NUR]G/IA-0148

International

Agreement Report

Assessment of RELAP5/MOD3.1

Using LSTF Ten-Percent

Main Steam-Line-Break

Test Run SB-SL-01

Prepared by

J. G. Oh, H. D. Lee, K. K. Jee, S. K. Kang/KOPEC

Y. S. Bang, K. W Seul/KINS

H. Kumamaru, Y. Anoda/JAERI

Korea Power Engineering Company Korea Institute of Nuclear Safety

150 Duckjin-Dong, Yusong-Ku P.O. Box 114

Taejon, Korea 305-353 Yusong, Taejon

Korea 305-600

Japan Atomic Energy Research Institute

Tokai-Mura, Naka-Gun

Ibaraki-Ken 319-1195, Japan

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

September 1998

Prepared as part of

The Agreement on Research Participation and Technical Exchange under the

International Thermal-Hydraulic Code Assessment and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
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owned rights.
NUREG/IA-.0148

International

Agreement Report

Assessment of RELAP5/MOD3.1

Using LSTF Ten-Percent

Main Steam-Line-Break

Test Run SB-SL-01

Prepared by

J. G. Oh, H. D. Lee, K K Jee, S. K Kang/KOPEC

Y. S. Bang, K W. Seul/KINS

H. Kumamaru, Y. Anoda/JAERI

Korea Power Engineering Company Korea Institute of Nuclear Safety

150 Duckjin-Dong, Yusong-Ku P.O. Box 114

Taejon, Korea 305-353 Yusong, Taejon

Korea 305-600

Japan Atomic Energy Research Institute

Tokai-Mura, Naka-Gun

Ibaraki-Ken 319-1195, Japan

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

September 1998

Prepared as part of

The Agreement on Research Participation and Technical Exchange under the

International Thermal-Hydraulic Code Assessment and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
NUREG/IA-0148 has been reproduced

from the best copy available.
ASSESSMENT OF RELAP5/MOD3.1 USING LSTF 10% MAIN

STEAM LINE BREAK TEST RUN SB-SL-01

Abstract

Results produced by the RELAP5/MOD3.1 computer code were compared with

the experimental data from JAERI's LSTF Test Run SB-SL-01 for a 10% main steam

line break transient in a pressurized water reactor. The code simulation for the base case

included a total of 189 fluid control volumes and 199 flow junctions to model the

transient two-phase flow phenomena. Also, a total of 180 heat slabs were used tc. model

the system heat transfer. The code predictions of the experimental results are generally

satisfactory for the trends of key parameters.

Sensitivity studies performed for the break discharge coefficient, the separator

drain line loss coefficient, and the number of steam generator nodes did not reveal any

strong dependencies. Nevertheless, optimal values of these parameters that led to the

lowest overall statistical error were obtained, and these values were subsequently used

in the "Base Case" analysis.

i
Table of Contents

Abstract ............................................................................................................................. i

Table of Contents ....................................................................................................... ii

List of Tables .................................................................................................................. iii

List of Figures ................................................................................................................ iii

Sum m ary of Study .......................................................................................................... vi

1. Introduction ................................................................................................................. 1

2. Experimental Facility Description

(Large Scale Test Facility ; LSTF) ........................................................................ 5

3. Test D escription (Experim ent SB-SL-01) ............................................................ 7

4. D escription of Code V ersion and Input D eck ..................................................... 9

4.1 Input M odelling ................................................................................................ 9

4.2 Initial and Boundary Conditions ..................................................................... 11

5. R esults from Base Calculation ........................................................................... 12

5.1 Break Junction Flow Behavior ....................................................................... 12

5.2 V essel Upperhead Voiding .............................................................................. 13

5.3 RCS and Secondary Pressure Response .......................................................... 14

5.4 Therm al Response ........................................................................................... 15

5.5 W ater Level Behavior .................................................................................... 16

5.6 Loop M ass Flow Behavior ............................................................................. 17

6. Discussion of Statistical R esults ........................................................................... 19

7. Sensitivity Studies ................................................................................................ 22

7.1 Break Junction D ischarge Coefficient (Cd) ................................................... 23

7.2 Separator drain line loss coefficient (K) ........................................................ 23

7.3 Steam Generator N odalization ...................................................................... 24

8. R un Statistics ............................................................................................................. 26

9. Conclusion and R ecom m endation ........................................................................ 27

R eferences ...................................................................................................................... 29

Tables .............................................................................................................................. 30

Figures ............................................................................................................................ 40

ii
List of Tables

3.1 Sequence of Events for Experiment SB-SL-01 ....................... 30

3.2 Initial and Boundary Conditions for Experiment SB-SL-01 .............. 31

5.1 List of Assessment Parameters (LSTF SB-SL-01) ........................ 32

6.1 The Mean Errors ( E ) for the Base Calculation (Cd=0.85, K=10.0, n=5) ... 33

6.2 The Root-Mean-Square Errors ( E nMS)

for the Base Calculation (Cd=0.85, K=I 0.0, n=5) ....................... . 34

6.3 The Standard Deviations of the Errors (a )

for the Base Calculation (Cd=0.85, K=10.0, n=5) ....................... . 35

7.1 The Root-Mean-Square Errors ( s) for the Sensitivity Study

on the Discharge Coefficient (Cd) ....................................... .. 36

7.2 The Root-Mean-Square Errors (e Rms) for the Sensitivity Study

on the S/G Drain Line Loss Coefficient (K) ................................ 37

7.3 The Root-Mean-Square Errors (e E s) for the Sensitivity Study

on the Number of Steam Generator Nodes (n) ............................... 38

8.1 Run Statistics Data in Base Calculation ................................... 39

List of Figures

2.1 Schematic View of ROSA-VI LSTF ........................................ 40

3.1 Three Core Power Control Curves for LSTF ............................. 41

4.1 RELAP5/MOD3.1 Nodalization of LSTF for MSLB Transient Assessmenl 42

5.1 Break Flow Rate .................................................................. 43

5.2 Break Void Fraction .............................................................. 44

5.3 Break Density .................................................................... 45

5.4 Vessel Upperhead Void Fraction ............................................. 46

5.5 Pressurizer Pressure ........................................................... 47

5.6 Volume Averaged Void Fraction of Three Upperhead Nodes ............. 48

5.7 Downcomer Flow Rate of Intact Loop ......................................... 49

5.8 Secondary Pressure of Broken Loop .......................................... 50

5.9 Secondary Pressure of Intact Loop .......................................... 51

5.10 Cold leg Temperature of Intact Loop ......................................... 52

iii
5.11 Cold leg Temperature of Broken Loop ......................................... 53

5.12 Hot leg Temperature of Intact Loop ............................................ 54

5.13 Hot leg Temperature of Broken Loop .......................................... 55

5.14 SG Steam Dome Temperature of Intact Loop ................................ 56

5.15 SG Downcomer Temperature of Intact Loop ................................. 57

5.16 SG Downcomer Temperature of Intact Loop (4 loops) ...................... 58

5.17 SG Steam Dome Temperature of Broken Loop ............................ 59

5.18 SG Downcomer Temperature of Broken Loop ................................ 60

5.19 SG Downcomer Temperature of Broken Loop (4 loops) ...................... 61

5.20 SG Boiler Bottom Temperature of Broken Loop .......................... 62

5.21 SG Boiler Middle Temperature of Broken Loop ............................. 63

5.22 SG Boiler Top Temperature of Broken Loop ................................. 64

5.23 SG Boiler Bottom Temperature of Intact Loop ........................... 65

5.24 SG Boiler Middle Temperature of Intact Loop ................................ 66

5.25 SG Boiler Top Temperature of Intact Loop ................................. 67

5.26 Average Temperature of RCS Broken Loop ................................ 68

5.27 Collapsed Liquid Level of Pressurizer ........................................ 69

5.28 Narrow Range of Liquid Level of S/G Intact Loop ........................... 70

5.29 Narrow Range of Liquid Level of S/G Broken Loop ........................ 71

5.30 Liquid Level of S/G Intact Loop ............................................. 72

5.31 Liquid Level of S/G Broken Loop ........................................... 73

5.32 Pump Suction Flow Rate of Broken Loop ..................................... 74

5.33 Pump Suction Flow Rate of Intact Loop .................................... 75

5.34 Downcomer Flow Rate of Broken Loop .................................... 76

5.35 Separator Flow Rate of Intact Loop ......................................... 77

5.36 Separator Flow Rate of Broken Loop ......................................... 78

7.1 Comparison of Break Flow Rate (Cd Value Sensitivity) ..................... 79

7.2 Comparison of Pressurizer Pressure (Cd Value Sensitivity) .................. 80

7.3 Comparison of Secondary Pressure of Broken Loop (Cd Value Sensitivity) 81

7.4 Comparison of Secondary Temperature of Broken Loop

(Cd Value Sensitivity) ............................................................. 82

7.5 Comparison of Break Void Fraction (Cd Value Sensitivity) ................. 83

7.6 Comparison of Secondary Pressure of Intact Loop (Cd Value Sensitivity) 84

7.7 Comparison of Secondary Temperature of Intact Loop

iv
(Cd Value Sensitivity) .......................................................... 85

7.8 Collapsed Liquid Level of Pressurizer (Cd Value Sensitivity) ............... 86

7.9 Downcomer Flow Rate of Intact Loop (Cd Value Sensitivity) ............ 87

7.10 Downcomer Flow Rate of Broken Loop (Cd Value Sensitivity) . ......... 88

7.11 Comparison of Break Flow Rate (K Value Sensitivity) ................... 89

7.12 Comparison of Pressurizer Pressure (K Value Sensitivity) ................. 90

7.13 Comparison of Secondary Pressure of Broken Loop (K Value Sensitivity). 91

7.14 Comparison of Secondary Temperature of Broken Loop

(K Value Sensitivity) ............................................................ 92

7.15 Comparison of Break Void Fraction (K Value Sensitivity) ............... 93

7.16 Comparison of Secondary Pressure of Intact Loop (K Value Sensitivity) ... 94

7.17 Comparison of Secondary Temperature of Intact Loop

(K Value Sensitivity) ............................................................ 95

7.18 Collapsed Liquid Level of Pressurizer (K Value Sensitivity) ............. 96

7.19 Downcomer Flow Rate of Intact Loop (K Value Sensitivity) ............. 97

7.20 Downcomer Flow Rate of Broken Loop (K Value Sensitivity) ............ 98

7.21 Comparison of Break Flow Rate (S/G Nodes Sensitivity) .................. 99

7.22 Comparison of Pressurizer Pressure (S/G Nodes Sensitivity) .............. 100

7.23 Comparison of Secondary Pressure of Broken Loop

(S/G N odes Sensitivity) .......................................................... 101

7.24 Comparison of Secondary Temperature of Broken Loop

(S/G Nodes Sensitivity) .......................................................... 102

7.25 Comparison of Break Void Fraction (S/G Nodes Sensitivity) ................ 103

7.26 Comparison of Secondary Pressure of Intact Loop

(S/G N odes Sensitivity) ......................................................... 104

7.27 Comparison of Secondary Temperature of Intact Loop

(S/G N odes Sensitivity) .......................................................... 105

7.28 Collapsed Liquid Level of Pressurizer (S/G Nodes Sensitivity) ........... 106

7.29 Downcomer Flow Rate of Intact Loop (S/G Nodes Sensitivity) ........... 107

7.30 Downcomer Flow Rate of Broken Loop (S/G Nodes Sensitivity) ......... 108

7.31 Steam Generator Nodalization Diagram for The Nodes Sensitivity Study .. 109

7.32 The Required CPU Time for Calculation (S/G Nodes Sensitivity) ........... 110

8.1 The Required CPU Time in The Base Calculation ............................ I11

V
Summary of Study

This report presents a comparison of RELAP5/MOD3.1 computer code results to

the experimental data from the Japan Atomic Energy Research Institute Large Scale

Test Facility(JAERI LSTF) Test Run SB-SL-01 for simulating a 10% main steam line

break (MSLB) transient in a pressurized water reactor (PWR). The LSTF volumes are

scaled at 1/48 of a typical 3423 MWt Westinghouse 4-loop PWR, however the height

and relative elevations of each component are full scale. The LSTF core consists of

1104 full length electrical heater rods, which are used to simulate PWR nuclear fuel

rods. Greater details about the LSTF design and the scaling used therein are provided in

Section 2, "Experimental Facility Description", and the procedures used are given in

Section 3, "Test Description".

The RELAP5/MOD3.1 nodalization for benchmarking the LSTF Test Run SB-SL01 included a total of 189 fluid control volumes and 199 flow junctions to simulate the

thermal-hydraulic phenomena. A total of 180 heat slabs were used to model the system

heat transfer phenomena through the heater rods, the steam generator tubes, the internal

structural components, and the vessel shell. Further details pertaining to the input

modeling can be found in Section 4, "Description of Code Version and Input Deck".

The code initialization via a steady state simulation is described therein as well.

