NUREG/IA-0183

International 

Agreement Report

Analysis of the LOBI 

Experiment Test BT 

the RELAP5/MOD3.z 

Prepared by 

J. Blanco, E. Moralo, R. Sanjuan, C. G6mez 

Union Fenosa Generacion S.A. 

Central Nuclear Jose Cabrera 

Madrid 

SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

June 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission

56 Using

Code
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NUREG/IA-0183

International 

Agreement Report 

Analysis of the LOBI 

Experiment Test BT-56 Using 

the RELAP5/MOD3.2 Code 

Prepared by 

J. Blanco, E. Moralo, R- Sanjuan, C. G6mez 

Union Fenosa Generacion S.A. 

Central Nuclear Jose Cabrera 

Madrid 

SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

June 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
ABSTRACT

This document has been drawn up by UNION ELIýCTRICA FENOSA, 

S.A. (UNION FENOSA), which participates in the Code Application and 

Maintenance Project (CAMP) as a member of UNIDAD ELECTRICA, 

S.A. (UNESA). The document presents a comparison between 

experimental data from the plant and data obtained by simulation with 

the RELAP5/MOD3.2 code of the experiment BT-56 which took place 

on 3rd July 1990 at the Joint Research Centre of the Commission of 

the European Communities, located in Ispra (Italy). 

The experiment originally scheduled for the LOBI-MOD2 facility was a 

Loss of Main Feedwater (LOFW), but multiple failures occurred, meaning that the evolution of the experiment deviated it from the 

scheduled sequence. 

The simulation was performed using the RELAP5/MOD3.2 code, on a 

Digital AlphaServer 2000 4/200 computer and the operating system 

DIGITAL-UNIX. 

The results show an acceptable agreement between the 

phenomenology observed during the experiment and that predicted by 

the code.

iii
TABLE OF CONTENTS 

A bstract ............... .................. ...... .... .. . .............. iii 

Table of Contents ..................................................... v 

List of Figures ....................................................... vii 

List of Tables ........................................................ x 

Executive Sum m ary ................................................... xi 

N om enclature ....................................................... xiii 

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

2. Description of the LOBI-MOD2 Facility .................................. 3 

3. Description of the Experiment ......................................... 9 

3.1 Initial pressurization prior to pump shutdown ......................... 11 

3.2 Primary overpressurization with the pumps shut down ................. 13 

3.3 Loss of inventory (LOCA) across the rupture disk ..................... 13 

4. Description of the Plant Model for RELAP5/MOD3.2 ....................... 16

V
5. Steady-State Simulation ............................................ 22 

5.1 Analysis of the Primary .......................................... 22 

5.2 Analysis of the Secondary ....................................... 24 

5.3 Steady-State Results ........................................... 24 

6. Transient Sim ulation ............................................... 26 

6.1 Characteristics of Transient Simulation ............................. 26 

6.2 Transient Simulation Results ..................................... 28 

6.3 Run Statistics ..... ........................................... 30 

7. Sensitivity Calculations ............................................. 30 

7.1 Coefficients of Discharge across the rupture disk ..................... 31 

7.2 Heat Transfer in the Steam Generators ............................. 32 

7.3 Coefficients of Discharge across the SG Relief Valve (BL) .............. 32 

8. C onclusions ....................................................... 33 

9. References ...................................................... 35 

APPENDIX I: Average Temperatures in Core Heat Structures .................. 36 

APPENDIX II: Simulations Graphics ...................................... 41

vi
LIST OF FIGURES

Figure 2.1. Heater rod (section and power distribution) 

Figure 2.2. Schematic diagram of the primary system 

Figure 2.3. Schematic diagram of the secondary system 

Figure 2.4. Intact loop steam generator 

Figure 3.1. Schematic diagram of secondary system operation in Test BT-56 

Figure 4.1. Schematic diagram of nodalization 

Figure 4.2. Reactor head 

Figure A.I.1. Average temperature in the wall of the core heat structures 

Figure A.I.2. Core nodalization

Figure A. 11.1. 

Figure A. 11.2. 

Figure A. 11.3. 

Figure A. 11.4. 

Figure A. 11.5. 

Figure A. 11.6. 

Figure A. 11.7. 

Figure A. 11.8. 

Figure A. 11.9.

0

Pressure in the intact loop cold leg 

Pressure in the broken loop cold leg 

Pressurizer pressure 

Pressure in the upper plenum 

Velocity in the intact loop hot leg 

Velocity in the intact loop cold leg 

Velocity in the broken loop hot leg 

Velocity in the broken loop cold leg 

Temperature in the intact loop hot leg

Temperature in the intact loop cold leg 

Temperature in the broken loop hot leg 

Temperature in the broken loop cold leg 

Core inlet temperature 

Core outlet temperature 

Temperature in the upper plenum 

Feedwater flow to the steam generators 

Volumetric steam flow in the steam lines 

Volumetric flow in the surge line 

Steam generators pressure 

Pressurizer vs upper plenum pressures

vii

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A.

11.10. 

11.11. 

11.12. 

11.13. 

11.14. 

11.15. 

11.16. 

11.17. 

11.18. 

11.19. 

11.20.
Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A.

11.37. Temperature in 

11.38. Temperature in

the core heat structures 

the core heat structures -

Le'd 

Le'j

11.21. Upper plenum vs steam generator pressures 

11.22. Core inlet density 

11.23. Density in the intact loop hot leg 

11.24. Density in the intact loop cold leg 

11.25. Density in the broken loop hot leg 

11.26. Density in the broken loop cold leg 

11.27. Temperature in the core heat structures - Le\ 

11.28. Temperature in the core heat structures - Le, 

11.29. Temperature in the core heat structures - LeN 

11.30. Temperature in the core heat structures - Le\ 

11.31. Temperature in the core heat structures - Le\ 

11.32. Temperature in the core heat structures - Le\ 

11.33. Temperature in the core heat structures - Le\ 

11.34. Temperature in the core heat structures - Le\ 

11.35. Temperature in the core heat structures - Le\ 

11.36. TemDerature in the core heat structures.A - I e%

tel 

eel 

,el 

,el 

,el 

el 

Jel 

'el 

'el 

'el 

'el

Collapsed level in the pressurizer 

Collapsed level in the steam generators 

Flow across the rupture disk 

Flow regime at the rupture disk 

Sens 1. Pressure in upper plenum 

Sens 1. Core heat structure temperature (level 4) 

Sens 1. Temperature in broken loop cold leg 

Sens 1. Steam generators pressure 

Sens 2. Pressure in upper plenum 

Sens 2. Core heat structure temperature (level 4) 

Sens 2. Temperature in broken loop cold leg 

Sens 2. Steam generators pressure 

Sens 3. Pressure in upper plenum

viii

/el 1

o~

2 

3 

4 

5 

6 

7 

8 

9 

10 

11 

12

Figure 

Figure

A. 

A.

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A. 

Figure A.

11.39. 

11.40. 

11.41. 

11.42. 

11.43. 

11.44. 

11.45. 

11.46. 

11.47. 

11.48. 

11.49. 

11.50. 

11.51.
Figure A. 11.52. Sens 3. Core heat structure temperature (level 4) 

Figure A. 11.53. Sens 3. Volumetric steam flow in the steam lines 

Figure A. 11.54. Sens 3. Steam generators pressure

ix
LIST OF TABLES 

Table 3.1. Configuration of the experimental facility 

Table 3.2. Sequence of main events during Test BT-56 

Table 5.3.1. Steady-state results 

Table A.. 1. Core structure adjustment temperature in the RELAP5 model 

0o

x
EXECUTIVE SUMMARY

This paper presents the results obtained from simulation with the 

RELAP5/MOD3.2 code of the experiment BT-56, which took place at the LOBI

MOD2 experimental facility on 3 rd July 1990. The experiment originally foreseen 

was the loss of main feedwater (LOFW), but multiple failures occurred which 

meant that the experiment deviated from the scheduled sequence. The sequence 

of events that finally characterized the transient was as follows: 

"* Loss of feedwater and turbine trip. 

"* Failure to open of the relief valve for the steam generator in the intact 

loop. 

"* Shutdown of the main coolant pumps, with loss of forced circulation in 

the primary. 

" Failure of scram (disconnection of electrical core) up to 24 s. 

"* Rupture of primary system rupture disk, located on the downcomer. 

Test BT-56 presented a sequence of particularly relevant thermalhydraulic 

events, as regards understanding of the interactions between components and 

assessment of the safety of Pressurized Water Reactors. The most outstanding 

phenomena that took place were as follows: 

"* Heating and pressurization of the primary system. 

"* Pressure mismatch between the primary system and the pressurizer. 

"* Core dryout at high pressure.

xi
"* Sudden depressurization of the primary after the break of the rupture 

disk (LOCA). 

"* Heat sink inversion. 

* Reverse flow in the hot legs and the core, up to primary 

depressurization. 

* Thermal stratification in the loops. 

The results of the experimental simulation using the RELAP5/MOD3.2 code 

reflect an acceptable reproduction of the phenomenology and plant behaviour.

e'J 

0

xii
NOMENCLATURE

AFW Auxiliary Feedwater 

AFWS Auxiliary Feedwater System 

AIS Accumulator Injection System 

ANM Annular-Mist 

BL Broken Loop 

CL Cold Leg 

ECCS Emergency Core Cooling System 

HL Hot Leg 

HPIS High Pressure Injection System 

IL Intact Loop 

JRC Joint Research Centre (of the European Communities) 

LOCA Loss of Coolant Accident 

LOFW Loss of Feedwater 

LPIS Low Pressure Injection System 

0 MCP Main Coolant Pump 

PL Primary Loop 

PZR Pressurizer 

RPV Reactor Pressure Vessel 

SG Steam Generator 

SL Secondary Loop 

SLG Slug

xiii
1. INTRODUCTION

This report has been drawn up by UNION FENOSA within the framework of the 

"Code Assessment and Maintenance Project" (CAMP). The objective of the 

report is to analyse the BT-56 experiment performed using the RELAP5/MOD3.2 

code. The experiment was carried out at the LOBI experimental facility and 

originally consisted of a Loss of Feedwater transient (LOFW) with turbine trip. 

