Text

Methods for Performing BWR Sys Transient Analysis
	ML20235H560
Person / Time
Site:	Peach Bottom
Issue date:	09/09/1987
From:	Willie Lee, Olson A, Rubino L; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20235H548	List: ML20235H547; ML20235H560;
References
	PECO-FMS-0004, PECO-FMS-4, NUDOCS 8710010101
	Download: ML20235H560 (200)
	v • d • e

{{#Wiki_filter:- _ _ _ _ _. I t METHODS FOR PERFORMING BWR SYSTEMS TRANSIENT ANALYSIS I! b'2 Prepared By : [! [ 0% 'Date A.M. Olson, Engineer-Supervisory l-Safety Analysis Group Fuel Management Section Reviewed By : II , b L/ f ) Date ' W.G. Lee, Senior Engineer Fuel Technology Branch Fuel Management Section Approved By : J ~ f ) Date L.F. Rubino Engineer-in-Charge Fuel Management Section J Operating License DPR-44 rnd DPR-56 Philadelphia Electric Company Nuclear Operations Nuclear Support Department 2301 Market Street Philadelphia, PA 19101 8710010101 870928 PDR ADOCK 05000277 l-P PDR L

i 1 I J DISCLAIMER l I i This document was prepared by Philadelphia Electric Company and is believed to be true and accurate to the best of its knowledge and information. This document and the information contained herein are authorized for use only by Philadelphia Electric Company and/or the appropriate 1 subdivisions within the U.S. Nuclear Regulatory Commission for review purposes. l l With regard to any unauthorized use whatsoever, l Phliadelphia Electric Company and its officers, directors, agents, and employees assume no liability or make any l warranty or representation with regard to the contents of this document or its accuracy or completeness. l I i lL-- _ _

j i ACKNOWLEDGEMENT The author would like to thank'EPRI for sponsoring the development, verification, and continued support of the RETRAN-02 computer code. The assistance of Joseph P. Waldman in performing some of the analysis in this report is also greatly appreciated. i i 1 ii I

ABSTRACT Philadelphia Electric Company (PECo) has developed a systems-transient analysis model for Boiling Water Reactors. The basis of the model is the PEco modified version of the RETRAN-02 computer' code. The ability of the model to accurately predict the course of reactor transients is demonstrated by comparisons to actual. plant data obtained from Peach Bottom Atomic Power Station Units 2 and 3. Qualification of the model includes simulation.of plant startup tests, the Peach Bottom Unit 2 safety / relief valve test, and the Peach Bottom Unit 2 turbine trip tests. The Nuclear Regulatory Commission test problem (Licensing Basis Transient) has also been evaluated. This model, though qualified primarily against ) Peach Bottom data, is applicable for analysis of other Boiling Water Reactors of similar design. ni L__-____--

. TABLE OF CONTENTS PAGE DISCLAIMER i ACKNOWLEDGEMENTS ii ABSTRACT iii TABLE OF CONTENTS iv LIST OF TABLES vil LIST OF FIGURES. viii' l.0 INTRODUCTION 1-1 1.1 Purpose 1-1 1.2 Brief; Description 1-1 1.3 Model Qualification 1-2 1.4 Model Application 1-3 2.0 MODEL DESCRIPTION 2-1 2.1 Model Geometry 2-4 2.1.1 Fluid Volumes, Junctions, and Heat Conductors 2-4 2.1.2 Steam Lines 2-4 2.1.3 Feedwater Lines 2-10 2.1.4 Reactor Vessel 2-11 2.1.5 Recirculation Loops 2-12 2.1.6 Core Region 2-12 '2. 2 Component Models 2-14' 2.2.1. Safety / Relief Valves 2-14 2.2.2 Steam Separators 2-15 4 2.2.3 Recirculation Pumps 2-16 2.2.4~ Jet Pumps 2-16 2.2.5 Core Hydraulics 2-19 2.3_ Trip Logic-2...................................... iv i

TABLE OF CONTENTS PAGE ) 2.4 Control Logic 2-20 i 2.4.1 Sensed Parameters and Miscellaneous Calculations 2-20. 2.4.2 Reactor Water Level Calculation 2-28 l 2.4.3 Feedwater Control System 2-29 2.4.4 Recirculation Control System 2-30 2.4.5 Turbine Electro-Hydraulic Control System 2-31 1 3.0 QUALIFICATION 3-1 3.1 Peach Bottom Startup Tests 3-2 3.1.1 Feedwater System Transients 3-3 4 3.1.2 Turbine Electro-Hydraulic Control } System Transients 3-11 ] 3.1.3 Reactor Recirculation Transients 3, 3.1.4 Conclusions 3-37 i 3.2 Peach Bottom Safety / Relief Valve Test 3-38 3.2.1 Test Description 3-38 3.2.2 Model Inputs 3-39 3.2.3 Results 3-41 3.2.4 Conclusions 3-42 3.3 Peach Bottom Turbine Trip Tests 3-46 3.3.1 Test Description 3-46 3.3.2 Initial Conditions and Model Inputs 3-49 3.3.3 Comparison to Test Data 3-51 3.3.3.1 Pressure Comparisons 3-51 3.3.3.2 Power and Reactivity Comparisons 3-65 3.3.3.3 Conclusions 3-91 l 1 v l

TABLE OF CONTENTS PAGE 4.0' NRC TEST. PROBLEM 4-1 4.1 Description of Licensing Basis Transient 4-1 4.2-Model Inputs 4-3 4.3 Results 4-6 4.4 Conclusions 4-8 5.0

SUMMARY

5-1

6.0 REFERENCES

6-1 l l l vi

LIST OF TABLES TABLE NUMBER TITLE .PAGE I ' Table. 2-1 ' Peach Bottom Volume Geometric Data 2-5 Table 2-2 Peach Bottom Junction Geometric Data 2-7 Table 2-3 Peach Bottom ~ Beat Conductor Geometric Data 2-9 . Table 2-4 Peach Bottom Trip Logic Definitions 2 Table 2-5 Peach Bottom Control Model Input Definitions 2-22 Table 3-1 Initial Plant Conditions For SRV Test 3-40 Table 3-2 Peach Bottom Turbine Trip Tests Initial Conditions 3-48 Table 3-3 Peach Bottom Turbine Trip Tests Summary of Initial Input Parameters 3-48 Table 3-4 Peach Bottom Turbine Trip Tests Summary of Normalized Core Average and LPRM Level Neutron Flux Peaks 3-66 Table 3-5 Peach Bottom Turbine Trip Tests 9tmmar7 < of. Core Average Peak Neutron Flux 3-66 Table 3-6 Peach Bottom Turbine Trip Tests Time of Peak Neutron Flux 3-68 i Table 3-7 Peach Bottom Turbine Trip Tests I Summary of Net Reactivities 3-68 ) Table 4-1 Initial Conditions for Licensing l Basis Transient 4-2 Table 4-2 Delayed Neutron Data for Licensing Basis Transient 4-5 vii L

i 'l LIST OF FIGURES .i l l -] 1 FIGURE NUMBER TITLE PAGE' l . Figure 2-1 Peach Bottom-RETRAN Model 2-2 Figure 2-2 Peach Bottom RETRAN Model ] Active Core Region 2-3 j Figure 2-3 Peach Bottom Jet Pump M-N Curve 2-18 Figure 2-4 RETRAN Control System Models Sensed Inputs 2-32 Figure 2-5 RETRAN Control System Models Miscellaneous System Logic 2-34 Figure 2-6 RETRAN Control System Models Sensed Narrow and Wide Range Water Levels 2-37 i Figure 2-7 RETRAN Control System Models Feedwater System 2-40 Figure 2-8 RETRAN Control System Models Recirculation Control System 2-45 Figure 2-9' RETRAN Control System Models Turbine EHC Control System 2-49 Figure 2-10 RETRAN Control System Models Point Kinetics Calculations 2-54 Figure 2-11 RETRAN Control System Models 1-D Kinetics Calculations 2-58 . Figure 1 PB2Cl Feedwater Startup Tests +6" Setpoint Change - 60% Power l-Change in Feedwater Flow vs. Time 3-5 I viii

j LIST OF FIGURES . FIGURE NUMBER TITLE PAGE I l i;c se 3-2 PB2C1 Feedwater Startup Tests j -6" Setpoint Change - 75% Power Change in Feedwater Flow vs. Time 3-6 1 Pigure 3-3' PB2Cl Feedwater Startup Tests +6" Setpoint Change - 75% Power Change in Feedwater Flow vs. Time 3-7 Figure 3-4 PB2C1'Feedwater Startup Tests -6" Setpoint Change - 95% Power Change in Feedwater Flow vs. Time 3-8 Figure 3-5 PB2C1 Feedwater Startup Tests +6" Setpoint Change - 95% Power Change in Feedwater Flow vs. Time 3-9 t Figure 3-6 PB2Cl Feedwater Startup Tests +6" Setpoint Change - 60% Power Reactor Water Level vs. Time 3-10 Figure 3-7 PB2C3 Turbine EHC Tests -6.6 PSI Setpoint Change - 25% Power C V. Error Signal vs. Time 3-13 Figure 3-8 PB2C3 Turbine EHC Tests -6.6 PS1 Setpoint Change - 25% Power Change in C.V. Position vs. Time 3-14 Figure 3-9 PB2C3 Turbine EBC Tests +8.9 PSI Setpoint Change - 25% Power C.V. Error Signal vs. Time. 3-15 -Figure 3-10 PB2C3 Turbine EHC Tests +8.9 PSI Setpoint Change - 25% Power Change in C.V. Position vs. Time 3-16 1. ( l L IX

LIST OF FIGURES i FIGURE NUMBER TITLE PAGE 1 1 Figure 3-11 PB2C3. Turbine EHC Tests -6.9 PSI Setpoint Change - 100% Power C.V. Error Signal vs. Time 3-17 i Figure 3-12 PB2C3 Turbine EHC Tests -6.9 PSI Setpoint Change - 100% Power Change in C.V. Position vs. Time 3-18 Figure 3-13 PB2C3 Turbine EHC Tests +6.8 PSI Setpoint Change - 100% Power C.V. Error Signal vs. Time 3-19 Figure 3-14 PB2C3 Turbine EBC Tests +6.8 PSI Setpoint Change - 100% Power i Change in C.V. Position vs. Time 3-20 Figure 3-15 'PB3Cl Recirculation Startup Tests 2 M-G Set Drive Motor Trip - 94% Power Reactor Power vs. Time 3-26 Figure 3-16 PB3Cl Recirculation Startup Tests 2 M-G Set Drive Motor Trip - 94% Power Core Plate Pressure Drop vs. Time 3-27 Figure 3-17 PB3C1 Recirculation Startup Tests 2 M-G Set Drive Motor Trip - 94% Power Reactor Pressure vs. Time 3-28 Figure 3-18 PB3Cl Recirculation Startup Tests 2 M-G set Drive Motor Trip - 94% Power Reactor. Water Level vs. Time 3-29 Figure 3-19 PB3Cl Recirculation Startup Tests 2 M-G Set Drive Motor Trip - 94% Power Feedwater Flow vs. Time 3-30 l X t

LIST OF FIGURES FIGURE NUMBER TITLE PAGE ) L Figure 3-20 PB3Cl Recirculation Startup Tests 2 M-G Set Drive Motor Trip - 94% Power Recirculation Loop 'A' Flow vs. Time 3-31 Figure 3-21 -PB3Cl Recirculation Startup Tests 2 M-G Set Drive Motor Trip - 94% Power Recirculation Loop 'B' Flow vs. Time 3-32 i L Figure 3-22 PB2C1 Recirculation Startup Tests 1 M-G Set Drive Motor Trip - 95% Power Reactor Water Level vs. Time 3-33 Figure 3-23 'PB2Cl Recirculation Startup Tests 1 M-G Set Drive Motor Trip - 95% Power Reactor Pressure vs. Time 3-34 Figure 3-24 PB2C1 Recirculation Startup Tests 1 M-G Set Drive Motor Trip - 95% Power Feedwater Flow vs. Time 3-35 l~ Figure 3-25 PB2C1 Recirculation Startup Tests l 1 M-G Set Drive Motor Trip - 95% Power Recirculation Pump Speed vs. Time 3-36 Figure 3-26 PB2C5 SRV Lifft Test Steam Dome Pressure Change Pressure Change vs. Time 3-43 Figure 3-27 PB2C5 SRV Lift Test Core Average Neutron Flux Normalized Neutron Flux vs. Time 3-44 Figure 3-28 PB2C5 SRV Lift Test Core Average Neutron Flux Point vs. 1-D Kinetics Normalized Neutron Flux vs. Time 3-45 l xi i