Sensitivity studies were performed with the RELAP5/MOD3.1 computer code to

investigate the effect of the break flow discharge coefficient (Cd), the steam separator

drain line loss coefficient (K), and the number of steam generator nodes (n) upon the

thermal hydraulic performance of the code. This sensitivity study revealed only a

marginal dependence of the predictive capabilities of RELAP5/MOD3.1 on these three

parameters. Nevertheless, the following optimal values of these parameters were

selected based on minimum overall statistical error: Cd = 0.85; K = 10.0; n = 5. A more

detailed account of the sensitivity analysis is given in Section 7, "Sensitivity Studies".

vi
The "Base Calculation" investigation of the 10% MSLB transient uses the values

Cd = 0.85, K = 10.0, and n = 5 for the code simulation. Comparison between the

RELAP5/MOD3.1 code predictions and the LSTF Test Run SB-SL-01 experinental

data are described in Section 5, "Results from the Base Calculation". The quantification

of these comparisons via statistical error evaluations are given in Section 6, "Discussion

of Statistical Results". Salient features of the results of these comparisons are pr.vided

below.

On the whole, the code-data comparisons based on gross features are reas:)nably

good. The code, however, appears to have difficulty in accounting for finer details

which leads to significant deviations from the experimental results in certain important

cases. As an example, the RELAP5/MOD3.1 code cannot predict steep variable &aanges

or sharp gradients. A manifestation of this occurs at the time of High Pressure Safety

Injection(HPSI), as outlined in the following summary description.

The break flow rate is predicted well by the code, except toward the end of the

transient. The hot leg temperature of the intact and broken loops, the cold leg

temperature of the broken loop, the steam generator (SG) downcomer temperal.tre of

the intact loop, and the average temperature of the reactor coolant system (RCS) 'broken

loop are all predicted reasonably well by the RELAP5/MOD3.1 computer code as

compared to the SB-SL-01 experimental data for a MSLB transient.

The code prediction of the cold leg temperature of the intact loop is in reasonable

agreement with the experimental data until the initiation of HPSI. Subsequently, the

code predictions and the experimental data show considerable divergence because the

RELAP5/MOD3.1 code cannot track the steep decrease in temperature with time which

occurs HPSI. A similar divergence occurs for the time variation of the pressurizer

pressure, secondary pressure of the intact loop, and the volume-averaged void fraction

of the three upperhead nodes, with respect to the initiation of HPSI. The code

predictions of the broken loop SG temperature diverge from the experimental values

vii
essentially from the very beginning of the transient. At the end of the transient, the

temperature discrepancy is about 11%. The collapsed liquid level of the pressurizer is

predicted rather poorly by the code following initiation of HPSI.

The computer code substantially underpredicts the time rate of change of the liquid

level, leading to an underprediction of the collapsed liquid level by about 27% at the

end of the transient. The RELAP5/MOD3.1 output relating to the intact loop

downcomer and separator flow rates also deviates considerably from the SB-SL-01

experimental data.

viii
1. Introduction

The objective of the investigation reported here was to evaluate the accuracy of the

thermal hydraulic computer code RELAP5/MOD3.1 in simulating a design basis MSLB

PWR transient. The output from this computer code was compared to the experimental

data from Test Run SB-SL-01 of the ROSA-IV Large Scale Test Facility (LSTF)

operated by Japan Atomic Energy Research Institute(JAERI). The experiment Test Run

SB-SL-01 was the simulation of the 10% main steam line break (MSLB)

Simulation of a postulated main steam line break (MSLB) is important from the

perspective of potential plant damage and related environmental consequences, although

the probability of occurrence of such an event is considered to be small. The closure of

the main steam isolation valves (MSIVs) following a MSLB will prevent blowdown of

the secondary plant system for all breaks except those for which the break is located

between the steam generator (SG) outlets and the MSIVs. As a result, only the broken

loop SG secondary side will blowdown during such an accident. This SG secondary side

acts as a heat sink owing to the decreasing pressure.

This may lead to excessive cooling and subsequent re-pressurization and

"pressurized thermal shock" of the reactor coolant system (RCS), which poses a

potential threat to the integrity of the reactor vessel (RV). Other potential damage stems

from the possibility of a "reactivity power excursion", that is, return to power via

aqddition of reactivity, as a consequence of the excessive cooling of the reactor coolant

system (RCS). Finally, the possibility of SG tube failure arises because lowering the SG

shell side pressure causes a greater pressure difference across these tubes. Computer

code simulations and calculations are needed for an examination and understarding of

these phenomena and the potential severity of the consequences of an accident since it is

I
not generally feasible to perform full scale tests. And for assessment and verification of

code calculations, it is imperative that the code calculations be compared against

integral system data. From the viewpoint of reactor safety, it is very important to verify

the predictive capability of thermal-hydraulic safety analysis codes for transient

accidents.

The LSTF was designed as a 1/48 scale, integral test/experimental facility for

simulating the response of a PWR to design basis loss-of-coolant-accidents (LOCAs)

and plant transients. The Test Run SB-SL-01, conducted during 1990 at the LSTF,

simulated experimentally a 10% MSLB transient in a PWR.

In modeling the 10% MSLB transient Test Run SB-SL-01 with RELAP5/MOD3.1,

it was recognized at the outset that the following variable parameters might affect the

code-data comparisons: the break discharge coefficient (Cd), the steam separator drain

line loss coefficient (K), and the number of S/G nodes (n). Hence, sensitivity studies

were performed with the RELAP5/MOD3.1 computer code using the following set of

values for the three sensitivity parameters:

Cd = {0.75, 0.80, 0.85, 0.90}; K = {0.0, 10.0, 50.0, 100.0}; n = {3, 5, 7, 9}.

The results of these studies did not indicate a pronounced dependence of the code

output on the sensitivity parameters. Nevertheless, based on certain statistical error

criteria or indices used, it was concluded that the following values of the sensitivity

parameters produced the minimum overall discrepancy between the computer code

predictions and the experimental data values: Cd = 0.85; K = 10.0; and, n = 5.

The "Base Calculation" investigation of the 10% MSLB transient used these

optimal values of Cd, K, and n. On the whole, the code-data comparisons from this

2
investigation are reasonably satisfactory. The code, however, has difficulty in predicting

the finer details of the physical phenomena involved in a MSLB transient. In turn, this

leads to significant deviations between the code output and the experimental data in

some important cases.

The break flows are predicted accurately by the code, except towards the end of

the transient. Also predicted reasonably well by the RELAP5/MOD3.1 computer code

are the following time variations: the hot leg temperature of the intact and broken loops,

the cold leg temperature of the broken loop, the SG downcomer temperature of the

intact loop, and the average temperature of the RCS broken loop.

The code prediction of the intact loop cold leg temperature is in reasonable

agreement with the experimental data until the initiation of high pressure safety

injection (HPSI). Subsequently, the code output and the experimental data diverge

significantly, as the RELAP5 computer code cannot model steep variable chamge or

sharp gradients associated with the initiation of HPSI. A similar situation, with respect

to initiation of HPSI, exists with the time variations of the pressurizer pressure,

secondary pressure of the intact loop, and the volume averaged void fraction of the three

reactor vessel (RV) upperhead nodes. The code predictions of the broken loop SG

temperature diverge from the experimental values from the very beginning of the

transient, culminating in a difference of 11% between the code output and the test data

values at the end of the transient. The pressurizer collapsed liquid 1hvel is

underpredicted by the code following HPSI, and at the end of the transient this

underprediction amounts to 27% of the experimental value. The intact loop downcomer

and separator flow rates produced by the code also deviate significantly from the test

values.

3
The LSTF experimental facility is described in Section 2, while the Test Run SBSL-01 is discussed in Section 3. The input deck modeling and initialization of the code

are discussed in Section 4. The results of the "Base Calculation" are provided in Section

5. The evaluation of statistical errors, based on difference between the code and

experimental data for various time values, is given in Section 6. A discussion of the

sensitivity studies regarding the optimal values of Cd, K and n can be found in Section 7.

The run statistics are given in Section 8.

4
2. Experimental Facility Description

(Large Scale Test Facility; LSTF)

The Rig Of Safety Assessment (ROSA)-IV Program's Large Scale Test Facility

(LSTF) is a test facility for integral simulation of thermal-hydraulic response of a

pressurized water reactor (PWR) during small break loss-of-coolant accidents

(SBLOCAs) and plant transients. The PWR core nuclear fuel rods are simulated by

using electrical heater rods in the LSTF.

A brief outline of the ROSA-IV program's LSTF is given below, with special

emphasis on its suitability for simulating the main steam line break transient.

The LSTF experimental facility was designed to model the thermal-hydraulic

phenomena in a PWR during postulated small break LOCAs and plant transients. The

general facility view is shown in Figure 2.1 [1].

The LSTF volumes are scaled at 1/48 of a typical 3423 MWt Westinghouse 4-loop

PWR reference plant. The height of each component and the relative elevations are full

scale, (i.e., they are equal to the corresponding components and elevations of the

reference plant) in order to correctly simulate the coolant flow under natural circulation

conditions. The flow areas are also scaled by 1/48 in the pressure vessel, and by 1/24 in

the two steam generators. However, the hot and cold legs are scaled so as to conserve

the ratio of the length to the square root of pipe diameter for the reference PWR. Such

an approach was used in order to simulate the flow regime transitions in the primary

loops. The core power is scaled by 1/48 at power equal to 14% of the reference four

loop PWR. The LSTF core consists of 1104 full length (3.66m) electrical heater rods

(including 1008 powered rods and 96 unpowered rods) placed in a 17 X 17 array, as in

the nuclear reactor core of the reference plant. Geometric parameters such as diameter,

rod-to-rod pitch, and rod heated length in the test core are equal to the corresponding

values in the reactor core, so as to preserve the heat transfer characteristics of the core.

5
The test core has a rated thermal output of 10 MW.

Generation of non-condensible gases in the core such as hydrogen is simulated by

injection of nitrogen gas into the core. The broken loop in which the steam line break

occurs, represents a single loop of the reference plant, while the other three loops are

represented by a single intact loop. The steam generators and secondary coolant systems

are designed to simulate both steady state and transient responses of the 2/48 scaled

steam and feedwater flows, as well as the scaled primary-to-secondary heat transfer in

the reference PWR. The 141 U-tubes in each steam generator are arranged in a square

array, and they consist of nine groups of U-tubes with different heights.

The inner diameter and wall thickness of each U-tube are 19.6 mm and 2.9 mm,

respectively. The pressurizer is located on the hot leg connected to the intact steam

generator. It is scaled in accordance with the facility volume scaling requirements. Also,

the height to diameter ratio of the pressurizer is conserved in relation to the reference

PWR.

The system break in the reference PWR is simulated in the LSTF by using a break

unit. The break unit consists of a venturi flow meter, a spool piece to measure the twophase break flow rate and density, a break orifice and a break valve. The break valve is

designed to open in less than 0.1 second.

6
3. Test Description (Experiment SB-SL-01)

Experimental run SB-SL-01 was conducted during 1990 in the LSTF, located at

the Japan Atomic Energy Research Institute (JAERI). This experiment simulated a 10%

main steam line break transient in a pressurized water reactor (PWR), and it was

initiated by manual operation at the beginning of the break.

The break was located in the main steam line of steam generator B. The break

diameter was 31.9 mm, which corresponds to 10% of the cold leg flow area. The

experiment SB-SL-01 was carried out at full power. Three core power control curves

were used for the LSTF experiments, as shown in Figure 3.1 [2]. The JAERI core power

curve, in which the delayed neutron fission power is estimated conservatively, was used

for the test run SB-SL-01.

The, reactor trip was initiated by manual operation at the beginning of the break,

rather than by reactor trip signal generation following the break. Automatic protective

actions taken during the early period of the main steam line break included closure of

MSIVs, steam generator auxiliary feedwater injection, and safety injection. The NMSIVs

closure was initiated by manual operation at the time of the break, rather than by safety

injection signal (SIS) generated on a low pressure of 4.24 MPa in the broken loop

secondary side. The auxiliary feedwater flow, delayed by the loss of off-site power,

started refilling the intact loop steam generator at 28 seconds after the break. The HPSI

was designed so as to change the injection location during the test. This system included

two pumps: a high pressure injection pump (PH) and a charging pump (PJ). The safety

injection flow was initiated manually at 1156 seconds, rather than by automatic SIS.

The sequence of main events for the test run SB-SL-01 is summarized in Table 3.1 [3].

Prior to the initiation of the experiment, LSTF facility was set to have a primary

system pressure of 15.52 MPa, a primary system hot leg and cold leg temperature of

598.1K and 562.4K, respectively, and a primary coolant flow rate of 24.3 kg/se.c. The

7
steam generator secondary side pressure and temperature were set to 7.3 MPa and

495.2K, respectively, and the feedwater flow rate was 2.74 kg/s. The initial test core

power level was 10 MW. Table 3.2 summarizes the initial conditions used for test run

SB-SL-01.

8
4. Description of Code Version and Input Deck

The RELAP5 code is based on a non-homogeneous and non-equilibrium model for

one dimensional, two-phase system. Recently, the RELAP5/MOD3 code development

program has been initiated to develop a code version suitable for the analysis of all

transient and postulated accidents in PWR systems. In this code assessment on the

MSLB, the unmodified released code version, RELAP5/MOD3.1 is used.

