NUREG/IA-0 118

* -. TD/SPB/REP/0 130

International

Agreement Report

Analysis o~f LOFT Test L5-1

Using RELAP5/MOD2

Prep ared by

S. Cooper

Nuclear Electric

Barnett Way

Barnwood, Gloucester, GM4 7RS

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

May 1993

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-0118

TD/SPB/REP/0130

International

Agreement Report

Analysis of LOFT Test L5-1

Using RELAP5/MOD2

Prepared by

S. Cooper

Nuclear Electric

Barnett Way

Barnwood, Gloucester, GL4 7RS

Off-ice of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

May 1993

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 saf ety 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.
NUCLEAR ELECTRIC jjTECHNOLOGY DIVISION

TD/SPB/REP/O 130

TITLE: Analysis of LOFT Test L5-1 Using RET.AP5/MOD2

AUTHOR/AFFILIATION: S. Cooper , Operational Performance (SW) &

Analysis Section, Station Performance Branch

ADDRESSEE: HSD/PPG

DATE: January 1991

SUMMARY:

RELAP5/110D2 is being used by Technology Division for the calculation of

certain small break loss-of coolant accidents (SfLOfCA) and pressurised

transients in the Sizewell 'B' PWR.

To assist in validating RELAP5/MOD2 for the above Application, the code

is being used to model a number of small LOCA ind pressurised fault

simulation experiments carried out in integrAl. test facilities. The

present report describes a RELAP5/MOD2 analysis of nn intermediate break

LOCA test in the LOFT facility. This test was designed to simulate the

rupture of a single 14 inch diameter accumiilator injection line in a

commercial PWR with a 25% break in the broken loop cold leg. Early in

the transient the pumps were tripped and tile TIPIS injection initiated;

towards the end of the transient, accumulator and 1,PtS injection began.

RELAP5/MOD2 gave reasonably accurate predictions ofT the system thermal

hydraulic behaviour but failed to accurately calctilate the core dryout

which occurred due to boil-off prior to acctimilator Iinjection. The error

is due to the failure to calculate the correct core void distribution

during this period of the transient. A separate calculction using the

RELPIN code using hydraulic data from the REI.AP5 analysis gave

significantly improved predictions of the core cdryout. However, the peak

clad t'emperature was underpredicted, it is believed thant the error is due

to the fact that the core liquid inventory in this boildown was

overpredicted in the RELAP5/MOD2 calculation.

A sensitivity calculation showed that when the core void distribution and

core inventory were well predicted, both REIJAP and REIPIN accurately

predict fuel clad temperatures.

The work in this report was carried out on behalf of PPG and HSD under

Agreement TSPBSW200l, PPG Task G212(22G).

KEYWORDS:

LOFT, RELAP5/MOD2

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LIST OF CONTENTS

SECTION TITLE

SUMMARY

1 .

2.

3.

4.

LIST OF CONTENTS

INTRODUCTION

TEST DESCRIPTION

DESCRIPTION OF THE CODE INPUT MODEL

COMPARISON OF RELAP5/MOD2 RESULTS

WITH EXPERIMENT

CALCULATION OF FUEL TEMPERATURES

DURING CORE UNCOVERY

SENSITIVITY STUDIES

Break Flow

Core Bypass Flows

Core Filler Gap Inlet Loss Coefficients

5.

PAGE

I.

2.

4.

.5.

5.

5.

6.

7.

7.

8.

8.

8.

10.

11.

6.

6.1

6.2

6.3

7.

8.

9.

10.

11.

DISCUSSION

CPU TIME

CONCLUSIONS

ACKNOWLEDGEMENT

REFERENCES

TABLES

LIST OF FIGURES

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1. INTRODUCTION

The RELAP5/MOD2 code, Reference 1, is being used by Nuclear Electric

for the calculation of Small Break Loss of Coolant Accidents (SBLOCA)

and pressurised transient sequences in the Sizewell 'B' PWR. To

validate the code for this purpose, it has been used to model

experiments of this type of transient carried out in various integral

test facilities. A number of these studies hanve been for experiments

carried out in the LOFT experimental reactor, Reference 2, and are

described in References 3, 4, 5, 6 and 7.

To assist in assessing the capability of RELAP5/110D2, the LOFT test L5-1

has been selected for analysis. This test was designed to simulate the

rupture of a single 14 inch diameter accumulator injection line in a

commercial PWR, equivalent to a 25% area break in the broken loop cold

leg. Early in the transient the pumps were tripped and the HPIS

injection initiated; towards the end of the transient, accumulator and

LPIS injection began.. It should be noted that for Sizewell 'B' analyses

a 25% break is classified as Thqrge, whereas in this report, as in the

external literature, this break size is referred to as intermediate.

The work reported here was performed by Technology Division under ICA

TSPBW2001, PPG Task G212, Subtask 22G.

2. TEST DESCRIPTION

The sequence of events in the test L5-1 is given in Table 2. *A brief

description of the test is given below. A more complete description

may be found in Reference 2.

The test was initiated from a steady state condition by operating a

quick-opening blowdown valve in the co~ld leg. The primary pressure fell

quickly, with the reactor trip occurring at. 0.1.7s when the hot leg

pressure fell below 14.2MPa. Coincident with r&!actor trip, the main

feedwater pump was tripped and the main stenm control valve started to

close. HPIS injection was initiated when the hot leg pressure reached

lO.6MPa, accumulator" injection at 1.66HPa and LPIS at l.OBMPa. The

primary coolant pumps were manually tripped in the test at 4s and the

coastdown completed by 19s when the flywheels wpre decoupled.

3. DESCRIPTION OF THE CODE INPUT MODEL

The code version used for the calculations reported here was RELAP5/MOD2

cycle 36.05 version E05. The input model was based on that previously

used by Technology Division for the analysis of LOFT small break tests

LP-SB-O1, LP-SB-02 and LP-SB-03, loss of feed test LP-FW-01 and the

loss -of -offs ite-power anticipated transient without trip test L9-4

(References 3, 4, 5, 6 and 7). The noding diagram is shown in Figure

1. Changes to the basic input deck were as follows.

1. Steady state controls were set for the loop mass flow, SG

pressure and SG level. These were deleted before the transient

was run.
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2. The broken loop was modified with the addition of the quick

opening break valve and the deletion of the components associated

with the steam generator simulator. This simulator is located

in the broken loop and is used for some tests but was detached

'from the facility for L5-1. Subcooled and two phase discharge

coefficients for the break, Cdl and Cd2 were both set to 1.0.

3. Feedwater to the steam generator was ramped down at the

appropriate rate commencing at the time of reactor trip. The

main steam isolating valve was set to start closing at the

prescribed rate at the time of the reactor trip. Components

representing the auxiliary feedwater were removed from the deck.

4. The reactor trip was set to occur when the intact loop hot leg

pressure fell below 14.2MPa.

5. Because the primary circuit puimps are coupled to the electric

motors via fluid drives, the effective moment of inertia is

difficult to estimate. For this reason the pump speeds were

input from the data up to 67s when the PCP-l pump was allowed

to spin freely and the PCP-2 pump was locked.

6. An accumulator and LPIS components were added to the data. These

systems were connected to the HPIS injection flow path which

discharges into the intact loop col.d leg. Pressure setpoints

for the accumulator, LPIS and HPIS injection were checked against

the test conditions.

7. Core power up to and following trip was specified from data from

the test as in the calculation in Reference 8.

After a steady state had been successfutlly achieved, the deck was

processed by PYGMALION, Reference 9, to provide a full input dataset

in which all the initial conditions were set at: their converged steady

state values. A summary of the calculated stendy state is given in Table

1. The calculated steady state is in excellent agreement with the data

with the exception of the steam generator feed, and hence steam flow.

