Text

Transient Analysis Methodology
	ML20077A461
Person / Time
Site:	River Bend
Issue date:	04/30/1991
From:	Jacqueline Thompson; GULF STATES UTILITIES CO.
To:	;
Shared Package
ML20077A459	List: ML20077A461; RBG-34939, Forwards EA-PT-91-0003-M, River Bend Station Plant Transient Analysis Methodology, for Review & Approval;
References
	EA-PT-91-0003-M, EA-PT-91-3-M, NUDOCS 9105100059
	Download: ML20077A461 (162)
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RlVER BEND STATION PLANT TRANSIENT ANALYSIS METHODOLOGY

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RIVER BEND STATION PLANT TRANSIENT ANALYSIS METIIODOLOGY April 1991 Principal Encineer Thomas W. Oliphant Contributors John P. Egan Lynn A. Leatherwood Stone S. Luo David R. Swope Reviewed: !

Jamds L. Thompson '

Supervisor - Thermal-Hydraulic Analysis Approved: h/,

Jo e h $ Miller b

D re tor - Engineering Analysis Gulf States Utilities Company River Bend Station P.O. Box 220 St. Francisville, LA 70775 O ,

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- IMPORTANT NOTICE tEGARDING CONTENTS OF THIS DOCUMINT PLEASE READ CAREFULLY Thie document wee propwed by Gulf States Utbee Company for the use of the U.S. Nuclew e Regulatory Commeseson in mettwo regwdeng the operstog Econes for the Rever Cend Station, To the best of the issuer's knowledge, thae document containe work performed in accordence with sound engeneerno practice ed is a true and occurote representation of the f acto.-
. The work reported hereen is the property of Gulf Steise Utaties Cornpeny, and any usage other then -
~g as described above se prehetweed. Othw then for the intended usage, nanthw Gulf States Utation i N Company, nor any of its e.wloyees or officere, nor any othw person acteg on its behell;
~o Makes any warranty or representation, exprese or irWed, with respect to the occuracy, templetenees or usefulnese of the
information contemed in this report, or that the use of any informetion, apperstue. method. or procese diecioned herein would ',
not ininnge pnvetely owned nghte; or !
O Aseumes any GebiGties with roepect to the use of. or for damagee
l. _.reeutting froen the me of, any information, apperatue, method. or
procese disclosed in this report.
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l Acknowledgements j
The authors. gratefully acknowledge the assistance of Dr.
D.A. -Prelewicz, B.J. Gitnick, ._and G.B. Peeler of i
.ScIENTECH, Inc. , Dr. A. Ancona' of Ancona &. Associates, and j
J.C. Chandler of John Elston Associates'in the completion .i
-of_this. project and the preparation of this report, ,
The authors acknowledge the contribution of Gulf States LUtilities _ (GSU). clerical- employees Jerri Fontenot and Nancy _ Scott and of.. student engineers. Kathy L.- Hugle of= l O . the- University of ~ Louisville and ' Eric- Ballon of the University of Florida.
y In addition,-the authors-acknowledge the support provided-by_the Core Analysis Group,.who provided-' input-data for
.ESCORE and the.one-dimensional reactor. kinetics used by RETRAN_.
The-authors gratefully._ acknowledge the'long-term finan-cial support-from Gulf. States Utilities; in particular,.
Mr.1 Jim Deddens,:Mr.-Ken-Suhrke, and Mr.?Mel Sankovich.-
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Table of1 Contents Acknowledgements . . . . . . . . . . . . . . . . iii
. Table of Contents . . . . . . . . . . . . . . . . iv List of Tables . . .. . . . . . . . . . . . . . vi
-List of Figures . . . . . . . .. . . . . . . . . vil
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . 1 1.1 '
SUMMARY
. . . . . . . . . . . . . . . . 1 1.2 METHODOLOGY APPLICATION . . . . . . . . . 3
1.3 CONCLUSION
S . . . . . . . . . . . . .. . 4
2.0 DESCRIPTION
OF COMPUTER PROGRAMS . . .. . .. 6 2.1 THE~FIBWR STEADY-STATE CORE THERMAL-HYDRAULIC CODE- . . . . . . . . . . . . .- 6 2.2 THE SIMTRAN-E CROSS SECTION CODE , . - . . 7 2.3 THE ESCORE FUEL ROD PERFORMANCE CODE . . 8 ..
'2.4 THE REBAL INITIALIZATION CODE . . . . . . 9
2 ~ 5 - THE RETRAN-02 ONE-DIMENSIONAL TRANSIENT CODE . . . . . . . . . . . . . . . .. 11 2.6 THE EDTRAN OUTPUT EDITOR , . .. . . . 12
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13.0- RBS-SYSTEM MODEL . . . . . . . . .. . . . . 15 3.1 ' SYSTEM NODALIZATION . . . . . . . . . . 1. 5 3.l.1 - Vessel Internals . . . . . . 16
'3.1.2 Core Reaion-. . . . . - . . . . 17 3.1.3 Recirculation Looos . . . . . 18 3.1.4 Main Steam Linga . . . . . . 19 3 1.5 Steam Bvoass Lines - . . . . . ,
20 3.1;6 Safetv/ Relief ValvenL-, . . .
20
'3.2 HEAT DEPOSITION . . . . ... .. . . . 21 3.2.1 -Powered Qqnductors=. . . . . - . = 21 -
3.2.2- Non-Powered Conductors . .. 22 3.2.3 Non-Conductina Heat Exchanaers . .. .: .- .. . . 22:
POWER GENERATION 22 3 ~. 3 . .. . . . . . . . .
3.4L. TRIP LOGIC. . . . . . . -. -. . . . . . . 23
3.5 CONTROL
SYSTEMS .. . . . . . . . . . . 24
'4.0 CROSS SECTION' GENERATION . . . .. ... . . . 47
'5.O CALCULATION OF FUEL' ROD GAP CONDUCTANCE .. 50.
6.0 COMPARISON TO RIVER BEND STATION TRANSIENTS 53 6.1 GENERATOR LOAD REJECTION . - . . . . . . 53 iv
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612- WATER-LEVEL INCREASE EVENT' . . .r. .. . -58 6.3 = WATER LEVEL SETPOINT~ STEP CHANGES .-.'. -- 62 7.0- , COMPARISON;TO PEACH-BOTTOM UNIT.2 TRANSIENTS 85 "7.1 - BENCHMARK DESCRIPTION ... . . . . . . . . 85 7.2- PEACH BOTTOM RETRAN MODEL .-. . . . . . - . 87 7.3 TURBINE TRIP TEST SIMULATION- . -
. . .-- . 88 I 7.3.1 Initial Conditions . . . -. .. 88
- 7.3.2 Transient Modelino- . . - . . . .89
7. 3. 3' - Simulation Results' . . . . - . 91 --'
7.4 LICENSE BASIS TRANSIENT MODELING . . . . 93 7.4.~1: Initial Conditions .- . . . - . _ 93 7'.4.2 Analytical Results .. .. . . . 94 .i 8.0= HOT-CHANNEL MODEL . . . . . - . . . . . . - . . . . .
124-REFERENCES . 129
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'I A CALCULATION OF= FUEL = ROD GAP-CONDUCTANCE .. . 131 B ACRONYMS USED IN THE TEXT . . . . . . . . . 149 ..
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List of Tables Egg Title Page 1.1 Design Features . . . . . . . . . . . . . . . 5 3.1 Control Volume Geometric Information . . . . 25 3.2 Junction Geometric Information . . . . . . . 27 3.3 Heat Conductor Information . . . . . . . . . 30 3.4 RETRAN Core Neutronic Region Information . . 31 3.5 RBS Trip Control Data . . . . . . . . . . . 32 6.1 Initial Conditions, RBS Load Rejection Tran-sient (Scram 8904) . . . . . . . . . . . . . 63 6.2 Sequence of Events, RBS Load Rejection Tran-sient (Scram 6304) . . . . . . . . . . . . . 64 6.3 Initial Conditions, RBS Water Level Increase Event . . . . . . . . . . . . . . . . . . . 65 6.4 Sequence of Events, RBS Water Level Increase Event . . . . . . . . . . . . . . . . . . . 66 7.1 Peach Bottom Turbine Trip Test Initial Conditions . . . . . . . . . . . . . . . . . 97 7.2 Peach Bottom License Basis Transient Initial' .
Conditions . . . . . . . . . . . . . . . . . 98 m
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(3 (j List of Figures No. Title Page 2.1 Transient Analysis Calculational Flow . . . 14 3.1 RBS System Nodalization . . . . . . . . . . 36 3.2 RBS Instrumentation Control System . . . . 37 3.3 RBS Water Level Control System . . . . . . 38 3.4 RBS Recirculation Flow Control System . . . 40 3.5 RBS Turbine Control System . . . . . . . . 42 3.6 RBS Feedwater Control System . . . . . . . 44 3.7 RBS Miscellaneous Edits Control System . . 45 6.1 Predicted vs Measured Neutron Flux, RBS Scram 89-04 . . . . . . . . . . . . . . . . 67 6.2 Predicted vs Measured Reactor Pressure, RBS Scram 89-04 . . . . . . . . . . . . . . . . 68 6.3 Predicted vs Measured Total Steam Flow, RBS Scram 89-0 . . . . . . . . . . . . . . . . 69 6.4 Predicted vs Measured Total Core Flow, RBS Scram 89-04 . . . . . . . . . . . . . . . . 70 '
6.5 Predicted vs Measured Feedwater Flow, RBS Scram 89-04 . . . . . . . . . . . . . . . . 71 6.6 Predicted vs Measured Vessel Water Level, fg RBS Scram 89-04 . . . . . . . . . . . . . . 72 6.7 Predicted vs Measured Recirculation Pump
() Speed, RBS Scram 89-04 . . . . . . . . . . 73 6.8 Predicted vs Measured Bypass Valve Posi-tion, RBS Scram 69-04 . . . . . . . . . . . 74 6.9 Predicted vs Measured Neutron Flux, RBS Water Level Increase Transient . . . . . . 75 6.10 Predicted vs Measured Narrow Range Water Level, RBS Water Level Increase Transient . 76 6.11 Predicted vs Measured Wide Range Water Level, RBS Water Level Increase Transient . 77 6.12 Predicted vs Measured Core Flow, RBS Water Level Increase Transient . . . . . . . . . 78 6.13 Predicted vs Measured Feedwater Flow, RBS Water Level Increase Transient . . . . . . 79 6.14 Predicted vs Measured Turbine Steam Flow, RBS Water Level Increase Transient . . . . 80 6.15 Predicted vs Measured Water Level, RBS Water Level Setpoint Change (+6") . . . . . 81 6 ," Predicted vs Measured Feedwater Flow, RBS
'4ater Level Setpoint Change (+6") . . . . . 82
, a Predicted vs Measured Water Level, RBS Water Level Setpoint Change (-6") . . . . . 83 6.18 Predicted vs Measured Feedwater Flow, R3S Water Level Setpoint Change (-6") . . . . . 84
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List of-Figures-(Continued)_
'ESA Title Eggg 7.1 Peach. Bottom RETRAN Model Nodalization . . 99 7.2 . Predicted vs Measured Core Power, Peach Bottom Test TT1 . . . . . . . . . . . . . . 100
'7. 3 Predicted vs Measured Upper Plenum Pressure, Peach Bottom Test TT1 . . . . . . 101 7.4- Predicted vs Measured Dome Pressure, Peach Bottom Test TT1 . . . . . .. . . . . . . . 102 7.5 Predicted vs Measured TSV Pressure, Peach Bottom Test TT1 . - . . . . . . . . . . .. . 103 7.6 Predicted vs' Measured Core Inlet Flow, A Peach Bottom Test TT1 . . . . . . . . . ..
104
'7. 7. Predicted vs Measured Core Average Power,-
Peach Bottom Test'TT2 . . . . . . -. . . . . 105 :
7.8 Predicted vs Measured Upper Plenum Pressure, Peach Bottom Test TT2 .. . . . 106 7.9 Predicted vs= Measured Dome Pressure, Peach
. Bottom Test TT2 . . . . . . . . . . . . . .- 107 .
7.10 Predicted vs-Measured TSV Pressure, Peach '
Bottom Test TT2 . . .. . . . . . . . . . . 108-7.11' Predicted vs Measured Core Inlet Flow, l
, a. Peach Bottom Test TT2-. . . . . . . .. . . 109 l~ { .
7.12 ' Predicted vs, Measured Core Average Power, Peach Bottom-Test TT3 .. . . . .. . . . . . 110 7.13' Predicted vs Measured Upper Plenum Pressure, Peach Bottom Test TT3 . . . . . . 111 7.14 Predicted vs Mensured Dome = Pressure, Peach Bottom Test TT3 . . . .. . . . . . . . . . . 112 7.15 Predicted vs Measured TSV Pressure,--Peach
' Bottom Test TT3 . . . . . . .. . . . . . . . 113 7.16 Predicted-vs-Measured' Core Inlet Flow, Peach Bottom Test TT3 .. . . . . .. . . . 114 17.17- Axial PowerLDistribution, Peach Bottom License Basis Transient'. .. . . . - . . . . 115
'7.18 Fuel Temperature Distribution, Peach Bottom; License Basis Transient .. .. . . . . . . . 116 7.19. Initial. Heat. Flux Distribution, Peach Bottom' License Basis. Transient
. .. . .. 117 7.20 Initial. Void Distribution, Peach Bottom
= License Basis Transient . . . . .- . . . . . 118 Core Average: Power, Peach Bottom-License
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7.21 Basis Transient . . . . .. . . . . . . . . . . . 119 7.22- Upper-Plenum Pressure,-Peach Bottom License Basis Transient . . . .. . . . .. . . .. 120 7.23 Core Inlet Flow, Peach Bottom License Basis-3 Transient . . - . . . . . . . . . . .. . . . 121
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EA-PT-91-0003-M List of Figures (Continued) h Title Pm 7.24' Heat Flux Distribution 9 t = 0.8, Peach Bottom License Basis Transient . . . . . . 122 7.25 Heat Flux Distribution @ t = 1.2, Peach Bottom License Basis Transient . . . . . . 123 8.1 RBS RETRAN Hot Channel Nodalization . . . . 128 l
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'1.0.. INTRODUCTION This report describes the analytical- tools and i'
models to be used by Gulf-States Utilities (GSU) for performing transient thermal-hydraulic analyses in support of River Bend Station- (RBS)'. Methods for steady-state core physics analysis have been submitted previous-l y ' '.
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.RBS is a boiling water reactor plant located in West Feliciana- Parish, Louisiana.- The plant.has !
a Boiling Water Reactor /6 (BWR) Nuclear Steam Supply' System _(NSSS)-
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.' i h designed by General Electric (GE), who currently provides .
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Le nuclear fuel and . analytical services. The architect-li Q j- engineer' functions were performed by Stone . & Webster Engineering . Corporation.- Design features'of the plant are shown in Table 1.1.
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1.1: .
SUMMARY
T h i s .- r e p o r t -- d e s c r i b e s : t h e . c o m p u t e r p r o g r a m s and finput's ' used to . perform analyses of rapid core wide 1 Anticipated Operational Occurrences-(AOos) and the design.
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basis overpressurization event. These tools will be used -
- in-analyses of AOos and the overpressurization event in supphrt of reload licensing and plant operational changes such as- increased core. flow. Analyses of loss of feedwater heating,_and control rod withdrawal error are -
performed using steady state methods previously submit-i ted'. Analysis of. loss of coolant accidents and local events such as control rod drop accidents are not
-included--in this report. ,
These -methods are qualified through benchmark --!
analyses of plant. data for RBS and Peach Bottom. ,
_ The computer. codes used in the transient analysis
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are-described'in Section 2.0. This'section_provides a
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p brief.' description.of.the calculations, performed by the.
f key _ software = tools'used in the GSU methodology.
l The RBS RETRAN system model is described in Section
. 3.0. LBasic characteristics of the RETRAN r.odeling-and-
- description of.the_ major. input: segments.are provided.
Supporting analyses covering nuclear cross sections i
and. fuel rod gap conductance:are; described-in Sections
. 4. 0. and -5. 0. The results of t.hese analyses have a strong-impact. on._the. end result. of the . thermal-hydraulic
. transient analysis.-
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t] .- Sections 6.0 and 7.0 report benchmark analyses l
against data from RBS plant data and the Peach Bottom l turbine trip tests. Benchmark analyses cover several RBS operational events, the Peach Bottom turbine trip tests, and a postulated License Basis Transient based on the Peach Bottom tests. These benchmarks demonstrate the capabilities of RETRAN and the GSU methodology to predict a broad spectrum of transient events.
The RETRAN hot channel model is described in Section 8.0. This model is used to predict the performance of potentially limiting assemblies in the_ core from system .
effects predicted by the system model.
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1.2 METHODOLOGY APPLICATION i
The methods described in this report will be used to meet licensing requirements for RBS. Additional informa-tion qualifying this application will be provided sepa-
! rately. The separate submittal will describe the methods
used in determining an operating limit critical power 1
ratio, calculation of uncertainty factors for the thermal l-
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V margin analysis,- and applications of the GSU methodology.
1.3 CONCLUSION
S The RBS RETRAN model provides acceptable representa-tion of plant behavior during transient events within the capabilities of the coding. System phenomena are predicted consistently with measured data by the system model..
Based on the results of the benchmark analysen, it ,
is concluded that the use of the RBS RETRAN model and fg associated code packages is acceptable for analyzing Aoos i s and- the overpressurization event for reload safety analysis, licensing, and plant operational support activities.
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1 Table 1.1 Design Features 1 1
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 Method Valve Flow Control Number of Jet Pumps 20 Number'of Recirculation Pumps 2 Number of Safety / Relief Valves 16 ,.
Number of Control Rods 145
-rg Number of Fuel Bundles 624 L);
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2.0 DESCRIPTION
OF COMPUTER PROGRAMS The computer programs used at GSU for reactor tran-sient analysis are described in this section. Figure 2.1 I shows how the programs are linked together.
2.1 THE FIBWR STEADY-STATE CORE THERMAL-HYDRAULIC CODE FIBWR a evaluates the steady-state thermal-hydrau-2 lic performance of BWR cores. FIBWR calculates the flow, ;
void, and pressurc distributions for multiple, parallel channels within the core by solving the one-dimensional equations for continuity, momentum, and energy. FIBWR was developed by Yankee Atomic Electric Company under a project sponsored by the Electric Power Research Insti-tute (EPRI).
The GSU-methodology uses FIBWR to develop initial l'
! core pressure and flow distributions for use in the RETRAN System Model and to develop a response surface to correlate hot bundle flow and core flow for hot channel l^
L analyses. FIBWR models have been developed for analyses
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k.J of Peach Bottom and RBS reported later in this report.
Both models have been benchmarked to plant data.
2.2 THE SIMTRAN-E CROSS SECTION CODE SIMTRAN-E' ("SIMTRAN") generates a one-dimensional cross sections file for use with the space-time kinetics option in RETRAN-02. SIMTRAN extracts nuclear data from restart files written by the three-dimensional nodal simulator code, SIMULATE-E. The nuclear data include ,-
macroscopic cross sections and pseudo-cross sections covering delayed neutron fractions and generation
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V factors. Use of SIMULATE-E for core analysis is de-scribed in a separate report.
SIMTRAN reads three-dimensional, two-group partial cross sections from SIMULATE-E restart files for enough control rod states to characterize the RETRAN problem to be- analyzed . Each SIMTRAN analysis requires a SIMULATE-E case with nominal control rod positions for the beginning of the problem. Whenever control rod movement is expected, additional SIMULATE-E cases with all control rods partially and fully inserted into the core are used,
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y i. SIMTRAN perturbs the SIMULATE statepoints according to user input to calculate the dependence of each cross section and pseudo-cross section on fuel temperature and moderator density. It then collapses the nuclear data from the SIMULATE three-dimensional representation to a one-dimensional axial cross section set using flux adjoint weighting. SIMTRAN spatially averages the fuel temperatures and moderator densities, correlating the SIMULATE tabular cross sections into polynomial functionJ of these independent variables at each axial node. These polynomials are of the form: .
.Dg = CF11 + CF2ndU + CF21dU3 + CF12dTg + CF22dUdTg + CF22dT gdU 2
. %iQ) where Ei = the fitted macroscopic cross section at node i; CF = coefficients calculated by SIMTRAN; l dU = the change in relative nodal modera-tor density; and
= the change in the square root of the dTr nodal fuel temperature.
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(_,)/ 2.3 THE EbCORL FUEL ROD PERFORMANCE CODE I
l ESCORE 8 calculater the steady-state, thermal-mechan-ical performance characteristics of a fuel rod. The J l
performance parameters calculated by ESCORE include fuel ;
pellet temperature distribution, fission gas release, gap conductance, internal pressure, cladding stress and l strain, power to melt, cladding corrosion, irradiation growth, pellet densification, and stored energy.
ESCORE is used in the GSU methodology to determine gap conductance values for use in the RETRAN system and ,
hot channel analysis.
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\J 2.4 THE REBAL INITIALIZATION CODE The- REBAL computer code' was developed to f acilitate the calculation of consistent thermal-hydraulic initial conditions for RETRAN models of BWR systems. REBAL is used to initialize the RBS RETRAN model for plant transients which are not initiated from rated power and flow conditions.
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Li REBAL solves the steady-state pressure loss and heat balance-equations for the reactor vessel, jet pump, and
. recirculation system, using an iterative procedure to solve for the reactor vessel water densities and-pres-sures at rated conditions. Subsequent cases retain calculated pressure loss coefficients which are held constant as reactor operating conditions vary. These coefficients include the effective separator dryer loss coefficient, the upper downcomer loss coefficient, and a l partial--loss coefficient for the non-nozzle portion-of
'the recirculation loop losses. ,
,REBAL determines the separator and jet pump charac-teristics consistently with known component performance L, curves and measured plant data over a broad range of operating conditions. The jet. pump M-N curve, core flow
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versus drive flow, and core and separator pressure drop
-versus flow relationships are preserved over a. wide range
.of power:and flow states.
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2.5 -THE RETRAN-02-ONE-DIMENSIONAL TRANSIENT CODE i
'RETRAN-027 ("RETRAN") is the transient thermal-hydraulic computer program used for analysis of A00s at GSU. The GSU methodology uses RETRAN in two different modes. RETRAN is exercised in the core average mode to determine system response to transient events and in the hot channel mode to evaluate the change in thermal margins. ,
'RETRAN was developed under EPRI sponsorship from 1
LRELAP4. RETRAN is a one-dimensional, homogenous equilib-rium mixture _ thermal-hydraulic model with provisions for [
W slip between the vapor and liquid phases. Power genera-Q-
tion may be accomplished through a user controlled table of power as a function of time, through salution.of the
-point -kinetics equation, or through a more rigorous solution of- the one-dimensional- neutron diffusion
i. equation. Other user-selectable options include a i momentum-mixing option for modeling_ of special components such as valves, separators, and jet pumps. RETRAN also L
lf -has a capability _ for extensive modeling of control L
systems.
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Devel'opment of RETRAN input is described in Section 3~.0 of_this report.
2.6 THE EDTRAN OUTPUT EDITOR EDTRAN' is a post-processor code developed by Public Service Electric & Gas Company of New Jersey. EDTRAN reads restart files generated by RETRAN and performs simple output manipulation _ functions. The. options l available in EDTRAN are EDIT, LINK, and TCPR. /
The EDIT-option extracts minor edit information from N the RETRAN restart file and prints it in a user-specified
[d~ format.
The _ LINK option links the RETRAN system and -hot channel models.- It~ extracts the core power, pressure, and flow as functions of time from the restart file and-
calculates the hot channel power and flow characteristics from user input. The hot channel flow is determined from either. a user

