NUREG/IA-0197

International 

Agreement Report 

Assessment of 

RELAP5/MOD3 .2-NPA3.4 

Against an Inadvertent Closure of all 

Three MSIV's in VANDELLOS-1I 

Nuclear Power Plant 

Prepared by 

C. Llopis, A. Tanarro, R.M. Fanegas 

TECNATOM 

Avda. Montes de Oca, 1, 

San Sebastian de los Reyes 

28700-Madrid 

SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

January 2001 

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
NUREG/IA-0197

International 

Agreement Report 

Assessment of 

RELAP5/MOD3.2-NPA3.4 

Against an Inadvertent Closure of all 

Three MSIV's in VANDELLOS-II 

Nuclear Power Plant 

Prepared by 

C. Llopis, A. Tanarro, R.M. Fanegas 

TECNATOM 

Avda. Montes de Oca, 1, 

San Sebastian de los Reyes 

28700-Madrid 

SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

January 2001 

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

The work consists in reproducing an actual inadvertent manual closure of all 

three main steam isolation valves transient that occurred at Vandell6s II Nuclear Power Plant 

(Tarragona, Spain), by means of the RELAP 5/MOD 3.2 code, thus becoming part of an 

assessment type exercise. 

The calculations have been performed by the plant technical support staff and 

TECNATOM S.A. in a jointly effort. Vandell6s NPP input model for RELAP 5/MOD 3.2, 

has been developed, starting from the input model for the former versions RELAP 5/MOD 

2 and MOD 2.5, by the licensee technical services department. The model was provided to 

TECNATOM S.A., along with the plant data, transient available records and associated 

information in order to perform the transient calculation and sensitivity studies. Finally, the 

results and conclusions were discussed and presented jointly by both organizations. 

The present report is mainly focused to the steady state and transient calculations 

and the findings and conclusions derived from them. The general purpose model provided 

by. the licensee, along with some little modifications implemented for the specific transient 

calculation, are described. A thorough description of the model build-up process is not 

included.

iii
ASSESSMENT OF RELAP 5 / MOD 3.2 AGAINST AN INADVERTENT CLOSURE 

OF ALL THREE MSIV's IN VANDELLOS U NPP 

TABLE OF CONTENTS 

ABSTRACT 

EXECUTIVE SUMMARY 

1. INTRODUCTION 

2. PLANT AND TRANSIENT DESCRIPTION 

2.1 Plant Description 

2.2 Transient Description 

3. CODE INPUT MODEL DESCRIPTION 

3.1 Reactor Vessel 

3.2 Reactor Coolant System Piping 

3.3 Steam Generator Primaries 

3.4 Reactor Coolant Pumps 

3.5 Pressurizer and Associated Systems 

3.6 Steam Generator Secondaries 

3.7 Main Feedwater System 

3.8 Auxiliary Feedwater System 

3.9 Main Steam System 

3.10 Emergency Core Cooling System 

3.11 Chemical amd Volumetric Control System 

3.12 Heat Structures 

3.13 Reaictor Kinetics 

3.14 Control System and Trips 

4. STEADY STATE CALCULATION AND RESULTS 

5. TRANSIENT CALCULATION AND ANALYSIS OF RESULTS 

5.1 Transient Calculation Description 

5.2 Analysis of Results 

6. RUN STATISTICS 

7. CONCLUSIONS 

FIGURES

V
FIGURES 

FIGURE 1 

FIGURE 2 

FIGURE 3 

FIGURE 4 

FIGURE 5 

FIGURE 6 

FIGURE 7 

FIGURE 9 

FIGURE 10 

FIGURE 11 

FIGURE 12 

FIGURE 13 

FIGURE 14 

FIGURE 15 

FIGURE 16 

FIGURE 17 

FIGURE 18 

FIGURE 19 

FIGURE 20 

FIGURE 21 

FIGURE 22 

FIGURE 23 

FIGURE 24 

FIGURE 25 

FIGURE 26 

FIGURE 27 

FIGURE 28 

FIGURE 29 

FIGURE 30 

FIGURE 31

Vandell6s II NPP model general nodalization for RELAP 5/MOD 3.2. 

Reactor Pressure Vessel nodalization 

Steam Generator nodalization 

Pressurizer and associated piping and valve nodalization 

Steam Lines nodalization 

Main Feedwater and Auxiliary Feedwater Systems nodalization 

Emergency Core Cooling Systems nodalization 

Reactor fission power 

Pressurizer pressure 

Nuclear power 

Pressurizer collapsed level

Loop 1 

Loop 2 

Loop 3 

Loop 1 

Loop 2 

Loop 3 

Loop 1 

Loop 2 

Loop 3

hot leg temperature 

hot leg temperature 

hot leg temperature 

cold leg temperature 

cold leg temperature 

cold leg temperature 

average temperature 

average temperature 

average temperature

Loops 1, 2, and 3 delta temperature 

RCS loop 1 mass flow rate 

RCS loop 2 mass flow rate 

RCS loop 3 mass flow rate 

Steam generator A pressure 

Steam generator B pressure 

Steam generator C pressure 

Steam generator A NR downcomer level 

Steam generator B NR downcomer level 

Steam generator C NR downcomer level

vi
FIGURE 32 

FIGURE 33 

FIGURE 34 

FIGURE 35 

FIGURE 36 

FIGURE 37 

FIGURE 38 

FIGURE 39 

FIGURE 40 

FIGURE 41 

FIGURE 42 

FIGURE 43

TABLES 

TABLE 1 

TABLE 2 

TABLE 3 

TABLE 4

Steam generator A steam mass flow rate 

Steam generator B steam mass flow rate 

Steam generator C steam mass flow rate 

Steam generator A MFW mass flow rate 

Steam generator B MFW mass flow rate 

Steam generator C MFW mass flow rate 

AFW mass flow rates 

Pressurizer relief valves mass flow rate 

Pressurizer safety valves mass flow rate 

Steam generator A relief and safety valves discharged mass 

Steam generator B relief and safety valves discharged mass 

Steam generator C relief and safety valves discharged mass

Vandell6s H NPP nominal conditions at 100% rated power. 

Vandell6s II NPP status at event initiation time 

Steady state calculation key parameter results 

Sequence of events

vii
EXECUTIVE SUMMARY

The work to which this report refers falls within the framework of the 

CAMP/SPAIN Project, the objectives being those referred to in the CAMP Program itself, 

in particular, achievement of the capacity required to apply calculation tools supporting the 

quantitative analysis of transients and operation procedures, as well as joint participation in 

the international efforts aimed to validate and experience the use of appropriate calculation 

codes. 

The work consisted in using the RELAP 5/MOD 3.2 code to reproduce an actual 

transient that occurred at Vandell6s NPP Unit 2, on 26th June 1989, as a part of an 

assessment exercise within the framework of the CAMP Program. 

Vandell6s II is a three loop Nuclear Power Plant (NPP) with 2775 MWt (982 

MWe), Westinghouse design, Pressurized Water Reactor (PWR). The NPP project was 

developed during the 1980s. The first criticality was reached on 14th November 1987 and, 

after its first connection to the grid on December, it began commercial operation on March 

1988. 

The transient to be simulated consisted in the inadvertent manual closing of all 

three main steam isolation valves, producing the reactor and turbine trips on steam generator 

B low-low level. Several specific plant conditions existing by the time of the event, including 

the operability of certain systems and some setpoint shifts, along with the operator actions 

taken affected the key parameter evolutions and have been analyzed. 

The agreement between calculation results and plant data was acceptable, taking 

into account the existing uncertainties in the registered plant data described in this report, and 

Vandell6s NPP model for RELAP 5/MOD 3.2 code has shown itself to be adequate for 

simulation of plant transients, such as the case dealt with in this study. Nevertheless, as it 

is described below, it is worth pointing out the convenience of evaluating the applicability 

of a general purpose model to the particular case, in order to modify or remove as needed 

the non-suitable sections. For simplicity, as well as computing efficiency, it may be even

ix
advisable to delete those parts which do not have a significant effect on the specific transient 

calculation and, on the other hand, may produce any kind of problem. For example, in the 

calculation described herein, this latter was the reason to remove the pressurizer relief tank 

section from the model. 

Concerning the transient calculation statistics, it was run on a Hewlett Packard 

Apollo Serie 700, Model 735 Workstation, with 99 MHz clock frequency. Maximum time 

step used throughout the transient calculation was 0.1 seconds, the minimum being 6.25 

milliseconds during a 2 seconds period at the transient first stage (figure 43). The 1200 

seconds transient calculation required 3405 C.P.U. seconds (figure 44), i.e. a 2.84 overall 

ratio of C.P.U. time to real time.

x
1 INTRODUCTION 

The work to which this report refers falls within the framework of the 

CAMP/SPAIN Project, the objectives being those referred to in the CAMP Program itself, 

specifically, achievement of the capacity required to apply calculation tools supporting the 

quantitative analysis of transients, operational procedures, etc..., and joint participation in 

the international efforts aimed to validate and experience the use of appropriate calculation 

codes. 

The work consists in using the RELAP 5/MOD 3.2 code to reproduce an actual 

transient that occurred at Vandell6s NPP, Unit 2, on 26th, June, 1989. It is a part of an 

assessment exercise within the framework of the CAMP Program. 

The report includes the fundamental key points attained during each of the 

different stages of the work: 

- Identification of the main variables involved in the transient and its different 

stages. 

- Study and review of the reference plant (Vandell6s II NPP) general purpose 

model for RELAP 5/MOD 3.2 code. 

- Steady state calculation. 

- Implementation of the transient specific features and boundary conditions for the 

transient calculation, along with the suitable modifications and simplifications 

to the provided model. Sensitivity studies on the affecting boundary conditions 

to define the most likely combined evolution of some plant parameters for which 

records are not available or the registered data seem not reliable enough.

I
PLANT AND TRANSIENT DESCRIPTION

2.1 Plant Description 

The transient occurred at Vandell6s II NPP, a commercial Westinghouse 

Pressurized Water Reactor (PWR) owned by two Spanish utilities, ENDESA and 

IBERDROLA. The Unit went first critical in 1987. 

