NU/EGIIA-0 172 

International 

Agreement Report 

Assessment of RELAP5/MOD3.2 

Against a Main Steam 

Isolation Valve Closure at 

TRILLO I Nuclear Power Plant 

Prepared by 

A. de Lucas, R. L. Perezagua 

Empresarios Agrupados, AIE 

Spain 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

February 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
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NUREG/IA-0172

International 

Agreement Report 

Assessment of RELAP5/MOD3.2 

Against a Main Steam 

Isolation Valve Closure at 

TRILLO I Nuclear Power Plant 

Prepared by 

A. de Lucas, R. L. Perezagua 

Empresarios Agrupados, AIE 

Spain 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

February 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
CONTENTS

Page No. 

EXECUTIVE SUMMARY 1 

ABSTRACT 2 

1. INTRODUCTION 3 

2. DESCRIPTION OF PLANT AND TRANSIENT 4 

2.1 PLANT DESCRIPTION 4 

2.2 TRANSIENT DESCRIPTION 5 

3. CODE INPUT MODEL DESCRIPTION 7 

3.1 PRIMARY CIRCUIT 7 

3.1.1 Reactor Pressure Vessel (RPV) 7 

3.1.2 Coolant Loops 7 

3.2 SECONDARY CIRCUIT 8 

3.3 CONTROL SYSTEM 9 

3.3.1 Primary Circuit 9 

3.3.2 Secondary Circuit 9 

4. STEADY STATE CALCULATION 11 

5. RESULTS 12 

5.1 SECONDARY SYSTEM 12 

5.2 PRIMARY SYSTEM 13 

6. COMPARISON OF RESULTS 15 

6.1 PRIMARY SYSTEM 15 

6.2 SECONDARY SYSTEM 15 

7. RUN STATISTICS 17 

8. CONCLUSIONS 18 

9. REFERENCES 19

i-1
LIST OF TABLES

Page No. 

Table 1: Main Characteristics of Trillo NPP 20 

Table 2: Comparison of Steady-State Values 21/22 

Table 3: Sequence of Transient 23

1-2
LIST OF FIGURES

Page No. 

Figure 1 : Trillo 1 NPP RELAP5/MOD3.2 Model 24/25 

Figure 2: PRIMARY SIDE PRESSURE PRESSURIZER 26 

Figure 3: COOLANT AVERAGE TEMPERATURE 27 

Figure 4:PRESSURIZER LIQUID LEVEL 28 

Figure 5: MAIN STEAM PRESSURE GENERATOR N01 29 

Figure 6: MAIN STEAM PRESSURE GENERATOR N02 30 

Figure 7: MAIN STEAM PRESSURE GENERATOR N13 31 

Figure 8: LIQUID LEVEL STEAM GENERATOR N01 32 

Figure 9: LIQUID LEVEL STEAM GENERATOR N02 33 

Figure10: LIQUID LEVEL STEAM GENERATOR N03 34 

Figurel 1: MAIN STEAM FLOW STEAM GENERATOR N01 35 

Figurel2: MAIN STEAM FLOW STEAM GENERATOR N02 36 

Figure13: MAIN STEAM FLOW STEAM GENERATOR N03 37 

Figure14: MAIN STEAM RELIEF VALVE STEAM GENERATOR N03 38 

Figurel5: MAIN STEAM SAFETY VALVE STEAM GENERATOR N03 39

1-3
EXECUTIVE SUMMARY

This report represents an assessment calculation carried out by Empresarios 

Agrupados, for the Relap 5/Mod 3.2 code with respect to a main steam isolation 

valve closure at Trillo I NPP. 

Trillo I is a PWR nuclear power plant designed by KWU-Siemens, FRG. It came into 

operation in 1988. 

Trillo I NPP has a nominal thermal reactor power of 3010 MWt and three cooling 

loops. The steam generators are U-tube (4086) type with a pre-heater. The rated 

generator output is 1040 MWe. 

The code was installed on a Silicon Graphics Onyx workstation at EA, and a post

test calculation on it has been carried out. 

The Trillo I model developed by EA with the Relap 5/Mod 3.2 consists of 273 control 

volumes, 297 junctions, 101 heat structures and 252 control variables. 

It can be concluded from this transient study that a reasonable agreement exists 

between the calculation results and the data obtained during the test (D-100-306).

1
ABSTRACT 

This report includes an assessment of RELAP5/MOD3.2 against a main steam line 

isolation at 99% power in Trillo I NPP (PWR, KWU-Siemens, 1040 MWe). 

The results used are those obtained during test D-100-306 (Ref. 3), performed August 19, 1988 at Trillo I NPP. 

The three plant cooling loops are simulated to take into account: 

i. A single loop containing a valve in which the disturbance occurs 

ii. The remaining two loops are simulated using two separate model loops 

The Trillo I model used comprises 273 control volumes, 297 junctions, 101 heat 

structures and 252 control variables. 

From this study it can be concluded that the post-test calculation shows a 

reasonable agreement with the actual plant data.

