NUREG/IA-0177

International 

_ •Agreement Report 

Assessment of a 

Reactor Coolant Pump Trip for 

TRILLO NPP with 

RELAP5/MOD3.2 

Prepared by 

M. E Lozano, R Moreno, PMSA 

A. L. Cedillo, M. N. Sanjurjo, C. N. TRILLO 

PMSA 

Sector Embarcaciones (Bloque 40), 17, I'D 

28760 Tres Cantos - Madrid 

SPAIN 

C.N. TRILLO 

Plaza Carlos Trias Bertran, 7, planta 70 

28020 Madrid 

SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

March 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-0177

International 

Agreement Report 

Assessment of a 

Reactor Coolant Pump Trip for 

TRILLO NPP with 

RELAP5/MOD3.2 

Prepared by 

M. E Lozano, P Moreno, PMSA 

A. L. Cedillo, M. N. Sanjurjo, C. N. TRILLO 

PMSA 

Sector Embarcaciones (Bloque 40), 17, 1

0

D 

28760 Tres Cantos - Madrid 

SPAIN 

C.N. TRILLO 

Plaza Carlos Trias Bertran, 7, planta 70 

28020 Madrid 

SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

March 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
ABSTRACT 

This document presents the comparison between the simulation results and the plant 

measurements during a Reactor Coolant Pump Trip in C. Trillo Nuclear Power Plant 

in different burnup conditions. 

This work is part of the Spanish contribution to the Code Assessment and 

Maintenance Program (CAMP) as a member of UNIDAD ELECTRICA, S.A. 

(UNESA). 

C. Trillo is a 3010 MWth three loop PWR Nuclear Power Plant designed by Siemens

KWU. Commercial operation started in August 1988, currently the Plant is running 

the 1 lb annual cycle. 

The simulation has been carried out with the RELAP5/MOD3.2 code, running on a 

HP 715/50 work-station. 

The transient main parameters have been correctly assessed, however predictions at 

the beginning of transient are more difficult as the reactor remains at a high power 

level after pump trip.

i
EXECUTIVE SUMMARY 

C. Trillo is a 3010 MWth three loop PWR Nuclear Power Plant of Siemens-KWU 

design located on the Alcarria region (Central Spain) on the Tajo river. 

The Thermal-hydraulic Analysis Group of C. Trillo has prepared a model of the plant 

using RELAP5/MOD3.2. This model, that has been checked by a wide process of 

model validation against start-up test data and transients includes the following 

characteristics: 

"* Three loops 

"* Detailed nodalization for Reactor Pressure Vessel 

"* Reactor point kinetic model with actual data of C. Trillo core for both BOC and 

EOC conditions (Two inputs set) 

"* Auxiliary systems simulation. 

"* Control, Limitation and Protection System. 

"* Secondary side from Feedwater Tank to Turbine Inlet including Bypass, Isolation, 

Relief and Safety Valves. 

The areas of application of Relap5 model for C. Trillo are the following: 

"* Analysis of transients occurred during operation 

"* Analysis of design modifications before their plant implementation 

"* Answer to specifics questions of different C. Trillo departments 

"* PSA project, including large, medium and small leaks, transients and Anticipated 

Transients Without Scram 

"* Answer to CSN (Spanish Nuclear Safety Council) questions 

"* Operators training: calculations of operation manual instructions. 

The transients calculated for this assessment are two Reactor Coolant Pump Trip at 

different core burnup conditions. Such transients, according to plant design did not 

result in reactor trip, instead power operation (at part load level) is continued due to 

the actuation of the limitation system that cut-back the permitted reactor and generator 

power to values compatible with two loop operation. 

The main conclusion was that all significant plant parameters have been correctly 

assessed, although predictions at the beginning of transients are more difficult as the 

reactor does not trip to zero power.

ii
TABLE OF CONTENTS 

Abstract ............. ............................................. 1i 

Executive Summary ............................................................. ii 

Table of Contents ............................................................... iii 

List of Tables ............................................................... iv 

List of Figures ..................................................................... v 

I. Introduction .............................................................. 1 

II. Plant Description .................................................. 2 

III. Model Description .................................................. 5 

III.a Primary System ............................................ 5 

III.b Secondary System .......................................... 8 

III.c Control, Limitation and Protection System ...... 9 

IV. Transient Description ............................................ 12 

V. Analysis Results .................................................... 13 

V.a Reactor Coolant Pump at BOC ....................... 14 

V.b Reactor Coolant Pump at EOC ....................... 15 

VI. Conclusions ............................................................ 15 

VII. Run statistics .......................................................... 16 

VIII. References ............................................................ 17

iii
LIST OF TABLES 

Table 1.-Technical Data of the Reactor Coolant and Pressurizer System. 

Table 2.- Variables identification in transient figures. 

Table 3.- Steady State Parameters. 

Table 4.- Run Statistics.

iv
LIST OF FIGURES (1/2) 

1.- C. Trillo Nodalization For Relap5/Mod3.2. 

2.- C. Trillo Nodalization (Secondary Side) 

3.- C. Trillo Nodalization (Feedwater System) 

4.- Primary Average Temperature. Plant Data (YTOOT101) 

5.- Secondary Pressure. Plant Data (SF1OC1 10) 

6.- Primary Pressure. Plant Data (YTOOP5 11) 

7.- Hot Leg Temperature, Loopl. Plant Data (YA10TO03) 

8.- Hot Leg Temperature, Loop 3. Plant Data (YA30TO03) 

9.- Cold Leg Temperature, Loop 1. Plant Data (YA10T955) 

10.- Cold Leg Temperature, Loop 3. Plant Data (YA30T955) 

Reactor Coolant Pump Trip at 130C 

1.1.- Reactor Coolant Pump Speed 

1.2.- Primary And Secondary Pressure 

1.3.- Reactor And Generator Power 

1.4.- Loop 1 Temperatures 

1.5.- Loop 3 Temperatures 

1.6.- Primary Average Temperature 

1.7.- Coolant Mass Flow Rate 

1.8.- Pressurizer Level 

1.9.- Steam Generator Mass Flow Rate ( Loop 1,2) 

1.10.- Steam Generator Mass Flow Rate (Loop 3) 

1.11.- Feedwater Mass Flow Rate 

1.12.- Steam Generator Level 

Reactor Coolant Pump Trip at EOC 

2.1.- Reactor Coolant Pump Speed 

2.2.- Primary And Secondary Pressure 

2.3.- Reactor And Generator Power 

2.4.- Loop 1 Temperatures 

2.5.- Loop 3 Temperatures 

2.6.- Primary Average Temperature 

2.7.-Coolant Mass Flow Rate

V
LIST OF FIGURES (2/2)

2.8.- Pressurizer Level 

2.9.- Steam Generator Mass Flow Rate ( Loop 1,2) 

2.10.- Steam Generator Mass Flow Rate (Loop 3) 

2.11.- Feedwater Mass Flow Rate 

2.12.- Steam Generator Level

vi
LIST OF ABREVIATIONS

ATWS Anticipated Transient Without Scram 

BOC Beginning of Cycle 

CSN Consejo de Seguridad Nuclear 

EOC End of Cycle 

KWU Kraftwerk Union 

PSA Probabilistic Safety Assessment 

PWR Pressurized Water Reactor 

RCL Reactor Coolant Loop 

RCP Reactor Coolant Pump 

RCS Reactor Coolant System 

RPV Reactor Pressure Vessel 

SG Steam Generator

vii
I. Introduction

The objective of present work is to evaluate the predictability of the Relap5/mod3.2 

/1/ on major thermal-hydraulic behavior during a Reactor Coolant Pump Trip in C. 

Trillo. Such a transient behavior includes the following phenomena: asymmetric loop 

behavior, control and limitation system actuation, short-term reactor power behavior 

owing to different moderator temperature coefficient, etc. To do this the calculation 

results are assessed and compared with the start-up test data D-100-303 /2/ (BOC 

conditions, first cycle) and with the RCP trip transient occurred on July 18, 1990 /3/ 

(near EOC conditions, 2nd cycle). 