LIST OF FIGURES l FIGURE NUMBER TITLE PAGE Figure 3-29 Turbine Trip Test 1 i Turbine Inlet Pressure l Pressure Rise vs. Time 3-53 l Figure 3-30 Turbine Trip Test 1 Steam Dome Pressure Pressure Rise vs. Time 3-54 Figure 3-31 Turbine Trip Test 1 Upper Plenum Pressure Pressure Rise vs. Time 3-55 Figure 3-32 Turbine Trip Test 2 Turbine Inlet Pressure Pressure Rise vs. Time 3-56 Figure 3-33 Turbine Trip Test 2 Steam Dome Pressure Pressure Rise vs. Time 3-57 Figure 3-34 Turbine Trip Test 2 Upper Plenum Pressure Pressure Rise vs. Time 3-58 Figure 3-35 Turbine Trip Test 3 Turbine Inlet Pressure Pressure Rise vs. Time 3-59 Figure 3-36 Turbine Trip Test 3 Steam Dome Pressure Pressure Rise vs. Time 3-60 xii

l l l LIST OF FIGURES FIGURE NUMBER TITLE PAGE Figure 3-37 Turbine Trip Test 3 Upper Plenum Pressure. Pressure Rise vs. Time 3-61 Figure 3-38 Turbine Trip Test 1 Upper Plenum Pressure Filtered Data Pressure Rise vs. Time 3-62 Figure 3-39 Turbine Trip Test 2 Upper Plenum Pressure Filtered Data Pressure Rise vs. Time 3-63 Figure 3-40 Turbine Trip Test 3 Upper Plenum Pressure Filtered Data Pressure Rise vs. Time 3-64 i Figure 3-41 Turbine Trip Te. 1 Core Average Neutron Flux Normalized Core Flux vs. Time 3-71 Figure 3-42 Turbine Trip Test 1 LPRM A Neutron Flux Normalized LPRM Flux vs. Time 3-72 i Figure 3-43 Turbine Trip test 1 LPRM B Neutron Flux Normalized LPRM Flux vs. Time 3-73 Figure 3-44 Turbine Trip Test 1 LPRM C Neutron Flux l Normalized LPRM Plux vs. Time 3-74 I l l ' xiii

i t LIST OF FIGURES FIGURE NUMBER TITLE PAGE ____-----==_ Figure 3-45 Turbine Trip Test 1 LPRM D Neutron Flux Normalized LPRM Flux vs. Time 3-75 Figure 3-46 Turbine Trip Test 2 Core Average Neutron Flux Normalized Core Flux vs. Time 3-76 l Figure 3-47 Turbine Trip Test 2 LPRM A Neutron Flux Normalized LPRM Flux vs. Time 3-77 Figure 3-48 Turbine Trip Test 2 LPRM B Neutron Flux Normalized LPRM Flux vs. Time 3-78 Figure 3-49 Turbine Trip Test 2 LPRM C Neutron Flux Normalized LPRM Plux vs. Time 3-79 Figure 3-50 Turbine Trip Test 2 LPRM D Neutron Flux Normalized LPRM Flux vs. Time 3-80 Figure 3-51 Turbine Trip Test 3 Core Average Neutron Flux Normalized Core Flux vs. Time 3-81 Figure 3-52 Turbine Trip Test 3 LPRM A Neutron Flux Normalized LPRM Flux vs. Time 3-82 Figure 3-53 Turbine Trip Test 3 LPRM B Neutron Flux Normalized LPRM Flux vs. Time 3-83 l xiv

_l l l L LIST OF FIGURES i FIGURE NUMBER TITLE PAGE Figure 3-54 Turbine Trip Test 3 LPRM C Neutron Flux Normalized LPRM Flux vs. Time 3-84 Figure 3-55 Turbine Trip Test 3 LPRM D Neutron Flux Normalized LPRM Flux vs. Time 3-85 Figure 3-56 Turbine Trip Test 1 Total Core and Scram Reactivity Reactivity vs. Time 3-86 Figure 3-57 Turbine Trip Test 2 Total Core and Scram Reactivity Reactivity vs. Time 3-87 Figure 3-58 Turbine Trip Test 3 Total Core and Scram Reactivity Reactivity vs. Time 3-88 Figure 3-59 Turbine Trip Test 1 Dome Pressure as Input Core Average Neutron Flux Normalized Core Flux vs. Time 3-89 Figure 3-60 Turbine Trip Test 1 Dome Pressure as Input Total Core and Scram Reactivity Reactivity vs. Time 3-90 Figure. 4-1 NRC Test Problem (LBT) Initial Core Power Distribution Relative Power vs. Core Height 4-9 t' l xv

i l l i f LIST OF FIGURES ] i FIGURE NUMBER TITLE PAGE Figure 4-2 NRC Test Problem (LBT) Initial Core Beat Flux Distribution Heat Flux'vs. Core Height 4-10 ' Figure 4-3 NRC' Test Problem (LBT) Initial Core' Void Distribution Void Fraction vs Core Height 4-11 Figure 4-4 NRC Test Problem (LBT) Core Power Core Power vs. Time 4-12 Figure 4-5 NRC Test Problem (LBT) Core-Heat Flux Heat Flux vs. Time 4-13 Figure 4-6 NRC Test Problem (LBT) Core Average Void Fraction Void Fraction vs. Time 4-14 Figure 4-7 NRC Test Problem (LBT) Core Average Fuel Temperature Fuel Temperature vs. Time 4-15 Figure 4-8 NRC Test Problem (LBT) Core Midplane Pressure Pressure vs. Time 4-16. Figure 4-9 NRC Test Problem (LBT) Total Core Flow Core Flow vs. Time 4-17 Figure 4-10 NRC Test Problem (LBT) Heat Flux Distribution at 0.8 Sec. Heat Flux vs. Core Height 4-18 xvi

\\ 1 i l LIST OF FIGURES l 1 ' FIGURE NUMBER ITITLE PAGE Figure 4-11 NRC Test Problem (LBT) Beat Flux Distribution at 1.2 Sec. Beat Flux vs. Core Beight 4-19 Figure 4-12 NRC Test Problem (LBT) Total Core Reactivity Reactivity vs. Time 4-20 Figure-4-13 NRC Test Problem (LBT) Scram Reactivity Reactivity vs. Time 4-21 i Xvii -_m

I' l.9-INTRODUCTION 1.lf Purpose This report' describes a transient analysis model for the Nuclear Steam Supply-System (NSSS) of a General Electric-(Reference 1) Boiling Water Reactor (BWR). This model, Ldeveloped by-the Philadelphia Electric Company (PECo), is based on the'PECo modified version of the RETRAN-02 computer code (Reference 2) and will be used to evaluate the-transient response of the NSSS to operational transients, both normal and abnormal. The model is intended to be used for providing plant operational support and for performing reload analysis and licensing calculations. 1.2 Brief Description RETRAN-02 is a complex, one-dimensional, thermal-hydraulic, transient analysis computer code developed by the Electric Power Research Institute (EPRI). It is a variable nodalization code requiring the user to input a control l I volume / flow path network / heat slab model of the system to be ) analyzed. The development of the input for the model presented in this report, representing the Peach Bottom Atomic Power Station, was based on as-built drawings and vendor specifications. The nodalization network chosen for the Peach Bottom model is 1-1 1 )

r based on extensive experience with the RETRAN-02 code and comparison of model predictions to experimental data and is applicable-in general for modeling the'NSSS of other similar BWRs. 1.3 Model Qualification The.RETRAN-02 computer code package is the result of an extensive code development effort sponsored by EPRI since 1975. It has undergone a comprehensive qualification and verification process in the last decade. Reports and conclusions-based on code predictions of various separate-l effects tests, system effects experiments and power reactor startup and special tests can be found in the base RETRAN-02 documentation (Reference 2). The Nuclear Regulatory Commission (NRC) Safety Evaluation Report (SER) for RETRAN-02 is contained in the documentation. RETRAN-02 has been widely utilized by utilities and consulting communities on a variety j of transient problems. In addition, further qualification of RETRAN-02 has been demonstrated by the Philadelphia Electric Company in this report through the application of RETRAN-02 to the analysis of: (i) The Peach Bottom Units 2 and,3 Startup Tests (ii)- The Peach Bottom Unit 2 Cycle 5 S/RV Test i l ,-2

(iii) The Peach Bottom Unit 2 Cycle 2 Turbine Trip Tests i (iv) The NRC Licensing Basis Transient The results of the above evaluations are presented in Sections 3.0 and 4.0 of this report.

l. 4 Model Application j

l The Philadelphia Electric Company RETRAN-02 model is designed to serve as a best estimate, general purpose,' systems analysis tool. It can be used for a wide range of purposes, including special tests, design changes, and operational transient evaluations. This model can also be used to analyze limiting transients for core reload design and licensing purposes. i 1 I l l-3 l ln L_ _ _______ _---__

2.0-MODEL' DESCRIPTION This section describes the Peach Bottom Atomic Power. Station (PBAPS) model developed for use with the PECo modified version of the RETRAN-02 computer code (Reference 2). This model was developed in such a way as to allow the analysis of a wide range of transients with only minor modifications to the model input. This development was based on many years of on going experience with the code and includes several l revisions of the model based on that experience. A diagram of the model nodalization selected for the PBAPS RETRAN-02 model is illustrated in Figures 2-1 and 2-2, including fluid volumes, junctions and heat conductors. A description of the primary inputs to the code is given in the subsequent sections. These descriptions should be viewed as ' typical' for the PBAPS model. The actual inputs used to model any particular transient may vary from the following description based on the nature of the transient and the previous experience gained in modeling it. Such variation will be noted where necessary. 2-1 I 1 _________._.____.___________.___9

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i .2.1 Model Geometry 2.1.1 Fluid Volumes, Junctions, and Heat Conductors The necessary geometric data used in calculating the fluid volumes, junctions, and heat conductors was obtained primarily from as-built plant drawings. The fluid volume nodes are, in general, defined as distinct regions within the primary system such as the steam dome or downcomer. Where further nodalization is required due to limits in code assumptions, these regions are divided into sub-regions (i.e., upper, middle and lower downcomers). j System components such as jet pumps, steam separators, and recirculation pumps are also typically described as single fluid volumes. A list of the key input parameters for the fluid volumes, i junctions, and heat conductors is presented in Tables 2-1 { \\ through 2-3. The basis for these inputs is found in the PBAPS Model Calculation Databook (MCD)(Reference 3). ) 2.1.2 Steam Lines i The four main steam lines are lumped into one line which is divided into six fluid volumes (see Figure 2-1). Two of the volumes model the steam lines inboard of the Main Steam I Isolation Valves (MSIVs). The second inboard volume is l l 1 l 2-4 l 1 l l

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connected to the junctions representing the safety / relief valves. The remaining four volumes model the steam lines from the MSIVs to the turbine stop/ control valves. The last outboard volume (Vol. 55) provides the pressure feedback signal to the turbine Electro-Hydraulic Control (EHC) system control logic. The turbine is not modeled as a fluid volume, however, the turbine EHC control logic calculates the turbine speed and also governs the steam flow to the turbine via a negative fill. The steam bypass system is modeled up to and including the condenser. Important junction parameters such as flow areas and flow pressure loss coefficients are based on GE steam bypass system design data and the modeling approach outlined in Reference 4. The bypass valve position is governed by the turbine EHC control logic. The bypass lines downstream of the bypass valves are modeled as passive heat conductors with an adiabatic boundary condition. 2.1.3 Feedwater Lines The feedwater lines are not modeled as fluid volumes. Instead, a fill junction controlled by the feedwater logic is used. The feedwater control logic is used to represent the feedwater heaters, turbines and pumps in a dynamic manner and to account for the piping delays due to the length of the l feedwater lines. } ) I 2-10 l f )