2
1. INTRODUCTION

Empresarios Agrupados, as a participant in the CAMP Espaha Project, has 

developed a model for Trillo NPP (PWR, KWU-Siemens, 1040 MWe). The main 

objective of this study is to use this model to analyse the transient caused by closure 

of one of the main steam line isolation valves at Trillo I NPP with the Relap 5/ 

Mod 3.2 code (Ref 1) installed on a Silicon Graphics Onyx workstation 

The results used are those obtained during test D-100-306 (Ref. 3), performed 

August 19, 1988 at Trillo I NPP. Table 2 gives a description of the main variables 

recorded during the test and their corresponding denomination. 

In this study, the above test results are compared with those obtained with the model 

developed with the Relap 5/Mod 3.2 code. It can be seen that the two sets of results 

largely agree. 

This study comprises the following steps: 

"* Plant modelling 

"* Steady state calculations 

"* Transient calculations 

"* Comparison with actual plant data

3
2. DESCRIPTION OF PLANT AND TRANSIENT

2.1 PLANT DESCRIPTION 

Trillo I nuclear power plant is a PWR design, designed by KWU-Siemens, FRG. It 

came into operation in 1988. It has a thermal reactor power of 3010 MWt. 

The reactor coolant system (RCS) consists of a reactor and three closed reactor cooling loops, connected in parallel to the reactor vessel. Each loop has a coolant 

pump and a steam generator. The RCS also has a pressuriser that heats the coolant 

electrically. 

The coolant (demineralised water at high pressure) circulates through the reactor 

core to remove the heat generated by the nuclear reaction. The heated coolant 

passes from the reactor vessel through the coolant loop pipe (hot leg) to the steam 

generators. Here, it releases its heat in a heat exchanger to generate the steam that 

goes to the turboalternator in the so-called "secondary circuit" of the thermal cycle. The cycle is completed when the coolant is pumped back to the reactor vessel 

through the so-called cold leg. 

The coolant pumps are vertical, electric motor powered, single piece, centrifuge type. The power supply to the pumps is designed to maintain sufficient flow to cool the reactor under all circumstances. 0 

The steam generators are vertical with 4086 U-tube units (incoloy material) in order 

to separate the reactor coolant from feedwater-steam circuit. 

The pressuriser is electrically heated and connected to one of the reactor coolant 

loops by the surge line and maintains the pressure in the RCS in normal operating conditions and controls pressure variations during plant transients and maintains 

system pressure within design limits in upset conditions. 

The thermohydraulic design of the RCS prevents nucleate boiling limit being reached in the fuel assemblies, and ensures that the steam generator size and 

reactor coolant pump (RCP) power are optimum. 

The reactor core arranged in the reactor pressure vessel is the power station's 

nuclear heat source.

4
The area of the core where fission occurs consists of an array of closely arranged 

identical fuel assemblies which contain the nuclear fuel. The water of the reactor 

coolant system circulates around the assemblies to moderate the reaction and 

remove heat. 

To control the nuclear reaction, part of each fuel assembly contains neutron

absorbing control rods which are moved axially by electro-magnetic drive units. 

During reactor trip, the fuel rod assemblies fall by gravity. 

Table 1 summarises the main characteristic data for Trillo I NPP. 

2.2 TRANSIENT DESCRIPTION 

The reference transient analysed in this report ("Main Steam Isolation Valve 

Closure") was carried out at Trillo I NPP in August 1988, as part of the hot 

operational test programme. The starting situation of the plant, before carrying out 

the test, was steady state at 99% power, with all control systems operating correctly 

in automatic mode. 

The plant logic for this type of transient is to cool the plant to a certain ratio (100 K/h) 

by opening a relief valve in the isolated generator, until the secondary pressure 

reaches 74 bar, at which point a new steady state (partial shutdown) is established. 

To minimise releases to the atmosphere as much as possible, a condenser bypass 

is also used. 

The transient began at 17:36:44, starting with manual closure of the main steam 

isolation valve of generator No. 3 (RA03-SO01). 

The steam flow and the level in the isolated generator decreased rapidly. The 

pressure in this generator increased to reach 85 bar, generating reactor protection 

signals YZ1 I and YZ16. These signals caused reactor and turbine trip, respectively, 

9 seconds after the transient began. 

The main steam relief valve of the isolated generator began to open as the pressure 

reached 85 bar. In spite of this, the pressure in the isolated generator continued to 

rise until it reached the set point for the safety valve (87.3 bar), 10.3 seconds after 

the start of the transient, causing it to open. This limited the pressure rise in the 

generator to a maximum of 88.7 bar, at which point the pressure began to fall again.

5
The safety valve of the isolated generator closed when the pressure in it dropped to 

82.5 bar. 

The level in the isolated generator began to decrease as soon as the isolating valve 

was closed. At 13 seconds after the start of the transient the level reached 10.2 m, at which moment the order was sent automatically to start the pumps of the RR System (startup/shutdown) to make up the level in the generator. 

Unlike generator 3, after isolation, pressure in the intact generators dropped as a 

consequence of the increased flow of main steam required from the two generators, 

as the turbine continued demanding full load. 

This depressurisation of the intact generators caused the level in them to increase 

by approximately 0.3 m. 

Steam pressure in the header dropped from the start of the transient, to a minimum of 61 bar. As a consequence of reactor and turbine trip, pressure in the intact 

generators recovered until it reached the set-point of the condenser bypass (80.4 bar). The condenser steam-dump stabilised the pressure in the header to 

80.4 bar. 