The Nuclear Steam Supply System (NSSS) consists of a reactor and three closed 

reactor coolant loops connected in parallel to the reactor vessel. Each primary loop is 

composed of a hot leg, steam generator tubes, crossover leg, main reactor coolant pump and 

cold leg. An electrically heated pressurizer is connected through a surge line to the hot leg 

of one of the loops. In addition, certain auxiliary systems, as the Chemical and Volume 

Control System (CVCS) and the Residual Heat Removal System (RHRS), are also connected 

to the coolant loops. 

The rated power of the reactor is 2775 MWt, electrical output being 982 MWe. 

The reactor core is made up of Zircaloy-clad slightly enriched uranium dioxide fuel rods in 

canless assemblies, various internal structures, reactivity control components and core 

monitoring instrumentation. The core is housed in a reactor vessel through which pressurized 

light water flows, acting both as moderator and coolant. 

48 Rod Cluster Control (RCC) assemblies are used for reactor control. They 

consist of clusters of 24 cylindrical absorber rods each. The control rods of cadmium-indium

silver (Ag-In-Cd) and a boron compound (B4C) neutron absorber in stainless steel (SS-304) 

clads extend full core length when fully inserted. Each control rod moves vertically in its 

own tubular guide thimble in selected fuel assemblies. Above the core, each cluster of 

absorber rods is attached to a spider connector and drive shaft, which is raised and lowered 

by a drive mechanism mounted on the reactor vessel head. Downward trip of the RCC is by 

gravity.

2

2
The Reactor Coolant System (RCS) has three similar loops, each connected to 

the reactor vessel. Each loop has a Reactor Coolant Pump (RCP) and a Steam Generator 

(SG). The system also includes a pressurizer With its relief fank, connecting piping and 

instrumentation lines needed for reactor control and protection. 

The Reactor Coolant System transfers the heat released in the core to the steam 

generator secondary coolant, where the turbine generator set driving steam is produced. The 

reactor coolant water is borated and demineralized and flows at a rate depending on the 

reactor thermalhydraulic condition. It also acts as neutron moderator and reflector. 

Reactor Coolant Pumps are vertical, centrifugal, Westinghouse model 100D, 

single stage and shaft-seal type. Each of them is driven by a single speed 7500 HP AC 

induction motor, mounted over the pump hydraulic section. The pump power supply system 

is designed in such a manner as to achieve a suitable water flow rate to cool the reactor core 

under any required condition. A fly wheel connected to the pump shaft provides additional 

inertia, thus increasing the pump coastdown time period. 

The Pressurizer is a electrically heated cylindrical vertical tank connected to one 

of the reactor coolant hot legs through a surge line. It maintains Reactor Coolant System 

pressure during normal operation, limits pressure variations during plant load transients, and 

keeps system pressure within design limits during abnormal conditions. 

The Steam Generators are Westinghouse design, model F, with 5626 vertical 

inconel U-tubes and an expanded upper section that houses integral moisture separation 

equipment to produce dry steam with a quality of at least 99.75 percent. 

Preheated Main Feedwater (MFW) enters the upper part of the steam generator 

downcomer through several J-nozzles localized in a feedwater ring. Circulation ratio on the 

secondary side of the steam generator is 3.27 at rated power. The auxiliary feedwater system 

includes a turbine driven pump and two motor driven pumps. They also inject water into the 

upper part of the downcomer.

3
The Secondary Side includes the usual Balance Of Plant (BOP) components: 

feedwater pumps, main steam lines, safety and relief valves, main steam isolation valves, 

turbine bypass and trip valves, main steam control valves, turbine, condenser, hotwell, heat 

exchangers, etc. 

Reactor coolant piping, reactor internals and all of the pressure containing and 

heat transfer surfaces in contact with reactor water are stainless steel clad, except the steam 

generator tubes and fuel tubes which are Inconel 600 TT and Zircaloy-4, respectively. 

The Reactor Control System maintains the programmed average primary coolant 

temperature by positioning the control rods (short-term reactivity control). Reactivity changes 

due to burnup of the fuel or neutron poisons (Xenon) build-up are compensated by the 

operator manually reducing primary system boron concentration. Primary pressure is 

controlled by means of the pressurizer spray and pressurizer heaters, while pressurizer level 

is controlled through CVCS makeup and letdown flow rates and pressurizer heaters. The 

turbine bypass shows the two classical aspects of this particular reactor type: average primary 

temperature control and secondary side steam pressure control. Level in the steam generator 

is controlled by means of the Feedwater System (FWS). 

Reactor Protection System (RPS) produces the reactor automatic scram when the 

values of some plant key parameters exceed their predefined setpoints. Plant safety is 

guaranteed by the RPS and the Engineered Safety Features, which include the containment 

cooling, spray, isolation and ventilation systems, the main steam lines and main feedwater 

isolation features, the emergency core cooling system (including three high pressure injection 

pumps, two low pressure injection pumps and three accumulators) and the auxiliary feedwater 

system. 

The nominal values of the main operating parameters at 100% rated power are 

listed in Table 1 below

4
Total core thermal power 2775 MWt 

NSSS total heat output 2785 MWt 

Heat generated in fuel 97.4 % 

Nominal RCS operating pressure 155.1 x 105 Pa 

Pressurizer water level 60 % 

Total primary coolant flow rate 13545 Kg/s 

Vessel inlet temperature 290.67 0C 

Vessel outlet temperature 326.22 °C 

RCS average temperature 308.44 °C 

Average temperature rise in vessel 35.56 °C 

Pump suction temperature 290.55 °C 

Steam generator shell side operating pressure 66.46 x 10' Pa 

Steam temperature 282.33 °C 

Steam generator narrow range water level 50 % 

Total steam flow rate (all three SG) 1538.5 Kg/s 

Total SG blowdown flow rate (all three SG) 47.78 Kg/s 

Total main feedwater flow rate 1586.3 Kg/s 

Feedwater temperature 223.89 °C 

Steam generator circulation ratio 3.27 

TABLE 1 : Vandell6s II NPP nominal conditions at 100% rated power 

5
2.2 Transient Description

On June 26, 1986, with Vandellos NPP Unit 2 ihi operation Mode 1 (power 

operation) with 989 MWe electrical load, and while performing maintenance works on the 

remote shutdown panel, a manual main steam isolation signal was inadvertently activated, 

producing the reactor and turbine trips on steam generator B low-low level. 

According to the Scram Analysis Report, by the time of the transient initiation, the 

following main control systems were available and performing their intended functions in 

their automatic modes : 

- Control rods 

- Turbine 

- Pressurizer level 

- Pressurizer pressure 

- Steam generator level 

- Main feedwater turbine driven pumps 

- Charging flow 

- Steam generator relief valves (due to on-going maintenance activities, isolation 

valves upstream of steam generator A and B relief valves were closed, and were 

reopened after the reactor trip) 

- Turbine bypass (in Tavg. mode) 

In addition, primary and secondary main parameters were within their operating 

normal values at the existing power level. Plant conditions at the time of the event were

6
Operation mode 1 (power operation) 

Nuclear power 100 % (2775 Mwt) 

Output power 989.7 MWe 

Rod position Control bank D 211 steps 

Boron concentration 156 ppm 

Pressurizer pressure 158.4 Kg/cm2 

Average temperature 308.6 °C 

Pressurizer level 58.8 % 

Steam generator level 50% 

Operating RCPs A, B, C 

Operating MFW turbine driven pumps A, B

TABLE 2 : Vandell6s II NPP status at event initiation time

7
According also to the Scram Analysis Report and the transient recorded data, the 

following actuations were manually or automatically initiated during the transient evolution: 

1. Inadvertent manual closure of all three main steam isolation valves (transient initiating 

event). 

2. Reactor and subsequent turbine trips on low-low steam generator B level. 

3. Turbine bypass valves opening as design (in Tavg. mode) 

4. Pressurizer PORV's opening and proper seating. 

5. Steam generator relief valves opening. 

6. Opening of the first safety valve in all three steam generators and the second one in 

steam generator C. Although all these valves seated again properly, the steam 

generator C first and second safety valves closed at a lower pressure than designed. 

As a result, a steam leakage through these valves was detected some 11 minutes after 

the scram. Steam generator C safety valves were observed close approximately 15 

minutes after the scram. 

7. MFW turbine driven pumps trip, MFW system isolation and subsequent Auxiliary 

Feedwater System actuation. 

Some uncertainties exist regarding operator manual actions and some other 

automatically controlled actuations during the transient, because of lack of associated 

registered data or further details addressing these particular issues in the Scram Analysis 

Report, namely

8
manual control of the pressurizer heaters in order to recover pressure conditions and 

resulting heater power evolution 

manual control of auxiliary feedwater mass flow rate and resulting flow evolution 

steam generator wide range level evolution (narrow range level was lost early in the 

transient) 

by the time of the inadvertent closure of the MSIVs from the remote shutdown panel, 

maintenance activities were being performed on this panel. In order to support these 

activities, the motor driven isolation valves upstreani of the steam generators A and 

B relief valves were closed and manually reopened several seconds after the scram, 

but the exact time for the reopening of these valves is not accurately known. 

as it has been mentioned above, according also to the Scram Analysis Report, the 

steam generator C safety valves 1 and 2 opened as a result of the steam generator 

pressure increase and seated again properly afterwards, but at a lower pressure than 

designed, the exact value of this pressure being unknown. As a consequence, a steam 

leakage through these valves was detected some 11 minutes after the scram. Steam 

generator C safety valves were observed close approximately 15 minutes after the 

scram. 

Finally, no credit has been taken for some registered data during certain stages 

of the transient because of lack of reasonableness. For example, plant data regarding steam 

flow rates after the main steam lines isolation seemed not reliable.

.9
CODE INPUT MODEL DESCRIPTION

The work described herein was carried out using the RELAP 5/MOD 3.2 code, 

running on a Hewlett Packard 735 workstation. Graphics results were obtained via a program 

developed by TECNATOM and adapted to RELAP 5 restart-plot file structure for graphics 

applications. 