According to KWU design the Upset Condition "Failure of one main coolant pump" 

does not result in reactor trip, but power operation (at part load level) is continued 

instead. 

Comparing the calculations with real data, overall transient behavior predicted by 

using Relap5/mod3.2 was in a general good agreement with plant data, and the 

following conclusions were obtained: 

"* Relap5/mod3.2 predicted well the sequence of events and the major phenomena 

during the transient. 

"* There is increased difficulty on plant assessment calculation when reactor scram 

does not occur and plant remains at a high power level. 

"* Short-term behavior depends on core bum-up conditions, as reactivity feedback 

depends on core burn-up too.

1
II.-Plant Description

C. Trillo is a Three Loop PWR Nuclear Power Plant designed by KWU-Siemens. 

Commercial operation started in August 1988 and Plant Dynamic Start-Up Test took 

place in the spring and summer 1988. 

C. Trillo has a nominal reactor power of 3010 MWth at an operating pressure of 158 

bar and an average coolant temperature of 310 'C. The reactor core consists of 177 

fuel assemblies, each containing 236 fuel rods in a 16x16 square array. All major 

plant components are similar to those of other PWR plants. The values of main 

parameters are listed in table 1. 

The reactor pressure vessel is the fixed point of the reactor coolant loops whilst the 

reactor coolant pumps and steam generators are flexibly mounted. The reactor 

pressure vessel internals are basically composed of the lower core structure with flow 

skirt for flow distribution, and the upper core structure containing the control rod 

guide assemblies. 

The three steam generators are designed as vertical U-tube bundle heat exchangers 

with preheaters. 

The reactor coolant pumps are single-stage centrifugal pumps with overhung impeller 

and vertical shaft. They are driven by an asynchronous motor through a gear 

coupling. A flywheel prolongs pump coastdown time. 

The pressurizer consists of a cylindrical shell, an upper and a lower hemispherical 

head. Its lower section is filled with water, the upper part with steam. The steam 

volume can be increased or reduced by heating the water or by spraying water into the 

steam cushion as necessary to regulate the operating pressure. The heating elements 

are flanged into the lower head. The nozzles for the spray lines are located in the 

upper head together with piping connections for the safety valves, the relief valve, and 

the connecting line to the RPV vent system. 

A too high pressure rise in the reactor coolant system is prevented by the two pilot 

operated safety valves which discharge into the pressurizer safety tank. A relief valve, 

electrically controlled, is provided in addition to the safety valves as a means of 

limiting pressure increase. Its actuating pressure is lower than that of the safety 

valves. Safety valves pressure is only reached in case of a highly improbable transient

2
(e.g. ATWS). 

The reactor coolant system is supported by the auxiliary systems. A number of the 

nuclear auxiliary systems are directly involved in reactor operation for coolant 

injection and extraction, clean-up, degasification, adjustment of coolant boric acid 

concentration for reactivity control and admixture of chemicals to the coolant. 

The emergency core cooling and residual heat removal system is of a three plus one 

train design connected to both cold and hot RCS legs. Each subsystem consists of: 

- One safety injection pump that takes borated water from the storage tanks and 

deliver it via a logic circuit to the corresponding hot leg of primary side. A rupture in 

the hot side causes the actuation of logic circuit to feed via the intact cold side line. 

- Two accumulators connected each to the hot ad cold side injection lines. 

- One residual heat removal pump that sucks water parallel to the safety injection 

pump from the water storage tanks or, after these are empties, from the containment 

sump. The injection into the primary in also through the hot and cold side lines. 

- Two borated water storage tanks. 

In addition to those, there is a fourth train with one safety injection pump and one 

residual heat removal pump with the corresponding water storage tanks, that works as 

a spare system for any of the other three subsystems. 

The steam generator is a standing U-tube bundle heat exchanger. The heating medium 

is the reactor coolant which enters the hot inlet chamber, passes through the U tubes 

and transfers the heat generated to the feedwater/steam cycle. 

Three 50% capacity feedwater pump units, one as standby, deliver the feedwater 

through the high pressure heater system to the steam generators. The feedwater inlet 

temperature to the SGs is 220 °C at full load. 

Each steam generator is assigned a separate feed station comprising low and full-load 

control valves. Each one of these main feedwater lines is divided on reaching each 

steam generator into three lines of 40%, 50% and 10%. The 10% line leads to a ring 

of sprinklers in the steam zone of the steam generator. The 40 % and 50 % lines lead to 

the economizer where feedwater is heated to just below boiling temperature, after 

leaving the economizer it mixes with the circulating water. Approximately 50% of the 

feedwater flows upwards through the preheater in a direction opposite to the primary 

coolant, and 40% flows downward. This so-called "split flow" design protects the

3
tube sheet against cold feedwater shocks and, on the other hand, grants a high thermal 

efficiency of the steam generators. The steam generation occurs under natural 

circulation. For this purpose the tube bundle is surrounded by a guide shell extending 

above tube bundle and connecting to moisture separators. At full power, the secondary 

main steam pressure is 68.5 bar. The water level in the steam generators is kept 

constant by the feedwater control valve station. 

The steam and feedwater system are designed such that the residual heat transported to 

the steam generators can be transferred to the circulating water in the condenser via 

the main steam bypass station after completed shutdown of the turbine generator unit. 

The main steam bypass station is capable of dumping approximately 60 percent of the 

full load steam flow to the turbine condenser in a controlled manner so that on turbine 

trip the reactor does not trip as well; in this case only controlled reactor power 

reduction is necessary. When main steam bypass station is not available, the main 

steam pressure rises are limited by main steam relief and safety valves. 

The start-up/shutdown feedwater system comprises two electrically driven pumps 

connected to the feedwater tank which supply the steam generators via the low load 

control valves during start-up and shutdown operation. Those pumps start up 

automatically if all the main feedwater pumps fail and, being connected to the 

emergency diesel, are also capable of shutting down the plant on loss of auxiliary 

power without resort to the emergency feedwater pumps. 

The reactor control systems are designed in such a way that the reactor process 

operates at its optimum operating point. The reactor control system actuates on control 

rods and boron/water injection. 

The core protection system initiates actions to prevent accident conditions when 

deviations occur which the process controls can not correct. The trip system monitors 

the process variables which are significant to the safety of the reactor plant and the 

environment for the purpose of prompt detection of accident conditions, measurement 

of data processing and automatic initiation of protective actions to keep plant 

conditions within safe limits. If accident conditions occur, measures to mitigate their 

consequences are initiated automatically. 

A significant feature of Siemens-KWU PWR plants is the so called Limitation System, 

which plays an intermediate role between Control and Protection Systems.

4
The limitation system has the following tasks: 

"* Holding of reactor power below set limits by limiting integral reactor power. 

"• Prevention of unacceptable local loading on fuel or fuel cladding by appropriate 

control of power distribution and limitation or reduction of reactor total power. 

"* Assurance of an adequate shutdown reactivity margin by limitation of control rod 

bank movement. 

"* Supervision of performance of control rod dropping after actuation of reactor 

scram. 

", Prevention of draining or overfilling the pressurizer by limitation of coolant 

inventory and maintenance of coolant pressure within set limits actuating heaters 

and spray valves. 

The limit settings and the countermeasures associated with these are selected on the 

basis of the related safety requirements such that conditions are no more severe than 

the initial conditions assumed in the safety analyses and such that process upset 

variables are returned within their normal operating values. 

IH.- Model Description. 

The C. Trillo model is shown in figures 1, 2 and 3. It consists of 303 volumes, 351 

junctions, 163 heat structures and 944 control variables. The three loops are simulated 

with actual data specific for each loop in order to have the possibility for simulating 

non symmetric transients and to understand the different loop behavior. 

Mll.a.- Primary System. 