The experiment underwent an unexpected sequence of malfunctions, presenting 

a phenomenology of special interest for the verification of the codes habitually 

used for PWR reactor safety assessment. 

The LOBI project (LWR Off-normal Behaviour Investigations) consisted of a high 

pressure experimental reactor, designed and operated at the Commission of the 

European Community's Joint Research Centre at Ispra (Italy). 

The reactor incorporates the essential primary components and secondary 

cooling systems of a four-loop PWR reactor, its design being scaled to obtain the 

same response during normal and off-normal operation. The construction of the 

reactor was commissioned in December 1979, and it started up in June 1982 

with the MOD1 configuration for the research of large-break LOCA phenomena, 

subsequently being modified to the MOD2 configuration for the simulation of 

small-break LOCA's and special transients. 

The main objectives of the LOBI experimental reactor were as follows: 

Identification and/or verification of the basic thermalhydraulic 

phenomenology postulated for LOCA (large or small-break) accident 

conditions in a PWR reactor, along with recovery procedures and 

procedures management in the event of an accident.

1
* Acquisition of an experimental database for the development and/or 

implementation of analytical and validation models independent of the 

thermalhydraulic codes used for PWR safety assessment. 

The BT-56 experiment was performed as part of the experimental series 

corresponding to program B, carried out under the auspices of the Program of the 

European Commission's Reactor Safety Research Commission, with independent 

contributions made by many of the industries and institutional bodies of the 

European Community member countries.

2
2. DESCRIPTION OF THE LOBI-MOD2 FACILITY

The LOBI experimental reactor is an integral high-pressure system representing 

1300 MWe KWU design PWR reactor, at a scale of approximately 1:712. The 

primary has two loops, one representing the three intact loops and the other 

representing the broken loop of the reference reactor. Each primary loop has a 

main coolant pump (MCP) and a steam generator (SG). The core is made up of 

64 electrically heated rods arranged in a square matrix of 8 x 8 inside the reactor 

pressure vessel (RPV). The rated power is 5.28 MW. The facility also includes 

other components, such as the lower plenum, the upper plenum, the annular 

downcomer and the reactor head external assembly. The pressurizer is connected 

to the hot leg of the broken or intact loop, depending on the experiment to be 

performed. In test BT-56, it was connected to the hot leg of the broken loop. 

The primary coolant system, which is shown schematically in Figure 2.2, 

operates under normal conditions of approximately 158 bar at a temperature 

between 294 and 326°C. 

In configuration MOD2, the Emergency Core Cooling System (ECCS) may be 

replaced with a High Pressure Injection System (HPIS), by Accumulator Injection 

System (AIS) and by the Low Pressure Injection System (LPIS). The HPIS and 

LPIS systems may be connected to the cold and hot legs or to a combination of 

both. These systems did not actuate at any time during the BT-56 transient. 

The heat generated in the core (5.28 MW) is removed from the primary loops via 

the secondary circuit, which is made up of a condenser, a cooler, the main 

feedwater pump and the auxiliary feedwater system (AFW). The operating 

conditions of the secondary coolant circuit are approximately 210'C for 

feedwater temperature and a pressure of 64.5 bar. The secondary cooling circuit 

is designed to operate at a pressure and temperature of 100 bar and 31 00C, 

respectively. Figure 2.3 shows a schematic drawing of the secondary circuit.

3
The LOBI-MOD2 facility is entirely scaled to maintain similarity with the reference 

plant under normal and off-normal operating conditions, as regards 

thermalhydraulic phenomenology. The scaling required a coefficient of 3:1 

between the thermal and hydraulic parameters of the intact and broken loop 

steam generators. 

The broken loop steam generator (BL-SG) has 8 U-shaped tubes (+ 1 spare), and 

steady-state heat transfer at nominal power is 1.32 MW, while the steam 

generator in the intact loop (IL-SG) has 24 tubes (+ 1 spare) and a heat transfer 

of 3.96 MW under steady-state conditions at rated power. Figure 2.4 shows a 

schematic drawing of the steam generator in the LOBI-MOD2 intact loop. 

Each steam generator consists of a single-piece cylindrical vessel, with an 

annular downcomer separated from the riser by means of a cylindrical sleeve 

supported by the tubesheet. Each generator has two steam separators. The U

0 shaped tubes are connected in a way as similar as possible to that used in the 

reference plant. There is a connection between the secondary and the primary 

water inlet or outlet at the tubesheet, in order to allow the steam generator tube 

rupture (SGTR) accident to be simulated. 

The core is made up of 64 electrical resistances simulating 64 fuel elements, 

with hollow tubes having an active heating length of 3.9 metres, a diameter of 

10.75 mm and a pitch of 14.3 mm. The thickness of the tube walls varies in 5 

steps, in the heated zone (Fig. 2.1), in order to produce a cosine distribution in 

axial heat flux. 

The measurement system consists of a total 470 measuring channels. This 

allows all of the relevant thermalhydraulic quantities to be measured in each of 

the components of the loops, vessel and steam generators.

4
EAD'BELIED4'

NICKEL-201 

NICKEL BRAID I

M429A]L 14RE I NICKEL-201

330 200 412 663 1750 663 412 2460

4,7 0,5 6,5 14,0 45,1 14,2 6,6 8,4

Q tot = 5.28 MW NORMAL OPERATING POWER

Figure 2.1. Heater rod (section and power distribution)

o

5

0

Q / Q tot (%)

w
,, IoIPump 

Sft"' 

2 Vessel H...S 

Fthage T pan 

Figure 2.2. Schematic diagram of the primary system

6
J• Feedw~ter 

I.,. pumps 

ImI 

from 

10 Stuiage Tank 

Feedwater Ptinp" 

Figure 2.3. Schematic diagram of the secondary system

7
Steam

Fine Separator 

Coarse Separator

Feed Water

Measurement 

Levels

Figure 2.4. Intact loop steam generator

8
3. DESCRIPTION OF THE EXPERIMENT 

The BT-56 experiment represents a transient of multiple inadvertent failures 

developing on the basis of an originally specified LOFW type transient. 

The configuration of the system is shown in Table 3.1. 

Pressurizer * Surge line connected to the hot leg of the 

broken loop 

* PORV in control mode at 162 bar 

* SRV setpoint 167/164 bar, with an orifice 

of 2.44 mm 

HPIS 9 Disconnected 

AIS * Disconnected 

AFW * Actuation on low level in the SG downcomer 

and PCS temperature. 

SG Control 0 Pressure control during the initial part of 

Test BT-56 with a pressure increase rate of 

0.07 bar/s 

* Following the Scram, foreseen at 59s, 

pressure control at 70 bar with SRV 

MCP seal water 0 Extraction of flow from upper plenum during 

steady-state phase, using the pressurizer 

level control system 

* Extraction of flow from lower plenum during 

the transient, using the seal water tank 

control system 

Containment * Atmospheric 

Table 3.1. Configuration of the facility during the experiment 

The sequence of events occurring during the experiment was as follows: 

Failure to open of the IL steam generator relief valve. Steam relief from 

the IL steam generator occurred across the BL steam generator relief 

valve. Isolation of triple-loop steam generator, with subsequent loss of 

heat transfer on closure of the steam line valves at 60 s. The closure 

sequence of the valves (foreseen and actual) is shown in Figure 3.1.

9

0•
"* Shutdown of the main coolant pumps (MCP's) due to loss of pressure in 

the seal system, with loss of forced circulation in the primary. 

"* Delayed core scram. 

* Loss of primary inventory due to rupturing of the safety disk. 

As a result of all the above, the sequence of thermalhydraulic phenomena was 

particularly relevant for understanding of the interaction between the 

components and validation of the codes used for the light water reactor safety 

assessment. 

The most important thermalhydraulic phenomena that occurred were as follows: 

"* Primary pressurization and heating. 

"* Compensation of steam generator pressures due to their having the 

same steam relief path. 

"* Primary and pressurizer pressure mismatch. 

"* Dryout of the core at high pressure. 

"* Loss of primary inventory from the downcomer. 

The sequence of events that followed during the BT-56 experiment is shown in 

Table 3.2.

10

0

Time (s) Event 

"* Closure of condenser valve and of the 

0 feedwater valves. 

"* IL steam generator SRV fails closed.
* Partial loss of heat removal in the 

secondary. 

9.70 * Primary SRV and PORV open at 170 bar. 

12.2 * Closure of primary SRV and PORV. 

16.3 * Opening of primary SRV and PORV. 

17.1 0 Main pump trip on low differential pressure 

in the seal water system. 

19.0 * Primary pressure greater than 86 bar. 

19.1 * Break of rupture disk (orifice of 24.7 mm). 

19.8 a Core dryout. 

20.5 * Closure of primary PORV and SRV. 

23.5 & Manual scram of the core 

33.0 * Core Rewet 

60.0 * Steam generator isolation 

100. 1 End of experiment 

Table 3.2. Sequence of main events during Test BT-56 

Analysis of the transient may be divided into the following phases: 

"* Initial pressurization prior to pump shutdown. 

"* Overpressurization of the primary with the pumps shut down. 

"* Loss of inventory (LOCA) across the rupture disk. 