A RELAP5/MOD3.1 input deck specific to experimental run SB-SL-01 was

created from the reference input deck of LSTF. This was accomplished via an iterative

process of modifications and updates, using input values and options appropriate to the

experimental data and analyses

4.1 Input Modelling

The RELAP5/MOD3.1 model of the LSTF facility for experimental run SB-SL-01

included 189 fluid control volumes and 199 flow junctions. The system nodalizition is

illustrated schematically in Figure 4.1. A total of 180 heat slabs (shown as shaded areas

in Figure 4.1) were used in the model to represent heat transfer in the steam generator,

reactor, primary system piping, and pressurizer.

The volumes in the reactor vessel numbered 100 to 156. Flow enters from the cold

legs into the vessel downcomer annulus in branch 104. The primary reactor vessel flow

path is downward through the downcomer (annulus 108) to the lower plenum (volumes

112 and 116). A portion of the inlet flow is bypassed directly to the hot legs through the

slip-fit between the core barrel assembly and reactor vessel wall at the hot leg nozzles.

Another portion of the inlet flow is diverted upward through volumes 101 and 100 to

the upper head (volumes 148 and 152). The upper plenum is represented by volumes

128 to 140. Volume 156 represents the guide tubes that route a portion of the core exit

9
flow from the upper plenum to the upper head. The reactor core hot channel (volume

124) is subdivided into nine equally spaced volumes. Heat structures were used to

represent the fuel pins, the major internal structures (thermal shield, core barrel wall,

guide tube walls, etc.), and the reactor vessel cylindrical shell and spherical heads. The

use of these heat structures allows simulation of inter-region heat transfer, such as heat

transfer between the core and bypass regions.

The steam generators are represented by nine volumes in the primary system and

thirteen volumes in the secondary system. Heat is exchanged between the primary and

secondary sides of the steam generator via the U-tubes, which are modelled as heat

structures. The multi-volume model for the steam generators is quite important for

simulating the heat transfer phenomena accurately.

The main and auxiliary feed water systems are modelled by time dependent

volumes (volumes 360 and 350 in the broken loop and volumes 560 and 550 in the

intact loop, respectively). The feed water and steam flow rates are regulated by 'control

logic' in order to maintain the required secondary. pressure and liquid level.

The pressurizer is modelled by surge line (volume 600) and by pressurizer vessel

(volume 610). The pressurizer vessel is divided into eight subvolumes.

The hot leg is modelled by seven volumes (volumes 202 to 216 in the broken loop

and volumes 402 to 416 in the intact loop, respectively), while the cold leg is modelled

by three volumes (volumes 244 to 252 in the broken loop and volumes 444 to 452 in the

intact loop, respectively). The pumps in each loop are represented by volumes 240 and

440 respectively.

The HPSI, LPSI and charging tank are modelled by time dependent volumes

numbered 760 to 780 in the broken loop and 720 to 750 in the intact loop, respectively.

The accumulator (volume 700) is connected to the cold legs of both loops via the

distributor (volume 710). The control valve in the accumulator line (valve 711 and 712)

serves to isolate the accumulator when the accumulator is empty.

10
On the secondary side of steam generator B, the 10 % main steam line break was

simulated by opening a valve (junction 915) at break time. The break diameter was 31.9

mm, and the discharge coefficient was adjusted to a value of 0.85 to match the break

flow rate, via results obtained from the sensitivity studies. In view of the importamce of

the primary system pressure calculation, pressure control systems associated ,aith the

pressurizer (such as heaters, sprays, PORVs, and SRVs) were included in the model.

The input deck for the base case calculation listed in Appendix A is based on the

RELAP5/MOD3 code manual [4].

4.2 Initial and Boundary Conditions

A RELAP5/MOD3.1 steady state simulation was performed in order to obtain

appropriate steady state system conditions prior to the initiation of a steam line break. A

comparison between the simulated initial test conditions thus obtained (from tlhe code

run) and the corresponding measured initial test conditions are given in Table 3.2. In

general, the agreement between the simulated and the measured initial test conditions is

satisfactory. However, the primary coolant pump speed is an exception. The

RELAP5/MOD3.1 prediction of the pump speed is lower than that measured in the test.

The reasons for the underpredicted coolant pump speed are not clear. However, this

discrepancy is believed to have negligible effect on the overall system transient. All

relevant information pertaining to the initial and boundary conditions was incorporated

in the steady-state and transient input decks for the RELAP5/MOD3.1 computer code.

11
5. Results from Base Calculation

The RELAP5/MOD3.1 computer code calculations were compared against the

experimental data from JAERI's LSTF test run SB-SL-01. The initial and boundary

conditions used in the computer code simulations were obtained from a steady state run.

The sequence of thermal-hydraulic events during the transient calculation are

summarized in Table 3.1. The corresponding experimental data values are also

summarized in this table for comparison. Also, the computer code simulations used a

break discharge coefficient value of 0.85, and a separator drain line loss coefficient

value of 10.0, for the base calculation.

The assessment of the RELAP5/MOD3.1 computer code against the experimental

data is described below in relation to several important physical parameters such as

break flow rate, pressurizer pressure and steam generator secondary side temperature.

Table 5.1 summarizes the list of assessment parameters. For each listed assessment

parameters, the identification numbers for the RELAP5/MOD3.1 calculation and the

corresponding LSTF SB-SL-01 measurement are provided in the first two columns,

respectively. The uncertainty in measurement associated with each assessed parameter

is given in column 3. The last column identifies the figure which shows a comparison

between the code calculation and the corresponding experimental data for each

assessed parameter.

5.1 Break Junction Flow Behavior

Figure 5.1 -Figure 5.3 show a comparison between the experimental data and the

calculated results for the steamline break junction. The computer code-test data

comparisons are generally good, except for about 50 seconds following initiation of the

transient. During the initial 50 seconds, the calculated break flow was overpredicted by

12
0.3 )- 1.1 kg/sec, which is equivalent to an overprediction of 5 - 16% in relatior. to the

experimental data. The code predicts a constant void fraction value of 1.0 at the break

junction, while the small deviation of the experimental data from this value could be

attributed to uncertainties associated with experiments and the processing of te:;t data.

The code predictions of the mass flow rate (Figure 5.1) and appropriately scaled mass

density variation (Figure 5.3) are essentially identical. This attests to the validity of the

results in Figure 5.2, namely that the* break flow void fraction is essentially c:nstant

throughout the transient.

5.2 Vessel Upperhead Voiding

For severe cooldown events such as MSLB, reactor vessel upperhead should be

modeled carefully such that any voiding in the upperhead is monitored. The vessel

upperhead modeling affects the system pressure significantly due to voiding effect [5, 6,

7]. Following the initial rapid depressurization, the process slows down as the vessel

upperhead reaches saturation and begins to flash.

Figure 5.4 shows RELAP5/MOD3.1 code output pertaining to void fraction

variation with time in the three vessel upperhead nodes. The bottom node is full of

water at all times, while the same holds for the middle node except between 1089

seconds and 1361 seconds (following the initiation of the transient) when the void

fraction varies between 0.0 and 0.14. The void fraction for the top node in the vessel

upperhead varies between 0.0 and 1.0 for times between 721 seconds and 3241 seconds,

reaching the peak value between 1097 seconds and 1265 seconds. During other times,

the top node in the vessel upperhead is full of water.

The volume averaged void fraction in the three vessel upperhead nodes is shown in

Figure 5.6. An examination of the code output and the experimental data from the plots

indicates significant differences between RELAP5/MOD3.1 predictions and the test

13
results. The experimental data always indicate presence of steam in a two-phase mixture,

while the code output cannot confirm the presence of any steam at all for the first 750

seconds of the transient. During this initial time frame (of 750 seconds), the

experimental data indicate a void fraction of about 0.05 (or 5%). Conversely, while the

experimental data indicate an essentially linear variation of the void fraction with time,

with values between about 0.025 and 0.04 for times between around 1350 seconds and

3500 seconds, the code predicts an essentially linear variation with values varying

between 0.24 and 0.0 for practically the same time interval. Again, the experimental

void fraction variation around 1350 seconds is very steep, while the predicted code

variation is very gradual.

5.3 RCS and Secondary Pressure Response

The RELAP5/MOD3.1 code prediction regarding the primary side pressure agrees

reasonably well with the experimental data for the initial 800 to 1000 seconds of the

transient as can be seen from Figure 5.5. Thereafter, while the rate of increase of

pressure is approximately the same for the code calculation and the experimental data,

the code overpredicts the pressure by about 1 MPa. According to Figure 5.5, a similarly

favorable comparison between code predictions and experimental data holds for the

secondary side pressure variation with time for the intact loop during the initial 800

seconds. Subsequently, however, the pressure for this case is underpredicted by about

0.4 MPa after about 1000 seconds. This underprediction of pressure by the code after

about 800 seconds appears to be related to the code overprediction of primary side

pressure after about 800 seconds (Figure 5.5), and the overprediction of intact loop

downcomer flow rate around this time (Figure 5.7).

On the other hand, the secondary side depressurization transients for the both

broken loop and intact loop are predicted quite well by the RELAP5/MOD3.1 computer

14
code in relation to the experimental variation (Figure 5.8 and 5.9, respectively). This

transient involves a monotonically decreasing variation of pressure with time, on

account of flow of steam through the pipe break.

5.4 Thermal Response

Following the closure of the main steam isolation valves (or MSIVs), the main

steam line break transient involves blowdown of the two-phase fluid mixture only on

the broken loop steam generator secondary side. Thus, the steam generator tubes in the

broken loop are cooled at a fast rate, thereby causing rapid cooling of the primary

system.

The intact loop cold leg temperature variation with time is shown in Figure 5.10.

The code predictions show excellent agreement with the experimental data until the

time of HPSI injection at around 1170 seconds into the transient. These comparisons are

significantly less satisfactory as the time rate of change of temperature for ihe code

prediction and the experimental data are substantially different. The code successively

overpredicts and underpredicts the experimental temperature values, leading to an

underprediction of about 45 C at 3500 seconds. The colder HPSI liquid causes a nearly

instantaneous steep change in the observed temperature, leading to lowering of the cold

leg temperature by about 541C. The RELAP5/MOD3.1 code underpredicts this

temperature drop by a substantial amount.

The broken loop cold leg temperatures involve a uniformly-decreasing monotonic

variation with time, as can be seen from Figure 5.11, on account of the blowdovn of the

secondary side of the isolated, broken loop steam generator. For this case, the

RELAP5/MOD3.1 predictions show excellent agreement with the experimental data.

The hot leg temperature variation of the intact and broken loops is shown in Figures

5.12 and 5.13, respectively. For both of these cases, the RELAP5/MOD3.1 predictions

15
show very good agreement with the test data, and the two transients are virtual

facsimiles of one another.

The intact loop steam generator steam dome temperature is shown in Figure 5.14.

The code output shows excellent agreement with the test data for the initial 800 seconds

approximately. Thereafter, the code output underpredicts the measured temperatures by

about 15 *C. The code predictions of the intact loop steam generator downcomer

temperatures are in reasonable agreement with the experimental data, as shown in

Figures 5.15 and 5.16, which indicate steady-state conditions around 1600 seconds.

The broken loop steam generator steam dome temperature comparisons are shown

in Figure 5.17, and these are not satisfactory. The code underpredicts the experimental

data, and the discrepancy gets progressively worse with time, culminating in an

underprediction of around 501C at the end of the transient. The broken loop steam

generator downcomer temperature comparisons are shown in Figures 5.18 and 5.19. The

code again underpredicts the data, but the discrepancy here is small.

The steam generator boiler temperature-time variations for the broken loop are

shown in Figures 5.20 -5.22 and those for the intact loop are shown in Figures 5.23-

5.25. The overall comparisons are satisfactory, but the RELAP5/MOD3.1 temperature

predictions are generally lower than the corresponding experimental data. The

discrepancy typically ranges between 5 IC and 10 IC, but for the top and middle of the

broken loop boiler, it is as large as 201C. The average temperature comparisons for the

RCS broken loop are shown in Figure 5.26, and these are quite satisfactory.

5.5 Water Level Behavior

The code-data comparisons for the collapsed liquid level of the pressurizer are

shown in Figure 5.27. The comparisons are excellent for the initial 1300 seconds of the

transient with respect to (i) the collapsed liquid level, (ii) the time when the pressurizer

16
becomes empty, and (iii) the time when the pressurizer starts to refill. However, beyond

the HPSI injection time when the refill process gets initiated, the comparisons are less

satisfactory as the code underpredicts the rate of refill significantly. Towards th-. end of

the transient, the code underpredicts the collapsed liquid level by about 1.8m.

Figures 5.28- 5.31 show the steam generator collapsed liquid level of the intact

and broken loops. The RELAP5/MOD3.1 comparisons with the experimental data are

reasonably satisfactory, although the code prediction typically underestimates the

experimental values.

5.6 Loop Mass Flow Behavior

Figure 5.32 and 5.33 show the flow rate variation on the primary side of the intact

and broken loops. Here, the RELAP5/MOD3.M flow rate predictions compare very well

with the test data. The large difference in the steady-state values of the flow rate for the

broken and intact loops are caused by insufficient mixing of primary side coolant by

natural circulation after the reactor coolant pumps are shut down.