supplied . hot channel flow _ split as' a function of time, . or with a : user supplied response surface.-- EDTRAN prints-the results of this option in the format of RETRAN hot-channel model input deck.
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EA-PT-91-0003-M O The TCPR option extracts power, flow, pressure, and enthalpy, from the RETRAN hot channel restart file and uses' the GEXL-Plus critical power correlation' to calcu-late Critical Power Ration (CPR) for specific transients, O
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EA-PT-91-0003-M Figure 2.1 Transient Analysis Calculational Flow
( SIMTRAN-E v J SIM U L AT E-E RESTART FILE 1-D CROSS SECTIONS RETRAN02
_ PRESSURE D SYSTEM FIBWR _ BOUNDARY CONDITIONS b BUNDLE FLOW N EDTRAN LINK RESPONSE SURFACE HOT CHANNEL DECK RETRAN02 OT CHANNEL BUNDLE RESPONSE L o L
l EDTRAN j TCPR DCPR,- CPR VS. TIME l- 1 Revision 0 Page 14 L
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-C 3.O RBS SYSTEM MODELING l This section describes RETRAN conventions used in l
modeling the NSSS and connected systems for RBS. In 4
. preparing the input for RETRAN the authors familiarized ;
themselves with the program limitations as outlined in the Safety Evaluation Report for RETRAN'. The separate hot channel model is described in Section 8.0.
3.1 SYSTEM NODALIZATION c fm The RBS system model encompasses t!.e reactor core, recirculation loops, main steam lines, turbine control i
j valves, condenser bypass line, and safety / relief valves.
As-built geometric and performance data were the primary source for.RETRAN input, where such data were unavail-able, design information was used. System model nodal-ization is shown in Figure 3.1. Volume and junction data l
are given in Tables 3.1 and 3.2.
The system model nodalization was based on the results of a series of parametric studies.- Separate studies were performed for the core region, the separa-l-
[s Revision 0 Page 15
EA-PT-91-0003-M p
( -
tor / dryer /downcomer region, the steam lines, and the recirculation lines. The nodalization shown in Figure 3.1 is sufficiently detailed to be in agreement with the results of finer nodalization schemes 0 Other benchmark studies" also demonstrate agreement with plant data measurements of pressure, flow, and water level.
3.1.1 Vessel Internals Feedwater enters the vessel via a rill junction ,
located at the middle downcomer, which is consistent with l q the physical location of the feedwater sporger. Sub-( \
' ' ~ '
/
cooled feedwater mixes with saturated water returning from the steam separators and dryers through the upper l downcomer. The combined flow passes into the . lower downcomer. The recirculation loops draw from this L region, passing a portion of the flow through the recirculation pumps and returning it to the vessel as jet pump drive flow. The_ recombined fluid is driven through the jet pumps into the lower plenum, from which it proceeds upwards to the core. The flow is again split at the core inlet, where most of the flow enters the active
/,
t i Revision 0 Page 16
.(,/-
EA-PT-91-0003-M I
core. The remainder enters the bypass region. The active core flow cools the nuclear fuel.
These flow components again converge at the core exit and enter the upper plenum. The liquid-vapor mixture travels upwards through the standpipes to the steam separators, where most of the liquid is rejected to the upper downcomer. The steam dryers strip additional liquid from the flow, allowing predominantly dry steam to enter the steam dome. The last volume is a water annulus which connects the dome to the middle downcomer; this region corresponds to the fluid outside the dryer seal ;
skirt. The water annulus and separator are modeled using the bubble rise option. The upper down comer is modeled using the nonequilibrium values option.
3.1.2 Core Reqi2D The core consists of an active core region and a bypass region. Three volunies are used to model the bypass, with two junctions representing leakage flow paths around the lower tio plate and direct from the lower plenum. The active core region consists of 25 Revision 0 Page 17 1
EA-PT-91-0003-M i
axial volumes. The algebraic slip option is used to provide for differences in the velocities of the vapor and liquid phases as they proceed up the channel. The slip option is used in the core region only, and is defeated elsewhere in the model. The profile fit option is used to provide for s..bcooled boiling for the neutronics calculations. Power is provided to the active core region volumes through heat conductors simulating the fuel cladding. Power generation within the fuel is calculated by solution of the time-dependent diffusion equation in one axial dimension. Power deposition to the ,.
bypass region is accomplished through a non-conducting heat exchanger. Heat conductors and power generation are described in Sections 3.2 and 3.3, respectively.
3.1.3 Recirculation Loons ,
The two physical recirculation loops are combined as a single analytical loop using six fluid volumes and seven junctions. The analytical loop consists of volumes and junctions representing a recirculation pump (with associated pump curves) , a flow control valve, a dis-Revision 0 Page 18
EA-pT-91-0003-M
(
'v charge line, the jet pump manifold, the jet pump risers, and the jet pumps. The recirculation pump suction and jet pump driven flow inlet are connected to the lower downcomer, while the jet pump drive flow is fed into the lower plenum. Valve position control and pump power supply (High Frequency Motor Generator (HFMG), Low Frequency Motor Generator (LFMG) or none) are accom-plished by control systems. Manuf acturer's data for pump performance was used to develop homologous pump curves.
3.1.4 Main Steam Lines The four main steam lines are lumped into a single equivalent line using 11 fluid volumes and 12 junctions.
The length of the longest steam line volume is approxi-mately 50 feet. The line consists of an exit volume frot.a the reactor vessel, a safety relief valve volume, a flow restrictor, two main steam isolation valves, a condenser bypass line, a turbine stop valve, and a turbine control valve. Flow exits the steam line via a negative fill which is administered by a control system.
l l
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Revision 0 Page 19
.. . . . . . . . ,. - - - - - - - _ . = . - .
EA-PT-91-0003-M (7
(%) 3.1.5 Steam Bvoass Lines The bypass modeling consists of a bypass valve, line volumes on either side of the valve, and the condenser.
3.1.6 Safetv/ Relief Valves All sixteen safety / relief valves are modeled as relief valves in five groups based on their setpoints.
Some valves are modeled as low-low set valves. Low-low ,
set valves include additional logic which serves to open and close the valves at lower pressures than the normal 3
relief setpoints. This serves to limit Safety Relief Valve (SRV) cycling to the low-low set valves rather than all of the valves. Relief valves exhaust from the second steam line volume downstream from the vessel into a large volume representing containment, rO Revision-O Page 20 V
- = _ . . - - = _ -- _ . _ _
EA-PT-91-0003-M
,CT C 3.2 HEAT DEPOSITION The RETRAN options for heat deposition used in the RBS model are powered heat conductors, nor.-powered heat conductorn, and non-conducting heat exchangers. This section describes these heat deposition options. Heat conductor data are shown in Table 3.3.
3.2.1 Powered Conductors Powered heat conductors are used to represent the p fuel rods. Fuel rod information included in the powered
\J heat conductor input includes material and geometry data for the fuel pellet, diametral gap, and cladding. The flux solution from the neutr'Aic model determines the power source in these conductors, which transmit heat to the fluid in the active core volumes. Material proper-ties for the fuel and clad where obtained from standard reference sources. A specific gap heat transter coeff1-cient was developed using ESCORE as described in Section 5.
Revision 0 Page 21
EA-PT-91-0003-M O 3.2.2 Non-Powered Conductors Non-powered conductors are used for the core shroud, upper plenum, lower plenum, and standpipes. These structures allow storage of heat but contain no power source.
3.2.3 Non-Conductina Heat Exchanaers The non-conducting heat exchanger logic provides ,
direct heat input to a region. This option simulates the power deposition in the bypass region, which is assumed to be 2% of the total core power.
3.3 POWER GENERATION Transient power generation; is calculated by space-time kinetics and the multi-state control rod model. The space-time kinetics option solves the time-dependent neutron dif fusion equation in one (axial) dimension. The multi-state control rod model maintains the initial axial I Revision 0 Page 22 m -_________.m_..___mm____m.--
EA-PT-91-0003-M i
if -
O control rod distribution during a scram, moving all of the control, rods together until full insertion is reached. These RETRAN options provide the most mechanis - '
tic pre.11ction of the axial power shape change during a BWR transient.
Data for the neutronic regions are shown in. Table
. 3 .' 4 . i 3.4 TRIP-LOGIC r
The RBS protective systems are characterized by 101 trips. These trips include reactor scram, turbine and feedwater trips,' recirculation pump trip and transfer, ;
and safety / relief valve logic. The RBS tW ps also 1
include logic for event initiation and control system
)
functions. RBS trips are shown-in Table-3.5. ,.
I
( Revision 0 Page 23-l
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( 3.5 CONTROL SYSTEMS The control systems used to simulate RBS include an instrumentation section (i.e. sensed process variable),
water level measurement, recirculation flow control section, turbine control section, feedwator/ level control system, and a miscellaneous edits section. The control schemes are shown in Figures 3.2 through 3.7.
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Table 3.1 Control volume Geornetric Information (Continued) i .
2 ,
10 M E lag goL .I)!OL g!! Flout rtoist Ormy gg OEstetP110m
210 0 0 938.080 5.100 5.100 5.100 '183.922 1s.303 31.078 uPPEa PLEutet -
l- 220 0 0 274.620 5.905 5.905 5.905 46.507 0.5298 - M .178 siamePIPts j 230 1 0 744.388 7.533 4.53187 T.533 102.790 0.2647 42.083 stEast sEPAAA1GES 232 0 0 - 784.951 7.406 7.406 T.406 .105.999 0.0833 52.021 STEAsl 999ER I
234 0 -0 2861.900 17.292 17.292 29.240 97.876 6.4780 52.021 wEssEL se E .
j 240 2 0 1426.611 12.391 3.10524 - 2.405 232.578 0.T460 42.083 uPPEg ammerumsgo i 244 3 .0 294.349 9.604 3.12899 9.604 29.607 1.0625 42.417 tanfEg ammats ,
i 250 0 0 1335.570 T.275 -7.275 7.542 177.079 1.9440 35.142 at00LE maammmmsg j l 256 0 0 1689.800 ' 24.017 24.01T 30.T16 48.502 1.4550 11.125 L0tdER Samanumre ;
300 0 0 '181.692 33.002 33.002 51.237 3.546 1.5025 -18.376 mECIRC suCTIem }
! 305 0 0 '64.100 2.418 2.418 9.015 3.546 0.7513 -15.958 mECIRC MSIP i 1 311 0 0- 157.210 20.736 20.7 % 44.333 3.546 1.5025 15.043 mECiet ptper sistianneE i 315 0 0 85.977 1.197 1.197 11.773 7.303 1.2116 5.693 DECIRC setaff0tP 320 0 0 152.184 20.980 ' 20.990 30.508 4.988 0.7953 6.890 RECanC elSER t j 325 0 0- 159.351 14.818 14.818 14.818 20.058 0.8772 10.938 JE7 Ptsu's 33.155 1.797
~
500 0 0 336.290 26.3 % 26.3 % 10.143 27.500 sTE m LinE vol. 1 SG1 0 0 204.787 1.797 1.797 20.190 10.143 1.797 26.391 stem time UUL. 2 '
502 0 0 126.706 1.797 1.797 12.492 10.143 1.797 26.164 sTE M LluE vol. 3 e l 504 0 0 510.526 2.318 2.318 50.333 10.143 1.797 25.643 stem LINE wet. 4 506 0 0 438.028 5.521 5.521 43.185 10.t43 1.797 21.919 stem Lame wot. 5 l 508 0 0 433.028 5.521 5.521 43.185 10.143 1.797 18.195 stem LluE vol. 6 i 510 0 0 438.028 5.521 5.521 43.185 10.143 1.797 14.471 stem time vol. T I l 511 0 0 438.028 5.521 5.521 43.185 10.143 1.797 10.747 stE m LINE vol. 8 512 0 0 229.787 3.257 3.25T 3.257 16.663 - 3.257 10.017 tup 8INE euEADER [
513 0 C 280.164 2.443 2.443 27.622 10.143 1.797 10.102 sTE m LINE vol. 9 t 514 0 0 101.610 7.982 T.982 10.017 10.143 1.797 3.917 STEmpt LINE wot.10 600 0 0 113.35G 3.948 3.948 78.364 1.446 0.227 12.151 STEmpt 87 Pass vol 1 6C2 0 0 156.780 17.307 17.307 151.190 1.037 0.190 -5.156 STEmpt 87 Pass WOL 2 650 0 1 10000.0 20.000 20.000 20.000 500.000 14.300 -14.750 CouDEksEW 900 0 2 1.43E+6 186.250 186.250 177.000 8080.000 101.000 -32.000 METAlep4ENT
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EA-PT-91-0003-M Table 3.3 Heat conductor Information VOLS 40ML MDMR DMEL DMER ClHit CMeR iD ISL ISR ASUL ASLS 0 0 0 0.0544 0 0.5 101 0 101 0 2417.71 24.3282 0 0.5 24.3282 0 0 0 0.0544 102 0 102 0 2417.71 0 0.5 2417.71 24.3282 0 0 0 0.0544 103 0 103 0 0 0.0544 0 0.5 0 2417.71 24.3282 0 0 104 0 ?O4 0 0 0 0.0544 0 0.5 105 0 105 0 2417.71 24.3282 0 0.5 '
2417.71 24.3282 0 0 0 0.0544 106 0 106 0 0 0.0544 0 0.5 l 0 24tT.71 24.3282 0 0 107 0 107 0 0 0 0.0544 0 0.5 I 108 0 108 0 2417.71 24.3282 0 0.5 )