The model was developed by the licensee technical services department staff and 

provided to TECNATOM for the transient calculation. It constitutes the result of a long 

standing task, during which a previous model for the older version of the code, RELAP 

5/MOD 2, was updated and adapted to RELAP 5/MOD 3.2 input requirements, and its scope 

was enlarged to support a wider range of applications. 

The nodalization and control logic were carried out so that it could be used to 

study a wide range of transients, the aim being to make a general purpose model available, 

with a range of application as large as possible. Therefore, they include components and 

control systems which do not affect this specific transient. As it shall be described later, 

some of these components, namely those representing the pressurizer relief tank and 

associated rupture disk and piping and the whole MFW model, were removed from the 

model for the transient calculation. Nevertheless, the originally provided complete model is 

described herein. 

Vandell6s II NPP model for RELAP 5/MOD 2.3 contains representative 

modelling for all major Nuclear Steam Supply System and Engineered Safeguard Features 

components, along with a substantial portion of the Balance of Plant systems. In general, the 

one-dimensional, non-equilibrium, two-velocity formulation was applied. Choking model was 

applied at all junctions where choked flow were expected to occur under some condition. 

Each loop has been modelled separately, including individual hot leg, steam generator, 

reactor coolant pump and cold leg modelling. Each of the two HPIS and LPIS trains were 

modeled separately. Figure 1 shows Vandell6s NPP model for RELAP 5/MOD 3.2 

nodalization scheme. The input model consists of 381 hydrodynamic volumes, 412 junctions, 

and 57 heat structures with 279 mesh points.

10

3
In addition, a complete general purpose control system was implemented. A 

interactive control variables set was also added to allow the code being run with the Nuclear 

Plant Analyzer software. 

Kinetics data corresponding to a EOL condition were introduced to use the 

RELAP 5/MOD 3.2 point kinetics model. 

3.1 Reactor Vessel 

Reactor vessel nodalization is shown in Figure 2. Flow enters from the cold legs 

into the reactor vessel downcomer, modeled by means of annulus component 504. The 

primary reactor vessel flow path is downward through this latter component to the lower 

plenum, component 506. A small portion of the vessel inlet flow is diverted upward through 

branch 502, which models the downcomer upper region above the cold and hot legs 

centerline elevation. It flows into the vessel upper head, single volume 520, through branch 

component 500. The junction between components 502 and 500 represents the upper reactor 

vessel bypass nozzles. Water flowing downward through the downcomer annulus into the 

lower plenum is subsequently splited between a main path through the reactor core, pipe 

component 510, and all the core bypass flow paths, modeled by means of pipe 512. 

The upper plenum is represented by branches 518, 516 and 514. In this latter, 

flow coming down from the upper head joins together with those coming up from the reactor 

core and core bypass. All the coolant flows exit then the vessel through the outlet nozzles 

towards the hot legs. 

The reactor core, modeled by meanr of pipe component 510, is divided into six 

active cells. Heat structures 5101 and 5111 represent the 41448 average fuel pins (157 fuel 

assemblies containing 264 fuel rods each) and a single hot rod respectively.

11
Reactor Coolant System Pinin2

The nodalization of the hot, cold and crossover legs for each primary loop is 

shown within the general nodalization scheme in Figure 1. All three hot legs are modeled by 

means of three branch components each, the only difference being the connection of the 

pressurizer surge line to the branch component 104 in loop 3. Pump suction or cross-over 

legs are represented by a pipe component with three cells. The cold leg of each loop are also 

modeled by means of three branch components. Accumulator discharge piping and both trains 

of the HPIS and LPIS are connected to the first branch component in each cold leg 

submodel. Pressurizer spray lines join also the first branch components of loops 2 and 3 with 

the pressurizer head. Finally, charging flow from the CVCS enters into the first branch of 

loop 2 cold leg, whereas the letdown is taken from the corresponding branch in loop 1. 

3.3 Steam Generator Primaries 

Nodalization schemes for all three steam generators are identical. Nodalization 

for steam generator C is shown in Figure 3. Flow from each loop hot leg enters into the 

associated steam generator inlet plenum, modeled by means of a branch component (branches 

170, 140 and 110 respectively). A pipe component (pipes 111, 141 and 171) , containing 14 

cells, represents the 5626 U-tubes of each steam generator. These tubes are lumped into a 

single component, which volume and flow area are the sum of the volumes and flow areas 

of the whole tube bundle, while length and elevation are those of an equivalent average tube 

and the hydraulic diameter is the one corresponding to a single tube. Finally, the outlet 

plenum is represented by another branch component for each steam generator (branches 172, 

142 and 112 respectively). 

3.4 Reactor Coolant Pumps 

Reactor coolant pumps are represented in the model through pump components 

178, 148 and 118, as it is shown in Figure 1. A complete set of homologous curves both for 

single or degraded flow are entered. Pump performance is in this way expressed in terms of 

the single phase data and the difference data using a two-phase multiplier that is a function

12

3.2
of void fraction. In addition, neither pump velocity table nor control trip were entered. 

Instead, a representative table of relative pump motor torque (pump motor torque/rated motor 

torque) was implemented. Thus, the code solves the torque-ineitiia equation to calculate the 

pump velocity, whether the pump is tripped or not (in the former case, considering a null 

value for the pump motor torque). 

3.5 Pressurizer and Associated Systems 

Vandellos II model nodalization for the pressurizer and its associated systems 

is shown in Figure 4. The pressurizer upper head is modeled with branch 406, and the 

pressurizer cylindrical body and lower head are modeled with 7-cell pipe 404. The upper 

head has 5 connections : one to the cylindrical body, one to the spray line, one to the 

common line from which two PORV valves take suction (pipe component 429) and one to 

each of the three safety valve inlet piping (branches 439, 441 and 442). The surge line is 

modeled with 5-cell pipe 400, which is connected to the pressurizer cylindrical body through 

single junction 417 and to loop 3 hot leg at the middle point of branch 104 component. 

Each of the model pressurizer spray lines which take suction from loops 2 and 

3 include a spray valve (servo-valves 410 and 420) and a single junction (414 and 424), these 

latter representing the continuous spray. They connect loop 2 and 3 cold legs (namely branch 

components 150 and 120) to the pressurizer head, by means of pipes 408 and 418 and branch 

components 412, 422 and 425. Governed by the associated control logic, the spray valves 

open in response to mild primary coolant overpressurization. 

Power-operated relief valves (PORVs) are modelled by motor valve components 

426 and 428. Operation of these valves is governed by means of the associated control 

system as a function of the primary pressure values and trends. On turn, pressurizer safety 

valves are modeled through servo-valve components 430, 432 and 434. The pressurizer relief 

tank was originally modeled downstream of both kind of valves by means of 6-cell pipe 427, 

with its rupture disk, valve 461, connecting it to the containment atmosphere, time dependent 

volume 462. Nevertheless, because it did not have a significant importance for the specific 

transient and some problems arose with the incondensable gas model calculation before the

13
PORV valves opening, all the mentioned components downstream the pressurizer safety and 

relief valves were removed from the model for the transient calculation. 

3.6 Steam Generator Secondaries 

The nodalization for the loop 3 steam generator is shown in Figure 3, and it is 

completely analogous to those corresponding to steam generators A and B, the only 

difference being the numbering pattern which is also specified within this figura. Modelling 

of the steam generator primary region was described previously in section 3.3. 

As it is shown in Figure 3, main feedwater enters the loop 3 steam generator 

downcomer annulus at branch component 220, where it is Eombined with the recirculation 

liquid flow returning from the separator, component 208, through the downcomer annulus 

branch 218. The combined flow descends through the downcomer, modeled by means of the 

vertical annulus component 222 with 4 nodes, and enters the boiler lower region through 

single junction 226, which represents the path from the downcomer to the boiler. 

Valve component 201 connects the lower node of the downcomer annulus 222 

to a time dependent volume, in order to model the steam generator blowdown. 

The boiler region surrounding the steam generator tube bundle is modeled by 

pipe component 200 with 7 hydrodynamic cells. There, the heat removed from the core by 

the primary coolant and transfered to the steam generator secondary side coolant is calculated 

by means of heat structure 1111, representing the tube walls. Axial cell distribution was 

made consistent between the primary and boiler regions to make the definition of the 

associated heat structure easier. In this way, the boiler cells correspond to the upflow and 

downflow hydrodinamic nodes of the tubes, except in the bend region, where the 

corresponding tube section is divided into two cells, surrounded by branch 200 seventh cell. 

The two-phase mixture exiting the boiler region flows through the mid-steam 

generator regions, branches 204 and 206, before entering the volume inside the separator 

cylinders, branch-separator component 208. The separation process in both the swirl vane

14
separators and the steam dryers is assumed to occur in this component. The separator 

performance is idealized, not allowing liquid fractions passing through the vapor outlet into 

the steam dome region unless the separator cylinders volume" is completely full of liquid 

(VOVER = 10"36). For the liquid fall back junction, recirculation flow may contain vapor 

fractions for liquid fractions in the separator volume lower than 0.15 (VUNDER) and is a 

single phase liquid flow otherwise. 

A separator bypass flow path, representing the volume between the outermost 

steam separator cylinders and the inside of the steam generator shell, is modeled by means 

of branch component 216. During normal steam generator operation, this region is stagnant 

and contains the quiescent liquid level. 

Steam coming out from the separator enters the steam dryer region, which 

volume is modeled through branch components 210 and 212. No separator features have been 

implemented into these components. Finally, the steam flows into the steam dome volume, 

single volume 214, and exits the steam generator through its steam outlet nozzle, sngljun 

224. 

All the above description is applicable to the nodalization of steam generators 

A and B. Only the associated component numbering scheme differs among them. 

