Text

Rev 0 to BWR-2 Transient Analysis Model Using Retran Code
	ML20235J143
Person / Time
Site:	Oyster Creek
Issue date:	09/03/1987
From:	Alammar M, Chin J, Furia R; GENERAL PUBLIC UTILITIES CORP.
To:	;
Shared Package
ML20235J123	List: ML20235J122; ML20235J143;
References
	TR-045, TR-045-R00, TR-45, TR-45-R, NUDOCS 8710010322
	Download: ML20235J143 (135)
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{{#Wiki_filter:_ _ _ _ _ _ _ - _ _ _ _ BWR-2 TRANSIENT ANALYSIS MODEL USING THE RETRAN CODE TR-045 Rey, O BA No.: 335400 M. A. Al ammar R. V. Furia Authors J. H. Chin l C. B. Mehta Date: September 3,1987 Approvals: 8 Uh $/J &7 Nuclear Analysis & Fuels Director Date GPU Nuclear i Upper Pond Road Parsippany, New Jersey 07054 hk PL P

TR-045 Rev. O Page 2 of 135 ABSTRACT A system transient model for the Oyster Creek Nuclear Generating Station is described. The model is based on the RETRAN-02 computer code u d primarily intended for analysis of reload transients. Model qualification included the simulation of nine startup tests. A representative application of the model to a typical Oyster Creek reload analysis (Cycle 10) with a comparison against vendor's results is also presented.

TR-045 Rev. O Page 3 of 135 BWR-2 TRANSIENT ANALYSIS MODEL USING THE RETRAN CODE TABLE OF CONTENTS Page

1.0 INTRODUCTION

9 1.1 Purpose................................................... 9 1.2 Brief Description of RETRAN............................... 9 1.3 Model Qualification....................................... 10 1.4 Model Application......................................... 11 1.5 Thermal Limits Evaluation................................. 11 2.0 THE BASIC RETRAN MODEL.......................................... 12 2.1 Model Geometry 14 2.1.1 Vessel Internals 14 2.1.2 Core Region........................................ 15 2.1.3 Recirculation Loops.............................. 16 2.1.4 Feedwater and Steam Lines.......................... 17 2.2 System Components......................................... 19 2.2.1 Recirculation Pumps................................ 19 2.2.2 Steam Separators................................... 19 2.2.3 Safety / Relief Valves............................... 20 2.2.4 Core Hydraulics.................................... 21 2.3 Trips and Control Models.................................. 22 2.3.1 Feedwater Control System........................... 22 2.3.2 Electric Pressure Regulator........................ 23

TR-045 Rev. O Page 4 of 135 TABLE OF CONTENTS (Cont.) Page 2.4 Steady State Initialization............................... 25 l 2.5 CPR Calculation........................................... 25 26 2.6 RETRAN Kinetics........................................... 3.0 STARTUP TESTS BENCHMARK......................................... 50 52 3.1 Pressure Regulator Test................................... 3.2 Level Setpoint Change..................................... 53 53 3.3 MSIV Closure Test......................................... l 3.4 Bypass Valve Test 54 3.5 Turbine Trip Test 55 3.6 Generator Trip Test 56 3.7 Recirculation Pumps Trip Test 57 3.8 Power-Flow Control Test 58 3.9 Isolation Condenser Test 60 4.0 REPRESENTATIVE RELOAD TRANSIENTS BENCEMARK...................... 81 4.1 Description of the Turbine Trip Without Bypass............ 82 4.1.1 Model Description......................................... 83 4.1.2 Sensitivity Studies....................................... 84 4.1.3 Turbine Trip Without Bypass Benchmark Results.............. 99 4.2 Main Steam Isolation Valve Closure Without Scram.......... 101 4.3 Feedwater Controller Failure (Max. Demand)........... 104

5.0 CONCLUSION

S..................................................... 134

6.0 REFERENCES

135

TR-045 Rev. O Page 5 of 135 LIST OF TABLIS Table Title Page 2.1 Description of Trip Elements......................... 28 2.2 RETRAN Boundary Conditions........................... 33 2.3 Volume Geometric Data................................ 34 2.4 Junctions Geometric Data............................. 36 2.5 Heat Conductor Geometric Data........................ 38 3.1 Comparison Between Plant & Model Response at 61 Beginning and End of Test Conditions 1 l l 4.1 TTWOBP Events Sequence............................... 106 4.2 Licensing Model Specific Parameters.................. 107 4.3 Code Options Used for Sensitivity Analysis........... 108 4.4 Summary of Sensitivity Analysis...................... 109 4.5 Summary of Code Options Used in Licensing Model...... 110 4.6 Hot Channel Parameters............................... 111

TR-045 Rev. O Page 6 of 135 LIST OF FIGURES Figure Title Page j 2.1 Oyster Creek RETRAN Noding Diagram................... 39 2.2 RETRAN Model of the Feedwater Control System......... 40 2.3 RETRAN Model of the Pressure Regulator Control....... 42 System - Steady State Conditions 3.1.1 Plant Us RETRAN - Pressure Setpoint Change Dome..... 62 Pressure 3.1.2 Plant Vs RETRAN - Pressure Setpoint Change Power.... 63 3.2.1 Plant Vs RETRAN - Level Setpoint Change Level Response 64 p 3.2.2 Plant Vs RETRAN - Level Setpoint Change Power Response 65 3.3.1 Plant Vs RETRAN - MSIV Closure Level Response....... 66 3.3.2 Plant Vs RETRAN - MSIV Closure Dome Pressure........ 67 3.4.1 Plant Vs RETRAN - Bypass Valve Test Dome Pressure... 68 3.4.2 Plant Vs RETRAN - Bypass Valve Test EPR Pressure.... 69 Signal L: 3.5.1 Plant Vs RETRAN - Turbine Trip Power Response....... 70 3.5.2 Plant Vs RETRAN - Turbine Trip Dome Pressure........ 71 3.6.1 RETRAN Vs Plant - Generator Trip Power Response..... 72 3.6.2 RETRAN Vs Plant - Generator-Trip Dome Pressure...... 73 3.6.3 Trip Timing & CV/BP Closure......................... 74 3.7.1 RETRAN Vs Plant - Pump Trip Tes' Pump Coastdown..... 75 3.8.1 RETRAN Vs Plant - Recirculation Flow (A)............ 76 3.8.2 RETRAN Vs Plant - Level Response.................... 77 3.8.3 RETRAN Vs Plant - Power Response.................... 78

TR-045 Rev. O Page 7 of 135 LIST OF FIGURES (Continued) Figure Title Page 3.8.4 RETRAN Vs Plant - Dome Pressure...................... 79 3.9.1 Isolation Consenser Test - Point & 1-D Kinetics Vs.... 80 Plent Power 112 4.1.1 TTWOBP - Power Response....,.......................... 4.1.2 TTWOBP - Dome Pressure............................... 113 4.1.3 TTWOBP - Relief Valve Flow........................... 114 4.1.4 TTWOBP - Level Response.............................. 115 4.1.5 TTWOBP - Core Flow................................... 116 4.1.6 TTWOBP - Avg Heat Flux............................... 117 4.2.1 MSIV ATWS - Normalized Power......................... 118 4.2.2 MSIV ATWS - Dome Pressure............................ 119 4.2.3 MSIV ATWS - Heat Flux................................ 120 4.2.4 MSIV ATWS - Core Flow 121 4.2.5 MSIV ATWS - Level Change (Downcomer)................. 122 4.2.6 MSIV ATWS - Feedwater Flow........................... 123 4.2.7 MSIV ATWS - Turbine Flow............................. 124 4.2.8 MSIV ATWE - Safety Valve Flow........................ 125 4.3.1 FWCF - Change in Downcomer Level..................... 126 4.3.2 FWCF - Power Response................................ 127 4.3.2A FWCF - Power Response, Biased to High Water Level TTP 128 4.3.3 FWCF - Dome Pressure Rise........................... 129 4.3.4 FWCF - Core Flow, Time Scale Biased to TTP........... 130

TR-045 Rev. O Page 8 of 135 LIST OF FIGURES (Continued) Figure Title Page \\ 4.3.5 FWCF - Change in Avg. Heat Flux, Time Scale Biased... 131 to TTP 4.3.6 FVCF - Bypass Flow - Time Scale Biased to TTP 4..... 132 4.3.7 FWCF - Relief Valve Flow - Time Scale Biased to TTP.. 133 \\

l TR-045 Rev. O Page 9 of 135

1.0 INTRODUCTION

1 l 1.1 Purpose The purpose of this report is to describe a systems analysis model of the Oyster Creek Nuclear Power Station's Nuclear Steam Supply l System (NSSS) which is a BWR-2, 1930 MWTH plant. The model is based on the RETRAN-02 MOD 4 Computer Code'*'. This model will be used to evaluate the plant's response to normal or abnormal transients. It will be used for operational and reload licensing support of the Oyster Creek plant in the areas where RETRAN has been reviewed and accepted by the NRC according to the applicable SERs. 1.2 Brief Description of RETRAN The RETRAN computer code is a one-dimensional thermal hydraulic code developed by EPRI as a tool for best estimate analysis of light water reactor systems. The Nuclear Regulatory Commission has reviewed RETRAN and issued a Safety Evaluation Repor (SER) which allows the code to be referenced in a licensing submittal. The l current code version used here is RETRAN 02-MOD 4 which reflects l l comments associated with the NRC & EPRI code reviews and suggestions 1 from code users. The code options and special models used are presented in Chapter 2, which have been kept the same throughout except as specifically noted. A detailed description of RETRAN is available in the literature and a knowledge of the code is assumed for the review of this document. The code is installed and controlled on GPUN computers using GPUN procedures which assures the required quality assurance / control.

TR-045 Rev. O Page 10 of 135 1.3 Model Qualification The Oyster Cieek RETRAN Model qualification was carried out according to the following program: \\ \\ .a. A benchmark against nine startup tests. b. A sensitivity study on the base licensing model for the limiting reload transient for Oyster Creek (Turbine Trip Without Bypass - TTWOBP) A benchmark against vendor's analysis of Cycle 10 reload c. transients ITWOBP, Feedwater Controller Malfunction, and MSIV Closure Without Scram. The nine startup tests were: plant response to level setpoint change, response to pressure setpoint change, five recirculation pumps trip, recirculation flow change, bypass valve test, MSIV closure test, generator trip (with bypass) Turbine trip with bypass, and isolation condenser test. The sensitivity studies involved perturbing more than 12 paramaters in a conservative manner and calculating their impact on the change of ACPR/ICPR where ACPR is the change in Critical Power Ratio (CPR) during the TTWOBP transient and ICPR is the initial CPR. Results from this study were used in establishing a conservative set of parameters, uncertainty margin, and a proper choice of code options to be used for the licensing model. .___m..

4 TR-045 ^ Rev. O s.. Page 11 of 135 .) 1.4 Model Application The model is primarily intended to analyze chapter 15 FSAR transients for reload applications. Conservative assumptions have been used towards this objective. 1.5 Thermal Limits Evaluation The parameter of interest in analyzing Chapter 15 FSAR transients for reload application is the critical power ratio. The critical o power at which the departure from nucleate boiling is calculated to occur has been adopted as the fuelecladding integrity safety limit. The uncertainties in monitoring the core operating state and in the procedures used to calculate critical power result in an uncertainty in the critical power. The CPR safety limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transitions a considering all uncertainties. The safety limit MCPR determined by GPUN utilizes the General Electric Company Thermal Analysis Basis, GETAB'*, which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the GE critical quality (X) boiling length (L), GEXL, correlation. A CPR of 1.07 is used by GPUN for the fuel claddirg integrity safety limit. l

TR-045 Rev. O Page 12 of 135 2.0 THE BASIC RETRAN MODEL The RETRAN model for Oyster Creek is based on the RELAP4 model developed by EXXON. It was then converted to RETRAN and expanded to contain the control systems, the steam lines, and the isolation condensers. The' model has been in use within GPUN for a number of years where it was further refined and expanded. The important recent refinements are the 1 separator /downcomer region, the steam lines re-noding, the core model and the isolation condenser model. The changes are basically a detailed noding of those regions in order to capture the important phenomena that take place during chapter 15 type transients. The present model consists of 61 volumes, 84 junctions and 40 conductors. The main noding diagram is shown in Figure 2.1. It includes a lower plenum, 12 active core volumes plus a core inlet and a core exit volumes, a core bypass, upper plenum, stand Pipes, separators, upper and lower downcomers and a steam dome. It also contains 8 nodes steam lines, steam manifold, steam bypass line, a condenser volume and a turbine chest volume. The steam system valves are independently modeled, i.e. Main Steam Isola-tion Valves (MSIVs), Turbine Stop Valve (TSV), Turbine Control Valve (TCV) and Turbine Bypass Valves (TBV). The five recirculation loops are modeled as two loops, representing four pumps and one pump respectively. i l The safety and relief valves are modeled as junctions, with contraction coefficients connected to a containment volume at atmospheric pressure. The Moody critical ficw option is chosen for all choked junctions. 1

l TR-045 Rev. O Page 13 of 135 The feedwater control system model is shown in Figure 2.2. It is a three element proportional plus integral (PI) controller requiring downcomer level, steam flow and feedwater flow for input. The pressure regulator-system is shown in Figure 2.3. It consists of the Electrical Pressure Regulator (EPR), associated control valve relays and servomotor, the nine bypass valves and the turbine. Its output is the control valve / bypass valves position that control the relevant junctions areas while output steam flow is calculated by the code. The data required for setting up the model was obtained from as built plant drawings, equipment manuals, field tests documents and operational procedures. A brief discussion of the major systems, and the rational behind the noding of the model is presented in the following sections.

i TR-045 Rev. O Page 14 of 135 2.1 Model Geometry In this chapter, a brief description of the model will be given. The modeling philosophy was built on GPUN experience in simulating a l wide range of plant transients over the years. The excellent agreement obtained with startup tests (Chapter 3) shows the model l high quality and robustness in simulating a wide range of scenarios. 2.1.1 Vessel Internals The region below the lower core support plate is represented as a single volume (lower plenum). Passive conductors are used to represent the lower vessel head and lower plenum internal structures. The upper plenum, standpipes and separators are modeled as separate volumes. A two-sided passive heat conductor is used to represent the upper plenum s'aroud between the upper plenum and the lower downcomer whiin single one-sided conductors attached to the standpipes and separators volumes are used to represent the standpipes and the separators. The area above the separators is divided into two volumes, one representing the space between the separators and the dryers inlet while the other represents the steam dome and the dryers area. Passive heat conductors attached to the latter volume are used to represent the vessel walls and the dryers materials.