A calculated heat balance using the measurements suggest that,

neglecting heat losses, the steam flow shotild be. 23.3±+2.0kg/s.

4. COMPARISON OF RELAP5/MOD2 RESULTS WITH EXPERIMENT

Primary and secondary pressure histories are shown in Figure 2. Primary

pressure falls rapidly until the subcooled break flow ends at 10s. The

measured pressure then falls more slowly uintil the loop seal clears at

about 50s. Up to this time, the RELAP5 calcuilated pressure is in good

agreement with the data. However, in the calculantion the loop seal does

not clear until about 64s and the pressure continiies to be held up until

that time. After the loop seal has cleared, the calculated pressure

falls more quickly than in the experiment and at 180s there is again

good agreement with the data. The time of initiation of accumulator

injection is, therefore well predicted. The accumulator and LPIS flow

maintain the pressure until the end of the transient at 300s.

There is good agreement between the calculated and measured steam

generator pressure. The slight difference may be attributed to either

a small error in the calculated inventory or a slight mismatch in the
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timing of the closing of the feedwater valve and the steam control

valve.

The calculated break mass flow is compared with experiment in Figure

3. The period of subcooled discharge is predicted to end at 13s. There

is good agreement in the main, with the flow being slightly

overpredicted from 40 to 80s, evidently as a result of the slight

overprediction in pressure associated with the error in predicting the

timing of the loop seal clearance, noted above. The primary system mass

inventory is shown in Figure 4; there is a slight discrepancy in the

initial inventory which is probably the result of approximations in the

input data but is within the uncertainty of the measured value which

was reported as typically ± 300kg In previous tests (Reference 4). After

allowance for this offset, the calculated inventory follows the

measurements accurately for the first 50s, after which the calculated

break flow is evidently too high. From this point onwards, Figure 4

suggests that the break flow Is slightly overpredicted until accumulator

injection commences. It appears that the largest overprediction in the

breakf low occurs during the two-phase discharge period from 50 to 80s.

Figure 5 shows the calculated collapsed liquid level for the loop seal

upstream and downstream sides. There are no readily comparable data

other than differential pressure measuirements which are also shown in

Figure 5 and have been arbitrarily scaled. Althoiigh the data are noisy,

it appears that the loop seal begins to clear at about 25s and is

completely clear at about 45s. In contrast, the calculation predicts

loop seal clearance late, starting at about 50s and completing at 64s.

The collapsed liquid level in the reactor pressure vessel is shown in

Figure 6. Figure 7 shows the collapsed liquid level in the core region.

Conductivity probe measurements of the local. void fraction in the core

are reported in Reference 2. These data have been used to derive the

measured collapsed liquid level curves shown in Figure 7. The large

uncertainties in the calibration constants for the probes have been used

to deduce the uncertainty bands shown in Figure 7. However, comparison

of these data with the known elevation of the core dryout, deduced from

thermocouple data, suggests that even the lower bound curve nay

considerably overestimate the collapsed core liquiid level, implying that

these curves should be considered only indicative of trends (see below).

The measurements suggest that RELAP5 underestimates the liquid inventory

in the core after loop seal clearing. After loop seal clearing, the

data suggest a significant recovery in the core liquid level which is

not captured in the RELAP5 simulation. The rate of core boildown

appears well predicted prior to accumulator injection after which the

rate of refilling of the core is underpredicted.

Figure 8 shows the RELAP5 calculated void fraction in the core. Note

that from l0s the void fraction in the top core volume is lower than

the volume below it. This reverse profile is accentuated by reverse

flow through the core which is most significant from 35 to 60s (see

Figure 9). During this period, liquid draining from the hot leg flows

down through the core and up the downcomer where it augments the flow

arriving at the break from the intact loop cold leg. The void gradient

in the core slowly increases and the top most core volume is completely

dry at 169s. From 80s onwards the calculated void fraction in the lowest

core volume exceeds 0.4. This is because of significant steam

generation due to the heat transfer from the metalwork in the lower

plenum, and flashing due to depressurisation. According to the RELAPS
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analysis a void fraction of 1.0 occurs only in the top half of the core

and liquid quickly reappears at these levels when accumulator injection

commences at 189s.

In Figures 10 to 13 the calculated fuel clad surface temperatures are

compared with the measured values. RELAPS predicts only a brief dryout

of the top half of the core, for a period of about 30s, whereas the

measurements indicate dryout extending over the whole core for a period

up to l05s. The predicted peak clad temperature is 533K compared to

the measured value of 700K.

Figures 14 and 15 compare measured and calculated HNSI, LHSI and

accumulator injection flow rates. The 1111S1 f low rate is accurately

modelled. The LHSI flow is calculated to begin 3s later than in the

experiment. The flow steadily increases to 2.5kg/s from an initial

level of l.lkg/s: data indicate an initial peak of 1.75kg/s falling to

about 1.0kg/s. The error in LHSI flow rate is probably due to the

calculated primary system pressure being underpredicted by about

2.0 lO5 Pa from 220s onwards. Accumulator injection (Figure 15) is

predicted to commence at 189s, compared with 185s seen in the test.

Comparison of the calculated accumulator flow rate with that deduced

from the accumulator level measurements indicates that the flow rate

is slightly underpredicted and the duration of the first period of

injection is also underestimated (24s compared to 40s). The accumulator

flow is predicted to cycle on and off whereas the measurements indicate

continuous injection.

5. CALCULATION OF FUEL TEMPERATURES DURING CORE UNCOVERY

It has been found in previous similar stuidios (References 10 and 11),

that owing to the coarse nodal representation of the core, RELAP5 is

generally unable to give an accurate representation of the core void

fraction distribution and core liquid level even when the core liquid

inventory is accurately calculated. Therefore, the RELPIN code

(Reference 12) has been developed. REI.PTN extracts core hydraulic data

from a RELAP5 calculation, calculates a new void fraction profile for

the core using a drift flux type model. Fuel clad temperatures are then

calculated assuming single phase steam cooling above the dryout front.

The RELPIN calculational route gives a more. conservative calculation

of fuel pin temperatures in the exposed par~t of the core than a

standalone RELAP5 analysis.

Fuel clad temperatures calculated using RETIPTN with core hydraulic data

from RELAPS are compared with the measuiremnnts in Figures 10 to 13.

Peak fuel clad temperature is now predicted to be 660K, compared with

the measured value of 700K. Dryout is predicted to commence at 133s

and the quench to be completed at 217s. The predicted heatup rat e is

in good agreement with the measurements, the slight differences are due

to the cell positions not coinciding exactly with the measurement

points. The calculated period of dryout is still slightly less than

the measurements indicate at all levels in the core, hence peak

temperatures are underpredicted. Figure 16 compares the dryout level

calculated by RELPIN with that deduced from thermocouple data and shows

quite clearly that the mixture level has been overpredicted.
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6. SENSITIVITY STUDIES

The RELAP5/MOD2 calculation, discussed above is felt to give a

reasonable simulation of the L5-1 test. However, two areas are evident

where the predictions may be usefully improved, these being the

calculated primary pressure and the calculated dryout level in the core.

Several sensitivity studies were carried out to investigate the

significance of uncertainties in key parameters. The base case

calculation described above was carried out on the Harwell Cray2

computer. The sensitivity calculations, along with a base case

calculation, were carried out on a SUN workstation using a slightly

different version of RELAP5/MOD2. It was established that the base case

calculations on the workstation and the Cray gave practically identical

results.

