NUREG/A-0022

Inteonaional

• Agreement Report

TRAC-PF1/MOD1 Post-Test

Calculations of the OECD

LOFT Experiment LP-SB-3

Prepared by

E. J. Allen, A. P. Neill

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-Hydrauflc Code Assessment

and Applcation Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
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by such third party Would not infringe privately owned rights.

Available from

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U.S. Government Printing Office

P.O. Box 37082

Washington, D.C. 20013-7082

and

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NUREG/IA-0022

• International Agreement Report

TRAC'PF1/MOD1 Post-Test

Calculations of the OECD

LOFT Experiment LP-SB-3

Prepared by

E. J. Allen, A. P. Neill

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 Applicatlon Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission

J

!
0

/
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, cold leg break, OECD LOFT Experiment

LP-SB-3 using the best-estimate computer code TRAC-PF1/MODl is

presented.

Descriptions of the LOFT facility and the LP-SB-3 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.

Results of the TRAC-PFl/MODl calculation are found to be

generally consistent with those reported, by members of the OECD

LOFT Program Review Group, in the LP-SB-3 "Comparison Report".

Overall trends with respect to pressure histories, minimum

primary system mass inventory and accumulator behaviour are

reasonably well reproduced by TRAC-PFl/MOD1. Prior to break

uncovery, the break mass flow rate is slightly over-predicted by

the TRAC critical flow model. (Subcooled and two-phase choked

flow multipliers of 1.0 were used throughout the calculation).

The most significant discrepancy is in the rate with which the

fuel rod cladding temperature rises during the core uncovery

phase of the transient. TRAC-PF1/MOD1, in common with other

codes, significantly over-predicts the rate with which the core

heats up. Contrary to experimental observations, conditions for

reflux condensation are not predicted by TRAC-PFl/MODl during

this-part of the transient. This, together with the underprediction of core density by TRAC's interphase drag model, is

considered a potential contributor to the poorly predicted rate

of core heat-up.

AEE7 - R 2275 (ii)
CONTENTS

SECTION 1 INTRODUCTION

2 THE LOSS OF FLUID (LOFT) TEST FACILITY

3 EXPERIMENT LP-SB-3

4 TRAC-PFI/MOD1

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

6 STEADY STATE CALCULATION

7 BOUNDARY CONDITIONS FOR TRANSIENT

CALCULATION

7.1 Decay Heat Data

7.2 Primary Pump Injection

7.3 Operational Setpoints

8 TRANSIENT CALCULATION

8.1 Introduction

8.2 CPU Usage and Time Step Behaviour

8.3 Chronology of Events

8.4 Mass Depletion Phase

8.4.1 Primary and Secondary System

Pressures

8.4.2 Break Mass Flow Rate, Break

Line Density and Primary

System Mass Inventory

8.4.3 Primary System Densities

8.4.4 Pumps Degradation

8.5 Core Boil Off Phase

8.5.1 Primary and Secondary System

Pressures

8.5.2 Break Mass Flow Rate and

Primary System Mass Inventory

8.5.3 Primary System Densities

8.5.4 Fuel Cladding Temperatures

PAGE

1

1

2

3

3

4

4

4

4

4

5

5

5

5

6

7

7

8

9

9

9

10

10

10

AEEW - R 2275 (iii)
CONTENTS (Continued)

PAGE

8.6 Core Cool-Down and Recovery Phase

8.6.1 Primary and Secondary System

Pressures

8.6.2 Fuel Cladding Temperatures

8.6.3 Accumulator Injection Flow

9 CONCLUSIONS

11

12

12

12

13

14

14

10

11

ACKNOWLEDGEMENTS

REFERENCES

APPENDIX I - MICROFICHE LISTING OF THE

TRAC-PFl/MODI INPUT DECK FOR

LP-SB-3

APPENDIX II - SERIES OF PICTURES SHOWING

PREDICTED SYSTEM CONDITIONS

THROUGHOUT THE LP-SB-3 TRANSIENT

APPENDIX III - COMPARISON OF BREAK MASS FLOW RATE

PREDICTED BY TRAC-PFl/MODI WITH

THAT IMPLIED BY THE HOMOGENEOUS

EQUILIBRIUM MODEL

AEEW - R 2275 (iv)
LIST OF TABLES

1 INITIAL CONDITIONS FOR EXPERIMENT LP-SB-3

2 OPERATIONAL SETPOINTS FOR EXPERIMENT LP-SB-3

3 CHRONOLOGY OF EVENTS FOR EXPERIMENT LP-SB-3

AEEW - R 2275 (v)

LIST OF FIGURES

1 LOFT SYSTEM CONFIGURATION FOR EXPERIMENT LP-SB-3

2 TRAC-PFI/MOD1 PRIMARY SYSTEM NODALISATION DIAGRAM FOR

LP-SB-3

3 TRAC-PF1/MOD1 REACTOR VESSEL NODALISATION DIAGRAM FOR

LP-SB-3

4 TRAC-PF1/MOD1 STEAM

DIAGRAM FOR LP-SB-3

5 LP-SB-3 CALCULATION

6 LP-SB-3 CALCULATION

7 LP-SB-3 CALCULATION

8 LP-SB-3 CALCULATION

(0-300 SECONDS)

9 LP-SB-3 CALCULATION

(0-6400 SECONDS)

10 LP-SB-3 CALCULATION

11 LP-SB-3 CALCULATION

12 LP-SB-3 CALCULATION

13 LP-SB-3 CALCULATION

ONSET

14 LP-SB-3 CALCULATION

ONSET

GENERATOR SECONDARY SIDE NODALISATION

- CPU USAGE

- TIME STEP SIZE

- PRIMARY SYSTEM PRESSURE

- SECONDARY SYSTEM PRESSURE

- SECONDARY SYSTEM PRESSURE

- BREAK MASS FLOW RATE

- BREAK LINE DENSITY

- PRIMARY SYSTEM MASS INVENTORY

- ILHL DENSITY AND STRATIFICATION

- ILCL DENSITY AND STRATIFICATION

15 LP-SB-3 CALCULATION - PRIMARY COOLANT PUMPS DIFFERENTIAL

16

17

18

19

PRESSURE

LP-SB-3 CALCULATION - PRIMARY SYSTEM - SECONDARY SYSTEM

PRESSURE

LP-SB-3 CALCULATION - PRIMARY AND SECONDARY SYSTEM

PRESSURES

LP-SB-3.CALCULATION - PEAK CLADDING TEMPERATURE (0.05 m)

LP-SB-3 CALCULATION - PEAK CLADDING TEMPERATURE (0.28 m)

AEEW - R 2275 (vi)
LIST OF FIGURES (Continued)

20 LP-SB-3 CALCULATION - PEAK CLADDING TEMPERATURE (0.53 m)

21 LP-SB-3 CALCULATION - PEAK CLADDING TEMPERATURE (0.81 m)

22 LP-SB-3 CALCULATION - PEAK CLADDING TEMPERATURE (1.25 m)

23 LP-SB-3 CALCULATION - PEAK CLADDING TEMPERATURE (1.57 m)

24 LP-SB-3 CALCULATION - ACCUMULATOR MASS FLOW RATE

25 LP-SB-3 CALCULATION - ACCUMULATOR MASS FLOW RATE AND

SECONDARY SYSTEM PRESSURE

26 LP-SB-3 CALCULATION - ACCUMULATOR MASS FLOW RATE AND

SECONDARY SYSTEM PRESSURE (STEAM BYPASS VALVE FLOW AREA

INCREASED)

AEEW - R 2275 (vii)
COMMERCIAL IN CONFIDENCE

1 INTRODUCTION

This paper describes a post-test calculation of the OECD LOFT

small, cold leg break experiment LP-SB-3 using the

"best-estimate" computer code TRAC-PFl/MOD1. Sections 2, 3 and 4

describe the LOFT facility, the LP-SB-3 experiment and the

version of TRAC-PFl/MODI used, respectively. Development of the

input model is detailed in Section 5 and the calculations

performed in 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 transient calculation. The main conclusions from

the analysis are summarised in Section 9.

