ML20079P121
| ML20079P121 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 10/31/1991 |
| From: | John Miller, Jacqueline Thompson GULF STATES UTILITIES CO. |
| To: | |
| Shared Package | |
| ML20079P109 | List: |
| References | |
| EA-PT-91-0003-S, EA-PT-91-0003-S-S01, EA-PT-91-3-S, EA-PT-91-3-S-S1, NUDOCS 9111120233 | |
| Download: ML20079P121 (209) | |
Text
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i EA-PT-91-0003-S ,
RIVER BEND STATION PLANT TRANSIENT ANALYSIS METIIODOLOGY .
, Supplement 1 Delta cPR Methodology and Additional Benchmarks Nonproprietary Versio-October 1991 Contributors John P. Egan Lynn A. Leatherwood Stone S. Luo t Thomas W. Oliphant
- David R. Swope
-7 :
Reviewed: ,/ / d%
James-I/. Thompson'
~
Acih--
Supervisor - Thermal-Hydraulic Analysis l
Approved:
Jos WhNs - \h Ofi _
Dir horS.QMiller-
- Engineering _ Analysis Gulf States _ Utilities Company.
River Bond Station
, P.O. Box 220 '
l St. Francisville, LA_70775
~
9111120233 911031 PDR ADOCK 05000458 P PDR
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EA-PT-91-0003-S IMPORTANT NOTICE REGARDING CONTENTS OF Tills DOCUMENT PLEASE READ CAREFULLY This document was prepared by Gulf States Utilities Company thr the use of the U.S. Nuclear Regulatory Commission in matters regarding the operating license for the River Bend Station.
To the best of the issuer's knowledge, this document contains work performed in accordance with sound engineering practice and is a true and accurate representation of the facts.
The work reported herein is the property of Gulf States Utilities Company, and any usage other than as described above is prohibitedi Other than for the intended usage, neither Gulf States Utilities Company, nor any of its employees or officers, nor any other person acting on its behalf:
o Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the informa-tion contained in this repo:1, or that the use of any information, apparatus, method, or procen disclosed herein would not infringe privately owned rights; or l
o Assumes any liabi!!les with respect to the use of, or for damages -
resulting from the use of, any information, apparatus, method, or process disclosed in this report, l
l EA-PT-91-0003-S t
Acknowledgements The authors gratefully acknowledge the assistance of Dr.
D. A. Prelewicz, D. J. Gitnick, and G. B. Peeler of SCIENTECH, Inc. in the completion of this project and the preparation of this report.
In addition, the authors acknowledge the support provided by the Core Analysis Group, who provided ..nput data for ESCORE and the one-dimensional reactor kinetics used by RETRAN.
For clerical support, the authors would like to thank Jerri Fontenot and Mancy Scott.
Also, the authors gratefully acknowledge the long-term support from Gulf States Utilities; in particular, Mr.
Jim Doddens, Mr. Ken Suhrke, and Mr. Mel Sankovich. This report is the result of a development effort begun in 1983 by Mr. T. M. Howe.
The authors also acknowledge the contribution of student engineers Kathy L. Ilugle-of the University of Louisville and Eric Ballon of the University of Florida.
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1 EA-PT-91-0003-S Table of Contents l l
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . 12 2.0
SUMMARY
. . . . . . . . . . . . . . . . . . . 15 3.0 COMPUTER PROGRAMS USED IN ACPR METHODOLOGY . . 18 3.1 SYSTEM TRANSIENT ANALYSIS . . . . . . . . 18 3.2 EDTRAN LINK TO HOT CHANNEL ANALYSIS . . . 20 3.3 RETRAN HOT CHANNEL ANALYSIS . . . . . . . 21 3.4 EDTRAN ACPR CALCULATION . . . . . . . . . 21
, 4.0 HOT CHANNEL MODEL . . . . . . . . . . . . . . 23 5.0 CRITICAL POWER RATIO METHODOLOGY . . . . . . . 29 6.0 UNCERTAINTY ANALYSIS . . . . . . . . . . . . . 35 6.1 SENSITIVITY STl')IES . . . . . . . . . . . 38 6.1.1 Time Step __and Nodalization Sensitivity . . . . . . . . . . 40 6.1.2 Nuclear Model Sensitivities . . 41 6.1.3 Heat Transfer Model Sensitivities . . . . . . . . . 43 6.1.4 Core Hydraulic Model Sensitivities . . . . . . . . . 44 6.1.5 Separator /Drver Hydraulic Model sensitivities . . . . . . . . . 44 6.1.6 Recirculation System Hydraulic Model sensitivities . . . . . . 45 6.1.7 Etla_tn Line Hydraulic Model Eensitivities . . . . . . . . . 45 6.1.8 Feedwater Flow Sensitivities . 46 6.2 RESULTS OF SENSITIVITY ANALYSES . . . . . 47 7.0 APPLICATION OF THE RBF' METHODOLOGY TO CALCULATION OF THERMAL MARGINS . . . . . . . . 68 7.1 FUEL VENDOR ANALYTICAL RESULTS . . . . . 69 7.2 CALCULATION OF THERMAL MARGIN REQUIREMENTS . . . . . . . . . . . . . . 70 7.2.1 Lpad Re%ction without Bypass . 71 7.2.2 Pressure Reculator Failure . . 79 7.2.3 Feedwater Controller Failure . 87 7.3 EVALUATION OF- OFF-NOMINAL OPERATING CONDITIONS . . . . . . . . . . . . . . . 96 7.3.1 Power-Dependent MCPR Limit . . 97 7.3.2 flow-Dependent MCER Limit . . . 97 4
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EA-PT-91-0003-S Table of Contents -t 7.4 PROTECTION OF PIPING AND COMPONENTS FROM ;
OVERPRESSURIZATION . . . . . . . . . . . 99 8.0 COMPARISON TO RIVER BEND STATION TRANSIENTS . 179
- 8.1 TURBINE RUNBACK EVENT . . . . . . . . . . 179 8.2 PRESSURE INCREASE TRANSIENT . . . . . . . 182 9.0 REVISED PEACH BOTTOM UNIT 2 TRANSIENT h COMPARISONS . . . . . . . . . . . . . . . . . 202
10.0 REFERENCES
. . ., . . . . . . . . . . . . . . 208 1
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i EA-PT-91-0003-S List of Tables No. Title Page 1.1 R.BS Design Features . . . . . . . . . . . . . 14 6.1 Sequence of Events for RBS Load Rejection without Bypass Transient . . . . . . . . . . . 51 6.2 Sequenco of Events for RBS Pressure Regulator Failure Transient . . . . . . . . . . . . . . 52 6.3 Sequence of Events for RBS Feedwater Controller Failure Transient . . . . . . . . . 53 6.4 Sensitivity Studies for RBS Lead Roj ou lon without Bypass (LRNB) Transient . . . . . . . 54 6.5 Sensitivity Studies for RBS Pressure Regulator Fail Downscale . . . . . . . . . . . . . . . . 57 6.6 Sensitivity Studies for RBS Feedwater Controller Failure (FWCF) Transient . . . . . 60 6.7 Adjustments to the RBS RETRAN Best-Estimate Model for Design-Basis Calculations . . . . . 63 6.8 Statistical Adjustment Factors for the RBS RETRAN Model . . . . . . .. . . . . . . . . . 64 7.1 Initial Conditions for Pressurization Events . 102 7.2 Initial Conditions for Overpressure Events . . 103 7.3 Sequence of Events for RBS Load Rejection without Bypass Transient - Design Basis . . . 104 7.4 Sequence of Ever.ts for RBS Pressure Regulator Failure Transient . . . . . . . . . . . . . . 105 7.5 Sequence of Events for RBS Feedwater Controller Failure Transient . . . . . . . . . 106 8.1 Initial Conditions, RBS. Turbine Runback Transient . . .. . . . . . . . . .. . . .- . 186 8.2 Sequence of Events, RBS Turbine Runbsek Transient (Scram 86-12) . . . . . . . . . . . 187 8.3 Initial Conditions, RBS Pressure Increase Transient . . . . . . . . . . . . . . . . . 188-9.1 Peach Bottom 2 EOC2 Turbine Trip Results . . . 204 6
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EA-PT-91-0003-S List of Figures No. Title Page 3.1 Block Diagram for CPR Methodology . . . . . . 22 4.1 RBS RETRAN Hot Channel Nodalization . . . . . 28 5.3 Graphical Illustration of CPR . . . . . . . . 34 6.1 Schematic Diagram of Transient-Specific Uncertainty Methodology . . . . . . . . . . . 65 6.2 Determination of Transie;1t-Specifia SAF . . . 66 6.3 Determination of OLMCPR . . . . . . . . . . . 67 7.1 RETRAN Load Rejection Event Response -
Neutron Flux . . . . . . . . . . . . . . . . 107 7.2 RETRAN Load Rejection Event Response -
Average Surface Heat Flux . . . . . . . . . . 108 7.3 RETRAN Load Rejection Event Roslonse - Core Inlet Flow . . . . . . . . . . . . . . . . . 109 7.4 RETRAN Load Rejection Event Responso -
Pressure Risa (psi) . . . . . . . . . . . . . 110 7.5 RETRAN Load Rejection Event Res:ponse -
Relief Valve Flow (%)_. . . . .. . . . . . . 111 7.6 .RETRAN Load Rajection Event Response -
Narrow Range Level (inches) . . . . . . . . . 112 7.7 RETRAN Load Rejection Event Response -
Vessel Steam Flow . . . . . . . . . . . . . . 113 7.8 RETRAN Load ' action Event Response -
Turbine Steam aw . . . . . . . . . . . . . 114 7.9 RETRAN Load Rejection . Event Response- -
Feedwater Flow . . . . . . . . . . . . . . . 115 7.10 RETRAN Load Rejection Event Response - Total Reactivity ($) . . . . . . . . . . . . . . . 116 >
7.11 Comparison of RETRAN' and GE LRNB - Base - ,
Average Surface Heat Flux . . . . . . . . . . 117_
7.12 Comparison of RETRAN and GE LRNB - Base -
Core Inlet Flow . . . . . . . . . . . . . . . 118 7.13 Comparison of RETRAN and ' GE LRNB - Base -
Pressure Rise (psi)-. . . . . . . . . . . . . 119 7.14 Comparison of RETRAN and GE ' LRNB - Base -
Relief Valve Flow-(%) . - . . . . . . . . . . . 120 7.15 Comparison of RETRAN and GE LRNB -
Base -
Vessel Level (inches) . . . . . . . . . . .. 121 7.16 Comparison of RETRAN and GE LRNB - Base -
Vessel Steam Flow (%) .. . . . . .. . . . . 122 1.17 Comparison.of RETRAN and GE LRND -
Base -
Total Reactivity ($) . . . . . . . . .. . . 123 7.18 RETRAN Pressure Regulator Failure Response -
Neutron Flux (%) . - , . . .. . .. . . . .. . 124 1 7
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EA-PT-91-0003-S List of Figures No. Title Page f
7.19 RETRAN Pressure Regulator Failure Response -
kverage Surface Heat Flux (%) . . . . . . . . 125 7.20 dETRAN Pressure Rugulator Failure Rosponse -
Core Inlet Flow (%) . . . . . . . . . . . . . 126 7.21 RETRAN Pressure Raqulator Failure Response -
Pressure Rise (p'si) .. . . . . . . . . . . . 127 7.22 RETRAN Pressure Regulator Failure Response -
Relief Valve Flov (%) . ._ . . . . . . . . . . 128 7.23 RETRAN Pressure Regulator Failure Response -
Vessel Level (inches) . . . . . . . . . . . . 129 7.24 RETRAN Pressure Regulator Failure Response -
Vessel Steam Flow (%) . . . . . . . . . . . . 130 7.25 RETRAN Prersure Regulator Failure Response -
4 Turbine Steam Flow (%) . . . . . . . . . . . 131 7.26 RETRAN Pressure Regulator Failure Response -
Feedwater Flow (%)_ . . . . . . . . . . . . . 132 7.27 RETRAN Pressure Regulator Failure Response -
Total Reactivity (S) . . . . . . . . . . . . 133 7.28 GE and RETRAN Pressure Regulator Failure -
Base - Average Surface Heat Flux- . . . . . . 134 7.29 GE and RETRAN Pressure Regulator Failure -
Base - Core InlC: Flow-(%) . . . . . . . . . 135 7.30 GE and RETRAN Pressure Regulator Failure -
Base - Pressure Rise (psi) . . . . . . . . . 136 7.31 GE and RETRAN Pressure Regulator Failure -
Base - Relief Valve Flow (%) . . . . . . . . 137 7.32 GE and RETRAN Pressure Regulator Failure -
Base - Vessel Level (incaa) . . . . . . . . 138 E 7.33 GE and RETRAN Pressure Regulator Failure -
Base - Vessel Steam Flow (%) . . . . . . . . 139 7.34 GE . and RETRAN Pressure Regulator Failure -
Base - Turbine Steam Flow . . . . . . . . . . 140 7.35 GE and RETRAN Pressure-Regulator Failure -
Base - Total-Reactivity . . . . . . . .. . . 141 7.36 GE and RETRAN Pressure Regulator Failure -
MOD - Average Surface Heat Flux . . . . . . . 142 7.37 GE and RETRAN Pressure Regulator Failure -
MOD - Core-Inlet Flow (%)_. . . . . . . . . . 143 7.38 GE and RETRAM. Pressure _ Regulator Failure -
MOD - Pressure Rise (psi) . . .. . . . . . . 144 7.39 GE and RETRAN Pressure Regulator Failure -
MOD - Relief Valve Flow (%) . . . . . . . . . 145 7.40 GE and RETRAN Pressure Regulator Failure -
MOD - Vessel Level (inches) . . . . . . . . . 146 d
-EA-PT-91-0003-S List of Figures No. Title Page 7.41 GE and RETRAN Pressure Regulator Failure -
MOD - Vessel' Steam Flo'.1 (O _. . . . . . . . . 147 7.42 GE and RETRAN Pressure-Ragulator Failure -
MOD - Turbine Steam-Flow (%)- . . . . _ . . . . 148 7.43 GE and RETRANLPressure_ Regulator Failure -
MOD -_ Total' Reactivity (() . .- . . ..... .. . 149 7.44 RETRAN Feedwater Controller Failure Response
- Neutron. Flux =(%) . . - - . . . .- . .. . . . .. 150 7.46 RETRAN Feedwater Controller _ Failure Response
- Average Surface Heat Flux (%):._. . . . . . 151-7.46 RETRAN Feedwater Controller Failure' Response
- Core Inlet Flow (%) . . . _ . . . - . . . .. .. . .
152-7.47 RERAN Feedwater Controller Failure Response
- Core Subcooling (Btu /lbm).. .- . .- . .: . . .-153 7.48 RETRAN Feedwater Controller Failure Response 4
--Pressure' Rise (psi) . . . . ... . . . . . . - . - .
154 7.49 RETRAN Feedwater Controller Failure Response -
- Relief Valv- Flow (%) , . . . . - , . .- . . .. 155 7.50 RETRAN Feedwater Controller Failure Response
- Bypass Flow (%) . . . . . . . . . . .- . . . . . 156 7.51 RETRAN Feedwater Controller Failure Response
- Vessel Level (inches) . . . . - . . . . . . . ..-157-
'7.52 RETRAN Feedwater Controller Failure Response
- Vessel Steam-Flow.(%).., .. .; .: . .. . . - . . 158 7.53 RETRAN Feedwater Controller Failure Response-
- Turbine Steam Flow (%)- .. . . .. . . ..
. . 159 7.54 RETRAN Feedwater Controller Failure Response
- Feedwater Flow (%) . . . .:.-. .- . . .. .. .- 160 7.55 RETRAN Feedwater Controller Failure Response.-
- Total. Reactivity _ ($) . .-. . . . . - . . - . - . . 161 7.56 Comparison of.-RETRAN'and GE FWFAIL -Base - H Average Surface-Heat Flux:(%) .- 1.-.-._ -
_. . L . .-. 162- l 7.57. Comparison ofLRETRAN and-GE FWFAIL - Base - -
Core Inlet Flow-(%) . . .. . . .1 . .. . . .. . 163 7.58 -Comparison of RETRAN.and-GE FWFAIL - Base -
Pressure RiseD(psi)L....._. .1 . c.- .: . . - _ . .
. 164-7.59- Comparison-'of-RETRAN and GE FWFAIL - Base -
-Relief Valve Flowi(%) . .:-...... .. . . . .
.-165 7.60 Comparison of.RETRAN and GE FWFAIL'- Base -
Vessel' Level'(inches): . .. . . = . . . . . . . 166-4
_7.61.-Comparison ~of_RETRAN and.GE FWFAIL - Lase -- ,
Vessel Steam Flow (%) . ...:. . . . . .. _. . . = . - 16 7 j 7.62 Comparison'of~RETRAN and GE FWFAIL - Base -
Total Reactivity.($) _ . - . .. . . - - . . . . . - . .. 168 19 N'
EA-PT-91-0003-S List of Figures No. Title Page 7.63 Comparison of RETRAN and GE FWFAIL - MOD -
Average Surface Heat Flux (%) . . . . . . . . 169 7.64 Comparison of RETRAN and GE FWFAIL - MOD -
Core Inlet Flow (%) . . . . . . . . . . . . . 170 7.65 Comparison of PLTRAN and GE FWFAIL - MOD -
Pressure Rise (psi) . . . . . . . . . . , . . 171 7.66 Comparison of RETRAN and GE FWFAIL - MOD -
Relief Valve Flow (%) . . . . . . . . . . . . 172 7.67 Comparison of RETRAN and GE FWFAIL - MOD -
Vessel Level (inches) . . . . . . . . . . . . 173 7.68 Comparison of RETRAN and GE FWFAIL - MOD -
Vessel Steam Flow (%) . . . . . . . . . . . . 174 7.69 Comparison of RETRAN and GE FWFAIL - MOD -
Total Reactivity ($) . . . . . . . . . . . . 175 7.70 Typical Power-Dependent MCPR Limit . . . . . 176 7.71 Typical Flow-Dependent MCPR Limit . . . . . . 177 7.72 MSIV Closure Overpressure Event - Bottom of Reactor Vesse) (psig) . . . . . . . . . . . . 178 8.1 River Bend Turbine Runback - Scram 86-12 -
EPIS Core Average Power . . . . . . . . . . 189 8.2 River Bend Turbine R" b.' - Scram 86-12 -
ERIS TCV Position . . . . . . . . . . . 190 8.3 River Bend Turbine a wr - Scram 86-12 -
ERIS BPV Position . . . . . . - . . . . . . . .