The relief control valve of the isolated generator was used to cool the reactor to 100 K/h, bringing the plant to hot standby, until pressure in the secondary reached 

74 bar. 

As regards the behaviour of the primary circuit, after the main steam isolation valve closed, pressuriser level rose by 10 cm to reach a maximum value of 7.51 m; coolant 

pressure rose to 157.62 bar and reactor power varied insignificantly until trip occurred. Power then fell rapidly to decay power, mean temperature fell from 308.60C to 2960C, which is zero load value, causing pressuriser level to reach 4.2 m. 

With partial shutdown, caused by the isolated generator relief regulation valve, due 

to the thermal coupling between the primary and the secondary, the mean 

temperature fell to 292.70 C; pressuriser level reached 3.57 m; coolant pressure reached 138.8 bar.

6
3. CODE INPUT MODEL DESCRIPTION

A simplified diagram of the of the developed model is given in Figure 1. In general 

terms, this model contains the following main parameters:

273 

297 

101 

667 

252 

1082 m3 

375 388 kg 

101

Control volumes 

Junctions 

Heat structures 

Mesh point structure heat 

Control variables 

Total system volume 

Total system mass 

Trip number

3.1 

3.1.1

The reactor is represented using 200-series components. The core, the downcomer 

and the bypass of the RPV are subdivided into nodes of uniform axial height. The 

core is divided into two parallel channels, each with eight axial cells. Cross-flow is 

allowed between channels. 

Radial subdivision of the core is resolved as one internal channel and another 

around it. 

From the point of view of heat structures, the axial power distribution is similar in 

both channels, so that each one generates 50% of the real power.

3.1.2 Coolant Loops

The model considers the three coolant loops of the primary circuit. 

Loop 1: 700-series components 

Loop 2: 400-series components 

Loop 3: 300-series components 

The pressuriser and the surge line (500-series components) are connected to 

loop 1. The volume control system extraction is taken from Loop 2 (tmdpjun 457).

7

PRIMARY CIRCUIT 

Reactor Pressure Vessel (RPV)
The injection to the volume control system (tmdpjun 772, 472 and 372) and the 

additional borate system (tmdpjun 777, 477 and 377) are connected to each loop. 

Apart from the connections to the different auxiliary systems of the primary, the three 

coolant loops of the primary circuit are identical. For these loops, the pipes, the three primary coolant pumps (RCPs), pump (components 365, 465 and 765), and the 

U-tubes for each steam generator (components 340, 440 and 740) are modelled. 

The U-tubes of the SG are modelled as 6 axial nodes. 

The pressuriser is constructed of vertical volumes (components 510, 520, 530, 535 

and 540). The relief valve YP10-S102 (component 575) and the two safety valves 

YP10-S191/192 (components 576 and 577) are connected to the top of component 535 to take into account its level in the pressuriser. The spray lines (component 552 

for the loop spray, 557 for the volume control spray and 559 for the borate addition 

system) are connected to the top of the pressuriser (component 540). 

3.2 SECONDARY CIRCUIT 

The secondary circuit considered in the model consists basically of three steam 

generators (600, 800 and 900 series components), the main steam lines (100-series components), up to the condenser and the turbine. The turbine stop and control valves are modelled with valve component 122. The train of valves to the condenser 

is modelled with valve component 119. The turbine and the condenser are not modelled explicitly, but are represented by time-dependent control volumes. The safety and the relief valves for each loop are also modelled (valve components 106, 131 and 165 for the relief valves and valve components 104, 129 and 163 for safety valves). Components 105, 107, 130, 132, 169 and 166 represent the atmosphere. The main steam isolation valves are represented by components 110, 135 and 170. 

The part corresponding to the feedwater system at the different generators is considered from the feedwater tank (tmdpvol 60) up to the generator connecting pipes (sngljun 90, 96 and 72). The heater train is modelled with pipe component 63. 

From the rise to the heater train, the system splits into the corresponding three 

feedwater loops. The full load feedwater control valves are represented with valve 

components 51, 90 and 66 (RL21/22/23-S002). These valves are controlled by control variables 949, 925 and 914 respectively. The low load control valves are 

modelled with components 20, 30 and 40 (RL21/22/23-S013). The associated 

control variables are 950, 926 and 915 respectively. 

The main and auxiliary feedwater pumps (RL and RR) are modelled as a time 

dependent junction (tmpdpjun 61).

8
3.3 CONTROL SYSTEM

In the model, the control systems and most significant limitations of the plant are 

simulated using the code's control block. The following controls are modelled: 

3.3.1 Primary Circuit 

Primary mean temperature control 

This control maintains the temperature in the primary in accordance with the 

load diagram, in operating situations that could cause (relatively small) non

equilibrium between the power generated in the reactor and that extracted in 

the generator, which would basically cause a temperature rise. The system 

regulates the insertion of banks and/or the lowering of the control rods to bring 

the mean temperature back to normal 

"* Control of reactor control rods 

This control maintains the core reactivity inventory during power operation and 

shutdown 

"* Coolant pressure control 

This system controls variations in coolant pressure. It includes activation of the 

sprays, electric heaters and relief and safety valves of the pressuriser 

"* Control of pressuriser level 

This control maintains coolant level by controlling the Volume Injection & 

Extraction System. The pressuriser reference level is established as a function 

of the inlet reactor temperature and the temperature rise through the core 

3.3.2 Secondary Circuit 

Steam Generator level control 

This control maintains the water inventory in the steam generators to ensure 

production of saturated steam. The level program is constant at any power and 

the set-point only changes during certain events (eg, tube break). It acts on the 

full and low load control valves of the feedwater system. The level control takes