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-3, 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-3 was in the cold leg of the intact loop

between the primary coolant pumps 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-3, the QOBVs and

the isolation valves remained closed (because the break was in

the operating loop). Broken loop spool pieces, with orifices to

simulate the steam generator and pump hydraulic resistances, were

installed for experiment LP-SB-3.

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.

AEEW - R 2275 1
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 HPIS was not used during experiment LP-SB-3.

3 EXPERIMENT .LP-SB-3

Experiment LP-SB-3 was conducted on 5 March 1984 in the LOFT

facility at the Idaho National Engineering Laboratory. LP-SB-3

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

consortium of countries under the auspices of the Organisation

for Economic Cooperation and Development (OECD).

The LP-SB-3 experiment simulated a 4.66 cm (1.84 inch) equivalent

diameter break in a cold leg pipe of a commercial PWR. The break

line piping and details of the break nozzle (0.88 cm diameter)

are shown in Figure 1. The transient was specifically designed

to achieve conditions which would allow an assessment of:

(i) The mechanism for core heat transfer under slow coolant

boil off conditions.

(ii) The effectiveness of operator initiated steam generator

feed and bleed recovery.

(iii) The effectiveness of accumulator injection when a small

pressure differential exists between the accumulator and

the primary system.

The experiment was initiated, from conditions representative of

those in a commercial PWR, by opening a valve in the Intact Loop

Cold Leg (ILCL) break piping. The primary system depressurised

rapidly until fluid saturation conditions were reached in the hot

leg at - 100 seconds. This resulted in a decrease in the primary

system depressurisation rate and, with void formation in the

fluid, a reduction in the break mass flow rate. The continued

operation of the primary coolant pumps homogenised the fluid and

the primary system void fraction steadily increased as fluid was

discharged out the break. The pumps were tripped after

1600 seconds when approximately 2800 kg of coolant inventory

remained in the primary system. After pump trip, the fluid

expelled from the break was predominantly steam. The HPIS was

not used during the experiment and the mass inventory was allowed

to decrease, uncovering the core.

As the core became increasingly void of fluid, the fuel cladding

started to heat up. The break was isolated, to prevent further

depressurisation of the primary system, when the cladding

temperature reached 835K. When the maximum cladding temperature

reached 988K, a steam generator feed and bleed procedure was

initiated. The energy removal through the secondary system

depressurised the primary system and the fuel cladding started to

cool. Accumulator flow commenced when the primary system had

depressurised to 2.79 MPa and the fuel cladding was gradually quenched, from the bottom, upwards. The LPIS, initiated at a

AEEW - R 2275 2
primary system pressure of 1.03 MPa, quickly refilled the core

region, thus terminating the experiment.

4 TRAC-PFI/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-PFI/MODI was developed for analysing

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

the calculations described in this paper was AEEW Version B02A

which contains the LANL updates to TRAC-PFl/MODl Version 12.7.

5 TRAC-PFlI/MOD1 INPUT MODEL FOR LP-SB-3

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

LOFT small break experiments was based on a TRAC-PFI/MODl large

break deck for the experiment LP-FP-I. The LP-FP-1 deck,

developed at AEEW, originated from the LANL input deck for

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

(1, 2, 3) were employed, where necessary, in producing the small

break deck.

The modifications, to the large break deck are described in

References 4, 5, 6 and 7 and included:

(i) Replacement of the three-dimensional vessel with a onedimensional representation.

(ii) Inclusion of the ILCL break.

(iii) Inclusion of the primary pump injection.

(iv) Renodalisation of the broken loop to provide a coarser

representation.

(v) Renodalisation of the ECCS to provide a coarser

representation.

(vi) Accurate repositioning of the ECCS injection point in the

ILCL.

With the exception of those modifications specifically required

to model the LP-SB-3 experiment, the input deck was identical to

that used for the TRAC-PF1/MODl analyses of LP-SB-1 (7) and

LP-SB-2 (8).

Figures 2, 3 and 4 show the nodalisation diagrams for the primary

system, the reactor vessel and the steam generator secondary

side, respectively.

A total of 41 components, 47 junctions and 182 cells were used in

the model.

A microfiche listing of the TRAC-PFl/MODl input deck for LP-SB-3

is contained in Appendix I.

AEEW - R 2275 3
6 STEADY STATE CALCULATION

AEEW Version B02A of the TRAC-PFI/MODI 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 410 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 flow) was also

run. In running the steady-state calculation, a total of

773 seconds of CPU time were used with an average time step size

of 0.096 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 speeds of the primary coolant

pumps.

The initial conditions predicted by TRAC-PF1/MODI for experiment

LP-SB-3 are compared with the measured data in Table 1. The

calculations produced stable initial conditions, within, or very

close to, the quoted experimental uncertainties, for all

significant parameters.

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 associated with analysis of the LOFT LP-SB-l

experiment, are reported (7) to be in reasonable agreement with

the available data.

7 BOUNDARY CONDITIONS FOR TRANSIENT CALCULATION

7.1 Decay Heat Data

Following reactor scram, decay heat data were specified to the

TRAC-PF1/MOD1 transient calculation by means of a "power versus

time" table. Data contained in Reference 9 were used throughout

the transient.

7.2 Primary Pump Injection

During experiment LP-SB-3, the primary coolant pump injection

system was set up to deliver a total flow of 0.12 is- 1 to the

primary coolant pumps (10). This was simulated, in the

TRAC-PF1/MOD1 model, by using "FILL" components to supply the

primary pump injection system with liquid at a constant rate of

1.63875 x 10-3 ms- 1 . The flow areas of the pump injection pipes

were 3.66131 x 10-2 m2 which implied an injection rate of

6.0 x 10-5 m3 s- 1 , or 0.06 ls-1, to each pump. The primary

coolant pump injection was inserted downstream of each pump.

7.3 Operational Setpoints

The operational setpoints (for reactor scram, main feedwater

shut off, MSCV closure and opening, primary coolant pump trip,

break valve closure, steam bypass valve opening, main feedwater

AEEW - R 2275 4
initiation, accumulator injection initiation and LPIS initiation)

measured during the experiment, and the way in which the

setpoints were specified in the TRAC-PFI/MODl calculation, are

given in Table 2.

8 TRANSIENT CALCULATION

8.1 Introduction

The transient calculation was restarted from the end of the

transient mode steady state calculation by opening the VALVE

component in the ILCL break piping. As for the steady state

calculations, AEEW Version B02A of TRAC-PFI/MOD1 - which

incorporates the code updates contained in LANL Version 12.7 -

was used for the transient 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-3

"Comparison Report" (11). For discussion purposes, results are

considered separately for each of the three distinct phases of

the transient ie mass depletion phase, core boil-off phase and

core cool-down and recovery phase.

8.2 CPU Usage and Time Step Behaviour

6400 seconds of elapsed transient were calculated, requiring

4702 seconds of CRAY-XMP CPU time (see Figure 5). This

corresponds to a CPU/real time ratio of 0.73. The userspecified minimum allowable time step throughout the calculation

was limited to 1 x 10-5 seconds. At - 300 seconds, the code

started to take very small time steps and it was found necessary

to reduce the user-specified maximum allowable time step size

from 0.5 seconds to 0.2 seconds for a short period, during which

the calculation "recovered" (see Figure 6). The average time

step for the calculation (problem time/total number of time

steps) was 0.32 seconds.

8.3 Chronology of Events

A comparison of the measured and predicted timings of significant

events is given in Table 3.

Experiment LP-SB-3 was initiated by opening the break valve in

the ILCL break line. After 9.2 seconds, the primary system had

depressurised to the reactor scram set point of 14.19 MPa. The

steam generator control system responded to the reactor scram by

isolating the feedwater flow to the steam generator and

preventing the steam flow out of. the steam generator. The main

feedwater pump was shut off at 9.4 seconds and the feedwater was

isolated by 10.8 seconds. The Main Steam Control Valve (MSCV)

started to close at 9.5 seconds and took - 11 seconds to become

fully closed. The timings of these initial events were

predicted, by the TRAC-PFI/MODI calculation, to within -

2 seconds.