24.3282 0 0 0 0.0544 109 0 109 0 2417.71 0.0544 0 0.5 24.3282 0 0 0 110 0 110 0 2417.71 0 0.0544 0 0.5 0 2417.71 24.3282 0 0 111 0 111 0 0 0 0.0544 0 0.5 112 0 112 0 2417.71 24.3232 0.054 , 0 0.5 24.3282 0 0 0 113 0 113 0 2417.71 0 0.5 24.3232 0 0 0 0.0544 114 0 114 0 2417.71 0.0544 0 0.5 24.3282 0 0 0 115 0 115 0 2417.71 0.0544 0 0.5 24.3282 0 0 0 116 0 116 C 2417.71 0 0.0544 0 0.5 0 2417.71 24.3282 0 0 117 0 117 0 0 0 0.0544 0 0.5 118 0 118 0 2417.71 24.3282 0 0.5 24.3282 0 0 0 0.0544 119 0 119 0 2417.71 0.0544 0 0.5 24.3282 0 0 0 120 0 120 0 2417.71 0 0.0544 0 0.5 2417.71 24.3282 0 0 121 0 121 0 0 0 0.0544 0 0.5 0 2417.71 24.3282 0 122 0 122 0 0 0 0.0544 0 0.5 123 0 123 0 2417.71 24.3282 0 C.5 24.3282 0 0 0 0.0544 124 0 124 0 2417.71 0.0544 0 0.5 24.3282 0 0 0 125 0 125 0 2417.71 0.911 0.911 44.18 7.291 0.17 1.464 4.675 4.001 241 141 256 43.225 4.675 4.001 12.5 12.5 606.196 100.039 0.17 1.464 242 142 256 593.093 4.675 4.001 1.196 1.198 56.842 58.098 9.588 0.17 1.464 244 144 256 i
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'EA-PT-91-0003-M Table 3.4'sRETRAN Core.Neutronic Region Inforsmation REGION 8ESW ' CORE CDRE REGION REGION 4 10 luTERwAL WottBE CoupuC10R El.EvutTION NEIGNT
.- 1 24 100 0 0.911" 0.911 2 ~ 12 101 1 1.411 ~ 0.500 -
l 3- 12 102 2 1.911 0.500 I, 4 12 103 - 3 2.411 - 'O.500 -
i 5 12 104 4 2.911 0.500 +
!~ 6 12 105- 5 3.411 0.500 l 7 12 .106 6 3.f t1 - 0.500 l 8 12 107 7 4.411 0.500 .
i- 9- , - 12 108 8 4.911 0.500
! 10 12 109 9 5.Att- 0.500 i 11 12 110 to 5.911 0.500
'- 12 12 - 111- 11 6.411 0.500-13 12 112 12 6.911 0.500 i -' 14 12 113 13 7.411- -0.500 l 15 12 114 14 7.911 0.500
! 16 12 115 15 8.411 0.500
' 17 12 116 16 8.911 0.500 18 12 117 17 9.411 0.500 i 19 12 118 18 9.911 0.500
[. 20 12 119 19 10.411 0.500 21- 12 120 20 -10.911 0.500 22 12 121 21 11.411 0.500 23 12 122 22 11.911 0.500
. 24 12 123 23~ 12.411 0.500 l 25 12 124 24 12.911 0.500 26 12 125 25 13.4N 0.500
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. CARD # - ]DT4P - IDSIG 131 Lug .SETFT MLAY ' SESCRIPflGR i- 060970 187 4 234 0 1127.7 0.& *sa-4 set orEE
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EA-PT-91-0003-M l O Figure 3.2 RBS Instrumentation Control System 4 FLOV
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EA-PT-91-0003-M ! } g~' / Figure 3.3 RBS Water Level Control System 17 s$ TOTAL LIQUID 40 VDLuwt no; __Y e
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EA-PT-91-0003-M Figure 3.3 RBS Water Level Control System (Continued) 14 } 'l0EI -130 l l -131
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6( 60 26 256 Revision' O Page 39 I 1
l l EA-PT-91-0003-M l l 1 e ( } V Figure 3.4 RBS Recirculation Flow Control System j - _, Ch 4 45 1 L T L _jux.s kh h* ; .~2 M -3 1
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MANUAL - AVID TRANCIER L ( i TRIP = DN 4 AUTC TRIP = Orr s) MANUAL I l m Revision 0 Page 40 1 l
EA-PT-91-0003-M
~
(s'") Figure 3.4 RBS Recirculation Flow Control System (Continued)
^
MANUAL - AUTO TRANSTER pggy ^ TRID = CN => AU/0 SENSCR
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f TABLC 17 ()
/N Revision 0 Page 41
1 l EA-PT-91-0003-M 1 CN, U Figure 3.5 RBS Turbine Control System
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@ PDESLUPE TRANSDVCER Y__ __.ree i l '. T I in -t::
LLG rNG d jT!** dPEE*,;URE !ETPT
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wATC f.a - LIMtTER : s y _ "8v . _L_. . : I 229 Id *III I4 ,=4 hat 4 hN UM M us rNG t D'II' LT t ; ;- SUM 3 A ,apLt 26 i l 46 231 7gp l-10 hPV C AP ACIIY LIM- p
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/ \ eV S On 0 Page 42 (
EA-PT-91-0003-M Figure 3.5 RBS Turbine Control System (Continued) tR!p t!M 3a TRIP I tcv LINtae!!Attoa. TO ZLR0 f dl IN Eh)
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U IN N bum RATE LIMITED 4 a 0 25 fa,Lt.35 gIAg LOAD DEMAND , Revision 0 Page 43 l l
EA-PT-91-0003-M i m a
; / , \ Figure 3.6 RBS Feodwater Control System l i
l i l i.E i POlN I SE I DC w'N SENSE D h ~ f Hi TEE DV A TCP ( - 10 ) - ? g [ INT FNu
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/"' g Revision 0 Page 44 (v)
EA-PT-91-0003-M O Figure 3.7 RBS Miscellaneous Edits Control System
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i Q Revision 0 Page 45
. . . , . . ~
EA-PT-91-0003-M l l ( N Figure 3.7 RBS Miscellaneous Edits Control System (Continued)
? !-l 9
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. . .m _ . _ . _ . . _ . _ _ . . _ _ . _ . _ _ . , _ . _ _ . _ _ _ _ _ _ . _ . _ _ . . . _ . . _ _ .
n EA-PT-91-0003-M
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~
jlh 4.7 CROSS SECTION GENERATION This section describes the procedure used.to obtain the cross section information used in the RETRAN core model. Polynomial expressions for the cross sections required by : the. dif fusion equation are prepared by-SIMTRAN l from SIMULATE-E restart files. The SIMULATE-E cross section data contains delayed neutron fractions and energy group inverse velocities calculated- by lattice ; i physics: codes. contributing to the SIMULATE-E-analysis. ; N The SIMTRAN calculation is described in Section-2.2.- ,
' Consistent with the-RBS RETRAN model, the SIMTRAN 3f 4 ;
model-includes-25' active core nsutronic regions and.two'-
- 5,,' ' - reflector regions ' THE RBS SIMULATE-E modeling -:also g -contains 25: active core regions,--eliminating the need for . axialDcollapsing or_ expansion of.the core analysis-data.
i For transient' analyses involving: control rod scram, . L L . threevseparate SIMTRAN calculations are performed.: - As-- notabin:Section'2.2,thebasecaseusescore-andsystem- , l : conditions consistent.with the. initial conditions to-be
, modeled- ini the:: transient analysis. -Secondary cases - --provide; data for partial ~ and ~ full insertion 1- of the control-rods.-
Revision.O Page 47-
w. m $ 1 -9 4- weg. e a e%e= 4 ger-+-4-. er ed-'eme Am--ekshewf 4-we-t -*-1,se 3eWed eene e ----- ---.ri-rtwa-*r-e m_- * * - - - h - *----------A - -
FA-PT 0 0 0 3 -M i l l l SIMTRAN. includes several additional weighting l features to allow flexibility in obtaining the spatially
-averaged nodal temperatures and densities. The code also contains provisions to adjust reflector cross sections so that boundary conditions between SIMULATE-E and RETRAN-02 will agree.-
A sensitivity study was performed to identify the optimum temperature and density weighting options, as well as the. reflector cross section adjustment option. These sensitivities were performed using the model for the Peach Bottom Atomic Power Station, and the_ data for ,. the - three turbine- trip tests conducted. The power distribution and the neutron flux response were used as the figure of merit. The result of'this study indicated-that the best'results achieved were with the reflector group 2 absorption cross section adjusted to match nodal
' power, the power distribution- is used as the' density-t weighting function, and an inverse distribution is used as the temperature weighting function'.
Previous util'ity experience with RETRAN using-L SIMTRAN-collapsed cross sections has shown overprediction of transient power peaks relative - to ' the Peach Bottom test data. In the GSU modeling, however, analyses with i I Revision 0 Page 48 o l l l'
EA-PT-91-0003-M
'f N - unmodified'SIMTRAN cross sections predict the poWor peaks accurately. .This agreement is due in part-to the use of delayed - neutron information from the ENDF/B-V library rather than the: ENDF/B-III -data coded in the original CASMoi and in part to EPRI-developed improvements in the multi-state control _ rod model . Analyses reported in this document are based on cross section information extracted - directly from SIMTRAN output without further normaliza-tion.
3 l 4 L. t Revision 0 Page 49
EA-PT-91-0003-M r~x
)
k' 5.0 CALCULATION OF FUEL ROD GAP CONDUCTANCE Gap conductance directly affects how rapidly a change in rod power will produce an associated change in
-rod surface heat flux. RETRAN requires hot channel and core average gap conductances to perform transient analyses.
RETRAN treats the gap as a material through which heat is conducted. The RBS RETRAN model implements the transient gap conductance using cold gap width and fill gas conductivity as a function of temperature. The ; methodology used by GSU to determine gap conductancis ('") v values for safety analysis is described in greater detail in Appendix A. The gap conductance values used in the RETRAN i
-analysis have two conflicting ef fects on the consequences of the transient. Low values are more conservative for system or core average effects and high values are more -
conservative for hot channel effects. Gap conductance values are calculated separately for these applications. The core average gap conductance affects the react'ivity feedback due to a change in moderator density and void fraction. During a rapid power increase Revision 0 Page 50 i
EA-PT-91-0003-M l n: _, x) transient, ' heat transferred to the coolant causes coolant voiding, which has a negative reactivity effect. Faster heat de' position in the water produces more negative void reactivity feedback, which mitigates the power increase.
' Lower gap-conductance values in the RETRAN system model delays'the. onset of corewide coolant voiding, resulting in a. larger power increase. In slow events, the gap- ,
conductance has a negligible effect because the heat flux - remains'approximately equal to the neutron flux through-out the transient.- The power response of the hot channel analysis-is -
- d r i v e n ~.b y ' b o u n d a r y conditions from the core -average system analysis. The gap conductance determines the heat !
flux response in-the. hot channel analysis.--
For pressurizatior. transients and other events whic:1 include' rapid- power increases, a larger value of hot channel = gap conductance- results in a larger prompt increase -- in - heat flux. Higher values of heat flux r result in more' severe thermal margin decrease during-transients,_as indicated by a larger value of Change in Critical' Power Ratio ( ACPR) .
The methodology described in Appendix A was used to 1 produce sample-gap conductances for RBS. Cycle 3. The Revision 0 Page 51
EA-PT-91-0003-M l-. t
't,d .
core average gap conductance was calculated to be 1218 2 i BTU /hr-f t *F and the hot channel gap conductance was l calculated to be 2315 BTU /hr-ft 2 _.7, l l l l l l l 1 1 \ t l l l l l l l ? l l t [ .' l l
/O V' Revision 0 Page 52 l
I 1.-. - -
EA-PT-91-0003-M O 6.0 COMPARISON TO RIVER BEND STATION TRANSIENTS The RBS RETRAN model is gaalified by comparison of calculated results to measured data from operational events and startup tests at the plant. As part of the Emergency Response Information System (ERIS), RBS is equipped with a transient data recorder which can digita11, store many plant variables continuously during plant operation. 6.1 GENERATOR LOAD REJECTION O RBS experienced a generator load rejection transient on December 1, 1989. The scram resulting from this event was termed Scram 89-04. The initial conditions for this transient are shown in Table 6.1. The timing of major events during the transient is shown in Table 6.2. RETRAN predictions are compared with ERIS data in Figures 6.1 through 6.8. With the reactor operating at 96% power, a turbine-generator trip resulted in an automatic reactor scram on Turbine Control Valve (TCV) fast closure. Concurrent Revision 0 Page 5 1
EA-PT-91-0003-M.. ! b - with the turbine trip, the recirculation pumps automati- ' cally transferred to the LFMG sets.- Immediately follow-. I ing the_ scram, reactor pressure rose rapidly to approxi- 4 mately 1120 psig, causing the four groups of safety relief valves, including the five low-low set SRVs, to i cycle open. The-turbine bypass' valses also opened. Water level initially _ decreased due to collapse of steam. voids resulting from the rapid pressure increase and reactor-scram. The narrow range level dropped to as
-low as 10" above instrument zero from an initial level of 37". Feedwater flow remained-relatively high.since the .
water -level = was well below the normal water level setpoint. When water -level is. below this setpoint, j sfeedwater regulator. valves are opened to increase feedwater flow. Water level then recovered, reaching the level 8_ high water level trip at around 50 seconds.- -This resulted in a feedwater pump trip. ' Figure 6.1 showst the. Average Power Range Monitor
-( APRM) flux response.- Since the bypass valves opened and the reactor scrammed,'neither the predictions nor the data:show any increase-in flux above the initial value.
Flux decreases rapidly following scram. _ Figure 6.2 shows the predicted reactor pressure . response compared to data I [ Revision 0 Page 54 a l
EA-PT-91-0003-M
?
from two channels. RETRAN predictions are in good 1
- agreement with thel peak pressure.- The predictions are in agreement. With the data regarding the number of SRVs which opened in the relief mode (nine). However, analysis.of tailpipe temperature r esponses indicates that several of the remaining seven valves opened briefly in the. safety mode-at pressures somewhat below the normal
(-; relief or safety-valve setpoints. To accommodate these P open safety valves, the opening of two valves in the safety. mode- is: included in the RETRAN modeling.
-Reactor pressure decreases rapidly following opening 1
of ; the E SRVs, which close -in. sequence- as the pressure
~
idecreases. :The pressure decrease is terminated when the L bypass closes at.around 21 seconds. These valves reopen for a brief period at around-33 seconds to relieve the subs'quent e pressure-increase. L Total . steam-_ flow -as measured , by' the t elbow flow
