Text

Large Break LOCA Analysis Methodology
	ML20066L069
Person / Time
Site:	Comanche Peak
Issue date:	12/31/1990
From:	Brozak D, Da Silva H, Tajbakhsh A; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	;
Shared Package
ML20066L067	List: ML20066L062; ML20066L069;
References
	RXE-90-007, RXE-90-7, NUDOCS 9102060251
	Download: ML20066L069 (119)
	v • d • e

._

1 l

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

)

I r

i i I

4

?

i k

t .....___ )i i 3

f i

i REACTOR ENGINEERING L 1

7 5

c r

f-

=

- - - a 1ot21 if,gg2cggg

P g3,300g gs .

TUELECTRIC

, _ . _ l

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

I i

i .

4 1

! RXE-90-007 i

! T 3

i .,

i LARGE BREAK LOSS OF COOLANT ACCIDEliT AllALYSIS l 1

I METHODOLOGY i i

l i DECEMBER, 1990 j ..

D. E. Brorak

! A. E. i:.jcakhsh i H. C. . dt. Silva, Jr.

i i

I I 4

i Reviewed: -./ N M Datet- N / '

' " V Whoo .-Choc .

Supervisor, LOCA Analysis ,

i

\

Approved: '% -

- b%21 Dato: /

Adsaf Husain Director, Reactor Engineering i

i i

t 5

1 I !

! .. l i

t

.A

..[

i

J

,-,_..-,-a,.

_ . . . . -- . , . . . . . . . . . . . - . . , , , , . . ,, _ e - . -_..~.........-,_.....__....,--a-.-.. .

DISCLAIMER 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 HEREUNDER, EXPRESSED OR IMPLIED, OR STATUTORY, OF ANY KIND OR NATURE WHATSOEVER, REGARDING THIS 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 atithorize 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.

11

~~ ~

ABSTRACT This report is presented to demonstrate the application of the USNRC-approved Advanced Nuclear Fuels (ANF) Corporation's large break (Emergency Cora Cooling Systems) ECCS Evaluation Model entitled EXEM/PWR, to the Comanche peak Steam Electric Station (CPSES).

This report contains a description of the EXEM/PWR 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 the LOCA-ECCS licensing analyses that comply with USNRC regulations contataed in 10 CFR 50.46 and Appendix c K thereto.

. In order to comply with a 10 CFR 50, Appendix X requirement, a full spectrum of large breaks, ranging from 0.6 to 1.0 discharge coefficients for Double-Ended Guillc. tine breaks

(DEG) and 1.0 for a longitudinal split breLk, .is chaminci.

Furthermore--in order to support the Technical Specificstion linear heat generation rate (LHGR) limit as a function of core height--all realistic potentially limiting axial power shapes g are considered, and analyses are presented for the chopped cosine and two top skewed axial power profiles.

i 111

?

o-D

_ , , . . . . . . . . . - - - - - - - ^ - - - ' - ~ ^ ^ - - - - ' - ~ ' - ' ' ~

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

a Finally-although higher poak clad temperatures / PCT) are

! usual'ly associated with beginning of cycle'(Boc) fun.because or tho higher sto ett o torgy-a fuel burnup study is ciso conducted. This it t'ano to confirm that the end of t,yclo (EOC) pin pressure,5-#ic.) are higher than'those encountered i earl.y in life and conse tuently foster a higher driving force

~

for rod burst-do indu k roruit in a lower PCT for the fuel under com " ion.

This mew.r.,uology-including all codos, input docks and conclusions reached wDhin this report-will be applied to subsequent fuel cy :len fee the comanche Peak Steam Electric Station Unit Ono and Unit *'o. Evaluations will be perfort,.od on 1.he basis of the cyclo-epecific parameters to verify that

, the results of the present analyses remain bounding.

4

.i l

} .

l p

iv

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

l l

TABLE OF CONTENTS PAGE DISCLAIMER . . . . . . . . . . . . . . . . . . . . . . . 11 ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . iii TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . V LIST OF TABLES . . . . . . . . . . . . . . . . . . . . Vii LIST OF FIGURES . . . . . . . . . . . . . . . . . . . Vili CHAPTER

1. INTRODUCTION , . . . . . . . . . . . . . . . . . . . . 1-1

2. DESCRIPTION OF THE METHOD . . . . . . . . . . . . . . 2-1

' 1 BACKGROUND . . . . . . . . . . . . . . . . . . . .

2-1 2.2 OVERVIEW OF THE METHOD . . . . . . . . . . . . . . 2-3 2.2.1 THERMAL-HYDRAULIC ANALYSIS . . . . . . . . . 2-3 2.2.1.1 BLOWDOWN , . . . . . .- . . . .. . . 2-3 2.2.1.2 END OF-BYPASS . . . . . . . . . . .- 2-4 2.2.1.3 REFILL . . . . . . . . . . . . . . . 2-4 2.2.1.4 EOTTOM OF-CORE RECOVERY (BOCREC) . . 2-5 2.2.1.5 REFLCOD . . . . . . . . . . . . . . 2-7 2.2.2 FUEL ROD THERMAL ANALYSIS . . . . . . . . . 2-9 2.2.2.1 BLOWDOWN . . .. . . .. . . . . . . 2-9 2.2.2.2 REFILL AND REFLOOD . . . . . . . . . 2-10

2.3 DESCRIPTION

OF THE MODELS . . . . . . . . . . . . 2-12 2.3.1 CPSES-1 RELAP4-EM SYSTEM BLOWDOWN MODEL , . 2-12 7

~

v

= - _ _ _ _ _ _ - - _ - - - _ - - _ _ - -

PAGE 2.3.1.1 VOLUMES, JUNCTIONS AND HEAT STRUCTURES . . . . . . . . 2 2.3.1.2 CORE POWER . . . . . . . . . . . . . 2-13 2.3.1.3 EMERGENCY CORE COOLING SYSTEMS-. . . 2-14 2.3.1.4 TRIPS AND DELAYS . . , . . . . . . . 2-15 2.3.2 RELAP4-EM HOT CHANNEL MODEL , , . . . . . . 2-15 2.3.3-ACCUM-SIS MODEL , . . . . . . . . . . . . . 2-17 2.3.4 RFPAC MODELS . . . . . . . . . . . . . . . . 2-17 2.3.4.1 CONTAINMENT . . . . . . . . . . . . 2-17 2.3.4.2 PREFILL . . . . . . . . . . . . . . 2-18 2.3.4.3 SHAPE /REFLOOD . . . . . . . .. . . 2-18 2.3.4.4 REFLEX , , . . . . . . . . . . . . . 2-19 2.3.5 TOODEE2 MODEL . . . . . . . . . . . . . . . 2-19 3.

BASE CASE ANALYSIS AND SENSITIVITY-STUDIES . . . . . . 3-1 3.1 BASE CASE ANALYSIS . . . . . . . . . . . . . . . . 3-2 4

3.2 SENSITIVITY STUDIES . . . . . . . . . . . . . . . 3-6 3.2.1 BREAK CPECTRUM . . . . . . . . . . . . .. . 3-6 3.2.2 AXIAL POWER SHAPE . . . . . . . . . . . . . 3-10 3.2.3 EXPOSURE . . . . . . . . . . . . . . . . . . 3-12

4. CONCLUSION m

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

5. REFERENCES

. . . . . . . . . . . . . . . . . . . . . . 5-1 APPENDIX: DESCRIPTION OF COMPUTATIONAL TOOLS . . . . A-1 vi

LIST OF TABLES TABLE PAGE 2.3.1 CPSES-1 NSSS Nodalization Summary . . . . . . . 2-21' 2.3.2 Summary of CPSES-1 RELAP4-EM System Model Volumes . . . . . . . .. . . . . 2-22 2.3.3 Density Reactivity Table . . . . . . . . . . . 2-26 2.3.4 Doppler Reactivity Table . . . . . . . . . . . 2-27 2.3.5 ECCS Flow vs. Pressure . . . , _. . . . . . . _ . '2-28 2.3.6 Time Delay for Each~ System . . . . . . . . . . 2-29 2.3.7 Fuel Assembly / Rod Data . . . . . . .. . . . . 2-30 3.1 Summary of CPSES-1 Large Break LOCA Accident Assumptions for Base Case and Sensitivity Studies . .. . . . . . - . 3 3.2 Summary Of Initial Conditions for CPSES-1 Large Break LOCA Base and Sensitivity Studies . . . . . . . . , . . . . . 3-15 3.3 Summary of Fuel Parameters for _ _

Base Case Large-Break LOCA Analysis . . . . .. 3 3.4 Sequence of Events for Base Case Large Break LOCA . . . . . . . . . . . . .. . 3-17_

3.5 Sequence of Events for Break-Spectrum Study (Chopped Cosine, BOC) . . . . . 3-18 3.6 Sequence of Events for Power Study (DEGB, CD=1.0,_BOC).. . . . . . . . . . . . .13-19 3.7 Sequence of Events for Burnup Study-(Top Skewed at 8.75',

DEGB, CD=1.0) .. . . . .. . . . . . . . . . . 3-20 4.1 Summary of Results for Base Case and Sensitivity Studies . . . .. . . . . . . . 4 A.1 Input and Output for the EXEM/PWR Methodology Computer Codes (Refer to Figure A.1) . . . . . A-15 vii

J LIST OF FIGURES FIGURE PAGE 2.2.1 Schematic Representation of the EXEM/

PWR ECCS Evaluation Model . . . . . . . . . . . 2-31 2.3.1 CPSES-1 RELAP4 System Blowdown Model . . . . . 2-32 2.3.2 RELAP4 Hot Channel Nodalization . . . . . . . . 2-33 2.3.3 ACCUM-SIS Nodalization .. . . . . . . . . . . 2-34 2.3.4 Reflex Nodalization for CPSES Unit 1 . . . . . 2-35 2.3.5 TOODEE2 Hot Rod Nodalization . . . . . . . . . 2-36 2.3.6 TOODEE2 Radial Nodalization . . . . . . . . . . 2-37 3.1 Axial Power Shapes . . . , _. . . . . . . . . . 3 3.2 Normalized Power . . . . . . . . . . . . . . . 3-22 3.3 Total Reactivity . . . . . . . . . . . . . . . 3-22 3.4 Downcomer Flow Rate . . . . . . . . . . . . . . 3-23 3.5 Average Core Inlet Flow Rate . . . . . . . . . 3-23 3.0 Average Core Mid Node Quality . . . . . . . . . 3-24 3.7 Core Inlet Subcooling . . . . . . . . . . . . . 3-24 3.8 Downcomer Liquid Mass Inventory . . . . . . . . 3-25 3.9 Total Break Flow . . . . . . . . . . . . . . . 3-25 3.10 RCS and Secondary Pressures . . . . . . . . . . 3-26 3.11 Containment Pressure . . . . . . . . . . . . . 3-26 3.12 Containment Spray System Flow Rate . . . . . . 3-27 3.13 Accumulator Flow Rate . . . . . . . . . . . . . 3-27 3.14 CCP and HHSI Pump Flow Rate . . . . . . . . . . 3-28 3.15 RHR Pump Flow Rate . . . . . . . . . . . . . . 3-28 viii

FIGURE PAGE 3.16 Mid Elevation Hot Assembly Heat Transfer Coefficient . . . . . . . . . . . . . 3-29 3.17 Hot Rod Temperature at PCT Node Elevation . . . 3-29 3.18 Core Flooding Rate . . . . . . . . . . . . . . 3-30 3.19 Mid Elevation Hot Assembly Zr/ Water Reaction Depth . . . . . . . . . . . . 3-30 3.20 PCT / Ruptured Node Cladding Temperature (Base Case) . . . . . . . . . . . . . . . . . . 3-31 3.21 PCT / Ruptured Node Clad Temperature (Cosine /CD=0.8) . . . . . . . . . . . . . . . . 3-31 3.22 PCT / Ruptured Node Clad Temperature (Cosine /CD=0.6) . . . . . . . . . . . . . . . . 3-32 3.23 PCT / Ruptured Node Clad Temperature (Cosine / Split /CD=1.0) . . . . . . . . . . . . . 3-32 3.24 Hot Rod Temperature at PCT Node Elevation (Skewed 0 8.75 ft.) . . . . . . . . . 3-33 3.25 PCT / Ruptured Node Clad Temperature (Skewed 0 8.75 ft.) . . . . . . . . . . . . . . 3-33 3.26 PCT / Ruptured Node Clad Temperature (Skewed 0 9.75 ft.) . . . . . . . . . . . . . . 3-34 3.27 PCT / Ruptured Node Clad Temperature (Exposure / Skewed 0 8.75 ft.) . ., . . . . . . 3-34 A.1 ANF EXEM/PWR Methodology for Large Break LOCA . . . . . . . . . . . . . . . . . . A-28 ix

CHAPTER 1 INTRODUCTION The present report describes the application of the USNRC-approved (Ref. 1.1) Advanced Nuclear Fuels (ANF, formerly Exxon Nuclear) Corporation's-large break ECCS Evaluation Model, entitled EXEM/PWR, to the Comanche Peak Steam Electric Station Unit One (CPSES-1).

The method is used to perform the LOCA-ECCS (Emergency Core Cooling Systems) licensing analyses that comply with USNRC regulations contained in 10 CFR 50.46 and Appendix K thereto.

The analyses presented in this report include a description of the EXEM/PWR methodology (Chapter 2), 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 exact compliance with the explicitly approved EXEM/PWR methodology. Regarding features of the calculation

. procedure which are " implied" in the approval, there is but one deviation: the thermal-hydraulic calculations represent the core region using five axial nodes (rather than the three shown in ANF's submittal). This deviation has been made in order to increase accuracy.

1-1

Three types of sensitivity studies are presented in Chapter 3.

The first is a break spectrum study. Large breaks ranging, from 0.6 to 1.0 discharge coefficients for Double-Ended Guillotine (DEG) hnd 1.0 for a longitudinal split break, are examined in order to comply with 10 CFR 50, Appendix K.

The second type of sensitivity study examines all realistic potentially limiting axial power shapes in order to support the LHGR limit as a function of height. This is done as follows: First the population of shapes is developed through the axial power distribution control analysis described in Reference 3.5. 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 the most likely to have the highest integrated power up to the PCT elevation are selected. Analyses are presented for the chopped cosine and two top skewed profiles. These are the most likely candidates to yield the highest PCT according to the criterior just descri. bed.

The third type of sensitivity (burnup study) consists of examining the EOC fuel condition for the most limiting break 1-2

and power shape as determined in the previous sensitivity studies (in which BOC fuel is used).

In Chapter 4, results from all these sensitivity studies.are used to establish the Design Basis Accident (i.e., most limiting LOCA case) for the EXEM/PWR 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 methodology, their interfaces, interrelation-ships, and respective inputs and outputs.

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

including all codes, input decks, inferences and conclusions--so that the above may be applied to the Comanche Peak Steam Electric Station Unit One and Unit Two for subsequent fuel cycle analyses and to address any applicable 10 CFR 50, Appendix K issues. Evaluations will be performed on the basis of specific parameters to insure that results of the present analyses remain bounding.

1-3

1 CHAPTER 2 DESCRIPTION OF THE METHOD 2.1 DACKGROUND In 1975, the NRC approved use of the Exxon Nuclear Company (ENC) WREM-based generic PWR ECCS Evaluation Model (Ref.

2.4). This LOCA Evaluation Model ic based on the NRC-developed Water Reactor Evaluation Model (WREM) (Ref.

