performance of the Plant Hatch ECCS and the ECCS parameters assumed in the generic determination of the statistical upper bound. (Proprietary Information Deleted) The plant variable uncertainty term reflects the sensitivity of the SAFER calculated PCT to uncertainties in decay heat, PLHGR, break flow, initial stored energy and the minimum temperature for film boiling. Since the Plant Hatch PCT results are higher than the generic study, the plant variable uncertainty term was evaluated on a plant-specific basis using the propagation of errors method described in Reference 2. (Proprietary Information Deleted) Results 5-2

NEDC-32720 By verifying that the Licensing Basis PCT for Plant Hatch is greater than the Upper Bound (95th percentile) PCT, the level of safety and conservatism of this analysis meets the NRC acceptance criterion. 5.3 ALTERNATE OPERATING MODE CONSIDERATIONS 5.3.1 Power Uprate The analyses presented in this document were performed to support Plant Hatch operation up to a bounding thermal power level of 2763 MWt, corresponding to 108% of the current licensed value of 2558 MWt. (Proprietary Information Deleted) The results of this analysis conservatively bound the cunent plant operation. 5.3.2 Extended Load Line Limit Analysis (ELLLA) and ARTS Plant Hatch has implemented ELLLA and ARTS. The higher rod line in the ELLLA region permits reactor operation at rated power for core flows below rated. (Proprietary Information Deleted) For the higher power level with its corresponding power / flow map, the flow-dependent MAPLHGR factors will be re-evaluated for the reload in which the power uprated conditions will be implemented. 5.3.3 Increased Core Flow The impact on LOCA results due to increased core flow (ICF) operation, corresponding to 105% of rated core flow, was evaluated for a thermal power of 2763 MWt using the same ECCS parameters as used for the rated core flow conditions. (Proprietary Information Deleted) Results 5-3

NEDC-32720 5.3.4 Single-Loop Operation (SLO) (Proprietary Information Deleted) 5.3.5 Feedwater Heater Out-of-Service and Final Feedwater Temperature Reduction (Proprietary Information Deleted) 5.4 MAPLHGR LIMITS Current GE BWR MAPLHGR limits (as a function of exposure) are based on the most limiting value of either the MAPLHGR determined from ECCS limits (PCT) or the MAPLHGR determined from fuel thermal-mechanical design analysis limits. The bounding MATLHGRs used in this Plant Hatch SAFER /GESTR-LOCA analysis (i.e., 13.9 kW/ft for GF9 and 13.4 kW/ft for GE13) are higher than the thermal-mechanical MAPLHGR limits for these fuel designs. This analysis establishes that for all GE9 and GE13 fuel designs at Plant Hatch, the MAPLHGR is not limited by LOCA/ECCS considerations. Results 5-4 l

NEDC-32720 Table 5-1 COMPARISON OF DBA RECIRCULATION SUCTION LINE BREAK RESULTS FOR PLANT HATCH UNITS 1 AND 2 Unit 1 Unit 2 Analysis Assumptions Single GE9 GE13 GE9 GE13 Failure PCT (*F) PCT ('F) PCT ( F) PCT ( F) l Nominal Diesel Battery (Ded. Diesel) Nominal Diesel Battery (Swing Diesel) Appendix K Diesel Battery (Ded. Diesel) Appendix K Diesel Battery (Swing Diesel) (Proprietary Information Deleted) ) e 9 Results 55 ?

NEDC-32720 Table 5-2

SUMMARY

OF RECIRCULATION LINE BREAK RESULTS FOR PLANT HATCH (IX2) Peak Local Peak Local Break Size and Nominal Oxidation APP.K Oxidation Recirc Break Location PCT (*F) (%) PCT (*F) (%) DBA, Suction 80% DBA, Suction 60% DBA, Suction DBA, Discharge 1.0 ft2, Discharge 0.5 ft2, Discharge

  • (3) 0.3 ft2, Discharge
  • (3) 0.1 ft2, Discharge 0.08 ft2, Discharge 0.05 ft2, Discharge (Proprietary Information Deleted)

(1) (Proprietary Information Deleted) (2) (Proprietary Information Deleted) (3) (Proprietary Information Deleted) Results 5-6

NEDC-32720 Table 5-3 PLANT HATCH RESULTS FOR NON-RECIRCULATION LINE BREAKS (1X2) - Break Size Core-Wide - Location Peak Local Metal-Water - Single Failure PCT (*F) Oxidation (%) Reaction (%) (Nominal) Note 1 2 0.27 f1 Core Spray Line Break Battery (Nominal)

  • (3)

Note 1 2 2.41 ft inside Containment Steamline Break Battery (Nominal) Note i 2 2.62 ft Outside Containment Steamline Break Battery (Nominal)

  • (2)

Note 1 2 0.72 fl Feedwater Line Break Battery (Proprietary Information Deleted) (1) (Proprietary Information Deleted) (2) (Proprietary Information Deleted) (3) (Proprietary Information Deleted) Resuhs 5-7

NEDC-32720 Table 5-4 EXTENDED LOAD LINE LIMIT ANALYSIS RESULTS COMPARISON FOR PLANT HATCH 0) Limiting LOCA: DBA - Recirculation Suction Line Break, Dedicated Diesel Failure Peak Cladding Temperature ( F) Analysis ELLLA Base Case Basis (% Core Flow) (% Core Flow) Nominal Appendix K (Proprietary Information Deleted) O) (Proprietary Information Deleted) Results 5-8 l