3.1. Initial pressurization prior to pump shutdown 

The steam generated in the SG's during the steady-state phase was collected a 

header and channelled to the condenser. At the beginning of the transient the 

steam generator feedwater valves closed. The feedwater flow in the intact loop 

decreased rapidly. The feedwater flow in the broken loop dropped to 8% of the 

rated value and remained at this value for approximately 36 seconds, at which 

moment it disappeared completely. 

Closure of the feedwater valves also implied closure of the condenser valve 

(simulating a turbine trip), at the same time as opening of the steam generator 

relief valves. However, the valve on the intact loop steam generator did not 

open, as a result of which steam could be dumped via the relief valve on the

11

aq
broken loop steam generator. The scheduled opening procedure and the 

procedure actually performed are shown in Figure 3.1. 

As a result of closure of the condenser valve and of failure of the steam dump, 

the secondary pressure and temperature increased, causing the temperature of 

the primary to increase also. This heating of the primary gave rise to an 

expansion in the pressurizer, causing the PORV and SRV setpoints to be reached. 

The opening of these two valves served to control pressurizer pressure, as a 

result of which this pressure decreased until the valves closed. 

Following the closure of the valves, the core was still active, so there was a 

second pressure increase which caused the valves to re-open. Pressurization of 

the primary, which in turn caused a reduction in the differential pressure between 

the seal water injection and the primary, caused the MCP's to trip as a result of 

the pump protection system. This protection system is equipped with brakes that 

0stop the pumps immediately. In less than half a second the pumps stopped 

completely, forced circulation suddenly being lost. This effect of sudden 

shutdown may be seen in the fluid velocity graphics. The core continued at full 

power.

12
3.2. Primary overpressurization with the pumps shut down.

As a result of the shutdown of the pumps, the mass flow in the primary loops 

decreased rapidly. On the secondary side, the loss of forced flow caused a drop 

in pressure, as a result of the decrease in primary-secondary heat transfer. This 

led to a sudden increase in primary pressure, with the pressurizer valves open 

and discharging steam. 

The flow across the pressurizer valves increased considerably. These valves were 

capable of controlling the pressure in the pressurizer, but not in the primary. 

Consequently, pressurizer pressure decreased and the pressure in the primary 

increased. This was due to the fact that the maximum flow along the surge line 

was lower than the maximum flow across the valves, as a result of which there 

was a mismatch between the pressurizer and primary pressures. 

All the absolute pressure measuring instruments went off scale, due to the 

calibration range being limited to 17.4 MPa. The only exception to this were the 

pressure measurements in the pressurizer and the upper plenum, where there 

were wide range devices (with KDG extension). 

3.3. Loss of inventory (LOCA) across the rupture disk. 

The increase in primary pressure led to the break of the rupture disk, causing a 

loss of inventory (LOCA) at high pressure and with the core at full power. 

The sudden depressurization that occurred as a result of the disk rupturing, in 

combination with the high temperatures in the core, caused the coolant to reach 

saturation temperature and to vaporize. The location of the break, along with the 

loss of forced circulation, caused flow reversal to occur in the hot legs and core. 

The phase change caused resistance coolant heat transfer to worsen drastically,

13
as a result of which the temperature in the resistances rose sharply, a 

temperature of 645.7'C being reached in the central part of the core. 

The manual scram of the core which occurred at 23.5 s (it was originally 

scheduled at 59 s) limited the maximum temperature reached in the resistances. 

Certain areas of the core, and the hot legs, underwent condensation and 

rewetting. 

The pressurizer dried out at 40 s, causing an increase in volumetric flow along 

the surge line, due to phase change. 

At 50 s reversal of the heat sink occurred. The primary pressure and temperature 

became lower than those in the secondary, as a result of which the secondary 

ceased to be the heat sink and began to transfer heat to the primary. From this 

moment onwards, there was stratification in the loops as a result of the presence 

of overheated steam and water in saturation conditions, as may be deduced from 

S the measurements taken by the different thermocouples 1 . 

At 60 s the steam line isolation valves closed. This implied a disconnection 

between the two steam generators, which until that time had shared the only 

open relief valve. The intact loop lost all steam dump capacity and could be 

cooled only via the primary system. The BL steam generator was now able to 

release more steam, thus increasing its depressurization. 

According to the scheduled sequence, steam generator dry out was foreseen at 

100 s; however, due to a series of unforeseen events, this did not occur. The 

AFW did not actuate at any time during the sequence. 

At the end of the loss of inventory phase, the primary circuit was completely 

depressurized, while the secondary continued to be pressurized.

14
To" CU CUMXdESM RNEU 

at 

TZ=E = t 

Fui a d of sAT 60 ST 

Figure :3.1. Schematic diagram of secondary operation during Test BT-56

15

PROCEDURE 

7

T THE WNDENSER

REAL BEHAVIOUR
4. DESCRIPTION OF THE PLANT MODEL FOR RELAP5/MOD3.2 

The RELAP5/MOD3.2 code was used for this analysis, running on a Digital 

AlphaServer 2000 4/200 computer under the DIGITAL UNIX operating system. 

With a view to creating the data input file for the LOBI-MOD2 facility, Dr. C. 

Addabbo of the Joint Research Center (Ispra) was requested to supply the 

information available on Test BT-56, and the following was received: 

" A disk containing the nodalization file for RELAP5/MOD2.5 plus another 

file with the nodalization for RELAP5/MOD3.0 of the LOBI-MOD2 

reactor. These files were prepared for Test BT-1 7 (Loss of Feedwater 

Transient, LOFW). 

"* A disk containing the measures obtained during the BT-56 experiment. 

0 * Reports on the experimental results for BT-56 (References /1/, /2/ and 

/3/). Subsequently, information was received on the control systems /4/ 

and on the characteristics of the main system valves /5/. 

The plant nodalization used in the present analysis is shown in Figure 4.1. The 

nodalization is organized on the basis of various categories, depending on the 

number of components: primary system piping and components of the intact 

loops (series 100) and broken loop (series 200), pressurizer and surge line (series 

400), reactor vessel (series 300), intact loop steam generator secondary side 

(series 500) and the same for the broken loop (series 600), emergency core 

cooling systems (series 700 and 800) and discharge lines and containment 

(series 900). 

Nodes 363, 364, 365, 375, 370 and 380 correspond to the vessel head, which 

is an external assembly, as may be appreciated in Figure 4.2.

16
The starting point used consisted of the inputs for BT-17 for RELAP5/MOD2.5. 

This file incorporated the control systems and plant trips established for this 

experiment, including the following among others: 

"* Steam generator level control by FW control 

"* PRZR level control by removal from the upper plenum and make-up via 

the pump seals 

"* Adjustment of circulation ratio by means of valves 520 and 620 

"* Power generated by the electrical resistances in the core and the 

pressurizer heaters 

"" Pressurizer pressure control system (Cooling Coils) simulated by a 

control volume isolated during the transient (tmdpvol 430 and valve 

425) 

"* Heat leakage throughout the system 

Comparison of this file with the general model for LOBI-MOD2 /3/ showed a 

series of changes made at the JRC for the simulation of experiment BT-1 7: 

"* Pressurizer located on broken loop 

"* More detailed nodalization in the steam generators 

"* Replacement of safety injection for the accumulators 

"* Added volumes and junctions for the simulation of leakages 

Although experiment BT-1 7 was very similar to BT-56, it was also a Loss of 

Feedwater scenario, there were significant differences between the two. This

17
implied an in-depth review of the Ispra model for the LOBI-MOD2 facility and its 

modification on the basis of References /1 /,/2/,/4/,/5/ and /7/, adapting it for the 

RELAP5/MOD3.2 version. The main modifications made to the original model 

were as follows: 

"* Initial temperature, pressure and flow conditions in the primary and 

secondary, under steady-state conditions (initialization). 

"* Differences in feedwater temperatures (bounding conditions). 

"* Activation and control of the pressurizer heaters and core electrical 

resistances (different sequence). 

* Change in reference levels and circulation ratios in the steam generators 

under steady-state conditions. 

* Adjustment of the by-pass of the upper part of the vessel and 

associated pressure loss coefficients. 

"* Change in the area of the pressurizer relief and safety valves (PORV and 

SRV), pressure loss coefficients (/l/ and /5/). 

" Adjustment of primary-secondary heat transfer by correcting the tube 

fouling factor (present in version 3.2 of the code). No steam generator 

tube bundle structure was incorporated, since the geometry of both the 

intact and broken loops was inadequate for simulation with this type of 

structure (few tubes, arranged in an annular configuration). 

" Adjustment of the temperatures in the walls of the core resistances 

through correction of the fouling factor in the heat structures. No 

bundle structure has been incorporated for reasons similar to those

18
presented above in relation to the steam generators (mesh of few 

tubes). 

"* Adjustment of pressure loss coefficients and discharge coefficients 

across the broken loop steam generator relief valve. 

"* Modifications to vessel head nodalization. 

"* Incorporation of detailed nodalization of the rupture disk and 

corresponding discharge line. 

"* Incorporation of steam line closure or isolation valves. 

"* Control of equipment operation in accordance with the transient 

sequence. 

co

19

Figure 4.2. Reactor vessel head

21

o

ixotatio• fkLon
5. STEADY-STATE SIMULATION

Adjustment of the steady state of the model developed for RELAP5/MOD3.2 of 

the LOBI-MOD2 reactor is the first step to correct simulation of the transient 

analyzed. 

This adjustment is based on the information provided by the wide range of 

instrumentation with which the facility is equipped during the 20 seconds prior 

to initiation of the transient. During this period, the system was in steady-state 

conditions with slight oscillation of its values. 

This section presents a brief analysis of the steady-state data for Test BT-56 /2/, 

along with the results obtained using the RELAP5 model following the 

adjustment. 

5.1. Analysis of the Primary 

In general there are few absolute pressure measuring devices in the system, 

although there are quite a number of differential pressure measurements, these 

having been used for adjustment of pressure losses under steady-state 

conditions. 