AEEW - R 2275 5
Following reactor scram, the secondary system pressure increased

sharply due to the isolation of the feedwater and the MSCV

closure. In order to keep the secondary system pressure between

6.35 MPa and 7.09 MPa, the MSCV cycled four times, during the

experiment, between 87.5 seconds and 1030 seconds. TRAC-PFl/MODl

predicted the MSCV to cycle only twice, at 27 seconds and at

124 seconds.

TRAC-PFl/MODl correctly predicted the time at which fluid

saturation occurred in the primary system ie at - 100 seconds.

An abrupt degradation in the differential pressure across the

primary coolant pumps was observed at 875 seconds, in the

experiment. Pump degradation was predicted-to occur between

700 seconds and 1100 seconds. In both the experiment and the

calculation, the primary coolant pumps were tripped at 1600

seconds. This resulted in the break uncovering at 1612 seconds

in the experiment and at - 1625 seconds in the calculation.

The start of core heat up (in the upper part of the core) caused

by the slow boiling off of liquid was detected, in the

experiment, at - 3800 seconds. This was predicted, by the

TRAC-PFl/MODl calculation, to commence - 260 seconds later. When

the highest indicated cladding temperature reached 835.2K, the

break was isolated. This occurred, some 50 seconds later than

predicted, at 4742 seconds.

The cladding temperature continued to rise until the highest

indicated value reached 988K, at which time the secondary side

feed and bleed operation was initiated by fully opening the main

steam control bypass valve. This action, which .occurred at

5415 seconds in the experiment and was predicted, by the

TRAC-PFl/MODI calculation, to occur at 4964 seconds, caused a

very rapid cooldown and depressurisation of the secondary system.

The maximum fuel clad temperature was measured at 5422 seconds

and predicted to occur at 4980 seconds.

In the experiment, the primary system pressure reached the

accumulator setpoint (2.79 MPa) at 5588 seconds. In the

TRAC-PF1/MODI calculation, accumulator injection was predicted to

commence at 5308 seconds. The LPIS was initiated (at a primary

system pressure of 1.03 MPa) at 6785 seconds in the experiment;

the core quickly refilled and the experiment was terminated at

6845 seconds. The calculation was terminated at 6400 seconds

when accumulator nitrogen was predicted to have entered the

system.

A series of pictures showing the predicted void fraction

distribution, the liquid and vapour velocities and the

occurrences of stratified flow conditions, throughout the system,

is presented in Appendix II.

8.4 Mass Depletion Phase

The mass depletion phase is defined as that part of the transient

between break initiation (0 seconds) and pump trip

(1600 seconds).

AEEW - R 2275 6
8.4.1 Primary and Secondary System Pressures

A comparison of the measured and predicted primary system

pressure histories is shown in Figure 7. The initial, rapid

subcooled depressurisation is well represented and the time at

which fluid saturation conditions occur (' 100 seconds) is also

well-predicted. Following the end of subcooled depressurisation,

and prior to pump trip, (100-1600 seconds) the overall primary

system pressure history is well reproduced; detailed

characteristics of the primary system pressure behaviour,

resulting from the cycling of the secondary side MSCV, are not

represented exactly, however.

Following isolation of the main feedwater and closure of the

MSCV, the rate at which the secondary system pressurises is

slightly over-predicted, as shown in Figure 8, and the cycling of

the MSCV is not accurately reproduced. During the experiment,

the MSCV cycled four times, between 87.5 seconds and

1030 seconds; it was predicted, by the TRAC-PFl/MODI calculation,

to cycle only twice - at 27 seconds and at 124 seconds. It can

be seen in Figure 8 that the setpoints at which the MSCV was

opened and closed were not consistent throughout the experiment.

Values of 7.09 MPa and 6.35 MPa (10), respectively, were

specified to the TRAC-PFl/MODl calculation. A further factor,

complicating simulation of this part of the transient, is the

leakage rate through the "closed" MSCV which is not specified

precisely in the experiment documentation. In the calculation,

to account for steam leakage, the minimum MSCV flow area was

restricted to 1.0% of its fully opened value. This corresponded

to a leakage rate of - 0.12 kgs- 1 at 7 MPa and is of the order

implied by Reference 10. As shown in Figure 9, the overall trend

in secondary system pressure history is reasonably well

reproduced and, at the time at which the primary coolant pumps

trip (1600 seconds), the predicted secondary side pressure is

- 0.1 MPa greater than the measured value.

Calculations presented by members of the OECD LOFT Program Review

Group also failed to reproduce precisely the experimental MSCV

cycling behaviour and, in line with the TRAC-PFl/MODl

calculation, slightly over-predicted the secondary system

pressure just prior to pump trip (1600 seconds).

8.4.2 Break Mass Flow Rate, Break Line Density and Primary System

Mass Inventory

A comparison of the measured and predicted break mass flow rates

is shown in Figure 10. Although well-predicted for the first

- 500 seconds of the transient, TRAC-PFI/MODl slightly

over-predicts the break mass flow rate after this time, until the

primary coolant pumps are tripped at 1600 seconds. Figure 11

indicates that the break line density is well reproduced during

the mass depletion phase, suggesting that the slightly overpredicted break mass flow rate between - 500 seconds and

1600 seconds is a consequence of the TRAC critical flow model.

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

throughout the calculation. (It appears that a multiplier of

AEEW - R 2275 7
- 0.8 would have given the best agreement over the 500 seconds to

1600 seconds period.) The break mass flow rate predicted by

TRAC-PFl/MODl (between - 500 seconds and 1600 seconds) is - 10%

higher than that implied by the Homogeneous Equilibrium Model

(12). Details are given in Appendix III. Since no ILCL

stratification, prior to pump trip, is observed in the

experiment, serious deficiencies in the TRAC-PFl/MODI branch

offtake model (highlighted during analysis of the LP-SB-I (7) and

LP-SB-2 (8) experiments) are not manifested in this calculation.

Consistent with the over-predicted break mass flow rate, the rate

at which the primary system mass inventory is predicted to

decline is more rapid than that measured (see Figure 12). It is

thought that the discrepancy between the measured and predicted

initial primary system mass inventories, as shown in Figure 12,

is caused by a difference in interpretation, between the

experiment and the calculation, of what constitutes the "primary

system". The predicted initial primary system mass inventory

includes all system components except the secondary side of the

steam generator component (component numbers 20, 21, 22 and 27),

the steam line valve (component number 23) and the ECCS upstream

of the valve in the accumulator line. A similar approach was

adopted for the TRAC-PFl/MODI calculations of LP-SB-1 (7) and

LP-SB-2 (8), in which the initial primary system mass inventories

were over-predicted slightly (by - 150 kg). The difference

between the initial mass inventory predicted for LP-SB-3 and

those predicted for LP-SB-1 and LP-SB-2 results mainly from the

inclusion, for LP-SB-3, of the broken loop steam generator and

pump simulators and a larger part of the ECC line. The initial

experimental mass inventory on the LP-SB-3 experiment data tape

is only slightly (- 100 kg) greater than that for LP-SB-l and

LP-SB-2 (13), despite the inclusion, in the LP-SB-3 experiment of

the broken loop simulator section.

The break mass flow rates predicted by members of the OECD LOFT

Program Review Group are, generally, in good agreement with the

experimental data. These calculations used subcooled choked flow

multipliers of 0.93 to 1.0 and two-phase choked flow multipliers

of 0.7 to 0.81.

8.4.3 Primary System Densities

A comparison of the measured and predicted ILHL and ILCL

densities are shown in Figures 13 and 14. The times at which

TRAC-PFl/MODl predicts stratification to occur are also

indicated. The ILHL density is very well-predicted by the

TRAC-PFl/MODl calculation; the density in the ILCL is slightly

over-predicted. Stratification is predicted in both legs when

the primary coolant pumps are tripped at 1600 seconds.

The calculations performed by members of the OECD LOFT Program

Review Group (11) consistently over-predicted the ILCL density

and, to a lesser extent, the ILHL density.