The downcomer flow rate of the broken loop is shown in Figures 5.34. The

computer code flow rate predictions compare favorably with those of the experimental

data, including the time of attainment of quasi-steady state conditions (-200 seconds).

During the early stage of the transient, the computer code-test data compari;ons are

good. However, following the rapidly changing depressurization prediction of the code

around 800 seconds and the start of HPSI injection (at 1156 seconds), the code

overpredicts the experimental data for the downcomer mass flow rate very sub-.tantially

(by an average of about 200%) for nearly 900 seconds. Subsequently, the code-data

comparisons are reasonably satisfactory, but the code underpredicts the flow rate by

about 1 kg/sec. A similar situation is observed for the code prediction of separator flow

rate in the intact loop, Figures 5.35, especially for the inlet flow and liquid discharge

17
flow. The code prediction of broken loop separator flow is shown in Figure 5.36,

indicating that a quasi-steady state flow rate of about 1 kg/sec is attained in about 500

seconds after the initiation of the transient.

18
6. Discussion of Statistical Results

For an objective evaluation and judgement of the agreement between th! values

calculated by RELAP5/MOD3.1I and the experimental data values, it is necessary to

define quantitative parameters that can be used as a measure of the accuracy of the code

predictions. The most important parameter used to evaluate the relative accuracy of the

code prediction in the present work is the fractional error, which is defined as the ratio

of the error of the calculated variable to the experimental value of the variables.

(Xi),d, - (X,)x p

(Xi)exp

In equation (1), Eiis the fractional error of the predicted value by the code, (X,) co, is

the value of the dimensional variable calculated by the code at time t = t, and (X,),,, is

the value of the corresponding variable obtained from the experimental data, -which is

assumed to be the true value.

In order to compare the accuracy of the predicted values, the mean error ( e ), the rootmean-square (RMS) error ( E s) and the standard deviation of the error, (aJ a) were also

computed using the following equations:

7 (2) N

-= [( .= 2)112

i=1 NI

Rouhl t 1/2 )u th1t (

Roughly speaking, the predicted values that lead to low estimates of E , E Ms.

19
and a . may be considered to be a good fit to the experimental data.

Tables 6.1 -6.3 give values for E, E Rms, and a , , respectively, for the 'Base

Calculation' (Cd-0.85 and K=10.0) as described in Figures 5.1 -5.36. The section on

'Discussion of Results' compares the RELAP5/MOD3.1 computer code output against

the relevant experimental data.

The entire transient was sub-divided into three sub-intervals, and corresponding

data values for various variables were accordingly categorized into three sub-sets. A

total of 125 time points were used in each of the two time intervals (0 sec, 1000 sec) and

(1000 sec, 2000 sec), and a total of 188 time points were used in the third time interval

(2000 sec, 3500 sec). Statistical characterization of the differences between the code

predictions and the corresponding experimental data were performed over each of the

three time sub-intervals as well as over the entire transient (consisting of a total of 438

points). All 'statistical' evaluations considered here involve only one time-dependent

sample function (for each variable), and the statistical error characterization is based on

time averages.

These statistical evaluations over the three sub-intervals as well as over the entire

transient are summarized in Table 6.1 - 6.3. These three tables for the 'Base Calculation'

show, respectively, the mean error, the root-mean-square error, and the standard

deviation of the errors between the code and data values for 28 different variables which

characterize the MSLB transient.

The results in Tables 6.1 - 6.3 quantify the 'Discussion of Result' approximately.

As can be seen from Tables 6.1 -6.3, the statistical error values for most cases are

generally reasonably low. However, there are several cases wherein the values obtained

for the statistical parameters are rather misleading. As indicated earlier, the main reason

20
for this lies in the division by small (close to zero) values of (X,),p for obtaining E, It

can be concluded that these statistical error values reflect or quantify the characteristics

of the corresponding code-data comparisons for the majority of the cases reasonably

well.

21
7. Sensitivity Studies

The sensitivity studies described here are related to any investigation into the

effect of certain parameters on code predictions. Optimal results of these parameters are

selected so that the differences between the code calculations and the experimental data

are minimized. The statistical error values are used as a guide in assessing the optimal

values of the sensitivity parameters. For the investigation reported here, the break flow

discharge coefficient (Cd), the separator drain line loss coefficient (K), and the number

of steam generator nodalization (n) are selected as the sensitivity parameters which have

a significant impact on the computer code predictions.

The sensitivity studies for these parameters are shown plotted in Figures 7.1 - 7.30.

The values of E RMs (i.e., the RMS or root-mean-square value) for the three classes of

sensitivity studies are presented in Tables 7.1 -7.3. This statistical summary of the

sensitivity studies was discussed in the preceding section. For each of the three classes

of sensitivity studies, ten cases are considered for the purpose of statistical error

evaluation. The error evaluations for the sensitivity studies were performed over the

time interval 0.0 to 1000.0 seconds for all variables.

For each of the ten cases considered, the statistical error values are of the same

order of magnitude for the three classes of sensitivity studies. As was the situation for

the base calculation, the majority of the cases in Table 7.1 - 7.3 reflect acceptable error

values. Table 7.1 for the discharge coefficient sensitivity study is based on the Cd

values of 0.75, 0.80, 0.85, and 0.90. Table 7.2 for the separator drain line loss

coefficients sensitivity study uses K values of 0.0, 10.0, 50.0 and 100.0. Table 7.3 for

the nodalization sensitivity study is based on 3,5,7 and 9 steam generator nodes.

In summary, the sensitivity studies do not indicate a pronounced dependence on

the break flow discharge coefficient (Cd), the separator drain line loss coefficient (K),

or the number of steam generator nodes (n). Nevertheless, Cd=0.85, K=10.0, and n=5

22
are considered the optimal values for the three parameters investigated here.

7.1 Break Junction Discharge Coefficient (Cd)

Break junction discharge coefficient (Cd) values of 0.75, 0.80, 0.85, and 0.90 were

chosen for this sensitivity study. A lowering of the Cd value led to a decrease in the

break junction flow, as is to be expected from the basic definition of this vaxiables.

Since different mass and energy blowdown rates result in different system

depressurization rates, the steam generator shell side temperatures and associated heat

removal rates are expected to be different for the four different discharge coefficient

cases.

For the important time variations of break flow rate and pressurizer pressure, the

value Cd=0.85 leads to the lowest RMS error value, as can be seen from Table 7.1. For

most of the remaining time variations of important physical parameters, such as the

break void fraction, there is very little to choose between the four values of Cd.

7.2 Separator drain line loss coefficient (K)

The steam separator provides practically 100% dry steam during normal operation.

During an accident scenario, such as the main steam line break (MSLB) transi.ent, the

flow rates through the separator could be several times larger than the design basis

value. The separator performance under conditions that exist in a MSLB transient is

very important because it can change water inventory in the secondary system. This is

the principal factor that affects the heat transfer from the primary to the secondary side.

A separator component was used to simulate the steam generator separator and an

annulus component was used to simulate the steam generator downcomer, such that an

internal recirculation flow is maintained.

23
The separator drain lines were modeled as a junction flowing into the downcomer

from the separator to avoid frothing conditions in the downcomer during steady state

initialization. However, this introduced an undesirable side effect during steam

generator inventory depletion transients. When steam was carried under from the

separator via the drain lines, the liquid was forced up the downcomer to the steam dome

via the separator by-pass.

This non-physical phenomenon could be suppressed by using a large loss

coefficient in the separator drain line. This is the approach historically undertaken with

RELAP5/MOD2 [8, 9]. In order to study the effect of internal recirculation flow on the

void fraction and water inventory in the secondary side, steam separator drain line loss

coefficients of 0.0, 10.0, 50.0 and 100.0 were chosen in conjunction with a break

junction discharge coefficient of 0.85.

For the two physical variables of break flow rate and pressurizer pressure, the loss

coefficient values K=0.0 and K=10.0 lead to nearly the same statistical error, as can be

seen from Table 7.2. For most of the remaining variations, it is again a toss-up between

these two K values with regard to optimality for minimum error. However, when

considering the time variations of break void fraction, secondary pressure and

temperature of intact loop, collapsed liquid level of pressurizer, and downcomer flow

rates of intact and broken loops, the value K=10.0 leads to a smaller error level than

K=0.0 with respect to the data-code comparisons.

7.3 Steam Generator Nodalization

It is seen from Table 7.3 that, with regard to the number of steam generator nodes

(n), the value n=5 leads to the lowest overall RMS error for variations of break flow rate,

pressurizer pressure, break void fraction, secondary pressure and temperature of intact

loop, and the downcomer flow rates of intact and broken loops. The data-code

24
comparisons for the four different nodalizations are almost the same in relation to the

variations of secondary pressure and temperature of broken loop.

Figure 7.31 shows the steam generator nodalization diagram for the nodflization

sensitivity study. The four different cases of nodalization show 3, 5, 7 and 9 nodes in

the steam generator, while the nodalization for other components remains uncha•'ged.

Figure 7.32 shows a plot of CPU time vs transient (or real) time. For a given

nodalization parameter n, this curve represents a linear relationship between the

transient (or real) time and CPU time. However, for a fixed value of the transient time,

the CPU time is a monotonically-increasing or convex function of the number of nodes

n. (In other words, the CPU time increase with n at a faster rate than that indicated by a

linear relationship).

25
8. Run Statistics

The SUN SPARC/station 20 workstation (at KOPEC) with OS Version 5.4 was

used as the main frame computer for the calculations described here.

Figure 8.1 presents a plot of the required CPU time vs transient time for the base

calculation. The user-specified maximum time step was 0.003906 second (in real time)

for the entire transient calculation. The run statistics from the major edit are

summarized in Table 8.1. The grind time can be calculated as follows:

CPU time(sec),

Number of time steps,

Number of volumes,

Transient real time,

CPU= 232736.40 - 28.15 = 232708.25

DT= 895880 - 0 = 895880

C= 189

RT= 3500 (sec)

Grind time = CPU X 1000 / (C X DT) = 1.3744 CPU msec/vol/step

26
9. Conclusion and Recommendation

The overall comparisons between the RELAP5/MOD3.1 computer code

predictions and the LSTF Test Run SB-SL-01 experimental data for a MSLB transient

are reasonably satisfactory. This code provides approximate, but useful, predictions

with regard to the variations of mass flow rate, void fraction, pressure, collapsed liquid

level, temperature, and system flow rate for such a transient. Some of the results of

comparison between the code output and the experimental data are quite good. However,

certain shortcomings of this code are also apparent, when the code-data compari sons are

examined more closely.

Sensitivity investigations with respect to the effects of the break discharge

coefficient (Cd), the separator drain line loss coefficient (K), and the number of nodes

in the steam generator (n) on code-data comparisons did not reveal a pronounced and

straightforward dependence. The parameter values Cd=0.85, K=10.0, and n:--5 were

nevertheless selected as optimal values, based upon the lowest overall statistical error

for the code-data comparisons.

The following suggestions are recommended for improving the performance of

RELAP5/MOD3.1 in relation to the experimental results obtained for simulations of the

MSLB phenomena:

(1) The RELAP5/MOD3.1 computer code (and previous versions of this code)

cannot properly account for short wavelength or high frequency physical phe:nomena.

This can be noted in the time variations of volume averaged void fraction of uPperhead

nodes (Figure 5.6), and in the substantial underprediction of the refill rate (for the

collapsed liquid level) of the pressurizer (Figure 5.27). A related case involves the

underprediction of variable changes occurring over a small time interval, as can be seen

in Figure 5.10 ('cold leg temperature of intact loop'), and in Figures 5.28, 530. The

27
inability of RELAP5 to predict high frequency phenomena can be also seen in Figure

5.34 ('downcomer flow rate of broken loop'), where the code output displays instability

in the form of highly oscillatory output, due to insufficient 'damping' of the

mathematical or numerical models.

Although a more comprehensive description of this modeling deficiency in

RELAP5 is beyond the scope of this investigation, it is suggested here that the interphase drag sub-model in the two-phase models of RELAP5 should be investigated

further for a practical implementation of problem dependent, optimal numerical

'damping'. Suitable modeling of the inter-phase drag in this code is expected to mitigate

the deficiencies noted above.

(2) The improvement of the void fraction and conservation of mass modeling in

RELAP5 would lead to a significant improvement in the predictive capabilities of this

code.

Although the prediction of break void fraction by the code (Figure 5.2) is

reasonable, the calculated values of void fraction for the three upperhead nodes (Figure

5.6) involve significant discrepancy for the entire transient. These shortcomings affect,

in turn, predictions of pressure variation (Figure 5.5 and 5.9), collapsed liquid level

variation (Figures 5.27-5.31), and mass flow rates (Figures 5.32 -5.36).

28
References

[1] The ROSA-IV Group, 1990 b, "ROSA-IV Large Scale Test Facility (LSTF) System

Descriptions, JAERI-M 90-176, Japan Atomic Energy Research Institute.