The set of temperature measurements available is much more complete than that 

of absolute pressure. Comparison of the temperature measurements of the two 

loops shows that the hot leg temperatures are different, due to the presence of 

the by-pass in the vessel. According to the information obtained from 

reference/7/, there are three different by-pass paths: 

* vessel head by-pass 

* by-pass at the elevation of the lines due to nozzle losses 

* by-pass due to the presence of 5mm orifices connecting the 

downcomer and the upper plenum.

22
From the experimental data /2/it may be deduced that the effect of the by-pass 

mainly affects the hot leg of the intact loop. Thus, it is observed that hot leg of 

the broken loop shows the same temperature as the upper plenum, while the 

intact loop hot leg temperature is 4 'C lower. The nodalization and pressure 

losses in the upper part and the head of the vessel of the model for 

RELAP5/MOD3.2 were adjusted such that the effect of the by-pass was 

practically all on the intact loop, correct temperature simulation being obtained 

(as shown in Table 5.3.1). 

Mention should be made also of the asymmetry of heat removal via the 

secondary, this being due to the fact that the steam generators are very different 

and that steam flow control is common to both. This control uses the cold leg 

temperatures of both loops weighed in terms of their flows, regulating the only 

steam exit valve to the condenser. All the above was taken into account in the 

simulation by means of the models of the steam generators and their turbine inlet 

control system. 

Flows converge on the reactor vessel at different temperatures: the cold leg 

flows of both loops and the flow resulting from the vessel head by-pass. The 

lowest temperature is the head by-pass, which is some 40 C lower than the cold 

leg temperature. This means that the fluid at the inlet to the downcomer 

presents different temperatures, depending on the angle of measurement. The 

angles at which the instrumentation is located at different elevations varies but, 

as may be seen, the fluid is mixed as it flows along the downcomer. In view of 

the single dimension nodalization performed by RELAP5, the simulation cannot 

reproduce this type of effect. 

The temperature in the walls of the core resistances is measured by means of 

thermocouples located at different elevations (levels). The adjustment of the 

temperatures in the walls of the core heat structures was obtained by adjusting 

the fouling factor for these structures. Determination of this parameter was 

based on the study attached in Appendix I.

23
5.2. Analysis of the Secondary

The pressure in the dome of the intact loop steam generator shows a slightly 

higher value than that in the broken loop, despite the two being connected to a 

common steam outlet. This is due to the different pressure losses presented by 

the steam lines, due in turn to the flows and diameters of the pipes being 

different. The pressure losses in the steam lines were adjusted in the model in 

order to take this effect into account. Mention should be made of the fact that 

during the last 20 seconds of the steady state the secondary pressure oscillated 

significantly. 

Although the feedwater for both steam generators comes from one same tank, 

the experimental data show that the temperature of this water is different in the 

two generators, as was considered in the simulation. The steam outlet 

temperature for both loops was, however, practically identical. Furthermore, it 

should be pointed out that the steam temperature measured by thermocouples 

TF87S and TF97S for each of the loops is 2 °C higher than the saturation 

temperature corresponding to the pressure measured by the respective pressure 

measuring (PA87S and PA97S). 

5.3. Steady-State Results 

The steady state is the starting point for analysis of the transient. The calculation 

performed was for 500 s with an algorithm for time control implicit in heat 

transfer and hydrodynamics (flag 7). Table 5.3.1 shows a comparison between 

the plant measurements for Test BT-56 and the steady-state results calculated 

using RELAP5/MOD3.2. The values specified by JRC for Test BT-56 showed a 

slight variation with respect to those obtained during performance of the 

experiment. 

Specified Test BT-56 RELAP5/MOD3.2 Units 

Primary System 

Pressure in upper plenum 15.8 15.87 15.85 MPa 

Core power 5.28 5.23 5.23 MW

24
Table 5.3.1. Steady-state results

25

Intact loop 

- Mass flow 21.0 20.9 20.9 Kg/s - Vessel inlet temperature 294 296 295.56 °c - Vessel outlet temperature 326 326 326.36 0C 

Broken loop - Mass flow 7.00 7.30 7.30 Kg/s - Vessel inlet temperature 294 296 296.96 0C - Vessel outlet temperature 326 329 329.97 0C 

Pressurizer 

- Liquid level 4.80 5.10 5.10 m - Temperature 346 348 350.77 0C 

MCP seal flow injection - Intact loop - 0.01 0.015 Kg/s - Broken loop 0.005 0.005 Kg/s - Temperature - 30 30 0C 

Secondary System 

Intact SG loop - Steam pressure in dome 6.45 6.52 6.52 MPa 

-Mass FW flow 2.00 2.00 1.96 Kg/s - Inlet temperature 210 207 206.85 0C - Outlet temperature 280 281 281.07 0C - Level in downcomer 8.00 8.07 8.07 m - Circulation ratio - 6.40 6.40 

Intact SG loop - Steam pressure in dome 6.45 6.51 6.52 MPa - Mass FW flow 0.67 0.76 0.73 Kg/s - Inlet temperature 210 204 204.05 0C - Outlet temperature 280 281 281.04 0C - Level in downcomer 8.40 8.45 8.45 m - Circulation ratio 4.3 4.3
6. TRANSIENT SIMULATION

This section describes the results of simulation of the transient, explaining the 

evolution of the main variables, such as pressures, temperatures and densities. 

The following sections contain information on the simulation performed and the 

results obtained, as well as on the statistics obtained from calculation CPU 

consumption. 

6.1. Characteristics of Transient Simulation 

The 500 s steady-state having been obtained with the model of the facility for 

RELAP5/MOD3.2, suitably adjusted, simulation of the accident sequence began. 

A 100 s restart transient type input was used for this new calculation, including 

the pertinent changes for adjustment of the transient. 

* The feedwater flow in both steam generators was adjusted during the 

transient. For this purpose a flow table was introduced based on time 

for each SG, in accordance with the plant measurements. This implied 

cancellation of the feedwater flow control used for steady-state 

conditions. 

Closure of the turbine valve and opening of the broken loop steam 

generator relief valve were simulated by means of time-activated trips. 

The coefficients of pressure loss for the relief valve are those referenced 

by JRC /5/. The discharge coefficients across this valve were adjusted 

in order to fit steam flow. 

* On the primary side, the dimensions of the PORV and the SRV were 

changed in accordance with the experiment documentation. The 

pressure loss coefficients were adjusted depending on the fraction of 

opening, in accordance with curve /1/. In order to make the valves more 

realistic, the decision was taken to simulate them by means of the

26
"Ifmotorvalve" model, which allows opening time to be added. The PORV 

was assumed to be the slower of the valves, the opening time included 

being considerably longer than that of the SRV. 

" Time-based tripping of the main pumps was simulated. The post-trip 

pump speed curve was simulated by introducing the pump revolutions 

per minute curve provided by JRC /2/ for the BT-56 experiment. The 

head losses associated with these pumps following shutdown were also 

simulated. 

" As regards tripping of the electrical core, a minor decrease in power 

was simulated prior to the trip. As from this moment, the electrical 

power drops suddenly, reaching zero almost instantaneously. All the 

above was simulated using a power versus time curve, with the values 

measured at JRC for the BT-56 experiment /2/. 

& The rupture disk was simulated by means of an instantaneously opening 

valve (tripvalve). The rupture disk discharge coefficients were adjusted. 

" The seal flows were simulated by means of time-based tables, in 

accordance with the measurements obtained during the BT-56 

experiment. 

"* Extraction from the upper plenum was cancelled at the beginning of the 

transient, in accordance with the experiment documentation. 

"* The HPI and ACC systems were cancelled out, since they did not 

actuate during the transient. 

"* The pressurizer pressure control system (Cooling Coils) was cancelled 

out, since it did not actuate during the transient.

27
" Pressure loss coefficients were introduced for reversed flow (K reverse) 

at the vessel inlet from the hot legs, as well as at the inlet to the BL 

pump, in order to adjust the velocities in the loops. 

"* The pressure losses in the surge line were adjusted in order to correctly 

simulate the mismatch between the primary and pressurizer pressures. 

" The transient lOOs simulation was performed using the calculation 

algorithm implicit in heat transfer and hydrodynamics time (flag 7), this 

ensuring correct heat transfer. 

6.2. Transient Simulation Results 

The graphs included in Appendix II show a comparison between the evolution of 

Test BT-56 and the results of the transient simulation performed using 

0 RELAP5/MOD3.2. The transient lasted 100 s, as a result of which the last 20 

seconds of steady-state conditions are represented in the graphs(480 s to 500 s) 

along with the 100 s of the transient (500 s to 600 s). 

The main results of this comparison are as follows: 

"* Analysis of the graphs shows that the simulation of the transient using 

RELAP5/MOD3.2 presents good agreement as regards the evolution of 

the heating and the primary pressure peaks during the initial phase of 

the transient (fig. 1 to 4). The times obtained for opening of the valves 

and rupturing of the rupture disk coincide with those measured during 

Test BT-56. 

" The depressurization of the primary following rupturing of the disk (after 

19.1 s) was correctly simulated, although it may be observed that the 

primary pressure predicted by RELAP5 is slightly higher than that 

measured in the facility, this meaning that the primary becomes the

28
heat sink 5 s later (fig 21). Nevertheless, the simulation presents a more 

pronounced decrease than that measured at the plant during the 

interval 55-70 s into the transient. This is due to an increase in the flow 

across the rupture disk (fig 41), caused by a change in phase (SLG to 

ANM) in the upper part of the downcomer (fig 42). 

" The pressure mismatch and drying out of the pressurizer were correctly 

simulated, and the time taken for the pressure to converge with that of 

the primary agreed with the experimental values (fig 20). 

