ML20073Q829

Small Break LOCA Analysis Methodology

ML20073Q829
Person / Time
Site:	Comanche Peak
Issue date:	05/17/1991
From:	Brozak D, Husain A, Tajbakhsh A TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20073Q827	List: ML20073Q826 ML20073Q829
References
RXE-91-004, RXE-91-4, NUDOCS 9106040015
Download: ML20073Q829 (60)

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I SMALL DREAK LOSS OF COOLANT ACCIDENT ANALYSIS METilODOLOGY MAY, 1991 i A. E. Tajbakhsh D. E. Droz a);

11 . C. da Silva, Jr.

Reviewed: f - 1 Date: - / Y/

/ ~U Whee M Choo Supervisor, LOCA Analysis Approved:

A- eruw Date: 5/l7}91 Ausjif liusain '

Director, Reactor Engineering

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DISCLAIMER j The information contained in this report was prepared for the specific requirement of Texas Utilities Electric Company (TUEC), and may not be appropriate for use in situations other

.than-those for which it was specifically prepared. TUEC  !

PROVIDES NO WARRANTY llEREUNDER, EXPRESSED OR IMPLIED, OR STATUTORY, OF ANY KIND OR NATURE WHATSOEVER, REGARDING TilIS REPORT OR ITS USE, INCLUDING BUT NOT LIMITED TO ANY WARRANTIES ON MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE.

By making this report available, TUEC does not authorize its use by others, and any such use is forbidden except with the prior written approval of TUEC. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warrants provided herein. In no event shall TUEC have any liability for any incidental or consequential damages-of any type in connection with the use, authorized or unauthorized, of this report or for the information in it.

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I ABSTRACT This report is presented to demonstrate the application of the USNRC-approved Advanced Nuclear Fuels (ANT) Corporation's Emergency Core Cooling Systems (ECCS) Evaluation Model EXEM pWR Small Break Model, to the Comanche Peak Steam Electric Station (CPSES).

This report contains a description of the EXEM PWR Small Break methodology which includes the computer codes, the details of the nodalization schemes, and the calculational procedures followed during all phases of the LOCA. The methodology is used to perform small break LOCA-ECCS licensing analyses that comply with USNRC regulations contained in 10 CFR 50.46 and 10 CFR 50, Appendix K. The method also satisfies the requirements of NUREG-0737, THI Action Item II.K.3.30.

In order to comply with a 10 CFR 50, Appendix K requirement, a full spectrum of small breaks, ranging from 4 to 8 inches in diameter, is examined.

Furthermore--in order to support the technical specification linear heat generation rate ( LHGR) limit as a function of height-- several potentially limiting power shapes are considered. Analyses are presented for the chopped cosine and two additional shapes, one of them most limiting according to lii

the vendor's FSAR analysos and a similar one which was  :

developed using the methods described in Reference 1.2.

1 The present methodology-including all codos, input docks and conclusions reached within this report-Vill be applied to subsequent fuel cycles for the comancho peak steam Electric I Station Unit One and Unit Two. Evaluations will be performed on the basis of the cyclo-specific paramotors to verify that the results of the prosont analysos remain bounding.

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4 TABLE OF CONTENTS l

PAGE 1

9 DISCLAIMER . . . . . . . . . . . . . . . . . . . . . . . 11 ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . 111 TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . v LIST OF TABLES . . . . . . . . . . . . . . . . . . . . vii LIST OF FIGURES . . . . . . . . . . . . . . . . . . . viii CilAPTER

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . 1-1 l

2. DESCRIPTION OF Tile METHOD . . . . . . . . . . . . . . 2-1 2.1 DETERMINATION OF INITIAL FUEL PARAMETERS . . 2-1 2.2 SYSTEM TiiERMAL-ilYDRAULIC RESPONSE ANALYSIS . 2-2 2.3 FUEL ROD T}{ERMAL RESPONSE ANALYSIS . . . . . 2-4

2.4 DESCRIPTION

OF TiiE MODELS . . . . . . . . . . 2-5 2.4.1 CPSES-1 ANF-RELAP NSSS MODEL . . . . . . 2-5 2.4.1.1 VOLUMES, JUNCTIONS AND llEAT STRUCTURES . . . . . . . . . 2-6 2.4.1.2 CORE POWER . . . . . . . . . . - . . . 2-9 2.4.1.3 EMERGENCY CORE COOLING SYSTEMS . . . 2-10 2.4.1.4 TRIPS AND DELAYS . . . . . . . . . . 11 2.4.2 TOODEE2 MODEL . . . . . . . . . . . . . . 2-13 y

- . _ . . - . _ . _ _ - . _ - , _ . _ _ _ . _ _ . . ~ . _ _ . . . _ . . _ . . . . . _ _ . _ _ - - . _ , . , _ _ _ . , _ _ . ,_ ._, - - - - . _ _ - - . . - -

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3. BASE CASE ANALYSIS AND SENSITIVITY STUDIES . . . . . . 3-1  ;

3.1 BASE CASE ANALYSIS . . . . . . . . . . . . . . . . 3-5 3.2 SENSITIVITY STUDIES . . . . . . . . . . . . . . . 3-11 3.2.1 BREAK SPECTRUM . . . . . . . . . . . . . . . 3-11 3.2.2 AXIAL POWER SHAPE . . . . . . . . . . . . . 3-18

4. CONCLUSION . . . . . . . . . . . . . . . . . . . . . . 4-1

5. REFERENCES . . . . . . . . . . . . . . . . . . . . . . 5-1 ,

APPENDIXt DESCRIPTION OF COMPUTATIONAL TOOLS . . . . A-1 l

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LIST OF TABLES TABLE PAGE 2.1 CPSES-L ANP-RELAP Nodalization Summary . . . . 2-15 2.2 Summary of CPSES-1 ANP-RELAP System Model Components . . . . . . . . . . . . 2-22 2.3 Density Reactivity Table . . . . . . . . . . . 2-28 0 2.4 Doppler Reactivity Table . . . . . . . . . . . 2-29 2.5 Scram Reactivity Table . . . . . . . . . . . . 2-30 [

2.6 ECCS Flow vs. Pressure . . . . . . . . . . . . 2-31 2.7 . Trips and Delays . . . . . . . . . . . . . . . 2-32 2.8 Fuel Assembly / Rod Data . . . . . . . . . . . . 2-33 2.9 Steam Generator Safety Valves Flow Rates . . . 2-34 f

3.1 Summary of CPSES-1 Small Break LOCA Accident Assumptions for Base Case and Sensitivity Studies . . . . . . . . . 3-18 3.2 Summary of Initial Conditions for CPSES-1 Small Break LOCA Base and Sensitivity Studies . . . . . . . . . . . . . . 3-19 3.3 Summary of Fuel Parameters for Base Case Small Break LOCA Analysis . . . . . . 3-20 3.4 Sequence of Events for Base Case Small Break LOCA . . . . . . . . . . . . . . . 3-21 3.5 Sequence of Events for Break Spectrum Study . . . . . . . . . . . . . . . . 3-23 3.6 Sequence of Events for Power Study . . . . . . . . . . . . . . . . . . . . . 3-23 4.1 Summary of Results for Base Case and Sensitivity Studies . . . . . . . . . . . . 4-4 A.1 Input and Output for the EXEM/PWR Methodology Computer Codes (Refer to Figure A.1) . . . . . A-8 l

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LIST OF FIGURES FIGURE -

PAGE 2.1 Schematic of ENC Small break Model . . . . . . 2-35 2.2 CPSES-1 ANF-RELAP SDLOCA Nodalization Diagram . 2-36 2.3 TOODEE2 Nodalization Diagram . . . . . . . . . 2-37 3.1 Axias Power Shapes for SDLOCA Analyses . . . . 3-24 3.2 Core Power . . . . . . . . . . . . . . . . . . 3-25 3.3 Primary and Secondary System Pressures . . . . 3-25 3.4 Intact and Broken Loop AFW and Steam Flows . . 3-26 3.5 Central Core Region Void Fractions . . . . . . 3-26 3.6 Average Core Region Void Fractions . . . . . . 3-27 3.7 Reactor Vessel Downcomer Water Level . . . . . 3-27 3.8 Lower Plenum Water Level . . . . . . . . . . . 3-28 3.9 Intact and Broken Loop SG Inventories . . . . . 3-28 3.10 Central Core Clad Temperatures . . .. . . . . 3-29 3.11 Intact and Broken Loop SG Downhill Collapsed Level . . . . . .. . . . . . . . . . 3-29 3.12 Accumulator Flow Rates . . . . . . . . . . . . 3-30 3.13 Collapsed Level in Central Core Region . . . . 3-30 3.14 Primary System Inventory- . . . . . . . . . . . 3-31 3.15 Break Flow Rate . . . . . . . . . . . . . . . . 3-31 3.16 Pumped ECCS Injection Flow Rate . . . . . . . . 3-32 3.17 -TOODEE2 Surface Clad Temperatures . . . . . . . 3-32

• 3.18 Core Power . . . . . . . . . . . . . . . . . . 3-33 3.19 Primary and Secondary System Pressures . . . . 3-33 3.20 Intact and Broken Loop AFW and Steam Flows . . 3-34 vili

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i 3.21 Central Core Region Void Fractions . . . . . . 3-34 i 3.22 Average Core Region Void Fractions . . . . . . 3-35  ;

3.23 RV Downcomer Water Level . . . . . . . . . . . 3-35 3.24 Lower Plenum Water Level . . . . . . . . . . . 3-36 3.25 Intact and Droken Loop SG Inventories . . . . . 3-36 3.26 Central Core Clad Temperatures . . . . . . . . 3-37 3.27 Intact and Broken Loop SG Downhill Collapsed Levels . . . . . . . . . . . . . . . 3-37 '

3.28 Accumulator Flow Rates . . . . . . . . . . . . 3-38 3.29 Collapsed Level in Central Core Region . . . . 3-38 3.30 Primary System Inventory . . . . . . . . . . . 3-39 3.31 Dreak Flow Hate . . . . . . . , . . . . . . . . 3-39 3.32 Pumped ECCS Injection Flow Rate . . . . . . . . 3-40 3.33 TOODEE2 Surface Clad Temperature . . . . . . . 3-40 3.34 Core Power . . . . . . . . . . . . . . . . . . 3-41 3.35 Primary and Secondary System Pressures . . . . 3-41 3.36 Intact and Broken Loop AFW and Steam Flows . 3-42 3.37 Central Core Region Void Fractions . . . . . . 3-42 3.38 Average Core Region Void Fractions . . . . . . 3-43 3.39 Reactor Vessel Downcomer Water Level . . . . . 3-43 3.40 Lower Plenum Water Level . . . . . . . . . . . 3-44 3.41 Intact and Broken Loop SG Inventories . . . . . 3-44 3.42 Central Core Clad Temperatures . . . . . . . . 3-45 3.43 Intact and Broken _ Loop SG Downhill Collapsed Level . . . . . . . . . . . . . . . . 3-45 3.44 Accumulator Flow Rates . . . . . . . . . . . . 3-46 3.45 Collapsed Level in Central Core Region . . . . 3-46 ix .

l 3.46 Primary System Inventory . . . . . . . . . . , 3-47 3.47 Break Flow Rate . . . . . . . . . . . . . . . . 3-47 3.48 Pumped ECCS Injection Plow Rate . . . . . . . . 3-48 3.49 TOODEE2 Surface Clad Temperaturec . . . . . . . 3-48 3.50 TOODEE2 Surface Clad Temperatures . . . . . . . 3-49 3.51 TOODEE2 Surface Clad Temperatures . . . . . . . 3-49 A.1 ANF EXEM PWR SBI4CA Computer Code Interfaces . A-15 P

X

CHAPTER 1 l l

i INTRODUCTION i

The present report describes the application of the USNRC-approved (Reference 1.1) Advanced Nuclear Fuels (ANP, formerly Exxon Nuclear) Corporation's ECCS Evaluation Model, entitled "EXEM pWR Small Break Model", to the Comancho Peak Steam Electric Station Unit One (CPSES-1).

The method is used to perform the Small Break LOCA-ECCS (Emergency ; ore Cooling Systems) licensing analyses that comply with USNRC regulations contained in 10 CFR 50.46, 10 CFR 50, Appendix K, and the requirements of NUREG-0737, THI Action Item II.K.3.30.

The analyses presented in this report includo a description of the EXEM PWR Small Break methodology (Chapter ),

including the details of the nodalization schemes and procedures followed during all phases of the LOCA, which is postulated to occur with the plant in normal operation. Each calculation is performed in enact compliance with the explicitly approved EXEM PWR Small Break methodology.

Regarding features of the calculation procedure which are

" implied" in the approval, there has been but one deviation:

the thermal-hydraulic calculations represent the average core i 1-1

-- - . . . . . . - - - , ~

u s 1 region using nine axial nodes (rather than the three shown in ANF's submitthl). This deviation has been made in order to increase accuracy.

Two types of sensitivity studies are presented in Chapter 3.

The first is a break spectrum study. Breaks ranging from 4 to 8 inches in diameter are examined in order to comply with 10 CFR 50.46 (a) (1) (i) .

The second type of ser.sitivity study examines all realistic potentially limiting axial power shapes in order to support the LHGR limit as a function of height and to satisfy the requirements of 10 CUR 50, Appendix K Part I, A, 1. This is donc as follows: First the pcpulation of shapes is developed through the axial power distribution control analysis described in Reference 1.2. Then, the shapes which are closest to the Technical Specification LHGR limit are selected. After that, the selected shapes are adjusted upward until the axial power shape curve touches the curve representing the Technical Specification LHGR limit as a function of core height. Finally, the shapes which are most likely o have the highest integrated power up to the PCT elevation are selected. Analyses are presented for the chopped cosine (Shape 2, Figure 3.1) and two additional shapes, one of them most limiting according to the vendor's FSAR analyses (Shape 1, Figure 3.1) and a similar one (Shape 1-2

d e 3, Figure 3.1) which was developed using the method described above. These are the most likely candidates to yield the highest PCT according to the criterion just described.

In chapter 4, results from these sensitivity studies are used to establish the mcst limiting small break LOCA case for the EXEM PWR Small Break methodology and to show compliance with the LOCA-ECCS criteria in 10 CFR 50, Appendix K for CPSES-1.

The Appendix provides a description of the codes used in the EXEM PWR Small Break methodology, their interfaces, interrelationships, and respective inputs and outputs.

The main objective of the work performed in connection with the present report is to obtain approval cf this methodology

-- including all codes, input decks, inferences and conclusions -- so that it may be applied to the Comanche Peak Steam Electric Station Unit One and Unit Two for subsequent fuel cycle analyses and to address pertinent licensing issues at any time. Evaluations will be performed on the basis of specific parameters to insure that results of the present analyses remain boundirig, whenever such analyses are warranted.

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CHAPTER 2 r

DESCRIPTION OF THE METHOD The present report describes the application of Advanced Nuclear Fuels' ECCS Evaluation model, entitled "EXEM PWR Small Break Model", to the Comanche Peak Steam Electric Station Unit One.

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The EXEM PWR Small Break methodology is illustrated -

O schematically in Figure 2.1. For presentation purposes the methodology can be said to embody three basic types of calculations: (1) Determination of Initial Fuel Parameters, (2) System Thermal-Hydraulic Response Analys!'. and (3) Fuel Thermal Response Analysis. These are discussed in the sections that follow. Details of the codes used in these calculations lncluding interfaces, interrelationships, inputs and outputs are provided in the Appendix.

