-NUREG/A-001 1

Intemnational

Agreement Report

TRAC-PF1 MOD1 Post Test

Calculations of the OECD

LOFT Experiment LP-SB-1

Prepared by

E. J. Allen

United Kingdom Atomic Energy Authority

Winfrith, Dorchester

Dorset, England

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

April 1990

Prepared as part of

The Agreement on Research Participation and Technical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report was prepared under an international cooperative

agreement for the exchange of technical information. Neither

the United States Government nor any agency thereof, or any of

their employees, makes any warranty, expressed or implied, or

assumes any legal liability or responsibility for any third party's

use, or the results of such use, of any information, apparatus product or process disclosed in this report, or represents that its use

by such third party Would not infringe privately owned rights.

Available from

Superintendent of Documents

U.S. Government Printing Office

P.O. Box 37082

Washington, D.C. 20013-7082

and

National Technical Information Service

Springfield, VA 22161
NUREG/IA-001 1

International

Agreement Report

TRAC-PF1 MOD1 Post Test

Calculations of the OECD

LOFT Experiment LP-SB-1

Prepared by

E. J. Allen

United Kingdom Atomic Energy Authority

Winfrith, Dorchester

Dorset. England

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

April 1990

Prepared as part of

The Agreement on Research Participation and Technical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report is based on work performed under the sponsorship of the

United Kingdom Atomic Energy Authority. The information in this

report has been provided to the USNRC under the terms of the

International Code Assessment and Application Program (ICAP)

between the United States and the United Kingdom (Administrative

Agreement - WH 36047 between the United States Nuclear Regulatory

Commission and the United Kingdom Atomic Energy Authority Relating

to Collaboration in the Field of Modelling of Loss of Coolant

Accidents, February 1985). The United Kingdom has consented to the

publication of this report as a USNRC document in order to allow

the widest possible circulation among the reactor safety community.

Neither the United States Government nor the United Kingdom or any

agency thereof, or any of their employees, makes any warranty,

expressed or implied, or assumes any legal liability of

responsibility for any third party's use, or the results of such

use, or any information, apparatus, product or process disclosed

in this report, or represents that its use by such third party

would not infringe privately owned rights.
SUMMARY

Analysis of the small, hot leg break, OECD LOFT Experiment

LP-SB-I using the "best-estimate" computer code TRAC-PFlI/MOD1 is

presented.

Descriptions of the LOFT facility and the LP-SB-1 experiment are

given and development of the TRAC-PF1/MOD1 input model is

detailed. The calculations performed in achieving the

steady state conditions, from which the experiment was initiated,

and the specification of experimental boundary conditions are

outlined.

Results of a "Base Case" transient calculation are found to be

generally consistent with those reported by other members of the

OECD LOFT Program Review Group. The experimental trends with

respect to pressure histories and minimumsystem mass inventory

are reasonably well reproduced by the TRAC-PFl/MODl calculation.

However, the inability of TRAC-PF1/MODI to account for main-pipe

stratification in determining fluid conditions in a side branch

leads to significant discrepancies between the measured and

predicted break line and hot leg densities and is identified as

the main reason for the poorly predicted break mass flow rate.

Implementation, via the TRAC-PFI/MODI control system, of

correlations for determining side branch quality as a function of

main-pipe stratified liquid level are shown to be effective in

improving the predicted hot leg and break line densities and

break mass flow rate. The remaining differences between measured

and predicted data are considered to be due to deficiencies in

the TRAC-PFl/MOD1 critical flow model and the sensitivity of the

break flow to the hot leg liquid level behaviour.

It is recommended that some means of accounting for the effect of

main-pipe stratified liquid level, in determining fluid

conditions in a side branch, be implemented in the TRAC-PFl/MODl

code.

It is also suggested that a closer examination of the factors

influencing the draining of the steam generator tubes is required

to resolve the observed discrepancies in hot leg liquid level

behaviour.

AEEW - R 2254 i

CONTENTS

PAGE

SECTION 1. INTRODUCTION 1

2. THE LOSS OF FLUID TEST (LOFT) FACILITY 1

3. EXPERIMENT LP-SB-1 2

4. TRAC-PFI/MODI 3

5. TRAC-PFI/MODI INPUT MODEL FOR LP-SB-I 3

6. STEADY STATE CALCULATION 4

7. BOUNDARY CONDITIONS FOR TRANSIENT 5

CALCULATIONS

7.1 Decay Heat Data 5

7.2 Primary Pump Injection 5

7.3 Steam Generator Secondary Side 6

Auxiliary Feedwater Flow

7.4 High Pressure Injection System 6

7.5 Operational Setpoints 6

8. BASE CASE CALCULATION 6

8.1 Introduction 6

8.2 CPU Usage and Time Step Behaviour 6

8.3 Chronology of Events 7

8.4 General Observations 8

8.4.1 Break Mass Flow Rate and Break 8

Line Density

8.4.2 Primary System Densities 8

8.4.3 Primary System Mass Inventory 9

8.4.4 System Pressure 9

8.5 Detailed Discussion 10

8.5.1 Break Mass Flow Rate and 10

Break Line Density

8.5.2 Running Speed 11

8.6 Summary 11

9. BASE CASE CALCULATION WITH EPRI CORRELATION 12

9.1 Introduction 12

9.2 General Observations 12

9.2.1 Break Mass Flow Rate and 12

Break Line Density 9.2.2 Primary System Densities 12

9.2.3 Primary System Mass Inventory 12

9.2.4 System Pressure 13

AEEW - R 2254 iii
CONTENTS (Cont'd)

PAGE

SECTION 9.3 Detailed Discussion 13

.9.3.1 ILHL Liquid Level Behaviour

9.3.2 Time .f Break Uncovery

9.4 Summary

10. BASE CASE CALCULATION WITH EPRI CORRELATION

AND INPUT MODEL MODIFICATIONS

10.1 Introduction

10.2 General Observations

10.2.1 Break Mass Flow Rate and

Break Line Density

10.2.2 Primary System Densities

10.2.3 Primary System Mass Inventory

10.2.4 System Pressure

10.3 Detailed Discussion

10.3.1 ILHL Liquid Level Behaviour

10.3.2 Choice of Correlation

10.3.3 Choice of Choked Flow

Multiplier

13

13

14

15

15

16

16

16

16

16

16

16

17

18

18

19

20

20

20

10.4 Summary

11. CONCLUSIONS

12. RECOMMENDATIONS

13. ACKNOWLEDGEMENTS

14. REFERENCES

TABLES

FIGURES

APPENDIX I -

APPENDIX II -

APPENDIX III -

MICROFICHE LISTING OF THE TRAC-PFI/MODI

INPUT DECK FOR LP-SB-l (USED FOR THE

STEADY STATE CALCULATION)

THE EPRI CORRELATION FOR BRANCHLINE FLOW

QUALITY AS A FUNCTION OF MAINLINE

STRATIFIED LIQUID LEVEL

EFFECT OF DEFICIENCIES IN THE

TRAC-PFI/MOD1 INTERPHASE DRAG MODEL ON

THE PREDICTION OF THE CORE DENSITY FOR

LP-SB-l

AEEW - R 2254 iv
CONTENTS (Cont'd)

APPENDIX IV - MODIFICATIONS TO LP-SB-1 BASE CASE INPUT

DECK FOR FINAL TRANSIENT CALCULATION

APPENDIX V - LP-SB-I BASE CASE CALCULATION WITH EPRI

CORRELATION AND INPUT MODEL

MODIFICATIONS - SERIES OF PICTURES

SHOWING PREDICTED SYSTEM CONDITIONS

THROUGHOUT THE TRANSIENTý

AEEW - R 2254 V
TABLES

1. INITIAL CONDITIONS FOR EXPERIMENT LP-SB-1

2. PRESSURIZER VOLUME

3. STEADY STATE PRESSURE DROPS

4. STEADY STATE ENVIRONMENTAL HEAT LOSSES

5. STEADY STATE CORE BYPASS FLOW RATES

6. TRANSIENT CALCULATION - DECAY HEAT POWER TABLE

7. TRANSIENT CALCULATION - STEAM GENERATOR SECONDARY SIDE

AUXILIARY FEEDWATER FLOW RATE TABLE

8. TRANSIENT CALCULATION - HIGH PRESSURE INJECTION SYSTEM

(HPIS) FLOW RATE TABLE

9. OPERATIONAL SETPOINTS FOR EXPERIMENT LP-SB-1

10. LP-SB-1 BASE CASE TRANSIENT CALCULATION - CHRONOLOGY OF

EVENTS

11. RESULTS FROM "STAND-ALONE" BREAK LINE MODEL

12. COMPARISON OF COEFFICIENTS USED IN VAPOUR PULL THROUGH AND

LIQUID ENTRAINMENT CORRELATIONS

AEEW - R 2254 vi
FIGURES

1. LOFT SYSTEM CONFIGURATION FOR EXPERIMENT LP-SB-1

2. LP-SB-1 PRIMARY SYSTEM PRESSURE AND TIMINGS OF SIGNIFICANT

EVENTS

3. TRAC-PFI/MOD1 PRIMARY SYSTEM NODALISATION DIAGRAM FOR

LP-SB-I

4. TRAC-PFI/MOD1 REACTOR VESSEL NODALISATION DIAGRAM FOR

LP-SB-1

5. TRAC-PF1/MOD1 STEAM GENERATOR SECONDARY SIDE NODALISATION

DIAGRAM FOR LP-SB-1

6. LP-SB-1 STEADY STATE CALCULATION -

7. LP-SB-1 STEADY STATE CALCULATION -

8. LP-SB-1 STEADY STATE CALCULATION -

ACROSS CORE

9. LP-SB-1 STEADY STATE CALCULATION -

10. LP-SB-1 STEADY STATE CALCULATION -

(ILHL) TEMPERATURE

11. LP-SB-1 STEADY STATE CALCULATION -

(ILCL) TEMPERATURE

12. LP-SB-1 STEADY STATE CALCULATION -

(ILHL) MASS FLOW RATE

13. LP-SB-1 STEADY STATE CALCULATION -

14. LP-SB-1 STEADY STATE CALCULATION -

15. LP-SB-1 STEADY STATE CALCULATION -

SIDE (SGS) LIQUID LEVEL

CPU USAGE

TIME STEP SIZE

TEMPERATURE DIFFERENCE

PRIMARY SYSTEM PRESSURE

INTACT LOOP HOT LEG

INTACT LOOP COLD LEG

INTACT LOOP HOT LEG

PUMP SPEED

REACTOR POWER

STEAM GENERATOR SECONDARY

16. LP-SB-1 STEADY STATE CALCULATION - STEAM GENERATOR SECONDARY

SIDE (SGS) TEMPERATURE (BOTTOM OF DOWNCOMER)

17. LP-SB-1 STEADY STATE CALCULATION - STEAM GENERATOR SECONDARY

SIDE (SGS) PRESSURE

18. LP-SB-1 STEADY STATE CALCULATION - STEAM GENERATOR SECONDARY

SIDE (SGS) STEAM MASS FLOW RATE

19. LP-SB-1 STEADY STATE CALCULATION - STEAM GENERATOR SECONDARY

SIDE (SGS) FEEDWATER MASS FLOW RATE

20. LP-SB-1 STEADY STATE CALCULATION - PRESSURIZER LIQUID

TEMPERATURE

21. LP-SB-1 STEADY STATE CALCULATION - PRESSURIZER PRESSURE

AEEW - R 2254 vii
FIGURES (Cont'd)

22. LP-SB-1 STEADY STATE CALCULATION - PRESSURIZER LIQUID LEVEL

23. LP-SB-1 STEADY STATE CALCULATION - BROKEN LOOP COLD LEG

(BLCL) TEMPERATURE

24. LP-SB-1 BASE CASE CALCULATION CPU USAGE

25. LP-SB-1 BASE CASE CALCULATION - TIME STEP SIZE

26. LP-SB-1 BASE CASE CALCULATION - BREAK MASS FLOW RATE

27. LP-SB-1 BASE CASE CALCULATION - BREAK UPSTREAM DENSITY

28. LP-SB-1 BASE CASE CALCULATION - BREAK MASS FLOW RATE,

COMPARISON WITH-OECD LOFT.PROGRAM REVIEW GROUP

29. LP-SB-1 BASE CASE CALCULATION - INTACT LOOP HOT LEG (ILHL)

DENSITY

30. LP-SB-1 BASE CASE CALCULATION - HOT LEG DENSITY, COMPARISON

WITH OECD LOFT PROGRAM REVIEW GROUP

31. LP-SB-1 BASE CASE CALCULATION - INTACT LOOP COLD LEG (ILCL)

DENSITY

32. LP-SB-1 BASE CASE CALCULATION - COLD LEG DENSITY, COMPARISON

WITH OECD LOFT PROGRAM REVIEW GROUP

33. LP-SB-1 BASE CASE CALCULATION - PRIMARY SYSTEM MASS

INVENTORY, COMPARISON WITH OECD LOFT PROGRAM REVIEW GROUP

34. LP-SB-1 BASE CASE CALCULATION - CORE VOID FRACTIONS

35. LP-SB-1 BASE CASE CALCULATION - PRIMARY SYSTEM PRESSURE

36. LP-SB-1 BASE CASE CALCULATION - PRIMARY SYSTEM PRESSURE,

COMPARISON WITH OECD LOFT PROGRAM REVIEW GROUP

37. LP-SB-1 BASE CASE CALCULATION - SECONDARY SYSTEM PRESSURE

38. LP-SB-1 BASE CASE CALCULATION - SECONDARY SYSTEM PRESSURE,

COMPARISON WITH OECD LOFT PROGRAM REVIEW GROUP

39. LP-SB-1 BASE CASE CALCULATION - ILHL VOID FRACTION,

STRATIFICATION ONSET AND BREAK FLOW

40. LP-SB-1 BASE CASE CALCULATION - EFFECT OF SIZE OF BYPASS

PATH ON RUNNING SPEED

AEEW - R 2254 viii
FIGURES (Cont'd)

41. LP-SB-1 BASE CASE CALCULATION WITH

UPSTREAM DENSITY

42. LP-SB-1 BASE CASE CALCULATION WITH

MASS FLOW RATE

43. LP-SB-1 BASE CASE CALCULATION WITH

LOOP HOT LEG (ILHL) DENSITY

44. LP-SB-1 BASE CASE CALCULATION WITH

LOOP COLD LEG (ILCL) DENSITY

45. LP-SB-1 BASE CASE CALCULATION WITH

PRIMARY SYSTEM MASS INVENTORY

46. LP-SB-1 BASE CASE CALCULATION WITH

PRIMARY SYSTEM PRESSURE

47. LP-SB-1 BASE CASE CALCULATION WITH

SECONDARY SYSTEM PRESSURE

EPRI CORRELATION - BREAK

EPRI CORRELATION - BREAK

EPRI CORRELATION - INTACT

EPRI CORRELATION - INTACT

EPRI CORRELATION -

EPRI CORRELATION -

EPRI CORRELATION -

48. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION - INTACT

LOOP HOT LEG (ILHL) AND VAPOUR PULL-THROUGH LEVELS

49. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION -

INTACT LOOP HOT LEG (ILHL) LIQUID LEVEL

50. EFFECT ON BREAK LINE DENSITY OF a) INTACT LOOP HOT LEG

(ILHL) LIQUID LEVEL (OR THE LEVEL AT WHICH VAPOUR PULL.:

THROUGH IS PREDICTED TO OCCUR) AND b) AMOUNT OF MASS IN

SYSTEM REQUIRED TO BE DISCHARGED PRIOR TO BREAK UNCOVERY

51. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION - VAPOUR

PULL-THROUGH EFFECT ON BREAK LINE DENSITY

52. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - BREAK MASS FLOW RATE

53. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - BREAK UPSTREAM DENSITY

54. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - INTACT LOOP HOT LEG (ILHL)

DENSITY

55. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - INTACT LOOP COLD LEG (ILCL)

DENSITY

56. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - PRIMARY SYSTEM MASS INVENTORY

57. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - PRIMARY SYSTEM PRESSURE

AEEW - R 2254 ix
FIGURES (Cont'd)

58. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - SECONDARY SYSTEM PRESSURE

59. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - INTACT LOOP HOT LEG (ILHL)

LIQUID LEVEL AND NATURAL CIRCULATION

60. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS, INTACT LOOP HOT LEG (ILHL) LIQUID

LEVEL, COMPARISON WITH OECD LOFT PROGRAM REVIEW GROUP

61. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - INTACT LOOP HOT LEG (ILHL) MASS

FLOW RATES

62. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - INTACT LOOP HOT LEG (ILHL) MASS

FLOW BALANCE (FLOW FROM VESSEL TO BREAK + FLOW FROM STEAM

GENERATOR TO BREAK - BREAK FLOW)

63. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - EFFECT OF CHOICE OF CORRELATION

ON BREAK LINE DENSITY

64. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - EFFECT OF CHOICE OF CORRELATION

ON BREAK MASS FLOW RATE

65. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - EFFECT OF CHOICE OF CORRELATION

ON INTACT LOOP HOT LEG (ILHL) LIQUID LEVEL

66. LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - EFFECT OF CHOICE OF CHOKED FLOW

MULTIPLIER ON BREAK MASS FLOW RATE

AEEW - R 2254 X
COMMERCIAL IN CONFIDENCE

1. INTRODUCTION

This paper describes post-test calculations of the OECD LOFT

small, hot leg break experiment LP-SB-1 using the "best-estimate"

computer code TRAC-PFI/MOD1. Sections 2, 3 and 4 describe the

LOFT facility, the LP-SB-l experiment and the versions of

TRAC-PFl/MODl used, respectively. Development of the input model

is detailed in Section 5 and the calculations performed i.

achieving the steady state conditions, from which the experiment

was initiated, are outlined in Section 6. The experimental

boundary conditions, and the way in which they are specified to

the code, are defined in Section 7. Section 8 describes the

"Base Case" transient calculation. The effects of implementing

correlations for predicting branch line quality as a function of

main branch stratified liquid level are examined in Sections 9

and 10. The effect of varying the choked flow multiplier is also

discussed in Section 10. The main conclusions and

recommendations from the analysis are summarised in Sections 11

and 12.