9
account of the following variables: feedwater and steam flows to the generator, 

water level in the generator, and power level 

Cooldown to 100 K/h control 

This control system acts on the secondary system relief valves in order to cool 

the primary system with a gradient of 100 K/h. Therefore, the 100 K/h shutdown 

valve (secondary relief valve) is simulated, and the discharge of this valve is 

modelled as tmdpvol, for each cooling loop 

Control of Secondary System Maximum Pressure 

This system limits the pressure rise in the secondary system to prevent 

continued actuation of relief and safety valves.When the main steam pressure 

rises above its setpoint value the bypass valves open to the condenser, which 

allow the pressure to be maintained below a specified value 

The operational setpoint of the main steam header pressure is 3 bar above the 

actual main steam pressure and varies as a function of the output in 

accordance with the part-load diagram. Variation is performed as a function of 

the pressure in the main steam header, and it is restricted at the low-load end 

to a maximum of 80.4 bar 

Control of Minimum Pressure in Secondary System 

This system limits main steam pressure drop in the secondary system. When 

main steam pressure drops below the setpoint minimum, the turbine control 

valve is actuated to maintain the steam pressure at an adequate value 

The operational setpoint of the main steam header pressure is 3 bar below the 

actual main steam pressure and varies as a function of the output in 

accordance with the part-load diagram.

10
4. STEADY STATE CALCULATION

The steady state of the plant at 99% power was calculated using the null transient 

method. The model was run for 300 seconds to ensure stabilisation of all model 

variables. 

Table 2 compares the values obtained with the model and those from the test 

records before the transient began.

11
5. RESULTS

This section describes the plant-model behaviour and the progress of the transient, using the RELAP5/MOD3.2 code. The plant data and records used are therefore 

those for the conditions before starting the test and are obtained from Ref 3. 

The plant records were quoted as percentages, so it has been necessary to process them to be able to compare them with the results obtained with the RELAP5/MOD3.2 

code. 

The initial boundary conditions in the model, and the control numbers required by Trillo I NPP are obtained from available project documentation 

The situation at the start of the transient is a stationary state calculated beforehand. 

Its main parameters are shown in Table 2, together with a comparison with plant 

records at the start of the test. 

5.1 SECONDARY SYSTEM 

The transient is started at t=0 seconds, closing the main steam isolating valve of 

generator 3 (trip 430). The isolating valve is represented by component valve 110, 

and closes linearly in two seconds. 

Isolation of generator 3 causes an increase in pressure and a drop in main steam 

flow. As a consequence, the level of this generator decreases, too. 

The drop in main steam flow in generator 3 switches trip 683 to true at 6.5 s, generating an order to open the low load feed valve (valve 040) and a signal to close 

to the full load feed valve (valve 066). 

When trip 683 occurs, the level control in generator 3 (control variable 958) switches 

to being controlled by the low load valve, which is limited to only opening 57% (to 

prevent uncontrolled fill-ups). 

The pressure in this generator reaches 85 bar at t=-10.1 s (trip 415). As a result, 

three signals are generated: 

* Reactor trip signal (trip 716) 

* Turbine trip signal (trip 716)

12
Open relief and relief control valve signal (trip 680). The relief control valve is 

modelled with valve component 106. Control variable 658 models its control 

parameters. This valve serves to cool to a ratio of 100 K/h. 

Pressure in generator 3 continues to rise, to reach 87.3 bar (trip 511) at t-1 1.5 s. At 

this point, the mains steam safety valve opens, modelled as valve 104 and controlled 

by variable 109. This valve stays open until the pressure in generator 3 falls back to 

below 82.3 bar (trip 512) and closes fully at t=-1 5 s 

The level in generator 3 (control variable 958) reaches 10.2 m at t=-13 s, generating 

an order to start the startup/shutdown system pumps. When the generator 3 safety 

valve closes, its relief control valve cools temperature down to 100 K/h, to reach 

74 bar at t=300 s 

As the isolation valve of generator 3 finishes closing, the other two generators 

increase their flow demand (the turbine follows the power control) until the pressure 

in the collector falls below 63 bar, causing low pressure control in the turbine. 

When reactor and turbine trip occurs, pressure increases in the steam header until it 

reaches 80.4 bar (trip 598) at t=-17.35 s, when the condenser bypass valves 

(represented by valve component 119 and control variable 937) open. These valves 

stay open until the pressure in the steam header falls back to their setpoint values 

(80.4 bar). In the model, this event occurs at t=148 s. From that moment, the plant is 

cooled through the relief valve of generator 3. 

At t=-300 s, the plant reaches a new steady state, corresponding to hot standby. 

5.2 PRIMARY SYSTEM 

Events in the primary circuit after isolation of generator 3 are as follows: Immediately 

after the isolation valve of generator 3 closes, the primary circuit heats up; the level 

in the pressuriser (control variable 510) increases to 7.5 m, and coolant pressure 

reaches 159.5 and mean temperature in the primary increases to 309.45 °C. 

