ML20095K541
| ML20095K541 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/18/1995 |
| From: | Brozak D, Da Silva H, Tajbakhsh A TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML19317C215 | List: |
| References | |
| RXE-95-001-NP, RXE-95-1-NP, NUDOCS 9512290080 | |
| Download: ML20095K541 (120) | |
Text
.
- O RXE-95-001-NP
.O i
O SMALL BREAK LOSS OF COOLANT ACCIDENT ANALYSIS METHODOLOGY O
DECEMBER,1995 A. E. Tajbakhsh D. E. Brozak H. C. da Silva, Jr.
O 4
Reviewed:
M Date: /
/
% Vee G.
oe Safety A alysis Manager O
/8!/f/9,f' Approved:
IW fdde-Date:
Mickey R.((llgore O
Nuclear Analysis a Fuel Manager O
Approved:
I-Date: L*fII AusafHusain Nuclear Engineering Manager
.O 9512290080 951219 O
PDR ADOCK 05000445 P
. ~ _. _.. _...
O O
lO DISCLAIMER lO The information contained in this report war prepared for the specific requirement of Texas Utilities Electric Company (TU Electric), aad may not be appropriate for use in situations other than those for which it was specifically prepared. TU Electric PROVIDES NO WARRANTY O
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 O
i PARTICULAR PURPOSE.
1 k
By making this report available, TU Electric does not c.ithorize its use by others, and any such use is forbidden except with the prior written approval of TU Electric. Any such written approval shall itself be deemed to incorporate the disclaimers ofliability and disclaimers of warrants provided herein. In no event shall TU Electric have any liability for any incidental or 1
consequential damages of any type in connection with the use, authorized or unauthorized, of O
this report or for the infonnation in it.
O ii O
O l
ABSTRACT O-This report is presented to demonstrate TU Electric's application of USNRC-approved Siemens O
Power Corporation's (SPC) Emergency Core Cooling Systems (ECCS) Evaluation Model EXEM PWR Small Break Model, to the Comanche Peak Steam Electric Station (CPSES).
O' This report contains a description of the EXEM PWR Small Break methodology which includes the computer codes, the details of the nodalization schemes, and the calculational O
procedures followed during all phases of the Loss-of-Coolant Accident (LOCA) analyses. The methodology is used to perform small break LOCA-ECCS licensing analyses that comply with i
USNRC regulations contained in 10 CFR 50.46 and 10 CFR 50, Appendix K. The method also O
satisfies the requirements of NUREG-0737, TMI Action Item II.K.3.30.
O In order to comply with a 10 CFR 50, Appendix K requirement, a spectrum of small breaks, ranging from 2 through 4 inches in diameter, is examined.
O Although higher peak clad temperatures (PCT) are usually associated with beginning oflife (BOL) fuel because of the higher stored energy, a fuel bumup study is also conducted. This is O
done in order to confirm that the end of cycle (EOC) pin pressures, which are higher than those encountered early in life and which foster a higher driving force for rod burst, do indeed result in lower PCT for the fuel under consideration.
iii O
O:
All system analyses were performed with ANF-RELAP using an explicit representation of the 4:
CPSES-14 loops. However, because CPSES-1 has no significant loop asymmetries, a model using the more customary 2-loop representation, with a broken loop and a lumped intact loop has also been developed. The limiting analysis was repeated using this 2-loop model and yielded essentially identical result? to the 4-loop model.
4; In order to further support the " robustness" of the findings, two additional types of sensitivity studies were performed. The first was a time step study demonstrating that all break spectrum results are converged. The second was the [
3-e:
The methodology presented herein - including all codes, results, input decks, inferrences and conclusions presented within this report - will be used to perform small break LOCA analyses and evaluations in compliance with 10 CFR 50.46 criteria and 10 CFR 50, Appendix K 4
requirements, for fuel cycle analyses and to address pertinent licensing issues, for Comanche Peak Steam Electric Station Unit One and Unit Two.
9:
e S
iv 9
O TABLE OF CONTENTS O
PAGE DI S C L AIMER..................................................... ii O
AB S TRA CT...................................................... iii TABLE OF CONTENTS
.............................................v LI S T OF TAB L ES................................................... vii
-O LIST OF FIGURES.............................
.................ix
- O CHAPTER 1.
INTRODUCTION.............
......................1-1 i
i 2.
DESCRIPTION OF THE METHOD............
............2-1
.O 2.1 DETERMINATION OF INITIAL FUEL PARAMETERS.......... 2-1 2.2 SYSTEM THERMAL-HYDRAULIC RESPONSE ANALYSIS...... 2-2 o
2.3 HOT ROD THERMAL RESPONSE ANALYSIS.................
2-3
2.4 DESCRIPTION
OF THE MODELS...........................
2-5 2.4.1 CPSES-14-LOOP ANF-RELAP NSSS MODEL............ 2-5 2.4.1.1 VOLUMES, JUNCTIONS AND HEAT STRUCTURES..... 2-6 2.4.1.2 CORE POWER.................................... 2-8 O
2.4.1.3 EMERGENCY CORE COOLING SYSTEMS............. 2-9 2.4.1.4 TRIPS AND DELAYS............................. 2-10 2.4.2 CPSES-12-LOOP ANF-RELAP NSSS MODEL........... 2-12
- O v
O
01 J
PAGE 1
el 2.4.3 TOODEE2 MODEL.................................. 2-15 1
3.
BASE CASE ANALYSIS AND SENSITIVITY STUDIES.................
3-1 1
3.1 B ASE CASE ANALYSIS........................................ 3-4 3.2 SENSITIVITY STUDIES........................................
3-10 3.2.1 BREAK SPECTRUM...........
.....................3-10 0
3.2.2[
]
....................3-16 3.2.3 TIME STEP........
3-16 3.2.4 TWO-LOOP VERSUS FOUR-LOOP ANF-RELAP MODELS.....
3-19 0
3.2.5 EXPOSURE STUDY......................................
3-20 4.
CONCLUSION...........
.....................................4-1 4
5.
REFERENCES...................................................
5 - 1 APPENDIX: DESCRIPTION OF COMPUTATIONAL TOOLS.................. A-1 G:
9-8.
G vi G
9)
LIST OF TABLES m
',/
l TABLE PAGE 2.1 CPSES-1 ANF-RELAP Nodalization Summary
. 2-17
'1 2.2 Summary of CPSES-1 ANF-RELAP System Model Components
........ 2-24
\\
2.3 Density Reactivity Table......
2-30 3
2.4 Doppler Reactivity Table
. 2-30 2.5 Scram Reactivity Table..
. 2-31 2.6 ECCS Flow vs. Pressure
. 2-32 3
2.7 Trips and Delays.
2-33 2.8 Fuel Assembly / Rod Data
. 2-34 2.9 Steam Generator Safety Valves Flow Rates 2-34 3.1 Summary of CPSES-1 Small Break LOCA Accident Assumptions for Base Case and Sensitivity Studies.
3-21 3.2 Summary ofInitial Conditions for CPSES-1 Small Break LOCA Base and Sensitivity Studies
. 3-22 3.3 Summary of Fuel Parameters for Base Case Small Break LOCA Analysis.....
3-23 3
3.4 Sequence of Events for Base Case Small Break LOCA...
3-24 3.5 Sequence of Events for Break Spectrum Study.....
3-25 3.6 Sequence of Events for [
]
3-26 9'~
3.7 Sequence of Events for 2-Loop to 4-Loop Comparison 3-27 3.8 PCT Summary for Break Spectrum Study 3-28 0
vii n
wM
O.
LIST OF TABLES (cont'd) e TABLE PAGE 3.9 PCT Summary for [
]...................
3 -2 8 4;
3.10 PCT Summary for Time Step Study................................. 3-29 3.11 PCT Summary 2-Loop Model Validation Study......................... 3-30 3.12 PCT Summary for Exposure Study................................... 3-3 0 g!
4.1 Summary of Results for Base Case and Sensitivity Studies..............
4-5 A.1 Input and Output for the EXEM/PWR Methodology Computer Codes (Refer to Figure A.1 )............................................... A-7 g
e l
9; 0-c' viii G;
O-LIST OF FIGURES O
FIGURE PAGE 2.1 Schematic of EXEM PWR Small break Model......................... 2-35 0
2.2 CPSES-14-loop ANF-RELAP Nodalization Diagram....................
2-36 2.3 CPSES-14-loop ANF-RELAP Nodalization Details.....................
2-37 2.4 CPSES-12-loop ANF-RELAP Nodalization Diagram......
2-38 g
2.5 TOODEE2 Nodalization Diagram.....
........................2-39 3.1 Axial Power Shape for SBLOCA Analyses
...........................3-31 O
3.2 Total Core Power 3 -3 2 3.3 Primary and Secondary System Pressures.............................
3-32 3.4 [
] Region Void Fractions..............................
3-33 0
3.5 [
] Region Void Fractions.....
3-33 3.6 [
] Region Void Fractions................................
3-34
'O 3.7 Upper Plenum Liquid Fractions....................................
3 -3 4 3.8 [
] Collapsed Water Level...............................
3-35 3.9 [
] Clad Temperatures..................................
3-3 5 g
3.10 Loop Seal Void Fractions...................................
3-36 3.11 Accumulator Mass Flow Rates......................................
3-3 6 0
3.12 Total Break Flow Rates...
3 -3 7 3.13 Total ECCS Flow Rates.........................
.................3-37 3.14 TOODEE2 Clad Temperatures for 3 inch Break........................ 3-38 g
ix 0
0:
LIST OF FIGURES (cont'd) e-FIGURE PAGE 3.15 Total Core Power.............
3-38 e:
3.16 Primary and Secondary System Pressures 3-39 3.17 [
] Void Fractions...............
3-39 3.18 [
] Region Void Fiactions...........................
3-40 g:
3.19 [
] Region Void Fractions..
. 3-40 3.20 Upper Plenum Liquid Fractions....................................
3-41 9;
3.21 [
] Collapsed Water Level............................... 3-41 3.22 [
] Clad Temperatures 3-42 3.23 Loop Seal Void Fractions.........................................
3-42 g
3.24 Accumulator Flow Rates.......................
3-43 3.25 Total Break Flow Rate..................
3-43 S'
3.26 Pumped ECCS Injection Flow Rate..................................
3-44 3.27 TOODEE2 Clad Temperatures for 4 inch Break.........................
3-44 3.28 ANF-RELAP Clad Temperatures for 2 inch Break.......................
3-45 g
3.29 ANF-RELAP Clad Temperatures for [
]
S ensitivity Study...............................................
3 -4 5 3.30 ANF-RELAP Clad Temperatures for Six Time Steps for Base Case g;
3 inch Bred <................
...............................3-46 3.31 ANF-RELAP Clad Temperatures for Three Time Steps for 4 inch Break......
3-46 3.32 ANF-RELAP Clad Temperatures for Two Time Steps for 2 inch Break.......
3-47 X
S-
,:0 LIST OF FIGURES (cont'd)
- O FIGURE PAGE 1
3.33 Comparison of ANF-RELAP Clad Temperatures for Two-Loop Versus
- O Fo ur-Loop Model................................................
3-47 j
i 3.34 TOODEE2 Clad Temperatures at BOL and EOC.......................
3-48 A.1 ANF EXEM PWR SBLOCA Computer Code Interfaces................... A-13 lO O
- O
.O
~O
- O
.O xi
.O
. - =
O 1
CHAPTER 1
!0
(
INTRODUCTION 0
The main objective of this work is to obtain approval of TU Electric's application of Siemens Power Corporation's (SPC) methodology - including all codes, all input decks, all results, all inferences and conclusions - so that it may be applied to the Comanche Peak Steam Electric
- O Station Unit One and Unit Two for fuel cycle analyses and to address pertinent licensing issues.
o o
i This report describes TU Electric's application of SPC's USNRC-approved (Reference 1.1)
Emergency Core Cooling Systems (ECCS) Evaluation Model, entitled "EXEM PWR Small 0
I Break Model", to the Comanche Peak Steam Electric Station Unit One (CPSES-1).
O The methodology is used to perform the Small Break LOCA-ECCS licensing analyses that comply with USNRC regulations contained in 10 CFR 50.46,10 CFR 50, Appendix K, and the requirements of NUREG-0737, TMI Action Item II.K.3.30.
O The analyses presented in this report include a description of the current EXEM PWR Small O
Break LOCA methodology (Chapter 2 and Appendix), including the details of the various nodalization. schemes and procedures followed during all phases of the LOCA analyses, which is postulated to occur with the plant in normal operation. Each calculation is performed in O
l-1 0
OI i
compliance with the explicitly approved EXEM PWR Small Break LOCA methodology.
Three principal computer codes are used. RODEX2 provides the initial fuel conditions. ANF-RELAP calculates the system thermal-hydraulic response including core boundary conditions.
TOODEE2 is used to calculate hot rod behavior. Five types ofsensitivity studies are presented in Chapter 3.
The first is a break spectrum study. Breaks ranging from 2 through 4 inches in diameter are O
examined in order to comply with 10 CFR 50.46 (a)(1)(i).
The second type of sensitivity is a time step study. These are performed for each of the cases within the break spectrum, in order to verify that a converged solution has been obtained.
The third type of sensitivity study identifies the bounding [
]. This study is required by the SPC Safety Evaluation Report (SER) (Reference 1.1) and consists of re-analyzing the most limiting break size with the [
O
].
All ANF-RELAP analyses discussed above and presented in Chapter 3 are performed using an explicit representation of the CPSES-14 loops. In the fourth type of sensitivity study, because CPSES-1 has no significant loop asymmetries, an ANF-RELAP model using the 9
industry-wide customary representation, a 2-loop model with a single broken loop and a lumped intact loop, has also been developed and is presented in Chapter 2. In Chapter 3, the e:
1-2 9
iO limiting analysis is repeated using this 2-loop model in order to show that it yields results O
essentially identical to the explicit 4-loop model.
The fifth type of sensitivity is a fuel exposure study. This is done in order to find out whether O
beginning of life (BOL) or end of cycle (EOC) result in lower PCT for the fuel under consideration.
O In chapter 4, key results from base case analyses and sensitivity studies are summarized.
Chapter 4 also summarizes how the most limiting small break LOCA case for the EXEM PWR Small Break methodology is determined, how the PCT is computed, and how compliance with the LOCA-ECCS criteria in 10 CFR 50, Appendix K for CPSES-1 and CPSES-2 is O
demonstrated.
The Appendix provides a description of the codes used in the EXEM PWR Small Break methodology, their interfaces, interrelationships, and respective inputs and outputs.