The model of the primary system includes all the main components of the system. In 

the reactor vessel elements the volumes corresponding to the downcomer, lower and 

upper plenum, the core, core bypass and the upper head are defined. In the 

nodalization of the upper part of downcomer (volumes 510 and 540), special care was 

taken to correctly simulate the steady-state mass flow rates in the connection with the 

hot legs (bypass) as well as with the upper plenum and upper head (unintentional leak 

paths), so that to obtain the requested values. The downcomer is represented by the

5
components 530 to 538. The lower plenum consists of the nodes 560, 570 and 571, 

while the core bypass is represented by node 578. 

The core was subdivided to two parallel channels that are interconnected by junctions 

to allow cross flow exchange. The external channel (15% of the core volume) is 

represented by nodes 581 to 588, while the larger channel (85%) consists of nodes 

591 to 598. The power and fuel are distributed in the same proportion. The exit of the 

core is represented by nodes 500 and 501 and upper plenum by volumes 511 and 512. 

Above the reactor coolant loops, the upper plenum consists of nodes 513 and 520, the 

vessel head of nodes 548, 549 plus the pipe 550. 

The fuel rods have been simulated as a cylindrical heat structure with an internal heat 

source whose power derives from point kinetics calculation. The total reactor power is 

calculated as the sum of prompt fission power plus decay power from fission 

fragments. 

The built-in data for fission products and actinides have been used, but data for 

delayed neutrons have been entered from specific kinetics calculation. 

The reactivity feedback model used assumes separability of feedback effects. Two 

tables, one defining reactivity as a function of coolant density, the other as a function 

of volumetric average fuel temperature, have been supplied using Trillo-specific 

(BOC and EOC) data. 

For the first case, simulating the start-up test, data from KWU input of transient 

analysis corresponding to the first core has been used /5/. 

KWU supplies, each reload, nuclear results for the final pattern calculated by the code 

MEDIUM. The data of end of second cycle (89-90) has been used for the calculation 

of case 2. 

Control variables simulating the control rods reactivity contribute to reactivity 

feedback calculations. 

The control rods reactivity is calculated by a simulation of control rods movements 

caused by temperature or power deviations, assuming that reactivity is a function of 

control rod position. 

The three loops are analogously simulated. The hot legs are modeled by volumes 100, 

200 and 300. The steam generator inlet chambers are simulated by nodes 110, 210 and 

310. The tubes volumes (120, 220 and 320) are divided into 8 cells each. Proper heat

6
structures are used to link the primary to the secondary side of steam generator where 

real data are used. 

Steam generator outlet is modeled with volumes 130, 230 and 330. Nodes 140, 240 

and 340 simulate the intermediate leg between seam generator and reactor coolant 

pump. 

The reactor coolant pumps are modeled by volumes 150, 250 and 350 by using 

specific homologous curves supplied by the vendor. Finally, the cold legs are modeled 

by volumes 160, 260 and 360. 

The pressurizer has been modeled by means of volume 410 divided into 3 cells, the 

upper head is the node 415 and bottom zone is represented by volume 402. Safety and 

relief valves are connected to the steam filled space (volume 415) and discharge into 

the relief tank dome (468). Volume 450 models the spray line, while the mass flow is 

controlled by the valve 451. The pressurizer heaters are modeled by means of a heat 

structure governed by control system. 

The following auxiliary systems are connected to the primary system: 

"* Volume Control System (TA). The volume control system is connected via three 

charging lines each to the cold leg. The extraction point is located between the 

steam generator and the reactor coolant pump. In addition, the volume control 

system is connected to a pressurizer spray line for the purpose of auxiliary spraying 

if necessary. 

"* Residual Heat Removal System (TH). The TH system is divided into four parallel 

subsystems. Three of them are directly assigned to the reactor coolant piping loops 

and connected via a hot and cold injection lines. Each of the three subsystem 

consists of 1 safety injection pump, 2 accumulators, 1 residual heat removal pump 

and 2 borated water storage tanks. In addition there are one safety injection pump 

and one residual heat removal pump with the corresponding water storage tanks, as 

a spare system for any of the other three. 

Extra borating system (TW). The TW system is connected to the reactor coolant 

piping with lines from the storage tanks of TH system and own borating tanks.

7
HI.b.- Secondary side

The three steam generator secondary sides are simulated identically, so only loop 1 is 

briefly described in the following: 

The downcomer (volume 600), modeled as an ANNULUS component, divided into 3 

cells is connected to volume 640 and to the bottom tubes volumes (602 and 604). The 

steam generator preheater is simulated by volumes 604 and 606. In its turn volume 

604 is connected to the 40% feedwater junction and located below volume 606, which 

is itself connected to the 50 % feedwater valve. 

The tube area is modeled by means of volumes 602 and 610. The first one is divided 

into two nodes and connects the tubes plate to the top of the preheater. It is thermally 

connected to the primary side by heat structures. Volume 610 covers from the top of 

volumes 606 and 602 to the steam separator inlet. 

Volume 620 models both steam separator and steam dryers. It is defined as a separator 

component and is connected to the riser, steam generator dome and downcomer. 

Volumes 635 and 636 simulate the area outside the separators and dryers. 

Volumes 660 and 670 account for the steam line lying from the outlet of steam 

generator to the inlet of the steam header (volume 910). A line containing the turbine 

stop and control valves connects volume 910 and 940, which represents the turbine. 

There is also a line that connects volume 910 to a time dependent volume which 

represents the condenser (920) through the bypass valve 915 (SF system). 

Each steam generator is provided with its own main steam , feedwater, emergency 

feed and blowdown lines (figure 2). The following secondary systems are associated 

with the steam generator: 

"* Main Steam System (RA). The steam generated is passed through three parallel 

main steam pipes (660, 760, 860) to the main steam header and from there to the 

turbine valves (912). The isolation (661, 761 and 861), relief (662, 762 and 862) 

and safety (663, 763 and 863) valves are simulated with their corresponding control 

system. 

"* Main Feedwater System (RL). The main feedwater pumps (32, 33, 56) deliver the 

feedwater from the tank to the steam generators via headers through the high 

pressure heaters. Each feedwater line is provided with a full load and low load

8
control valve. The required control, check and isolation valves are simulated 

(Figure 3). 

"* Startup and Shutdown System (RR). The startup and shutdown pumps supply the 

steam generator via the low load valves of feedwater system (62, 72, 82) bypassing 

the high pressure heaters. 

"* Emergency Feedwater System (RS). This system supplies the steam generators in 

the event of malfunctions accompanied by failure of the main feedwater and startup 

and shutdown system. 

"* Steam Generator Blowdown (RZ). This system is used to maintain a certain water 

quality in each steam generator and to detect heating tube leakage as well. 

II.c.- Control, Limitation and Protection Systems 

The more important systems regarding the dynamic behavior of the plant have been 

simulated. A brief description of each of them follows: (for a detailed description see 

Ref /4/) 

* Control System 

e Average Coolant Temperature Control (KMT) . In steady-state operation the 

setpoint for average coolant temperature traces the appropriate characteristic 

on the part-load diagram. The average coolant temperature is measured in 

all reactor control loops and processed by a controller. The control 

deviation is formed by comparing with the setpoint and regulates control 

rods movements. 

e Secondary Maximum Pressure (PSOMAX). Its mission is to limit pressure 

increases in secondary side. When pressure gets larger than setpoint (not 

constant but function of plant conditions) the condenser bypass opens 

keeping pressure below a specific value avoiding safety and relief valves 

actuation. 

o Turbine Control. The turbine control valve is governed by the demanded 

power and the secondary minimum pressure setpoint. 

* Feedwater. This system controls feedwater flow as a function of both steam 

flow and steam generator level.

9
"* Coolant Pressure Control. This system maintains the controlled variable, 

coolant pressure, within specified limits of the setpoint by the use of the 

pressurizer heaters and spray valves to control the water/steam volume in 

the pressurizer. 

"* Pressurizer Water Level Control. The task of this system is to maintain 

pressurizer water level at a preset setpoint under all operating conditions by 

using the charging pumps and the HP reducing station. 

Limitation System. 

"* Permitted Reactor Power (PERL). The "permitted reactor power" signal 

sets the level to which reactor power is limited depending on the 

instantaneous plant condition. The following events are taken into account 

explicitly: Fast shutdown of the reactor, steam generator tube leak, 

malfunction of one reactor coolant pump, faults in feedwater supply, power 

density limitation and faulty control rod position measurement. 