3.7 Main Feedwater System 

The feedwater system included in the general purpose model is shown in Figure 

6. It extends from the main feedwater header, time dependent volumes 300 and 310, from 

which both feedwater pumps take suction, to the steam generator feedwater inlet nozzles. The 

two main feedwater pumps are explicitly modeled by means of pump components 301 and 

311. Check valves at the outlets of these pumps are represented by check-valve components 

303 and 313 respectively. Discharges from the two main feedwater pumps are combined 

before flowing through the high pressure heater, which primary and secondary sides are 

modeled by means of branch component 321 and time dependent volume 306 respectively, 

thermally connected through heat structure 3221. A bypass line around this heater is also

15
modeled, with the associated bypass isolation valve, valve component 341. Components 344, 

345, 347, 355, 357, 365 and 367 model the main feedwater lines to the individual steam 

generators. A main feedwater control valve is modeled for each of these lines, valves 346, 

356 and 366. Finally, check valves 348, 358 and 368 are introduced upstream of the 

connection between these lines and those from the auxiliary feedwater system, to prevent any 

reverse flow through the main feedwater lines. 

Both feedwater flows, main and auxiliary, when operating, enter the steam 

generators through the corresponding feedwater nozzles (single junctions 373, 363 and 353). 

3.8 Auxiliary Feedwater System 

The auxiliary feedwater turbine driven pump is modeled through time dependent 

volume 370 and time dependent junction 371. This latter injects feedwater flow into a header, 

single volume 359, under demand of the associated control system, and depending on the 

velocity of the pump driving turbine and on the discharge pressure in the header. Auxiliary 

feedwater flow splits among three lines to the three steam generators. Each of these lines 

contains a check valve (valves 324, 325 and 326) and an isolation valve (valves 377, 376 and 

375). Downstream of the isolation valves, these lines connect to the ones from the auxiliary 

feedwater motor driven pumps. 

Feedwater make-up from the two auxiliary feedwater motor driven pumps is 

modeled separately by means of time dependent volumes 350 and 360 and corresponding time 

dependent junctions 351 and 361. Under demand, flow is injected into the common discharge 

header, single volume 354, depending also on the existing pressure in it. Water flows then 

towards the steam generators through three separate lines, each of them provided with a 

check valve (check valves 329, 328 and 327) and an isolation valve (servo-valve components 

374, 369 and 364).

16
3.9 Main Steam System

Main steam system modeled extends from the steaffi generator outlet nozzles to 

the turbine first stage and steam dump lines common discharge header, including all the 

associated valves. Steam extractions taking suction from the steam header are also modeled, 

along with those from steam lines 1 and 3 to the auxiliary feedwater turbine driven pump. 

The wholw system is shown in Figure 5. 

Single junctions 284, 254 and 224 represent the steam generator outlet nozzles, 

connecting the steam dome of each steam generator to the associated steam line. Reduced 

flow areas and added flow losses are defined for these junctions in order to represent the 

steam flow restrictors at those locations. 

The first section of the steam lines between the steam generator outlet nozzle 

and the connections to the relief and safety valves are represented by 3-cell pipe components 

660, 630, and 600 respectively, with different lengths from each other since all three steam 

lines are not completely symmetrical. Branch components 662, 632 and 602 model the next 

steam line sections. All lines to the steam generator relief and safety valves are connected 

to the outlet of these branch components. 

Each steam line is provided with a relief valve (motor-valves 668, 638 and 608) 

and five separately modeled safety valves, which are represented by means of five servo

valve components. 

Main steam isolation valves are included downstream branch components 662, 

632 and 602 through motor-valve components 664, 634 and 604. The last section of the main 

steam lines are represented in the model by pipe components 666, 636 and 606, which drive 

the steam flow from each steam generator to the common steam header, branch component 

654.

17
From the steam header, three flow paths are modeled. The first one models the 

normal operating steam flow path to the turbine, represented by a pressure boundary 

condition through time dependent volume 711. This line includes the turbine control and stop 

valves, which are lumped together into a single valve component, servo-valve 710. 

All four steam dump lines to the condenser are explicitly and separately 

modeled, each of them with its associated valve, servo-valve components 780, 781, 782 and 

783. Each of these lines discharge into a separate pressure boundary condition, time 

dependent volumes 785, 786, 787 and 788. 

During normal operation condition, a small portion of the main steam flow 

entering the steam header is diverted through a third steam header outlet line which 

represents several steam consumptions, including those to the moisture separator reheaters 

and the main feedwater turbine driven pumps. The associated flow rate is controlled in the 

model by valve components 650 and 641 and is discharged into time dependent volume 642. 

In addition, steam supply to the auxiliary feedwater turbine driven pump is also 

modeled by means of trip-valve components 622 and 620, which take suction from the inlet 

of branch components 662 (steam line 1) and 602 (steam line 3) and discharge into branch 

component 624. Then, this steam flow passes through servo-valve 626 to a common pressure 

boundary condition, time dependent volume 628. 

3.10 Emergency Core Cooling System 

The model of Vandell6s II ECCS includes two High Pressure Injection System 

(HPIS) trains, two Low Pressure Injection System (LPIS) trains and three nitrogen

pressurized accumulators. HPIS and LPIS models are showed in Figure 7. 

Both HPIS and LPIS take suction from the refuelling water storage tank, 

modeled by means of time dependent volume 810. Under demand and depending on the 

ECCS pumps condition, emergency core cooling water exits this pressure boundary

18
condition, passes through a common line, branch 816, and is diverted between two lines, 

branch components 852 and 854, each supplying flow to a HPIS train and a LPIS one. 

The two HPIS pumps are separately modeled by pump components 813 and 819, 

each provided with a check valve, valve components 815 and 821, at their discharge. 

Downstream of these valves, both trains join together in a shared flow path, provided also 

with a check valve, valve component 823, before being splited into three different pipes, 

each driving the emergency coolant flow to one loop hot leg and equipped also with a check 

valve, valve components 829, 830 and 831. 

The LPIS trains are quite similar to those of the HPIS described above. Each 

of them consists of an independent pump, pump components 836 and 841, with a check valve 

downstream, check-valve components 838 and 843. Both trains share a common flow path, 

modeled through branch component 845, from which three lines to each RCS loop cold leg 

are modeled, each provided with a check valve, valve components 846, 847 and 848. 

The last sections of the injection lines to the three RCS loop cold legs, branch 

components 832, 833 and 834, are shared between the HPIS and LPIS. 

The model for the accumulators and associated piping and valves can be seen 

in the general nodalization draw showed in Figure 1. All three accumulator models are 

essentially identical, each of them comprising an accumulator component (806, 803 and 800), 

and two discharge pipes which connect the pressurized tanks to the three RCS cold legs 

through a check valve, (valve components 862, 860 and 817) and a isolation valve (motor 

valves 808, 805 and 802). These latter get open manually or on safety injection signal. 

3.11 Chemical and Volumetric Control System 

As it is shown in Figure 1, CVCS makeup is modelled through two time 

dependent volume components (460 and 463), discharging into a common line (single volume 

458) through two regulating valves (servo-valves 459 and 464, respectively). Flows coming 

from both boundary conditions show different boron concentrations, and get mixed in the

19
shared discharge line which drives them to the loop 2 cold leg after passing through an 

isolation valve, valve component 455. 

CVCS letdown, on turn, is modeled through a flow boundary condition, time 

dependent junction 450, which takes suction from RCS loop 1 cold leg and discharges to time 

dependent volume 440. The discharge flow is estimated depending on the discharge orifices 

manually opening and on the pressurizer level. 

3.12 Heat Structures 

Fuel rods are modeled through heat structures 5101 and 5111. Heat structure 

5101 represents the average rod condition, which is applied to all the 41448 fuel rods in the 

core (157 fuel assemblies containing 264 fuel rods each), with 4354.8 m2 heat transfer area. 

Heat structure 5111 represents a single hot rod, with a hot channel factor of 1,43. Both heat 

structures are defined with 6 axial nodes (one per each associated hydraulic node) and 5 

radial mesh points. The first two intervals between mesh points represent the fuel (U0 2), 

their power source being calculated by the point kinetics model. The third interval represents 

the gap and the last one the clad (Zircaloy-4). 

Heat structures 1711, 1411 and 1111 model the heat transfer between the 

primary and secondary coolant through the tube walls of steam generators A, B and C 

respectively. All of them are defined with 14 axial nodes and 5 radial mesh points and 

represent the 5626 tubes in each steam generator (no steam generator tube plugging is 

considered), with a total heat transfer area of 5085 m2 per steam generator. The convection 

and boiling correlations for vertical tube bundles without cross-flow were selected both for 

the inner and outer tube wall surfaces. 

Pressurizer heaters are modeled through heat structure 4041, which transfer the 

heater power calculated by control variable 178 to the coolant in the pressurizer model (pipe 

component 404) two lower cells. No heat losses through the pressurizer walls are considered.

20
Finally, Main Feedwater Heat Exchanger is modeled by means of heat structure 

3221, which calculates the heat transfer from time dependent volume 306 to the feedwater 

in branch 321. This heat structure was deleted, along with the rest of the main feedwater 

model provided, when this latter was simplified by directly introducing the actual transient 

main feedwater flow recorded data by means of boundary conditions. 

All the above mentioned heat structures include 57 axial nodes and 279 radial 

mesh points. 

3.13 Reactor Kinetics 

Concerning the core, the code point kinetics model is used, with feedback due 

to moderator density and volumetric average fuel temperature, introduced through separate 

tables and affected by the associated volume density and fuel temperature weighting factors 

respectively. Decay of standard fission products is also considered. Total reactor power 

(fission power plus fission product decay power) is 2775.41 MWt. EOL condition values for 

the delayed neutron fraction (0) and prompt neutron generation time (A), along with those 

for the delayed neutron precursor yield ratios (f) and decay constants (Xi), are used. 

Selected fission product data are those defined in the 1979 ANS Standard for 

235

U. Initial conditions for fission product are the ones corresponding to operation at the 

defined power level for a 1000 hours period time. 

Feedback reactivity ($) due to moderator density (kg/m3) and fuel temperature 

('K) are introduced by means of tables of these parameters versus reactivity ($), RpLO] and 

RF[TFi]. In addition, for each specific hydrodynainic cell a volume weighting factor Wj is 

applied to the moderator density reactivity table resulting value, while analogous fuel 

temperature WFi weighting factors defined for the associated heat structure axial node are 

applied to the fuel temperature reactivity table.