" The effects of drying out of the primary and of flow reversal in the hot 

legs and core were correctly simulated (fig. 5 to 8). Mention should be 

made of the fact that the plant QF reference meters measure the 

velocities of two-phase, non-stratified flows and that they do not 

measure direction (this measure is always positive). 

* The temperatures in the legs provided by the simulation are lower than 

the temperatures measured at the plant, as from 50 s into the transient 

(fig 9 to 12). This is due to the effect of the higher depressurization 

contemplated by the code for this interval. 

"* Following reversal of primary-secondary heat transfer, a temperature 

step is observed the broken leg hot leg (fig 11), following steam 

generator temperature. This effect is smoother in the simulation with 

RELAP5, due to lower heat transfer, and also occurs 5 s later. 

" The thermal stratification that occurs in the legs as from 60 s into the 

transient, caused by the presence of superheated steam and liquid 

under saturation conditions, is qualitatively reproduced with RELAP5. 

The temperature obtained with RELAP5 for the superheated steam 

present in the hot legs (fig 9,11) shows a lower value than the 

experimental data. This effect could be due to the lower secondary to 

primary heat transfer in the simulation calculations.

29
" The core temperatures obtained with RELAP5 (fig 27 to 38) are lower 

as from 50 s into the transient, due fundamentally to the previously 

described hot leg temperature effect. 

" The analysis of densities shows that the simulation presents a good 

adjustment in the legs (fig 23 to 26), although the core rewet shown is 

lower than that which occurs in the plant (fig 22). 

" From analysis of the results it would appear to be the case that the 

secondary-primary heat transfer calculated by RELAP5 is lower than 

that actually occurring. This is due to the complexity of the 

phenomenon simulated, with the presence of steam inside the tubes 

and reversed flows in the hot legs. There is also high sensitivity to flow 

velocities in the legs. 

(-4 

e The evolution of the pressures (fig 19) and of the volumetric steam flow 

in the secondary (fig 17) were correctly simulated. 

6.3. Run Statistics 

The transient base case was performed on a Digital AlphaServer 2000 4/200 

platform using the DIGITAL UNIX operating system. 

The time elapse control algorithm used for the simulation with RELAP5/MOD3.2 

is implicit in heat transfer and hydrodynamics (flag 7). 

The maximum time step allowed was 0.05 s for the first 18 seconds of the 

transient and 0.001 s for the rest, although the simulation reduced this 

parameter as from 50 s. The CPU consumption time was 13773 seconds, giving 

a ratio of real time to CPU time of 138/1. 

7. SENSITIVITY CALCULATIONS

30
Several sensitivity calculations were performed to study the importance of 

certain phenomena in the response of the model. The phenomena analyzed were: 

heat transfer in the steam generators and coefficients of discharge across the 

rupture disk and the BL steam generator relief valve. The graphs with the main 

results are included in Appendix II. 

7.1. Coefficients of Discharge across the Rupture Disk 

In this sensitivity calculation, an analysis was made of the response of the model 

to the coefficients of discharge. The rupture disk coefficients for the base case 

were adjusted as recommended for the code /8/. 

Variations in the subcooled and superheated discharge coefficients causes slight 

changes in the evolution of primary pressure and temperature, but variations in 

the two-phase discharge coefficient produces significant effects. The sensitivity 

calculations included here consists of reducing the two-phase flow discharge 

coefficient: 

Coefficients of discharge Subcooled liquid Two-phase flow Superheated steam 

across rupture disk 

Base Case 0.8 1.2 1.0 

(recommended) 

Sensitivity 0.8 0.9 1.0 

Variations are observed in the evolution of the primary pressure (fig. 43 vs fig.4) 

as a result of the change in the two-phase discharge coefficient (from 20 to 70 

s), without any significant influence on the secondary (fig. 46 vs fig.19). 

A significant effect is observed on the temperature of the electrical resistances in 

the core (fig. 44 vs fig. 30), with the maximum value reached increasing and 

cooldown delayed during the period of two-phase flow discharge. Differences are 

also observed in the cold leg temperatures (fig 45 vs fig. 12) after heat sink 

reversal.

31
7.2. Heat Transfer in the Steam Generators

The calculation performed consists of increase the heat transfer in the steam 

generators with respect to the base case. For this purpose, the value of the 

fouling factor was adjusted to 1.0 (steady-state and transient). The effect of 

fouling during reversal of the heat sink causes slight changes over heating of the 

steam in the legs (fig. 49 vs fig. 12) and the core (fig. 48 vs fig.30), without 

any significant influence on the secondary (fig. 50 vs fig. 1 9). 

7.3. Coefficients of Discharge across the SG Relief Valve (BL) 

In this sensitivity calculation, an analysis was made of the response of the model 

with varying coefficients of discharge across the BL steam generator relief valve. 

Variation in the subcooled and two-phase flow discharge coefficients causes 

slight effects on the secondary evolution, so the sensitivity calculations consists 

on change the steam flow discharge coefficient. 

Coefficients of discharge Subcooled liquid Two-phase flow Superheated steam 

across relief valves 

Base Case 0.8 1.2 0.925 

Sensitivity 0.8 1.2 1.1 

The main differences are observed in the steam flow across the relief valve (fig. 

53 vs fig. 17) and the pressure (fig. 54 vs fig. 19) in the steam generators, 

without any significant influence on the primary (fig 51, 52).

32
8. CONCLUSIONS

The LOBI-MOD2 reactor BT-56 experiment was unusual due to the multiple 

malfunctions that occurred during its performance, which caused a major 

deviation in the sequence with respect to the original approach. 

The unexpected evolution of the sequence and the variety of events and 

phenomena that coincided made the result especially interesting for 

understanding of plant accident phenomenology, as well as for verification of the 

behaviour of the thermalhydraulic simulation codes. 

The information obtained from the LOBI-MOD2 experimental facility has made it 

possible to adequately develop and adjust the model corresponding to the facility 

for RELAP5/MOD3.2. The calculations performed using the RELAP5 code have 

served to clear up some of the uncertainties encountered during transient 

analysis and development of the model, as well as to explain and confirm the 

0 phenomenology deduced from the experimental measures. In addition, the 

measures obtained from the facility have made it possible to verify the behaviour 

of the code when reproducing a transient as rapid and complex as the one dealt 

with here. 

The simulation performed using RELAP5/MOD3.2 have faithfully reproduced the 

evolution of pressures as a result of closure of the condenser valve and 

shutdown of the primary pumps, as well as depressurization due to opening of 

the pressurizer valves and the break of the rupture disk. 

The entire chain of sequences occurring during the experiment has also been 

correctly simulated: loss of forced circulation, rapid primary pressurization, 

depressurization of the secondary, high-pressure dryout and rewetting of the 

core. The time taken for the pressurizer to empty has been adequately 

calculated, as has the mismatch between the pressurizer and the primary circuit 

and subsequent convergence.

33
Likewise, RELAP5/MOD3.2 has acceptably reproduced heat transfer from the 

core and the evolution of temperatures in the walls of the rods. It should be 

observed that the three-dimensional temperature distribution effects in the core 

region cannot be simulated with RELAP5, since it is a single dimension model. 

The phenomena that present the greatest uncertainty and a certain deviation 

with respect to what was deduced from the experimental data are: flow across 

the rupture disk (strongly influenced by the phase change occurring in the 

downcomer), primary-secondary transfer (with reversal of the heat sink) and 

condensation in the primary, causing core rewet. 

A sensitivity analysis was performed on those parameters of the model showing 

uncertainties, with significant sensitivity observed in the evolution of pressures 

and temperatures in the core with respect to adjustment of the factors of 

discharge across the rupture disk. Mention should be made of the fact that the 

values assigned to the coefficients of discharge for better adjustment of the 

sequence are acceptable with respect to those referenced in the literature. 

Judging by the results obtained, the RELAP5/MOD3.2 code demonstrates the 

capacity to satisfactorily reproduce a transient of these characteristics, showing 

adequate behaviour in simulation of the associated phenomenology.

34
9. REFERENCES

* /1/ "Quick look report on LOBI-MOD2 Test BT-56". Commission of the 

European Communities. JRC. No.: 4356. LQC 92-71. February 1994. 

* /2/ "Experimental data report. LOBI-MOd2 Test BT-56". Commission of 

the European Communities. JRC. No.: 4352. LEC 93-73. December 

1993. 

* /3/ "RELAP5 Base Input Data for LOBI-MOD2". Commission of the 

European Communities. JRC. No.: 4030. LPC 85-18. February 1985. 

* /4/ "LOBI-MOD2 Control Systems" Technical Note No. I. 06.C1.86.10, 

R. Wampach, February 1986 

* /5/ Technical information on LOBI valves. 1613/308/87/BD-GFDS/bs. 

Referenced to TPTF-Meeting "LOBI Valves Characteristics". Ispra, 

10/9/1987. 

* /6/ "Test Specification LOBI-MOD2 Test BT-15/BT-16" C.Addabbo, 

LOBI. June 1, 1990. 

* /7/ "Analysis of the Core Bypass in LOBI-MOD2 Test BC-04" 

Commission of the European Communities. JRC. No.: 4330. October 

1990. 

* /8/ "RELAP5/MOD3.2 Code Manual. Volumen V: User's Guidelines". 

NUREG/CR-5535-V5. June 1995.