2.1 DETERMINATION OF INITIAL FUEL PARAMETERS Calculationn are required to determine initial fuel conditions for both ANF-RELAP and TOODEE2. For ANF-RELAP these are peak stored energy in the form of fuel temperatures. For TOODEE2 these also include: fission gas 2-1

inventory, gap width, crack and plenum volume. These calculations are performed using the RODEX2 code. This code is also part of the EXEM/PWR methodology (Reference 2.7) currently used in performing large break loss-of-coolant accident analyses that comply with 10 CFR 50.46 and Appendix K thereto.

Four RODEX2 calculations are normally performed. The first is to determine the time of peak stored energy. The next two calculations determine the initial fuel temperatures to be used to initialize ANF-RELAP fuel temperatures in each of the two core regions, using the previously determined maximum stored energy and the power in each region. The fourth calculation sets the initial conditions for TOODEE2, 2.2 SISTEM THERMAL-HYDRAULIC RESPONSE ANALYSIS The system Thermal-hydraulic response is analyzed using ANF-

RELAP, a modified version of RELAPS/ MOD 2, INEL Cycle 36.02.

The ANF-RELAP system model used for CPSES-1 is described in detail in Section 2.4.1. The initial conditions of the ANF-I RELAP fuel rod model, i.e. stored energy are determined in a i

separate set of calculations using RODEX2, as described in Section 2.1 and in the Appendix. The RELAPS/ MOD 2 code is described in detail in Reference 2.2. In addition to less 2-2

significant changes and corrections, RELAPS/ MOD 2 has been modified in three major ways to produce ANF-RELAP:

l (1) The Moody critical flow model (Reference 2.3) substitutes the RELAPS/ MOD 2 critical flow model during two-phase discharge in order to comply with the related requirement of 10 CFR 50, Appendix K, Section I, C, 1, b.

(2) The RELAPS/ MOD 2 mixture level calculation is modified with the objective of producing a two-phane letal more suitable for the TOODEE2 fuel rod tnermal analys's.

(3) A counter-current flow limitation (CCFL) constitutive equation based on a Kutateladze formulation with constants adjusted on the basis of the S-UT-08 test (Reference 2,4) is made available for use instead of the mechanistic interphase drag models in vertical junctions which can be selected by the user.

The ANFmRELAP calculation providtes the thermal-hydraulic boundary conditions for the fuel thermal response analysis, which is performed using the TOODEE2 code (Reference 2.5).

Seven sets of parameters are supplied from ANF-RELAP, as shown in Figure 2.1:

(li Saturation temperature entering the central core.

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(2) Vapor mass flow rate exiting the central core.

(3) Normalized core power.

(4) Position of the two-phase contral core mixture level.

(5) Average quality of the fluid below central core mixture level.

(6) Mass flow of liquid entering the central core.

(7) Temperature of the liquid entering the central core.

2.3 FUEL ROD THERMAL RESPONSE ANALYSIS TOODEE2 (Reference 2.5) is used to calculate the hot fuel rod heat up during the entire accident. It is part of the original WREM package approved by the NRC (Reference 2.6) and is also part of the EXEM/PWR methodology (Reference 2.7) currently used in performing large break loss-of-coolant accident analyses that comply with 10 CFR 50.46 and Appendix K thereto.

TOODEE2 is a two-dimensional, time-dependent fuel rod thermal and mechanical analysis program. TOODEE2 models the fuel rod as radial and axial nodes with time-dependent heat sources.

Heat sources include both decay heat and heat generation via reaction of water with zircaloy. The energy equation is solved to determine the fuel rod thermal response. The code considers conduction within solid regions of the fuel, 2-4

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radiation and conduction across gap regions, and convection and radiation to the coolant and surrounding rods.

The outputs of TOODEE2, viz., peak clad temperature, percent local cladding oxidation and percent pin-wide cladding oxidation are compared to the 10 CFR 50.46 (b) (1) through (3) criteria. Regarding (3), if pin-wide oxidation is less than 1% it is concluded that the criteria of less than 1%

core-wide oxidation (3) is met.

2.4 DESCRIPTION

OF THE MODELS l 2.4.1 CPSES-1 ANF-RELAP NSSS MODEL l l

1 The Comanche Peak Steam Electric Station consists of two l Westinghouse pressurized water reactors. Both units are typical four-loop plants with a rated thermal power of 3411 l

MWt each. '

The CPSES-1 ANF-RELAP NSSS model results from a considerable amount of engineering insight and experience gained over the past four years and incorporates:

a. Information from the most recent plant drawings, design basis documents, vendor documents, Technical Specifications and Final Safety Analysis Report.

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b. Careful consideration of the guidelines set forth by ANF for the application of their methodology (Reference 2.8).

This section describes the ANF-RELAP base input model for the Comanche Peak Steam Electric Station Unit One (CPSES-1). The discussion of this model is divided into the following sub-sections:

1. Volumes, junctions and heat structures

2. Core power

3. Emergency core cooling systems

4. Trips and delays 2.4.1.1 VOLUMES. JUNCTIONS AND HEAT STRUCTURES Figure 2.2 shows the CPSES-1 nodalization diagram for the ANF-RELAP base input mJdel which is comprised of 114 regular volumes, 15 time dependent volumes, 124 regular junctions, 15 valve and time dependent junctions, 18 active heat structures representing nine axial nodes in two different regions of the core and 60 passive heat structures. Table 2.1 identifies the volumes, junctions, and heat structures associated with each region or system and provides their node number for cross-referencing with Figure 2.2. Table 2.2 summarizes key 2-6

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parameter values for each of the CP3ES-1 ANF-RELAP NSSS model components.

Three of the four loops are assumed to be identical and are modeled as one intact loop with appropriately scaled input.

The pressurizer is connected to this loop following ANF .

1 procedures (Reference 2.8). These represent the " unbroken" loops. The " broken" loop is a single loop.

The reactor core is divided into two concentric radial sections: a central region and a peripheral region. The peripheral annulus produces approximately 70% of the power '

and the central region 30%. ?3oth , the central region and the poripheral annulus are divided into nine axial volumes.

Crossflow betweets these concentric regions is taken into consideration by flow junctions in the ANF-RELAP model which are established as shown in Figure 2.2. The fuel in the core is represented by twc sets of heat structures. One set represents the fuel in the peripheral region and is associated with those fluid volumes. The other set of heat structures represents the hotter assemblies in the central region. All core fuel assembly heat structures are divided into nine axial elements.

Steam generator models include both primary and secondary sides. A detailed nodalization of the steam generator 2-7

secondary has been implemented in order to insure realistic heat transfer behavior across the steam generator tubes.

Steam generator pressure relief is obtained by simulation of the safety valves only (Table 2.9), i.e. credit is not taken for the heat removal capability of the steam dump and bypass syr ve percent of the steam generator tubes are ass. . ..,ged. This ansumption is made to support the potential .<eed for operation under these circumstances and is a conse,vative assumption for fewer obstructed tubes.

Reactor coolant pumps are modeled using Westinghouse homologous curves in the single phase regime combine,1 with homologous difference and multiplier curves for the CE-EPRI tests in the two-phase regime. The CE-EPRI reactor coolant pump data were reviewed and found to be applicable to CPSES reactor coolant pumps.

The containment is represented by a time dependent volume (TMDPVOL) with constant atmospheric pressure. This is done for simplicity.

2.4.1.2 CORE POWER The total core power during transients is determined by the point reactor kinetics model in ANF-RELAP. Conservative 2-8

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input data are entered for this model in order to compute the fission power and decay heat per 10 CFR 50, Appendix K requirements. The model accounts for the reactivity effects associated with scram, change in moderator density and in fuel temperature. The effects are evaluated on a core average, cycle specific basis using the reactor physics methodology and associated uncertainty factors presented in Reference 2.9 to assure conservatism. For the analyses presented herein, reactivity feedbacks representative of the CPSES-1 core have been selected and are shown in Tables 2.3, 2.4 and 2.5 for moderator density effects, fuel temperature effects and scram, respectively.

All core power is conservatively assumed to be generated in the fuel, i.e. none is deposited in moderator, cladding, or passive heat structures. This power is distributed according to the nodal power factor (NPF) entered for each active heat structure that represents a portion of UO2 fuel.

2.4.1.3 EMERGENCY CORE COOLING SYSTEMS The CPSES ECC system is arranged into four subsystems: (1) the high head charging / safety injection, (2) intermediate 2-9

head safety injection, (3) low head residual heat removal injection, and (4) accumulators, l

There are two safety injection trains. Each train contains one centrifugal charging pump, one intermediate head safety injection pump, and one low head residual heat removal pump with associated piping, valves, controls, and instrumentation.

' Since loss of offsite power is asrumed to occur coincidentally with the reactor trip, only one train is represented in the present NSSS model. The other train is removed on the assumption that one diesel generator train fails to start. This assumption is made in order to satisfy the single failure criterion, as discusced in Chapter 3.

All pumped systems take suction from the refueling water storage tank (RWST) during the injection stage. In the present analyses the RWST water temperature is taken at the maximum value (120 degrees F) allowed by the Technical Specifications. This is conservative since it minimizes heat removal by sensible heat transfer to injected fluid.

The flow per each loop versus pressure values for each injection system, which are given in Table 2.6 were derived from the values given in Reference 2.10. These values 2-10

conservatively underestimate injection into the RCS because they incorporate spillage of injection to the containment while inventory depletion associated with spillage is already accounted for in the ANF-RELAp break flow calculation. In order to provide additional margin, the injection capacities of Reference 2.10 are reduced by 2% in deriving the values of Table 2.6.

The system contains four accumulators, one per loop. The three intact loop accumulators are lumped. The minimum Technical Specifications (Reference 2.11) tank water volume (6119 gals. per tank) is used. The accumulators are modeled using a two-volume PIPE component (as opposed to the ACCUM component) for simplicity, per ANF methodology.

2. 4.1. 4 TRIPS AND DELAYS The following trips and delays are used:

1. Reactor trip occurs on a low pressurizer pressure signal (1860 psia, Reference 2.8) plus a 2 second delay for signal processing. The 2.4 second rod travel time is accounted for by the scram reactivity (Table 2.5).

2. The reactor coolant pumps (RCP) are tripped at reactor trip, as discussed in Chapter 3. This trip 2-11

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occurs because at reactor trip offsite power is assumed lost.

3. Steam flow isolation is initiated at the time of reactor trip with the turbine stop valves taking 0.5 seconds to close (Table 2.7). The steam dump and bypass system is not credited. The safety valves operate as shown in Table 2.9.

4. Main feedwater isolation begins at the time of "S" signal and the 2 second signal processing time (Table 2.7).

5. SI Actuation Signal occurs on a low pressure "S" signal (1715 psia, Reference 2.8) followed by a 2 second delay for signal processing.

6. The delays for each of the pumped safety injection systems are given in Table 2.7.

7. Accumulators inject at set pressure (648 psia) without delay.

8. Available auxiliary feedwater (1 turbine , 1 motor-driven) pumps are assumed to be up and running 60 seconds after the "S" signal. Actual AFW injection 2-12 l

is delayed for another 98 seconds, conservatively accounting for the flow travel time down the piping.

One motor-driven AFW pump is assumed lost due to the unavailability of offsite power compounded with the failure of one diesel generator to start (single failure).

2.4.2 TOODEE2 MODEL TOODEE2 is used to calculate the temperature distribution in the hot rod. Table 2.8 summarizes the fuel geometry data used in the TOODEE2 model.

The present TOODEE2 model divides the fuel rod into 24 axial and 10 radial nodes.

The first and last axial nodes are identified as the bottom and top of the fuel rod, respectively. The TOODEE2 hot rod axial nodalization diagram is shown in Figure 2.3.

The fuel pellet is divided into 8 radial rings (nodes) in which the last radial line location includes the gap. The first inner fuel pellet is node 2, and gridline 1 is identified as the pellet centerline. The last gridline is identified as the cladding outer radius. The cladding is 2-13

l divided into 2 radial rings as required by EXEM PWR Small Break Model. The radial nodalization scheme is also shown in Figure 2.3.

The thermal-hydraulic bour 'ary conditions for the TOODEE2 calculation are those associated with the central core region of the ANF-RELAP model, as described in Section 2.1 and in the Appendix. The power history for TOODEE2 is also obtained from ANF-RELAP.

The initial conditions of the fuel rod, i.e. stored energy, fission gas inventory, gap width, rod plenum and crack volumes are determined in a separate set of calculations using RODEX2, as described in Section 2.] and in the Appendix.

2-14

. - - . - . . . - .. - _ .. . - .~ .-

4 +

TABLE 2.1 l

CPSES-1 ANF-RELAP NSSS NODALIZATION

SUMMARY

l l

l l

N0 DING NUMBER OF NUMBER OF NUMBER OF '

REGION OR SYSTEM DIACRAM VOLUMES JUNCTIONS HEAT STRUCTURE l NUMBER ACTIVE TMDPVOL ACTIVE TMDPJUN ACTIVE PASSIVE j REACTOR VESSEL (RV) l RV DOWNCOMER (DC) j o Upper DC 100 1 0 0 0 0 1 o Middle DC 104 1 0 3 0 0 1 i o Bottom DC 108 5 0 4 0 0 5 RV LOWER PLENUni (LP) ,

1 o Bottom LP 116 1 0 0 0 0 1 l 1

o Middle LP 120 1 0 3 0 0 1 l

o Top LP 124 l' 0 3 0 0 1 l

RV CORE BYPASS & BARREL / BAFFLE (BYPASS 1 o Bottom Bypass 128 3 0 2 0 0 3 RV CORE ACTIVE FUEL REGION o Hot Central 130 1 0 2 0 1 0 Core # 1 i o Hot Central 132 1 0 2 0 1 0 j Core # 2 -

o Hot Central 134 1 0 2 0 1 0 Core # 3 o Hot Central 136 1 0 2 0 1 0 Core # 4 o Hot Central. 138 1 0 2 0 1 0 Core # 5 o Hot Central 140 1 0 2 -0 1 0 Core # 6 o Hot Central 142 1 0 2 0 1 0 Core # 7 '

o Hot Central 144 1 0 2 0 1 0 l Core # 8 o Hot Central 146 1 0 1 0 1 0 Core # 9 l

2-15 l

l l

TABLE 2.1 (Cont'd)

CPSES-1 ANF-RELAP NSSS NODALIZATION

SUMMARY

N0 DING NUMBER OF NUMBER OF NUMBER OF REGION OR SYSTEM DIACRAM VOLUMES JUNCTIONS HEAT STRUCTURE IMiDJJL 4.CTIVE TMDPVOL ACTIVE TMDPJUN ACTIVE PASSIVE REACTOR VESSEL (RV) (Cont'd) o Average 150 1 0 1 0 1 0 Core # 1 o Average 152 1 0 1 0 1 0 Core # 2 o Average 154 1 0 1 0 1 0 Core # 3 o Average 156 1 0 1 0 1 0 Core # 4 o Average 158 1 0 1 0 1 0 Core # 5 o Average 160 1 0 1 0 1 0 Core # 6 o Average 162 1 0 1 0 1 0 Core # 7 o Average 164 1 0 1 0 1 0 Core # 8 o Average 165 1 0 0 0 1 0 Core # 9 RV UPPER PLENUM & GUIDE TUBES (UP. cts) o Bottom UP 166 1 0 5 0 0 1 o Guide Tubes 170 1 0 0 0 0 1 o Middle UP I 173 1 0 1 0 0 1 o Middle UP II 174 1 0 3 0 0 0 o Top UP 178 1 0 0 0 0 1 RV UPPER HEAD (UH) o Bottom UH 180 1 0 0 0 0 1 o Middle UH 181 1 0 4 0 0 1 o Top UH 182 1 0 0 0 0 1 2-16