IR-045 Rev. O Page 15 of 135 The downcomer (1:nside vessel barrel and outside of the core 4 shroud and separators) is divided into two volumes. The 1 upper section of the downcomer (UDC) is the region above the { t feedwater sparger to the top of the separator. This section utilizes the "non-equilibrium" option of RETRAN. Use of a i non-equilibrium model in this volume is important for pressurization transients since an equilibrium model representation would not allow the steam in the volume to superhe'at thus causing an underprediction of the pressurization rate. The lower section of the downco.ner (LDC) is the region between the feedwater sparger to the recirculation loops suction side. This is the subcooled region of the downcomer due to the entering feedwater through the spargers. The feedwater sparger is a ring with rectangular cross-section and one-inch holes through which feedwater is discharged into the downcomer. 2.1.2 Core Region The reactor core is represented by 12 active (fueled) core volumes and two nonfueled volumes. In addition, a core bypass volume is represented. A heat conductor is used in each of the 12 active volumes while two-sided passive heat

TR-045 Rev. O Page 16 of 135 conductors are used between che bypass volume and the lower downcomer and core volumes to represent core shroud and fuel bandles cans. The transient variatione in flow, enthalpy 1 and pressure' in a fuel bundle are obtained using a RETRAN l l " Hot Channel" model for use in calculating tb3 change in l critical power ratio. A hot chcanel deck is set up for each fuel bundle geometry type and consists of 24 active fuel volumes and two unheated bundle volumes with a conductor simulating the fuel rod segment in the volume. A three volume average core with an average core inlet and core outlet volumes are used with a three volume bypass region to fully represent the hot channel core model. A time dependent upper and lower plenums volumes driven by the system model respence output are used as boundary conditions. Average fuel conductors and average core to bypass conductors representing fuel bundle cans are also used. 2.1.3 Recirculation Loops The five recirculation loops at OC are modeled as two loops representing one and four loops. Each loop consists of a suction volume, pump and discharge volume. The two loop

TR-045 Rev. O Page 17 of 135 modeling is used to analyze a single recirculation pump trip. The RETRAN ' transport delay' option is used in the recirculation loop piping, the lower downcomer and lower plenum to allow representation of fluid temperature movement as a front. The walls of the loops piping are modeled as one-sided passive conductors. 2.1.4 Feedwater and Steam Lines The feedwater lines beginning at the regulating valves, immediately downstream the three electric feed pumps, and passing through the high pressure heaters to the reactor vessel are modeled as one volume with a positive RETRAN fill junction at the inlet. The fill junction flow is controlled by the feedwater control system while the inlet enthalpy is calculated by RETRAN during steady state initialization. The RETRAN transport delay option is used to represent the proper time delay fc r feedwater enthalpy changes to reach the reactor vessel. The two steam lines are represented individually from the steam dome to the high pressure turbine through the four control valves and to the main condenser through the nine bypass valves. Each steam line is represented by eight volumes divided by the Main Steam Isolation Valves (MSIVs).

TR-045 Rev. O Page 18 of 135 The steam lines inboard of the MSIVs are divided into four volumes with the third volume centered around the safety and' relief valves. Those valves are modeled as junctions con-nected to a containment volume. The outboard sections of each steam line is divided into four volumes with the last I one connected to a common steam manifold volume. The Turbine Stop Valves (TSVs) and the Turbine Control Valves t (TCVs) are modeled separately by single valves with the TCV connected to the high press turbine chest volume. The nine Bypass Valves (BPVs) are represented by a single valve connecting the bypass line (from the steam manifold) to a containment volume representing the main condenser. The areas of the TCV and the BPV are controlled by the pressure regulator control system, while the flow to the turbine / condenser is calculated by RETRAN.

TR-045 Rev.-0 Page 19 of 135 2.2 System Components 'Some areas of the model require special consideration during input preparation. The modeling of these special components can signifi-cantly influence the behavior of operational transients. These components are described below. 2.2.1 Recirculation Pumps The OC recirculation pumps are Byron / Jackson type which are close to the Bingham pumps modeled in RETRAN. The latter option is used with the m3 ment of inertia, friction factors and rated speed corresponding to actual pumps data. The model pump performance was benchmarked against actual pumps trip testo where the required parameters, especially coastdown time, friction coefficients and flow were tested. All Reactor Protection System (RPS) pump trips at Oyster Creek trip the M/G set and the pump-motor system. 2.2.2 Steam Separators The steam separators play a significant role in simulation of a BWR since t'ney couple the steam dome and the core which

TR-045 Rev. O Page 20 of 135 are the primary areas of interest. The 151 individual steam separators are modeled as a single component using RETRAN two region separator option, which describe the variation in j j I carryunder as a function of separator inlet quality and downcomer mixture level. This model is used with a zero initial carryover since the dryers are accounted for in the separator model, while the carryunder was set to 0.1 percent for best estimate analyses which was the carryunder (cu) measured at full power during the startup test program'*'. A conservative value of 0.25% for the carryunder fraction is used for licensing applications. The separator inertia used, which has the largest impact on system response during the limiting pressurization transients, was supplied by the vendor and applied at 1 separator inlet junction. Loss coefficients for the various junctions were calculated in detail from vendor drawings using existing eagine. ring practices. i l 1 i 2.2.3 Safety / Relief Valves Oyster Creek has five electromatic relief valves (EMRVs) and 16 spring loaded safety valves (SVs). The EMRVs are l J arranged in two groups of 2 and 3 valves each while the l 1 1 i l s

TR-045 Rev. O Page 21 of 135 safety valves are arranged in 4 groups of 4 valves each. The valves of each group have a common setpoint and each group is represented by a junction connecting the steam line to a containment (sink) volume in the RETRAN model. The area of the junction is taken as the flow area of the valve times the number of valves being modeled. When the valve is-opened with the steam line pressurized, the junction flow becomes choked and the Moody critical flow option is chosen in RETRAN for choked flow calculation. Contraction coefficients are used on valve junctions to get the specified flow at the reference pressure. 2.2.4 Core Hydraulics The core region was modeled using the standard single stream, compressible flow form of the momentum equation (MVMIX=0) and the Baroczy two-phase friction multiplier (JTPMJ=0). The frictional loss coefficients were calculated from detailed vendor drawings and using existing engineering practice. The model pressure drop across the lower core tie plate was benchmarked agninst actual plant measurement while core mid plane pressure WP.J compared to vendor's supplied data Loss coefficients for the internal core junctions were determined to give the required core pressure at the l given vendor's bypass flow distribution.

TR-045 Rev. O Page 22 of 135 2.3 Trips and Control Models The Oyster Creek RETRAN model has the required Reactor Protection System (RPS) trips and other trips to simulate various transients and equipment failure. Table 2-1 shows a listing of trips used. The control models used represent the two control systems at Oyster Creek, namely the feedwater controller and the pressure regulator. The various parameters and settings of the different' control blocks were taken,either from actual test conditions or vendor's data. A description of the two systems is given below. 2.3.1 Feedwater Control System The feedwater controller allows for either a one-element mode where the controller output is only a function of the difference in setpoint and sensed level or a three-element mode which is normally used and adds en additional steam-feed mismatch to the icvel error. The Lnree-element mode is the one modeled' which basically consists of a lead-lag compensation on the total error followed by a 1

TR-045 Rev. O Page 23 of 135 proportional plus integral (PI) controller. The primary controller constants are the steam-feed mismatch gain, the proportional band and the reset rate. Those parameters were based on actual plant settings while the rest are based on system description documents and field settings. The three electric feedwater pumps are not modeled but a RETRAN positive fill junction is used to represent feedwater flow as controlled by the control system. Figure 2-2 gives the control block diagram for the system. The sensed liquid model is measured using RETRAN Liquid Level calculation adjusted to reflect area changes in the separator / standpipes / steam dome regions. This adjustment is implemented using control blocks and a volume to height table. 2.3.2 Electric Pressure Regulator (EPR) The four turbine control valves and the nine bypass valves are represented as single outlet junctions l respectively. The EPR controls the respective

-n .: /,, j 3, t r;' ) ,.( 1 ,e >l 3 f . :Tk-04 5 } i. ^h ^ p 'T Rev. O l ,/ Page 24 of 135 i i ? ,/ junction area's while the code calculates the relevant flow. 4 a .) /, The turbine hydraulic model consisi;s of a lag and lead / lag l I e 1 .g i ? blocks the output of which represeMs the hydraulic torque j / which is assumed to be equal to the load torquef The 1 s i g 1 1 t f ? 7, [ turbine inertia, the spbed governor and tie load limit are ,9 4 4 f l >r ( A loss of load Erip is modeled by included in the model. r setting the load torque to zero upon actijatian of the a / trip. The EPR controller starts by developing a pressure / f error signal between t!,e steam' mani5cid pressure.(vol. 217) t f i i t / /jl: / t \\ 0 . i,> ,end the pressure setpoinh. This error signal'is then used 4 L ,3 ) 4 1 1!,.< p i to modulate the control valve re' lay which in turn controls a \\ l' secondary valve reAay and the control valve servGrr9 tor, y 4 1 .i Sis error, signal also controls the nine bypass valves l >/ ? dapending on the status of the control valves,' which is also i i l t i> controlled by the turbine load demand signal. l [ The overall system bioch diagram is shown in irigure 2.3 / 1 'i where the rest of the parameters were taken from vendor f t document:s. j s Oyster Creek does not,ase an automatic load f %1owing h j \\ capacity, therefore these is no elaborate recirculation f ( control system connectoq 'to turbine /EPR controller. ./ l' v g g 1 ,e j ./ A l k. 5 n ,.1 i e l[.[ .\\ i sW t

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Rev. 0

.Page 25 of'135 2.4 Steady State Initialization The RETRAN steady state initialization option'was used to initialize' the model. The boundary conditions'specified were dome pressure, steam manifold pressure and: lower downcomer enthalpy. Mass flow-rates at' measured locations were used as-boundary: conditions, namely-recirculation flow, steam flow and feed flow while; core bypass' flow; fraction was based on vendor's documentation. ;A typicalJset of boundary conditions is shown in' Table 2-2. The lower downcomer. enthalpy is adjusted to give the required carryunder. A listing'of geometric parameters is shown in Tables 2.3-2.5. i 2.5 Critical Power Ratio The critical. power ratio (CPR) is calculated for each transient that is analyzed. The RETRAN hot channel model, described in section-2.1.2, provides the time dependent pressure, flow and.enthalpy for-use with the GEXL correlation. The hot channel model is driven j using time dependent normalized power and upper'and lower plenum conditions (pressure, enthalpy, etc.) from the RETRAN system model. a j l 1 ] ) l i

1 TR-045 11ev. O Page 26 of 13S. The hot channel'uses'a fixed set of peaking. factors and initia* hot channel flow to. place.the hot channel on.the CPR operating' limit. The hot channel flow being obtained from the core 3-D simulator model. CPR'is calculated using a GPUN developed code, RACE'**l, which used the time dependent' parameters from the RETRAN hot channel calculation. RACE uses the inlet enthalpy and bundle average flow 1 l at each time step along with the power shape used in the hot channel to calculate the nodal quality and boiling length. The critical ~ power is calculated at each time step and the ACPR for the transient is determined. 2.6 RETRAN Kinetics The RETRAN-02 MOD 4 code has both a point and-1D kinetics model. The i input generation for both of these models is described in reference 10. The use of point-or ID kinetics for a given transient depends upon the adequacy of the kinetics model for that' transient. Typically, pressurization transients are analyzed using 1D kinetics i due to the changes in the axial power. shape that occur during the course of these events. For Oyster Creek, these are the most limiting, transients analyzed with RETRAN. Most other transients'can. j be analyzed with the point kinetics model.- The description of the l transient will state if 1D kinetics were used. i i j ._--._L__-_-_._________O

TR-045 Rev. O Page 27 of.135 The kinetics input were calculated to reflect the nature of the analysis being performed. The kinetics input for the startup tests i were calculated from " core models" set up to match core conditions at the time the tests were run. Scram times used in the kinetics calculation were based on measurements taken during the tests. This t I was consistent with the best estimate model approach used in these analyses. The sensitivity studies and comparison to the vendor's licensing analysis introduced conservatism in the analyses. i Transients were run using kinetics. input calculated at EOC with all rods fully withdrawn and the reactor at full power and full flow. 1 I The EOC exposure distribution was based on a Haling depletion of the l cycle to decrease the scram bank worth at the beginning of the scram. The scram bank insertion time line used a maximum allowed 1 scram times. This is consistent with the vendor analysis which facilitates ccmparison and provides acceptable conservatism for licensing analyses.