[2] The ROSA-IV Group, 1989, "Supplemental Description of ROSA-IV LSTF with

No. I Simulated Fuel Rod Assembly", JAERI-M 89-113, Japan Atomic Energy

Research Institute.

[3] The ROSA-IV Group, 1990 a, "Quick Look Report for LSTF Test SB-SL-01",

Japan Atomic Energy Research Institute.

[4] The RELAP5 Development Team, 1995, "RELAP5/MOD3 Code Manual",

NUREG/CR-5535.

[5] Lin, C. L., Chao. J., Chiu, C., 1984, "A Boron Model for the Main Steam Line

Break Transient", Proceedings of International Nuclear Power Plant Thermal

Hydraulic And Operations Topical Meeting (China), pp 22-24.

[6] Tang, Jan-Ru, Wang, Song-Feng, Yuann, Ruey-Ying, 1986, "An Analysis of Steam

Line Break Accident for Maanshan Units 1 & 2 Using RETRAN-02/MOD.'-. Code",

Proceedings of Second International Nuclear Power Plant Thermal Hydraulics And

Operations, pp 15-17.

[7] Liang, Kuo-Shing, Wang, Song-Feng, 1986, "Systematic Evaluation of the Main

Steam Line Break Accident for a Typical Westinghouse Three Loop Plant Using

RELAP5/MOD2 Code", Proceedings of Second International Nuclear Power Plant

Thermal Hydraulics And Operations, pp 15-17.

[8] Rogers, J. M., 1989 a, "An Analysis of MB-2 100 % Steam Line Break Test T2013 Using RELAP5/MOD2", AEEW-R2476, UKAEA Atomic Energy

Establishment, WINFRITH Reactor Systems Analysis Division.

[9] Rogers, J. M., 1989 b, "An Analysis of Semiscale Mod-2C S-FS-1 Steam Line

Break Test Using RELAP5/MOD2", UKAEA Atomic Energy Establishment,

WINFRITH Reactor Systems Analysis Division.

29
Table 3.1 Sequence of Events for Experiment SB-SL-01

Measured Calculated

Event Data Data

(seconds) (seconds)

Transient initiations 0 0

Reactor scram 0 0

MS1V closure 2 2

Steam generator feedwater valve 2 2

closure

Turbine throttle valve closure 0 2

Auxiliary feedwater injection 28 58

High pressure safety injectionsteam 1156 1156

PORV open 3312 2905

-30-
Table 3.2 Initial and Boundary Conditions for Experiment SB-SL-01.

Parameters Units LSTF RELAP5,MOD3.1

(Measured) (Calculated)

Core power MW 10.000±0.044 10.(00

Pressurizer pressure MPa 15.520±0.080 15.5;90

Hot leg fluid temperature K 598.100±4.396 599.400

Cold leg fluid temperature K 562A.00±4.134 564.100

Primary coolant flow rate kg/sec 24.3000±02668 24.6)00

Pressurizer liquid level m 2.7000±0.0086 2.64.00

Primary coolant pump speed rpm 800±4 768

Primary coolant flow control valves - fully open fully open

SG secondary pressure MPa 7.3000±0.0393 7.3(00

SG secondary liquid level m 10.3000±0.0329 11.8165

SG feedwater temperature K 495.2000±3.8477 495.4.000

SG feedwater and main steam kg/sec 2.7400±0.0442 2.7500

flow rates

- 31 -
Table 5.1 List of Assessment Parameters (LSTF SB-SL-01)

Description Calculatiou Measurement Uncertainty Figure

Break Flow Rate mflowi 91500 FE560A-BU 0.7686 ka/sec 5.1

Break Void Fraction void 32801 SP-VOLUME Not Found 5.2

Break Density rho 32801 N/A N/A 53

Pressurizer Pressure v 61004 PE300A-PR 0.1078 MPa 5.4

Vessel Upperhead Void Fraction voidg 14401 WA N/A 5.5

voidg 14801

vodg 15201

Volume Averaged Void Fraction of Three rho 14401 SB-SL-01 Not Found 5.6

Upperhead Nodes rho 14801

rho 15201

Secondary Pressure of Broken Loow v 31601 PE450-SGB 0.0539 MPa 5.7

Secondary Pressure of Intact Low v 51601 PE430-SGA 0.0539 MPa 5.8

Hotleg Temperature of Intact Loon tempf 40001 TE02OD-HLA 33070 K 5S9

Coldlea Temnerature of Intact Looo temof 45201 TEOSOD-C•A 3370 K 5.10

SG Steam Dome Te=nerature of Intact Loon temof 51601 TE-245C-SGA 3.1080 K 5.11

SG Downcamer Te=merat"re of Intact Loon teef 6"0008 TE432-SGA 31080 K 5.12

SG Downcomer Temperature of Intact Loop tempf 50008 TE431-SGA 33080 K 5.13

(4 loops) TE432-SGA

TFA33-SGA

TFA34-SGA

Hotlea Termorature of Broken Loon tenxf 20001 TE170D-HI.B 3.3070 K 5.14

Coldleu Temrxratu'e of Broken Loow taof 25201 T -22 CLB .- 33M70 K 5.15

SG Steam Dome Tezcn'ature of Broken Loon tanof 31601 TE-245C-SGB 3.1080 K 5.16

SG Downcoimer Tenoerature of Broken Loon temvf 3M005 TE472-SGB 3.1080 K 5.17

SG Downcrmer Temperature of Broken Loop temp 3D005 TE471-SGB 3.1080 K 5.18

(4 loops) TE472-SGB TE473-SGB

TE474-SGB

SG Boiler Bottom Temperature of Broken Loon temuf 30401 TE-099C-SGB 3.1080 K 5.19

SG Boiler Middle Temperature of Broken Loop temof 30403 TE-150C-SGB 3.1080 K 5.20

SG Boiler Too Temterature of Broken Loot ternuf 30405 TE-192F-SGB 3.1080 K 5.21

SG Boiler Bottom Temoerature of Intac LooD ternuf 50401 TE-099C-SGA 3.1080 K 5.22

SG Boiler Middle Temperature of Broken Loop temnf 50403 TE-150C-SGA 3.1080 K 5.23

SG Boiler ToD Temoerature of Broken Loot tempf 50405 TE-178C-SGA 3.1080 K 5.24

Average Temperature of RCS Broken Loop tempf 20001 TE240--HLB 3.3070 K 525

temof 25201

Collapsed Liouid Level of Pressurizer cntrvar 610 LE280-PR 0.0544 m 5.26

Narrow Range of Liquid Level of cntrvar 508 LE440-SGA 0.0192 m 527

S/G Intact Loo

Narrow Range of Liquid Level of cntrvar 308 LE46--SGB 0.0192 m 528

S/G Broken Loow

Liquid Level of S/G Intact Loop cntrvar 512 LE441-SGA 0.0352 m 529

LE442-SGA

LE430-SGA

Liquid Level of S/G Broken Loop cntrvar 312 LE461-SGB 0.0352 m 530

LE462-SGB

LE450-SGB

Pump Suction Flow Rate of Broken Loop mflowj 24001 FE160B-LSB 0.0174 kg/s 5.31

FE06OA-LSB 1.0710 kg/s

Pump Suction Flow Rate of Intact Loop mflowj 44001 FE020B-LSA 0.0174 kg/s 532

FE02DA-LSA 1.0710 kiz/s

Downcomer Flow Rate of Broken Loop mflowj 30100 FE471-SGB 01169 kg/s 533

FE472-SGB

FE473-SGB

FE474-SGB

Downcomer Flow Rate of Intact Loop mflowj 50100 FE431-SGA 0.1129 kg/s 535

FE432-SGA

FE433-SGA

FE434-SGA

Separator Flow Rate of Broken Loop mflowj 3001 N/A N/A 537

mflowj 30802

mflowi 30803

Separator Flow Rate of Intact Loop mflowj 50801 N/A N/A 5.38

mflowj 50802

mflowi 50803

-32-
Table 6.1 The Mean Errors (E) for the Base Calculation (Cd=85, K=10.0, n=5)

Figure Time Range (sec) _____ No. No. Parameters ~~2~~.3O i1rne

N oO0 "t< ID00 1000 • t <2000 2000< t. :3 50 0( fu e)<3:

1 5.1 Break Flow Rate -0.0107 0.0080 0.1175 0.0474

2 5.2 Break Void Fraction -0.0024 0.0106 0.09113 0.0723

3 5.4 Pressurizer Pressure -0.0028 0.0763 0.0709 0.0499

4 5.6 Volume Averaged Void Fraction of Three - 52755 15681 19991

Upperhead Nodes -5 5_

5 5.7 Secondary Pressure of Broken Loop -0.1553 -0.1733 -02163 -0.1994

6 5.8 Secondary Pressure of Intact Loop -0.0009 -0.1165 -0.1196 -0.0848

7 5.9 Hotleg Temperature of Intact Loop -0.0000 -0.0092 -0.0070 -0.0056

8 5.10 Coldleg Temperature of Intact Loop 0.0021 0.0163 -0.0422 -0.0128

9 5.11 SG Steam Dome Temperature of Intact Loop -0.0021 -0.0239 -0.0,3 -0.0196

10 5.12 SG Downcomer Tenmeamture of Intact Loop -0.0103 0.0038 0.0029 -0.0006

11 5.14 Hotleg Temperature of Broken Loop -0.0001 -0.0124 -0.0178 -0.0112

12 5.15 Coldleg Temperature of Broken Loop -0.0081 -0.0087 -0.0134 -0.0105

13 5.16 SG Steam Dome Temperature of Broken Loop -0.0395 -0.0861 -0.1091 -0.0826

14 5.17 SG Downcomer Temperature of Broken Loop -0.0193 -0.0135 -0.0145 -0.0156

15 5.19 SG Boiler Bottom Temperature of Broken Loop -0.0173 -0.0140 -0.0180 -0.0166

16 5.20 SG Boiler Middle Temperature of Broken Loop -0.0167 -0.0366 -0.0362 -0.0308

17 5.21 SG Boiler Top Temperature of Broken Loop -0.0233 -0.0392 -0.0401 -0.0348

18 5.22 SG Boiler Bottom Temperature of Intact Loop -0.0018 -0.0163 -0.0.174 -0.0127

19 5.23 SG Boiler Middle Temperature of Intact Loop -0.0012 -0.0208 -0.0236 -0.0164

20 524 SG Boiler Top Temperature of Intact Loop -0.0033 -0.0176 -0.0.162 -0.0129

21 525 Average Temperature of RCS Broken Loon -0.0034 -0.0117 -0.0 L70 -0.0116

22 5.26 Collapsed Liauid Level of Pressurizer -0.5160 -3.0766 -02308 -1.1393

23 5.27 Narrow Range of SG Level of Intact Loop 02396 -03400 -02382 -0.1521

24 5.28 Narrow Range of SG Level of Broken Loon -0.8151 -0.9153 -0.9188 -0.8881

25 5.31 Pump Suction Flow Rate of Broken Loop 0.5014 0.0336 0.0501 0.1806

26 5.32 Pump Suction Flow Rate of Intact Loop 0.5991 0.3029 11.6543 5.0488

27 5.33 Downcomer Flow Rate of Broken Loop -0.3198 -0.4050 -02366 -0.3214

28 5.35 Downcomer Flow Rate of Intact Loop 0.0451 2.5183 -0.8304 0.3779

The number of data points used for evaluation 125 125 183 438

-33-
Table 6.2 The Root-Mean-Square Errors (e cms)

for the Base Calculation (Cd=0.85, K=10.0, n=5)