' meters -upstream -of. the - SRVs - is shown. in Figure. 6.3. . Predicted and measured: steam flow both drop rapidly .
following fast closing of the TCVs. The subsequent increase-is due to opening ofs-the..SRVs and the bypass. The'. steam flow decreases.as rellef valves close during
- the pressure ' decrease period. , Step- changes -in flow Revision 0 Page 55
3 EA-PT-91-0003-M l -
~
indicate the'. times when groups of valves closed. The flow measurement at . low flow is inaccurate due to the
-i type of instruments-used and the calibration' range. The predicti'on provides a more realistic representation beyond 20 seconds, when the steam flow is low. The predicted increase of flow at around 40 seconds occurred [
when.the bypass reopened.
Figure 6.4 shows the measured and predicted core ; flow response. Both show a rapid decrease in core flow following transfer to the LFMGs. The prediction is in excellent-agreement.from above 30% of rated flow. The . flow measuring instruments are .less accurate at low flow. Feedwater flow response is- shown in Figure 6.5. Both measurements and predictions show a general decrease in.feedwater flow and-then a rapid decrease following feedwater pump trip on a high level 8 at around 50. seconds. Differences between the predictions and plant. data are due to the dif ferences in predicted narrow range water level shown . in Figure 6.6, Water level, with steam / feed flow' mismatch,--is the -main-determinant of feed flow magnitude. Water level-is one of the more difficult variables to predict accurately. L .
'- Revision 0 Page 56 , , - ,e, -c - w --
EA-PT-91-0003-M
. m s.
In addition, the feedwater control system had two trains in auto and one in manual at the time this event occurred. The RETRAN base model is set up for three trains in auto mode, with an approximation used for the 2 auto /1 manual mode. An earliez benchmark" for a similar event, which also occurred while the feedwater system was in the 2 auto /1 manual mode, the feedwater flow was forced to model-measured data. More accurate predictions of water level were obtained in that case. For this benchmark, feedwater flow response is included in the prediction to provide a more general simulation of . l- the plant response. The overall water level response trend is predicted well. However, the prediction does not drop as low as the plant data and recovers somewhat more slowly. The first water level recovery is over-predicted by the code. Figure 6.7 shows excellent agreement in recircula-tion pump speed including speed following pickup of the LFMGs. This shows that the RETRAN recirculation pump model is a very good representation of the actual component. l l. l t
/
Revision 0 Page 57 F
EA-PT-91-0003-M
;Qi .
v Figure _6 8 shows predicted and measured bypass valve position. ThelRETRAN prediction shows'a shorter period of valve reopening. Simulatior of this load rejection transient shows that . RETRAN can predict plant performance during this type of event with good accuracy. . Predictions- are especially good during the_first-second or so following initiation of the event. This is the. time period during which-Maximum Average Planar Linear Heat Generation Rate (MCPR) will occur for similar but more severe events such-as those analyzed for-the thermal limits determination. . L , 6.2 WATER LEVEL INCREASE EVENT RBS experienced a- water level increase transient on June 18,.1987. The reactor was-operating at approximate-ly 70: percent power and 54.5-percent-core flow whenna
= battery inverter f ailed. -Initial conditions for'this transient--are shown-in Table 6.3.
I The ' inverter - f ailure interrupted power' to panel-l
-1VBN-PNLO1B1, which in turn locked the feedwater regulat-i j- ing valves in positi'on. In addition to the loss of power O Revision 0 Page 58
l EA-PT-91-ooo3.g l (- _ O to the feedwater controller, a recirculation flow control valve (FCV) runback sigr.a] was generated causing both valves to begin to close. FCV "A" completed the runback; however, the hydraulic power unit serving FCV "B" f ailed, causing the valve to stop short of the full runback. Recirculation Pump "B" attempted to transfer to slow speed immediately; however, the LFMG for the pump did not engage and the pump coasted down to zero speed. This concurrent core flow reduction caused a reduction in power and steam flow, which was not offset by a decrease in feedwater flow because the feedwater regulating valves , were locked in position. Water level rose to near the A () Level 8 setpoint, where the reactor protection system generated a scram signal. The indicated level did not exceed Level 8 on all water level instrumentation, so the turbines and feedwater pumps did not trip initially. "'he turbine and one feedwater pump were manually tripped following the reactor scram. Following the reactor scram, water level decreased and rose'again, eventually exceeding-the Level 8 trip setpoint and tripping the feedwater pumps. The timing of major events during the transient is shown in Table 6.4. / \_ U Revision 0 Page 59
.. . . . -. . - - . _ . .... - - . -. - .~ .. - - . - - - . . - . . - - - . - . _ . .
r EA-PT-91-0003-M Predicted plant performance is compared with ERIS-data in Figures =6.9 through 6.14.- The RETRAN results agree with measured plant data - available ' for the. event. RETRAN predicts a slightly- < slower level recovery than the data; th'is difference is 'I probably caused by modeling asymmetric recirculation loop conditions with a single, composite recirculation loop. Instruments recording narrow range. water level and core-
< flow draw power from the f ailed bus, so these indications are unreliable 'during the first 21 seconds -of the -
transient. . When power.was_ restored, the instrumentation- . resumed--functioning. Figure 6.9 _shows the neutron flux during_ the a.c transient. The predicted flux follows a similar trend as , that exhibited by the data, except that the flux. level is higher-before scram occurs. This is most likely-due to L uncertainties-in the core flow measurement when_one pump- . i. L , Lis running.and the other pump is! idle. Figures 6."10 and 6.11 show the water level response -
'as- -determined from: the narrow range 'and wide-- range instrumentation.- -The rise in -level -following the-
-recirculation. flow decrease matches the observed level
, -v e r y -- w e l l .- The _ subsequent drop in level following 1 n (! cO L l: Revision 0 Page 60 wr- -e so wc--~ e-- -- .,w- ~ s -w-
i EA-PT-91-0003-M h. D. reactor scram also matches the observed data. The level recovery.following the scram is somewhat delayed in the RETRAN calculation. As discussed ~above, this is felt to be aue to the limitations in modeling asymmetric loop behavior with.a single composite loop model. Figure 6.12 shows the core flow response during the transient. The-initial portion of the transient agrees s with data,-however, the flow at this point was forced to agree with the measured flow. The flow coastdown following the trip of the remaining pump dif fers from the
~ data by approximately 7% of rated flow. This is~likely -
due to :the ' imitations in modeling . asymmetric loop I
' behavior with a single composite loop model. ~ Figures 6.13 and 6.14 show the feedwater and turbine -
steam flows during the transient. The responses obtained agree very well .with the measured data. Feedwater remains constant due-to the faedwater regulating valves locking.in position. Eventually one of the-pumps-is. tripped, and feedwater flow decreases to very nearly the sam'e value as calculated by RETRAN. -Steam flow initially
-drops in response to the power decrease associated with the. recirculation flow decrease.
Following the scram, Revision 0 Page 61
EA-PT-91-0003-M (3 . w) the calculated flow decreases in a manner similar to that observed. 6.3 WATER LEVEL SETPOINT STEP CHANGES RBS Startup Test 1-ST-23A, which was performed on May 24, 1986, included step changes in the reactor water level setpoint. The RETRAN model was benchmarked to the feedwater flow and reactor narrow range water level response to these changes. The unit was operating at near rated conditions for the test. . RETRAN predictions are compared with ERIS data in {,, Figures 6.15 through 6.18 for the +6" and -6" level setpoint tests. For the +6" level setpoint change, both the feedwater flow and water level response agree closely with plant data. For the -6" test, it appears that a slightly larger step change than intended was initiated (approximately 7.8"). The plant feedwater response is much slower for the negative change. RETRAN predicts that-changes in feedwater flow will take place at the same rate as for a positive step. However, the overall water level response is comparable. A Revision 0 Page 62
EA-PT-91-0003-M O
~ . Table 6.1 Initial Conditions, RBS Load Rejection Tran-sient-(Scram 8904) y Plant Data RETRAN Power- 96.8% 96.8%
Core Flow 96.6% (81.6 MLB/HR) 96.6% Steam Flow 96.1% (12.0 MLB/HR) 96.8%- Feedwater-Flow 96.1% (12.0 MLB/HR) 96.1% Dome Pressure 1032.0 1032.0 Water Level 37.0" 37.0"
' Recirculation control 2 in Manual 2.in Manual Pressure Regulator' Full Arc Full Arc -Feedwater Control 3 Element 3-Element 2 Auto, 1 Manual 2 Auto, 1 Manual .-- f
m.
\ '
f-- . 1' l l. l l Revision 0- Page 63
-- m . _ . _ - _ _ _ - .m. + _ _ _ _ ,_ ,__ ~ w
EA-PT-91-0003-M 1
-S ,
Table 6.2 Sequence of Events,.RBS' Load _ Rejection Tran-sient (Scram 8904) P1 ant. RETRAN Data TIME Load Rejection 0.00 0.000 TCV. Fast Closure 0.10 0.107 Scram 0.12 0.177= EOC HFMG Trip 0.14 0.247 BPV Fast Open 0.10 (0%) 0.207 (0%) 0.8 (100%) 0.482 (100%) SRV-1-- Open - (1) 1.4 1.78
SRV-2-.Open (1) 1.5 1.78 SRV-3 Open (3) 1.5 1.92-SRV-4 OpenL(4) 1.5- 1.92 p -Peak-Pressure 2.0 2.2 -
y ": -SRV-4 Closed 4.5 5.97 O SRV-3 Closed 11,8 9.38 ' i> e-SRV-2 Closed 14.7 10.05 L SRV-1 Closed- 20.8 10.76 l: L l-p- l t D b
- Revision 0 Page 64-e . . . - . r- , n . - -- _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _
EA-PT-91-0003-M O ~
-Table-6.3 Initial Conditions, RBS Water Level Increase Event ' Plant Data MM Power (%NBR) 70.0 70.0 '
Core Flow -( % NBR) - 54.5 5 4 ~. 5 Steam Fl'ow (%NBR) 86.2 67;2" Feedwater Flow (%NBR) 67.2 67.2 Narrow Range Water Level (in) 35.7 35.7 Recirculation Flow Control 2 loops.in MANUAL Pressure Regulator full arc Feedwater Controller Three-Element
-lv(9 'Non-condensible gas build-up in condensing chambers of . steam flow' instrumentation for feedwater control resulted in:an. error in the indicated. steam flow.
Revision 0 Page 65
EA-PT-91-0003-M o - Table 6.4 Sequence of Events, RBS Water Level Increase Event TIME (sec) Battery Inverter Failure 0.0
-FCV Runback Initiated 0.0 Recirculation-Pump-B.. Tripped O.0 -Reactor Scram on High Level 37.0 Recirculation Pump A Tripped 43.0 Feedwater Pump A Tripped 54.0 Manual Turbine Trip 84.0-(Not Simulated)
Feedwater Pumps B & C Tripped 86.0 (Not Simulated) i e r O i I: 8
-Revision 0 .Page 66
EA-PT-91-0003-M O - Figure 6.1 Predicted vs Measured Neutron Flux, RBS Scram 89-04 RETRAN BENCHMARK OF SCRAM 89-04 0 - O I I I I I 2
3 E 8 e -
~
X D _X9 - - O LL8 Z _JO LL o E
'T &
F- Q. O E w - _ Zu {0 x - (O F-
- W W
0 0
^ 1 1 1 1 o c0 10 20 30 90 50 60 TIME (SEC)
O Revision 0 Page 67
.c 6 -ERIS' REACTOR. PRESSURE (PSIG)-
y .a00' a50 ' 900 950 1000 1050 1100 1150 1200 m g . . i i i i i i i
.g " ERIS g- . REACTOR PRESSURE (PSIG) o 800 e50 W 950 10E) 1050 1100 1150 1200 D. r * .i i i i i i i i , .RETRAN REACTOR PRESSURE (PSIG) p e50 900 950' IGI) 1050' 1100 1150 1200 . . . . . - - . i MT tp N -
U) O 2 03 5 - 3 4, 3 m o Z U KL z e.
~ ~
2 o, y E o, g _8 -
$ $ E e g- 8 :
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$ u m ' - .
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EA-PT-91-0003-M ,-^(. . L.) Figure 6.3 Predicted vs Measured Total Steam Flow, RBS Scram 89-04 i RETRAN BENCHMARK OF SCRAM 89-04
.g 3 i i i i i E'.
s _. sE . 39
~ '3 S f r r -
/~'g
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,_ i i AAAAAh-- t A Ani. _ _ A .
o c0 10 20 30 40 50 60 TIME (SEC)
. Revision 0 Page 69
r I i EA-PT-91-0003-M 4 i g Figure-6.4- l Predicted _vs Measured Total Core Flow,
.RBS' Scram-89-04.
l l
l lRETRAN BENCHMARK OF SCRAM 89-04
]
m 8.- l i. .i- i- i ]- p i I
.-- - -- s - g- g_ ! t .r. _r- h. _ <
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TIME 1SEC) I -' Revision 0 Page.70 4
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w *v w y - w r, -ynv = We e t---e 4- & .r e ---4 ..,,:o, - - _ s__w -_-._., - . _ __,____,_m_ _ . , _ . _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ , _ _ _ . _ , _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ . _ . _ _ _
__ ._ ., _ _. - - - _ _ . _ . = _ . . - , _ . . _ _ . . _ _ _ _ _ . . - . _ . - _ _ _ , i , EA-PT-91-0003-M LO , Figure 6-5 . Predicted'vs-Measured Feedwater Flow, RBS Scram 89-04' RETRAN' BENCHMARK OF SCRAM 09 ~ m m i i i i i
~
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Revision 0 Page 71
. :. . , a.. ,- , - .- ,. , , -. . , , -,
~
wh g .' # ,' N ERIS NARO RANGE: LEVEL C(INCHES) " O 10 20 30 90 50 60 g :'
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tn O 10 20 .30 *40 . 50 60 r- a .a a 4 i s # , O ERIS NARD RANGE LEVEL A(INCHES) * *
, 0 10 20 30 90 s0: s0 m i i i a i a i RETRAN NR WATER LEVEL' [ INCHES) -
g 10 20' 30 90 50 60
' S r'O2-o.