6.1 Break Flow

Although a number of factors influence the calculated rate of fall of

the primary pressure, the break flow rate is often found to be the key

parameter. Frequently measurements of the break flow show large

fluctuations and it is difficult to quantify accurately the primary

inventory at any time. RELAP5 calculations were carried out with a

range of assumed values for Cdi and Cd2 and the results showed that the

course of the L5-1 transient was relatively insensitive to this

parameter. Adoption of a value of 1.2 for Cd1 and Cd2 resulted in the

calculated primary pressure falling slightly more rapidly than in the

experiment after 70s and also resulted in the time of clearing of the

loop seal being advanced from 64s to 60s, n minor improvement. The

time of core uncovery was similarly advanced but there was no change

in the severity of the predicted uncovery. The more rapid calculated

depressurisation also resulted in the predicted time of accumulator

injection being brought forward to 158s, in worse agreement with

experiment than was achieved in the base case.

6.2 Core Bypass Flows

In the LOFT pressure vessel there are several flow paths that allow flow

to bypass the core. The four most significant paths are modelled in

the RELAP5 simulation. The bypass path throiigh the filler gap allows

the primary fluid to flow from the cold leg nozzle to the hot leg nozzle

(Figure 1). Similarly there i s a leakage path from the downcomer to

the hot leg nozzle. There are several. leakage paths from the lower

plenum to the upper plenum, which are luimped together as a single core

bypass in the RELAPS model (component 235). Finally, in the broken

loop, leakage between the hot leg and the cold leg through the ref lood

assist bypass valves is represented using junction 375.

In setting up the steady state calculation the flow resistances for the

four bypasses were adjusted to give flows agreeing approximately with

those quoted in Reference 13.

A sensitivity calculation was carried out in which all of the steam

leakage paths, i.e. all except the one designated "core bypass", were

closed. This brought the calculated time of loop seal clearing into

agreement with experiment. However, there was no improvement - in the
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predicted depressurisation rate after 40s, indicating that in the base

case calculations the late clearing of the loop seal was not solely

responsible for the errors in the depressuirisation rate. Although the

minimum elevation of the collapsed liquid level in the RPV during

boildown was predicted to be lower by only 50mm in the sensitivity

calculation, the peak clad temperature was higher by about 40K. This

sensitivity of peak clad temperature to the calculated core collapsed

liquid level highiights an inherent uncertainty in the RELAF5 predicted

peak clad temperatures in boildown transients of this type.

6.3 Core Filler Gap Inlet Loss Coefficients

In a previous analysis of LOFT test L5-1 using RELAPS/MOD2 carried out

at the Korea Nuclear Safety Centre (Reference 14), peak clad

temperatures were better predicted than in the present analysis.

Examination of the -KNSC input deck indicated a number of differences

from the deck used in the present analysis. Each difference has been.

systematically investigated and only one change has been identified

which improves the present predictions.

In the base case calculation the loss coefficients at the inlet at the

top of the filler gap (volume 223, Figure 1) Are set at 52.0 which is

the value used by the INEL originators of the deck. The KNSC calculation

used a value of 15.0. A sensitivity calculation was carried out in which

these loss coefficients were set to zero. Timings of the main events

are shown in Table 2. The predicted pressure, break flow and primary

inventory were similar to the base case. The collapsed liquid level

in the RPV, Figure 17, shows a slightly more severe iincovery than the

base case while the core void profile, Figure 18, shows that from 170s

until recovery the core is completely empty. Consequently, RELAP

predicts a core heat-up and peak clad temperatuires which are in much

better agreement with the data, Figures 19-22. A RELPIN calculation

is also shown which overpredicts peak clad temperatures. This is due

to the delay in predicting accumulator injection, Figure 23. This

Figure also suggests that the uncovery is overpredicted.

The significant difference between the base case calculation and this

sensitivity calculation was an increase i~n the. flow out of the top of

the filler gap from 40s onwards. This Increase in flow was drawn mainly

from the downcomer and reduces the steam flux from the lower plenum into

the core, and from 110s to 150s there was A liquitd flow out of the bottom

of the core (c.f. a steam flow into the core. during the same period of

the base case). The combined effect of the lower collapsed liquid level

and the reduction in the steam flow through the core, resulted in a more

severe core dryout.

Although it is likely that the overprediction of core uncovery could

be reduced by adjustment of the 'loss coefficients at the filler gap

inlet, this has not been attempted.
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7. DISCUSSION

The present base case analysis of the LOFT intermediate break test L5-1

used RELAP5/MOD2 cycle 36.05 version E05. The analysis gave a

satisfactory representation of the depressurisation rate and system mass

inventory but peak clad temperatures were about 150K lower than

measured. Previous analyses using RELAPS/MODI (References 8 and 13)

also successfully predicted the main features of the transient and gave

a better prediction of peak clad temperatures.

The present calculation showed a uniform core void fraction distribution

during the boildown period as seen in previous similar analyses with

RELAP5/MOD2 (Reference 10). This tends to confirm earlier conclusions

that RELAP5/MOD2 cannot be used to reliably calculate the core dryout

elevation, and core heatup above the dryout level, for core boildown

sequences using the current nodalisation. Failure to calculate the

correct void fraction distribution and dryout hehaviour is believed to

be largely due to numerical approximations in the implementation of

interphase drag modelling and as a consequence of representing the core

by a small number of nodes (Reference 10). Croxford (Reference 15)

showed a significant improvement in the predicted void profile in going

from 6 to 24 cells. Such fine noding would be impractical for plant

calculations and the RELPIN approach has been developed as an

alternative. The RELPIN calculation described above (using 12 cells)

provided a better prediction of the dryout and peak clad temperatures.

Peak clad temperatures were still underpredicted, but this is due to

an overestimate of the liquid inventory in the core during boildown in

the RELAP5 calculation.

A sensitivity calculation showed that the core void fraction was very

sensitive to the flow in the RPV filler gap. This is a feature unique

to the LOFT test facility and its complex geometry and an absence of

measurements prevent it being modelled accurately. In the sensitivity

calculation, in which the flow resistance of the filler gap was reduced,

the modified flow conditions in the filler gap resulted in the core

uncovery being slightly overpredicted. Note that in this case the

RELPIN predictions for the fuel clad temperatures were very good, the

overprediction of peak clad temperatures being due to the late

prediction of accumulator injection. In prnctice, the flow resistance

of the filler gap is not known with any confidence. Overestimation of

*this parameter may well be responsible for the underestimation of the

extent of the core uncovery in the base case cnlctilation.

8. CPU TIME

The base case calculation was performed uising RELAP5/MOD2 cycle 36.05,

UK version E05, on a Cray 2 computer. The calculation used 607s of CPU

time for the 300s of problem time. The repeat of the base case

calculation on a SUN Sparc workstation required a CPU time of 6640s, for

300s problem time.
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9. CONCLUSIONS

1. This report has described the resulIts of a RELAP5/MOD2

calculation of LOFT test L5-1, which simulated a 25% cold leg

break loss of coolant accident with operation of the high head

safety injection CHHSI), low head safety injection (LHSI) and

accumulator injection systems.

2. The calculation gives a reasonable simulation of the Loft L5-1

test. Depressurisation rate, break flow, primary coolant

inventory and emergency core cooling system injection flow rates

were all well predicted.

3. RELAP5/M0D2 did not give a realistic prediction of the liquid

distribution within the core during the boildown. Consequently

the fuel temperature excursion due to uncovery was not predicted

by the code. Failure to calculate the correct void fraction

distribution and dryout behaviour is believed to be due to

numerical approximations in the implementation of interphase drag

modelling and as a consequence of representing the core by a

small number of nodes.

4. A supplementary calculation using the RELPIN code using core

inventory data from RELAP5/MOD2 gave significantly improved

predictions of fuel clad temperatures. Residual underprediction

of peak clad temperatures is apparently due to an overprediction

of liquid inventory in the core during boildown in the

RELAP5/MOD2 calculation.