2.8),

s In 1976, the ENC PWR model was updated resulting in the ENC WREM-II Evaluation Model (Ref. 2.9). The ENC-WREM-II model differs from the ENC-WREM model in four areas: (a) flow reduction due to blockage during reflood at rates less than 1 in/sec, (b) FLECHT multipliers for low reflood rates, (c) ice condenser containment pressure, and (d) hot wall delay.

In 1979, WREM-II was updated, leading to the WREM-IIA model (Ref. 2.1).- The WREM-IIA differs from the WREM-II only with respect to evaluation of the reflood portion of the LOCA transient. During this portion of the transient, the RELAP4-EM/ FLOOD (WREM-II) calculation is replaced by a similar calculation using REFLEX.

2-1

In July 1986, the NRC accepted the EXEM/PWR (Ref. 2.14) large break ECCS Evaluation Model for referencing of related licensing topical reports. EXEM/PWR is based on ENC WREM-IIA PWR ECCS EM (Ref. 2.1).

EXEM/PWR updates WREM-IIA in four phases of the transient '

calculation (a) stored energy and fission gas release models are revised in the fuel rod model in the RODEX2 code, (b) the NUREG 0630 clad rupture / blockage and a-new fuel rod model are added to the RELAP4-EM system blowdown calculation, (c) leakage flow from upper plenum to downcomer is allowed, also new split break and core outlet enthalpy models are used lk along with a revised carryout rate fraction correlation in the REFLEX code for the reflood period, and (d) the heatup model in TOODEE2 includes'a revised steam cooling model, NUREG-0630 clad rupture / blockage, a revised radiation heat transfer model, and a revised reflood heat transfer correlation.

The present report describes the application of Exxon Nuclear Company's large break ECCS Evaluation model, entitled EXEM/PWR, to the Comanche Peak Steam Electric Station Unit One.

2-2

l 2.2 OVERVIEW OF THE MET 4QD The EXEM/PWR methodology is illustrated schematically in Fig.

2.2.1. The accident is divided into three phases: blowdown, refill, and reflood. These phases are separated by two key events: End-of-Bypass (EOBY) and Bottom of Core Recovery (BOCREC). For presentation purposes, it is also appropriate to distinguish two types of calculations performed over these periods: Thermal-Hydraulic and Fuel Rod Thermal Analysis.

These are discussed in the sections that follow.

2.2.1 THERMAL-HYDRAULIC ANALYSIS 2.2.1.1 BLOWDOWN The analysis of the large break LOCA begins with the hydraulic analysis of the blowdown phase, noted as Step (1) on Fig. 2.2.1. RELAP4-EM computes the thermal-hydraulic conditions of the primary and secondary systems during the depressurization following the LOCA. The RELAP4-EM system model used for CPSES-1 is described in detail in Section 2.3.1. The RELAP4-EM system calculation determines the time dependent boundary conditions for the blowdown portion of the hot channel calculation. These are: (a) the core inlet and outlet plenum conditions and (b) the core power level.

RELAP4-EM system calculation also provides the End-of-Bypass 2-3

time (EOBY), mass and energy releases to the containment up to EOBY, and initial system conditions for the reflood analysis, i

\ 2.2.1.2 END-OF-BYPASS The time at which downward flow through the downcomer is sustained for at least one second, less the time for accumulator fluid to flow from the intact cold leg injection t

point to the downcomer, is the calculated time for End-by-Bypass. This time signals the end of the blowdown as well as the start of the refill period.

2.2.1.3 REFILL The rate at which the ECCS fluid is injected into the primary system intact recirculation lines during refill is determined by the ACCUM-SIS calculation (Stop (3) in Fig. 2.2.1). This calculation uses a RELAP4-EM model which is essentially identical to the ECCS portion of the RELAP4-EM system blowdown model. The ACCUM-SIS calculation determines the ECCS flow rates to the cold legs after the End-of-Bypass period (EOBY). The intact loop ECCS boundary conditions for the ACCUM-SIS calculation are taken from the RELAP4-EM system blowdown calculation up to EOBY and assumed to be constant and equal to the containment pressure at EOBY thereafter.

2-4

Therefore, this calculation repeats the system calculation out to EOBY.

The determinatjon of the containment backpressure for the refill period is done by ICECON/ CONTEMPT-LT (Ref. 2.5), which is included in the RFPAC code.

The power generated in the core during the refill and reflood portions of the transient is calculated using a one-volume RELAP4-EM model and the FISHEX code, as shown in Step (4) in Fig. 2.2.1. The RELAP4-EM code is used to calculate the delayed fission contribution to the normalized decay power.

Since RELAP4-EM computes total power, the fission contribution is obtained within FISHEX by subtracting fission product decay heat from the RELAP4-EM total power. Then the 20% multiplier is applied only to the fission product decay heat and not to the actinide decay heat, in compliance with the 10 CFR 50, Appendix K requirements.

2.2.1.4 Bottom of Core Recovery (BOCREC)

Following the EOBY as determined in the RELAP4-EM system blowdown calculation, downflow is calculated in the downcomer region of the reactor vessel. ECC water injected into the intact loops of the reactor will flow to the lower plenum under the influence of gravity forces. The time at which the 2-5

water level reaches the bottom of the active fuel is called the Bottom of Core Recovery (BOCREC) and signals the start of the reflood portion of the transient.

The time to begin reflood, the ECCS flow rates to be used in the reflood analysis, and the temperature at which the ECCS fluid enters the core at the start of reflood are calculated in PREFILL, which is also a part of the RFPAC code (Step (5) in Fig. 2.2.1). The initial and boundary conditions to the PREFILL code are obtained from RELAP4-EM system blowdown results, the intact loop ACCUM-SIS calculation and the ICECON/ CONTEMPT-LT calculation. The phenomena' addressed by PREFILL are: (a) hot wall delay period, (b) free-fall' delay time, (c) oxtended accumulator flows, (d) open channel flow spill, and (d) core inlet subcooling.

The start of reflood (BOCREC) is calculated by integrating in time the allowed flow rate of the ECCS water to the appropriate intact cold leg volume frac' n, to the lower plenum, and to the downcomer volume below the core inlet until they become liquid full. The time required for the ECCS water to fall from the bottom of the cold leg pipe to the core inlet (i.e., the free-fall delay time) is added to the time needed to fill the volumes listed above, yielding the actual BOCREC time.

2-6

When the ECCS fluid is injected into the downcomer, the fluid experiences a hot wall delay. Steam upflow created at the hot walls limits the downflow of ECCS fluid in the downcomer.

During the hot wall delay period, the level in the downcomer may rise above the bottom of the broken loop cold leg, and liquid can flow out the break. In this situation, the break flow is calculated by a hydraulic model which includes open channel flow. If the ECCS flow is higher than the maximum flow allowed by the hot wall phenomenon then the allowed flow into the system is adjusted to account for the spillage. The adjusted flow rates are the ones used in the previously described integration process which determines BOCREC.

2.2.1.5 REFLOOD This calculation considers the rate of reflooding of the reactor core (Step (S) in Fig. 2.2.1) and establishes core fluid conditions for the heatup calculations. The REFLEX code is used to perform the reflood analysis. In the ANF reflood calculations, the initial fuel rod temperatures for the average core are used. These are obtained from the RELAP4-EM hot channel calculation at EOBY (Step (2) in Fig. 2.2.1). The SHAPE /REFLOOn code calculates the fuel rod temperatures at BOCREC with the assumption of adiabatic heatup.

2-7

The REFLEX program calculates core reflood rates. This program is built upon a RELAP4 skeleton. The RELAP4 system equations are simplified-in REFLEX in the interest of computational speed as follows:

The core neutronics, transient heat conduction and critical flow tables are omitted.

Acceleration pressure losses are omitted in the flow equations. Mass accutuulation and gravitational losses are i also omitted in all systems components except in the core and downcomer nodes and in the cold leg piping to the break during the accumulator discharge phase.

The_ fluid state equations are based on analytical fits to property tables over a limited pressure range,10-100 psia.

This method is faster than the previous table look-up process.

The numerical scheme of RELAP4 is replaced for the flow calculation by the linear theory method (Ref. 2.10), using a Gauss-Jordan elimination method (Ref. 2.11).

The core outlet enthalpy is conservatively assumed to be determined by steam generator secondary temperature and 2-8

containment pressure in order to yield a conservatively high upper plenum pressure for reflood.

2.2.2 FUEL ROD THERMAL ANALYSIS The fuel rod thermal analysis encompasses the three time-periods outlined above, viz. blowdown, refill, and reflood, using two computational tools, viz. RELAP4-EM hot channel and TOODEE2.

2.2.2.1 Hlowdown i

T::e RELAP4-EM computer program is also used to perform the Hot Channel analysis which is identified as Step (2) in Figure 2.2.1. It is used: (a) to calculate the heatup transient during the blowdown phase, (b) to establish the temperature profile and extent of the metal-water reaction at the End-of-Bypass (EOBY) for the Fuel Rod Thermal Analysis described in Section 2.2.2, and (c) to provide average core, hot assembly, and hot rod cladding and fuel temperatures for the reflood calculation. Boundary conditions from the system blowdown calculation are used in performing these calculations. The RELAP4-EM Hot Channel model used for CPSES-1 is described in detail in Section 2.3.2.

2-9

2.2.2.2 Refill and Reflood The rod thermal analysis during the refill and reflood period is performed by TOODEE2 computer code. TOODEE2 uses the EOBY temperatures from the hot channel analysis (Step (6) in Fig.

2.2.1) and performs an adiabatic heatup, except for radiation, which continues until BOCREC.

The reflood rates, as calculated in REFLEX, provide the remaining boundary conditions to complete the hot rod temperature analysis from BOCREC through the reflood period ,

until core quench.

TOODEE2 is a two-dimensional,_ time-dependent fuel rod element 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 zircalloy. The energy equation is solved to determine the fuel rod thermal 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 2-10

the cladding (due to the differential pressure across the clad) and the cladding temperature, the code determines whether the clad has swelled and ruptured. -Whenever rupture is determined and the flooding rate drops below 1 in/sec, only steam cooling is allowed downstream of the ruptured node. This is in compliance with the related Appendix K requirement. Th3 effect of clad strain on_pellot-to-clad gap heat transfer and on the thinning of the oxide layer on the outside of the cladding is considered. Once fuel rod rupture is determined, the code calculates both inside and outside metal water heat generation. Fuel rod rupture reduces the subchannel flow area at the rupture and diverts flow from the hot rod subchannel to neighboring subchannels.

Flow recovery ib allowed above the rupture. The effect of flow diversion on heat transfer to the coolant is accounted for. The TOODEE2 code calculates heat transfer coefficients as a function of fluid condition or via reflood data-based correlations.

The outputs of TOODEE2, viz. peak clad temperature, percent local cladding oxidation and percent pin-wide cladding s

oxidation are compared to the 10 CFR 50.46 criteria (if pin-wide oxidation is less than 1% it is concluded that the crjteria of less than 1% core-wide oxidation is met).

2-11

2.3 DESCRIPTION

OF THE MODELS 2.3.1 CPSES-1 RELAP4-EM SYSTEM BLOWDOWN MODEL The Comanche Peak Steam Electric Station consists of two Westinghouse pressurized water reactors. Both units are four-loop plants with a rated' thermal power of 3411 MWt.

This section describes the RELAP4-EM system blowdown base input model for the Comanche Peak Steam Electric Station Unit one (CPSES-1). The components of this model are as follows:

1. Volumes, junctions, and heat structures

2. Core power

3. Emergency core cooling systems-

4. Trips and delays 2.3.1.1 VOLUMES, JUNCTIONS AND HEAT STRUCTURES Figure 2.3.1 shows the CPSES-1 nodalization diagram for the base input model which is comprised of 50 volumes, 74 junctions and 50 heat structures. Except for the number of axial core nodes, the model is identical to that approved by the NRC in connection with EXEM/PWR (Ref. 1.1). Table 2.3.1 identifies the particular volumes, junctions, and heat structures associated with the important regions or systems.

2-12

Table 2.3.2 summarizes the most important parameters of the CPSES-1 NSSS mode) volumes and junctions. These parameters were calculated using information from the most recent plant drawings, design basis documents, vendor documents, Technical Specifications and Final Safety Analysis Report.

2.3.1.2 CORE POWER The total core power during transients is determined by the point reactor kinetics model in RELAP4-EM. Conservative input data are entered for this model in order to compute the fission power and decay heat per 10 CFR 50, Appendix K. The model accounts for the reactivity effects associated with the 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 factor presented in References 2.18 to assure conservatism. For the analyses presented herein, reactivity feedbacks representative of the CPSES-1 core have been selected and are shown in Tables 2.3.3 and 2.3.4 for moderator density effects and fuel temperature effects, respectively. Scram reactivity is conservatively neglected in the model.

2-13

2.3.1.3 EMERGENCY CORE COOLING SYSTEMS The ECCS system is arranged into four subsystemst (1) the high head charging / safety injection, (2) intermediate head safety injection, (3) low head residual heat removal injection, and (4) accumulators (Fig. 2.3.3). 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, valvec, controls, and instrumentation. Only one train is represented in the present NSSS model. The other train is taken out by the single failure criterion in compliance with 10 CFR 50, Appendix K. 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 its minimum value_(40 degrees F) in order to minimize the containment back-pressure. The flow versus pressure values for each injection system, which are given in Table 2.3.4, reflect spillage of injection to the broken loop. The injection capacities were obtained from Ref.3.1.

The system contains four accumulators, one per loop. The minimum accumulator set pressure is used in all calculations in this report. A sensitivity study using the highest accumulator set pressure allowed by Technical Specifications 2-14

yielded insignificant differences in the fuel temperatures.

t Accumulator water temperature is assumed to be 90 degrees F for consistency with the initial containment temperature.

The minimum Technical Specifications (Ref. 3.4) tank water volume (6119 gals.) is also used.

2.3.1.4 TRIPS AND DELAYS The following trips and delays are used in the blowdown model:

1. Reactor coolant pumps trip at time of break.

2. Steam flow is isolated at time of break.

3. Main feedwater is isolated at time of break.

4. SI signal is generated at time of high containment pressure.

S. The delays following the SI signal for each of the pumped safety injection systems are given in Table 2.3.6.

6. Accumulators inject at the minimum accumulator set pressure.

2.3.2 RELAP4-EM HOT CHANNEL MODEL a

This model is used for the determination of the thermal response of the hot rod during blowdown. The hot channel nodalization diagram for the chopped cosine axial power shape 2-15

calculations is shown in Fig. 2.3.2. The nodalization of the hot rod heat structures may vary for other power shapes.

Figure 2.3.2 shows that five fluid' volumes are used to reprecent the average core, five fluid volumes for the hot channel and one fluid volume for each (inlet and outlet) plenum. Five heat structures are used to model the average.

core, five to model the hot assembly and twenty-four to model ,

the hot rod. Crossflow between the average core and the hot i

channel is represented as' required in 10 CFR 50, Appendix K.

The present nodalization differs from EXEM/PWR (Ref. 2.1) in the number of core axial nodes. EXEM/PWR utilizes only three, while the present calculation uses five volumes. This !

is done in order to increase accuracy.

The lower and-upper plenum volumes-in the hot channel calculation are time dependent volumes. Their pressures and properties are read from a file containing their values.

This file is generated in a previously performed system blowdown calculation. The power level is also read from the-system blowdown. All the initial conditions.for the hot channel calculation are set identical to those of the corresponding system ulowdown, t

2-16 1

l l

2.3.3 ACCUM-SIS MODEL t

The objective of the ACCUM-SIS calculation is to determine the ECCS flow rates to the lumped intact loop cold leg and to 1

the containment after EOBY.