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r EA-PT-91-0003-M l O - Figure 6.7 Predicted vs Measured Recirculation Pump Speed, RBS Scram 89-04 i 1 RETRAN BENCHMARK 0F SCRAM 89-04 N l- l i i i i i
~ - x- kl 'El ~ - El 1
x-w x-w I O O O W.- ' W W W W - 1 W{ 1 W{
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-o -o. c0 10 -20 33 TIME (SEC)
O Revision 0 Page 73 e re. a, *--,_ .-..wv,. 4 ..,~,...>,,,,_.,v., . , ~ , - 4,- - . . - . . . . . - . _ - ,,...,,.-...,...~...wm-.- - - - . . - - . - - - - - , . .
EA-PT-91-0003-M O Figure 6.8 Predicted vs Measured Bypass Valve Posi-tion, RBS Scram 89-04 RETRAN BENCHMARK OF SCRRM 89-04 _g- g .-;. . ; i i- i r --- d e4 , 8 ~8 g r J H- m . o- ,
~
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l l t I l ! 20 30 40 50 60 o c0- 10 ilME (SEC) O -Revision 0 Page 74
EA-PT-91-0003-M
,r ' ,
t
's Figure 6.9 Predicted vs Measured lieutron Flux, RBS Water Level Increase Transient LEVEL TRRNSIENT 870618 - NEUTRON FLUX (1 NBR) g- g i i i i i i r-g~ g -
E ,B (n ci R 58 ~ $s U 0:: t 8 8 I u- 1 1 l l
. l o c0 20 90 60 00 100 120 190 160 TIME (SEC)
Revision 0 Page 75
EA-PT-91-0003-M O G Figure 6.10 Predicted vs Measured Narrow Range Water Lovel, RBS Water Level Increase Tran-sient LEVEL TRANSIENT 870618 - NRRROV RANGE LEVEL (IN)
~
g g i i i i i i S 8
? .? ' . ;)
t'h g V O _e _ 5R CR x i J S S !
, f ! I 1 1 I I o cc 20 40 60 00 100 120 140 16(
TIME (SEC) t f ( Revision 0 Page 76
EA-PT-91-0003-M q R_j Figure 6.11 Predicted vs Measured Wide Range Water Level, RBS Water Level Increase Tran-olent LEVEL TRANSIENT 870618 - VIDE RRNCE LEVEL (IN)
~
i l g g 1 I / I i 1 S S 9 (% e is . O g e - _ wR ys g g f I I I I f I o c0 20 90 60 80 100 120 140 160 TIME (SEC) e f Revision 0 Page 77
EA-PT-91-0003-M (' . q rigure 6.12 Predicted vs Measured Core flow, RBS Water Level Increase Transient LEVEL TRANSIENT 870618 - CORE FLOW (IN8R)
@~ @
I i i i i i i g - g -
~
8 ,8 in b a: _ 3 ~
$Y 8 (V
A
~ ? ? ~ ~' . f I f t l j o c0 20 40 60 00 100 120 140 160 TIME (SEC)
I'"; N_J Revision 0 Page 78
EA-PT-91-0003-M r w Figure 6.13 Predicted vs Measured Feedwater Flow, RBS Water Level Increase Transient LEVEL TRANSIENT 870618 - FEEDVATER FLOV (1NBR) R- R i i i i i i i g - g - 8 ,8 ' 4 g] 7
\j b _E - .
58 G8
L.
r e R R i i i i ( i i i o c0 20 90 60 80 100 120 190 160 TIME (SEC) { i \,- Revision 0 Page 79 l
l 1 I EA-PT-91-0003-M l
,~, 'w' rigure 6.14 Predicted vs Measured Turbine Steam riow, RBS Water Level Increase Transient LEVEL TRANSIENT 870618 - TUR61NE STERM FLOV (1 NBR)
R. r R i i i i i i i 8 8 _ _ ~ E 8 7 =
/ .m h] -- _ cr: - ~ $8 CS x
R R i i
.. /\ A i i i o c0 20 40 60 80 100 120 140 160 l
TIME (SEC) i l f) O Revision 0 Page 80 l
r-0: O I. ',' I Y !: a i i i i- M M
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Figure 6.16 Predicted vs Measured Feedwater Flow, RBS Water Lovel Setpoint Change (+6") LEVEL SETPOINT CHANGE (+6 INCHES) O~ U i i l i I n n I I N N (D Q _J _) bo - bo - - a a - W W
.q O O i <
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$ zd C
(O %
~ & % L1)
W % 0_ o i I I I I T T-5 0 5 10 15 20 25 TIME (SEC) O Revision 0 Page 82 l
EA-PT-91-0003-M O Figure 6.17 Predicted vs Measured Water Level, RBS Water Level Setpoint Change (-6") LEVELSETPOINTCHRNCE(-6INCES) i~ $ i i i i i i i N N i i a ~O M9 W9
Wo Io I O O Z Z M w
W e 9 g9 - W7 07 ~ e ~s 0 Z i i Z C V E I l I O O J J9 -39 - uma e - gi
% Z Z
Z g g .-
~ %
Ei..F9 W r- Wo i ti i o o I
f. f I I l l l I l l 7 2.0 2.5 ' 5.0 7.5 10.0 12.5 15.0 17.5 20.0 22.5 25 l TIE (SEC) l l
I (,,) l Revision 0 Page 83
\
EA-PT-91-0003-M i O -
. Figure 6.18 Predicted vs Measured Feedwater Flow, RBS Water Level Setpoint Change (-6")
LEVELSETPOINTCHANCEl-6 INCHES) 0" O i i i i I I I i I N N 0 : y-O. 0 O ]-- W W 0- 0- ' Z Z-E E 1-I o Io (
'U; - 0; .)1 )i D. O J- J. -.
LL lL ' 3 I' 3 -
u. o - u. o -
N N ?
' Z' E
(0 % ' H.' g: . gj W E i o, o l I I 1 I I I 1 11.5 20.0 22.5 25. i %.0 2.5 5.0 7.5 10.0 12.5 15.0 TIMEISEC) i, r l 1 LO j . Revision 0 Page 84 l l; L. l .
EA-PT-91-0003-M ! n w 7.0 COMPARISON TO PEACH BOTTOM UNIT 2 TRANSIENTS . As a'part of the methods qualification, GSU calcu-lational results were compared with experimental data from Peach Bottom Atomic Power Station, Unit 2 (" Peach Bottemh). These comparisons demonstrate the validity of GSU-methods for analysis of -fast pressurization tran-sients. 7.1 BENCHMARK DESCRIPTION . Three turbine - trip transient experiments were . performed at Peach Bottom in April 1977'at the end of , Cycle 2".. I'he purpose of thene tests was to' investigate ,
- the effect on-neutron flux of-a pressure transient in the reactor core following:a turbine trip. .The three tests were identified as Tests TT1, TT2, and TT3; four stabili-1 ty -tests were also performed during this period.
Additional instrumentation was installed to gather more ' detailed information than is normally available and some t normal protective features _were disabled to allow a more
- severe transient'than would normally be expected. The-O Revision 0 Page 85 r-a- e w , -- <-+-&-, ,r- --- --+-wr-,--e --*-,-r~re*---,--twrw++-,*--r-a.,--ww-*e--vie.-*--r- ,w*-r=wm-r--we~*--e-er--'- wv----
~ . . _ . . - _ _ _ . _ _ _ - . _ . _ . _ _ _ _ _ _ . . _ . _ . . _ . _ . _
EA-PT-91-0003-M O results ' of - ~ these three tests are useful for reload analysis.methodo qualification. The tests were each initiated by manual turbine trip. The normal reactor scram on turbine stop valve position was disabled, allowing a higher power spike consistent with design basis analyses. Reactor scram was
-initiated ' by high neutron flux at a reduced .setpoint.
_ condenser bypass was allowed to initiate normally on turbine stop valve position. In addition to the EPRI-sponsored tests, the Peach Bottom benchmark knalyses also co.ared a License Basis + Transient (LBT)-problem developed by Brookhaven National i
. Laboratory u to . test the capabilities of transient computer codes to evaluate conditions more severe than -those observed in the tests.- Because the postulated LBT-transient' required . assumptions outside the . normal. GSU -methodology, the LBT benchmark-is included as-a demon- -l stration of capabilities rather than methods.
.i . l O Revision'O Page 86
1 1 EA-PT-91-0003-M l 7.2 PEACH BOTTOM RETRAN MODEL The RETRAN model of Peach Bottom was developed with the same techniques as were used for River Bend. A nodalization diagram for the Peach Bottom system model is shown in Figure 7.1. Geometry. and configuration were - obtained primarily from the EPRI report" as supplemented by other sources",", Since the Peach .Sottom benchmarks- only cover a limited range of f ast pressurization events, the control systems are much simpler than those in the River -Bend a model, which cover all enticipated operating modes. The Peach Bottom model includes simple controls to close the. , turbina stop valves,.to initiate scram on either time or high power, to open the bypass valves, to adjust bypass valve and turbine stop valve loss : coefficients as a
- function. of position, and to' set feedwater flow and 7 enthalpy.
1 L l L .C)? h- Revision 0 Page 87
-.----._,---.---...-_,.~,_.......,_,,.__,,,,,,,,-,,.,._.,,n,,.._.-,.-,,.,, . ,,_..,..,,,,,,.,n.,,,,.~,...,,,.,,...,,,,,n. ,,.,.,.,_,e s
I t EA-PT-91-0003-M J o 7.3 TURBINE TRIP TEST SIMULATION
.)
This section describes the analysis of the test data obtained during the th' tee tarbine trip tests. i l 7.3.1 Initial conditions Initial statepoint conditions are determined from published data. RETRAN . initialization requires addition- , al. data, . sufficient to ' calculate initial conditions , everywhere within the system. RETRAN initial conditions for Tests TT1, TT2,-and.TT3 are summarized in Table 7.1. ! Plant Data. Actual plant data available includes
. core power and flow, done pressure, - steam flow, water - level', recirculation flows and recirculation'and jet pump specifications. - This'intormation-was obt ined from,the literature" H " or' from test data : available on floppy disk-through EPRI. A steam-flow-feedwater temperature !
relationship and a core flow-recirculation flow relation-: , ship were determined from the'available plant data. [ O - Revision 0 Page-88 I y w. + a *v w-+ ,. e*cw, w.m.,-n-+=.e-<-r-me- --+nr w..--.e-r -.,,v.-cw--e =,-.----e .--re.- - , , = , . . ...,-.,.e.~+---,w-w-.--.s -, e4.- -r ,e
EA-pT-91-0003-M O Calculated Data. Core bypat,p flow and loss er i-cients were calculated by FIBWR. REBAL was u to determine the remaining necessary initial conditions. Core neutronics data for each case was calculated by SIMULATE-E and processed f or input to RETRAN by SIMTRAN. The methodology used for this data transfer in described in Section 4.0 and in Reference 1. Initial values f or the control system were calculat-ed for each case and 10 second null-transient runs were made to verify that the system was initialized to a self-consistent steady state. . O 7.3.2 Transignt Modelina Several parameters are case dependent. These are generally items that control event timing, specify trip setpoints, or control boundary conditions. They are items that were set by the plant operators for the tests, or that were measured during the tests. However, separator inertia is condition dependent, specifically of separator inlet quality. Thus it varies with each case to the extent that quality varies. O Revision 0 Page 89
l EA-PT-91-0003-M G The high flux scram setpoint was set at the plant dif ferently for each test. Alternatively, the actual scram time as recorded can be input. Rod position as a function of time, and hence, speed, was also recorded. Turbine Stop Valve (TSV) and Bypass Valve (BPV) positions were recorded during each test. Changes in feedwater flow and recirculation pump speed were not considered significant during the period modeled. These were assumed constant. Separator inertia is determined independently for each case from General Electric ODYN Qualification . Report". The inertia is a function of inlet quality and does not effect the initialization of RETRAN. Best results were obtained by applying half the total inertia at the separator inlet, half of that part attributed to the standpipes at the standpipe inlet, and the remainder to the separator liquid return junction. The inertia of the steam outlet of the separators was based on the geometry of the steam region in the separators. The area of the steam region is calculated from the geometry of the separators and the water layer thickness obtained from General Electric ODYN Qualification. 1 r b Revision 0 Page 90
--w ____ _ _ _ _----_ _ ___-_ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ . - - _ - _ - _
1 EA-PT-91-0003-M O - 7.3.3 Simulation Results 4 The results of the Test TT1 benchmark are shown in l Figures 7.2 through 7.6. The results of the 'lest TT2 benchmark are shown in Figures 7.7 through 7.11. The results of the Test TT3 benchmark are shown in Figures 7.12 through 7.16. Each set of figures provides predic-tion versus measurement comparisons of core average 3 power, upper plenum pressure, dome pressure, TSV "A" pressure, and core inlet flow, calculated at the jet pump exit. .. The power comparisons show general agreement vith A the published data - in both timing and magnitude. The ! Test TTI simulation (Figure 7.2) shows an.overprediction Lof = the magnitude of the power peak, but the timing-of the , peak and the initial power ' rise are -predicted very closely. The Test TT2 power-spike (Figure 7.7) is in very close' agreement with the' data. Although the power trace follows the data closely, the predicted Test TT3 ! - power peak (Figure-'7.12) -is later and slightly lower than L the data. - The greater width'of'the predicted. impulse i-indicates'it is conservative relative to the data, i i l. Revision 0 Page 91 , re er e- sa P-e-wwe- N ew-a=+we w ww --wvw-w--w-ww-twwv--w-w r-e+ m h t 4-
EA-PT-91-0003-M v The measured data for upper plenum pressure exhibits some fluctuation. The Test TT1 simulation (Figure 7.3) overpredicts the upper plenum pressure af ter the initial pressure surge is stopped. The Test TT2 (Figure 7.8) and TT3 (Figure 7.13) predictions follow the data trends closely. The dome pressure data were also smoothed for comparison with predictions. The TT1 analysis (Figure 7.4) overpredicts the data after the initial pressure surge. As with the upper plenum pressure, the Test TT2 (Figure 7.9) and TT3 (Figure 7.14) follow the data trends , closely. p As was the case with the other pressure data, some '] scatter in the TSV pressure data made smoothing necessary before reasonable graphic comparisons could be made. The predictions for Test TT1 (Figure 7.5), TT2 (Figure 7.10), and TT3 (Figure 7.15) all track the smoothed data. The " measured" core flow data for the three tests were calculated from recorded jet pump differential pressures. The recorded signals show a large amount of noise, which wcs first filtered. The average of the four filtered signals was then used in the solution of the fundamental differential equation relating flow to O V Revision 0 Page 92
EA-PT-91-0003-M i
.Q-V differential pressure and inertia. The resulting flows were plotted with an additional time adjustment of 0.25 l
s. to account for the process instrumentation delay )
discussed in Section 9, Reference 12. The predictions l for Test TT1 (Figure _7.6), TT2 (Figure 7.11) , and TT3 ! (Figure 7.16) - show similarity of timing of the initial l rise and of peaks and valleys, though the magnitudes of >