"* Permitted Generator Power (PERG). This setpoint is intended for limiting 

generator power and is itself a function of PERL. 

"* Reactor Power Limitation Module (L-RELEB). The momentary reactor 

power level is continuously compared with the permitted reactor power 

level. The difference signal is compared with a series of limit signals with 

step-raised settings that actuates countermeasures of increasing intensity. 

"* Reactor Power Limitation on Reactor Coolant Pump Malfunction (PUMA

RELEB). A few seconds after malfunction of one reactor coolant pump the 

permitted reactor and generator power levels are decreased and control rod 

insertion (STEW-PUMA) is actuated. 

"* Reactor Power Limitation on Basis of Feedwater Mass Flow (SPEISE

RELEB). In the event of malfunction of main feedwater pumps in operation 

at high reactor power levels, permitted reactor and generator power are 

quickly reduced to prevent reactor trip caused by low steam generator level 

or high primary pressure. 

"* Primary- Secondary Overpower (KOL-RELEB). This system prevents any 

increase in power which would otherwise lead to steam dump to the 

atmosphere. Limit signal results in an increase block for permitted reactor

10
power level and in the actuation of the countermeasures derived from this. 

"* Average Coolant Temperature Limitation System (KMT-RELEB). Its 

mission is to limit temperature excursions that the control system can not 

handle. A series of three limit signals with step-raised settings actuates 

countermeasures of increasing intensity where required. 

"* Reactor Power Limitation on High Energy Content in Primary Loop 

(LOOP-RELEB). For comparison with the preset limits for this system the 

pressurizer water level, coolant pressure and positive coolant pressure 

gradient are summated, appropriately weighted. On actuation of LOOP

RELEB control rod movement command are issued. 

Protection System. A list of Reactor Protection System actuation is presented 

bellow: 

"• Reactor Trip (RESA). Reactor trip actuation and turbine trip actuation are 

automatically effected by the reactor protection system on the basis of the 

same criteria. Trip is effected if any of the limit values is fulfilled. 

"* Cut out of the Reactor Coolant Pumps. 

"* Feedwater Delivery Interruption. 

"* Closing of Main Steam Isolation Valve. 

"* Actuation Criteria for Emergency Core Cooling System (injection). 

"• Automatic Cooldown by 100 K/h. 

"* Closing of Isolation Valve upstream of Main Steam Safety Valve 

"* Main Steam Safety and Relief Valves (opening/closing) 

"* Pressurizer Relief and Safety Valves (opening/closing)

11
IV. Transient Description

The event "failure of one main coolant pump" is detected by the decreasing speed of 

the pump motor. As soon as pump speed falls below 94%, the reactor and generator 

power are reduced to about 40 and 30% resp. by the limitation system. The reduction 

of reactor power is performed by means of the so-called STEW-synchronous, i.e. a 

predetermined number of control rod elements fall into the core in order to reduce the 

reactor power to values near 40%. In the long term the reactor power remains at a 

level of about 30%, corresponding to the turbine power. 

The decreasing pump speed causes a corresponding flow reduction that results in flow 

reversal in the affected loop, this also results in a reduction in the core flow rate 

leading to a decrease of pressure loss over the vessel, and as a consequence to a 

higher delivery rate of the running pumps. 

After complete coastdown of the affected pump a new steady state condition is 

established, whereas the elevated flow in the other loops is distributed at vessel inlet 

into a major part for core flow and a minor part streaming backward through the idle 

loop. 

The heat removal from primary side mainly takes place in the steam generators of the 

intact loops whereas the heat transferred in the affected loop is small in accordance 

with the new conditions. 

The main steam flow rates in the intact steam generators follow the trend of the heat 

transfer rates. The main steam bypass station opens only for a short time (20 s) until a 

new equilibrium between reactor and generator power is reached. 

The primary coolant average temperature is reduced in correspondence with a 

temperature setpoint which is automatically adapted to a 2-loop operation condition. 

The control rod injection initiated automatically warrants the continuation of power 

operation in the event of a main coolant pump failure. This automatic power reduction 

proves as sufficiently fast and efficient with respect to both time and magnitude of 

power cut. With exception of the control rod injection due to limitation system 

actuation only control system intervenes during the event and manages the plant to 

attain a new part load steady state condition. The reactor protection system does not 

respond at all.

12
The general transient behavior, described above, is similar both at BOC and EOC 

cases except for the earlier phase before final stabilization. 

The different reactivity coefficients have an influence on the first seconds of the 

transient when the control rods fall into the core. At EOC conditions the higher and 

more negative value of moderator temperature coefficient implies a stronger rise of 

reactor power, and consequently also in other plant variables (pressure, 

temperature...). This behavior is not observed when the transient occurs at BOC 

conditions with a moderator coefficient near zero. 

Figures 4 to 10 present the comparison of different plant variables at BOC and EOC 

conditions. 

V.-Relap5/Mod3.2 Simulation 

The transient has been simulated with the C. Trillo model presented on figure 1. All 

actions, except the initiating event (loop 3 RCP trip) are due to automatic actions of 

control and limitation systems. 

The initial steady-state conditions were obtained from new transient run up to 200 

seconds each. Two steady-state calculations were performed for BOC and EOC 

conditions to simulate the particular conditions of the plant when the pump trips were 

initiated. The main difference between the initial states was the primary average 

temperature, so the calculated steady-states have to be adapted, by the modification of 

temperature setpoint to the real conditions measured in plant. 

The results obtained from the steady-state calculations are compared with the 

measured initial conditions in Table 3. 

Two set of figures are presented. The first corresponds to BOC conditions, with data 

of the start-up test D-100-303 and the second compares the transient simulation at 

EOC conditions with the data of the transient occurred on July, the 18± 1990. 

Comparing the calculations results with the plant data, it was found that the 

Relap5/Mod3.2 code predicts well the RCP trip transient behavior and the timing of 

sequence of events.

13
V.a .- Reactor Coolant Pump Trip at BOC.

Figure 1.1 presents the initiating event, trip and coastdown of loop 3 reactor coolant 

pump. 

The insertion of control rod elements is actuated by the limitation system which 

detects the signal "pump speed < 94 % and reactor power > 65 %". Selected pairs of 

control rod elements drop into the core reducing the reactor power to approximately 

40%, as can be seen in figure 1.3. 

The subsequent behavior of reactor power depends on primary average temperature, 

because control system is active during the whole transient. Figure 1.6 presents the 

comparison between calculated and measured temperatures: general behavior is 

similar, but some minor differences appear when temperature oscillates near to the 

setpoint. In the final part of the transient both temperatures converge to a common 

value defined by the two loop part load diagram. 

The calculated and measured temperatures in loops 1 and 3 are presented in figures 

1.4 and 1.5, an overall good agreement was obtained for these variables. 

The transients of pressurizer water level (figure 1.8) and coolant pressure (figure 1.2) 

reflect the temperature transients, the general behavior is very similar and in the final 

stabilization converge to the same value limited by the control system. 

Figure 1.3 presents values of thermal corrected reactor power, this variable is not 

measured directly in the plant but calculated as a function of thermal reactor power 

(sum of the temperature rises of the three loops) and neutron flux signal. This signal 

approaches the thermal reactor power during steady state and the neutron flux signal 

during fast power transient. 

The Relap calculated thermal corrected power uses the thermal power calculated as 

sum of temperature rises and the fission power calculated by the code. The results do 

not compare well because of plant treatment of data for this case with one loop 

inactive, is different to Relap variables.

14
V.b .- Reactor Coolant Pump Trip at EOC.

The simulation corresponds to a real transient occurred in the plant on July 18, 1990, 

the core conditions coincide to a near EOC state. 