35
APPENDIX I: AVERAGE TEMPERATURES IN CORE HEAT STRUCTURES

Level - 1 

No of TH32D301 (*C TH35D301 (

0CQ AVERAGE (0 C) 

thermocouples 

2 314 318 316 

Level- 2 

No of TH34D402 (

0

C) TH35F402 (

0

CQ AVERAGE (0 Q) 

thermocouples 

2 323 322 322,5 

Level - 3 

No of TH32C103 (TC) TH34E103 (

0C) AVERAGE (0 C) 

thermocouples 

2 320 322 321 

Level - 4 

No of TH37C404 (°0) TH37G404 (00) AVERAGE (0 C) 

thermocouples I 

2 326 327 326,5 

Level - 5 

No of TH35D105 (00) TH37F105 (0C) AVERAGE (0 C) 

thermocouples I 

2 341 335 338 

Level - 6 

No of TH34D106 (1C) TH35F106 (01 AVERAGE (00) 

thermocouples II

cq
2 344 346 345

Level - 7 

No of 

thermocouples 

4

Level - 8 

No of 

thermocouples 

4

TH32C407 (0C) TH34E407 (0C) 

353 354

TH32A208 (°C) TH32E208 (0C) 

342 349

-k

+

TH37A407 (0C) TH37E407 (OC) 

341 345

TH34G2086(C) TH35E208 (0C) 

352 355

AVERAGE (0 C) 

348,25

+

AVERAGE (0 0) 

349,5

Level- 9

In,-

N1Jo 

'thermocouples

6 ITH31G209 (OC) TH34B209 (°C) TH35D209 (°C) 

352 341 350

TH37F209 (0C) 

351

TH38C209 (°C) AVERAGE (0 C) 

345 349

Level - 10

No 

thermocouples

I- 6

of TH31A21o (0C) TH32F210 (0C) 

349 352

TH33G210 (0C) 

352

TH34D210 (

0

C) 

355

TH35B210 (0C) 

343

TH36G210 (0C) AVERAGE (o C) 

356 351

Level - 11 

No of TH31B311 (0C) TH32C311 (8C) TH33H311 (00) TH34E311 (°C) 

thermocouples 

6 346 348 349 350

TH37E311 (OC)

343

TH38F311 (C) AVERAGE (0 C) 

335 345

Level - 12 

No of TH32E312 (

0

C) TH34G312 (1C) TH35A312 (°0) TH35E312 (0C) 

thermocouples 

6 351 351 335 349 345

TH38H312 (0C) AVERAGE (0 Q) 

332 344

Level - 17

I

I

i

of TH31 C209 ( C)

356

o.

TH36B3312 (°C)

I

I
No of TH33C217 (°C) 

thermocouples 

1 336 

Level - 18 

No of TH31D618 (QC) TH35H518 (OC) AVERAGE (0 C) 

thermocouples 1 

2 335 338 336,5 

Level - 19 

No of TH33E119 (

0

C) TH36D319 (

0

C) AVERAGE (0 CQ 

thermocouples 1 

2 312 333 322,5 

* Values of measurements in Test BT-56 LOBI-MOD2 /2/. 

(N
Length (m) Avg, temp. (IC) Measure level 

0 316 Level - 1 

0,16 322,5 Level - 2 

0,44 321 Level - 3 

0,66 326,5 Level - 4 

1,16 338 Level - 5 

1,66 345 Level - 6 

2,16 348,25 Level - 7 

2,66 349,5 Level - 8 

3,16 349 Level - 9 

3,38 351 Level - 10 

3,66 345 Level - 11 

3,82 344 Level - 12 

4,71 336 Level - 17 

5,51 336,5 Level - 18 

6,17 322,5 Level - 19

* The point of origin was taken at the first thermocouple level

level-1

Heated zone 

Upper plenum

level-1 2 

level-1 6 

level- 19

a

39
AVERAGE TEMPERATURE IN THE WALL OF CORE HEAT 

STRUCTURES

AVG. TEMP. (-C)

360 

350 

340 

330 

320 

310 

300 

290

0

0,16 0,44 0.66 1,16 1,66 2,16 2,66 3,16 3,38 3,66 3,82 4,71 5,51 6.17 

LENGTH (m)

Figure A.II.1. Average temperature in the wall of core heat structures 

coI NMMEAT M•Uc unE $ THE RELAP MODEL

A.II.2. Core nodalization

40

a
APPENDIX II: SIMULATIONS GRAPHICS

41
BT56 Transient Comparison LOBI/RELAP5. Base case

Figure A.II. 1.- Pressure in IL Cold Leg

10.0 ....... : ....... ....... I .......

0.0 4 

20.0 

15.0 

10.0 

5.0

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

80 490 500 510 520 530 540 550 560 570 580 590 60 

Time (s) 

Figure A.II.3.- Pressurizer Pressure

0

Figure A.ll.2.- Pressure in BL Cold Leg

20.0

*-* LOBI-PA16 

O---ORe ap5

15.0 

"10. 

(. 

5.0

80 490 500 510 520 530 540 550 560 570 580 590 600 Time (s) 

Figure A.ll.4.- Pressure in Upper Plenum

[ LOI-PA40KDG 

10 Relap5

15.0 

10.0 

5.0

480 490 500 510 520 530 540 550 560 570 580 590 80 

Time (s)

20.0 

15.0

N,:

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

I * LOBI-PA38KDG ---OReap5

480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

^^

* LOBI-PA26 

1---ORelap5 I

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

S. .. .., .. . ....... I.......

..4 t I I

S....... •........:-....... ........ t.......

I...Li.L..!

...L .• .... . .. ... .... ......... .. ... r". .. .. . .... .. .. . . . .. 

... ...... .. . ... ....... ....... .............. .
Figure A.II.5.- Flow Velocity in HL

BT56 Transient Comparison LOBI/RELAP5. Base case 

IL Figure A.II.6.- Flow Velocity in CL IL

8.0 

6.0 

4.0 

2.0 

0.0 

-2.0 4,

8.0

*TLOBI-QFI2V 

G----O Relap55

LOBI-QF22V 

O0-----ORe ap5

0.........4........................ .......................... 

................... iiiii ... ------- ---

6.0 

6 4.0 

0 

UL 

2.0

I.

B0 490 500 510 520 530 540 550 560 570 580 590 6 

Time (s) 

Figure A.IT.8.- Flow Velocity in CL BL

O0

U."

480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

80 490 500 510 520 530 540 550 560 570 580 590 6 

Time (s) 

Figure A.II.7.- Flow Velocity in HL BL

I I

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

43.

8.0 

6.0 

4.0 

2.0 

0.0 

-2.0 48

S..... ....... .. • .............. . . . .. . .. . . ... .. . . ....... I... ..

I * G TOBI-QM24V -Relap5 I

I

590 600

• . . . .. . . .. . .,.......... .. . .t....... 4 ....... 

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

S... .. .. ...... ...... ......

0.0 4; 

8.0 

6.0 

"Q 4.0 

0 

2.0

0O 490 500

S.....r....... .. . ....... &.. 

0---ORelap5 -- LOBI-QF14 
BT56 Transient Comparison LOBI/RELAP5. Base case

00.0 

00.0 

00.0

0.n-n

Figure A.1I.9.- Temperature in HL IL

480 490 500 510 520 530 540 550 

•- mTime (s) •:: Figure A.II. 11.- Temperature in HL BL

Figure A.l.l10.- Temperature in CL IL

560 570 580 590 6G

* *LOBI-TFI IH000 

X-MXLOBI-TFIiHi80 

0-0 RelapS-gas 

0-ORelap5-liq

400.0

5 300.0 

C, 

E 

(D 

F- 200.0

0 80 490 500 510 520 530 540 550 

Time (s) 

Figure A.I. 12.- Temperature in CL BL

560

-LOBI-TF16l1000 

LOBI-'171TI6llS0 

G -ORelap5-gas 

0-ORelap5-liq

*-- LOBI-T2 IH000 I 

*-9( LOBI-T2IH1iS0 

----ORelap5-gas 

0-ORelap5-liq

p300.0 

E 

HT 200.0

utLu 00.0 1 480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

.- A I

-LOBI-TP26H1000 

xLOBI3-1T126H 180 

G-ORelap5-gas 

O-ORelap5-lq

480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

00.0

00.0 

00.0

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

............... ............... ....... ........ .... .................. I ...... I ................ 

. . . . . . . . . . 

. . . . . . . . . . 

. . . . . . . . . . 

. . . . . . . . . . .

S.... ................ ..... r........, ........ 1........r..... .. ....... =. . . .

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

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

. ........ . i. .. . ...... -- - - - -- . .

I
BT56 Transient Comparison LOBI/RELAP5. Base case

Figure A.II.13,- Core Inlet Temperature Figure A.I. 14.- Core Outlet Temperature 00.0 

00.0. 

00.0

480 490 500 510 520 530 540 550 560 570 580 590 6( 

Time (s) Figure A.I. 15.- Upper Plenum Temperature 

00.0 

00.0480 490 500 510 520 530 540 550 560 

Time (s)

"•-- IOBI-TM 36V 165 

SRelap5-gas 

-- ORelap5-liq

400.0

& 300.0 

0) 

E 

-. 200.0 

100.0 0 4

- LOBI-'IV38H000 

--.-- Relap5-gas 

0-ORe ap5-lq

0 

U.

2.0 ........ . ....... ....... ....... 

1.0 _. 

0 .0 .... .....

-1.0 489

80 490 500 510 520 530 540 550 

Time (s) 

Figure A.II. 16.-BL-,IL Feed Water Flow

*"* LOBI-TtP37V 165 

*-.. RelapS-gas 

ORelap5-liq --

560 570 580 590 600 

- LOBI-QM63EIDO 

-LOBI-QM64EFDO 

O--ORelap-IL 

0--- Re ap-BL

57058 50 9049 50051055 2 3 4 56 57 58 59 0 

520 530 540 Time (s)

4a,

S........•........1 ....... .......

* MW f* I L90 W 6 QM -4

:.. 2 ..--5 :-•; .- ... ==• ....... , ...... ,,...... •,....... •........ 