'the changes are'.different. However in no case does the 5
mismatch exceed 84 of the initial value of flow. The predicted behavior of the core inlet flow is consistent with the phenomenology.- . 7.4' LICENSE BASIS TRANSIENT MODELING This section. describes the demonstration analysis i covering the Peach Bottom License Basis Transient. k 7.4.1 Initial conditions Initial. conditions for the LBT analysis. were
. generated to be consistent with the' Brookhaven analysis.
Cross sections for this analysis were generated from a O Revision 0 Page 93
...;. _ .__._ . .__ _ a_ ._- _ . _ . _ - _ _ ._ _ _ _._.. _ .- .._
EA-PT-91-0003-M O
\v/
two-cycle Haling depletion rather than the stepwise depletion used for the turbine trip test benchmarks. The original Haling distribution was not nearly as bottom peaked as that used by Brookhaven. To obtain better agreement with this power shape, the Haling depletions were re-run using a lower value of subcooling and a larger albedo value for the lower reflector. RETRAN initialization was accomplished consistently with the turbine trip test analyses. Initial conditions for the LBT analysis are summarized in Table 7.2. h'A 7.4.2 Analvtical Results The analytical results of the License Basis Tran-sient benchmark:are shown in Figures 7.17 through 7.25. Comparisons are made to GE and BNL initial conditions of axial power shape, fuel temperature, void distribution, and clad surface heat flux. Transient comparisons to GE and BNL results included power, core pressure, core flow, and axial clad surface heat flux. The calculated axial power distribution (Figure 7.17) is in good agreement with both the GE and the BNL l O Revision 0 Page 94 l
EA-PT-91-0003-M , O' v curves. The calculated initial fuel temperatures (Figure 7.18) agree well with the GE data. The initial heat flux (Figure 7.19) is a little lower than both the GE and the BNL data. Since the power and the power distribution are the same, this difference can probably be attributed to a difference in the type of fuel modeled which would lead to a difference in heat transfer area. The calculated initial void distribution (Figure 7.20) is higher than both the GE and the BNL curves. This may be due to differences in void models or in core bypass flow. The calculated. transient power-(Figure 7.21) curve - is higher and narrower than the GE curve, and both higher
.and wider than the BNL curve. The area under the calculated curve appears'to be similar to that under the GE curve, indicating a similar total energy release.
I Since CPR calculations are more sensitive to the total energy of an impulse than to the height'of the impulsa, the calculated results are not unconservative. The transient pressure calculation (Figure 7.22) shows. good agreement to both the~GE and the BNL curves.
The calculated core inlet flow (Figure 7.23) shows an initial rise which appears to be due to the initial scram induced power decrease. This flow surge is turned around l
\ Revision 0 Page 95 l l
~
l l 6 i o .. ._ _ _ _ . _ _ _ . _ _ . _ . _ . . _ . _ _ .
EA-PT-91-0003-M r~ *
!]T in both the calculated and the GE curves by the power impulse .due to void collapse. Since the GE impulse occurs first, the GE flow decreases first. By about 1 second, the calculated flow appears to be following a similar trend to the GE data, about 5% lower.
The clad surface heat flux at 0.8 and 1.2 seconds (Figures 7.24 and 7.25) is similar in shape to both the GE and BNL curves, reflecting the changes during the transient due to scram and void collapse.. l l l l l l
/
Revision 0 Page 96
l l EA-PT-91-0003-M l I (~, 1 \ N-- Table 7.1 Peach Bottom Turbine Trip Test Initial Conditions TT1 TT2 TT3 Power, MW. 1562 2030 2275 Core flow, lbm/s 28139 23028 28306 Dome pressure, psia 991.6 976.1 986.6 Core inlet enthalpy, Btu /lbm 528.0 518.3 522.7 Water level, inches 23.0 29.2 29.8 Steam flow, lbm/s 1638 2171 2470 Recirculation flow, lbm/s 9090 7652 9378 APRM high power trip, % rated 85 95 77 k (_) l I f- . Revision 0 Page 97 l l L
EA-PT-91-00C3-M
^
f ^'3
\J Table 7.2 Peach Bottom License Basis Transient Initial Conditions Power, MW. 3441.2 Core flow, lom/s 28472.2 Dome pressure, psia 1034.
Core inlet enthalpy, Btu /lbm 522.7 , Water level, inches 28.0 Steam flow, lbm/s 3900 Recirculation flow, lbm/s 9500 ; I^h. L si Revision 0 Page 98 l l t -
l EA-PT-91-0003-M O - Figure 7.1 Peach Bottom RETRAN Model Nodalization II l
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4. EA-PT-91-0003-M f&[ . Sb Figure 7.7 Predicted vs Measured Core Average ' Power, Peach Bottom Test TT2 PEACH BOTTOM 2 E00 2 TURBINE TRIP 2 o i i i i
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Predicted vs Measured Core Inlet' Flow, Peach ~ Bottom Test TT2 PERCH BOTTOM 2 E0C 2 TURBINE TRIP 2 8 _. @ g g i i i- i 1 1 1 .1 o _
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O O O MEASURED CORE AVERAGE POWER 2 s.o s.o 7.0 l' O.o 1.0 2.0 3.0 9.o i i 1n 5 i i i i e e *
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Figure 7.14 - Predicted vs Measured Dome Pressure,- Y Peach Bottom Test'TT3 , 4 e PERCH BOTTOM 2 EOC2 TURBINE. TRIP 3 O,. O, :
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TIME Revision 0- Page 114 i - {. : ll . l .. . . .
u- , EA-PT-91-0003-M I rs4 g l i Figure 7.17 Axial Power Distribution, Peach Bottom License Basis Transient BWR Licensing Basis Transient Relative-Power @ t- = 0.0 1.6 1.4 - - [' ~ ' . ,
~.
1.2 - - - 3 .< ..
- ym j ~ ., y ij . Relative Power 0.8 --l '*}.
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. . . . . . . BNL 0.4 -- + RETRAN . 0.2 - -- '
O l l l l
-0 0.2 0.4 0.6 0.8- -1 Fractional Core Height Revision 0 Page 115
l EA-PT-91-0003-M l
-Figure 7.18 Fuel. Temperature-Distribution, Peach ~ Bottom License Basis Transient BWR Licensing Basis Transient -Initial Fuel Temperature 2000 1800 - - ..***..,
[ ', 5 1600 - - l- ' 1400 --l 1200 - -
.e..
( Average Fuel- , Temperature (F) 3, , 800 - - GE 'o
*BNL. ----*-- R ETR AN -
40o . . 200 - - 4 0- l l l l 0 0.2 . 0.4 0.6 0.8 1 Fractional Core Height l I . Revision 0 Page 116. ,
EA-PT-91-0003-M Figure 7.19 Initial Heat Flux Distribution, Peach Bottom License Basis Transient BWR Licensing Basis Transient Clad Surface Heat Flux @ t = 0.0 1 8 7 . j' [ ,' ., ,
5--;
L Clad Surface - Heat Flux '~
/ (Watts /m sq.
L(']v- x <10E 5)- GE 3- -
. . . . . . . BNL - ^
RETRAN t
.1 - -
\. 0- l l l l l l> 0 2 4 6 8 10 12-- Axial Distance from Bottom (ft.) 1 Revision 0 Page 117
EA-PT-91-0003-M 4 Figure 7.20 Initial Void Distribution, Peach Bottom License Basis Transient 4 BWR Licensing Basis Transient Initial Void Distribution 0.8 0.7 - - g 7 ......
. 0.6 - - /y /
[,..*,,..,...a***.......
O.5 - - .
Average Void 0.4 - - / GE
n. ~ Fraction . '
,.. ......* BNL 0.3 - - ,.. "'- RETRAN . 0.2 - - ,. ~^-- NEUTRONIC VOID 0,1 - -
l l' l. l 0 0.2 - 0.4 0.6 0.8 - 1 ; Fractional Core Height i ti Revision 0 Page 118 o l L
i . I . l l-BNL. CORE AVERAGE POWER E . 0.0 1.0: 2.0 ' 3.0 ~ 9.0 5.0- 6.0 7.0' 8.0~ .' . I d m i 1 I- 1 I -1 1 I $-
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i et O u O- 8. .
I B *tf f' CD -8
'T _ --t - . t* S-m -
r e . og -O O P y e t-
=
g 3 .o i g W O s 9 -O i
.y u p 7 I I I 1 1 I '
a O 3 i i _ ,=...,s _.- -, _ . . . . . . _ - ..
.- - . . . . . . . . . - . . ~ . - - . . . . . - . . .. . .. . . . - . . - - . . .
7 EA-PT-91-0003-M
/%).
a* ' l Figure 7.24 Heat Flux Distribution @~t_= 0.8, Peach I Bottom' License Basis Transient BWR Licensing Basis-Transient Clad Surface Heat Flux-@ t = 0.8 8 i 7- -
; ..... ..., t 6- - }
5-- % , Clad Surface N-
~ Heat Flux .
t
; (Watts /m sq,4 ~.,
E - ; x 10 .5 I 3 .. GE
. . . . . . . Bu t ----o-- RETRAN l \
1
t. ,
l' oi ; ; ; ; ; l' O 2 4 6 8 10 12 Axial Distance from Bottom (ft) 1 l J- ) L -Revision 0 Page 122
, - . .. .. . -.. -. ... . . = . - . - _ . . . - - - - .
1 EA-PT-91-0003-M !
. ,= 'u- Figure 7.25- Meat Flux Distribution 9 t = 1.2, Peach I . Bottom License Basis Transient BWR Licensing Basis Transient !
Clad Surface-Heat Flux @ t = 1.2 f 8 ' W . =.... 7-- - _ ,_. '/ . 6- .l
\ ., s . ., y N
6--, '. . Clad Surface '. Heat Flux \. p) (Watts /m 'sq. 4 ~
~
GE.
L/ x 10'.5) . . . . . . . Bu t 3- -
+ -- RETRAN 2- -
1- -- 0 .l l l l l' O 2 4: 6 8 10 _12 Axial Distance from Bottom (ft) t t n O "-i 1 " "" ' '
>:3',-i j j LEA-PT-91-0003-M c
,rq 8.0 HOT CHANNEL MODEL- ./ _
i
'The overall system transient response,is calculated -
with the RETRAN computer code. In the system model, the core is_ represented by-a. single, average power channel. Tha results of:the system calculation neutron- flux, , core pressure, core flow and inlet enthalpy as a function' of ~ time during the transient are used .as boundary: j I conditions _ for the RETRAN hot - channel model.- The hot channel is executed: separately from the systems model to
' allow a different model= to be developed for each mechani-cally.uniqueifuel design in the RBS core and to facili- - tate changes. to _- initial- conditions such as radial and ;
V axialEpower distribution, gap conductance,-etc., which
- may(differ-from those_used for the system-(core average). . calculation- .-- - The RETRAN hot channel model geometry represents the-
" ~ 1 in schannel : portion. of one - fuel - bundle and the. upper - ; plenum. Figure 8.1=ista.noda'lization-diagram'of-the RBS
, hot channel model.' This model consists of 27 volumes, 27 - ' junctions, 25 c conducting- heat exchangers and 25 non-conducting heat exchangers.- The non-conducting heat t
Revision:O Page 124
EA-PT-91-0003-M s fW V exchangers model the direct power deposition to the
~ coolant. 'The forcing functions (boundary conditions)- for this model come from the RETRAN systems transient calculation. ,
The hot-channel rod conduction power is. proportional to the . transient neutron flux; as is the power fraction directly deposited in the' active coolant due to. gamma -
' heating and neutron slowing down. The upper plenum in 'theshot channel model is a RETRAN time-dependent volume, set to the pressure versus time from the system calcula-tion. -The inlet.to the hot channel is a RETRAN positive .
fill, so;that the inlet flow and enthalpy may be sp6ci-g fied.- The-inlet enthalpy is ' taken . directly f rom the i,j : systems -calculation;- the core flow from the _ systems
. calculation _isLmultiplied by the fraction of the total flow.whichnenters the active--zone of the hot channel.
e The1 flow distribution fraction-is calculated by;a response -surface fit to a series 1of parametric FIBWR core hydraulic > - calculations . . FIBWR . calculates the multi-channel'-inlet flow distribution, including leakage flows 1
- and water _ tube flows. The independent parameters of this:
responsetsurface= fit are core power, flow, pressure and-radial--peaking fac' tor. - The core flow distribution was
-Revision 0 Page 125
4 - ,( - EA-PT-91-0003-M M d found to be insensitive to. the calculated change--in inlet enthalpy.- By using _ an inlet flow forcing procedure, ; rather' than a plenum-to-plenum pressure drop forcing function, the potential for inconsistencies between the
~
systems _model and the hot channel model from differences
' in pressure drop models, flow regime options, nodaliza-tion, inertia and time step selection are minimized.
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 represen- ..tative value - of 2315 BTU /hr-f t ..F was used in the hot 2
channel analysis.-- This value was determined by the RBS
' ESCORE model to be typical of high-powered assemblies of
_ N] 5. the current fuel- design which may be on thermal limits at - end-of-cycle conditions. In accordance with the GETAB" procedure for B6R/6, a standard 1.4 middle-peaked axial power distribution is used- for hot channel analysis. Sensitivity studies have shown that ACPR is somewhat dep'endent on-axial power shape, with_a top peaked-power distribution yielding a l higher ACPR.- However, because a
' high power assembly will . also _ have high _ voids, the probability is low that an assembly near thermal limits will be peaked.above.mid-core.
Revision'O Page 126
.., - . . ~ . . .. . . . - . - . . . . . ~ . .-.. . . ~ - - - - - . - . . . . - - - -
c ,
EA-PT-91-0003-M w
\' : Output from1the' RETRAN hot channel model >is; pro-i i
cassed by EDTRAN to determine ACPR.- This calculation is -
,desc'ribed in-a separate report, ) i '; [j ' - )
t i ll l l. l:
;v- ,
I - .- I lb ' .i
' Revision'0 .Page 127 i i . _ .. _ --
d EA-PT-91-0003-M O i V Figure:8.1 RBS RETRAN Hot Channel Hodalization
%1 A 210 -t 126 i "! 125 $N[2 " 1 YN'h, 23 s j ,,24 3 ggg, N
l 122 hkh
! 121 @M2 Nl 120 hkh * 'l 119 $$fA N
II N 118 hkkh . w 137 gg/g *
, 1*Nl Nm 116 $ @f j$3 Kjfj'
(' 14 114 $$g l l 27 Volumes
$ 3 ~. 333 gg " 112 N[d pf/f] 25 Conductors "l 111 kkd N
9 110 Mf/) A
%s g /7gfj 27 Junctions .
7 "l 100 k[// 0 3o7 ([//g A 25 Non Conducting N' 106 f/8[/, Heat Exchangers 4 Nw 105 @[/A jo4 fggy L 2 s, 103 (f8h
,,, gg4 " 101 fIf/) /
g k101 i Revision 0 Page 128
" n+ *e-'-"e' e--wre- v -- w W w P-+TW-F 9 w -- e -r *s - -m ,-eiw,ee+w n'-w+a' -+n-i w w--P==ts*-- -+*--+-t:=e-im!-vt+- r'
EA-PT-91-0003-M p
9.0 REFERENCES
1. C.H. Greene, et. al., " Steady State Core Physics Methods: for BWR Design and Analysis," EA-CA 0001-M, Gulf States Utilities Company (1991).