191 8.4 River Bend Turbine Runback - Scram 86-12 -
ERIS Dome Pressure . . . . . . . . . .. . . 192 8.5 River Bend Turbine Runback - Scram-86-12 -
ERIS Steam Flow . . . . . . . . . . . . . . . 193 i
8.6 River Bend Turbine Runback - Scram 86-12 -
ERIS NR Water Level . . . . . . . . . . . - - . . 194 8.7' River Bend Turbine Runback - Scram 86-12 -
ERIS.Feedwater Flow . . . . . - . . . . . . . . . 195 8.8 River Bend Turbine Runback - Scram 86-12 -
ERIS Core Flow . - . . . . . . . . . . . . . - . . 196 8.9 RETRAN Analysis of Pressure Regulator Event
- ERIS Average TCV Position (%) . . ., .. . 197 8.10 RETRAN Analysis of Pressure Regulator Event
- ERIS; Neutron Flux (%) . ... '. - .. .. . . . 198 8.11 RETRAN Analysis of. Pressure Reghfator Svent
- ERIS Reactor Pressure (psig). . . .. . . . . 199 i 8.12 RETRAN Analysis of Pressure Regulator Event
- ERIS Narrow Range Level (inches) . . . . . 200 8.13 RETRAN Analysis of Pressure Regulator Esent
- ERIS Total Core Flow (MLB/hr) . . . . . - . . 201
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EA-PT-91-0003-S List of Figures No. Title Page 9.1 Revised Predicted vs. Measured Core Average Fower, Peach Bottom Test TT1 . . . . . . . . 205 9.2 Revised Predicted vs. Measured Core Average Power, Peach Bottom Test TT2 . . . . . . . . 206 9.3 Revised Predicted vs. Measured Core Average Power, Peach Bottom Test TT3 . . . . . . . . 207 11
EA-PT-91-0003-S 1
1.0 INTRODUCTION
This report is a supplement to the Gulf States Utilities (GSU) topical report, which describes and 8
qualifies the methodology developed to perform transient thermal-hydraulic analyses in support of River Bend Station (RBS). Thia supplement describes the methods and procedures used to evaluate changes in thermal margins during rapid core-wide Anticipated Operational Occurrences (AOOs) such as generator lead rejections and increased feedwater flow events. These methods will be used to analyze the consequences of AOOs that may occur during plant operation, support reload design and licensing, and evaluate the safety impact of proposed technical specification changes.
RBS is a boiling water reactor (BWR) plant located in West Feliciana Parish, Louisiana. The plant has a Boiling Water Reactor /6 Nuclear Steam Supply System (NSSS) designed by the General Electric Company (GE), which currently provides nuclear fuel and analytical services. The architect-engineer functions were performed by Stone & Webster Engineering 12
EA-PT-91-0003-S ,
-i
. Corporation. Design features of the plant are shown in l-Table 1.1.
The methodology presented in this supplement is based, in part, on.information considered proprietary to GE which has been licensed by GSUffor RBS use. To diligently safeguard-the licensed. methodology, two'
- versions of this report will-.be issued.- The report-number of the proprietary version.will have a- ,
terminating identifier "SP". A nonproprietary version--
l will also be issued:and have a terminating-identifier
!' S " . Several paragraphs will be deleted in the-nonproprietary version and-the following text.will'be l
inserted:." PROPRIETARY INFORMATION DELETED".
Information-considered proprietary should'not>be released into,the public domain.
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EA-PT-91-0003-S Table 1.1 RDS Design Features Plant Name River Bend Station Plant Type BWR/6 Rated Thermal Power (MW,) 2894 Rated Core Flow (Mlbm/hr) 84.5 Rated Steam Flow (Mlbm/hr) 12.45-Recirculation Flow Control Valve Flow Control Number of Jet Pumps 20- !
Number of Recirculation Pumps 2 Number of Fuel Assemblies 624 j' 14 i
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EA-Pt-91-0003-S 2.0
SUMMARY
This document desc3/ibus * , -
esed to predict the change in C-itica. r '
.c gdCPR) of the most limiting wsi.mboles ;u t to '-
during a rapid, core-wide transient. ....a ly s . " thermal margin changes during slow transients (e.g., loss of feedwater heating), a,d core transients (e.g., control rod withdrawal error) are performed using steady-state methods previously submitted. 2 Methods to perform accident analyses, such as the loss of coolant accident and control rod drop, have different evaluation methodologies and consequence criteria, and are not discussed in this report.
The computer _ programs used to calculate ACPR and the data flow between programs are described in Section 3.0. In Section 4.0, the GSU hot channel model is described. Section 5.0 presents a discussion of the Critical Power Ratio (CPR) concept and the GSU ACPR methodology.
Section 6.0 describes the serisitivity studies 15 l
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EA-PT-91-0003-S performed to determine the uncertainty _ allowance factors for each type of Aoo event. These studies demonstrate that the cycle-specific results provide a level of conservatism equivalent to 95 percent probability, at a 95 percent confidence level or higher.
Section 7.0 presents examples of thermal margin analyses, and comparisons of the results from applica-tions of the GSU methodology to vendor calculations.
Section 8.0 presents the results of additional system transient benchmark calculations, which have been completed since the original methodology document submittal.
Section 9.0 presents updated comparisons and figures for the Peach-Bottom EOC2 turbine trip tect benchmarks. After the submittal of the original topical report', a small change in the GSU core physics L
procedures necessitated the regeneration of the reactor kinetics data used for the Peach Bottom transients.
The value for reactivity uncertainty used in the 16
EA-PT-91-0003-S overall uncertainty calculations (presented in Section 6.0) is based on these revised benchmark calculations.
The system response for the Peach Bottom benchmarks did not change due to the small change in the GSU Core Physics procedures and therefore the system parameters comparison given in the original reports are valid.
The results of the benchmark analyses demonstrate the good agreement of the GSU methodology with Peach Bottom and RBS plant-measured data for system transient parameters such as core power, pressure, and flow. The GSU methodology provides a conservative basis for ACPR analysis of Aoos for reload design and licensing analysis, safety evaluation of technical specification changes, and plant operational support activities. GSU proposes to quote the' thermal limits based on calculations with conservative (design basis) inputs.
These design basis ACPR results bound the 95/95 percent-probability / confidence-level ACPR results obtained from the statistically adjusted best-estimate calculation.
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EA-PT-91-0003-S TMH 7TER PROGRAMS USED IN ACPR METHODOLOGY calculation of ACPR response is perfoi . ~ al calculational steps. Figure 3.1 presents a diagra ,of the calculational flow in the GSU methodology. The computer programs shown in Figure 3.1 are described below.
3.1 SYSTEM TRANSIENT ANALYSIS The first step is to calculate the core response to the AOO event using the'RETRAN02 systems model 3
previously described'. The RETRAN system transient analysis predicts the core average response to the transient event initiator. In addition to the base system model, RETRAN requires a nuclear cross section/ kinetics data file prepared by SIMTRAN-E (see discussion of SIMTRAN-E procedures in Reference 1).
Core flow and pressure drop is calculated by the FIBWR4 code. The output from this step is a RETRAN restart file containing time-dependent core parameters for use as boundary conditions in the hot channel analysis.
The boundary conditions (forcing functions) for the 18
EA-PT-91-0003-S RETRAN hot channel model are extracted from the RETRAN restart file by the EDTRAN computer code. EDTRAN is a small data transfer code developed by the Public Service Electric and Gas Company to read and process RETRAN restart files.
Specific, time-dependent parameters that are passed to the hot channel analysis from the system analysis include neutron flux, core inlet flow, core inlet enthalpy, and core-exit pressure.
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EA-PT-91-0003-S 3.2 EDTRAN LINK TO HOT CHANNEL ANALYSIS The LINK option in EDTRAN calculates hot channel l 1
boundary conditions and prepares a RETRAN hot channel input deck. Hot channel power is defined as a function of core power and radial-peaking factor. Axial power shape in the hot channel is assumed to be a 1.4 middle peaked cosine (see Section 4.0) and is constant for each timestep. Hot channel flow rate is the total core flow multiplied by a hot / average channel. flow factor.
This factor is specified in the LINK input as a response surface fit to parametric FIBWR calculations in which core power, total core flow, presuure, inlet enthalpy and radial peaking factor were varied.
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EA-PT-91-0003-S 3.3 RETRAN HOT CHANNEL ANALYSIS The next step, referred to as " hot channel" analysis, is to evaluate in detail the thermal response of a high power, potentially limiting fuel assembly.
In the GSU methodology, a separate RETRAN02 model is used to determine the thermal response of the hot channels. This model is described in detail in Section 4.0. The RETRANO2 hot channel model calculates time-dependent heat-flux and coolant state parameters in the het channel. These parameters are then passed to the EDTRAN ACPR calculation step.
_3.4 EDTRAN ACPR CALCULATION The General Electric GEXL-PLUS correlations is used to predict the change in critical power ratio (ACPR) for the hot channel, as described in Section 5.0. The EDTRAN code.(TCPR module)-calculates ACPR.
Results extracted from the hot channel analysis restart.
file are used to calculate the transient ACPR and the
! ratio ACPR/ICPR, referred'to as-the RCPR.
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EA-PT-91-0003-S Figure 3.1 Block Diagram for CFR Methodology
[SIMTRAN-E v J SIM U L AT E-E RESTART FILE 1-D CROSS SECTIONS-RETRANO2 PRESSURE D SYSTEM f
FIBWR BOUNDARY CONDITIONS N EDTRAN BUNDLE FLOW LINK.
RESPONSE SURFACE HOT CHANNEL DECK-4 ,-
RETRANO2 QOT CHANNEL BONDLE RESPONSE
< r EDTRAN TCPR h DCPR, CPR VS.- TIME 22'
. . . _ . , . . . _ _ _. - ..m. - ..-s..
1 EA-PT-91-0003-S i
1 I
4.0 HOT CHANNEL MODEL ;
Analysis of the thermal response of the high-powered, most limiting assemblies in-the core during a- ,
system transient event'is referred to as " Hot Channel-Analysis". In the'GSU methodology, a separate RETRAN02 -
model is used totdetermine.the' thermal response.of the hot' channels. The overall system transient response is n also calculated with the RETRAN computer code.- In the:-
system model, the core is represented by a single,- ,
average-power channel representing 1 624 fuel assemblies.
The time-dependent results of the system calculation are.the neutron; flux,icore' pressure,-core flow-Land--
inletLenthalpy. These all'are forcingifunctions-(boundary conditions) f for_ subsequent hot: channel '
calculations. 'The hot 5 channel _'is executedeseparately from the systems-model to allow a differentfmodelito be
- developed 1 for_ each: mechanically uniqueifuel design :in-
+
the RBS core and to-facilitate changes.in initial
-conditions such as radial and'axielipower distribution, gap conductance,-pressure loss coefficients,!!which may differ from.those used for_the system'(core' average) calculation.
23'
b EA-PT-91-0003-S The RETRAN hot channel model geometry represents the in-channel portion of one fuel bundle and the upper plenum. Figure 4.1 is a nodalization diagram of the RBS hot channel model. This model consists of 27 volumes, 27 junctions, 25 conducting heat exchangers and 25 nonconducting heat exchangers. The conducting heat exchangers model the fuel rods, and explicitly include the pellet /elad gap and the cladding. The nonconducting heat exchangers model the direct power depositior. to the coolant.
The forcing functions (boundary conditions) for this model are obtained-from the RETRAN systems transient calculation, and input to the hot channel model is prepared by the EDTRAN code (LINK-module) in the following manner: The hot. channel rod conduction power is proportional to the system transient neutron flux; as is the power directly deposited in the active coolant due to gamma heating and neutron slowing down.
The allocation of this energy is based on a user-supplied power deposition factor. The upper plenum in the-hot channel model is a RETRAN time-dependent volume (TDV), reproducing the pressure versus time response I
24 l l
EA-PT-91-0003-S from the system calculation. The inlet junction to the hot channel is a RETRAN positive fill, so that the time-dependent inlet flow and enthalpy may be specified (flow forced solution). The core inlet'enthalpy is taken from the lower plenum enthalpy in the systems calculation. The hot channel inlet flow is obtained by multiplying the lower plenum flow from the RETRAN systems calculation-by the fraction of the total core flow that enters the active zone of the hot channel; this flow distribution fraction is calculated by a response surface fit to a series of parametric FIBWR core hydraulic calculations.
FIBWR4 is a separate code that calculates the multichannel flow distribution, including nypass i
l (leakage path) flows and water tube flows. The FIBWR l
results are fit to the following core parameters:
thermal power, flow, pressure, inlet enthalpy and radial peaking factor. .The hot channel / core flow distribution was found to be accurately represented by a second order polynomial response surface fit. By using-FIBWR-calculated flow ratios, the potential is minimized for inconsistencies between the systems model 4
25 s . s
_ - - - -r,,- ,
EA-PT-91-0003-S and the hot channel model due to differences in pressure drop models, flow regime options, and core nodalization.
In addition to the above forcing functions, the axial power shape and gap conductance must be specified for the hot channel model. An axially averaged representative value of 2315 BTU /hr-f t 2 *F was used in the GSU hot channel analysis. This value was determined by the RBS ESCORE' model to be typical of high-powered assemblies of the current fuel design, which may be at thermal limits at end-of-cycle conditions. In accordance with the GETAB6 procedure for BWR/6, a standard 1.4 middle-peaked axial power distribution was used for hot channel analysis.
Sensitivity studies (see discussion in Reference 6,.
Appendix V) have shown that ACPR is dependent on axial power shape, with a top peaked power distribution yielding a higher ACPR. However, it is unlikely that an assembly near thermal limits-will be peaked above mid-core.
26
EA-PT-91-0003-S Output from the RETRAN hot channel model is processed by the EDTRAN code (TCPR modt41e) and entered into the ACPR calculation. EDTRAN reads the RETRAN restart file and selects the hot channel parameters needed to evaluate the GEXL-PLUS correlation: nodal thermodynamic equilibrium enthalples, nodal flows, system pressure and saturated water properties at the system pressure. The GEXL-PLUS calculation is described in Section 5.0.
i 27
EA-PT-91-0003-S Figure 4.1 RBS RETRAN Hot Channel Nodalization 210 i L 210 126 25 N 24 125 25 124 24 23 123 23 22 N 122 22 21 N 121 2T 20 N 120 20 19 N 119 19 .
18 N 118 .18 17 N 117 17-16 15 N 116 16 115 15 14 334 .34 3
113 13 N 112 12 11 10 N 111 -11 110 -10 8N 109 9-8N 108 8- l l 27 Volumes 7N 107 7 6N 106 6. l l 25 Conductors 5N 105 5 4N 104 4 N 27 Junctions 103 -3 2
102 2 N 25 Non Conducting I N 101 1 Heat Exchangers 101 5
28
)
EA-PT-91-0003-S 5.0 CRITICAL POWER 2ATIO METHODOLOGY The thermal limit that defines fuel coolability is the critical power ratio (CPR), Critical power is defined as the power at which the transition from nucleate boiling to film boiling will occur if the pressure, flow, inlet enthalpy and axial power shape are hald constant. The transition to film boiling is commonly assumed as the point of cladding failure from overheating; hence the critical power is the maximum power at which an assembly may be operated without overheating the cladding. The CPR is defined-as the ratio of the critical power to the existing bundle power.
EBS plant technical specifications require that the CPR be maintained at or above the operating limit minimum CPR (OLMCPR) during normal operation. The-OLMCPR is analytically determined-to ensure that, during any anticipated operational occurrence (AOO),
the CPR does not violate the Safety Limit CPR (SLCPR).
The SLCPR is defined as the CPR at which, during sustained operation at the SLCPR, at least 99.9 percent 29
. .- - . . ..-. -._- .- . - ... .. . - . . . . . . . -..- . . . . . . - - - . -.. - . - . ~. - - . ...-
-1
' EA-PT-91-0003-S of the fuel rods in the core can be expected not to undergo a boiling transition. The SLCPR'is determined through a statistical analysis that accounts for plant measurement uncertainties'and the uncertainty in--the correlation itself. In the GSU methodology and in plant operation, the critical power is determined by s
the GEXL-Plus correlation. The measurement 5
uncertainties and the GEXL-Plus correlation -
uncertainties are independent of safety analysis methodology, so the current SLCPR valuefof'1.07>is appropriate for use with the GSU methodology.
PROPRIETARY INFORMATION DELETED 4
30
.-4., ,. .,.o + .- , ,,,n.,,-%., , 4 + y , .w.-- -
-w-- . . - ,. .- ,
l EA-PT-91-0003-S The CPR is evaluated as illustrated in Figure 5.1. Shown in the figure is the bundle average thermodynamic equilibrium steam quality distribution over the length of the bundle for the critical and the operating bundle powers and the corresponding GEXL-Plus correlation line. As seen in the figure, the critical power is that bundle power at which the quality curve 6
becomes tangent to the correlation line.
PROPRIETARY INFORMATION DELETED l
.31
=. - -- . . .-
EA-PT-91-0003-S i
Using the GSU methodology, the change in CPR (ACPR) during an AOO is calculated for the most limiting fuel assembly in the core, referred to as the hot channel. In the analytical procedure, the hot
-channel is assumed to be operating at the technical specification OLMCPR at the start of the transient.
The initial CPR (ICPR) and the ACPR for each transient are then determined by iterative calculations performed using the RETRAN hot channel and EDTRAN computer codes (as ::hown in Figure 3.1). If the minimum CPR (including an allowance factor for uncertainty) during the AOO event violates the safety limit CPR, the initial assembly power for the hot channel must be reduced. By an iterative process, the initial hot channel power can be found which results in a minimum.CPR that does not violate the SLCPR:
OLMCPR = SLCPR + ACPR + Uncertainty Allowance (5.1)
It is convenient to express the. severity of an AOO event.in relative units'ACPR/ICPR. Dividing 32
.n..+ - - - - + -
1 EA-PT-91-0003-S It is convenient to express the severity of an AOO event in relative units ACPR/ICPR. Dividing equation (5.1) by the initial CPR (ICPR), and noting that the ICPR determined for the most severe AOO event will equal the desired OLMCPR:
1 = SLCPR/OLMCPR + ACPR/ICPR + UAF/ICPR (5.2)
Where, UAF is the uncertainty allowance factor.
By defining the statistical adjustment factor (SAF) as the ratio UAF/ICPR, and the ratio ACPR/ICPR as RCPR, Equation (5.1) becomes:
OLMCPR = SLCPR/( 1- (RCPR + SAF)]- (5.3) l The calculations to determine appropriate values for the SAFs are-presented-in Section 6.0. Us_ of these SAFa to set-the OLMCPR (equation 5.3) ensures a 95 percent probability at a 95 percent confidence level of not violatingithe SLCPR during an AOO event.
33 I
EA-PT-91-0003-S ;
Figure 5.1 Graphical Illustration of CPR Tangent Pont v
Cnbcal Quauty Vews L _~
\
u hne Average -
Steam Quahty (X)
X Verses L Corresponding _
l To Openung Bundle Power l~
(:
g .