TR-045 Rev. O Page 28 of 135 Table 2.1 Description of Trip Elements-Trip ID Action Cause of Trip Actuation 01 Problem End Time Transient Time > Setpoint 02 Initiate Fill Tables Transient Time > Setpoint 03 Trip Recirculation Pumps Transient. Time > Setpoint, or Volume (104) (Steam Dome) Pressure > Setpoint, or Control block-143 (Level TAF) < Setpoint 06 Initiate Scram Control block - 143 (Level,TAF) < setpoint, or Volume (104) (steam dome) Pressure > setpoint, or Normalized flux > setpoint, or Trip #12 (Turbine Trip) activated, or Trip # 15 (Partial MSIV closure) activated. 07 Core Spray Pumps On Control block -143 (Level, TAF) l < setpoint

t

\\

'r - TR-045 Rev.10~ Page 29 of 135 . Table 12.1 Description of. Trip Elements (Cont'd) Trip'ID Action Cause of Trip Actuation i l l ~ l 08' Core Spray Injection Valve Volume (104).(Steam' Dome): Pressure < setpoint. 09 Turn ADS on

Volume '(102) ~ (upper plenum)

Liquid Level ('setpoint

10 Open Two Relief Valves Volume (104) (Steam Dome) Pressure

> setpoint -10 close Two Relief Valves Volume (104) (Steam Dome) Pressure l < setpoint-11 Turn Emergency Condenser On Volume-(104) (Steam Dome) Pressure > setpoint, or-Control block-143 (Level, TAF)< l setpoint. I i 12 Turbine Trip Control block-143 (Level', TAF)': I > setpoint or Junction 290 (TCV) mass floe < setpoint 4

TR-045 Rev. 0-Page 30 of 135 Table 2.1 Description of Trip Elements (Cont'd) Trip ID Action Cause of Trip Actuation 13 Open Three Releif Valves Volume (104) (Steam Dome) Pressure ( setpoint -13 Clos Three Belief of Valves Volume (104) (Steam Dome) Pressure < setpoint. i 15 Close MSIVS Volume (104) (Steam Dome) 1 Pressure > setpoint, or ] l Control block-143 (Level, TAF) l j l < setpoint l 16 Close Turbine Bypass Valve Transient Time > setpoint 17 Isolate Emergency Condenser Junctions 120, or 122 (Isolation Condensers inlet) mass flow > setpoint 31 Open Safety Valves Bank #1-Volume (104) (Steam Dome) Pressure > setpoint

TR-045 Rev. O Page 31 ofL135 Table 2.1 Description of Trip Elements (Cont'd) Trip ID Action Cause of Trip Actuation -31 Close Saf"ety Valves Bank #1 Volume (104) (Steam Dome) Pressure < setpoint 32 Open Safety Valves Bank #2 Volume (104) (Steam Dome) Pressure > setpoint. -32 Close Safety Valves Bank #2 Volume (104) (Steam Dome) Pressure < setpoint. 35 Open Safety Valves Bank #3 Volume (104) (Steam Dome) Pressure > setpoint. l 1 -35 Open Safety Valves Bank #3 Volume (104) (Steam Dome) Pressure < setpoint. 36 Open Safety Valves Bank #4 Volume (104) (Steam Dome) Pressure > setpoint. -36 Close Safety Valves Bank #4 Volume (104) (Steam Dome) Pressure < setpoint.

TR-045. Rev. 0 j Page 32 of 135 l 1 Table 2.1 Description of Trip Elements (Cont'd) T l l 1 Trip ID Action Cause of Trip Actuation i ) i 33 Open Emergency Condenser #1 Shell Side Valve l Volume (413) (shell volume) press > setpoint ) -33 Close Emergency Condenser #1 Shell Side Valve .i Volume (413) (shell volume) Press <.setpoint l l 36 Open Emergency Condenser #2 Shell Side Valve j ) 1 Volume (433) (shell volume) l Pressure > setpoint -36 Close Emergency Condenser #2 Shell Side Valve Volume (433) (shell volume) l Pressure < setpoint l

TR-045 Rev. O Page 33 of 135 Table 2.2 RETRAN Boundary Conditions 1930 MW (Thermal) Power Feedwater Flow 7.25 MLBM/Hr Dome Pressure 1034.7 psia Turbine Inlet Pressure 964.7 psia Lower Downcomer Enthalpy 518.26 Btu /lbm Upper Downcomer Liquid Level 160 inches IAF j n Recirculation Flow 61 x 10' lbm/hr Core Bypass flow fraction 12.3%

l I

l 1 i ) 'l

o TR-045 Rev. O Page 34 of 135 l TABLE 2.3 VOLUME GEOMETRIC DATA VOL FLUID FLOW BOTTOM VOLUME NOTES DESCRIPTION VOLUME AREA ELEV HEIGHT l (FT3) (FT2) (FT) 001 53.496 53.48 16.35 1.09 NONE CORE INLET SECTION 002 59.692 59.562 17.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION 003 59.692 59.562 18.44 1.00 ENTH. TRNSPT, ACTIVE CORE SECTION 004 59.692 59.562 19.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION 005 59.692 59.562 20.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION 006 59.692 59.562 21.44 1.00 ENT3. TRNSPT. ACTIVE CORE SECTION 007 59.692 59.562 22.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION l 008 59.692 59.562 23.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION 009 59.692 59.562 24.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION 010 59.692 59.562 25.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION 011 59.692 59.562 26.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION-012 59.692 59.562 27.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION 013 59.692 59.562 28.44 1.00 ENTH. TRNSPT. ACTIVE CORE SECTION 014 85.032 71.456 29.44 1.19 NONE CORE OUILET SECTION 015 807.49 56.670 16.70 14.24 NONE CORE BYPASS 101 1786.4 107.0 0.0 17.0 TRNSPT DELAY LOWER PLENUM I 102 783.64 207.09 30.63 3.80 NONE UPPER PLENUM I 103 1149.5 287.375 44.51 4.1 NONE SEPARATOR / DRYER AREA 104 2996.1 194.0 48.41 15.44 NONE STEAM DOME 105 1997.4 145.0 10.83 25.03 TRNSPT DELAY LOWER DOWNCOMER 106 180.56 30.24 34.41 5.99 NONE STAND PIPES 107 448.5 60.06 40.38 4.23 FUNCT SEPER. SEPARATOR 108 1413.27 211.3 35.36 9.25 NON-EQ. PRZ UPPER DOWNCOMER 201 71.0 2.535 33.30 15.97 NONE NORTH STEAM LINE 202 72.35 2.535 33.31 15.975 NONE SOUTH STEAM LINE 203 71.0 2.535 17.35 15.97 NONE NORTH STEAM LINE 204 72.35 2.535 17.36 15.975 NONE SOUTH STEAM LINE 205 71.0 2.535 1.40 15.97 NONE NORTH STEAM LINE 206 72.35 2.535 1.405 15.975 NONE SOUTH STEAM LINE 207 71.0 2.535 -14.55 15.97 NONE NORTH STEAM LINE 208 72.35 2.535 -14.55 15.975 NONE SOUTH STEAM LINE 209 74.06 2.535 -15.86 1.32 NONE NORTH STEAM LINE 210 74.06 2.535 -15.86 1.325 NONE SOUTH STEAM LINE 211 74.06 2.535 -17.17 1.32 NONE NORTH STEAM LINE 212 74.06 2.535 -17.17 1.325 NONE SOUTH STEAM LINE 213 74.06 2.535 -18.47 1.32 NONE NORTH STEAM LINE 214 74.06 2.535 -18.47 1.325 NONE SOUTH STEAM LINE 215 74.06 2.535 -19.78 1.32 NONE NORTH STEAM LINE ________m_

TR-045 Rev. O Page 35 of 135 TABLE 2.3 (CONT'D) VOLUME GEOMETRIC DATA VOL FLUID FLOW BOTTOM VOLUME NOTES DESCRIPTION VOLUME AREA ELEV HEIGHT (FT3) (FT2) (FT) 216 74.06 2.535 -19.78 1.32 NONE SOUTH STEAM LINE 217 363.23 5.66 -19.78 19.92 NONE STEAM MANIFOLD 218 328.7 1.755 -19.78 11.49 NONE BYPASS STEAM LINE 219 1.E+6 1.E+3 -19.78 1.E+3 TIME DEPEND CONDENSER, ATMOSPH. '] 220 1.E+6 1.E+3 -19.78 1.E+3 TIME DEPEND TURBINE 221 67.27 7.46 -5.05 1.34 NONE SIM LINE(TSV-TCV) 301 675.40 12.44 -27.08 41.09 TRNSPT DELAY RECIRC LOOPS, SUCTION 302 201.60 12.44 -24.92 3.30 NONE RECIRC PUMPS 303 438.60 12.44 -22.91 29.16 TRNSPT DELAY RECIRC LOOPS,DISCHGE 304 168.85 3.11 -27.08 41.09 TRNSPT DELAY RECIRC LOOP, SUCTION 305 50.4 3.11 -24.92 3.30 NONE RECIRC PUMP-306 109.65 3.11 -22.91 29.16 TRNSPT DELAY RECIRC LOOP,DISCHGE 401 122.22 0.74 44.61 30.47 NONE EMERG COND A.STMLINE f 402 63.78 1.29 66.8 3.78 ENTH TRNSPT EMERG COND A TUBE 403 36.78 0.634 54.91 12.25 TRNSPT DELAY EMERG COND A, RETURN 404 40.63 0.498 7.47 47.85 TRNSPT DELAY EMERG COND A, RETURN J 405 3038.56 433.0 67.97 7.0 NONE EMERG COND A.SHELL 421 122.22 0.74 44.61 30.47 NONE EMERG COND B,STMLINE 422 63.78 1.29 66.8 3.78 ENTH TRNSPT EMERG COND B, TUBE 423 36.78 0.634 54.91 12.25 TRNSPT DELAY EMERG COND B, RETURN 424 40.63 0.498 7.47 47.85 TRNSPT DELAY EMERG COND B,RETUPJi 425 3038.56 433.0 67.97 7.0 NONE EMERG COND B,SHELL 601 1273.06 3.944 -34.91 70.0 TRNSPT DELAY FEEDWATER LINE j 1 l I i i l l l J