21
Finally, scram reactivity is introduced through table 850, whereas control rod 

and boron reactivity are assessed and introduced through control variables 140 and 749 

respectively. 

3.14 Control System and Trips 

The model has been completed by including the following plant control and 

protection systems by means of RELAP 5 trip and control variable provided scheme 

- Average temperature (rod control) 

- Pressurizer pressure and level 

- Steam generator level (main feedwater valves and turbine driven pumps control 

system) 

- Auxiliary feedwater control system 

- Steam generator relief and safety valves 

- Turbine and steam dump control 

- Reactor trips and AMSAC 

- Turbine trips 

- Permissives 

- Safety injection 

The complete control and protection systems include 753 control variables, 247 

variable trips, 372 logical trips and 101 general tables.

22
STEADY STATE CALCULATION AND RESULTS

The steady state at 100 % rated power level conditions was directly calculated 

with the model provided by the licensee and described above. The steady state condition 

achieved was subsequently used as the initial condition for the transient, by means of the 

code dump-restart feature. Key plant parameter values obtained at the end of the steady state 

calculation, which took 2000 seconds problem time (5186 seconds CPU time consumption 

with a maximum time step size of 0.1 seconds), are listed below in Table 3 and compared 

with the actual plant recorded data and nominal values. These latter are taken from the Final 

Safety Analysis Report, the Process Diagrams, and the document WTP-ENG-TN-81-001 

"Model F Steam Generator Thermal Hydraulic Report", while the actual plant data are those 

obtained as time averaged values of the recorded data just before the transient initiation. 

As it can be seen, a overall good agreement was achieved between the steady 

state calculation results and the actual plant conditions by the time of the event. It must be 

noted however that these recorded data are not completely consistent (for example, the 

recorded values of the core thermal power, temperature rise in vessel and total primary mass 

flow rate do not match the vessel thermal balance). Therefore, because of lack of data to 

analyze the plant instrumentation measurement errors, the steady state condition reached was 

considered an acceptable compromise initial condition for the transient calculation. However, 

as it is described later, the small discrepancies between some of the steady state calculation 

results with respect to the plant data could be the reason for some differences between the 

recorded and the calculated transient evolutions of certain variables, in particular those 

corresponding to the steam generator secondary side pressure. In addition, it should be 

mentioned that the discrepancy between the calculated and measured reactor coolant flow 

values may arise, to a considerable extent, from the fact than volumetric flow rates are 

measured at the plant and then compared to a reference 100 % value to obtain the percentage 

readings. Because no exact data were available about this reference value neither regarding 

under which conditions it was defined (in order to relate volumetric and mass flow rates), 

the mass flow rate included in table 3 reference values column was taken as 100 % mass 

flow rate, but the accuracy of this assumption may not be assured.

23

4
REFERENCE 

VARIABLE RECORDED DATA RELAP 5 MOD 3.2 

VALUE 

Total core thermal power (MWt) 2775 2788 (N0049: 100.5%) 2794.93 (RKTPOW / 106) 

NSSS total heat output (MWt) 2785 2804.87 (CNTRLVAR 921 +922 +923) 

Heat generated in fuel (%) 97.4 100.0 (HS 5101, 5111) 

Nominal RCS operating abs. pressure (bar) 155.1 156.6 (P0480) 155.16 (P 406010000) 

Pressurizer water level (%) 60 59.15 (L0480) 59.87 (CN'T'RLVAR 146) 

Total primary coolant flow rate (Kg/s) 13545 13441.15 (Avg. F0400/20/40 99.23%) 14115.2 (MFLOW 18801 +15801 +12801) 

Vessel inlet temperature (°C) 290.67 292.33 (Avg. T0650/70/90) 292.04 (Avg. CNTRLVAR 47, 45, 43) 

Vessel outlet temperature (QC) 326.22 324.87 (Avg. T0651/71/91) 326.22 (Avg. CNTRIVAR 46, 44, 42) 

RCS average temperature (°C) 308.44 308.6 ([Th,+Tc,,j/2) 309.13 (Avg. CNTRLVAR 112, 111, 110)

TABLE 3 : STEADY STATE CALCULATION KEY PARAMETER RESULTS

24
""" REFERENCE 

VARIABLE RECORDED DATA - RELAP 5 MOD 3.2 F VALUE 

Average temperature rise in vessel (*C) 35.56 32.54 (T1,

0 -Tco1d) 34.18 (Avg. CNTRLVAR 533, 523, 513) 

Pump suction temperature (°C) 290.55 291.77 (Avg. TEMPF 176, 146, 116) 

SG shell side operating abs. pressure (bar) 66.46 67.7 (Avg. P0400/20/40) 67.92 (Avg. P 66001, 63001, 60001) 

SG narrow range water level (%) 50.0 50.8 (Avg. L0400/20/40) 50.0 (Avg. CNTRLVAR 257, 234, 210) 

Total steam flow rate (Kg/s) 1538.5 1530.4 (F0405 +F0425 +F0445) 1544.37 (MFLOW 66001 +63001 +60001) 

Total SG's blowdown flow rate (Kg/s) 1 15.39 17.49 (MFLOW 26100+23100+20100) 

Total main feedwater flow rate (Kg/s) 1586.3 1558.4 (F0403 + F0423 + F0443) 1561.8 (MFLOW 26600+35600+34600) 

Feedwater temperature (°C) 223.89 224.63 (Avg. T7503/04/05) 223.89 (Avg. TEMPF 37201, 36201, 35201) 

Steam generator circulation ratio (-) 3.27 3.31 (CNTRLVAR 701)

TABLE 3 : STEADY STATE CALCULATION KEY PARAMETER RESULTS (Cont.)

25

i i
In addition to the small discrepancies observed between the calculated and the 

recorded parameter values, a significant feature dealing with the void fraction distribution 

in the boiler cells which affected the steam generator secondary mass distribution was 

observed. The void fraction increased from volume to volume up to the boiler region (7-cell 

pipe 200) fifth cell, where it dropped below that of the fourth cell, increasing again when 

ascending across the remaining upper cells. It was observed that the code predicted a slug 

flow regime in the first five cells and then an annular-mist flow in the upper ones. 

Consequently, it is thought that the described anomalous result had to do with the flow 

regime transition. It could be produced by the different interphase frictions between both 

regimes, which would lead to less liquid being carried over in the upper cells towards the 

separator volume. 

It is worth to point out that, when investigating this circumstance, the same code 

behaviour was also observed in the typical PWR model provided with the code as a sample 

problem (typpwr. i). Other sources consulted in this regard also referenced this matter but did 

not provide suitable solutions.

26
TRANSIENT CALCULATION AND ANALYSIS OF RESULTS

5.1 Transient Calculation Description 

Transient calculation was conducted by means of a "restart" calculation, with 

transient option, starting from the conditions of the last dump generated in the steady state 

calculation. Given that this calculation was also performed with the transient option, a 

intermediate calculation was carried out with the steady-state option and 1 second calculation 

time. In this way, a significantly shorter restart-plot file containing the desired transient 

initial conditions in its first dump was generated, and problem time was automatically reset 

to zero at the beginning of the transient calculation. 

Thus, the initial transient conditions are those previously obtained from the 

steady state calculation, which are described in chapter 4 above. Conditions and perturbances 

constituting the event root causes and plant specific status by the time of the event, along 

with some ad-hoc suitable modifications and simplifications to the provided model are 

introduced in this restart input file. The calculated problem time was 1200 seconds, including 

a 154 seconds initial period under 'hnull transient" conditions (steady state or no perturbance 

conditions). 

The initiating event was implemented by substituting the interactive input 

variables in the MSL's isolation valves governing logic (used for the manual closing of these 

valves when running the code with the Nuclear Plant Analyzer software), for time tables, 

which produced the valves closing at time 154 seconds into the calculation, closing time 

being 1 second. 

Operator assumed manual actions and specific plant conditions affecting the 

transient evolution were implemented in the following way 

three time tables were defined, in order to apply a time dependent factor to the 

auxiliary feedwater mass flow rate calculated by the model. It allowed to 

simulate the operator manual control over the AFW makeup.

27

5
a similar time dependent factor was introduced to prevent the opening under 

demand of the steam generator A and B relief valves for a defined time period 

after the reactor trip. The objective was to reproduice the effect of the isolating 

valves downstream of these relief valves which, according to the Scram 

Analysis Report, were closed by the time of the trip to support maintenance 

activities and were reopened afterwards. 

trips and control variables defining the pressure close setpoints of steam 

generator C safety valves 1 and 2 were modified, in order to model the reported 

shift of these setpoints. 

Some model modifications and minor corrections were also implemented for the 

transient calculation through the restart input file. In addition, several suitable simplifications 

were introduced, the most significant being : 

because the model region downstream of the pressurizer safety and relief valves 

did not have a significant importance for the specific transient and, during the 

initial runs, some problems arose with the incondensable gas model calculation 

even before the PORV valves opening, all the involved components within this 

region were removed from the model and replaced by five time dependent 

volumes representing the pressurizer relief tank conditions at valve discharges. 

when first running the transient with the MFW model and associated control 

logic included in the provided model, calculation results and recorded data 

showed significant discrepancies regarding MFW mass flow rates evolution 

during the transient first stages. These discrepancies were attributed to some 

modifications introduced in the plant MFW control logic parameters (gains) 

after the event date (June 26, 1986), which were, on turn, implemented into the 

model itself, resulting in a different response of the updated model with respect 

to the old plant MFW control system. Given that the model of the MFWS and 

associated control logic corresponding to the parameter old values was not 

directly available to TECNATOM by the time calculation activities were carried

28
out, and both old and current MFW models had been extensively tested by the 

licensee, it was concluded that additional assessment on this subject was 

worthless and, consequently it was decided not to implement again the old 

model. Instead, MFW model was substituted for three time dependent volume 

components associated to a time dependent junction each, in order to directly 

introduce the MFW mass flow rate registered data as a boundary condition. 