I Verses L Comspondmg -
,. To Cntical Bundle Poew -
l Boiling langth (L) ;
4-
- d. _
.'34 :
4 4
4 .i s
)
P EA-Pr ol-0003-0 ,
6.0 UNCERTAINTY ANALYSIS This section describes-the calculation of uncertainty allowance factors for the G5U analysis methodology. -These allowance factors are included in .
the Operating Limit Minimum Critical Power Ratio (OLMCPR) to cover computer code and analytical ~model uncertainties._ A number of parametric sensitivity +
studies were performed to quantifyLthe: component of-'
uncertainty.in'the best-estimate system transient and hot channel models-in relativeJACPR/ Initial CPR (RCPR) units. Each componentiwas perturbed bycan amount judged to be the^ uncertainty in'the parametet at 95' percent probability and'95 percent' confidence level'.
The changes.in RCPR.due'to these. perturbations were combined statistically'toLobtainLan:overallL95/95"(95 percent probability with a 95Lpercent confidence. level)
RCPR.-
4 l
For a technical specification or reload license-support ~ application'~ the 0LMCPR will actuallyfbe determined using a conservative RETRAN model;with; design basis-inputs.for suchfparameters asLscram speed, t.
35
EA-PT-91-0003-S q 1
l I
time delays and safety / relief valve (S/RV) setpoints.
The end results of the scnsitivity calculations are transient-specific statistical adjustment factors (SAFs), which compensate for uncertainties (e.g.,
RETRAN code and model uncertainties, choice of input <
parameters, kinetics data preparation uncertaintles) in the design basis calculation.
The design bauls model includes the following conse vative feateares not found in the best-estimate ,
model
- Reactor-protection system trip and S/RV setpoint allowance for instrument accuracy, calibration error and drift
- S/RV flow area reduced to 90 percent of nominal area, technical specification stroke times for the S/RVs
!
- Technical specification stroke and delay times for the TCVs (turbine control valves) and TSVs (turbine stop valves); valvo position versus .me -
curve assumed-linear
- Technical specification control rod insertion speeds and maximum RPS (scram) delay times;.
different insertion curves:for pressurization and overpressurization events.
\
f 36 g w *e,7y e e vway w y %w,e- y y yym_c w-y'-- y w
EA-PT-91-0003-S River Bond Station, which is a BWR/6-class reactor, has a fast t' ram design feature. Because of the fast scram, a much lower sensitivity to scram speed is noted for 1::)S than for older BWRs. Therefore, the technical specification scram speed has been usod 3
instead of the fastor, as-measured control rod insortion rato for tha design basis calculations. This approach is conservativo, however, it greatly simplifies application of the 95/95 RCPR and the SAFs.
The SAFs were determined from the design basis RETRAN calculation and 95/95 RCPR as follows:
SAF = RCPR um - RCPR&,,, w,;,og (6.1) for each type of transient analyzed. The reported OLMCPR and ACPR were dotormined from the conservative RETRAN calculation and the SAFs as follows:
OLMCPR = SLCPR/( 1 - (RCPR &,,. u,;,og + SAF)) (6.2) o 37 I
EA-PT-91-0003-S This procedure ensures that the Safety Limit CPR (SLCPR) w!11 not be violated during an operational transient event.
Figures 6.1 through 6.3 present an overview of the procedure utilir,ed to determine the uncertainty allowance for each transient category.
6.1 SENSITIVITY STUDIES This section presents the results of sensitivity studies of the RBS transient analysis methodology.
These studies quantify the components of uncertainty for the load rejcetion without bypass (LRNB), pressure regulator f ailure downscale (PRDF), and feodwater controller failure (FWCF) to maximum demand. The uncertainty components include the nuclear model, the heat transfer model, the core hydraulics model, the separator,' dryer hydraulics model, the recirculation system hydraulics model, ari the steam line hydraulics-model.
I 38
1 EA-PT-91-0003-S The sequence of events for these three transients are shown in Tables 6.1 through 6.3. The load .
rejection without bypara transient is started by a turbine control valve fast closure trip with the bypass valves disabled. The pressure regulator failure (downscale) transient is initiated by a rapid reduction in steam demand from 100 percent to o percent. The feedwater controller failure transient is initiated by an instantaneous increase in feedwater demand with the foodwater flow rapidly increasing to the maximum foodwater pump capacity.
Critical Power Ratio (CPR) sensitivity is quantified using the variable RCPR, which is defined as the ratio of transient ACPR to initial CPR. Other sensitivities were evaluated, including changes in core power, surface heat flux, and dome pressure.
The individual parameter variations used and the resulting sensitivities are summarized in the tabular results (Tables 6.4 through 6.6). In general, the parameter of interest was perturbed two standard deviations from its nominal value by an amount judged 39
)
i EA-PT-91-0003-S ,
to be at a 95 percent confidence level.
6.1.1 Time Sten and Nodalization Sensitivity During the development of the GSU RETRAN model, a number of nodalization and time stop sensitivity studies were performed. Nodalization studies woro performed for the core region, the ;
separator / dryer /downcomer region, and the steam lines / bypass. These studies guided the choice of nodalization selected, prompting changes in RBS model core nodalization from 12 to 25 nodes and a general reconfiguration of the separator / dryer /downcomer nodalization scheme. The final nodalization scheme was determined-by the benchmark studies to be sufficiently discrete to produce results which are consistent and phenomenologically correct.
Time step studies'were performed to identify.the maximum time step.for that portion of the transient where the neutron flux and power are rapidly changing.
In general, a time step limit of 0.001 seconds was found to replicate the'results from calculations with 40 i
. ~ . . . . .. - , . -
EA-PT-91-0003-S finer time steps. During load rejection transients, however, the width of the flux spike, and therefore the resulting ACPR, has a small sensitivity to time step.
For the load rejection transient an additional uncertainty factor was therefore included for time step size.
6.1.2 Nuclear Model Sensitivities Nuclear model sensitivities relate to the generation of nuclear heat inside the fuel rod and the fraction of heat deposited directly to the active coolant and the bypass in the RETRAN core model.
The reactivity uncertainty associated with the use of CASM0/ SIMULATE /SIMTRAN-generated RETRAN kinetics data in the GSU methodology was determined from the benchmark analyses based on the Peach Bottom turbine trip tests..
A 21 percent uncurtainty allowance-was chosen by examining the difference in peak excess reactivity inferred from inverse point kinetics in the measured 41
g . . -_ .
l EA-PT-91-0003-S ;
I l
data and GSU RETRAN calculations. As can be seen the table below, the results indicate a 95 percent confidence upper bound (from X2 test) on the standard
! deviation between measured and calculated peak reactivity.
Peak Excess Reactivity (St ,
Inst Data Calculation % Difference TT1 0.804 0.843 4.85 TT2 0.784 0.799 1.91 TT3 0.825 0.818 -0.85 v Average 1.97 I l Standard Deviation 2.85 95% Confidence Standard Deviation 12.57 The RBS model uses two-group, cross-section constants for one-dimensional kinetics parameters. The thermal group absorption cross section (Ea2) was adjusted to produce a 21 percent change in peak reactivity to account for variations in the predicted -
void and doppler coefficients and-scram Worth. Other.
variables chosen for this sensitivity analysis included the decay heat fraction,-the direct moderator heating fraction, and the bypass heating fraction.
42
EA-PT-91-0003-S 6.1.3 Ecat Transfer Model Sensitivities Heat transfer model uncertainties impact the time required for transfer of heat out of the fuel rod and into the coolant. Variables selected for sensitivity analysis were the core average (system model) and hot channel gap conductance. The CPR was found to be-insensitive to an increase in the number of radial nodes in the fuel rod, or to a perturbation in the fuel rod thermal. Studies performed to develop the methodology for averaging the gap conductance showed no significant sensitivity to use of an axially varying gap conductance as compared to an axially uniform gap conductance.
6.1.4 Core Hydraulic Model Sensitivities Core hydraulic model sensitivities are associated with the distribution of coolant and pressure within the core. Variables selected for the final sensitivity analysis were core bypass flow fraction, orifice and grid loss coefficients, and total core pressure loss.
43 I
r
..-----. . - - . _ ..-. ._ . - . - - - - _ - - _ = _ - _ . - - . . - . . _ - ..-
1 EA-PT-91-0003-S l
Additionally, correlation parameters in the subcooled boiling model and algebraic slip (drift flux) model were investigated.
l 6.1.5 Egnarator/Drver Hydraulic Model Sensitivities separator / dryer hydraulic-model sensitivities are associated with the coupling of the core model and the steam dome. Variables selected for sensitivity analyses were the carryunder fraction, the separator inertia, the total separator pressure loss, and the I
fraction of the separator pressure loss assigned to the inlet. No sensitivity was noted for changes in the non-equilibrium interphase heat transfer coefficient for the separated upper downcomer volume, the steam flow between the upper downcomer and the steam dryer, or the initial water level.-
6.1.6-Recirculation System Hydraulic Model sensitivities l The recirculation system hydraulic model sensitivities related to the response of the i
. recirculation system to core and pump parameters.
, 44
EA-PT-91-0003-S Variables selected for sensitivity analyses were pressure distribution, jet pump M ratio, jet pump N ratio, time delay for recirculation pump trip to the Low Frequency Motor Generator (LFMG), and recirculation pump inertia.
6.1.7 Steam Line Hydraulic Model Sensitivities steam line hydraulic model sensitivities are associated with the dynamic response of the steam lines. Variables selected for sensitivity analyses were steam line pressure losses, steam line inertia, I and turbine control valve stroke time. No sensitivity was noted for changes in safety / relief valve capacity and safety / relief valve setpoints. For the FWCF event, the bypass valve stroke time and bypass flow capacity were also varied.
45
)
kf h -h
l i
EA-PT-91-0003-S Freliminary nodalization studies showed that the current steam line with 11 nodes is sufficiently discrete to produce a converged delta CPR. This c
nodalization was also used for the-Peach Bottom turbine trip tests. No additional uncertainty component was included for steam line nodalization.
6.1.8 Feedwater Flow Sensitivitiqa Feedwater flow response is relatively slow, so that uncertainties in the feedwater performance do not affect the delta CPR in the pressure regulator failure and load rejection transient analyses. For the FWCF event, the variables selected for sensitivity analyses included maximum feedwater flow and the associated time constant for feedwater flow increase.
46 ms um i
.- _.-_- . . - . . - _ =- _ _ . . .. . . . . . . . . .
EA-PT-91-0003-S 6.2 RESULTS OF SENSITIVITY ANALYSES
,ensitivity studies were performed for the three most limiting RBS USAR7 transients, which will be analyzed with the GSU RETRAN-system and hot channel '
models. These transients are the load rejection without bypass (LRNB), feedwater controller failure (FWCF) and pressure regulator downscale' failure (PRDF). )
1 I
RETRAN/EDTRAN delta CPR analyses were performed to determine the effects of the variations described )
above. The results of the sensitivity studies are shown in Tables 6.4 through 6.6, which summartm- the parametric variations and main output sensitivities of each of the evaluated transients.
The RCPR calculated for the load rejection without bypass transient was most sensitive to scram and recirculation pump trip timing variables. The RCPR calculated for tho' pressure regulator fallare transient was most sensitive to steam line pressure losses and l core bypass flow fraction. The RCPR calculated for the feedvater' flow controller-failure was most sensitive to 47
EA-PT-91-0003-S initial water level and total pressure losses in the i
dryor-separators.
The results of the sensitivity studies were used <
\
in the determination of SAFs for future analysis.
The SAFs link the deterministic results of conservativa design-basis calculations to best-ostimate results that have been statistically adjusted to bound 95 percent of expected outcomes with a 95 percent confidence factor.
Design-basis calculations are made using the benchmarked RETRAN model with the input modifications described in Table 6.7. While this calculation is j still basically a best estimate calculation, the use of conservative scram and setpoint inputs makes it appropriate for use in design basis analysis.
Statistical adjustment factors are derived for each transient _from the results of the sensitivity .
studies. Best-estimate results comprise as the base case'in this study. Becauae the BWR/6. control rod i
48
EA-PT-91-0003-S scram times are much faster than those of older BWRs, scram speed are treated in a bounding fashion (at Technical Specification values). The remaining input parameters are assumed to be normally distributed, so that the RCPR distribution is also normal and the standard deviation can be calculated as the square root of the sum of the squares of the individual differences.
In a normal distribution, a one-sided tolerance limit bounding 95 percent of the data points 10 established at the mean plus 1.645 standard deviations.
With 95 percent confidence statistics, an RCPR tolerance limit established at this point encompasses 95 percent of the random statistical combinations of uncertainties with 95 percent confidence. Thu differ-ence betwoon this 95/95 RCPR value and the value of RCPR calculated with design-basis inputs is the SAF for the transient. Table 6.8 summarizes the SAF
. calculation. As can be seen in the table, the design basis calculations exceed the 95/95'RCPR values. llence the SAFs calculated using equation (6.1) would be negative. To be conservative, GSU proposes to use a j
49 l
EA-PT-91-0003-S zero value for the SAFs, (e.g., to quoto ACPR results from the design-basis calculation without any corrections or reduction factors).
1 SO
- u. -. .-
EA-PT-91-0003-S Table 6.1 sequence of Events for RDS Load Rejection without Bypass Transient Time Event 0.00 100% power & 200% flow steady state 0.20 Load rejection initiated turbine control valve fast closure (turbine trip) ;
0.17 Roactor scram by turbine trip )
0.24 Recirc pump trip to LFMG 0.82 Peak power 0.94 Peak RCPR I
1.18 SRV group 1 (one valve) open 1.18 SRV group 2 (one valve) open .
1.36 SRV group 3 (three valves) open 1.36 SRV group 4 (four valves) open 1.55 SRV group 5 (seven valves) open 1,55 Recirc pump ATWS trip by high pressure 1.59 Peak pressure i 1.80 End of simulation 5
51
EA-PT-91-0003-S Table 6.2 sequence of Events for RBS Pressure Regulator Failure Transient Time Event .
1 0.00 100% power & 100% flow steady state 0.01 Pressure regulator fail downscale 1.84 Reactor scram by APRM high trip 1.94 Peak power 2.05 Peak RCPR 3.31 SRV group 3 (one valve) open l 3.31 SRV group 2 (one valve) open 3.45 SRV group 3 (threu valves) open 3.45 SRV group 4 (four valves) open 3.52 Peak pressure 3.60 SRV group 5 (seven valves) open !
3.72 Recirc pump ATWS trip by high pressure 5.00 End of simulation 4
t f
52
= _ _ . . _ . .,. . . -
EA-PT-91-0003-S J
l Table 6.3 Sequence of Events for RDS Feedwater controller Failure Transient Ilan Event.
0.00 100% power & 100% flow steady state Feed water controller fail 14.50 Peak power ;
14.64 TSV trip by high water level i 14.65 FWP trip by high water level i 14.69 TCV trip by TSV trip Reactor scram by high vator level 14.74 DPV start opening 14.78 Peak RCPR 1 Recire pump trip to LFMG by TSV trip 16.10 SRV group 1 (one valve) open 16.10 SRV group 2 (one valve) open 16.23 SRV-group 3 (three valves) open SRV group 4 (four valves) open 16.26 Peak pressure 16.32 SRV group 5 (seven valves) open 16.38 Recirc pump ATWS trip by high pressure 18.00 End of simulation 53
S
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EA-PT-91-0003-8 Table 6.4 sensitivity studies for.RBS Load Rejection without Bypass (LRNB)
I Transient (Continued) (Proprietary Information Deleted)
Description Change From To Pbs er(%) Q"(%) Pres.(psi)
CORE HYDRAULICS Leakage Flow -10% 12.4 % 11.1 % 163.30 . 101.65 1177.6S Pressure Loss Dist. (Orifice / Bundle) 10 % 29% FJ1% 26K/74% 160.61- 101.25 I177.57 Core Total Pressure losses + 10% 25. I', psi 27.68 p<i 163.13 101.55 I177.63 SEPARATOR / DRYER HYDRAULICS ,
Separator Pressure Drop + 10 % 5.4 psi 5.9 psi 156.21 100.87 I177.21 i
Pressure Loss Dist.(Ent/ Exit) 10 % 85 % /15 % 75 % /25 % 159.49 101.18 1177.39 Carryunder +25% ' O.0019 0.0025 158.56 100.98 I177.69 Separator Inertia -20% 0.157 fri 0.126 fr' 158.01 100.80 7177.33 I
RECIRC SYSTEM HYDRAULICS Jet Pump M Ratio . +5% 2.319 2.435 170.16 102.13 1177.82 Jet Pump N Ratio 4 15 % 0.149 0.171 - 158.53 101.07 1I77.50 RC Pump Trip Delay , +20 % 0.14 sec 0.168 see 165.59 101.74 1177.72 RC Pump Inc.tia - + 10 % 31900 35090!b-ft2 163.65 101.54 1177.56 L
l Flow Inertia +20% 26.3 ft* 28.0 ftd 160.16 101.22 1177.54 55
EA-PT-91-0003-S Table 6.4 Sensitivity Studies for RBS Load Rejection witI2 cut Bypass (LRNB)
Transient (Continued) (Proprietary Information Deleted)
Description Clunge From To Power (%) Q"(%) Ihs.fp4)
J -_i STEAM LINE IIYDRAULICS Steam Line Pressure Drop + 10 % 40 psi 44 pai l 158.02 101.05 1177.78 Steam Line Inertia -7 % 27 fr' 25.07 ft-' 160.68 101.53 1179.50 TCV Stroke Time -20 % 0.19 sec 0.152 see 171.65 102.22 I178.04 2
( A RCPR = . 0219 Note 1: Ia 2adjusted to increase peak excess reactivity 25%
(from S.410 to S.513) for the load rejection event.
56
EA-PT-91-0003-8 Table 6.5 sensitivity studies for RBS Pressure Regulator Fail Downscale Transient (PRDF) (Proprietary Information Deleted)
Dewription Change From - To Pbwer(%) Q"(%) Pres.(pQ BASE CASE N/A N/A N/A 128.35 102.62 1152.76 TIME STEPS Maximum Time Step -50 % .001 .0005 128.20 102.54 1152.76 NUCLEAR MODEL Peak Excess Reactivity . 25 % Note i Note 1 128.13 102.62 1152.46 Initial Thermal Power . 2% 100 % 102 % 130.82 106.69 1155.33 Promp* Power to Active Coolant -25 % 2% 1.5 % 128.06 103.11 1152.93 HEAT TRANSFER -
Gap Conductance (Core Avg) -30% 1207 845 127.84 102.1I i153.73 Gap Conductance (Hot Channel) +30% 2288 2975' 128.35 102.62 I152.76
- VOID MODEL
- C, (Algebraic Slip C, Coef.) + 10 % 1.41 1.55 127.45 102.54 1152.70
~Subcooled Voids -63 % 1.1 % 0.4 % 127.88 102.62 I152.90 4
.I i 57
EA-PT-91-0003-S Table 6.5 Sensitivity Studies-for RBS Pressure Regulator Fail Downscale Transient (PRDF) (Continued) (Proprietary Information Deleted)
DescripGon Change Frum To l Power (%)Q*(%) Pres.(pd)
CORE IIYDRALLICS Inkage Flow -10% 12.4 % 11.1 % 129.58 102.72 1152.71 Pressure loss Dist. (Orifice / Bundle) 10 % 29%f71 % 26 %/74 % 127.70 102.55 1152.68
, Core Total Pressure losses + 10 % 25.16 psi 27.68 psi 129.07 102.73 1152.73 SEPARATOR / DRYER IIYDRAL11CS Separa'or Pressure Drop + 10 % 5.4 psi 5.? psi 128.01 102.53 I153.20 Pressure less Dist. (EntfExit) 10 % 85%Il5% 75 % 125 % 127.65 102.55 I152.59 Carryunder - +25 % 0.0019 0.0025 128.33 102.60 1153.15 Sepantor Inertaa -20 % ' .0157 fr' O.126 fr' 128.29 102.56 1152.84 RECIRC SYSTEM JIYDRAULICS Jet Pump M Ratio +5% 2.319 2.434 128.25 102.58 1152.53 Jet Pump N Ratio + 15 % 0.149 0.171 f 128.31 102.57 1152.82 RC Pump Trip Delay +20% 0.14 sec 0.168see 128.35 102.62 1152.76 RC Pump Inertia + 10% 31900 35090th.ft2 128.35 102.62 1152.76 Flow Inertia ' +20 % 26.3 fr' 23.0 ft4 12Ts.30 102.57 r I152.82 58
.I 4-4'.