TR-045 Rev. O Page 36 of 135 l TABLE 2.4 JUNCTIONS GEOMETRIC DATA JUNC CONNECT FLOW ELEV INERTIA LOSS NOTES DESCRIPTION VOLUME AREA-HEIGHT COEFF FROM TO (FT2) (PT) (1/FT) 001 1 2 53.48 17.44 'O.01696 0.4688 HORIZ' CORE INLET. 002 2 3 59.692 18.44 0.01675 0.4688 HORIZ ACTIVE CORE.VOLS-003 3 4 59.692 19.44 0.01675 0.4688 HORIZ ACTIVE CORE VOLS. 004 4 5 '59.692 20.44 0.01675 0.4688 HORIZ ACTIVE CORE,VOLS 005 5 6 59.692 21.44 0.01675 0.4688 HORIZ ACTIVE CORE VOLS 006 6 7 59.692 22.44 0.01675 0.4688 HORIZ " ACTIVE CORE VOLS 007 7 8 59.692 23.44 0.01675 0.4688 HORIZ ' ACTIVE CORE VOLS H 008 8 9 59.692 24.44 0.01675 0.4688 HORIZ . ACTIVE CORE VOLS 009 9 10 59.692 25.44 0.01675 0.4688 HORIZ ACTIVE. CORE VOLS 010 10 11 59.692 26.44 0.01675 0.4688 HORIZ ' ACTIVE CORE VOLS 011 11 12 59.692 27.44 0.01675 0.4688 HORIZ ACTIVE CORE VOLS 012 12 13 59.692 28.44 0.01675 0.4688. HORIZ ACTIVE CORE VOLS 013 13 14 59.692 29.44 0.01688 0.4688-HORIZ. CORE OUTLET-014 14 102 71.42 30.63 0.02731 4.09 HORIZ CORE-UPPER PLNM 015 15 102 56.67 30.74 0.1422 10.0 HORIZ BYPASS-UPR PLNM 016 1 15 1.4423 16.89 0 09056 -1.0 HORIZ CORE INLET-BYPS 101 101 15 0.6384 16.70 0.09056 -1.0 HORIZ ~ LWR PLNM-BYPASS 102 101 1 11.377 16.35 0.2019 4.00 HORIZ LWR PLNM-CORE 103 102 106 30.24 34.42 0.108 0.426 HORIZ-UPR PLNM-STDPIPE 104 103 104 145.0 48.51 0.3209 2.96 HORIZ SEP/DRYE-DOME 105 108 105 52.6 35.76 0.096 -1.0 _VRT/NSLP UPR-LWR DNCMRS 106 105 301 12.44 14.00 3.180 1.045 VERT LWR DNCMR-RECIRC 107 105 304 3.110 14.00 12.700 1.045-VERT LWR DNCMR-RECIRC 108 104 201 2.535 49.25 5.58 0.116 VERT DOME-STM LINE 109 104 202 2.535 49.27 5.69 0.116 VERT DOME-STM LINE 110 106 107 26.95 40.39 0.7419 1.616 VRT/NSLP STDPIPE-SEPRTR 111 107 108 24.01 44.31. 0.047 -1.0 VRT/NSLP SEP-UPR DNCMR,STM 112 107 108 26.68 40.50 0.098 1.335 VRT/NSLP SEP-UPR DNCMR, LIQ-113 108 103 56.4 44.61 0.036 0.187 VRT/NSLP~ UPR DNCNR-DRYR 120 103 401 0.4417 45.01 111.6 4.43 VERT VESL-EC STM LINE 122 103 421 0.4417 45.01 111.6 4.43 VERT VESL-EC STM LINE 201 201 203 2.535 33.30 11.04 0.2025 VERT N. STM LINE 202 202 204 2.535 33.31 11.26 0.2025 VERT S. STM LINE 203 203 205 2.535 17.35 11.04 0.0992. VERT N. STM LINE 204 204 206 2.535 17.36 11.26 0.0992 VERT S. STM LINE 205 205 207 2.535 1.40 11.04 0.2025 . VERT N.,STM LINE 206 206 208 2.535 1.405 11.26 0.2025 VERT S. STM LINE 207 207. 209 2.535 -14.55 11.27 -1.0 VERT N. STM LINE-208 208 210 2.535 -14.55 11.38 -1.0 VERT S. STM-LINE 209 209 211 2.535 -15.86 11.5 0.66 HORIZ N. STM LINE l 210 210 212 2.535 -15.86 11.5 0.66 HORIZ S. STM LINE 211 211 213 2.535 -17.17 11.5 0.04 HORIZ N. STM LINE l 4 _L______--_-

I -l TR-045 Rev. O Page 37 of 135 f TABLE 2.4 (CONT'D) JUNCTIONS GEOMETRIC DATA JUNC CONNECT FLOW ELEV INERTIA LOSS NOTES DESCRIPTION VOLUME AREA HEIGHT COEFF FROM TO (FT2) (FT) (1/FT) 212 212 214 2.535 -17.17 11.5 0.04 HORIZ S. STM LINE 213 213 215 2.535 -18.47 11.5 0.04 HORIZ N. STM LINE l 214 214 216 2.535 -18.47 11.5 0.04 HORIZ S. STM LINE 215 215 217 2.535 -19.78 7.36 0.1625 HORIZ STMLINE-MANIFOLD 216 216 217 2.535 -19.78 7.36 0.1625 HORIZ STMLINE-MANIFOLD 217 217 218 1.75 -17.39 58.82 2.010 HORIZ MANIFOLD-BYPASS 1 218 218 219 1.4 -8.30 3.15 0.0 CHOKED BYTASS-CONDENSER 291 217 221 6.36 -5.05 2.49 1.0 VERT MANIFOLD-STMLINE 290 221 220 5.58 -5.05 1.8 -1.0 VERT STMLINE-TURBINE 240 203 219 0.181 18.35 1.0 0.0 CHOKED 2 RELIEF VALVES 241 204 219 0.271 18.35 1.0 0.0 CHOKED 3 RELIEF VALVES 250 203 219 0.362 18.35 1.0 0.0 CHOKED SAFETY VALVES #1 251 203 219 0.362 18.35 1.0 0.0 CHOKED SAFETY VALVES #2 252 204 219 0.362 18.35 1.0 0.0 CHOKED SAFETY VALVES #3 253 204 219 0.362 18.35 1.0 0.0 CH0KED SAFETY VALVES #4 301 301 302 12.44 -24.92 2.83 0.891 HORIZ SUCTION-PUMP 302 302 303 12.44 -22.91 2.115 1.045 HORIZ PUMP-DISCHARGE 303 303 101 12.44 6.25 2.135 1.90 HORIZ RECIRC-LWR PLNM ) 304 304 305 3.110 -24.92 11.32 0.891 HORIZ SUCTION-PUFT i 305 305 306 3.110 -22.91 2.115 1.045 HORIZ PUMP-DISCHARGE 306 306 101 3.110 6.25 2.135 -1.0 HORIZ RECIRC-LWR PLNM 401 401 402 1.411 70.57 130.76 2.35 VRT/NSLP STMLINE-TUBE,EC A 402 402 403 0.6342 67.15 64.89 4.04 VRT/NSLP TUBE-RETURN,EC A 403 403 404 0.4985 55.31 127.49 1.04 VRT/NSLP RETURN, EC A 404 404 304 0.371 7.87 90.48 2.5 VRT/NSLP RTURN-RECIRC,EC A 405 405 219 4.27 74.96 0.013 1.0 VERT SHELL OUTLET, A l 406 0 405 1.0 68.07 0.0 0.0 VERT SHELL INLET, A 407 404 304 0.001 7.87 90.48 -1.0 VERT RTRN-RECIRC,LKGE 408 405 219 0.01 74.96 0.013 -1.0 VERT SHELL LKGE 409 0 405 1.0 68.07 0.0 0.0 VERT SHELL INLT, LKGE 421 421 422 1.411 70.57 130.76 2.35 VRT/NSLP.STMLINE-TUBE,EC B 422 422 423 0.6342 67.15 64.80 4.04 VRT/NSLP TUBE-RETURN,EC B 423 423 424 0.4985 55.31 127.49 1.04 VRT/NSLP. RETURN, EC B 424 424 301 0.371 7.87 90.48 2.5 VRT/NSLP RTURN-RECIRC,EC B 4 25 425 219 4.27 74.96 0.013 1.0 VERT SHELL OUTLET, B 426 0 425 1.0 68.07 0.0 0.0 VERT SHELL INLET, B 427 424 301 0.001 7.87 90.48 -1.0 VERT-RTRN-RECIRC.LKGE 428 425 219 0.01 74.96 0.013 -1.0 VERT SHELL LKGE 429 0 425 1.0 68.07 0.0 0.0 VERT SHELL INLT, LKGE i 601 0 601 1.0 -34.91 0.0 0.0 HORIZ FEEDWATER INLET 602 601 105 1.988 35.09 27.87 4.994 VERT FW LINE-VESSEL 701 0 102 0.61 31.42 0.0 0.0 HORIZ CORE SPRAY

TR-045 Rev. O { Page 38 of 135 i l TABLE 2.5 HEAT CONDUCTOR GEOMETRIC DATA HEAT VOLUME GE0 METRY SURFACE AREA VOLUME CONDUCTOR COND ON : TYPE LEFT RIGHT DESCRIPTION LEFT RIGHT (FT2) (FT2) (FT3) j 001 0 2 CYLIND 0.0 4416.13 45.61S FUEL RODS 002 0 3 CYLIND 0.0 4416.13 45.615 FUEL RODS 003 0 4 CYLIND 0.0 4416.13 45.615 FUEL RODS 004 0 5 CYLIND 0.0 4416.13 45.615 FUEL RODS 005 0 6 CYLIND 0.0 4416.13 45.615 FUEL RODS j 006 0 7 CYLIND 0.0 4416.13 45.615 FUEL RODS 007 0 8 CYLIND 0.0 4416.13 45.615 FUEL RODS 008 0 9 CYLIND 0.0 4416.13 45.615 FUEL RODS j 009 0 10 CYLIND 0.0 4416.13 45.615 FUEL RODS 1 010 0 11 CYLIND 0.0 4416.13 45.615 FUEL RODS 011 0 12 CYLIND 0.0 4416.13 45.615 FUEL RODS 012 0 13 CYLIND 0.0 4416.13 45.615 FUEL RODS 013 2 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 014 3 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 015 4 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 016 5 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 017 6 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 018 7 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS j 019 8 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 020 9 15 CYLIND 954.8 978.26 6.456 FUEL CAN-BYPASS l 021 10 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 022 11 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 023 12 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 024 13 15 CYLIND 954.8 978.26 6.458 FUEL CAN-BYPASS 101 0 101 SLAB 0.0 584.4 446.4 VESS2L, LOWER PLNM 102 0 101 SLAB 0.0 845.3 64.4 LOWER PLNM INTERNALS 103 15 105 CYLIND 832.68 847.46 109.2 CORE SHROUD 104 105 0 CYLIND 1400.46 0.0 853.9 VESSEL, LOWER DNCMR 105 102 105 CYLIND 329.96 345.3 30.72 UPR PLNM-LWR DNCMR 106 0 104 LLAB 0.0 862.0 532.0 VESSEL, STM DOME 107 0 104 SLAB 0.0 2200.0 93.0 DRYERS, STM DOME 121 108 0 CYLIND 510.52 0.0 311.2 VESSEL, UPR DNCMR 125 0 108 CYLIND 0.0 1568.52 36.18 STAND PIPES 131 107 0 CYLIND 2430.04 0.0 12.57 SEPARATORS 301 301 0 CYLIND 1381.26 0.0 126.36 RECIRC, SUCTION 303 303 0 CYLIND 908.91 0.0 83.52 RECIRC, DISCHARGE 304 304 0 CYLIND 345.31 0.0 31.84 RECIRC, SUCTION 306 306 0 CYLIND 220.91 0.0 20.63 RECIRC, DISCHARGE 401 402 405 SLAB 1579.06 1753.81 13.88 EC A TUBES 421 422 425 SLAB 1579.06 1753.81 13.88 EC B TUBES

TR 045 { 2 "$ E ~ >:I Rev.0 9 s, Q (M !-+ -! (-)-E g M2 2 8 + ! =55 5 l > @E : n E M ~ 2 g 2 = 'j ~ v n g ~-~v a /. 3 n y ~ D E -}-Y =gr! ,I E4 3 c e H E 5 "C g ( 3--() 8 5(' ) ) W'r s E \\ g A i g d ,.,(GD) i ; - E' -dD E 3" FM 3 g v M 5. m 8 m g E E W w s g 't W o r. s w cn g1 g . C M EI >0 I oz j W 5-2 _' d sI; g = c e*,r= g c, !a, a c =; =;., d m _; e W .e ' EE w 4

g. *-

~ We g$ F s N> a W T o .o i = -. ~ 2 5 E" g g g; 9 p. g r M i O y w h A -5 .E a E. s,z g' w 3 s ..e C-g w (V5 S,, a i U <( .2 = 2 i / G( C) ~ (> ]~ ( = u M 5 l D

M S N U A S O N C S O C 0 9 9 8 M M U U S S DEH EC FT A M M 5 A 8 1 A S M EI M = TM T U " S S O N T S OC F T N 8 8 1 A E F M L L A S = P B T E A L U S S T S r ( 4 1 2 8 N V E B M O L C N E H A S A V C T U f R E N P TL I IR S E N T I R L O \\ E' T V g# E A6 7 L 8 1 8 M M = L O T U n S E 2 3 4 V 4 G M W N 1 E 1 S O L O L C N y U L F AU F S TCA + 1 = L $w S E N V O E C "g Lk

WO, LF W F 1 S 0 V GE 1 NKOW FROTL SF = 0 0 L R E C 0 O 5 V E 0 9 1 1 M R E 1 M = L E S = R / VE E U u S F LK L n S L N E A O N V O A S O R V R T C T C O = S T ag = 2 9 9 9 M 1 G = L U n SN L S O C RE 7 6 L 9 9 M L OR U n TN O S C l P j 0 M 0 OR F e

T ( L )5 A 3 N M 3 G IS u U M R O O S R R F R ( E 3 5 4 1 0 0 = 0 NG 0 1 I = 2 M S NO C = LA ) 2 7 N M G 2 IS U M R O O S R R F R E ( E 1 RUT 1 M SN = I S O U L S E R P N T S O P E C =S ng = D L M O A F EI TN SA M