As a consequence of the reactor trip, the above-mentioned control variables for the 

primary fall immediately, and pressure in the primary falls to approximately 8 bar; 

mean temperature falls by about 50C, and pressuriser level (due to coolant density 

reduction) falls by half a meter from its maximum level. 

After this, the behaviour of the primary depends on the thermal coupling that is 

established with the secondary circuit. This means that the average coolant

13
temperature also follows the cooling ratio to 100 K/h. Mean temperature reaches 292.4°C, pressuriser level 3.84 m, and coolant pressure stays at 138.5 bar once the 

partial shutdown is over (300 s into transient).

14
6. COMPARISON OF RESULTS

This section gives a qualitative and quantitative comparison of the data recorded in 

the real test in the plant and the predictions obtained with the RELAP5 model. 

Table 3 shows a comparison over time of the sequence of events recorded in the 

test and those predicted by the model. 

6.1 PRIMARY SYSTEM 

Figure 2 shows both the recorded signal YR50-P401 and the model variable 

p 540 01 0000 representing the pressure in the pressuriser. The figure shows that 

the prediction fits satisfactorily with the data recorded in the plant. In general, the 

estimated pressure follows the same trend as the measurements. The maximum 

error between the prediction and real data is calculated as less than 4%. Taking into 

account the errors inherent in the measuring apparatus and the user-effect, this 

value is considered acceptable. 

Figure 3 shows a comparison between the average coolant temperature recorded in 

the test (YT10-T101) and corresponding one in the model, control variable 606. As 

with pressure, there is a good agreement between both values. The maximum error 

between the two parameters is less than 0.7% throughout the analysis. This 

temperature error could be caused by a slight difference in the residual heat in the 

fuel. 

Figure 4 shows a comparison between the record of the level in the pressuriser 

(YR71-L401) and model control variable 510. Overall, the trend is similar with both 

values. The maximum error between them is about 7%. This is due to the higher 

mean temperature predicted by the model. 

6.2 SECONDARY SYSTEM 

Figures 5, 6 and 7 show a comparison between the pressure recorded in each of the 

three steam generators (YB10/20/30-P951) together with the model's associated 

variables. 

There is a good agreement between the values given by the code and those 

recorded in the plant. The maximum error is less than 3% in the isolated generator, 

and less than 1% in the intact generators.

15
Figures 8, 9 and 10 show the levels recorded in the steam generators (YB10/20/30

L951) together with the corresponding control variables of the model (960, 959 and 

958). Overall, there is a good agreement between the levels. Although there are 

slight differences, the recorded and calculated values do generally agree. The 

maximum error between them is 4%, which is acceptable considering that the 

variable calculated by the code is collapsed level and the recorded level is 

calculated as differential pressure. 

Figures 11, 12 and 13 show the variation in steam flow from the generators both 

recorded in the test (RA01/02/03-F901) and the model variables (mflowj 160, 124 

and 100). There is a reasonable similarity between the measured and the calculated 

values. 

The behaviour of the relief and safety valves is shown in figures 14 and 15. Control 

variable 658 shows the position of the relief valve RA03-S005 of the isolated 

generator, with control variable 109 representing the position of safety valve RA03

S002 of the same generator. There is good agreement between these two calculated 

values and those recorded in the test.

16
7. RUN STATISTICS 

The calculation has been performed using the RELAP5/MOD3.2 code, running on 

ONYX (Silicon Graphics) YP25 1CPU Work-Station. 

The main characteristics of the hardware are the following: 

"* Control processor MIPS-R10000 

"* Clock frequency 194 MHz 

"• Performance 

MIPS 1000 

MFLOPS 400 

SPECint92 300 

SPECfp92 600 

"* Main memory 64 Mb 

"* Disk storage 4.59 Gb 

"* Operating system IRIS 6.2 

Transient calculation took 300 seconds problem time, and 565.5 seconds of CPU 

time consumtion, with a maximum time step size of 0.1 seconds. The ratio of CPU 

time to real transient time was consequently 1.88.

17
8. CONCLUSIONS

The transient corresponding to closure of the main steam isolation valve in steam 

generator 3 of Trillo I NPP has been simulated using the RELAP5/MOD3.2 code. 

Except for certain discrepancies listed below, the results in general compare 

favourably with the plant measurements (test D-1 00-306): 