Figure Time Rarwe (sec)___

No. Figure Parameters ogt<3o No. Ort<1000 1000Ot<2O)0 2O00•t<MCM5

1 5.1 Break Flow Rate 0.0270 0.0159 0.1437 0.0939

2 5.2 Break Void Fraction 0.0135 0.0110 0.0117 0.0120

3 5.4 Pressurizer Pressure 0.0173 0.0777 0.0778 0.0655

4 5.6 Volume Averaged Void Fraction of hree 0.9335 6.8087 2.8789 4.1526

Upperhead Nodes

5 5.7 Secondary Pressure of Broken Loop 0.1661 0.1751 02468 0-2067

6 5.8 Secondary Pressure of Intact Loop 0.0257 0.1173 0.1198 0.1013

7 5.9 Hotlez Temperature of Intact Loop 0.0027 0.0095 0.0074 0.0072

8 5.10 Coldleg Temperature of Intact Loop 0.0032 0.0226 0.0518 0.0360

-9 5.11 SG Steam Dome Temperature of Intact Loop 0.0046 0.0241 0.0284 0.0227

10 5.12 SG Downcomer Temperature of Intact Loop 0.0150 0.0052 0.0031 0.0087

11 5.14 Hotlez Temperature of Broken Loot 0.0025 0.0130 0.0179 0.0137

12 5.15 Coldleg Temperature of Broken Loop 0.0090 0.0092 0.0135 0.0112

13 5.16 SG Steam Dome Temperature of Broken Loop 0.0446 0.0868 0.1091 0.0884

14 5.17 SG Downcomer Temperature of Broken Low 0.0202 0.0135 0.0148 0.0162

15 5.19 SG Boiler Bottom Temperature of Broken Loop 0.0182 0.0141 0.0181 0.0171

16 5.20 SG Boiler Middle Temperature of Broken Loop 0.0177 0.0373 0.0379 0.0332

17 521 SG Boiler Ton Temuerature of Broken Loop 0.0251 0.0394 0.0404 0.0364

18 5.22 SG Boiler Bottom Temperature of Intact Loop 0.0055 0.0168 0.0175 0.0149

19 523 SG Boiler Middle Temperature of Intact Loop 0.0034 0.0213 0.0237 0.0193

20 524 SG Boiler Top Temperature of Intact Loop 0.0047 0.0178 0.0163 0.0145

21 5.25 Average Temnerature of RCS Broken Loon 0.0038 0.0121 0.0170 0.0131

22 5.26 Collapsed Liquid Level of Pressurizer 0.6301 13.9244 02712 7.4569

23 527 Narrow Ranre of SG Level of Intact Loon 2.1236 0.3410 02887 1.1657

24 528 Narrow Range of SG Level of Broken Loop 0.8436 0.9153 0.9188 0.8969

25 531 Pump Suction Flow Rate of Broken Loop 2.5046 0.0362 0.0530 1.3714

26 5.32 Pump Suction Flow Rate of Intact Loop 2.4229 20.2147 107287 69.6087

27 5.33 Downcomer Flow Rate of Broken Loop 0.4001 0.4359 0.8085 0.6163

28 5.35 Downcomer Flow Rate of Intact Loop 2.4101 3.3852 0.8375 2.2890

The number of data points used for evaluation 125 125 188 438

- 34 -
Table 6.3 The Standard Deviations of the Errors (a,)

for the Base Calculation (Cd=0.85, K=10.0, n=5)

No. Figure Parameters Time Range (!sec) (ful r3ge

No. O:5t<10D0 l000:5t<2DOCO 20•00:t'.5D Urne

1 5.1 Break Flow Rate 0.0248 0.0138 0.0327 0.0811

2 52 Break Void Fraction 0.0133 0.0030 0.0029 0.0097

3 5.4 Pressurizer Pressure 0.0171 0.0144 0.0321 0.0425

4 5.6 Volume Averaged Void Fraction of Three 0.6947 4.3043 2.4144 3.6397

Upperhead Nodes

5 5.7 Secondary Pressure of Broken Loop 0.0587 0.0244 0.0157 0.0543

6 5.8 Secondary Pressure of Intact Loop 0.0256 0.0142 0.0365 0.0556

7 5.9 Hotleg Temperature of Intact Loop 0.0027 0.0023 0.0024 0.0044

8 5.10 Coldleg Temperature of Intact Loop 0.0024 0.0157 0.0300 0.0337

9 5.11 SG Steam Dome Tempr-ature of Intact Loop 0.0041 0.0029 0.0013 0.0116

10 5.12 SG Downcomer Temperature of Intact Loop 0.0109 0.0036 0.0011 0.0087

11 5.14 Hotleg Temperature of Broken Loop 0.0025 0.0039 0.0012 0.0078

12 5.15 Coldleg Temperature of Broken Loop 0.0039 0.0028 0.0017 0.0037

13 5.16 SG Steam Dome Temperature of Broken Loop 0.0208 0.0108 0.0027 0.0315

14 5.17 SG Downcomer Temperature of Broken Loop 0.0057 0.0015 0.0029 0.0044

15 5.19 SG Boiler Bottom Temperature of Broken Loop 0.0057 0.0016 0.0019 0.0038

16 5.20 SG Boiler Middle Temperature of Broken Loop 0.0058 0.0073 0.0113 0.0126

17 5.21 SG Boiler Top Temperature of Broken Loop 0.0114 0.0039 0.0044 0.0106

18 5.22 SG Boiler Bottom Temperature of Intact Loop 0.0051 0.0042 0.0014 0.0078

19 5.23 SG Boiler Middle Temperature of Intact Loop 0.0032 0.0042 0.0014 0.0101

20 5.24 SG Boiler Tor Temperature of Intact Loot 0.0033 0.0024 0.0016 0.0066

21 5.25 Average Temperature of RCS Broken Loot 0.0018 0.0030 0.0011 0.0060

22 526 Collapsed Liquid Level of Pressurizer 0.3616 13-9803 0.0744 7.3694

23 5.27 Narrow Range of SG Level of Intact Loop 2.1100 0.0263 0.0161 1.1558

24 5.28 Narrow Range of SG Level of Broken Loop 02171 0.0023 0C022 0.1250

25 5.31 Pump Suction Flow Rate of Broken Loop 2.4539 0.0137 0.C172 13594

26 5.32 Pump Suction Flow Rate of Intact Loop 2.3477 202124 106.653 69.4254

27 5.33 Downcomer Flow Rate of Broken Loop 02404 0.1611 0.7633 0.5259

28 5.35 Downcomer Flow Rate of Intact Loop 2.4097 2.222 0.1086 22576

The number of data points used for evaluation 125 125 18 438

-35-
Table 7.1 The Root-Mean-Square Errors (ews)

on the Discharge Coefficient

for the Sensitivity Study

(Cd)

(K=10.0, n=5)

No. Figure No. Parameters Discharge Coefficient

Cd=0.75 Cd-0.80 Cd=0.85 Cd=0.90

1 Figure 7.1 Break Flow Rate 0.0435 0.0280 0.0270 0.0394

2 Figure 7.2 Pressurizer Pressure 0.0232 0.0196 0.0173 0.0189

3 Figure 7.3 Secondary Pressure of Broken Loop 0.0751 0.1226 0.1661 02048

4 Figure 7.4 Secondary Tempmrature of Broken Loop 0.0349 0.0400 0.0446 0.0494

5 Figure 7.5 Break Void Fraction 0.0133 0.0134 0.0135 0.0136

6 Figure 7.6 Secondary Pressure of Intact Loop 0.0227 0.0258 0.0246 0.0274

7 Figure 7.7 Secondary Temprature of Intact Loop 0.0043 0.0045 0.0046 0.0051

8 Figure 7.8 CollaDsed Liquid Level of Pressurizer 0.5327 0.5778 0.6301 0.6721

9 Figure 7.9 Downcomer Flow Rate of Intact Loop 2.4534 Z4114 2.4101 2.4205

10 Figure 7.10 Downcomer Flow Rate of Broken Loon 0.4333 0.4328 0.4001 0.4405

The number of data points used for evaluation 125 125 125 125

-36-
Table 7.2 The Root-Mean-Square Errors (cfms) for the Sensitivity Study

on the S/G Drain Line Loss Coefficient (K)

(Cd=0.85, n=5)

No. Figure No. Parameters Loss Coefficie nt

K=0.00 K=10.0 K=W0.0 K=100.0

1 Figure 7.11 Break Flow Rate 0.0259 0.0270 0.(828 0.0338

2 Figure 7.12 Pressurizer Pressure 0.0169 0.0173 0.0174 0.0175

3 Figure 7.13 Secondary Pressure of Broken Loop 0.1657 0.1661 0.1661 0.1672

4 Figure 7.14 Secondary Termeraturc of Broken Loow 0.0443 0.0443 0.'445 0.0445

5 Figure 7.15 Break Void Fraction 0.0135 0.0135 0.0151 0.0136

6 Figure 7.16 Secondary Pressure of Intact Loop 0.0264 0.0248 0.(232 0.0224

7 Figure 7.17 Secondary Tem -erature of Intact Loop 0.0046 0.0046 0.C043 0.0043

8 Figure 7.18 CoUansed Liquid Level of Pressurizer 0.6252 0.6262 0.E274 0.6285

9 Figure 7.19 Downcomer Flow Rate of Intact Loop 2.6412 Z4101 2A705 2.4000

10 Figure 7.20 Downcomer Flow Rate of Broken Loop 0.4377 0.4006 0A278 0.4400

The number of data points used for evaluation 125 125 125 125

- 37 -
Table 7.3 The Root-Mean-Square Errors (Ews) for the Sensitivity Study

on the Number of Steam Generator Nodes (n)

(Cd=0.85, K=10.0)

No. Figure No. Parameters Number of S/G Nodes

n=3 n=5 n=7 n=9

1 Figure 7.21 Break Flow Rate 0.0511 0.0270 0.0371 0.0389

2 Figure 722 Pressurizer Pressure 0.0154 0.0173 0.0179 0.0175

3 Figure 7.23 Secondary Pressure of Broken Loop 0.1404 0.1661 0.1684 0.1676

4 Figure 7.24 Secondary Temnrature of Broken Loop 0.0410 0.0443 0.0450 0.0449

5 Figume 7.25 Break Void Fraction 0.0167 0.0135 0.0153 0.0157

6 Figure 726 Secondary Pressure of Intact Loop 0.0312 0.0248 0.0253 0.0245

7 Figure 7.27 Secondary Temperature of Intact Loop 0.0054 0.0046 0.0046 0.0043

8 Figure 7.28 Collapsed Licnid Level of Pressurizer 0.6031 0.6262 0.6309 0.6327

9 Figure 7.29 Downcomer Flow Rate of Intact Loop 3.3190 2.4101 2A527 4Z3797

10 Fi•gue 7.30 Downcomer Flow Rate of Broken Loon 0.8223 0.4006 0.4361 0.4376

The number of data points used for evaluation 125 125 125 125

-38-
Table 8.1 Run Statistics Data in Base Calculation

Transient time CPU time Attempted Repeated Last Courant

(sec) (sec) ADV ADV DT DT'

0 28.75 0 0 3.125E-2 0

500 33481.12 127880 0 3.906E-3 4.74466E-3

1000 66705.25 255880 0 3.906E-3 4.46076E-3

1500 99842.10 383880 0 3-906E-3 4A9261E-3

2000 132695.70 511880 0 3.906E-3 4.51401:E-3

2500 166225.70 639880 0 3.906E-3 4.53266S-3

3000 199659.60 767880 0 3.906E-3 4.54726-3-3

3500 232736.40 895880 0 3.906E-3 4.56085E.-3

- 39 -
I AInt 1 e 

IQmi6) P"

Iii'too (e' Vallelmo

11111~m Villerlin

Iii I _ _ _ 

it ~ l l ~ p3i

0=1 

r.ra.rtn~o A~

~) ~n~inyIntr line

21 SIt>m geFio - -:~ 2s ni-tMm foao(l

Figurc 2.1 Schematic View of ROSA-VI LSTI
Tim! (s)

Figure 3.1 Three Core Power Control Curves for LSTF

- 41 -
OCC( VALVE K.!AM~Ht HCA"R

AFW AJW

CONTlAINMENT VOLUME ?OR tWWOWEMNA4. MEAT LOW9

mIrw

towNCOVEA Pro

SYOAM CENCRATOP W - Volume

Junction

Heat Structure

Time Dependent Volume

PPR$ SSUIZER

ROIS)

I

I

UPPER HtAD

3w (2)

RCs PUMP

IIICS)

CmOSSOVER tip ACC UMUiATOR CROSSOVER rio

Broken Loop Intact Loop

Figure 4.1 RELAP5/MOD3.1 Nodalization of LSTF for MSLB Transient Assessment
-o- FE560A-BU

- RELAP5/m3.1 (mflowj 91500)

7

6 U,

S5

o• 4

0

LA) 3 , U,

cuu

2

0

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.1 Break Flow Rate
-o- SP-VOLUME

RELAP5/m3.1 (voidg 32801) (Cd=O.85, K=10.0) 1.2

1.0

0.8

Y V -

- - - - - - - - - - - - - - - - - - -- o

C

0

0.6 -t 1- -t 1- t 4

0.4

0.2

0.0

4 4. 4 4. 4

4 4. 4 4. 1 4

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.2 Break Void Fraction
- RELAP5/m3.1 (rho 32801)

45 ' ___________ - -

40

35

30

~25

Z. 20

15

10

5

0

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.3 Break Density
* -- _ RELAP5/m3.1 - BOTTOM (voidg 14401)

-o- RELAP5/m3.1 - MIDDLE (voidg 14801)

-A-m RELAP5/m3.1 - TOP (voidg 15201) (Cd=0.85. K=10.0) 1.2

1.0

0.8

.-

0

CU

U-

*0 0.6

* * * *.4~- : : i ~. t ~ I I I I

I

I

0.4

0.2

0.0

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.4 Vessel Upperhead Void Fraction
-o- PE300A-PR

--. RELAP5/m3.1 (p 61004)

18 ,

PORV

OPEN

16

cu 14 a.

U) u) 12 "• • •HPSI

10

8

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.5 Pressurizer Pressure
-o- RUN SB-SL-01

-• RELAP5/m3.1 (volume average of rho 14401, 14801 and 15201) (Cd=0.85, K=10.0)

0.5

0.4

I T I~ ¶ Y

1- 1 1-

0.3 ,KU

C

0

(U

LL

0.2

I

I

0.1

0.0

-0.1

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.6 Volume Averaged Void Fraction of Three Upperhead Nodes
-o- FE431+FE432+FE433+FE434 (SGA)

-- RELAP5/m3.1 (mflowj 50100)

25

20

~15

0~1

S 5 ,

= 10 - _ _ -_-""" _ _ _

0

-5

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.7 Downcomer Flow Rate of Intact Loop
-o- PE450-SGB

---- RELAP5/m3.1 (p 31601)

8

7

6

-. 5

IL

4

.. . , , . . U,,

............ .,, . ... ...