TABLE 2,1 (Cont'd)

CPSES-1 ANF-RELAP NSSS NODALIZATION

SUMMARY

N0 DING NUMBER OF NUMBER OF NUMBER OF REGION OR SYSTEM DIAGRAM VOLUMES JUNCTIONS HEAT STRUCTURE NUMBER ACTIVE I@PVOL ACTIVE IliDER H ACTIVE PASSIVE INTACT LOOP (IL) PRIMARY IL_h0T LEGS (HLel o IL HL #1 210 1 0 1 0 0 1 o IL HL's # 2&3 214 2 0 1 0 0 2 o IL HL to SG 217 0 0 1 0 0 0 IL STEAM CENERATORS (SGs) o IL SG U-Tubes 220 10 0 9 0 0 8

& Inlet / Outlet Plena IL CROSS-0VER LEC'S (XLGs) o SG to RCP Inlet 223 0 0 1 0 0 0 o IL XLG 226 3 0 2 0 0 3 IL REACTOR COOLANT PUMPS (RCPs) o IL RCP 230 1 0 2 0 0 0 IL COLD LEG'S (CLs) o IL CL # 1 236 1 0 1 0 0 1 RCP Side IL CL # 263 240 2 0 1 0 0 2 RV Side INTACT LOOP SECONDARY IL HAIN AND AUXILIARY FEEDWATER (MFW & AFW) o IL MFW Source 302 0 1 0 0 0 0 o IL MFW Flow 303 0 0 0 1 0 0 o IL AFW Source 304 0 1 0 0 0 0 o IL AFW Flow 305 0 0 0 1 0 0 2-17

TABLE 2.1 (Cont'd)

CPSES-1 ANF-RELAP NSSS NODALIZATION

SUMMARY

NODING NUMBER OF NUMBER OF NUMBER OF REGION OR SYSTEM DIAGRAM VOLUMES JUNCTIONS HEAT STRUCTURE NUMBER ACTIVE IliDEYDL ACTIVE TMDPJUN ACTIVE PASSIVE INTACT LOOP SECONDARY (Cont'd)

IL SC Vessel o IL SG Downcomer 310 4 0 3 0 0 0 o IL DC to Boiler 315 0 0 1 0 0 0 o IL SG Boiler 330 5 0 4 0 0 0 o IL SG Separator 350 1 0 3 0 0 0 o IL SG Steam 360 1 0 0 0 0 0 IL SC STEAM LINE AND SAFETY VALVES o IL Steam Line 365 0 0 0 1 0 0 o IL Steam Sink 370 0 1 0 0 0 0 o IL Safety Valve 375 0 0 0 1 0 0 o IL S.V. Steam 380 0 1 0 0 0 0 Sink BROKEN LOOP (BL) PRIMARY BL NOT LECS (HLs) o BL HL # 1 410 1 0 1 0 0 1 o BL HLs # 2&3 414 2 0 1 0 0 2 o BL HL To SG 417 0 0 1 0 0 0 BL STEAM CENERATOR (SG) o BL SG U-Tubes 420 10 0 9 0 0 8

& Inlet / Outlet Plena BL CROSS-OVER LEC (XLC) o SG to RCP Inlet 423 0 0 1 0 0 0 2-18

TABLE 2,1 (Cont'd)

CPSES-1 ANF-RELAP NSSS NODALIZATION

SUMMARY

N0 DING NUMBER OF NUMBER OF NUMBER OF

-REGION OR SYSTEM DIACRAM VOLUMES JUNCTIONS HEAT STRUCTURE

_ NUMBER ACTIVE TMDPVOL ACTIVE TMDPJ1TH ACTIVE PASSIVE l

BROKEN LOOP (BL) PRIMARY (Cont'd) i I

1 o BL XLG 426 3 0 2 0 0 3 l I

BL REACTOR COOLANT PUMP (RCP)

BL RCP 430 1 0 2 0 0 0 BL COLD LEG (CL) o BL CL # 1 436 1 0 1 0 0 1 RCP Side o BL CL # 2 440 1 0 1 0 0 1 Middle Volume o BL CL # 3 441 0 0 0 0 1 RV Side BROKEN LOOP SECONDARY BL MAIN AND AUXILIARY FEEDWATER (MFW & AFW1 o BL MFW Source 502 0 1 0 0 0 0 o BL MFW Flow 503 0 0 0 1 0 0 o BL AFW Source 504 0 1 0 0 0 0 o BL AFW Flow 505 0 0 0 1 0 0 BL SG VESSEL o BL SG DC 510 4 0 3 0 0 0 o SG DC to Boiler 515 0 0 1 0 0 0 o BL SG Boiler 530 5 0 4 0 0 0 o BL SG Separator 550 1 0 3 0 0 0 o BL SG Steam 560 1 0 0 0 0 0 Dome 2-19

-. . - _ - - - . - . . _ . - - ~ . . . . - ~ - . . _ . . . . .-

TABLE 2,1 (Cont'd)

CPSES-1 ANF-RELAP NSSS NODALIZATION SUFMARY NODING NUMBER OF NUMBER OF NUMBER OF REGION OR SYSTEM DIAGRAM VOLUMES JUNCTIONS HEAT STRUCTURE NUMBER ACTIVE TMDPVOL ACTIVE TMDPJUN ACTIVE PASSIVE BROREN LOOP SECONDARY (Cont'd)

BL SC STEAM LINE AND SAFETY VALVES o BL Steam Flow 565 0 0 0 1 0 0 o BL Steam Sink 570 0 1 0 0 0 0 o BL Safety Valve 575 0 0 0 1 0 0 o BL S.V. Steam 580 0 1 0 0 0 0 Sink PRESSURIZER (PRZR) o PRZR Surge- 603 0 0 1 0 0 0 Line Flow o PRZR Surge-Line 610 1 0 1 0 0 0 o PRZR Tank 620 6 0 5 0 0 6 ACCUMULATORS (ACCUM) o IL ACCUMs 700 2 0 1 0 0 0 o IL ACCUM Surge 703 0 0 1 0 0 0 Line & Flow 704 1 0 0 0 0 0 705- 0 0 0 1 0 0 o BL ACCUMs 720 2 0 1 0 0 0 o BL ACCUM Surge 723 0 0 ~. 0 0 0 Line & Flow 724- 1 0 0 0 0 0 725 0 0 0 1 0 0 EMERCENCY CORE COOLING SYSTEM (ECCS)

T&CS o IL RHR Source. 748 0 1 0 0 0 0 o IL HHSI Source 749 0 1 0 0 0 0 o IL CCP Source 750 0 1 0 0 0 0 2-20

. . - - - . -. .- . . - . .. - - . - __.~. . _ . _ - ._ - - - - .. . _ . . .

TABLE 2.1 (Cont'd)

CPSES-1 ANF-RELAP NSSS NODALIZATION

SUMMARY

N0 DING NUMBER OF NUMBER OF NUMBER OF REGION OR SYSTEM DIAGRAM VOLUMES JUNCTIONS HEAT STRUCTURE NUMBER ACIlYE TMDPVOL ACTIVE TMDPJUN ACTIVE PASSIVE ECCS (Cont'd.1 o IL CCP Flow 751 0 0 0 1 0 0 o IL HHSI Flow 752 0 0 0 1 0 0 o IL RHR Flow 753 0 0 0 1 0 0 o BL RHR Source 768 0 1 0 0 0 0 I i

o BL HHSI Source 769 0 1 0 0 0 0 1 o BL CCP Source 770 0 1 0 0 0 0 o BL CCP Flow 771 0 0 0 1 0 0 i o BL HHST Flow 772 0 0 0 1 0 0 o BL RHR Flow 773 0 0 0 1 0 0 BREAK & CONTAINMENT DEEAK o Break. Junction 805 0 0 0 1 0 0 Valve o Containment 810 0 1 0 0 0 0 2-21

TABLE 2.2

SUMMARY

OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS COMPONENT AREA LENGTH. VOLUME VERT. ANGLE ELEV. CHANGE ROUGH. HYD. DIAM. FLAGS NUMBER TYPE (FT*2) (FT) (FT*3) (DECREES) (FT) (FT) (FT)

I 100 SNGLVOL 26.65 5.85 155.92 -90.0 -5.85 0.00 1.53 100 104 BRANCH 20.74 2.29 47.49 -90.0 -2.29 0.00 0.98 100 108-1 ANNULUS 34.07 3.93 133.89 -90.0 -3.93 0.00 1.57 100 108-2 ANNULUS 33.24 4.00 132.96 -90.0 -4.00 0.00 1.52 100 108-3 ' ANNULUS . 33.24 4.00- 132.96 -90.0. -4.00 0.00 1.52 100 108-4 ANNULUS 33.24 4.00 132.96 -90.0 -4.00 0.00 1.52 100 108-5 ANNULUS 33.63 4.05 136.21 -90.0 -4.05 0.00 1.42 100 116 SNGLVOL 47.91 2.96 141.80 90.0 2.96 0.00 7.81 100 i

120 BRANCH 114.12 2.96 337.81 90.0 2.96 0.00 12.05 100 i

124 BRANCH 96.86 4.23 409.73 90.0 4.23 0.00 11.11 100 128-1 ANNULUS 25.08 4.00 100.31 90.0 4.00 0.00 0.008 100 '

128-2 ANNULUS 25.08 4.00 100.31 90.0 4.00 0.00 0.008 100 128-3 ANNULUS 25.08 4.00 100.31 90.0 4.00 0.00 0.008 100 .

130 BRANCH 15.32 1.33 20.38 90.0 1.33 5.E-6 0.0365 000 132 BRANCH 15.32 1.33 20.38 90.0 1.33 5.E-6 0.0365 000 l 134 BRANCH -15.32 1.33 20.38 90.0 1.33 5.E-6 0.0365 000 136 BRANCH 15.32 1.33 20.38 90.0 1.33 5.E-6 0.0365 000 138 BRANCH' 15.32 1.33 20.38 90.0 1.33 5.E-6 0.0365 000 ,

r 140 BRANCH 15.32 1.33 20.38 90.0 1.33 5.E-6 0.0365 000 142 BRANCH 15.32 1.33 '20.38 90.0 1.33 5.E-6 0.0365 000 l

144 BRANCH 15.32 1.33 20.38 90.0 1.33 5.E-6 0.0365 000 p

t TABLE 2.2 (Cont'd) -

SUMMARY

OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS COMPONENT AREA LENGTH VOLUME VERT. ANGLE ELEV. CHANGE ROUGH. HYD. DIAM. FLAGS NUMBER TYPE __ (PT 2) (FT) (FT-3) (DEGREES) (PT) (FT) (FT) 146 BRANCH 15.32 1.33 20.38 90.0 1.33 5.E-6 0.0365 000 ,

t 150 BRANCH 35.75 1.33- 47.55 90.0 1.33 5.E-6 0.0365 000 i

152 BRANCH 35.75 1.33 47.55 90.0 1.33 5.E-6 0.0365 000 l 154 BRANCH 35.75 1.33 47.55 90.0 1.33 5.E-6 0.0365 000 156 BRANCH 35.75 1.33 47.55 90.0 1.33 5.E-6 0.0365 000 158 BRANCH 35.75 1.33 47.55 90.0 1.33 5.E-6 0.0365 000 160 BRANCH 35.75 1.33 47.55 90.0 1.33 5.E-6 0.0365 000 Y

U 162 BRANCH 35.75 1.33 47.55 90.0 1.33 5.E-6 0.0365 000 l' 164 BRANCH 35.75 1.33 47.55 90.0 1.33 5.E-6 0.0365 000 I 165 BRANCH 35.75 1.33 47.55 90.0 1.33 5.E-6 0.0365 000 '

166 BRANCH 84;84 1.28 108.60 90.0 1.28 0.00 0.0365 100 170 SNGLVOL 16.84 13.29 223.78 90.0 1.28 0.00 0.25 100 173 BRANCH 84.84 2.405 204.04 90.0 2.405 0.00 10.39 100 174 BRANCH 84.84 2.42 205.31 90.0 2.42 0.00 10.39 100 178 SNGLVOL 84.84 2.55 216.34 90.0 2.55 0.00 10.39 100 180 SNGLVOL 64.84 2.15 182.41 90.0 2.15 0.00 10.20 100 181 BRANCH 125.48 2.37 297.38 90.0 2.37 0.00 11.04 100 182 BRANCH 80.26 4.53 363.60 90.0 6.90 0 00 11.04 100  !

I 7

I

[

TABLE 2.2 (Cont'd)

SUMMARY

OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS .

COMPONENT AREA LENGTH VOLUME VERT. ANGLE ELEV. CHANGE ROUGH. HYD. DIM. c' LAGS NUMBER TYPE (FT-2) (PT) (FT-3) (DEGREES) (FT) (FTl (PT) 210 BRANCH 13.74 5.20 71.46 0.0 0.00 1.5E-4 2.42 100 214-1 PIPE 13.77 8.88 122.31 0.0 0.00 1.5E-4 2.42 100 214-2 PIPE 14.32 7.59 104.52 0.0 0.00 1.5E-4 2.42 100 220-1 PIPE 63.52 7.91 502.41 90.0 7.91 1.5E-4 5.19 100 220-2 PIPE 31.38 9.06 284.31 90.0 9.06 5.0E-6 0.0553 100 220-3 PIPE 31.38 7.25 227.52 90.0 7.25 5.0E-6 0.0553 100. >

220-4 PIPE 31.38 7.25 227.52 90.0 7.25 5.0E-6 0.0553 100 -

220-5 PIPE 31.38 4.44 139.32 90.0 4.44 5.0E-6 0.0553 100  !

! 220-6 PIPE 31.38 4.44 139.32 -90.0 -4.44' 5.0E-6 0.0553 100 -

i 220-7 PIPE 31.38 7.25- 227.52 -90.0 -7.25 5.0E-6 0.0553 100

, N 220-8 PIPE 31.38 7.25 227.52 -90.0 -7.25 5.0E-6 0.0553 100 t L 220-9 PIPE 31.38 9.06 284.31 -90.0 -9.06 5.0E-6 0.0553 100 '

220-10 PIPE 63.52 7.91 302.41 -90.0 -7.91 1.5E-4 5.19 100 226-1 PIPE 15.72 10.32 162.23 -90.0 -10.32 1.5E-4 2.58 100'  ?

226-2 PIPE 15.72 10.50 165.06 0.0 0.00 1.5E-4 2.58 100 i 226-3 PIPE 15.72 4.51 70.90 90.0 4.51 1.5E-4 2.58 100  :

230 PUMP 32.04 7.36 235.80 90.0 5.81 00' )

236 BRANCH 12.36 7.14 88.26 0.0 0.00 1.5E-4 2.29 100 ,

240-1 PIPE 12.36 15.76 194.88 0.0 0.00 1.5E-4 2.29 100 ,

240-2 PIPE 12.36 2.20 27.192 0.0 0.00 1.5E-4 2.29 100 ji 302 TMDPVOL 300.00 10.00 3000.00 0.0 0.00 0.00 0.00 100  !

i 304 TMDPVOL 300.00 10.00 3000.00 0.0 0.00 0.00 0.00 100 i 310-1 PIPE 584.53 7.85 4588.59 -90.0 -7.85 1.5E-4 15.72 100 310-2 PIPE 17.01 7.25 123.33 -90.0 -7.25 1.5E-4' O.34 100 310-3 PIPE 17.01 7.25 123.33 -90.0 -7.25 1.5E-4 0.34 100 310-4 PIPE 17.01 9.06- 154.11 -90.0 -9.06 1.5E 0.34 100' L

h

TABLE 2.2 (Cont'd) *

SUMMARY

OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS.