2. THE LOSS OF FLUID TEST (LOFT) FACILITY

The Loss of Fluid Test (LOFT) facility, at the Idaho National

Engineering Laboratory (INEL), is a 50 MW(t) Pressurised Water

Reactor (PWR) system designed to simulate the major components

and system responses of a commercial PWR during Loss-of-Coolant

Accidents (LOCAs) or operational transient accidents. The

experimental'assembly is instrumented in order that system variables can be measured and recorded during transients. The

facility is comprised of five major subsystems - the reactor

vessel, the operating (intact) loop, the "broken" loop,' the

blowdown suppression system and the Emergency Core Cooling System

(ECCS). The configuration of the major LOFT components, for

experiment LP-SB-l, is shown in Figure 1.

The operating (intact) loop simulates three loops of a commercial

four-loop PWR and contains a steam generator (of vertical, U-tube

design), two primary coolant pumps (in parallel), a pressurizer,

a venturi flowmeter and connecting piping. The break location

for experiment LP-SB-1 was in the hot leg of the intact loop

between the steam generator and the reactor vessel.

The broken loop consists of a hot leg and a cold'leg connected to

the reactor vessel and the blowdown suppression tank header.

Each leg contains a Quick-Opening Blowdown Valve (QOBV), a

recirculation line, an isolation valve and connecting piping.

The recirculation lines provide a small flow from the broken loop

to the intact loop and are used to maintain the broken loop fluid

temperature at approximately the core inlet temperature prior to

experiment initiation. During experiment LP-SB-1, the QOBVs and

the isolation valves remained closed (because the break was in

the operating loop). The broken loop spool pieces, with orifices

to simulate the steam generator and pump hydraulic resistances,

were not installed for experiment LP-SB-1. These were replaced

by a straight piping spool piece.

AEEW - R 2254 1
The LOFT reactor vessel has an annular downcomer, a lower plenum,

lower core support plates, a nuclear core (containing 1300 fuel

rods) and an upper plenum. The downcomer is connected to the

cold legs of the operating and broken loops, and the upper plenum

is connected to the hot legs of the operating and broken loops.

The LOFT ECCS consists of two accumulators, a High Pressure

Injection System (HPIS) and a Low Pressure Injection System

(LPIS). Each system is designed to inject scaled flows of

emergency core coolant directly into the primary coolant system.

The accumulators and LPIS were not used during experiment LP-SB-l

and scaled HPIS flow was directed into the Intact Loop Cold Leg

(ILCL). Volume scaling of the HPIS flow was based on the

assumption that only one of the three charging pumps and one of

the three HPIS pumps, in the reference plant, were available.

3. EXPERIMENT LP-SB-l

Experiment LP-SB-1 was conducted on 23 June .1983 in the LOFT

facility at the Idaho National Engineering Laboratory. LP-SB-l

was the second in a series of experiments, sponsored by a

consortium of countries under the auspices of the Organisation

for Economic Cooperation-and Development (OECD), designed to

address small break issues raised as a result of the accident at

Three Mile Island in 1979.

The LP-SB-l experiment simulated a 7.6 cm (3 inch) equivalent

diameter break in a hot leg pipe of a commercial PWR. LP-SB-1

was one of a pair of experiments aimed at addressing the effects

of early and delayed pump trip on system behaviour. The primary

coolant pumps were tripped early in experiment LP-SB-1 and pump

trip was delayed in experiment LP-SB-2. The following objectives

were defined for the two small break experiments (1):

i) Determine system transient response characteristics for

hot leg small break LOCAs with early and delayed pump

trip and break size of 7.62 cm (3 inch) equivalent

diameter.

ii) Determine the system mass inventory, mass distribution,

and core heat transfer characteristics when pumps are

shut off under high system void conditions (LP-SB-2).

iii) Provide integral nuclear system data for assessing the

ability of computer codes to predict system response

during a small break LOCA.

iv) Obtain data which can be used to investigate emergency

core coolant distribution, thermal mixing and effect on

core coolant mass inventory.

v) Provide data for evaluating the usefulness of accident

diagnosis techniques in identifying small hot leg break

LOCA characteristics.

AEEW - R 2254 2
Experiment LP-SB7 l was initiated, from operating conditions

representative of those in a commercial PWR, by opening a valve

in the Intact Loop Hot Leg (ILHL) break line. The primary side

pressure history and the timings of significant events during the

experiment are shown on Figure 2. The primary coolant system

pressure decreased to the reactor scram and feedwater trip

setpoint (14.57 MPa) in 1.4 seconds. The Main Steam Control

Valve (MSCV) was closed manually upon verification cf reactor

scram. The main feedwater was isolated at 3.8 seconds and the

MSCV was fully closed at 15.4 seconds. The primary coolant

system pressure decreased rapidly to the primary coolant pump

trip setpoint (11.12 MPa) at 24.6 seconds and the pressurizer

indicated zero liquid level at 34.6 seconds. The HPIS setpoint

signal of 8.24 MPa was reached at 41.4 seconds and the system

pressure had declined to fluid saturation in the break line at

57.5 seconds. Auxiliary feedwater was manually initiated at 1.1

minutes and turned off at 31.1 minutes. Following break uncovery

at 11.9 minutes, the high quality steam flow out of the break

caused a further acceleration in the primary coolant system

depressurisation rate. The primary system pressure fell below

the steam generator secondary pressure at approximately 18

minutes. The minimum primary coolant system mass inventory was

reached after 37 minutes at which time the HPIS flow rate

exceeded the break flow rate and primary coolant system refill

began. The experiment was terminated, after one hour, when the

primary coolant system pressure had fallen to 2.5 MPa. The

liquid inventory in the reactor vessel remained at least 1.5 m

above the top of the core during the transient and sufficient

cooling was present to keep the fuel cladding temperatures close

to the saturation temperature of the fluid in the reactor

vessel.

4. TRAC-PFl/MOD1

TRAC (Transient Reactor Analysis Code) is an advanced "bestestimate" computer code, developed at the Los Alamos National

Laboratory, for analysing transients in thermal hydraulic

systems. Specifically, TRAC-PFl/MOD1 was developed for analysing

postulated accidents in PWRs. The versions of the code used for

the calculations described in this paper were Version B02A and

Version B02C which contain the LANL updates to TRAC-PFl/MOD1

Version 12.7.

5. TRAC-PFl/MODI INPUT MODEL FOR LP-SB-l

The development of a TRAC-PFl/MODl input model for analysis of

the LOFT.small break experiment LP-SB-l was based on a

TRAC-PFl/MODI large break deck for LP-FP-1. The FP-l deck,

developed at AEEW, originated from the LANL input deck for

experiment L2-3. Additional published data on the LOFT facility

(2,3,4) were employed, where necessary, in producing the small

break deck.

The required modifications to the large break deck are described

in References 5 and 6 and include:

AEEW - R 2254 3
i) Replacement of the three-dimensional vessel with a

one-dimensional representation.

ii) Removal of the accumulator and the LPIS.

iii) Removal of the broken loop steam generator and pump

simulators.

iv) Inclusion of the ILHL break.

v) Inclusion of the steam generator secondary side

auxiliary feedwater.

vi) Inclusion of the primary pump injection.

Additional modifications by Neill (7) - to the position of the

ECCS injection in the ILCL - and Pelayo (8) - to the steam

generator recirculation ratio and nodalisation - were also

incorporated in the LP-SB-I input deck.

Figures 3, 4 and 5 show the nodalisation diagrams for the primary

system, the reactor vessel and the steam generator secondary

side, respectively. A total of 36 components, 42 junctions and

147 cells were used in the model.

A microfiche listing of the TRAC-PFl/MODI input deck for LP-SB-l

(used for the steady state calculation) is contained in

Appendix I.

6. STEADY STATE CALCULATION

Version B02A of the TRAC-PFI/MOD1 code - which incorporates the

updates contained in LANL Version 12.7 - was used for the steady

state calculations. Steady state mode calculations were run for

550 seconds and, in order to determine system conditions during

transient mode code operation, a short period (50 seconds) of

transient mode "steady state" (ie with no BREAK in the circuit)

was also run. In running the steady state calculation, a total

of 680 seconds of CPU time were used with an average time step

size of 0.12 seconds (see Figures 6 and 7). It was found that

the calculation converged to a steady state more readily when the

maximum allowable time step was reduced from 1.0 seconds to

0.1 seconds.

The calculations were performed with control systems governing

the behaviour of the steam generator secondary side steam and

feedwater mass flow rates and the speed of the primary coolant

pumps.

The initial conditions predicted by TRAC-PFl/MODl for experiment

LP-SB-I are compared with the measured data in Table 1. The

calculations produced stable initial conditions, within the

quoted experimental uncertainties (see Figures 8-23*), for all

The steady decline in primary system pressure and pressurizer

temperature, pressure and liuid level, at 550 seconds (as

indicdted 'In Fiqures 9 20, 21 and _22 repectivelyl is due to

the differing treatment, by TRAC-PF1 /MOD1, of the PRESSURIZER

component when in the transient mode, as opposed to the steady state mode, of code operation.

AEEW - R 2254 4
significant parameters-with the exception of the steam volume and

the liquid level in the pressurizer. The figures predicted for

these quantities were outside the uncertainties of the

experimental data due, it is thought, to differences between the

experiment and the TRAC-PFI/MOD1 input deck in the interpretation

of the pressurizer geometry. The pressurizer volume implied by

the LP-SB-1 Experiment Analysis and Summary Report (EASR) (9),

the TRAC-PFl/MOD1 large break input decks and the LOF•

specification (2) differ as shown in Table 2. Efforts to resolve

the discrepancies were unsuccessful. The approach adopted, in

performing the steady state calculations, was to specify the

initial pressurizer liquid volume as quoted for the experiment,

but to allow the initial steam volume to be outside the

uncertainties of the experimental data.

The magnitudes of the steady state pressure drops around the

primary circuit, the environmental heat losses from the system

and the core bypass flow rates, obtained from the TRAC-PFl/MODl

calculations, were in reasonable agreement with the available

LOFT data as shown 'in Tables 3, 4 and 5, respectively.

Achievement of these initial system conditions is discussed in

detail in References 5 and 6.

7. BOUNDARY CONDITIONS FOR TRANSIENT CALCULATIONS

7.1 Decay Heat Data

Following -reactor scram, decay heat data were specified to the

TRAC-PFI/MODl transient calculation by means of a "power versus

time" table. In deriving the table, the approach adopted by Hall

and Brown (14) was followed and two sources of information were

used. During the first 250 seconds of the transient, when both

neutron and fission power were expected to be present, data were

taken from the RELAP5/MODI input deck used for the pre-test

prediction of LP-SB-I (15). When fission power only was present,

the data contained in Reference 16 were appropriate and these

data were used from 250 seconds until the end of the transient.

The power table used in the TRAC-PFI/MODI calculation of LP-SB-1

is reproduced as Table 6.

7.2 Primary Pump Injection

During experiment LP-SB-,, the primary coolant pump injection

system was set up to deliver a total flow of 0.095 is. to the

primary coolant pumps (9, 17). Thiswas simulated in the

TRAC-PFl/MODl model by using "FILL" components to supply the

primary pump injection system with liquid at a constant rate of

1.2974 x 10-3 ms- 1 . The flow areas of the pump injection pipes

were 3.6613 x 10-2 m2 which implied an injection rate of 4.75 x

10-5 m3 s- 1 , or 0.0475 s 1 , to each pump.

AEEW - R 2254 5
7.3 Steam Generator Secondary Side Auxiliary Feedwater Flow

In experiment LP-SB-l, the steam generator secondary side

auxiliary feedwater flow, of 0.5 is-1 (17), was manually

initiated at 63.4 seconds and turned off at 1864.8 seconds (9).

This was simulated, in Qhe TRAC-PFI/MODI calculation, by using a

"FILL" component to provide feedwater to the secondary side of

the steam generator at a flow rate of 0.061673 ms- 1 during the

required period. The flow area of the auxiliary feedwater system

pipework was 0.0081073 m2 implying a flow rate of 5,x 10-4 m S-1

or 0.5 1s-1. The "time versus velocity" table used is given in

Table 7.

7.4 High Pressure Injection System

The HPIS was initiated in experiment LP-SB-1 when the ILHL

pressure had fallen to 8.24 MPa (9). The table of HPIS flow rate

against Primary Coolant System (PCS) pressure, used in the

TRAC-PFl/MOD1 calculation, was derived from that given in the

Experiment Specification Document (17) and is reproduced in

Table 8.

7.5 Operational Setpoints

The operational setpoints (for reactor scram, main feedwater shut

off, MSCV closure, primary pump trip, HPIS initiation and

auxiliary feedwater initiation) measured during the experiment,

and the way in which the setpoints were specified in the

TRAC-PFl/MOD1 calculation, are given in Table 9.

8. BASE CASE CALCULATION

8.1 Introduction

The initial transient calculation, termed the "Base Case

Calculation", was restarted from the end of the transient-mode

"steady state" calculation. The FILL component, originally

attached to the break line, was replaced by a BREAK component in

order to initiate the transient. As for the steady state

calculations, Version BV2A of TRAC-PFl/MOD1 - which incorporates

the code updates contained in LANL Version 12.7 - was used for

the Base Case Calculation.

In this Section, the TRAC-PFl/MODl predictions are compared with

the experimental data and with the results presented - by members

of the OECD LOFT Program Review Group - in the LP-SB-I

"Comparison Report" (18).

8.2 CPU Usage and Time Step Behaviour

Four thousand seconds of elapsed transient were calculated,

requiring 11,332 seconds of CRAY X-MP CPU time (see Figure 24).

This corresponds to a CPU/real time ratio of 2.8. The

user-specified minimum allowable time step throughout the

calculation was 10-5 seconds. The maximum time step was limited

to 0.5 seconds for the first 1500 seconds of the transient.

AEEW - R 2254 6
Shortly after this time, the code attempted to reduce the time

step below the minimum allowable value. It was necessary to

reduce the maximum permitted time step to 0.1 seconds, for the

remainder of the transient, to enable the calculation to proceed

(see Figure 25). The average time step for the calculation

(problem time/total number of time steps) was 0.08 seconds. The

very small cell (7.238 x 10-6 M3 ) representing the leakage path

between the reactor vessel and the core barrel nrozzles (bypass

path 4 - component number 83) was the dominant component

responsible for limiting the time step size throughout the

transient.

8.3 Chronology of Events

A comparison of the measured and predicted timings of significant

events during the LP-SB-1 transient is given in Table 10.

The experiment was initiated by opening the valve in the ILHL

break line. The primary coolant system pressure decreased

rapidly to the reactor scram and main feedwater trip set point of

14.57 MPa and, following a 2 second delay, closure of the main

steam control valve was initiated. Isolation of the main

feedwater took 2.4 seconds. The timings of these initial events

were predicted, by the TRAC-PFI/MOD1 calculation, to within -1

second.