2.1.4 Reactor Vessel A single volume is used to model the steam space above the steam separators. The downcomer. region is divided into three volumes. The upper downcomer models the region surrounding the steam separators and includes the normal steam-water interface. This volume is modeled using the RETRAN 'non-equilibrium' option to allow superheating of the steam above the steam-water interface during pressurization events. The middle downcomer models the region surrounding the standpipes and is the volume where the feedwater flow and the liquid flow from the steam separators mixes. The lower downcomer models the region surrounding the core shroud and jet pumps. Plow to the recirculation loops and jet pump suctions are from this volume. 1 A single volume is used to model the fluid region below the core support plate (lower plenum). The upper plenum region above the upper guide plate and the standpipes are both modeled au single volumes. Two-sided, passive heat conductors are used to model the material of the shroud head and the standpipes. A single volume is used to model the internal volume of the 211 steam separators. t l l l 2-11 l o____-_________

Since the liquid flow from the separators, which includes any-steam carryunder, enters the sub-cooled middle downcomer, the effects of carryunder can be ascertained without introducing steam bubbles into the liquid downcomer. 2.1.5 Recirculation Loops The two recirculation loops are modeled separately, with one recirculation control system model driving both loops. Each recirculation loop is modeled with three fluid volumes '(suction, pump, discharge). Each loop drives ten jet pumps lumped as one. A more detailed description of the recirculation pumps and jet pumps is provided in Section 2.2. l 2.1.6 Core Region Twenty-four fluid volumes are used to model the active (i.e., fueled) region of the core. Additionally, single volumes are used to model an unheated-core inlet region and core oytlet region..The entire core bypass region is modeled with one fluid volume. Twenty-four heat conductors are used to represent the reactor fuel, one per active volume. A standard, cylindrical, three region representation of the fuel rods is used with ten nodes in the fuel, one node in the gap and four nodes in the cladding. The material conductivity and heat capacity for I l 2-12 i w____-_-_-__-

the 002 fuel and the Zircaloy cladding are taken from Reference S. The gap thermal conductivity is set to yield a constant heat transfer coefficient throughout the core. The fuel channels are modeled with twenty-four passive heat conductors corresponding to each active fluid volume. Additionally,.the core barrel is modeled as a single passive heat conductor. The calculated water density and fuel temperature from each active core volume and heat conductor is used to provide-feedback to each of the twenty-four neutronic regions used in the one-dimensional kinetics calculation. The two unheated core volumes provide feedback for the two neutronic regions associated with the core reflector regions. A RETRAN non-conducting heat exchanger model is used to model i the addition of direct heating to the core bypass volume. A constant fraction of 0.017 of the total core power is used for the core bypass heating. Direct heating of the active core volumes is modeled in RETRAN in a way which allows the value of the direct heating to vary with the water density in a given volume with the core average value typically about 0.02 of the total core power. These values of direct heating are based on calculations performed by the fuel vendor (Reference 6). 2-13 1

2.2 Component Models l

The transient behavior of a BWR is often governed by the characteristics of its various components (i.e., pumps, separators, etc.). In addition, the components may require special consideration or inputs when modeled with RETRAN. A description of the major component models in the PBAPS RETRAN model is given in this section. 2.2.1 Safety / Relief Valves Peach Bottom Units 2 and 3 each have eleven safety / relief valves separated into groups of three or four valves with each group at a common setpoint. Additionally, there are two unpiped spring safety valves at each unit. Each group of valves at a common setpoint is modeled with a single fill junction with an area equivalent to the sum of the valve throat areas. When the valves are opened, the flow through the valve is governed by a fill table relating flow to pressure. This table is based on a standard ASME coefficient of discharge calculational method which is described in Reference 3. 2-14

'2.2.2 Steam Separators The steam separators are a very important component in the system model and have a substantial effect on transient simulation since they couple two regions of prime importance, the steam dome and core region. Since the separators have very complex thermal-hydraulic characteristics, the main-emphasis in modeling them is to achieve the proper coupling between the two regions as opposed to a detailed thermal-hydraulic calculation. There are 211 individual separators at each of the Peach Bottom units. The separators are modeled as a single component using the RETRAN mechanistic separator model. This model utilizes a non-equilibrium calculational method and determines carryunder and carryover as a function of separator inlet quality and downcomer liquid level. In the PBAPS RETRAN model, the values for carryunder and carryover are held to a constant value based on manufacturer's data (Reference 7). The separator input parameters which have the most significant affect on system response are the separator inlet inertia and the pressure drop across.the separators. The i separator inlet inertia is a function of inlet quality and is determined from manufacturer's data (Reference 6). The Pressure loss coefficients are such that the separator 2-15 L----------__---

pressure drop agrees withLthe vendor equation describing pressure drop as a function of inlet quality. 12.2.3 Recirculation Pumps The RETRAN centrifugal pump model is used to model the Peach' Bottom recirculation ~ pumps. Actual pump data is used to ' input pump performance parameters in the normal operating quadrant with the remaining quadrants based on built-in curves for a pump of similar specific speed. Typically, most j l abnormal operational transients do not result in pump i l operation outside of the normal quadrant. Rated values for ] J pump _ flow, head, and torque are based on actual pump data as is-the pump moment of inertia. A control system model which simulates the recirculation j motor-generator (M-G). sets supplies a pump motor torque to the centrifugal pump model. This control system is described in detail in Section :2.4. 2.2.4 Jet Pumps Each recirculation loop in the PBAPS RETRAN model drives ten jet pumps lumped as one. The RETRAN jet pump model option (momentum-mixing) is utilized to simulate the momentum j exchange between the jet pump drive flow and suction flow in 1 l 1 2-16 L1 _ -_ ---- - -

the jet pump throat. A single-fluid volume'is used to.model each lumped jet pump. In order to properly predict the jet pump behavior, it:is important that the jet pump M-ratio and N-ratio (M-N) characteristics are modeled correctly. The M-ratio is the ratio of driven-flow (suction' flow) to drive flow. The N-ratio is the ratio of specific energy increase of the suction flow to the specific energy decrease in the drive flow. The M-N characteristic is a curve of.N-ratio as a function of M-ratio. To determine the PBAPS jet pump M-N characteristic, a series of steady-state plant data points were taken at various power / flow conditions.' This data was then applied to a RETRAN sub-model of the recirculation loops and jet pumps to determine the M-N values at each point. A second RETRAN sub-model was then used to determine a set of jet pump inlet and outlet junction flow areas and flow pressure loss coefficients that best-fit the previously determined M-N values. Figure 2-3 illustrates the calculated PBAPS jet pump M-N characteristic curve. Reverse flow pressure loss coefficients (for reverse jet pump flow) are based on currently recommended values by industry-wide RETRAN users. 1 O 2-17 L_

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'2.2.5 Core Hydraulics Changes in core flow due to changes in core inlet conditions (i.e., enthalpy), core power level, reactot prressure, or any combination of these must be predicted accurately for proper transient simulation. To assure an accurate core hydraulic representation, values for flow pressure drop coefficients and two phase friction multipliers are set to match values developed with steady-state thermal-hydraulics models (Reference 8) and benchmarked against plant data. Initial values of core bypass f3ow and total core (plenum to plenum) pressure drop are determined by steady-state thermal-hydraulic calculations and input to RETRAN. The RETRAN algebraic slip option is used to account for differences in in-core phase velocities. 2.3 Trip Logic RETRAN provides switching type control elements (i.e., trips) which allow for the actuation of various process events such au the tripping of a pump or a valve. These actuations may be accomplished either directly, by specifying the process variable trip setpoint or indirectly by specifying the time at which a particular trip is to occur. This trip logic is i used in the PBAPS RETRAN model to simulate the Reactor Protection System (RPS) and to initiate various transients and equipment actuations or failures. Table 2-4 provides a I 2-19

listing of the trip logic in the PBAPS RETRAN model. This trip logic can be easily expanded to incorporate any required additional trips. 2.4 Control Logic l RETRAN provides continuous type control system elements (such as summers, lags, etc.) whose outputs are continuous functions of their inputs. These control elements can be used to model various plant systems, their controllers and various components of the plant not modeled explicitly with volumes (i.e., turbine generator). All RETRAN variables available for editing are available as control element inputs. The control inputs utilized in the PBAPS RETRAN model are listed in Table 2-5. 2.4.1 Sensed Parameters and Miscellaneous Calculations The RETRAN control logic elements can be used to simulate the effects of instrument time constants on sensed inputs to the Reactor Protection System (RPS) or other control systems. This type of use of control logic is illustrated in Figures 2-4 and 2-5. Another use of the control logic is the manipulation of a process variable to a desired form as in the calculation of direct core bypass heating, illustrated in Figure 2-4. 2-20

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The control logic can also be used to perform calculations on various process variables to obtain a desired output edit. This type of calculation is performed for the generation of various reactivity output edits for 1-D reactor kinetics and is illustrated in Figure 2-11. Process feedback variables for point reactor kinetics calculations are also generated by control logic and are illustrated in Figure 2-10. 2.4.2 Reactor Water Level Calculation Two types of level instrumentation are simulated in the PBAPS RETRAN reactor water level calculation. These are: (1) a I narrow range (0-60"), temperature (by pressure) compensated instrument (NR GEMAC) which is used to control reactor water level via the feedwater control system and provides RPS trip signals, (2) a wide range (-165" - +50") instrument (WR YARWAY) which is not compensated and provides RPS and Emergency Core Cooling System (ECCS) trip signals. Many parameters and phenomena are involved in accurately predicting reactor water level. A detailed description of these phenomena and the PBAPS RETRAN reactor water level calculation is found in the PBAPS MCD (Reference 3), and an illustration of the calculation is found in Figure 2-6. i i i I i l 2-28 1 1 1 1 l

2[4.3- "Feedwater-Control System -The complete PBAPS feedwater system including turbines, pumps, heaters and the controller is modeled using control elements. The modeling is based on a first principals approach and is described in detail in the PBAPS MCD (Reference 3). The model assumes three condensatc pumps and three feed pumps in operation. The. controller logic allows for one-element or three-element control. Controller output in one-element control is a function of the difference in reactor level and a level setpoint. Three-element control, which is normally used, adds a steam flow to feed flow mismatch component to the level error to produce a total error signal. All controller settings and gains are based on actual plant settings. The controller signal is used to modulate the rate of steam flow to the feedwater turbines which drive the feedwater pumps. Feedwater pump speed and reactor pressure are used to i determine feedwater flow. Feedwater enthalpy is determined as a function of main turbine steam flow with appropriate l delays and time constants to simulate feedwater heater heat transfer rates and piping delays. Figure 2-7 illustrates the { PBAPS RETRAN model of the feedwater control system. } I 2-29 l

'2.4.4 L Recirculation Control System Variations'in core flow at PBAPS are accomplished by changen in. recirculation pump speed. Recirculation pump speed is controlled via hydraulically coupled motor generator-(M-G) . sets which' provide.a variable frequency power supply to each of the recirculation pump motors. A detailed description of the PBAPS RETRAN recirculation control system model is found in the PDAPS MCD (Reference'3). The PHAPS RETRAN recirculation control system model can simulate either automatic or master manual (normal operating mode) operating modes.. All controller settings and gains are based on actual plant settings. Only one M-G set is simulated and the calculated pump motor torque is used to drive both recirculation pumps. To model transients with asymmetric loop behavior, a separate (but identical) 1 recirculation control system can be added to the model. I l Figure 2-8. illustrates the PBAPS RETRAN model of the recirculation control system. i l 1 d f f 2-30 l lE______-- 1