COMPONENT AREA LENGTH VOLUME VERT. ANGLE ELEV. CHANGE' R015H. HYD. DIAM. FLAGS NUMBER TYPE (FT-2) (PT) (FT-3) (DEGREES) (PT) (FT) (FT) 330-1 PIPE 162.97 9.06 1476.54 90.0 9.06 1.5E-4 0.0972 100 330-2 PIPE 166.43 7 75 1206.63 90.0 7.25 1.5E-4 0.0972 100 330-3 PIPE 166.43 7.25 1206.63 90.0 7.25 1.5E-4 0.0972 100 330-4 PIPE 155.37 4.44 692.52 90.0 4.44 1.5E-4 0.0972 100 330-5 PIPE 261.99 3.41 89I.37 90.0 3.41 1.5E-4. 10.54 100 350 SEPARATR 143.30 23.74 3401.97 90.0 23.74 0.00 7.80 100 360 SNGLVOL 383.5? 9.73 3732.15 90.0 9.73 1.5E-4 12.76 100 370 TMDPVOL 300.00 10.00 3000.00 0.0 0.00 0.00 0.00 100

[ 380 TMDPVOL 300.00 10.00 3000.00 0.0 0.00 0.00' O.00 100 410 BRANCH 4.58 5.20 23.82 0.0 0.00 1.5E-4 2.42 100 414-1 PIPE 4.59 8.88 40.77 0.0 0.00 1.5E-4 2.42 100 414-2 PIPE 4.59 7.59 34.84 0.0 0.00 1.5C-4 2.42 100 420-1 PIPE 21.17 7.91 167.47 90.0 7.91 1.5E-4 5.19 100 420-2 PIPE 10.46 9.06 94.77 90.0 9.06 5.0E-6 C.0553 100 420-3 PIPE 10.46 7.25 75.84 90.0 7.25 5.0E-6 0.0553 100 420-4 PIPE 10.46 7.25 75.84 90.0 7.25 5.0E-6 0.0553 100 420-5 PIPE 10.46 4.44 46.44 90.0 4.44 5.0E-6 0.0553 100 420-6 PIPE 10.46 4.44 46.44 -90.0 -4.44 5.0E-6 0.0553 100 l

420-7 PIPE 10.46 7.25 75.84 -90.0 -7.25 5.0E-6 0.0553 100 420-8 PIPE 10.46 7.25 75.84 -90.0 -7.25 5.0E-6 0.0553 100 420-9 PIPE 10.46 9.06 94.77 -90.0 -9.06 5.0E-6 0.0553 100 420-10 PIPE 21.17 7.91 167.47 -90.0 -7.91 1.5E-4 5.19 100 l 426-1 PIPE 5.24 10.32 $4.0B -90.0 -10.32 1.5E-4 2.58 100 l

426-2 PIPE 5.24 10.50 55.02 0.0 0.00 1.5E-4 2.58 100 426-3 PIPE 5.24 4.51 23.63 90.0 4.51 1.5E-4 2.58 100 i 430 PUMP 10.68 7.36 78.60 90.0 5.81 00 l

l

TABLE 2.2 (Cont'd)

SUMMARY

OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS-COMPONENT AREA LENGTH VOLUME ~ VERT. ANGLE ELEV. CHANGE ROUGH. HYD. DIAM. FLAGS.

NUMBER TYPE '(FT-2) (PT) (FT 3) (DEGREES) (FT) (FT) (FT)

• 436 BRANCH- 4.12 7.14 29.42 0.0 0.00 1.5E-4 2.29 100 440 BRANCH 4.12 15.76 64.96 0.0 0.00 1.5E-4 2.29 100 441 BRANCH 4.12 2.20 9.064 0.0 0.00 1.5E-4 2.29 100 502 TMDPVOL 100.00 10.00 1000.00 0.0 0.00 0.00 0.00 100

[

504 TMDPVOL 100.00 10.00 1000.00 0.0 0.00 0.00 0.00 100 510-1 PIPE 194.84 7.85 1529.53 -90.0 -7.85 1.5E-4 15.72 100 510-2 PIPE 5.67 7.25 41.11 1.5E-4

~

-90.0 -7.25 0.34 100 510-3 PIPE 5.67 7.25 41.11 -90.0 -7.25 1.5E-4 0.34 100 510-4 PIPE 5.67 9.06 51.37 -90.0 -9.06 1.5E-4 0.34 100 530-1 PIPE $4.32 9.06 492.18 90.0 9.06 1.5E-4 0.0972 100 1

• 530-2 PIPE 55.48 7.25 402.21 90.0 7.25 1.5E-4 0.0972 100 530-3 PIPE 55.48 7.25 402.21 90.0 7.25 1.5E-4 0.0972 100 '

530-4 PIPE 51.99 4.44 230.84 90.0 4.44 1.5E-4 0.0972 100 530-5 PIPE 87.33 3.41 297.79 90.0 3.41 1.5E-4 10.54 100 550 SEPARATR 47.77 23.74 1133.99 90.0 23.74 0.00 7.80 100 560 SNGLVOL 127.86 9.73 1244.05 90.0 9.73 1.5E-4 12.76 100  ;

570 TFDPVOL 100.00 10.00 1000.00 0.0 0.00 0.00 0.00 100 580 TMDPVOL 100.00 10.00 1000.00 0.0 0.00 0.00 0.00 100 i 610 BRANCH 0.683 67.49 46.10 90.0 27.89 0.00 0.683 100 620-1 PIPE 36.58 7.627 279.00 90.0 7.627 0.00 6.82 100 620-2 PIPE 36.58 7.627 279.00 90.0 7.627 0.00 6.82 100:

620-3 PIPE 36.58 7.627 279.00 90.0 7.627 0.00 6.82 100 620-4 PIPE 36.58 7.627 279.00 90.0 7.627 0.00 6.82 100

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TABLE 2.3  !

l DENSITY REACTIVITY TABLI i l

del {SITY lbm/ft*3 REACTIVITY ($) _

0.62 -54.65 6.24 -32.46 12.49 -17.63 18.73 -9.38 I 24.97 -4.56  :

31.21 -1.82 37.46 -0.47 43.87 0.00 46.82 0.15 i- .

49.94 0.30 62.43 0.60 ,

i l

t I

i I

l l

l l

2-28 l':

t TABLE 2.4 DOPPLER REACTIVITY TABLE TEMPERATURE (F) REACTIVITY ($).

200.0 1.691 400.0 1.283 600.0 0.919 800.0 0.589 1000.0 .0.284 1200.0 -0.000 1400.0 -0.267 1600.0 -0.519 1800.0 -0.759 2000.0 -0.988 2200.0 -1.207 '

2400.0 -1.417 2600.0 -1.620 2800.0 -1.816 3000.0 -2.006 3200.0 -2.189 3400.0 -2.367 3600.0 -2.541 3800.0 -2.709 4000.0 -2.874 i

2-29

l P

TABLE 2.5  ;

SCRAM REACTlVITY TABLE a

TIME (SEC) REACTIVITY ($)

0.00 0.00 0.48 0.00 0.96 -0.053 1.44 -0.133 1.92 -0.400 2.16 -0.000 l 2.40 -1.813 2.64 -3.413 2.88 -4.800 ,

3.12 -5.120 3.36 ~5.227 3.60 -5.333 1.E6 ~5.333 ,

t i

1 2-30

= , _ .-

TABLE 2.6 1

ECCS PLOW PER EACil LOOP VS. PRESSURE (98%)

' RCS CCP llPSI RilR TOTAL PRESSURE (1bm/sec) (1bm/sec) (1bm/sec) (Ibm /sec)

(psig) 0 13.42 19.73 128.54 161.29 100 12.86 19.01 16.95 48,82 120 12.75 18.86 0.0 31.61 200 12.30 18.26 0.0 30.56 1

300 11.72 17.50 0.0 29.22 400 11.13 16.69 0.0 27.82 i 500 10.54 15.84 0.0 26.38 3 600 9.92 14.92 0.0 24.84 700 9.29 13.98 0.0 23.27 800 8.65 12.96 0.0 21.61 900 7.98 1*.82 0.0 19.80 1000 7.30- 10.58 0.0 17.88 1200- 5.88 7.41 0.0 13.29 1300 5.13 5.19 0.0 10.32 1400 4.35 1.40 0.0 5.75 1500 3.55 0.0 0.0 3.55

- 1600 2.71 0.0 0.0 2.71 1800 0.84 0.0 0.0 0.84 2000 0.0 0.0 0.0 0.0 2100 0.0 0.0 0.0 0.0 2-31 y- ,cw --,-+,--..ym-

, s t

TABLI 2.7 TRIPS AllD DELAYS l

ACTION TRIPS AND DELAYS (sec) l Lo-Przr Pressure signal RCS 0 1860 psia Reactor trip 2 sec after Lo-Przr signal "S" signal Pressurizer 0 1715 paia Reactor Coolant Pumps trip Reactor Trip Main Steam Isolated 0.5 sec after Reactor Trip Main Feedwater Isolated 2 sec after "S" signal Auxiliary Feedwater Injects 158 sec after "S" signal SI Actuation Signal 2 sec after "S" signal l

Charging Pump Injects 17 sec after "S" signal l HPSI Pump Injects 2? sec after "S" signal l-l Accumulators Inject RCS 0 648 psia l

l L

I 2-32

_ . ~ . . _ . . _ _ _ . . _ . . - . . _ _ _ _ _ . _ - - - _ _ _ _ _ _ _

TABLE 2.8 I FUEL ASSEMBLY / ROD DATA PARAMETER VALUE Outer Diamotor of Fuel Rod -0.374 in Active Fuel Height 144.0 in No. of Fuel Assemblics 193 No. of Fuel Rods /Assy 264 No. of Guido Thimbios/Assy 24 No. of Instr. Tubes /Assy 1 Cladding Thickness 0.0225 in Diametral Gap 0.0065 in ,

Outer Dia. of Guide Thimble 1 0.482 in 2-33

I

[

TABLE 2.9 STEAM GENERATOR SAFETY VALVES FLOW RATES ,

Secondary Pressure SV Flow Rato (psla) (Ibm /sec) 0.0 0.0 1200.0 0.0  ;

4 1236.0 0.0  ;

1236.1 248.1 1246.3 248.1 1246.4 498.3 1256.6 498.3 1256.7 750.5 1266.9 750.5 1267.0 1004.8 1287.5 1004.8 1287.6 1263.2 4

2000.0 1263.2 i

i i

l 2-34 1

wg, , ,

6 ,

! 1 1

l l

I Initial- Fuel- Dimensions Gas inventory l Conditions initial Fuel Temperatures x

(RODEX2) s

\

'\

Initlol Stored - i Energy \

\ l Ue \Mt i

System Response .

Fuel Response Analysis =

(TOODEE2)

(RELAPS/M002) Normolized Power  :

Mixture Level

-- Fluid Temperatures Moss Flow Rates Quality Y

PCT

-Clod Oxidation Rupture-Flow Blockage Figure 2.1 Schematic of ENC Smoll Break Model i

2-35

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A) Top Skewed B) Chopped Cosine Figure 2.3 TOODEE2 Nodalization Diagrams 2-37

CHAPTER 3 BASE CASE ANALYSIS AND SENSITIVITY STUDIES Regulations pertaining to small break loss-of-coolant accident analyses require the investigation of the impact of variations in several method- and plant-specific issues on the LOCA consequences.

)

)

Method-specific issues are suggested throughout 10 CFR 50.46, Appendix K thereto and in NUREG-0737 II.K.3.30, and are addressed in Chapter 5 and Appendix A of Reference 2.1. The present work constitutes an application of an approved Evaluation Methodology, using method-specific parameters as prescribed by the method developers (Reference 2.8). Hence, the effect of variations in method-specific parameters within the bounds of methodology recommendations has already been ascertained in Reference 2.1 and sensitivity studies for these variables are not repeated here.

The plant-specific issues which warrant investigation are given in the following passages from 10 CFR 50.46, Appendix K thereto and NUREG-0611, along with the approach taken in addressing each one.

10 CFR 50.46 (a) (1) (1) , requires that "a number of postulated 3-1

. i loss-of-coolant accidents of different sizes, locations and other properties" be calculated in sufficient amount "to provide assurances that the most severe postulated loss-of-coolant accidents are calculated." In compliance with this requirement, a break spectrum study has been conducted and the results are reported in Section 3.2.

10 CFR 50, Appendix K, Part I, A, (1) states: "A range of power distribution shapes and peaking factors representing power distributions that may occur over the core lifetime shall be studied and the one selected should be that which results in the most severe calculated consequences for the spectrum of postulated breaks and single failures analyzed."

A power shape study has been conducted and the results are reported in Section 3.3, in compliance with this requirement. .

10 CFR 50, Appendix K, Part I, D, (1) states: "an analysis of possible failure modes of ECCS equipment and their effects on-ECCS performance must be made. In carrying out the accident evaluation, the combination of ECCS subsystems assumed to be operative shall be those available after the most damaging '

single failure of ECCS equipment has taken place." The

-limiting single failure for the small break loss-of-coolant accident analyses in the CPSES-1 FSAR has been-determined by

{ the NSSS vendor (Reference 3.1). It is the loss of one ECCS injection train. Unless a common cause is established, the i

3-2

I e .

loss of one ECCS injection train involves multiple failures of ECCS equipment and therefore is not a single failure. The 1 required common cr.use is the loss of power to the train. In order to arrive at this condition consistently, it must be l assumed that both the preferred 345 KV and the alternate 148 i KV offsite power sources are lost and that one emergency i diesel generator fails to start. Hence, the most damaging single failure of ECCS equipment postulated for the present study is the failure of an emergency diesel generator to start. Offsite power (which is not ECCS equipment) unavailability is postulated .in order to make the single failure meaningful in a consistent manner, i.e. the diesel generator is:not needed if either the preferred 345 KV or the alternate 148 KV offsite power sources are available. Thus, one high head centrifugal charging pump, one intermediate head safety injection pump and one-low head residual heat removal (RRR) pump (which is not challenged in these analyses) as well as all four accumulators are available to mitigate the accident and are credited in all the calculations.

Two additional conservatism are incorporated into all of the calculations in this work. ._

The first is in the initial power level, which is taken to be 3636 MWt. This power level includes both a 1.02 multiplier 3-3

.-...~-_.--...--.-_a._....

a

l to account for calorimetric error and an increase of 4.5%

above the licensed power level of 3411 MWt, representing a margin potentially available.

The second is that five percent of the steam generator tubes are assumed plugged. This assumption is made to support the potential need for operation under these circumstances and is a conservative assumption when f ewer tubes are actually obstructed.

3.1 DASE CASE ANALYSIS This section presents licensing analysis results for a 6.0 inch diameter break in the discharge line of the Reactor Coolant Pump. The axial power shape used for this base case is that determined by the vendor in the FSAR analysis (Reference 3.1) as the most limiting and is shown as power shape 1 in Figure 3.1. The fuel rod exposure which maximizes stored energy is calculated by RODEX2 and occurs at 678.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Fuel parameters used in this base case are consistent with this exposure.