In the experiment, the main steam control valve was fully closed

at 15.4 seconds. Although not documented in the experiment

specification, the steam flow bypass valve was opened onceduring

the experiment, at - 30 seconds, when the secondary side pressure

exceeded - 6.7 MPa. This was simulated, in the TRAC-PFl/MODl

calculation, by allowing the main steam control valve to reopen.

The primary coolant system pressure continued to decrease rapidly

and reached the primary coolant pump trip set point (11.12 MPa)

after 24.6 seconds and the HPIS initiation set point (8.24 MPa)

after 41.4 seconds. The timings of both these trips were very

well predicted by the TRAC calculation.

After 43 seconds, the primary coolant pumps had coasted down to

their flywheel uncoupling frequency (12.5 Hz). TRAC predicted

this to occur some 3 seconds later than in the experiment.

Fluid saturation in the break line, signalling the end of

subcooled blowdown, occurred at 57.5 seconds in the experiment.

This was predicted to occur some 7 seconds later in the

calculation.

The auxiliary feedwater was initiated at 63.4 seconds and turned

off at 1864.8 seconds. Identical timings were used for the TRAC

calculation.

A further acceleration in the experimental primary coolant system

depressurisation rate occurred when the break started to uncover

at 715 seconds. Prior to this time, and following the end of

subcooled blowdown, the break mass flow rate was under-predicted

AEEW - R 2254 7
by TRAC and the start of break uncovery was not calculated until

1200 seconds.

The time at which the primary coolant system pressure fell below

the secondary system pressure (1077 seconds in the experiment)

was over-predicted by some 300 seconds.

Minimum primary system coolant mass inventory is estimated to

have occurred at between 1800 seconds and 2200 seconds in the

experiment. The predicted time of minimum primary system mass

inventory was - 1680 seconds (defined as the time at which the

HPIS plus pump injection mass flow rates exceeded the break mass

flow rate).

The experiment was terminated at 3668 seconds when the primary

coolant system pressure had fallen-to the termination criterion

of 2.487 MPa. This was predicted to occur over 400 seconds

earlier, at 3227 seconds, in the TRAC calculation.

8.4 General Observations

8.4.1 Break Mass Flow Rate and Break Line Density

The measured and predicted break mass flow rates and break

upstream densities are shown in Figures 26 and 27, respectively.

(Subcooled and two-phase choked flow multipliers of 1.0 were used

throughout the calculation). It can be seen that, prior to the

time of measured break uncovery (- 700 seconds), TRAC-PFl/MODl

under-predicted the break line density and the break mass flow

rate, The time at which the break was predicted to uncover

occurred some 500 seconds later than in the experiment. Figure

28 indicates that, with the exception of the GRS

(W Germany - DRUFAN 02) results, all participants of the OECD

LOFT Program Review Group (18) also under-predicted the break

mass flow rate during the first - 700 seconds of the transient

and over-predicted the time of break urfcovery.

8.4.2 Primary System Densities

A comparison of the measured and predicted ILHL densities is

shown in Figure 29. 1- nt't tr the exnerimental behaviour,

the TRAC-PFl/MODI calculation predicted that the ILHL emptied

completely at - 1500 seconds. A similar trend was observed by

participants of the OECD LOFT Program Review Group (18), as shown

in Figure 30.

A comparison of the measured and predicted ILCL densities is

shown in Figure 31. In common with the results from the OECD

LOFT Program Review Group (18) (see Figure 32), the TRAC-PFl/MODl

calculation predicted that the ILCL started to empty several

hundred seconds later than in the experiment. The predicted rate

at which the ILCL emptied was more rapid than that observed

experimentally.

AEEW - R 2254 8
8.4.3 Primary System Mass Inventory

TRAC-PFI/MODl predictions of the primary system mass inventory*

are compared with those of the OECD LOFT Program Review Group and

with the measured data in Figure 33. The predicted rate of

primary system mass depletion, prior to the time of experimental

break uncovery (- 700 seconds), was lower than that measured.

Although not accurately known from the experimental data, minimum

primary system mass inventory was estimated to have occurred at

between 1800 and 2200 seconds (9). The time of minimum primary

systemmass inventory (defined as the time at which the HPIS plus

pump injection mass flow rates exceeded the break mass flow rate)

was predicted by TRAC-PFl/MOD1 to have occurred earlier, at - 1700 seconds. The calculated minimum primary system mass

inventory appeared to be in reasonable.agreement with the

measured data. The core void fractions, calculated by

TRAC-PFl/MODI, and shown in Figure 34, indicate that, as observed

in the experiment, no core uncovery was predicted.

8.4.4 System Pressure

A comparison of the measured and predicted primary system

pressure histories is shown in Figure 35. As indicated by the

correctly predicted timings of the primary coolant pump trip and

the HPIS initiation, the initial rapid subcooled depressurisation

was well represented. Following the end of subcooled blowdown,

and prior to the measured time of break uncovery, the slow rate

of depressurisation was reasonably well reproduced. The increase

in the rate of depressurisation, due to uncovery of the break,

was predicted to occur Later (at ~ 1200 seconds) than in the

experiment.. This led to a slight over-prediction of primary side

pressure, for a while, following the measured time of break

uncovery. As shown in Figure 36, the calculations performed by

the participants of the OECD LOFT Program Review Group also

predicted the rate of subcooled depressurisation well. Following

the end of subcooled blowdown, and prior to the time of measured

break uncovery, all calculations predicted a lower than measured

pressure plateau. .Consistent with their late predictions of the

time of break uncovery, all- calculations, with the exception of

GRS, over-predicted the timing of the increased rate of

depressurisation.

The overall secondary side presqure history is reasonably wellpredicted as shown in Figure 37 and compares favourably with the

calculations performed by participants of the OECD LOFT Program

Review Group (see Figure 38). Following "closure" of the MSCV,

* In calculating the predicted primary system mass inventory, all

flows into/out of the primary circuit during the transient-were

added to/subtracted from the initial primary system mass. The

initial primary system mass included all system components

except the secondary side of the steam generator component

(component numbers 20, 21, 22 and 27) and the steam line valve

(component number 23).

AEEW - R 2254 9
its minimum flow area was restricted to 0.35% of its fully-opened

value to account for the steam leakage which occurred during the

experiment.

8.5 Detailed Discussion

8.5.1 Break Mass Flow Rate and Break Line Density

The ILHL void fraction calculated by TRAC-PFl/MODl is shown in

Figure 39;, the time at which stratification in the hot leg is

predicted to occur is also indicated. Figure 39 shows that when

single phase conditions.were predicted in the hot leg, ie

subcooled liquid (4 60 seconds) or high quality steam

(> 1400 seconds), the mass flow rate in the break line was well

predicted. However, during stratified flow conditions,

TRAC-PFl/MODI takes no account of the hot leg liquid level in

determining the density of the fluid in the break line. The

density assumed, at the entrance to a branchline, is the volume

weighted density of the steam and liquid phases in the main pipe.

The break line density is therefore under-predicted when the

stratified level is above the break and over-predicted when the

level is below the break.

In order to gain an indication of the effects of TRAC-PF1/MODl's

inability to account for hot leg flow stratification on break

line density, relative to any deficiency in its critical flow

model, a TRAC-PFl/MODI calculation was performed on a "stand-alone" model of the LP-SB-l break line. The calculation,

in which the density of the fluid being fed to the break line was

altered, indicated that, in order for the TRAC critical flow

model to predict the measured break mass flow rate, the

calculated break line density would need to be similar to that

measured (see Table 11) ie at - 500 seconds, the reported break

flow was calculated by TRAC-PFl/MODI using a break line density

- 4% greater than that measured experimentally. The break line

density and break mass flow rate predicted by the Base Case

Calculation, at 500 seconds, were- 20% less than those measured.

The tentative conclusion from this calculation was that the

poorly predicted break mass flow rate, observed in the Base Case

Calculation, was mainly due to TRAC-PFl/MODI's inability to

account for the effects of hot leg flow stratification on break

line density, rather than deficiencies in its critical flow

model. (The RELAP5/MOD2 analysis of LP-SB-l (14), in which the

difficulty in accounting for the effects of hot leg flow

stratification had been overcome, found the critical flow model

to be inadequate for low quality conditions).

In order to comprehensively assess the effects of TRAC-PFl/MODl's

inability to account for hot leg flow stratification in

determining fluid conditions in the break line, correlations for

predicting branchline flow quality as a function of mainline

stratified liquid level were implemented in the LP-SB-I input

deck and the Base Case Calculation was re-run as described in

Sections 9 and 10.

AEEW - R 2254 10
8.5.2 Running Speed

The dominant component responsible for limiting the time step

size throughout the Base Case Calculation was the very small cell

(7.238 x 10-6 M3 ) used to represent the leakage path between the

reactor vessel and the core barrel nozzles (Bypass Pach 4 -

Component Number 83). This component was responsible for the

code "crashing" at - 1500 seconds (see Section 8.2). On

increasing the volume of this component, and decreasing that of

its neighbouring PLENUM component, by 3.5 x 10-3 M 3,

significant improvements in the running speed of the code were

realised and it was no longer necessary to reduce the maximum

allowable time step in order for the code to continue running.

Increasing the volume of Bypass Path 4 reduced the CPU to problem

time ratio from 2.15 to 1.32 between 500 and 1500 seconds (see

Figure 40).

8.6 Summary

1) The following points summarise the general observations made

on the TRAC-PFl/MODl Base Case Calculation of LP-SB-I. The

findings tended to be consistent with those of the OECD LOFT

Program Review Group (18).

i) Prior to the time of measured break uncovery, the break

line density, break mass flow rate and rate of primary

side mass depletion were under-predicted.

ii) The time at which the break uncovered was

over-predicted.

iii) In contrast to the experimental behaviour, the ILHL was

predicted to empty.

iv) The time at which the ILCL started to empty was

predicted to occur later than in the experiment. The

predicted rate at which the cold leg emptied was more

rapid than that measured.

v) The minimum primary. systemimass inventory was in

reasonable agreement with the measured data - as

observed in the experiment, no core uncovery was

predicted.

vi) The overall trends in primary and secondary system

pressure histories were reasonably well reproduced.

2) The tentative conclusion from a TRAC-PFl/MODl calculation

using a "stand-alone" model of the LP-SB-l break line was

that the poorly predicted break mass flow rate was mainly a

consequence of TRAC-PFl/MODl's inability to account for the

effects of hot leg flow stratification on break line density

(rather than deficiencies in its critical flow model).

3) The dominant factor responsible for limiting the time step

size throughout the Base Case Calculation was the very small

volume (7.238 x 10-6 m3 ) of the cell representing the

AEEW - R 2254 11
leakage path between the reactor vessel and the core barrel

nozzles. Increasing the size of this cell by 3.5 x 10-3

_m3 led to an improvement in the CPU to problem time ratio of

- 40% over a 1000 second period.

9. BASE CASE CALC"JLITION WITH EPRI CORRELATION

9.1 Introduction

In order to take account of the hot leg stratified liquid level

in determining the fluid conditions in the break line, EPRI

(Electric Power Research Institute, USA) correlations (19) for

predicting the liquid levels at which the onset of vapour pull through and liquid entrainment occur and the resulting branchline

flow quality were implemented in the LP-SB-1 input deck. The

correlations are described in Appendix II. The transient was

re-run using Version B02C of the TRAC-PFl/MODl code which employs

a branch offtake quality model for stratified flow.

9.2 General Observations

9.2.1 Break Mass Flow Rate and Break Line Density

As shown in Figure 41, implementation of the EPRI correlation

greatly improved the predicted break line density. for the initial

500 seconds of the transient. Although this led to an improved

break mass flow rate over this period, the break flow still

tended to be under-predicted, as shown in Figure 42. Break

uncovery occurred earlier than in the Base Case Calculation but

still significantly (- 300 seconds) later than in the

experiment. The discrepancies between the measured and predicted

break line densities prior to - 150 seconds are a consequence of

the offtake model only operating when fluid conditions in the

main pipe are fully stratified. The divergence of the measured

and predicted break line density and break mass flow rate at

500 seconds is-discussed in Section 9.3.

9.2.2 Primary System Densities

As shown in Figure.43, implementation of the EPRI correlation

greatly improved the predicted ILHL density behaviour. Contrary

to the Base Case Calculation, and in line with the experimental

measurements, no emptying of the ILHL was predicted by the

revised calculation.

Figure 44 indicates that although some improvement to the time at

which the ILCL started to empty was achieved by the revised

calculation, the rate at which the cold leg emptied was still

more-rapid than that observed experimentally.

9.2.3 Primary System Mass Inventory

As shown .in Figure 45, the improvement (over the first

500 seconds) in the predicted break mass flow rate (on

implementing the EPRI correlation) caused the primary system mass

to deplete slightly more rapidly than in the Base Case

AEEW - R 2254 12
Calculation. However, because the ILHL did not empty in the

revised calculation, the minimum primary system mass inventory

remained - 200 kg higher'than that predicted by the Base Case

Calculation.

9.2.4 System Pressure

The differences in the break mass flow rate predicted by the Base

Case and the EPRI Correlation Calculations are manifested in

slightly different primary and secondary pressure history trends,

as shown in Figures 46 and 47, respectively.

9.3 Detailed Discussion

9.3.1 ILHL Liquid Level Behaviour

The ILHL liquid level and the level at which the EPRI correlation

predicts the onset of vapour pull through to occur (h ) are shown

in Figure 48. When the ILHL liquid level falls below h (ie

after - 500 seconds), vapour pull-through is invoked in the EPRI

correlation and the quality in the break line increases. As was

seen in Figures 41 and .42, there follows a consequent decline in

break line density and break mass flow rate which clearly

represents a departure from the experimental behaviour.

The predicted ILHL liquid level is compared with that measured

during the experiment in Figure 49. Although well predicted

during the initial part of the transient, the calculated level

fell by - 2 cm at - 500 seconds. The experimental data show the

liquid level to have remained constant (at - 7 cm above the

centre line) from - 200 seconds until the decline to break

uncovery started at - 700 seconds. This difference in ILHL

liquid level behaviour is significant because the EPRI

correlation predicts vapour pull-through to commence when the

liquid level is - 6 cm above the centre line. The difference,

between the calculation and the experiment, in the behaviour of

the ILHL liquid levels is thus responsible for the observed

discrepancies in break line density and for the predicted decline

in break mass flow rate at - 500 seconds.

9.3.2 Time of Break Uncovery

The likely effect, on the break line density (and the time at

which the break uncovers), of the ILHL liquid level (or the level

at which vapour pull through is predicted to occur) is shown in

Figure 50a. The effect of the amount of system mass required to

be discharged, prior to break uncovery, is shown in Figure 50b.

For the current calculation, it seems likely that, had the liquid

level been correctly predicted (or the level at which vapour

pull-through is assumed to occur had been lower), the time at

which the break uncovered would still have been:over-predicted by ~ 200 seconds - see Figure 51 which shows the TRAC-PF1/MODl

predicted break line density when the EPRI correlation is

modified to reduce h .

AEEW - R 2254 13
Three factors have been identified which would contribute to the

time of break uncovery being over-predicted:

i) prior to the time at which vapour pull through is

predicted to occur, the break mass flow rate is

under-predicted (by the TRAC-PFl/MODl critical flow

model) by - 0.5 kgs- 1 . After 700 seconds (the time at

which the break started to uncover in the experiment),

this would result in a system mass excess in the

calculation of - 350 kg and could account for the time

of break uncovery being over-predicted by

120 seconds;

ii) the volume of the Reflood Assist.Bypass Line, implied

by the TRAC-PFl/MODl input deck, was found, after

reference to the LOFT specification document (2), to be

too large by - 0.21 m3 . This could account for the

time of break uncovery being over-predicted by

60 seconds;

iii) the TRAC-PFI/MODI interphase drag model causes the

density in the core to be underestimated by - 10% (20).

This could cause the time of break uncovery to be over-predicted by - 20 seconds (see Appendix III).

9.4 Summary

Accounting for the hot leg stratified liquid level in determining

the fluid conditions in the break line, by implementing the EPRI

correlation for branchline flow quality, resulted in the

following:

i) the predicted break line density was greatly improved

for the first 500 seconds of the transient;

ii) although the predicted break mass flow rate was

improved during the first 500 seconds of the transient,

it still tended to be under-predicted;

iii) although break uncovery was predicted to occur earlier

than for the Base Case Calculation, it was still

significantly (- 300 seconds) later than in the

experiment;

iv) a difference, between the calculation and the

experiment, in the behaviour of the ILHL liquid levels

was responsible for discrepancies in the break line

density and for a decline in the predicted break mass

flow rate at - 500 seconds;

v) had the ILHL liquid level been correctly predicted and

no such decline in break mass flow rate occurred at

500 seconds, the time at which the break uncovered

would still have been over-predicted by - 200 seconds.