Finally, a set of control variables were introduced to convert some calculated 

parameters into the units used in the plant data, in order to make their comparison easier. 

In addition, integral type control variables were introduced to evaluate the secondary valve 

discharged masses. 

5.2 Analysis of Results 

Transient results are shown in figures 8 to 44, most of which showing the 

calculation result plots (in wide line) compared with the associated plant instrumentation 

recorded data (in narrow line) for the fundamental variables. This set of variables was chosen 

based on their importance to the transient evolution. Additionally, it must be pointed out that 

the variables evolution introduced as boundary conditions, i.e. the main feedwater flow to 

all three steam generators, are also obtained from plant data. Consequently, the associated 

analytical and recorded plots completely match each other. 

Table 4 shows the sequence of the most important simulation events with 

RELAP 5/MOD 3.2

29
1 PLANT 

EVENT PATRELAP 5!MOD 3.2 

RECORDED DATA 

MSL isolation 154 s. 154 s. 

PRZ. PORV's 1 and 2 opening/closing 160 s. / 167 s. 

SG's 1, 2 NR low-low level (16 %) 162.4 s. 

Reactor trip (on SG NR low-low level) 160 s. 162.4 s. 

AFW initiation (100 % AFW flow: 94 162.4 s. 

Tm/h per SG aprox.) 

160.5 s. 164 s. 

PRZ. maximum pressure (164.4 Kg/cmr2) (164.4 Kg/cm2) 

SG's 1, 2 relief and safety valves 

opening/closing 

relief valves (SG-1, SG-2) 164 / 386, 416 s. 

safety valves 1 166 / 188 s. 

safety valves 2 166 / 184 s. 

safety valves 3 168 / 180 s. 

SG 3 relief safety valves opening/closing 

relief valve 162 / 282 s. 

safety valve 1 (reduced setpoint) 166 / 630 s. 

safety valve 2 (reduced setpoint) 166 / 196 s. 

safety valve 3 168 / 180 s. 

AFW flow step reduction (from 94 Tm/h 

per SG to 83 Tm/h aprox.) 

AFW flow step reduction (from 83 Tm/h 

per SG to 59 Tm/h aprox.)

TABLE 4 : Sequence of events

30
Event root disturbance, close of all three steam line isolation valves, is 

introduced at time 154 seconds into the calculation, thus interrupting the steam flow to the 

turbine (figures 31, 32 and 33) and resulting in a sudden steam generator pressure increase 

(figures 25, 26 and 27), up to the secondary safety and relief valves open setpoints. In 

particular, according to the calculation results, relief valves and safety valves 1, 2 and 3 in 

all three steam generators get opened between 162 and 168 seconds into the calculation 

(figures 40, 41 and 42). 

One of the main discrepancies between the analytical results and the transient 

reported data is the number of safety valves which were opened. As it has been described, 

according to the calculation results, secondary pressure increases up to the third safety valve 

open setpoint, thus producing three valves to lift in all three steam generators, while, 

according to the Scram Analysis Report, only the first valve in steam generators A and B and 

the first and second valves in steam generator C got open in the actual transient. The reason 

for this discrepancy may be, to a considerable extent, the higher secondary pressure obtained 

in the steady state calculation and used as initial condition for the transient calculation, 

leading to a calculated pressure maximum which slightly exceeds that recorded in the real 

transient. 

As a consequence of the steam flow interruption, the steam generator function 

as RCS heat sink becomes degraded, thus producing primary coolant pressure (figure 10) to 

increase, reaching a maximum value of 164.4 Kg/cm2 at 164 seconds time, and pressurizer 

both PORV's to lift (figure 38). 

Degradation of heat removal by the steam generators has a significant effect on 

cold leg temperatures (figures 15, 16 and 17), which experiment a steep increase and show 

a maximum value of 303 °C at 170 seconds, when the reactor trip stops this temperature 

excursion. Hot legs temperature increase (figures 12, 13 and 14) is quite lower. 

In addition, MSL isolation leads to a MFW flow reduction (figures 34, 35 and 

36) which, on turn, results in steam generator downcomer narrow range level to decrease 

(figures 28, 29 and 30). At time 162.4 seconds into the calculation, steam generator A and

31
2 low-low level setpoint (16 %) is reached, thus producing the reactor being automatically 

tripped (figure 9). This condition (1/3 steam generator low-low level) also causes the AFW 

initiation at time 162.4 seconds (figure 37) 

Reactor trip leads to the turbine trip and the turbine valve closing as well as the 

turbine bypass valve opening in average temperature mode. Both automatic actions have no 

effect over the RCS and steam generator secondary side conditions since Main Steam Lines 

are isolated. 

On reactor trip, RCS pressure and temperature increasing trends are removed, 

and the overall system conditions become controlled by the secondary side conditions as far 

as these latter determine the steam generator efficiency as RCS heat sink. 

Therefore, from reactor trip on, secondary side thermalhydraulic conditions are 

determined by the combined effects of the auxiliary feedwater mass flow and the secondary 

side valves opening. In this sense, as it was described in item 2.2, significant uncertainties 

existed regarding the AFW mass flow rate evolution and the secondary valves performance. 

Namely 

no recorded data were available on the auxiliary feedwater mass flow rate evolution 

under manual control, neither could it be estimated through sensitivity studies on the 

WR level evolution because of lack of recorded data on this latter (narrow range level 

was lost early in the transient) 

by the time of the inadvertent closure of the MSIVs from the remote shutdown panel, 

maintenance activities were being perform-ed on this panel. In order to support these 

activities, the motor driven isolation valves upstream of the steam generators A and 

B relief valves, were closed and manually reopened several seconds after the scram, 

but the exact time for the reopening of these valves is not accurately known. 

steam generator C safety valves 1 and 2 opened as a result of the steam generator 

pressure increase and seated again properly afterwards, but at a lower pressure than

32
designed, the exact value of this pressure being unknown. As a consequence, a steam 

leakage through these valves was detected some 11 minutes after the scram. Steam 

generator C safety valves were observed close approxiimiately 15 minutes after the 

scram. 

Therefore, sensitivity studies were focused on adjusting the secondary pressure 

evolution to the recorded data through the combined effect of manual step reductions of the 

AFW flow rate and the shift of steam generator C safety valves 1 and 2 close setpoints. In 

this way, steam generator calculated pressures were eventually adjusted to the recorded 

values by reducing the mentioned setpoints in 2.5 bars (safety valve 1 close setpoint becomes 

lower than relief valve setpoint), and introducing two step reductions to the AFW flow rates 

at time 555 seconds (from 94 Tm/h per SG to 83 Tm/h aproximately) and 755 seconds (from 

83 Tm/h per SG to 59 Tm/h aproximately) respectively. On the other hand, steam generator 

A and B relief valves were considered operables from the reactor trip on, that is, it was 

assumed that the isolation valves upstream of the mentioned relief valves, which were closed 

to support maintenance activities as it has been described above, were reopened on reactor 

trip. 

The maneouvre to manually reduce the AFW flow rate, is justified in order to 

prevent a safety injection actuation on low pressurizer pressure (setpoint : 128.018 bar) 

because of an excessive RCS cooldown through the steam generators.

33
6 RUN STATISTICS

The calculation has been performed using the RELAP 5/MOD 3.2 code, running 

on a Hewlett Packard Apollo Serie 700, Model 735 Workstation. 

"nj Control processor PA-RISC 7100 

"[] Clock frequency 99 MHz 

sj Performance: 

SPEC marks 89 147 

MIPS 124 

MFLOPS 40 

SPECint92 80 

SPECfp92 150 

"[] Main memory 48 Mby 

"[] Disk storage 2 Gby 

"[ Operating system HP-UX 9.03 

Steady state calculation took 2000 second problem time, and 5186 seconds CPU 

time consumption with a maximum time step size of 0.1 seconds, the ratio of C.P.U. time 

to real time consequently being 2.59. 

Concerning the transient calculation statistics, the maximum time step used 

throughout the transient calculation was 0.1 seconds, the minimum being 6.25 milliseconds 

during a 2 seconds period at the transient first stage (figure 43). The 1200 seconds transient 

calculation required 3405 C.P.U. seconds (figure 44), i.e. a 2.84 overall ratio of C.P.U. 

time to real time. It must be taken into account that, for the transient calculation, both the 

model regions downstream pressurizer safety and relief valves and that corresponding to the 

whole MFW system were removed, thus simplifying the model to a considerable extent. In 

fact, when first running the transient previously to the MFW section being removed from the 

model, C.P.U time consumption accounted for 3803.4 seconds (about 10 % higher)

34
7 CONCLUSIONS

The agreement between calculation results and plant data was acceptable, taking 

into account the existing uncertainties in the registered plant data described in this report, and 

Vandell6s NPP model for RELAP 5/MOD 3.2 code has shown itself to be adequate for 

simulation of plant transients, such as the case dealt with in this study. Nevertheless, as it 

is described below, it is worth pointing out the convenience of evaluating the applicability 

of a general purpose model to the particular case, in order to modify or remove as needed 

the non-suitable sections. For simplicity, as well as computing efficiency, it may be even 

advisable to delete those parts which do not have a significant effect on the specific transient 

calculation and, on the other hand, may produce any kind of problem. For example, in the 

calculation described herein, this latter was the reason to remove the pressurizer relief tank 

section from the model. 

The assessment of the code/model set reliability and goodness has resulted limited 

to a certain extent because of lack of some significant plant records, which made necessary 

to impose estimated combined evolutions of the associated parameters as .boundary conditions 

to reproduce the overall plant performance. 

A significant feature dealing with the void fraction distribution in the boiler cells 

which affected the steam generator secondary mass distribution was observed during the 

steady state calculation. The void fraction increased from volume to volume up to the boiler 

region (7-cell pipe 200) fifth cell, where it dropped below that of the fourth cell, increasing 

again when ascending across the remaining upper cells. It was observed that the code 

predicted a slug flow regime in the first five cells and then an annular-mist flow in the upper 

ones. Consequently, it is thought that the described anomalous result had to do with the flow 

regime transition. It could be produced by the different interphase frictions between both 

regimes, which would lead to less liquid being carried over in the upper cells towards the 

separator volume. It is worth to point out that, when investigating this circumstance, the 

same code behaviour was also observed in the typical PWR model provided with the code 

as a sample problem (typpwr.i). Other sources consulted in this regard also referenced this 

matter but did not provide suitable solutions.