NEDC-32720 Table 5-5 INCREASED CORE FLOW RESULTS COMPARISON FOR PLANT HATCH (O Limiting LOCA: DBA - Recirculation Suction Line Break Dedicated Diesel Failure Peak Cladding Temperature ( F) Analysis ICF Base Case Basis (105% Core Flow) (100% Core Flow) Nominal Appendix K (Proprietary Information Deleted) (D (Proprietary Information Deleted) Results l 59 l

NEDC-32720 Table 5-6 SINGLE-LOOP OPERATION RESULTS COMPARISON FOR PLANT HATCHC0 Limiting LOCA: DBA - Recirculation Suction Line Break Dedicated Diesel Failure Peak Cladding Temperature ( F) Analysis SLO Base Case Basis (Power / Flow) (Power / Flow) Nominal Appendix K (Proprietary Information Deleted) (O (Proprietary Information Deleted) Results 5-10

NEDC-32720 f Table 5-7 FEEDWATER TEMPERATURE REDUCTION RESULTS COMPARISON FOR PLANT HATCH 0X2) Limiting LOCA: DBA - Recirculation Suction Line Break, Dedicated Diesel Failure 'f Peak Cladding Temperature ( F) Analysis FWTR Base Case Basis (-70 F) (ELLLA) Nominal Appendix K (Proprietary Information Deleted) O) (Proprietary Information Deleted) (2) (Proprietary Information Deleted) Results 5-11

....... -. - ~.. 3 s ? NEDC-32720 2 4 1 i '] t 1 s 1 M 3 l (Propri; tryInformation Deleted) 1. i -1 ) e i 1 1 2 i-1 i 1 4 .J 1 d s l 5 E i e a 4 a ,e 1 i i. \\ Figure 5-1. Nominal and Appendix K LOCA Break Spectrum Results Results 5-12

l GE ProprietaryInformation NEDC-32720P

6.0 CONCLUSION

S LOCA analyses have been performed for Plant Hatch using the GE SAFER /GESTR-LOCA Application Methodology approved by the NRC. These analyses were performed to demonstrate conformance with 10CFR50.46 and Appendix K and thus establish a revised licensing basis for Plant Hatch with the GE SAFER /GESTR-LOCA methodology. The Plant Hatch SAFER /GESTR-LOCA results presented in Section 5 demonstrate that a j sufficient number of plant-specific PCT points have been evaluated to establish the shape cf 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 DBA. Table 6-1 summarizes the key SAFER /GESTR licensing results for Plant Hatch. The analyses presented are performed in accordance with NRC requirements and demonstrate conformance with the ECCS acceptance criteria of 10CFR50.46. Therefore, the results I documented in this report may be used to provide a new LOCA Licensing Basis for Plant Hatch. The results are valid for fuel designs with comparable geometry to the fuels analyzed and for MAPLHGR and PLHGR values less than or equal to those shown in Table 4-2. l With the explicit verification that the Licensing Basis PCT for Plant Hatch is greater than l the Upper Bound (95th percentile) PCT, the !cvel of safety and conservatism of this analysis j meets the NRC approved criteria. Therefore, the requirements of Appendix K are satisfied. l I O 9 t ~~ Conclusions 6-1

GE ;4oprietaryInformation NEDC-32720P Table 6-1 SAFER /GESTR-LOCA LICENSING RESULTS FOR PLANT HATCH 1. Limiting Fuel Type GE13 2. Limiting Break DBA Suction 3. Limiting Failure Dedicated Diesel Battery 4. Peak Cladding Temperature (Licensing - 1688'F Basis) 5. Estimated Upper Bound PCT (95% <1600 F Probability PCT) 6. Maximum Local Oxidation < 1.2 % 7. Core-Wide Metal-Water Reaction <0.1% f f Conclusions 6-2

GEProprietaryInformation NEDC-32720P

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. (Proprietary Information Deleted) 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. ' 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. 5. Letter, D.G. Eisenhut (NRC) to R.L Gridley (GE), " Safety Evaluation Report on Revision of Previously Imposed MAPLHOR (ECCS-LOCA) Restriction for BWRs at Less Than Rated Core Flow", May 19,1978. 6. "GE13 Compliance with Amendment 22 of NEDE-240ll-P-A (GESTAR-II)", NEDE-32198P, December 1993. 7. "Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis", NEDC-31376P, General Electric Company, December 1986. 8. " Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Edwin I. Hatch Nuclear Plant, Units 1 and 2", NEDC-30474P, December 1983. 9. Letter (MFN 023-90), R.C. Mitchell (GE) to R.C. Jones (NRC), " Reporting of Changes and Errors in ECCS Evaluation Models", June 13,1990. l l l References 7-1 l l

~. GEProprietaryInformation NEDC-32720 d APPENDIX j SYSTEM RESPONSE CURVES FOR NOMINAL AND APPENDIX K DBA RECIRCULATION LINE BREAKS t Included in this Appendix are the system response curves for Plant Hatch. Table A-1 contains the figure numbering sequence for the nominal and Appendix K DBA recirculation breaks. 4 (Proprietary System Response Results Deleted) j i l 4 e A-1

hl r GENuclearEnergy l 175 Curtner Avenue SanJose, CA 95!25 4 i d j ) ? I l l 1 I \\ P i s j t 1 i 1 a l l \\ r 4 9-i 1 l I r}}

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