2. A.F. Ansari, et. al., "FIDWR - A Steady-State Core Flow Dis,tribution Code for Boiling Water Reactors,"
EPRI NP-1923, Electric Power Research Institute (1981).
3. A. P. Ansari, et. al., "FIBWRt A Steady-State Core Flow Distribution Code for Boiling Water Reactors, Code Verification and Qualification Report,", EPRI NP-1923, Electric Power Research Institute (1987)

4. J. A. McClure and G. C. Gose, "SIMTRAN-Et A SIMULATE-E to RETRAN-02 Datalink", EPRI NP-5509-CCM, Electric Power Research Institute (1987).

5. I. B. Fiero, "ESCORE: The EPRI Steady-State Core !
Reload Evaluator Code", EPRI NP-4492-CCMP, Volumes I and II, Electric Power Research Institute (1986). l 6. B.J. Gitnick, " Consistent Initialization of Pres-
\ sure and Flow Parameters for BWR Transient Analysis at Off-Rated Power and Flow Conditions," Nuclear Technology, January 1991, pp. 92-104.
7. J.H. McFadden, St. al., "RETRAN A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," EPRI NP-1850-CCM, Volumes I, II, and III, Electric Power Research Institute (1981).

8. C. S. Brennan, "EDTRAN: A Program for the Editing, Linking and Post Processing of RETRAN Computer Runs", PSE&G NFU-0090, Public Service Electric and Gas of New Jersey (1988).