5. The predicted core uncovery was found to he sensitive to the flow

behaviour of the RPV filler gap. Since this feature is unique

to the LOFT test facility, it is unlikely that plant calculations

will exhibit similar behaviour. Errors i~n estimating the filler

gap resistance may be responsible for the underestimation of the

extent of the core uncovery in the REI,AP5/MOD2 calculation.

10. ACKNOWLEDGEMENT

The author would like to acknowledge Mr C Harwood who carried out the

preliminary dataset preparation for this work.

11. REFERENCES

1. NUREG/CR-4312, RELAP5/MOD2 Code Manual Volumes 1 and 2

V.H. Ransom et al.

December 1985

2. NUREG/CR-2398, Experimental Data Report for LOFT Intermediate

Break Experiment L5-1 and Severe Core Transient Experiment L8-2

D.B. Jarrell and J.M. Divine

November 1981
NUCLEAR ELECTRIC -9- TECHNOLOGY DIVISION

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3. GD/PE-N/544, RELAP5/MOD2 Calculat.ions of OECD LOFT TEST LP-SB-01

P.C. Halal and G. Brown

July 1986

4. GD/PE-N/606, RELAP5/MOD2 Calculations of OECD LOFT TEST LP-SB-02

P.C. flall

October 1987

5. GD/PE-N/535, RELAP5/MOD2 Calculations of OECD LOFT TEST LP-SB-03

C. Harwood and G. Brown

April 1985

6. GD/PE-N/597, RELAP5/MOD2 Calculation of OECD LOFT Test LP-FW-01

M.G. Croxford, C. Harwood and P.C. Hall

April 1987

7. GD/PE-N/721, RELAP5/MOD2 Analysis of L.OFT Experiment L9-4

M.B. Keevill

December 1988

8. EGG-LOFT-6004, Posttest Calculation of LOFT Intermediate Break

Experiment L5-1 and Severe Core Trnnsinnt 1.8-1

T. Chen and S.M1. liodro

September 1982

9. PWR/RUG/P(90)34, Implementation of PYGIIAITON for use on CRAY2,

W.J. Sibley

May 1990

10. GDIPE-N/725, Analysis of Semiscale Test S-T,1H-l Using RELAP5/MOD2,

P.C. Hall and D.R. Bull,

February 1989

11. TD/SPB/REP/0066, International Standard Problem 26: Analysis of

JAERI LSTF SB-CL-18 (5% Cold Log Break) Using RELAP5/MOD2,

S. Cooper and A.D. Froushan,

June 1990

12. ITD(A)/PERF/182, A User Guide for the RELPIN Program,

N. Musto,

November 1989

13. NUREG/CR-3406, RELAP5 Assessment: LOFT fntermediate Breaks L5-1

and L8-1,

J.L. Orman and L.M. Kmetyk,

August 1983

14. KAERI/KNSC Report, Assessment of RfEI.iAP5/110D2 using LOFT

Intermediate Break Experiment L5-1.

E.J. Lee, B.D. Chung and H.J. Kim,

January 1990

15. NUREGfIA-0014, Analysis of the THETIS Boildown Experiments Using

RELAP5/MOD2,

M.G. Croxford and P.C. Hall

July 1989
NUCLEAR ELECTRIC -10- TECHNOLOGY DIVISION

TD/SPB/REP/0 130

Table 1 - Initial Steady State Conditions

Parameter Measured Value RELAP

Core power (MW) 45.9+±1.2 45.9

Hot leg pressure (MPa) 14.93+±0.08 14.92

Hot leg temperature (K) 579.1+ 0.9 579.4

Cold leg temperature (K) 552.3+±0.9 551.5

Mass flow rate (kg/s) 308.2+ 4.0 308.2

Pressuriser level (in) 1.13 +0.03 1.14

SG secondary temperature (K) 537.8+ 0.8 538.5

SG pressure (MPa) 5.05+±0.06 5.11

SG l~evel (in) 3.22+ 0.02 3.22

SG feed flow (kg/s) 25.3+ ±0.06 23.86

Table 2 - Sequence of Events

Event Measiirement Base case Sensitivity

(fill gap)

Time (s) Time (s) Time (s)

Cold leg QOBV opened 0.0 0.0 0.0

Reactor trip (14.2MPa) 0.17+0.01 0.05 0.10

HPIS trip (lO.2MPa) 0.4+0.1. 0.5 0.45

Subcooled break flow ended 10.5 +0.5 13.0 13.0

Pressuriser indicated empty 15.5 +0.5 11.0 11.0

Loop seal cleared 50.0 4-5 64.0 64.0

Primary pressure below secondary 53.0 ±E 1 .0 41.3 41.6

Clad temperature excursion started 108.4+ 1.0 169.0 141.0

Accumulator trip (l.66MPa) 185.8+0.5 1.89.0 199.0

LPIS flow initiated 201.0 + 0.5 204.0 212.8

Clad Quench complete 213.0 + 1.0 200.0 280
NUCLEAR ELECTRIC -11- TECHNOLOGY DIVISION

TD/SPB/REP/O 130

List of Figures

I. RELAP5/H0D2 Noding Diagram for Calciulation of LOFT L5-1

Base Case Calculation

2. Primary and Secondary Pressure

3. Break flow

4. Primary fluid inventory

5. Loop seal collapsed liquid level

6. RPV collapsed liquid level

7. Core collapsed liquid level

8. Core void fractions

9. Flow through core

10. Fuel clad surface temperatures - (Elevation 1.33m)

11. Fuel clad surface temperatures - (Elevation 0.91m)

12. Fuel clad surface temperatures - (Elevation 0.63m)

13. Fuel clad surface temperatures - (Elevation 0.49m)

14. HHSI and LHSI flows

15. Accumulator flow

16. Core Dryout Level

Sensitivty Case Calculation

17. RPV collapsed liquid level

18. Core void fractions

19. Fuel clad surface temperatures - (Elevation 1.33m)

20. Fuel clad surface temperatures - (Elevation 0.91mn)

21. Fuel clad surface temperatures - (Elevation 0.63m)

22. Fuel clad surface temperatures - (Elevation 0.49m)

23. Core Dryout Level
605

430

125 COMPi

r624 COMP

(ý -PUMP

DNENT

)NEN T

INj

STEAM

GENERATOR

LX-BRANCH WITH

Ea JUNCTIONS

222 JUNCTION OR VALVE

PRESSURISER V72HEAT STRUCTURE

515

SPRAY

LINE REACTOR

VESSEL

510 440

N)

350

b3i

162o 224

FIGURE 1 FIGUE RELAP5/MOD2 1 Noding Diagram for Calculation of LOFT L5-1
Nuclear Electric - TOPAS Graphic*

X101 1

(0

(0

CL

-------- ---- ------- ---- .......---.----------------------.......----..------.... .4.. ............. -

--------------.--.---. .4.................. ..... a .------.-. . . . .---.. . . .

3.. ............. . ....... . .............. ... ...... 4. ....'.---- ýk..... ..----- -- ------------ ..------. 1'--- ........................----

2......... *----------- 4 ------------------ ------ A A ----- ....... ...... .. . . i----- . ..---

-4..--- .----------------------........... ...... 4..... ........... I-------------I............

0- -------------------------- . ....... . ................ .............. 4...................-----........-...........

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Time (s)

-- A

;0

0

NI

CD

Us

I1 0001-- PE..PC.002* 1E6 P5200 1 PE..SGS..OOI*1IE6

Figure 2. Primary and Secondary Pressure
Nuctoor Etactic - TOPAS Graphics

1 1 4 4 4 1 4 4 4 4 1 4 1

140.