35
Note. - some minor errors were detected in the code associated documentation, namely in the 

Volume II of the code manual (User's Guide and Input Requirements) 

- At the end of the Hydrodynamic Components describing cards section (item 

A7.11.6 : Accumulator Tank Initial Fill Conditions, Standpipe/Surgeline 

Length/Elevation and Tank Wall Heat Transfer Terms), there are several non

corresponding paragraphs (from "The volume flags entered in W8 ... 'to the end 

of the section) 

In the Reactor Kinetics section, items A12.8.6 (Cards 30001701 through 

30001799, Volume Weighting Factors) and A12.8.7 (Cards 30001801 through 

30001899, Heat Structure or SCDAP Component Weighting Factors), cards are 

described that should be used if the TABLE3 or TABLE4 options are selected, 

and not otherwise, as it is read.

36
LiL 

4NJ 

ITI 

I:I~ u

37
(3 

SNG~LUN

rD .

SRN C A 5i

(N 

"'5 

AN 

(3) 

k3) 

Lii 

0

1)

FIGURE 2: REACTOR PRESSURE VESSEL NODNLJZATION 

38

- /i
SNGLjuN A LINEAS 

24 DE VAPOR 

VSNGLVOL 2!4 

I 

BRANCH 21 2Z 

BRANCH 210 

BRANCH 216 SEPARATR 

208 

1L.GUA DE B8RANCH 218 CORRESPON 

CI I BRANCH VOLUMENE 

BRANETA 2 206 GENERADORE' 

a.V.1 I .v 

170 140 

BRANCH 204 171 I 141 

"-172 142 iP! P7 200(7) 260 6 230 7 8 0 (7 26- i 234. 

6 (6) 268 2.36 

3L12 (3)12 

2 13 (2)27 24 280 250 

SNGLJUN•1 14 (1) 282252 

22'6 

DE RAMA BRANCH BRANCH A RAMA 

CALIENTE FRfA 

FIGURE 3 :.STEAM GENERATOR NODALIZATION

39

-A
VALVULAýS 0CE AL;\/.0 

426 ______ 

4.33 CDscc DE

VAL2VULAýs OE SECURICAC 0 

439 ~I 4 

430 

-- I 27 

__ _ _ 4? 4 3 TANCUE: 

_ I IALIVIC 

L 438

4-5 

G4.-±-2 ES2AC 

NCL -UN 

400 :~ 

LINEA CE: COMPENSAC:CN (LAZO 2

-NGL-UN PP 

INPA N CPH 

420 - H LAZO 3 

2___ 2 4:18 

412 IDE LAZO 2 

- * 4 i 2 0 

ROC:ACO CCONTINUC 

FIGURE 4: 

PRESSURIZER AND ASSOCIATED PIEPING 

AND VALVES NODALIZATION

40

T.VDPV0DL
VAPOR A TURE3INA I URBI3 H~A I~II?,AIICII PlrE . 710 _ _ _ _ 

-- 70 > ,j- 757 76 _ _ 1

ii

FIGURE 5 : STEAM LINES NODALIZATION

DESCARGA AL- C:ONDENSADOR 

_____ 76 ] -775 -tq><j--[ 7135 

761 7f76 -"-1<=186 

___0 781 790 T-4i i6=2 L 7827_ 

17613-778 -L><1j- -3J 

650 652 641 

CONSUMOS VAPOR VARIOS I~ O 

.VULA VAL VUL AS 

ivin SEGUR IDAD 

A1)PVOl. IMOPVOL 

ý1 76 718(~ 2,T722 724 

A715X717 719 721 7 222j 

6RANCII 602 

604 

662 

600 

VAPOR A TUIRBOE30MBA -JUN 224 ACUA ALIMENTACION AUXILIAR : ul __ ý ,620 

INERADOR VAL VC VAPOR 3 21[<-62 

622 626 7)OPO

795 

0 

0
A.A.PRHIC JPAL

(300 502 

_00 1q 16 50

3\l 10) 31 - sill I i

A.A.AUXII IA!?

TURBOHO M BA

Q370)

57 1

MOl 0130M13AS 

329 ____-- 5-74 

351 3 28 369 

VF-4-jf 358 j t2

.50 4 327 354 64

505

3~ 1I3

530 

318 320

15

I3 11-I I ý

J 321 

CALI-N IlAO 

VALVOLJA 

0)1 C01l II

3 '12

32 3

j - 322.. 

ORI') 

s~ 346 X 556 X 

518/ 358/ 568/ (;1NIPAl)OR VAPOR )

'I -''a373 -~ 

A GF1N1MAIDuR VAPOR< 2

A (4 NHPAOR

.1 ,,..!

VAPOR 3 

- 3.55- 1

FIGURE 6: MAIN FEED WATE R AND AUXILIARY FEEDWATER SYSTEMS NODALIZATION

324 371 

A' ~ 354 f--j [ 

3525 376 

326 3 7_ 

L f, I1 i.

2
829

815

823

LPI

I--.-...

______ _L 4 - L AZO 3 m

-LAZO I

RA/IAAS FRIAS

FIGURE 7: EMERGENCY CORE COOLING SYSTEMS NODALIZATION

TANQUE DE 

RECARGA

H-IPI
PRZ. "IFS7~ (Kol.4 OOOL.O 3.005,OI 7.0OOOO 01l 1.0500E.02

01 7.5000Z.01 1 .001P.+02 1.25009+02

r0 IA *~I I 

0

1. 7500E.02

YflSSXN POWZN (h
LOOP I HO. TKU. (C) 

02.5000E.02 2.7000C.02 2.900003.02 3.10002+02 3.30009.02 3.50009.02

4, 

N

N 

04

U HH - N 

- i 'Ii ' 

I 

i

I 

I

'a

N

Pazl. LCVEL M'4 

D0003,01 6.0000C.01 8.00003,01 1.00003,02

I

V 

3 

*

I 

I I

'0

9 

0 

'N 

p
E.0? 3 HL TZ"~. (C) LO 2 HL TD(. (C) ~25OJUO~2.00U,2 2.9000U.OJ 3 .SOZ2 10OOU0 2.OS~,2.5000E.02 2.1000Z.02 OO~O 3OOZg,02 .. 3.3000E.02 3.5000E.02 

H 45 N 

N 31 

.. .. . ..
LOOP 2 CL TW (C) 

2.*5050501C0 2.7000Z.02 2.90005,02 3.10005.02 3.3000R.02 3.5000%5,02

LOOP I CL TE" (C) 

2.5000M+0323.70009.03 2.9000Z+02 3.1000E.03 3.3000t.02 3.5000.05,0

0 _ 

-_ .-- - __ 

______ I _____ ______ ________ ________ I ' 0 I ___ - ____ _____ _____ Off 

'9 It 0 b0 

_____ N 0 

I' 

ii Cii 

-J .4 

N N 

N N 

ii Ii 

I.. I

N N 

Ii l-. 9 lb ii lb 0 ______ ______ ______ 0 (3 

N - N 

U g 

I I
LOOP I AVG. TIMP. (C) 

2.5000-,02 2.?000E.02 2.9000E.02 3. 1000Z.02 3.30009.02 3.5000E,02 

~- - --------------

I •. I4' ,. I

n9

LOOP 3 CL TIMP. (C) 

2 5000e,02 2.7000-,02 2.9000-.02 3.1000r.02 3.3000E.02 3.5000E,02 

C, o

9 

CC 

0 

15 

p
LOOP I AVO. TilW. (C) 

:12.5000K.02 2.7000E.02 2.9000t.02 3.10001.02 3.3000M.02 3.5000X.02

H

I0

(I 

U

I *1. p 

4 La 

I .1 C 

tf I !

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9 

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V IT

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RCS -,pi .2,3 d.1t,-t=P* & * (C) 

-- ~1,.-5 3 3 

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Tecnatom s. a

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- ct.lr-40 4

1 70400 

.0 000E.00 1.5000E+02 3.00008.02 4.50008,02 6.0000M.02 7.50008.02 9.0000-002 1.00008.03 1..2000E.03 

PX.AC0OR TIM8(E C)3C 

50
2 

S

.0000t.00 1.5000Z.02 3.00002.02 4.300002.02 6.00002+02 

R2ACTOR TIM (3IC)

7.5000C.02 9.00002,02 1.03000.03 1.2000Z+03

t~VS XNDV2RTZ13 104qU02. CLOSMZ (26-6-09) 

XtCs loop 3 note Obo rat. m'. m=L0IPL9N?) 

-~ -E -

.0000C.00 2.3000Z.02 3.0000C.02 4.0000Z.02 6.0:0002.02 7.50002.02 9.00002+02 1.05000203 1.20002,03 

.2C2R ?Dm (S=C

51

Tecnatom s. a. (j

VAIM443LO-NPP 

yiu'= 23

2' - & 

Tecnatom s.a.I

W.M3LLOS-U00 

Tig=- 24

E3 a_ a a a __

ICOIVd I0O.V'.'TMf W.JOL CLO=0YZ (26-6-89) 

RCS loop 2 =So flom rate (%.) (221JJ/?LUMn 

outrva-4042 70420
00E.00 1.5000E.02 3.OOOOE.02 4.5000E.02 6.0000E.02 7.500GE.02 9.OOOOZ.02 1.0500E.03 1.300

52

vxýzlzos-"p 

riý.. 25

mrv. :ýTý X&210" CýSý (26-4-89) 

SO-L 

- ý1- 4osl POAaO

.00

RTACTOR T=ý (S=)

OL-03

VAMZLLOS-UPT 

riý. 26

"rVE ýT=T IOUFaAL CLOSý (26-9-83) 

SC-2 P-.- (xq/..2) (R=-%P/FLAN-) 

-týr- -4082 F042ý

.0000R.00 1.500OZ.02 3.0000Z.02 4.5000E.02 6-00002.02 7.5000E.02 9.OOOOX-02 1.050OX-03 1.200OZ-03 

ýR Tl= (SEC)

Tecnatom s.aTeenato7n s.m Oct
Tecnatoin s.m 

v7jwz=s-NPP 

xsxv. xwv7mr-2rr wmAL cwmx (26-6-agi 

?ig-. 27 

30-3 I)r*.s=* (II/Cm2) (PMM/PLRXT) 

S s 

c.t.1- 4003 P0440 

s R 

.00009.00 1.5000E.02 3.0000X.02 4.SOOOZ+02 6.0000X+02 ?.SOOOZ+02 9.OOOOZ.02 1.0500B.03 1.20002.03 

RT-%=R TnO (SICC) 

Teenatom s.a. 