O
-0 0
1-3
- O CHAPTER 2
'O DESCRIPTION OF THE METHOD O
This report describes the application of USNRC-approved, Siemens Power Corporation's latest ECCS Evaluation model, entitled "EXEM PWR Small Break Model" (Reference 2.1), to the Comanche Peak Steam Electric Station Unit 1.
O The EXEM PWR Small Break methodology is illustrated schematically in Figure 2.1. For O
Presentation purposes the methodology can be said to embody three basic types of calculations:
(1) Determination of Initial Fuel Conditions (RODEX2), (2) System Thermal-Hydraulic Response (ANF-RELAP), and (3) Hot-Rod Thermal Response and Cladding Heatup 0
(TOODEE2). These are discussed in the sections that follow. Additional details of the codes used in these calculations including interfaces, interrelationships, inputs and outputs are o
provided in the Appendix.
2.1 DETERMINATION OF INITIAL FUEL CONDITIONS
.C Calculations are required to determine initial fuel conditions for both ANF-RELAP and TOODEE2. [
O
]. These calculations are performed using the RODEX2 code. This code is also part O
2-1 O
O of the EXEM/PWR methodology (Reference 2.7) currently used in performing large break loss-of-coolant accident analyses that comply with 10 CFR 50.46 and Appendix K thereto.
[
e
].
e.
2.2 SYSTEM THERMAL-HYDRAULIC RESPONSE ANALYSIS The system thermal-hydraulic response is analyzed using ANF-RELAP, a modified version O'
ofRELAP5/ MOD 2.
1 The explicit 4-loop ANF-RELAP system model used for CPSES-1 is described in detail in
- j, Section 2.4.1 and the lumped 2-loop model is described in detail in Section 2.4.2. The initial conditions of the ANF-RELAP fuel rod model,i.e. [
], as mentioned in Section 2.1 and described in the Appendix. The RELAP5/ MOD 2 code is described in detail in Reference 2.2. In addition to less significant changes and corrections, RELAPS/ MOD 2 has been modified in three major 8,
ways to produce ANF-RELAP:
l 1
o l
l 2-2 l
e
1O
[
1O iO
- O l
lO l
l
.]
!O l
The ANF-RELAP calculation provides the thermal-hydraulic boundary conditions for the fuel i
thermal response analysis, which is performed using the TOODEE2 code (Reference 2.5).
O l
[
l l
l 10
].
l l
2.3 HOT ROD THERMAL RESPONSE ANALYSIS
'O TOODEE2 (Reference 2.5) is used to calculate the hot fuel rod heatup during the entire l
accident. It is part of the original WREM package approved by the NRC (Reference 2.6) and l
O is also part of the TU Electric Large Break LOCA methodology (Reference 2.7) currently used 2-3 O
O l
l in performing large break loss-of-coolant accident analyses that comply with 10 CFR 50.46 and Appendix K thereto.
TOODEE2 is a two-dimensional, time-dependent fuel rod thermal and mechanical analysis program. TOODEE2 models the fuel rod as radial and axial nodes with time-dependent heat sources. Heat sources include both decay heat and heat generation via reaction of water with zircaloy. The 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.
6 The outputs of TOODEE2, namely: peak clad temperature, percent local cladding oxidation and percent pin-wide cladding oxidation are compared to the 10 CFR 50.46 (b) (1) through (3) criteria. Regarding (3), if pin-wide oxidation is less than 1% it is concluded that the criteria ofless than 1% core-wide oxidation is met, e-
[
ei e-
.]
e l
2-4 e.
')
2.4 DESCRIPTION
OF TIIE MODELS J
2.4.1 CPSES-14-LOOP ANF-RELAP NSSS MODEL The Comanche Peak Steam Electric Station has two Westinghouse pressurized water reactors.
Both units are typical 4-loop plants with a rated thermal power of 3411 MWt each.
)
l The CPSES-1 ANF-RELAP NSSS model reflects a considerable amount of engineering J
insight and experience and incorporates:
a.
Information from the most recent plant drawings, design basis documents, vendor
~)
documents, Technical Specifications and Final Safety Analysis Report.
3 1
- b. Careful consideration of the guidelines set forth by SPC for the application of their methodology (Reference 2.8).
D This section describes the explicit 4-loop version of the ANF-RELAP base input model for the Comanche Peak Steam Electric Station Unit 1 (CPSES-1). The discussion of this model is C) divided into the following sub-sections:
l.
Volumes, junctions and heat structures 2.
Core power 3.
Emergency core cooling systems
'O 4.
Trips and delays 2-5 c
O, l
l Since there are no significant loop asymmetries only loop-l will be discussed in this section, i.e., where the corresponding information for the other three loops is the same, the redundant information is omitted.
o 2.4.1.1 VOLUMES, JUNCTIONS AND HEAT STRUCTURES Figure 2.2 shows the CPSES-1 nodalization diagram for the ANF-RELAP entire explicit 4-8 loop base input model. Table 2.1 identifies the volumes, junctions, and heat structures associated with the reactor vessel, pressurizer, loop-l and other systems. It also prosides node numbers for cross-referencing with Figure 2.3, which, in order to allow better visibility, shows S.
only loop-l and the reactor vessel. Table 2.2 summarizes key parameter values for the reactor vessel, pressurizer, loop-l and other systems of the CPSES-1 ANF-RELAP NSSS explicit 4-loop base input model.
[
S:
S S
e-2-6 e
~. -..
l) l l
l
?
l l
l l
!)
l Steam generator models include both primary and secondary sides. An appropriately detailed 1
i nodalization of the steam generator secondary has been implemented in order to insure realistic l
heat transfer behavior across the steam generator tubes.
l D.
l Steam generator pressure reliefis obtained by simulation of the safety valves only (Table 2.9),
i.e. no credit is taken for the heat removal capability of the steam dump and bypass system nor the atmospheric relief valves. Five percent of the steam generator tubes are assumed plugged.
This assumption is required by the methodology. It is made in these analyses in order to support the potential need for operation under these circumstances and is a conservative l
assumption for fewer obstructed tubes.
o l
Reactor coolant pumps are modeled using Westinghouse homologous curves in the single phase regime combined with homologous difference and multiplier curves for the CE-EPRI l
tests in the two-phase regime. The CE-EPRI reactor coolant pump data were reviewed and i
t j
found to be applicable to CPSES reactor coolant pumps.
- O 2-7 O
=
O The containment is represented by a time-dependent volume (TMDPVOL) with constant O
atmospheric pressure.
2.4.1.2 CORE POWER The total core power during transients is determined by the point reactor kinetics model in ANF-RELAP. Conservative input data are entered for this model in order to compute the fission power with a 1.02 multiplier and decay heat with a 1.2 multiplier, per 10 CFR 50, Appendix K requirements. The model accounts for the reactivity effects associated with scram, change in moderator density and in fuel temperature. The effects are evaluated on a core average, cycle specific basis using the reactor physics methodology and associated uncertainty factors presented in Reference 2.9 to assure conservatism. For the analyses presented herein, reactivity feedbacks representative of the CPSES-1 core for cycle 5 have been selected and are shown in Tables 2.3,2.4 and 2.5 for moderator density effects, fuel temperature effects and scram, respectively.
All core power is conservatively assumed to be generated in the fuel, i.e. none is deposited in moderator, cladding, or passive heat structures. This power is distributed according to the nodal power factor (NPF) entered for each active heat structure that represents a portion of UO fuel. End of cycle convertion ratios are used to maximize actinide decay heat.
2 9
9 2-8 9
l 0 l
2.4.1.3 EMERGENCY CORE COOLING SYSTEMS O
The CPSES ECC system is arranged into four subsystems: (1) the charging / safety injection, (2) high head safety injection, (3) low head residual heat removal injection, and (4) accumulators.
O 1
l There are two safety injection trains. Each train contains one centrifugal charging pump, one O
high head safety injection pump, and one low head residual heat removal pump with associated piping, valves, controls, and instrumentation.
O Loss of offsite power is assumed to occur coincidentally with the reactor trip. One diesel l
generator train is removed on the assumption that it (one train) fails to start. Therefore, only lO one train of safety systems are represented in the present NSSS model. This assumption is made in order to satisfy the single failure criterion, as discussed in Chapter 3.
!O All pumped systems take suction from the refueling water storage tank (RWST) during the injection phase. In the present analyses the RWST watec temperature is taken at the maximum lO value (120 degrees F) allowed by the Technical Specifications. This is conservative since it minimizes heat removal by sensible heat transfer to injected fluid.
lO The pumped ECCS mass flow rates for each loop, versus pressure, for each injection system, which are given in Table 2.6, were derived from the values given in Reference 2.12.
O I
2-9 l
!O l
l
O!
The system contains four accumulators, one per loop. The minimum Tecimical Specifications (Reference 2.11) tank water volume (6119 gals. per tank) is used. The accumulators are modeled using a two-volume PIPE component (as opposed to the ACCUM component), per SPC methodology.
9; 2.4.1.4 TRIPS AND DELAYS
'O The following trips and delays are used:
1.
Reactor trip occurs on a low pressurizer pressure signal (1860 psia) plus a 2 O:
second delay for signal processing. The 2.4 second rod travel time is accounted i
for by the scram reactivity (Table 2.5).
Oi 2.
The reactor coolant pumps (RCP) are tripped at reactor trip, as discussed in Chapter 3. This trip occurs because at reactor trip offsite power is assumed lost.
The RCPs cannot operate after a reactor trip if offsite power is not available.
3.
Steam flow isolation is initiated at the time of reactor trip with the turbine stop 01 valves taking 0.5 seconds to close (Table 2.7) following a 2 second signal processing delay. The steam dump and bypass system and the atmospheric relief valves are not credited. The safety valves operate as shown in Table 2.9.
8!
Ol 2-10 9)
O 4.
Main feedwater isolation begins at the time of"S" signal plus a 7 second delay O
which includes 2 seconds for signal processing (Table 2.7).
l 5.
St Actuation Signal occurs on a low pressurizer pressure "S" signal (1715 psia)
O plus 2 seconds.
.O 6.
The delays for each of the pumped safety injection systems are given in Table 2.7.
l' 7.
Accumulators inject at set pressure (603 psia) without delay.
lO 8.
Available nuxiliary feedwater (1 motor-driven pump) is assumed to be up and l0 running 60 seconds after the "S" signal. " Cold" AFW injection is delayed for 1
)
another 140 seconds, conservatively accounting for the flow travel time down the piping. During these 140 seconds, AFW is delivered at the higher main feedwater
- O temperature.
One motor-driven AFW pump is assumed lost due to the unavailability of offsite power, compounded with the failure of one diesel O
generator to start (single failure). Turbine-driven auxiliary feedwater (TDAFW) is not credited because it is difficult to demonstrate that it would be automatically activated on a steam generator Lo-Lo level signal prior to quenching of the fuel.
,O However, preliminary calculations show that TDAFW shifts the most limiting break to a larger size (6 inches) and results in lower peak clad temperatures by
- O approximately 250 F. Thus, although the TDAFW is not considered in any of the 2-11 0
1 I
O present analyses, it might be in future analyses, if adequate justification for its 9-availability can be demonstrated.
2.4.2 CPSES-12-LOOP ANF-RELAP NSSS MODEL The purpose of developing the 2-loop model is simply to obtain the same results as the 4-loop model but in considerably less computation time. Therefore, the 2-loop model duplicates the O
4-loop model in every possible way.
The 2. loop version of the ANF-RELAP CPSES-1 NSSS model is derived from the 4-loop version in a direct manner: three of the four loops are assumed to be identical and are modeled as one lumped loop with appropriately scaled input. The pressurizer is connected to the intact lumped loop following usual modeling practices. The lumped loop represents the " unbroken"
- l l
or " intact" loops. The " broken" loop remains the single loop-l described in Section 2.4.1 and shown in Figure 2.3. Figure 2.4 shows the nodalization diagram for the entire 2-loop model.
The data in Tables 2.1 and 2.2 also apply to this model except for the lumped loop data, which are not listed in order to avoid redundancy. The loop lumping is done using the standard modeling practice for deriving the lumped loop input from the single loop input as summarized below:
9:
(1)
Component flow areas are three times the area of the corresponding single loop component.
S 2-12 S-
lO (2)
Component lengths are identical in the lumped loop and in the corresponding l
O single loop component.
(3)
Component fluid volumes are three times the volume of the corresponding single j
loop component.
0 (4)
Azimuthal angles remain zero.
(5)
Inclination angles are identical in the lumped loop and in the corresponding single
.O loop component.
O (6)
Elevation changes are identical in the lumped loop and in the corresponding single loop component.
I O
(7)
Wall roughnesses are identical in the lumped loop and in the corresponding single loop component.
0 (8)
Hydraulic diameters are identical in the lumped loop and in the corresponding single loop component.
(9)
Control flags are identical in the lumped loop and in the corresponding single loop O
component.
2-13 O
0:
(10) Initial Conditions are identical in the lumped loop and in the corresponding single loop component.
l (11) Junction flow areas are three times the area of the corresponding single loop junctions.
l O
l (12) Forward and reverse loss coefficients are identical in the lumped loop and in the corresponding single loop junctions.
O.
l (13) Junction flags are identical in the lumped loop and in the corresponding single l
loop junctions.
l O'
l (14) Junction initial mass flow rates are three times the flow rates of the corresponding l
single loopjunctions, whereas velocities are the same.
(15) The surface area factors for the heat conductors are three times the area factors of 8
the corresponding single loop heat conductors.
(16) All control system parameters are identical in the lumped loop and in the 0:
corresponding single loop.
O 2-14 f
O
'O All features of the 4-loop model described in Section 2.4.1, including core power, emergency O
core cooling systems, trips and delays, are preserved in the 2-loop model and therefore need not be repeated.
O 2.4.3 TOODEE2 MODEL TOODEE2 is used to calculate the temperature distribution in the hot rod. Table 2.8
.O summarizes the fuel geometry data used in the TOODEE2 model.
O O
j The first and last axial nodes are identified as the bottom and top of the fuel rod, respectively.
The TOODEE2 hot ro-d axial nodalization diagram is shown in Figure 2.5.
- O
[
O J
O 2-15 O
O The thermal-hydraulic boundary conditions for the TOODEE2 calculations are those associated with the [
] region of the ANF-RELAP model, as described in Section 2.1 and in the Appendix.