- EA-PT-91-0003-S 3'.
]' Table 6.5 Sensitivity Studies fer RBS Pressure Regulator Fail Downscale i Transient (PRDF) (Continued) (Proprietary Information Deleted) n Description . f Change From To g Pbwer(%) Q*(%) Pres.(psi)
STEAM LINE IWDRAULICS Steam Line Pressure Drop + 10 % 40 psi 44 psi 127.60 102.54 I152.75
't Steam Line Inertia -7 % 27 ft-' 25.07 fr' 127.70 102.55 1152.57 TCV Stroke Time -20 % 0.19 sec 0.152 see 128.35 102.62 1152.76 2
( ARCPR =.0039 Note 1: Za2 adjusted to increase peak excess reactivity 25%
(from S.410 to $.513) for the load rejection event.
1 1
4 l
s s
59
Fi .
EA-PT-91-0003-8 Table 6.6 Sensitivity Studies for RBS Feedwater Controller Failure (FWCF)
Transient.(Proprietary Information Deleted)
Dewription Change From To Pbwer(%) Q"(%) 1%s.(psi)
! EASE CAFE . N/A N/A N/A IM.12 102.83 1162.48 TIME STEIS Maximum Time Step -50 % .001 .0005 104.13 102.83 1162.95 NUCLEAR MODEL Peak Excess Reactivity 21 % - Note I Note i 104.11 102.83 1162.93 Initid Thermal Power 2% 100 % 102 % 106.32 107.21. I165.35 Pronpt Power to Active Coolant -25 % 2% 1.5 % I M.14 103.37 1162.77 IIEAT TRANSFER
~
Gap Conductarxe (Core Avg) -30% 1207 845 102.35 102.82 1164.40 Gap Conductance (Hot Channel) +30 % 2288 2975 105.12 102.83 1162.48 VOID MODEL C, (Algebraic Slip C Coeff + 10 % - 1.41 1.55 104.14 102.83 1162.51 Subcooled vsds 43% 1.1 % 0.4 % 104.15 102.84 1162.57 60
EA-PT-91-0003-8 Table.6.6 Sensitivity Studie 7 for RBS Feedwater Controller Failure (FWCF)
Transient (Contir"e4) (Proprietary Information Deleted)
Description Cha m Fruen To Iwer(%) Q*(%) Pres.(psi)
CORE HYDRAULICS Leakage Flow -10% 12.4 % 11.1 % 104.24 102.88 1162.60 Pressure less Dist. (Onfice/ Bum!!e) -10% 29 %/71 % 26%B4% 104.14 102.82 1162.52 Core Total Pressure losses + 10 % 25.16 psi 27.68 psi 104.17 102.85 1162.75 SEPARATOR / DRYER HYDRAULICS Separator Pressure Drop + 10 % 5.4 psi 5.9 psi 104.I3 102.80 I162.89 Pressure less Dist. (Ent/ Exit) 10 % 85 % /15 % 75 %/25 % 104.13 102.73 1162.44 Carryunder +25 % 0.0019 0.0025 104.20 102.86 1162.81 Separator Irm% -20 % ' .0157 ft4 0.126 ft ' 104.12 .102.83 I162.48 RECIRC SYSTEM HYDRAULICS Jet Pump M ratio +5%- 2.319 2.435 104.12 102.83 1162.41 Jet Pump N ratio + 15 % 0.1549 0.171 104.13 102.83 1162.48 RC Pump Trip Delay '
+ 20% 0.14 sec 0.168 see 104.12 102.83 1162.55 RC Pump L.citia + 10 % 11900 35090th.ft2 104.12 102.83 1162.35 Flow Inertia +20 % 26.3 ft~' 28.0 ft4 104.12 102.83 1162.59 1
61
EA-PT-91-0003-8 Table 6.6 Sensitivity Studies for RBS Feedwater Controller Failure (FWCF)
Transient'(Continued) (Proprietary Information Deleted)
Deciption From To Pres.(psi)
Changre - l Power (%)Q"(%)
' l STEAM LINE HYDRAULICS .
i Steam Line Pressure Drop + 10% 40 psi 44 psi 103.99 102.81 I161.97 Steam Line Inertia -7 % 27 ft4 25.07 ft4 104.12 102.83 1160.77 TSV Delay Time +6%- 0.094 sec 0.1 see 104.12 102.83 1165.23 SRV Valve Capacity (area) -10% 0.11441 0.10297 104.12 192.83 I162.74 SRV Pressure Setpoint + 10 % Note 2 Note 2 104.12 102.83 1239.67 FEEDWATER FLOW Maximum Runout Flow +5% 122 % 128 % 105.58 102.59 1161.77 Ramp Rate -50 % ; 0.333 sec 0.166 see 104.13 102.83 I.*'
g/{ nRCPR2 =.0033 Note 1: Ea2 adjusted to increase peak excess reactivity 25%
(from $.410 to $.513) for the load rejection event.
Note 2: SRV setpoints for five groups of safety relief valves were. increased by 10 percent.
62
EA-PT-91-0003-S i
l Table 6.7 Adjuntments to the RDS RETRAW Best-Estimate Model for Design-Dasis calculations
- Reactor Protection System (RPS) setpoints Pro changed to analytical limits.
- Tech Spec values are used for scram insertion speed. Even more conservative values are used for pressurization and overpressurization analyses to account for the effect of high reactor pressure.
!
- APRM thermal circuit lag is raised to 6.6 seconds (Tech Spec maximum value).
- Opening setpoints for SRVs are increased to analytical limits. Closing setpoints are a?sumed to be 98 percent of the opening netpoint.
- TCV stroke time is reduced to-the Tech Spec '
value of 0.15 seconds. ,
- TSV stroke time is reduced to the Tech Spec
, value of 0.10 seconds.
l 63 l
l
i.-
l l l '-
I
-i i
- =
f i
s- . ..
' 4 EA-PT-91-0003-S "
-1 i
i
'?
i Table 6.8 Statistical Adjustment. Factors for the RBS RETRAN Model .
i
- i t ..
- PROPRIETARY.INFORF . TION DELETED . .
L i
.t i
-f t
.. }
t L
,)
3 T - ,I a
t b
h
'. }
64 -i
)
- t-.
f e- s- , , , , < .~...n , n. , . - . ~ , , ,,
EA-PT-91-0003-S Figure 6.1 Schematic Diagr e of Transient-Specific Uncertainty Methodology SCHEMATIC DIAGRAM OF UNCERTAINTY METHODOLOGY ,
REACTIVITY UNCERTAINTY FROM PEACH UOTTOM TURSINE TRIP TESTS PEACH BOTTOM PEACH BOTTOM MEASURE 0 6p CALC LATED 6 p OmER MODEL UNCERTAINTIES I
I (ONE SHOWN .
IT TYPICAL. OF OTHERS) a "'
I DISTRIBUTION 2a VALUE -
} If UNCERTAINTY J
2a VALUE ESTIMATED do GSU RETMN MOD 81
- __
if If RETRAN RETRAM RETRAN WITH WrrH BEST 2a PERTURSED 2a 6p PERTUmeAT10N 1f If If -If if I
2a VALUE 2a VALUE 6 RCPR 6 RCPR I -I.
t t STATISTICAL CONVOLimON (NORMAL DISTRSUTION)
NOTE: THIS PROCEDURE if' IS RUN FOR EACH CATEGORY OF TRANSIENT 95/97 TO 05TAIN AN ADO UTCERTAINTY
^
ALLOW NC 65 I
l 1
1
% . . . . l
EA-PT-91-0003-S Figure 6.2 Determination of Trane. lent-Specific SAF DETERMINATION OF TRANSIENT SPECIFIC SAF 95/95 RCPR h n SAF
^
E d Design Basis RCPR E
O d Uncertainty Allowance 11 E
O Best Estimate RCPR 0
66
~
l EA-PT-91-0003-S Figure 6.3 Determination of OLMCPR DETERMINATION OF OLMCPR OLMCPR h
Design Basis ACPR For Worst AOO s o
i=
1 k o Uncertainty Allowance (= SAF'OLMCPR) 1.07 SLMCPR
' (0.1% hpoetation of BT) 1,00 -================ (50% hoectation of BT) 67 I
I l
l EA-PT-91-0003-S 7.0 APPLICATION OF TFE RDS METHODOLOGY TO CALCULATION OF THERMAL MARGINS This section describes the application of the River Bend Station (RBS) methodology to the calculation of thermal margins for reload safety analysis, Technical Specification amendments and other operating license support activities.
With each change in core configuration (a new fuel type or rearrangement of existing fuel types), the core and plant operating limits are reevaluated..
Design basis events are reanalyzed for the planned core configuration to determine the need to change operating limit values. For the original safety analysis of RBS, many different Anticipated Operational Occurrences (AOOs) were analyzed to determine the OLMCPR. For subsequent reloads, the fuel vendor determined that analysis of five limiting events was sufficient to establish the maximum change in CPR so that an adequate thermal margin can be maintained.
68 1
EA-PT-91-0003-S 1
The events analyzed for subsequent reloads include the generator load rejection without bypass (LRNB), pressure regulator failure downscale (PRDF),
loss of feedwater heating (LFWH), feodwater controller failure to maximum demand (FWCF), and control rod withdrawal error (CRWE) .
The LFWH and CRWE eo ..ts are evaluated with steady state reactor physics methods discussed in 2
another report. Methods discussed in this report will be used to determine MCPR limits for LRNB, PRDF and FWCF transients. This chapter-compares GSU's LRNB, PRDF, and FWCF analyses and vendor results'.
7.1 FUEL VENDOR ANALYTICAL RESULTS Reload 2, Cycle 3 was chosen as the basis for.
this sample application of the methodology. Transient response and ACPR results were typical of those
, calculated by the vendor for the most recent cycles.
69 1
l 4
i EA-PT-91-0003-S The fuel vendor's GEMINI methodology "" was 8
used to perform the Reload 2 analyses, which assume end of cycle 3 exposure conditions. Initial conditions for the plant transient analyses are listed in Tables 7.1 and 7.2.
System transients that establish the OLMCPR were analyzed at rated power conditions. The vendors' analyses accounted for uncertainties by adding statistical adjustment factors to the calculated ACPR i2 The fuel vendor calculated ACPR values were 0.04 for PRDF, 0.07 for LRNB, and 0.06 for FWC". The
.MCPR-Safety Limit used in these analyses (and all analyses in this report)-was 1.07.
7.2 CALCULATION OF THERMAL MARGIN REQUIREMENTS This section presents the results of using the 1
RBS methodology to analyze LRNB, PRDF, and FWCF l transients, which are potentially limiting for RBS. As 70 J
i EA-PT-91-0003-S for the vendor results described above, RBS Reload 2, Cycle 3, was chosen as the reference cycle for this calculation.
7.2.1 Lppd Reiection without Bypass A rapid pressurization transient is evaluated for each reload application because-events of this type are potentially limiting. The load rejection without bypass (LRNB) was the pressurization event evaluated for RBS.
Fast closure of the turbine control valves (TCVs) is initiated whenever electrical grid disturbances result in significant loss of electrical-load on the '
generator.- The TCVs close as rapidly as possible to-prevent excessive overspeed of the turbine-generator 1 rotor. When the TCVs close, the steam flow is abruptly reduced, causing an increase in system pressure. The reactor scrams on a signal from the TCV controller in 71
__m____ _._m.______ _ _ _ . - . - - _ -
EA-PT-91-0003-S f
anticipation of this pressure surge, which results in a rapid power increase in the core.
For purposes of the analysis, the TCVs operated in the full arc mode and had a full stroke closure time of 150 msec. In the full arc mode, the TCVs normally operate in a partially closed position,- so an abbreviated closure time of 72.6 msec was used in the LRNB analysis, based on a 48.4 percent open initial TCV position. This faster closure time resulted in a more abrupt steam flow stoppage and a more severe transient.
The turbine trip transient is very similar to the load rejection transient. In this event, the turbine stop valves (TSVs) were closed by the turbine controller. The TSVs were. fully open during operation and closed in 100 msec. Because this rapid closure causes a large pressure transient, the Reactor Protection System (RPS) commands a reactor scram and recirculation pump trip (transfer to the LFMG) when TSV 72 4
l
EA-PT-91-OOO3-S motion is sensed. Although the TSVs close more rapidly than the TCVs, the assumed initial valve position gives the load rejection event a more abrupt pressure transient under analytical conditions.
The RBS RETRAN model was used to evaluate the load rejection transient. System response for a typical load rejection transient based on RBS Cycle 3 operating conditions and core configuration is shown in Figures 7.1 through 7.10. A sequence of events for the LRNB transient is shown in Table 7.3.
The transient neutron flux is shown in Figure 7.1. The effects of recirculation pump trip and scram decreased the neutron flux initially, allowing a lower reactor power when the pressure wave reached the core at about 0.3 seconds. Rapid pressurization induces strongly positive void reactivity due to collapsing voids in the core, which results in a rapid-flux increase to a maximum of approximately 230 percent.
73 l
1
EA-PT-91-0003-S Negative reactivity from inserting control rods and doppler overcame the positive void reactivity and reversed the power trend at appro::imately 0.8 seconds.
The transient core average surface heat flux is shown in Figure 7.2. The heat flux was driven by the neutron flux with a time delay induced by the thermal 11 inertia of the fuel. The heat flux decreased during I D
the initial neutron flux decrease and then rose slowly 3 during the power tinnsient to a maximum of approximately 106 percent. N The transient core inlet flow-is shown in Figure 7.3. The pressure wave reached the core through two separate paths. The dryers and separators delay-the pressure wave slightly, so the first effects were realized through the downcomer and lower plenum. The increased lower plenum pressure resulted in a temporary increase in core inlet flow, which initiated the power increase by reducing the core voids. As the pressure 74 i
. . - _ _ - _ - _ _ _ _ _ _ - - - _ - _ _ - _ - _ _ _ - _ _ _ .i
EA-PT-91-0003-S wave arrived at the top of the core, this flow increase was quickly reversed. The slow reduction trend over the rest of the transient was driven by coastdown of the recirculation pumps.
The transient core presrure is shown in Figure 7.4. Following the initial pressure wave transport delay between the TCV and the core, the pressure rose rapidly by approximately 160 psi. At 1.6 seconds, the pressure trend reversed from the effects of the relief valves. When the relief valves reseated, the pressure again rose. By this time, core power was very low and no further transient effects were analyzed.
The transient relief valve flow is shown in Figure 7,5. The core-pressure during the transient event forced the va]ves to open at.approximately 1.5 seconds and remain open until the pressure returned to l l
! near normal at about 4.0 seconds.- 1 l j 75
l i
1 EA-PT-91-0003-S The transient water level is shown in Figure 7.6.
The initial pressure pulse forced liquid from the downcomer into the lower plenum and core, reducing the indicated water level. Reduced recirculation loop flow also decreased the sensed level. As the power level decreased later in the transient, liquid was again fed into the downcomer region from the core outlet via the dryers and separators, so the water level rose.
The transient vessel steam flow is shown in Figure 7.7. The pressure wave reached the vessel at 0.3 seconds, abruptly stopping the steam flow. As the power rose from the increased reactivity, steam production overcame the increased pressu. and forced fluid out of the vessel and into the suppression pool through the SRVs. The fluctuations in-steam flow for the remainder of the transient occurrence indicated damped, harmonic oscillations in the steam lines.
The transient turbine steam flow is shown in 76
l EA-PT-91-0003-S Figure 7.8. The rapid decrease shows the effect of the TCV as the event initiator.
The transient feedwater flow is shown in Figure 7.9. The three motor operated feedwater pumps were-assumed to continue operation throughout the transient.
The transient total reactivity la shown in Figure 7.10. An initial drop in reactivity was expected from recirculation pump trip and control rod movement before the pressure wave reached the core.' The reactivity reacned a peak at approximately 0.7 seconds, at which time the negative scram reactivity and negative reactivity from Doppler broadening reversed the transient and created a strongly negative reactivity gradient for the remainder of the transient occurrence.
Selected comparisons with vendor calculations are shown in Figures 11.11 to 11.17. The GSU RETRAN-results are the same as shown in Figures 11.1 to 11.10.
77 m.--_______-_-_______._-_
EA-PT-91-0003-S In the second set of figures, however, the fuel vendor calculated results for this event are shown for comparison purposes. . The minor difference in results is believed to be due to-the following factors:
The initial water level was 34.8 inches in the vendor analysis and 36 inches in.the GSU analysis. The vendor used a value one-half way between the high and low level alarms; GSU used the nominal value for initial water level.
The vessel steam flow at the elbow-flow meter location was in very good agreement overall.
liowever, the vendor calculations showed a somewhat greater damping'of the flow oscillations, when the relief valves were open between 1.5 and 4.0 reconds.
The ACPR calculated for this transient was 0.10.
The value calculated by the fuel' vendor for the same 78
I EA-PT-91-0003-S trensient was 0.07.
7.2.2 Pressure Reculator Failure A pressure regulator failure transient is evaluated for each reload application because events of this type are potentially limiting in thermal margin effects.
This event is similar to the load rejection transient because it commands a closure of the TCVs.
This event usually causes a less severe pressurization because the TCVs close in the normal rather than the fast mode. The event is potentially limiting because the reactor scram is delayed until a high flux or high pressure setpoint is reached.
the When the pressure regulator fails downscale, pressure controller senses a pressure lower than actual system pressure and commands a decrease in steam flow The TCV to return the pressure to its setpoint value.
closes slowly, allowing the core to pressurize until 79 o
EA-PT-91-0003-S the event is terminated by the high flux trip.
The RBS RETRAN model was used to evaluate the pressure regulator failure transient. A typical-pressure regulator failure transient based on RBS Cycle 3 operating conditions and core configuration is shown in Figures 7.18 through 7.27. A sequence of events for the PRDF transient is given in Table 7.4.