The earlier under-prediction of break mass flow rate,

the volume of the Reflood Assist Bypass Line being too

AEEW - R 2254 14
large and the core density being under-predicted by the

TRAC-PFl/MOD1 interphase drag model were identified as

factors contributing to the, late prediction of break

uncovery;

vi) in line with the experimental measurements, and

contrary to the Base Case Calculation, no emptying of

the ILHL was predicted to occur;

vii) although there was some improvement to the time at

which the ILCL started to empty, the rate at which it

emptied was still more rapid than that observed

experimentally.

10. BASE CASE CALCULATION WITH EPRI CORRELATION AND INPUT

MODEL MODIFICATIONS

10.1 Introduction

In Section 9, the difference in behaviour of the ILHL liquid

level, between the experiment and the TRAC-PFI/MODI calculation

employing the EPRI correlation, was identified as being

responsible for the observed discrepancies in the break line

density and for the decline in break mass flow rate at - 500

seconds. Had the liquid level been correctly predicted, however,

it was thought that the time at which the break uncovered would

still have been over-predicted by some 200 seconds. One of the

factors identified as contributing to this was the volume of the

Reflood Assist Bypass Line specified in the TRAC-PFl/MODl input

deck. This was found to be - 0.2 m3 too large and thought to

account for the time of break uncovery being over-predicted by

60 seconds.

Prior to investigating the behaviour of the ILHL liquid level in

more detail, it was decided to re-run the calculation with the

volume of the Reflood Assist Bypass Line corrected. Two minor

errors, identified in the Base Case input model, were also

corrected at this stage, ie:

i) correction of GRAV terms in Lower Plenum;

ii) correction of effective cell lengths (and associated

FRICs) for PLENUM components.

Also, the volume of Bypass Path 4 - responsible for limiting the

time step size in the Base Case Calculation - was increased (and

the volume of its neighbouring PLENUM component decreased). All

modifications made to the Base Case input deck are detailed in

Appendix IV.

In this Section, the calculation employing the modified input

deck is used to investigate the behaviour of the ILHL stratified

liquid level. The influence, on the ILHL level, of the choice of

correlation used for determining the level at which vapour pull

through occurs is discussed and the effect of varying the choked

flow multiplier at the end of the break line is also examined.

AEEW - R 2254 15
A series of pictures showing, at 200 second intervals, the

predicted void fraction distribution, the liquid and Vapour

velocities and the occurrences of stratified flow conditions

throughout the system is shown in Appendix V.

10.2 General Observations

10.2.1 Break Mass Flow Rate and Break Line Density

As expected, and as shown in Figure 52, decreasing the volume of

the Reflood Assist Bypass Line reduced the time taken for the

break to uncover. A similar "shift" - see Figure 53 - is

observed in the predicted break line density.

10.2.2 Primary System Densities

The ILHL and ILCL densities, predicted using the modified input

deck, follow the same trend as in the previous calculation. As

expected, however, the reduction in the size of the Reflood

Assist Bypass Line causes events to occur slightly earlier - see

Figures 54 and 55.

10.2.3 Primary System Mass Inventory

The primary system mass inventory, predicted using the modified

input deck, follows a similar trend to the previous calculation.

The effect of reducing the size of the Reflood Assist Bypass Line

is observed as a constant deviation prior to the time of break

uncovery. The system empties to the same level in both

calculations, however, and the size of the Reflood Assist Bypass

Line does not therefore influence the minimum primary system mass

inventory - see Figure 56.

10.2.4 System Pressure

As shown in Figures 57 and 58, altering the size of the Reflood

Assist Bypass Line has very little effect on the primary and

secondary system pressure histories.

10.3 Detailed Discussion

10.3.1 ILHL Liquid Level Behaviour

Figure 59 indicates that the predicted decrease in ILHL liquid

level (which now occurs just after 400 seconds) coincides with

the time at which natural circulation in the ILHL is calculated

to have ceased. In the experiment, natural circulation ended at

- 500 seconds (9); however, the experimental hot leg liquid level

(calculated from the density measured by the middle beam of the

hot leg densitometer) showed no decline at this time (see

Figure 49). In line with the TRAC-PFl/MODl calculations, the

predictions of the OECD LOFT Program Review Group (18) show the

ILHL density and liquid level to be lower than that measured

between - 400 and - 700 seconds (see Figures 30 and 60). -Figure

61 indicates that between 550 seconds and 950 seconds, liquid is

predicted to flow back from the steam generator towards the

AEEW - R 2254 16
break. Experimental measurements suggest (9) that there was some

flow from the steam generator to the break after . 400 secondsand this is considered responsible (9) for maintaining the hot

leg level at - 7 cm above the centre line until - 700 seconds.

The delay in the time at which liquid starts to drain back from

the steam generator following the end of natural circulation, in

the TRAC-PFI/MODI calculation, appears to be linked to the

difference in behaviour of the ILHL liquid levels.

A mass flow balance on the ILHL, taken from the TRAC-PFI/MODI

calculation is shown in Figure 62. It is clearly seen that the

declines in the predicted ILHL liquid level (at - 400 seconds and

at - 800 seconds - see Figure 59) correspond to the periods

during which the break flow exceeded the sum of~the flow from the

vessel to the break and the flow from the steam generator to the

break.

Factors governing the flow between the steam generator and the

break have not been fully investigated. The magnitude and

direction of this flow, however, are clearly significant in

determining the ILHL liquid level behaviour (and hence the break

flow).

10.3.2 Choice of Correlation

Although the experiments performed in the Two-Phase Flow Loop,

(TPFL) at INEL confirmed the forms of previously proposed

correlations for predicting the mainline liquid levels at which

vapour pull-through and liquid entrainment occur, different

values for the constants used in the correlations were

recommended (see Table 12). The TPFL experiments found the

liquid level range, over which vapour pull through and liquid

entrainment occurred, to be greater than previously reported..

It is found that implementation of the CATHARE correlation

(C = 0.62), in the TRAC-PFl/MODl *input deck, improves the

agreement between the calculated and measured results, with

respect to break line density and break mass flow rate, as shown

in Figures 63 and 64. (Further improvement is realised.when a.

correlation with no vapour pull-through (h .= 0.142) is used).

The CATHARE correlation predicts the onset of vapour pull through

to occur when the ILHL liquid level is - 1 cm below that..assumed

by the EPRI correlation (see Figure 65). For experiment LP-SB-I,

the discrepancies between the measured and predicted ILHL liquid

levels become less significant, therefore, when the CATHARE, as

opposed to the EPRI, correlation (for the level at which vapour

pull-through commences) is used. Had the ILHL level been

correctly predicted, however, it appears that the EPRI

correlation would have performed satisfactorily.

The correlation for the level at which vapour pull-through occurs

- implemented in a modified version of RELAP5/MOD2 (24) - is of

the same form as the EPRI correlation, but a value of 0.75 is

chosen for the coefficient C . Use of this coefficient would be

expected to produce vapour pull-through at a main pipe liquid

level between those predicted by the EPRI and CATHARE

correlations.

AEEW - R 2254 17
The EPRI correlation for the branchline quality is found (19) to

over-predict the branchline void fraction for main-pipe liquid

levels just below the level at which vapour pull-through

commences. This is considered (19) to be due partly to the fact

that the correlation does not tend to zero as the vapour pullthrough level is approached. The correlation for the branchline

quality implemented in the modified version of RELAP5/MOD2 (24)

is different from that proposed by EPRI. It has the advantage of

tending to zero as the vapour pull-through level is approached

and is found (19) to provide a better fit to experimental

branchline void fraction data immediately after the onset of

vapour pull-through.

The likely effect, therefore, on the current calculation, of

implementing the correlations used in the modified version of

RELAP5/MOD2, rather than the EPRI correlations, would be to delay

the time at which vapour pull-through occurred and, for ILHL

liquid levels just below the vapour pull-through level, to lessen

the reduction in the break line density, ie to provide a slightly

better fit to the experimental data.

10.3.3 Choice of Choked Flow Multiplier

Since it could be argued that, had the break mass flow rate been

correctly predicted during the initial part of the transient, the

observed discrepancy in ILHL liquid level behaviour may not have

arisen, calculations were performed in which it was attempted to

reproduce the experimental break mass flow rate more accurately.

The break mass flow rates predicted by TRAC-PFl/MOD1 using

2-phase choked flow multipliers of 1.0 (Base Case with EPRI

Correlation and Input Modifications) and 1.2 are shown in Figure

66. Although the initial experimental break mass flow rate is

well reproduced using a choked flow multiplier of 1.2, the drop

in predicted break flow at - 400 seconds (characteristic of

previous calculations) is still noticeable, confirming that the

ILHL liquid level behaviour is not correctly reproduced despite

the initial break mass flow rate being well-represented.

10.4 Summary

1) As expected, reducing the volume of the- Reflood Assist

Bypass Line improved the time at which the break uncovered.

2) Natural circulation in the ILHL and the time at which liquid

starts to drain back from the steam generator are important

factors in determining the behaviour of the ILHL liquid

level. The predicted delay between natural circulation

ceasing and liquid starting to drain back from the steam

generator (not observed in the experiment) is thought to

account for the difference in behaviour, between the

experiment and the calculation, of the ILHL liquid levels

and hence the discrepancies in break mass flow rate.

AEEW - R 2254 is
3) The discrepancies between the measured and predicted ILHL

liquid levels become less significant when the CATHARE, as

opposed to the EPRI, correlation for the level at which

vapour pull-through commences is used. Use of the CATHARE

correlation therefore improves the agreement between the

measured and predicted break mass flow rates. Had the ILHL

level been correctly predicted, however, it is thought that

the EPRI correlation would have performed satisfactorily.

4) Use of a 2-phase choked flow multiplier (at the end of the

break line) of 1.2 indicated that the ILHL liquid level

behaviour is not correctly predicted despite the initial

break mass flow rate being well represented.

11. CONCLUSIONS

The main findings of the TRAC-PFl/MODl analysis of the OECD LOFT

experiment LP-SB-l were as follows:

1) Very small cells were identified as having a detrimental effect on the running speed of the calculation. It was

demonstrated that increasing the volume of a particular cell

reduced the CPU to problem time ratio by - 40%.

2) The results of the TRAC-PFl/MODl Base Case Calculation of

LP-SB-I tended to be consistent with those reported by

members of the OECD LOFT Program Review Group. The overall

trends with respect to pressure histories and minimum system

mass inventory were reasonably well represented by the

TRAC-PFl/MODl calculation. The inability of TRAC-PFl/MODl

to account for the main branch stratified ,liquid level in

determining fluid conditions in a side branch led to

discrepancies between the measured and predicted break line

and ILHL densities and was found to be the main reason for

the observed differences in break mass flow rate.

3) Implementation of an EPRI correlation, for determining side

branch quality as a function of main-pipe stratified liquid

level, was 'effective in improving the predicted break line

density early in the transient. The break mass flow rate

was also improved, although it remained slightly lower than

that measured.

4) A difference, between the TRAC-PFl/MODl calculation and the

experiment, in the time at which liquid started to drain

back from the steam generator was identified as a possible

cause for small discrepancies in the ILHL liquid level

behaviour. The effect of these discrepancies, on-the

break line density and break mass flow rate, became less

significant when the CATHARE correlation (for predicting the

level at which vapour pull through occurs) was implemented,

rather than the EPRI correlation.

AEEW - R 2254 19
12. RECOMMENDATIONS

It is suggested that:

1) A means of accounting for the effect of main pipe stratified

liquid level in determining fluid conditions in a side

branch should be implemented in TRAC-PFl/MODI.

2) A closer examination of the factors influencing the draining

of the steam generator tubes is required to resolve the

observed discrepancies in intact loop hot leg liquid level

behaviour.

13. ACKNOWLEDGEMENTS

The guidance of Dr C G Richards (AEEW) and the support of

Sr F Pelayo (CSN, Spain), during the course of this work, are

gratefully acknowledged.

14. REFERENCES

1. PETERS M D, Quick Look Report On OECD LOFT Experiment

LP-SB-I, July 1983 (OECD LOFT-T-3204)

2. REEDER D L, LOFT System and Test Description, July 1978

(TREE-1208, NUREG/CR-0247)

3. KEE E J, SCHALLY P J, WINTERS L, Base Input for LOFT RELAP5

Calculations, July 1980 (EGG-LOFT-5199)

4. GRUSH W H, TANAKA M, MARSILI P, Best-Estimate Prediction for

the OECD LOFT Project Small Cold Leg Break Experiment

LP-SB-3, November 1983 (OECD LOFT-T-3005)

5. ALLEN E J, Internal Document

6. ALLEN E J, Internal Document

7. NEILL A P, Internal Document

8. PELAYO F, TRAC-PFl-MODI Post-Test Calculations of the OECD

LOFT Experiment LP-SB-2, March 1987, (AEEW - R 2202)

9. MODRO S M, FELDMAN E M, NEWMAN N and PETERS M D, Experiment

Analysis and Summary Report on OECD LOFT Nuclear Experiments

LP-SB-I and LP-SB-2, May 1984 (OECD LOFT-T-3205)

10. BIRCHLEY J, Private Communication

11. JOUSE W C, Internal LOFT Reactor Vessel Core Bypass Flows,

November 1981 (LO-17-81-044)

12. ALEMBERTI A, PROTO G, Experiment Analysis and Summary Report

on OECD LOFT Nuclear Experiment LP-SB-3 (Draft),

December 1985

AEEW - R 2254 20
13. BIRCHLEY .J, Private Communication

14. HALL P C, BROWN G, RELAP5/MOD2, Calculations of OECD LOFT

Test LP-SB-1, July 1986 (GD/PE-N/554 (REV)

15. CHEN T H, MODRO S M, Best Estimate Prediction for OECD LOFT

Experiment LP-SB-1, May 1983 (OECD LOFT-T-3203)

16. McPHERSON G D, Decay Heat Data for OECD LOFT Experiments,

October 1985 (Letter to OECD LOFT Programme Review Group

Members)

17. MODRO S M, CHEN T H, OECD LOFT Project Experiment

Specification Document, Hot Leg Small Break Experiments

LP-SB-1 and LP-SB-2, April 1983 (OECD LOFT-T-3201)

18. ANODA Y, OECD LOFT Experiment LP-SB-1 and LP-SB-2

Calculation Comparison Report, July 1986 (Draft)

19. ANDERSON J L, BENEDETTI R L, Critical Flow Through Small

Pipe Breaks, May 1986 (EPRI NP-4532)

20. ARDRON K, CLARE A, Interphase Drag Calculations in the

RELAP5/MOD2 and TRAC-PF1/MOD1 Thermal Hydraulic Codes

(GD/PE-N/557 (REVISED)

21. MEMPONTEIL A, Super Moby Dick - First Results of Phase

Separation in Tees, Seminaire CATHARE, 5-6 Novembre 1985,

Grenoble

22. Safety Code Development Group, Energy Division, LANL,

TRAC-PFI/MODl: An Advanced Best-Estimate Computer Program

For Pressurised Water Reactor Thermal-Hydraulic Analysis,

July 1986, (NUREG/CR-3858, LA-10157-MS)

23. BOYACK B E, STUMPF H, LIME J F, TRAC User's Guide

(NUREG/CR-4442, LA-10590-M), November 1985

24. ARDRON K H, BRYCE W:M,-Internal Document

AEEW - R 2254 21
TABLE 1

INITIAL CONDITIONS FOR EXPERIMENT LP-SB-1

MEASURED (2) TRAC PREDICTED

PRIMARY COOLANT SYSTEM

Core AT (K)

Hot Leg Pressure (MPa)

Cold, Leg Temperature (K)

Mass Flow Rate (kgs-1)

REACTOR VESSEL

Power Level (MW)

STEAM GENERATOR SECONDARY SIDE

Liquid Level (m)

Water Temperature (K)

Pressure (MPa)

Mass Flow Rate (kgs- 1 )

PRESSURIZER

Liquid Volume (W3 )

Steam Volume (m 3 )

Water Temperature (K)

Pressure (MPa)

Liquid Level (m)

BROKEN LOOP

Cold Leg Temperature (K)

18.5

15.00

557.2

483.1

1.7

0.08

1.5

3.2

48.8 + 1.2

3.12

535.2

5.53

25.79

0.625

0.377

615.8

15.06

1.072

+,0.05

+ 3.6

+ 0.05

T 0.77

+ 6.001

+ 0.001

T 8.2

+ 0.11

T 0.002

18.9

15.0

557.7

483.1

48.8

3.12

536.6

5.57

25.57

0.624 0.308a

615.2

14.98

1.384a

557.7 555.7 + 6.3

a These figures are not within the uncertainties of the experimental

data possibly due to a difference, between the experiment and the

TRAC input deck, in the interpretation of the pressurizer geometry.