...... .I ......... ..L....-....... .. ....... i ...... ....... ....... " ". . . .. . . . L . .. . . . . ..... . . -....... , ....... ....... ,....... ,. . ... , ..... . . . .

3,0

0 570 580 59D 600 490 500 510 550 560 670 580 590 600
BT56 Transient Comparison LOBI/RELAP5. Base case

Figure A.ll. 17.- IL-BL Steam Volumetric Flow

0.08 

0.06 

0.04 

0,02 

0.00

30 490 500 510 520 530 540 550 560 570 580 590 6 

Time (s) 

Figure A.ll. 19.- IL-BL Steam Generators Pressure

-LOBI-QV65UVX 

-- x LOBI-QV66HVX 

3Relap5-IL 

Re ap5-BL

< 0,010 

E 

0 

E 0o

10.0. .-- "LOBI-PA97S 

.LOBI-PA87S 

9.0 ....... ....... I .................... I....... ............. I ............................ I -O RelapS-IL 

0-O Relap5-BL 

8. ------- . ........ ... .. --- -....... ....... •....... i......... ....... 

6.0 

5.0 . ..................... , ....... I ....... 

6.0 ............... !....... -- ---------...... ...... ...... .. .. . .... ...... 

4.0 .............. ......... • ...... .......................... ". . . ... . . ... . . - 40 : 

4.0 ....... : . . . . ... . ...... , ....... ....... :-...... " ... . .I ....... I - - - - -- -I- - -

2.0 480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

41 

20.0

15.0 

(. 

(0 

5)1.0

Figure A.11.18.- Volumetric Flow in Surge-Line

80 490 500 510 520 530 540 550 560 570 580 590 6& 

Time (s) 

Figure A.ll.20.- Pressurizer vs Upper Plenum Pressure

"--"LOBI-QV4|V SO---ORe ap)5 I

0

-LOBI-PA40KDG 

-T.OBI-PA38KDG 

O--ORelap5-PZR 

O-ORelap5-Upper Plenum

480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

... .. ................. ........... ; .............................. .•....... ;....... ....... 

S....... ......... ...... --.........

--i ....... ...i....... ..*1. --------... .. . • : ... .. ...... ... ... ... ... ... .......... , ..... ....... I........ ............... ........ 

.S....... .... S.L------ - -..-.. . . t... . 

:crcEIII

N ........ .

0015•

S................ . ,. - -. .--......... . t ....... .! , .......1 ....... • ...... " i " '

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

ý1114N.

I I I • I I
BT56 Transient Comparison LOBI/RELAP5. Base case 

Figure A.II.21.- Upper Plenum Vs SG BL Pressure Figure A.H.22.- Core inlet Density

20.0

Figure A.II.23.- IL HL Density 

00.0 

00.0 ............... - ........ .......... 

00.0 ............... ... . 

0.0 ....... ....... ................ ..... .. .... .................. 

00.0 

480 490 500 510 520 530 540 550 560 570 580 590 60 

Time (s) 480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

800.0 

*- *LOBI-PA38KDG 

n-- LOBI-PA66KDG, 

0--0 RelapS-SG BL 

0 Relap5-Up r Plenum S600.0 -- - - -............ ....... .. . . ....... • ......... ............. ....... , ... . , .. . 

< F: .... 40.0 .. ... ... , .. ... , . ... ....... ;....... ,....... ,........ ,......... ....... ,....... ,....... 

6 00.0 . . . . . . . . - -

ZI, 

0.0 ....... ............... , ....... .

-200.0 

0 480 490 500 510 520 530 540 550 560 570 580 590 60 

Time (s) Figure A.ll.24.- IL CL Density 

-- * T.OBI-DDI IHOnTA I . 

x--x LOBI-DDIIHPER 

-0-ORelap5 

600.0 ------- .. . . . . .. ... . ....... . .... .. . . .. . . . . . . ....... ' ....... . . . . 

E¢ 400.0 ....... U) ,....... I ....... ,....... .............. 

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

.... ......

.....

....

t- 200.0 ....... ....... ............... ......... ........ . 

0.0 ....... ,....... •....... ,....... ,....... ,....... ,....... ,..... ....... 

-200.0

O----ORe S-- I,OBR-DS-35VD1Aj p 

4--*T.OBI-DD 16HDIA 

M-4 LOBI-DDI6HPER 

G---ORe ap5

30
BT56 Transient Comparison LOBI/RELAP5. Base case

Figure A.I1.25.- BL HL Density 

00.0 

00.0 -------.------ . .I . - .. -....... .................. ........... 

00O 

0,0 ------- ......... •........ r ------- ......... •...... . . .. I, . . . . . . 

0 

00.0 480 490 500 510 520 530 540 550 560 570 580 590 60o 

! Time (s) 

Co Figure A.I1.27.- Core Heat Structure Temperature - Level I

00.0

00.0 

00.0

480 490 500 510 520 530 540 550 580 570 580 590 600 

Time (s)

Figure A.H1.26.- BL CL Density

*--*LOBI-DD21HDIA 

4-c LOBI-DD2111PER 

3-E Re ap5 

C 

O0) 

t

o,1 C7

800.0 

600.0 

400.0 

200.0 

0.0

480 490 500 510 520 530 540 550 560 570 580 590 6 

Time (s) 

Figure A.ll.28.- Core Heat Structure remperature - Level 2

- LOBI-TH32D301 

LOBI-TT135D301 

0-- Relap5

" 400.0 

E 

E

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

............... ....... ....... .................. I ........ I ....... ....... ....... 

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

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

eu...u

480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

-T LOBTI-DD26HDII 

)0 

"'--- LOBI-TH34D402 

SLOBI-'1113517402 

G---ORe ap5

•vgov @

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

wi Jý*' 

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

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

S. . ........ ., , ....... • ..... J ...' ...... •....... ,....... ,....... ,....... ,....... ........ . . 

J ......,, ....... i ....... i ...... ....... ... .. ..... .......

I
BT56 Transient Comparison LOBI/RELAP5. Base case 

Figure A.11.29.- Core Heat Structure Temperature - Level 3 Figure A.I.30.- Core Heat Structure Temperature - Level 4

00.0 

00.0 

00.0

480 490 500 510 520 530 540 550 560 570 580 590 600 

F1Time (s) 

Figure A.ID.l.- Core Heat Structure Temperature - Level 5 00.0 

00.o 

00.0 .. . . -- - - -- - - - - - - - . . . . . . . . . . . 00.0 ------- . .. . ....... . ..... .. ... ....... .. . .• ....... . ....... I ....... ....... I •. . . 

00.0 

00.0 

480 490 500 510 520 530 540 550 560 570 580 590 60D 

Time (s)

'-- LOBI-TH32CI03 I 

- LOBI-T-134E103 

G-.ORelap5

500.0 

01 

S400.0 

C) 

E 

a) 

I

300.0

S..............

--* IOBI-TH37C404 

LOBI-TH37G404 

O -- Rel a pS 

*-- LOBI-TII34DI06 

LOBI-TH35FI06 

-G--ORe apS

480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

e

. .

IB .

200.0 s i 1 480 490 500 510 520 530 540 550 560 570 580 590 6C 

Time (s) 

Figure A.II.32.- Core Hteat Structure Temperature - Level 6 700.0 *LOBI-T1135DI05 

LOBI-TH37FI05 3-i-O Rela 5 

600.0 .. . . .. . . .. . . -- - - .. . . .. . . .. . . .. . . .. . . -- -

500.0 ........ .... .. J....... t..... .,. . . ....... I ....... , ....... ....... ....... , ....... ....... 

o,500.0 

(l1 

E 490o.0 

300.0 

200.0

^^^

S 1.. ... .. . : . ....... i....... -------. .... .i ....... ii.... ..........
0 000

(1) 

0 

L6 

CL 

w 

0 

0 

C, 

am 

E 

0 

C: 

CD 

L

(o

0 

6 

0

00 

0 

0 

C0 

0 

00 

E 

0 

2 

0 

0 

0 

00 

)0 

0) 

0 

0 

ri-) 

0 

0 

00 

N 

0 

2 0 

0 00 

E 

0 

2 

0 

C) 

00 

0) 

2 

0 U 

0 

0 

00

6 o. o oo q q v 8 8

I 

CL) 

E 

50

000000 

IU 1 0

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

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

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

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

------------ : ------------- :--------- :-------

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

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

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

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

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

............ : ------------ ------------ : ----- .... I ------------

(or) ejiedwe.L

AESSSS 

LU III ----------- ----------------------- -- ----- . ............ 

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

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

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

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

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

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

.................... ...... : ------------ : ............. 

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

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

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

..... .....

0 

12 

o0 

0 

00

q 0 0 0 0 q 

0 0 0 0d 0 o 

(0.) ein-jedwej 

PIu I 

S. . . .... ÷........... •........... •............ : --- --- -- --- --- .. 

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

8 

C, 

o4)

9 9 o•

t3 

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

S................• ............... . . .-- ...... -- -................ 

. .0 

0 ---- ------ ---- ---- --- - --- --- --- - -- --- --

--------------------- ------------------- 0 

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

0 - - - -- - - -- -- -0-- - - -- - - - - -- - - - - - -

0

E 

*1=

J•
BT56 Transient Comparison LOBI/RELAP5. Base case

Figure A.II.37.- Core Heat Structure Temperature - Level 11 00.0 

50.0

50.0 1

80 490 500 510 520 530 540 550 

Time (s) Figure A.I1.39.- Pressurizer Collapsed Level

400.0 

---- XLOBT-TH3IB311 SLOBI-T1132C31 I 

G-B LOBI-TH33H311 

0- OLOBI-TH34E311 

A-A LOBI-TH37E311 

4- 1.OBI-TH38F311 350.0 

0--- Relap5

9 

C) 

0) 

E 

0)

570 580 590 600

Figure A.1I.38.- Core Heat Structure Temperature - Level 12

4 I 

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

S. ..... i ....... ,• ....... .......4....................... .......

j%5-Bj

Time (s)

6.0 

4.0 

2.0

I0-ORelap5 I

E

480 490 500 510 520 530 540 550 560 570 580 596 600 

Time (s)

S....... ; ..... .. i i .i .Iii i i i iii ... ..