J' }W .c ,,it .2.4.5 Turbine Electro-Nydraulic Control System Steam' flow to the turbine at PBAPS is regulated by modulation of steam admission control valves. Themodula[ ion.is accomplished.by means of an electro-hydraulic control system [ using the valve upstrenn, pressure as controlling feedback. The PBAPS RETRAN turbine EHC system model is a highly accurate representation of.the actual plant system. A-( detailed description of this model is found in the PBAPS MCD (Reference 3). l The turbine EBC components included in the model are a-simulation of the turbine generator for speed feedback, the speed control unit, the load, control unit, the pressure-control unit, the control-valve control unit, the bypass l valve control unit, and the automatic load followintj. unit. 1 [ Normal plant operation is in the manual mode (no automatic j l l i load following). All controller settings and gains are based l L l i on actual plant settings. Figures 2-9.i13ustrates the'PBAPS ] RETRAN model of the turbine EHC control system. i 2-31

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p _g 0 3 41 2 3 4 4 4 4 4 , 3 l f 3 f 3 f 3 1 i - M M M M M U U U U U S S S S S g 0 = S 4 2 S ' S 2 3 S S N N 2 2 2 N N N C E E C E C E C E A D A D A D A D D 5 2 1 S 9 8 7 6 5 N 3 3 3 3 3 3 3 3 a 3 a 3 a 3 a O I M M M M M S U U U U U L T S S S S S E A r D L O U 'Y 9 S S 7 S 6 5 5 S 0 M C T 1 N ' N 1 N 1 4 1 N E E E t E 1 SI C C C C L D ^ D - L A N A D A D A A D E 2OC D R E 2 ET S 2 G A R N C R 1 E I V U O A T G C E E R I 8 9 0 1 32 - N N O F 2 ( 3 f 3 ,<3 '3 . 2 3 3 f C 3 l AI K M M M M M R U U U U U T S S S S S T E = N R I 0 S O S S 1 N 1 N 1 N 8 N N 1 S 2 S P E E E C E E C C C D D A A D A D A D 3 1 1 7 6 5 4 3 2 2 2 2 2 3 3 n 3 n. 3 A 3 A M M M M M U U U U U S S S S S S S S S 7 N 5 N 4 N 3 N E E E E C D C D C D C D A A A A 01 0 1 I[

1;; c 1 F ( EG E A R R U E T V A A RE E P 2 R W O E 3 4 5 6 7 7 6 r 3 r 6 6 f 6 6 C T 6 3 3 3 3 y 3 i 3 6 6 3 W M M M M 4 U U U U U S S S S S G = = i 0 P M P P P P 3 2 M W M M E 2 M E C E E C E C T A T T A T ^ T A 2 4 0 0 1 1 2 5 0 9 8 7' 6 6 L 5 5 5 6 i 3 a 3 J 3 n 3 n 3 3 J M M M M W W S U U-U U U U - L S S S S S S JE = . TD E NO R P 8 P 6 P s P 4 P T 1 M P 1 W i M 1 M U M OM C C E C E C E A E E . C R T A A A A T T T T T E L ( P O M 0 E R T T 6 5 . 1 9-9 - T E G 2N AR .O E . EC VA R 15 f 5 l 5 5 f 5 t 5 2 3 4 5 6 E UN R 3 i 3 i 3 U 3 i 3 i 3 O GA C M M W M M M I R U U U U U U FT S S S S S S E = = = R P P P 1 8 M 9 M W 1 P M M P P W E E E C E E E C T C T T A T T T A A 0 2 9 9 0 9 8 7 6 5 5 4 L 4 4 4 4 3 ^ 3 J 3 A 3 n 3 R 3 M M M M M M U U U U U U S S S S S S r = P. 6 M W 4 M 3 M 2 M 1 E P P P P P P M M E E E E E E C T T C T T C T C T T A A A A 8 0 8 C8 5 3 2 8 w/n ( ll l ll lll

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M M P p R R L O O N N XU 4 l 6 f 5 5 i 5 i 5 N 4 L L 4 d U G U M M X . X U G U D" L V L TX A F X F M U M U M E M R L R L R R R P F P F O O O L N L C N 5 g 2 4 S L E ) .T D N O O M C L ( x O u S T 9 n 0 T I 5 6 1 R ID D 5 t 5 1 T E N E n 2N M M L a M R U c U O R P S W E C L P A R UN GA R ~ I FT E X X R " U U A L l. F F M M g M R R p R P O O L N N x X u U 3 5 7 8 n R 5 5 5 5 T 5 ii 5 I 5 5 N N M L n L M M U a U 4C U U W c M S S . X G X G U A B C" A V U = V L A L C X F X A F X M X x U M M U R R U M U M U L E E M R L P P R M R L L R L O R P F R R P F F F P F I L C C L O O L L C N N dOO4 OO 2 N cs n

3.0 QUALIFICATION The qualification of the PBAPS RETRAN model includes comparison of a variety of predicted rectlts to experimental test data. This test data includes plant startup test results, the Unit 2 Cycle 5 safety / relief valve (SRV) test, and the Unit 2 Cycle 2 turbine trip tests. In addition, the qualification work includes application of the PBAPS RETRAN model to the NRC Licensing Basis Transient (LBT) (Section 4.0). The results of these test predictions are intended to demonstrate Philadelphia Electric Company's proficiency in the application of.the RETRAN code for performing operational support and licensing calculations and should be viewed as a supplement to the large amount of previous qualification effort presented in the RETRAN documentation (Reference 2) as reviewed by the NRC. The neutron kinetics inputs utilized in the RETRAN analyses were developed using PEco's versions of the three-dimensional core simulator code SIMULATE-E (Reference 9) and the l L three-dimensional to one-dimensional collapsing code SIMTRAN-E (Reference 10). These inputs include void and doppler reactivity tables and the core power distribution for point (zero-dimensional) kinetics and axial dependent cross-section sets and the initial power distribution for one-dimensional kinetics. The methods used to develop these inputs will be described in a reload safety evaluation 3-1

methods report to be submitted for NRC review at a later date. 8.1 Peach Bottom Startup Tests Prediction of the PBAPS startup test data was performed to benchmark the RETRAN model control system logic described in Section 2.4. The results presented in the following sections indicate the high level of accuracy achieved in modeling the PBAPS control systems. Accurate representations of the plant control systems are highly desirable as their operating characteristics often have a strong impact on the transient behavior of a BWR. All the tests analyzed in the following sections with the exception of the turbine EBC tests were performed and recorded during the initial startup at PBAPS Units 2 and 3. The turbine EHC tests were performed at Unit 2 during cycle three. All measured data was taken by normal plant instrumentation and reduced from strip chart recordings. This method of data acquisition and reduction can lead to uncertainties in the measured data. Much effort was taken to keep these uncertainties to a minimum. 1 4 l 3-2

1 1 D.l.1 Feedwater System Transients 1 1 1 In order to verify that the PBAPS feedwater system conformed to all system design specifications and response criteria, a series of feedwater system transient tests were performed during plant startup. These transient tests were initiated by reducing or increasing the reactor water level setpoint in a stepwise fashion at various reactor power levels. A change in the level setpoint creates a change in the feedwater flow 4 demand and results in an increase or decrease in feedwater flow until the reactor water level reaches the new setpoint. A level setpoint change transient is effective for verifying the modeling of the feedwater turbine pump dynamics. i Several of these transient tests, covering a wide range of reactor power levels, were analyzed. To eliminate possible sources of process feedback error (i.e., RETRAN calculated water level) from the calculation, thereby making it a test of the feedwater control logic only, the recorded feedwater controller output data was used to drive a stand-alone model of the turbine pump dynamics logic. The stand-alone model utilized the controller output data to calculate feedwater flow. Comparisons of the calculated and actual feedwater flows for the five startup tests analyzed are illustrated in Figures f 3-1 through 3-5. The excellent agreement of these test ) i } 1 3-3 t

predictions with the measured data demonstrates the accuracy of the feedwater control logic over a wide range of operating cor.31 tionc. To further qualify the PBAPS RETRAN model, the feedwater transient at 60% power was reanalyzed using the full system model. The actual reactor level setpoint change inferred from the feedwater controller output and reactor water level data was used as input to the model. The calculated feedwater flow was essentially the same as in the previous calculation and is not presented again. The calculated water level is compared to the data in Figure 3-6. Agreement with the data is excellent throughout the calculation. This demonstrates the accurate representation of the reactor feedwater system and the reactor water level calculation which provides the process. feedback to the feedwater system. I l f I 3-4 i

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1 3.1.2 Turbine Electro-Hydraulic Control Transients A series of turbine EBC system transient tests were performed j i at Unit 2 during cycle 3 for the purpose of insuring that the I turbine EBC settings allowed for a stable response under transient conditions. The tests were performed by reducing or' increasing the turbine EBC pressure setpoint in a stepwise fashion at'various reactor power levels. A change in the pressure'setpoint causes a transient change in turbine ) control valve position until the turbine inlet pressure changes sufficiently to account for the change in the setpoint. This section presents an analysis of the four turbine EHC tests performed. ( A pressure setpoint change transient is effective for j verifying the very complex transfer functions used to simulate the turbine EHC pressure control unit. The recorded ( I data was first used to determine the control valve servo gain l used in the RETRAN model. The recorded control valve error signal was used to drive a model of the control valve servo, j l A servo gain was chosen that best matched the recorded i control valve position data. The servo gain was found to vary depending on the test power level and whether the control valve was stroking open or closed. An average value was chosen for the RETRAN model. Once the control valve servo gain was determined, a comprehensive analysis of the turbine EHC logic was performed. To eliminate possible i ) 3~11

sources of process feedback error (i.e., RETRAN calculated turbine inlet pressure), thereby making it a test of the turbine EHC logic only, the pressure'setpoint change was combined with the sensed pressure data to produce a demand signal. This' derived demand signal was used to drive a stand-alone model of the turbine EHC control logic. The demand signal.ie modified by the lead / lag network and resonance compensator in the pressure control unit to produce .a' control valve position demand signal. ) .l The control valve demand signal, in combination with the l feedback control valve position, is used by the control valve l control unit model to produce a control valve positioning error signal wh'ich modulates the control valve position. Comparison of the calculated and actual control valve error signals and valve positions is illustrated in Figures 3-7 1 through 3-14. The differences between the predicted and measured data are due primarily to the use of the constant, 1 average control valve servo gain. The good agreement of these test predictions with the measured data demonstrates the accuracy of the more important portions of the turbine EHC control logic over a wide range of operating conditions. 4 i l 3-12

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3!_,3 Reactor Recirculation Transients A series of reactor recirculation system transient tests were performed during plant startup. These tests were primarily trips of one or both motor-generator (M-G) set drive motors. The transient data taken during the Unit 2 tests is limited in that two key parameters, core power and core flow, were not recorded. In addition, either M-G set speed or recirculation pump flow was recorded during each test, but never both at the same time. The Unit 3 transient test data includes core power and core flow (core plate pressure drop) but only one test was available for analysis. This section presents the analysis of a two M-G drive motor trip performed at Unit 3 at near full power (94% NBR) and a one M-G drive motor trip performed at Unit 2 at near full power (95% NBR). Sufficient data was not available to generate the exact reactor kinetics parameters for the Unit 3 test, therefore, kinetics parameters developed for the Unit 2 test were used. Since the Unit 2 and Unit 3 initial cores were identical and the two state points are nearly identical, this is a reasonable approximation, l-i I; i: 3-21 ( l;