The ACCUM-SIS calculation is essentially an application of RELAP4-EM. The nodalization diagram for this' calculation is -l given in Fig. 2.3.3. The input is identical to that of the ;

i system volumes. The cold legs are time dependent volumes

~

with pressures set by the previous blowdown calculation.

2.3.4 RFPAC MODELS As previously described, RFPAC combines the four codes used to perform the refill and reflood thermal-hydraulic analyses l (ICECON/ CONTEMPT-LT, PREFILL, SHAPE /REFLOOD, and REFLEX) and i

j eliminates the need for data transfer between codes. The input for each of these codes is described in detail in Ref.

2.15.

2.3.4.1 CONTAINMENT ICECON/ CONTEMPT-LT calculates the containment pressure 1

l response. The containment model is constructed so as to conservatively minimize containment pressure-for the reflood 2-17

calculations. The initial containment pressure is taken ab 14.7 psia, temperature at 90 F, and relative hamidity at 100%. The containment volume 1used is 3.063E6 ft3 . The spray system uses two spray pumps, sc as to maximize containment heat removal. This model includes the maximum flow rates, minimum water temperature, and rated heat removal capanity for the fan coolers, which also maximizes containment heat removal.

2.3.4.2 PREFILL s

The PREFILL code calculates (a) the time to beginning of reflood, (b) the ECCS injection flow rates for the re"ill I' analysis, and (c) the temperature at which ECCS fluid entern the core at the start of reflood. The transient specific input to this code is obtained from the RELAP4"EM-blowdown results, the ACCUM-SIS results and-ICECON results. The geometrical input involves a rearrangement of the information derived for the RELPAP4-EM system model.

2.3.4.3 SHAPE /RRELOOD

]

The SHAPE /REFLOOD calculation begins at BOCREC as determined by PREFILL. It uses the average core fuel and cladding temperatures from the RELAP4-EM hot channel calculation at E0BY to determine the average rod temperature at the peak i

2-18

y power location at BOCREC time for use in the Fuel cooling 1 Test Facility (FCTF) reflood correlatiors. The power shape. .

is transient specific; however, an evenly spaced 24 step i axial profile is used.

2.3.4.4 BEFLEX f 4

The nodalization diagram 'for REFLEX is shown in Fig. ;2.3.4.

The present model uses 26 volumes and 24 junctions to represent the primary system. The REFLEX model is obtained-by collapsing RELAP4-EM volumes as seen by comparison of

}

Figures 2.3.1 and 2.3.4. The intact and broken loop secondary sides are represented by 3 and 2 volumes, respectively. The core bypass flow area is inc'luded in the downcomer annulus area for downcomer liquid level calculations as prescribed in Ref. 2.1. The angle between-the cold leg and the ECCS line penetration is-45 degrees.

The ECCS mixing pressure drop penalties for this case are 0.6 psi during accumulator injection and 0.15 psi afterwards.

2.3.5 TQODEE2 MODEL TOODEE2 calculates the temperature distribution in the hot 4 I rod during refill and reflood. TOODEE2 calculations.begin at end-of-bypass-(EOBY). Only radiation heat transfer is allowed during the refill period. Only steam cooling is 2-19

}

i

_, , . y , , . _ . . .._-_ _...__. ,

l

.- pi i- j i'

/ y r U i

l allowed de.<0ctream.of th6 ruptursd node following:cladL L rupture for- reflood rates less -than one: inch per seconc',, j 1

Tablo 2.3.7 summarizes.the fuel'_ geometry dataeused in tho' TOODEE2 model. ,

'l 4

The.present-TOODEE2 Modell divides the fuel-rod into 24 exial d and 10 radial nodes.- .t :

?

. . . 1 The f1rst and last axial nodas'are identified!as the bottom-A

. . J :

and top of~the fuel rod,'respectively, 'The axial. ;

nodalization'of.the heat-structures-for the hot, rod in the~' -

q e

'IOODEE2 model is identical- to thtbof the hot rod' ini the -

RELAP4-Di Hot Channel modCl L(Fig s 2. 3. 2) .: The.TOODEE2. hot, a rod axial nodalization diagram for the chopped cosine-:' axial} >

]

power shapo:ca!cn.iations is shoMn vin Fig.;2.3.5. : The 4

~

nodalization may v'ary for other power; shapes'. ~Different axial. nodalizationsf aro discussed in. the sectic'as Edescribing

the calculations to which they ap;.aly.

, The ' fuel ~ pellet -is divided into 4 radial rings - (nodes) -in. l

~

r whlvh the 1ast radial line location / includes the gap. The- :

-r first inner funi pellet-is node:2, ani gridline-1..is. --

l identified as thie pollet henterline. ' The :last gridlir:e'is I idaantaftled .as the clad outer radius. .The^dladdingfis= divided j

inte.2, radial rings as require'dfby EXEM/PWR.. The :Yadiali .

~ nodalization scheme its shown' in- Fig. - 2 ; 3. 6.

c I 2-20 d i

I

l l .

. ~ -- --, . ~ . , , ....i . , , w -..--.....e.~.Jm,

I j TABLE 2.3.1 I

CPSES-1 NSE3 Nodalization Summary I

! component j Qescrintion Volume No.

Downcomer 27, 28 Lower Plenum 29, 50, 51 Average Core 30 to 34

!!ot Assembly 35 to 39 Core Bypass 40 Upper IIcad 1 Upper Core 5?

Upper Plenum 2 Guide Tubes 53 Containment 4i Intagt_Lagn Broken Loon ,

RCPs 12 24 Ilot Log 3 15 Intermediate Log 10, 11 22, 23 Cold Leg 13, 14 25, 26 S/G - Primary 4 to 9 16 to 21 S/G - Secondary 47 48 Accumulator 43 45 SI Discharge Line 44 46 Pressurizer 41 Surgo Line 42 Total = 53 Heat conductor Descrintion conductor No.

Avorage Core 5

llot Assembly 5 S/G per loop 4 Containment 5 RCS Piping 31

--Total = 50 l Fill Junction Junction lio, Descrintion Jntact Loon . Broken' Loon Centrifugal Charging Pumps 69. 70 Safety Injection Pumps 71 72 Low Pressure Injection Pumps 73 74.

Main Feedwater 65 66 Auxiliary Feodwater- 67 68 Steam Line Valve 64 63 Tota). = 12 l

2-21

TA9LE 2.3.2

SUMMARY

OF CPSES-1 RELAP4-EM SYSTEM MODEL VOLUMES l i

i' VOLUME ' 4 010N VOLUME VMUMt FLOW ARIA HYDRAULIC ELEV.

NUMBER DESCRIPfl0N (Ff3) LENGTH (FT2) DIAMtitR (FT)-

(if) (FT) 01 UPPER HEA0 892.2414 9.8460 90.6749 1.9476 30.9/5 02 UNDtt PLENUM 672.7352 7.9750 1.0+06 1.5991 23.0000 52 UPPit CORE 74.6550 1.2769 1.0+06 L.0704 21.7231 53 OGIDE TV6tl 220.3825 13.2900 16.0829 0 3372 23.0000 f 03 Hot Ltc 298.1295 3.6457 13.7607 2 5282 25.7083 04 $0 INLtt 538.6653 -7.9114 68.0871 5.3756- 27.6802-05 $0 TUBtt 422.7393 13.4737 31.3752 f .0553 35.5916 i 06 $0 TUBts 422.7393 14.5852 31.3752 L.0553 49.0653 l 07 SG TUBts 422.73Y3 14.5852 31.3752- 0.0553 49.0653 1 08 $G fusts 422.7393 13,4737 31.3752 0.0553 35.5916 7.9114 68.0871 5.3756 27.6802 j 09 50 OUfLif $38.6653 10 IN1tRM. Ltc 231.7245 5.7917 15.7242 2.5833 15.3125 l 11 luftRM. ttC 166.4935 4.0415 15.7242 2.5833 15.3125 i 12 PUMP 5.8000 7.3615 32.0316 3.6871 ii.1042 <

13 COLD Ltc 5.0970 2.2917 12.3741 2.2917 25.7709 14 COLD tr.G 5.0970 2.2917 12.3741 3.0124 25.7709 15 Hot Ltc 99.3765 3.6457 4.5869 2.5282 25.7083 l' 16 $0 INLt1 179.5551 7.9114 22.6957 5.3756 27.6802 17 $G TUBis 140.9131 13.4737 10.4584 0.0553 35.5916 i 18 $G TUBis 140.9131 14.5852 10.4584 0.0553 49.0653 '

19 SG TUBES 140.9131 14.5852 10.4584 0.0553 49.0653 20 SG TUDtt 140.9131 13.4737 - 10.4584 0.0553 35.5916 21 50 Outttf 179.5551 7.9114 22.6957 5.3756 27.6802 22 INitRM. LEG 77.2415 14.0415 5.2414 2.5833 15.3125 23 INTERM. LfC $5.4995 5.7917 5.2414 2.5833 15.3125 i

t l

l 2-22

TABLE 2.3.2 (Continued...) ,

SUMMARY

OF CPSES-1 RELAP4-EM SYSTEM MODEL VOL"MES VOLLMt Rt010N YOLUp! VOLUMt FLOW ARtt. HYORAULIC (LIV.

NUMBER OtSCRIPf!CW (713) LENGTH (F12) DIAMtitR (FT)

(if) (FT) 24 PUMP 78.6000 7.3615 10.6772 3.6871 21.1042 25 COLD Ltc 51.6990 2.2917 4.1247 2.2917 25.7709 26 r%D Ltc 51.6990- 2.2917 4.1247 3.0124 25.7709 27 OPPER OWNCOMER 392.6160 14.0000 1.0*06 1.4145 19.9167 28 ( NtR 00WNtnMER 479.1362 14.3333 35.6714 1.6363 5.5834 50 LOWER HEAD 120.2742 2.5126 47.8684 3.3216 0.4292-51 LOWER PLENUM 460.6664 3.5000 1.0+06 5.1105 2.0834 29 L CORT SUPi PLI 335.9651 4.1397 1.0+06 0.0691 5.5834 30 CORE 1 AVG 122.0097 2.4000 50.8738 0.0363 9.7231 31 CORE 2 AVO 122.0097 2.4000 50.8738 0.0363 12.1231 32 CORE 3 AVO 122.0097 2.4000 50.8738 0.0363 14.5231 33 CORT 4 AVG 172.0097 2.4000 50.8738 0.0363 16.9231 34 CORE 5 AVG 122.0097 2.4000 50.8N8 0.0363 19.3231 35 CORT 1 HOT 0.6350 2.4000 0.2646 0.0365 9.7231 36 CORE 2 Hot 0.6350 2.4000 0.2646 0.0365 12.1231 37 ChRE 3 HOT 0 6350 2.4000 0.2646 0.0365 14.5231 38 CORE 4 HOT 0.6350 2.4000 0.2646 0,0363 16,9231 39 CORE 5 HOT 0.6350 2.4000 0.2646 0.0365 19.3231 40 BYPAS$ 298.5298 13.3750 22.3200 0.7762 9.3750 41 PRES 5URIZER 1836.2393 30.5397 36.7823 6.8434 55.3308 42 PZR SURGE Likt 46.6806 27.8893 0.6827 0.9323 27.4415 ,.

43 ACCUMULATOR ll 4050.0000 10.8152 226.9008 9.8132 33.f775 t '

44 DISCH Likt IL 95.4600 7.8067 1.2528 0.7292 25.7709 45 ACCUMULA10R BL 1350.0000 10.8152 75.6336 9.8132 42.9908 46 OlsCH Likt DL 40,0400 17.2200 0.4176 0.7292 25.7709 47 SitAM GthtRA10R 17862.0000 41.8300 169.3512 0.1234 35.5916 48 $1 TAM GENERATOR 5954.0000 41.8300 56.4504 0.1234 35.5916 49 CONTAINMENT 3.063+06 299.00 10244.1500 114.21 +31.0000 1

I 2-23 l

- - . . - - m.

v. ,- -- e-- -- - e e -v-ve'"vrt- - v v y-e f r*--4

1 I

TABLE 2.3.2 (Continued...)

SUMMARY

Or e?SES-1 RELAP4-D4 SYSTEM MODEL JUNOTIONS JUNCil0N JUNCfl0N ELEV (FT) L/A ARIA FORWARD REVERSE HYDRAULIC NUMBER LOCAf l0N (fla1) (Fit) LOSS CotF LOSS CotF DIAMtitR 57 DWNCMR/VHEAD 33.9167 0.1808 0.6981 1.4946 1.4722 0.1667 01 UHEAD/ GUIDE 36.2900 0.4550 0.5199 6.84023 7.09124 0.4617 60 UPCORE/ GUIDE 23.0000 0.4748 11.9831 0.7321 0.6669 3.9061 4 61 UPCORE/UPLNm 23.0000 0.0s42 28.8708 1.7018 1.4852 6.0629

, 62 GUIDE /UPLNM 24.2391 0.5103 11.5647 1.34902 1.34902 3.8373 1

02 UPLtNUN/HL 26.9167 0.7834 13.7607 0.2424 0.4844 2.4167 03 HL/50 26.5238 0.8239 13.7607 0.3292 0.2272 2.4167 04 SG/TUBis 35.5916 0.2728 31.3752 1.8828 2.6029 3.6491 05 TUBES /TUBtB 49.0653 0.4294 31.3752 1.0 07 .1.0 07 3.M91 06 1UBES/TUBis 61.4666 -- 0.4294 31.3752 4.48907 4,48907 3.6491 07 TUBts/lVBis 49.0653 0.4294 31.3752 1.0 07 1.0 07 3.6491 08 TUBts/$0 35.5916 0.*728 31.3752 2.6029 1.8828 3.6491 09 50/li 20.5238 0.5267 t*,7242 0.4485 0.5419 2.5833 10 iL/lL 16.6042 0.8053 15.7242 1.0 07 1.0 07 2.5833 11 IL/RCP 21.1042 0.4905 15.7242 0.'591- 0.1591 2.5833 12 RCP/CL 26.9167 0.6602 12.3741 1.0 07 1.0 07 2.2917 13 CL/CL 26.9167 0.9239 12.31 1 1.0-07 1.0 07 2.2917 14 CL/DWNCMR 26.9167 0.4541 12.3741 1.29431 0.4645i 2.2917 15 UPLENUM/HL 26.9167 2.3502 4.5869 0.2424 0.4844 2.4167 1' 16 HL/$0 28.5238 2.4716 4.5869 0.3292 0.2272 2.4167 1? SC/fuBis 35.5916 0.8185 10.4584 1.8828 2.6029 3.6491 18 TUBES / TUBES 49.0653 1.2883 10.4584 1.0 07 1.0 07 3.6491 19 TUBES / TUBES 61.4666 1.2883 10.4584 4.48907 4.48907 3.6491 20 TUBES / TUBES 49.0653 1.2883 10.4584 1.0 07 1.0 07 3.6491 21 TUBts/$0 35.5916 0.8185 10.4584 2.6029 1.8828 3.6491 22 $0/lL 28.5238- 1.5801 5.2414 0.4485 0.5419 2.5833 23 IL/tt 16.6042 2.4159 5.2414 1.0 07 1.0-07 2.5833 24 !!/ACP 21.1M2 1.4714 5.2414 0.1591 0.1591 2.5833 25 RCP/CL 26.9167 1.9806 4.1247 1.0 07 1.0 07 2.2917 26 BREAK VALUE 26.9167 2.7716 4.1247 1.0 07 1.0 07 2.2917 27 CL/DWNCHR 26.9167 1.4523 4,1247 1.29431 0.46451 2.2917 28 U/L DWNCMR 19.9167 0.4010 35.6714 1.0 07 1.0 07 6.7393 29 OWNCMR/LPLN 5.5834 0.2702 26.6891 0.3552 0.0826 5.8294-58 LHEAD/LPLNM 2.0834 0.0402 82.0641- 0.0000 0.0000 10.2219 59 LPLNM/LCSP 5.5834 0.1327 49.9264 0.6628 0.6960 7.9730 I

l I

i i

I l

i-2-24

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

TABLE 2.3.2 (Continued...)