G:
[
e:
)
i 4:
i O!
t l
e:
l l
Ol eI O
i 2-16 9
O TABLE 2.1 CPSES-1 ANF-RELAP NSSS NODALIZATION
SUMMARY
Region Noding Number of Number of Number of
-OR-Diagram Volumes Junctions Heat Structures O
System Number Active TMDP Active TMDP Active Passive REACTOR VESSEL (RV)
RV DOWNCOMER(DC) i O
o Bottom oc 108 5
0 4
0 0
5 RV LOWER PLENUM (LP) o Bottom LP 109 1
0 0
0 0
1 O
o Middle LP 110 1
0 3
0 0
1 o Top tP 111 1
0 4
0 0
1 RV CORE BYPASS & BARREUBAFFLE (BYPASS)
O o Bottom Bypass 128 3
0 2
0 0
3 RV CORE ACTIVE FUEL REGION O
O O
2-17 O
O TABLE 2.1 (Cont'd)
CPSES-1 ANF-RELAP NSSS NODALIZATION
SUMMARY
Region Noding Number of Number of Number of
-O R-Diagram Volumes Junctions Heat Structures Active TMDP Active TMDP Active Passive REACTOR VESSEL (RV)(cont'd)
~
O O
- l M
O O
O-O 2-18 O
O TABLE 2.1(cont'd)
O CPSES-1 ANF-RELAP NSSS NODALIZATION
SUMMARY
Region Noding Number of Number of Number of
-OR-Diagram Volumes Junctions Heat Structures O
Active TMDP Active TMDP Active Passive REACTOR VESSEL (RV)(cont'd)
(~
- O i
O M
M M
M lO O
O O
O 2-19 0
01 TABLE 2.1(cont'd)
CPSES-1 ANF-RELAP NSSS NODALIZATION
SUMMARY
Region Noding Number of Number of Number of
-O R-Diagram Volumes Junctions Heat Structures System Number Active TMDP Active TMDP Active Passive g:
REACTOR VESSEL (RV)(cont'd) 9:
~
~
4; UPPER PLENUM & GUlDE TUBES (UP,GTs) o Bottom UP 166 1
0 6
0 0
1 o Guide Tubes 170 1
0 0
0 0
1 o Middle UP 1 173 1
0 1
0 0
1 9
o Middle UP 11 174 1
0 5
0 0
1 o Top UP 178 1
0 0
0 0
1 UPPER HEAD (UH) e o Bottom UH 180 1
0 0
0 0
1 o Middle UH 181 1
0 5
0 0
1 o Top UH 182 1
0 0
0 0
1 LOOP-1 PRIMARY g
LOOP-1 HOT LEG (L1 HL) oL1HL#1 410 1
0 2
0 0
1 o Li HL's # 2&3 414 2
0 1
0 0
2 0
o L1 SG Inlet &
422 1
0 3
0 0
0 Outlet Plena 426 1
0 2
0 0
0 0
2-20 S
O TABLE 2.1 (Cont'd)
O CPSES-1 ANF-RELAP NSSS NODALIZATION
SUMMARY
Region Noding Number of Number of Number of
-OR-Diagram Volumes Junctions Heat Structures T"""
- **i
I"""
O LOOP-1 PRIMARY (cont'd)
L1 STEAM GENERATOR (SG)
IL So U-Tubes 424 8
0 7
0 0
8 O
L1 CROSS-OVER LEG (XLG) oLIXLG 460 4
0 4
0 0
4 O
L1 REACTOR COOLANT PUMP (RCP) oLIRCP 475 1
0 2
0 0
0 0
Lt COLD LEG (CL) o lL CL # 1 480 1
0 1
0 0
1 RCP Side o IL CL # 2 490 1
0 1
0 0
1 O
Middle o IL CL # 3 495 1
0 0
0 0
1 RV Side LOOP-1 SECONDARY O
L-1 MAIN AND AUXILIARY FEEDWATER(MFW & AFW) o L1 MFW Source 502 0
1 0
0 0
0 o L1 MFW Flow 506 0
0 0
1 0
0 o L1 AFW Source 520 0
1 0
0 0
0 o L1 AFW Flow 525 0
0 0
1 0
0 0
2-21
.O
O TABLE 2.1 (Cont'd)
CPSES-1 ANF-RELAP NSSS NODALIZATION
SUMMARY
)
Region Noding Number of Number of Number of
-OR.
Diagram Volumes Junctions Heat Structures System Number Active TMDP Active TMDP Active Passive g
LOOP-1 SECONDARY (Cont'd)
L1 SG Vessel o L1 SG Downcomer
$10 4
0 3
0 0
0 0
o Ll DC to Boiler 515 0
0 1
0 0
0 o L1 SG Boiler 540 5
0 4
0 0
0 o L1 SG Separator 560 1
0 3
0 0
0 gi o L1 SG Steam Dome 570 1
0 0
0 0
0
/
L1 SG STEAM LINE AND SAFETY VALVES o L1 Steamline 575 0
0 0
1 0
0 o L1 Steam Sink 580 0
1 0
0 0
0 o L1 Safety Valve 585 0
0 0
1 0
0 0
o L1 S.V. Steam 590 0
1 0
0 0
0 Sink PRESSURIZER (PRZR -is connected to LOOP-4 Hot Leg in explicit 4-loop model)
O o PRZR Surge-Line 603 0
0 1
0 0
0 o PRZR Surge-Line 610 1
0 1
0 0
0 o PRZR Tank 620 6
0 5
0 0
6 O
ACCUMULATORS (ACCUM) o L1 ACCUM 720 2
0 1
0 0
0 0
2-22 9
O TABLE 2.1 (Cont'd)
O CPSES-1 ANF-RELAP NSSS NODALIZATION
SUMMARY
Region Noding Number of Number of Number of
-OR-Diagram Volumes Junctions Heat Structures O
Active TMDP Active TMDP Active Passive ACCOMULATORS (ACCOM) (cont' d) o L1 ACCUM Surge-735 0
0 1
0 0
0 Line & Flow 730 1
0 0
0 0
0 EMERGENCY CORE COOLING SYSTEM (ECCS)
1 0
0 0
0 o Ll CCP Source 750 0
1 0
0 0
0 O
o Ll CCP Flow 770 0
0 0
1 0
0 o IL HHSI Flow 775 0
0 0
1 0
0 BREAK & CONTAINMENT O
BREAK o Break Junction 805 0
0 1
0 0
0 Valve O
CONTAINMENT o Containment 810 0
1 0
0 0
0 0
O l
1 2-23 0
Table 2.2
SUMMARY
OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS Component Area length Volume Inclimation Elevation Surface D,
Fings Numberfrype (ft')
(ft)
(ft')
(Degrees)
Change (ft)
Roughness (ft) 100 BRANCII 1333 5.85 77.96
-90.0
-5.85 0.00 1.53 100 102 BRANCil 1333 5.85 77.96
-90.0
-5.85 0.00 1.53 100 104 BRANCH 1038 2.29 23.78
-90.0
-2.29 0.00 0.98 100 106 BRANCII 1038 2.29 23.78
-90.0
-2.29 0.00 0.98 100 108-1 ANNULUS 33.24 3.93 133.89
-90.0
-3.93 0.00 1.57 100 108-2 ANNULUS 32.42 4.00 129.68
-90.0
-4.00 0.00 1.45 100 108-3 ANNULUS 32.42 4.00 129.68
-90.0
-4.00 0.00 1.45 100 108-4 ANNULUS 32.42 4.00 129.68
-90.0
-4.00 0.00 1.45 100 108-5 ANNULUS 32.42 4.05 136.01
-90.0
-4.05 0.00 1.42 100 109 SNGLVOL 47.44 2.96 140.41 90.0 2.96 0.00 7.77 000 110 BRANCH I12.92 2.96 33425 90.0 2.96 0.00 11.99 000 Iii BRANCII 95.91 4.23 405.7I 90.0 4.23 0.00 11.05 000 128-1 ANNULUS 25.08 4.00 10031 90.0 4.00 0.00 0.008 000 128-2 ANNULUS 25.08 4.00 10031 90.0 4.00 0.00 0.008 000 128-3 ANNULUS 25.08 4.00 10031 90.0 4.00 0.00 0.008 000 2-24 I
O e
G G
G S
8 8
8 8
l O
O O
O O
O O
O O
O O
i 1
Table 2.2 (cont'd)
SUMMARY
OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS i
i I
t Component Area Length Volume inclination Elevation Surface D,
Fings I
Numberffype (ft')
(ft)
(ft')
(Degrees)
Change (ft)
Roughness (ft) j I
i e
I l
t I
I E
r
?
k k
amu 2-25 h
i
Table 2.2 (con ('a)
SUMMARY
OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS Comipoment Area Length Volume Indiention Elevation Surface D
Rags y
Number / Type (ft')
(ft)
(ft')
(Degrees)
Change (ft)
Roughness (ft) i f
I 166 BRANCH 84.84 1.28 108.60 90.0 1.28 0.00 0.0399 000 170 SNGLVOL 16.84 13.29 223.78 90.0 12.98 0.00 0.25 100 173 BRANCH 84.84 2.405 204.04 90.0 2.405 0.00 1039 000 174 BRANCH 84.84 2.42 20531 90.0 2.42 0.00 1039 100 178 SNGLVOL 84.84 2.15 182.41 90.0 2.15 0.00 1039 100 180 SNGLVOL 81.72 2.96 241.88 90.0 2.96 0.00 10.20 100 181 BRANCH 125.48 237 297.38 90.0 237 0.00 12.64 100 2-26 I
O e
e e
e e
e e
e e
e
O O
O O
O O
O O
O O
O Table 2.2 (cont 4)
SUMMARY
OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS Component Ares 12egth Volume Inclination Devation Surface D
Mags y
NumberfType (ft')
(ft)
(ft')
(Degrees)
Chante (ft)
Roughness (ft) 182 BRANCH 80.26 4.53 363.60 90.0 4.53 0.00 10.11 100 410 BRANCil 4.58 5.20 23.82 0.0 0.00 1.5E-4 2.42 000 414-1 PIPE 4.59 8.88 40.77 0.0 0.00 1.5E-4 2.42 000 414-2 PIPE 4.59 7.59 34.84 0.0 0.00 1.5E-4 2.42 000 422 BRANCII 21.17 7.91 167.47 90.0 7.91 1.5E-4 5.19 000 424-1 PIPE 10.46 9.06 94.77 90.0 9.06 5.0E-6 0.0553 000 424-2 PIPE 10.46 7.25 75.84 90.0 7.25 5.0E-6 0.0553 000 424-3 PIPE 10.46 7.25 75.84 90.0 7.25 5.0E-6 0.0553 000 424-4 PIPE 10.46 4.44 46.44 90.0 4.44 5.0E-6 0.0553 000 424-5 PIPE 10.46 4.44 46.44
-90.0
-4.44 5.0E-6 0.0553 000 424-6 PIPE 10.46 7.25 75.84
-90.0
-7.25 5.0E-6 0.0553 000 424-7 PIPE 10.46 7.25 75.84
-90.0
-7.25 5.0E-6 0.0553 000 424-8 PIPE 10.46 9.06 94.77
-90.0
-9.06
.5.0E-6 0.0553 000 426 BRANCH 21.17 7.91 167.47
-90.0
-7.91 1.5 E-4 5.19 000 460-1 PIPE 5.24 7.67 40.19
-90.0
-5.81 1.5E-4 2.58 000 460-2 PIPE 5.24 7.07 37.05
-45.0
-4.50 1.5E-4 2.58 000 460-3 PIPE 5.24 3.52 18.44 0.0 0.00 1.5E-4 2.58 000 4604 PIPE 5.24 7 07 37.05 45.0 4.50 1.5E-4 2.58 000 475 PUMP 10.68 7.36 78.60 90.0 5.81 00 2-27
Table 2.2 (cont'a)
SUMMARY
OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS Component Area Length Volume inclimation Elevation Surface D
Hags 2
y Numberfrype (ft )
(ft)
(ft')
(Degrees)
Change (ft)
Roughness (ft) 480 BRANCH 4.12 7.14 29.42 0.0 100 1.5E-4 2.29 000 490 BRANCH 4.12 15.76 64.96 0.0 0.00 1.5E-4 2.29 000 495 BRANCH 4.12 2.20 9.064 0.0 0.00 1.5E-4 2.29 000 502 TMDPVOL 100.00 10.00 1000.00 0.0 0.00 0.00 0.00 100 504 TMDPVOL 100.00 10.00 1000.00 0.0 0.00 0.00 0.00 100 510-1 PIPE 194.84 7.85 1529.53
-90.0
-7.85 1.5E-4 15.72 100 510-2 FIPE 5.67 7.25 41.11
-90.0
-7.25 1.5E-4 034 100 510-3 PIPE 5.67 7.25 41.11
-90.0
-7.25 1.5E-4 034 100 510-4 PIPE 5.67 9.06 5137
-90.0
-9.06 1.5E-4 034 100 540-1 PIPE 5432 9.06 492.18 90.0 9.06 1.5E-4 0.0972 000 540-2 PIPE 55.48 7.25 402.21 90.0 7.25 1.5E-4 0.0972 000 540-3 PIPE 55.48 7.25 402.21 90.0 7.25 1.5E-4 0.0972 000 540-4 PIPE 51.99 4.44 230.84 90.0 4.44 1.5E-4 0.0972 000 540-5 P1PE 8733 3.41 297.79 90.0 3.41 1.5E 10.54 000 560 SEPARATR 47.77 23.74 1133.99 90.0 23.74 0.00 7.80 100 570 SNGLVOL 127.86 9.73 1244.05 90.0 9.73 1.5E-4 12.76 100 2-28 O
e e
o e
e e
e e
e e
i
o o
O O
O O
O O
U u
u Table 2.2 (cont'd)
SUMMARY
OF CPSES-1 ANF-RELAP SYSTEM MODEL COMPONENTS Component Area length Volume inclination Elevation Surface D,
flags Numberfrype (ft')
(ft)
(ft')
(Degrees)
Change (ft)
Roughness (ft) 580 TMDPVOL 100.00 10.00 1000.00 0.0 0.00 0.00 0.00 100 590 TMDPVOL 100.00 10.00 1000.00 0.0 0.00 0.00 0.00 100 610 BRANCH 0.683 67.49 46.10 90.0 27.89 0.00 0.683 000 620-1 PIPE 36.58 7.627 279.00 90.0 7.627 0.00 6.82 000 620-2 PIPE 36.58 7.627 279.00 90.0 7.627 0.00 6.82 000 620-3 PIPE 36.58 7.627 279.00 90.0 7.627 0.00 6.82 000 620-4 PIPE 36.58 7.627 279.00 90.0 7.627 0.00 6.82 000 620-5 PIPE 36.58 9.704 354.97 90.0 9.704 0.00 6.82 000 620-6 PIPE 36.58 9.704 354.97 90.0 9.704 0.00 6.82 000 720-1 PIPE 22.93 1.50 34.390 90.0 1.50 0.00 5.40 000 720-2 PIPE 80.46 16.35 1315.61 90.0 16.35 0.00 5.40 000 730 SNGLVOL 0.418 81.12 33.91
-90.0
-10.17 0.00 0.73 100 745 TMDPVOL 1.00 10.00 10.00 90.0 10.00 0.00 0.00 01 750 TMDPVOL 1.00 10.00 10.00 90.0 10.00 0.00 0.00 01 810 TMDPVOL 100.00 1.00 100.00 0.0 0.00 0.00 0.00 100 2-29
O TABLE 2.3 e:
DENSITY REACTIVITY TABLE (p=0.0044) 3 DENSITY (Ibm /ft )
REACTIVITY (S) g 0.62
-185.331 6.24
-76.88 12.49
-34.001 18.73
-8.68 24.97 0.619 31.21 1.219 37.46 0.469 42.10 0.00 43.70
-0.161 46.14
-0.661 49.94
-1.461 62.43
-3.751 0.