The transient behavior of a two M-G set drive motor 4 trip is caused by the decrease in the recirculation pump speed and flow. The rate of the decrease is determined by the total rotational inertia of the M-G sets and the recirculation pump flow characteristics. The gradual decrease in.the recirculation drive flow results in a reduction of the driving head produced by the jet pumps and a subsequent reduction in core flow. This,-in turn, results in an increase in core voids and leads to a reduction in core power. The reduction in core power results in a reduction of reactor steam flow and pressure. The increase in core voids also displaces fluid from inside the core region to the downcomer and causes a swell in the reactor water level. The combination of the level swell and reduction in steam flow results in a decrease in feedwater flow demand and a decrease in feedwater flow. The final reactor statepoint is a natural circulation condition 1 with no contribution to the core flow driving head from the recirculation pumps (elevation head only). A single M-G set drive motor trip is characterized such that only one recirculation loop is in operation. This results in an increased flow through the operating loop l due to a reduction in the hydraulic resistance sensed l I l 3-22 l l l. ______-____a

by'the loop. The reduction in the hydraulic resistance is'due to the decrease in the driving head produced by the tripped loop. As the driving head of the tripped loop decreases, the head from the operating loop eventually causes a flow reversal across the jet pumps of the tripped loop. The M-G set trip transient tests are effective for verifying the characteristics of the reactor recirculation pump and jet pump behavior. The pump coastdown data can be used to verify the total rotational inertia of the system. In addition, the tests also exercise the other major plant control systems such as the feedwater system and the turbine EBC system. l The predicted results of the Unit 3 two M-G set drive motor trip test are presented in Figures 3-15 through 3-21. Figure 3-15 compares the predicted core power to the measured data. The core power trend is predicted accurately. The final core power is overpredicted by i approximately 4% NBR. This may be due, in part, to the approximation of the reactor kinetics. The predicted core plate pressure drop (Figure 3-16) compares well l I with the measured data and demonstrates that the recirculation loop (including jet pumps) dynamic characteristics are modeled accurately. The reactor 3-23 L____________ l

l pressure (Figures 3-17) is slightly underpredicted. The predicted trends agree well with the measured data. A comparison of the predicted reactor water level to the measured data is presented in Figure 3-18. The I data exhibits some oscillatory behavior which is not d predicted by RETRAN. With this exception, the predicted water level agrees well with the measured data. Reactor water level is typically one of the more .I difficult parameters to predict accurately for a BWR i because of the complex geometries and processes 4 ) involved in the calculation. The predicted feedwater i flow is presented in Figure 3-19. The feedwater flow is strongly influenced by the reactor water level and steam flow as these are the process feedback signals to the feedwater control system. Any errors in the i . prediction of these feedback parameters (especially ] I water level) will produce errors in the feedwater flow prediction. The predicted feedwater flow is in good agreement with the data. The predicted recirculation drive flows are presented in Figures 3-20 and 3-21. Agreement with the measured data is excellent. This demonstrates that the M-G set coastdown characteristics and the recirculation pump flow characteristics are modeled accurately. I The predicted results of the Unit 2 one M-G set drive motor trip are presented in Figures 3-22 through 3-25. 3-24

-1 No core power data is available for comparison to the i calculation. Figure 3-22 compares the predicted reactor water level to the measured data. Again, some oscillatory behavior is present in the data. With this exception, the reactor water level is predicted accurately throughout the transient. The predicted reactor pressure is presented in Figure 3-23. Agreement with the measured data is good with a slight overprediction of the pressure after 15 seconds. l l The predicted feedwater flow is compared to the measured data in Figure 3-24. The feedwater flow trends and magnitudes are predicted accurately. The 1 oscillations in the data (due to the oscillating measured water level) are not predicted. Figure 3-25 compares the predicted recirculation pump speed to the data. Agreement throughout the transient is excellent. This again demonstrates the accuracy of the M-G set and recirculation pump models. Overall, the agreement of the predicted results with the measured data for both tests is good. The predicted results of the recirculation system parameters (core plate pressure drop, recirculation l pump speed and flow) are excellent. l J 3-25 1

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l 341.4 Conclusions i The results of the plant startup test predictions demonstrate the accurate representation of the major plant control systems (feedwater, turbine EBC, recirculation) in the PBAPS RETRAN model. Furthermore, the results also demonstrate that the recirculation system component models (i.e., recirculation i pumps, jet pumps) are adequately represented in the system model and can accurately predict the transient behavior of the plant. 1 l i l l l 3-37

1 I l 3.2 Peach Bottom Safety / Relief Valve Test i This section compares the PBAPS Unit 2 Cycle 5 safety / relief valve (SRV) experimental test data with RETRAN predicted results. This comparison is intended to establish the accuracy and validity of the RETRAN point kinetics model for j i use in predicting plant transients other than fast i l pressurization events. 2.2.1 Test Description j In August, 1980, during the PBAPS Unit 2 BOCS startup, a series of safety / relief valve (SRV) actuation tests were performed. The tests consisted of 13 single SRV actuations and three double SRV actuations. Experimental test data from a double SRV actuation on August 18 was selected for use in benchmarking the PBAPS RETRAN model. This test was performed at 42.4% rated core thermal power and 57.2% rated core flow. To initiate the test, two SRVs were opened simultaneously causing a depressurization and a decrease in core power. Sufficient time was allowed to establish a new equilibrium state. The SRVs were then closed one at a time (approximately three seconds apart) causing a pressurization and power increase after which the initial steady-state condition was reestablished. The RETRAN predictions of steam dome pressure change and core neutron flux are compared with the experimental test data. ) 1 3-38 I

I r3.2.2. LModel Inputs-( The RETRAN base model described in Section 2.0 was used for this analysis. The initial plant conditions used as input to-RETRAN for the SRV test analysis are listed in Table 3-1. The SRV actuation in.RETRAN was controlled by time (SRV . actuation is normally controlled by pressure).to simulate the i manual opening and closing of the valves during the test. 1 f The base model has the ASME conservative 0.9 multiplier on rated SRV flow. This' conservatism was removed for this i l l analysis to obtain a best estimate calculation. The SRV actuation' transient is a relatively slow transient when compared to transients such as a turbine trip. The slow rate l ] and magnitude.of pressurization of this transient produce 1 virtually no change in the core axial power distribution, therefore, the RETRAN point neutron kinetics model was-utilized. l i 3-39

TABLE 3-1 4. . INITIAL PLANT CONDITIONS FOR SRV TEST ' Core Thermal Power 1397.27 MW Steam Dome Pressure 1011.92 psia Core-Inlet Enthalpy 522.0 BTU /lbm Core Pressure Drop 13.42 paid Active Core Flow 54.118 Mlbm/hr Core Bypass Flow 4.482 Mlbm/hr Recirculation Drive Flow 15.408 Mlbm/hr Steam Flow 5.202 Mlbm/hr i 3-40

3.2.3 Results The calculated change in steam dome pressure is compared to the measured data in Figure 3-26. The two SRVs were opened at 0.3 seconds causing an initial depressurization followed by a slight repressurization to a new equilibrium state as the \\ turbine EHC control system responded to the change in steam flow at the turbine control valves. The first SRV was reclosed at 14.3 seconds followed by the second SRV at 17.5 seconds. This results in two distinct pressurization phases followed by a gradual return to the initial reactor state. The agreement of the RETRAN predicted pressure with the data is excellent in each phase of the transient. Figure 3-27 compares the calculated neutron flux with the measured data. The initial depressurization causes the neutron flux to drop to approximately 85% of its initial value. The neutron flux then reaches a new equilibrium value after a slight overshoot as the reactor pressure stabilizes. The subsequent SRV closures each cause an increase in neutron flux as pressure increases. The neutron flux then returns to its initial equilibrium value. ~ 1 RETRAN accurately predicts the measured neutron flux response throughout tne transient. The peak values of neutron flux for the depressurization and first pressurization match the data 3-41 l 1

i i very well. The neutron flux for the secord pressurization is overpredicted. This can be attributed to a slight I overprediction of the pressure increase during the second SRV O I -closure. To check the accuracy and applicability of the RETRAN point kinetics model, a second, identical calculation was performed utilizing the RETRAN one-dimensional kinetics option. A comparison of the calculated neutron flux (point kinetics versus one-dimensional kinetics) is shown in Figure 3-28. Agreement between the two options is excellent, validating the point kinetics calculation against the higher-order one-dimensional calculation. 2.2.4 Conclusions The PBAPS RETRAN model with point kinetics accurately predicts the transient response of the SRV test. The agreement between the predicted and measured data is excellent. The results of the core neutronics calculation are shown to be virtually identical to the results of the one-dimensional core neutronics calculation. The RETRAN point kinetics model is valid for'use in predicting the response of transients that are not severe overpressurization events and do not result in appreciable changes in core axial power distribution. i 3-42 4

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3.3 Peach Bottom Turbine Trip Tests The model predictions in Sections 3.1 and 3.2 demonstrate.the accuracy and abilities of most of the elements in the PBAPS RETRAN model. However,'the ability to accurately predict the thermal hydraulic and neutronic behavior for rapid pressurization events must also be demonstrated. This type of event is typically limiting from a reload licensing standpoint for BWRs, thus it is important to demonstrate a high degree of accuracy in the modeling techniques used to predict such events. To establish the accuracy of the RETRAN model and methods, an analysis of the three pressurization transient tests conducted at PBAPS Unit 2 during the end-of-cycle 2 was performed. 3.3.1 Test Description In April of 1977, in conjunction with the General Electric Company (GE), three pressurization transient tests were performed at PBAPS Unit 2. These ;.ests (hereby referred to as TT1, TT2, and TT3) were performed near the end of operating Cycle 2. In order to obtain the most accurate data possible for I verification of modeling techniques, special instrumentation was installed to monitor important process parameters. In 3-46

addition, the tests were conducted in such a manner (i.e., delayed scram times, etc.) as to best reproduce typical end-of-cycle licensing conditions. A detailed description of each test'can be found in Reference 11. Table 3-2 lists the initial reactor power and core flow for each test. These values were obtained from the process computer P-1 edit taken prior to each test. The test conditions were such that the pressurization resulted in c significant positive neutron flux transient. Each test was initiated by manually tripping the main turbine which resulted in rapid closure of the turbine stop valves. l I 3-47

TABLE 3-2 PEACH BOTTOM TURBINE TRIP TESTS INITIAL CONDITIONS TEST POWER CORE FLOW (MW) (% NBR) (Mlbm/hr) (% NBR) TTl 1562 47.4 101.3 98.8 TT2 2030 61.6 82.9 80.9 TT3 2275 69.1 101.9 99.4 l l l l l ~ TABLE 3-3 l PEACH BOTTOM TURBINE TRIP TESTS

SUMMARY

OF INITIAL INPUT PARAMETERS 1 TT1 TT2 TT3 Cora Thermal Power (MW) 1562.0 2030.0 2275.0 i 1 Total Core Plow (Ibm /sec) 28139.0 23028.0 28306.0 Coro Bypass Flow (1bm/sec) 2202.0 1856.0 2372.0 Core Pressure Drop (psid) 22.0 16.8 23.2 Stoca Dome Pressure (psia) 994.0 986.0 993.0 Core Inlet Enthalpy (DTU/lbm) 528.31 519.80 523.62 Stoca Plow (lbm/sec) 1638.0 2171.0 2470.0 l 3-48

3.3.2 Initial Conditions and Model Inputs The standard PBAPS RETRAN model described in Section 2.0 was used with initial conditions based on available plant data. Values for core power, core flow, core inlet enthalpy and reactor pressure were based on process computer P-1 edits taken before each transient test. Initial steam flows were calculated by a heat balance using measured values of reactor power and feedwater temperature. The calculated values agree l with the P-1 values (feedwater flows plus control rod drive flow) to within 0.1 percent. The core bypass flow and pressure drop were calculated for each test with PEco's version of the SIMULATE-E (Reference 9) computer code. Recirculation flows and pump speed were l l adjusted to be consistent with reactor conditions. Initial l water levels were input to match the data for each test. Additional data was used to specify other RETRAN inputs. These include the Turbine Stop Valve (TSV) position vs. time signal and the Turbine Bypass Valve (BPV) position vs. time signal. The TSV position signal for TTl failed so the averaga of the TT2 and TT3 signals was used. 3-49