SUMMARY

OF CPSES-1 RELAP4-EM SYSTEM MODEL JUNCTIONS l

JUNCil0N JUNCil0N ELEV (FT) L/A ' AREA FORWARO REVERSE HYDRAULIO NUMBER LOCAtl0N (if*i) (FT2) LOSS COEF LOSS COEF DIAMETER 30 LC$P/1Av0 9.7231 0.0418 50.8738 4.5720 5.0613 8.0483 31 1/2AVO 12.1231 0.04 72 50.8738 1.4020 1.4020 8.0483 32 2/3AVO 14.5231 0.'0472 50.8738 1.4020 1.4020 8.0483 35 5AVC/UPCR 21.7231 0.0361 50.8738 1.4020 1.4020 8.0483 33 3/4AVO 16.9231 0.0472 $0.8738 1.4020 1.4020 8.0483 4/5AVO 19.3231 0.0472 50.8738 1.4020 1.4020 8.0483 ,

36 LCSP/1 HOT 9.7231 4.5533 0.2646 4.5720 5.0613 0.5804 37 1/2H01 12.1231 9.0703 0.2646 1.4020 1.4020 0.5804 38 2/3 HOT 14.5231 9.0703 0.2646' 1.4020 1.4020- 0.5804 39 3/4H01 16.9231 9.0703 0.2646 1.4020 1.4020 0.5804 40 4/5H01 19.3231 9.0703 0.2646 1.4020 1.4020 0.5804 41 5 HOT /UPCR 21.7231 4.5476 0.2646 1.00065 0.91189 0.5804 42 LCSP/Byps5 9.3750 0.2870 5.3294 43.3410 46.2161 2.6049 43 BYPSS/UPCR 22.7500 0.2613 3.7661 21.8091 22.1510 2.1898 44 CROS$FLW 1 10.9231 0.4167 1.6592 9.5220 9.5220 1.4535 45 CROSSFLW 2 13.3231 0.4167 1.6592 9.5220 9.5220 1.4535 46 CRossFLW 3 15.7231 0.4167 1.6592 9.5220 9.5220 1.4535 47 CROS$FLW 4 18.1231 0.4167 1.6592 9.5220 9.5220 1.4535 48 CeossFLW 5 20.5231 0.4167 1.6592 9.5220 9.5220 1.4535 49 PR2R/$URCE $5.3308 50.7563 0.6827 0.8675 1.3377 0.9323 50 $URCE/HL 27.7711 50.8541 0.6827 0.7017 3.2479 0.9323 51 Af/ATOL 33.5775 30,4552 1.2528 3.9754 3.9754 0.7292 52 Af0L/CL 25.7709 30.9223 1.2528 2.4044 2.4044 0.7292 53 AT/Aftt 42.990 114.9168 0.4176 4.0102 4.0102 0.7292 54 ATOL/CL 25.770 116.3182 0.4176 2.4044 2.4044 0.7292 55 CL/CNikMNT 16.9'67 1.5340 4.1247 1.00 0.50 2.2917 50 CL/CNihMNT 26.9167 1.2668 4.1247 0.50 1.00 2.2917 65 MFW FILL 40.5916 0.0000 3.0000 0.0000 0.0000 1.1284 '

66 MFW FILL 40.5916 0.0000 1.0000 0.0000 0.0000 1.1284 67 Aux FILL 73.5916 0.0000 3.0000 0.0000 0.0000 1.1284 68 AUX FILL 73.5916 0.0000 1.0000 0.0000 0.0000 1.1284 69 CCP/Filt 26.6873 0,0000 3.0000 0.0000 0.0000 1.1284 70 CCP/FitL 26.7913 0.0000 1.0000 0.0000 0.0000 1.1284 71 HHP / FILL 26.6873 0,0000 3.0000 0.0000 0.0000 1.1284 72 HHP / FILL 26.7913 0.0000 1.0000 0.0000 0.0000 1.1284 73 RHR/ FILL 26.6873 0.0000 3.0000 0.0000 0.0000 1.1284 74 RHR/ FILL 26.7913 0.0000 1.0000 0.0000 0.0000 1.1284 64 TSV FILL 95.7583 0.0000 3.0000 0.0000 0.0000 1.1284 63 ftv flLL 95.7583 0.0000 1.0000 0.0000 0.0000 1.1284 F

2-25

-4,- g , % -, <y=, . - . - . ~ , . . , - , ,, , , 4_ ,,,.-,_m._ .,.,,,,.n _ . , . _ , .m_.v

TABLE 2.3.3 DENSITY REACTIVITY TABLE i

NORMAL DENSITY REACTIVITY ($)

10.01 ~54.65 0.1422 -32.46 0.2845 -17.63 0.4267 -9.38 ,

0.5690 -4.56 0.7112 -1.82 0.8535 -0.47 3 1.0000 0.00 1.0669 0.15 1.1380 0.40 i 1.4225 0.6u .j l

l i

1 2-26 -.

l

~

i l

! TABLE 2.3.4 4

I i DOPPLER REACTIVITY TABLE I

1 i TEMPEA70RE (F) REACTIVITY ($)

200.0 1.691 400.0 1.283' 600.0- 0.919 a 800.0 0.589 1000.0 0.284 1200.0 +0.000 5

1400.0 -0.267 1600.0 -0.519 3

1800.0 -0.759-2000.0 -0.988 2200.0 -1.207

2400.0 -1.417'

, 2600.0 -1.620 .

2000,0 -1.816

3000.0 -2.006 :~

!' 3200.0 -2,189 3400.0 -2.367 3600.0 -2.541 l 3800.0 -2.709

4000.0 -2.874

)

1 e

4 G

I 2-27

-f i

l i.

. , - , , , , .-- n ,.,-,...A, ,, , , , . _ . ., .- , . . 1-- - -...!

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

2 i

f e

TABLE 2.3.5 ECCS FLOW VS. PRESSURE RCS CCP (1) llPSI (1) RHR (1) TOTAL ,

PRESSURE (ibim/sec) (1bm/sec) (1bm/sec) (1bm/sec) 1 (psia) 0.0 13.70 20.26 131.13 165.09 14.7 13.70 20,26 -131.13 165.09 i 34.7 13.5G 20.13 123.27 156.98-54.7 13.47 19.99 114.80 148.26 114.7 13.13 19.60 34.60 67.33 154.7 12.90 19.22 0.00- 32.12 214.7 12.55 18.66 31.21 414.7 11.37 16.79 28.16 614.7 10.13 14.53 24.66 1014.7 7.45 8.57 16.02 1614.7 2.77 0.00 2.77 2814.7 0.00 0.00 _

4 h

w 2-28 l

.. _ , . - .. . = , . . . ,.

-,__-..;. , _ _ . ~ , _ _ ..-,_-..w,-.,...,.__, - -

l TABLE 2.3.6 TIME DELAY FOR EACH SYSTEM ACTION TIMF DE W A M R SI S M CINT REACHED (sec, SI actuation signal 2 Charging pumps up to speed 17 (Fill Table 1 initiated)

HPSI pumps up to speed 22 (Fill Table 2 initiated)

RHR pumps up to speed 27 (Fill Table 3 initiated)

Containment Spray 34 i

l 2-29

= - - r y p _

y. .-i,ar.tv y-w-- v- t m;v

TABLE 2.3.7 FUEL ASSEMBLY / ROD DATA PARAMETER VALUE Outer Diameter of Fuel Rod 0.374 in Active Fuel lleight 144.0 in .

No. of Fuel Assemblies 193 No. of Fuel Rods /Assy 264 No. of Gulds Thimbles /Assy 24 No. of Instr. Tubos/Assy 1 Cladding Thickness 0.0225-in Diametral Gap 0.0065 in Outer Dia of Guide Thimble 0.482 in i

l

.3 l 2-30

, , , e-,, -ew-w

. .n.w., . - , , - - . . p

FIG. 2.2.1 BCHEMATIC REPRESDrrATION OF THE EXEM/ PIER ECCS EVALtrATION HODEL EVDIT A PIPE END OF BOTTOM OF BREAK BYPASS CORE RECOVERY Dlowdown Pe r i I_I Eeffood Regime @

Containoent] 5. ICECON/ CONTEMPT-LT (RFPAC)

CON BREAK FIDW

~

ECCS SPILL AND TAIN-AND ENERGY ENERGY VS. TIME MENT VS. TIME PRES-

3. REIAP4 - EM ACCUM - SIS SURE VS.

,o Primary & ECCS FIDW AND TIME 8 Secondary DIERGY VS. TIME N Systems

1. REIAP4 - EM 5. FRETILI.

SISTEM EOBY (RFPAC)

Average TIME ,

Core AND TIME OF BOREC SYSTEM IDWER FIDILM DITitAlrI

. Ilot CONDITIOf4S ,

Assembly 5. REFLEX -

(RFPAC)

INPUT TO - -

FCTF CORREIATION AVERAGE POWER CORE FUEL ROD VS. TIME BCs flot 5. S!! APE /REFIDOD (RFPAC) VS.

Rod 2. REIAl'4 - EM AND CIAD TIME TEMIS ItOT CflAlstIEL 4. FISifEX Ilot Assembly POWut vs. TIME If0T ROD 3 Average Cof4DITIONS Core l ADIABATIC l !!EATUP

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

e t

l FIGURE 2.3.1

.. - CPSES1 RELRP9 SYSTEM BLOWDOWN MODEL ..

t N

l s s%

1 1,,

. .. F.

p ::::::n f O ~::.

\ / -

\ _/

1

, ,,,g ..u .. u. .

g

[#4h /Mh

@ t

,c. > 0 0-

, p .i.m w " .

s -

sa t JL; q.

6 O , a eu.

@ j @ j" u g .

3

. O L, L

^, ,

t ih i@ @ ' I (U --9 3 g , y N O u--.

GOu

_ - i ,,

) , , .- , - -

yy . --a. ,

is .O .

,,9 -

-7 ,y, , M *,n ,(,',.. . .

-l G ~4* - '.i-*. i. . l.. t @ :

y g

(" ',1 ' 3n jyff l i py dug I '

'j' e.

, wz ,g; c

- . - 9 _::

~

<,, ../ 4, <.,

O O O kb r

1..

h' ,0 i.

$/? $ & .

s .-f;G e l3 t .I 0 s f% 9 - -

c a ~r-wn }'

4 ,"h g

'" g

--- ~~

gj , 3.cys y v g i, . -- > r., s z.

l

,0 /

ll ,0 . , ,

i j k, '" O '"

i4 --

i5 ve6 erit

"* Juut iIDe,

A o ... -

w_.-- s.& + na e

l 2-32 l-

'yge - yr p 9y -

p a er- U-'wy -'-T-'----fvm srer r inpr P H m -!u-N9-r 7

!, L !j E,f, il .f'il ;\ f.}!l1 i! , (l1! { ;

i!_ .

.=

0 1

_. 3_

, 6 4

_ 3

. m3

. 3

_ sr _

a o

C C Wt o c

o ptc* ,.

. 1 :

r C C wc*"

t _

o o _ .

v .

aa

-b .g

. o e a( M t

l I : - ; : : !

- s- .

O I

l gd$ -

h? , M [* '

? g , 2 - I

_ I s ?I I! ~}

pUE hl l .

3

I h

T [s FE , { _ - c A _.

Z z " _

, 7 shl s

I _

,8g

_ L _.

A _

D g :. : ,

,.l O Y .,! .

f , .

3 o . . = r - 6

~7s K ' 38 g F

L i

n rl '

> - 4 .

E ~

p"v l

i N

A O

R s i d

H

3 ,

e C

u s

5

@ Q _ h - *'

M w

e T k U

O u n _. s N

E S

H t

./ L 4

P A

L t

e e

t r

e v

s M ;

4

.eA > L 2

A 1

P R

E o

w T

)

p o

~

E ;

w y. .f* ,

b ye t B

R - '

2 3

2 E

f AM ; "b R

';1s;W ds A - ~-

E g 2 - l 8

4 8 i U -

,f 9.mP - Err y, G .

I F 1

' e , 7

- s

.i, , ,. -

Y .

e2 6

,y ' s N k a ' 7 V s 7 7'

z ye e n

b - F s i n- n8v ' i s

o I_

C 3 J s 5

Q-

@ @ M d ,

u U e.

o H p u '

f t p _

A t t

o 5 .

e e

t A h

N n .

a -

2

$ c C t s 4 -> w r o r w )

u -

yp -

t A

.- Wl 7 ,~, rf -

'h "

C -

, 14 - ,. n i+ , ~ i $ jl ' !~ ['I:l ,, , i i

FIGURE 2.3.3 ACCUM-SIS NODALIZATION O

l aCturiutA108 thus

\

l Nu -

T CCP Fgli l

l "

i

@- ml! Fall "Y $ Awp Fall t

INTACT 08t EROKEN LOCP COLD LE6 G

l 2-34 l-i I -

_. , - y .

FIGURE 2.3.4 REFLEX NODALIZATION FOR CPSES UNIT 1 (w/o mn worrits) _ ..

m_ <

cc

. . . . . ..n M4'M co-i, i.

e,1{-- ..s,n[.3r u. y.. @

. i se n ep ..s c p - -

a, pa_ _. .qtc..

m m. m ca em m m m m c.e- ,,

ii -

_/

m cm, (fl.

ru ,

< j 5" _ Q >

\

) ( "

}

.s h 86kW6d1 telt (ttel) 4m E 3 no.. ..

a n . ...n.. .... .

C:3 u....., n.. ....

. n . , .n...

- .m. n o .n ,...

2-35

- - - . . - - - . - . - . ~

l 4

2 FIGURE 2.3.5 T00DEE2 ItOT ROD llODALIZATIOli 12.o f t 12.0 FT 24 7.8 (N N00C 23 3.0 IN N00C 22 3.0 (H NCDC -

34 23.8lN N00C 21 3.0 (H N00C to 3.0 IN N00C 19 3.0 IN N00C ie 3.0 lH N00C

30IN N00E

9.6 rT 9.6 FT 16 3.0 IN N00C 15 3.0 IN N00C 14 3.0 IN N00C 23 - 22.s IN N00C t3 3.0 IN N00C

' t2 3.0 IN N00C i1 . 3.0 (H N00C 9o 3.0 IN N00C 9 3.0 lH N00C 22 3.0 IN NCDC ._

5 3,0IN N00C 7.2 iT 2t. 3,0 IN N00g 7.2 FT ----.

. i,3iN uo0g ,

20 2.4 174 N00C 3.0 IN N00C 19 3.0 (N N00C ' 3.0IN N00C

'S 3.01H N00C '

3.0 IN N00C 97 3.0 iN N00g t6 3.0 IN N00C 15 3.0 IN ::0DE

$4 3.0 IN Noct -

3 19.8 IN N00C

t3 3.0 IN N00C -

12 3,0 IN N00C 4.5 FT 19 2.4 IN N00C 4.3 (T -

to 3.0 IN N00C 9 3.0 lie N00C 8 3.0 IN N00C ,

7 3.0 IN N00C 6 3.0 IN N00C 2 25.8 IN N00C

$ 3.0 IN N00C 4 3.0 IN N00C 3 3.0 IN N00C 2 4.8 IN N00C 2,4 FT 2.4 TT l 25.5 IN N0DC t 28.8 IN N00C o.o ri o.o rT A) Chopped cosino B) Top Skewed 1

l l

2-36 l l

'l l

i i ,

l l

Fig. 2.3.6 1

1 TOODEE2 Radial Nodalization r

i Fuel Centerline l -> Radial Direction

, l-2 I l l l l l l8 - + - gap l l 10 l Node No.