I

0 1s)

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.8 Secondary Pressure of Broken Loop
-o- PE430-SGA

- RELAP5/m3.1 (p 51601)

10 _ _ _.. ....-- - .. .... u

ao

10

aj)

ELI 4

2

0

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.9 Secondary Pressure of Intact Loop
-o- TE08OD-CLA

-- RELAP5/m3.1 (tempf 45201)

650 , ,

600 HPSI

550

500

CL.

E

I-450 P

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.10 Coldleg Temperature of Intact Loop
-o-- TE220D-CLB

-- m RELAP5/m3.1 (tempf 25201)

650

600

550

* 500

(D

400

350

0 500 1000 1500 2000 2500 3000 3500

Timm to%

Figure 5.11 Coldleg Temperature of Broken Loop
-o- TE020D-HLA

.... RELAP5/m3.1 (tempf 40001)

650 .. ....... ....-- - -

6006 0 0

550

L...

* 500

C L ... .. ...... . , • .....

-.- 450

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.12 Hotleg Temperature of Intact Loop
-o- TE170D-HLB

---- RELAP5/m3.1 (tempf 20001) (Cd=0.85, K=10.0) 650

600

550

CL)

E

a,

I

500

450

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.13 Hotleg Temperature of Broken Loop
-o- TE-245C-SGA

........ RELAP5/m3.1 (tempf 51601) (Cd=0.85, K=10.0) 650

600

550

E,

L..

500

450

C

**fl--...- ...

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.14 SG Steam Dome Temperature of Intact Loop
-o- TE432-SGA

RELAP5/m3.1 (tempf 50005) ICd=085. K=10.0n 650

600

550

C',

E

ci)

01

500

450

r r - ----------.-

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.15 SG Downcomer Temperature of Intact Loop
TE431-SGA ----- TE432-SGA

---- TE433-SGA -- ,- TE434-SGA

.RELAP5/m3.1 (tempf 50005) (Cd=0.85, K=10.0) 650

600

550

a)

E

a,

500

450

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.16 SG Downcomer Temperature of Intact Loop

(Downcomer is comprised of four loops)
-o- TE-245C-SGB

RELAP5/m3.1 (tempf 31601)

650 t,,-V o0, r,- IU.UJ

600

550

" 500

CL

E

(1)

S450

400

350

0 500 1000 1500 2000 2500 3000 3500

*r,, /M

lisle%, ,`

Figure 5.17 SG Dome Temperature of Broken Loop
-o- TE472-SGB

........... RELAP5Im3.1 (tempf 30005) (Cd=0.85, K=10.0) 650

600

550

!

a,

E

500

450

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.18 SG Downcomer Temperature of Broken Loop
TE471-SGB -'- TE472-SGB

,-nTE473-SGB -i- TE474-SGB

- RELAP5/m3.1 (tempf 30005)

650

600

550

E

a'

500

450

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.19 SG Downcomer Temperature of Broken Loop

(Downcomer is comprised of four loops)
o-TE-099C-SGB

RELAP5Im3.1

650

600

550

EI

4-'

500

450

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.20 SG Boiler Bottom Temperature of Broken Loop
-o-- TE-150C-SGB

-- RELAP5/m3.1 (tempf 30403)

650 , ,, , ..

600

550

E 50 ....

4 5 0 ...........

400 .

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.21 SG Boiler Middle Temperature of Broken Loop
--o-TE-192F-SGB

RELAP5/m3.1 (tempf 30405) (Cd=0.85, K=10.0) 650

600

550

(D

C',

E

1)

500

450

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s).

Figure 5.22 SG Boiler Top Temperature of Broken Loop
-o- TE-099C-SGA

........ RELAP5/m3.1 (tempf 50401) (Cd=O.85, K=10.0) 650

600

550

E

1)

500

450

n. ii. . .

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.23 SG Boiler Bottom Temperature of Intact Loop
-o-- TE-1 50C-SGA

-- RELAP5/m3.1 (tempf 50403) (Cd=0.85. K=1O.O1 650

600

550

CL

E)

500

450

.n~ ... . . ........ . ....

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.24 SG Boiler Middle Temperature of Intact Loop
-o- TE-178C-SGA

RELAP5/m3.1 (tempf 50405) (Cd=0.85, K=10.0)

650

600

550

a)

E

a)

I

I

500

450

-~ p t~.

400

350

0 500 1000 1500 2000 2500 3000 3500

Tima tO

Figure 5.25 SG Boiler Top Temperature of Intact Loop
-o- TE240-HLB

-- RELAP5/m3.1 (average of tempf 20001 and 25201)

650 ___________'...

600

500

CL I. .1 ..........

E 4...."-...... .- 

4 0........ • -- m•

400

350

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.26 Average Temperature of RCS Broken Loop
-o- LE280-PR

- RELAP5/m3.1 (cntrlvar 610)

5

4

3

-aj

Vr

in

I

2

1

0

-1

0 500 1000 1500 2000 2500 3000 3500

TirnA (s~)

Figure 5.27 Collapsed Liquid Level of Pressurizer
-o- LE440-SGA

--. RELAPS/m3.1 (cntrlvar 508) (Cd=0.85, K=10.0) 3

2

'I,

-D

1

-4

0

-1

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.28 Narrow Range of SG Level of Intact Loop
-o- LE460-SGB

- RELAP5/m3.1 (cntrlvar 308) (Cd=0l AA K=1fl fl'

3

2

_in

._-

1

........... 0 0 0 0 C

,--8

0

-1

0. 500 1000 1500 2000 2500 3000 3500

I IISU~~ ~O)

Figure 5.29 Narrow Range of SG Level of Broken Loop
- LE441-SGA a--- LE442-SGA

-- LE430-SGA - RELAP5/m3.1 (cntrlvar 512) (Cd=0O85, K=10.0) 16

14

12

10

8

6

0---

A rlI r

!

I

4

2

0

0 500 1000 1500

Time

2000

(s)

2500 3000 3500

Figure 5.30 Liquid Level of SG Intact Loop
-LE461-SGB

LE450-SGB

LE462-SGB

RELAP5/m3.1 (cntrlvar 312) (Cd=O.85, K=10.0) 16

14

12

10

(D,

I1 -J

B

6

4

2

0

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.31 Liquid Level of SG Broken Loop
-o- FE160B-LSB (LOW)

-,•- FE160A-LSB (HIGH)

-RELAP5/m3.1 (mflowj 24001) (Cd=0.85, K=10.0) 50

40

a)

cc)

0

CU

30

20"

10

0

-10

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.32 Pump Suction Flow Rate of Broken Loop
--o- FE020B-LSA (LOW)

-A- FE020A-LSA (HIGH)

-RELAP5/m3.1 (mflowj 44001) (Cd=0.85, K=10.0) 50

40

• 30

!

I

(D3

0

CU

20

10

0

-10

0 500 1000 1500 2000 2500 3000 3500

Time% (s

Figure 5.33 Pump Suction Flow Rate of Intact Loop
-o- FE471+FE472+FE473+FE474 (SGB)

=RELAP5/m3.1 (mflowj 30100)

25 _ 'U-U.Ou 1 (\ I U.UJ

20

• 15

~10

0

-L

',,

0

-5

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.34 Downcomer Flow Rate of Broken Loop
.--- RELAP5Im3.1 (Inlet Flow: mflowj 50803)

-,-- RELAP5/m3.1 (Steam Discharge: mflowj 50801)

-A- RELAP5/m3.1 (Liquid Discharge: mflowj 50802)

40

30

42

CM

20

a)

0

ir 10

U/)

0 mON

-10

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.35 Separator Flow Rate of Intact Loop
-o- RELAP5/m3.1 (Inlet Flow: mflowj 30803)

---- RELAP5/m3.1 (Steam Discharge: mflowj 30801)

-A,_ RELAP5/m3.1 (Liquid Discharge: mflowj 30802)

60 ,.,,_-v.,.,,.,, ,',- ,vJ.J

50

40

30

o 20

09

0

-10

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 5.36 Separator Flow Rate of Broken Loop
FE560A-BU

RELAP5/m3.1 (Cd = 0.75) -,- RELAP5/m3.1 (Cd = 0.80)

SRELAP5/m3.1 (Cd = 0.85) RELAP5/m3.1 (Cd = 0.90) 8

7

44

0)

LL

to

U,

cu,

0 50 100 150 200 250 300 350 400 450 500

Tima (cQ

Figure 7.1 Comparison of Break Flow Rate

(Cd Value Sensitivity)
-PE300A-PR

RELAP5/m3.1 (Cd = 0.75)

RELAP5/m3.1 (Cd = 0.85)

RELAP5/m3.1 (Cd = 0.80)

--- RELAP5/m3.1 (Cd = 0.90)

17

16

15

(L

02 14

I 13

'IL

12

11

10

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.2 Comparison of Pressurizer Pressure

(Cd Value Sensitivity)
-- PE450-SGB

0 RELAP5/m3.1 (Cd = 0.75) RELAP5/m3.1 (Cd = 0.80)

-- RELAP5/m3.1 (Cd = 0.85) .--- RELAP5/m3.1 (Cd = 0.90)

SI T I IT - - 8

7

6

5

I

0.0

I.,.

a) L..

C,)

Cl) a)

L. ci~

4

3

2

1

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.3 Comparison of Secondary Pressure of Broken Loop

(Cd Value Sensitivity)
TE-245C-SGB

RELAP5/m3.1 (Cd = 0.75)

-RELAP5/m3.1 (Cd = 0.85)

RELAP5/m3.1 (Cd = 0.80)

- RELAP5/m3.1 (Cd = 0.90) (K = 10.0) 580

560

540

520

a2)

E

FT

S........ ..

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

I 500

!

480

460

440

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.4 Comparison of Secondary Temperature of Broken Loop

(Cd Value Sensitivity)
-- • SP-VOLUME

RELAP5/m3.1 (Cd = 0.75)

RELAP5/m3.1 (Cd = 0.85)

RELAP5/m3.1 (Cd = 0.80)

- RELAP5/m3.1 (Cd = 0.90)

a

0

LL

1.05

1.04

1.03

1.02

1.01

1.00

0.99

0.98

0.97

0.96

0.95

1

0 100 200 300 400 500 600 700 800 900 1000

Time (.R)

Figure 7.5 Comparison of Break Void Fraction

(Cd Value Sensitivity)
PE430-SGA

RELAP5/m3.1 (Cd = 0.75) RELAP5/m3.1 (Cd = 0.80)

RELAP5Im3.1 (Cd = 0.85) -C,- RELAP5/m3.1 (Cd = 0.90)

9.0

8.5

8.0

('3

a)

I...

CO

U)

I

I

7.5

7.0

6.5

6.0

0 100 200

Figure 7.6

300 400 500 600 700 800 900

Time (s)

Comparison of Secondary Pressure of Intact Loop

(Cd Value Sensitivity)

1000
-- TE-245C-SGA

0 RELAP5/m3.1 (Cd = 0.75) RELAP5/m3.1 (Cd = 0.80)

-- RELAP5/m3.1 (Cd = 0.85) --- RELAP5/m3.1 (Cd = 0.90)

575

570

565

"L 560

E

555

550

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.7 Comparison of Secondary Temperature of Intact Loop

(Cd Value Sensitivity)
LE280-PR

-a- RELAP5/m3.1 (Cd = 0.75) RELAP5/m3.1 (Cd = 0.80)

RELAP5/m3.1 (Cd = 0.85) ---- RELAPS/m3.1 (Cd = 0.90) (K = 10.0) 3.0

2.5

2.0

1.5

-J 1.0

0.5

0.0

-0.5

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.8 Collapsed Liquid Level of Pressurizer

(Cd Value Sensitivity)
-- FE431+FE432+FE433+FE434 (SGA)

RELAP5/m3.1 (Cd = 0.75) RELAP5/m3.1 (Cd = 0.80)

RELAP5/m3.1 (Cd = 0.85) --- RELAP5/m3.1 (Cd = 0.90) (K = 10.0)

ý0,

4-0 ('3

0

LL

U,

CO

(U

18

16

14

12

10

8

6

4

2

0

-2

-4

*~....-. *~* .... .... . .. ....