"* It is observed that both the coolant pressure and temperature show slight 

differences from the test values. This must be related to the difference between 

the plant power and the calculated reactor power 

"* There may be differences, since magnitudes of the variables given by the 

calculation are compared with real values measured by sensors featuring 

uncertainty in their response and signal processing 

"* In the behaviour of the secondary, the differences between the model and the 

test are more significant and are probably due to the fact that during the test 

the isolating valve of the affected generator did not close completely. This 

would justify the difference in time between closure of the bypass valve in the 

test and in the model, and the steam generator fill times

18
9. REFERENCES 

1. RELAP5/MOD3.2, NUREG/CR-5535 

2. Trillo document, Cierre v6lvula de aislamiento de vapor principal. Test No. 

D-100-306. August 1988 

3. KWU-Siemens document, Reactor Coolant and Pressurizer System. R232/222 000, 

Rev e. April 1993 

4. Trillo documents, Dossier System: RA, RL/RR, TA, TW, YA, YP. 

5. KWU-Siemens document, Volumes of the Reactor coolant and Pressuriser System. 

R1 0/861E2036

19
Table 1 

Main Characteristics of Trillo NPP 

Thermal reactor power 3010 MWt 

Net electrical output 1040 MWe 

Fuel U0 2, sintered 

Number of assemblies 177 

Fuel rod assembly 236 

Active length of fuel rods 3.4 m 

Outer diameter fuel rod 10.75 mm 

Cladding tube material zircaloy 4 

Absorber material Ag-In-Cd 

Number coolant loops 3 

Reactor operating pressure 158 bar 

Coolant flow rate 16.684 kg/s 

No. Steam generators 3 

No. Tubes in SG 4086 

Tube material Incoloy 800 

Average tube length 20.49 m 

No. coolant pumps 3 

Type Single-state centrifugal pump 

Discharge head 101 m 

Design flow rate 5.292 kg/s 

Pressuriser operating pressure 156.5 bar 

Total nominal PRZ electrical 1638 kW 

heating capacity 

Total pressuriser volume 45 m3

20
Table 2 

Comparison of Steady-State Values 

Variable Recorded RELAP5 

Signal Value Variable Value 

Reactor power YTOO-XI 11 2979.9 MW rktpow 2982.5 MW 

Real pressuriser level YR71-L401 7.35 m cntrlvar 510 7.32 m 

Required pressuriser level YR71-L411 7.41 m cntrlvar 517 7.43 m 

Coolant pressure YR50--P401 154.26 bar p 540010000 154 bar 

Reactor inlet temperature loop 3 YA30-T955 293.80C tempf 390010000 293.116 0C 

Reactor inlet temperature loop 2 YA20-T955 -- tempf 490010000 293.1240C 

Reactor inlet temperature loop 1 YA10-T955 293.350C tempf 790010000 293.118 0C 

Heating range loop 3 YA30-T951 300C cntrlvar 515 32.270C 

Heating range loop 2 YA20-T951 31.2 0C cntrlvar 514 32.270C 

Heating range loop 1 YA1O-T951 31.20C cntrlvar 513 32.270C 

Steam Generator 3 pressure YB30-P951 67.87 bar p 650010000 67.723 bar 

Steam Generator 2 pressure YB20-P951 67.5 bar p 850010000 67.72 bar

21
Table 2 (cont'd) 

Comparison of Steady-State Values

Variable Recorded RELAP5 

Signal Value Variable Value 

Steam Generator I pressure YB10-P951 66.9 bar p 950010000 67.725 bar 

Steam flow outlet SG3 RA03-F901 540 kg/s mflowj 100000000 544.28 kg/s 

Steam flow outlet SG2 RA02-F901 540 kg/s mflowj 124000000 544.73 kg/s 

Steam flow outlet SG1 RA01-F901 522 kg/s mflowj 160000000 543.92 kg/s 

Steam Generator 3 level YB30-L951 12.07 m cntrlvar 958 11.949 m 

Steam Generator 2 level YB20-L951 11.9 m cntrlvar 959 11.949 m 

Steam Generator 1 level YB10-L951 11.9 m cntrlvar 960 11.949 m 

Coolant average temperature YT10-T11 308.15 0 C cntrlvar 606 308.30C 

SG3 inlet water flow RL23-F901 547.31 kg/s mflowj 720000000 547.80 kg/s 

SG2 inlet water flow RL22-F901 564.81 kg/s mflowj 960000000 542.71 kg/s 

SG1 inlet water flow RL21 -F901 550.0 kg/s mflowj 980000000 543.45 kg/s

22
Table 3

Sequence of Transient (seconds)

Plant-test RELAP Model 

Prediction

SG3 isolation valve begins to close 

SG3 isolation valve fully closed 

SG3 full and low control valves begin 

to close and open, respectively 

Reactor and turbine trip 

SG3 Relief Valve begins to open 

SG3 Safety Valve opens 

SG3 Safety Valve closes 

Turbine bypass opens 

Turbine bypass closes 

SG3 partial shutdown completion

23

Event

0.0 

2.0 

6.0 

9.0 

9.0 

10.3 

14.75 

16.0 

100 

300

0.0 

2.0 

6.5 

10.1 

10.1 

11.5 

15.0 

17.35 

148 

300
Figure 1. Trillo

1, 26-C.

.xt. TA

I NPP RELAP5/MOD 3.2 Model

TH ALTA 

FRESIMN

LOOP 1 LOOP 2

24

LOOP 3

I.. - :-

I': 

T-1
FEEDWATER 

TANK

Figure 1 (cont'd) Trillol RELAP5//MOD3.2 Model 

RL2"-SO13 n FW-LOOP 3 

... 5. 4 65 I--tq6 71 SG-3 RL23-S002 68 70 72 

PJL22SO1-3 ' 8 f-FW-LOOP 2 

58 93s95 0 . S G -2 

R2-0292 94 9 

oR"21-S713 FW-LOOP I 

"1• • 50 -.I• •62 F 54 " ' 5 [ sRL21-S002 53- 5 98 

25
Figure 2 

PRIMARY SIDE PRESSURE PRESSURIZER 

MSIV CLOSURE TRILLO I NPP

0 25 50 75 100 125 150 175 200 225 250 275 

TIME (SEC) 

26

180 

170 

160 

S150 

ILu 

140 

(I) 

w 

ILZ 130

300

- P 540010000 

120 

110 

100
LZ 

( 03S ) 3W11

00 qzz

909 JeAIJU 0

1,01, -01,1 ------

09z 0ON SL L 091I 9U 1 001 9L 09 9z 0

i I i i i i i ii

......... ......... .................. V . .. .......... ...... ........... .......... .... ................ ................... ........... ................. .................... .. .......... . ............... 