11 Gridline No 1 2 3 4 5 6 7 9 i

l I

l 4 1

2-37 l

\

CHAPTER 3 BASE CASE ANALYSIS AND SENSITIVITY STUDIES 10 CFR 50, Appendix K requires the investigation of the -i impact of variations in several method- and plant-specific issues on the LOCA consequences.

The method-specific parameters requiring investigation are (a) nodalization and (b) time step. Such studies.are conducted for methodology development and approval. The l present work constitutes an application of an approved methodology using time step and nodalization as prescribed l therein. Hence, the effect of variations in-these parameters l

within the bounds of methodology recommendations has'already been ascertained to be negligi' ale, and sensitivity studies for these variables are not repeated here.

According to 10 CPR 50, Appendix K, the plant-specific _ issues-i which must be examined are (a) break spectrum (location, size, and type), (b) axial power' shape, and (c) fuel _ type and exposure. These are the sensitivity studies: examined in this chapter.

't j

3-1 y e _- -..-e,y--- # y _ -.,m._, +.,-r 4 ge e- +

3.1 DASE CASE ANALYSIS This section presents licensing analysis results for a Double-Ended Guillotine (DEG) break in the discharge line of the Reactor Coolant pump. The chopped cosine axial power shape used for this base case is shown in Fig. 3.1. The fuel rod exoosure which maximizes stored energy is calculated by RODEX and occurs at 613.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 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.

The major assumptions are that a DEG break occurs at 0.05 seconds with coincident loss of offsite power. The initial power level in taken to be 3636 MWt. This power level ;

includes both a 1.02 multiplier to account for calorimetric error and an increase of 4.5% above the licensed power level of 3411 MWt, representing a margin-potentially available.

ECCS injection into the broken loop is lost, and is postulated to spill directly to the containment. One pumped injection train is assumed lost due to failure of a diesel generator to start. This is the postulated single failure as requsred by 10 CFR 50, Appendix K. Thus,.one high head i

centrifugal charging pump, one intermediate head safety i.

3-2 l

E injection pump and one low pressure high flow residual heat removal pump along with three accumulators are available to mitigate the accident. Containment pressure ja minimized in accordance with Branch Technical Position CSB 6-1 (Ref. 3.2),

" Minimum Containment Pressure Model for PWR ECCS Performance i Evaluation." Minimization of containment pressure is done by minimizing initial pressure and-temperature and maximizing free volume and heat sinks. -Furthermore, containment safeguards are also assumed to function as designed while consistent with the single failure; i.e., only one train of j containment sprays is available. The other is taken out by l the diesel-generator failure postulated above. The fan coolers are disabled on the SI signal as per design. !

i-i Five percent of the steam generator tubec are assumed plugged for this analysis. This assumption is made to support the ,

potential need for operation under these circumstances and is a conservative assumption for fewer obstructed tubes.

1 Table 3.4 summarizes the timing of significant events for this case. This tablo should assist in the review of the l following figures, which present key results.

i Figures 3.2 and 3.3 show reactor power and net reactivity following the accident during the system blowdown phase. The !

reactor power decreases rapidly due to-negative reactivity ,

3-3

from core voiding. Between 4 and 5 seconds the power spikes 1

mildly because of an increase in reactivity, which in turn is caused by an increase in the liquid fraction in the center of ,

the core (Fig. 3. 6) . The increase in power results from a temporary coolant accumulation in that region, which is associated with a second flow reversal this time towards the normal flow direction, following the first reversal caused by the cold leg break location (Figs. 3.4 and 3.5). Beyond this time, core power follows the RELAP4-EM decay heat values

, (which represents 1.2 times 1971 ANS Decay Heat Standard).

l Figures 3.6 and 3.7 show mid-core average quality and coro inlet subcooling, respectively. Both figures indicate that core flashing takes place around 1 second. Again the quality falls between 4 and 10 secends due to the flow reversal discussed above and evidenced in Figs. 3.4 and 3.5. Shortly after accumulator injection (at 15 seconds, Fig. 3.13) the mid-core quality drops quickly. The quality increases back to 1.0 at 24 seconds.

Figure 3.8 shows the downcomer liquid inventory. The downcomer remains full until 4 seconds. As shown in Fig.3.4, ;i this time period corresponds to the period of flow reversal caused by the break. After that the downcomer is quickly ;

depleted until 3 seconds after the accumulators begin to ,

inject, when it once again begins to fill quickly.

3-4 i

i I

Figure 3.9 shows the total break flow. The flow rapidly accelerates to two-phase critical flow (Moody model) in lesu than 0.1 second at the pump discharge. Rapid depressurization and flashing limit the initial break flow rates. The break flow rate gradually diminishes as volumes upstream of the break become void.

i Fig. 3.10 and 3.11 show system and containment pressures respectively. Superimposed on the' primary pressure is the seconday pressure showing that the heat transfer direction is reversed at approximately 8.0 seconds. The containment pressure peaks to about 36 psia, 19 coconds into the blowdown. Tho pressure turns around at this time due to steam condensation on equipment and concrete surfaces.

Containment spray comes into play only at approximately 34 i seconds, injecting at a constant rate thereafter (Fig. 3.12).

ECCS flow rates are presented in Figs. 3.13 through'3.15.

The accumulators begin to inject at 15 seconds and are empty at 44 seconds. The available centrifugal charging _ pump begins to discharge at 18 seconds and the intermediate head.

safety injection pump at 23. The low pressure injection system comes on at approximately 28 seconds.

i 3-5

- - , - . , - - - - - - - . _.f,.. 4 y , er,.4- 4 %.

(

1 Figure 3.16 shows the heat transfer coefficient at the peak clad temperature (PCT) node. Heat transfer is abruptly degraded as the core flashes at approximately one second into the accident. Tho blowdown clad temperatures at the PCT node are presented in Fig 3.17.

The coro flooding rates are shown in Fig. 3.18. The flooding rate does not drop below one inch per second until 100 seconds. The PCT time is approximately 60 seconds.

The metal reaction depth at the hot spot is shown in Fig.

3.19.

The PCT node clad temperature-history is shown in Fig. 3.20.

The PCT is calculated to be 1959 'F in node 10 (Fig. 2.3.5),

4.7 ft above the bottom of the core. It is coincident with the ruptured node.

3.2 SENSITIVITY STUDigg 3.2.1 BREAK SPECTRUM l r

t l

l The most 1imiting break location has been determined in previous studies for this (Ref. 3.1) and other similar plants (Ref. 3.3) to be in the cold leg at the reactor coolant pump.

discharge. This determination results primarily from the 3-6

l l

loss of ECCS flow to th6 core associated with it. Therefore, this cold leg break location remains most limiting for the present evaluation and a worst break location search need not be repeated. This most limiting break location is the one considered in all cases discussed throughout this work.

According to the approved ANF EXEM/PWR methodology, the break f size is the first consitivity issue' addressed, holding l constant the axial power shape and the fuel exposure. The rationale for addressing break size first is that system j thermal-hydraulic behavior during the blowdown period is largely affected by break size but is nearly independent.of power shape and fuel exposure. Therefore the most limiting size for this shape and exposure will also be the most limiting size for other shapes and exposures.

l l The break spectrum study is conducted first for-the r

, guillotine type break with chopped cosine power shape and beginning of life (BOL) fuel. The reason for performing the break spectrum calculations with the other two parameters fixed to the values cutlined above is tha,t the large break LOCA analysis of record (Ref. 3.1) shows the most limiting break as a Double-Ended Guillotine type with chopped cosine axial power shape, and BOL fuel.

3-7

(-

i l Three break-sizes are examined by giving to the break dischargo coefficient the values of 1.40 (base case, Section l 3.1), 0.8 and 0.6, respectively.

Split type breaks are analyzed following the guillotine-type breaks. Those analyses are expected to yield lower peak clad temperatures and are done to confirm this expectation for i

CPSES-1. Thorofore, only the 1.0 discharge coefficient is oxamined for the longitudinal splits. It is noted that in I

EXEM/PWR the split break area is twice the ma.;imum pipe. area, i

as in the DEG.

The accident assumptions for this and other studies 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.

4 The sequenco of events for the break spectrum study is

( summarized in Table 3.5.

l The result of this study is that the most limiting break is a p Double-Ended Guillotine with a 1.0 discharge coefficient located in the main coolant pump discharge. Future studies

} will be performed using 1.0 as the limiting discharge coefficient.

3-8 l

s;

3.2.1.1 DIq.cD=1.0 This is the base case calculation described in Section 3.1.

The PCT is calculated to be 1959 'F in node 3 0 at 4.7 f t above the bottom of the core.

l 3.2.1.2 QLG CD=0.0 i

The results of this calculation are quite similar to those of the base case (DEG CD=1.0, Section 3.1), during the various i stages of the thermal-hydraulic analysis. However, during the fuel rod thermal analysis, the PCT node and the ruptured node do not coincide for this calculation, as shown in Fig.

3.21. The PCT is calculated to be 1870 'F in node 21 (7.3 ft above the bottom of the core per Fig. 2.3.5) and the ruptured ;

node is node 16 (at 6.1 ft).

3.2.1.3 DEG CD;,RzG This calculation is nearly identica: to the one discussed above (DEG CD=0.8). The PCT node and the ruptured node do not coincido for this calculation either, as shown in Fig. {

3.22. The PCT is calculated to be 1768 'F in node 21 in this ;

case (7.3 ft above the bottom of the core per Fig. 2.3~.5) but i the ruptured node is node 10 (at 4.7 ft).

3-9

,. iay =

i 3.2.1.4 SPLIT CD=L Q n

The longitudinal split break calculation shows results similhr to both the 0.8 and the 0.6 DEG. The PCT node and the ruptured node do not coincide for this calculation either, as shown in Fig. 3.23. The PCT is calculated to be 1 1901 'F in node 21 in this -case (7.3 ft above the bottom of the core per Fig. 2.3.5) and the ruptured node in node 10 (at 4.7 ft).

3.2.2 AXIAL POWER SHAPE 4

The axial power shape study is performed to support the I technical specification linear heat generation rate (LHGR) )

limit as a function of core height. This study is performed ,

for the most limiting break determined in the break spectrum !

study (DEG CD=1.0, Section 3.2.1)-and at the burnup yielding '

the highest stored energy. The maximum stored energy-occurs I

at 613.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 j 3.5. For that analysis a prescribed series of load follow cases are modelled which provide the maximum. variation in l

3-10

, , . .________.__m,.

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

h I

axial shapes achieved within the allowed operating conditiolle.

The selection of the axial power shapes to be examined is a two-step process. The first s'op is selecting the power shapes which are closest to the Technical Specification 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 Specifications ,

a (Ref. 3.4) limit at that location. These power shapes are shown in Figure 3.1.

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

The conclusion to be drawn from the axial power shape study is that the most limiting power shape-is the profile which peaks at 8.75 ft, show.n in Figure 3.1. This result will be used in all other studies in the future.

+

3.2.2.1 LOP PEAKED AT 8.75 FT AND CD=1.0 As expected, there is minimal difference between blowdown results for this calculation and those for the chopped cosine and CD=1.0. Even clad temperatures for this period are 3-11 ,

,, _ ,,..-m. ,

nearly identical, as can be seen by comparison of Figs. 3.24 and 3.17. The ruptured node and-the FCT node coincide in this calculation ,i well (Fig. 3.25) and correspond to node 13 (B.7 fw above thr bottom of the core per Fig. 2.3.5). The calculated PCT is 2034 F.

3.2.2.2 TOP PEAKED AT 9.75 AMD CD=1.0 The PCT for this calculation is 1983 'l' and occurs at node 22 i

(11.0 ft above the bottom of the core per Fig. 2.3.5). The ,

rupture occurs at node 17 (9.7 ft above the bottom of the 9

core per Fig. 2.3.5). Figure 3.26 shows the clad temperaturec for this case.

3.2.3 EXPOSURE s

The exposure study is done to support operation to EOC burnup levels. It is done because pin pressure increases with o, ,isure, and higher pin precsures increase the driving force for rod burst, with the attendant effect of raising peak clad temperatures. It should be noted, however, that the stored energy effect tends to dominate the pin pressure-effect so that a lower peak clad temperature is expected at EOC.

Nevertheless, this study is done to confirm that that is indeed the case for the fuel under consideration.

3-12

The sequence of events for the burnup study is' summarized in Table 3.7.

i The clad temperatures are shown in Fig. 3.27.

l The conclusion from the burnup' study l1s that'all burnups are bounded by the beginning of cycle (613.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months exposure) 1 condition for the present fuel, since the two extremes (maximum stored energy and maximum pin pressure) have been examined. This exposure (613.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) will be_used in future 1 1

studies unless fuel changes warranting a re-evaluation of this assumption occur.

l l

3-13

TABLE 3.1

SUMMARY

OF CPSEd-1 LARGE BREAK LOCA ACCIDENT ,

ASSUMPTIONS FOR BASE CASE AND SENSITIVITY STUDIES !

1 i

' 1. The initial power level is 104.5% above the 1.02 (calorimetric error factor) x 3411 MWt (i.e., 3636 MWt).

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

3. Reactor coolant pump discharge line break occurs at 0.05 i seconds.

4. Loss of offsite power occurs coincident with-break at 0.05 seconds.

5. Failure of one diesel generator to start. takes out one I high head centrifugal charging pump, one intermediate head safety injection pump and one low head high flow residual heat removal pump. This is the single failure assumption as required by 10 CFR 50, Appendix K.

6. No credit is given for reactor scram.

7. Three accumulators inject into intact loops on demand. [

8. One high head centrifugal charging' pump, one intermediate head safety injection pump and one low head high flow residual heat removal pump inject on demand after the appropriate delays.

9. Containment pressure is minimized in accordance with branch Technical Position CSB 6-1 (Ref. 2.15), " Minimum Containment-Pressure Model for PWR ECCS Performance Evaluation." Minimization of containment pressure is done by minimizing initial pressure and' temperature and maximizing free volume and heat sinks. Furthermore, containment safeguards are also assumed to function as designed while consistent with the single. failure; i.e.,

only one train of containment sprays is_available. The other is taken out by the diesel-generator failure postulated in assumption 5 above. .The fan coolers are disabled on the SI signal as per design.

10. Passive heat structures are included.

11. No credit is given for Auxiliary Feedwater.

3-14 I

l

TABLE 3.2 1

SUMMARY

OF INITIAL CONDITIONS FOR CPSES-1 LARGE BREAK LOCA BASE CASE AND SENSITIVITY STUDIES DfSCRIPfl0N VALUE Core Power 3636 MWt Power Upgrade MultipLler 1.045 Power Colorimetric Uncertainty Multiplier 1.02 Reactor Coolant Punp Heat 20 MWt (approx)

Power shapes Analysed -Chopped Cosine- I

. Top skewed D 8.75 f t Top skewed Q 9.75 ft Peak Linear Power (fratudes 102% factor)' .. .

Base case (Fig. 3.1) 13.16 KV/ft-Top Peaked at 8.75 ft (Fig. 3.1) 12.71 KW/ft.

Top Peaked at 9.75 ft (Fig. 3.1) 12.54 KW/ft i

Total Peaking factor, F g Base Case (Fig. 3.1) 2.32' Top Peaked at 8.75 ft (Fig. 3.1) 2.24 Top Peaked at 9.75 ft (Fig. 3.1) 2.21.