TABLE 2.4 I
DOPPLER REACTIVITY TABLE (p=0.0044)
TEMPERATURE ('F)
REACTIVITY (S) 400.0 3.481 650.0 3.481 800.0 2.841 1000.0 2.051 3
1200.0 1.311 1400.0 0.612 1450.0 0.441 1585.0 0.000 1600.0
-0.049 g
1800.0
-0.679 2000.0
-1.289 2200.0
-1.879 2400.0
-2.449 2600.0
-3.009 g
2800.0
-3.549 3000.0
-4.079 2-30
lO
- O TABLE 2.5 SCRAM REACTIVITY TABLE (p=0.0044) l O
TIME (SECONDS)
REACTIVITY (S) 0.00 0.00 0.48 0.00 0.96
-0.053 1.44
-0.133 0
1.92
-0.400 2.16
-0.800 2.40
-1.813 2.64
-3.413 2.88
-4.800 O
3.12
-5.120 3.36
-5.227 3.60
-5.333 1.E6
-5.333 O
O O
.O O
2-31
- O
1 O
i i
i j
TABLE 2.6 8-i ECCS FLOW FOR EACH LOOP VS. PRESSURE RCS CCP HHSI g
PRESSURE (lbm/sec)
(lbm/sec)
(psig) 0 11.27 18.42 100 10.96 17.75 120 10.89 17.61 8-200 10.64 17.05 300 10.33 16.32 400 10.00 15.57 500 9.68 14.75 600 9.34 13.89 8
700 9.00 12.97 800 8.65 12.02 900 8.23 10.98 1000 7.90 9.68 1200 7.11 6.89 1300 6.71 5.05 1400 6.30 1.98 1500 5.87 0.0 1600 5.34 0.0 1800 4.02 0.0 2000 2.26 0.0 2100 1.25 0.0 0:
O' O
2-32 0
~..
10
'O TABLE 2.7 TRIPS AND DELAYS O
ACTION TRIPS DELAYS (Sec) o Lo-Przr Pressure signal RCS @ 1860 psia o Reactortrip Lo-Przr signal 2.0 o "S" signal PRZR @ 1715 psia o Reactor Coolant Pumps trip Reactor Trip C
o Main Steam Isolated Reactor Trip 2.5 o Main FeedwaterIsolated "S" signal 7.0 o Auxiliary Feedwater Injects "S" signal 200.0
_v o SI Actuation Signal "S" signal 2.0 o Charging PumpInjects "S" signal 17.0 O
o HPSI Pump Injects "S" signal 22.0 o AccumulatorsInject RCS @ 603 psia O
.O O
2-33 O
O TABLE 2.8 FUEL ASSEMBLY / ROD DATA I
PARAMETER VALUE e
j o Outer Diameter of Fuel Rod 0.360 in o Active Fuel Height 144.0 in o No. of Fuel Assemblies 193 9
o No. of Fuel Rods /Assy 264 o No. of Guide Thimbles /Assy 24 o No. ofInstr. Tubes /Assy 1
o Cladding Thickness 0.025 in 8
o Diametral Gap 0.0065 in O
TABLE 2.9 STEAM GENERATOR SAFETY VALVES FLOW RATES Secondary Pressure (psia)
SV Flow Rate (1bm/sec) 0.0 0.0 1200.0 0.0 1236.0 0.0 1236.1 248.1 8l 1246.3 248.1 1246.4 498.3 1256.6 498.3 1256.7 750.5 1266.9 750.5
- l 1267.0 1004.8 1287.5 1004.8 1287.6 1263.2 2000.0 1263.2 4
2-34 6
i I
f I
I l
i f
i i
i t
l
(
l
)
Figure 2.1 Schematic of TU Electric's Small Break Model l
I l
O u., a t..,
Fl l
l
.:.-, "n... T **
n..J.
l=l l
l
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l
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l~l 8
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44b3 4
-S ru rm 3
w-L,
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L..a 1 Leep 2 i
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g g m.m c m
Figure 2.2: ANF-RELAP 4-Loop SBLOCA O %% rw v Nodolization Diagram Q r-tw va rw m.
MW.mw.:w m7#
8
O 0
~ y w /7,/
575 rd M
Loop 1 182 685 p
m f _ _ _ _ _.-- - -
'1' 151 1
_ _ _ _ p i._
t i
1-170 l
l 37, 1
Proprietary:
1
~
Sto-t l }510-1 RV Upper l
374 l
& Widdle i
Downcomer
\\\\
//
O
[
its l
7 ig e
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Q sie I
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17.g48 I
0 m'
I f 735 7
7 Proprietary:
/
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3 I
RV Core Active Fuel Region i
3 I
3 4 475 o
8 I
4 1
4 as 8
4so.3 n
111 O
Hect Conductors i
i l
l Primary / Secondary Fluid Volumes I
Time-dependent Flutd Volumes Fluid Junctions O
'N
/
Figure 2.3: Enlorged Vlow of Reactor Yessel
\\
/
and Loop 1 Nodollzotion Ologram j
k\\h4 /[
O
O' M
e O,
el e
s m
ie t
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f8
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R i d M
z x
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i i
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28 ST e e, a
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4 zo 8
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,,,d,
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re t
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-.-...s-..
e S
s
=
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4:
110 ll 1
l H
H O
4
!O
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O
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._M M
!o Figure 2.5:
T00DEE2 Nodalization Diagram
,,)
'~
CHAPTER 3 BASE CASE ANALYSIS AND SENSITIVITY STUDIES O
Small break loss-of-coolant accident analyses frequently require the investigation of the impact of variations in several method-and plant-specific issues on the LOCA O
consequences.
O Method-specific issues are suggested throughout 10 CFR 50.46, Appendix K thereto, and in NUREG-0737 II.K.3.30, and are addressed in Reference 2.1. The present work constitutes TU Electric's application of SPC's approved Evaluation Methodology (EM),
O using method-specific parameters as prescribed by the method developers (Reference 2.8).
Hence, the effect of variations in method-specific parameters within the bounds of O
methodology recommendations has already been ascertained in Reference 2.1 and sensitivity studies for these variables need not be repeated here.
O There are, nevertheless, three exceptions: (1) a [
] is mandated in Reference 1.1 and, (2) a time step study conducted even O
though the threshold for this requirement per Reference 1.1, was not reached. In addition, (3) because CPSES has no significant loop asymmetries, an ANF-RELAP model using the more customary representation, where the intact loops are lumped leading to a 2-loop O
representation, is demonstrated, in another sensitivity study, to yield essentially identical 3-1 0
O!
4 i
results to the explicit 4-loop model. It is TU Electric's intention to utilize this 2-loop el l
model in future analyses while CPSES continues to show no significant loop asymmetries.
e, The plant-specific issues which warrant investigation are given in the following passages from 10 CFR 50.46, Appendix K thereto and NUREG-0611, along with the approach taken in addressing each one.
gl 1
i i
10 CFR 50.46 (a)(1)(i), requires that "a number of postulated loss-of-coolant accidents of 9l different sizes, locations and other properties" be calculated in sufficient amount "to provide assurances that the most severe postulated loss-of-coolant accidents are calculated." In compliance with this requirement, a break spectrum study has been gl conducted.
el Although higher peak clad temperatures (PCT) are usually associated with beginning of life (BOL) fuel because of the higher stored energy, a fuel burnup study is also conducted.
This is done in order to confirm that the end of cycle (EOC) pin pressures, which are
,l higher than those encountered early in life and which foster a higher driving force for rod burst, do indeed result in lower PCT for the fuel under consideration.
9:
10 CFR 50, Appendix K, Part I, A, (1) states: "A range of power distribution shapes and peaking factors representing power distributions that may occur over the core lifetime shall be studied and the one selected should be that which results in the most severe 3-2 e:
i D
i l
l 9
calculated consequences for the spectrum of postulated breaks and single failures analyzed."
i i
[
i l
O o
i lO l
}
o 10 CFR 50, Appendix K, Part I, D, (1) states: "an analysis of possible failure modes of ECCS equipment and their effects on ECCS performance must be made. In carrying out 0
the accident evaluation, the combination of ECCS subsystems assumed to be operative l
shall be those available after the most damaging single failure of ECCS equipment has 1;0 taken place." The limiting single failure for the small break loss-of-coolant accident analyses in the CPSES-1 & 2 FSAR has been determined by the NSSS vendor (Reference O
3.1). It is the loss of one ECCS injection train. Unless a common cause is established, the loss of one ECCS injection train involves multiple failures of ECCS equipment and 3-3
..O
O' therefore is not a single failure. The required common cause is the loss of power to the g
train. In order to arrive at this condition consistently, it must also be assumed that both the preferred 345 KV and the altemate 148 KV offsite power sources are lost and that one 8
emergency diesel generator fails to start. Hence, the most damaging single failure of ECCS equipment postulated for the present study is the failure of an emergency diesel generator to start. Offsite power (which is net ECCS equipment) unavailability is e
postulated in order to make the single failure meaningfull, i.e. the diesel generator is not needed if either the preferred 345 KV or the alternate 148 KV offsite power sources are 9I available. Thus, one motor driven auxiliary feedwater pump, one high head centrifugal charging pump, one intermediate head safety injection pump and one low head residual heat removal (RHR) pump (which is not challenged in these analyses) as well as all four
,j accumulators are available to mitigate the accident and are credited in all the calculations.
9' One additional conservatism is incorporated into all of the calculations in this work. That conservatism is that five percent of the steam generator tubes are assumed plugged. This assumption is made to support the potential need for operation under such circumstances and is a conservative assumption when fewer tubes are actually obstructed.
3.1 BASE CASE ANALYSIS This section presents licensing analysis results for a 3.0 inch diameter break in the discharge line of the Reactor Coolant Pump. The axial power shape used for this base case is that determined as described in Section 3.0 as most limiting and is shown in Figure 3-4 O
O O
3.1. The fuel rod exposure which maximizes stored energy is calculated by RODEX2 and i
occurs at 605 hours0.007 days <br />0.168 hours <br />0.001 weeks <br />2.302025e-4 months <br /> for the [
] and 1214 hours0.0141 days <br />0.337 hours <br />0.00201 weeks <br />4.61927e-4 months <br /> for the
[
). Fuel parameters used in this base case are consistent with this exposure, j
O The accident assumptions are summarized in Table 3.1 and the initial conditions are l
O summarized in Table 3.2. Key fuel rod parameters are summarized in Table 3.3. Table 3.4 summarizes the timing of significant events for this base case.
]
O
[
'O
]. When the neutronics models are activated at the start of the transient, there is a small reactivity imbalance which causes the power to rise. This effect is not significant because it is a slight linear increase (0% to 5%) over a short period of time (25 to 30 seconds). Furthermore, the effect is in the conservative direction and is therefore c nsidered acceptable. Following this brief ramp, the power is soon seen (Figure 3.2) to O
drop off rapidly, at the time of reactor trip, due to the negative reactivity associated with control rod insertion. Reactor trip is activated by the pressurizer low pressure signal.
O After that, reactor power tapers off according to decay heat.
Figure 3.3 shows the primary and the secondary pressures and is used as a road map in the O
following discussion of system performance during this accident. The four accident 3-5 O
O periods (marked I through IV) in this figure have the following characteristics:
g Period I - Depressurization:
e:
The accident period marked I in Figure 3.3 corresponds to the early rapid depressurization which follows break opening. From the secondary side standpoint, period I includes: (1) the early pressure rise due to steam production in the steam generators while the main steam lines are isolated and the steam dump and bypass system is assumed to be inoperable and (2) part of the period where the steam generators are discharging through el, the safety valves, J
Period II - Voiding:
Period I ends and period II begins when a substantial production of steam begins in the core and slows down the depressurization rate. This substantial steam production begins when the bottom of the core starts to boil. This indicates that the whole core is boiling.
Thus, the onset of period II occurs when the lowest core nodes begin to develop a significant void fraction. This occurs at the same time, around 210 seconds, for [
] At this time then, the entire core is boiling, resulting in a large production of steam. The effect of this steam production is to reduce the net depressurization rate of the primary system. That in tum leads to the nearly flat primary system pressure trace, which characterizes period II, as seen in Figure 3.3. During period II, water is held up in the upper plenum (Figure 3.7) by the steam generated in the core.
3-6
l O
O l
] Near the end of period II, as steam production decreases, j
because less water is available, due to liquid boil offin the core, the broken loop seal cleair (Figure 3.10). This allows the depressurization rate to increase, by clearing a vent path from the upper plenum to the break. The loop seal clearing also temporarily disrupts O
the [
]: (1) flash i
either due to the depressurization associated with the clearing or by flowing back into the core and flashing there and/or, (2) to exit via break. The [
],
)
O still in period II, but ends due to dry-out,just prior to the onset of period III. There is also an intermediate heat up and quenching of the clad (Figure 3.9), driven by redistribution f fluid in the core, induced by the loop seal clearing.
O Period II from the secondary side point of view has two distinct behaviors. In the first part of period II all secondary pressures remain stable near the safety valves' set points. This is because in period I the steam generators' safety valves have opened due to the steam dump and bypass system unavailability, in order to discharge the steam produced. The O
atmospheric relief valves (ARVs) are not credited. The early part of period Il continues this behavior. In the second part of period II, steam generator pressures in two loops O
begin to drop following the primary. Since this is also the secondary's behavior in period III, it is discussed in the next section.