AEEW - R 2275 8
8.4.4 Pumps Degradation

The measured and predicted differential pressures across the

primary coolant pumps are shown in Figure 15. The differential

pressure decreases steadily after fluid saturation conditions are

reached due to void formation in the primary system. During the

experiment, oscillations in the pump differential pressure -

associated with a change from symmetric flow through both pumps

to asymmetric flow through only one pump - were observed between

- 700 and - 875 seconds (10). TRAC-PFl/MODl, in common with

other codes (11), was not able to reproduce this situation. The

pump characteristics employed in the LP-SB-3 input deck were

those reported for the TRAC-PFl/MODI calculation of LP-SB-2

(8 (calculation "A")). In the experiment, pump degradation was

considered to have occurred at 875 seconds; TRAC-PFl/MODl

predicted this to occur between 700 seconds and - 1100 seconds.

The calculations performed by members of the OECD LOFT Program

Review Group (11) tended to over-predict the time at which the

pumps degraded (by up to - 700 seconds).

8.5 Core Boil-Off Phase

The core boil-off phase is defined as that part of the transient

between pump trip (1600 seconds) and steam generator feed and

bleed initiation (5415 seconds - experiment, 4964 seconds -

calculation).

8.5.1 Primary and Secondary System Pressures

As shown in Figure 7, following primary coolant pump trip, the

primary system pressure history is well-predicted prior to the

time at which the break is isolated (4742 seconds - experiment,

4688 seconds - calculation). Figure 16 shows the measured and

predicted differences between primary and secondary system

pressures. Figure 17 shows the primary and secondary system

pressure histories as predicted and as measured. Although the

trends in primary and secondary system pressure histories are

reasonably well reproduced by the TRAC-PFl/MODl calculation, the

measured and predicted results (most noticeably for the secondary

system) appear to diverge slightly after - 3500 seconds. The

predicted difference between the primary and secondary system

pressures, immediately prior to break isolation, is significantly

larger than that measured. Following break isolation, the

primary side pressure rises but is predicted to remain below that

of the secondary system until the secondary side feed and bleed

is initiated. This represents an important departure from the

experiment in which, following break isolation and prior to feed

and bleed initiation, there was a - 500 second period when the

primary system pressure was above that of the secondary system

during which reflux condensation could occur. This difference

between the measured and predicted relative primary and secondary

system pressures has important consequences for the rate at which

the core heats up, as discussed in Section 8.5.4.

AEEW - R 2275 9
8.5.2 Break Mass Flow Rate and Primary System Mass Inventory

Following pump trip (at 1600 seconds) the break uncovers and, as

shown in Figure 10, the break mass flow rate is well-predicted up

until the time at which the break is isolated (4742 seconds - experiment, 4688 seconds - calculation).

As shown in Figure 12, the residual primary system mass

inventory, predicted by TRAC-PFl/MODl, at the time of pump trip

(1600 seconds) is in good agreement with the experimental value

of - 2800 kg. It should be recognised, however, that the

agreement is, to some extent, fortuitous because the overpredicted break mass flow rate prior to this time and the overpredicted initial inventory act as compensating errors. The

residual mass (- 1500 kg) at the time at which the break is

isolated is well-predicted and reflects the accurate reproduction

of the break mass flow rate following pump trip. In line with

the experimental procedure, primary pump injection was turned off

in the calculation following pump trip; the primary system mass

inventory therefore remained constant following break isolation

until the accumulator flow was initiated.

8.5.3 Primary System Densities

Following pump trip, the liquid in the system falls to the lowest

parts of the circuit, ie the reactor pressure vessel and the loop

seals. As shown in Figure 13, the hot leg density rises

immediately after the pumps are tripped due to liquid draining

back from the "up-side" of the steam generator tubes. In the

experiment, the instrument measuring the loop seal density

(positioned in the piping leading from the steam generator

outlet) becomes immersed in liquid following pump trip and

remains so until the end of the transient. In the calculation,

the presence of liquid is not predicted in this part of the

circuit until - 3200 seconds (although liquid is present in the "pump side" of the loop seal prior to this time)- *see

Appendix II. Reasons for this discrepancy have-not been fully

investigated. However, factors which might adversely influence

the predicted behaviour have been identified, ie incorrect

pressure imbalance across the loop seal and insufficiently

detailed nodalisation of the vertical sections of the loop seal.

In general, the calculations submitted by members of the OECD

LOFT Program Review Group (11) correctly predicted the loop seal

density following pump trip (but over-predicted the density prior

to pump trip).

8.5.4 Fuel Cladding Temperatures

The predicted peak fuel rod cladding temperatures (at each of the

six levels implied by the TRAC-PFl/MODl core noding scheme) are

compared with the experimental data (at the nearest corresponding

elevation) in Figures 18-23. A fuel rod power peaking factor of

1.6 (as implied in an AEEW input deck for LOFT test L2-6) was

assigned to the peak rated rod in the TRAC-PFl/MODl calculation.

The following observations are made:

AEEW - R 2275 10
(i) The time at which the core starts to heat up is overpredicted by '- 260 seconds (see Figure 23). The times at

which the break is isolated (peak clad temperature of 835.2

K) and the steam generator feed and bleed is initiated

(peak clad temperature of 988 K) are under-predicted by

- 50 seconds and - 450 seconds, respectively.

(ii) The predicted rate at which the core heats up is greater

than that measured. Agreement is considerably better at

the top of the core (see Figure 23) than at lower

elevations and agreement worsens following break isolation

(see, for example, Figure 20).

(iii) As measured experimentally, no heat-up is predicted at the

bottom of the core (see Figure 18). The heat-up at the

0.28 m elevation, observed in the experiment, was not

predicted by the TRAC-PFl/MODl calculation in the cell

representing elevations from 0.15 m to 0.42 m (see

Figure 19).

The above observations are consistent with those presented, by

members of the LOFT Program Review Group, in the LP-SB-3

"Comparison Report" (11). It is widely reported (11, 14, 15)

that an important factor, contributing to the over-predicted

.- rates of care heat-up, is the inability of the codes, using a

,,:l-dimensional vessel representation, to correctly predict the

distribution, throughout the core, of reflux condensation

returning to the vessel. However, in the TRAC-PFl/MODI analysis,

an important discrepancy exists between the measured and

predicted differences in primary and secondary pressure. During

the core heat-up phase in the calculation, the primary side

pressure is below that of the secondary and it is unlikely,

therefore, for the presence of reflux condensation to be

predicted. It is suggested, therefore, that in the TRAC-PFI/MOD1

calculation, it is the lack of reflux condensation, rather than

its poorly predicted distribution throughout the core, that might

lead to the heat-up rate being over-predicted. This certainly

appears to be the case following break isolation when, in the

experiment, the primary side pressure becomes greater than that

of the secondary and there is a significant reduction in the rate

at which the core continues to heat up. In the calculation, the

primary side pressure does not exceed that of the secondary until

the steam generator feed and bleed is initiated and no reduction ..in the rate of heat up is predicted following break isolation.

Although not fully investigated, it is thought that the tendency

of TRAC-PFl/MOD1 to under-predict the core density, during

certain conditions, due to deficiencies in its interphase drag

model (16) could result in the core drying out at a greater rate

than that measured and might contribute to the rate of core

.heat-up being over-predicted.

8.6. Core Cool-Down and Recovery Phase

The core cool-down and recovery phase is defined as that part of

the transient between steam generator feed and bleed initiation

AEEW - R 2275 1i
(5415 seconds - experiment, 4964 seconds - calculation) and

experiment termination (6845 seconds).

8.6.1 Primary and Secondary System Pressures

As shown in Figure 9, steam generator feed and bleed initiation

results in a rapid decrease in secondary system pressure. The

corresponding decrease in primary system pressure is shown in

Figure 7. The rates at which the primary and secondary system

pressures are predicted to decline are slightly under-predicted.

Sensitivity of the pressure histories to the size assumed for the

secondary side steam bypass valve, used during feed and bleed

operation, and the consequent effect on the accumulator flow

rate, is discussed in Section 8.6.3.