The accident assumptions are summarized in Table 3.1 and the initial conditions are summarized in Table 3.2. Key fuel rod parameters are summarized in Table 3.3. Table 3.4 summarizes 3-4 I

the timing of significant events for this base case.

Figure 3.2 shows the core power dropping of f quickly at first i due primarily to the combined effects on reactivity of moderator voiding and control rod insertion and then tapering off according to decay power.

Figure 3.3 shows the primary and the secondary pressures and is used as a road map in the following discussion of system performance during this accident. The four accident periods (marked I through IV) in this figure'are characterized by their primary system depressurization rates and have the l

following characteristics:

k Period I - Denressurization:

The accident period marked I in Figure 3.3 corresponds to the rapid depressurization which follows break opening. From the secondary side standpoint, period I corresponds to a pressure rise due to steam production in the steam generators while the main steam lines are isolated and the steam dump and bypass system is not operational, as indicated in Figure 3.4.

Period II - Voidina:

Period I ends and period II begins when a substantial -

production of steam begins in the core and slows down the 3-5

i

. . I l

depressurization rate. This substantial steam production is evidenced in the rapid increase in the void fractions at mid l I

core elevation for both the central (Figure 3.5) and average l (Figure 3.6) core regions, shortly followed by a similar '

behavior in the corresponding lower nodes. The effect of this steam production is to reduce the depressurization rate of the primary system, resulting in the nearly flat system pressure which characterizes period II, as seen in Figure 3.3. Period II from the secondary side point of view begins when the pressure stabilizes near the safety valvo's set point. During period II the steam generator safety valves ,

open in order to discharge excess steam produced due to (a) the absence of feedwater between the main feedwater trip time and the time the auxiliary feedwater reaches the steam generators in combination with (b) the steam dump and bypass system unavailability. During period II there is still some heat transfer between the primary and the secondary, but this

ends with period II.

Egriod III - Heat uo l

The end of period II and beginning of period III is brought about by the end of significant steam production in the core, l

l 1.e. dryout. Thus, the end of period II and beginning of period III can be inferred from the time at which the downcomer and lower plenum levels begin to fall, as shown in Figures 3.7 and 3.8. The dropping of-these levels imply the 3-6

b* <

l i l T

core is dry and therefore steam production has essentially j ceased. Therefore, period III is characterized by an i

increased depressurization rate which is due to the lack of '

steam generation to compensate for the energy discharge 1

through the break, conversely, from the secondary side point 1

of view, period III is characterized by slow depressurization, as auxiliary feedwater injection removes '

sensible heat and raises steam generator inventories (Figure 3.9). At this time however, the primary can no longer benefit from the cooler secondary, since the primary pressure has dropped below the secondary pressure at the beginning of the period and the secondary is lost as a heat sink.  !

It is in period III that the fuel experiences its temperature excursion as shown in Figure 3.10. The clad temperature of Figure 3.10 starts to rise slightly before the beginning of period III because its axial location dries out slightly

- before steam production ceases throughout the core. With the exception of a small perturbation in the broken loop, due to injection from the SI pumps, the loop seals are clear at the beginning of period III and remain clear throughout the accident, as evidenced by the steam generator tubes' water levels shown in Figure 3.11. Therefore, loop seal plugging has a small effect on the peak clad temperature for the base

. case analysis (6 inch break). The effect of loop seal plugging is more pronounced for the 4 inch break and is discussed in that section.

3-7

_ _ _ _.m. _ . _ . _ ____...._..-. _- _ ._ u__.__ _ ___ . . . _ - . . _ . . - - __ _ _ . _ _ _ . _ _ _

Period IV - Recovervt Period III ends when the system pressure reaches the accumulator injection pressure. At that time, as shown in Figure 3.12, the massive injection of water reinstates steam production in the core causing the depressurization rate to level off again and marking the onset of period IV.  ;

Accumulator injection causes key water levels to rise (Figures 3.7, 3.8 and 3.13) and arrests the primary inventory depletion (Figure 3.14). The clad temperatures begin to turn around at accumulator injection time. Figure 3.10 shows the clad temperature histories one node above, one below and at i

the PCT location as calculated by ANF-RELAP. The rods are I

quenched from the bottom up with node 7 quenching first and node 9 last.

3.2 SENSITIVITY STUDIES 3.2.1 BREAK SPECTRUM The most limiting break location has been determined in previous studies for this (Reference 3.1) and other similar plants (Reference 3.2) to be in the cold leg at the reactor coolant pump discharge. Therefore, this cold leg break location remains most limiting for the present evaluation and a worst break location search need not be repeated. This 3-8

_ . . _ _ . . ~ . . . _ . ~ _ - . . _ _ _ . . . _ . . . -_ . . _ _ _ . . _ . _ _ _ _ _ _ _ . _ , _ _ . . - _ . _ _ . . , _ . - . ~ . - . _ _ _ _ _ _ _ . . . - . .

i most limiting break location is the one considered in all l

cases discussed throughout this work.

4 According to the approved ANF EXEM PWR Small Break Model, the break size is the first sensitivity issue addressed, holding constant the axial power shape. The rationale for addressing break size first is that system thermal-hydraulic behavior is largely affected by break size and less dependent on power shape and consequently the break size is a first order effect, while the power shapo is second order.

The break spectrum study is conducted using the power shape determined by the vendor _in the TSAR analysis (Reference 3.1) as the most limiting and at the burnup yielding the highest t

stored energy. The fuel rod exposure which maximizes stored energy is calculated by RODEX2 and occurs at 678.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Fuel parameters used in this base case are consistent with this exposure. In conclusion, the same power shape and exposure used for the base case analysis are used in the break spectrum study.

Three break sizes are examined, namely: 6 inch (base case),

t 4 inch and 8 inch, the last two sizes resulting in lower peak clad temperatures than the base case.

The accident assumptions for this and other studies are 3-9

p summarized in Table 3.1 and the initial conditions are summarized in Table 3.2. Key fuel rod parameters are summarized in Table 3.3.

The sequence of events for the break spectrum study is summarized in Table 3.5.

The result of this study is that the most limiting break is a 6 inch break located in the reactor coolant pump discharge.

The power shape sensitivity study will be performed using the limiting break size.

6 inch Break s

This is the base caso calculation described in Section 3.1.

The PCT is calculated to be 1812 'F in node 20 at 10.875 ft above the. bottom of the core. The clad temperature history as calculated by the TOODEE2 code at the node where the PCT occurs is shown in Figure 3.17.

4 inch Break The results of this calculation are similar to those of the base case (Section 3.1) so that only the key differences are pointed out here. The PCT is calculated to be 1585 'F in 3-10

e- .

node 20 in this case, 10.875 ft above the bottom of the core.

i The clad temperature history as calculated by the TOODEE2 code at the node where the PCT occurs is shown in Figure 3.33.

Figure 3.18 rhows the power behavior subject to the same mechanisms which control the base case calculation.

Figure 3.19 shows the primary and the secondary pressures.

The same four accident periode (also marked I through IV in 1

this figure) are used in the following discussion of the 4 j inch break.

1 Period I - DeDressurization:

As in the base case the accident period marked I in Figure 3.19 corresponds to the depressurization of the primary system due to the break while the secondary pressure rises to the safety valve's-set point. There are no major distinctions between system behavior during this period between the 4 inch break and the 6 inch base case except, of course, that the depressurization rate is lower.

Period II - Voidino:-

As in the base case, period I ends and period II begins when a substantial production of steam begins in the core and slows down the depressurization rate. This substantial-steam 3-11

_ _ _ _ . _ , _ . . ~ - _ - . _ _ _ _ - _ _ _ - - _ _ . _ . _ _ _ _ _ _ . - _ . . _ _ _ _ _ _ - . ._ _ ._ _ _ . _ .-. _

production is evidenced in the rapid increase in the void fractions at mid core elevation for both the central (Figure 3.21) and average (Figure 3.22) core regions, shortly followed by a similar behavior in the corresponding lower nodes. The 6 inch base case discussion for this period applies to the 4 inch break as well.

Period III - Heat un:

As in the 6 inch discussion, the end of period II and beginning of period III is also inferred when the downcomer and lower plenum levels begin to fall, as shown in Figures 3.23 and 3.24. The dropping of these icvels imply the core is dry, steam production has essentially ceased and the primary system pressure begins to drop significantly again.

i It is also in period III that the fuel experiences its temperature excursion as shown in Figure 3.26. As in the base case, the clad temperature of Figure 3.26 starts to rise slightly before the beginning of period III because its axial 1

location dries out slightly before steam production ceases throughout the core. However, unlike the base case, in the 4 l inch break there is an intermediate quenching of this clad location, which is driven by redistribution of fluid in the l

core induced by clearing of the loop seals (Figure 3.27).

The quench is also seen as a small spike in the pressure curve at the end of period II. For the 4 inch break the loop 3-12

seals are also clear at the beginning of period III, but when the pressure drop associated with the end of significant steam production comes about and the high head safety injection pumps begin to discharge, son.e of that ECC water flows back accumulating in the broken loop seal. This accumulation can be inferred from the steam generator tube's water levels, shown in Figure 3.27. This effect is significant for the 4 inch break because it reduces the rate of depressurization of the system, thereby reducing pumped .

l injection and delaying accumulator injection. Another ,

contributor to the reduction of the primary depressurization l rate is the bottled up secondary. In fact the secondary is i being cooled by the primary during period III, as evidenced by the pinch point in the pressure histories (Figure 3.19) around 750 seconds.

Period IV - Recoveryl As in the base case, period III ends when the system pressure reaches the accumulator injection pressure. At that time, as shown in Figure 3.28, the massive injection of wats.r reinstates steam production in the core causing the depressurization rate to level off again and marking the onset of period IV. Accumulator injection causes key water levels to rise (F.igures 3.23, 3.24 and 3.29) and arrests the primary inventory depletion (Figure 3.30). The clad ,

temperature begins to turn around at accumulator injection 3-13

---t - ,, - ,---r-n..m -....-,..,z e-c......,,--- - . -m---r.,-.,-,-,- . - . . . - - +

,_,r-----.---m--n.,1,,_._,--,,4 .

.--.-,.._,e. . -,..gw.wwe, y-

_ _ _ _ . . - - - _ _ _ _ _ _ _ _ _ ___.. _ . . _ m _ _ _ _ . _ _ . - . _ _ _ _ . . _ _ _ . _

time. Figure 3.26 shows the clad temperature histories one node above, one below and at the PCT location as calculated by ANF-RELAP. The rods are quenched from the bottom up with node 7 quenching first and node 9 last.

Conclusion The same conclusion drawn for the base case applies to the 4 inch calculation. The pumped injection flows (Figure 3.31) cannot keep up with the break flow (Figure 3.32) during periods I and II. Still, the accumulator injection pressure is reached well before the clad temperatures are too high and the temperatures are effectively turned around. In compliance with ANF methodology, the clad temperatures are recalculated using the more conservative TOODEE2 code and are shown in Figure 3.33.

8 inch Break This calculation is similar to the base case calculation as evidenced in Figures 3.34 through 3.48. The difference in results is primarily due to the faster depressurization associated with the 8 inch break, the earlier accumulator injection, higher ECC flow rates and lower PCT. The PCT is calculated to be 1750 'F in node 22 in this case, 11.625 ft above the bottom of the' core. The clad temperature history 3-14

_ . - _ _ _ . . _ . _ . . - . _ . _ _ _ _ _ . _ . _ - .____.______m_. _ _ _ _

i as calculated by the To0DEE2 code at the nodo where the PCT occurs is shown in Figure 3.49. .

l 3.2.2 AXIAL POWER SHAPE The axial power shape study is performed to support the Technical Specification linear heat generation rate (LHGR) limit as a function of core height. This study is performed I

for the most limiting break determined in the break spectrum '

study (6 inch, Section 3.2.1) and at the burnup yielding the highest stored energy. The maximum stored energy occurs at 678.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> when maximum fuel densification occurs, resulting in the maximum gap width.

The population of axial power shapes is developed through the power distribution control analysis described in Reference 3.3. For that analynis a prescribed series of load follow cases are modeled which provide the maximum variation in axial shapes achieved within the allowed operating conditions.

The selection of the axial power shapes to be examined is a two-step process. The first step is selecting the power 3-15

~ - _ _ - _ _ _ - - _

shapes which are closest to the Technical Specifications limit curve for each elevation. The second step is selecting power shapes which have the highest integral power up to the PCT elevation. The selected shapes are subsequently renormalized so that the peak LHGR matches the Technical

• Specification (Reference 3.4) limit at that location.

Analyses are performed for the chopped cosine (Shape 2, Figure 3.1) and two additional shapes, one of them most limiting according to the vendor's FSAR analyses (Shape 1, Figure 3.1) and one (Shape 3, Figure 3.1) developed using the method described above.

These are the most likely candidates to yield the highest PCT according to the criterion just described. All these power shapes are shown in Figure 3.1.

The sequence of events fur the axial power shape study is summarized in Table 3.6.

The conclusion to be drawn from the axial power shape study is that power shape 3, shown in Figure 3.1 is the most limiting. This result will be used in all other studies in the future.

POWER SHAPE 1 (6 inch. BOLL This is the base case calculation described in Section 3.1.

The PCT for this calculaticn is 1812 'F and occurs at node 3-16

3 21, 11.25 ft ubove the bottom of the core. The TOODEE2 calculation for the surface clad temperature at the node where the PCT occurs 10 shown in Figure 3.17.

CHOPPED COSINE (6 inch. BOL).

l l

The PCT for this calculation is 1578 'F and occurs at node 20, 10.875 ft above the bottom of the core. The system performance throughout the accident is nnarly identical to

' that described in Section 3.1 for the base case so that discussion need not be repeated here. The TOODEE2 calculation for the surface clad temperature at the node where the PCT occurs is shown in Figure 3.50.

POWER SHAPE 3 (6 inch, BOL)

The PCT for this calculation is 1837 'F and occurs at node 21, 11.25 ft above the bottom of the core. The system performance throughout the accident is nearly identical to that described in Section 3.1 for the base case so that discussion need not be repeated here. The TOODEE2 calculation for the surface clad temperature at the node i

where the PCT occurs is shown in Figure 3.51. This is the most limiting combination of break size and power shape.

l l

l 3-17 E _ _. _ _ _ _ _ ._ _._ _ - _ - _ _ - - - - - - . - - - - - - - - - _ - - - - - - - - - - -

TABLE 3.1

SUMMARY

OF CPSES-1 SMALL BREAK LOCA ACCIDENT ASSUMPTIONS FOR BASE CASE AND SENSITIVITY STUDIES

1. The initial power is 3636 MWt. This is 4.5% plus 2% for calorimetric error above the licensed power level.

2. 5% of the steam generator tubes are plugged.

3. Break in reactor coolant pump discharge occurs at 0.0 s.

4. Reactor trips due to a Lo-Pressurizer pressure signal.

5. Loss of offsite power coincides with reactor trip.

6. The reactor coolant pumps (RCP) are tripped at reactor trip since RCp cannot operate without offsite power after a reactor trip.

7- Steam flow isolation is initiated at the time of reactor trip. The steam dump ar.d bypass system is not credited.