120-

100-

s0-

'A

a)

C,

)

0

4-

0)

0)

0)

-- -- -- --

.. . . . . . . . . . . .. ... . . - - - - - -- - - - - - - - - - - -- - - - - - . . . . . .. .. . . . . - - - - - --- - - - - -- - - - - -- - - - - - -- - - - -

... .. -------------- ---- -------- ------- -- ----------------- -------- -------- ------------- -------------------------------------

* r4 ---- ---.------

DATA

. .. . . .. . .. . . .. . -- - - - - . . . . .- - - - - - -- -

20-

n.

0 20

lF 6WJN3600

40 60

FR-BL-BRK

80 100 120 140 160 180 200 220 240 260 280 300

Time (s)

--i

1-1

Lo

0

0

Figure 3. Break fRow
NucLear Etectric - TOPAS Graphics

4 + 4 1 A ± A _______ I _______

5000-

4500-

4000-

3500-

- 3000-

(0 2500-

--------

-- - -- - - -- - - - -

- I I

------------

6 -----------

------------

------------ ............

... ....... .. ............

------------ ------------

.... ...... .. ............

------------

......... ..

------------

-----------

T I T 1 1 +

------------ -----------

----------- ----------

------------ -----------

----------- - ---------

01

---------- ----------- - --------- ýý; ----------------------------------------------- ------------ ------------------------ i ................

N r~RE LAP

-- - - --- - -- - - -

------------ ----------- 4-

1500-

1000-

500-

0

- --- --- . . . . . .

- -- - - - - - - --- 

-- -- -- -

------ ------ ----- - .--

DAT I .... ......

l-* I I I

20

V

CNILA?

40 60

EXPPRMASS

80 100 120 140 160 180 200 220 240 260 280 300

Time (s)

-4i

0i

a;

IN

0

0 Figure.4. Primary fulud inventory
Nuclear Elsetric - TOPAS Graphics

3.5- -- --- .. . . . . . . . . . . . . . . . . . .. . . . .

' I S I S * S S* . S a * * S

a. . , -

S * . S * S

S * S , S *

* S S S S * a S S 5

S * S

0

(D

3

-D

CO

-L

-O

0

(U

3.U- i- -i-----------------------S 4 S S - . .

* S * S S

S S S S * S

* S * S S , ,

S * S * S * S * S * S

S * S * S S S

S * S * S * * S S S . S S * ,

* . . S . S . S

S 5 5 5 . . S . S * S

* . S S * S S.

2.5,------------------------------------------------------------------------------------------------------------------ * S S

* * S * S * S * S * S S

S S * S * S . S - S S 5 S S * S * S

- S . * S * S

S S * S S S * S

S * S S * . S S S

S * S S * . S

20 ... L~J.-4---±-------------- - .S .J. .a~ S S S * S S * S * S * S S *

S S S * S * . S * . S S S S

S . S * S *

'i: : *

S S S

::.\ : ;P.ELAP * . * . *

f .5* 5 . - S *S S

- . -55 IS S S S S S

'A J. J .

~1S S

1.0* ~ I I--------------4. V- S S

~, S * I: ~ N

~--~\- ~

- S S S

* S S flJ.. I~I..

- A.~A~& ~SwS44SS'1 ~~'eSbqS ~ vy

0'5

-0.5- I-i

I Y4 - yr-v., %, 

v

- - -. :v

DATA ------------- ----- ---

-V

-------- --- L 4 ------------- -------------

1 r -r t

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

-4i

1-D Wa'

CO

"I

M~

0

Downside

CNTh~hý

Downs ide

LVL-PC-027

upside

CNTRLVAR6

Upside

LVL-PC-.028

Time (s)

Figure 5. Loop seaL coLlapsed Liquid Leve L
NuceLar ReLctr~ic - TOPAS Graphics

0)

0~

U

7.0-

6.5.

6.0-

5.5.

5.0-

Top of RPV

- --------- ...........

---------- ------------

............. -----------

------------

-----------

------------

-----------

............

------------

-------------

............

............

............

------------

------------

------------

-----------

------------

------------

------------

------------

------------

------------

----------

-----------

----------

-----------

----------

4.3 -------------------. i . . .. . . .... . . . 4 ----.---------------------.-- .-------------.-.------

3.0.......----......... .i....... E . A ..- ------------ ................ A..------------------------------- -----

--- --- - -- -... - - . -. .- . ------ ------------ - - - - -- - - - - -- -- - - - - - - -- .- -- .--..- - ----.. . . - - ..... I -------

3.5............................ ... r.- -------------------------.--- -------------------------------

0.5-

0.0.

-Bottom ofCore ------

.............

a - - -- - - .. . . .. . . .. . . .

- -- -- - -- - -- - -- -

--- --- -I -- - - - -a--- - - - - - - - -

a *

I tt T 1 T

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Time (s)

--q

0

N,

0

0U

CNTRLVAR3

Figure 6. RPV collapsed Liquid LeveL
Nuctear Retctr-tc - TOPAS Gra.phics

I 4 9 4- 9 4- 9 9 + 4 4-

0. 9

1- Top Of Core

- ----. -- ---------- .--------.

3-

L

T:

I'

/

/

-rI

.-I--- ---- 

.....

/

. 7 i ---

DATA .....- .

--- ---

------------

/:

....................................................................................... ............ i

I

I

:1 17-- - -- - -

---- ------- ----------- ------------ --- ---------------- ---------------------------------------

L

0

M

C

L

LL.

0.6 -

------

-- --

-- -- -- --

--- --

0.

-------- 

------

- ---------- . ..... . ------------ - M..------ ---- ------- - -

0.4- -.------

------

---

------------

------------

---j \

-/ ---K

----4------* --- 

RELAP .....

___-If\ ----I' J

---------

---------------- ------------------------ F ----------- 4-

0.34- ------ ..... .... ------ ----- -------- ! ............. %:--------------------- ------------ ------------

A 2 ±- ----- ------------ ------------ I-------.------I------4- - 9---------------- t--------------------t.----------- -------------

0.I T

-4- ------ --------....... ........ ------- -------------- -------- ------- --- ----------- -------------- ...... ------------ ......

Bottom Of Core

A A I -~-.- T I -t I 1- I I 1- t I

0

20 40

Upper Estimate

CPLUFS60

Me an

CLEVEL

30 100 120 140 160 180 200 220

Lower Estimate

CM INUSREA

Figure 7. Core coLLapsed Liquid LeveL

240 260 280 300

T Ime (s)

C13

rn

0

C.A

W
Nuceýea Etectric - TOPAS G~-phlcs

I I

C

%4-

0

1.0.

0.9.

0.8.

0.7

0.6

0.5

0.4

0.3

0.2

0. 1

0.0

- .----- ........

- .------..........

---- -- ---- -- -

I

--- RELAP --

--- --- -- - -- .-- 

-----------

............

-----------

------------

-----------

ýO

............

............

. ... .. ..----. --I -+-- --------------- ------------ -- --- - .-- 

--------- ------- ------- --- ----- ----------- A; -------

------------ ---- ---- ------

/ 'P.' .. 

-/-- -- - - - - - - - - - - - -- - - - -- - - - - - - - - - - - A I -

.. . .. . .. . -1- - - - - - - - - - - - - - - - - - - - - --- - - - . . . . . . . .

------- I ----- .... ------ ------------

--- .F.......... ..... ........

---- --- --- -- ------ ------------------ ----------

-- -- - -- -- - - - - - - - - - - - - - - - - - - - - - - * - - - - - -- - - - - -- - - * - - - - - - - - - - - - - -

------ ------ ----- ------------ ----- I ------- ------------

------------ - ------------ - ----------- --- -----------

-------- ..... .----- .------- .-------

_ __/ --- --------- _ ----- ------ -------

-- -- - ---- - - - - - -- - - - - - - - - - - -

I.