Accumulator Water Volune 6119 gals /Accum Accumulator lover Gas Pressure 623 psig Accumulator Water Temperature 90 'F Safety injection Puuped Flow Table 2.3.4 contairvnent Parameters Table 3.1; Item 9 Refueling Water Storage Tank Temperature 40 'F Initial Loop Flow 9743 lbm/sec vessel inlet Temperature 539.9 'F-Vessel outlet Ternperature 622.8 'F Reactor Coolant Pressure 2250 psia Steam Pressure 940 psia Steam Generator Tube Plugging Level 5%

Fuel Parameters Cycle 1,. Table 3.3

.t 3-15

. .. . . ~. .- . -

I TABLE 3.3 ~

.i

SUMMARY

OF FUEL PARAMETERS FOR BOC and EOC LARGE BREAK LOCA ANALYSIS l PARAMETERS Fuel Rod Geometry Data Table 2.3.7 Elginninn of Cvele (BOC)-

Time to Maximum -i Stored Energy Exposure 613.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Fuel Rod Composition:

Averace Core Hot Assembly Not hxi Gram Moles 0.02326 0.02327 0.02327 :

Helium fraction 0.96796 0.96793 0.96793 Argon fraction 0.00000 0.00000 0.00000 Hydrogen fraction 0.00000 0.00000 0.00000 Nitrogen fraction 0.03200 0.03200 0.03200 .

Krypton fraction 0.00001 0.00001 0.00001 xenon fraction -0.00004 0.00006 0.00006 Effective cold 5.923 5.979 5.981 plenun length (in)

Dish volume (in3 ) 0.1419 0.1416 0.1414 End of Cvele (EOC)

T{me to Maxinun Fuel Din Pressure 7542.0 hours0 days 0 hours 0 weeks 0 months Fuel Rod Composition:

Average Core Hot Assembly Hot Rod Grom Holes 0.02329 0.02487 0.02520' Hetium fraction 0.96703 0.9050. 0.89349 Argon fraction 0.00000 0.00000 0.00000 Hydrogen froctien 0.00000 0.00000 0.00000 Nttrogen fraction 0.03197 0.02994 0.02954 Krypton fraction 0.00013 0.00852 0.01018 Xenon fraction 0.00087 0.05590 0.06680 Effective cold 6.253 6.104 6.095 plenum length (in)

Dish voluw (in3 ) 0.1397 0.0837 0.0819 3-16

'4 TABLE 3.4 SEQUENCE OF EVENTS FOR BASE CASE LARGE BREAK LOCA EvtWT TIME (SECONDS)

1. Break opena 0.05

2. Loss of offsite power 0.05

3. Main feedwater isolated 0.05

4. MSIVs close 0.05

5. High contalrvnent pressure Hi 1 signal 1.07 6.Accumulatorinjection,intactLoop 14.90

7. Centrifugal charging pumps inject 18.07

8. End of aBypass 22.66

9. Safety injection pm ps inject 23.07

0. Time of sustained downfall 24.88

11. Low pressure pumps inject 28.07

12. Bottom of Core Recovery 37.91

13. Rod burst 40.01 3

14. Acetsnulator empty 43.85.

15. Peak clad temperature reached 60.06 S

3-17

TABLE 3.5 SEQUENCE OF EVENTS FOR BREAK SPECTRUM STUDY (CdOPPED COSINE, BOC) flME (SECONOS) _

EVENT DOUBLE

ENDED CUILLOTINE SPLIT CDs1.0 CD=0.8 C0=0.6 CD=1.0 .

1. Break opens 0.05 0.05 0.05 0.05

2. Loss of offaite power 0.05 -0.05 0.05 =0.05

3. Mein feedwater isolated 0.05 0.05 0.05 ' 0.05 -

4. Msive close 0.05 0.05- 0.05 0.05

5. High containment pressure 1.07 1.14 1.30- 1.12 HI 1 algnal

6. Accunulator injection 14.90 15.10 16.70 15.k0

7. Centrifugal charging punps inject 18.07 18.14 18.30 10.12

8. End of* Bypass 22.66 23.13 25.29 22.91

9. Safety injection punpa inject .23.07. 23.14 23.30. 23.12

10. Time of sustained downfall .24.88 25.34 '27.50 25.15

11. Low pressure purps inject -28.07 28,14 28.30_ 28.12

.l

12. Bottom of Coro Recovery 37.91 38.43: 40.63 38.19

13. Rod burct 40.01 47.10 52,19 .41.ro

14. Accumutator empty 43.85 43.95 =45.65 _44.15.

15. Peak clad temperature reached 60.06 .72.43' 73.29 73.01 3-18 l

j i

TABLE 3.6 SEQUENCE OF EVENTS FOR POWER SHAPE STUDY (DEG,-CD=1.O, BOC)

TIME (SECONo$)

Emi CHCPPED COSINE TOP SKEWED

~

-TOP $KEVED (BASE CASE) AT 8.75' AT 9.75'

1. Break opens 0.05 0.05 0.05

2. Loss of of fsite power 0.05 0.05 0.05

3. M, sin feedwater isolated 0.05 0.05: 0.05

. 4. Mslys close '0.05 0.05 0.05 t

5. High contairvnent pressure algnal 1.07 1.07 '1.07

6. Acetsnulator injection, intact loop 14.90 14.90 14.90

7. Centrifugal charging pumps inject 18.07 18.07 ;18.07

8. Fnd of Bypass 22.66 22.64. 22.64

9. Safety injection pumps inject 23.07: 23.07 23.07'

10. Time of sustained downfall 24.88- 24.86 24.86

11. Lcw pressure ptanps inject -28.07 28.07. .28.07

12. Bottom of Core Recovery 37.91 - 37.93 37.91

13. Rod burst 40.01 44.69 44.99

14. Accunutator empty - 43.85 43.70 43.80

15. Peak clad temperature reached 60.06 72.84 - 218.44 3-19

TABLE 3a7 SEQUENClE OF EVENTS FOR BURNUP STUDY (TOP SKEWED AT 8.75', DEG, CD=1.0). ,.

I i

EVENT TIME (SEc0NDS)

BOC EOL I

1. Break opens 0.05 10.05

2. Loss of offelte power 0.05 0.05

3. Main feedwater isolated 0.05 0.05 4 MSlvs close 0.05 0.05

5. High containment prestore HI 1 signal 1.07 1.07 ;-

6. Accunulator injectico, intact loop 14.90 14.90

7. Centrifugal charging punps inject 18.07 18.07

8. End of Bypass- 22.64 22.64

9. Safety injection punps inject 23.07 23.07

10. Time of sustained downfall 24.86 24.86 i

11. Low pressure pumps inject 28.07 28.07 s

12. Bottom of Core Recovery 37.93 37.93

13. Rod burst 44.69 51.14- '

14. Accumulator empty 43.70 43.70

15. Peak clad temperature reached 72.84 72.24 i

l i

l l

I l

l l

l 3-20 i 1

l

FIGURE 3.1 AXlAL POWER SHAPES 1.5 A,j.

f4

/ - ,!k j \

f .3 s .2

? / /\ \ \- .

o

/ / / \ y\

I .0

/ \ '\. \

o 09

-+' \ \

?;

0.8 Y{,W \

5 i 0.7 ?

/ {-

4 O.6

///'.

0.5

\\

" H/ \\

"' I DI \\

gi E2 > , , , , ,

0 0.2 0.4 0.6 0g ,

N 12 O 6 00 R PEAK g75 R P 0 -9.75 R PCAg l

I 3-21 I

l 1

Comanche Peak Steam E!ectric Station Unit 1 l

, 1 , ,

6 e i E i $ . .

, a >

i 2

ii .. !

l $

o y

i 9-

. -, . ,, J .. ,, ,. u vi.. (s.. ..:

Figure 3.2 Normallied Power (Dase Case)

Comanche Peak : Steam Electric Station Unit 1

)

?

E n '

a s a

.- ?

l t

w 8

9. ..' .' i, i. A 4 J n

n. . i ..... . >

Figure 3.3 Total Reactivity (Dase Case)

--a---

3-22

Ccmanche Peak Steam Electric Station Unit 1 l , , . , , .

=

s

=

l s

a l '

e c 1

~

t l J i j a . -

x I. .' . . n d 4 ,, ,; n n.. . m.i Figure DA Downcomer Flow Rate (Dase Case) l.

Comanche Peak Steam Electric Station : Unit 1

l 4

s s

a sI -

a i .

m 1

c m i t 3 .

y l I

t- ! g l l 0 - r l }

l-

,t l.

. .' . n i.

n.. < ..... . i

,o ,,

I 1 Figure 3.5 Average Cole inlet flow Rate (D&1e Case)

I l' 3-23

Comanche Peak Steam Electric.. Station Unit 1

' ~

3 !

5 e -

5 a

I e, - -

0 j r

n - .

I 1 i q a i e4 4- 4 17 66 49 it it - 33 time p... 4.)

rigure 3.6 Average Core Mid Node Quality (Dane Case)-

Comanche Peak Steam Electric Station Unit 1 0 a

$ 8 a

3 H

c E

2 I ~

i g .

~

2 3

E- A i

/

$ 8 .

E 2

I. . .

i. i. a

,, vi.. n. 43 A

Figure 3.7 Core Inlet Subcooling (Base Case) 3-24

- _ - _ - . - _ _ _ = _ _ _

.. , - . . .. . . - . _ .. . =-. .- . .

Comanche Peak Steam Electric Station Unit 1 i i p s y G

a O

l .

5

, -l E

g '

t E

8 a -

D E

' ^

b e

.e i e n is a n n u tim. (s... 4.i Figure 3.5 Downcomer Liquid blass Inventory (Dase-Case)

Comanche ' Peak Steam Electric Station Unit 1 l , , ,

f .

i !

O l

n 2

c

l. - .

i 5 l - -

1 5 l - -

1 i 1 1 3

.o e e is is a n a u tim. (s... 4.)

l Figure 3.9 Total Dreak Flow (Dase Case) l l 3-25

l-o l

Cornanche Peak Stearn Electric Station Unit 1

[ , , , .

7 1l '

E -

{ -

B t

b ! '

N- . . . . .

4 >r E

I (g

n 43 I

~ 1 g .i ts

.. . . ,, i. n ,, ,, u time (s... 4.)

Figure 0,10 ItCS & Secondary Pressures (Dese Case) g l

Cornanche Peak Steam Electric Station- Unit 1 A

7. '

a, a t .

E U

5 a O

$ ~

a 6

R E

.B n B

E

-u x. . .

tim. (s... 4.) .

Figure atti Coritainment Pressure (Dane Case) 3-26

l l

i Comanche Peak ' Steam Electric Station- Unitf1 r i-

. I .,

a d

- i.

4 1 _

c -

7 j g I

5 e i

.9 j

o i -

n.. ...u .i .

1 Figure 3.12 Containtnent Spray System Flow Rate (Base Case) i Comanche Peak Steam Electric- Station - Unit 1 i-l-

7 h . .

N -

E a

~

g .

3 0

m o

Io -

c 6.

3

,1 I~

n E

8 I ~

1 i

.- *o so se se no ice m- no ie n.......,

1 l Figure 3.13 Accumulator Flow Rate (Base Case) 3-27

4 Comanche Peak Steam . Electric Station Unit 1-g . , , ,

d I '

H E

a O -

3 0 '

g 8 i c

3 .

g r 4

2 h

v i 1 1 j {

.e ao is so so ice ne no ti . ts... 4.)

Figure 3.14 CCP & lillSI Pump Flow Rate (Dase Case)

Comanche Peak Steam Electric Station Unit 1-

~

n $

k E

a C I .

1 2

$ g -

c -

o. E E

E 8

, g

=

0 I 1 i g

- io . u so se ino no . n, is, -

tim. (s.. 41 Figure 0.15 EllR Pump Flow Rate (Date Case)

.=

3-28

Comanche Peak Stearn Electric Station Unit 1 3

3 E h t'

g S :

a 2 , e m . :

i 3

j/A

,~, i h $k .

j m

'b (s J,y to Y

.3

~

%e < e no ss O n n n vi.. i s... 4.>

Figure 3.16 Mid Elevation 1(ot Assembly llent Transfer Coeff (Itane Case)

Comanche Peak Steam Electric Station Unit 1 l ! -

e 4 ! -

a 2 E

^

3 gg .. .

d I. i .' n .. A 4 J u tim ts... 4.)

Figure 3.17 liot Rod Temperature At PCT Node Elevation (Dase Case) 3-29 i

-l j

I l

Comanche Peak Stearn Electric Station- Unit I- <

s

. W 3

5 -

Y s -

8

(

c I

n.. i. ..i Figure 3.18 Core Flooding Rate (Base Case)

Comanche Peak. Steam Electric ' Station Unit 1 l.'

i i i , , ,

& 3 .

o .

u

6. ' !

3 a i E f bN l ti . (s...... i Figure 3.19 Mid Elevation Hot Assembly- Zr/ Water Reaction. Depth (Dase Case) 3-30

.l t

Comanche Peak Steam Electric Station Unit 1

g. -

C 1 -

, A U

$ l .

g 3

=*

g .

2 -

L e

E i -

.a

.S -

E-

/

U 1 .

l.. .. .... .. . .. .. .. .... ... .

tim.1s. 4.i Figure 3.20 PCT / Ruptured Node Cladding Temperature-(Base Case)-

Comanche Peak Steam Electric Station Unit 1 i

C A e

S g .

! ^

t.

e E i -

ne

.e e' -

c g .

l ... ... .... .. .. ..... .. .... .... , , . .

n.. is... 4. >

Figure 3.21 PCT / Ruptured Node Clad Ternperature (Cosine /CD=0.8) .

3-31

~ _-.

-i 1

i Comanche Peak Steam Electric: Station Unit 1. .!

! : i 1

$ f 4

8 g -

t a

.k Ei -

/ ,

e4

.e . s- -

2 U n l.. ... .... ... u.. .. .. .. .... .. ..

ri.. o .....

-4 Figure 3.22 PCT / Ruptured Node Clad Temperature (Costne/CD=0.0)

Comanche Peak Steam Electric Station Unit 1 i

c . A U

e. g .

r.

e e

E i -

a =s i

U n.

I .. .

ri.. o.. ..i Figure 3.23 PCT / Ruptured Node Clad Temperature (Cosine / Split /CD=1.0).

l 3-32 1

__.__ _ ____.-___ __m.. ___

4 Comanche Peak Steam Electric Station Unit'l I , , ,

e a ! -

.a E

2 E.

n l

~

0 f t ~

f 1 1 i I. . . ,, ,. .. ... ,. . n

n. . .. ... . i Figure 3.24 Hot Rod Temperature At -PCT Node Elev. (Skewed 0 0.7S f t)

Comanche Peak Steam Electric Station Unit 1, I

In. 2 .

a e

S g .

D t 7 t.

o E

e i -

u j i -

1 0 g .

l ..r .... .. .... .... ..... , , . . . . . . . . ..... . . . . . . . .

n.. c. ....>

Figure 3.25 PCT / Ruptured Node Clad Temperature (Skewed O 8.75 f t)

)

3-33 l

Comanche Peak Steam Electric Station Unit 1 C t . .

. A U

O l -

[ ,

5 e t -

1.

H E l -

w

.!! [

c g . .

i ... ... .... ,... .... ,,... m.. -......... ... . . .

tin. is... 4.)

Figure 3.26 PCT / Ruptured Node Clad Temperature (Skewed 0 0.75 ft)'

Comanche Peak Steam Electric Station-. Unit- 1 i

C .