Y R 1 E L SS D APY B = 1 = I S 1 6 4 N 4 = 4 MG 0 L O 1 1 C 1 U1 U r r S= M ng ) L W S T 1 S O N 2 A L I r P F O 1 6 P O Y S 1 T B S P E A )8 ( O I N R P 4 2 4 T Y M IB B 1 U k R U M s T S OR E (F N l G T R N I E O W P O N P l O 5 IT I T SO P N E I VLAV 4 G LL 1 3 + kI l

3 T }3 N I O T( 4 3 G N F EM L T 3 E 3 D M S O Y U M L Y S S ALE S R = S S S A A P Y P 2 B 3 Y G B N F l 1 3 1 M = U L S N S O C

S J A P P B 1 Y 7 B O M i T U n 0 6 M S 1 U S 0 0 3 2 6 6 T T N N 0 S 7 I I R T O T N O 0 M I 0 O V 4 3 R 5 5 E S G G E N N V 1 L 6 G F F A V S N S A F P Y B 5 4 4 4 M M U a U ( 2 5 M S S U n S 2 1 E 6 E V 3 1 V M 1 L - L = 9 = A A ) V U a S V 5 E 3 S 3 S N S V 4 N S S O S M L O A C A M A C P P V U a Y O Y + S B R B S S F nN ( AP YB

OR P 1 NUS T 5 1 T ) N I 0 V 2-C O O T T 8 1 T 4 3 1 2 I 1 1 = 0 = 1 1 M L G 1 M NG IL D U U t L t I 4 J s M O S M A L = 7 1 1 1 0 G G 1 A L = L L S N O E C N IBR l 6 0 U 1 1 M G T U n A S L L: T A N M I O R P O T ) N E 0 ) 3 S 1 V C = M M S O N O R R O F F ( C (

N IO L P O D RE N l T VWA NLO M OALE CVFD 32 D A EN E T g LI ZOR I N TA A C W I M N JR U O O J L N - F 2 Y 2 A G LER N E F VLAV = L O R T N O 1 2 C N I M DNA M ED 0 Y 2 M A LE U ^ R D S EEP S = 3> 8 1 QO T M N O I R P F(

O O T R (F ( 8 2 M U L J S ) 2 r ' OT Y ( A LE 7 R 2 E T VL N AV I L O RTN OC Y RA 6 D . 2 N G O C N E F S 52 M U L J S l

)8 4

1 M O R(F

i O T ( E C N: NU IB J R U V T C O O ) L O T T 8 RE 2 MT VA r RNLE 1 OOAR O CV T ( R O TO M O V 0 R 3 E T S E N VL I AV L OR TN OC 9 2 G N F ) 82 M OR F( ( l j,

t- ~ TR-045 Rev._0-Page 50 of 135' 3.0 'MODEL QUALIFICATION It 'is not intended here to qualify RITRAN.,as. a code because this has already been done through UGRA and reviewed by the NRC resulting.in M-corresponding SER. The. objective here is to qualify the Oyster Creek RETRAN model with respect to the noding schenie used, code options exercised, control systems simulation,.and the adequacy'of calculational procedures for the different parameters required by the code. Nice diverse startup tests have.been utilized for.this purpose, thus coveringi nearly all aspects of the model.*Best estimate values for the core-bypass fraction (10%), separator inlet inertia (calculated based.on'. as built drawing), and steam carryunder (0.1% based.on plant measurements during startup tests) were used. The same noding diagram as-shown l l in Figure 2.1 was used for all startup tests. The startup tests were chosen to benchmark the following: Control Systems benchraark and model stability: - the level setpoint change test and the pressure setpoint test were chosen. Liquid level model benchmark: the level setpoint change was used for small level changes and the MSIV closure test:was. chosen for severe level drop after a void collapse in a reactor scram. Steam line model-benchmark: Turbine. trip with bypass and generator-trip with bypass. Also these tests benchmark'the turbine'model and ~ the different timings between stop valve. control valve and bypass valves. --.__--.-_---__-f

TR-045 Rev. O Page 51 of 135

Bypass valve sizing: Bypass valve test to benchmark the bypass contraction coefficient.
Isolation Condensers Benchmark:

Isolation condenser test.

Recirculation pumps benchnerk: Recirculation pumps trip and the power flow control tests were chosen to benchmark the pumps coastdown characteristics and model response to recirculation flow variations along the power flow line.

The overall objectives of these benchmarks are the following: 1. Prove Model Stability: The goal is to show that the RETRAN model is as stable as the plant when subjected to the same stability tests applied to the plant during the startup tests and using the same l acceptance criteria applied to those tests. The basic criterion used was that the decay ratio, defined as the ratio between successive amplitudes of the same sign of important plant parameters, ir less than unity.'** 2. Prove that the RETRAN model will respond adequately to mild and severe transients during power operation and post trip over the time scale expected in reload transients. The adequacy of the response is based on an acceptance criterion of 15% error margin between RETRAN response and the plant for the maximum change in a parameter. This error margin was based on the following considerations:

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Page 52"ofi135

- 3 g i .I The plant 1s$tartup tests results were available on. analogue-- 1 a. a T-5F chartascords.. The percentage;e"tror associated with" reading. d i 'A L'; . ( ) parameters offlthe charts was" assumed to be 10%.- q A t 9 .t.

Ai y

[ 'b.- RETRAN" calculation of a certain parameter is' regarded as, acceptable-if it is within 5% error margin.from the.true value' T '6,i , c 'h, f' of the plant parameter. n- .T C

e.

The above' values were based on engineering; judgment. The details!cf'the -startup test benchmarks and the acceptance criterin' tests were" documented.. in reference'(5).. All benchmarks-passed'the acceptance criteria-for the primary parameters of interest. / su A summary description of each test and important results will le n presented below. 3.1 Pressure Regulator Test (STP19) j. The purpose of this benchmark.is to detentine the model response to t pressure regulator setpoint chasge and to deinonstrate the stability 1 m.,. l of the model power void feedbackfloop to pressure perturbations <sf, m The. test involved a 10 Psi decrease in Electrical Pressure Regulator.- 6 (EPR)setpointfollowedbyasteadystatefor6bseconds:andthen11 psi increase, at 100% power (1600 MW)$ .a g p. b g-y 4 sh .I .1 ____._______1___________________ f, ,]

n;,r i ( 3 ' TR-045 Rev. 0: w Page-53 of 135 A' comparison.between RETRAN.and' plant data is shown on Figures 3'.1.1 [q and 3.1.2, 'where it ' can be 'seen' that the model tracks-plant data quite'well and' clearly passes the stability acceptance test. J u , ) 3i2. Level:Setpoint Change (STP21) u The purpose of:this, benchmark is to determine the mode 1' response ~to' ~ 1 level _setpoint change and to demonstrate the stability-of'the model. for changes in inlet subcooling initiated by changes'in the feedwater system.. The test involved a level setpoint decrease lof'10 inches at 100% l i J i power level in a step fashion and monitoring plant response. 1 A comparison between RETRAN and plant data for level-and power is shown on Figures 3.2.1 and 3.2.2 where the stability acceptance criteria are clearly met. 3.3 MSIV closure Test (STP12) The purpose of this benchmark is to check the model response.to full isolation and to test the model level behavior in a post scram' q conditions. This test was not a pressurization test because the- 'i 1 MSIV closure time was approximately-8 sec., therefore.this benchmark j 'I .I A ._-.m_..__..-__-_-___.--_____m__ ___m._ _.--_9._..______.m_______.______ m_m.-.-_.________.-m__________1.____._

( f I ) 'TR-045. Rev. 0 Page 54;ofs135l' ~ ' main' objective is to test the model void. collapse"after a scram-resulting in a severe level drop and to test.the pressure behavior. following the MSIV isolation. 1 l-l The' test proceeded by manually closing the MSIVs and monitoring plant response..' Figures 3.3.1-2 show'a comparison.between the model: and plant output. Level response (Figure 3.3.1) shows good agreement throughout and clearly passes the 15%' acceptance criterion, while the pressure behavior is shown in Figure 3.3.2 where the model does show the same' trend as the plant up to 30-35 sec. when plant conditions become unclear as to the status of the feedwater system which would impact the cooldown rate (pressure behavior). The margin between RETRAN ~ and plant pressure response is believed to be due to measurement errors on the wide range reactor pressure recorder used for this test. This test clearly demonstrated the ability,of the model to track severe level changes very well. 1 3.4 Bypass Valve Test (STP20) J 4 The objective of this benchmark was to establish a' contraction . i coefficient on the choked bypass junction (j218) and also to qualify the bypass model of the pressure regulator control' system. 4 u.

>y<, TR-045 .Rev. 0 Page-55'of.135. The test proceeded by' manually opening one bypass-valve and leaving it open for about 40 see then closing it, with the pressure regu-lator:(EPR) in control.. Plant response'shows a slight decrease:in' dome pressure (3 psi) due to high initia1' steam flow'after which the ' control valve closed a little to maintain pressure; when the bypass valve was closed a slight increase in dome pressure (~3 psi) for:a short time until the control valve opened to maintain pressure. Plant data used for comparison were dome pressure and EPR pressure signal only because changes in other plant. parameters were small and-within the 10% error bounds which made it unsuitable for use. s Figures 3.4.1-2 show model to plant comparison where the stability-I criteria and the 15% margin are clearly satisfied. The conclusion of this benchmark is that the correct contraction coefficient ~has' been chosen and the bypass model control settings used predict plant behavior in an excellent manner. 3.5 Turbine Trip Test (SIP 17) The primary objective of this benchmark was to qualify the separator model, the steam lines model and the bypass openingLtime. 10ne dimensional kinetics was used in the simulation. The test. proceeded. by manually tripping the turbine with the emergency trip switch'from 4 i' h

9 TR-045 Rev. O Page 56 of 135 the control room. The bypass valves opened in 0.1 to 0.2 seconds after a 0.3 second delay following the trip. The Simulation was carriel out by closing the Turbine Stop Valve (TSV) which then initiated a fast Turbine Control Valve (TCV) closure and a fast bypass valve opening. A special control module was built that would close the TCV and place a full bypass open demand signal upon a TSV closure signal. A comparison between model and plant response for. power and dome pressure is shown in Figures 3.5.1-2 where the model shows an acceptable conservative response. RETRAN power and pressure peaks are approximately 25-27% higher than plant output, f which is clearly higher than the 15% acceptance criterion. The main I reason for this deviation, is the lack of measurement error on turbine stop valve closure time and bypass opening time which are the two most critical parameters that determine power and pressure behavior. The test acceptance is based on the conservative nature of the response. l l 3.6 Generator Trip Test (STP18) The objective of thic benchmark was to qualify the turbine model of the pressure regulator control system and the timing between control valve closure and bypass valve opening, when the control valve closure is under the normal speed relay. One dimensional kinetic was used for this simulation. The test proceeded by manually i i

.TR-045 Rev. 0 Page 57 of 135 tripping the generator off the line by opening the main breakers. =The turbine speed governor. closed the control valves.and;the pressure regulator opened the bypass valves within 0.5 seconds. The stop valves closed by.a no-load. turbine trip before the turbine overspeed trip (110%).was reached. :The reactor scrammed on.high' APRM neutron flux 0.9 sec. after the generator trip and the APRM-(#5)' peaked at' 150% of rated ~ power 1.1' seconds af ter-the trip. The simulation was initiated by a ' loss of load trip and figures 3.6.1 and 3.6.2 shows a very good comparison.between the model and test data for power and pressure while Figure 3.6'3 show thel timings between trip initiation, control valve and bypass closures. The trip sequence obtained by the model is the same as the test;'namely, a scram on'high flux and a no-load turbine trip (20% steam-flow). .j ~ 1 I The 15% acceptance criterion is clearly satisfied. i l. l 3.7 Recirculation Pumps Trip Test (STP14) i The objective of this test was to benchmark the pumps coast down time and the frictional torque coefficients as a function of pump speed. 'The Reactor Protection System (RPS) pump trip signals at J Oyster Creek (Low-Low & Hi Press.) result in one' kind of pump trip ~ y which is the tripping of the whole power train frem the M/G set to the pump' motor, and the inertia involved is the total loop inertia,- i.e. M/G set motor, fluid coupling, generator, pump motor and the' pump itself. ~l l 1 _ _ - - - _ -=-