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.9 Downcomer Flow Rate of Intact Loop

(Cd Value Sensitivity)
FE471-SGB

RELAP5/m3.1 (Cd = 0.75)

RELAP5/m3.1 (Cd = 0.85)

RELAP5/m3.1 (Cd = 0.80)

- RELAP5/m3.1 (Cd = 0.90)

25

20

d)

c7)

0

cu

U)

15

10

5

0

-5

0 100 200 300 .400 500 600 700 800 900 1000

Time (s)

Figure 7.10 Downcomer Flow Rate of Broken Loop

(Cd Value Sensitivity)
-- FE560A-BU

---- RELAP5/m3.1 (K = 0.0)

RELAP5/m3.1 (K = 50.0)

RELAP5/m3.1 (K = 10.0)

-RELAP5/m3.1 (K = 100.0)

8

7

(0

0)

0

co

6

5

4

I

3

2

0 50 100 150 200 250 300 350 400 450

Time (s)

500

Figure 7.11 Comparison of Break Flow Rate

(K Value Sensitivity)
....... PE300A-PR

RELAP5/m3.1 (K = 0.0) RELAP5/m3.1 (K = 10.0)

-.-- RELAP5/m3.1 (K = 50.0) -o-- RELAP5/m3.1 (K =100.0)

17

16

15

cu

CL

M)

14

13

I.

8

12

11

10

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.12 Comparison of Pressurizer Pressure

(K Value Sensitivity)
PE450-SGB

0 RELAP5/m3.1 (K = 0.0)

•-- RELAP5/m3.1 (K = 50.0)

-- ,RELAP5/m3.1 (K= 10.0)

.--- RELAP5/m3.1 (K = 100.0)

8

7

6

CU

(L

5

I0 4

3

2

1

0 100 200 300 400 500 600 700 800 900

Time (s)

1000

Figure 7.13 Comparison of Secondary Pressure of Broken Loop

(K Value Sensitivity)
TE-245C-SGB

- RELAP5/m3.1 (K = 0.0) RELAP5/m3.1 (K = 10.0)

-- RELAP5/m3.1 (K = 50.0) - RELAP5/m3.1 (K = 100.0)

580

560

540

C)

E

a)

520

500

480

460

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.14 Comparison of Secondary Temperature of Broken Loop

(K Value Sensitivity)
SP-VOLUME

RELAP5/m3.1 (K = 0.0) -,- RELAP5/m3.1 (K = 10.0)

RELAP5/m3.1 (K = 50.0) --- RELAP5/m3.1 (K = 100.0)

T I I I 1 1 -

a

0

(U

1.05

1.04

1.03

1.02

1.01

1.00

0.99

0.98

0.97

0.96

0.95

!.

0 100 200 300 400 500 600 ý 700 800 900 1000

Time (s)

Figure 7.15 Comparison of Break Void Fraction

(K Value Sensitivity)
PE430-SGA

RELAP5/m3.1 (K= 0.0) -,-RELAP5/m3.1 (K= 10.0)

RELAP5/m3.1 (K = 50.0) -.---- RELAP5/m3.1 (K = 100.0)

8.5 I IT

8.0

7.5

(0 7.0

13.

I

!

6.5

6.0

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.16 Comparison of Secondary Pressure of Intact Loop

(K Value Sensitivity)
TE-245C-SGA

RELAP5/m3.1 (K = 0.0) RELAP5/m3.1 (K = 10.0)

---- RELAP5/m3.1 (K = 50.0) -----. RELAP5/m3.1 (K = 100.0)

575.

570

565

E

a) I560

555

550

0 100 200 300 400 500 600 700 800 900 1000

Timrs 1Q1

Figure 7.17 Comparison of Secondary Temperature of Intact Loop

(K Value Sensitivity)
LE280-PR

RELAP5/m3.1 (K = 0.0)

RELAP5/m3.1 (K = 50.0)

-- ,RELAPS/m3.1 (K= 10.0)

RELAP51m3.1 (K = 100.0) (Cd = 0.85)

3.0

2.5

2.0

1.5

a,

-J

0*

I 1.0

I

0.5

0.0

-0.5

I

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.18 Collapsed Liquid Level of Pressurizer

(K Value Sensitivity)
FE431+FE432+FE433+FE434 (SGA)

RELAP5/m3.1 (K = 0.0) -,- RELAP5/m3.1 (K = 10.0)

RELAP5/m3.1 (K = 50.0) ---- RELAP5/m3.1 (K = 100.0) (Cd = 0.85)

C,)

0)

C',

18

16

14

12

10

8

6

4

______ I _ ______________________________ ______________________________ ______________________________ ______________________________ ______________________________ ______________________________

I

•D

,,-3

I

2

0

-2

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.19 Downcorner Flow Rate of Intact Side

(K Value Sensitivity)
FE471-SGB

RELAP5/m3.1 (K = 0.0) -,-RELAP5/m3.1 (K = 10.0)

-- RELAP5/m3.1 (K = 50.0) ----- RELAP5/m3.1 (K =100.0) (Cd = 0.85) 25

20

Cl)

0)

0

Uto

cl)

('o

15

10

5

0

-5

0 100 200 300 400 500 600 700 800 900

Time (s)

1000

Figure 7.20 Downcomer Flow Rate of Broken Side

(K Value Sensitivity)
-- FE560A-BU

RELAP5/m3.1 (Case 1)

RELAP5/m3.1 (Case 3)

RELAP5/m3.1 (Case 2)

---- RELAP5/m3.1 (Case 4)

8

7

(I

I

!

0

LL

cu

2'

6

5

4

3

2

0 50 100 150 200 250 300 350 400 450 500

Time (s)

Figure 7.21 Comparison of Break Flow Rate

(S/G Nodes Sensitivity)
PE300A-PR

RELAP5/m3.1 (Case 1)

RELAP5Im3.1 (Case 3)

RELAP5/m3.1 (Case 2)

RELAP5/m3.1 (Case 4)

17

16

15

(U 14

13 I

8

!

()

(L

u)

13.

12

11

10

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.22 Comparison of Pressurizer Pressure

(S/G Nodes Sensitivity)
PE450-SGB

RELAP5/m3.1 (Case 1) RELAP5/m3.1 (Case 2)

RELAP5/m3.1 (Case 3) - RELAP5/m3.1 (Case 4)

8

7

6

CU

5

4 0

3

2

1

0 100 200 300 400 500 600 700 800 900

Time (sW.

1000

Figure 7.23 Comparison of Secondary Pressure of Broken Loop

(S/G Nodes Sensitivity)
TE-245C-SGB

RELAP5/m3.1 (Case 1) RELAP5/m3.1 (Case 2)

RELAP5/m3.1 (Case 3) - RELAP5/m3.1 (Case 4)

580

560

540

CL

E

tJ

520

500 CD

480

460

440

0 100 200 300 400 500 600 700 800 900

Time (s).

1000

Figure 7.24 Comparison of Secondary Temperature of Broken Loop

(S/G Nodes Sensitivity)
SP-VOLUME

RELAP5/m3.1 (Case 1) RELAP5(m3.1 (Case 2)

RELAP5/m3.1 (Case 3) ----- RELAP5/m3.1 (Case 4)

CU

0~

LL

1.05

1.04

1.03

1.02

1.01

1.00

0.99

0.98

0.97

0.96

0.95

0

CA

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.25 Comparison of Break Void Fraction

(S/G Nodes Sensitivity)
-- PE430-SGA

RELAP5/m3.1 (Case 1) RELAP5/m3.1 (Case 2)

RELAP5/m3.1 (Case 3) 0---- RELAP5/m3.1 (Case 4)

8.5

8.0

12 7.5

U) 7.0

I-.

6.5

6.0

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.26 Comparison of Secondary Pressure of Intact Loop

(S/G Nodes Sensitivity)
TE-245C-SGA

RELAP5/m3.1 (Case 1) RELAP5/m3.1 (Case 2)

-- RELAP5/m3.1 (Case 3) - RELAP5/m3.1 (Case 4)

575 i i . ,

570

565

ID.

CL 560

U- E

I-.

555

550

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.27 Comparison of Secondary Temperature of Intact Loop

(S/G Nodes Sensitivity)
LE280-PR

RELAP5/m3.1 (Case 1) RELAP5/m3.1 (Case 2)

-w- RELAP5/m3.1 (Case 3) - RELAP5/m3.1 (CAse 4) (Cd = 0.85; K = 10.0) 3.0

2.5

2.0

1.5

Q-i 1.0

0.5

0.0

-0.5

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.28 Collapsed Liquid Level of Pressurizer

(S/G Nodes Sensitivity)
.......- FE431+FE432+FE433+Fe434 (SGA)

RELAP5/m3.1 (Case 1) RELAP5/m3.1 (Case 2)

RELAP5/m3.1 (Case 3) ---- RELAP5/m3.1 (Case 4) (Cd = 0.85; K = 10.0) 18

16

14

12

10

0

-3-,

cu

0

LL

U)

CU

8

6

4

2

0

-2

-4

11 MIA I

.........

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.29 Downcomer Flow Rate of Intact Loop

(S/G Nodes Sensitivity)
............ FE471-SGB

-o- RELAP5/m3.1 (Case 1) RELAP5/m3.1 (Case 2)

-A-- RELAP5/m3.1 (Case 3) --- RELAP5/m3.1 (Case 4)

25

20

U)

U,

0

LLU)

15

10

5

!

w

!

0

-5

0 100 200 300 400 500 600 700 800 900 1000

Time (s)

Figure 7.30 Downcomer Flow Rate of Broken Loop

(S/G Nodes Sensitivity)
ise 3 0 SEPARATOR

360 3N (31

MrENEI74O9

F `9*

Number of Nodes u 3

$00 SEPARATOR

w~w F '- 9'

DOWNCO . . . . ST AM

'I"OENRATOR

No

It2 20 SOA4 O

l~ ot j ot OAO

301 1 (Si

Nubro.Nd 5-

Number of Nodes a 5

AFrw

I I

EX TERNIM• 3 11 1 1 7

DOWN+COMEIR - . - STEAM

pot 4 II t CE ONEATOR

SM 9) IS S I$

to I 1 t

21? $ 01

Number of N~ods a g

A"W

SEPARATOR

GMEtRATOR STEAM

Number of Nodes a 7

Figure 7.31 Steam Generator Nodalization Diagram for The Nodes Sensitivity Study
----- Case I Case 2

--- ,-- Case 3 -- Case 4

(j,

E

P

0

80000

70000

60000

50000

40000

30000

20000

10000

0

0

0 100 200 300 400 500 600 700 800 900

Time (s)

Figure 7.32 The Required CPU Time for Calculations

(S/G Nodes Sensitivity)

1000
250000

200000

o 150000 (D(/)

E

D 100000

03

I50000

0

0 500 1000 1500 2000 2500 3000 3500

Time (s)

Figure 8.1 The Required CPU Time in the Base Calculation

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT tý UMBER

(2-8W) (Assgned b•, NRC, Add VoL, Supp., Rev. NRCM 12, BIB LIOGAPHIC DATA SHEET and Addend.=m Numbers. If any.)

3201.3202 BBIGAHCDT HE

(See inst s o the reverse) NJREGiIA-0148

2. TITLE AND SUBTITLE

Assessment of RELAP5/MOD3.1 Using LSTF Ten-Percent Main Steam-Line-Break Test Run

SB-SL-01 3. DATE REPORT PUBUSHED

September 1998

4. FIN OR GRANT NUMBER

D6227

5. AUTHOR(S) 6. TYPE OF F EPORT

J. G. Oh, H. D. Lee, K K. Jee, S. K. Kang/KOPEC

Y. S. Bang, K. W. Seul,/KINS

H. Kumamaru, Y. Anoda/JAERl 7. PERIOD COVERED (inclusive Dates)

8. PERFORMING ORGANIZATION -NAME AND ADDRESS (7f NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and meAilig address;f contractor,

provide name and mailing oddress.)

Korea Power Engineering Company Korea Institute of Nuclear Safety Japan Atomic Energy Research Institute

150 Duckjin-Dong, Yusong- Ku P.O. Box 114 Tokai-Mura, Naka-Gun

Taejon, Korea 305-353 Yusong, Taejon Ibaraki-Ken 319-1195, Japan

Korea 305-600

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, Te Same as above' if conractor, provide NRC Division, Office or Region, U.S. NuA,&erRegulatory Commission.

and mailing address)

Division of Systems Technology

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washinaton. DC 20555-0001

10. SUPPLEMENTARY NOTES

G. Rhee. NRC Proiect Manager

11. ABSTRACT (200 wrds or less)

Results produced by the RELAP5/MOD3.1 computer code were compared with the experimental data from JAERI's LSTF Test

Run SB-SL-01 for a 10% main steam line break transient in a pressurized water reactor. The code simulation for the base case

included a total of 189 fluid control volumes and 199 flow junctions to model the transient two-phase phenomena. Also, a total of

180 heat slabs were used to model the system heat transfer. The code predictions of the experimental results are generally

satisfactory for the trends of key parameters.

Sensitivity studies performed for the break discharge coefficient, the separator drain line loss coefficient, and this number of steam

generator nodes did not reveal any strong dependencies. Neverthless, optimal values of these parameters that led to the lowest

overall statistical error were obtained, and these values were subsequently used in the "Base Case" analysis.

12. KEY WORDS/DESCRIPTORS (List words orphrases that will assist researchers in locating the report.) 13. A ,AILABIUTY STATEMENT

RELAP5/M OD3.1 unlimited

JAERI LSTF 14. s:-CURrTY CLASSIFICATION

main steam line break test (Th;: Page)

assessment unclassified

small-break LOCA (loss-of-coolant-accident) (This Repo)

pressurized water reactor unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-89)
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