8. Main feedwater isoletion is initiated-on an "S" signal.

9. Failure of one diesel generator to start takes out one high head centrifugal charging pump, one intermediate head safety injection pump, one RHR pump and one motor-driven AFW pump. This is the sing 30 failure assumed for compliance with 10 CFR 50, Appendix K, Part *..

10. One high head centrifugal charging pump, one intermediate head safety injection pump inject on demand after the appropriate delays, at conservative flow rates.

11. One turbine- and one motor-driven AFW pumps are credited, but their injections are conservatively delayed in order to account for flow travel time.

12. All accumulators inject on demand.

3-18

4 .

TABLE 3.2

SUMMARY

OF INITIAL CONDITIONS FOR CPSES-1 SMALL BREAK LOCA BASE CASE AND SENSITIVITY STUDIES DESCRIPTION VALUf Core Power 3636 MWt Power Upgrade Multiplier 1.045 Power Calorimetric lincertainty Multiplier 1.02 Power Shapes Analyted Chopped 2osine Shape 1 Fig. 3.1 Shape 3 Fig. 3.1 Peak Linear Power (includes 102% factor)

Chopped Cosine (Fig. 3.1) 13.16 KV/ft Shape 1, Top Peaked at 9.25 ft (Fig. 3.1) 12.59 KW/ft Shape 3, Top Peaked at 10.75 ft (Fig. 3.1) 12.36 KW/ft total Peaking f actor, FTg Chopped Costne (Fig. 3.1) 2.32 Shape 1, Top Peaked at 9.25 ft (Fig. 3.1) 2.22 Shape 3, Top Peaked at 10.75 ft (Fig. 3.1) 2.18 AccLanulator Water Voltrie per Tank 6119 gals Acctsnulator Cover Gas Pressure 648 psia Acctsmlator Water tenperature 90 *F Safety injection Ptsuped Flow Table 2.6 Refueling Water Storage *.ank feeperature 120 'F Initial Loop Flow 9680 ltra/see vessel inlet Tenperature w; "F Vessel Outlet Tenperature 627 *F Reactor Coolant Pressure 2280 psia Pressuriter Water Volume 1116 ft 3 Steam Pressure 894 psia Auxitlary feedwater Flow per SG 42.3 lb/sec Steam Generator Tube Plugging Level 3%

Steam Generator Safety Valves Set Points And Flows Table 2.9 -

_ Fuel Parameters Cycle 1, Table 3.3 3-19

. i TABLE 3.3

SUMMARY

OF FUEL PARAMETERS FOR BASE CASE SMALL BREAK LOCA ANALYSIS PARAMcmts  ! vAttrs Fuel Rod Ceonetry Data Table 2.8 T t.se to Maalsun Stored Erergy Exposure 678.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> fuei Rod ComposItlon:

Averaoe Core Hot Central Core Gram moles 0.02327 0.02327 Hellun fraction 0.99495 0.99493 Argon fractton 0.00000 0.00000 Hydrogen fraction 0.00000 0.00000 Nitrogen fraction 0.00500 0.00500 Krypton fraction 0.00001 0.00001 Xenon fraction 0.00004 0.00006 Effective cold 5.889 5.938 plerta tength (in) 8 Dish wlune (in ) 0.1419 0.1411 Average fuel tevnp.

i at peak stored energy (*F) 1668 2104 l

l i

L l

3-20

TABLE 3.4 SEQUENCE OF EVENTS FOR BASE CASE SMALL BREAK LOCA (6 inch, Power Shape 1, BOC; EVENT TIME (SECOND$)

1. Break opens (period I begins) 0.0

2. Reactor Trip Signal 3.2

3. RCP tripped 5.2

4. MSiv closed 5.7

5. "S" signal 8.6

6. MFW isolated 10.6

7. Centrifugal charging *unps inject 25.6

8. Mid elevation avg core bolts (period 11 begins) 26.0

9. Safety injection ptrps inject 30.6

10. Critical Heat Flux reached at PCT node 106.0

11. smet of Lower plenta depletion (period Ill begins) 163.0

12. Steam Generator tubes drained 165.0

13. Loop seats begin to fitt N/A

14. Auxiliary Feedt.ater injects 166.8

15. Accumulator injectic,(period IV begins) 329.9

16. Peak clad terperature reached 346.0

17. Peak clad tenperature node quenched (period IV ends) 480.0

18. Calculation ends 600.0 3-21

TABLE 3.5 SEQUENCE OF EVENTS FOR BREAK SPECTRUM STUDY (Power Shape 1, BOC)

TINF (SECONDS)

INENT 6 inch 4 inch 8 inch

1. Break opens (perlod I tegins) 0.0 0.0 0.0

2. Reactor Trip Signal 3.2 6.2 2.1

3. RCP tripped 5.2 8.2 4.1

4. MSIV closed 5.7 8.7 4.6

5. "S" Signal 8.6 13.6 4.9

6. NFV isolated 10.6 15.6 6.9

-7. Centrifugal charging ptaps inject 25.6 30.6 21.9

8. Mid-el avg core tells (period 11 begins) 26.0 47.0 7.B

9. Safety injection pumps inject 30.6 35.6 26.9

10. Critical Heat Flus reached at PCT mde 106.0 735.0 68.0

11. Lower plentan depletion tegins (period III) 163.0 441.0 80.0

12. Steam Generator tutes drained 165.0 400.0 80.0

13. Loop seals tegin to fill N/A 400.0 N/A

14. Aux. Feedwater injects 166.8 171.8 163.1

15. Acetmalator injects (perlod IV begins) 329.9 918.7 170.9

16. Peak clad temperature 346.0 932.0 186.0

17. Peak node quenched 480.0 1152.0 326.0

18. Calculatton ends 600.0 1400.0 400.0 3-22

TABLE 3.6 SEQUENCE OF EVENTS FOR POWER SHAPE STUDY (6 inch, BOC)

TIMF ($fCONDS)

EVENT POWER SHAPE I CH0PPE0 POWER (BASE CASE) COSINE SHAPE 3

1. Dreak ogens (period I tegins) 0.0 0.0 0.0

2. Reactor Trip Signal 3.2 3.2 3.1

3. RCP tripped 5.2 5.2 5.1

4. MSIV closed 5.7 5.7 5.6

5. "S" Signal 8.6 9.1 8.5

6. NFV isolated 10.6 11.1 10.5

7. Centrifugal charging ptmps inject 25.6 26.1 25.6

8. Mid elev. avg core bolt s (period II begins) 26.0 26.0 26.0

9. Safety injection ptmps inject 30.6 31.1 30.6

10. Critical Heat Flus reached at PCT node 106.0 106.0 106.0

11. Lower plerta oeptetion begins (perlod !!!) 163.0 163.0 163.0

12. Steam Generator tubes drained 165.0 165.0 165.0

13. Loop seats begin to f tll N/A N/A N/A

14. Ax. Feedwater injects 166.8 166.7 167.3

15. Acetnutator inject (period IV begins) 329.9 331.6 326.7

16. Peak clad temperature 346.0 356.0 346.0

17. Peak rude quenched -480.0 480.0 480.0

18. Calcutatlon ends 600.0 600.0 600.0 3-23

t l L i

i l

Figure 3.1 SBLOCA Axial Profiles CPSES Unit 1 Cycle 1 $

A ,

1.4 - / / V~ ,

i.3 -

j y og-- 3y i .2 -

p f ,' x x g. .

'- ~

/ / /  % \\ t 1 f4/ \ \\

\\\\

l f /;7~ \ -

~

! "- / /  %

!  ! / // \ \L t 3

j O.5

/ p  %

\

\4 '

O.4 - v  ;

i O.3 - -

0.2 0.1 -'

f 0.0 i a a i i

• 3 ' '

f

-0 0.2 0.4 . O.6 0.8 1 tiormolized C ,re Elevation Shope 1 i 0- Cosine + Shape 3 0

~

L

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1

,, 6 1NCH COLD LEG BREAK o9 7?

m C9 yw cio rd N 8

e 2-8 d9 s=

9 0

0.0 200.0 40'0.0 000.0 TIME ISEC)

Figure 3.2 Core Power c a 6 INCH COLD LEG BRERK G;e&

a.

-2' ya ,

a M;

'M 9 D-CP 174010000 y8 O--oP 360010000 CL $ I o .

II R h q g' N 9 #

H"3e IV es -

w w.

E9 E= . .

cr 0.0 200.0 400.0 600.0 TIME (SEC)

Figure 3.3 Primary and Secondary Systam Pressures 3-25

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 9

6 INCH COLD LEG BRERK C$

r e a3 o a

.o 5$- a e u-r MZtsilETE88 5-e8 it??tsiDa8E8 w~

D uo ibW

-o-ec o c o c o o c e o c o c a a o

=~= ==.=====. = .

a g :. . . . .

n x , ,.

0.0 200.0 400.0 600.0 TIME (SEC)

Figure 3.4 -intact and Broken Loop AFW and Steam Flows 6 INCH COLD LEG BRERK 9

a /. q .

de i j

.[,

g g- a x

Q. ., j

.2, c e r ( l' '

8 d- I z o-3 q g" k hlgf b ## -

U!!!E!!!$!$$

oq oe g ,

a--6vma nemm 0.0 200.0 400.0 600.0 TIME (SEC)

Figure 3.5 Central Core Region Void Fractions-3-26

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 6 INCH COLD LEG BRERK 9

s un a>

a fv ' vp j 'l

'j i

}

-w j (  ;

th o- Jh 4 r ,

E5 a c  % lh ,

M%j^4

~

E b h t/ p 3 t>-avmca 15ccicom Ee j iau8i88lM8188s 0:

0.0 20'0.0 100.0 600.0 ,

TIME (SEC) l Figure 3.6 Average Core Region Void fractions i 6 INCH COLD LEG BRERK U-s" d

>e dT ba 10 s 5 )

8 s. 4/(

5 \ t b9 gw 4

i E

89*

x 0.0 200.0 400.0 600.0 TIME ISEC)

Figure 3.7 RV Downcomer Collapsed Water Level 3-2'/

! COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 6 INCH COLD LEG BRERK

(*e -

-s

. . . 0o D e-

_] ] Y-

- am j_

M cn -

o-O a- . m F

z b

a. e - {

U S

8) ' 0.0 20'0.0 40'0.0 600.0 TIME (SEC) figure 3.8 Lower Plenum Water Level

.- 6 INCH COLD LEG BRERK oo 70

- r Oo 3-0 CNTRLVAR 43 gi- 0- CCNTRLVAR 42 .~

, m c.n 0 . -

- T J,7

• rn b C"!

B n-L ce W o-2o e C

w -lrOOOOOc 000000 OMOOOOO ui o O.0 200.0 400.0 600.0 TIME (SEC)

Figure 3.9 Intact and Broken Loop SG Inventories 3-28

. . , , . . , - - - - - - , - - ,-m,-_

4 .

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 6 INCH COLD LEG BRERK

'ad i

mR M dr

.~ buyA

a. g ez .a h13 5i dX J5' Ig z9

-c r,7

/-

S8 Y].

0 5

a $

f/[F t%

xd -Qi b ld3"IE !E!EU g .

emn.rr omwm c .  % .y.

s: 8 CC o i

l a_ 0.0 200.0 400.0 600.0 TIME (SEC)

Figure 3.10 Central Core Cladding Surf ace Temperatures 6 INCH COLD LEG BRERK l H9

'S ~

d g %

o

)

] d --

[

3

_3 1

D--O CNTRLVfiR 15 0-0 CNTRLVfiR 14 O

9 b JW Q dr 'N Z

b 00 g d w Pi. y; - ;_ .

W 0.0 200.0 400.0 600.0 TIME (SECl figure 3.11 Intact and Broken Loop SG Downhill Collapsed Level 3-29

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 a 6 INCH COLD LEG BRERK k?

E~ i J o. P

,o.

2 8 o

i d9o 1 I

i .k n

i,dil g S. ( ia 8a lM i

I I i IjI[

cn g l Y ,

d I wL d9

!58 s?. . .

0.0 200.0 400.0 000.0 TIME (SEC)

Figure 3.12 Accumulator Flow Rates 6 INCH COLD LEG BRERK so

'4 -

d d9 cn -

l 0,; .

fc pM&

8 t vi n ./f 8e N o er / N/ g d ,

I 59 5

u 0.0 200.0 400.0 600.0 TIME (SEC)

Figure 3.13 Colicpsed Level in Central Core Region i

3-30

l l

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1

. 6 INCH COLD LEG BREAK 00, I$

rC

$N' J9 x o-O* g o g W"nW -

C E.

c cT9 \

\

59 \

E 'y 9 N^

o '

0.0 200.0 400.0 600.0 TIME (SEC)

Figure 3.14 Primary System inventory a 6 INCH COLD LEG BRERK S

y E

an8-19 3E g ca z9 C 8~

ES a OO

!9 c g _.

Hm

'xM q k\h& & gypf i a

0.0 200.0 400.0 600.0 TIME (SEC)

Figure 3.15 Break Flow Rate 3-31

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 o

6 INCH COLD LEG BRERK R

m- y S

sm /

d-

'/

$ /

d /

_9 /

m 8-

_a T r Jo

.J E

09 O 0.0 20'0.0 40'0.0 600.0 TIME ISEC)

Figure 3.16 Pumped ECCS Injection Flow Role 6 INCH COLD LEG BREAK 9 , , , , i i d

e! -

S C s -

I ~

El -

x -

Y 1

D8-c.

5 w g -

k o $

5 e .o o ss. , u s. s its. ass.e ans. : ns.1 eis. $7s.s sse.s ses.s TIME - SECONDS Figure 3.17 T00DEE2 Surf ace Cladding Temperatures 3-32 l

-COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1

, 4 INCH COLD LEG BRERK 09 T?

u,

[9 xW 2

clo Mi o

CL Uh-8 i

.J

$o o

d-9 -  ;

O.0 300.0 600.0 900.0 1200.0 1500.0 TIME (SEC)

Figure 3.18 Coro Power c q 4 INCH COLD LEG BREAK

'v' 8 FG c (J o te d LJ #

CC e.

g-h 0-OP 174010000 1 MP E0010000 co -

II O$ - ? WO%k;s,

[ ,

III u) o L g g. 'f" Itg IV

>-. O D"**+Ng (E s *uoisig E" -

cr 0 .0 300.0 600.0 900.0 1200.0 1500.0 TIME ISEC) figu ro - 3.19 Primary and Secondarf System Pressures 3-33

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 4 ]NCH COLD LEG BRERK C5 r* -

.o b

_J "

L r NdrISO !S$S$fiO anruw; sewa:nco cc 9 m nrton; 57 m om co uo_

b U9  !

E{

m  !

}'

11 -

a w i

c0000000000e0eee0eo

• i !

} p p ,

.. :. c. s o c. a . . o a a a a a ert 0.0 300.0 60'O.0 900.0 12b0.0 1500.0 TIME (SEC)

Figure 3.20 Inicct and Broken Loop AFW and Steam Flows i INCH COLD LEG BRERK 9

" ~ "

C u .

~.

_J ,

~ co i j gi g d' i

e O U. _ .