( 'TR-045' .Rev. 0 Page 58 of 135 Figure'3.7.1 shows an acceptable comparison between'model'and plant-output for total recirculation flow,'where the 15% acceptance criterion'is clearly met. .3.8 Power-Flow Control Test (STP15) The purpose of this benchmark was to determine the model' response to changes in recirculation flow. This' response will.be in the form of appropriate power-flow behavior as compared to plant data and model stability to perturbations in recirculation flow. Two tests were involved here, the first test (A) proceeded by decreasing recirculation flow by.15%,'from 100% rated conditions, in l about 30 seconds and then 70 seconds later, the flow was increased by approximately 14% in about 30 seconds. The second, test'(B) was carried out by reducing the flow by 20%,1from rated conditions,-in I about 40 seconds at about the same rate as A.- l Figure 3.8.1 shows a comparison for the recirculation flow between l model and plant output for test (A). Table 3.8.1 shows a comparison between model output and main plant parameters at beginning and end of flow changes, since chart" r t recorders outputs were not available. i

TR-045 Rev. 0 -Page 59 of 135 Figure 3.8.2 shows level response for Test B while Figures 3.8.3, 3.8.4' show power, and dome pressure comparisons with plant data, where an acceptable response can be seen. The primary parameter in these tests was reactor power which clearly passed the 15% acceptance criterion for Test A (Table 3.1), but for Test B the error margin was 20% which was l higher than the acceptable limit. The main reason for the deviation was due to lack of test data on actual recirculation flow behavior and the same flow behavior as for Test A was used for the simulation because test document'*) implied approximately the same rate which may not be quite f the case as may be seen from Figure 3.8.3. The other three parameters, steam flow, dome pressure, and level are secondary parameters because a l change in recirculation flow immediately affects core void content which produces the initial reactivity change. The percentage error in those parameters ranges from 20-40% and due to small changes involved, e.g. 10-15 psi out of 1000 psi and approximately 2" of water and the unknown nature of the instruments errors (especially steam flow) those parameters were not used for the test criterion. However, the order of magnitude of these quantities relevant to plant output is an indicator of the goodness of the model. 1 l

T.

e TR Rev. 0-Page 60 of.135

~ 3.9 Isolation condenser Test-(STP 13) The objective of this benchmark was to qualify the isolation. condenser model. 13e actual. test was conducted by cutting in one isolation condenser with the~ plant. operating.at 200 MWt (11% of rated power) with full recirculation' flow, manual feedwater' control and with the Electrical. Pressure. Regulator (EPR) controlling-pressure,through the bypass valves. 'The RETRAN model was- ~ initialized at 202 MWt with 1015 psia dome pressure,~ manual-feedwater and EPR in service. t The result of the benchmark is shown'on Figure'3.9.1 for. power,' - where a comparison between point and.1-D kinetics'is also'shown. It can be seen that the model tracks plant data quite.well, butrits heat removal capacity is slightly lower than plant data which is in l-a conservative direction. The model was deliberately built to' remove power at design capacity while actual plant' data-showed at least 20% over design capacity. The test showed model stability'to sudden initiation of the isolation condenser.by passing:the stability test. It also showed the same trend as'the' plant and'the required 3%. heat removal capacity calculated as the' difference between power generated and power removed to the condenser (or power e removed through tube /shell sides).. The 1-D kinetics gives a closer response to the plant output during the initial' stage'only and it becomes very close to the point kinetics when the condenser initiation'is completed. Therefore, point'k'inetics[is quite' adequate in simulating long, term operation of the isolation ~ condenser. Ittis concluded that the test was acceptable.. J j

a ,i ,TR-045 ) -Rev. 0-Page 61 of 135-1 Table'3.1 l Comparison Between Plant ~& Model Response at Beginning'and End~of Test Conditions:

l TEST A Steam Flow *.

Power Change *' ' Change Press. Change - Level Change Plant /Model Plant /Model ' Plant /Model Plant /Model I t- . 0.15'/+.2' -12%/-11%. -11.9%/-14.3%- -11' psi /-15 psi, + +11.3%+12.5% +11.2%/+14.3% -+10 psi /+12. psi 1-0.2'/-0.16' j p TEST B- -15%-18% -16.5%/-15% -13: psi /-20 psi- +0.1'/+0.25' l

% change = Change Rated

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] 1 l i TR-045 l Rev. O Page 81 of 135 4.0 REPRESENTATIVE RELOAD TRANSIENTS BENCHMARK I d i The Cycle 10 reload transients'" as performed by GE were selected. to f .be used for benchmarking the Oyster Creek licensing model. -These transients were: the turbine trip without bypass' (TTWOBP), main steam j isolation valves closure without scram'(MSIV ATWS), and the feedwater I controller failure to maximum demand (FWCF). The. licensing boundary. ) conditions, initial parameters, setpoints, etc. as stated in reference have'been used'for the ifcensing mode. J This chapter will present a comparison between the RETRAN model and the. GE Cycle 10 reload analyses. A sensitivity study was performed for the limiting transient (TTWOBP) on a number of parameters in order to establish changes in model response due to possible tolerances'in certain i parameters. This response is measured in terms of its impact on the Critical Power Ratio (CPR) using a RETRAN Hot Channel Model (Section 2.1.4) and the GEXL Correlation. A final RETRAN base model was-established after the sensitivity studies were completed.where upper J limit values of certain parameters were used in the base model to assure conservative response. The CPR variabilities due to system parameters 1 perturbations were statistically combined to obtain an overs 11 CPR'

j multiplier which will be used on the base model calculation result to-

{ assure a 95/95 MCPR limit. i l j

.l TR-045 Fev. O Page 82 of 135 4.1 Description of the Turbine Trip Without Bypass Transient (TTWOBP) i The TTWOBP is one of the pressure increase category transients normally analyzed for Oyste'r Creek reload analysis. It produces l l the most limiting CPR for this category. In this transient the f turbine is tripped and it is assumed that the steam bypass valves which will normally open to relieve pressure fail to operate. The scram signal is received from the 10% stop valves closure signal. The transient is initiated by closure of the turbine stop valves. Once these valves begin to close, steam flow leaving the vessel decreases cauwing reactor vessel pressure to increase, while the core is still generating power. This rise in pressure causes a reduction in the core voids which result in a core power increase which continues until the new vcids generated by higher power, the Doppler reactivity feedback, and the scram reactivity override this' positive effect and begin to reduce the power. The pressure continues to increase until the Electromatic Relief Valves open. The 16 safety valves are kept closed as has been the case for previous reload analysis ('). The rise in power is followed by a rise in fuel temperature and fuel rod surface heat flux which results in a decrease in Critical Power Ratio (CPR). The basic sequence of events as modelad in RETRAN are shown in table 4.1 i i

\\ ( l l TR-045 1 Rev. O Page 83 of 135 ) 4.1.1 Model Description The same noding diagram as shown in Figure 2.1 was used for this transient and all other transients and startup tests. The core initial conditions and setpoints were the same as used in Oyster Creek reload analysis for Cycle 10 which l were based on OPL-3 inputs. The basic licensing model q parameters are the,same as those used for startup tests i except for those listed in table 4.2. Initial conditions that characterize certain events will be listed separately. For this transient, a linear closure rate was used for the turbine stop valve which closed in 0.1 second. The important code options used are listed in' table 4.3. These j l l are the same options used for the startup tests except as noted in the table. The default convergence criteria have i been used throughout except as noted otherwise. l 1 l l

f*. e '<t .1 I' TR-0451 j .Rev. O. Page 84 off135. 4.1.2 Sensitivity Studies 1 L A number of. sensitivity studies were carried out using this transient in order to identify the CPR1variabilityidueLto uncertainties in system parameters. Unc e. tainties. with': reasonable bases will be' included-in the statistical-analysis while the rest.will not and'in such cases the most' limiting values of such parameters will be used in the base licensing model to assure a conservative response. ~ ~ Before doing any input parameter. sensitivity analysisi.a, number of modeling and code options were exercised in order to optimize CPU run time and arrive at a converged solution. Those were: 1. Time Step Convergence Studies.. d Time step studies were carried out by starting-with 0.01 sec..for a maximum time step then reduced to 0.005-followed by 0.002 and 0.001 sec. 1The difference between ] the last two cases was less.than 0.3%, change in peak power and hence a 0.002 second maximum time step was adopted for the first 2 seconds.of the transient where ~ l the severe changes were taking. place, then1 relaxed tof 0.005 for the remainder of.the transient. J i d l a

TR-045 Rev. O Page 85 of 135 2. NSHAPE Frequency For One Dimensional Kinetics. For the calculation frequency of the flux shaping factor, the default option of a calculation every time step (plus at the minor edit frequency) was first used followed by NSHP=5,10 for the whole transient. It was l found that the default option is to be used for the 1 first 2 seconds while NSHP=10 may be used for the remainder without any noticeable impact on the transient as compared with the default option'throughout. 3. COURANT Limit. For the Courant limit sensitivity, the default option l (0.3) was used for the first 2 seconds then relaxed to f I 1.0 for the remainder with no noticeable difference against the default values when used throughout the transient. 4. Local Condition Heat Transfer Model for Conductors in Non-Equilibrium Volumes. The local condition heat transfer model was first used in volumes with phase separation (separator and upper downcomer) where 1 ft/ stack conductor element was used in order to more accurately represent heat conduction in le -______-

q Tk-045 Rev. O Page 86 of 135 such volumes. The stacks were then replaced by one sided single conductors representing separator structure, standpipes and vessel walls respectively', where average. volume properties are used. No noticeable .i l difference was seen.in important parameters. This is l l probably due'to the short term nature of the transient. j j 5. Separator Initial Mixer Level. The last area where sensitivity studies were carried out q before input parameter perturbation analysis was the i separator initial mass inventory. This is changed by varying the-initial mixture level, which was initially, set at approximately the same elevation as the upper downcomer mixture level. The mixture level was reduced by half with very little effect on peak power, but it reduced the required CPU time. However, when using a maximum time step of 0.002, no effect was seen when the mixture level was changed. The separator initial mixture level finally used was based on the assumption that the separator mixture density is approximately the l same as the standpipe average mixture density during steady state'***, and from this the initial mass I inventory can be calculated. l l l l

I ', =TR-045~ 'Rev. 0~. l Page.87. of 135 - The outcome of the above study result'd in using the following, e parameters / options: ~ 1..-Maximum time step of 0'.002 sec.Lduring' severe changes beyond which it can be relaxed.. 2. Default value for NSHP in 1-D kinetics during severe changes.

(first 2. seconds) beyond which,it can be relaxed when neutron flux becomes small.

3. Courant limit can be relaxed to 1 outside the severe ~ transient region (2 sec.).' 4. Local Condition. Heat Transfer Option not needed. 5. Separator mixture' level is established using an initial mass inventory based upon standpipe mixture density.- The objective of using the above parameters / options is to minimize CPU time for this transient without compromising accuracy. Items 1, 2, & 3 above should be revisited'for each. transient as different values are needed for different scenarios, i.e., above sensitivity-studies need'to be repeated.1-y l -l I --._.-_a___

TR-045 Rev. O Page-88 of 135 The local; conditions; heat transfer option (Item 4).'is not'neede'd- - i I for.short term transients, i.e., the typical reload. transients } -(TTWOBP, MSIV ATWS, FWCF) because,-as seen from the sensitivity. study, the heat. transfer time' constant between.the' fluid and.,tihe.

f heat: structure is longer than the transient time of interest and I-a detailed accurate representation-as offered by local-I conditions heat transfer model,: which is CPU ' intensive,: is not -

. j The separator mixture' level calculation (Item'5) ' necessary. should be used all the' time for all' transients because it improves CPU time and is based on.NRC approved' assumption. . i The RETRAN Licensing Model was' built around data obtained from. vendor's documents, systems diagrams, P& ids and various catalogues. It is postulated that deviations exist between ' I actual plant parameters and calculated parameters due to manufacturing tolerance, complex geometrics, etc. Hence,'a sensitivity analysis is required to assess'the impact of such. uncertainties on limiting parameters in a limiting reload 3 t \\ t I transient. ~ l The limiting CPR reload transient for Oyster Creek is:the TTWOBP(, hence the impact on MCPR will be investigated using' l the RETRAN system model for this transient followed by a RETRAN-r

TR-045 Rev. O Page 89 of 135 Hot Channel analysis using upper and lower plenums boundary conditions as obtained from tne system model. The outpuc of the Hot Channel model is'used to calculate the initial steady state CPR j and the maximum change in CPR (dCPR)'during the transient using the GEXL correlation. The difference'between tne ratio of the. 1 maximum ACPR and the initial CPR (ICPR) for the perturbed and unperturbed conditions (ARCPR) is used as the measure of the impact of the perturbation. A list of parameters chosen for sensitivity study is the following:

Core:

(a) Direct Moderator Heating I (b) Cap Conductance (c) Subcooled Void Profile Fit (d) Core SP

Separators:

(a) Carryunder (b) Inlet Inertia (c) AP Across the Separators ) l \\ l _____________________J

[ .TR-045 Rev. 0 ) Page 90 of 135

St.eam Lines i

(a) Steam Lines AP (b) Volume J [ (c) Inertia i

Recirculation Loops Volume
Vessel (a) Steam Dome Volume (b) Lower Plenum Volume (c) Lower Downcomer Volume
Nuclear Parameters Sensitivity In the following sections, a discussion of each perturbation level for each of the above parameters and its justification will be presented.