AEEW -. R.2254.
TABLE 2

PRESSURIZER VOLUME

REFERENCE PRESSURIZER VOLUME (m3)

LP-SB-1 EASR (9) 1.002

TRAC-PF1/MOD1 INPUT DECK 0.932

TABLE XXI 0.96

LOFT SPECIFICATION (2)

TABLE A-2 0.931

AEEW - R 2254
TABLE 3

STEADY STATE PRESSURE DROPS

AVAILABLE

COMPONENT/ TRAC-PFl/MODl CALCULATED TRAC PRESSURE

CELL PRESSURE DROP. -hpg DROP DATA

SCALED TO

PART OF REACTOR SYSTEM MASS FLOW

RATE OF

FROM TO (MPa) (kPa) 483.1 kgs- 1

(kPa)

Cold Leg 6/2p 7/9 15.1742-15.1548 19.4 19.9(10)

Inlet Nozzle 7/9 86/1 15.1548-15.1191 35.7 42.1 48.4(1)

Filler Inlet Port

Exit to Anulus

Circum. Flow to Downcomer

Downcomer 86/1 86/8 15.1191-15.1355 -16.4 6.9 7.8(11)

Turn and Mix 86/8 89/1 15.1355-15.1059 29.6 29.6 30.8(11)

Lower Core Support

Lower End Boxes 89/1 87/1 15.1059-15.0637 42.2 26.7 27.9(11)

Fuel Pins

Upper End Boxes

Upper Stack Region 87/1 1/4 15.0637-14.9922 71.5 58.8 53.7(10,11)

Core Barrel and Reactor Vessel Nozzles

Hot Leg

Steam Generator 1/4 3/is 14.9922-14.7650 227.0 224.7(10)
NJi TABLE 4

STEADY STATE ENVIRONMENTAL HEAT LOSSES

ENVIRONMENTAL HEAT LOSSES (kW)

.9 - -

ASSUMED FOR.

RELAP5 ANALYSES

OF LOFT

EXPERIMENTS

(12)

ESTIMATED FROM

LOFT

EXPERIMENTS

L9-1 AND L3-3

MEASURED WITH

ALL FLUID IN

PRIMARY COOLANT

SYSTEM AT A

TEMPERATURE OF

555K

TRAC-PF1/MOD1

PREDICTIONS

(13) (13)

Primary Coolant 143

System

170

Reactor Pressure 89

Vessel

Pressurizer 6 6

Steam Generator 20 19

Secondary Side

TOTAL 196 200 + 100 248 257.
TABLE 5

STEADY STATE CORE BYPASS FLOW RATES

TRAC-PFl/MODl Predictions Available

Component/Cell No Data(ll)

Mass Flow Rate % Primary Mass % Primary Mass

(kgs-I) Flow Rate Flow Rate

Paths 1, 2 and 3

Lower Core Support 89/is 16.93 3.5 3.5

Structure, Lower End

Box and Gauge Hole

Bypasses

Path 4

1.28 - 3.54

Outlet Nozzle Gap 83/1 13.23 2.7

Path 5

Core Barrel 79/1 0.20 0.04 0.04

Alignment Key

Reflood Assist 31/3s 25.66 5.3 5.25

Bypass Valve

(RABV)
TABLE 6

TRANSIENT CALCULATION - DECAY HEAT POWER TABLE

Power at Time t

Time After Scram (15) Power

(t, seconds) Initial Power (at time t)

(48.8 r, MW)

(r)

0.0 1.0 48.8000

0.15 0.88 42.9440

0.3 0.76 37.0880

0.6 0.58 28.3040

0.85 0.176 8.5888

1.0 0.122605 5.9831

1.3 0.1 4.8800

2.0 0.087420 .4.2661

4.0 0.075788 3.6985

7.0 0.064 3.1232

10.0 0.060012 2.9286

25.0 0.046738 2.2808

65.0 0.035 1.7080

100.0 0.031546 1.5394

250.0 0.025210 1.2302

a Neutron and Fission Power

b Fission Power

AEEW - R 2254
TABLE 7

TRANSIENT CALCULATION - STEAM GENERATOR

SECONDARY SIDE AUXILIARY FEEDWATER FLOW RATE TABLE

Transient Time

(Seconds) Liquid Velocity (ATime After Reactor Scram (ms- 1 )

(1.4 Seconds) + 50 seconds)*

0.0 0.0

111.9 0.0

112.0 0.061673

1913.4 0.061673

1913.5 0.0

4000.0 0.0

* Calculation was run in transient mode steady state

for 50 seconds.

AEEW - R 2254
TABLE 8

TRANSIENT CALCULATION - HIGH PRESSURE INJECTION

SYSTEM _•PIS) FLOW RATE TABLE

PCS Pressure HPIS Flow Rate (17) HPIS Flow Rate for

(MPa) (Is-.) TRAC-PFI/MOD1

Calculation*

(ms-1)

8.70 0.3155 0.0527

8.36 0.3155 0.0527

7.67 0.3918 0.0654

6.98 0.4883 0.0815

5.60 0.6031 0.1007

4.22 0.7022 0.1178

3.53 0.7583 0.1266

2.15 0.8505 0.1420

0.08 0.9564 0.1597

* flow area of HPIS piping = 5.9892 x 10-3 m2

AEEW - R 2254
7A 9

c6'ERAni~aL SE1iNTs POR EnyERIruxI-SB

Measured Diring Erperiment (9) Specified to TRAC-PFI/MOD1 Calculation

Action

Reference Setpoint Reference Setpoint

(Component/celU no)

Small-Break %Xlve Time 0 Seconds Time 0 Seconds

Opened

Reactor Scramnmed IBL Pressure 14.57±0.03 MPa ILIL (99/1) Pressure 14.57 MPa

Main Feedwater IUML Pressure 14.57±0.03 MPa IULL (99/1) Pressure

Shut Off

Main Steam Control "U•pon Verification of Reactor 2 seconds after 14.57 MPa

Valve Started to Scram" Reactor Scram + 2 seconds

Close

Primary 0lolant - 11.12 MPa ILCL (7/5) Pressure 11.12 MPa

PLprs Tripped

HPIS Flow ILHL Pressure 8.24±0.03 MPa ILHL (99/1) Pressure 8.24 MPa

Initiated

Auxiliary Feed- Time After 62±0.2 seconds Time After Reactor 62 seconds

water Initiated Reactor Scram

Scram

Auxiliary Feed- Time After 1801.4±0.8 Time After Initiation 1801.4 seconds

water Terminated Initiation seconds of Auxiliary Feedof Auxil- water

iary Feedwater

Experiment IULL Pressure 2.487_+0.001MPa IU!L (99/1) Pressure 2.487 MPa

Terminated

AEEW - R 2254
IP-SB-I BASE CASE TWSIET CKtrXATIct - CRFJLOGY OF EVERM

TIME AFTER EXPERIMENT

INITIATIO C(s)

EVENT

MEASURED (9) TRAC PREDICTED

Small-break valve opened 0.0 0.0

Reactor scrammed 1.4 ± 0.05 0.5

Main feedwater shut off 1.4 ± 0.2 0.5

Main steam control valve started to close 3.4 ± 0.2 2.5

Main feedwater isolated 3.8 ± 0.05 2.6

Main steam control valve fully closed 15.4 - 0.2 54 a

Primary coolant pumps tripped 24.6 ± 0.2 25.7

Pressurizer liquid level below indicating range 34.6 ± 0.4 5 3 b

HPIS flow initiated 41.4 ± 0.2 41.1

Primary coolant pump 1 coastdown completed 42.6 + 0.2 46.4

Primary coolant pump 2 coastdown completed 43.0 ± 0.2 46.4

Subcooled ,blowdown ended 57.5 ± 0.2 6 5c

Auxiliary feedwater initiated 63.4 _ 0.2 63.4

Break started to uncover 715 - 3 -'1200

Primary system pressure became less than 1077 ± 10 1380

secondary system pressure

Auxiliary feedwater shut off 1864.8 ± 0.8 1864.8

HPIS flow rate exceeded break flow rate 1998.0 - 200 1820.0

HPIS + pump injection flow rate exceeded break - 1680.0

flow rate

Experiment termination criterion reached 3668 ± 2.0 3227

a The opening of the steam flow bypass valve during

main steam control valve.

the transient was simulated using the

b Defined as the time when a = 1.0 in bottom cell of pressuriser (level 4 0.01 m after

- 40 seconds).

c Defined as the time when Tliq m Tsat in the break line.

AEEW - R 2254
TABLE 11

RESULTS FROM STAND-ALONE.BREAK LINE MODEL

LF-SB-1 'Stand-Alone' Base Case

Experiment Break Line Calculation

Model

Break Mass Flow 4.0 4.0 3.3

Rate (kgs- 1 )

Break Line Density 670 700 570

(kgm- 3 )

AEEW - R 2254
TABLE 12

COMPARISON OF COEFFICIENTS USED IN VAPOUR PULL

THROUGH AND LIQUID ENTRAINMENT CORRELATIONS

EPRI (19) SMOGLIE (19) CATHARE (21) ZUBER (19)

Ce 0.62 0.69 0.62 0.687

C 0.82 0.75 0.62 0.687

he 1 C m 2 0.2

D 2 D gPg (Pf - Pg)

h p mf2 0.2 -- +-- [ ]

D 2 D gpf (Pf - Pg)

where

he = liquid level at which liquid entrainment begins (m)

hp = liquid level at which vapour pull-through begins

(m)

D = mainline internal diameter (m)

Ce = coefficient. as above

C = coefficient, as above

P

mg = gas mass flow rate in branchline (kgs- 1 )

;f = liquid mass flow rate in branchline (kgs- 1 )

g = acceleration due to gravity (ms- 2 )

Pg = gas density (kgm- 3 )

Pf = liquid density (kgm- 3 )

AEEW - R 2254
V~i

4~

Intact loop Broken loop

01 Aý __I

FIGURE 1

LOFT SYSTEM CONFIGURATION FOR LP-SB-1
EXPERIMENT LP-SB-1

15

12.5

0

a1..

U,

U)

0

L.

aBreak initiated (0 s)

Reactor scram (1.4 s)

Primary coolani

Pressur

HPIS

I I a

CL Press. (PE-?C-0O6) Uncert = LO.O8 MPa

10

-2000

t pumps tripped (24.6 s)

izer emptied (34.6 s)

flow initiated (41.4 s)

-1500

CL

I

7.5 I-

-1000

Subcooled blowdown ended (57.5 s)

! ! i

5

15

LI

0 20 40 60 80 100

Time (s)

Primary system pressure (0 to 100 s).

EXPERIMENT LP-SB-1 J 'I I

CL

10

I CL Press.(PE-PC-006) I2

Uncert = = 0.088 MPo

1.'

Break uncovered (715 s)

Primary pressure less than .

secondary pressure (1077 ýS)

Experiment terminated (3668 s).

I I 0

000

00o

0~

00

)00 5

0

0 1000 2000

Time (s)

3000 4000

Primary system pressure (0 to 4000 s).

FIGURE 2

LP-SB-1 PRIMARY SYSTEM PRESSURE AND TIMINGS

OF SIGNIFICANT EVENTS

AEEW - R 2254
Component No Component Name

u~l

99

1

8

2

3

ILHL & Break Pipe

Intact Loop Hot Leg

Pressurizer

Steam Generator

Pump Suction

Component No

4

61

5

63

6

7

Component Name

Pump No 2

Pump No 2 Injection Tee

Pump No 1

Pump No 1 Injection Tee

Pump Discharge

Intact Loop Cold Leg

8 STEAM GENERATOR

2

HPIS

PUMP INJECTION

0 61

6 I.L.C.L.

1 PUMP SUCTII U63

PUMP INJECTION

3

BREAKLINE

FIGURE 3a

PRIMARY SYSTEM NODALISATION DIAGRAM FOR LP-SB-1

(A) INTACT LOOP
tzij Component-No .... Component Name

41 Broken Loop Cold Leg 31 Broken Loop Hot Leg

P. A. B. V

r

8

41

B. L C. L

31

B. L. H. L.

FIGURE 3b

PRIMARY SYSTEM NODALISATION DIAGRAM FORýLP-SB-l

(B) BROKEN LOOP
M~

K)

ul

79 Vessel Representation for "Svstem Mimic"

'C: Vessel Representation for "System Mimic" (See Appendix V)

81 80

80

I

81

I

80

82 IjiL

87

86 ý85

Component Nc

79

80

81

82

83

84

85

86

87

88

89

90

88

Component Name

Bypass Path 5

Upper Plenum

(D'comer)

Upper Plenum (core)

Plenum Component

(D'comer)

Bypass Path 4

Plenum Component

(core)

Downcomer Bypass

Downcomer

Pipe Above Core

Core

Lower Plenum and

Core Bypass

Lower Plenum

FIGURE 4

32

1

35

2

86

2

4

6

7

87

2 I

86

2

84I I84

5

4

3

3 3

5

6

°°

B°

2

1

85

2

4

5

6

8

4

6

89

I

I

i

901F ' 1: 2 :

90

4

LJ2

TRAC-PFI/MODI REACTOR VESSEL NODALISATION DIAGRAM FOR LP-SB-1
Ln

.P.

MAIN STEAM LINE 23

21

(7MAIN FEEIJVATER

27 Cells Numbered Component No

14-19 20

28,29,31 21

21-24,26 22

1-6 23

27

Component Name

22 ....AUX FEEEIVATER

Boiler

Steam Dome

Downcomer

Steam Line Valve

Main and Auxiliary

Feedwater Tee

FIGURE 5

TRAC-PF1/MOD1 STEAM GENERATOR SECONDARY SIDE NODALISATION DIAGRAM FOR LP-SB-1
TIE F•OLWVING ARE PLOTTED AGAINST REACTOR TIME

TOTAL CPU TIME

600

500

400

0

u

0i

V7

300

200

100

0 50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME • , SECONDS

FIGURE 6 - CPU USAGE

LP-SB-I STEADY STATE CALCULATION

THE FOLLOVING ARE PLOTTED AGAINST REACTOR TIME

TIME STEP SIZE

(n'

REACTOR TIME , SECONDS

FIGURE 7 - TIME STEP SIZE

LP-SB-1 STEADY STATE CALCULATION

AEEW - R 2254
THE FOLLDVING ARE PLOTTED AGAINST REACTOR TIME

FUNCTION

30

28

26

24

22

K

20

18

16

14

12

II] 1 I I I I I I I I I I

0 50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME , SECONDS

FIGURE 8 - TEMPERATURE DIFFERENCE ACROSS CORE

LP-SB-1 STEADY STATE CALCULATION

THE FOLLOVING ARE PLOTTED AGAINST REACTOR TIME

PRESSURE

0. 155E 08

O. 154E 08

0. 153E 08

0. 1SZE 06

2. 151E 08

0. 1SOE 08

0. 149E 08 .

0. 14SE 08 0 50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME , SECONDS

FIGURE 9 - PRIMARY SYSTEM PRESSURE

LP-S8-I STEADY STATE CALCULATION

AEEW - R 2254
THE FOLL17VING ARE PLOTTED AGAINST REACTC0A TIM~E

LIGUID TEMP'ERATURE

tv

a

REACTOR TIME , SECONDS

FIGURE 10 - ILHL TEMPERATURE

LP-S8-1 STEADY STATE CALCULATION

THE FCLLDVING ARE PLOTTED AGAINST REACTOR~ TIME

LIOUID TEMPERATURE

~OCr

566

564

562

560

di

0~

558 I

0 50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME , SECONDS

FIGURE 11 - ILCL TEMPERATURE

LP-SB-1 STEADY STATE CALCULATION

AEEW - R 2254
THE FOLLOWING ARE PLOTTED AGAINST REACTOR TIME

SIGNAL VAR. NO. a

0 50 100 ISO 200 250 300 350 400 450 500 550

REACTOR TIEE SECONDS

FIGURE 12 ILHL MASS FLOW RATE (KG/S7

LP-SB-1 STEADY STATE CALCULATION

THE FOLLOWING ARE PLOTTED AGAINST REACTOR TIME

SIGNAL VAR. NO. 7

3~Lr

348

346

34"

342

340

338

336

334

332

0 50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME , SECONDS

FIGURE 13 - PUMP SPEED IRAD/5)

LP-S8-1 STEADY STATE CALCULATION

AEEW - R 2254
TNE FOLLDVING ARE PLOTTED AGAINST REACTOR TIME

REACTOR POER

0.50E 08

0. 45E 08

0. 40E 08

O. 35E 08

0. 30E 08

0.25E 08

0. 20E 08

0. 15E 08

0. 1OE 08

O. 50E 07

O. OOE O0 I 1 I I I I I I I I i

0 50. 100 150 200- 250 300 350, 400 450 500 550 600

REACTOR TIMEE O SECONDS

FIGURE 14 - REACTOR POVER

LP-SB-1 STEADY STATE CALCULATION

THE FOLLDVING ARE PLOTTED AGAINST REACTOR TIME

SIGNAL VAR. NO). I

2.