O 3-7 O
Cl Period III - Heatup:
g:
The end of period II and beginning of period III starts with the end of significant steam I
production in the core caused by shortage ofliquid, i.e. the onset of dryout. The end of 4:
period II and beginning of period III can be determined from the time at which the core collapsed level reaches the mid core height of 6 ft, indicating the top part of the core has dried out. This can be seen in Figure 3.8. Another indicator is when the top of the core gl void fractions jump to large values e.g. 0.9, also indicating dry out conditions there, as shown in Figures 3.4,3.5 and 3.6. Thus, period III or the heat up period beginsjust before e
the hot rods enter into critical heat flux (CHF, Figure 3.9). The dropping of the collapsed core level to mid height (6 ft, Figure 3.8) and the rate at which it is dropping are indications that the core is drying out quicidy and that steam production has become very low.
Period III is characterized by an increased depressurization rate from the compounded effects of: (1) the previously cleared loop seal (Figure 3.10) and, (2) the lack of steam generation which had been compensating for the energy discharge through the break and keeping the pressure fairly constant in period II.
From the secondary side point of view, period III (and the second part of period II), is characterized by two depressurization rates: one, ahnost non-existant for the broken loop 1 (cleared loop seal) and the loop 4 with the pressurizer and, another, following the primary, for the other two loops (2 & 3). Loops 2 and 3 depressurize because they receive 3-8 O
iO
- O auxiliary feedwater. Loops 1 and 4 do not receive auxiliary feedwater due to the failure of 1 motor driven pump resulting from the single failure of one diesel generator. This is why loops 1 and 4 do not depressurize, while loops 2 and 3 do, as shown in Figure 3.3.
O It is during period III that the fuel experiences its temperature excursion as shown in O
Figure 3.9. The clad temperatures of Figure 3.9 start to rise right at the beginning of period III, because that is by definition when these axial locations dry out.
O Period IV - Recovery:
O Period III ends when the system pressure reaches the accumulator injection pressure. At that time, shown in Figure 3.11, the injection of accumulator water marks the onset of period IV. Accumulator injection causes the core collapsed water level to rise (Figure 3.8)
'O and clad temperatures to begin turning around. Figure 3.9 shows the clad temperature histories one node above, one below and at the PCT [
] location as calculated by ANF-RELAP. The rods are quenched from the bottom up with [
O
]. Steam produced in the rod quench process changes the primary system depressurization rate, which again becomes flat, as shown in Figure 3.3.
O Finally, Figures 3.12 and 3.13 show break flow and pumped injection flow, respectively.
These show that pumped injection flow overcomes break flow in the middle of period IV, ndi ating stable recovery is underway.
O 3-9
'O
_ _ _. _ _ _ ~
O!
3.2 SENSITIVITY STUDIES g
3.2.1 BREAK SPECTRUM The most limiting break location has been determined in previous studies for this l
e\\
(Reference 3.1) and other similar plants (Reference 3.2) to be in the cold leg at the reactor i
coolant pump discharge. Therefore, this cold leg break location remains most limiting for i
the present evaluation and a worst break location search need not be repeated. This most e,
limiting break location is the one considered in all cases discussed throughout this work.
i 0\\
According to the approved EXEM PWR Small Break Model, the break size is the first sensitivity issue addressed. The rationale for addressing break size first is that system thermal-hydraulic behavior is largely affected by break size and less dependent on other issues. Consequently, the break size is a first order effect, while the others are second order.
e The break spectrum study is conducted using the [
9
] It is the same power shape used for the base case and the discussion on how it is obtained has been given in Section 3.0.
- i Three break sizes are analyzed in detail, namely: 3 inch (base case),4 inch, and 2 inch.
Larger sizes (6 inch and 8 inch) were found to be less limiting in preliminary calculations and therefore are not discussed in this document.
3-10
'O l
o 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. The sequence of events for the break spectrum study is summarized in Table O
3.5.
O The result of this study is that the most limiting break is the 3 inch break located in the reactor coolant pump discharge. The 4 inch and the 2 inch breaks result in lower peak clad temperatures than the base case. The 2 inch break shows no significant clad heatup.
O The other sensitivity studies use the limiting 3 inch break.
3 inch Break O
This is the base case calculation described in Section 3.1. The ANF-RELAP PCT is calculated to be 1705.1 F in [
]above the bottom of the core. The clad O
temperature history as calculated by the TOODEE2 code at the node where the PCT occurs is shown in Figure 3.14. The TOODEE2 PCT is 1779.8 F in [
O
] The difference in elevations is due to the rupture of the TOODEE2 node corresponding to the highest powered node at the [
] as also shown in Figure 3.14. TOODEE2 initial fuel conditions for this run correspond to O
beginning of cycle (BOL).
4 inch Break O
The calculated system behavior for this case is similar to the base case (Section 3.1),
3-11 j
O i
i although event durations are somewhat shorter due to the larger break size. The PCT is e
also lower. The ANF-RELAP PCT is calculated to be 1479.7'F also in [
l
] above the bottom of the core. The clad temperature history as calculated by the TOODEE2 code at the node where the PCT occurs is shown in Figure 3.27. The TOODEE2 PCT for the 4 inch case is 1571.7 F in [
] above the top of the core. Since the highest powered node did not rupture, the ANF-RELAP and the TOODEE2 PCT occur at the same elevation. TOODEE2 initial fuel conditions for this run also correspond to beginning of cycle (BOL).
l Si Figure 3.15 shows the power behavior subject to the same mechanisms of the base case calculation.
,1 Figure 3.16 shows the primary and the secondary pressures. The same four accident periods (also marked I through IV in this figure) are used in the following discussion of the 4 inch break.
1 el Period 1 - Depressurization:
As in the base case the accident period marked I in Figure 3.16 corresponds to the depressurization of the primary system due to the break while the secondary pressure rises to and remains at the safety valves' set point. There are no major distinctions between system behavior during this period between the 4 inch break and the 3 inch base case except that the depressurization rate is higher for the larger break.
3-12 2
8
.O
- o Period II - Voiding
As in the base case, period I ends and period II begins when a substantial production of steam starts in the core and slows down the depressurization rate. This substantial steam
'O production also begins with the formation of void at the bottom core elevation. This indicates the entire core is boiling. It occurs at the same time, around 105 seconds, for [
O
] The 3 inch base case discussion for this period applies to the 4 inch break as well. In this case two loop seals clear, the broken loop 1 and
- O loop 2 (only loop 1 clears in the 3 inch case), in the middle of the period, most likely as a result of a larger break size. Loop seal clearing, as in the 3 inch case, also leads to an
- O increase in the primary system depressurization rate. The [
], which is temporarily disrupted by the loop seal clearing, and reinstated until the onset of period III (Figure 3.20). As in the base case, in the 4 inch 1
break there is also an intermediate heatup and quenching of the clad, driven by redistribution of fluid in the core, induced by the loop seal clearing (Figure 3.22).
- O Secondary side behavior is similar to the 3 inch case.
Period III - Heatup:
As in the 3 inch discussion, the end of period II and beginning of period III occurs when the core collapsed level drops be.!ow the mid core elevation of 6 ft (Figure 3.21). The dropping of this level to about 6 ft means the top half of the core is dry, and steam O
production has been substantially reduced. The jump in top core elevations' void fractions 3-13 O
O to high (0.9) values also signals the onset of dryout in this case. The primary system g
pressure continues to drop significantly as two loop seals remain clear and steam production is low. It is also in period III that the fuel experiences its temperature e
excursion as shown in Figure 3.26. For the 4 inch break the loop seals are also clear before the beginning of period III. Secondary side behavior is also similar to the 3 inch case.
Period IV - Recovery:
As in the base case, period III ends when the system pressure reaches the accumulator injection pressure. At that time, shown in Figure 3.24, the injection of accumulator water l
marks the onset of period IV. Accumulator injection causes the core collapsed water level
,j to rise (Figure 3.21) and clad temperatures to begin to tum around. Figure 3.22 shows the clad temperature histories one node above, one below and at the PCT [
] location as calculated by ANF-RELAP. The rods are quenched from the bottom up with [
l
] Steam produced in the rod quench process l
changes the primary system depressurization rate, which again becomes flat, as shown in l
l Figure 3.16. Finally, Figures 3.25 and 3.26 show break flow and pumped injection flow, respectively. Pumped injection flow overcomes break flow also in the middle of period IV, indicating stable recovery is underway.
The same conclusion drawn for the base case applies to the 4 inch calculation. The pumped injection flows (Figure 3.26) cannot keep up with the break flow (Figure 3.25) 3-14 4
d 10 i
!O during periods I, II and III. Still, the accumulator injection pressure is reached well before the clad temperatures are too high and the temperatures are effectively turned around.
O 2 inch Break This calculation is somewhat different from the base case calculation. The difference is
- O that there is no period III, i.e., no heatup period. As in the 3 inch base case, the voiding period II is also interrupted when the broken loop seal clears. At that time, the same
'O perturbations observed in the 4 inch and 3 inch cases occur here as well. The core
.I collapsed level dips and recovers, the {
] is interrupted and reinstated, there is a brief spike in clad temperatures (Figure 3.28) and the primary system pressure
.O begins to drop more quickly. The important difference between this case and the other two is that, the increased depressurization rate associated with loop seal clearing drops the O
break flow to less than the total pumped ECCS flow rate before any core heatup takes place. Thus, the transient is essentially over after the loop seal clears, in the sense that the Possibility of heatup is eliminated. There is never a sustained heatup of the clad for the O
2 inch break, only the spike associated with the clearing of the loop seal, which is shown in Figure 3.28. Since there is no sustained heatup for the 2 inch break, a TOODEE2 O
calculation is unecessary.
Table 3.8 provides a summary of the PCTs for the break spectrum study. Figures 3.30,
.O 3.31 and 3.32 summarize clad temperatures' histories for this study.
3-15
-0
Ol 3.2.2[
]
9 The [
.] This study is performed for the most limiting break determined in the break spectrum study (3 inch, Section 3.2.1). [
l 1
4 Oi The sequence of events for the two sensitivity cases are summarized and compared to the nominal case in Table 3.6. Figure 3.29 overlays the calculated ANF-RELAP clad temperatures for all three cases.
gl The conclusion, as seen in Figure 3.29, is that there is little difference in clad temperature history associated with these [
] for the CPSES model. In any case, [
] used in the base case calculation are the most limiting, as indicated in the PCT summary of Table 3.9.
e 3.2.3[
]
[
e; e
3-16
'O
=O l
Nevertheless, preliminary calculations revealed that clad temperature profiles were not O
converged if the maximum time step was too large. Therefore, it became necessary to find the largest time step at which results were essentially unaffected by further reductions in l
- 0 time step. The objective of this time step study was to fmd that optimum time step value.
This study was performed for all three breaks in the break spectrum study. The main O
convergence criterion was a visual inspection of the behavior throughout the transient, of 1
the most sensitive variable: the clad temperature. In addition to this visual criteria, in rder to be deemed " converged", a run must also exhibit the same sequence of events of O
a smaller time step run. For example, if accumulator injection precedes the PCT in the smaller time step, this must also be the case for a larger time step to be acceptable. Thus, O
similar clad temperature histories are considered necessary but not sufficient conditions.
Finally, although not a requirement, a maximum time step that was consistent throughout the break spectrum study was felt to be desirable if reasonably achievable.
0 Figure 3.30 shows six time step runs (0.05, 0.25, 0.010, 0.005, 0.0025 and 0.00125
-O seconds) for the base case 3 inch break. The three smallest show nearly identical results and event sequences. Thus it is concluded that a maximum time step of 0.005 seconds is adequate. It should be noted that three larger time steps of: 0.050,0.025 and 0.010 v
seconds were also tried. The two largest show discrepancies in the visual comparison to 3-17
,O
Ol the three smallest. The 0.01 seconds time step appears acceptable but was rejected here because: (a) The PCT occurred prior to accumulator injection and (b) this time step is too large for the 4 inch break described below.
O Figure 3.31 shows four time steps (0.010,0.005,0.0025 and 0.00125 seconds) for the 4 inch break. The three smallest are extremely close until significantly after the peak clad temperature has been reached. In this case, the 0.010 second time step did not meet the visual convergence criterion, as shown in Figure 3.31.
O Figure 3.32 shows two time steps (0.010 and 0.005 seconds) for the 2 inch break. Except 4
around the spike associated with the loop seal clearing, these are identical. Therefore, for the sake of consistency with the 3 and 4 inch cases, and considering there is no sustained heatup for this case, the 0.005 second time step is considered adequate for the 2 inch break also.
l l
In summary, all cases were certainly converged at 0.005 seconds. The base case 3 inch break might have been called converged at 0.010 seconds, except that the PCT occurred prior to accumulator injection in that case. The 4 inch break was clearly not converged at 0.010 seconds. The 2 inch break seems converged at 0.010 seconds. Therefore, the 0.005 second time step utilized is adequate for applications of the CPSES model.
Ol Although converged at 0.005 seconds, the actual numerical value of the PCT can vary 3-18 9;
l l,..
)
i i
h somewhat as the time step is reduced further. This can be interpreted as a convergence band. If the band is to the right of(i.e. higher than) the PCT, there could be a concern that the " actual" PCT would be larger. This issue was examined. For the limiting break, this D
variation is concluded to be between 20 F and 300F. [
l p
.]
[
] studies will only be conducted in future applications if:
D-(1)[
] apply or, l
(2) if breaks larger than 4 inch are analyzed, or g
(3)if[
] are utilized.
l l
i l
~
3.2.4 TWO LOOP VERSUS FOUR LOOP ANF-RELAP MODEL O
All analyses discussed up to this point in Chapter 3 have been conducted using the fully l
explicit four loop ANF-RELAP model discussed in Section 2.4.1.
l lO However, CPSES has no significant loop asymmetries and the fully explicit four loop model is cumbersome to execute, particularly at the small time steps required.
O l
l Therefore, it is reasonable to expect that the industry-wide conventional 2-loop l
representation, with a broken loop and a lumped intact loop, would yield results 3
substantially identical to the fully explicit four loop model.