...........

|1(I II

i Q•v

II II

00.0 [

4 560
BT56 Transient Comparison LOBI/RELAP5. Base case 

Figure A.II.41.- Rupture Disk Flow Figure A.II.42.- Rupture Disk Flow Regime 

14.0 ....... ....... __ ........... ..... ........ ....... ....... ....... ....... 

13.0 ------- ....... I ....... -- ....... i .. ..i....... I ....... ....... !........ I....... ....... ....... 12.0 -- - - --- -.... -I... ......... •....... --- -- ....... ....... ....... , ....... ....... 1....... 

. . ....... .... ........ . .. . .1 .... .0 ....... .. . .. . ... ...... ....... ........ I -- -- ; ------- . .... ... . . . .; . . . 148.0 13.0.. 0, -- - - .. . . . .. . . .

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

. . . .. ... ... . 8. -------.. .. ............ .. .. ......... ... I.. . .. .......... ...... ............. .... 

.4 1O1.0~ 

.. .. .... .. .. ..... .. .. .. . .... .. ... . .. ------- c.. c . . ... . . .. ,. . . ... . . .. ., . . . . . . 

.. 0 ................ ................ ........ ...... 

5 . .. .. . .. . . . . . .. . .- 1. .. . . . . - -

°.0o 

•8o~4. o.o o • 8 0 5 o • ..... o o ------ 5 ------- 0 •8 ----- • ..... ----- ...... o 

.00 480 4ý 50P V10 50 520 530 540 550 560 570 580 590 600 48 9 0 10 520 530 540 550 560 570 580 590 600 

Time (sl Time (s)

(-',
(S) awli 009 069 099 ON9 099 099 01'9 009 On9 01 009 061' 081,

,j15-gd 1a

SL8Vd-I19Q1 

SL6Vd-1SO1*-

9 dlj1G

t4)VDLEHI.-1HOt x-K 

1'Ot'3LCIU.-1801 -- *

]

01'l 

0-9 

0.9 

09'6

oiflssaid 610 W1u99 ua m ) S is -l'1I -917-1- atniU 

(S) aWl.j 009 069 099 OL9 099 099 001'S 09 ON 0Ms 009 061' 091 

0,009 

..... I----1--- 0,001 

17 [OAO' [ ainila9IdwajL aijni ImH aioZ) -t~t7 Vi v91l~ia

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

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

0

(D 

3 

*0

.9 069 0

-I[ I ui Wainvoduwaj -- gt71rlV oin~la (S) owij L 

19 0L9 099 099 019 069 Oa9 0M9 009 061' 9v.

[. AI!AI!!SUOS -9dV13U/I8O1 uosuj-dwoo luo!sujiI ggLg

09g

O'OZ

tunuoll.ioddfl ut ainssald -EI7-11-V 91n2ld

Sdvjo'do----o 

IDU)18EVd-IEIOI*

1 O'Z

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

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

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

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

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

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

............... bi8-gdupjao--e 

08 1H90dL-IHOI 1

00OH9MdL-15Ol *--

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

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

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

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

10-01
BT56 Transient Comparison LOBI/RELAP5. Sensitivity 2

Figure A.II.47- Pressure in Upper Plenum 

.. . . .......... .. . .• ....... ... ........ I ...... -......... ....... ,•....... ',....... t....... 

180 490 500 510 520 530 540 550 560 570 580 590 60 

Time (s) 

Figure A.I1.49.- Temperature in Cl. BL

Figure A.I1.48- Core Heat Structure Temperature - Level 4

*t--. IBI-PA3 8KDG 

10-0- Relap5

500.0 

S400.0 

C. 

E 

F

300.0

20.0 

15.0 

10.0 

5.0 

0.0 

V1 4 

4t

0,0 

00.0 

00.0 

00.0

0

* LOBI-TP26H000 

180 SLOBI-TM26H 0-O RelapS-gas 

0---ReIa5-lia

ID 

(L 

qj) 

(L

480 490 500 510 520 530 540 550 560 570 580 590 600 

Time (s)

480 490 500 510 520 530 540 550 560 570 580 590 6( 

Time (s) 

Figure A.1I.50- IL-BL Steam Generators Pressure

6.0 

9.0 .......-- ....... I . . .. I . .......... ....... ....... I ,........ ....... ....... . r I ....... ....... I 

7.08 1 - .---... ,...i•1....... t...... ....... " ....... : ....... ....... J .... ....... "" :..... .... il....... .. 

5.0 ... . 

4.0 

32 0......... 0..... ....... .. . .... ....... .............. ....... ........ . ....... ..... 

2.0

4E--LOBI-TH37C404 

x-- LOBI-TH37G404 

IG--O Relap5

0

SLOBI-PA97S 

* LOBI-PA87S 

0-0 Rclap5-IL 

*-----RelapS-BL

40U 49U 500 51U 52U 530 540 550 560 570 580 590 600 

Time (s)

9Afln n

* I I

SOD, a

.......- ....... i ...... •....... ......... .......
BT56 Transient Comparison LOBI/RELAP5. Sensitivity 3

Figure A.IT.51- Pressure in Upper Plenum

480 490 500 510 520 530 540 550 

C"Figure Fu A.II.53- IL-BL Steam Volumetric Time (s) Flow

S . ............... ...... . .... ..

Figure A.ll.52- Core Heat Structure Temperature - Level 4

[LOBI-PA38KDG reo Relap5

600.0

U 

E) 

E 

0)

560 570 680 590 600

10.0 *-- LOBI-QV65HV 

x--- LOBI-QV66HVX 

0-S RelapS-IL - RclapS-BL

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

8.0

E

0) 

U) 

n1 M)

6.0

4.0

180 490 500 510 520 530 540 550 56 

Time (s) 

Figure A.ll.54- IL-BL Steam Generators Pressure

510 520 530 540 s500 550 560 570 

Time (s)

20.0 

15.0 

10.0 

5.0

.. I . . ... ....... ..... 

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

-.. .I....... . -. ......... . .. (.......

0.0

0.06 

0.04 

0.02 

0.00

UIU

o8 ,9u Du 5 00 50 50 530 540 600 560 570 580 590 600 

Time (s)

^^^

2•.0 480 580 590 600 490

SLOBI-TH1370404 *- * ,OB T-TH370404 0--ORelap5 

O0 

4" * LOBI-PA97S 

x----x LOBI-PA87S 

0--OReiap5-UL 

0---ORela p5-BL

U.UO
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (2-89) NRCM 1102, (Assigned by NRC, Add Vol., Supp.. Rev.. 3201. 3202 BIBUOGRAPHIC DATA SHEET and Addendum Numbers, If any.) 

(See insbructions on the ,evwse) 

2. TITLE AND SUBTITLE NUREG/IA-0183 

Analysis of the LOBI Experiment Test BT-56 Using the RELAP5/MOD3.2 Code 3. DATE REPORT PUBLISHED 

June 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

J. Blanco, E. Moralo, R. Sanjuan, C. Gomez 

7. PERIOD COVERED (inclusive OaDs) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office orRegion, U.S. Nuclear Regulatory Commission, end mailing address; if ontbact• provide name and mailing address.) 

Union Fenosa Generacion S.A. 

Central Nuclear Jose Cabrera 

Madrid 

SPAIN g. SPONSORING;~ (ARAMI7ATLiIJ _ MAIl= AdT -l•^',•m . . - .-.. -.-. , .. (Ifl.. . esabv~fofufa ..... rvd R Dvso.Olo aRgoUS NcerRgllayC-msin - aSPONd OR OIZ -................. z (I NR. b" *Sam as aov, i t , povi NRC Disio, of', or Regadn, U.S. Nuclss. RegulatoyCommi

Division of Systems Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTFR

11. ABSTRACT (200 words orAess) 

This document has been drawn up by UNION ELECTRICA FENOSA, S.A. (UNION FENOSA), which participates in the Code Application and Maintenance Project (CAMP) as a member of UNIDAD ELECTRICA, S.A. (UNESA). The document presents a comparison between experimental data from the plant and data obtained by simulation with the RELAP5/MOD3.2 code of the experiment BT-56 which took place on 3rd July 1990 at the Joint Research Centre of the Commission of the European Communities, located in Ispra (Italy). The experiment originally scheduled for the LOBI-MOD2 facility was a Loss of Main Feedwater (LOFW), but multiple failures occurred, meaning that the evolution of the experiment deviated it from the scheduled sequence. The simulation was performed using the RELAP5IMOD3.2 code, on a Digital AlphaServer 2000 4/200 computer and the operating system DIGITAL-UNIX. The results show an acceptable agreement between the phenomenology observed during the experiment and that predicted by the code. 

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assW reserchers in ioating tMe eport) 13. AVAILABILITY STATEMENT 

RELAP5/MOD3 unlimited LOBI 

14. SECURrrY CLASSIFICATION 

(This Page) 

unclassified 

(Rhs Report 

unclassified 

15. NUMBER OF PAGES 

16. PRICE 

NRC FORM 335 (2419)

10. SUPPLEMENTARY NOTES
Federal Recycling Program 

I I I li
NUREG/IA- 0183 ANALYSIS OF THE LOBI EXPERIMENT TEST BT-56 

USING THE RELAP5/MOD3.2 CODE

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

JUNE 2000