2.

2.

2.

2.

1.

1.

I.

1.

0 50 100. 150 200 250 300 350 400 450 500 550

REACTOR TIME , SECONDS

FIGURE 15 - SGS LIQUID LEVEL (M)

LP-SB-1 STEADY STATE CALCULATION

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST RF-ACTO7R TIME

LIOUIO TEIMPERATURE

1ý

Ludi

Q

0 50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME , SECONDS

FIGURE 16 - SGS TEMPERATURE (BOTTOM DOWNCOMER)

LP-SB-1 STEADY STATE CALCULATION

THE FO.LOVING ARE PLOTTED AGAINST REACTOR TIME

SIGNAL VAR. NO. 6

7000000

6800000

6600000

6400000

6200000

6000000

5800000

5600000 -. , A Z, ,-4 r. _________

54000 1 0 50 100 150 200 250 300 350 400 450 500 55 600

REACTOR TIME , SECONDS

FIGURE 17 - SOS PRESSURE (PA)

LP-SB-I STEADY STATE CALCULATION

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST REACTOR TIME

SIGNAL VAR. NO. 14

50 100 150 200 250 300 350 400 450 500 550

REACTOR TIME SECONDS

FIGURE 18 - SGS STEAM MASS FLOW4 RATE (KG/SI

LP-SB-1 STEADY STATE CALCULATION

THE FCLLOVING ARE PLOTTED AGAINST REAC7OR TIME

MIASS FLOWJ PATE

0 50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME , SECONDS

FIGURE 19 - SGS FEEDVATER MASS FLOV RATE

LP-SB-l STEADY STATE CALCULATION

AEEW - R 2254
THE FCLLOVING ARE PL.OTTED AGAINST REACTOR TIME

LIQUID TEMP'ERATURE

616.4

616.2

616.0

615.8

615.6

615.4

615.2

615.0

614.8

614.6

`11ý

0 50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME * SECONDS

FIGURE 20 - PRESSURIZER LIQUID TEMPERATURE

LP-SB-1 STEADY STATE CALCULATION

THE FOLLOVING ARE PLOTTED AGAINST REACTOR TIME

PRESSURE

0.1514E 08

0.1512E 08

0.1510E 08

0.150&E 08

0.1506E 08

1504E 08

0.1502E 08-

0. 1500E 08

0.1498E 08

0.1496E 08

50 100 150 200 250 300 350 400 450 500 550 600

REACTOR TIME , SECONDS

FIGURE 21 - PRESSURIZER PRESSURE

LP-SB-I STEADY STATE CALCULATION

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST REACTOR TIME

VATER LEVEL

1.394

1.392

1.390

1.388 -

EnJ

I.

1.

I.

1.

1.

1.

386

384

382

380

378 -

376 -

0 50 100 150 200 250 300 350 400 450 500 550 6(O0

REACTOR TIME , SECONDS

FIGURE 22 - PRESSURIZER LIOUID LEVEL

LP-SB-I STEADY STATE CALCULATION

TFFFOLLDVING ARE PLOTTED AGAINST REACTOR TIME

LIQUID TEMPERATURE

l.,J

REACTOR TIME , SECONDS

FIGURE 23 - BLCL TEMPERATURE

LP-55-1 STEADY STATE CALCULATION

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST REACTOR TIME

TOTAL CPU TIME

i ZWULI

10000

8000

tU0 6000

4000

2000

0 500 1000 1500 2000

REACTOR TIME

2500 3000 3500 4000

SECONDS

FIGURE 24 - CPU USAGE

LP-SB-1 BASE CASE CALCULATION

"THE FOLODING ARE PLOTTED AGAINST REACTOR TIME

TIME STEP SIZE

u,

U)

REACTOR TIME , SECONDS

FIGURE 25 - TIME STEP SIZE

LP-SB-I BASE CASE CALCULATION

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST TIME

FR-PC-S03 *MASS FLOW RATE

u

Luj

000 1500 2000 2500 3000 3500

TIMT E , SECONDS

FIGURE 26 - BREAK MASS FLOW RATE

LP-S8-I BASE CASE CALCULATION

THE FO.LLVING ARE PLOTTED AGAINST TIMlE

OE-PC-SO'.B *MIXTURE DENSITY

1000

8O00

Boo

600

4

400

200

0

in

TIME , SECONDS

FIGURE 27 - BREAK UPSTREAM DENSITY

LP-SB-I BASE CASE CALCULATION

AEEW - R 2254
ol- 3:

0

M

.0

3:

0

(n

0

5

1003

Time (s)

FIGURE 28 - BREAK MASS FLOW RATE, COMPARISON WITH OECO GROUP

LP-SB-I BASE CASE CALCULATION

THE FOLLOVING ARE PLOTTED AGAINST TIME

DE-PC-002A * DE-PC-002B DE-PC-002C D MIXTURE DENSITY

BOTTOM BEAM

MIDDLE BEAM

TRAC BASE CASE

EXPT. TOP BEAM

TIME , SECONDS

FIGURE 29 - ILIIL DENSITY

LP-SB-I BASE CASE CALCULATION

AEEW - R 2254
loco aaIS

X N I RA-R/MCDiI

A j JAEhR5/m A z C r.

6Cor __ I NE: - (Z5/MCDi2

rco-

,cw~- P, -cCZI

LR~ +c 7 0%

-7

C,,

C

C-.

p.,

0*

0 500 1000 1500

Time (s)

FIGURE ZO - ILHL DENSITY, COIMPARISON VJITH OECO REVIEV GROUP

LP-SB-1 BASE CASE CALCULAT;CN

2000

ThE FOLLDVING ARE PLOTTEO AGAINST TIME

GE.-PC-OOIA ICE-PCC0IB ,DE-KC-0OIC

tIIXTUPE OENSITY

TPAC BASE CASE

r r

a.,

EXPT. BOTTOM zBE.M

-EXPT. MICCLE BEAM

-EXPT. TOP BEAM

TIIIE , SECONDS

FIGURE 31 - ILCL DENSITY

LP-SB-I BASE CASE CALCULATION

AEEW - R 2254
1000

Boo

a,,

600

400

C

a,

200

0

0 500" 1000 1500 2000

T ime (s)

FIGURE 32 - JLCL DENSITY, COMPARISON 1/1TH OECD REVIEW GROUP

LP-S8-1 BASE CASE CALCULATION

6000

V)

U,

4000

2000

'I,ý

0

0 500 1000 1500 2000

T i.me (s)

FIGURE 33 - PRIMARY MASS INVENTORY, COMPARISON WITH OECO GROUP

LP-SB-I BASE CASE CALCULATION

AEEW - R 2254
THE FOLLCVING ARE PLOTTED AGAINST REACTO7R TIME

VAPOR FRACTION

TRAC BASE CASE

CELL S (TOP)

CELL 3

CELL 2

I (BOTTOM)

0 500 1000 1500 2000 2500

REACTOR TIME , SE

FIGURE 34 - CORE VOID FRACTIONS

LP-SE-1 BASE CASE CALCULATION

THE FOLLOVING ARE PLOTTED AGAINST TIME

PE-PC-002 ,PRESSLFE

3000 3500 4000

CONDS

0. 20E 08

0. IBE 08

0. 16E 08

0.t14E 08

0. 12E 08

0. IOE 08 8

z

0. 8OE 07

0.6DE 07

0. 4OE 07

0. 20E 07

0. ODE 00

TIME

FIGURE 35 - PRIMARY SYSTEM PRESSURE

LP-S8-1 BASE CASE CALCULATION

AEEW - R 2254

, SECONDS
o 10'

l,

05

0~

C0

CD

0 500 1000 1500 2000

T i me (s)

FIGURE 36 - PRIMARY PRESSURE, COMPARISON

LP-SB-I BASE CASE CALCULATION

WITH OECD GROUP

THE FOLLOWING ARE PLOTTED AGAINSI TIM1E

PE-SCS-001 *SIGNAL VAR. NO. 6

8000000

7000000

6000000

5000000

4000000

3000000

2000000

1000000

0

TIME , SECONDS

FIGURE 37 - SECONDARY SYSTEM PRESSURE

LP-SB-I BASE CASE CALCULATION

AEEW - R 2254
10

S

(A

a.

6

0

L

C,,

0

4

2

0 500 1500 Z00

T ime (s)

FIGURE 38 - SECONDARY PRESSURE, COMPARISON WITH OECD GROUP

LP-SB-1 BASE CASE CALCULATION

TIE FOLLOVING ARE PLOTTED AGAINST REACTOR TlME

VAPOR FRACTION .STRATIFIED VOL FLAG *IASS FLOV RATE

FR-PC-503

12

10

8

0

4

2

0 1000 1500 2000 2500 3C

REACTOR TIME , SECONDS

FIGURE 39 - ILHL VOID FRACTION, STRATIFICATION I BREAK FLOW

LP-SB-1 BASE CASE CALCULATION

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST REACTOR TIME

TOTAL CPU TIME

IL•J

ovvw

5500-

5000 -

4500F

4000

3500

3000

2500-

2001500' -"

1000

500

0 I I I I f I

- RA BASE CASE

-'•BYPASS PATH MODIFIED

O 200 400 600 800 1000 1200 1400 1600 1800 2000

REACTOR TIME , SECONDS

FIGURE 40 - EFFECT OF SIZE OF BYPASS PATH ON RUNNING SPEED

LP-SB-I BASE CASE CALCULATION

THE FOLLOVING ARE PLOTTED AGAINST TIME

OE-PC-SO4B ,MIXTURE DENSITY

1000

800

600

r

4 400

200

0

800 1000 1200 1400 1600 1800 2000

TIME , SECONDS

FIGURE 41 - BREAK UPSTREAM DENSITY

LP-SB-l BASE CASE CALCULATION VITH EPRI CORRELATION

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST TIME

FR-PC-S03 *MASS FLOWJ RATE

In

TIME , SECONDS

FIGURE 42 - BREAK MASS FLOW RATE '

LP-SB-I BASE CASE CALCULATION WITH EPRI CORRELATION

THE FOLLOVING ARE PLOTTED AGAINST TIME

DE-PC-002A -DE-PC-002B ,DE-PC-002C

MIXTURE DENSITY

EXPT. BOTTOfl BEAM

r

MIDDLE BEAM

CASE

BEAM

TIME , SECONDS

FIGURE 43 - ILHL DENSITY

LP-SB-I BASE CASE CALCULATION VITH EPRI CORRELATION

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST TIME

DE-PC-O0lA ,DE-PC-O0IB *DE-PC-OOIC

MIXTURE DENSITY

IRAC BASE CASE

TRAC EPRI

r

-s

r

EXPT. BOTTOM BEAM

EXPT. MIDDLE BEAM

TOP BEAM

TIME , SECONDS

FIGURE 44 - ILCL DENSITY

LP-SB-I BASE CASE CALCULATION VITH EPRI CORRELATION

THE FOLLDVING ARE PLOTTED AGAINST REACTOR TIME

CONTROL OLK ID -?9,MASS INVENTORY

6000

5500

5000

4500

4000

3500

3000

2500 v

2000

1500

1000

5oo

n

0 200 400 600 800 1000 1200 1400 1600 1800 2000

REACTOR TIME * SECONDS

FIGURE 45 - PRIMARY SYSTEM MASS INVENTORY

LP-SB-I BASE CASE CALCULATION VITH EPRI-CORRELATION

0

AEEW - R 2254
THE FOLLOVING ARE PLOTTED AGAINST TIME

PE-PC-O02 #PRESSURE

20C 0 .2E De

C. .16E 06

14 0. 14E 08

12 0.12E 08

EXPT. TRAC BASE CASE

8 2 0 0 0 00 1. 80E 07

4 -O. 4DE 07

0 8 . .1 0. ODE 00

FIGURE 46 - PRIMARY SYSTEM PRESSURE

THE FOLLOVING APE PLOTTED AGAINST TIME

PE-SGS-O0I ,SIGNAL VAR. NO. 6

8 8000000

77000000

6 76000000

5 -5000000

TRAC EPRI

4 -4000000

3 TRAC BASE CASE 3000000

.2 2000000

I • 1000000

Ol 0

0 200 400 600 800 1000 1200 1400 1600 1800 2000

TIME , SECONDS

FIGURE 47 - SECONDARY SYSTEM PRESSURE

NEEW - R 2254 LP-SB-I BASE CASE CALCULATION WITH EPRI CORRELATION
THE FOLLOWING ARE PLOTTED AG-AINST REACTOR TIME

CONT'ROL BLK ID -33,CONTROL BLK 10 -52

ri,.

EPRI HP'

EPRI ILut LEVEL

600 800 1000 1200 1400 Ih

REACTOR TIME , SECONDS

FIGURE 48 - ILHL AND VAPOUR PULL THROUGH LEVELS

LP-SB-I BASE CASE CALCULATION VITH EPRI CORRELATION

THE FOLLOWING ARE PLOTTED AGAINST REACTOR TIME

CONTROL BLK ID -33,LIOUID LEVEL IN K L.

TRAC EPRI ILL LEVEL

ILHL LEVEL

REACTOR TIME , SECONDS

FIGURE 49 - MEASURED AND PREDICTED ILHL LEVELS

LP-SB-1 BASE CASE CALCULATION VITH EPRI CORRELATION

'AEEW - R 2254
d

2'correct behaviour"

breakline

density

ILHL level

decreasing (or

hp increasing)

time

FIGURE 50a

Effect, on Breakline Density, of the ILHL Licuid Level

(or the Level at which Vapour Pull-Through is Predicted

to Occur)

j

x \

breakline

density

mass in system (required

to be discharged prior to

break uncovery) increasing

- "correct"'behaviour

time P

FIGURE 50b

Effect. on Breakline Density, of the Amount of Mass

Effect, on in the System Required to be Discharged Prior to '2

time

Break Uncovery I

AEEW - R 2254
THE FOLLOWING ARE PLOIIED AGAINST TIME

DE-PC-SO4B ,MIXTURE DENSITY

1.0

uL

0.8

rO fi I

0.2

0.0I , I

0 200 400 600 800 1000 1200 1400

TIME 5 5ECONDS

FIGURE 51 - VAPOUR PULL THROUGH EFFECT ON BREAK LINE DENSITY

LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION

1000

EPRI
TVIEFOLLOW 1N1

FR-PC-5S03,

ARE PLOTTED AGAINST TIME

,MASS FLOW RATE

12

ul ,P. 10

Li

w

(n

Nl

EXPT.

IRAC 13ASE CASE

TRAC [PR[

TRAC EPRI + MODS

0 500 1000 1500 2000 2500 3000 3500 4000

TIMtE , 5ECOND5

FIGURE 52 - [3REAK MASS FLOW RATE

LP-SB-1 BASE CASE CALCULATION WITH EPRI CORR[LN' 9 INPUT MOOS
THE FOLLOWI1NG ARE PLOTTED AGAIN5T TIMlE :01

DE-PC-5048 ,MIXTURE DENSITY

1.0

0.

0.

N)-

LD 0.

EXPT.

TRAC BASE CASE

14)

II

"I

TRAC EPR!

IRAC [PRI + MOOS

0.

0.