3-19
- O
0l i
In order to test this hypothesis a 2-loop ANF-RELAP model was developed as described e
in Section 2.4.2. Figure 3.33 compares clad temperature histories as calculated with the 2-loop and 4-loop models, for the 3 inch base case. Both used [
]
S There are no significant differences in the transient as calculated with either model. The ANF-RELAP PCT is 1705.1 F for the 4-loop and 1709.70F for the 2-loop. TOODEE2 PCTs are given in Table 3.11.
e; i
Therefore, as a result of this finding, TU Electric intends to utilize the two-loop model in 9'
future applications of this methodology.
3.2.5 EXPOSURE STUDY This exposure study is done by performing an additional TOODEE2 calculation using the same ANF-RELAP boundary conditions as the base case 3 inch run, but for which RODEX2 initial fuel parameters are generated at EOC instead of the BOL conditions used for the base case.
9:
Figure 3.34 compares TOODEE2 BOL and EOC temperature histories for the base case 3 inch break. The highest clad temperature corresponds to the BOL case. The lowest set of curves in Figure 3.34 correspond to the ruptured node. PCTs for this sensitivity study are summarized in Table 3.12.
O 3-20 4
O l
iO TABLE 3.1
SUMMARY
OF CPSES-1 SMALL BREAK LOCA ANALYSIS
- g ASSUMPTIONS FOR BASE CASE AND SENSITIVITY STUDIES 1.
The initial power is 3479 MWt, which is 2% above the licensed power level of 3411 MWt, to account for calorimetric measurement uncertainty.
O 2.
5% of the steam generator tubes are plugged.
3.
Break in reactor coolant pump discharge occurs at 0.0 s.
4.
Reactor trips due to a Lo-Pressurizer pressure signal.
5.
Loss of offsite power coincides with reactor trip.
6.
The reactor coolant pumps (RCP) are tripped at reactor trip since RCP cannot operate without offsite power after a reactor trip.
7.
Steam flow isolation is initiated at the time of reactor trip. The steam dump and bypass system is not credited.
8.
Main feedwater isolation is initiated 7 seconds after "S" signal.
O I
9.
Failure of one diesel generator to start takes out one high head centrifugal charging pump, one intermediate head safety injection pump, one RHR pump and one motor-driven AFW pump. This is the single failure assumed for compliance with 10 CFR 50, Appendix K, Part D.
O
- 10. One high head centrifugal charging pump, one intermediate head safety injection pump inject on demand after the appropriate delays, at conservative flow rates.
- 11. One of the two motor-driven AFW pumps is credited, but injection is conservatively
'O delayed in order to account for flow travel time.
- 12. All accumulators inject on demand.
O 3-21 0
O TABLE 3.2
SUMMARY
OF INITIAL CONDITIONS FOR CPSES-1 SMALL BREAK LOCA BASE CASE AND SENSITIVITY STUDIES DESCRIPTION VALUE o Core Power 3479 Mwt o Power Calorimetric Uncertainty Multiplier 1.02 o Power Shape Analyzed
[
]
g o Peak Linear Power [
](includes 102% factor) 13.12 KW/ft o Fraction of heat deposited in fuel 0.974 o Total Peaking Factor, F (flat segment of K(z))
2.42 n
o Total Peaking Factor, F, [
]
[
]
O o Accumulator Water Volume per Tank 6119 gals o Accumulator Cover Gas Pressure 603 psia o Accumulator Water Temperature 150*F o Safety injection Pumped Flow Table 2.6 O
o Refueling Water Storage Tank Temperature 120*F o Initial Loop Flow 9661 lbm/sec o VesselInlet Temperature 566 'F o Vessel Outlet Temperature 626'F g
o Reactor Coolant Pressure 2280 psia o Pressurizer Water Volume 1116 fP o Steam Pressure 914 psia o Auxiliary Feedwater Flow to each of SGs 2 & 3 29.6 lb/sec Oi o Auxiliary Feedwater Flow to each of SGs I & 4 0.00 lb/sec o Steam Generator Tube Plugging Level 5%
o Steam Generator Safety Valves Set Points & Flows Table 2.9 o Fuel Parameters Table 3.3 g'
O' 3-22 9
O O
TABLE 3.3
SUMMARY
OF FUEL PARAMETERS FOR BASE CASE SMALL BREAK LOCA ANALYSIS O
PARAMETERS VALUES Fuel Rod Geometry Data Table 2.8 Time to Maximum Stored Energy Exposure Fuel Rod Composition:
I 1
[
O O
]
Average fuel temperature 1532 2236 2245 O
at peak stored energy ( F)
~O O
3-23 0
O TABLE 3.4 SEQUENCE OF EVENTS FOR BASE CASEI SMALL BREAK LOCA EVENT TIME G
(SECONDS)
- 1. Break opens (period I begins) 0.0
- 2. Reactor Trip Signal 25.1 0
- 3. RCP tripped 27.1
- 4. MSIV closed 27.6
- 5. "S" Signal 34.3 9
- 6. MFW isolated 41.3
- 7. Centrifugal charging pumps inject 51.3
- 8. Safety injection pumps inject 56.3 9
- 9. Entire core boils (period II begins)
-210
- 10. Auxiliary Feedwater reaches SGs 2 & 3 234.3
- 11. Broken Loop i seal clears
~730 O'
- 12. Critical Heat Flux at PCT node (period 111 begins)
~1050
- 13. Accumulator injection (period IV begins) 1824
- 14. Peak clad temperature reached 1824
- 15. Pumped ECCS flow exceeds break flow
~1600
- 16. Calculation ends 2100.0 el l
3 inch break, [
], maximum ANF-RELAP time 2
step of[
],4-loop explicit ANF-RELAP model, beginning of cycle gl exposure.
3-24 l
- l
Q 3
TABLE 3.5 SEQUENCE OF EVENTS FOR BREAK SPECTRUM 2 STUDY J
TIME (SECONDS)
EVENT 3 inch 4 inch 2 inch
- 1. Break opens (period I begins) 0.0 0.0 0.0 3
- 2. Reactor Trip Signal 25.1 12.6 66.4
- 3. RCP tripped 27.1 14.6 68.4
- 4. MSIV closed 27.6 15.1 68.9 3
- 5. "S" Signal 34.3 21.2 75.9
- 6. MFW isolated 41.3 28.2 82.9
- 7. Centrifugal charging pumps inject 51.3 38.2 92.9 3
- 8. Safety injection pumps inject 56.3 43.2 97.9
- 9. Entire core boils (period 11 begins)
~210
~105
~450
- 11. Broken Loop 1 seal clears
~730
~397
~2014
- 12. Critical Heat Flux at PCT node (period III begins)
~1050
~660 No Heatup
- 13. Accumulator injection (period IV begins) 1824 889 N/A O'
- 14. Peak clad temperature reached 1824 898 No Heatup
- 15. Pumped ECCS flow exceeds break flow
~1600
-1000 Early
- 16. Calculation ends 2100.0
~1000 2400.0 0
All cases: nominal [
], maximum ANF-RELAP time 2
g step of(
],4-loop ANF-RELAP explicit model, beginning oflife fuel exposure.
3-25 O
Ol TABLE 3.6
- l SEQUENCE OF EVENTS FOR I
] STUDY 2 TIME (SECONDS)
Ol EVENT
[
]
[
]
l
]
- 1. Break opens (period I begins) 0.0 0.0 0.0
- l
- 2. Reactor Trip Signal 25.1 25.2 25.0
- 3. RCP tripped 27.1 27.2 27.0
- 4. MSIV closed 27.6 27.7 27.5 9l
- 5. "S" Signal 34.3 34.5 34.3
- 6. MFW isolated 41.3 41.5 41.3
- 7. Centrifugal charging pumps inject 51.3 51.5 51.3 0!
- 8. Safety injection pumps inject 56.3 56.5 56.3
- 9. Entire core boils (period II begins)
~210
~210
~210
- 10. Auxiliary Feedwater reaches SGs 2 & 3 234.3 234.5 234.3 9i l1. Broken Loop 1 seal clears
~730
~730
~730
- 12. Critical Heat Flux at PCT node (period III begins)
~1050
~1050
~1050
- 13. Accumulator injection (period IV begins) 1824 1810 1802 O
- 14. Peak clad temperature reached 1824 1806 1804
- 15. Pumped ECCS flow exceeds break flow
~1600
~1600
~1600
- 16. Calculation ends
~2100
~2100
~2100 0
All cases: 3 inch break, maximum ANF-RELAP time step of[
],4-g-
2 loop explicit ANF-RELAP model, beginning oflife fuel parameters for ANF-RELAP, TOODEE2 runs not needed.
3-26 O'
O O
TABLE 3.7 SEQUENCE OF EVENTS FOR 2-LOOP TO 4-LOOP COMPARISONd i
O TIME (SECONDS) 4-LOOP 2-LOOP EVENT MODEL MODEL O
- 1. Break opens (period 1 begins) 0.0 0.0
- 2. Reactor Trip Signa!
25.1 25.0
- 3. RCP tripped 27.1 27.0 0
- 4. MSiv closed 27.6 27.5
- 5. "S" Signal 34.3 34.2
- 6. MFW isolated 41.3 41.2 0
- 7. Centrirugal charging pumps inject 51.3 51.2
- 8. Safety injection pumps inject 56.3 56.2
- 9. Entire core boils (period Il begins)
~210
-210 O
10 Auxiliary Feedwater reaches SGs 2 & 3 234.3 234.3 i1. Broken Loop 1 seal clears
~730
~712
- 12. Critical Heat Flux at PCT node (period 111 begins)
~1050
~1050 O
- 13. Accumulator injection (period IV begins) 1824 1780
- 14. Peak clad temperature reached 1824 1786
- 15. Pumped ECCS flow exceeds break flow
~160c
~1600 i6. calcuiauon enas 2i00.0 2ioo.o O
O l
All cases: 3 inch break, nominal [
], maximum ANF-4 RELAP time step [
], beginning oflife fuel exposure.
3-27 O
O TABLE 3.8 PCT
SUMMARY
FOR BREAI5 SPECTRUM STUDY 5 BREAK SIZE ANF-RELAP Pt \\'('F)
TOODEE2 PCT ('F) e (INCIIES) 3.0 1705.1 1779.8 4.0 1479.7 1571.7 9
2.0 NO HEATUP N/A O
TABLE 3.9 PCT
SUMMARY
FOR [
] STUDY' e
[
[ANF-RELAP PCT ('F)]
[TOODEE2 PCT ('F)]
I
[
]
1705.1 1779.8 e
[
]
1664.3 N/A
[
]
1610.6 N/A e
O All cases: [
], maximum ANF-RELAP time 5
step of[
],4-loop ANF-RELAP explicit model, beginning oflife fuel exposure.
All cases: 3 inch break, maximum ANF-RELAP time step of [
],4-5 g
loop explicit ANF-RELAP model, beginning oflife fuel parameters for ANF-RELAP, TOODEE2 runs not needed.
3-28 O
'O O
TABLE 3.10 l
l PCT
SUMMARY
FOR [
]'
3 INCH BREAK (SEE ALSO FIGURE 3.30)
MAX ANF-RELAP [
]
ANF-RELAP PCT ('F)
TOODEE2 PCT ('F)
[
]
1747.2 1815.2
[
l 1736.0 1758.9 O
[
]
1726.6 1783.4
[
]
1705.1 1779.8
[
]
1722.6 1809.5
[
]
1732.4 1795.2 4 INCH BREAK (SEE ALSO FIGURE 3.31)
O
[
]
ANF-RELAP PCT (8F)
TOODEE2 PCT ('F)
[
]
1625.1 1705.1
[
l 1479.7 1571.7 O
[
]
1487.7 1579.1
[
]
1459.6 1553.1 2 INCH BREAK (SEE ALSO FIGURE 3.32)
.O
[
]
ANF-RELAP PCT ('F)
TOODEE2 PCT ('F)
[
]
NO HEATUP N/A
[
l NO HEATUP N/A
- O All cases: [
],4-loop ANF-RELAP explicit 7
model, beginning oflife fuel exposure.
O Not " converged".
8 Optimum time step.
8 3-29 O
O e
TABLE 3.11 PCT
SUMMARY
FOR 2-LOOP MODEL VALIDATION STUDY '
ANF-RELAP NODALIZATION ANF-RELAP PCT ('F)
TOODEE2 PCT ('F) 4-LOOP (FIGURE 2.2) 1705.1 1779.8 2-LOOP (FIGURE 2.4) 1709.7 1792.8 e
e.
TABLE 3.12 e
SUMMARY
FOR EXPOSURE STUDY" EXPOSURE ANF ". LAP PCT ('F)
TOODEE2 PCT ('F)
BEGINNING OF LIFE 1705.1 1779.8 END OF CYCLE 1705.1 1745.7 4
e All cases: 3 inch break, [
], maximum ANF-20 RELAP time step [
], beginning oflife fuel exposure. See also Figure 3.33 All cases: 3 inch break, [
],4-loop ANF-22 RELAP same run at BOL. See also Figure 3.34 3-30 0
1 i
O i
O 3 INCH COLD LEG BAEAK i
1 I
c c3 i
0 r
O i
i i
y_
..:.............e.
c w
si:
O 9
- m.
I w"
c:
c)
C)
Q(
<g m
$ 0 00 4 '2 00 8'4.00 t '2 6. 0 0 1'68.00 210 00 0
TIME (SEC)
- 10 Figure 3.2 Total Core Power O
3 INCH COLO LEG BAEAK cm l
e-o9 i
1 O
f%
9",..
i v
~ n i
ct I
l tS--4D P 174010000 mm C"
k l
i e--e P 570010000 WP 571010000
....)..
P 572010000 w
M- --M P 5730'0000 j
wa-II O
c1. ~
I=0C" =;;;;:.-
- ZCCCCCCC i
> - =
i
. 4... -
Z...........
.Z.....
g O.-
l
<c 11 o "
.d 3r Z
IV o
Sg i
i
=
0 ma T
B 5
5 5
c5 0.00 42.00 84.00 126.00 168.00 210.00 10' TIME (SEC)
^
C
<1 C"
O Figure 3.3 Primary & Secondary System Pressures O
i Ol 9
3 INCH COLO LEG 8AEAK l
o c.
Ln '
w S'
-.a
~
======6==.__
0 cc 5 5-4DvoiDG 122c10000 CL.
O-4D vo l OG 117010000.
. 4..
O M volOG 112010000 Q
9 M
5,
- 2: 1 1 --
,.---,,,f Q1 0 00 4 '2 00 8'4.00 1'26.00 168.00 2'10.00 TIME (SEC)
- 10' Figure 3.4 Void Fractions e
e1 3 INCH COLO LEG BAEAK 4
en wJ" l
l l
u Q
- =.'=====?==_
S-4D vo t DG 140010000
....j...