A U

Q t .

e I-t c.

F E i -

u

.9 E U {

l ... n.. .. .... .... ..... m. . . . . . . , , , . . ... . . . .

tun. n. 4.)

Figure 0.27 PCT / Ruptured Node Clad Temperature (Exposure / Skewed 0 0.7S f t) 3-34

q CllAPTER 4 CONCLUSION l

The USNRC-approved (Ref. 4.1) ANF Corporation's large break I ECCS Evaluation model entitled EXEM/PWR has been applied to_ j the Comancho Peak Steam Electric Station Unit One (CPSES-1).

Each calculation has been performed in exact compliance with the explicitly approved EXEM/PWR methodology. Regarding features of the calculation procedure which are " implied" in -

the approval, there has been but one deviation:'the I'

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

Seven calculations have been presented with two objectives:

1. To demonstrate Texas Utilities' ability to properly apply EXEM/PWR (Ref. 1.1);.and i

2. To demonstrate the development of up-to-date input decks and conclusions which are in compliance with 10 CFR 50, Appendix K. Together, the codes, input decks and conclusions will be applied to subsequent fuel s

4-1 1

l cycles for the Comanche Peak Steam Electric Station Unit One and Unit Two. Evaluations will be performed i to verify that the results of the present analyses remain bounding.

Table 4.1 summarizes 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 (b) (1).

2. The total cladding oxidation at the peak location is under the 17% limit specified in 10 CFR 50 -(b) (2) .

3. The hydrogen. generated in the core by cladding oxidation is less than the 1% limit established by 10 CFR 50 (b) (3).

4. Only hot channel rods experience clad rupture. The average core region undergoes only minor dimensional changes, but no clad ruptures are calculated to-occur there. Thus, the coolable geometry criterion of 10 CFR 50 (b) (4) is satisfied.

4-2 l

1

t F

5. The available ECCS is successfully initiated,and the h core is well cooled in less than 200 seconda.

Therefore, the calculations comply with the long-term P c.ooling criterion of 10 CFR 50 (b) (5).

i Regarding the various sensitivity _ studies it has baen found:

1. The most limiting break is a Double-Ended Guillotins rupture of the main coolant pump discharge line uith a _j discharge coefficient of 1.0. l..

2. The most limiting Power Shape is the top skewed profile peaked at the elevation of 8.75 ft, which'is shown in Figure 3.1.

3. The most limiting exposure occurs at 613.8 hourr, and is coincident with maximum stored-energy in the fuel.

Texas Utilities will use the EXEM/PWR methodology inc2uding all codes, input decks, results, conclusions, and application procedures presented in this report to perform large break -

LOCA analynes and evaluations in compliance with 10 CFR 50 e criteria and 10 CFR 50, Appendix K requirements, for both ;

Comanche _ oon Steam Electric Station Unit one and Unit 1No.

4 4-3

. -_-_2

i TABLE 4.1

SUMMARY

OF RESULTS FOR BASE CASE AND SENSITIVITY STUDIES i

DOUBLE

ENDED 1 AX1 AL POWER t#, APE 8 CUILLOTikE BREAK CD CHOPPED COSINE (4) TUS PEAltEU A18.75 f t TO PEAKED AT 9.75 i FT I EXPOSURE

80t (5)

BOL (5)

EOL (6)

BOL (5) 1 0 6 374g op 3933 og 1.0 1959 8 (1) 2634 7

> 4.14 % (2) 5.38 % 1 76 % 3.40 %

0.56 % (3) 0.72 % l 0.46 % 0.69 %

0.8 IP'O */ E91Lii 2.31 %

0.43 % (1) PEAK CLAD TEMPERATURE -(DEGREES F)

(2) PERCENT LOCAL CLAODING OXICATI0li e 1768 0F O.6 (3) PERCENT CODE

WIDE OXIDATi^N 1.27 % (4) BASE CASE 0.30 % (5) MAXIHUM STORED ENERGY (6) MAXIMUM PIN PRESSURE-LONGITUDINAL SPLIT I AXIAL PCVER SHAPE 1 BRCAK CD CHOPPED COSINE ,

_I EXPOSURE

BOL (5) 1.0 1901 'F 2.83 %

0.52 %

il i

4-4

- _j

I q

l

' CHAPTER-5' W

1 REFERENCES >

phauter'1: IB 1.1 : Advanced ~ Nuclear'FuelsiCorporation "USNRC.'s Safety.

Evaluation'of Exxon-Nuclear: Company's-LargeiBreak ECCS- 'q Evaluation.ModelLEXEM/PWR;and Acceptance;forf Rcfarencing of.-Related Topical ~ Reports"f JulyL1986.

A

.CJ1gpter 2 :.

2.1- . Exxon Nuclear-Company,-"WREM-Based' Generic PUR ECCS-Evaluation;Model.-Update. ENC: WREM-IIA",-XNiNF-78-30'(A),.

May'19.79,7 2.2 K..R'. Merkx etLal.,f."RODEX2 Fuel Rod Thermal-Mechanical. 2j Response Evaluation?Model",-XN-NF-81-58f Revision 2, _

February-1983.

2. 3- K.--V. Moore'and W.::H.LRettig,1"RELAP-4','A Computer' Program for-Transient Thermal-HydraulicLAnalysis"',

ANCR-1127, Decemberi1973. .

~

l 2.4 ExxonsNuclear. Company, e'!WREM Based 1GenericL PWR)ECCS Evaluation'Model",3XN-75-41,.TV.1,2 and_ Sups.. 1-6, 197.5;

, s!

2.5 'L..L. Wheat et al.,<" CONTEMPT-LTlAlComputer-Program:for7 PredictingiContainment Pressuro-Temperature;ResponseLto;

. a Loss of: Coolant: Accident",1 TID;4500,?Junei1975.:

2.6 W.1V. Kayser, "PREFILL -fADComputer,Programyton

-Calculate.RefloodiParameters?and7 Flow RatesiforJthe'. ENC:

'WREM-Based LOCA ECCS' Analyses",-XN-NF-CC-4'4FDecember ,

1977. '

2.7 'W. V. Kayser, " REFLEX? -:PWR Reflood Computer Program -

Users Manual -WREM-IIA", XN-NF-CC-53,;. March-1984.. .

1 a i-

> l

, ..4,,6 ,- . - , ~ - s. . a .. ~ . ~ , _ .- . _ - s .. M

l l

2.8 Nuclear Regulatory Commiss.4nn, Division of Technical Review, "WREMt Water Roa0.1 Evaluation Model",

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

2.9 Exxon Nuclear Company, "WREM-Based Generic PWR ECCS 1 Evaluation Model Update ENC WREM-II", XN-76-27, July i

1976; XN-76-27, Supplement 1, September 1976; and XN-76-27, Supplomont 2, November 1976.

2.10 R. W. Jeppson, " Analysis of Flou in Pipe Networks", hDD Arbor Sciengg, Ants Arbor, Michigan,1976.

2.11 B. Carnahan, H. A. Luthor, and J. O. Wilkin, " Applied Numerical Methods", John Wiloy and Sons, Inc., New York, 1969.

2.12 Electric Power Research Institute, " Mixing of Em2rgency Core Cooling Water with Steamt 1/3 Scale Test and Summary", EPRI-294-2, June 1975.

2.13 Division of Technical Review, Nuclear Regulatory Commission, "TOODEE2 A Two Dimensional Time Dependent Fuel Element Thermal Analysis Prograu", NUREG-75/057, May 1975.

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

2.15 M. M. Giles ot. al., "RFPACt A Computer Program for PWR Refill-Reflood Analycis", Users Manual, 'NF-1164 (P),

May 1990.

2.16 USNRC, " Minimum Containment Pressuro Model for PWR ECCS Performance Evaluation", Branch Technical Position CSB 6-1.

2.17 Comanche-Peak Steam Electric Station Unit One, Technical Snecificationg.

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

Chantcr 3:

3.1 Comanche Peak Steam Electric Station Unit One, "Finnl Safety Analysis Report", Section 15.6, Amendment 78.

January.15, 1990.-

5-2

f 3.2 USNRC, " Minimum Containment. Pressure Model for PWR ECCS Performance Evaluation", Branch Technical Position CSB 6-1.

3.3 USNRC, " Water Reactor Evalua91on Modal (WREM): PWR Nodalization and Sensitivity Studies", - Technical Review U.S. Atomic Energy Commission, October 1974.

3.4 Comanche Peak Steam Electric Station Unit One, ,

Technical snecifications.

3.5 " Axial Power Distribution Control Analysis and

. Overtemperature and overpower. Trip Setpoint Methodology", RXE-90-006-P, to be published.

Ch.aD1 a 1.1.

4.1 Aoyanced Nuclear Puols Corporation, "USNRC's Safety Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluati( n Model EXEM/PWR and Acceptanco for Referencing of R4?.ated Topical Reports," July 1986.

5-3 i i

.r

APPENDIX DESCRIPTION OF THE COMPUTATIONAL TOOLS The EXEM/PWR Eve _luation Model utilizes four basic computer Co0sBt

1. RODEX2

2. RELAP4-EM

1. SYSTEM BLOWDOWN

2. HOT CHANNEL

3. ACCUMULATOR-SIS

3. RFPAC

1. PREFILL

2. REFLEX

3. ICECON/ CONTEMPT-LT

4. SHAPE /REFLOOD

4. TOODEE2 5

These basic codes address the various stages of the LOCA calculation as discussed in Section 2.2 and illustrated in Fig. 2.2.1. The codes, their interfaces, interrelationships and respective inputs and outputs are sunmarized in Fig. A.1 and Table A.1. The function of each code is described in the following sections.

d A-1

A.1 RODEXa RODEX2 is used within the EXEM/PWR framework to provide initial conditions for the RELAP4-EM system blowdown ,

calculation. These conditions are (a) stored energy in the fuel, (b) gap gas composition, and (c) rod internal pressure.

The stored energy is input iteratively by adjusting the fuel .

rod gap dimension in the RELAP4-EM system initialization calculation until the calculated stored energy matches the RODEX2 value. During the RELAP4-EM hot channel calculation, the built-in RODEX2 fuel models are activated so that gap adjustments are not needed there.

RODEX2 describes the thermal-mechanical performance c: fuel during its operational lifetime preceding the LOCA. The determination of stored energy for the LOCA analysis requires a conservative fuel rod therinal-mechanical model that is capable of calculating fuel and cladding behavior, including the gap conductance between fuel and cladding as a function of burnup. The paramotors affecting fuel performance, such as fission gas release, cladding dimer.sional changes, fuel densification, swelling, and thermal expanrion are accounted for.

RODEX2 provides an integrated evaluation procedure for considering the effect of varying temporal and spatial power A-2

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 i

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 all axial nodes. Deformation of the fuel and cladding and gas release are calculated using shorter time steps 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 aeformation calculations include consideration of densification, swelling, instantaneous plastic flow, creep, cracking and A-3

thermal expansion for the fuel pellet, and also consideration of creep, irradiation induced growth, and thermal expansion for the cladding.

A.2 EELAP4-EM

{

A.2.1 SYSTEM BLOWDOWN b

This code (Ref. 2.3) has been abundantly discussed in the literature. Only specific features of the EM version are briefly summarized in this section.

b The fluid dynamics portion of the RELAP4 program solves the fluid mass, energy, and momentum equations. There is a choice of several fluid momentum equations in RELAP4. All of these are one-dimensional approximations and differ in the mathematical treatment of momentum flux. The form used in EXEM/PWR is the Incompressible Mechanical Energy Balance equation where the fluid dynamics in the vicinity of an area change is treated as incompressible. This formulation is used 4.n all one-dimensional flow paths throughout the system piping and core. In the case of plena, the plenum area is specified as arbitrarily large, resulting in an equivalent stagnation volume treatment of the plenum. This modeling effectively eliminates the momentum flux portion of the equation as prescribed in Ref. 2.3. Similarly, in the caso of A-4

crossflow, all momentum flux terms are deleted in the crossflow direction by assuming no area change (an open lattice) within the crossflow path.

A.2.2 HOT CHANNEL This is not another code but an application of the RELAP4-EM code discussed above, to the hot channel. Boundary conditions from the system blowdown calculation are used.

The main outputs from this calculation are the hot rod temperature distribution and oxide layer thickness at the End-of-Bypass. These are used to initialize the heatup calculation performed in the TOODEE2 code for the refill and reflood periods. The (RELAP4 'M) hot channel calculation is necessary in order to adequately account for crossflows in the hot assembly and for the hot rod which is not represented in the system blowdown calculation.

A.2.3 ACCUM-SIS ACCUM-SIS is also an application of RELAP4-LM. The ACCUM-SIS calculation determines the ECCS flow rates to the cold legs after the end-of-bypass period (EOBY). The broken and intact loops are modeled, including accumulators, high, intermediate, and low pressure injection systems. Explicit modeling of the injection systems' piping is not done. These systems are A-5

moduled as fill junctions at the accumulator lines (Fig.

2.3.3). The broken loop flow is assumed to be lost to the containment and is included in the ICECON/ CONTEMPT-LT input.

The intact loop ECCS boundary conditions for the ACCUM-SIS calculation is taken from the RELAP4-EM system blowdown calculation up to EOBY and assumed to be constant and equal to the containment pressure at EOBY thereafter.

A.3 RFPAC mi RPPAC combines the four codes usod to perform the refill and reflood thermal-hydraulic analyses (ICECON/ CONTEMPT-LT, PREFILL, SHAPE /REFLOOD, and REFLEX) and eliminates the need for dr.ta transfer between codes. In the context of the overall EXEM/PWR methodology, RFPAC serves as a bridge between the RELAP4 and supplying fluid boundary conditions to TOODEE2.

A.3.1 ICECON/ CONTEMPT-LT ICECON is essentially the same program as CONTEMPT-LT (Ref.

2.5). This is a computer program developed to describe the thermal-hydraulic b aavior of reactor containment systems subjected to postulated accident conditions.

A-6

The code calculates the interrelated effects of reactor system blowdown, heat transfer, atmosphere leakage, safeguards system operation, pressure suppression system response, and miscellaneous mass and energy additions.

The code is used in the EXEM/PWR framework to provide containment pressure as a boundary condition for the PREFILL and REFLEX codes, which are used in reflood calculations.

The mass and energy releases to the containment are input from the RELAP4-EM system calculation during the blowdown stage and from the broken loop ACCUM-SIS calculation after that, as described in Section 2.2.

A.3.2 PREFILL i

The time between the system blowdown period as defined by the End-of-Bypass (E0BY) and the beginning of reflood when the water level reaches the bottom of the core (DOCREC) is the refill portion of the LOCA transient. The PREFILL code calculates the time to start of reflood and the flow of ECCS fluid to the core during reflood.

The phenomena addressed by PREFILL are (a) hot wall delay period, (b) free-fall delay time, (c) extended accumulator flows, (d) open channel flow spill, and (d) core inlet subcooling.

A-7

A.3.3 SJIAPE/REPLOOD SilAPE/REFLOOD uses the average core fuel and cladding temperatures at End-of-Bypass from the RELAP4-EM hot channel calculation to determine the average rod temperatures at the peak power location at the time of BOCREC for use in tho Fuel Cooling Test Fucility (FCTF) reflood correlations in the REFLEX code. '

Injection of subcooled ECC water is possible. Steam condensation in the intact loops is accounted for, and spillage to the break from the downcomer is based on gravi-tational head forces developed between the downcomer and the break when the downconer is full to the cold leg pipe level.

The steam-weter interaction pressure loss penalty during pumped ECC injection is reduced to an injection-angle independent value of 0.15 psi, based upon EPRI data (Ref.