• l g

> b

[ A %W o

i Wd w*:

J O' N y g O -O v010G 130010000 g O C+-Ov010G 138010000 a-6 VO100 196010000 d t ,=

0.0 360.0 600.0 900.0 12b0.0 1500.0 TIME (SEC)

Figure 3.21 Central Core Region Void fractions 3-34

~_- - - - . . - - . - _ - .. __- -.

l COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 i }NCH COLD LEG BRERK 9

m e p 9 o 1 # .,

d cn l bd- i r

y E /

09 0 j ym_

l W,'%

pbc pA s ,

8 .bg) Iyf u ' f

$af, 1 - .I"'

i

[ g p n *'k % TM8lE ylhthlE8!EE8 W h--A voicG w,nt onna b.0 350.0 60'0.0 9d0.0 12bO.0 1500.0 TIME ISEC) figure 3.22 Average Core Region Void Fractions i INCH COLD LEG BRERK C3

.m d

>9 d S-a d9 a c-5 do os ce s

d 9

o n- lr 5

89

$ 0.0 300.0 6do.O 9d0.0 12b0.0 1500.0 TIME (SEC)

Figure 3.23 RV Downcomer Collapsed Water Level 3-35

COMANCHE PEAK STEAM ELECTRIC STATION LINIT 1 4 ]NCH COLD LEG BRERK

[ ctg __

=

r d>

NK en

\ I l

j ct  %

$$ (

d*

3- .ru i

i] i

( \

CL l i b9 '

@*=0.0 J 300,0 600.0 E00.0 1200,0 1500,0 TIME (SEC)

Figure 3.24 Lower Plenum Water Level

, 4 INCH COLD LEG BREAK oo I$

r CD o d d- }-0 CNTRLV AR 43 i

(

• M CNTRLV6R 42 o -

E(~sw#

eo 3PI-u.

B z --

Eo a 5-O~. ss <mwS W'p>c 09 l .

0,0 300.0 6JO.0 900.0 1200.0 1500,0 i TIME ISEC1 Figure 3.25 Intact and Broken Loop SG Inventories 3-36

i .

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 4 ]NCH COLD LEG BRERK "0d cs ?

8-O D - f' SS '

\

U a C S S ill!88;!! A f VcNITL'1P 130100011 i*

co g-2-

5 ' "

92.

a8 ..

• a xa , i e

ne andeg g

}

,, p tw noonn c+

N '0. 0 30'0.0 60'0.0 900.0 12b0.0 1500.0 TIME (SEC)

Figure 3.26 Central Core Cladding Surf ace Temperatures 4 INCH COLO LEG BRERK H9

'S~

d

> :30 kJ o ,

C3 d .

g M CNTRLVAR 15 O---O CNTRLiPR 14 x

d o

O o, J R- a

• • • C 9

g

- crfp>ecevf' So cv A O.0 300.0 600.0 900.0 1200.0 1500.0 TIME (SEC) figure 3.27 Ir.tcet and Broken Loop SG Downhill Collapsed Level 3-37

. . . = _ .

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1

, 4 INCH COLD LEG BRERK ms U*

03 o J p_

g* i 09 d k- l ro

~

o l,

9_. I i. b I J O C C.

O!

H i . . . .

0.0 300.0 600.0 900.0 1200.0 1500.0 TIME (SEC)

Figure 3.28 Accumulator Flow Rates 4 INCH COLD LEG BRERK Un_

d

>9 ST i o l M9 DN \

& c~

J

\

\

YT j' pd%f u

5 i l 0 R_ l l d lE 9

' 5 o

" 0 .0 300.0 6d0.0 90'O.0 1200.0 1500.0 TIME ISEC)

Figure 3.29 Collapsed Level in Central Core Region 3-38

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1

, 4 ]NCH COLD LEG BRERK OO T$

r9 3 ?~

J9 g ?-

H U 3 ..

2: n

O

$<Y C \

Eo \ l 5

0.0 300.0 6d0.0 -960.0 12b0.0 1500.0 TIME (SEC)

Figure 3.30 Primary System Inventory a 4 INCH COLD LEG BRERK i

~

C9 s_as--

-;9 i

3 g_

u-E9 -

h $k

_a 9 55 @ ~'

E c

%d4ALA 0.0 3d0.0 6d0.0 9d0.0 12b0.0 1500.0 TIME (SEC)

Figure 3.31 Break Flow Rate 3-39

.. .- . . . . . - . . . . . _ -_. - ~ . - . -

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 4 INCH COLD LEG BRERK en -

b.

m J9 28~

8-J L.

9 en 8-1 CD

+

Jo

-d.

ib H

O H 0, a

'0. 0 360.0 60'0.0 9d0.0 12U0.0 1500.0' TIME (SEC)

Figure 3.32. Pumped ECCS Injection Flow Rote 4 INCH COLD LEG BREAK m  ! . . . . , , , ,

O yj - -

S-

.c g . .

w .

-l -

A 5: 8 - -

a. -

s- g

~

2E gg . .

5 o : , , , , ,

~.... o. . . - . . . . .... .s. . . . . . . , . . . . . , . . , . . . i n.. . , , . . . . o.. . .

TIME - SECONDS Figure 3.33 T00DEE2 Surf ace Cladding Temperature 3-40

i COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1

, 8 1NCH COLD LEG BRERK 00 T5 r9

$ ?-

J9 m o --

S~

55,

>r ..

=

9 QW c

E9 gr q g N

9 o

0.0 85.0 160.0 240.0 320.0 400.0 TIME (SEC1 Figure 3.46 Primary System Inventory 0 8 INCH COLD LEG BRERK 8

o C

E9 a g-

.o

$_J o x?.

5  %

59 \

d \

" l~

E 9 ( W VM % gvo w o

0.0 Ob.0 16'0.0 21'O.0 320.0 400.0 TIME (SEC)

Figure 3.47 Break Flow R0tc 3-41

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1

. 8 INCH COLD LEG BRERK oo T?

u, h9 cc w 2

oE9 Nw E

WT 8

d s R. o o

5 ^

0.0 85.0 16'0.0 210.0 32'0.0 400.0 TIME (SEC)

Figure 3.34 Core Power c q 8 INCH COLO LEG BRERK tn o Id o

$N H (A " '

J 0--OP 174010000 yq 0-oP 360010000 Q- 8 y$ II h{

]Sid p

III w

I Wo \ IV

.o j-

- ,"Se % m  :

E9 5 . . . .

oc 0.0 60.0 160.0 210.0 320.0 400.0 l TIME (SEC) l Figure 3.35 Primary and Secondary System Pressures 3-42

l COMANCHE PEAK STEAM ELECTRIC STATION l) NIT 1 l

, 8 INCH COLD LEG BRERK  ;

qg - _ _ _.

r~

ca 9 MoossG3GeosesesG4 dI S

'e r g-b net.$$IES Mq cltsji%s88E p s-dg m :. = uu a . = :. :. :. a = a .

Eg-e

t 9 go -- -- - - '.' _ - , _ _ _ ,

0.0 80.0 160.0 240.0 320.0 400.0 TIME (SEC)

Figure 3.36 Intact and Broken Loop AFW and Steam flows 8 INCH COLD LEG BRERK 9

w

~

f yq d

-m et M8!E !!ElEE h tr--AV010G 146010000

_b & i i)p )1 ,

5 VG I ( .

o y_ II f' $ ,

o d , .

M j_ o o 8 ( / "

l1 ae .

tphhygN 1 o&- , , , , ,

0.0 B0.0 160.0 -210.0 320.0 400.0 TIME (SEC)

Figure 3.37 Central Core Region Vold Fractions 3-43

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 8 INCH COLD LEG BRERK 9 --

w - =e - ,

p;g gg

.x ,

de [  !

s e- j .

E -

1 k f

h { l'f6f k [k U R.. / Ph t 8 u

I

/ I l . a b (

0 Bb.0 160.0 240.0 320.0 400.0 TIME (SEC)

Figure 3.38 Average Core Region Void Fractions 8 INCH COLD LEG BRERK U1p d

l h-  ! y Of f

<c a Yg l

}

(i 3+

8 4j!

230 8+

W ,

8aR b 0.0 Bb.0 16'0.0 240.0 320.0 400.0 TIME (SEC) figure 3.39 RV Downcomer Collapsed Water Level I

3-44 l

i 1

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 8 INCH COLD LEG BRERK

[9a

• ' Y 8,

u R g cn -

d o.

og.

5

-U m g s-59 8

_.] " 0.0 85.0 16'0.0 210.0 32'0.0 400.0 TIME tSEC1 Figure 3.40 Lowsr Plenum Water Level

,,. 8 INCH COLD L.EG BRERK Oo S

r 3--O CNTRLVAR 43 D#pai .

O O--O CNTRLv AR 42 ea&&

C c fa- O C.

u, "

Eno aW 3

es E

ao 5

a 5-d D! ..cs-n_n_m_.s_mmn,- - meccco m 0.0 05.0 16'0.0 240.0 320.0 100.0 TIME ISEC1 Figure 3.41 Intact and Broken Loop SG Inventories 3-45

m, , ,, . .

I l

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 q 8 INCH COLD LEG BRERK 8 ,6 '

o 4 BWE";li;ini:

//n c " sn, a*.= ~ m a n g g. p Q- "

H9 en

[/# ,M '

=- @- ,

- i.

SR  ! [

5k- .' .

o 1 e oi / -

h 00

[ 6Mwikggggg  ;

E 0.0 65.0 16'0.0 240.0 3do.0 400.0 TIME ISEC)

Figure 3.42 Central Core Region Cladding Surf ace Temperatures

-8 INCH COLD LEG BRERK so L. g -

._I W, '

S O 8&

m 0--O CNTRLVAR 15

) C+-O CNTRLyf9 14 o

C3 R

-jW -

E 5

o Y

to ,

...~.~....-....ua _- .-

. .. .m W 0.0 60.0 160.0 240.0 320.0 400.0 TIME (SEC)

Figure 3.43 Intoci and Broken Loop SG Downhill Collapsed Level 3-46

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 0 8 ltJCH COLD LEG BRERK

([*

1:

~

$9

.M- .f (j't '

ily) y g- a e, .

L a '

g& ,I <

h-de 1 0 .

! /

C f. t d

S9o- I /

R

!5 ?

HT-0<0 00.0 16'0.0 210.0 32'0.0 100.0 TIME (SEC)

Figure 3.44 Accumulator flow Rates 8 itJCH col.D LEG BRERK U A d

d or h$

n.

d~ 4'00Y l J s

LJ c[fI

$9 On se 5

o 0.0 00.0 160.0 210.0 320.0 400.0 TIME tSEC1 Figure 3.45 Collapsed Level in Contral Core Region 3-47

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 8 INCH COLD LEG BRERK P

( .

E /

• /

. g- /

s /

e o Ub' be

- g_

!5 reo 0.0 80.0 16'0.0 240.0 350.0 400.0 TIME lSEC)

Figure 3.48 Pumpad ECCS Injecilon Flow Rate 8 INCH COLD LEG BRE AK

,g , , , , , ,

O ei -

~

S ag. ~

i

$l -

E-a

~

i .

e E K-SI ~

5

" 1. . . i.. . .... n... ..... ...., J. . , J.. . o.. J.. . n. . .

Tiut - stconos figure 3.49 T00DEE2 Surf ace Cladding Temperatures 3-48

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 6 INCH COLD LEG BREAK L

WE -

S a 7 .

I ~ ,--

w ,./

N $ .

/ -

3- / x a!-

a. /

/ \

8 /

~s '\ /'

g wa 4w-g3 .

• 1 d /,. .

E ... i. . . ,;. . , , . . /,. . /.. , n. , .:. . J,. . . ....

TlW[ - SECONDS Figure 3.50 T00DEE2 Surf ace Cladding Temperature 6 INCH COLD LEG BREAK

[ , , , , , ,

O W!

o y - !k .

N, f ~

$! ;f,/

~

. 7 r i

$?

n_

[ ]  %

f k

o j' C o

E AJ (f\

~

O g

..  %~ :=~.

5 0

g.. .... ii... ii. .

,n.

n. . ,

T'uE - SECONDS figure 3.51 T00DEE2 Surf ace Cladding Temperatures 3-49

-.6

. sma-4 au.- m a = = um - r uu m.m. -

m -'.4%,_+a- ~a- - m a . +.4.  % m m..m a &&L A %ha m me  % A.h a h 2. A A_ ,a..a.

.1 O *

,l i

t.

S W

,t 1

t t

t 5

s k

i I

1 4

- f l

l t-(

,.43+- s-cy-o,,w--y. oy_ e y ---mvy--y, .r,m,-,w 3,n,- ,y.,m.e,. ~ , . - .m.,r-_-_y,_,% ....my.,-..,r--ry.-,,,, - - ,,m..c-..-.,..w,_,,,._,m-a , , - - . - we. .e .,, - _ , - , -em..m,.m-4-ye.,..,~e,-.

CHAPTER 4 CONCLUSION The USHRC-approved (Reference 4.1) ANF Corporation's ECCS Evaluation model entitled EXEK PWR Small Break Model has been applied to the Comancho Peak Steam Electric Station Unit one (CPSES-1).

Each calculation has been performed in exact compliance with the explicitly approved EXEM PWR Small Break methodology.

Regarding features of the calculation procedure which are

" implied" in the approval, there has been but one deviation:

the thermal-hydraulic calculations represent the average core region using nine axial noden (rather than the three shown in ANF's submittal). This deviation has been made in order to increase accuracy.

Five calculations have been presented with two objectives:

1. To demonstrate Texas Utilities' ability to properly apply EXEM PWR Small Break Model (Reference 1.1); and

2. To demonstrate the development of up-to-date input decks and conclusions which are in compliance with 10 CFR 50.46 and Appendix K thereto. Together, the 4-1

codes, input decks and conclusions will be applied to subsequent fuel cycles for the Comanche Peak Steam Electric Station Unit One and Unit Two. Evaluations will be performed to verify that the results of the present analyres remain bounding.

Table 4.1 sumraarizes the analyses and their key results. In each of the cases presented in this report, the calculated results show the following:

1. The calculated peak clad temperature is lower than the 2200 degrees F peak clad temperature limit set forth in 10 CFR 50.46 (b) (1) .

2. The total cladding oxidation at the peak location is under the 17% limit specified in 10 CPR 50.46 (b) (2) .

3. The hydrogen generated in the core by cladding oxidation is less than the 1% limit established by 10 CFR 50.46 (b) (3) .

4. No clad rupture is seen in these analyses. The average core region undergoes only minor dimensional changes, no clad ruptures are calculated to occur there. Thus, the coolable geometry criterion of 10 CPR 50.46 (b) (4 ) is satisfied.

4-2

5. Following accumulator injection, the rods are fully quenched, the pressure level is near the Ri!R pumps set point and the core is well cooled thereafter.

Therefore, the calculations comply with the Jong-term cooling criterion of 10 CFR 50.46 (b) (S) .

Regarding the sensitivity studies it has been found:

1. The most limiting break is a 6 inch break in the main coolant pump discharge line.

2. The most limiting power profile is Power Shape 3, which is shown in Figure 3.1 and which corresponds to the most limiting Small Break LOCA shape as determined by the methods of Reference 3.3.

Texas Utilities will use the EXEM PWR Small Break model including all codes, input decks, results, conclusions, and application procedures presented in this report to perform small break LOCA analyses and evaluations in compliance with 10 CFR 50.46 criteria and 10 CFR 50, Appendix K requirements, for both Comanche Peak Steam Electric Station Unit One and Unit Two.