At end of cycle (EOC) the reactor behavior is significantly modified by coolant 

temperature changes since the temperature reactivity coefficient is more negative than 

at beginning of cycle, then the influence of coolant temperature changes on reactor 

power is larger. The more negative moderator temperature coefficient leads to a 

reactor power rise after the first reduction (Figure 2.3), due to average temperature 

decrease (figure 2.6), that reaches the limiting value of permitted reactor power (not 

observed in the previous case) causing a new insertion of control rods. Unfortunately 

plant records for reactor power or control rods movements are not available and it is 

not possible to know exactly the sequence and the precise chronology of events, but 

by looking at the average temperature behavior it can be deduced that a second 

control rod insertion has occurred in plant transient. 

Figure 2.2 presents the evolution of primary and secondary pressure and their 

correspondent Relap calculations that show a coherence with the temperatures 

behavior (figures 2.4, 2.5 and 2.6). The largest difference between measured and 

calculated data occurs after the transition to reversal flow in the affected loop. 

VI.-Conclusions. 

The Relap5/mod3.2 was assessed using C. Trillo real data for two different reactor 

coolant pump trips. An overall good agreement was obtained although some minor 

differences were observed. 

The results demonstrated that the C. Trillo nodalization with the control, limitation 

and protection system could successfully describe the widespread behavior of plant 

transients and the following conclusions were obtained: 

* Relap5/mod3.2 predicted well the transient behavior during the pump trip, 

including variables such as the primary and secondary pressure, secondary mass 

flow rate, and primary temperatures. The calculation results also predicted well the 

sequence of events and the major phenomena during the transient, such as the

15
asymmetric loop behavior, the pump coastdown and the primary to secondary heat 

transfer. 

" C. Trillo model for Relap5/Mod3.2 has shown to be a valuable tool to analyze 

Plant transients and to explain the behavior of main variables as in the 2nd case that 

a new control rod insertion can be deduced by the code results. 

" There is increased difficulty on plant assessment calculation when reactor scram 

does not occur and plant remains at a high power level. 

VII.- Run statistics. 

The calculations were carried out with the RELAP5/MOD3.2 code, running on a HP 

715/50 work-station. 

The run statistics are presented in the table 4.

16
VIII.- References 

/1/ Relap5/Mod3 Code Manual - Code Structure, System Models and Solution 

Methods. NUREG/CR-5535, EGG-25596, Idaho Falls, Idaho, June 1995. 

/2/ Commissioning Result Report D-100-303 

/3/ Informe de Incidente Operativo: Disparo de una bomba de refrigeraci6n principal. 

Julio 1990. 

/4/ Limitation Systems in the Reaktor-Leittechnik. R 193-82/e 11 h 

/5/ Description of the Computer Model LOOP 7 for Calculation of the "Anticipated 

Transients without Scram" (ATWS) RI 1/103/75 E

17
Table 1.-Technical Data of the Reactor Coolant and Pressurizer System. (1/4)

Principal Data 

Thermal Reactor Power 

Power transferred by the steam generators 

Number of reactor coolant pumps 

Total coolant flow through RPV at full load 

Normal operating pressure at the RPV outlet 

Coolant temperature at zero load hot 

Coolant temperature between RPV and SG at full load 

Coolant temperature between RCP and RPV at full load 

Coolant temperature between SG and RCP at full load 

Mean coolant temperature of full load 

Pressure loss in the reactor coolant system at full load 

Water volume incl. full load water volume of pressurizer 

Test pressure for hydrostatic pressure testing 

Temperature for hydrostatic pressure testing 

- min. 

- max. 

Reactor Pressure Vessel 

Design pressure 

Design temperature 

Normal operating pressure 

- Inlet 

- Outlet 

Operating temperature 

- Inlet 

- Outlet 

Water volume 

Pressure loss across RPV at full load 

Fuel

3010 MW 

3028 MW 

3 

16602 kg/s 

158 bar 

296.7 °C 

325.8 °C 

294.5 °C 

294.3 °C 

310.0 °C 

6.6 bar 

300 m3 

227.5 barg 

45 0C 

55 °C 

176 bar 

350 °C 

161.1 bar 

157.3 bar 

294.5 °C 

325.8 °C 

123.4 M3 

2.8 bar 

Sintered U02

18
Table 1.-Technical Data of the Reactor Coolant and Pressurizer System. (2/4)

Number of fuel assemblies 

Number of fuel assemblies with control assemblies 

Total number of fuel rods 

Number of guide thimbles per fuel assembly 

Arrangement 

Overall length of fuel rods 

Active length of one fuel rod at full load 

Outside diameter of fuel rods 

Reactor Coolant Piping System 

Operating pressure on discharge side of RCP at full load 

Design temperature 

Design pressure 

Inside diameter 

Water Volume (three loops) 

Flow velocity between RPV-SG at full load 

Flow velocity SG-RCP and RCP-RPV 

Coolant flow in cold leg of RCL 

Steam Generators 

Number 

Power transferred per SG at full load 

Heat surface of the heating tubes per SG approx. 

Number of heating tubes per steam generator 

Primary Side 

Flow per steam generator at full load 

Operating pressure of medium approx. 

Inlet temperature at full load 

Outlet temperature at full load 

Design temperature 

Design pressure 

Pressure loss at full load

177 

52 

41772 

20 

Square lattice 

4185 mm 

3414.6 mm 

10.75 mm 

161.2 bar 

350 °C 

176 bar 

750 mm 

29.6 m3 

18.8 m/s 

17.0 m/s 

7.43 m 3

/s 

3 

1009.3 MW 

5400 m2 

4086 

5534 kg/s 

160 bar 

325.8 0C 

294.3 °C 

350 °C 

176 bar 

2.4 bar

19
Table 1.-Technical Data of the Reactor Coolant and Pressurizer System. (3/4)

Volume of one steam generator 

Inner diameter of tubes

Secondary side 

Flow per steam generator at full load 

Main steam pressure in dome at full load 

Max. steam moisture at steam generator outlet at full 

load 

Feedwater inlet temperature at full load 

Design temperature 

Design pressure 

Normal (controlled) water level 

Volume of one steam generator 

Reactor Coolant Pumps 

Number 

Flow per pump at full load 

Pressure increase at full load 

Power consumption at 30 bar and 50 'C approx. 

Power consumption at 157 bar and 297 'C approx. 

Operating temperature at full load 

Design temperature 

Design pressure 

Inlet and outlet parts, inside diameter 

Speed 

Water volume per pump 

Pressurizer 

Total available volume 

Water volume at zero load (hot) 

Steam volume at zero load (hot) 

Water volume at full load 

Steam volume at full load

36.8 m3 

19.6 mm 

552.8 kg/s 

67.5 bar 

0.25 % 

220.1 °C 

350 °C 

88.3 bar 

12. m 

175 m3 

3 

7.49 m3

/s 

6.6 bar 

9518 kW 

7126 kW 

294.4 °C 

350 °C 

176 bar 

750 mm 

1480 minY1 

2.45 m

3

45.2 m3 

13.8m 3 

31.4m 3 

27.5 mn3 

17.7 m3

20
Table 1.-Technical Data of the Reactor Coolant and Pressurizer System. (4/4)

Operating temperature 346 °C 

Design temperature 362 °C 

Design pressure 176 bar 

Operating pressure 157.5 bar 

Internal diameter in the cylindrical section 2200 mm 

Total nominal pressurizer heating capacity 1638 kW 

Total number of heater rods 78 

heating capacity per heater rod 21 kW

21
Table 2.- Variables identification in transient figures. (1/2)

Keyword 

YTOOP511 

YT00X 111 

YA10TO03 

YA20TO03 

YA30TO03 

YA10T955 

YA20T955 

YA30T955 

YA10TO01 

YA20TO01 

YA3T0O01 

YTOOT101 

YTOOL951 

SF1OC110 

RAO1F901 

RA02F901 

RA03F901 

RL21F002 

RL22F002 

RL23F002 

YB1OL951 

YB20L951 

YB30L951 

P 200_01 

CNTRL 925 

CNTRL 056

Variable 

PLANT DATA 

Primary Pressure 

Thermal Corrected Reactor Power 

Hot Leg Temperature (Loop 1) 

Hot Leg Temperature (Loop 2) 