2.12). During the accumulator injection period the penalty is 0.6 psid.

Log-mean-temperature heat exchanger thermal balance equations are used for the heat transfer occurring in the steam generators instead of the RELAP4-EM conservation equations.

This is faster and is justified by the slow thermodynamic A-8

changes occurring in the steam generators during reflood as compared to blowdown.

A.3.4 REFLE4 The REFLEX program calculates core refloor "Lis program is built v,'n a RELAP4 skeleton, che RELAP4 system equations are simplified in REFLEX in the interest of computational speed.

The system modeling detail and sophistication required for the blowdown calculation is not required for the somewhat slower reflood process. The code utilizes a quasi-steady state solution of the mass, momentum and energy equations for PWR reactor systems. Specific models were developed for the system, core downcomer annulus, ECC mixing location, and steam generators. An equation of state was developed to provide fluid properties. The modifications made to the original RELAP-EM/ FLOOD code to produce REFLEX are as described in the following paragraphs.

The core neutronics, transient heat conduction and critical flow tables are omitted.

Acceleration pressure losses are omitted in the flow equations. Mass accumulation and gravitational losses are A-9 i

also omitted in all system components except in the core, downcomer nodes, and in the cold leg piping to the break during the accumulator discharge phase.

The fluid state equations are based on analytical fits to property tables over a limited pressure range,10-100 psia.

This method is faster than the previous table look-up process.

The numerical scheme of RELAP4 is replaced for the flow calculation by the linear theory method (Ref. 2.10), using a Gauss-Jordan elimination method (Ref. 2.11). [

The core outlet enthalpy is conservatively assumed to be determined by steam generator secondary temperature and containment pressure in order to yield a conservatively high upper plenum pressure for reflood.

A.4 TOODEE2 TOODEE2 calculates the time dependent temperature distribution in the hot rod during the refill and reflood portions of the LOCA. The TOODEE2 calculation begins at End-of-Bypass (EOBY).

A-10

TOODEE2 is a two-dimensional, time-dependent fuel rod element thermal and mechanical analysis program. T00DEE2 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 zircalloy. The energy -

equation is solved to determine the fuel rod thermal 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. Whenever rupture is determined and the flooding rate drops below 1 in/sec, only steam cooling is allowed downstream of the ruptured node. This is in compliance with the related Appendix K requirement.

The effect of clad strain on pellet-to-clad gap heat transfer and on the thinning of the oxide layer on the outside of the cladding is considered. Once fuel rod rupture is determined, the code calculates both inside and outside metal water heat generation. Fuel rod rupture reduces the subchannel flow area at the rupture and diverts flow from the hot rod subchannel to neighboring subchannels.

A-11

Flow recovery is assumed above the rupture. The effect of flow diversion on heat transfer (both convection and radiation) to the coolant is accounted for. The TOODEE2 code calculates heat transfer coefficients as a function of fluid condition or via reflood data-based correlations.

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 criteria (if pin-wide oxidation is less than 1% it is concluded that the criteria of less than 1% coro-wide oxidation is met).

A.S DATA PREPARATION AND TRANSFER TOOLS The EXEM/PWR methodology also includes four additional codes for preparing data and transferring results between the basic codes described above:

1. FISHLX

2. SHAPE

3. BLOCK

4. BLOWDOWN-ICECON A.5.1 FISHEX FISHEX is used to determine the normalized power (P(t)/P(0))

following the End-of-Bypass, for use with REFLEX and TOODEE2.

A-12

The code accounts for fission and decay heat, including heat from actinide decay. The 20% overpower factor included in the RELAP4 calculation is not applied in FISHEX to the decay heat from actinides, since this extra 20% of the decay heat from actinidos is not required by 10 CFR 50.46 Appendix K.

A.5.2 EHhEE SHAPE automates the building of portions of input decks to RELAp4, RODEX2, and TOODEE2. The code prepares input related to the axial power profile. The SHAPE code can alter and re-normalize a given axial power shape to a prescribed axial peaking factor. It then generates the power fraction input i

data for RELAP4 and the axial power factors for input to the RODEX2 and TOODEE2 codes. The code can also set up certain blocks of input data to RELAP4, viz. reactiv.it-l coefficient data, core heat slab datt. 'oro section data, and core geometry data cards.

A.5.3 DLOCK BLOCK generates the clad swel31ng and rupture tables for input to the RELAP4 code. The swelling and rupture model used in BLOCK is taken from TOODEE2, The model is based on data with temperature rates of 0.0 to 28.0 degree C per A-13

second and can interpolate the data between these two ramp rates.

1 l

4 The 0.0 degree C per second ramp rate is the most conservative becaur' this rate leads to swelling and rupture

) at lower cladding hoop stresses. The 0.0 degree C per second

, value is used in the present LOCA/ECCS analyses. The tables generated by BLOCK are valid for calculating the fuel rod pre-rupture strain in RELAP4.

A. 5. 4 BLOWDOW!{-LCE,Qll BLOWDOWN-ICECON is a data transfer tool. This code reads output files from the RELAP4 system calculation and transfers information to ICECON/ CONTEMPT-LT and to FISHEX. The RELAP4 mass and energy releases to the containment are written in the appropriate format for CONTEMPT-LT use. Time-dependent reactivity calculated by RELAP4 is also read by BLOWDOWN-ICECON and converted into the appropriate format for input to FISHEX.

A-14

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

BLOCK INPUT:

(1)* Number of fuel rods per assembly Number of instrument tubes per assembly Number of guide tubes per assembly Inside diameter of cladding outside diameter of cladding Rod pitch Outside diameter of instrument tube outside diameter of guide tube Cladding temperature ramp rate OUTPUT:

(2) Rupture and blockage tables SHAPE PUNCH INPUT:

(3) 24 point axial power profile (Reactor Physics)

(4) Tech Spec peaking factor (5) Renormalized 24 point axial power profile to Tech Spec peaking factor OUTPUT:

(6) 101 point axial power profile with Tech Spec peaking factor The numbers in this table correspond to the numbers in Figure A.1.

A-15

TABLE A.1 (Continued...)

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

SHAPE j INPUTt (6) 101 point axial power profile with Tech Spec peaking factor (7) Adjusted axial peaking factor (tech spec) at peak node Axial nodalization to be used in RELAP4, TOODEE2, or RODEX2 Bundle geometry data (RELAP4)

OUTPUT (8) Reactivity weighting factors Power fraction data Core heat slab data Core section data Core geometry (9) Power fraction data Core heat slab data Core section data Core geometry (10) Axial core power factors Peak axial power location (11) Axial power factors Axial grid locations A-16

TABLE A.1 (Continued...)

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

RODEX2 INPUTt

{

(11) Axial power factors Axial grid locations (12) Description of fuel, e.g. geometry, density, enrichment, etc.

Cladding type and dimensions Initial mole fractions of fill gas i Spring dimensions Hydraulic diameter, area, mass flux Axial nodalization (13) Core power history [

Average core L Hot assembly Hot rod A-17

TABLE A.1 (Continued...)

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

RODEX2 OUTPUT (14) 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)

Hot rod dish + crack volume (at exposure of interest)

Hot rod variables (at exposure of interest) to calculate cladding diameter and cold gap width Hot rod mole fractions (at exposure of interest)

Hot rod radially averaged density (at exposure of interest)

Cladding + fuel surface roughness (15) Hot rod, hot assy, and average core gram-moles of gas (at exposure of interest)

Hot rod, hot assy, and average core mole fractions of gas (at exposure of interest)

Hot rod radially averaged density (at exposure of interest)

Fuel model data cards, fuel density, and flux depression (at exposure of interest)

Cold plenum length + dish volume (at exposure of interest) used to calculate cold plenum volume (16) Hot assy and average core gram-moles of gas (at exposure of interest)

Hot assy and average core mole fractions of gas (at exposure of interest)

Cold plenum length + dish volume (at exposure of interest) used to calculate cold plenum volume A-18 E . . - ______.__m_ _ . _ _ _ _ . . _ , _

TABLE A.1 (Continued...)

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

RELAP4-SYSTEM INPUT (2) Rupture and blockage tables (8) Reactivity weighting factors Power fraction data Core heat slab data Coro section data Coro geometry (16) Hot assy and average core gram-moles of gas (at exposure of interest)

Hot assy and average coro mole fractions of gas (at exposure of interest) cold plenum length + dish volume (at exposure of interest) used to calculato cold plenum volume (17) NSSS definition (e.g. geometry, pump data, heat slabs, etc.)

ECCS definition (e.g. accumulator volume, SI flow rate, etc.)

Containment definition Fuel data Neutronics data A-19

TABLE A.1 (Continued...)

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

RELAP4-SYSTEM

{

OUTPUT:

(18) Core inlet and outlet plenum data as boundary conditions Core power data EOBY time (19) EOBY time (20) Break mass and energy out to EOBY time Liquid remaining in the primary system at EOBY 2

Reactivity versus time EOBY time rL (21) RELAP4 ECCS model input (22) Cold leg pressures (intact and broken loops) to EOBY time as boundary conditions Containment pressure at EOBY time EOBY time Time when the high containment pressure SIAS setpoint is reached (23) Maximum downcomer or lower plenum slab temperature (24) Steam generator secondary pressure and liquid mass A-20

TABLE A.1 (Continued...)

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

RELAP4 HOT CHANNEL INPUT:

(2) Rupture and blockage tables -

(9) Power fraction data -

Core heat slab data Core section data Core geometry (15) Hot rod, hot assy, and average core gram-moles of gas (at exposure of interest)

Hot rod, hot assy, and average core mole fractions of aW gas (at exposure of interest)

Hot rod radially averaged density (for all core nodes at exposure of interest) jl Fuel model data cards, fuel density, and flux depression (at exposure of interest)

Cold plenum length + dish volume (at exposure of interest) used to calculate cold plenum volume 1

(18) Core inlet and outlet plenum data as boundary conditions Core power data EOBY time OUTPUT (25) Fuel average temperatures and cladding temperatures for the 5 average core nodes (26) Punch file created containing hot rod temperature distribution for the 24 axial node input Punch file created containing the oxide layer thickness for the 24 axial node input Hot rod internal pressure at the EOBY Hot rod cladding, hot assembly cladding, and average core cladding temperatures at the EOBY for calculation of radiation model sink temperatures for the 24 hot rod --

axial node input.

A-21

TABLE A.1 (Continued...)

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

BLOWDOWN ICECON INPUT:

(20) Break mass and energy out to EOBY time Liquid remaining in the primary system at EODY Reactivity versus time EOBY time OUTPUT:

(27) Punch file created containing break mass flow rate and enthalpy versus time (28) Punch file created containing reactivity versus time ACCUMULATOR-SIS INPUT:

(21) RELAP4 ECCS model input (22) Cold leg pressures (intact and broken loops) to EOBY time as boundary conditions Containment pressure at EOBY time EOPi time Time when the high containment pressure SIAS setpoint is reached OUTPUT:

(29) Broken loop ECCS flow rates and enthalpy to containment after EOBY (30) Intact loop ECCS flow rates and enthalpy after EOBY A-22

TABLE A.1 (Continued...)

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

CONTEMPT-LT INPUT:

(27) Punch file created containing break mass flow rate and enthalpy versus time (29) Broken loop ECCS flow rates and enthalpy to containment after EOBY OUTPUT:

(31) Containment pressure and temperature response versus time FISHEX INPUT:

(28) Punch file created containing reactivity versus time (32) Effective delayed neutron fraction divided by prompt neutron generation mean lifetime U230 atoms consumed per U235 atoms fissioned OUTPUT:

(33) Punch file created containing normalized power versus time A-23

TABLE A.1 (Continued...)

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

PREFILL INPUT:

(23) Maximum downcomer or lower plenum slab temperature (30) Intact loop ECCS flow rates and enthalpy after EOBY OUTPUT:

(34) BOCREC time ECCS injection rates after BOCREC Temperature of ECCS fluid entering the core E

SilAPE/REFLOOD INPUT:

(25) Fuel average temperatures and cladding temperatures for the 5 average core nodes at EOBY (33) Punch file created containing normalized power versus time (39) Core power Core average linear heat generation rate OUTPUT:

(35) Average rod temperature at the peak power location at BOCREC time for use in the FCTF reflood correlations A-24

TABLE A.1 (Continued...)

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

REFLEX IPPUT:

(10) Axial core power factors-Peak axial power location (24) Steam generator secondary pressure and liquid mass (31) Containment pressure and temperature response versus -

time

, (34) BOCREC time ECCS injection rates efter BOCREC Temperature of ECCa fluid entering the core (3 5) Average rod temperature at the peak power location at BOCREC time for Use in the FCTF reflood correlations (36) Primary system geometry and loss coefficients based on the RELAP4 system deck l OUTPUT:

(37) Core coolant conditions versus time (core inlet flow, saturation temperature, effective core inlet ficoding rato, and quench height)

BOCREC time Time when instantaneous reflood rate drops below 1 inch /sec 1

A-25 i

J

-. .. : . ._ L _

TABLE A.1 (Continued...)

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

TOODEE2 IllPUT (14) 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)

Hot rod dish + crack volume (at exposure of interest)

Hot rod variables (at exposure of interest) to calculate cladding diameter and cold gap width (used for geometric definition of hot rod and blockage data)

Hot rod mole fractions (at exposure of interest)

Hot rod radially averaged dcnnity (at exposure of interest)

Cladding + fuel nurfaco roughness (19) EOBY time (26) Punch file created containing hot rod temperature distribution for ~

the 24 axial node input Hot rod internal pressure at the EOBY Hot rod cladding, hot assembly cladding, and average core cladding temperatures at the EODY for calculation of radiation model sink temperatures for the 24 hot rod axial nodo input (33) Punch file created containing normallred power versus time (37) Core coolant conditions versus time (core inlet flow, saturation temperature, effective core inlet flooding rate, and quench height)

BOCREC time Time when instantaneous reflood rate drops below 1 inch /see BOCREC time Core saturation temp at DOCREC Temperature of ECC water at BOCREC (38) Axial power factors (from SHAPE)

Axial grid locations (from SHAPE)

Hot rod ALHGR (from SHAPE)

Additional blockage data 9

A-26

i TABLE A.1 (Continued...) ;

INPUT AND OUTPUT FOR THE EXEM/PWR METHODOLOGY COMPUTER CODES (refer to FIGURE A.1) 4 TOODEE2 OUTPUT:

(40) Peak cladding temperature Percent local cladding oxidation Percent pin wide cladding oxidation 9

L

.4 D

o

@ A-27

. ___ _j

FIG.-A.1 RNF E X E t1/ P W R METHODOLOGY FOR LARGE_BRERK_LOCR

@J g_ .

SHAPE ;-

SHAPE PUNCH h '

b :

R00EX2 Q >

BLOCK 4 -

. ,I._ h b b. .

RELAPH h '

N RELAP4 HOT SYSTEM Q ; CHANNEL 4

O 4 :- g en et ec

w . -

E A C C U M - S 1,o.

BLOWOOWN 1CECON

...... ... )...

.........]...

l PREFILL 7

CONTEMPT i

FISHEX g

l f ....L...._...,

i $

i I, SHAPE l

l 3< 7n l

l g ,

i REFLOOD l

l ......................... .......l .l

f

. REF

LEX l4 '

Q l

RFPRC O >

h T00DEE2 se v

A-28

ML20066L069

Contents

Text

=

2.3 DESCRIPTION

2.3 DESCRIPTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Navigation menu

ML20066L069

Text

=

2.3 DESCRIPTION

2.3 DESCRIPTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Navigation menu

Search