Because the current PBAPS RETRAN feedwater control logic model is not designed to simulate the rapid feedwater flow excursions that occurred during the tests (due to the opening of the high pressure steam admission valves to the feedwater turbines), the measured feedwater flow was input to RETRAN as a function of time for each test. Two key input parameters are the steam separator pressure drop and inlet inertia. These parameters have a strong impact on the pressurization rate in the core upper plenum and thus affect the magnitude and timing of the core neutron flux response. The steam separator pressure drop specified for each test was based on the measured values. The separator inlet inertia for each test was based on manufacturer's data (Reference 6). l The control rod scram time and speeds were inferred from the measured control rod position relay outputs for each test. It was assumed that all control rods moved at the average speed. The values of the primary parameters needed to specify the initial conditions for each test are summarized in Table 3-3. i 3-50

i '3.3.3-Comparison to Test Data L3.3.3.1 Pressure Comparisons 'The RETRAN predicted pressures at the turbine inlet, steam dome, and core upper plenum are compared to the measured data in Figures 3-29 through 3-37. The predictions have been-corrected for sensor and sensing line delays based on information provided in Reference 11. The measured data was taken directly from the data tape and has not been filtered to remove sensing line resonances. The accurate prediction l-of the propagation of the pressure wave from the turbine stop 1 valves to the reactor steam dome demonstrates that the steam line dynamic characteristics are accurately represented by the RETRAN steam line model. The initial pressure oscillation in the steam dome is slightly over predicted for TT1 and accurately predicted for TT2 and TT3. Overall -agreement between the predicted and measured pressure data is excellent. 1 l A comparison of the RETRAN predicted core upper plenum (core exit pressures to the filtered (to remove sensing line resonances) measured data for the first 2.0 seconds of each test is presented in Figures 3-38 through 3-40. The predictions have been corrected for sensor and sensing line delays. To accurately predict the core neutronic response (i.e., transient power), it is essential to correctly predict the core upper plenum pressure response. As indicated by the 3-51

figures,:there is excellent agreement between the predicted and' measured upper plenum pressure'for TT2 and TT3. The RETRAN predicted pressure for TTl is slightly higher than the measured data. The initial pressurization rates and general trends are predicted well for each test.- The overall agreement between.the RETRAN predicted pressures and the measured data is.very good. The largest differences occur for TT1. These differences are discussed later in this section. l l J .l 4 l i 3-52 i i i 1 1

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0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 7 6 5 4 3 2 1 n_m_ vow _g, eu ): m m Du _ O a w vu l l l ,,l .;ltl(l

3 e r u g i ~ F _~

- - i :.

i4 E R 2U A T T S N A ~ - A S S D i3 ..::.:.:=; :::: R i ...G e EE TR 2 T T E T R P r-P R M I i T U N E E N L I B P RR , 2 UE - %,N TP P I' U l y l i 1 ' j g p-

i i

...O ,1 I / r ', r. r { l lH',,7 M ^ ~ ^ ^ 0 0 0 0 0 0 0 0 0 o 0 0 0 0 0 0 7 C 5 4 3 2 1 mIOv c m.x cL a v m a)u _ ) ) i U o wiwcc

3 e r u g i F f: - - :.:.::.::: .;4 E R r 3 U A-T T S A N-S S D A-R-

...:::.;3 E E 3

T-T R T E-T R-P P T I R E T L E N NI E p I B i N .1 ..:::.;2 RI 1

1 U B i

TR U i T l s r l [ i,, j i,F I -7 .,,t ,:...::...:::.;1 l i, l iii, i ii i l i s I l l l 3, 1 ei E i i 't-4 0 0 0 0 o 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 0 8 6 4 2 2 4 6 1 1 m.v e m._ a e' a m " eL - a Q w ee

3 e r u ~ g ~ ~ F ~ i ~ ~ . ~. ..-....:.... :.:.:s ...e . G 6 l E 3R U A T A N-T S S D A S R- .:...e G. l E E 3 T-T T E-R T R-P P I RE 's, T M EO N D I B M 7, i l A R U E T T S

.r

. e. e i 1 a r t s r 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 8 7 6 5 4 3 2 1 m mgv a) m a-a'3 m m wL _ a wiec

l 1 1 m s m l m u .= 3 .en k Lg g \\ l j m i, E i es l F # l i i Q z,l mW o 4, _m x,........... m V) W, . pm 1, s w, m H Me O \\ = = L a e v) 2 v g H D M .i e z ,c E W E.W % ~. 1 r A ~p m ...........s... .......................,-u gg ........ N u, 3W FA f ' Q< ~. n 3 ~, i c =:- ~_ i -c T, c a O 4 s 6 l 6 4 6 i O. O. O. O. O. O. O. O. O O O O O O O O O O co 5 O T M N (!sd) es!g e;nsseJd 3-61

3 er u g o i F .9 i: : :.- E R 1U s TS SS A EE T s s TRA P PD IRMD TUE ~ ~i NR E EE N LT IBP L I F R R UE TP P 1 U f 9 A T N A A D R T 1 T E T R 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 O 7 6 5 4 3 2 1 mm_ von _TOL 0 o mt _ O i. O t w,O~

,l 2 9 ~ ~ 3 3 e. r u / g i F p ,5 .** - :',... :... i...:.. :.. ......:....i::. ..y 1 e E R 2U T S s S S A E E T ,s TRA %~ ) c PD ~ P e S R MD I ( T UE . i....:.:... .,1 e s NR m E EE N LT i T B PL I IF RR UE TP r P i. r U A s, 1, T N r,g r, A A i D 74 ,5 R 2 T 0 T E T R 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 7 6 5 4 3 2 1 s_.mavm.- Ot-3 m m DL1 wiOw

3 e r u g i F s .......:.... i :". . =. o E o. R A. 3U g T S y m S S A E E T TRA s PD s P R MD I ......:....i- : 9. :. .s..6 =:. ".. .,1 T UE NR E N LT EE 'p B P L I I F

o, RR UE e

TP s p P r * ',r r U 'a A 8 r T r A N_ r 9 r D A.. r t R_ .. i. p. 3 T_ s T E_ T R-0 0 0 0 Q 0 0 0 0 0 0 0 0 Q 0 0 0 0 8 7 6 5 f 3 2 1 m._ av & }y Y Ou 3 D v DL 11 . _O )aeOu 1 l

~ 1 '3.3.3.2 Power and' Reactivity Comparisons Figures 3-41 through 3-55 compare the predicted core average. neutron flux to the measured average of the LPRM signals for i each test. Also compared is the predicted neutron flux response to the average of the'LPRM signals at each LPRM level (A,B,C, and D) in the core. A summary of the predicted and measured' neutron flux peaks is-given in Table 3-4. l The RETRAN predicted neutron flux response is in excellent agreement with the measured data for TT2 and TT3. The magnitude and the timing of the core average neutron flux peak and the area under the flux peak are predicted very accurately. The neutron flux response for TTl is significantly overpredicted. Table 3-5 presents a summary of peak core average neutron flux and area under the flux peak for each test. 1 A summary of the time of peak neutron flux is presented in Table 3-6. The predicted neutron flux response for TT2 and TT3 at each LPRM level is also in excellent agreement with the data. This demonstrates that RETRAN accurately predicts I the change in the core power distribution during each transient. The overprediction of the TTl neutron flux response is discussed later in this section. I 3-65

L i l' I I TABLE 3-4 i PEACH BOTTOM TURBINE TRIP TESTS 1 I

SUMMARY

OF NORMALIZED CORE AVERAGE AND { .l l LPRM LEVEL NEUTRON FLUX PEAKS i i ,l CORE _A _B _C _D AVG. I i CALC 4.92 7.06 8.48 8.20 7.11 TTl DATA 3.48 4.47 5.23 5.59 4.83 %DIFF. 41.4 57.9 62.1 46.7 47.2 CALC 3.66 4.87 5.09 4.72 4.66' ) TT2 DATA 3.51 4.50-4.86 5.03 4.52 l %DIFF. 4.3 8.2 4.7 -6.2 3.1 1 1 f CALC 4.22 5.63 5.99 5.53 5.37 TT3 DATA 3.68 4.90 5.49 5.58 4.94 j %DIFF. 14.7 14.9 9.1 -0.9 8.7 j TABLE 3-5 PEACH BOTTOM TURBINE TRIP TESTS

SUMMARY

OF CORE AVERAGE PEAK NEUTRON FLUX l PEAK NEUTRON FLUX (NORM) AREA UNDER PEAK CALC. DATA %DIFF. CALC. DATA %DIFF. TTl 7.11 4.83 47.2 1.165 0.891 30.8 I TT2 4.66 4.52 3.1 0.742 0.737 0.7 TT3 5.37 4.94 8.7 0.694 0.680 2.1 4 l L l 3-66

l. l The calculated net reactivity, scram reactivity, and net reactivity implied by the data are presented in Figures 3-56 through 3-58. The implied net reactivity was calculated 4 utilizing the inverse point kinetics equation. The kinetics parameters (Betas and Lambdas) calculated for each test were input to the equation with the average LPRM signal input as the amplitude function. A summary of the calculated and implied net reactivities is presented in Table 3-7. The. implied data indicates that the net reactivity turns (slope becomes negative) before scram occurs for each test. However, while the neutron flux turns before scram occurs for TTl and TT2, the neutron flux for TT3 turns after the scram occurs. Thus, the peak neutron flux and area under the peak for TT3 are sensitive to the scram delay time. The peak net reactivity is overpredicted by approximately 1.5% for both TT2 and TT3. This is due to the slight overprediction of the upper plenum pressure at the time of peak reactivity. The peak core average neutron flux for TTl is overpredicted by 49%. However, the peak net reactivity is overpredicted by only 12%. The disparate overprediction of the neutron flux is due to the extreme sensitivity of the j flux to changes in reactivity when the net reactivity is I close to $1.00. i 3-67 )

. TABLE 3.l PEACH' BOTTOM TURBINE TRIP TESTS TIME OF PEAK NEUTRON FLUX TIME (SEC) CALC. DATA TT1 .762 .768 TT2 .726 .726 TT3 .696 .702 l TABLE 3-7 \\ 1 PEACH BOTTOM TURBINE TRIP TESTS

SUMMARY

OF NET REACTIVITIES PEAK REACTIVITY TIME OF PEAK lj,}. (SEC) CALC. DATA CALC. DATA l TTl 0.862 0.770 0.726 0.732 l TT2 0.777 0.767 0.690 0.690 TT3 0.831 0.816 0.672 0.666 ) l' i. l 3-68

1 The large overprediction of the neutron flux for TTl is caused by the overprediction of the first pressure oscillation in the steam dome and upper plenum. At the time of peak reactivity, the predicted upper plenum pressure is approximately 2.9 psi higher than the measured data. Because 1 the neutron flux.is very sensitive to the net reactivity at I this time, it is substantially overpredicted. To determine the magnitude of the net reactivity f overprediction due to the.overprediction of the upper plenum pressure, an inferred steam dome pressure was used as a boundry condition'to drive the RETRAN model. The inferred steam dome pressure was obtained by calculating the ratio of the measured TT1 and TT2 steam dome pressures at each' data point and then multiplying the predicted TT2 steam dome pressure by the ratios. This technique is shown to be reasonable by virtue of the fact that the measured steam dome pressure ratios remain constant to within a few percent throughout the important portion of the transient (0-1 sec). This technique also produces less distortion of the steam ~ dome pressure than attempting to filter the measured data. The pressure driven calculation results in a peak neutron flux of 5.12 (6.0% high) and an area under the flux peak of 0.937 (5.2% high). The calculated TT1 neutron flux and net reactivity responses using the inferred steam dome pressure are illustrated in Figures 3-59 and 3-60. The calculated peak net reactivity is $0.786 (2.1% high). This analysis l l 3-69 l l

verifies that the reactivity components (void, doppler, and scram) implied by the cross section data. input to TTl are accurately represented. The overprediction of the first pressure oscillation for TT1 is not fully understood. The failure of the TSV signal for TT1 creates some uncertainty in the bypass valve opening delay and TSV closing characteristics. Alternative calculations with smaller _ bypass valve delays have resulted in improved prediction of the first pressure oscillation but have also resulted in the underprediction of the pressure later.in the transient. The loss of the anticipatory i full-open bypass valve signal (due to TSV signal failure) may have resulted in less than 100% bypass valve position depending on the settings of the load limit and maximum combined flow at the time of the test. The overprediction of the first pressure oscillation is not observed for TT2 or TT3. Both these tests show a sharp reduction in the pressurization rate in the upper plenum at approximately 0.55 sec. The predicted reduction in the pressurization rate is not as severe for TT1 (see Figure 3-38 through 3-40). It is possible that this is low qutlity phenomena (low power, high core flow) which is not observed at high power (high quality) conditions. 3-70

TURBINE TRIP TEST 1 CORE AVG. NEUTRON FLUX Figure 3-41 9.0

8. O _

TT1 DATA

7. 0 -

i' R..E_T R.A..N...