8.6.2 Fuel Cladding Temperatures

Following steam generator feed and bleed initiation, the core

takes slightly longer to quench in the calculation then in the

experiment. As shown in Figures 20, 21 and 22, the enhanced rate

with which the core is predicted to cool, following the

commencement of accumulator flow (- 5300 seconds) is very well

reproduced. However, in the calculation, as a result of the

slightly under-predicted primary side depressurisation rate,

there is a relatively longer period (approximately twice as long)

between feed and bleed initiation and accumulator flow

commencement (during which the rate of cooling is less rapid).

8.6.3 Accumulator Injection Flow

Accumulator flow is predicted to commence - 280 seconds earlier

than in the experiment. As shown in Figure 24, its behaviour is

reasonably well reproduced. The oscillatory accumulator flow,

reported in other calculations (11),- is not observed in the

TRAC-PFl/MODI analysis. Previously reported peaks in the

accumulator flow rate predicted by TRAC-PFl/MODl were found to be

the result of incorrect mass flow rates being attributed, by

TRAC, to "signal variables" used for plotting. (The problem

arises when the void fraction in the bottom cell of the

accumulator is % 0.0).

At - 6400 seconds, accumulator nitrogen was predicted to have

entered the system and the calculation was terminated. Finer

nodalisation of the bottom of the accumulator (to ensure a = 0.0

in the bottom cell) could have prevented this situation.

The size of the steam bypass valve used during the feed and bleed

operation was not specified in the available documentation. The

effect, on the secondary side depressurisation rate and the

accumulator flow rate, of increasing the valve area by - 70%

(from 2.46 x 10-4 m2 to 4.093 x 10- m2 ) is shown in Figures 25

and 26. The larger flow area produced a more rapid secondary

side depressurisation and, consequently, a greater accumulator

flow rate, in better agreement with the experimental data.

AEEW - R 2275 12
9 CONCLUSIONS

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

experiment LP-SB-3 were as follows:

Mi) The calculation ran with a CPU to real time ratio of 0.73

and an average time step of 0.32 seconds.

(ii) Despite the break line density being well-predicted,

TRAC-PFl/MODI slightly over-predicted the break mass flow

rate prior to pump trip. This is thought to be a

consequence of deficiencies in the TRAC critical flow model

- the predicted break mass flow rate during this period was

10% greater than that implied by the Homogeneous

Equilibrium Model. In general, the calculations reported

(by members of the OECD LOFT Program Review Group) in the

LP-SB-3 "Comparison Report" correctly reproduced the

experimental break mass flow rates by applying appropriate

break flow multipliers (of < 1.0). (Subcooled and twophase choked flow multipliers of 1.0 were used throughout

the TRAC-PFl/MOD1 calculation).

(iii) Overall trends with respect to primary and secondary system

pressure histories were reasonably well-predicted.

Differences between the experiment and the calculation in

the relative magnitudes of the primary and secondary system

pressures, however, resulted in conditions for reflux

condensation not being predicted during the core heat-up

phase of the transient. This was contrary to the

experimental behaviour and was thought to contribute -

particularly following break isolation - to the predicted

rate of core heat-up being significantly greater than that

observed experimentally. Although not fully investigated,

under-prediction of the core density, due to deficiencies

in the TRAC-PFl/MODl interphase drag model, has been

identified as a potential contributor to the poorly

predicted rate with which the core temperature rises. The

calculations presented in the LP-SB-3 "Comparison Report"

also significantly over-predicted the rate of core

heat-up.

(iv) Minimum primary system mass inventory was reasonably well

reproduced. As observed in the experiment, no uncovery of

the bottom of the core was predicted by the TRAC-PFl/MODI

calculation.

(v) The predicted time between steam generator feed and bleed

initiation and the core being quenched was slightly longer

than that observed experimentally. The main reason for

this was thought to be a delay, relative to the experiment,

in the time at which the accumulator flow was predicted to

commence. This was caused by the predicted rate of

depressurisation during feed and bleed operation being

slower than that measured and was shown to be rectified

when the flow area of the steam bypass valve used during

the feed and bleed operation (not specified in the

available documentation) was increased.

AEEW - R 2275 13
(vi) During the core recovery phase, the mass flow rate from the

accumulator was reasonably well reproduced. Severe

oscillations in the accumulator flow, reported in the

calculations performed by members of the OECD LOFT Program

Review Group, were not observed in the TRAC-PFl/MODl

calculation.

(vii) Accumulator nitrogen was predicted to have entered the

circuit at - 6400 seconds and the calculation was

terminated at this time, ie before initiation of the Low

Pressure Injection System. The bottom cell of the

accumulator was partially full at this time and finer

nodalisation of the bottom of the accumulator could have

prevented the premature ingress of nitrogen.

The calculation reported should be considered a "Base Case", the

results of which have not been exhaustively analysed. With the

exception of the rate at which the rod cladding temperatures rose

during the core uncovery phase of the transient, agreement with

the experimental data was considered reasonable. A fuller

investigation of factors influencing the rate with which the core

heats up (including the effect of deficiencies in the

TRAC-PFl/MODl interphase drag model) would be desirable.

10 ACKNOWLEDGEMENTS

The guidance of Dr C G Richards during the course of this work is

gratefully acknowledged.

11 REFERENCES

1 REEDER D L, LOFT System and Test Description, July 1978,

(TREE-1208, NUREG/CR-0247).

2 KEE K J, SCHALLY P J, WINTERS L, Base Input for LOFT RELAP5

Calculations, July 1980 (EGG-LOFT-5199).

3 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).

4 ALLEN E J, Internal Document.

5 ALLEN E J, Internal Document.

6 NEILL A P, Internal Document.

7 ALLEN E J, TRAC-PFl/MODI Post-Test Calculations of the OECD

LOFT Experiment LP-SB-l (AEEW*- R 2254).

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

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

9 McPHERSON G D, Decay Heat Data for OECD LOFT Experiments,

October 1985 (Letter to OECD LOFT Programme Review Group

Members).

AEEW - R 2275 14
10 ALEMBERTI A, PROTO G, Experiment Analysis and Summary

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

December 1985 (T-20000-XX-140-020).

11 ALEMBERTI A, Comparison Report on OECD LOFT Experiment

LP-SB-3, August 1986 (T-20000-XX-140-021).

12 HALL, CZAPARY, Tables of Homogeneous Equilibrium Flow

Parameters for Water in SI Units, September 1980

(EGG-2056).

13 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).

14 HARWOOD C, BROWN G, RELAP5/MOD2 Calculation of OECD LOFT

Test LP-SB-03, April 1986 (GD/PE-N/535)..

15 GUNTAY S, RELAP5/MOD2 Assessment: OECD-LOFT Small Break

Experiment LP-SB-3, Gemeinschaftsberict Nr 13, April 1986.

16 ARDRON K H, CLARE A J, Assessment of Interphase Drag

Calculations in the RELAP5/MOD2 and TRAC-PFl/MOD1 ThermalHydraulic Codes, November 1986 (GD/PE-N/557 (Revised)).

17 Safety Code Development Group, Energy Division, LANL,

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

for Pressurised Water Reactor Thermal-Hydraulic Analysis,

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

AEEW - R 2275 15
TABLE 1

INITIAL CONDITIONS FOR EXPERIMENT LP-SB-3

PRIMARY COOLANT SYSTEM

Measured (10)

50.3 ± 1.2

482.6 ± 2.6

Power Level (MW)

Primary Coolant Mass Flow

Rate (kgs- 1 )

Pressurizer Pressure (MPa)

Pressurizer Liquid Level (m)

ILCL Temperature (K)

ILHL Temperature (K)

STEAMIGENERATOR SECONDARY

SIDE

Liquid Level (m)

Pressure (MPa)

Liquid Temperature (K)

Mass Flow Rate (kgs- 1 )

ECCS - ACCUMULATOR A

15.26

1.115

556.6

576.6

0.21

5.58

535.6

26.67

2.84

303.7

2.04

±

±

±

±

0.11

0.06

1

1

Predicted

50.3

482.2

15.20

1.15

558.1

577.6

0.21

5.57

536.0

26.40

2.84

303.7

2.04

+ 0.005

± 0.06

± 3.6

± 0.77

± 0.1

±6

± 0.01

Pressure (MPa)