I 20

F

V010G23006

40 60

VOID0G23005

80 100 120 140 160 180 200 220

V01DG23004 VOI0G23003 V0IDG23002

Figure 8. Core Void fractions

240 260

V010G23001

280

T ime

300

(S)

-_4

c-n

a

0

0ll
Nuceuar Electr-ic - TOPAS Graphic*

I 4 4 4 4 4 4 4

IAt% I

0)

-a

X-

- - - - - -- - - - - - - - - - - - -- --.- - --- - - - - - - . .. .. . -- -* - - - - - --- -- - -- - - - - - - - - - - ---.- -. - -.- - - a-- -- - - - - - -- - - - -

70 ---------------- ........ ............. ......------- ------------- ------- 4----- ------ 4 ----

60 - ------ - -- R P .... -------------....... 4--------------------------- ----- 4--------------I------4 -

so ----- - ...... ............. ....... I4..... ......- 4------ ----- ------------- ------------4

40 - - ----- .................... --- ........................... 4...... ..... ------- ------------

-- --0- -.-.-------.........

*444 4 4

-. .R P -------

..... ---------- ---- ------....... *2 4 ..... .----- 4 ------------ --- ......... ------------4-----

- - -4-- - - - --- --4-- - - - - -- - - - - - - -- - - - - - -- 43 - - - - - - -- -4 - - - -- -- - - - - -- . . . . . .. - - - - - -- - - - -

540 -- - ----- ------- I ........ r -------------------------------

-4-- - - - *r - - - - -- -- - - - - - -- -- - - - --- - - - - --- - - - - - -- - -- - - - - - - - - - . . . . . . - - - - --- - - - - - - - - -- - - - - - - - - - -- - - - - -- - - - - -- - ------ ------ - ------ ------------------ 4----- 

PC

4 4 4 4 4 F 4- + F F F + 4

0 20 40

Outlet

MFLOUJ2~i4600

60

Inlet

MFLOVJ22501

80 100 120 140 160 180 200 220 240 260 280

T ime

300

(S)

-4i

;0

0

0-

Figure 9. FLow through core
Nuctoor- ELoctric - TOPAS Graophice

(D

L

a.)

CL

E

CD

0 20 40 60 80 t00 120 140 160 180 200 220 240 260 280 300

Time (s)

---

CD)

P1-

Co

f~6~ EER0E1 TE-2H1 3.049

Figure 10. FueL cLad surface temperatures - (Elevation 1.33m)
Nuecteaw Etieti-ic - TOPAS Graphics

4 4 4 1 1 t t 4 1 t 4

L

CL

E

CT --------------------- T----------------------------- '9- - -

64 - ------------- ------------- ------------- ------ -- DAT --- -9---- ..... --- ......-----

62 - - -- - - - -- - - - - - - - - - - .. . . . . . . . . . . . . . . . . . . . . -- - - --- - - - -- - - - - - - - - - - - - - - - - - - - - - -- - - - - -

68----................ ............. ..................---- --------------- -------------------------

660-- ..... ...... -.................................

640-------- ----- -- ------ DAT --------

500 -........ ---------- ------ ........ ....... .......................---- ---- ---

560 ------------ j..........-----------t---------------------------- ------------------------- 4----I------- .....--------------- 

------- L.....-- ------------------ *---------=' ----- ----- ------

------ ---- ----- - ----. --- ...... -----------

0 20 40 60 80 200 120 140 160 180 200 220 240 260 280 300

Tume (s)

--I

CO

;0

M~

iEFFP230041 0 TE-4H13.037 TE-3B10.037 CLADTEMiP7 v

F igure 11'. Fue L c Lad surface temperatures - (E levat ion 0. 91 m)
Nuctear Etect,-c - TOPAS Graephics

-t * 4- 4 4- 4 + 4 4 + 4

~I~9I*~

Q

L

L

CD

E

* S5 5 - - - C , - V . --- '-- -- 64 - - - -- - - - --- - - - .. . . .. . . .

--- -- - - ............

62 - - .. ----- -------------- I------------- ----- ....... - --------------

SF 

------------ 9

560-------------------------------- - t * - ---------------- ------------- T......A- ----------------------------------

5 -6 - - . . .. . . . . . . . - - -- - - - - - - - - -- - - --.. 5 - -- - - - - - - -- - - - - - - - - - --- - - - - - - - -- - - - - - - -

5 -- 4 - -- - -- -- - - - .. .. . -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . . .- - - - - - - - - - - -

5 - -- 2 - - - - -- - - - -- -- - - -- - - - - - - - - -- -- .. . .. . . .. . . -- - -- - -- -- - -- - - - -. . . . .

-- -- -- - .v . ........ 4 8 -- - - - - -- - - - - - -- - - - - -- - - - - - .. . . .. . . - - - -- - - - -- - - - - - - - - - - - - - -

N.)

[,An~~~~ -4-------------------------------- ~~~ -;............

RELAP

- - - - - - - - - - - - - - - - - - - - - - - - - - - ----------

---- -- . . . - - -... - - - - - - - - - --.5

0

* 5 5 5 i * S * S * S 5 9 5

S 5 5 Si- *1.

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Time (s)

--A

U)

IN,

C)

RFFHMP23001 0 TE.2H1-4-028 -E.2H1-5-026 CLADTEtIP5 C)

Figure 12. FueL cLad surface temperatures - (ELevation 0.63m)
Nuclear Electric - TOPAS Graphic*

I 4 4 I I + I + 4 4. 4 + 4 4

700-

680.

660-

640-

620-

* 6 * I * I I 8 I I 0 * 8 * 8 ......... L......A.... 4 8 * 0

-- ,t * a.- 8 * *8~......8~*~~

* I * 8 8 8 * 8

I..........~...4....................I...................I................... I.................*1 8 8 * 8 * I 8 * 8

8 * 8 8 8

8 * 8 8 * 8 * 8

8 * 8 * 8 * 8 * 8 * 8 8 * 8 * 8 * 8 * * 8

* 8 * 4 8 * 8 6 .8.

-.......- t-...............1~................t,....................................................................8 * 8 * 8 8 * 8 8

8 * 8 * 8 8 * 8

8 * 8 8 8 * 8 * 8

8 * 8 8 * 8 8

* 8 * 6 * 8 8

* 8 * 8 * 8 * 8 * 8 * 8

..................... *6*......F............."1'.-4 8 8

* 8 8 * 8 * 8

* 8 8 * 8 8

L......8 * 8 * 8......8 * 8 .3

.3........3........8..... .3.......3.............

* * * DATA 6 * 8 * 8 8

* g 8 8 8 8

8 * * 8 * 8 S * 8 * 8 * 8 8 * 8 * 8 

L

a)

0.

~~. . . ..................... I........... .... . . -------------------------------...... . . . . . ......------ 

RELPIN.-O'

540..... .1.. -1 - --------------------- E P

520 - ------ -------------------

500................ 4 i

1.An ------------------------. . -----------------. . ........... ~......e---- RE LAP -----.--..

- -- -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

440]

0

. 4- - ----- - -

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

1ime (S)

-l

m

0

C)

C)

T

-ET F230076621 o TE..2H1 3.021 CLAOTEMP4

F igure 13. Fue L c Lad sur face t emperat ures - (E levat ion 0. 49m)
tNjctoor Etactric - TOPAS Gr~aphics

4- U -4 ------- J------- ----- --------- ------ 1 ------- i ............ i----------- - .--- i ------------ ------------ - ...-.