0 500 1000 1500 2000 2500 3000

TIME , SECOND5

3500 4000

FIGURE 53 -

LP-SB-1 BASE

BREAK UPSTREAM DENSITY

CASE CALCULATION WITH EPRI CORRELN & INPUT MO5
IHE FOLLOWING ARE PLOTTED AGAINST TIME

upl

DE-PC-002A

MIXTURE DENSITY

,OE-PC-002B ,DE-PC-002CP03

.*

*3

2.-

EXPI. BOTTOM BEAM

)

EXPT. MIDDLE BEAMI

)

2o

Iw

Ll

TRAC EPRI + MOOS

IRAC EPRI

-TRAC BASE CASE

-EXPT. TOP BEAM

TIME , SECOND5

FIGURE 54 -

LP7-SB-1 BASE

ILHL DENSITY

CASE CALCULATION WITH EPRI CORRELN & INPUT MODS
TIHE FOLLOWING AIRE PLOTTED AGAINST TIME

:p'

tL3

A.

DE-PC-OOIA

MIXTURE DENSITY

,DE-PC-0011B ,DE-PC-OOIC

1.0

0.8

0.6

1000

if)

*

*

U, r

TRAC BASE CASE

IRAC EPRI

IRAC EPRI + MOOS

I2

U)

EXPT. BOTTOM BEAM

EXPT. MIDDLE BEAM

EXPT. TOP BEAM

0.4

0.2

0.0

TIME , SECONDS

FIGURE 55

LP-SB-1 BAS

- ILCL DENSITY

E CASE CALCULATION WITH EPRI CORRELN & INPUT MODS
THE FOLLOWING ARE PLOTTED AGAINST REACTOR TIME

CONTROL BLK ID -79,MASS INVENTORY

til

6000 6000

5500 5500

U'

5000 TRAC EPRI 5000

45001 4500

4000- 4000

3500- 3500

3000 3000

2500 2500

2000o ............... o2000

C / I-- _ EXPT.

0 500 1000 1500 2000 2500 3000

REACTOR TIME , SECONDS

FIGURE 56 - PRIMARY SYSTEM MASS INVENTORY

LP-SB-I BASE CASE CALCULATION WITH EPRI CORRELATION & INPUT MODS
THE FOLLOWING ARE PLOTTED AGAINST TIME

PE-PC-002 ,PRESSURE

t'J

ul

0. 20E

0. 1OE

0. 16E

0. 14E

0. 12E

0. 1OE

0. 80E

0. 60E

0. 40o

0. 20E

0. OOE

08

08

08

08

08

0B

07

07

07

07

00

a

C*

*-

z

0 500 1000 1500 2000 2500 3000 3500 4000

TIME , SECOND5

- PRIMARY SYSTEM PRESSURE

3ASE CASE CALCULATION WITH EPRI CORRELN & INPUT MODS

FIGURE 57

LP-SB.-1 E
THE FOLLOVING ARE PLOTTED AGAINST TIME

Lfl

PE-SGS-001 ,SIGNAL VAR. NO. 6

8000000

7000000

6000000

5000000

4000000 a

3000000

2000000

1000000

0 0 500 1000 1500 2000 2500 3000 3500 4000

TIME , SECONDS

FIGURE 58 - SECONDARY SYSTEM PRESSURE

LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELN & INPUT MOOS
T3f FOLLOVING ARE PLOTTED AGAINST REACTOR 7rnE

CONTROL. BIX 10 -33,LIQU!D VELOC17Y

10

8

6

4

2

0

REACTOR TIME , SECONOS

FIGURE 59 - ILHL LIQUID LEVEL AND NATURAL CIRCULATION

LP-SB-1 BASE CASE CALCULATION VITH EPRI CORRELATION & INPUT MODS

LP-SB-1

0.3

CD

0.2

0.1

0.8

0.6 V0..4 *•

0.2

0.0 - 0.0

0 500 1000 1500 2000

Time (s)

FIGURE 60 - ILHL LEVEL, COMPARISON WJITH OECD REVIEWJ GROUP

LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELATION & INPUT MOOS

AEEW - R 2254
THE FOLLOWING ARE PLOTTED AGAINST REACTOR TIME

MASS FLOW RATE

45

40

35

30

25

20

15

u

I

I EXIT FROM VESSEL

0 -

ENTRY' TO STEAM GENERATOR 1

0 200 400 600

REACTOR TIME

800 1000

, SECONDS

1200 1400

FIGURE 61 - ILHL MASS FLOW RATES

LP-SB-I BASE CASE CALCULATION VITH EPRI CORRELN & INPUT MODS

TIE FOLLOWING ARE LOTTED AGAINST REACTOR TIME

FUNCTION

0 200 400 600 800 1000 1200 1400

REACTOR TIME , SECONDS

FIGURE 62 - ILHL MASS FLOW BALANCE

LP-SB-1 BASE CASE CALCULATION WITH EPRI CORRELN & INPUT MODS

AEEW - R 2254
til

K)

THE FOLLOWING ARE PLOTTED AGAINST TIME

DE-PC-504B ,MIXTURE DENSITY

*

-*f

r_

TRAC HP=O. 142

-TRAC BASE CASE

TRAC EPRI + MODS

EXPT.

TRAC CATHARE

0 200 400 600 800 1000 1200 1400

TIME , SECONDS

FIGURE 63 - EFFECT OF CORRELATION CHOICE ON BREAKLINE DENSITY

LP-SB-I BASE CASE CALCULATION WITH EPRI CORRELATION & INPUT MODS
THE FOLLOWING ARE PLOTTED AGAIN5T TIME

ti2

t.P

FR-PC-503 ,MASS FLOW RATE

6.0

U

ILo

•r•

EXPT.

.TRAC HP=O. 142

*TRAC BASE CASE

*TRAC EPRI + MODS

TRAC CATHARE

0 200 400 600 800 1000 1200 1400

TIME . SECONDS f

FIGURE 64 -

LP-SB-I BASE

EFFECT OF CORRELATION CHOICE ON BREAK MASS FLOW

CASE CALCULATION WITH EPRI CORRELATION & INPUT MODS
Lfl

TfHE FOLLOWING ARE PLOTTED AGAINST REACIOR TIME

CONTROL BLK ID -33,CONTROL BLK ID -32

0.30

0.28

0.26

0.

(ft

0.22

0.20

0.18

0.16

0. 14

0.12

0.10

-EPRI + M005 HP

-CATHARE HP

EPRI + MODS ILHL LEVEL

CATHARE ILHL LEVEL

REACTOR TIME , SECONDS

FIGURE 65 -

LP-SB-1 BASE

EFFECT OF CORRELATION CHOICE ON ILHL LEVEL

CASE CALCULATION WITH EPRI CORRELATION & INPUT MOOS
THE FOLLOWING ARE PLOTTED AGAINST TIME

FR-PC-503 ,MASS FLOW RATE

ti'

A.

6.0

5.5

5.0

4.5

tL) u

L3

V

4.01

3.5

3.0

2.5

2.0

2

1.0

0 50 100 150 200 250 300 350 400 450 500 550 600

TIME , SECONDS

FIGURE 66 - EFFECT OF 2-PHASE MULTIPLI.ER CHOICE ON BREAK FLOW

LP-5B-1 BASE CASE CALCULATION WITH EPRI CORRELN & INPUT MOOS
APPENDIX I

MICROFICHE LISTING OF THE TRAC-PFI/MODI INPUT DECK FOR

LP-SB-1 (USED FOR THE STEADY STATE CALCULATION)

(See inside back cover pocket for microfiche)

AEEW - R 2254

APPENDIX I I

THE EPRI CORRELATION FOR BRANCHLINE FLOW QUALITY AS A

FUNCTION OF MAINLINE STRATIFIED LIQUID LEVEL

Experiments performed in the Two-Phase Flow Loop (TPFL) at the

Idaho National Engineering Laboratory (INEL), during 1984,

investigated the liquid entrainment/vapour pull-through phenomena

of a tee with a large mainline to branchline diameter ratio and

the effect of these phenomena on the flow rate in the branchline

(19). As a result of the experimental work, correlations

(recommended for use in codes) for predicting the liquid levels

for the onset of vapour pull-through and liquid entrainment and

the resulting branchline flow quality were derived as follows:

he 1

D 2

Ce

D

Cp

D

* 2

9g( 0.2

gPg (Pf - Pg)

hp

D

1

2

f 2 0.2 p (]

gf(Pf - Pg)

X = 0ifh >hpp

hhe

X = exp [Cx ( A

bp he

if he 4h < hp

X 1 lif h< he

where he = liquid level at

(i)

= liquid level at

(W)

which liquid entrainment begins

which vapour pull-through begins

D = mainline internal diameter (m)

Ce = constant obtained from experimental data (0.62)

C = constant obtained from experimental data (0.82)

;g= gas mass flow rate in branchline (kg s-1)

0

mf =liquid mass flow rate in branchline. Ckg s-1 )

g

AEEW - R 2254

= acceleration due to gravity (ms- 2 )
Pg gas density (kg m- 3 )

*-pf = liquid density (kg m- 3 )

x = branchline flow quality

Cx= constant obtained from experimental data (- 3.4)

Implementing the correlation for branchline flow quality as a

function of stratified liquid level involved modifying the

LP-SB-I input model as shown in Figure AII.I. A VALVE component

was introduced in order to control the amount of steam entering

the branchline and Version B02C of the TRAC-PFl/MODl code -

employing a branch offtake quality model for stratified flow -

was used.

AEEW - R 2254
99 ILHL.

99 ILL

4

I

I

I

I

I

3 2

BREAK

(a) Base Case Calculation

97 ILHL 99

2 I3

I ,1

'steam-only' junction!' 2 i'"i

(FRIC = 1.0E20) 1 ___

Quality control

valve

I

I

I 2

Offtake model applied

here

3 q5

4

BREA

(b) EPRI Correlation Calculations

FIGURE AII.1

Nodalization Modification for Implementation of Breakline

Quality Control

AEEW - R 2254

APPENDIX III

EFFECT OF DEFICIENCIES IN THE TRAC-PFI/MODI INTERPHASE DRAG

MODEL ON THE PREDICTION OF THE CORE DENSITY FOR LP-SB-l

An assessment of the interphase drag correlation used for

modelling vertical two-phase flows in TRAC-PF1/MOD1 is described

in Reference 20. The assessment compared void fractions

calculated by the code with those predicted by standard

correlations in order that an estimate might be made of the void

fraction errors likely to arise, in a particular application, due

to deficiencies in the code's modelling of interphase drag.

The plot of percentage error in the density predicted by the code

(when compared with that predicted by the Wilson-Rooney

correlation) as a function of void fraction (20) is reproduced in

Figure A.III.1. (The Figure is applicable to zero liquid flow

rates - under which circumstances errors were found to be

largest).

In order to gain an indication of the likely magnitude, and

effect, of errors in the densities predicted by TRAC-PF1/MODl for

LP-SB-1, Figure A.III.1 was applied at a particular point in the

transient. Figure A.III.2 shows the fluid conditions in the core

(component number 88) and the pipe above the core (component

number 87) at - 800 seconds as predicted by the "Base Case + EPRI

Correlation" calculation of LP-SB-1.

In cell numbers 87/1, 87/2, 88/5 and 88/4, the fluid conditions

are such that the error in density (defined as (P 2 - P1)/Pi x

100%, where P, = density calculated by Wilson Rooney correlation

and P2 = density predicted by the code) is - - 10%, ie the

density in these cells is underestimated, by TRAC-PF1/MODI, by

- 10%. For the remaining cells (88/1, 88/2 and 88/3) the density

error is negligible.

Had the density in parts of the core and the pipe above the corje

not been underestimated, the additional gravitational head

available to balance the fluid in the downcomer would have been:

(hp) x g x 10%

= (l.446x486+0.476x544+0.457x564+0.457x591) x 9.81 x 0.1

- 1461 kg m-1 s-2

This would be equivalent to increasing the density in the top 1461

cell of the downcomer by 1461 kg m- 3 ie 103 kg m- 3 . 9.81 x 1.446

Therefore, had the density not been underpredicted, additional

fluid of the order of

(0.46762x486+0.15393x544+8.0499x10- 2 x564+8.0499x10- 2 x591) x 0.1 +

0.19192 x 103*

- 60 kg

* X (Vol x Ap)

AEEW - R 2254
would have remained in the vessel. It would have been necessary to expel less fluid from the break and break uncovery could be expected to'have occurred - 20 seconds earlier.

AEEW - R 2254
FIGURE AIII.I

ERRORS IN MEAN TWO-PHASE MIXTURE DENSITY FOR j = 0

in

'U

+. 100

80

60

40

20

0-

-20

-So

-50

-to

-100

(a) D h0.01m

Pzlo MPa

20 40 60 80

Oag (-/.)

PA& MPG

PI,1Po KEY "-

-- RELAP S I MOD 2

--- TRAC -- PF1/ MOD I

AEEW - R 2254
FIGURE A.III.2

FLUID CONDITIONS IN THE CORE AND THE PIPE ABOVE THE CORE

AT - 800 SECONDS

Jg

(ag Ug) Ug CELL

j I

Ujz (1-a9) Ut

0.225 0.6054

0.14

1.446

0.3761

0.476

0.10 0 .3512

0.09

0.08

0.05

0.02

0.01

0.457

0.3481

... 0.457

0.3436

0.3457

0.3463

0.3467

pm = 486 kgm-3

n a = 0.372 87/2

IVL = 0.46762 m3

pm = 544 kgm-3

m a = 0.2925 87/1

VgL = 0.15393 m3

pm = 564 kgm-3

m a = 0.265 88/5

V8L = 8.0499 x 10-2 m3

pm = 591 kgm-3

m a = 0.2275 88/4

1 V8L = 8.0499 x 10-2 m3

x

a = 0.131 88/3

V8 L = 4.0337 x 10-2 m3

a = 0.06781 88/2

V8 L = 4.0161 x 10-2 m3

a = 0.01798 88/1

= 5.3724 x 10-2 m3

0.2583

0.1184

0.09877

0.09487

0.08604

0.08109

0.07751

0.07588

0.16

0.07

0.07

0.07

0.07

0.07

0.07

0.07

Pressure - 6.3 MPa

Hydraulic diameter = 0.0122 m

AEEW - R 2254
APPENDIX IV

1

MODIFICATIONS TO LP-SB-l BASE CASE INPUT DECK FOR FINAL

TRANSIENT CALCULATION

Correction of GRAVs in Lower Plenum (Component Number 90)

GRAV

Base Case Final .

Calculation Calculationi

Main Branch -1.0, 0.0, 1.0 -1.0, -0.182245, 1.0

Side Branch 0.0, 1.0 -0.0824089, 1.0

2 Correction of Effective Cell Lengths for Plenum Components

(component Numbers 82 and 84)

EFFECTIVE PLENUM-SIDE CELL LENGTHS

Component - Base Case Final

Calculation Calculation2

82 0.142,0.0705,0.0705, 0.284,0.141,0.141,

0.142,0.142,0.0705 0.284,0.284,0.141

84 0.142,0.329,0.142, 0.284,0.658,0.284,

0.142,0.329,0.329 0.284,0.658,0.658

1 data obtained by running TRAC-PFI/MOD1 with elevation data and

allowing the code to compute the GRAV terms-.... : .

2 the code expects the total height (and width) of the plenum and

not half the height (and half the width) as was :specified -for

the Base Case Calculation. -

AEEW - R 2254
APPENDIX IV

MODIFICATIONS TO LP-SB-l BASE CASE INPUT DECK FOR FINAL

TRANSIENT CALCULATION

Correction of GRAVs in Lower Plenum (Component Number 90)

GRAV

Base Case Final

Calculation Calculation1

Main Branch -1.0, 0.0, 1.0 -1.0, -0.182245, 1.0

Side Branch 0.0, 1.0 -0.0824089, 1.0

2 Correction of Effective Cell Lengths for Plenum Components

(Component Numbers 82 and 84)

EFFECTIVE PLENUM SIDE CELL LENGTHS

Component

Base Case Final

Calculation Calculation 2

82, 0.142,0.0705,0.0705, 0.284,0.141,0.141,

0.142,0.142,0.0705 0.284,0.284,0.141

84 0.142,0.329,0.142, 0.284,0.658,0.284,

0.142,0.329,0.329 .0.284,0.658,0.658

1 data obtained by running TRAC-PF1/MOD! with elevation data and

allowing the code to compute the GRAV terms.

2 the code

not half

the Base

expects the total height (and-width) of the plenum and

the height (and half the width) as was specified for

Case Calculation.