Temperature (K)

Liquid Level (m)

AEEW - R 2275
TA1BLE 2

OPERATIONAL SETPOINTS TOR UMENT LP-SB-3

Measured During Experiment Specified to TRAC-PFI/MODl

(10) Calculation

Action

Reference

Reference Setpoint (component/ Setpoint cell no)

Small Break

Valve Opened

Reactor Scrammed

Main Feedwater

Shut Off

MSCV Begins to

Close

MSCV Begins to

Open

Primary Coolant

Pumps Tripped

Break Valve

Closed

Steam Bypass Valve Opened

Main Feedwater

Initiated

Accumulator

Injection

Initiated

LPIS Initiated

Time

ILHL Pressure

ILHL Pressure

Secondary System

Pressure

Secondary System

Pressure

Primary Coolant

System Mass

Inventory

Remaining

Maximum Cladding Temperature

Maximum Cladding

Temperature

Maximum Cladding Temperature

IUH Pressure

IL=L Pressure

0 secs

14.19 MPa

14.19 MPa

6.35 MPa

7.09 MPa

2800* kg

835.2K

988K

988K

2.79 MPa

1.03 MPa

Time

ILHL Pressure

(1/4)

ILHL Pressure

(1/4)

Initially:- ILHL

Pressure (1/4)

During MSCV

Cycling:-

Secondary

Pressure (21/2)

Secondary

Pressure (21/2)

Time

Maximum Cladding

Temperature of'

Supplementary Rod

Maximum Cladding

Temperature of

Supplementary Rod

Maximum Cladding

Temperature of

Supplementary

Rod

ILHL Pressure

(1/4)

ILHL Pressure

(1/4)

0 secs

14.19 MPa

14.19 MPa

+0.2 secs

14.19 MPa

+0.29 secs

6.35 M4Pa

7.09 MPa

1600 secs

835.2K

988.OK

988.OK

2.79 MPa

1.03 MPa

A A J. J

* It had been intended to trip the primary coolant pumps when 2000 kg of primary

mass inventory remained.

AEEW - R 2275
'IBLE 3

CIMRlEMMGY OF EVEI' FOR '1PERII2T LP-SB-3

Time After Experiment

Event Initiation (seconds)

Measured (10) Predicted

Small break valve opened 0.0 0.0

Reactor Scrammed 9.21 ± 0.01 11.22

Main feedwater pump shut off on scram signal 9.41 ± 0.01 11.42

signal

MSCV started to close 9.5 ± 0.2 11.52

Main feedwater valve isolated 10.81 ± 0.01 12.22

MSCV closed 21 ± 0.2 24.7

Pressurizer level below indicating range 67 ± 3 84.41

First time MSCV cycled opened 87.5 ± 0.2 26.7

Subcooled blowdown ended 98.5 ± 0.5 100

Pumps differential pressure degradation 875 ± 3 700 - 11002

observed

last time MSCV cycled open 1030 ± 0.2 124.4

Primary coolant pumps tripped 1600 ± 2 1600

Break uncovered 1612 ± 5 1624

Primary coolant pumps decoupled from 1626 ± 1 1650

flywheels

Start of core heat up 3800 ± 50 40603

Break isolated 4742 ± 2 4688

Steam generator feed and bleed initiated 5415 ± 5 4964

Maximum cladding temperature reached 5422 ± 1 4980

Accumulator injection initiated 5588 ± 3 5308

Cbre quenched 5800 ± 50 55604

LPIS initiated 6785 ± 2 -

Experiment terminated 6845 ± 2

1 Defined as the time at which the liquid level in the pressurizer 4 0.01 m.

2 As implied by Figure 15.

3 Defined as the time at which Tmax peak rod - Tsat > 2K.

4 Defined as the time at which Tmax peak rod - Tsat 4 2K.

AEE3 - R 2275
APPENDIX I

MICROFICHE LISTING OF THE TRAC-PFI/MOD1 INPUT DECK

FOR LP-SB-3

The microfiche can be found on the inside of the back

cover of this report.

APPENDIX II

SERIES OF PICTURES SHOWING PREDICTED SYSTEM CONDITIONS

THROUGHOUT THE LP-SB-3 TRANSIENT

The following series of pictures shows, at selected 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 2, 3 and 4.

AEEW - R 2275

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

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 0.0

VLIQ --

(10.0)

VVAP --

(10.0)

VOID FRACTION

Pi .8

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TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 107.9

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WVA? -ý

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

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 210.4

VLIQ -- k

(10.0)

WVA --

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VOID FRACTION

<= 0.88

I

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TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 300.6

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TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 500.4

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F-F17
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S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 210.4

VLIQ --

(10.0)

vvAP--

( 10.0)

VOID FRACTION

I':= 0~"'. 88,!hi::•"'

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

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 300.6

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( 10.0)

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(10.0)

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F--.7-]
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 500.4

(10.0)

VVAP-+

(10.0) m

VOID FRACTION

.11
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 1000.'

VLIQ --

(10.0)

WAP ---

(10.0)

VOID FRACTION

1
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION

4.

Ti 1513.4

VLIQ -- +

( 10.0)

VVAP -- +

( 10.0)

f •

VOID FRACTION

I; 0.A
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION

.-

Ti 2002.(

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( 10.0)

VVAP --- *

( 10.0)

VOID FRACTION

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

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 2514.

iVLIQ -

( 10.0)

WAP ---

( 10.0)

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

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 3024.f

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(10.0)

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10.0)

VOID FRACTION

<;i• 0.88•iii•;i
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 3505.2

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(10.0)

VVAP ---

(10.0) 7ZZZZZZZJ

1.4

VOID FRACTION

F--018

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

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 4000.7

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(10.0)

VVAP -- +

( 10.0)

W4

VOID FRACTION

I~1* * * ~
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 4515.

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(10.0)

VVAP ----

(10.0)

VOID FRACTION

I

A:.~'.:i:
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 5013.(

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(10.0)

VVAP -4

(10.0)

77-71

VOID FRACTION

IT

-4
S.M.A.R.T. SYSTEM MIMIC FOR ANALYSIS OF REACTOR TRANSIENTS

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION Ti 5500.E

VLIQ --

( 10.0)

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( 10.0)

[1 177

I

*~ *~.-

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

TITLE OF FRAME:- LOFT TEST LP-SB-3 TRAC CALCULATION

FTi 6019.5

VLIQ --

(10.0)

WAP -- +

( 10.0)

VOID FRACTION
APPENDIX III

COMPARISON OF BREAK MASS FLOW RATE PREDICTED BY TRAC-PFlI/MOD1

WITH THAT IMPLIED BY THE HOMOGENEOUS EQUILIBRIUM MODEL

As described in Section 8.4.2, although well-predicted for the

first - 500 seconds of the transient, TRAC-PF1/MOD1 slightly

over-predicted the break mass flow rate after this time, until

the primary coolant pumps were tripped (and the break uncovered)

at 1600 seconds. The fluid conditions in the last cell of the

break line, at 1000, 1200 and 1400 seconds, are given in Table

AIII.l. The predicted break mass flow rates at these times are

compared with those implied by the Homogeneous Equilibrium Model

(HEM) (12) in Table AIII.2. The TRAC-PFl/MODI predicted values

are found to be - 10% higher than the HEM figures. Figure AIII.l

- taken from the TRAC Code'Manual (17) - confirms that

TRAC-PFl/MOD1 is expected to over-predict mass fluxes (under

these conditions) when compared with the Homogeneous Equilibrium

Model.

un~om

120000.

E

800000.