Hot Leg Temperature (Loop 3) 

Cold Leg Temperature (Loop 1) 

Cold Leg Temperature (Loop 2) 

Cold Leg Temperature (Loop 3) 

Average Temperature (Loop 1) 

Average Temperature (Loop 2) 

Average Temperature (Loop 3) 

Primary Average Temperature 

Pressurizer Level 

Secondary Pressure (Header) 

Main Steam Mass Flow Rate (Steam Generator 1) 

Main Steam Mass Flow Rate (Steam Generator 2) 

Main Steam Mass Flow Rate (Steam Generator 3) 

Feedwater Mass Flow Rate (Steam Generator 1) 

Feedwater Mass Flow Rate (Steam Generator 2) 

Feedwater Mass Flow Rate (Steam Generator 3) 

Steam Generator 1 Level 

Steam Generator 2 Level 

Steam Generator 3 Level 

RELAP DATA 

Primary Pressure 

Thermal Reactor Corrected Power 

Permitted Reactor Power

22
Table 2.- Variables identification in transient figures. (2/2)

Keyword 

CNTRL 010 

CNTRL 012 

CNTRL 014 

CNTRL 011 

CNTRL 013 

CNTRL 015 

CNTRL 467 

CNTRL 468 

CNTRL 469 

CNTRL 484 

CNTRL 140 

CNTRL 004 

MFLOWJ 160 

MFLOWJ 260 

MFLOWJ 360 

PMPVEL 150 

PMPVEL 250 

PMPVEL 350 

P 910_01 

MFLOWJ 660 

MFLOWJ 760 

MFLOWJ 860 

CNTRL 061 

CNTRL 2165 

CNTRL 2265 

CNTRL 2365 

CNTRL 724 

CNTRL 729 

CNTRL 734

Variable 

Hot Leg Temperature (Loop 1) 

Hot Leg Temperature (Loop 2) 

Hot Leg Temperature (Loop 3) 

Cold Leg Temperature (Loop 1) 

Cold Leg Temperature (Loop 2) 

Cold Leg Temperature (Loop 3) 

Average Temperature (Loop 1) 

Average Temperature (Loop 2) 

Average Temperature (Loop 3) 

Primary Average Temperature 

Primary Average Temperature (Setpoint) 

Pressurizer Level 

Coolant Mass Flow Rate (Loop 1) 

Coolant Mass Flow Rate (Loop 2) 

Coolant Mass Flow Rate (Loop 3) 

Pump Speed (Loop 1) 

Pump Speed (Loop 2) 

Pump Speed (Loop 3) 

Secondary Pressure 

Main Steam Mass Flow Rate (Steam Generator 1) 

Main Steam Mass Flow Rate (Steam Generator 2) 

Main Steam Mass Flow Rate (Steam Generator 3) 

Generator Power 

Feedwater Mass Flow Rate (Steam Generator 1) 

Feedwater Mass Flow Rate (Steam Generator 2) 

Feedwater Mass Flow Rate (Steam Generator 3) 

Steam Generator 1 Level 

Steam Generator 2 Level 

Steam Generator 3 Level

23
Table 3.- Steady state parameters. (1/2)

BOC PLANT BOC RELAP EOC PLANT EOC RELAP 

Primary pressure (MPa) 15.795 15.85 15.89 15.87 

Average Temperature (°C) 308.65 308.0 310.05 310.3 

Cold Leg Temperature, Loop 1 (°C) 293.01 292.34 294.75 294.72 

Cold Leg Temperature, Loop 2 ('C) 292.34 294.72 

Cold Leg Temperature, Loop 3 ('C) 292.77 292.45 295.13 294.83 

Hot Leg Temperaure, Loop 1 ('C) 322.24 323.6 328. 325.83 

Hot Leg Temperaure, Loop 2 (°C) 323.6 325.83 

Hot Leg Temperaure, Loop 3 ('C) 322.16 323.6 326.16 325.82 

Pressurizer Level (m) 7.60 7.59 7.78 

Secondary Pressure (MPa) 6.72 6.61 6.967 6.775 

S.G. 1 Mass Flow Rate (kg/s) 531.59 529.15 550.46 

S.G. 2 Mass Flow Rate (kg/s) 530.8 550.71 

S.G. 3 Mass Flow Rate (kg/s) 524.48 527.41 547.86 

S.G. 1 Feedwater Mass Flow Rate (kg/s) 553.8 529.24 547.16 

S.G. 2 Feedwater Mass Flow Rate (kg/s) 553.3 529.18 548.12
Table 3.- Steady state parameters. (1/2)

BOC PLANT BOC RELAP EOC PLANT EOC RELAP 

S.G. 3 Feedwater Mass Flow Rate (kg/s) 546.0 527.16 547.16 

S. G. 1 Level (m) 11.96 11.96 11.98 

S. G. 1 Level (m) 11.96 11.98 

S. G. 1 Level (m) 12.05 11.97 11.97
Table 4.- Run Statistics.

CASE 1, BOC Conditions 

Transient time (s) 1500. 

CPU time (s) 5405.8 

Total number of time steps 10000 

CPU time /Transient time 10.81 

Number of active volumes 303 

CASE 2, EOC Conditions 

Transient time (s) 500. 

CPU time (s) 15482.8 

Total number of time steps 110000 

CPU time /Transient time 10.96 

Number of active volumes 1303

26
t

Figure 1.- C. Trillo Nodalization (Primary System)

-b1
STEAM GENERATOR

00

Figure 2.- C.Trillo Nodalization ( Steam/Water cycle)
6Z

C) 

* a

ci)

-I- - -- I

CDI 

CD 

CD 

C44 

oo 

c-i F - - - - - -

I3. I I 

__..

) 

) 

)
C.TRILLO. REACTOR COOLANT PUMP TRIP 314

312 

310

308 

306

0 50 100 150 200 

TIME (s) 

FIG.4.- PRIMARY AVERAGE TEMPERATURE, YTO

250 300 350 400

C.TRILLO. REACTOR COOLANT PUMP TRIP

50 100 150 200 2.50 300 350 

TIME (s)

400

FIG 5.- SECONDARY PRESSURE. SFIOCI 10

30

E.-

304 

302 

300 

90

85 

so

w 

U, 

U,

75

70

65 

60

0
C.TRILLO. REACTOR COOLANT PUMP TRIP

0 50 100 150 200 250 300 350 

TIME (s) 

FIG.6.- PRIMARY PRESSURE, YTOOP511

C.TRILLO. REACTOR COOLANT PUMP TRIP

150 

145 

340 

330 

320 

310 

300 

290 

280 

270 

260 0 50 100 150 200 

TIME (s) 

FIG.7.- HOT LEG TEMPERATURE. YAIOTO03

250 300 350

31

165 

160

155

Cd, 

Cd,

BOC 

*EOC 

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

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

4'00 

00

44
(.TRTIfLL RPAC'"fl flflfl AMrTPThDr~DT'•D

340 ! 

BOC 

EOC 3 3 0 . . .. . ................................................................. I ........................ .......... ............. ........................ ........................ ................... 

3 2 0 -.. . . . . .. . . .,. . . . . . . . .. . . . . . . .. . .. . ...... ................... ...... ................... ... ........... ............ 

3 1 .0 ..... ..... ... ... ... .......... ... ............ ... ... ... ... .. 

, ,3 0 0 . ........................ .. .I .................. :........................ ........................ ........................ ........................ i....................... i.................. 

310. 

2 7 .3 . . . . . . . . . . ... . . . . . .. .......... . . . ... . . . . . . . . . . . .. . . . . . .I........................ .. . . . . . . . . .. . . . . . . . . . 

290 

280 

270C 

2 705 ....................... ,................................................. 4 . .......................... ........................ ................ ....... ................ ....... ..................... 

260 I I . . . 0 50 100 150 200 250 300 350 400 TIME (s) 

FIG.8.- HOT LEG TEMPERAThRE, YA3OTOO3 

G.TRILLO. REACTOR COOLANT PUMP TRIP 

310 I 

: BOG -

: E~~OG .. 