The transient neutron flux is shown in Figure 7.18. A peak flux of approximately 135 percent was reached at about 2 seconds after event initiation. The 11ux increase was terminated by a high thermal flux trip at 1.87 seconds. The flux decreased rapidly as the control rods were inserted.
Figure 7.19 shows the transient core-average surface heat flux, which reached.a peak value of 103-percent at about 2.1 seconds after event initiation. The reactor scram terminated.the heat flux 80 i
i
)
l l
EA-PT-91-0003-S l
rise by rapidly reducing the heat generation rate.
The transient core inlet flow is shown in Figure 7.20. Initially, the cor6 inlet flow increased.
As in the LRNB, the pressure wave reached the core through two separate paths: through the dryers and separators, and through the downcomer and-lower plenum.
The increased lower plenum pressure resulted in an increase in core inlet flow, which' initiated the power increase by reducing the core voids. The core flow was reduced at about 3.6 seconds as the recirculation pumps underwent coastdown.
The transient core prescure rise-is shown in Figure 7.21. After a delay due to propagation of'the pressure wave from the TCV to the reactor vessel, the pressure rose rapidly by approximately 140 psi.: At approximately 3.8 seconds, the pressure trend reversed I when the relief valves opened. When the relief valves
! rescated at about 6 seconds, the pressure again rose.
1
! 81 4 1
l
'l I
EA-PT-91-0003-S The pressure rise continued until about 8.5 seconds when the relief valves reopened, and pressure again decreased. Core power was very low by this time and the thermal transient was complete.
Figure 7.22 shows transient relief valve flow.
Steam pressure during the transient event forced the valves to open at approximately 3.6 seconds and remain open until the pressure returned to near normal at about 6.0 seconds. Following. closure of the relief valves the pressure again increased and reopened the valves starting at about 8.5 seconds. The step increases in flow reflect the opening of groups of valves with different opening-setpoints.
The transient water level is shown in Figure 7.23. The-initial pressure increase forced liquid from the downcomer into the lower plenum and core, reducing the indicated-water level. Reduced.
recirculation loop flow following recirculation pump 82
i EA-PT-91-0003-S trip also decreased th( sensed level. As the power level decreased later, liquid was again fed into the downcomer region from the core outlet through the separators, so the water level recovered slightly.
The transient vessel steam flow is shown in Figure 7.24. Following the steam line transit delay, the pressure wave reached the vessel rapidly decreasing the steam flow. Continued steam production in the vessel raised the pressure, causing the relief valves to open at about 3.5 seconds. The vessel steam flow then returned to the original value until the valves closed at about 6.0 seconds. -Increasing pressure following valve clocure caused-the valves-to reopen at about 8.5 seconds again increasing the steam > flow.
The transient turbine steam flow is shown in Figure 7.25. The flow decreased over a 3-second period as the TCVs closed'due to the zero demand-signal-from-the assumed failed controller.
83
t EA-PT-91-0003-S The transient feedwater flow is shown in Figure 7.26. The feedwater pumps continued operation, but the regulator velves first closed'somewhat in response to the decreased steam flow. At about 4 seconds, the feedwater flow began to increase in response to the low water level.
The transient total reactivity is shown in Figure 7.27. Initially, the reactivity increased due to pressurization-driven void collapse and core inlet flow increase. The reactivity reached a peak at approximately 2.0 seconds, at which time the scram and revoiding caused reactivity to decrease. The dimenishing effect of the scram is offset by-the-pressurization-driven void collapse, resulting in a slight increase in reactivity at about 3 seconds.
Subsequently, the void collapse rate decreased so that the reactivity rapidly decreased-for the remainder of the transient event.
t 84
i EA-PT-91-0003-S Comparisons with vendor calculations are shown in Figures 7.28 to 7.35. The GSU RETRAN results are the same as shown in Figures 7.18 to 7.27. In the second set of figures, however, the fuel-vendor-calcu ated results for this event are shown for comparison purposes. Differences in results are due to the following factors:
The difference in timing of the pressure and other responses was due to the manner of modeling the pressure regulator failure. The G5U model assumes that the regulator begins to output a zero demand signal at time zero.. The TCV servo then responds to this zero demand by closing the valve. There is e lag caused by the valve servo response characteristics. The vendor model shows
.less lag. This is-apparent from Figure 7.34 which shows a basically linear decrease in turbine steam flow over a 2.5 second period for the vendor response.
85
-,-w i
EA-PT-91-0003 S l
Figures 7.36 to 7.43 show the same comparison of vendor and GSU responses for the PRDF event. In this case,_the GSU model_was modified to have a linear decrease in TCv position comparable to the 1 l
l vendor model. The vendor initial water level of l i
34.8 inches was also used. The timing of events 1
l I
can be seen to correspond much more closely, !
showing that the differences in timing apparent in Figures 7.28 to 7.35 are due to the different regulator models. Despite these differences in timing, the responses that affect MCPR, such as peak thermal flux, peak pressure,' core flow at time of peak heat flux were not affected. .The only effect was a time shift in the response.
The initial water level was 34.8 inches in the vendor analysis and 36 inches in the GSU analysis. The vendor used a value one-half way j between-the high- and low-level alarms; GSU used the nominal value for. initial water level.
86 i
o
EA-PT-91-0003-S I
)
The GSU methods calculated a ACPR value of 0.03 for this event. The value calculated by the fuel vendor for the same transient was 0.04.
7.2.3 Feedwater Controller Failure A water-level increase trar.sient was evaluated for each reload application be:?.use events of this type are potentially limiting in thermal margin effects.
For BWR/G plants, a feedwater controller failure to maximum demand is the most severe event of this type.
Such an event is initiated by a failure of the feedwater control system which results in generation of a maximum demand signal.. The feedwater flow quickly increases-to the runout value,-and the water level increases until a high level-8 turbine trip occurs.
The increased core inlet subcooling due to the high feedwater flow causes the power level to increase prior to the turbine crip, so that the trip-is initiated from 87
EA-PT-91-0003-S a condition above rated power.
This type of event differs from a turbine trip because it is initiated from an above-rated power condition with a high water level. Bypass valves are assumed to be operational.
The RBS RETRAN model was used to evaluate a i
feedwater controller failure transient. A typical feedwater controller failure transient based on RBS Cycle 3 operating conditions and core-configuration is shown in Figures 7.44 through 7.55. A sequence of-events for the feedwater controller failure transient
-is shown'in Table 7.5.
The transient neutron flux is shown in Figure 7.44. The flux increased approximately linearly over the period up to the turbine trip. The flux increase was terminated by a scram on turbine trip.
The increasing pressure and core flow caused a 88 s
EA-PT-91-0003-S mcmentary reversal in the flux decrease sufficient to create a peak flux of approximately 129 percent just after the trip time of 18.9 seconds.
Figure 7.45 shows the transient core average surface heat flux, which reached a peak value of.
102.6-percent at the time of trip. The reactor scram terminated the heat flux rise by rapidly reducing the heat generation rate. Because the rate of power increase was slow for this event prior to the scram, the thermal flux increase followed the neutron flux increase much more closely than for faster events, such as the LRNB and PRDF.
The transient core inlet flow is shown in Figure 7.46. After a momentary decrease and then increase due to pressure wave effects, the core inlet flow decreased due to tripping of the recirculation pumps.
89'
LA-PT-91-0003-S Figure 7.47 shows core inlet subcooling versus time. The subcooling increased slowly as the water level increased due to the increase of feedwater flow.
The pressure rise following the trip caused the subcooling to increase. Subcooling followed the pressure changes which caused the saturation enthalpy to change.
The transient core pressure rise is shown in Figure 7.48. The pressure changed very little during the period when the water level was increasing prior to the turbine trip. Following the trip, the pressure rose rapidly-by approximately 130 psi. At approximately 21.0 seconds, the pressure trend reversed-when the bypass and relief valves opened. When the relief valves reseated at about 23 seconds, the pressure again rose. The pressure rise continued until about 28 seconds when the reduced power and bypass valve flow caused the pressure to begin' decreasing.
- Core power was very low by this' time and the thermal 90
EA-PT-91-0003-5 transient was complete.
Figure 7.49 shows transient relief valve flow.
Steam pressure during the transient event forced the valves to open at approximately 20.5 seconds and remain open until about 23.0 seconds, Figure 7.50 shows bypass valve flow during the transient event. When the valves opened in response to increased pressure demand at about 19.0 seconds, the flow initially spiked to a value above the 10 percent NBR nominal value. However, it quickly adjusted to about 10 percent NBR for the remainder of the event.
91 a-
EA-PT-91**0033-S 4
The transient water level is shown in Figure 7.51. Water level first increased due to the increased feodwater flow. Wher the high level-8 trip was reached at about '" 0 seconds, the pressure .
incre:ge forced liquid from the dcwncomer into the lower plenum and core, reducing the indicated water j level 4 Reduced recirculation loop flow following l recirculation pump trip and feedwater pump trip on level-8 also decreased the sensed level.
The transient vessel steam flow is shown in Figure 7.52. Steam flow remained relatively constant during the period when the water-level was increasing prior to the trip. The turbine trip caused the stop valves to close rapidly which decreased _the vessel-steam flow. The steam flow oscillated at a low value until the relief and bypass valves opened to relieve the pressure. increase. When the relief valves closed, the steam flow decreased _to tr.atch the bypass valve flow of 10 percent NBR.
92 .
(
L i
_ ..-. .. _ __ _ __ _ _.-_- ~ . _ _ _ ._____ _ _ _ _ . . _ _ ..._ _ _ -
l EA-PT-91-0003-S ,
The transient turbine steam flow is shown in Figure 7.53. The flow remained essentially constant at ;
the initial value until the turbine trip. The flow went to zero over a 0.1-second period as the turbine stop valves closed.
The transient feedwater flow is shown in ,
Figure 7.54. The flow increases rapidly from rated to ,1 116 perc1nt rated, which is the runout value, and j remained at this value until the turbine trip on high level-8. This signal tripped the foodwater pumps. The l feodwater flow coasted down to zero over a period of approx 4 stely 3 seconds.
The transient total reactivity is shown in Figure 7.55. . The reactivity increased slowly as the water level increased due to the increased subcooling !
caused by the high feedwater flow. At the trip, reactivity initially decreased oce to insertion of
. control rods. A slight increase following the initial 93 ,
D *We- -C-
EA-PT-91-0003-S decrease was due to the positive reactivity effects of pressurization-induced void collapso and increased core flow. Subsequently, the pressure and core flow both decreased so the reactivity rapidly decreascC for the remainder of the transient.
Comparisons with vendor calculations are shown in Figures 7.56 to 7.62. The GSU RETRAN results are the same as shown in Figuren 7.44 to 7.55. In the second set of figures, the fuel-vendor-calculated results for this event are shown for comparison purposes.
Differences in results are due to the following factors:
2
- The difference in timing of the pressure and other responses was due to the different scram times predicted by the two methods. Figure 7.60 shows that there were differences in the initial levels, level setpoint for the trip, and also the filling rata. The runou'. flows were also 94
EA-PT-91-0003-S slightly different. These small differences did not affect-the predicted-heat flux at the time of the trip (see Figure 7.56) which was the key variable for thermal limits determination in this event.
The event was simulated again with the feedwater
'l runout flow rate set so that the high level-8 trip would occur at the same time for the GSU and fuel vendor calculations. Also, the bypass flow was modeled as a negative fill. This produced bypass flow predictions similar to those of the vendor model in the sense that there was no overshoot of tne flow above the 10 percent bypass capacity when the valves opened.
Figures 7.63 to 7.69 show the results of that case.
Note that the results are much closer in this case.
The AcPR calculated for this transient was 0.10.
The value calculated by the fuel vender for the same transient was 0.06.
95 9
i
_ - _=
EA-PT-91-0003-S 7.3 EVALUATION OF OFF-ilOMINAL OPERATIliG CONDITIO!iS In previous subsection, analyses contributing to the calculation of the rated MCPR operating limit were described. Additional analyses were performed to validate or redefine as necessary the power- and flow -
dependent MCPR and MAPLl!GR limit curves.
All of the transient analyses performed during the evaluation of a proposed reload core configuration were used to validato existing off-nominal operating limits. The initial conditions and final results of each analysis either confirmed or invalidated the existing curves on the basis of CPR and heat flux effects-e 96
EA-PT-91-0003-S 7.3.1 Power-Denendent MCPR Limit A function defining the operating limit MCPR as a function of core power is established by a statistical evaluation of the Control Rod Withdrawal Error transient in the fuel vendor's standard safety analysis report". The GSU evaluat! ' of the control Rod Withdrawal Error transient is described in the core 2
analysis topical report.
A typical power-dependent MCPR limit functian is shown in Figure 7.70.
7.3.2 Flow-Dependent MCPR Limit A function defining the operating limit MCPR as a function of core flow is established through analysis of slow-power increase transients. Core power is assumed to ascend a constant rod line on the power-flow map until the physical limitation of core flow-is 97
EA-PT-91-0003-S reached; the MCPR at this point is not allowed to violate the SLCPR.
A transient was evaluated with a quasi-steady-state FIBWR analysis. A constant rod line was selected that terminated at the power level associated with the high-flux scram and flow corresponding to the physical limitation of the recirculation system. Power peaking was selected so that core MCPR was equal to the safety limit MCPR at the termination point. Several power-flow statopoints were evaluated along this constant rod line and core MCPR values were recorded.
This analysis defined a MCPR versus flow-limit function, which protects the safety limit MCPR during a i slow, unprotected flow increase transient. Transients that proceed more rapidly terminated at lower power levels by the-APRM flow-biased' scram. A typical' flow-dependent MCPR limit curve is shown in Figure 7.71.
98 1
N'Mr -*~" * -r&---**"-V F P --P W + *"' " - * =* r "
l l
l l
EA-PT-91-0003-S 7.4 PROTECTION OF PIPING AND COMPONENTS FROM OVER-PRFSSURIZATION Application of Section III of the ASME Code" requires that piping and components within a qualified system be protected from overpressurization during the occurrence of a postulated accident. This code mandateu that accident analyses assume the most adverse.
anticipated conditions and the unrelated failure of the most critical active component.
According to the code, the maximum pressure in components cannot exceed 110 percent of design pressure. For purposes of this analysis, the reactor vessel was considered a component. The RBS reactor vessel design pressure was 1250 psig, so the pressure-under maximum overpressure accident conditions, could not exceed 1375 psig.
The single-failure requirement changes the nature
- 99 I
- -- ,, r,- w --
,- g - - - -
l i
EA-PT-91-0003-S l
of the most severe overpressurization transient.
Although the load rejection transient is more severe when all protective systems function correctly, failure of the reactor to scram in response to the containment isolation signal makes the containment isolation event more severe. The ASME maximum overpressurization accident analysis assumes a contair ment isolation signal with concurrent failure of direct reactor scram logic. Reactor scrat. terminates the accident when either the flux or the pressure reaches the trip setpoint. The containment isolation signal causes the main steam isolation valves (MSIV) to close over a 3-second period. This cessation of the steam flow preasurizes the reactor vessel, causing void collapse I
- and increased power generation. The power increases until the high neutron flux trip scrams the reactor.
Pressure. increases to open the safety valves. Steam flow from the safety valves terminates the overpressure event.
I 100
I EA-PT-91-0003-S Initial conditions for che overpressuro protection analysis are listed in Table 7.2. RETRAN results of a hypothetical ASME overpressurization transient event are shown in Figure 7.72. This figA e shows pressure at the bottom of the reactor vessel versus time. The GSU methods resulted in a maximun vessel pressure of 1242 psig. The fuel-vendor analyses of record showed a maximum vessel pressure of 1248 psig for the same transient event. Peak pressure was the only variable of interest.
101
____--_-_m m_--_ - _ - . . _ _ _ _ _ _ _ .
a EA-PT-91-0003-S Table 7.1 Initial conditions for Pressurization Events Paraineter Units Value Core Power Hw 2,894 Core Flow Mlb/hr 84.5 Dome Pressure psig 1,040 Core Leakage Flow % 12.79 Steam Flow Mlb/hr 12.45 s TCV Position % open 48.4
- Reactor Water level Inches above zero 36.0 Steam Line Pressure Drop psi 40.0 Core Inlet Enthalpy Btu /lb 527.8 102 l
EA-PT-91-0003-S Table 7.2 Initial Conditions for Overpressure Events Parameter Unito Valun I Core Power MW, 2951.9 Cotc Flow Mlb/hr 84.5 Dome Pressure psig 1042.7 Core Leakage Flow % 12.31 Steam Flcw M1b/hr 12.74 TCV Position % open 65.1 Reactor Water level Inches above zero 36,0 Steam Line Pressure Drop psi 41.8 Core Inlet Enthalpy Btu /lb 527.9 103
-- . _ _ _ _ - _ _ _ _ _ . - - . - _ . - - _ _ _ _ = _ - _ _ - - _ _ _ _- _
EA-PT-91-00,3-S I
Table 7.3 sequence of Events for RDS Load Rejection without Bypass Transient - Design Basis Ijag Event 0.00 100% power & 100% flow steady stato 0.Gv Load rejection initiated turbine control valve fast closure (turbine trip) 0.07 Reactor scram by turbine trip
- 0. le Recirculation pump trip to LFMG 0.72 Peak power 0.83 Peak RCPR 1.31 SRV group 1 (one valve) open 1.47 SRV group 2 (eight valves) open 1.48 Recirculation pump ATWS trip by high pressure 1.55 SRV group 3 (seven valves) open 1.60 Peak pressure 3.94 SRV group 3 closed 4.19 SRV group 2 closed 4.39 SRV group 1 closed 6.00 End of simulation 9
104
.y , , _- -. . , - . -
EA-PT-91-0003-S Table 7.4 Sequence of Events for RDS Pressure Regulator Failure Transient Timg Event 0.0 100% powar & 100% flow steady state 0.01 Pressure regulator fail downscale 1.87 Reactor scram by APRM high trip 1.98 Peak power 2.08 Peak RCPR 3.48 SRV group 1 (one valve) open 3.60 SRV group 2 (eight valves) open 3.68 Recirculation-pump ATWS trip by high pressure 3.70 Peak pressure 3.72 SRV group 3 (neven valves) open 5.87 SRV group 3 closed 6.05 SRV group 2 closed 6.21 SRV group 1 closed 8.50 SRV group 1 reopen 8.87 SRV group 2 reopen 10.0 End of simulation 105
l EA-PT-91-0003-S Table 7.5 sequence of Ovents for RDS Feedwater controller Failure Transient Time Event r 0.0 100% power & 100% flow steady state 0.0 Feedwater controller fail 18.85 (L-8) trip by high water level 18.86 FWP trip by high water level 18.90 TCV trip (fast closure) 18.91 Reactor scram by high water level 28.95 BPV start opening 18.99 RC pump trip by TSV trip 19.65 Peak power 19.70 Peak ROPR 20.43 SRV group 1 (one valve) open 20.58 Recirculation pump ATWS trip by high pressure 20.59 SRV group 2 (eight valves) open 20.G0 SRV group 3 (seven valves) open 20.6 Peak pressure 22.46 SRV group 3 close 22.60 SRV group 2 close 22.74 SRV group 1 close 30.0 End of simulation d
106
I EA-PT-91-0003-0 Figure 7.1 RETRAN Load Rejection Event Response - Neutron Flux REIR;N L0iG REJEC110N EVENT RESPONSE i i i i e i i i i i g a
~
.B b'
n i -
u.g 5
E S .
zg -
f {
s s
l 8 8
i i e i 3
, , , , I ca.3) .50 1.T l.E !.00 2.50 LM L50 9.00 9.2 1.30 5.50 4.00 l.