,t i, ,f.,., ,s. i .. i

6. 0 -

i.>t>- f x,. i. _9 t i g i. r i q) g g g o

5. 0 ~

'-*'l-i,

(,) o ,I i, i. (D g u i f,Q.........................l.

4............

..f. -o E t o u. O Z

3. 0 -

~. 4 8 g

2. 0 -

.g

~

g 6 g ,b g g* 4 1.0 g g O l g w t . ~ - -. m 0.0 i . 4 i i i 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 1 3-71

TURBINE TRIP TEST 1 LPRM A NEUTRON FLUX Figure 3-42 9.0

8. 0 -

~- TT1 D ATA 7' 0 - R.E.TR A N... e s

6. 0 -

X V. 3 t a.,, 0 tr 5.0-...........!................ i....... ...).......... Q. A f l. "Q ,t l Q) s. i, N r e

4. 0 -

,.n-r-

o , i g g E l t k g O Z p g 3,0-s.****e ........+....... .J. .g.- p..... f p g p f e 2.0-v- n g t 1.0 ^ : e i s 4, "*. 4 1 00 _^"j_ __^g 1 I I i 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 3-72

l l l 1 l 1 i TURBINE TRIP TEST 1 l LPRM B NEUTRON FLUX Figure 3-43 9.0 8.0 TT1 DATA ~ 7'Q q............. R_ _E.T. _R_A_N_ _ _ i I l ,f.. t,

. s

6. 0 -

s>V r-l- .? 1 x s, 3 \\ e'. i L .f g 5 l I g ( $,Q- .....p.., ..g.. .4 ..j.. c ,e l I g t a ,8 O 1 W s = 4.0 - e - 0 i l E 2 ,1 g I k 1 i O f g t 3.0 - ~- - t e q t \\ g

2. 0 -

~. s, -. g g g g-s T

1. 0 -

8 4 ) \\ 4 0.0-i i 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 4 3-73 ___-______-_________________a

TURBINE TRIP TEST 1 LPRM C NEUTRON FLUX Figure 3-44 9.0 ~- ~ ~ -,'. l.,'- ~ ~- 8.0- ~ ~ ~-' " ~- TTI DATA R _E.T. _R.A..N _i ' ' ' '! !' I '. 7' 0 - ~- - '- ' ~

6. 0 -

~ ~ ~ : ' ' ~ :' ',' V. \\ ' ' '. ' ' ' ?. ~ ~ ' - x _2 u_ i, s s S. 0 - ',i i a. g .o. . N__ i,. 4, o _...................... o E i m o Z

3. 0_

4 g \\,

2. 0 --

s, % e s. \\ h a e s, 1.0 ~ j [ i 0.0 O 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) l i 1 i 3-74 i

TURBINE TRIP TEST 1 LPRM D NEUTRON FLUX Figure 3-45 9.0 1 8.0-

s, i

i ,u. l TT1 DATA l: i J'Q - 8.e. 4.........s......, ....s... R..E.T.R A N. i li ; f. 1

6. 0 -

s t X ,s e . 2 i, u. 2 i t 1 M 5.0-v- c. ~ J i. i g ,e gp n N p

4. 0 --

i-D t E L I i O \\ o ~I. ~ 3.0_ g g \\ .b b 2.0-4 + / g i e r s 1.0 ~ 8 l k j h h h 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 3-75 L___________________________________

TURBINE TRIP TEST 2 CORE AVG. NEUTRON FLUX Figure 3-46 7.0 1 6.0- ~ '.t:' ' ~'.'~~ t'~ TT2 DATA i i ';'~; 'i~1

5. 0 -

R..E.T R.A N... ,! \\. X 3 g q)

4. 0 -

' ' ' ~ '.. ' ~ ~. ~ ' ' -. ~ u O o c O .N_ -o 5.0- ~.. ~.. ~.......... ~.......~.a.~,.~.. ~ ~ ~. ~ ~. ~. E 1 g o t 2 b 6

2. 0 -

'i'-

~~;

1-8 6 b 8 g 6 1.0 ^ .~vs.- s, g 0.0 i 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 3-76

t TURBINE TRIP TEST 2 LPRM A NEUTRON FLUX Figure 3-47 7.0

6. 0 -

.l + t TT2 DATA i. i

s''i>-

5.0-RETRAN x _s. a y

4. 0 -

e o T / i a) r .r._a n W. WIN O 5,' o _ ........]... j... E L. C) Z .n. .i. .3_ .i h .t g e b y 1 1 1.0 y s e 1 l ~ ~ ~ ~ 4 0.0- ) i i i i O 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) ) 1 l 3-77 i =

~- , _.. -,y,_ l m, ,,e ,I, j V ys /

,9 j

s 2m-A 5 ,s. i,-, t TURBlNE TRIP TEST 2 LPRM B NEUTRON FLUYe / r / 'f Figure 3-48

7. 0 -

?.- ~. ~ ~ ~ ? ,r e i s s,o _ ..e TT2 DATA i

5. 0 -

+'+--

~ - R_ _E_T_R_A_N_ _ _ 8'- t g. x - t t s ,t 4 u_ r y g f - Q s f,,., .,L... .I-... 5>'.. 4 ( >;..... p.. r e 1 a_- J I* a c e w.,., ,\\, N . \\ -, O ; 3.0' ', r*'.' v.' o-E~ c t t s 1 O "l 1 g .s .,1, 2 0 -' i-1- f s, s s e 1. 0

~. v s, 1 e .f v. \\. g r, j.- e [s 0. 0.rv- .w i i i i l p 'O 0.25 0.50, 0.75 1 1.25 1.50 /.llme (Sec) s ',Y. .* 4 / 3-78 1 1, t- ^ y rr s e h tI s i k

l TURBINE TRIP TEST 2 4 LPRM C NEUTRON FLUX l Figure 3-49 1 7.0 i f {

6. 0 -

l 1 i TT2 DATA i i I ....[h'...... .i............}... i $, Q - l1 ) R..E.T.R.A.N... 1 t ) X 3 1 1 L }

4. 0 -

g c CL J l

I I

.q I

D f

o .N 1 i 0

3. 0 -

l E 1 L. 1 0 25 l l l i i-

s-: 2. 0 -

s, 1 j s s s. g

s.

i j 4 1.0 ^ ~. .s s s s 4 l l 4 \\ 0.0 i i i i i 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 3-79 l

fo TURBINE TRIP. TEST 2 LPRM D NEUTRON FLUX Fig u re. 3-50 7.0 i,

6. 0 -

'-l TT2 DATA i i i h -)

5. 0 -

R..E.T.R.A.N.. 5 X ,t s., ed ,t .i -L .s .i y i 4,0 - .............................., r... a, x Q l .g . i g ,. a U . g q) ,t . i .bJ. i

==- -) -] 0

3. 0 -

E t t i i C) r i \\ Z s e r k

2. 0 -

1 4 l l I 4 s g. -L b-1.0 = 0 4 1 0.0 l i i i 0 0.25 0,.50 0.75 1 1.25 1.50 l 1 Time (Sec) i \\

1-b0 1

l l I

TURBINE TRIP TEST 3 CORE AVG. NEUTRON FLUX J l. Figure 3-51

7. 0

. 6. 0 - " " " " ' ".'" ~ ~ ~ " :- " " " ' <."~'!"""'9"""""- [',. TT3 DATA i ,,i. i

5. 0 -
' ' ' r d ~

' ' ; ' ' ' ' a '. ~ ~. R_.E. _T_R_A_N_ _ _ x _3_ h. u. i . o> 4

4. 0 - ~'.

, ' ' ' ' '. ~ ~ ' ' '. - ' ' ' ' ' - g u o o v. o. N

3. 0 -

o E u o z

2. 0 - " " " " "> " " " " " ": - " " " :" " " " " i " " " " " 4 " " " " " "

i.0 e b e g ^ un t [ .e h h h h h 0 0.25 0.50 0.75 1 1.25 1.50 1 Time (Sec) 3-81 L________------__

TURBINE TRIP TEST 3 LPRM A NEUTRON FLUX Figure 3-52 7.0 6.0-t. q TT3 DATA i. i i i

5. 0 -

'l'-

'i R..E.T.R.A.N... : x s u. .s } 3 m .'a. 4.0- ~- Q. ,. i. g i. f g i y 1. e N_ 0 3.0 - E u O g e 8 I

2. 0 -

.5 i .il e i-4 8 e a + 1.0 = u-W e 6 9 2 g l 0.0 m-- j i i i i i 0 0.25 0.50 0.75 1 1.25 1.50 1 1 Time (Sec) l 3-82 L_______---

\\ L l TURBINE TRIP TEST 3 i PRM B NEUTRON FLUX Figure 3-53 7.0

6. 0 -

~?;-V--f-- ,o, TT3 DATA l 1 i. i i-m 'i, ; -

5. 0 -

R_E_T_R_A_N_ _ _ l x_, ,i u. .s

4. 0 -

x a g z aa .__a.

3. 0 -

- ~. ~ ~ r. l 3 0 E t_ O z g... .....l. },Q _.............l.... 1.0 - -~ ^^ ^ ^ ^ ^ ^ g q j s j l 0.0 1 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) I 3-83 i l I i 1 i 1

TURBINE TRIP TEST 3 LPRM C NEUTRON FLUX Figure 3-54 7.0

6. 0 -

- ? ,r ,3., l, Ii TT3 DATA R_E_T_R_A_N. _

+-

s'O-X r 2 u 40-x a. l ._I

l T

.I 1 q) .N e i 1 l Q 3,Q ..s.t . s.. .y.. 4 E e e L O z e s 5-- 2.0 - i

)

- - - i- - - - - 6 5 6 -=- ^.^ ^. 1.0 v - * - - * - ' s e b 6 1 j 0.0-i i i i i 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 3-84 _ ________ _ _ a

TURBINE TRIP TEST 3 LPRM D NEUTRON FLUX Figure 3-55 7.0 6.0- -} j i, t TT3 DATA i l i ,l-i -) 5.0-R_ _E_T_R_A_N_ _ _ I x __3 u s 1,....:.......................... f,Q _............ i e \\ a. ,f g g g) .i ,s .N 4 3 i d,. - en ? 1 a 3.0- ~- I E t.

i O

~ l l z s 4 I

2. 0 -

i s l -s I \\ r s 4 a s f g y 1.0 L 3 3 F =-,, _

l-e.

0.0 ) i i i i i 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 3-85 i i i

i j TURBINE TRIP TEST 1 1 1 TOTAL CORE AND SCRAM REACTIVITY Figure 3-56

1. 0 -

3 I TTIDATA <t 1 i r' - l'.u R. _E T_._ T_ O T_ A_ L_. \\

0. 8 -

,' + s 1 R E T. S.C R A.M. e ,1 j 0.6-e 9 a 0.4 - m M v 1 b_ '1 0.2 - +- O ,0 q, ,,,,., u,,m,ia..,u.al q m 0'0 . 7 l7 l'F F Il i" ] i \\. e: ' 1. - 1 - 0. 2 - n a g I .I \\. 3 t

1

-0. 4 - t. b 4 \\ n \\. l 1 g l - 0. 6 i 0 0.25 0.50 0.75 1 1.25 1.50 l Time (Sec) 3-86 a______---.

TURBINE TRIP TEST 2 TOTAL CORE AND SCRAM REACTIVITY Figure 3-57 1.0 TT2 DATA - l l .l.. R.ET. T.O. TA. L.