.........j............
O "- S-ev01DG 13$010000.
M votDG 130010000 O
I 7; Q
- - =_-
9 C3
^ ^
^
C, j ^^.
^._ A o,,00- - - -
0.
42.00 84.00 126.00 1'68.00 210.00 TIME (SEC)
- 10' w
)VoidFractions Figure 3.5
O O
3 INCH COLO LEG BREA_K e
tv O
m-t.u
- :: =,= = = = = = = :._
Oc M votDG 160010000 C.
. O--e VO I CG 155010000.
..;...........j....
C M votDG 150010000 O
O A.
.a a_. ; m __ a o
- i. _..
O o,
_, r_
_2 o..
0.00 42 00 84.00 125.00 168.00 210.00 w
TIME (SEC)
- 10' Figure 3.6 Void Fractions O
3 INCH COLD LEG BAEAK O
o C
E f
LA 25 CD O
s'
- ; cr2
....w...~.."'Jn...
2....................;..
y.....
<r e
15--4!3 VO f OF 166010000 m
tl O-@VOIDF 173010000 C
M VolOF 174010000
-m om
'Ei**********'**
06-
'D O
1 l
f h
a.
. e2 D
$1 _ d = = = N = = = = :!: = =
0.00 42.00 84.00 126.00 168.00 210.00 TIME (SEC)
- 10' O
Figure 3.7 Reactor Vessel Upper Plenum Liquid Fractions O
O e
3 INCH COLD LEG 8AEAK ea m
l
+
u um wm
.....L*'
y,..
- ~ ' '
- **~
w
.1 J a s
T' i
G a
w w~m l
a..
4 u
y Ou e
G
=
'O.00 4'2.00 8'4.00 1'26.00 1'68.00 210.00 iIME (SEC)
- 10' Figure 3.8 Collapsed Level e
3 INCH COLD LEG BAEAK 9
- a a=
" es
< co c) m u -
.i-O s-eHrteup 12s iO 10 0 s S--e H TT EMA 129101108 a
W HTTEMP 129101208
":iE w
H c3
=
g g.7
- ........j........r-O %___
_ _ J _i._ __ _
i i
4 7
i i
i
=
6.,
e 0.00 42.00 84.00 126.00 168.00 210.00 TIME (SEC)
- 10' 9
Figure 3.9 Cladding Temperature Profiles O
l O
O 3 INCH COLD LEG BAEAK e
N C
O!C C C C C'O C C C C $ C C o<
A@
C
. J..
. J..
. j.
Q C9-4D vo l OG 450030000 C
e-4D vo l OG 461030000 O
W VOIOG 452030000
- voiOG 463030000 a
-8 v g,_
. j..
'o o.
8 cs
'O "l:: = = = : :i = = = =: ::
- : : :t : : : : 4 : : ::: :
0.00 42 OC E4 00 126.00 168.00 210.00 TIME (SEC)
- 10' Figare 3.10 Loop Seal Vapor Void Fractions O
3 INCH COLD LEG BAEAK O
c.)
O s
en N
w
~
2 co es O
w
......J..
.* M uPLOWJ 735000000'
- i'*
cs_...
g M MALOWJ 736000000 g
M MFLOWJ 737000000
- MF L OWJ 738000000 y
'e 7............!.............:....
l W
p =_
O 2
===_;===============4
w go 0
3 l
ue U.
< 0.00 42.00 84.00 126.00 168.00 210.00 TIME (SEC)
- 10' O
Figure 3.11 Accumulator Mass Flow Rates O
1 O '1 u
l i
O' 3 INCH COLD LEG BAEAK
- o o cm e
<: ev O
(
-e va o 3 o..
ca ec =
g O*
a ca
- u. v
.s...
..r..
u
<t w
L-co a
_ h cr e O
e a
s s
J 0 00 42 00 84.00 126.00 168.00 210 00 TIME (SEC) 10' o
+
Figure 3.12 Total Break Mass Flow Rate 4
3 INCH COLD LEG BAEAK O
o C.
o C.3 u,
j
- Em-e g
a -
mg................
se o
a
- u. c.,
m a
gm l
(.n.;
U gl1 uJ
.c o
O"
?
?
O.00 4'2.00 8'4.00 1'26.00 1'68.00 210.00 10' TIME (SEC) 01 Figure 3.13 Total ECCS (single train) Mass Flow Rate 4
O l
l 3 i>CH COLO LEG Ostf AC (7000Ett PCT PflOFILE) l 1 9CT #EIN
=
[
sn Wwt 88 o
a w;
3; 59:
we 7
O E
f-
$s i
()
L.
.n.
.w.
m.
,.s..
,n..
TiWE - SECONOS Figure 3.14 3-inch CLB TOODEE2 Peak Cladding Temperature O
4 INCH COLO LEG BAEAK
.O o
o.
4 m
co '
f
.O gj.
.............j...........j..............:...............j.............
e tu 1
=m om Q-r.
ua O
c c) u
<4 O 0.00 2'8.00 5'S.00 8'4.00, i12.00 140.00 H
TIME (SEC)
- 10 0
Figure 3.15 Total Core Power O
~.
O !
4 INCH COLO LEG BAEAK cm Ql 8
s
" n m
l (n m S--E P 174010000 C. " j e--e P 570010000 I" '-
A--*P 571010000 (M
' j" -
y-p 572010000 II j
j
>+---x P 573010000 w
[ r.
====_______=== : --
=-
km
==
CQ.
,.e
.e.
e ca i 8.I z
g if IV O
Ua III we V) g c5*0.00 2'8.00 5'S.00 8'4.00 1'12.00 140.00 TIME (SEC)
- 10' x
c=
9l cz.
c.
Figure 3.16 Pritaary & Secondary System Pressures 01 4 INCH COLO LEG BAEAK 9,
w" 5-E VOIDG 122010000 9-CVOIDG 117010000 a
M VOIDG 112010000 j
i LL.
@g i
i O
- o. g.
o
-........j
..g...........
O m
.f...
- g.....
4:
I
_L i
.m
..[
i i.
o=.
~
o-0.00 28.00 56.00 84.00 112.00 140.00 w
TIME (SEC)
- 10' 4
Figure 3.17 Void Fractions 9
^
O O
4 INCH COLD LEG BAEAK w
i u,_
n (3-HDvoIOG 140010000
%/
[
e --e vo l DG 135010000 l
g h volOG 130010000 i
ggi C"
Q-o m
a, O
l O
g
. i.
=;
l l
l I
f e
o,
,,,j s ^^^
r O'I 05 2'8.00 5'6.00 8'4.00 1'12.00 1d0.00 TIME (SEC)
- 10' i
w Figure 3.18 Void Fractions O
4 INCH COLD LEG BAEAK O
=
N 5-4!D vo l DG 160010000 e--e vo t DG 155010000 a
M votDG 150010000 u.
O d
O a,m
- a. g_
- 9...
..g....
O O
r.-
e I
T g.
.q-m 9
V l
. # 5, ^ ^ ^-
o 0.00 28.00 56.00 84.00 112.00 140.00 TIME (SEC)
- 10' O
Figure 3.19 Void Fractions O
O O
4 INCH COLD LEG BAEAK eo
^
O W
5 i
i i
p ~.
h _ :_
u g.
z.,,
..g CI u
(!>-E vo l OF 166010000 O
e*--e vo l DF 173010000 O
M VO!DF 17401000C
-m Dm 06-
'L
- i'
- i*
- t*-
l
} l) c.
.c M... i, D*
n_______
O e
0.00 28 00 56.00 84.00 112 00 140.00 TIME (SEC)
- 10' Figure 3.20 Reactor Vessel Upper Plenum Liquid Fractions e
4 INCH COLD LEG BAEAK O
e O.
m
~
H u.
w am O
w $a y,..........
W a
o w
I gm I
- a. m 4
a O
a O
U l
%o m
i i
i i
0.00 28.00 56.00 84.00 112.00 140.00 v
TIME (SEC)
- 10' O
Figure 3.21 Collapsed Level O
O O
4 INCH COLD LEG BAEAK c2 c3 G
- c)
- e O
i L,
i
=
mo LL -
i w ~
S-E3HTTEMA 129101008 e -1D MT TEMP 129101108
- t M HTTEMD 129101200 l
ct 35 u.J O
-g i
i o
_ _ _ _ _ _ _ L _ ::: _
_a u
m o
O c,
m s
s e
4 0.00 28 00 56.00 84.00 112.00 140.00 TIME (SEC)
- 10' w
Figure 3.22 Cladding Temperature Profiles O
4 INCH COLD LEG BAEAK O
=
t%
S-EIVOIDG 460030000 O-e vDI DG 461030000 M yotDG 462030000 d
- C CC CC C C C C C CCC CC-
- votDG 463030000
<c O
ce ug.............j.,
.........j.............j.............
o o
a=
g g_
....... j..
...j........... j..
.j..........
O w
ct 8 c, i
i i
wo
..L A w
o' 30 ---- f8 30 s's.00 s'4.00 1's2.00 140.00 10' TIME (SEC)
O Figure 3.23 Loop Seal Vapor Void Fractions O
O e
4 INCH COLD LEG BAEAK eo O
C~
3i m
l l
l l
':E m e
CD C w6
- Op--E MF L OwJ 735000000
M M MFLOwJ 736000000 O
w MFLowJ 237000000 J
+--+ MF L OwJ 738000000 gj
'o o
m
..i.........
m o..
- 2 se g,
________m_______m________-_.
o a
ue g
UT i
i i
i
< 0.00 28 00 56.00 84.00 112.00 140.00 TIME (SEC)
^10' Figure 3.24 Accumulator Mass Flow Rates 9:
4 INCH COLD LEG BAEAK O
o c= =
o
<aN
(
m
} -
ca a
,~
09
....g...........,i i
g.
i G
u<
u; CE cn o.
g
$"e i
0.00 28.00 56.00 84.00 112.00 140.00 TIME (SEC)
- 10' O+
9' Figure 3.25 Total Break Mass Flow Rate e
-O O
4 INCH COLD LEG BAEAK e
e, ea O
-m l
l l
l 2m l
cu m a -
g.
a:
CD n
a U
- u.,.3, n
wg y
L) l W
J w*
l 4
e O
t
=
?
Q*
se i
0.00 28.00 56.00 84.00 112.00 140 00 TIME (SEC)
- 10' Figure 3.26 Total ECCS (single train) Mass Flow Rate
'O 4 INCM COLO LEC 6AEAK (T000fE2 PCT PROFILE) l t.
SCT m I
J 2
3 e aanmao m m
,/l w
8=
=
s 8
m:
3:
O
=9: f er i_l d
O y
t I
t t
t I
1 i
I 5.
- u..
.m.
TIE SECONOS O
Figure 3.27 4-inch CLB TOODEE2 Peak Cladding Temperature O
O e
2 INCH COLD LEG BAEAK 1
- cs c3 "
en
- w
==
e
~
I
~
- u. g w
cs.
~
M HTTEup 129101008 a
l M HTTEMP 129101108 y
M HTTEMP 129101208 w
m gm
$4 O m_
..g....
. c..
%=====: = = = = = = = = = = = := = = ; = =; _j a
)
u
_===-
m ca.
c3 V
0.00 4'8.00 9'6.00 1'44.00 1'92.00 240.00 TIME (SEC) 10' 1
Figure 3.28 Cladding Temperature Profiles 0;
1 l
3 INCH COLO LEG BAEAK O
- e co =
e
<= -
cs
~
- u. h_
w O
I a
- E w
g c3
=.
l a g_
a i
u L
j j
g 1
l 7
o.
e, m
a a
e i
0.00 42.00 84.00 126.00 168.00 210.00 10' TIME (SEC) w O
Figure 3.29 Clad Temp. for O
O O
3 INCH COLD LEG BREAK
- ao a*o
<=-
O o
Ao
- u. g
..... f...
.. f..
- 4..
g_..
o.
1 w
p ne v
=
a g..
..i..
..........f...
. i..
y u
L j
.o o.
C3 Q
"0.00 4'2.00 8'4.00 1'26.00 1'68.00 210.00 TIME (SEC)
- 10 w
Figure 3.30 Cladding Temperature at Different Ats
- O 4 INCH COLD LEG BAEAK O
- =
o=
t
-a
<=-
e
-o
- u. g Q
"g -
............f.....
......f...............~..
.. {..
c.
y w
go
=
g g.
...g...............g.
...g..
a O
h_
kI O
s o
e "O.00 2'O.00 4'O.00 6O.00 8'O.00 100.00 TIME (SEC)
- 10' O
Figure 3.31 Cladding Temperature at Different Ats O
- ~.
O O
2 INCH COLD LEG BAEAK
- e coo o
<m-O o
m ug
..4..............j.........
..j......
g_.........
CL s
w ge o
O o_
a U
I.
m
=.
a g
v v
s a
u 0.00 48.00 96.00 144.00 192.00 240.00 TIME (SEC)
- 10' w
Figure 3.32 Cladding Temperature at Different ots O
3 INCH COLD LEG BAEAK O
o o=
o
<= -
o
~
- u. R............::.................::...............::...
O S
ct s
w
~=
=.
a g.
a u
L e
o o.
.e m
0.00 42.00 84.00 126.00 168.00 210.00 TIME (SEC)
^10' O
O Figure 3.33 Cladding Temp. for 4-
& 2-Loop Models e
I O
ie cao teo saur tramier act owns)
O i
C_ i S
'O a
gi j
168c G
5:
ge 7
O
?
QE I
a L_.--
l oc EOC 5
l O
.=.
TIE - SECON05 Figure 3.34 TOODEE2 PCT Profiles at BOC & EOC
- O O
O O
O O
~_
O O
CHAPTER 4 CONCLUSION O
The USNRC-approved (References 1.1 and 2.1) SPC's ECCS Evaluation model entitled EXEM PWR Small Break Model has been applied to the Comanche Peak Steam Electric O
Station Unit One (CPSES-1).
.O Each calculation has been performed in close compliance with the explicitly approved i
EXEM PWR Small Break methodology. [
O
}. This was done in order to demostrate that the 2-loop model yields results which are essentially identical to the 4-loop model. TU O
Electric intends to use the 2-loop model in future calculations.
Six calculations, excluding [
] studies, have been O
presented with two objectives:
1.
To demonstrate TU Electric's ability to properly apply EXEM PWR Small O
Break Model(Reference 1.1); and
.O 4-1 0
_._..~-.._
O 2.