TIME ISEC) i 107 l
v - - .
EA-PT-91-0003-S Figure 7.2 RETRAN Load Rejection Event ReJponse - Average Surface Heat Flux R{TRAN LOT REICTM ERNi RfTb3Y
^
' i l I I I I I I I l N
n
, #4 X
J ~ '
J k@ -
F G
W I
ko -
TU 0
E
~
Zo gm t
F W
. K 1 I I I 1 l _, i I I I !
d.T .T l.T l.9 2.T 2.9 3.T 3.2 1.2 1.2 5.T 5.2 6.1 Id {$0l 108 4
n, , - , , ,
EA-PT-91-0003-S Figure 7.3 RETRAN Load Rejection Event Response - Core Inlet Flow REIRAN LOAD REJECil0N EENT RESPONSE S i i i i i i i i i , i r
88 .-
r
.L h
W Y ..
l 8
U
~
\~sw%
es C
1 x l
l
~
\ .
I l I I I I t _ i f f f _
l d2 .2 1.2 1.50 2.2 f.50 3.2 3.50 9.2 9.M 5.T 5.2 6.2 11E & cl 109 s
xw-
i EA-PT-91-0003-S Figure 7.4 RET.AN Lotd Rejection Event Response - Pressure Rise (psi)
ElRflN LOR)EICTION[YENT ESPONSE 8 I i i i i i i i i i i N
n M
o (L
wo -
w-n w
x W
No -
inB N
x /
L z -
Eg -
xe F
W x
1 I I i i i i i 1 i i d.T ,9 1.T 1.9 ?.T 2.2 3.T 3.2 9.T 4.2 5.T 5.2 6.T .
Ilf(SEC}
110
EA-PT-91-0003-S Figure 7.5 RETRAN Load Rejection Event Response - Reluf valve Flow (%)
RETRANLOADREJEC110NEVENTRESPONSE 3 i i i i i i i i i m
~
J O e
J d
d e
7 CS E
E I f i 1 I I I I i d.2 .M 1.2 1.E ?.T 2.9 3.E 3.50 9.2 4.50 5.3 5.2 6.T 11HEISECl 111
.i
=l
. _ _ i
i i,
+
EA-PT-91-0003-S Figure 7.6 RETRAN Load Rejection Event Response - Narrow Range Level (inches)
REIRfiN LO W REJEC110N EVENT RESPONSE i i i i i i i i i i g
W 5 Y G
~
bo, 9 X s
, za z
Z E
E I 1 1 1 1 1 1 1 1 1 1 CC.T .2 1.2 1.2 2.2 2.2 3.T 3.2 4.2 9.2 5.:0 5.2 6.2 11E ISEC) 112 i
ii
l l
[
l EA-PT-91-0003-S i
Figure 7.7 RETRAN Load Rejection Event Response - Vessel Steam Flow RETRRN LO@ REJECil0N EVENT RESPONSE R i i i i i i i i i i i g -
U
~3 08 w- 7 E
a .
.S 0
kS E
E
, j l I ! I I I I i i t I d.T .50 1.T 1.9 2.*T !.T 3.00 3.50 9.00 9.50 5.00 5.50 6.00 ilf ISEC) 113
EA-PT-91-0003-S l
1 Figure 7.8 RETRAN Load Rejection Event Response - Turbine Steam Flow 1
REIRRN LO T RE JC110N :N RESPONSE l i i i i i i i i i 4 S i U
8 , -
as
- o. -
W fn ,
Z 5
5
$3 m
b z
~
1 1 I t f f I f 1 1 1
.9 1.E !.T 2.2 3.00 3.2 1.00 9.2 5.:0 5.2 6.00 d.2 1.:D 11ME (:iEC) 114 s
e . -
EA-PT-91-0003-S Figure 7.9 RETRAN Load Rejection Event Response - Feedwater Flow RETRAN LOT REJC1104 fYENT RESPONSE i i i i i i 8 i i i 0
N -
5
>g 3'
L z
d -
a -
3p o
u L
z c
$S u
z I I I i 1 I f i 1 l g 1,3 t,@ g,g 3,z 3,g 9.2 9.2 5.2 S.2 6.2 d.2 .2 1.2 il E ISEC) 115 I
l I
l I
T T-
EA-PT-91-0003-S
(
Figure 7.10 RETRAN Load Rejection Event Response - Total Reactivity ($)
i l I I I I I I I i 9 I I I I N
n m
w b
O E
W ~
%0 f
E F
D F
Z3 - -
E-F W
0 1 I I 1 1 I I I I I I
)
%s a la la is 13 32 33 - ta 13 sa sa sa IK l101 116 t
s EA-PT-91-0003-S Figure 7.11 Comparison of RETRAN and GE LRNB - Base - Average surface Heat Flux CMWISH EliR IN) CE LMBP - BASE i i i i
{~j i i i i i i i e -
n ~
e w X
- 30 g y;0 -
J. L.
J k h E
6- W G I W
Ig 1g 3 -
, E
~
'a'
]
t n -
e 3 s in 0
> g 0 E Y csu .
"E W
l3 E O 6-
.. U r
, 1 1 l 1, 1- I i 1 I I 1
- o - d.2 . 9 1.E 1.9 f.2 2.2 3.2 3.5 1.2, 1.3 1.2 $.T 6.T
-Tif(SEC) 117-m . . D
EA-PT-91-0003-S Figure 7.12 Comparison of RETRAN and GE LRND - Daae - Core Inlet Flow CMHR153 Of RETRAN R0 DE LROSP - BASE g , g -- i i i- i i i i i i i i j
~
- U g - g 3- J~
o u.
J
- u. s ct W
j 4
F J j j, ~ b, ,.
-~-
g, ~
tc h W
I D
o I m '- .' .
csc o
u _g .
WS ES O
r E E I i i I I i 1 1 I i 1 o d.T .2 1.2 1.T 2.T !.2 3.2 1.2 4.2 9.2 5.2 5.T 6.I ilE EC) 118 1
I
I l
EA-PT-91-0003-S i
Figura 7.13 Comparison of RETRAN and GE LRND - Dasa -
Pressure Rise (psi)
. l l
CONIS) Of EM R0 CE LRN - BASE j~ j i i i i i i i i i i i H
- 0
- n.
0o -wo -
LW O
~~ W" 0
W H 0 z w
Z W W8 58 --a_
- z. n ,
J O '
, E
[
z
/ /
/
WS to 0 F W
- x
. I I I I f 1 1 1 1 1 1 o d.T a 1.T 1.s 2.x ts 3a 3.s <x $.s s.x s.s 3.x M (SEC) 119 4
, _ . - ..._ . _. _ _. _ . ._.._ _. - . ..._.. . _ _ . _ . .... . . . ~ . . . _ . . . . . _ . . _ _ . . _ - _ _ _ . _ . - . _ . _ _ _ _ _ _ _ . . _ _ . -
1 EA-PT-91-0003-S -
i Figure 7.14 Comparison of RETRAN and GE LRNB - Base - Relief ,
Valve Flow (%)
CTARISM & REIRfW fND 21.RGP - BASE e
~
$." Q i i i i i i i i i i I 4
3 .
e4
- 3 o -Do - -
30 JU N, 0" L" y L- W W J
>. C
-J- .> .
Em e
>e u. e L'
L w W J y= -
J E
[Y _,
~s as E
i i E O H l W
E E $
- :=
CE l hCSU -
l l
-l' ,, _g ji t t i' I i- 1 1-o da .9 12- : 1.2 ' 23 - 2.2 3.1 -- 3.2 94 1.9 54 5.2 - 6@-
- ilMEISEC1 w- .- -. --,\ - , , n .'e 4 me
-re .e t. --,v.v e 5 ,+ . e- o- - . e- e y
__-z__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - --
EA-PT-91-0003-S h
Figure 7.15 Comparison of RETRW and GE LRNB - Base-- Vessel Level-(inches)
COMPAR15N Of RETH FND DE LRV0BP BfE 4 --
g - '. g i i i i i i i i i e i S S B- .
3 E W c ~
'be ho ~*
2' U >
W d
> . w -. .
"US OS E csu a .
W T L1 C G1
$~
'Z z - W -
2 a2 W E 0 g
{-
- dE -
- ~ ~
. - e e i 1 1- i l' I I i t I t-o da .2 1.5 1.2 2.2 . f.2 3.00 3.2 - - 4.3 9.2 ' 5.00 - .5.2 6.3
=TIMEISECl
'121' O -
- - -- _ ._.___i._____.___________
EA-PT-91-0003-S Figure 7.16 Comparison of RETRAN and GE LRNB - Base - Vessel Steam Flow (%)
CCfARISM OF RETRIN RND GE LR'M - BASE I l l l l l l l l
- l g - g ,
~
=3,
_ e . -
.\ en
- 3 GE
{
38 ~3 y -
0" to d r
- {
de c A csc -
mo Jo fT u .
J fn
$ O - i n >
W
>O
- u $*O o b W
- 1I 1 21 1 i t i i i .
o d2 .S 12 1.2 !@ 2.2 3.00 3.50 42 q.T 5.T 5.50 E.I TIMEISEC) 122
EA-FT-91-0003-S Fig".tre 7.17 Comparison of RETRAN and GE LRNB - Base - Total Reactivity ($)
4 CMHRISM ETE N CE LMBP - BASE.
9r N
9N i i i i , , ,
M e w 0 ~po
- F >
> b w 0 F E O W Eo -to - - -
J c
J F G D o e-csv .
F9 -z9 .
a- ce - -
.dI ZI O F g
E 9-.9 I l l' ! ! ! ! I 'I i 1 7 4.2 a la la 23 '2a 33 .
3.s . 1.m -- 13 5[x - 5s s.I
_ ' TIE (SECl 123.
EA-PT-91-0003-S Figure 7.18 RETRAN Pressure Regulator Failute Response -
Neutron Flux (%)
REIRANPRES5UREREGULATORFAILURERESPONSE i i i l i I i i i o
N te "o - -
X$
a Z /
o -
e a
W Z
Z C
' ~
o WW Z
! ! ! ! f l l l 01.0 1.0 2.0 3.0 4.0 5.0 - 6.0 7.0 8.0 9.0 10.0 IlME(SEC}
124
~
EA-PT-91-0003-S Figure 7.19. RETRAN Pressure Regulator Failure Response -
Average Surface Heat Flux (%)
RETRAN PRESSURE REGULATOR FAILURE RESPONSE S I i i F i i i i C
U X'
3 u8- _
E b
^
n G --
C-
'l [@
E u.
~
~f f I f t t g ., ,-
' cD.0 : 1.0 2.0 - 3,0 4.0'- 5.0 6.0 , 7.0 8.0 9.0 10.0 ,
. TIME (SECl- -4 125
_ -__-__.m.a___._____ _- - _ - _
EA-PT-91-0003-S Figure 7.20 RETRAN Pressure Regulator Failure Response - Core Inlet Flow (%)
RETRRN PRESSURE REGlD' TOR FRILURE RESPONSE
! ' ' I i i i i , ,
n W
E,8 -
lL z .
r _
E 8
Z
~
3 0
x
. 1.0 2.0 3.0 4.0 5.0 6,0 7.0 e.0 9.0 10~e '
TIME (SECl 126
EA-PT-91-0003-S Figure 7.21 RETRAN Pressure Regulator Failure Response -
Pressure Rise (psi)
REIRRN PRESSURE REGULAIOR fall.URE RESPONSE 8 i i i i i i i i i N
m M
Q. -
"o -
U-0 U
To -
Jo -
0~
0 ill Z
a.
Z -
(r .
W
(
1 I I I I I i l l d.0 1.0 2.0 3.0 1.0 5.0 6.0 7.0 8.0 9.0 10.0 IlMEISECl 127 v.
EA-PT-91-0003-S Figure 7.22 RETRAN Pressure Regulator Failure Response - ,
Relief Valve Flow (%)
RETRAN PRESSURE REGULRTOR FAILURE RESPONSE i i i i i i i i i
, 8 U
u_ -
d LL.
d
~~
- d e
- Z G3 E
-!J r-N I I i 1 i f f I cD.0 1.0 2.0 3.0 9.0 5.0 6.0 7.0 8.0 9.0 10.0 TIME (SEC) 128
EA-PT-91-0003-S Figure 7.23 RETRAN Pressure Regulator Failure Response -
vessel Level (inches)
RETRANPIESSUREREGULATORFAILURE' RESPONSE i
g i i i i i i i 5 >
ta 5
50
~
_r -
d aPn -
U 0
z -
E"a ru e
e g _
l i I I t 1 1 ;
d.0 1.0 2.0 3.0 4.0 g
5.0 6.0 7.0 B.O 9.0 10.0 TIME (SEC1
{
129
l EA-PT-91-0003-S' Pigure 7.24 RETRAN Pressure Regulator Failure Response -
vessel Steam Flow (%)
RETRAN PRESSURE REGULATOR FAILURE RESPCNSE 8 I i i i i i i i i 0 _
N 38 -
- u. -
k u \
in -
E 8 !
W
[s s
~
h +
u g .
' I ' I i 1 i i i i c0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 1 TIME ISEC!
130
.e 4
_ ~ - _ _ _ _ _ _ _ - - _ . - _ _ - _ _ _ _ _ _ _ _ . _ _
- . . . . - . . ~ . - _ - - .. - . . ~ . .. .. - - - . . , . . - . - . - . . - -. - _
b r
i
-)
EA-PT-91-0003-S Figure 7.25 PETRAN Pressure Regulator Failure Response -
Turbine Steam Flow (%)
. RETRAN PRESSURE RECULATOR FAIUJRE RESPONSE ,
S i i i _1 i i i i i U
N
,, i.
a 0
u.
__i g -
r E
-W w
ui -
WE z
m ,
e a
g_ -
zg 1 Em H-g x +
~
.0 l
1 I- I t , i r- -i- ,
p- cf).0 - 1.0 - - 2.0 ~ 3.0 4.0 5.0 . 6.0 ' 7.0 8.0 9.0 10.0 2
. TIME (SE01' 131-1-
3
, a
? .
J>
d
_,. _ .h.- ,,-.a, . --._,,.r-. . , - ~ e,.ww+w,v,- err -
.%c , , - - - . - y --.y[-,% -.c-, g.m t. y y-yw,-
EA-PT-91-0003-S Figure 7.26 RETRAN Pressure Regulator Failure Response -
Feedwater Flow (%)
RETRAN FRESSURE REGULATOR FAILURE RESPONSE S i i i i i i i i i g -
~
gg %
E Fa 30 o
U u
Z 1 - .
h3 I td l
I I 1 1 f f i l c0. 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 TIME (SEC) 132 O
m
EA-PT-91-0003-S Figure 7.27 RETRAN Pressure Regulator Failure Response - Total Reactivity (S)
REIRflN f'RESSURE REGllATOR fAlLURE RESPONSE 9 l 1 I I I I I I T N,
e m
fa s
r o
E w
go - _
U J
E t-o Z9 - -
E-t' H
w Z-o i i ! ) i i i I I i
%.0 1.0 2.0 3.0 4.0 5.0 E.0 '7.0 8.0 9.0 10.(
IlMEISECl 133 s
EA-PT-91-0003-S Fip. a 7. B GE and RETRAli Pressure Regulator Failure - Base -
Average Surface Heat Flux
, PRESS. REG, FAIL 00VN SCALE - BASE
,h' ! ' i I i i i , ,
~
E ~E 1 O csc ,,
g .
N bdi
~
se -
d s GE N,
I ~
p ~hp _
E S S o o E d Eg zg _
al E c:
A A
- I I I f I l l q ,
d.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 TIME ISEC) 134
EA-PT-91-0003-S:
Figure 7.29 GE and RETRAN Pressure Regulator Failure - Bass -
-core Inlet Flow (%)
- PRESSiREG. Fall 00VNSCALE-BASE
@~, 1 I i- 1 I i. I i .I .
[
- w
, N
- GSU "O - 30 - - -
Q 00 3* J" O - IL J
- u. H g-d.
.g U- Z =
EO A
o -r W
h N I I~ f I f f I l' 'I 3-o- d).O - .l.0 2.0 ~ 3.0 9.0 5.0 - - 6.0 7.0 - ~ 8.0 9.0- 10.C
' TIME ISEC):
135l l
u..
~
t* -
g ny-- e ar-- ? 4 wy**-ss r^
_. _. . _ _ _.. - _ _ . . .- ~ _ . _ . _ _ _ . . _ . . _ _ _ _ - . . - - . . _ . . . - . . - _ _ . _ . . _ . . _ . . _ _ -_
?
EA-PT-91-0003-S Figure 7.30 GE and RETRAN Pressure Regulator FaiJtte.- Base -
Pressure Rise (psi) t
- PRESS. REG FAIL 00VN SCflE - BASE
~ 8 f- 8 1. i i i i i i i .1 N N N
- n
~ L no -wo -
a,0 0
.g-(n c su a
u -
-n x w
't W
.i' e- n-a tn '
in w .A (n E Y u a. CE x
- a. 2 g -
US tS e
1 e i W
e
,,- t ! I 'I t- I' 'i i o d.0 1.0 2.0 .. . 3.0 - 4.0 5.0 6.0 7.0 8.0 9.0 10.0 TlHE(SECl
.. 136 O
.m. , . , . , . - , .,y-r ,, . , , . , y . , . - <
a EA-PT-91-0003-S Figure 7.31 GE and RETRAN Pressure Regulator Failure - Basu -
Relief Valve Flow (%)
PRESS. REG. fall 00VNSCALE-BASE g- 'g , , i i i r -- i i i
~
2 2
- 'l 38 - S8 0"
J 2
- u. "
~
D) ~
U y cE 4
-o-- csu -
$N u. $
u LL U # 'N ;
J % ~
~ ~
8 W { '
Z 1
0 $
i l
1 i ! /
f 1 1 1 1
_ ! 1 6.0 7.0 B.0 9.0 10.0 i 2.0 3.0 4.0 5.0 o cc.0 1.0 I
TIME (SEC) 137 l.
I S
{. _
EA-PT-91-0003-S Figure 7.30 GE and RETRAN Eressure Regulator Failure - Base -
Vessel Level (inches)
P PRESS. REG. Fall 00VN SCALE - BASE
~.
g g i I i i I i i i j f'
3 .
b W ~
U
~
hjo o -
7 ~#
Z m
W W
> - J _
h S JO W
J (0 + CSU W (A in W CE >
cn >
W zo +
>o -
W" o
E" r
W x
S 9
\ ..