4-3

TABLE 4.1

SUMMARY

OF RESULTS FOR BASE CASE AND SENSITIVITY STUDIES BREAK $12L Atl AL POWER $HAPE (IWCHt$)

$NAff 1 (4) CMOPF(D CO$fb[ $ NAP [ ]

6.0 1812 *F (1) 1578 *F 1837 'F 3.12 % (2) 0.593 % 3.45 %

0.289 % (3) 0.193 % 0.3?9 %

4.0 1585 *F 0.685 % " 108 3

0.M7 %

(1) PEAK CLAD TEMPERATURE 8.0 1750 'F (DEGRits F) 2.214 % (2) PERCENT LOCAL CLADDlhG 0.252 % 0x tD Af im (3) FERCENT CoeE WIDF 0xtDAi!ON (4) $EE FIG. 3.1 FOR $ Hart $

4-4

t .

CHAPTER 5 REFERENCES Chapter 1:

1.1 Advanced Nuclear Fuels Corporation, "USNRC's Safety Evaluation of Advanced Nuclear Fuels' Small Break LOCA Evaluation Model ANF-RELAP and Acceptance for Referencing of Topical Report XN-NF-82-49, Revision 1,"

July 1988.

1.2 " Axial Power Distribution Control Analysis and Overtemperature and Overpower Trip Setpoint Methodology," RXE-90-006-P, to be published.

Chapter 2:

2.1 Exxon Nuclear Company, " Evaluation Model EXEM PWR Small Break Model," XN-NF-82-49, Revision 1, June 1986.

2.2 V. H. Ransom, et. al., "RELAP5/ MOD 2 Code Manual, Volume 1: Code Structure, Systems Models, and Solution Methods," NUREG/CR-4312, EGG-2396, August 1985.

2.3 F. J. Moody, " Maximum Flow Rate of a Single Component Two-Phase Mixture," J. Heat Transfer, Trans. ASME, 87, pp 134-142, February, 1965.

2.4 G. G. Loomis, " Summary of The Semiscale Program (1965 -

1986)," NUREG/CR-4945, EGG-2509, July 1987.

2.5 Division of Technical Review, Nuclear Regulatory Commission, "TOODEE2: A TVo Dimensional Time Dependent Fuel Element Thermal Analysis Program," 1;UREG-75/057, May 1975.

2.6 Nuclear Regulatory Commission, Division of Technical Review, "WREM: Water Reactor Evaluation Model,"

NUREG-75/056 (Revision 1) May 1975.

2.7 Advanced Nuclear Fuels Corporation, "USNRC's Safety Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluation Model EXEM/PWR and Acceptance for Referencing of Related Topical Reports," July 1986.

5-1

2.8 B. D. Stitt, " Advanced Nuclear Fuels Corporation l

Procedure for PWR Safety Analysis Calculations, Section 10: Small Dreak LOCA Analysis," ANF-1238 (P), June, 1989.

2.9 TU Electric, " Steady State Reactor Physics Methodology," RXE-89-003-P, July 1989.

2.10 D. S. Huegel, C. M. Thompson, " Comanche Peak Unit 1 Accident Assumptions Checklists," WCAP-12368, August 1990, Revision 1.

2.11 Comanche Peak Steam Electric Station Unit One, Technical Specifications.

Chapter 3:

3.1 Comanche Peak Steam Electric Station Unit One, " Final Safety Analysis Report," Section 15.6, Amendment 78, January 15, 1990.

3.2 USNRC, " Water Reactor Evaluation Model (WREM): PWR Nodalization and Sensitivity Studies," - Technical Review U.S. Atomic Energy Commission, October 1974.

3.3 TU Electric, " Steady State Reactor Physics Methodology," RXE-89-003-P, July 1989.

3.4 Comanche Peak Steam Electric Station Unit One, Technical Specifications.

Chanter 4:

4.1 Advanced Nuclear Fuels Corporation, "USNRC's Safety Evaluation of Advanced Nuclear Fuels' Small Break LOCA Evaluation Model ANF-RELAP and Acceptance for Referencing of Topical Report XN-NF-82-49, Revision 1,"

July 1988.

5-2

i .

APPENDIX DESCRIPTION OF Tile COMPUTATIONAL TOOLS The EXEM PWR Small Break Model consists of three basic computer codes:

1. RODEX2

2. AN P-RE LAP

3. TOODEE2 The codes, their interfaces, interrelationships and respective inputs and outputs are summarized in Figure A.1 and Table A.1. The function of each code is described in the following sections.

A.1 EQDEK2 RODEX2 is used within the EXEM PWR Small Dreak Model f ramework to provide initial conditions for the ANP-RELAP and TOODEE2 calculations, as illustrated in Figure A.1 and Table A.I.

RODEX2 describes the thermal-mechanical performance of fue) during its operational lifetime preceding the LOCA. The determination of stored energy for the LOCA analysis requires a conservative fuel rod thermal-mechanical model that is A-1

r-, . .-

l capable of calculating fuel and cladding behavior, including the gap conductance between fuel and cladding as a function of burnup. The parameters affecting fuel performance, such as fission gas release, cladding dimensional changes, fuel densification, swelling, and thermal expansion are accounted for.

RODEX2 provides an integrated evaluation procedure for considering the effect of varying temporal and spatial power histories on the temperature distribution, inert fission gas release, and deformation distribution (mechanical stress-strain and density state) within the fuel rod. The surface conditions for the fuel rods are calculated with a thermal-hydraulic model of a rod in a flow channel. The gap conductance model includes the effects of fill gas conduction, gap size, amount of fuel cracking and the fuel-cladding contact pressure.

The calculational procedure of RODEX2 is a time incremental procedure so that the power history and path dependent processes can be modeled. The axial dependence of the power and burnup distributions are handled by dividing the fuel rod into a number of axial segments which are modeled as radially dependent regions whose axial deformations and gas releases are summed. Power distributions can be changed at any time and the coolant and cladding temperatures are readjusted at A-2

all axial nodes. Deformation of the fuel and cladding and gas release are calculated using shorter time ntops than those used to define the power generation. Gap conductance calculations are made for each of these incremental calculations based on gas released through the rods and the accumulated deformation at the mid point of each axial region within the fueled region of the rod. The deformation calculations include consideration of densification, swelling, instantaneous plastic flow, creep, cracking and thermal expansion for the fuel pellet, and also consideration of creep, irradiation induced growth, and thermal expansion for the cladding.

A.2 Alif_RELAE ANF-RELAP is a modified version of RELAPS/ MOD 2, INEL Cycle 36.02. The RELAP5/ MOD 2 code is described in detail in Reference 2.2. RELAP5/ MOD 2 has been modified in three major ways to produce ANF-RELAP:

(1) The Moody critical flow model (Reference 2.3) substitutes the RELAP5/ MOD 2 critical flow model during two-phase discharge in order to comply with the related requirement of 10 CFR 50, Appendix K, Section C.

A-3

(2) The RELAPS/ MOD 2 mixture level calculation is modified with the objective of producing a two-phase level, more suitable for the TOODEE2 fuel rod thermal analysis.

(3) A counter-current flow limitation (CCFL) constitutive equation based on a Kutateladze formulation with constants adjusted on the basis of the S-UT-08 test (Reference 2.4) is made available for use instead of the mechanistic interphase drag models in vertical junctions which can be selected by the user.

The ANF-REIAP model is dec 'ribed in Section 2.4.1. Initial thermal-hydraulic conditions are determined using LOOPT (Section A.4.1) followed by initialization calculations which include a null transient run. Initial fuel rod stored energy is determined using RODEX2 (Section A.1). The gap width is adjusted until ANF-RELAP fuel temperatures match.

The ANF-RELAP calculation provides the thermal-hydraulic boundary conditions for the TOODEE2 code, as shown in Table A.1 and Figure A.1.

A.3 TOODEE2 TOODEE2 is a two-dj.mensional, time-dependent fuel rod element thermal and mechanical analysis program. TOODEE2 models the A-4

fuel rod as radial and axial nodes with time dependent heat sources. Heat sources include both decay heat and heat generation via reaction of water with zircaloy. The energy equation is solved to determine the fuel rod thermal

,i response. The code considers conduction within solid regions of the fuel, radiation and conduction across gap regions, and convection and radiation to the coolant and surrounding rods,  ;

respectively. Radiation and convective heat transfer are assumed never to occur at the same time at any given axial node. Radiation is considered only until the convective heat transfer surpasses it. Based upon the calculated stress in the cladding (due to the differential pressure across the clad) and the cladding temperature, the code determines whether the clad has swelled and ruptured. Once fuel rod rupture is determined, the code calculates both inside and outside metal water heat generation.

The outputs of TOODEE2, viz., peak clad temperature, percent local cladding oxidation and percent pin-wide cladding oxidation are compared to the 10 CFR 50.46 (b) (1) through (3) criteria. Regarding (3), if pin-wide oxidation is less than 1% it is concluded that the criteria of less than 1%

core-wide oxidation is met.

A-5

A.4 DATA PREPARATION AND TRANSFER TOOLS Also used with the EXEM PWR Small Break model also are 3 additional codes for obtaining input information and/or transferring results between the basic codes described above

1. LOOPT

2. SHAPE /PWR (SHAPE. PUN)

3. TIFRO A.4.1 LOOPT This code is used to determine initial thermal-hydraulic conditions for ANF-RELAP. Flows, pressure drops and temperatures are used to initialize the ANF-RELAP steady-stato deck. These conditions are not necessarily the initial conditions for the accident because ANF-RELAP initialization includes steady-state as well as a null transient calculation prior to initiation of the LOCA calculation.

i A.4.2 SHAPE /PWR (SHAPE. PUN)

SHAPE automates the building of portions of input decks to ANF-RELAP, RODEX2, and TOODEE2. The code prepares input related to the axial power profile. The SHAPE code can alter A-6

and re-normalize a given axial power shape to a prescribed axial peaking factor. It then generates the power fraction input data for AliF-REIAP and the axial power f actors for input to the RODEX2 and TOODEE2 codes.

A.4.3 IIIRQ TIFRO (IOODEE2 Input from BEIAP Qutput) takes the information necessary to prepare a table of mixture lovels and steam flow rates for TOODEE2 input f rom the AliF-REIAP restart file.

A-7

TABLE A.1 IllPUT A11D OUTPUT roR Tilt EXEM/PWR METitoDOLOGY COMPUTER CODES (refer to FIGURE A.1)

Sil APE . Pull IllPUTI (15* 24 point axial power prof 11e (Itonctor Physico) renormalized to Toch Spec peaking factor OUTPUT (2) 101 point axial power profile with Toch Spec peaking foctor

• The numbers in thin table correspond to the numborn in Figure A.1.

A-8

i TABLE A.1 (Continued...)

INPUT AND OUTPUT FOR TIIC EXEM/PWR METilODOLOGY COMPUTER CODES (refer to FIGifRE A.1)

SilAPE INPUT:

(2) 101 point axial power profile with Tech Spec peaking factor (3) Tech Spec axial peaking factor at peak node Axial nodalization to be used in ANF-RELAP, TOODEE2, or RODEX2 Bundle geometry data ( ANF-RE LAP)

OUTPUT:

(10) Core power and weighting fractions calculated from axial peaking factors (4) 24 even node axial power profile (7) 24 uneven node axial power profile 24 uneven axial grid locations A-9

TABLE A.1 (Continued...)

INPUT AND OUTPUT FOR Ti!E EXEM/PWR METilODOLOGY COMPUTER CODES (refer to FIGURE A.1)

RODEX2 _

INPUT:

(4) 24 even node axial power profile (12) Description of fuel, e.g. geometry, density, enrichment, etc.

Cladding type and dimensions Initial mole fractions of fill gas spring dimensions 11ydraulic diameter, area, mass flux Axial nodalization Core power history for:

- peripheral core region

- hot contral region OUTPUT:

(9) Hot rod cold plenum length (at exposure of interest) used to calculate cold plenum volume llot rod gram-moles of gas (at exposure of interest) llot rod dish + crack volume (at exposure of interest) llot rod variables (at exposure of interest) to calculate cladding diameter and cold gap width llot rod mole fractions (at exposure of interest)

}{ot rod radially averaged density (at exposure of interest)

Fuel temperature profile (8) Average fuel temperatures for the hot central region and for the peripheral region. The gap width is adjusted in ANF-RELAP to match these stored energies.

A-10

1 TABLE A.1 (Continued...)

INPUT AND OUTPUT FOR Tile EXEM/PWR METilODOLOGY COMPUTER CODES (refer to PIGURE A.1)

LOOPT INPUT:

(5) Steady-state thermal-hydraulic paramotors.

i OUTPUT:

(6) Adjusted steady-stato thermal-hydraulic parameters (e.g. for tube plugging percentage, pressure drcpo)

(16) Coolant channel mass flux A-11

o .

l TABLE A.1 (Continued...)

INPUT AND OUTPUT FOR THE EXEM/PWR METHODOLOGY COMPUTER CODES (refer to FIGURE A.1)

AN F-RE LAP INPUT:

(10) Core power and weighting fractions (8) Average fuel temperatures for the hot central region and for the peripheral region. The gap width is adjusted in ANF-RELAP to match these stored energies.

(11) NSSS information (Table 2.2)

ECCS, SG AFW, safety valve flows (Tables 2.6 and 2.9) }

Trips and delays (Table 2.7)

Fuel rod / assembly information (Table 2.8)

Neutronics information (Tables 2.3, 2.4, 2.5)

OUTPUT:

(13) Saturation temperature entering the central core Vapor mass flow rate exiting the central core Normalized core power Position of the two-phase central core mixture level Average quality of the fluid below central core mixture level Mass flow of liquid entering the central core Temperature of the liquid entering the central core A-12

. s TABLE A.1 (Continued...)

INPUT AND OUTPUT FOR Tile EXEN/PWR METi!ODOIDGY COMPUTER CODES (refer to FIGURE A.1)

TIFRO INPUTt (13) ANF-RELAP restart file (See ANF-RELAP Output (13))

OUTPUT:

(14) Saturation temperature entering the central core Vapor mass flow rato exiting the central core Normalized core power Position of the two-phase central core mixture level Average quality of the fluid below central core mixture level Mass flow of liquid entering the central core Temperature of the liquid entering the central core A-13

TABLE A.1 (Continued...)

INPUT AND OUTPUT FOR Ti!E EXEM/PWR METilODOLOGY COMPUTER CODES (refer to TIGURE A.1)

TOODEE2 INPUT (9) Hot rod cold plenum length (at exposure of interest) used to calculate cold plenum volume Hot rod gram-moles of gas (at exposure of interest) liot rod dish + crack volume (at exposure of interest) llot rod variables (at exposure of interest) to calculate cladding diameter and cold gap width (used for geometric definition of hot rod and blockage data) llot rod mole fractions (at exposure of interest) liot rod radially averaged density (at exposure of interest)

Cladding + fuel surface roughness (7) 24 uneven axial nodo power profile 24 uneven axial grid locations (14) Saturation temperature entering the central core Vapor mass flow rate exiting the contral core Normalized core power Position of the two-phase central core mixture level Average quality of the fluid below central core mixture level Mass flow of liquid entering the central core Temperature of the liquid entering the central core OUTPUT:

(15) Peak cledding temperature and time Percent local cladding oxidation Percent pin wide cladding oxidation Rupture Node location and time A-14

1 V

! SHAPE. PUN l 2 3 3, s i 1,P y il

"'l

' i l SHAPE /FWRi lRODEX2 l LOOPT 10 8 11 6 V 89 Y r

q

=

ANF-RELAP 13 7 9 u

Ti?RO 14 V

J TOODEE2 15 i

Figure A-1 ANF SELOCA Computer Code Interfaces A-15

. _ _ _ _ _ _ . _ _ _ _ _ - _ _ _ _ -