a. Core Parameters Sensitivities The core parametors that were perturbed were: the direct l moderator heating fraction, the fuel gap conductance, subcooled profile fit, and core pressure drop (AP). The best estimate direct moderator heating fraction as reported in reference 12 is 3.3%, which is equally partitioned between the in-channel moderator and bypass. A 25%'***

q TR-045 l Rev. 0-f Page 91 of 135 I l reduction in this fraction resulted in an increase in RCPR 1 (Table 4.4). It was decided to use a conservative 2% value for the direct moderator heating for the licensing model as used by reference (14). The decrease in direct moderator heating fraction results in more heat generated within the fuel which drives the fuel temperature up thus increasing l the transient surface heat flux and the ARCPR. The fuel gap conductance was reduced by 50% from a base value of 1000 BTU /HR.FT*.*F resulting in an' increase in ARCPR due to. ] increasing fuel temperature causing an increase in transient heat flux (Table 4.4). Because the model uses a constant gap conductance during the transient and the inherent conservatism in such a representation (**', a gap conductance of 1000 BTU /HR.FT*.*F may be used based on references (11, 13). A conservative value of 500 BTU /HR.FT* will be used for Cycle 10 core because of unavailable data on the EXXON VB fuel which made nearly 2/3 l of Cycle 10 core (the remainder is GE 8X8). Because the. uncertainty in this parameter is not documented, it will not be included in the statistical analysis and a conservative value will be used which is fuel cycle dependent. 1 l l

I TR-045 Rev. O Page 92 of 135 The base model used for the above analysis did not include the subcooled void profile fit option. The next step was to test the impact of this model. When this model was included f a reduction in ARCPR was obtained because of the increase-l l in neutronic voids produced by this model (Table 4.4). Hydraulic void collapse during the pressurization phase. .I .j resulted in a smaller percentage of neutronic voids collapsed and hence smaller reactivity increase causing a reduction in peak power. Neutronic voids are the sum of the hydraulic voids and the voids generated by the profile fit model for the subcooled region of the core. The subcooled void profile fit option is used in the licensing model in order to capture the subcooled void contribution during severe transients. The last core parameter studied was the core AP which was increased by 1.5 psi representing approximately a 10% error in measuring pressure drop across the lower core plate. This was done by adjusting core inlet loss coefficient and the downcomer loss coefficient to maintain the same loop pressure drop. This increase l 1 resulted in a very small increment in ARCPR (Table 4.4). b. Separators The separators parameters that were perturbed were the

TR-045 Rev. 0-Page 93 of 135 separator carryunder, inlet' inertia and separator pressure drop. The carryunder was reduced by 50% which resulted in a small increase in ARCPR due.to an increase in power because pressure wave through downcomer/ recirculation loops reach the core earlier due to a reduced. upper downcomer " cushion" at small carryunder values (Table 4.4). The separator inlet junction inertia (standpipe to separator) .I was increased by 100%. That resulted in an increase in -l 1 ARCPR. The best estimate value was 0.237 and the j perturbed was 0.47 while the vendor supplied inertia was 0.749. It was decided not to include this variability in the statistical analysis and to use the vendor's supplied inertia and to apply all of it to the inlet junction because this is the most conservative approach as shown in reference (14). The separator pressure drop (between standpipes and upper downcomer) was reduced by 0.5 psi by adjusting the separator inlet loss coefficient in order to represent a 10% l l uncertainty margin. The impact was an increase in ARCPR due to a reduced attenuation of the pressure wave as it j reaches the core thus producing a higher power (Table 4.4). 1

W< q h ~TR-045' -Rev. O Page 94 of 135 c. Steam Lines The ateam lines. parameters that were perturbed were a' steam line'AP (dome to steam chest). uncertainty'of.10%, the steam lines volumes and the; total-inertia., A 10% decrease 'in AP was assumed which resultedtin'an increase'in'ARCPR-due to a decrease in the; pressure wave attenuation'as it ~ reaches the core'thus~resulting in a higher void collapse and higher. power.(Table 4.4). A 5% uncertainty in volume-(representation a reduction) was assumed due to error in measuring pipes lengths.. This reduction resulted in an increase in ARCPR'because the ~ pressure wave will reach the core faster.with a re'uced d volume and'an earlier positive reactivity insertion will.be achieved (Table'4.4). A 7% increase in total' steam line inertia was assumed which resulted in an increase in ARCPR (Table 4.4)$ 'The g l perturbations in the steam lines-parameters were carried out x independently and in conservative directions only with reference (14) used as the basis.' l I'-

.. l L! l .TR-045 i Rev. O j Page 95 of 135 j It may be argued that reducing the volume by assuming a variability in pipes lengths (pipes diameters'have very -3 small tolerances) 1.e., a decrease results in a decreasing l inertia which has an opposite effect on the ARCPR which will tend to reduce the effect of an increasing ARCPR 1 caused by a reduction in volume. It can be seen that the analysis above represents the worst combination of parameters that result in the most conservative approach. l d. Recirculation Loops j A 5% uncertainty in recirculation loops volume is assumed in j a conservative direction (reduction). This resulted in an increase in ARCPR because the pressure wave reaching the core through this path will arrive earlier due to the reduced inventory thus resulting in an increase in integrated power (Table 4.4). e. Vessel The vessel parameters perturbed were the steam dome volume, lower plenum volume and lower downcomer voluem where a 10% reduction in each parameter was assumed based on errors in

.1 -TR-045 Rev.'0 j LPage 96 of 135 calculating vessel internals of.different' geometries. A j reduction inLvolume in each of the above parameters resulted l 1 in an increase in dRCPR because the' pressure. wave will reach-the core sooner due to reduced inventory. The' steam ,i I' dome volume reduction has more impact than the.other. j a l-parameters as shown in Table 4.4. -] 'I f. Nuclear Parameters' Sensitivities The uncertainty in the RETRAN.1D' kinetics input'was: determined from the' uncertainties in the' calculated core-power distribution. This was~done by running the.TTWOBP transient'using a-Haling power distribution as a base case. j The power distribution was then adjusted-to reflect!the. uncertainties in the power distribution for both the peak i node and average axial' power shape. 'The uncertainties in .J the power distribution were from gamma scan comparisons made with the 3-D core. simulator model. The TTWOBP was-rerun' I using the kinetics input calculated with the. adjusted ~ power 1 l l shape.- The ARCPR between the two cases is' taken as the l f. uncertainty in the kinetics input. -i 1L The uncertainty in the RETRAN Point kinetics input was l l' l. determined b'y applying a multiplier.to the void and Doppler-reactivity inputs to RETRAN. A multiplier of 0.75 was used 'for the void reactivity and a 1.1 multiplier'was used for the Doppler reactivity. Transients that are analyzed with i

q 1 TR-045 Rev. O Page.97 of 135 the point kinetics (non-pressurization transients) were run with and without.the multipliers. The ARCPR between the two cases is attributed to the uncertainty in the point kinetics input. Since the uncertainty in the RETRAN point kinetics input for non-pressurization transients is less than the uncertainty for the ID kinetics input, the uncertainty for the ID kinetics input will be used'for the overall uncertainty in RETRAN. The scram reactivity in the point kinetics model uses a 0.8 multiplier for conservatism since the uncertainties are not explicitly considered. Table 4.4 summarizes the results of the above analyses. ' Parameters with no known and accepted uncertainty band were eliminated from the statistical analysis and their most conservative values were used, those were gap conductance and separator inlet inertia. The subcooled void profile fit is a model option and no uncertainty analysis is required. The uncertainty in the remaining 12 parameters was used in developing a conservative CPR multiplier. Those are marked by an asterisk (*) in Table 4.4 and were assumed to represent upper limit values. The calculations (volumes, inertias, etc.) were based on accepted engineering judgment. Because upper limit parameter uncertainties were used, it is reasonable to assume l l that the overall uncertainty obtained as being an upper bound or .the 95/95 limit*'.

,.c 1

TR-045 Rev. O.

Page 98 of~135 The overall uncertainty (A) is given by: l 1. 'A = (' '

2 ( ARCPR i ) *f ' * =,0j 042 -

i=1-l- 'and;the minimum CPR is'given by: L MCPR'= 1.042 (SLCPR +'ACPRe) i where. 7 SLCPR;= Safety Limit CPR (1.07). ACPR = Calculated ACPR-Another sensitivity study was done on the Dynamic Slip Versus . It.was found that the-the Algebraic Slip Options'in the-code. 'd latter option is quite stable (especially for the hot channe1 model) and produced more conservative results; Because the GE code (0DYN) uses a drift flux formulation,'the Algebraic Slip Option is more appropriate.to use for comparison.' This option. will therefore be used for all reload analyses.'~A summary.of the final code options to be used for the Cycle 10. benchmark and f l future reloads is shown on Table 4.5. L- - -

l TR-045 Rev. O Page 99 of 135 4.1.3 Turbine Trip Without Bypass Benchmark Results The results of the TTWOBP transient of Cycle 10 are shown in j q Figures (4.1.1-4.1.'6). The power response comparison is shown in Figure 4.1.1 where it is' evident that the RETRAN i peak is higher but in' general GE's results show a slower i rate thus the total energy content, which is proportional to- ) the area under the curve, is different. This is reflected in the pressure response comparison shown in Figure 4.1.2 I where RETRAN'results are approximately 50 psi lower than GE. The reason for this deviation is the different methods used in the generation of the one-dimensional kinetic parameters. However, the impact on the CPR is more conservative as will be seen later. The level response j (Figure 4.1.4) shows a reasonable agreement for the first second then it starts to depart. This is because GE ODYN model'**) does not include the feedwater piping which is included in the RETRAN model. A reduction in steam flow on turbine stop valve closure will result in immediate reduction in feed flow until the total error signal of the feedwater controller reverses the trend. This reduction in feed flow causes the continued level drop after the initial drop due to void collapse, while the presence of feedwater piping in RETRAN causes a delay in the arrival of the

? .J 1 TR-045 Rev. O' x Page 100 of 135 reduced feedwater to the. vessel due to. transport time in the r 'j piping.. It is therefore. concluded that level discrepancy lis' j .\\ due to modeling differences; Figure 4.1.5 shows the core: j inlet flow while Figure 4.1.6 is the heat flux comparison. The core inlet flow behavior is dependent on recirculation 1 pump.coastdown which'was benchmarked against actual' plant' data (Section~ 3) and the. higher coastdown rate shown by RETRAN is more conservative because it results in a higher. heat flux due to reducedEcooling capability which has a l ' higher impact on the safety limit CPR. The.CPR; calculated from this-transient using the Hot Channel parameters shown i in Table 4.6 is more conservative than GE, as shown in'the ~ i following comparison. 6 GE 8X8 EXXON VB GE Results RETRAN Results GE Results-RETRAN Results Initial MCPR 1.32-1.317 1.29 1.291 i ACPR 0.25 0.291 0.22 0.294 Operating MCPR 1.38 1.43 1.35 1.44 +.

._w__.

l} h TR-045 Rev. 0 Page 101 of 135 .l f I Main Steam Isolation Valves Closure Without Scram ) ll 4.2

k I

J This transient represents the plant. response.to the simultaneous' closure of-all main steam isolation valves (MSIV) with the failure of the' reactor protection system. This transientcis analyzed to determine'the adequacy of the' safety valves to prevent: vessel:over pressurization.' CPR is not calculated for this transient. This' I transient proceeds by closing the MSIVs in 3 seconds and keeping' the five relief valves closed, but allowing the safety valves to open and the recirculation pumps to trip on the-high dome pressure trip setpoint, as was implemented in reference (9). The model used for this analysis is exactly the same as.for'the 1 TTWOBP with the same 1-D kinetics. file. The contraction j coefficient for the safety valves was set to deliver rated flow at k rated pressure according to OPL-M *) ratings. The same options as for the TTWOBP were used except for-the maximum time step wherel a value of 0.005 was used'for the first 5 sec. followed by a 0.1 I sec. since a converged solution was obtained for a 0.002 sec. AT MAX. The MSIV flow area versus time characteristics was based ] on vendor's supplied data which gives a non-linear behavior'with the valve 98% closed in 2 seconds. Figure 4.2.1 - 4.2.6 show a comparison between RETRAN and vendor's data for a number of parameters. f Figure 4.2.1 shows a comparison between the neutronic power response where the first peak is due to void collapse and subsequent positive reactivity insertion because of the pressure l = _ --

TR-045 Rev. O Page 102 of 135 I wave generated by the MSIV closure while the small second peak is { due to TSV closure on turbine trip at low steam flow. The amplitude and position of the two peaks were found to be dependent upon the closure characteristics of the'MSIV, especially for the second peak. The other differences in the shape and timing are due to differences in kinetic parameters between GE'l-D kinetics and GPUN methods based on EPRI's CPM-SIMULATE-SIMTRAN codes. The overall comparison between the two results is quite acceptable. ~ Figure 4.2.2 shows the dome pressure behavior where the RETRAN peak output is 16 psi below GE's followed by a 50 psi difference at 8 This effect is due to the reduced energy content of the sec.. power plot (Figure 4.2.1) caused by different kinetics approach. Figure 4.2.3 shows a comparison between relative average heat flux. The same trend can be seen by the two methods and the differences are due to differences in power profiles discussed earlier plus differences between conduction, heat transfer models and solution techniques between RETRAN and vendor's methods since heat flux is a function of power transferred to coolant.