AEEW - R 2254
6 Renodalisation of Reflood Assist Bypass Line 6

(i) Broken Loop Hot leq (Component Number 31)

Base Case Final

Calculation Calculation

Side Branch RAD 0.10795 m 0.111 m

" " TH 0.02858 m 0.0255 m

" DX 1.389 m,0.814 m, 1.687 m,0.726 m,

5.1044 m .1.849 m

" VOL 0.054 m3,0.0314 m3 , 0.0658 m3 ,0.0283 m3

0.1986 m 0.0721 m 3

" FA R3 0.0388 m2 , 0.02119 m2 R3 0.039 m2 , 0.013 m2

GRAV 0.0,0.2242,0.1912,0.0 0.0,1.0,0.0,0.0

" HD R3 0.2223 m,l.41E-3 m R3 0.2228 m,1.41E-3 m

(ii) Broken Loop Cold Leg (Component Number 41)

Base Case Final

Calculation Calculation

Side Branch RAD 0.10795 m 0.111 m

" TH 0.02858 m 0.0255 m

" DX 0.885 m,7.2834 m 2.129 m,3.27 m

" VOL 0.03033 m3 40.2768 m3 0.0830 m3,0.1275 m3

FA R2 0.0388 m2 ,0.02119 m2 R2 0.039 m2 ,0.013 m2

GRAV 0.0,0.199,0.0 1.0,0.0,0.0

HD R2 0.2223 m,l.41E-3 m R2 0.2228 !m 1.41E-3 m

6 Size of Reflood Assist Bypass Line decreased (2).

AEEW - R 2254

APPENDIX V

LP-SB-1 BASE CASE CALCULATIONS WITH EPRI CORRELATION AND

INPUT MODEL MODIFICATIONS - SERIES OFýPICTURES SHOWING

PREDICTED SYSTEM CONDITIONS THROUGHOUT THE TRANSIENT

The following series of pictures shows, at 200 second intervals,

the predicted void fraction distribution, the liquid and vapour

velocities and the occurrences of stratified flow conditions

throughout the system. The nodalisation may be compared with

that of Figures 3a, 3b, 4 and 5. The pictures were produced from

the final calculations - Base Case + EPRI Correlation + Input

Modifications - and include the break line quality control valve

(see Figure AII.1).

AEEW - R 2254
tri)

yi'

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME3 - TRAC PREDICTION OF LOFT SB-i (BASE + EPRI CORRELN + INPUT MODS) Ti 0.1

yL!o --4

5.00)

WAP

( 5.00)

VOtw R~CTIMI

I I

I

I
~tJ

IL3

.Al.

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-1 (BASE + EPRI CORRELN + INPUT MODS) Ti 229.2

VL.a -- 4

(5.00)

VAP

( 5.00)

VOID FRACTION

I I

I U-I

I
t:4

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-i (BASE + EPRI CORRELN + INPUT MODS) Ti 409. 6

( 5.00)

V.AP

(5.00)

VOID FR~ACTION

... .O... .. ..

F-F-T1
S. M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MODS)

tx.J

Ur'

Ti 626. 2

VL!O - (5.00)

WAP -+

(5.00)

VOIt FRACTION

I .... I
(Il

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-1 (BASE + EPRI CORRELN + INPUT MOOS) Ti 802.'0

VL[ I a

5.00)

WVAP -

( 5.00)

VOID FRACTION
Ln

l-4

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-i (BASE + EPRI CORRELN + INPUT MODS) Ti 1012.9

VLtIQ -

VVAP--

( 5.00)

voto niFRcrow

8 89..
J.-

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-I (BASE t EPRI CORRELN t INPUT MODS) Ti 1223.9

Vy..IQ

( 5.00)

VYAP -+

(5.00)

VOID FRACTION
LnI

a.b

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC-PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MODS) Ti 1402.2

V I a ----

(5.00)

VVAP

(5.00)

.1

VOID FlRACTO

I

[1
9-I

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MODS) Ti 1613.2

YvIr -- +

( 5.00)

VVAP +

( 5.00)

VOIO FRACTION
9'

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAMEs- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MOOS) Ti 1824.7

(5.00)

VYAP --

(5.00)

'VOID FRACTION

I .1]
t'i S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MODS)

tjLn

Ti 2010.5

VL IO -4

(5.00)

VVAP

( 5.0o0)
t!I

ul

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAMEt- TRAC PREDICTION OF LOFT SB-I (BASE t EPRI CORRELN + INPUT MODS) Ti 2221. 9

VLW "--+

( 5.00)

VVAP

(5.00)

voto FRAcriom rI-- % i0

I1
tri

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-i (BASE + EPRI CORRELN + INPUT MODS) Ti 2403. 1

( 5.00)

VYAP

( 5.00)

VOID FRACTION
till

N)

.4.

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAMEo- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MODS) Ti 2614. 5

V Ito ---- +

(5.00)

VVAP -- +

(5.00)

VOID FRACTION
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAMEo- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN t INPUT MODS) Ti 2825. 6

(5.00)

VVAP

(5.00)

VOID FRiACrwN

F~r..8B

LA

0-4
u-I

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN +.INPUT MODS) Ti 3006. 8

VLIa --- +

(5.00)

VWAP

(5.00)

VOID FRACTION

I .... I
tri

K)

NI

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MODS) Ti 3217. 7

VL 10

1 5.00)

VVAP

1 5.00)

VOID FRACTION

I77]-
t:'1

NJ

N; .

(11

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MODS) Ti 3428. 6

VLIO ----

(5.00)

VVAP --

( 5.001

VeOI FRACTION

F I ].--

I

F
tnl S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TI Ti 3600. 7

(5.00)

WAP

(5.00)

voro FRACrioN

I
-4

N1

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME,- TRAC PREDICTION OF LOFT SB-1 (BASE + EPR[ CORRELN + INPUT MODS) Ti 3811.5

VLIOa -4+

(5.00)

VYAP

(5.00)

VOID FRACTION

FAi
In

S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME#- TRAC PREDICTION OF LOFT SB-I (BASE + EPRI CORRELN + INPUT MODS) Ti 3932. 0

VLIOa -4+

1 5.00)

VVAP

(5.00)

VOID FRACTION

FI- 0R
DISTRIBUTION

PWdR HEAT TRANSFER AND HYDRAULICS WORKING GROUP

Mr

Dr

Mr

Mr

Mr

Dr

J Fell

M J Bird

P Dore

I Brittain

K G Pearson

D B Utton

Mr K Poutledge

Mr

Mr

Dr

Mr

Dr

Dr

Mr

Mr

Dr

D

P

P

M

L

T

D

B

M

Coucill

C Fall

R Farmer

W E Coney

C Daniels

J Haste

W McLaughlin

Chojnowski

El-Shanawany

AEEM

AEEW

AEEW

AEEW

AEEW

NIX

NNIC

BNF

CEGB

CEGB

CEGB

UKAEA

UKAEA

UKAEA

CEGB

CEGB

166/A32

338/B41

114/B40

201/A32

333/B41

c/o Reports Section, Library, Booths Hall,

Chelford Road, Knutsford, Lheshire WAl6 9QZ

c/o Reports Section, Library, Booths Hall,

Chelford Road, Knutsford, Cheshire

Springfields Works, Salwick, Preston, Lancs

Barnett Way, Barnwood, Gloucestershire

HSD, Courtenay Iouse,. 18 Warwick Lane,

London EC4P 3EB

CERL Kelvin Avenue, Leatherhead, SurreyBldg 392, AERE Farwell

Springfields NPDL, Salwick, Preston, Lancs

CLF 125, SRD Culcheth, Warrington

Marchwood Engineering Labs, Marchwood, Southampton

HSD, Nuclear Safety Branch, Courtenay House, 18

Warwick Lane, London EC4P 3EB

ADDITIONAL

AEE WINFRITH, DORCHESTER, DORSET, UDT2 8DH

Mr M

Dr I

Dr G

Dr P

Mr R

MrR

Dr A

Mr A

Miss

Dr C

Dr D

DrA

Dr S

DrJ

H Butterfield

H Gibson

R Kinmber

Coddington

Potter

Herbert

J Wickett

Neill i

E Allen

Richards

A Williams

L Nichols

R Kirnnersly

N Lillington

309/B41

236/A32

221/A32

.270/A32

209/A32

172/A32

214/A32

267/A32

266/A32

260/A32

210/A32

102/A50

203/A32

217/A32

(3

(3

(5

ccpies)

copies)

copies)

AE, - R 2254 i
UKAEA, SAFETY AND RELIABILITY DIRECrORATE, WIGSHAW LANE, CaYCHEIz, WARRINGTCN

Mr R N H McMillan CLF 112

Dr A A Debenham CLC 002

Mr P Clough

UKAEA, RISLEY NPDE, WAMRRNGTON, WA3" 6AT

Reports Library

Mr A Parry E417

UKAFA, .AERE HAR/ELL, DIDCOT, OXON, DXIl ORA

Dr D Hicks G35, Bldg 77

Mr M R Everett Bldg 329

Dr G F Hewitt TH 392.7

Mr P Black TH 392.4

UKAEA, RISLEY, NPDL, WARRINGTON, WA3 6AT

Dr W E Gardner 3/103

UKAEA SPRINGFIELDS NPDL, SALWICK, PRESTON, PR3 ORR

Mr R D Stacey Bldg 321

Mr A J Manley

UKAEA, WINDSCLE NL, SELLAFIELD, SEASCALE, COMBRIA

Mr A Garlick

BNFL, SPRINGFIELDS WORKS, SALWICK, PRESTO, PR3 OXJ

Mr D A Howl Bldg 382

Mr J P Cooch Bldg 382

NNC, BOOTH'S ALL, CHELFORD ROAD, KNUTSFORD, "IESIRE

Library Reports Section - 4 copies

Mr K Routledge

Dr K R Barry

Mr R J Tinsley

NNC Risley,,Wairington Road, Risley, Warrihgton

Dr D Briney

CEGB, *COMPUTING BUREAU, 85" PARK STREET, 'LCNDCT SEI

Mr L Wilson

CEGB, HEALTH AND SAFETY DEPARMENT, NUCLEAR 'SAFETY BRANCH, 18 WARICK LANE

LaMNON, EC4 4EB

Mr J R Harrison

AEEW - R 2254 ii
CEGB, PROJEC' MA1AGE4ENT TEAM, BCOTHS HALL, CQELFORD ROAD, KNUTSFORD, CHESHIRE

Mr M E -Durham

Dr M M Chestnutt

Mr P George

Dr R Garnsey

CEGB, CERL, KELVIN AVENUE, LEAT2HEAD, SURREY

Dr G C Gardner

Mr N Newman

CEGB, BEELEY NUCLEAR LABORATORIES, BERKELEY, GLOUCESrUSHIRE, GL13 9PB -

Dr D Witherington

Dr S J Board

Ms J P Longworth

Mr N Buttery

CEGB, GDCD BARNWý)D, BARNEIT A GLOUCESTER, GLOUCESTEREHIRE, GL4 7RS

Mr C Harwood

Mr I Pugh

Mr P D Jenkins

Mr K H Ardron

Mr J R D Jones

CEGB, FARCHOO'D ENGINEEMING IABORATORIES, SOUtTHPTOTN, FAPSHIRE, S04 4ZB

Dr B Chojnowski

SOUIH OF SCOMTXND ELErCITY BOARD, CATiCART HOUSE, SPEAU STREET, GLASGOW,

SCOTLAND

Mr D R Reeks

NUCLEAR NSTALLAIONS 'INSPECTORATE,- HELTH AND SAFETY EXECUTIVE, ST PETERS

HOUSE, STANLEY PRECINCT, BALLIOL ROAD, BOOIT.LE, L20 3LZ

Mr C Potter

Mr J F Campbell

Dr S A Harbison

CONSEJO7.DE SGURIDAD NUCLEAR, C/ SOR'ANGEL DE LA CR=Z 3, 28020 MADRID, SPAIN

Sr F Pelayo

AE=.% - R 2254 iii
LOFT PROGRA•E! REVIEW GROUP

TECHNICAL RESEARCH CNTER OF FINLAND (VTr), NUCLEAR ENGINEING LABORATORY,

POB 169, SF-00181 HELSINKI 18, FINLAND

Mr Heikke Holmstrom

EIDGENSSISCHES INSTITUT FUR PEAKTOREORSCHEG,-cH-8303 'AR--INGEN,

SWITZERLAND

Mr 0 M Mercier

OSTERREICHISCHES FORSCUNMSZETRMO SEIBERSDORF GESEfLSCHAJT za, .C/OMG

IDAHO INC, PO BOX 1625, IDAHO FALLS, ID 83415, USA

Dipl Phys S Michael Modro

UNIDAD ELECTRICA SA, C/ FRANCISCO GERVAS 3, E-28020 MADRID, SPAIN

Mr Jose Puga

OFFICE OF'NUCLEAR REACTOR REGULATION, US NUCLEAR REGULATORY CCMMISSION,

WASHINGTON DC 20555, USA

Dr Brian Sheron

JAPAN AT=IC EERGY RESEARCH INSTITUTE,' TOKA-1URA, NAKA-GUN, IBARAKI-KE

219-11, JAPAN

Dr Kanji Tasaka

US DEPARThMT OF DŽENGY, OFFICE OF NUCLEAR POWER SYSTMIS, MS B-107, WASEIUTCON

DC 20555,-USA

Mr D F Giessing

.GDCD, CaN7RAL ELECTRICITY GENERATING BOARD, BARE=T MAY, BARnwD, GLoCCESTER,

GL4 7RS, UK

Mr P D Jenkins

EPRI, PO BOX 10412, PALO*ALTO, CA• 94304, USA

Mr M Merilo

NTE NAZIONALE "ENEIE ALTERNATIVE, CRE CASACCIA, SP ANGUILLARESE KM 1 + 300,

CP NR 2400, 1-00100 ROME, ITALY

Dr Giorgio Palazzi

AEW.1 - R 2254 iv
STUDSVIK RIERGITEKIK AB, S-61182 NYKOPING, SWED&-

Mr 0 Sandervaag

US NU=AR REJATOFY Cd?^SSION, WAS1MNGTCO DC 20555, USA

Dr Robert Van Houten

EG&G, IDAHO INC, PO BOX 1625, IDAHO FALLS, ID 83415, USA

Dr D Croucher

OECD NUCt3AR TERGY AG-•CY, 38 BOULEVAD SUCHET, F-75016 PARIS, FRANCE

Mr Ralph Caruso

LIBRARIES:

Oounreay

Harwell

Risley

Springfields

Windscale

Winfrith

2

2

4

2

2

4

*U.S. GOVERNMENT PRINTING OFFICE: 1990--267-419/50793

AE-E, - R 2254 V

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12-89) (Asigned by NRC. Add Vol., Supp.. Rev.,

NRCM 1102. and Addendum Numbers, If any.)

3201,3202 BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/IA-001 1

2. TITLE AND SUBTITLE AEEW-R 2254

TRAC-PFl/MODl Post-Test Calculations of the OECD LOFT

Experiment LP-SB-l 3. DATE REPORT PUBLISHED

MONTH J YEAR

Apri l 1990

4 4. FIN3RANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

E. J. Allen Technical

7. PERIOD COVERED llnclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address,- if contractor, provide

name and mailing address.)

United Kingdom Atomic Energy Authority

Atomic Energy Establishment, Winfrith

Dorchester, Dorset DT2 BDH

United Kingdom

9. SPONSOR ING ORGANIZATION - NAME AND ADDR ESS (If NRC, type "Same as above"; if contractor. provide NRC Division, Office or Region. US. Nuclear Regulatory Commission,

and mailing address,)

Office of Nuclear Regulatory Research

U. S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

Analysis of the small, hot leg break, OECD LOFT Experiment LP-SB-I. using the

"best-estimate" computer code TRAC-PFl/MODl is presented.

Descriptions of the LOFT.facility and the LP-SB-I experiment are given and development

of the TRAC-PFl/MODl input model is detailed. The calculations performed in achieving

the steady state conditions, from which the experiment was initiated, and the

specification of experimental boundary conditions are outlined.

12. KEY WORDS/DESCR!PTORS (LIst words or phrases that will assist researchers In locating the report.) 13. AVAILABILITY STATEMENT

Unl imi ted

TRAC-PFI, LOFT, ICAP Program 14.SECURITY CLASSIFICATION

(This Page)

Unclassified

(This Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-89)

~scz-~

F

lit rnvr

-a.- -I '4 III

F'

*1*

IP

44

4 4.

41

*

tl

4'L

* Aflt. 

Ii5:

ti

'I

4' t

L qr .ý d

~#~s~~tb$

I.

I

it

ft.

,n*

Iqn,

'A

at.

I

r

Er",',.

&

Is

ol4

'I

f 4' 4

f4
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C, 20555

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE. $300

SPECIAL FOURTH-CLASS RATE

POSTAGE & FEES PAID U USNRC 

1

PERMIT No. G-67 J