9. General Electric Company, "GEXL-PLUS Correlation Application to BWR/2-6 Reactors, GE6 through GE8 Fuel", NEDC-31598P, (1988), Proprietary.
Revision 0 Page 129
i EA-PT-91-0003-M
10. D.A. Prelewicz, et. al., "BWR Water Level Model-ing," Proceedings: Sixth International RETRAN Conference, Electric Power Research Institute L(August 1990), p. 15-1,
~
11. J.S. Miller, et. al., "RETRAN Simulation of a BWR/6 Load Rejection Transient," Fifth Proceedings of Nuclear Thermal-Hydraulics, ANS Winter Meeting (November 1989), pp. 350-358, 12.- L.A. Carmichael and R.O. Niemi, " Transient and Stability Tests at Peach Bottom Atomic Power Sta- -
tion Unit 2 at End of Fuel Cycle 2," EPRI NP-564, Electric Power Research Institute.(1978).
13. M.S. Lu, et al., " Analysis of Licensing Basis Tran-sients for a BWR/4," BNL-NUREG-26684, Brookhaven National Laboratory (1979).

14. K. Hornyik and J. A. Nasser, "RETRAN Analysis of the Turbine Trip Tests at Peach Bottom Atomic Power ..
Station Unit 2- at the_End of Cycle 2", EPRI
NP-1076-8R, Electric Power Research Institute (1979).
.15. A.M. Olson, " Methods for Performing BWR Systems Transient Analysis," PEco-FMS-0004, Philadelphia-Electric Co. Par.y (1987).
l16. -" Qualification of the One-Dimensional Core Tran-sient: Model for Boiling Water Reactors," NEDO-24154, Volume I, General Electric Company (1978). 17 . - " General Electric Thermal Analysio Basis; Data, Correlation, and Design Application," NEDO-10958A, General Electric Company (1977). i Revision.O Page 130-(
A*..,e - -.a< ~LA.. r+..- -m%->4,- _ e f4,2-. 4 .u J . - . _ +4*,4- e -mA + ".M= maJ.l-4 a_Ae#.a3 Ea.JJm ah.-ah4 .*_.A.e 2 - A .m.LA--a_a .A_.Aa.uam.,__a_s%a,. EA-PT-91-0003-M O I A APPENDIX A CALCULATION OF FUEL ROD GAP CONDUCTANCE k Revision 0 Page 131 _ . - . , , . - - , , _ , -.-_.,-.. .-_ _ _,,-,., _ __..__- 7 - . , _ .,-.. _. _,,,. y,.
i EA-PT-91-0003-M
,/^
List of Tables A-1 Example'of RBS Cycle Dependent Power Distribution Used to Determine Average Gap Conductance . . . . . . . . . . . . . . . . 145 List of Figures A-1 Gap conductance contribution by ruel Rods in the 5-6 kW/ft Range . . . . . . . . . . . . 147 x l Revision 0 Page 132
4 l EA-PT-91-0003-M I CALCULATION OF FUEL ROD GAP CONDUCTANCE Thls appendix describes the use of the ESCORE computer program to calculate fuel rod gap conductance for use in RETRAN system and hot channel analyses. The following sections describe the preparation of ESCORE cases. c 1.0 ' ESCORE CALCULATION METHODOLOGY This section describes the use of the ESCORE comput-er program to evaluate fuel rod thermal-mechanical ef- t
. facts. The following sections describe the selection of s axial power shapes and the determination of the remainder of.the ESCORE input.
1.1 AXIAL POWER SHAPE IN ESCORE CALCULATIONS c Fuel rod gap conductance: is axially variant, and each nodal value is slightly dependent on the other nodes in the fuel-rod _via the gas gap temperature. While an
'l Revision 0 Page 133
t EA-PT-91-0003-M axially dependent gap conductance could be calculated in RETRAN, thermal-mechanical coupling between nodes is not 3 available in the current coding. A utility RETRAN study demonstrated that an axially averaged gap conductance produces nearly the same total heat deposited in the coolant as a function of time as the axially variable gap conductance for typical pressurization events. Selection of conservative gap conductance values for the core average application makes a greater dif ference in the transient consequences than the use of the axially averaged value. ESCORE calculated values of axially , averaged gap conductance are used in the RETRAN system p model. Another study in Reference 1 determined the effect of using an axially averaged gap conductance in the RETRAN hot channel model. For various axial power distributions examined in the study, the axially averaged gap conductance produced equal or higher calculated ACPRs than the axially variable gap conductance. An axially averaged gap conductance is therefore used for the hot channel analysis, since higher values of ACPR indicate more severe thermal margin effects during transients. Revision 0 P=1ge 134
' l c
EA-PT-91-0003-M l i
/
1.2 ESCORE INPUTS
'This section describes the preparation of input for the ESCORE gap conductance calculation. Separate input streams are required for system modeling and hot channel modeling because the conservatisms required for each application of the gap conductance are different.
Input parameters which are well characterized physical properties (such as yield strength) are taken at " established, nominal values. Parameters which-are. controlled Within fabrication ,. tolerances are also taken at nominal-values when used in the ESCORE analysis. Use of nominal values is appropri-ate because both the core average and hot bundle calcula-tions represent statistical evaluations of a large number of fuel rods and pellets; the mean values of these parameters will be the-nominal. values for the material lots. - These parameters include grain size, pellet O.D. , cladding I.D., and rod. pre-pressure. Other parameters which are not as strongly.charac-terized- and do not strongly affect the results as l
. demonstrated by parametri c analyses performed.by othera are'also' set to their nominal values. These parameters ' Revision 0 Page 135
s ; EA-PT-91-0003-M include-fuel pellet densification, fast flux to Linear Heat Generator Rate (LHGR) factor, system pressure, and resonan'ce escape probability. Axial power distribution and power history strongly affect the results and are chosen at conservative values. 2.0 POWER HISTORY IN ESCORE CALCULATIONS Fuel rod thermal-mechanical effects are strongly dependent on the irradiation history-of the fuel rod. , These effects include pellet phenomena (densification, cracking, swelling, and fission gas release), cladding phenomena (embrittlement, creepdown,-and pressurization), and ' interaction phenomena (pellet-cladding . interaction and axial stack growth). E 2.1 CORE AVERAGE GAP CONDUCTANCE The core average ' gap ~ conductance. determines how , changes in total core power causes changes in heat . deposited to- the coolant.- The high power bundles K 1 Revision ~_0 Page 136 u L l l-ll 1 - , .6,-u..-,_..._,--_ . . _ . . . , . -- _ - . . . . . _ . . . . . . . _ _ . - . _ . . . . , _ _ . . . . . _ . _ - , _ _ . . - _ . , .
EA-PT-91-0003-M i V contribute mota to the core thermal response than low power bundles because high power bundles generally have higher gap conductances than low power bundles. Using all of tae bundles in the core and number weighting the bundles underestimates the core average void response. The gap conductance of an individual fuel rod generally increases with burnup because the width of the gap decreases as the fuel pellet swells and the cladding creeps down toward the pellet. This effect is offset to some extent by decreased conductivity of the fill gas caused by the release of fission products from the , material lattice. On a core average basis, pellet cladding contact does not occur and the over prediction
,O" of contact conductance by the ESCORE code is not a concern.
The core average gap conductance is based on a statistical weighting of the number of fuel rods oper,at-ing within a certain power range at various burnup points during each cycle for the various type fuels in the core (i.e. fresh, once burned, and twice burned) . The rod average gap conductance is determined for each power interval and then weighted by the fraction of the rods in this interval. Corevide power and exposure distributions Revision 0 page 137 1
/
i EA-?T-91-0003-M L-for this analysis are determined in the nuclear design 2 analysis.
-in initial burnup interval of 0-3.7 GWd/T is used because this is near the end point of densification where the minimum gap conductance value would be attained.
Additional calculations cover subsequent burnup incre-ments of approximately 4 GWd/T to a maximum exposure of' , approximately 12 GWd/T for the fresh fuel. This defines four statepoints from beginning to end of cycle. Table A-1 presents the power and power history information necessary to determine the core average gap conductance. .; The gap conductance value at each kW/ft value within (~ a given statepoint box is determined for each burnup t interval or statepoint. This gap conductance value is. ' then weighted ~ by the relative number of rods in each , increment to define an overall average value for a l statepoint box. That is, considering the fresh fuel, the burn' history, as the calculation proceeds from left to right, is made by assuming that the fresh fuel burns at the-average value of all the fresh fuel in the core. j g This procedure is then repeated in burning the history
.from statepoint 2 to statepoint 3-at the average value L between each statepoint. In the case of the once burned .
p t Revision'O Page 138 [
. _ - . . _ _ , _ _ . - - . .- . _ - _ _ - - . . - - - , . . + . - . - - - -e--, -......m.---.,-,4-. . . - . . . - . . , -
EA-PT-91-0003-M
,/mq k) fuel which has a prior burn history before the start of statapoint 1, it is assumed that the prior history is the average obtained by the fresh fuel which was in the previous cycle. The once burned fuel is then burned from statopoint 1 to statepoint 3 at the average value of all the once burned fuel between each statepoint. Similarly, in the case of twice burned fuel which has a prior burn history before the start of statepoint 1, it is assumed that the prior history is the average obtained by the once burned fuel in the previous cycle. The twice burned fuel is then burned from statepoint 1 to statepoint 3 at .
the average value of all the twice burned fuel between each statepoint. Finally at the end of each burn point, (] the power level is stepped to the particular power level of interest. To further illustrate this point a typical core average gap conductance power histogram is shown in Figure A-1 for fuel rods in the 5-6 kW/ft range. i Using the increment-averaged power for the state-point to statepoint power history and prior cycle averages for the once burned and twice burned fuel results in a very smooth power history for the analysis. Path-dependent effects are minimized by this lack of l power fluctuations, resulting in a conservatively low i Revision 0 Page 139 [V} l w
d l EA-PT-91-0003-M i l l l l kJ value of the calculated core average gap conductance. This entire analysis was performed using an end of cycle bottom peaked Haling axial power profile to establish the trend in core average gap conductance with burnup. Since the trend revealed increasing gap conductance with burnup, the calculation was not performed for statepoint 4. 2.2 Hot Channel Gao Conductance The power history in the hot channel gap conductance calculation is biased to maximize pellet and relocation cracking without excessive fission gas release. Tho , bundle is modeled as a single fuel rod. The rod operates with the peak node at the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit for the first .001 GWd/T using a top peaked axial power shape. The rod then operates with the peak node at the MAPHLGR limit for the period .001 .002 GWd/T using a middle-peaked axial power shape. This power history produces substantial pellet cracking and relocation without significant fission gas release. The LHGR is then reduced to the bundle average value and then maintained constant to the second statepoint. The power l is then maintained at the average value between state-l ( Revision 0 Page 140 L 1
EA-PT-91-0003-M l hq.- point 2 and 3 and so on until the bundle reaches the end ' of cycle. The axial power profile at the beginning of the stritepoint is used across the burn interval and is then changed to the axial power profile at the next statopoint. This procedure is then repeated until statepoint 4 is reached. Although-the bundle average gap conductance for the once and-twice burned level may be slightly higher than that obtained for the fresh fuel, it is the fresh fuel that operates at linear heat rates that are of concern for thermal margin evaluations. Only the fresh fuel is . considered for determination of the maximum gap conduc-Lance. Several potential hot bundles are-tracked across O the burn cycle to ensure that the maximum gap conductance ; bundle has been identified. 3.0 ESCORE TECHNICAL EVALUATION CONCERNS This'section' describes the use of the ESCORE program for gap conductance analysis. The text specifically
-addresses technical concerns voiced during the ESCORE regulatory review 2, Revision 0 Page 141 's.
1 EA-PT-91-0003-M 3.1 Fuel' Rod Radial Power Distribution Using extreme void distributions (such as a constant value of 0% and 80% along the entire rod), the code was , allowed to calculate the resonance escape probability, P.
- The effect on the predicted gap ' conductance was very small (<3%). Any uncertainty introduced by the resonance - escape- probability is more than accounted for in the assumed gap conductance uncertainty of 25% _that was used in the ACPR calculation.
u I , 3.2 Power-to-Melt and Auxiliary Power Calculation
- These features of the ESCORE computer code were not used in determining the core average and hot channel gap s
conductance. ,
. 3.3 Fuel Red Temocrature Distribution.
In this . area, the overprediction of . pellet clad contact conductance was addressed. Based on the proce- ' l l Revision 0 Page 142
#vr-'ge .- u w-.E 5, , wr ..way. wu, .+.,.,-c...-%,, , .y -.rww.,, ,www- r - m e -wer-w e** m s t s . m ' w v e- e m m vw'd- -r = N ,-=evw tteu ,+W+-w+ =-w-y- vi'w~.-- r v m n'-- - - - * * -
EA-PT-91-0003-M ("s () dures used in the determining the core average and hot bundle gap conductance, the fuel was not predicted to experience pellet clad contact; although in some cases the predicted hot gap width was as small as 0.15 mil. The overprediction of gap conductance due to pollet clad contact conductance is not a concern for this ESCORE application.
4.0 REFERENCES
1. C.E. Dodge, " Application of Reactor Analysis Methods for BWR Design and Analysis," PL-NF-90-001, Pennsyl-g~s vania Power & Light Company (1990).
b) (
2. A.C. Thadani, " Safety Evaluation of Topical Report EPRI NP-5100, 'ESCORE - The EPRI Steady-state Core Reload Evaluator Codet General Description,'", U.S. Nuclear Regulatory Commission (1990) .
l l l l l l [] Revision 0 Page 143 l v l l
.- .- . . . . - . - - - . . - ~ - . - - . - - . - _ - ~ . . ~ -
i EA-PT-91-0003-M
\ .
Table'A-1 Example of RBS Cycle Dependent Power Distribution
Used-to Determine Average Gap Conductancel 4
7 Fresh Fuet
- Rod Power SOC NJeer of Rods (DC (kW/ft) $tatepoint 1 Statepoint 2 Statepoint 3 statepoint 12 232 0 0 0 '
2*3 1144 80 0 0 3*4 64 368 0 0
.4*5 368 1072 240 0 $*6 1136 1t28 1736 1832 6*7 2376 1848- 2376 1744 7*8 2304 2728 3040 4568 8*9 2848 4340 4752 5248 9 10 2848 1876 1296 48 10 11-- 120 0 0 0 Total Nueer of Rods 13440 13440 13440 13440 Exposure Averste (cwd/f) 0.1 3.77 7.53 11.36 $tetopoint Instantaneous Aversee Power 7.1 7.3 7.6 7.6 Burn Average Power
Statopoint to statepoint

7.2 7.45 7.6 .. .,
i once 8urned Fuel f Rod Power BOC Wuter of Rode IOC (kW/ft) Statepoint 1 Statepoint 2 - statepoint 3 tietepoint 4 3*4 96 96 160 256 4*5 600 488 832 608 5*6 ;1088 1384 1544 1268 6*7 2816 3600 4312 5484 7*8 3760 4520 3224 2562 8*9 1808 80 96 0 Totet Water of Rode 10168 10168 10168 10168 I Exposure Average (GWd/f) 12.38 15.83 19.15 22.44 - Statopoint. instantaneous Average Power 6.9 6.8 6.4 6.2 Burn Averste Power.* Statopoint to Statepoint
6.85 6.6 6.3 -
Prior Power 7.4 * . * !
' Power history and distribution taken' from core follow L
analysis output p Revision'0 Page 144
. 'se,- . . - . - ~ . - ,,rv,.. + , , , .. . ~ . . . . , . , .,...-1, ,,,,_m., .-.....-r,,-,..,.m-,-,_, w, , -
I EA-PT-91-0003-M C Table A-1 Example of RBS Cycle Dependent Power Distribution Used to Determine Average Gap Conductance (Continued) Twice Burned Fuel Rod Power BDC Nueer of Roos EOC (kW/ft) StatepoInt 1 Statepoint 2 Statepoint 3 tietepoint 4 i*2 444 116 396 476 23 3996 3880 3884 3764 34 1996 2424 2036 2052 4*$ 2168 2188 2364 940 5*6 2506 2640 2704 4332 6*7 3056 3072 3064 3068 7*8 464 312 164 0 Total NJeer of Rods 14632 14632 14632 14632 Exposure Average (GWd/T) 19.62 22.07 24.28 26.51 Statepoint Instantaneous Average Power 4.0 4.6 4.4 4.1 Burn Aversee Power
Statepoint to $tatepoint

4.3 4.5 4.25 Prior Power 7.0 * *

TOTAL NUMBER OF FUEL RODS IN CORE: 38240 -
0 l l i (%) Revision 0 Page 145 y,/
l EA-PT-91-0003-M Figure A-1 Gap Conductance Contribution by Fuel Rods in the 5-6 kW/ft Range Fresh Fuel - Statepoint 1 5.5 kW/ft 0.1 Burnup (GWd/T) Fresh Fuel - Statepoint 2 7.2 , 3.5 kW/ft 3.77 3.78 Burnup (GWd/T) Fresh Fuel - Statepoint 3 7.45 7.2 5.5 kW/ft 3.77 7.53 7.54 Burnup (GWd/T) Revision 0 Page 146
....._____...-_-..____........________._._.__m_... . _ _ . . _ _ _ _ . . _ . _ - . _ _
4 4
-EA-PT-91-0003-M ,
Figure.A-1 Gap Conductance Contribution by Fuel Rods in the 5-6 kW/ft Range Once Burned - Statopoint 1 7.4 5.5 kW/ft 12.38 12.39 + Burnup (GWd/T) Once Burned - Statepoint 2 7.4 i 6.85 ,
. . 5.5 .kW/ft O
12.38 15.83- 15.84 Burnup (GWd/T) Once Burned - 8tatopoint 3
.7 . 4 6.85 .'
6.6
-i -
5.5 kW/ft 12.38 15.83 19.15 19.16 BurnupF(GWd/T) l l. O : Revision-0 Page 147 i :- l
EA-PT-91-0003-M l l Figure A-1 Gap Conductance Contribution by Fuel Rods in the 5-6 kW/ft Range Twice Burned - Statepoint 1 7.0 5.5 kW/ft 19,82 19,83 Burnup (GWd/T) Twice Burned - Statopoint 2 7.0 5.5 4.3 kW/ft O 19.82 22.07 22.08 Burnup (GWrt/T) Twice Burned - Statopoint 3 7.0 5.5 4.5 4.3 kW/ft 19.82 22.07 24.28 24.29 Burnup (GWd/T) O Revision 0 Page 148
1 l EA-PT-91-0003-M I l O l 1 I i l APPENDIX B ACRONYMS USED IN THE TEXT \ ;. l-1 l' LO Revision 0 Page 149
- . - ~ . . _ . _ - . _ , . . _ _ . - _ _ -
. _ . _ . . . _ _ _ _ _ _ . _ _ _ . _ . ~ . . _ _ . . _ _ . . _ _ _ . . _ . _ _ _ _ _ . _ _ _ _ . _ _ . -
EA-PT*91-0003-M J - ACRONYMS USED IN THE TEXT e ACRONYM - SIGNIFICANQE , ADO Anticipated Operational occurrence APRM Average Power Range Monitor BVP- Bypass Valve BWR Boiling Water Reactor CPR Critical Power Ratio ACPR Delta-CPR (Change in Critical Power Ra-tio). EPRI Electric Power Research Institute f - ERIS Emergency Response Information System j ] FCV Flow Control Valve GE' General Electric Company. GSU Gulf States Utilities Company HTMG. High Frequency Motor Generator ICPR Initial Critical Power Ratio LBT License Basis Transient { LFMG Low Frequency Motor Generator LHGR Linear Heat Generation Rate MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio . MSIV Main Steam' Isolation Valve NBR Nuclear Boiler Rating O Revision 0 Page 150 e t 4 ~-er*+ w w- *-m., s vm +,wme,w r *- + - ,4c wsice -=E.
EA-PT-91-0003-M HSSS Nuclear Steam Supply System PhP Pressurized Water Reactor RBS River Bend Station RCPR Ratio of Critical Power Ratios SRV Safety / Relief Valve TCV Turbine Control Valve TSV Turbine Stop Valve
'4 Revision 0 Page 151 -- -- -- - - - - _ -}}

ML20077A461

Text

1.0 INTRODUCTION

SUMMARY

1.3 CONCLUSION

2.0 DESCRIPTION

3.5 CONTROL

SUMMARY

1.3 CONCLUSION

2.0 DESCRIPTION

9.0 REFERENCES

4.0 REFERENCES

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