3 - .- -- - - - - - - - -- - - -- - - - - - - - - - - - - .. . . . . .. . . . . - -- - - - - - - - ----. . . . . . . . . . . . .. . . . .. .I- - -- - - - -- - - - - -- - - - -

3 -- . - ---- -- -- --- - - - - - - - - - - - - -- - - - - -- - - - - - - - - - - - - - - - -- . . . . . . . . . . . .. -- --- - - - - - - - - - - - - - - - - - S * C* S *S * .S *

2 - . - -- - - - - -- - - - - - -- - - - - -- - - - - - - -- -0-- - -- - - - - -- - - - - - -- - - - - - - -- - - - - - -- - - - - - - - - - - - - - - - -

2. -----....... .............................. -------------------- t-------------

*~ * DATA 

3.4--- - - - ------------............------------- 1.5-j~~~~~~REA .. 4 ----- ----------- I-----------..... ..................----------- ... 4--------------.- a)

0,

N)

01

1.0- ------ DATA

V,

1.- S..

*5, 'SI

.55 5 S S

LI

Nf .. .. .... S- - - --

RE LAP I0.

0.

----- ------- . 5...... .. . . -- --- 1 S-'---- ...... - -- ---- ---------------------------- S * S

i. __________ * 4 4 4 4 4 + I 4 + I

0 20 40 60 80 100 120 .140 160 180 200 220 240 260 280 300

-- I.

~0

1-D

('A

0)

HHSI

fFLOWJ40

HHS I

FT-.PI 28-.104

LHS I

IfOVJ64500

LHS I Time (s)

ffFT..P1 0.05

Figure 14. HHSI and LHSJ f~ows
NucLear Etectric - TOPAS GraphIcs

* I .8. 4 1 -1 + F I I F t

7; n -I- .......... ......................... i ------------- _ ......... A ............ ........... l.. .......... ; ........... i ------------ ; ------------------------- ; ------------------------- i ------------

'Jý % -L - ---------- L ........... ; ----------- ........... i ............ ! ........................ ! -----------

DATA - -- -- -

(0

a20.1

17.'

15.1

12.

10.1

7.5

5.0

2.5

0.0

............... --. ........ ---- -- - -------- ............

5 - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - -

. . ....... - ------------- ----

------ ------ -------- ..... -------

- . v.....L......

P- - IU I * *

V *

0 20

F['CV, 6200

40 60 s0 100 120 140 160 180 200 220 240 260

Time (S)

-4i

CD,

1-1

0

;0

00..ECC-.01A*1252.5

F igure 15. Accumu Lat or F Low
Nuctear Reuctric - TOPAS Grap~hics

C

0

CD

(D

ILL

1.8

1.6

1.4

1 .2

0. E

0. e

0.4

0.;

Top of Coret

.........-------

................. RELPIN ---------- -- I - ----

- 4 - : - --- - ---

----- - - - - - - - - - - -. /- 

- - - - - - - - - - - -

-- - - - - - -- --- - - - - - . .........

-- -- ------- +

- --- --. ---- -- ----- - ----------- ------- - --- --- .... . . ...........--- -------- ----- -- .....

Dryout level deduced from

Bottom of Core thermocouple measurements . .. . . . . . . . . . . . . . .. .. . . . . . .. . . . . ..... . .. . . . . . . . . . - -.5- - - - -- - - -. .. - - - - -

0 20

OUENCHFRONT

40 60 80 100 120 140 160 180 200 220 240 260 280

Ti me

300

(S)

-4

CU

0

U4

C0 Figure 16. Core Elryout LeveL
NucLear Etectric - TOPAS Graphics

+ 4- 4 I + 4 4 4- 4 -4 4. 4 4.

-U

- -- - - - - -- *. - - - - -- - - - - - - - - - - -- - - - - - - - - - - -- - - . . . . . .. . . . .I . . . . . ... . . . . ------------ ------ ------

6. -- -- --- - -- - - - - - -T - - -- - -- - - - - -. . . . . ... . . . . . . . . . . . . . .. . . . . . . . . . . . .. . . . . . . . . . . . . .. . . . . .. . . . . .

6.0..... ------- ........... .... ....... 5 .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . . . .. . - - - - -

5. ------------------- ------- .......................... ............ ............ ............ ....

4. .................. 4...... ......... .... .... ............ ------------ .....

4 * .5 5 .- . . ..- .. -- - - - . .. . . . . .. . . . . -- - - --- - - - -- . . . . .. . . . . ....- - -- - - - -- - - - . .. . . . . .. . . . . .. . . . . .. . . . . . . . . .

3 .. . . .. . . . . . . . . . . .. - - - . . . . . . . . .

-- -- - ------- -------- --- ------ ------------------- .............

3. - ------------ ... . .. . . ....... ----- ------ ........... ------------------ 2. ---- ------ ----- ------- ------------------------------------ ----- ------------ a ------------

00

'. U I ............ .. ~~~.......... . . . . . . ---------- . . . .*---------4 

'a

S.

RELAP

Kor...... . . ...................... ....... ------ --------=- . ---............................... . . ..-----------

0.5.

0.0.

...... .. -------. ----

...................

RELAP

Sensitivity to Filler Gap Inlet Flow

..........

.... .~..t------- . .------- ------- ....... ............ . -I.-

0 20 40 60 80 100 120 140 160 ISO 200 220 240 260 280 300

Time (s)

-4-A

C:)

"I.

M

0 CNTRLAR3F. 2 CNrRLVAR3I 1

.Figure 17. RPV coL..apsed Liquid LeveL
Nuctear, ElLectric - TOPAS Gr-aphics

C

0

0.

L

0.4

0

0.3

0.2

0. 1

P1-

LA

0

0 20

V170GT3506

40 60 80 100 120 140 160 180 200 220 240 260 280 300

T Ime (s)

VOIDG23005 VOIDG23004 v01G30 V010G2iO 3002 V01G2 I0

Figure 18. Core Void fractions
1&jctea Eletric - T04'AS Graphic*

800-

760-

............... -------. . ....... .. . -- . ..-- - -- - -

---------- - -------------------------------- ~ R L.................. -- - - - - -- - - - - - -- - - - - -- - - - - - - - - - -i . . - - - - -- - - - - - - - - - - - - - -- - - - - -

L

CL

(ZU................... . 4 ............ ... U. . .......... .4....I............ .......... *.. . ..--. ..-----.-----..-- +. .........

. .. ....4..... ............. .................. 4...... . . . .

-- -- -- .. . . . . .. . . . . . ... -- - -- - --- ... . .. . .. . . .. .. . ..

* * ** DATA

............ I.................. .'..............I-... ...... ..... .. . ..... . .. . . . . I........... - ------ T ------ -- ,----------

U 4U 4 U RE *

64 - - - -- - -- - - - -L. . . . . .U.. . . . . . . . .. . . . . . . .I .. . . . . . . . . . . -- - - - - - - -- - - -- - - - - -- - - - -

- ----- W ..... ............. ............. ------ ------ ..... ...... ... ....... ------ -----

. . . -- -. - - -a-- ............ - - - --- ...... .... :* . .. .. .... L -- - -U-- - - - - - -- - - - - - - - - -

52 --------------------------------------........... d..... .... . '. .. ~ .. ----- ---- -. ----- ----

. . .. .4. . . . ... .4. . .... ........... --------------------- 44U 4------- ------------------- -------- - ----------------- -------.----

W

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Time (s)

-4

LO

aS

C13

M

C) iTEFF23To00(51 0 TE..21-1 3.049 CLAOTEI¶P9

F igure 19. Fue L c Lad sur face t emperat ures - (E Levati1on 1. 33m) C
rNjctow- Etecte-ic - TOPAS G.-ephice

I -- I

800-

760.

720-

680-

Sensitivity to Filler Gap Iniet Flow~

..........

-----------

------------

------------

------------

------------

------------ ...........

------------ f -----------

............ r -----------

----------- ...........