,,r., ~y

.TR-045 Rev. Page.103 ofE135-Figure 4.2.4' represents a comparison between. total core inlet' flow 3where the same trend can be seen.and the differences between the' picts is due to'the different methods.

'l i Figures 4'.2.5,6 show the level and feedwater flow behavior..The + level response shows excellent agreement while feedwater flow ~shows.. a deviation which_is attributed.to;the~different feedwater. ] ie controller' settings on the two models. The. vendor model is'. based j - on design data. which do not reflect the: actual controller settings that:are represented in the RETRAN model. -The.feedwate'r-controller j g settings are not included lin the'OPL-3 parameters.- The basic-conclusion here is that.both models recover level after 8 seconds. Figures 4.2.7-8 show a. comparison for turbine steam flow and safety valves flow which show very good agreement. The objective of this transient in a reload. analysis is to-determine the margin to the vessel pressure limit of.1375 psig. The margin'" for' Cycle 10 was 77 psi while RETRAN shows a margin of 93 psi for the peak vessel pressure.. Although less conservative ~ than the vendor result, the RETRAN result is still' conservative l based on startup test comparison and will insure a safe' margin to. the vessel pressure limit. l I l

TR--045 Rev. 0.' Page 104.of.135 l k 4.3 Feedwater Controller Failure (Max. Demand), J A This transient represents the plant' response to the failure of the.-- d ~ 'l I feedwater. control system in the' maximum demand position with. d initial level at the low level alarm setpoint. The maximum ) l ] l-feedwater flow usedLis 120% of' rated: flow "', with the.same ramp time as used by the vendor"'. Exactly.the same model was used' as before including one-dimensional kinetics.'and the only-i difference is the maximum time' step'of 0.1 sec. during the level' and power increase followed by a 0.002 second during the turbine l trip phase. The transient proceeds by ramping the'feedwater flow from its rated value.to 120% in 0.1 sec. with the feedwater s controller disabled. The increase in core inlet subcooling due to the increase in feedwater flow causes positive reactivity insertion which leads to a power increase. The downcomer. water level will increase due to steam feed mismatch until the.high water level l turbine trip setpoint is reached when a turbine trip is initiated lI' through a turbine stop valve closure (TSV). The high water. level setpoint is reached after 32.75 sec. in RETRAN as-compared to 32.3 1 sec. according to vendor's results, Figure 4.3.1. The. remainder of l-b the transient is characterized by a power increase due to void collapse following TSV closure with'the bypass valves opening followed by relief valves opening thus terminating the. transient.- J. Figure 4.3.2 shows a comparison between I .I l L j i

l i TR-045 j Rev. 0 1 Page 105 of 135 l RETRAN and vendor's data for neutron power where very good agreement can be seen. Figures 4.3.3-7 show a comparison for dome j pressure, core inlet flow, relative average heat flux (relative to I steady state conditions), bypass flow and relief valves' flow. The j results show acceptable comparison between the two methods. In Figure 4.3.1, RETRAN shows a' larger level drop due to void collapse than -0DYN (20 inches compared to 10 inches). The level qualification against startup test (Section 3) showed that the level change during void collapse could be as much as 55 inches within 10'sec. (Figur'e 3.3.1) during the MSIV closure test. The 20 inch drop seems more plausible than the 10 inch drop considering the rapid turbine stop valve closure of 0.1 sec. compared to the ~ approximately 8 sec. MSIV closure time during the startup test. In any event, the level response has no impact on CPR calculation which shows a more conservative result than GE. The CPR values calculated by RACE using the hot channel parameters in table 4.6, were within less than 1% of GE results as shown below. The intial CPR is slightly different from the TTWOBP because of different initial subcooling due to a reduced initial water level. GE 8X8 EXXON VB GE Results RETRAN/ RACE GE Results RETRAN/ RACE Initial MCPR 1.32 1.312 1.29 1.285 ACPR 0.20 0.203 0.18 0.198 Operating 1.33 1.319 1.32 1.318 Limit MCPR It is therefore concluded that the analysis performed by RETRAN is acceptable.

) s 1

1

.j tj LTR-045' a- - Rev.L0. Page 106.of.1135 '8' [} l :- _ 1 Table:4.1; s y.. lq l 'TTWOBP Events. Sequence * -l l 5 .r . Events' Time <1 0.0.. Steady; State.Initialir.ation 1.E-6 Turbine Stop. Closure'(TSV) i 0.01 Scram Initiated by.101.TSV. Closure 1 a 0.07 RPS' Scram Solenoid De-energized: 1 0.1. Turbine.Stop' Valves Fully; Closed .0.27 Control Rods Begin^~to'. Enter Core. Zone- ] 5.6 Control Rods.arel Fully Inserted 8.0 Transient Terminated" -l 3 I i

Safety valves are assumed closed. only relief _ valves open.

1 -l <j l -l a i'l l I i a l o i "{

TR-045 Rev. O Page 107 of 135 Table 4.2 Licensing Model Specific Parameters Core Power 1930 MWTH f Dome Pressure 1020 PSIG Vessel Steam Flow 7.25E6 LB/HR Control Rod Insertion Table 67A-0PL-3 Bypass Flow Fraction 12.3% Kinetics (One-Dimensional) End of Cycle 10 Initial Control Valve Position 82.7% Open Separator Carryunder 0.2% l l

-s -v-J TR-045 [Rev. O Page 108 of 135 Table 4.3 Code Options Used for Sensitivity Analysis ISFLAG 1 Dynamic Slip l JSST 0 Steady State Initialization Used IHTMAP 1 Combination Forced Convection and Free Convection Maps, q with Condensation 1 INEX <L 1 Iterative Solution Techniqup "1 N3TK 0 Local Conditions Not Used (used during start'up tests). IVOID 0 Void from Profile Fit Equation not used i NSHAPE 1 Flux spatial shape function calculated every time step until past the flux peak where it is gradually'celaxed COURANT 0.0 Default used (0.3) until the later part of the transient I j when it is relaxed (past severe disturbances' regions). ) . d i l 1 a ,t l ) l \\ l i

i 4 TR-045 i Rev. O Page 109 of 135 1 s 1 Table 4.4 i 1 l Summary of Sensitivity Analysis Perturbation ARCPR Core: Direct Moderator Heating: Reduced 25% +0.0178* Gap Conductance Reduced 50% +0.0085 j Void Profile Fit Included -0.0462 AP Increased By 1.5 psi +0.00427* j Separators: Carryander Reduced 50% +0.005* Inlet Inertia Increased by 100% +0.0185-AP Reduced By 0.5 psi +0.0064* Steam Lines: AP Reduced 10% +0.0142* 2 Volume Reduced 5% +0.0106* Inertia Increased by 7% +0.0163* Recirculation Loops: Volume Reduced By 5% +0.0149* Vessel: Steam Dome Volume Reduced 10% +0.011* Lower Downcomer Volume Reduced 10% +0.002* Lower Plenum Volume Reduced by 10% +0.002* Nuclear Sensitivity: +0.021* Included in calculating the CPR uncertainty (A) given in Section 4.1.2(f). 1 l l 1

.h q TR-045l { LRev.,0

j Page~110 of,135
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l 4 . Table 4.5 .) Summary of Code Options Used in Licensing'Model l ISFLAG 21 . Algebraic Slip-1 d 'JSST 'O Steady State Initialization IHTMAP" 1 Combined' Forced Convection and Free? Convection Maps,. 'with Condensation: INEXPL 1 Iterative Solution Technique 1 >j 1 NSTK 0 No Local. Conditions Heat. Transfer j 1 IMCL.R 10 Groeneveld.5.9 Heat' Transfer Correlation IVOID 1 Subcooled Void Profile' Fit with Default Option ] (Transient Behavior'According.to Water Density Ratio, -l Relative to. Steady. State). '1 i NFIT 1 Rod Bundles Parameters Used ~ NSHAPE 1 as in Table 4.3 COURANT 0.0 as in Table 4.3 h o u l 1 i ai .) Ej c; .l

T3-045 Rev. O Page ill of 135 Table 4.6 Hot Channel Parameters EOC l 9.E EH Peaking Factors: Local 1.20 1.28 i Radial 1.738 1.650 I' Axial .1.40 1.40 1 R-Factors 1.051 1.098 Bundle Power (MWt) 5.839 5.553 Bundle Flow (10' lb/hr) 91.13 90.75 ? I i

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4w6i uomM a ;g, 8 N 7 A I I R N T Y E D R O PTT 6 o I 1 t d. e s ) a C B i 5 E e I l l a S c ( S e E m. i M T 4 l I I T w o l D F s E s 3 a l I S p y P B. A L F C 2 E I l W F 6 3 4 i ! ! e I 1 r f::? i ug i elll F 0 0 0 0 0 O 0 0 0 0 0 5 0 5 2 1 1 $=mJ. B 3m )04n.>.D = I ,l1 ,lL

L r 18-079 8eA' O ' deSe LCC 01 LCS t [ n e e l I Z } 4-L )2 2 4-mo 5 32 O 4-4- i g m g I j .a r i = O ne m 2 l M ne ) ^ e 3 m d I E l ( E ( 4= o C m n ~ ~~ i, M 2 m g y y ,.....,. *D r I N M n' h M,w a Em 12 i = m 5 5 e m l 8 E 2 3 1 n a a S/MBL, WOLF SEVLAV FEILER I l

i j b .TR-045' l 1 j Rev. O Page 134-of.-135 5.0'~ Conclusion i P-Thistopical.reportpresented.an'applicationmethodologyusingRhTRAN-02' l .1 MOD 4 for reload safety analysis for Oyster l Creek. It presented a model- / j qualification program that consisted of-nine plant startup tests showing.. 1 the robustness'of the model'in simulating a wide range of plant-transients using the same'noding diagram...A sensitivity analysis was d I then. carried out to arrive at a conservative safety margin that would. encompass system pararueters uncertainties at the 95/95' upper bound' -{ l limit. Finally..in order to show that'the application methodology works, it was used to analyze Oyster Creek Cycle 10 reload transients (TTWOBP, 'i MSIV ATWS, FWCF) and to compare the results against the vendor's-(GE) 'J i J analysis for that Cycle. Excellent agreement was obtained. showing that using RETRAN-02 MOD 4 with the above methodology gives the required capability to do reload analyses. o ) l' 1 1

TR-045 Rev. O Page 135 of 135

6.0 REFERENCES

1. Dyster Creek Nuclear Generating Station, updated Final Safety Analysis Report, 1984 2. RETRAN A Program for Transient Thermal-Rydraulic Analysis of Complex Fluid Flow Systems - EPRI NP-1850-CCMA, 1984. 3. EXXON Nuclear Company. "The EXKON Nuclear Company WREM - Based NJP - BWR ECCS Evaluation Model & Application to the Oyster Creek Plant". KN-75-55, Revision 2, Aug 1976. 4. M. J. May, "0yster Creek Startup Tests " NEDE-13109, 1970. 5. M. A. Alammar, "0yster Creek RETRAN Model Startup Tests Benchmark," GPUN TDR 824, 1986. 6. P. J. Woods, "0yster Creek Finai Transient Analysis Report," GECR-5596, 1968. 7. M. A. Alammar, "0yster Creek Feedwater and Pressure Regulator Control Systems RETRAN Models," TDR 415, 1983. l 8. " Guidelines for Generating OPL-3 Inputs," NEDE-22061, 1982. l 9. " General Electric Reload Fuel Applications for Oyster Creek," l NEDD-24195, 1983 (Cycle 10). 10. " Methods for the Generation of Core Kinetics Data for RETRAN-02," I GPUN TR-033, 1987. 11. "One Dimensional Core Transient Model," NEDO-24154, 1978. l 12. " Station Nuclear Engineer Manual," NEDO-248100, 1986.

13. " Methods for the Analysis of Boiling Water Reactors. A Systems Transient Analysis Model (RETRAN)," YAEC-1233, 1981.

14. S. L. Forkner, et. al. "BWR Transient Analysis Model Utilizing the l RETRAN Program," TVA-TR81-01, 1981. 15. " General Electric Thermal Analysis Basis: Data, Correlation and Design Application," NEDE-10958-A, January 1977. 1 16. R. V. Furia " RACE: A Routine to Analyze Transient CPR Events," GPUN TDR 655, 1985. l l l

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ML20235J143

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6.0 REFERENCES

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ML20235J143

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1.0 INTRODUCTION

5.0 CONCLUSION

6.0 REFERENCES

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6.0 REFERENCES

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