V72nMCLTAS-M 

xsrv. Xiou7vzstrzn 102M)LL CLOST= (26-6-69) 

rig- 2B 

so-I Im Le"I M (RZLAP/PLMT) 

mtzlv=-257 L0400 

Sot 

$3 

.000OZ-00 1.5000E+02 3.00008-02 4.30DOZ+02 6.0000Z.02 7.500OZ-02 9.0000Z.02 1.0500E.03 1.20002.03 

Rr-%CTOR TIM (3=)

53
Tecnatom s.m MCI 

v7LNDELLos-up: 

Fig- 30

KSrV. !ZMDVERTZW NaNVAL CL0907X (26-6-89) 

30-3 M (RXIJý/VLkUT) 

-= 210 L0440

ýlk Tý (SW)

.0000r.00 1.50009.02 3.0000r,02 4.5000E.02 6-0000E.02 

RZXCýR TZ= (SLC)

54

Tecnatom s.a. IT 

vx=ýS-Kpv 

rlý- 29

KSý.. l2aZýT=" CLOSý: (26-6-89) 

SQ-2 T-ol 

-trIv-234 L0420

7.30001ý.02 9.OOOOE.03 I.OSOCZ.03 1.20003-03
Tecnatom s.m T 

rig=. 31

i

55

MSXV- ngMVZRTIM MANCAL CLIOSUAZ (26-6-89) 

S-G- I At- flý CTA/h) (MLAP/PLAW) 

0 0 

-trl--4071 P0405

I

REAMM TXM (SEC)

Tecnatom sm I 

VANDEZ"s-a" 

rig- 32

KgrVs XMJNMTZNT NUMAL CZOOSU= (26-6-891 

B.G. 2 st.m glý (Ta/h) (RZL"/PrAxm) 

6 6 =trl- 4072 "429

.0000E.00 Z.5000E.02 3.OOOOZ.02 4.5000Z.02 6.0000C.02 7.500OZ.02 9.000OZ.02 1.050OZ.03 1.2000E+03 

SZACTOX TZM (9=)
4.50009.02 G-OOOOZ.02 

ýR Tl= (9=)

Tecnatom s.a. T,

Tecnatom s.a. I

I

77,

56

vxlrDjm.Loa-w7v 

riq-. 33

xqýs 12vLDV=Tmrr )GUMAL Cý81= (26-6-89) 

9.0. 3 st- flý (ýAfh) (PXLANIPLANT) 

a 

-"1- 4073 T0445

x

PJMCTOR Tna (SZC)

VANDXLLOS-Npl 

rig- 34

xSIT9 1ý MA2;MLL cLosultz (26-6-89) 

8-0. 1 MrW fl- (TUAL) (RZLKP/Pl.&=) 

e 0 r'0403

00M.00 1.5000M.02 3.0000M.02 7.5000t.02 9ý.OOOOX.02 1.0500R.01 1,2000Z+03
K5rva nomv=TMT 53N0WI. CZ.070R2 (26-4-99) 

B.0. 2 KM 11-O (TU/b) CXZXJ.7/VLAM5) 

~001.x-052 7042i3

.4-

U 

- - flAfli tflOt

.0000E*00 1.30003+02 3.00003*02 4.5000M+02 0.O000302+0 

flZW"O 0XM (SMC)

WZVC ZZMVVSRT7 306MCM cLoSog= 426-6-09) 

S.0. 3 W" fla (IT&h5 (M3.OpmLAn) 

=oa~O4033 70443

p. 

- - OSOOKOS 120 

.00003*00 1.50003*02 3.0000Z*0

03,03

53L08-UO

Tigoo 34

03*03

7.ZACSR TD (S=C)

57

Tecnatom s.a..T

Pig-0 35

A

0 

ho

Tecnatom &a. I

1.50009+02 3.0000U.02

7.3000 02 9.000oz. 2 1 0500Z.03 1. 200CE.03

.0000S.00
I Teenatom s.a- act

T

58

Tecnatom s.a, I 

VX==.Lo8-wv7 

Fiý. 37

XSIVA MMLDVýTXNr WJrUJLL CLOSURZ (26-6-89) 

"W 

E3 cntrl-4061 -ý1-4062 -11.

ýi..

VXNDZLLOS-9r, 

Fiý 38

Xsýs IMU77ZRT=T IMNUAL CUMMý (26-6-09) 

P-.-L.- .. Ii.f -1-. - fl- -t. (Xql.) 

6 El .fl-J-42600000 fl..J-42900000

4.50001.02 6-0000r.02 7-50001ý.02 9.0000M.02 1.0500t.03 

ti..

1.2000Z.03

!j,
Marva fmvlD.DV~?wr OOJWI CLOSU3I (2-6-4s0m) 

s flo.*-43000000 .flowl-43200000 aflol*-43400000 

.00002,00 1.50003.02 3.00003.02 4.50003.02 6.00002.02 7.5000Z.02 9.0000Z.02 3..05003,03 1.20003.02 

Tee at m ..

Teenaatom sa. 

IWOY. nMYU3UIT 201UthL CLOSW3Z (26-6-89) 

Vigor 40 3.G.1 relief and safety ýaa diaohargod mass (2qix) 

* 4tr1..r4210 -trlwýa4211 oorrl..-421.2 r=trl..-4213 ootrl,.-4214 cotr.1--4Z0.5

.00009.00 1.3000+02,0 3.00003.03 4.50003.02 G.00003Z.02 7.50002,02 9.00003.02 -1.0500X+03 1.2000Z.03 

ti...

59

A 

e e G-
trrll--T-1

P- ;.Tt.. Vý', 

-S u! 0 " v tj o a

09 

.. T!4 I

E0.200or'l E0.100so-T ro.zoooo,6 r0+2000s, L to-XGOOO'9 rcý+X0000*t ro.20ODS-T 00-20 1','

-- T- Aa.". 17,_ 7.TT.. C-0-6 

-D-S 2uojVuaaL 

EO-200or-T co-zooso .1 to-20000,6 zo;ooog.L to-20000.9 to-30009, to-Roooo-C to-ZOODS'T 00-20000' 

NO
XSZV. nl.VUrTZN1 IDRUMX. CLOIt (26-4-89) 

44-0 =tz1..-4233

.00003,00 1.00003,02 3.0000Z.02 4.500003.02 6.00003.02 7.3000Z.02 9.0000Z.02 1.0500Z03*0 21.2000Z+03 

tio.

Tecna torn &a.I

MASZY. XXKVV20UIT 393853. CWSOWZ (26-6-19)

* vmtim.-Q catn12 -4235 

.0000M.00 1.5000M.02 3.0000M.02 4.3000Z+02 6.00003.02 

tin.

7.50003*02 9.00003402 1.05003*03 1.2000K.93

61

Tecnatorn s.a.T

71g.~. 43

II

6'

ULIX&OS-u0 

liqux 4"

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

(2-89) (Assigned by NRC, Add Vol., Supp., Rev., NRCM 1102. BIBLIOGRAPHIC DATA SHEET and Addendum Numbers, if any.) 

3201.,3202 BBIGAHCDT HE 

(See instructions on the reverse) NUREGIIA-0197 

2. TITLE AND SUBTITLE 

Assessment of RELAP5/MOD3.2-NPA3.4 Against an Inadvertent Closure 

of all Three MSIV's in VANDELLOS-Il Nuclear Power Plant 3. DATE REPORT PUBLISHED 

MONTH YEAR 

January 2001 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

C. Llopis, A. Tanarro, R.M. Fanegas Technical 

7. PERIOD COVERED (inclusive Dates) 

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

provide name and mailing address.) 

TECNATOM 

Avda. Montes de Oca, I, 

San Sebastian de los Reyes 

28700-Madrid, SPAIN

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above'; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

and mailing address.)

Division of Systems Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less) 

The work consists in reproducing an actual inadvertent manual closure of all three main steam isolation valves transient that 

occurred at Vandellos II Nuclear Power Plant (Tarragona, Spain), by means of the RELAP5/MOD3.2 code, thus becoming part of an assessment type exercise. The calculations have bee performed by the plant technical support staff and TECNATOM 

S.A. in a joint effort. Vandellos NPP input model for RELAP5/MOD3.2, has been developed, starting from the input model for 

the former versions RELAP5IMOD2 and MOD2.5, by the licensee technical services department. The model was provided to 

TECNATOM S.A., along with the plant data, transient available records and associated information in order to perform the* 

transient calculation and sensitivity studies. Finally, the results and conclusions were discussed and presented jointly by both 

organizations. The present report is mainly focused on the steady state and transient calculations and the findings and 

conclusions derived from them. The general purpose model provided by the licensee, along with some little modifications 

implemented for the specific transient calculation, are described. A thorough description of the model build-up process is not 

included.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

MSIV Closure unlimited 

RELAP5/MOD3.2 14. SECURITY CLASSIFICATION 

(This Page) 

unclassified 

(This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89)
Federal Recycling Program
tN UI(IIt/lA-019i/ OLI ALL TREEMSIV'S INVANI)ELO-'II NLAR POWRN I I CPLANT)fU 

OF ALL THREE MSIV'S IN VANI)ELLOS-II NUCLEAR POWER PLANT

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300

JAP%,•U2-ýl L OWx