3 0 5 .. . . . . . . .. ... .. . . . . . . .. ... . . . . . .. .. . . .... . . . . . . . . . . .. . . .. . . . . .. . . .. . . . . .. . . . . .. . . . .. . . . ... . . . . . .. .. . 

30 .......... . -.= ...... -.-- ---- " '•- .----- -......................................................... : =....-- .-.--- 

a-l 

-2 8 0 .. . .. . . .. . . .. . . .. . . .. . . .. .. .. . . .. . . .. . . 

0 01 01 02 0 5 0 5 0

TIME (s) 

FIG.9.- COLD LEG TEMPERATURE, YA I T955

32
C.TRILLO. REACTOR COOLANT PUMP TRIP

310 

BOC 

EOC 

3 0 5 .... . ....................................... ... . ....................................................................... ................................................................... 

305 

P1.0-CL LE TEMPERATURE, YT5 

b -2 9 5 a ....................... .. ..................... :........................ '.........................• ........................ ........................ ........................ '...................... 

•" 2 0 50 10 15 20 25 30:5 0 

i i:T:ME (S) 

FIG 1. COL LE TEPRAUE YAT5

33
200 

150

100

Lt•

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC 

I ! 150 

PMPVEL 150 PMPVEL 250... 

PMPVEL 350. 

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

50 H50 100 150 200 250 

TIME (s)

300 350 400 450

FIG.1.1L- REACTOR COOLANT PUMP SPEED

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

160

140 I120 

100 

80

0 50 100 150 200 250 

TIME (s) 

FIG. 1.2.- PRIMARY, SECONDARY PRESSURE 

34

300 350 400 450

0

0

I q

500

S..... YT00P51 I ~~~...... .....i . . . . . . .i ... ... ......... 2 .... : z " " .......... ...... .. i................... . ................... i ...... .... .. ..... 'P -"0.1. . . . . 

----------- " ...------- -"------ F m ----

-- P 91001 ..........

"..................................2 t2 2 :?: : ........ ....

60 FI I I

500

.!.. . . . . . .

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

......................... . .............. ....................................... a ................................................... .. : ...................

I

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

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

....................................... .................................................... ................... ....................................... ................... ................
C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

I, /A

20 

0

0 50 100 150 200 250 

TIME (s)

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

................................ .................. .............. .... ....................................... a 

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

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

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

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

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

300 350 400 450 500

FIG. 1.3.- REACTOR AND GENERATOR POWER 

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

340 

320

300

ul,

280

260

0 50 100 150 

FIG. 1.4.- LOOP I TEMPERATURES

200 250 

TIME (s)

300 350 400 450 500

35

LIV loo

80

60

0.

40

. .......

-YA10TO03 

YA10T955 

CNTRL 0 V 1 .......... 

CNTRL 467

................................... .. i................... I ................... ..................... ...... ... . .................................. ................... ................ 

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

....... ... ...... !.................... ................... ................... ..... ............ ................... ...................... ............... ..................... ................ . 

. .. . ............... .... ..... ........ ................... ...... ....... ................. i ........ ................... ............ ! ............ ...... ........... . . . ......... ....... 

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

tY I I OO I 

CNTRL 925 

CNTRL 061 ----

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

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

I 0
C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

340-

320

300

280 

260

0 50 100 150 

FIG.1.5.- LOOP 3 TEMPERATURES

200 250 300 

TIME (s)

350 400 450 500

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC 

.YTOOTI10i 

CNTRL 140 :i •I~~TRL4&4.:",. 

I I I I I I I 

. 5 100....... 150....... 200.. 250.... 3...350 . 400.450.500

0 50 100 150 200 250 300 350 400 450 500 

TIME (s) 

FIG. 1.6.- PRIMARY AVERAGE TEMPERATURE 

36

C 

LL

* . . YA30T003 

.YA30TO55 

CNTRL 01:2 CNTR 3.......... 

CNTRL 468 ..... 

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

... ..: .• ... . ,. ........... .. .... ............... : ...... ..... ::::: ................... • ................... .................... •................... •............... .... •................. 

."• -•....-.-; ".. - - - - .... IL%...."" . _ - -- _ _.='.•a.•.. .................. .................... .......... ...... ..................... ............... .. 

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

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

. ... . .... .. . . .. . .. . . . .. .. .!.. . . .. . .. . .. . .. . . . . . .. .. .. . . . . . .. . . . .. . . . . . . . .. .. . . ... . . . . .. . . . . . .

C 

I320 

315 

310 

305 

300 

295 

290 

285 

9'2"

280
8000 

7000 

6000 

5000 

4000 

3000 

2000 

1000 

0 

-1000 

JnAA

0 50 100 150 200 250 

TIME (s) 

FIG. 1.7.- COOLANT MASS FLOW RATE

300 350 400 450 500

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

0 50 100 150 

FIG.1.8.- PRESSURIZER LEVEL

200 250 300 350 400 450 500 

TIME (s)

37

0 

CIO

C.TRILLO. REACTOn COOLANT PUMP TRIP BOC 

MFLOWJ 160 

MFLOWJ 260 

................. ...... ........... M LW -..... 

.................... i ....... ............ • ...... ............. ................... i . ................... ................... ................... ................... ................. 

. ........ . 

0 50 100 150 2. 250 . .i. .. . . . . . . . . . . . . . . . .. . . . . .. . . . .. . . . .. . . . . .. . . . . .. .. . . .. . . . . . . . . . .. .. . . .. .. . . . . . . .. . . . . 

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

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

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

10 

9 

8 

7 

6 

5 

4

-- •VVV
600 

500

0 50 100 150 200 250 

TIME (s) 

FIG.1.9.- S.G. MASS FLOW RATE (LOOP 1,2)

300 350 400 450

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

600 

500 

400

0 50 100 150 200 250 

TIME (s) 

FIG.1.10.- S.G. MASS FLOW RATE (LOOP 3)

300 350 400 450 500

38

:RA01F90i 

MFLOWJ 660 

MFLOWJ 760 ----

S. ............... ................. ..... ............. ;................... •........................ ................... ................... ................... ................... ................ 

IIII I I I

400 -

0 

Li

I I I I I I I

300 

200 

100 

tA

500

0<

300 

200 

100 

n

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

V
200

100

0

600 

500 

400

0 50 100 150 200 250 

TIME (s) 

FIG. 1. 11.- FEEDWATER MASS FLOW RATE

300 350 400 450

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

13 

12 

I1 

10 

9 

8 

7 

6 

5

YB20L951 . . .B30L,951 .• .,. ........... ............ ..-- 

. . ,: - .

. .

.

.

.

.

.

.-

.

.

.....

.

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

........ 

......... , - - - .. 

- ... .............. . .... . ..L..9. ... ...... C N T l A.T ..-.-.-.-.-...... .÷ • :'-= • 2 • ; ........ i .... .......... i ...................... . i......... .= =......_.:....• 

"...................... 7- --- --- CNTRL 734 .... 

..." .• '' • .• ............ . .................. .................... i................... ................... ................... ................... i .................. • ............. 

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

! I I i I I

0 50 too 150 200 250 

TIME (s) 

FIG. 1.12.- STEAM GENERATOR LEVEL

300 350 400 450 500

39

RL21F002 

RL22F002 

RL23F002 S.. .. ~~CNTRL 2165 ...... '.. ........... .................... i ................... ................... i................... ................... ................... ................. C N T Ri•L 2 16 5 ...... ..... 

CNTRL 2365 

S ......... .. ............ .................... ................... .............. 5......... 0......... ...... ................. ...... .. ...... --. -. ...... . 

.. ................. :i ..."" ".......... i................................................................................... ........... ........ ................... ...................................... 

......- . . . . .. . . . . .. . . .. .. . . . . . .. . . . ... . . . .. .. . . . ... . . . . ... . . .. .. . . . . . .. . . .. \. . . .. . .. .... .. .. . . .