. I I 1 I f f I t i o c0.0 1.0 2.0 3.0 9.0 5.5 6.0 7.0 8.0 9.0 10.0 TIME (SEC!
138 i
s
EA-PT-91-0003-S Figure 7.33 GE and RETRAN Pressure Regulator.Failurs - Base -
Vessel Steam Flow (%)
PRESS. REG.FRil.00VNSCRLE-BASE-I I I l l l I I h.
. h.
4 d
g
-g -
U U
g! ~ $!
d I Nk ~
e I w C F w . (n . .
yp -p GE--O= M SU I N <
W E N -
as \
(a 5 w
i
.}
.m t
$ M
.- I t t i 1: t t- i i n- d.0 1.0 - 2.0 3.0 - .- 4.0 .5.0' 6.0 ~ 1.0 -- 8.0 9,0 10.0 '
TIMEISEC) 139 t
4 .
__ - _____-_2. m-_______'_._' __.___.____mL____.
EA-PT-91-0003-S Figure 7.34_ GE and RETRAN Pressure Regulator Failure - Base -
Turbine steam Flow
. PRESS. REG. FAIL 00VN00RLE-BASE sg -
i i i i > i i n .
E.~ E 0
94 3
a8 3 %.
u.g o >
- u. r E
I w-E- H W _
me wr z
w -
z- to g' o - csc z- -
O g ~ zg
' /
ct
~
u e o H w'
x E E
, I I I ~ 1' i 1 1- 1 i n cc.0 - 1.0 2.0 3.0 1.0 - 5.0 ' 6.0 - 7.0 8.0- 9.0 10.0
-TIME.ISEC1 140' i -
. == ___m_____ -______._________.__.__m_. __._._m___-...m_ -___..___.___.____-._._.m_._m.
- m. - _ . .
EA-PT-91-0003-S Figure 7.35 GE and RETRAN Pressure Regulator Failure - Baso -
Total Reactivity
- PRESS.REO. Fall 00VNSCALE-BASE 9F 9 I I I I I I I I I N N m
,, w o -
s w E
> F
- 0 F E o -
0 gd g
gd ^ ^&,k J F E O F F 3 - Z3 -
- E-LLl ' Z' O F L1J
. Z j 9- 9 I I I I
! ! 1 7 %.0 1.0 2.0 3.0 1.0 5.0 6.0 - 7.0 8.0 9.0 10.0 TIME!SEC) 141 4
- , , n -,
. . ._ _- . - - - - . ~ - - . - . . . . . . - . - _ - . - . .. - . - . . .- . . -- - . _ . = . . . . - - . . . _ . - . . .
EA-PT-91-0003-S Figure 7.36 GE and RETRAN Pressure Regulator Failure - MOD - ?
Average Surface Heat Flux
, PRESS. REG.FRll00VNSCALE-MOD g- g i i i i i i i i i
~
p = p =-
2 2 4
-- w .
9 .
~
.X ~ '
xg - 33 a- tu -
.J LL - H E
F w C I W -
r ,e .
tu ? KC a
E VI a
v1 a c- .a s "8 l$8 .ct b
w' .m -1 o s
.y 8 y - csv
~
I -1 1 ! I' I~ 'I i i i -o- ca.O - 1.0 - 2.0 - -3.D ' 4.0 -. 5.0 5.01 7.0 - 8.0 : - 9.0 : 10.0 l !!ME-ISECl 142- i 4
4
.i .
f S' a
4 e- '-- m, as 1- -m=., , , ,,-,a m , -e.w-' ,. ,ser, e m,n ,g.g - y,--=p yy 'y s* q y er y , w y -t ' p s' w
i i i i i i w
~C
~
E esu -
t
.p.
w tt ,
"O "30 '
3 3
~
O 32 CE h.
i J 1 s-W-
- J t W w z -
J- m
[
M z# r b.)
i T w a l x U 1 o U- Z- .
US . g@ z ~
o r l W z
.0 C
~
.- ! ! l~ l
~
'l l g- g .-
o cD.0 ' _ ! .0 2.0 3.0 4.0 - - 5.0 - 6.0 ' 7.0 8.0 9.0 -10.0 i
TIME (SEC) 143 1
l . +'
l-l
- p. -~,- , m y re., - , s , . > . - w w . , - . . , F + =
EA-PT-91-0003-S Figure 7.38 GE and RETRAN Pressura Regulator Failure - MOD -
Pressure Rise (psi)
PRE 35. REG, FRll. 00VN SCRLE - N00 8'
N 8N i i i i i i i i i C
t _.
[ . -
_- m-(0 =:>--cg u -
g I
g f
csu 5
w8 - 58 x- tn -
J (A (D 10 (a t u (L t
Q. Z W8 -$8 h
I I I I I I i i l o d.0 1.0 2.0 3.0 9.0 5.0 6.0 7.0 8.0 9.0 10.0 IIHE(SEC) 144 s
--.-----,....5 EA-PT-91-0003-S Figure 7.39- GE and RETRAN Pressure Regulator Failure - MOD -
Relief Valve-Flow (%)
= PRESS. REO. FRil. 00W SCALE - M00 8~
S i i i ! ! i. 1 i- i g - g -
~
U w
a w 3-
,g Dg =
csu --cm -
o- ' u. - o c, -
- u. t
-W J
> c J .> . . -
$$ u.I p -
- u. - Y J{
- w. g .} .g e J z-E z -
8 m E8 a -
W z
.. C U i
j) , ,- ..,
,- Q-.
o' cD.0 1.0 - 2.0 3.0 ' 1.0 5.0 6.0 ' 7.0 8.0 . ' 9.0 -- -!S.0 TIME-ISEC)
-145
~
EA-PT-91-0003-S Figure 7.40 GE and RETRAN Pressure Regulator Failure - MOD -
vessel Level (inches)
PRESS. REG. FRIL 00'.'N SCALE - M00 I I I I l l l l 1 G
b
~
? dY U J d
> .J _
~
$8 J R w
J c)
N $
m >
~
- Z .
- R crg csc d
w 5l ' f T
~
S 9 p-GE
. I I I i 1 i t 1 o d.0 1.0 2.0 3.0 4.0 5.6 6.0 7.0 33 9.0 10.t TIME (SEti 146 i
s
_. _ - - . _ . . __. , . - _ _ _ _ _ . - . _ . _ . . . - _ . . _ . . _ . . . _ . . . _ _ . _ . _ _ . . _ - - . . . _ _ _ _ - . ~ . . _ , _ _
EA-PT-91-0003-S Figure 7.41 GE and RETRAN Pressure Regulator Failure - MOD -
Vessel Steam Flow (%)
PRESS. REG. fall DOWN SCALE - M00
!~ ! 8 8 8 i- a i i i. .-I
~ N-
~
N H -
n -
e w 3 38 ~ 38 -
o- u_ ~
E -
I w -
E F W _ fn ,
yp gp .e-- C '. -
U ..
J ui -
/
u U1 I
u) y G5U .
i vi > .!
U z
>8 C8
~
e e
.O H I ,
g.
E
. .- .; ~
'N N .
I y
_ i , .. , , ,- , ,- ,-
o d.0 1.0 2.0 3.0 - : 1.0 . - 5.0 6.0 7.0 - - 8.0 9.0 L 10.0:
TIME ISEC).
- [147'.
l
...- , , , . . , , , , , , , , , - . . , . . . . . . , - . . . . +-.,..:. - -
a EA-PT-91-0003-S
\ e
] - - -
Piqure 7.42 ch c.id RETRAN Pressure Regulator Failure - HOD -
Turbine Steam Flott (%)
, - PRESS. REO. fall D M SCALE - MOD i
Q' S T I i i i i I 1
~ ~
f 4 e -
g 3
- 3
~
a@-~ u. @-
3 t (a r ,
a +
lh f .O - .
g i op wp z
U -
Z C)
- T
- D D I e-J - -
O h3 U Z 0 V-W T
, N $
C S U --C=
l I
! ! ! 1... . -- ! ! 4 o cc, 1.0 2.0 3.0 9.0 5.0 LG 7.0 2.0 9.0 10 4 TIME (SEC) 148 E
+
O
.1
N EA-PT-91-0003-S Figure 7.43 GE alid RETRAN Pressure Regulator Failure - MOD -
Total Reactivity ($)
PRESS. REC. fall D M SCALE - H30 9-N 9
N i i i i i i ! i i e
"o - -
y- -
r >
r
> r w U ,
F E a,v to to -]s -
I J a '
J F
- ct a o
- c. W
&.O s 0, -
xs \
g g csv
. El x N 9- 9 ' ' A
! ! ! 1 ! ! 1 7 %.0 1.0 7.0 3.0 4.0 5.0 6.0 7.0 n.0 9.0 13.0 IlME($[Cl 149
~
EA-PT-9.-0003-S Figure 7.44 RETRAN Feedwater Controller Failure 11esponse -
Neutron Flux (%)
TIRM f[fMi[R STR1L{R TAllR EPJ,T i i i i i i , ,
! i i i , i ,
r
~~__
5' J
w E
V i
!x W
z l l l l l I l l l d ! 1 1 I 10 12 li 16 il 3 D h 5 3 I IIf (SECl 150 a
-- w ,_
i EA-PT-91-0003-S Figure 7.45 RETRAN Feedwater Controller Failure Response -
Average Surface Heat Flux (%)
E!R3 f((M![R C31R1LER fall!E ESTI g i i i i i i i i i i i i i i d
i
~
ut
?
a f5 1
tJ i
t I 1 1 1 1 I i ! I I I 1 I I d ! 4 I I 2 12 M $ 18 3 3 h 3 3 I Ili ($[Cl i
l 151 l
l
EA-PT-91-0003-S' Figure 7.46 RETRAN Feedwater controller Failure Response -
Core Inlet Flow (%)
R[iR'h ff[Mi[R C3IR1L(R IAllWE kSPXil i i i i i i i i i i i i g i i 1
A.
BB -
1 d
i "i
y 0
R E8 W
1 1 1 1 I I I I i 1 1 1 1 1 d -! t i i 10 !! li Il II E 3 h 3 3 E 1 K 15Etl 152
EA-PT-91-0003-S Figure 7.47 RETRAN Feedwa ar Controller Failure Responsa -
Core Subcooling (Dtu/lbm)
EINI((.YI[$ $Id3Ihlld[kE I i l i I I i i I I I I I i l
I e
J
\
3 q ~
e-J 0
0 0
o I
D 0
{l 1
h kl I I I I I f I I I I t 1 l l d 2 i i i Il 12 11 16 11 3 D N 3 3 I Ililid l
l 153 l
,n.- P
EA-PT-91-0003-S Figure 7.48 RETRAN Feedwater Controller Failure Response -
Pressure Rise (psi)
R[H f[lMilR (EP111R FRILE UM i i i i i i i i i i i i g i i
~
E .
"O '
L s
E -
a I
i 3
1 t
i . , , i i , ,
d I t I I to 12 19 16 Il 3 !! N 3 3 I lli EC) 154
.s
~
EA-PT-91-0003-S Figure 7.49 RETRAN Feedwater Controller Failure Response -
Relief Valve Flow (%)
EM fl[Mi[R C51R1L{R IRlliR EDfi g i i i i i i i i i i i i i i e
l I
U_
)
b ~
a-i k
d d
x t
( i gi 1 I i I _ J_ i f I f I1 1 J l i d ! 9 i i la li 12 N 18 3 H N 3 3 I ilt 15ECl 155 9
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ - - _ - - - - - - - - --------J
EA-PT-91-0003-S 4 Figure 7.50 TETRAN Feodwater Controller Failure Responne -
Bypass Flow (%) '
IIR%I((D6l[R WIR1L[R IAlllA ESPX[
o i i i i i _i i i i i i r i ;
e U
J .
it g a
'1 C
1 20 T
e Y
2 1 I I I I I I i 1 1 1 1 1 ! , , ,
d I 1 I I Il 12 M 16 Il 3 3 $ 5 3 I 111 (SECl 156
.. . . _ . . . . . _ . _ . _ _ _ . . . _ _ . . _ _ . _ . _ _ _ _ . _ . _ . _ . _ _ _ _ _ _ . _ _ _ . . __ _J
EA-PT-91-0003-S Figure 7.51 RETRAN Feedwater controller Failure Response -
vessel Level (inches)
EM f((Mi[R C*itiL[R FRILM ESiM i i I 4 I I I I i i i i 1 I i h
3 Y
Y 0
J J
0 h
tt g ,
3 I i l I I 1 I i t 1 l 1 1 i d ! t I i 18 12 19 5 11 3 D h 5 3 I Ili 15[Cl 157 h- ___ .--____m___.____m_ ._
EA-PT-91-0003-S Figure 7.52 RETRAN Feedwater Controller Failure R1sponse -
Vessel Steam Flow (%)
Tin fl[Vi[R CAIR1LlR fEllf ESPXiE g , , i i i i i i i i i<r >
' \
W n -
J W
a b
x
! tJ T
k
, , , , , , , , , n , , , ,
I i I a 2 3 D N 3 3 I a 2' e 12 11 18 Ilf (3[Cl i a i
158
4 EA-PT-91-0003-S e
Figure 7.53 RETRAN Feedwater Controller Failure Response -
Turbine steam Flow (%)
i EM f[ DATER C3iR1L[R fAILE ESPXI g- i i i i i i i i i i i i i 7 1
4 h I
3 O
s8
= w I
C d
m
~
Wf l
e x
o-t t I i 1 i i 1 I I i i t i i d I i l l Il !! ~ N 3 Il 3 I h 3 Il I Ili (Ett -
159 i
h ,
- - - --__-_,_--___-_.____.m. ____ _m._ _..m__
EA-PT-91-0003-S i
Figure 7.54 RETRAN Feedwater Controller Failure Response -
Feedwater Flow (%)
l l
En mna rmnia nium um
~
C 1 l l l i i i i i I i i I 6 t
' ~
3E L
T d
3 o(
W W
L Es Li l %
! I I f f i ! ! I i l 1 I i d i i i i 10 12 N 16 II E !!
- N 5 N I M ISEtl 160 s
EA-PT-91-0003-S Figure 7.55 NETRAN Feedwater controller Failure Responsa -
> ttal Reactivity (S)
$N[EI3 @IU10[$IN $E i i F
w
> l s
b u
O
> \
O F
O -
Z .
a-F LO I
o I i i i f I I I I i i ! 1 !
Il 3 # N 3 3 I
't ? t i i to 12 11 16 lif ITU 161
. _ . - r .--m. ,.
EA-PT-91-0003-S Figure 7.56 Comparison of RETRAN and GE FWFAIL - Base -Average #
surface Heat Flux (%)
, Of41Ih I EIG % 7 fHil- Fl gp -
i i i i i i i i i i i- i i i i
.- l.
H
,, w Y
x A xo
,e - 3g 1
\/ p* _csc ",
J .
A l- 1 E
$ y CE#
u
' F F 1
11 3
10 0 0 E F
u e
, O F W
E I
. N ,,
I 4
i ! ! i i i e i t i i i i ! i 2 d 2 1 I I 10 !! H 13 il I R N 2L 21 I IITlEl 162 s
EA-PT-91-0003-S Figure 7.67 Comparison of RETRAN and GE FWFAIL - Dase - Core Inlet Flow (%)
CM419 I RIUh H) 2 ffRil. - TE g,g , , , i , , , , , , , , 4 ,
I )
- csu , ;
/\3 i
_, 3, _ .
r D
LL p
l
.) CE It F U
H J U .2
(
N W"I
~
s T
W D T U O
O Z w$ 2S \
a n W
T
! ! ! ! ! l l l l l l ! ! !
o o ! 1 i I E il N il 11 E 22 h 3 N I Ili (IU 163
1 EA-PT-91-0003-S r
Figure 7.58 Comparison of RETRAN and GE FWFAIL - Dase a ,
Pressure Rise (psi)
MTIS$ 10H N E ff AIL - 53 I i i i !
3',h~ l i i i I i 1 1 1 1 1
(~ [- '
CE .%
- C5U C
- E
[g -8
=
y-id n e
- > . at \ -
e \
0 W O E E
1 2 h)$ kE O F
-4
- 1 1 . _. .
o o ! 9 ! I E !! N !g 13 5 !! N N 3 I M (SEC1 164
's a
-~ ----.a_._--.____ _ _ _ _ _ _ _ _ _ _ , _ _ _ _
EA-PT-91-0003-S Figure 7.59 Comparison of RETRAN and GE FWFAIL - Base - Relief Valva Flow (%)
CXTl3 7 ElG4 %1 fHil- Ef
. l I i l i I l I I I i 1 i I &
h.' f.
w
- g. .
7 ecsc '
g . 8g . ct -.c g g.
C d t s
4 i,
\ ! g L d 5 d J K u
h h
, d w
^
4 1 1 1 1 1 I 1 i }
I
, f i 1 1 1 1 I I U M 16 1: E D d 5 3 I o d t 9 E Ili (1C) 165
EA-PT-91-0003-S Figure 7.60 Comparison of RETRAN and GE F13F.~tIL - Base - Vessel Level (inches)
Of7!ih T RIR % 2 ff AR - Fl 3 I I I I I I I i g I I I I T CE l
" ~
/
, n. ..
c
_ /
. : . /
2 :
s .8 .
9 w3 '
8 i
- e'!: -
8 E w
7.
N L I I f I ! ! 1 I i 1 ! ! ; 1 o d i t i I 3 12 H 11 11 E 2 h 3 3 I lli (SEC!
166
~
e
/
EA-PT-91-0003-S i
Figure 7.61 Comparison of BETRAN and GE FWFAIL - Base - Vessel Steam Flow (%)
CMW1304I f![in % CliJRIL - Fl - i c--T - ;
i i i i i i i i i
gg_ ' . i i i
3 t
- t. ~ t. ct -e ., .
r o csu 3
- 3 .
I' --
o Og' o- 6 il I r li
,e T W j (1 t-W. m I !
O J 1 {,
W -
.) O u o ,
a u l 0 > j l,
U.',R' !S f ( !
e, l o i W i I _
t t w. !
, , , , , , i , ( , , , , J E 22 3 5 3 I i E 12 N il 11 o d ! t 1 IM IEl 167-
^
,s
EA-PT-91-0003-S Figure 7.62 Comparison of RETRAN and GE FWFAIL - Base - Total .
Reactivity ($)
e e
M31 ETH ElfDA M D' I i i i i i i i 3 l l l 1 I I ;
h n j h
- ~
- a. .
- s. ,. !
- w
> F
- U F 2 A D W {t
-go -
U E J % cst C i d
f b
0 F
ct --c=
0 9 e0 2,0 .
c.
J' 2'
, O L W
t o, 0, f I I ! I I I I I d 4 {_, . 1 i 1 2
- i t a t! li li is E a a 3 3 I ili(El 168
EA-PT-91-0003-S Figure 7.63 Comparison of RETRAN and :C FWPAIL - MCP - Average surface Heat Flux (%)
s ,
CXTl3 I R[ip im 0[ ff A!L - G gg i i -i i i i ; -- ! i
-r-s X
3g . ._ _
a- u.