GO=

yI00

TRAC-PF1

. .urne (RELAP4/MOD6)

HEM

(Tuq=560 K, Psat=7.1 MPa)

40000-

7 7 L4.107

2fvvýnWHO I 8.0.106 I.040 1.20

Stagnation Pressure (Pa)

~.1I•,,,07

FIGURE AIII.I

AEEW - R 2275
TABLE AIII.1

FLUID CONDITIONS AT END OF BREAK LINE

Time Pressure Temper- Liquid Vapour Void Liquid Vapour Average- Quality

ature Density Density Fraction Velocity Velocity Density

(seconds) (MPa) (K) (kgm- 3 ) (kgm- 3 ) (ms- 1 ) (ms-I) (kgm- 3 )

I

1000 7.14616 560.4 738.5 37.37 0.3350 54.05 56.85 503.62 0.0261

1200 7.11658 560.1 739.0 37.20 0.4389 61.20 64.07 ' 530.98 0.0396

1400 7.07897 559.8 739.7 36.98" 0.4969 66.16 69.08 ' 390.52 0.0490
:P

TABLE AIII.2

Break Mass Break Mass % by which

Time Flow Rate Flow Rate TRAC-PFI/MODI

Predicted by Implied by over-predicts

TRAC-PFl/MODI HEM break mass flow compared with HEM

(seconds) (kgs- 1 ) (kgs- 1 )

1000 1.6578 1.4996 9.5

1200 1.6070 1.4386 10.5

1400 1.5748 1.4028 10.9

FIGURES

In Figures 7-26, the following line types are used:

TRAC-PFI/MODI Calculation

------------ Experimental Data

AEEW - R 2275

Inlicl loop Intact loop roken Wtoo

Mr

tL1

A

1

2.64 cm

0.88 cm

T (0 5 .3 in .)

Break nozzle insert

To DST

Figure 1 LOFT configuration for cold leg intact loop small

break Experiment LP-BB-03
Component No Component Name Component No

1

8

2

3

4

61

5

Intact Loop Hot Leg

Pressurizer

Steam Generator

Pump Suction

Pump No 2

Pump No 2 Injection Tee

Pump No 1

63

6

71

7

13

14

15

Component Name

Pump No 1 Injection Tee

Pump Discharge

ILCL, Break Line + Nozzle

ILCL + ECC Line

ECC Line

Accumulator Line Valve

Accumulator

8

PRESSURIZER 2 ACCUMULATOR

STEAM. GENERATOR 15

i

z

PUMP INJECTION

n 61

7

1 ILML

PUMP SUCTION

14

3 INJECTION

BREAK LINE

ECC LINE

FIGURE 2 ACCUMULATO

LINE VALVE PRIMARY SYSTEM NODALISATION DIAGRAM FOR LP-SB-3

(a) INTACT LOOP
Component No Component Name

31 Broken Loop Hot Leg

41 Broken Loop Cold Leg

~2ZZ~~RABV

41

S

BLCL

.: 2 : 3 I :5:6

FIGURE 2

PRIMARY SYSTEM NODALISATION DIAGRAM FOR LP-SB-3

(h~ TU~flTCF.M T.OOP

(b) BROKEN LOOP
Vessel Representation for "System Mimic" 79 (See Appendix 11)

Vessel Representation for "System Mimic"

80

32

81

I

80

lIIIL

!

84

87

2

I

82 -q--

Component No

79

80

81

82

83

84

85

86

87

88

89

90

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

1

35

2

1

86

2 I

5

5

6

H

4

5

6

7

0

1

86

2

3

4

5

6

7

8

1

85

2

3

3

4

5

2 6

I

It 2

90

4

FIGURE 3 - I 

TRAC-PFL/MOD1 REACTOR VESSEL NODALISATION DIAGRAM FOR LP-SB-3
MAIN STEAM LINE

21 MAIN FEEDOATER

27

22 .. AUX FEEDVATER

23

Cells Numbered Compone

14,19 20

28,29,31 21

21-24,26 22

I-6 23

27

FIGURE 4

it No Component Name

Boiler

Steam Dome

Downcomer

Steam Line Valve

Main and Auxiliary

Feedwater Tee

TRAC-PF1/MODl STEAM GENERATOR SECONDARY SIDE NODALISATION DIAGRAM FOR LP-SB-3
THE FOLLOWING ARE PLOTTED AGAINST REACTOR TIME

rTOTAL CPU TIME t.!1

tTj

6000

5500[

1WInfrithi

C-)

w

U)

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

REACTOR TIME , SECONDS

FIGURE 5

LP-SB-3 CALCULATION

CPU USAGE
TIHE FOLLOWING ARE PLOTTED AGAINST REACTOR TIME

TIME STEP SIZE

1Winfrithl

14)

-4

1.

0.8

0.6

z

0

u

0.4

0.2

0.0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

REACTOR TIME , SECONDS

FIGURE 6

LP-SB-3 CALCULATION

TIME STEP SIZE
THE FOLLOWING ARE PLOTTED AGAINST TIME

PE-PC-002 ,PRESSURE

lwinfrith

-4

t'

0. 20E

0. 18E

0. 16E

0. 14E

I0. 12E

* 0. fOE

0. 80E

0. 60E

0. 40E

0. 20E

0. OOE

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME , SECONDS

FIGURE 7

LP-SB-3 CALCULATION

PRIMARY SYSTEM PRESSURE
THE FOLLOWING ARE PLOTTED AGAINST TIME

ulj

1Winfrithl

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

8000000

7500000

7000000

* 6500000

*

-N

6500000

6000000

5500000

5000000

300

TIM-CE , SECONDS

FIGURE 8

LP-SB-3 CALCULATION

SECONDARY SYSTEM PRESSURES (0-300 SECONDS)
THE FOLLOWING ARE PLOTTED AGAINST TIME 1Wlnfr'ithi

t:'1

t~3

-4

U'

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

8000000

7000000

6000000

I I I I I I I I I I I I I

50000001

N

z

4000000

I I I I I I I I I I I I I

8000000

7000000

6000000

5000000

4000000

3000000

2000000

1000000

0

3000000

2000000

1000000

A

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME , SECONDS

FIGURE 9

LP-SB-3 CALCULATION

SECONDARY SYSTEM PRESSURES (0-6400 SECONDS)
THE FOLLOWING ARE PLOTTED AGAINST TIME JW In f r I7h

FR-PC-BPK ,CONTROL BLK ID -72

1010 I I I I S 10

6 6

cif

(_3

V 4 4

2

0

II

'A

~

I-..

*1 I- [ 1 I I - I I I I .

2

0 I J I I I

0 500 2000 2 I 3 I 4 4 "" 

I I I I a a . . I _

O 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000.

TIME , SECONDS

FIGURE 10

LP-SB-3 CALCULATION

BREAK MASS FLOW RATE
THE FOLLOVING ARE PLOTTED AGAINST TIME

til DELPC-SO5B ,MIXTURE DENSITY

I 1000

N)

N)

-4

ul

800

600

*

*

N

c.:i

400

200

0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME 5ECONDS

FIGURE 11

LP-SB-3 CALCULATION

BREAK LINE DENSITY
THE FOLLOVING ARE PLOTTED AGAINST'TIME

PCS-MASS IFUNCTION

1Winfrith

I

N)

U1%

*

*

*

*

*

*

*

*

6500

6000

5500

5000

4500

4000

3500

3000

2500

2000

1500

1000

500

0

TIME , SECONDS

FIGURE 12

LP-SB-3 CALCULATION

PRIMARY SYSTEM MASS INVENTORY
N~

THE FOLLOWING ARE PLOTTED AGAINST TIME

DE-PC-205 ,IIIXTURE DENSITY STRATIFIED VOL FLAG

I

lwinfrithl

1. 0 •,• .|.