C 300

C2

500

C.TRILLO. REACTOR COOLANT PUMP TRIP BOC

I I I
C.TRILLO. REACTOR COOLANT PUMP TRIP EOC 

200 

PMPVEL 1•0 

PMPVEL 250 

PMPVEL 350 ...... 

"10 . ................................................................................... . ................ 

50 

0 

0 50 100 150 200 250 300 350 400 450 500 

OTNE 

FIG2 21- REACTOR COOLANT PUMP SPEED 

C.TRILLO. REACTOR COOLANT PUMP TRIP EOC 

160 ............. ........................ .............. ........... P 200 

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

60 

0 50 100 150 200 250 300 350 400 450 500 

TIME (s) 

FIG.2.2 PRIMARY, SECONDARY PRESSURE

40
120 

100 

80

60

0

40 

20 

0 

340 

320 

300

280

260

0 50 100 150 200 250 300 350 400 450 500 

TIME (s) 

FIG.2.3.- REACTOR AND GENERATOR POWER

C.TRILLO. REACTOR COOLANT PUMP TRIP EOC

IYA10TO03 

YA10T955 .................. ...................................... ....................... .................................. ....................... ................. ff o o0 ....... 

CNTRL 010 .......... r•'M"I'P l mil ..

CNTRL 467 

€............ ........................ •~. . . . . . ... . ...... . . . ...... ... ... .... . .-.. '.............."".. ...... .....

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

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

I, 

S... ....... ................... .................. .................. ........................K.................................................... ....................................

I I I I |

0 50 100 150 200 250 

TIME (s)

I0I 4 I 300 350 400 450 U00

FIG.2.4.- LOOP I TEMPERATURES

41

C.TRILLO. REACTOR COOLANT PUMP TRIP EOC

I
C.TRILLO. REACTOR COOLANT PUMP TRIP EOC

0 50 100 150 200 250 300 350 400 450 500 

TIME (s) 

FIG.2.5.- LOOP 2 TEMPERATURES

C.TRILLO. REACTOR COOLANT PUMP TRIP EOC 

:YTOOTIOi 

CNTRL 140 

.................................................... ..... ................. ................ K. CnTR .494.77.... 

.. .. .................. .................. ................... ................... ................... ....................................... i..................C T L . . .. - .-. _ 

a a .. .......................................................... 

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

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

0 50 100 150 200 250 

TIME (s) 

FIG. 2.6.- PRIMARY AVERAGE TEMPERATURE 

42

300 350 400 450 500

340 

320

300

LL

280 

260

320 

315 

310 

305 

300 

295 

290 

285 

280
8000 

7000 

6000 

5000 

4000 

3000 

2000 

1000 

0 

-1000 

Infm

0 50 100 150 200 250 

TIME (s) 

FIG.2.7.- COOLANT MASS FLOW RATE

300 350 400 450

C.TRILLO. REACTOR COOLANT PUMP TRIP EOC

0 50 100 150 

FIG.2.8.- PRESSURIZER LEVEL

200 250 300 350 400 450 500 

TIME (s)

43

C.TRILLO. REACTOR COOLANT PUMP TRIP EOC 

MFLOWJ 160 

. MFLOWJ 260 ................. ..................... ...... .......... .......... .......... ........................................ ........... ........M , , i . . ..: • : 

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

i ...... 

I 

. .................................... 5 0.. 0..................................... ....................................................... ................ ................... ................ 

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

j . . . .. . . . . . . . . . . . .. . . . . . .. . .. . . . . . . .. . . . . .. . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . .. . . . .. . . . .L. . . . .. . . 

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

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

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

........... " ........................ ... ................. ...................................... .......................................... .................. ................... :................. 

- i --

C 

ci,

"-'V'VV

500

10 

9 

8

7

r.l~ 

..j

6 

5

4
C.TRILLO. REACTOR COOLANT PUMP TRIP EOC

0 50 100 150 200 250 

TIME (s)

300 350 400 450 500

FIG.2.9.- S.G. MASS FLOW RATE (LOOP 1,2

C.TRILLO. REACTOR COOLANT PUMP TRIP EOC 

MFLOWJ 860 

.................. ................... .................. ................... .................... •................... .................... ,................... * ................... >................. 

S.. .............. .................................... i..................... ................... .................... • ................... .................... ................... ;.................. 

.** ~ .. ... . . :........ .

0 50 100 150 200 250 

TIME (s) 

FIG.2. 10.- S.G. MASS FLOW RATE (LOOP 3)

300 350 400 450 500

44

600 

500 

400

0 

La 

cn

300 

200

100 

0

600 

500 

400 

300 

200

0 

LL.

100 

0
C.TRILLO. REACTOR COOLANT PUMP TRIP EOC

0 50 100 150 

TIME (s) 

FIG.2. 11.- FEEDWATER MASS FLOW RATE

C.TRILLO. REACTOR COOLANT

CNTRL 72"4 

CNTRL 729 

CNTRL 734 ..... 

0 50...5 ..... 0.... .....

..... ............ .'......................................................... ,-.,-... ............... ................... :.........•.'.......-:.". ..-- ......... ..:•................... ... ................. 

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

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

0 50 100 150 200 .250 

TIME (s) 

FIG.2.12.- STEAM GENERATOR LEVEL 

45

200 250 300 

PUMP TRIP EOC

300 350 400 450 50I0

600 

500 

400

0 

-J 

LL. 

U,

300 

200

100 

0

14 

12

10

-3 

UI

8 

6
NRC FORM 336 US. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-8W (AssIgned by NRC, Add Vol., Supp., Rev., 

NRCM 1102, and Addendum Numbers, If any.) 

3201, 3202 BIBLIOGRAPHIC DATA SHEET 

(See ,ftbCtb on therevevse) 

2. TITLE AND SUBTITLE NUREGA1A-0177 

Assessment of a Reactor Coolant Pump Trip for TRILLO NPP 

with RELAP5/MOD3.2 3. DATE REPORT PUBYSHED 

March 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

M.F. Lozano, P. Moreno, PMSA Technical 

A.L Cedillo, M.N. Sanjurjo, C.N. TRILLO 

7. PERIOD COVERED (Onclusive Dabes) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (if NRC, provide Divion, Ooke or Region, U.S. Nuclear Regulatoy Co-miion-, wd mailing addres•, if uornkir, 

provide name and mailng addess) 

PMSA C.N. TRILLO 

Sector Embarcaciones Plaza Carlos Trias Bertran, 7, planta 

38760 Tres Cantos - Madrid 28020 Madrid 

SPAIN SPAIN 

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, bpe 'Same as above', if conhctor, provide NRC Divison Oflice or Region, U.S. Nuclear Regule•ry Commission, 

and mailing address.) 

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 r200 words or kw) 

This document presents the comparison between the simulation results and the plant measurements during a Reactor Coolant 

Pump Trip in C. Trillo Nuclear Power Plant in different bumup conditions. This work is part of the Spanish contribution to the Code 

Assessment and Maintenance Program (CAMP) as a member of UNIIDAD ELECTRICA, S.A. (UNESA). C. Trillo is a 3010 MWth 

three loop PWR Nuclear Power Plant designed by Siemens-KWU. Commercial operation started in August 1988, currently the 

plant is running the 11th annual cycle. The simulation has been carded out with the RELAP5/MOD3.2 code, running on a HP 

715/50 work-station. The transient main parameters have been correctly assessed, however predictions at the beginning of 

transient are more difficult as the reactor remains at a high power level after pump trip.

12. KEY WORDS/DESCRIPTORS (List wods orphrases that wil assist ,esearchers in ocating he report.) 13. AVAIL.ABITrIY STATEMENT 

RELAP5/MOD3.2 unlimited 

Reactor Coolant Pump Trip 14. SECURITY CLASSIFICATION 

(This Page) 

unclassified 

(This Repot) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-8)
Federal Recycling Program
NUREG/IA-0177 ASSESSMENT OF A REACTOR COOLANT PUMP TRIP FOR 

TRILLO NPP WITH RELAP5/MOD3.2

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

MARCH 2000

FIRST CLASS MAIL 

POSTAGE AND FEES PAID 

USNRC 

PERMIT NO. G-67