To demonstrate the development of up-to-date input decks and conclusions
- ~
which are in compliance with 10 CFR 50.46 and Appendix K thereto. Together, the codes, input decks and conclusions drawn from these calculations will be Si applied to perform subsequent fuel cycle analyses for the Comanche Peak l
Steam Electric Station Unit One and Unit Two.
O 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:
9.
1.
The calculated peak clad temperature is lower than the 2200 F peak clad temperature limit set forth in 10 CFR 50.46 (b)(1).
2.
The total cladding oxidation at the peak location is under the 17% limit i
specified in 10 CFR 50.46 (b)(2).
l 3.
The hydrogen generated in the core by cladding oxidation is less than the 1%
Gl limit of 10 CFR 50.46 (b)(3).
O' 4.
The average core region undergoes only minor dimensional changes, no clad l
l mptures are calculated to occur there. Thus, the coolable geometry criterion of 1
10 CFR 50.46 (b)(4)is satisfied.
4:
4-2 4)
,e
.m
~-,, --
.w-,
-,-,n
+
- O O
5.
Following accumulator injection, the rods are quenched, the pumped ECCS flow exceeds the break flow and the core is well cooled thereafter. Therefore, the calculations comply with the long-term cooling criterion of 10 CFR 50.46 (b)(5).
lO Regarding the sensitivity studies it has been found:
1.
The most limiting break is a 3 inch break in the main coolant pump discharge O
line.
O 2.
The [
] revealed little sensitivity to these [
] for the CPSES model. [
O
]
O 3.
The optimum time step for these calculations is demostrated to be 0.005 seconds.
O 4.
Although converged at 0.005 seconds, the actual numerical value of the PCT can vary slightly as the time step is reduced further. This can be interpreted as O
a convergence band. For the limiting break, this variation is concluded to be 4-3 0
._. - - - -. ~
O' between 20 F and 30 F.
[
.]
5.
The two-loop ANF-RELAP model yields results which are basically the same as those obtained with the explicit four-loop model. Peak clad temperature histories are nearly identical. Numerically, the two-loop ANF-RELAP PCT is 1710 F and the four-loop PCT is 17050F. The corresponding TOODEE2 PCTs 9
are 1780 F for the 4-loop and 1793 F for the 2-loop.
TU Electric will use the EXEM PWR Small Break model including all codes, input decks, results, conclusions, and application procedures presented in this report to perform small break LOCA analyses and evaluations in compliance with 10 CFR 50.46 criteria and 10 i
9I CFR 50, Appendix K requirements, for both Comanche Peak Steam Electric Station Unit One and Unit Two.
9 e
ei 4-4 9i
-O O
TABLE 4.1
SUMMARY
OF RESULTS FOR BASE CASE, 2-LOOP MODEL AND EOC CASE EXPOSURE SENSITIVITY 2-LOOP l
BREAK SIZE ANF-RELAP (INCHES)
BASE CASE END OF CYCLE MODEL
-0 (BOL)
(EOC) 1779.8 0F (1)-
1745.7 F 1792.8 F 3.0 1.947 % (2) 2.513 %
2.061 %
0.303 % (3) 0.310 %
0.327 %
O 1571.7 F NOTES-4.0 0.379 %
0.044 %
ALL RESULTS FROM TOODEE2:
(1) PEAK CLADDING TEMPERATURE 2.0 NO HEATUP (2) PERCENT LOCAL CLAD OXIDATION O
(3) PERCENT CORE-WIDEi2 OXIDATION O
O
" hot pin value is used as an upper bound for the core-wide O
value.
4-5 0
.O
'O CHAPTER 5 REFERENCES O
Chapter 1:
1.1 Letter,G.M.Holahan (USNRC) to R.A.Copeland, Siemens Power Corporation (SPC), " Acceptance for Referencing of the Topical Report XN-NF-82-49(P)(A),
Revision 1, Supplement 1, Exxon Nuclear Company Evaluation Model Revised O.
EXEM PWR Small Break Model'(TAC No. M83302)," October 3,1994.
Chapter 2:
.O 2.1 Siemens Power Corporation, " Exxon Nuclear Company Evaluation Model Revised EXEM PWR Small Break Model," XN-NF-82-49 (P) (A), Revision 1 Supplement 1, May 1992.
2.2 V. H. Ransora, et. al., "RELAP5/ MOD 2 Code Manual, Volume 1: Code O
Structure, Systems Models, and Solution Methods," NUREG/CR-4312, Rev.1, March 1987.
2.3 F. J. Moody, " Maximum Flow Rate of a Single Component Two-Phase Mixture,"
J. Heat Transfer, Trans. ASME, 87, pp 134-142, February,1965.
2.4 G. G. Loomis, " Summary of The Semiscale Program (1965 - 1986),"
N1. REG /CR-4945, EGG-2509, July 1987.
2.5 Division of Technical Review, Nuclear Regulatory Commission, "TOODEE2:
- O A 'Jwo Dimensional Time Dependent Fuel Element Thermal Analysis Program,"
NI: REG-75/057, May 1975.
2.6 Nuclear Regulatory Commission, Division ofTechnical Review, "WREM: Water Reactor Evaluation Model," NUREG-75/056 (Revision 1) May 1975.
2.7 Advanced Nuclear Fuels Corporation, "USNRC's Safety Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluation Model EXEM/PWR and Acceptance for Referencing of Related Topical Reports," July 1986.
.O 5-1 O
O-2.8 P. Salim, " Guidelines for PWR Safety Analysis Section 10.0: Small Break O'
LOCA Analysis," EMF-1238 (P), Rev.1, Siemens Power Corporation, October, 1994.
2.9 TU Electric, " Steady State Reactor Physics Methodology," RXE-89-003 P, July 1989.
2.10 D. S. Huegel, C. M. Thompson, " Comanche Peak Unit 1 Accident Assumptions Checklists," WCAP-12368, August 1990, Revision 1.
2.11 Comanche Peak Steam Electric Station Units 1 and 2, Technical Soecifications.
2.12 Westinghouse Letter, J.L. Vota (LV) to W. J. Cahill (TUE) " Comanche Peak Units 1 & 2 Safety Evaluation for Reduced ECCS Flow to Prevent Charging /SI and HHSI Pump Runout During Recirculation," WPT-13963, September 25, 1991.
Chapter 3:
3.1 Comanche Peak Steam Electric Station Units One And Two, " Final Safety Analysis Report," Section 15.6, Amendment 78, January 15,1990.
3.2 USNRC, " Water Reactor Evaluation Model (WREM): PWR Nodalization and Sensitivity Studies," - Technical Review U.S. Atomic Energy Commission, October 1974.
S' 3.3 TUElectric, " Power Distribution Control Analysis And Overtemperature N-16 and Overpower N-16 Trip Setpoint Methodology," RXE-90-006-P-A, June, 1994.
O O
9 5-2 0
3 3
APPENDIX J
l DESCRIPTION OF THE COMPUTATIONAL TOOLS 1)
The EXEM PWR Small Break Model consists cf three basi: computer codes:
- 1. RODEX2
- 2. ANF-RELAP 3
- 3. TOODEE2 The codes, their interfaces, interrelationships and respective inputs and outputs are summarized in Figure n.1 and Table A.1. The functit>n of each code is described in the
)
3 following sections.
e A.1 RODEX2 1
i RODEX2 is used within the EXEM PWR Small Break Model framework to provide initial nditi ns for the ANF-RELAP and TOODEE2 calculations, as illustrated in O
Figure A.1 and Table A.I.
O RODEX2 describes the thermal-mechanical performance of fuel during its operational lifetime preceding the LOCA. The determination of stored energy for the LOCA analysis requires a com.ervative %l rod thermal-mechanical model that is capable of calculating O
fuel and cladding behavior, including the gap conductance between fuel and cladding as a function of burnup. The parameters affecting fuel performance, such as fission gas O
A-1 gO v
O 1
i i
i release, cladding dimensional changes, fuel densification, swelling, and thermal expansion are accounted for.
O RODEX2 provides an integrated evaluation procedure for considering the effect of varying temporal and spatial power histories on the temperature distribution, inert fission gas release, and deformation distribution (mechanical stress-strain and density state) 9 within the fuel rod. The surfece conditions for the fuel rods are calculated with a thermal-hydraulic model of a rod in a flow channel. [
O l
ej l
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 Oi power and bumup distributions are handled by dividing the fuel rod into a number of axial segments which are modeled as radially dependent regions whose axial deformations and gas releases are summed. Power distributions can be changed at any time and the coolant 9;
and cladding temperatures are readjusted at all axial nodes. Deformation of the fuel and cladding and gas release are calculated using shoner time steps than those used to define O
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. [
e; A-2 9!
O O
0
}
O A.2 ANF-RELAP ANF-RELAP is a modified version of RELAP5/ MOD 2, INEL Cycle 36.06. The RELAP5/ MOD 2 code is described in detail in Reference 2.2. RELAP5/ MOD 2 has been O
modified in [
] major ways to produce ANF-RELAP:
O
[
O
,O
!O O
]
A-3
- O
Oi The ANF-RELAP model is described in Section 2.4.1.
Initial thermal-hydraulic conditions are determined using LOOPT (Section A.4.1) followed by initialization calculations which include a null transient run. Initial fuel rod stored energy is determined O'
using RODEX2 (Section A.1). [
]
9 The ANF-RELAP calculation provides the thermal-hydraulic boundary conditions for the TOODEE2 code, as shown in Table A.1 and Figure A.I.
G.,:
i A.3 TOODEE2 TOODEE2 is a two-dimensional, time-dependent fuel rod thermal and mechanical ej analysis program. TOODEE2 models the fuel rod as radial and axial nodes with time-I i
dependent heat sources. Heat sources include both decay heat and heat generation via O
reaction of water with zircaloy. The code considers conduction within solid regions of the j
fuel, radiation and conduction across gap regions, and convection and radiation to the coolant and surrounding rods, respectively. Based upon the calculated stress in the e
cladding (due to the differential pressure across the clad) and the cladding temperature, the code determines whether the clad has swelled and ruptured. Once fuel rod rupture is Oi
- determined, the code calculates both inside and outside metal water heat generation.
The outputs of TOODEE2, namely: peak clad temperature, percent local cladding ej i
A-4 Si I
i--
..-,m m.-
O i
O oxidation and percent pin-wide cladding oxidation are compared to the 10 CFR 50.46 (b)
(1) through (3) criteria. Regarding (3), if pin-wide oxidation is less than 1% it is concluded that the criteria ofless than 1% core-wide oxidation is met.
O A.4 DATA PREPARATION AND TRANSFER TOOLS O
Also used with the EXEM PWR Small Break model also are 2 additional codes for obtaining input information and/or transfening results between the basic codes described above:
O 1.
LOOPT 2.
SHAPE /PWR (SHAPE. PUN)
O A.4.1 LOOPT This code is used to determine initial thermal-hydraulic conditions for ANF-RELAP. It 0
is needed because actual plant data, design data, or other safety analysis data are not necessarily available for the initial conditions desired. For example,5% steam generator
.O tube plugging. Flows, pressure drops and temperatures are used to initialize the ANF-RELAP steady-state deck. These LOOPT conditions are not exactly the initial conditions for the accident because ANF-RELAP initialization includes steady-state as well as a null O
transient calculation prior to initiation of the LOCA calculation.
O A-5 10
O-A.4.2 SHAPE /PWR (SHAPE. PUN)
OI SHAPE automates the building ofportions ofinput decks to ANF-RELAP, RODEX2, and TOODEE2. The code prepares input related to the axial power profile. The SHAPE code O.'
can alter and re-normalize a given axial power shape to a prescribed axial peaking factor.
It then generates the axial power factors for input a the RODEX2 and TOODEE2 codes and power fractions for ANF-RELAP.
O.
l 1
O O\\
O Oi Ol 0-A-6 9:
O O
TABLE A.1 INPUT AND OUTPUT FOR THE EXEM/PWR METHODOLOGY COMP' UTER CODES (refer to FIGURE A.1)
[
]
O INPUT:
(1)*
[
]
O OUTPUT:
(2)
[
]
I O
The numbers in this table correspond to the numbers in Figure A.I.
O O
O 1
- O A-7 O
O e
TABLE A.1(Cont'd)
INPUT AND OUTPUT FOR TIIE EXEM/PWR METHODOLOGY COMPUTER CODES (refer to FIGURE A.1)
SHAPE /PWR INPUT:
(2)
[
]
(3)
[
]
OUTPUT:
O (10)
[
]
(4)
[
]
e (7)
[
]
e e
e A-8 e
D 3
TABLE A.1(cont'd)
INPUT AND OUTPUT FOR THE EXEM/PWR METHODOLOGY COMPUTER CODES (refer to FIGURE A.1) d r
RODEX2 INPUT:
3 (4)
[
]
(12)
[
3 0
l OUTPUT:
O (9)
[
O
]
o (8)
[
1 O
A-9 O
O TABLE A.1(Cont'd) e INPUT AND OUTPUT FOR THE EXEM/PWR METHODOLOGY COMPUTER CODES (refer to FIGURE A.1)
G,
[
]
INPUT:
(5)
[
]
9 OUTPUT:
(6)
[
(15)
[
]
9 9
O 91 1
Ol A-10 l
O
- O l
O TABLE A.1(Cont'd)
INPUT AND OUTPUT FOR THE EXEM/PWR METHODOLOGY COMPUTER CODES (refer to FIGURE A.1)
O ANF-RELAP INPUT:
0 (10)
Core power and weighing fractions (8)
[
]
'O (11) o NSSS information (Table 2.2) o ECCS, SG AFW, safety valve flows (Tables 2.6 and 2.9) o Trips and delays (Table 2.7) o Fuel rod / assembly information (Table 2.8)
O o Neutronics information (Tables 2.3,2.4,2.5) 1 OUTPUT:
O (13)
I j
1 O
l i
O l
1 1
l O
A-11 O
)
O TABLE A.1(cont'd)
INPUT AND OUTPUT FOR THE EXEM/PWR METHODOLOGY COMPUTER CODES (refer to FIGURE A.1)
)
e' TOODEE2 INPUT:
(9)
[
g e-
}
(7)
[
g
)
(13)
[
e:
l e1 OUTPUT:
(14)
[
e1 l
e:
A-12 e
O O
O 1
O O
O
- O
\\
O O
Figure A-1 TU Electric's SBLOCA Analysis Computer Code Interface
.O (Numbers in Circles Conespond to those m Table A-1)
A-13 O
. _