0.8

0.6

0.4

0.2

0. 0 r.• t-.ir••Azt4Zdc

2

cr

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME , SECONDS

FIGURE 13

LP-SB-3 CALCULATION

ILHL DENSITY AND STRATIFICATION.ONSET
THE FOLLOIV]NG ARE PLOTTED AGAINST TItlE

DE-PC-1O5 .MIXTURE DENSITY .1

Wtlnfrithl W STRATIFIED VOL FLAG

t')

'.3

-4ul

I

I',)

(_0

1.0

0.8

0.6

0.4

0.2

0. 0 ,, r ak'--'L

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME E , SECONDS

l-IGUll~u 14

LP-SB-3 CALCULATION

ILCL DENSITY AND STRATIFICATION ONSET.
THE FOLLOVING ARE PLOTTED AGAINST TIME lWinfrithi

POE-PC-O01 PUMP DELTA-P

600000

t550000

ul

500000

450000

400000

350000

300000-

250000-

200000- I

150000-

100000-

500000

0:

-500001 , , , , I , , ,

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME , SECONDS

FIGURE 15

LP-SB-3 CALCULATION

PRIMARY COOLANT PUMPS DIFFERENTIAL PRESSURE
THE FOLLOWING ARE PLOTTED AGAINST TIME

FUNCTION ,CONTROL BLK ID -88

1WInfrithl

N)

1 i •~~ff I&uIJiJuuIj .. - I I I a I I ~ I I I I

8000001

600000

400000

I I

I I

Ii t II

I ~ IL

4vv ~ tvI; ,j

I - I I I ~~II

200000-

1~

1%

I' I I I I

-200000

-400000

-600000

I I I 1 I I I I ~ I ~

'I

I I

-8000001

I I I I I I I I I I I I

•utuuuti0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME , SECONDS

FIGURE 16

LP-5B-3 CALCULATION

PRIMARY SYSTEM - SECONDARY SYSTEM PRESSURE
THE FOLLOWING ARE PLOTTED AGAINST TIME UWinfrithl

PE-SGS-001 ,SIGNAL VAR. NO. 6,PE-PC-002

PRESSURE

80000009 , , , , 8000000

7000000 "-' ' -7000000

6000000 6000000

5000000 5000000

4000000 • _•,..le- 4000000

Z'.

3000000. 3000000

2000000- 2000000

1000000 1000000

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME , SECOND5

FIGURE 17

LP-SB-3 CALCULATION

PRIMARY AND SECONDARY SYSTEM PRESSURES.
THEFOLLOWING ARE PLOTTED AGAINST TIME lWinfr'ithi

TE-5HOS-002 ,CLADDING TEMP -FINE

tI,

t%)

U'

1000

950

I

I I I I I I | I I I

900 -

850

8001-

750

(

w•

0J

700 1-

650

6001-

550

500

450

400

- - ----

-S

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME ., SECONDS

FIGURE 18

LP-SB-3 CALCULATION

PEAK CLAD TEMPERATURES (0. 05M)
THE FOLLOWING ARE PLOTTED AGAINST TIME Wunfrithl

9~ TE-5G06-01I ,CLADDING TEMP -FINE

1000 I I I I I I I I I I I I I

950 -

900 I

850 [

800

750

.6

w

C3

700 I

650

600

550

500

450

/. nn

A

'I

I I I I I I I I I I I

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIMtE , SECONDS

FIGURE 19

LP-SB--3 CALCULATION

PEAK CLAD TEMPERATURES (0. 28M)
THE FOLLOWING ARE PLOTTED AGAINST TIME

Tip

1Winfrithl

TE-5106-021 ,CLADOING TEMP -FINE

-L

wC3

TIME SECONDS

FIGURE 20

LP-SB--3 CALCULATION

PEAK CLAD TEMPERATURES (0. 53M1)
THE FOLLOWING ARE PLOTTED AGAINST TIME IWinfrlthq

TE-5J08-032 ,CLADDING TEMP -FINE

It'.

1

w,v

0 500 1000 1500 2000 2500 3000 3500 4000 4500'5000 5500 6000

TIME S SECONDS

FIGURE 21

LP-SB--3 CALCULATION

PEAK CLAD TEMPERATURES (0.81M1)
THE FOLLOWING ARE PLOTTED AGAINST TIME :Rt.

tn

Winfr'Ith

TE-5HO7-049 ,CLADDING TEMP -FINE

w.

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIM-CE , SECONDS

FIGURE 22

LP-SB-3 CALCULAT ION

PEAK CLAD TEMPERATURES (1.25M1)
THE FOLLOWING ARE PLOTTED AGAINST TIME 1W InfrI th

TE-5GO6-062 ,CLADDING TEMP -FINE

1000,

.'950-'

U'/

900 /

850 I/

I

800 I

I

750 I

I I•

700 / *..

0 ' 650 I

600-

550-

500'

450-

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000

TIME , SECONDS

FIGURE 23

LP-SB-3 CALCULATION

PEAK CLAD TEMPERATURES (1.57M)
.THE FOLLOWING ARE PLOTTED AGAINST TIME

FR-ECC-OIA ',MASS FLOW RATE

WInfrithl

tll

1~)

-4

U'

4.

3.5

3. 0

2.5

2..0

LUuN

13,

1.5

1.0

0.

0.

-0.

-1.01

5000 5200 5400 5600 5800 6000 6200 6400

TIME , SECONDS

FIGURE 24

LP-SB-3 CALCULATION

ACCUMULATOR MASS FLOW
THE FOLLOING ARE PLOTTED AGAINST TIME lWinfrithl

FR-ECC-O1A

SIGNAL VAR. NO.

,PE-SGS-001

6

,MASS FLOW RATE

U'

4.0

3.5

3.0

2. 5

2.0

1.5

8000000

7000000 -

6000000 -

5000000 -

4000000

N,

z

3000000 -

2000000 -

1000000 -

8000000

7000000

6000000

5000000

4000000 L)

(5-

1.0

0.5

0. 0

-0. 5

3000000

2000000

1000000

-1.01,

5000

0 0

5200 5400 5600 5800 6000 6200 6400

TIME , SECONDS

FIGURE 25

LP-SB-3 CALCULATION

ACCUMULATOR MASS FLOV RATE AND SECONDARY SYSTEM PRESSURE
THE FOLLOWING ARE PLOTTED AGAINST TIME

I

t.j

-4

U'

FR-ECC-OIA

SIGNAL VAR. NO.

,PE-SGS-001

6

,MASS FLOW RATE

4.0

3.5

3.0

2.5

2.0

8000000

7000000

6000000

5000000

4000000

IWlnfrithI

8000000

7000000

6000000

5000000

uIul

EU .v

1.5

1.0

c'j

0.5

0.0

3000000 -

2000000

1000000 -

0 -

4000000

3000000

2000000

1000000

-1.01 0

5000 5200 5400 5600 5800 6000 6200 6400

TIME , SECONDS

FIGURE' 26

LP-SB-3 CALCULATION

ACCUM MASS FLOW RATE AND SEC PRES (ST BYPASS VALVE AREA INCREASED)

DISTRIBUTICN

PWR HEAT TRANSFER AND HYDRAULICS MRKING GROUP

Mr

Dr

Mr

Mr

Mr

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J Fell

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AEEW

AFEW

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NNC

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c/o Reports Section, Library, Booths Hall,

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AEEW - R 2275 vi
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-89) fAaslgned 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-0022

2. TITLE AND SUBTITLE AEEW-R 2275

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

Experiment LP-SB-3 3. DATE REPORT PUBLISHED

MONTH j YEAR

Apri 1 1990

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

......... "Technical '' 

E. J., Allen, A. P. Neill. .. ... 7. PERIOD COVERED (Inclusive Dares)

B. 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 addres) I :" -" .. ' ..... "

United Kingdom Atomic Energy Authority

Atomic Energy Establishment, Winfrith

Dorchester Dorset DT2 BDH

United Kingdom

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, typ sam as aboe- if contractor, provide NRC Div•ison, Offie or Region, U.S Nuclear Regulatory Commisin,

and mailing address. .

Office of Nuclear Regulatory Research

U. S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT 1200 words or less)

Analysis of the small, cold leg break, OECD LOFT Experiment LP-SB-3 using the

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

Descriptions of the LOFT.facility and the LP-SB-3 experiment are given and development

of the TRAC-PFl/MODl input model is detailed. Thecaluclations 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 IPTORS (List words or phrases that will assist researchers In locating the report.) 13. AVAILABILITY STATEMENT

Ilni imiltrl

14. SECURITY CLASSIFICATION

TRAC-PFl, OECD LOFT, Small break (This Page)

15. NUMBER OF PAGES

16. -PRICEI .. . .

NRC FORM 335 (2-89)




UNITED STATES

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BRANCH CHIEF

REACTOR & PLANT SYSTEMS BR

NL/N-353

WASHINGTON DC 20555