' b c
h W E 1 .
I c o---- GE p CSU W p
h, n O 0 E O h w a 0 F W
2 I i ! ! I t 1 I t l l l 9 ;
O d I 1 l l l$ ll 11 16 il 3 3 h 3 3 3 111 GCl 1G9
-_ - -_-- ._m_______.__-_._m_ . _ _ . _ . _
EA-PT-91-0003-S Figure 7.64 Comparison of RETRAN and GE FWFAIL - MOD - Core Inlet Flow (%)
Of' Tis 3 ElR H)I fER!L - G g ,g i i i i i i i i i i i i i i
. f.u ,
,, GEV g s' l t
,J " L' 3" a O k i F
s $
u z J i!
a "E
W E
O U Z
\
W .Eg a r W
2
- t I
I 1 I I I t i I i 1 I 1 i t o d ! 4 6 1 5 7 N 18 11 3 3 3 3- 3 I IIT (10 170
EA-PT-91-0003-S Figure 7.65 Comparison of RETRAN and GE FWFAIL - MOD -
Pressure Rise (psi)
CM'T!D 3 i013 % 0[ ff AR - G g ,3 i i i 4 i i i i. 1 i i 1 7 l
E' K' . .
R C
[8
' w~ E.
i uj 0 .< . tr t3 # m m W b $
[ ,2 wS t$
0 F W
/
csv I E a
__-j $
- 1 - 1-+ r d ~] N l l l l l I o d 2 9 i i E !? N ll il 'A 3 29 A ll I
(*
Ili (3[Cl 171
)
EA-PT-91-0003-S Figure 7.65 Comparison of RETRAN and GE FWFAIL - MOD - Relief Valve Flow ('t)
~
i I I I I I I I i I l i i l
. .h.
l l
}
~
)
g! j! ct -c
'}
J L W J >
k C HCSU t
W J
- W S gS D
x
, l
- ! ! 1 1
! !. ! ! 11 1 4 ! 4 1 >
o d I 9 I I le 12 H $ 11 E a a a a I IIT ITC) 172 l
l 1
1 1
EA-PT-91-0003-S Figure 7.67 Comparison of RETRAN and GE FWFAIL - MOD - Vessel Level (inches)
C4WS T UR % 7 IFAll G i i i i i i i i i i i i i i g g GE
\
3 s
3 a W Gs0
, !?
!?
M
=
J J
s .$
$R J l '
J E W W Ifl W W -
l Yh W r
[
a r W
t j 3 3 I I I l l I I I 1 I t ! t ]
O d I $ I I E 12 li Il 11 20 2 $ 3 3 }
M (SEC) 173
EA-PT-91-0003-S Figure 7.68 Comparisch of RETRAN and GE FWFAIL - HOD - Vessel Steam Flow (ik)
CX'Ti2 i TIR E 7 if All - G I I 3 I I !
I l l l I I I I I
- f. ,
t i
4 (A t i
- GSU 98
" 3 ._. e . _ _:
- e-i 2
e ,
a E W j G F 4 u _M i cI J I W
J O W @
0 W O >
W 8 E W
E i , , , , , , , , ,
3 3 3 3 3 I I I N tl !!
o o ! 1 E 12 Ilf ITC) 174
. , . - , . .- -1
I EA-PT-91-0003-S l
I l
Figure 7.69 Comparison of RETRAN and GE FWFAIL - MOD - Total l Reactivity (S) 1 e
e OfW M EU M ffAll-2 9' i i I I I I I I i i i ! I I i
. a 0
i 6
~
l 9 't 9 v .
F >
M M
> F
~ 0 F E O W ^ - ^
j
,a 29 go r j y .__ c t J F U
g 0
f cst-cm F3 29 .
g' ji 0 F
- W E
l
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k EA-PT-91-0003-S Figurc 7.70 Typical Power-Dopendent MCPR Limit i.M -- __
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184-
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EA-PT-91-0003-S Figure 7.72 MSIV Closure Overpressur Event - Bottom of Reactor Vessel (psig)
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EA-PT-91-0003-S 8.0 COMPARISON TO RIVER BEND STATION TRANSIENTS 8.1 TURBINE RUNBACK EVENT RBS experienced a turbine runbac) followed by a scrcm on December 26, 1986. This event was caused by a less of stator cooling while ope rating at near rated conditions. The initial conditions for this transient are shown in Table 8.1. .The timing of major events is shown in Table 8.2. RETRAN predictions are compared with ERIS data in Figures 8.1 through 8.8.
With the reactor operating at 99 percent power, a loss of atator cooling due to a failed temperature control valve caused a 5.9 percent reduction of the turbine load reference in 2.7 seconds. This was followed at approximately 19-second intervals by further reductions of the same magnitude. The second reduction was followed immediately by a partial closing of the turbine control valves (TCV) and a partial opening of the bypass valves (BPV). The third reduction caused an additional partial closure of the TCVs, which caused the BPVs to open fully. Dome pressure began to rise slowly. The fourth-reduction resulted in 179
EA-PT-91-0003-S a further partial closure of the TCVs. With full bypass capacity j
already being used, steam flow began to <1ccline and pressure began to rise more rapidly, as did power.
Approximately 60 seconds after the transient event began, a ,
high-flux scram occurred. Power and pressure dropped rapidly, followed by closure of the BPVs and then-the TCVs. Most other parameters remained approximately constant until the rapid rise in pressure and subsequent scram on high flux. Following the scram, there was a large decline in water level and first
]
fluctuations, and then a decline in core and f.adwater f'.ow..
Water leval recovered at about 80 seconds. '
Figure 8.1 shows a comparison of the-average power range monitcr (APRM) flux response to the RETRAN calculation.- RETTtAN conservatively overpredicts tne flux rise. The scram time and subsequent flux response showed. excellent agreement. Figures 8.2 and 8.3 show comparisons of the IUEPRAN calculated TCV and BPV ;
positions with the ERIS' data. Comparisons of the resulting dome pressure and steam flows are shown in Figures 8.4 and 8.5. The 180 I . _ _
i
l:
I EA-PT-91-0003-S initial decline in calculated steam flow was not as rapid as i indicated in the ERIS data, probably due to the larger calculated power (flux) rise and larger calculated dome pressure rise. And -
although the RETRAN calculation sitowed a final steam flow of zero, ERIS showed a final flow of 3 a percent. This was due to the lack of certain plant design features in the RETRAN model (steam jet air ejectors) and accounts for the difference ir. dome pressure after 70 seconds.
Calculated and measured narrow range (NR) water levels are shown in Figure 8.6. Good agreement is seen until the RETRAN-calculated steam flow and dome pressure! began _to depart from those in the ERIS data. Similarly, good agreement is seen in the feedwater flow-comparitsa, shown in Figure 8.7, until the RETRAN-calculated water level departs from that in the ERIS data.
Finally, Figure 8.8 shows a comparison of core flow. The RETRAN calculation shows a conservatively _ smaller rise just after the scram and more rapid subsequent' decline.
181
EA-PT-91-9003-S Simulation of this RBS transient event showed that the GSU RETRAN model accurately predicts this type of slow beginning, and also the fast-changing transient with some conservatism where needed, specifically in the pressure, power, and core flow at the time the MCPR occurs.
B.2 PRESSURE INCREASE TRANSIENT RBS experienced a pressure transient on April 30, 1988. The event was caused by the pressure regulator when control was transferred from Channel A to r.hannel B. No scram resulced from this event. The initial conditione for the event are shown in 3 Table 8.3. RETRAN predictions compared with measured data are shown in Figures 8.9 through 8.13.
Wit.t the re.4ctor operating at 100 percent power and 9-percent core flow, the electro-hydraulic control (EHC) system transferred control fror. Channel A to B. At the time of the transfer, the demand signal generated by B was ~20 percent lower than that generated by A. This caused the control valves to 182 l
EA-PT-91-0003-S reduce flow area until the system stabilized. In response to the area reduction, the reactor pressu.s cnd neutron flux increased.
The pressure and flux were restored to their initial values by the system.
The EHC system is primarily responsible for modulating the 4 turbine control valves (TCV) to maintain pressure in the vessel, and to regulate steam flow to the turbines. The portion of the system that regulates pressure-(pressure regulator) is comprised of two channels. One of these channels controls the pressure, while the other serves as a backup. The system has a
" supervisory" function, which monitors the system, and will initiate a transfer on conditions that indicate a failure in the controlling channel. To initiate the transfer, two conditions must be satisfied. The first condition is satisfied when the absolute difference between the two channels exceeds a set value of approximately 17 percent. The second' condition is an 1-electronically lagged average controller output determining which channel-is'the farthest away from- he average value. The channel with the largest difference is assumed to be the bad channel.
183 f.
EA-PT-91-0003-S The B channel hcd been trending downward over a period of time prior to the event. When the absolute difference condition was met, Channel A was further from the lagged average than Channel B. This caused the transfer of control from Channel A to Channel B.
To simulate the event, the pressure regulator control system described in " River Bend Station Plant Transient Analysis Methodology"' was modified to include the two channels, with a trip included to initiate the transfer.
Following the transfer, the valves beoan to close in response to the lowe.r demand signal. At the same time, pressure began to increase, which in turn increased the demand signal.
Neutron flux peaked at ~116 percent of rated in ~2 second',, and pressure peaked to 1034 psig in -A seconds.- The control system functioned to restore pressure and flux to their initial values.
Other parameters such as water level and core flow wera not-significantly affected.
184
EA-PT 91-0003-S Figure 8.9 shows the average TCV position versus time during the event. The RETRAN results agree reasonably well with the measured data. Figure 8.10 shows :he APRM response for the event. The results also agree reasonably well. The data indicated some sort of ringing that i likely due to modulation '
in one of the control systen.s. Figure 8.11 shows the pressure response, which again agrees well with data. Finally, Figures 8.12'and 8.13 show the narrow range level and core f1cw responses, respectively. As stated earlier, there is little impact on these parameters from the event. \
3
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185
EA-PT-91-0003-S Table 8.1 Initial Conditions, RBS Turbine Runback Transient (Scram 86-12)
Plant Data RETRAH 1
Power 99.0 % 99.0 %
Core Flow 97.8 % 97.8 %
Feodwater Flow 99.9 % 99.9 %
Dome Pressure 1018.6 psia 1018.6 psia Water Level 36.9 in. 36.9 in.
e
/
186 y
y w ' .-
EA-PT-91-0003-S B
Table 8.2 Sequence of Events, B (Scram 86-12) 3 Turbine Runback Transient Plant Data RETRAN Ls_erqodrd IF_eco. nds) 5.9 % load reduction 0.0 0.0 6.1 % load reduction 18.6 18.6 6.1 % load reduction 37.5 37.5 6.0 % load reduction 56.2 56.2
,j Scram
....; 59.8 60.1
_b BPV closure 66.9 67.5 TCV closure 67.5 67.5 6.0 % load reduction 75.1 75.1 l
l m
l l
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EA-PT-91-0003-C Table 8.3 Initial Conditions, RBS Pressuce Increase Transient '
Plant Data BETRAN l
Power 100.0 %
100.0 %
Core Flow 96.0 % 96.0 %
Feedwater Flow 99.9 %. 99.9 %
Dome Pressure 1025 psia 1G25 psia Water Level 36.5 in. 36.5 in.
188 0
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EA-PT-91-0003-S i
Figure 3.1 River Band Turbine Runback'- Scram 86-12 - ERIS Core Average Power RIVER BEND TURBINE RUNBACK SCRAM 86-12 R .i i . , , i S
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Figure R.2 River Bend Turbine Runback - Scram 86-12 -
ERIS TCV Position RIVER BEND TURBINE RUNBRCK SCRAM 86-12 I i i I i 7
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ERIS BPV Position RIVER BEND TURBINE RUNBACK SCRRM 86-12 i i l 9 i i I
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EA-PT-91-0003-S Figure 8.4 ' liver Bend Turbine Runback - Scram 86 ERIB Dome Pressure RIVER BEND TUR8INE RUNBACK SCRAM 86-12 f-- i i i i .
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EA-PT-91-0003-S Figure 8.5 River Bend Turbine Runback - Scram 86-12 -
ERIS Steam Flow RIVER BEND TURBINE RUNBRCK SCRAM 86-12 e
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h EA-PT-91-0003-S Figure 8.6 River Bend Turbine Runback - Scram 86-12 -
ERIS NR ,
Water Lovel RIVER BEND TURBINE RUNBRCK SCRAM 86-12 )
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EA-PT-91-0003-S Figur'2 8.7 River Bend Turbine Rur.back - Scram 86-12 - ERIS Feedwater Flow RIVER BEND TURBINE RUNBACK SCRAM 86-12 h
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-195 b
e EA-PT-91-0003-S Figure 8.8 River Bend Turbine Runback - Scram 86-12 -
ERIS Core Flow a
RIVER BEND TURBINE RUNBACK SCRAM 86-12 I i i i I
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EA-PT-91-0003-S Figure S.9 RETRAN Analysis of Pressure Regulator Evsnt - ERIS Average TCV Position (%)
RETRAN 1NALYSIS OF-PRES REGULATOR FVENT R~
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EA-PT-91-0003-S f
Figure 8.10 RETRAN Analysis of Pressure. Regulator-Event - ERIS Neutron Flux (%)
RETRAN RNALYSIs OF PRES REGUL9 TOR EVENT R- R i i i i i i
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EA-PT-91-0003-S B
g Figure 8.11 RETRAN Analysis of Pressure Regulator Event - ERIS Reactor Pressure (psig) c RETRAN RNRLYSI:3 0F PRES REGULRTOR EVENT o o ~
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EA-PT-91-0003-S i
Figure 8.12 RETRAN-Analysis of Pressure Regulator Event - ERIS Narrow Range Level (inches)
RETRAN ANALYSIS OF PRES REGULATOR EVENT i i i g g e i m m en u, ,
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E1.-PT-91-0003-S Fjgure 8.13 RETRAN Analysis of Pressure Regula.or 6 Event - ERIS Total Core Flow (MLB/hr)
RETRAN ANALYSIS OF PRES REGULRTOR EVENT i i g- g i i i m
8~
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EA-PT-91-0003-S 9.0 REVISED PEACH BOTTOM UNIT 2 TRANSIENT COMPARISONS f
After the submittal of the original topical report , a 3 small change in the GSU core physics procedures necessitated the regeneration of the reactor kinetics data used for the Peach Bottom turbine t rip transients. To assess any impact these changes might have. the transient calculations were redone. The resulting power curves (Figures 9.1, 9.2, and 9.3) were compared as before with published data. In each case, the peak power l showed a small increase over the previous calculation. The other system plots included in Rev. O were virtually unchanged and are not included here.
l l
A comparison of CPR calculations by GSU (using GEXL-Plus) with calculations by GE (using the original GEXL correlation) for l the Peach Bottom 2 turbine trip tests are shown on Table 9.1.
l The table shows the initial power / flow conditions,-the initial CPR (ICPR), the ratio of measured ACPR to ICPR (RCPR-M), the calculated RCPR (RCPR-C) and the error in RCPR-C (ARCPR). The
" measured" RCPR-M was determined by using measured boundary 202 i
I
1 I
T -PT-91-0003-S conditions driving a hot channel calculation of ACPR; the
" calculated" RCPR-C used boundary conditions from the GSU RETRAN calculation. The CPR uncertainty calculations discussed in Section 6.0 use the latest core physics and transient calculations.
4 A
(
203 I
EA-PT-91-0003-S Table 9.1 Peach Bottom 2 EOC2 Turbine Trip ACPR Results PROPRIETARY INFORMATION DELETED 204 m.h. A ea -
EA-PT-91-0003-S Figure 9.1 Revised Predicted vs. Measured Core Average Power, Peach Bottom Test TT1 PEACH BOTTOM 2 EOC2 TTl 9
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e 9
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cb.00 .80 1.50 2.40 3.20 '9.00 TIME (S) 205 i
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EA-PT-91-0003-S Figure 9.2 Revised Predicted vs. Measured Core Average Power, Peach nottom Test TT2 PEACH BOTTOM 2 EOC2 TT2 g ,
r-e
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.80 1.60 2.90 3.20 9.00 cb.00 TIME (S) i l
I 206 ls l
EA-PT-91-0003-S Figure 9.3 Revised Predicted vs. Measured Core Average Power, Peach Bottom Test TT3 PEACH BOTTOM 2 E0C2 TT3 O i i i i e-e a ,_ -
e Z
yo - -
om ,
CL 1) g MEASURED + M ALCULATED o
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i cb.00 .00 1.E0 2.40 3.20 4.00 TIME (S) 1 207
l i
EA-PT-91-0003-S
10.0 REFERENCES
- 1. " River Bend Station Plant Transient Analysis Methodology," EA-PT-91-0003-M, April 1991, transmitted to the NRC by RBG-34939 dated May 2, 1991.
- 2. " Steady State Core Physics Methods for BWR Design and Analysis," EA-CA-91-0001-M, January 1991.
- 3. EPRI NP-1850-CCM-A, "RETRAN-02 -
A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," June 1987.
- 4. A. F. Ansari, et. al., "FIBWR: A Steady-State Core j Flow Distribution Code for Boiling - Water l
Reactors," EPRI NP-1923, Electric Power Research Institute (1987).
- 5. Letter from J. S. Charnley (GE) to C. O. . Thomas (NRC) " Amendment 15 to General Electric Standard i Application for Reactor Fuel (GESTAR II)", January 1
23, 1986.
- 6. " General Electric Thermal Analysis Basis; Data, Correlation, and Design Application," NEDO-10958A, General Electric Company (1977).
- 7. River Bend Station Updated Safety Analysis Report.
208 l
EA-PT-9:-0303-S
- 8. Appendix A to Operating License NPF-47, "Tec' 1 Specifications River Bend Station."
- 9. " Supplemental Reload Licensing Submittal for River Bend Station Reload 2, Cycle 3," GE 23A5934, Rev. O, October 1988.
- 10. Letter, C. O. Thomas (NRC) to J. S. Charnley (GE),
" Acceptance for Referencing of Licensing Topical Report MEDE-24011-P-A, Rev. 6, Amendment 11, General Electric Standard Application for Reactor Fuel," November 5, 1985.
- 11. Letter, G. C. Lainas (NRC) to J. S. Charnley (GE),
" Acceptance for Referencing of Licensing Topical-Report,"- NEDE-24011-P-A, "GE Generic Licensing Reload Report," Supplement to Amendment 11, March 22, 1986.
- 12. Letter, J. S. Charnley (GE) to M. W. Hodges (NRC) ,
" GEMINI ODYN Adjustment Factors for BWR/6,"
July 6, 1987. ,
- 13. "GESSAR II: Nuclear Island Design," GE 22A7007 Rev. 6, Appendix 15B.
- 14. "ASME Boiler and Pressure Vessel Code," American Society of Mechanical Engineers, 1971 Edition, Summer 1973 Addenda.
l 209 l
l
.l l
i