------------------------

------ a-----------r-----* - 44

------ I........ .......... ------------- -----

------------ L- ----------.---

---------------- ;---------1 ------------------- -*I..------------- -- 

-- -- --- - - D 4 T ZZ - . . . . ..- - - - - - - - - - - - - - -

WA

L

(D

4 I A .* I i,

600-....-.........-....-----------.-.....-.... -.----

----------- .....----------- '--- .....-------- -------- 1------ ------ i L----J4----. ----- 600--------------a -.-- - ............... .. t!.4------- .

7 - -----

------ -- ---------------------------- ---

- ---- - - - - - - - - - - - - - -- ---------------

4 40------ --------------------- t-------.- L ------------------------. --------------- ,---- .-----.. -------..--------------------- *,-- I -

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Time (s)

C CLAOE 7

1-1

CfD

a

CO

M

C) FFHP2000410 TE.4H1-3-.037 TE-3BI O-037

F Igure 20. Fue L c Lad sur face t emperat ures - (E Levati1on 0. 91 m)
NucL.ar ELactric - TOPAS Graphics

800-

760..

720-

680-

640-

L

'' 600~

L

E

(1 560-

520.

Sensitivity to F

............

------------

------------

............

------------

L ...........

------------

L -----------

............

------------

............

------------

------------

------------

------------

.............

iller Gap Inlet Flow------

...........

.. ........ . .. . .. . .. .. .. .. . .. . ..

................... ------- I............

........ r..------. ................

...... ------ a------ .....

----- - - -- - - - - - - - -- - - - - - - -

------......... ..--------. .------- -.------

--------....... ---------- R A -E----. --.............---- ...... 

------ - - -- -------- .---- ....... .I. . .......... ...- ----

4 ..... ~ -- ---- .... RELA ----- .----

.......... ----.- -------.

------ ----- ------ ~ . . ..... ** ---------

-- --- -----------. -- - - - - - - - - -- . . . .... L - - - - - -- - - - - -j- - - - -

"3

"3

---------- I ------------------------ 4.

.----------

--------.-. ------.

............

............

------------

L.. . . . . .a... . . . . . a.. 

- --- a-- a

.... a.

-a----- - - - -

480-

440.]

0

---. -- --- - -- -- -- -- -- - --. - -- -- -- - -- -- ---. a a aa aT

-- - --- -- - - - - - a - - - - - -- a - - - - - a - - -- - -a

----------- ------------

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

--i

CD

U)J

N

;0

M~

W

Time (s)

-EtIP2T300031 0 TE.2H-14.028 TE.21-15.026 CLAOTEtIP5

Figure 21. FueL cLad surface temperatures.- (ELevation 0.6 3m)
Nuacteav Etacti-ic - TOPAS Graphics

800

760.

t ~ Sensitivity to Filler Gap Inlet Flow * S * S

- ------ - ---- ------ --------- ------- ------------------------- -- ---. --- -----

-S---- ---. . .----------.. .-.-. .- ------ A1------------------- ---------- -

/

0

L

4..

CO

CL

E

(D

------------------- 4.............I------------------------------ --------- I..............- ------#--- ------------- I............ ........... ......-

64 ------ ------------------------------ ------------------------ L-------------------D TA------------------ ..........------------------------ ---------

- - - - - - - - - - - -.-- - - - - - -

-- ............... -------- 5- - - - - S. - -

---------------------- ------------- ----------- ------------ ---------- I---------- - 4- ----- I...... ----- I------------------

680 --- -------...... ----- ...........-...... .. ..... .-.....------ L------ -------- --------

-T.

540 - ------------ ------- ---- DAT --- -- --- -- ----------

-- - - - - -- -- -- - --- ---- A I, -

--- ----------------------- ----------- ----------------.

C,,

C,,

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

-4i

U~i)

0,

;0

04

C0

Time (s)

fERMP2300021 0 TE-.2H1-3.021 CLAOTEt1P4

Figure 22. Fuel clad surface temperatures - (Elevation 0.49m)
Nucloar Electric - TOPAS Graphlcs

C

0

4-,

-0

LuI

1.8-

1.6-

1.4-

1.2-

1.0-

0.8B0.6-

0.4-

0.2-

0. 0.-

Top of Core

........ ---------------- -------- -------- - - - - REA

4 ----------- -----------------

-- .- - -- -- - -- -- -. .. .. . . .. . . . -- ..---- - r - - - - - T.. .. . . -- - - - -

.........------......... U .........------------

-- Sensitivity to Filler Gap Inlet Flow ---- I-------------

........... ......................... A......4I--- -----------

------------

------------

------------

------------

-- - -p - -- --- -

----------- ------------

---------- -----------

---------- ----- ------

--------- -..... .....

--------- -- ----- ---

------------

----- -----

---- ------ CA)

----------- ----------------------- I -------- I ---- ! --- -------- "

/4- 

& ------------ ; ------------

......................

----------- Drycut level deduced from

thermocouple measurements

4-.

---- ----- --- --- --- --- --- ---

* 1~

0 20

QUENCHFRONT

40 60 80 100 120 140 160 180 200 220 240 260 280

T ime

300

(S)

CD

U)

"I

C)

Figure 23. Core Dryout LeveL
RM 335

3201,3202

U.S. NUCLEAR REGULATORY COMMISSION

BIBLIOGRAPHIC DATA SHEET

IS.. instructions on the reversal

2. TITLE AND SUBTITLE

Analysis of LOFT Test L5-1 Using RELAP5/MOD2

1. REPORT NUMBER lA~suoned by NRC. Adid Vol.. Supo.. Rev.. MCI Addendlum Numbers, if any.)

NUREG/IA-0118

TD/SPB/REP/0130

3. DATE REPORT PUBLISHED

MONTH 1 tEAR

May 193

4. FIN OR GRANT NUMBER

L2245

5. AUTHOR (S) 6. TYPE OF REPORT

S. Cooper Technical Report

7. PERIOD COVERED tometuivue Datato

nowan. ,d mailin addnvLI

Nuclear Electric

Barnett Way

Barnwood, Gloucester, GMA 7RS

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC. type "Same asabove". if contracor. pro~v ids NRC Division. Offic or Region. U.1 Miclee, Regulatory Com~mission.

and mailing 0"4141j

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 vvoros of leaw)

The RELAP5/MOD2 code, Reference 1, is being used by Nuclear Electric for the calculation of Small Break Loss

of Coolant Accidents (SBLOCA) and pressurized transient sequences in the Sizewell "B" PWR. To validate the

code for this purpose, it has been used to model experiments of this type of transient carried out in various integral

test facilities. A number of these studies have been for experiments carried out in the LOFT experimental reactor,

Reference 2, and are described in References 3, 4, 5, 6, and 7.

To assist in assessing the capability of RBLAP5/MOD2, the LOFT test L5- 1 has been selected for analysis. This test

was designed to simulate the rupture of a single 14 inch diameter accumulator injection line in a commercial PWR,

equivalent to a 25% area break in the broken loop cold leg. Early in the transient the pumps were tripped and the

HPIS injection initiated; towards the end of the transient, accumulator and LPIS injection began. It should be noted

that for Sizewell "B" analyses a 25% break is classified as large, whereas in this report, as in the external literature,

this break size is referred to as intermediate.

12. KEY WOROSIDESCR IPTOR S Wist wot* or prmar that will assist trsabahrr in locating the report.) 13. AVAILABILITY STATEMENT

LOFT Unlimited

RELAP5/MOD2 14. SECUAITY CLASSIFICATION

ICAP Program IThis pa"i

Unclassified

(Vift fiesoort

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 12439)


Federal Recycling Program

NUREG/IA~-O118 ANALYSIS OF LOFT TEST L5-1 USING RELAP5/MOD2 MAY 1993

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

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