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

.............. .......... ................ ....... .. ............... V .................... r- - .......... ..... ........ .. .......... ............ ...... ............ ........... ............... ......... ........................ 

...... ........ .......... ................................. ............ .............. ..................... .. ....................... .. .. .7 ' ...... .. .. .................. .......... - ................ 

...................... .. ..... .. ...................... V - 1 ..................... .................... ............... ............................ - ................................... ........... -.......... ..... ............ .............. 

................................ ............................. ................................. ................ ........... ................................. .......... - ............ ...................... .......... .......... ... .. .... ..... ""7 ................. .............. 1 .... . ............. ....... . 

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

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

ddN L o07•1i 3uIrlSO13 AISIN 

3InlvmLdWl 3!OVM3AV INVO003 

C eJnBL-j

96Z 

-V6Z 

96Z 

. -oq 

008 

90 

-I 

vcC3 

0ii 

g0s 

M~E
Figure 4 

PRESSURIZER LIQUID LEVEL 

MSIV CLOSURE TRILLO I NPP 

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

l.........

0 25 50 75 100 125 150 175 200 225 250 275 

TIME (SEC) 

28

10 

9 

8 

7 

6 

5 

4 

3 

2 

1 

0

300

,> -J 

w_ 

m

--- cntrlvar 510 

- - YR71-L401 
Figure 5

MAIN STEAM PRESSURE GENERATOR N01 

MSIV CLOSURE TRILLO I NPP

85 

80 

75 

S70 

w 

a. 

60 

55 

50

0 25 50 75 100 125 150 175 200 225 250 275

TIME (SEC) 

29

-P 950010000 

------ YB1 0-P951

300
Figure 6 

MAIN STEAM PRESSURE GENERATOR N02 

MSIV CLOSURE TRILLO I NPP

85 

80 

75 

cu 

S 70 

w 

U) 65 

ILl 

60 

55 

50

I,

4

I - I I

0 25 50 75 100 125 150 175 200 225 250

. ............ ................................. ........ ....... .; .............. ... ....... .. .......... -- - -- - -- -

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

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

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

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

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

275 300

TIME (SEC)

30

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

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

.......... 

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

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

-- P 850010000 

YB20-P951 

I

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

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

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

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

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

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

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

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

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

.......... ..............
Figure 7 

MAIN STEAM PRESSURE GENERATOR N03 

MSIV CLOSURE TRILLO I NPP

90 

85 

80 

1- 75 

%0= UJ 

w 

70 

U) 

w 

0. 65 

60 

55 

50

0 25 50 75 100 125 150 175 200 225 250 275 

TIME (SEC) 

31

"- P 650010000 

------- YB30-P951

300
Figure 8 

LIQUID LEVEL STEAM GENERATOR N01 

MSIV CLOSURE TRILLO I NPP

13 

12 

_11 

%-J w 

wj 10 

-j 

-j 9 

8

0 25 50 75 100 125 150 

TIME (SEC)

32

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

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

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

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

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

7 -I. I

175 200 225 250 275 300

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

YBIO-L951 

• •'-cntrlvar 960

...........

. 

..

..........

..... ..... 

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

.............
(03S) 3WILl

M~ 001 9L 09 9z 0

696 JeAIJUO 

MI6"-O•O, ------

I I I I I I I I I I

ddN I O7-II1 :IrlSO13 AISI 

ZoN •l :N3 IV31S -l36A31 alnf1"l 6 ejn!.Ij

00S 9LZ 0o0 9L I 091

L 

10 

% 

oi r

I

S. ............. ..................... ... i............................ .................. ..... ......... i .................... .......... .. ................................... .. ................................. ... .... .... .. .... ..... .. .... ...... .. .. .. ... .. .......... ..... 

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

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

.... .............. ............... ... .... ............ ...... ... ............... ............. ... .......... ... .......... . ... .........................
Figure 10 

LIQUID LEVEL STEAM GENERATOR N03 

MSIV CLOSURE TRILLO I NPP

13 

12 

_-11 

-J 

-J 

8 

7

0 25 50 75 100 125 150 175 200 225 

TIME (SEC) 

34

250 275

-cntrlvar 958 

------ YB30-L951

300
(33s) 3vWI1

00L

.061- 0 1 -----

000000091% rMou06 09 OL 09 09 01t OL

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

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

I I I 

... .......... .......... .............. .................... . . :.......... .. .. ... ... ............... ....... .... .. ... ....... ... .................. ...... ...................................... . ........................... ... .................... 

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

.... ... ...... ... .... .... ... .... ..... ... ..... ... ... ... ....... ... ... ..... ... ... ... .. .... ... ... ... ... ... ... ... ... ..... ... ..... ... ,.. 