0. 8 -

R ET. S. C R A.M. 0.6 - 9 0.4 - i -) m 4 v 3 h. i 0.2 - l r-bg + o I. i c ,(' i cp hlkMlhalbllhil 0.0 .....,y.....,,r i i i ki i i \\ ~ 0. 2 - -ii ) i. 1 t i-i: l t (i. e. l i. - 0.4 -

4. -

1 4 ? d t \\ .i 11, j t t - 0. 6, ~ i i i i 0 0.25 0.50 0.75 1 1.25 1.50 1 Time -(Sec) l 3-87 l l l l l _-____--______-___-_____ _ - _ a

i TURBINE TRIP TEST 3 TOTAL CORE AND SCRAM REACTIVITY Figure 3-58 1.0 TT3 OATA R t!._IgIg. o.3_ R ET. S.C R A.M. j 0.6 - .l I i

.i>-

0.4 - m i g v .............. p...., ....,.7...........',...........,... '5 0,2 - "I q 1 o j :' O ci - tr L 1, b, in: 1. Ah O.o k k, Ij i I h 9"

k. :

6 0,2 - 7 ) l i- ~ 0.4 - i 1 \\ -0.6 i 3 0 0.25 0.50 0.75 1 1.25 1.50 l Time (Sec) 3-88 l 1 l

i TURBINE TRIP TEST 1 3 DOME PRESSURE AS INPUT CORE AVG. NEUTRON FLUX i Figure 3-59 l 7.0 J L ?- V- -t-

6. 0 -

TT1 DATA i .,U!- i-

5. 0 -

R..E.T.R..A.N... l 'i X ~ 1 5 L gp u

4. 0 -

o O y e .N }.. 0 5,0-E O Z 5 \\

2. 0 -

+ i 4 s. \\< s a-1.0 s 1 t 0.0 i i i i 0 0.25 0.50 0.75 1 1.25 1.50 Time (Sec) 3-89

TURBINE TRIP TEST 1 DOME PRESSURE AS INPUT TOTAL CORE AND SCRAM REACTIVITY Figure 3-60

1. 0 - l TT1 DATA

... i........i.. R. E T... T_ O_ T_ A_ L_,

0. 8 -

g R ET. S.C R A.M. i- +-

0. 6 -

s 't - e-0.4 - s. m w v t i

j i

i '5 o,2 _ . y.. j - i ~~ o O / i O H \\. ai,a u a aa m a n at o,o _ _ _ g;... _. _.... Mj\\Wf19lIIIl9lI'" j 'i i' .\\ j p i, i t

i O,r -

) - 0.2 - ) . i i{ J l \\- -0.4 - \\ i-i i \\, k i l i j - 0. 6 F' i i i 0 0.25 0.50 0.75 1 1.25 1.50 J Time (Sec) 3-90 l l I i

3.3.3.3 Conclusions The excellent agreement of the turbine trip predictions with the measured data demonstrates that the PDAPS RETRAN model can accurately predict the course of events of rapid i overpressurization transients and is applicable with a high degree of confidence to the evaluation of limiting transients for reload analysis and licensing. l l l i l i i I l 3-91

14.0 NRC Test Problem (Licensing Basis Transient) ) This section presents an analyais of the NRC Test Problem. 4 The NRC Test Problem is a turbine trip without bypass transient at PBAPS Unit'2. This particular transient is commonly referred to as the Licensing Basis Transient (LBT). The NRC has requested that Philadelphia Electric Company (PEco) perform this analysis in order to facilitate their evaluation of the transient analysis methods presented in l this report. This section describes the analysis and provides comparisons of results obtained by PECo to the results obtained by other organizations (Reference 12). 4.1 Description of Licensing Basis Transient (LBT) The LBT is defined as a turbine trip without bypass flow event with no recirculation pump trip for PBAPS Unit 2 at the end-of-cycle (EOC) 2. The initial conditions of the LBT are specified in Reference 12 with additional information obtained from References 11 and 13. A complete listing of the initial conditions is given in Table 4-1. Some inputs to the model (e.g., stop valve closure rate) were developed based upon the assumption that a licensing basis calculation is required. I l 4-1 L_______-----__--_-----

TABLE 4-1 l INITIAL CONDITIONS FOR THE LICENSING BASIS TRANSIENT PARAMETER INITIAL VALUE Core Power (MWt) 3440.0 Total Core Flow (Mlm/hr) 102.'S Core Bypass Flow (Mlbm/hr) 9.62 Core Inlet Enthalpy (BTU /lbm) 522.8 Turbine Steam Flow (Mlbm/hr) 14.04 Steam Dome Pressure (psia) 1034.0 Core Exit Pressure (psia) 1045.0 Core Inlet Pressure (psia) 1069.0 Total Recirculation Flow (Mlbm/hr) 34.2 2 Core Average Gap Conductivity (BTU /hr-ft OF) 1000.0 SAFETY RELIEF VALVE SETPOINTS (OPEN/CLOSE) 4 Valves (psla) 1090.8/1070.8 4 Valves (psia) 1100.9/1080.9 3 Valves (psia) 1111.0/1091.0 2 Valves (psia) 1242.0/1222.0 4-2

4.2.

Model Inputs' The PBAPS Unit 2 Cycle 1 end-of-cycle exposure distribution was determined with a step-wise cycle depletion using PECo's version of the 3-D simulator code SIMULATE-E (Reference 9). Actual plant conditions and rod patterns during the cycle were utilized. Cycle 2 was then depleted to EOC using a Haling analysis at 100% power and flow conditions. The' initial power distribution and nodal cross sections for the LBT were then i l determined with SIMULATE-E at EOC and collapsed to 1-D for RETRAN analysis by PEco's version of the 3-D to 1-D collapsing l code SIMTRAN-E (Reference 10). The methods used to develop these inputs will be described in a reload safety evaluation j L methods report to be submitted for NRC review at a later date. The RETRAN model described in Section 2.0 was used for this analysis with the following exceptions to conform to specific model inputs for the LBT described in Reference 13. The turbine bypass system was inactivated and the safety relief valve setpoints, delay and stroke times were changed to those defined { in Reference 12. The PBAPS steam lines described earlier were replaced with the GE steam line nodalization described in Reference 6 to be consistent with other calculations. The length and volume of the steam lines have a significant effect on the timing and magnitude of the pressure wave to the reactor vessel. L i L 4-3 l 1

The core region loss coefficients were altered to allow RETRAN to converge on the pressure distribution specified as initial conditions in Reference 12. The recirculation system loss coefficients and areas were altered to allow RETRAN to converge on the total recirculation flow reported in Reference 13. The reactor kinetic values for the six group delayed neutron fractions and decay constants (Beta and Lambda) were obtained from Reference 13 and are listed in Table 4-2. l 1 1 i l l 1 l 4-4

TABLE 4-2 DELAYED NEUTRON DATA FOR LBT

DELAYED GROUP YIELD FRACTION DECAY CONSTANT (SEC ~1) l' O.000207 0.0127 2

0.001163 0.0317 3 0.001027 0.1150 4 0.002222 0.3110 5 0.000699 le4000 l 6 0.000142 3.8700 TOTAL: 0.005460 ) 1 i 4-5 l

4.3 ~ Results Figures 4-1 through 4-3 show the initial axial distributions of the core power, heat flux and void fraction. These distributions agree well with the initial distributions reported by General Electric (GE) and Brookhaven National Laboratory (BNL).- RETRAN predicts slightly more power in the bottom of the core as indicated in Figure 4-1. RETRAN predicts slightly higher void fractions in the top half of the core as compared to the GE and BNL void distributions. This is most likely due to differences in the void models (GE Dix vs. EPRI Void). The transient core average power, heat flux,'and void fractions are plotted in Figures 4-4 through 4-6, respectively. The RETRAN transient core power and heat flux agree well with the GE q l results. The BNL transient power response is similar in magnitude but differs from the RETRAN and GE calculations in timing. The.BNL transient heat flux is dramatically different from both the RETRAN and GE calculations. This may be attributed to apparently large differences in fuel pin models as j indicated by the large differences in the transient core average fuel temperature shown in Figure 4-7. I l l I 4-6 ........_.__.._.___.__.._...__._.__._.._._._J

The initial core average void fraction is slightly larger than .the GE and BNL void fractions. RETRAN also calculates a larger percent change in void fraction than GE or BNL during the transient due to the larger increase in core mid-plane pressure between 0.6 and 0.9 seconds. The transient core mid plane pressure and core inlet flow are plotted in Figures 4-8 and 4-9. The RETRAN core mid plane pressure rises more quickly than the GE and BNL results but then peaks at about 1.0 seconds, while the other calculated pressures continue to increase. This is due to the rapid, successive opening of the second and third safety / relief valve groups near 1.0 seconds. The RETRAN core flow is similar to that reported by GE with RETRAN predicting essentially the same flow on the I j 1 first peak and less flow on the second peak. The overall trends l cre very similar. The BNL core flow trend agrees with both the RETRAN and GE results but differs substantially in magnitude. Both the core pressure and core flow responses are highly censitive to the timing of safety / relief valve actuation. l The RETRAN axial heat flux distribution at 0.8 seconds and 1.2 seconds compares well with the GE and BNL results as shown in Figures 4-10 and 4-11. The change in the axial RETRAN heat flux distribution as well as the magnitude is in reasonable agreement with GE and BNL'results. i 4-7 l t

The RETRAN transient total core reactivity (Figure 4-12) is similar to the BNL calculation in trend and magnitude. The peak reactivity occurs slightly earlier as evident by the power peaks in Figure 4-4. The RETRAN transient scram reactivity (Figure 4-13) is stronger due to the higher initial power in the bottom of the core. No reactivity data is available for a comparison to the GE results. 4.4 Conclusions The methods used in analyzing the LBT are the same as those .being proposed for use in performing reload design and licensing calculations at PECo. The results of PECo's analysis are consistent with the trends and magnitudes reported by GE. Good agreement is attained for core power and heat flux, core inlet flow, and core pressure. This is expected since the two methods are very similar, both utilizing 1-D core neutronics models. The differences between the RETRAN and BNL calculations are generally larger. Although not presented here, the results of i PECo's analysis are quite similar to the LBT results reported by the Tennessee Valley Authority (Reference 13). A significant ] l portion of the differences in the results may be attributed to d differences in the models and inputs used for evaluating the LBT by the other organizations. Not knowing all the input data used by the other organizations, it is difficult to conclude the 1 f precise reasons for the differences in the results. 4-8

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5.0

SUMMARY

The results of the analyses performed in this report demonstrate the ability of the Philadelphia Electric Company RETRAN mndel to accurately predict the course of a wide variety of transient events and that the model is applicable with a high degree of confidence to the evaluation of normal and abnormal transients for plant operational support and core reload analysis and licensing. 5-1

6.0 REFERENCES

1. ' Updated Final Safety Analysis Report - Peach Bottom Atomic Power Station Units 2 and 3', Vols. 1-9, Rev. 4, January, 1986. 2. J.H. McFadden et. al., 'RETRAN: A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems', Vols. I-III, Rev. 3, EPRI NP-1850-CCM, June 1987. 3. Philadelphia Electric Company, 'MCD-001: RETRAN Peach Bottom Model', Rev. 2, September, 1987. 4. Karl Hornik, Joseph A. Naser, 'RETRAN Analysis of the Turbine Trip Tests at Peach Bottom Atomic Power Station Unit 2 at the End of Cycle 2, EPRI NP-1076-SR, April, i 1979. 5. Warren Lyon, 'WREM: Water Reactor Evaluation Model (Revision 1)', NUREG 75/056, May, 1975. 6. General Electric, ' Qualification of the One -Dimensional Core Transient Model for Boiling Water Reactors', Volume 1, NEDO-24154, October, 1978. 7. General Electric, 'GE Reactor System and Plant Data Sheet', 257HA218, May, 1969. 8. Philadelphia Electric Company, ' Steady-State Thermal Hydraulic Analysis of Peach Bottom Units 2 and 3 Using the FIBWR Computer Code, PECo-FMS-001, February, 1985. 9. Philadelphia Electric Company, ' Philadelphia Electric Company's SIMULATE-E: A Nodal Core Analysis Program for Light Water Reactors', PECo-FMS-CCM-001, August, 1987. 6-1

10. Philadelphia Electric Company, ' Philadelphia Electric Company's SIMTRANE-E: A SIMULATE-E to RETRAN Datalink', PECo-FMS-CCM-002, September, 1987. 11. General Electric, ' Transient and Stability Tests at Peach Bottom Atomic Power Station Unit 2 at End of Cycle 2', EPRI NP-564, June, 1978. 12. M.S. Lu, et. al., ' Analysis of Licensing Basis Transients for a BWR/4', BNL-NUREG-26684, September, 1979. 13- 'BWR Transient Analysis Model Utilizing the RETRAN Program', TVA-TR81-01-A, April 7, 1983. 6-2}}

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