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

ddN I O7ifIl a=llSO-13 AISIW 

6oN Mi0VM=N=IO WV31S MO" IWV3:lS NIVIl 

14 ojn16 id

0

0

00 L 

OOM 

-I 

m 

00I -n 

r

0 

0017 

G) 

CO) m 

009

009 

OOL
(33s) uilll 

0OL 06 09 OL 09 09 017 O0 Oz M 0 

--------- •------------........-- ---- 

......... ........ .. ......... .......................... .. ..... ....................... .......... .... ... , .... OOi 

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

S.. .. ........ ............... . . ...• . . .... ... ... .... .... ... .... . . .; .. . . .. ..... .......... ..... .... ... .. ... ... ... .. .... .. .... ... ... ... .. ....... ..;... . ' ~ ~ ~~~ ~~ ~~~ :".................. ................. ' ... . ........................ ........... ......... .................. .. ................................. O O I 

'OOS 

Ic ........... ..................................................................................................

...................... ................ . . .. ........... .. ...... . ............................. . .......................... . .......................... ............... ........ .... .... ........ .. .... .......... .... ... ............................... . 0 0 " " 

1,0 6 :l Z O ..... ,0 a .. 0 " 

.. ......................... ..... . ................... .....' ....... .... ... ..... • ... . .............. ........... ........... ....... ....... ... .. ..... ....... ... ... .... .. ..... ... ......... ..... ....... ... ....... .. .. ... ............ ....... . .. ..... ....... ....... 0 0 1 

S009 

OOL ddN t oi1l~iJ. 3ifSO-13 AISW~ 

ZoN NjO1VH3N3E WV31S M011 W~V31S NIVW% 

ZI* oJn6i!j
Figure 13 

MAIN STEAM FLOW STEAM GENERATOR N03 

MSIV CLOSURE TRILLO I NPP

10 20 30 40 50 60 70 80 

TIME (SEC) 

37

600 

500 

0 400 LU 

300 0 -J 

U

w I- 200 

100 

0

90 100

-itflowj 100000000 

RA03-F901 

0
Figure 14 

MAIN STEAM RELIEF VALVE STEAM 

MSIV CLOSURE TRILLO 1 NPP

GENERATOR N03

45 

40 

35 

,--30 

e 25 

0 

U) 20 

-j 

-j 

LL 15 

10 

5 

0

0 10 20 30 40 50 60 70 80 90 100 110

TIME ( SEC) 

38

cntrlvar 658 

RAO3-S005 

120
MAIN STEAM

Figure 15 

SAFETY VALVE STEAM GENERATOR N03 

MSIV CLOSURE TRILLO I NPP

0 2 4 6 8 10 12 14 16 18 

TIME (SEC) 

39

---- cntrlvar 109 

- - - RA03-S002

20

120

100

0ý 

wj 

0L

80 

60 

40

20 

0
NRC FORM 336 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., SupP., Rev., 

NRCM 11o BIBOGRAPHIC DATA SHEET and Addendum Numbers, If any.) 

(See mnsfucftwol on the revue) 

2. TITLE AND SUBTITLE 

NUREGAA-0172 

Assessment of RELAP5/MOD3.2 Against a Main Steam Isolation 3. DATE REPORT PUBLSHED 

Valve Closure at TRILLO I Nuclear Power Plant MONTH YEAR 

February 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

A. de Lucas, R.L. Perezagua Technical 

7. PERIOD COVERED tlncluswe Dela) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (f NRC povide Divison, OnceoRegion, U.S. NucwReguIlatoy Comnimson, and rmang address, if conacw, 

provdeme And -e ad afsJ) 

Empresarios Agrupados, AlE 

Spain

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (D NRC, type 'ame as above^; if conbactor, provide NRC Division, Office or Region, U.S MicaWRegulator/ ComnmSon, 

and m-Ng ad*-.)

Division of System Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTRACT (Mo wads rams) 

This report includes an assessment of RELAP5/MOD3.2 against a main steam line isolation at 99% power in TRILLO I NPP 

(PWR, KWU-Siemens, 1040 MWe). The results used are those obtained during test D-100-306 (Ref. 3), performed August 19, 

1988 at TRILLO I NPP. The three plant cooling loops are simulated to take into account i. A single loop containing a valve in 

which the disturbance occurs; ii. The remaining two loops are simulated using two separate model loops. The TRILLO I model 

used comprises 273 control volumes, 297 junctions, 101 heat structures and 252 control variables. From this study it can be 

concluded that the post-test calculation shows a reasonable agreement with the actual plant data.

12. KEY WORDS/DESCRIPTORS (Lait wrds oraphnwas Ut wift e,•it rWesea in bw&e Me fwora) 13. AVAILASLUY STATEMENT 

unlimited 

RELAP5 14. SEOtuRrrYCLASsIATION 

LOCA RHR (5WA P,•,) 

unclassified 

(ni Rep" 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-M8
Federal Recycling Program
NUREG/IA- 0172 ASSESSMENT OF RELAP5/MOD3.2 AGAINST A MAIN STEAM FEBRUARY 2000 

ISOLATION VALVE CLOSURE AT TRILLO I NUCLEAR POWER PLANT

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL 

POSTAGE AND FEES PAID 

USNRC 

PERMIT NO. G-67