NUREG/IA-0119

ICSP-AS-BOUT-R

International

Agreement Report

Assessment and Application of

Blackout Transients at

Asco Nuclear Power Plant

with RELAP5/MOD2

Prepared by

F. Reventos, J. S. Baptista, A. P. Navas, P. Moreno

Asociacion Nuclear Asco

C/ Tres Torres, 7

08017 - Barcelona

Spain

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

June 1993

Prepared as part of

The Agreement on Research Participation and Technical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report was prepared under an international cooperative

agreement for the exchange of technical information. Neither

the United States Government nor any agency thereof, or any of

their employees, makes any warranty, expressed or implied, or

assumes any legal liability or responsibility for any third party's

use, or the results of such use, of any information, apparatus product or process disclosed in this report, or represents that its use

by such third party Would not infringe privately owned rights.

Available from

Superintendent of Documents

U.S. Government Printing Office

P.O. Box 37082

Washington, D.C. 20013-7082

and

National Technical Information Service

Springfield, VA 22161
NUREG/IA-0119

ICSP-AS-BOUT-R

International

Agreement Report

Assessment and Application of

Blackout Transients at

Asco Nuclear Power Plant

with RELAP5/MOD2

Prepared by

F. Reventos, J. S. Baptista, A. P. Navas, P. Moreno

Asociacion Nuclear Asco

C/ Tres Torres, 7

08017 - Barcelona

Spain

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

June 1993

Prepared as part of

The Agreement on Research Participation and Technical Exchange under the International Thermal-Hydraulic Code Assessment

and Application Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report documents work performed under the sponsorship of the Consejo De

Seguridad Nuclear of Spain. The information in this report has been provided

to the USNRC under the terms of an information exchange agreement between the

United States and Spain (Technical Exchange and Cooperation Agreement Between

the United States Nuclear Regulatory Commission and the Consejo De Seguridad

Nuclear of Spain in the field of reactor safety research and development,

November 1985). Spain has consented to the publication of this report as a

USNRC document in order that it may receive the widest possible circulation

among the reactor safety community. Neither the United States Government nor

Spain or any agency thereof, or any of their employees, makes any warranty,

expressed or implied, or assumes any legal liability of responsibility for

any third party's use, or the results of such use, or any information,

apparatus, product or process disclosed in this report, or represents that

its use by such third party would not infringe privately owned rights.
ABSTRACr

The Asociaci6n Nuclear Asc6 has prepared a model of Asc6 NPP using

RELAP5/1/1OD2. This model, which include thennalhydraulics, kinetics and

protection and controls, has been qualified in previous calculations of

several actual plant transients.

The first part of the transient presented in this report is an actual blackout and one of the transients of the qualification process. The results are

in agreeennt with plant data.

The second part of the transient is a hypothetical case. It consists in

re-starting a prinary pump and assume a new black-out.

The phenacenology prediction of this second part has been useful from the

operation and safety point of view.

iii

TABLE OF CONTENS

Page

ABSTRACT ................................................... iii

EXZT E SUMMARY .......................................... vii

1. i117ODUCTION ........................................... 1

2. PLANT AND TRANSIEN'T DESCRIPTION ....................... 4

2.1 PLANT DESCRIP•ON .................................. 4

2.2 DATA ACQUISITION AND ANALYSIS SYSTEM

DESCRIPTION ..................................... 4

2.3 TRANSIENT O........................ 4

3. MODEL DESCRIPTION ..................................... 6

3.1 THERL-HYDRAULIC MODEL ............................ . 6

3.2 KINETIC MODEL ....................................... 8

3.3 CONTROL AND PROTECTION SYSTE4S MODEL ............. 8

4. STEADY STATE CAICULATION ...................... 12

5. TRANSIENT CALCUIATION AND COMPARISON VERSUS

AC=UAL DATA ........................................... 13

6. APPLICATION: TRANSIENT DESCRIPTION .............. .... o 15

7. RUN STATISTICS ........... ............... * 17

v
TABLE OF CONTENTS (CONT.)

Page

8. CONCLUSIONS ...................................... .. 18

9. o... ................................. 19

INDEX OF TABIES ..................................... ...... 21

INDEX OF FIGURES ............................................ 32

vi
EC=I=VE SU02ARY

Ascb Nuclear Power Plant is a nuclear station with two PWR of 930 W4e of

Westinghouse design.

TIe Tn:: a.d1htydraulic analysis group of the Asodiaci6n Nuclear Asc6 (ANA) has

prepared i :d )el of the plant using RELAP5/MOD2. This model includes

thenralhydraui .s, kinetics and protection and controls.

ANA's commitment with the International Code Assessment and Application

Program (ICAP) is the participation with two cases.

One of the transients selected for this purpose is the "Assessment and

application of Black-out Transients at Asco NPP wi-: RELAP5/MOD2".

This transient has been chosen mainly because it is on-line with one of the

most important goals of the Thermalhydraulics analysis group of ANA, this is

to provide engineering support to plant operation.

After a successful assessment of an actual black-out transient, a

hypothetical case is simulated in order to generate information about the

application scenario (re-start and re-trip a Reactor Coolant Pump). The main

conclusions of the analysis are the following:

- First part: close agreement between results and data.

- Second part: the primary system recovers natural circulation. Core massflow is always positive during all the transient.

- Relap5/I.ld2 Asco Model is a valuable tool to analyze plant transients.

vii

FOREWORD

This report has been prepared by Asociaci6n Nuclear Asc6

in the framework of the ICAP-UNESA Project.

The report represents one of the application calculations

submitted in fulfilment of the bilateral agreement for cooperation in thermalhydraulic activities between the Consejo

de Seguridad Nuclear of Spain (CSN) and the United States

Nuclear Regulatory Commission (USNRC) in the form of Spanish

contribution to the International Code Assessment and Applications Program (ICAP) of the USNRC whose main purpose is

the validation of the TRAC and RELAP system codes.

The Consejo de Seguridad Nuclear has promoted a coordinated Spanish Nuclear Industry effort (ICAP-SPAIN) aiming to

satisfy the requirements of this agreement and to improve

the quality of the technical support groups at the Spanish

Utilities, Spanish Research Establishments, Regulatory Staff

and Engineering Companies, for safety purposes.

This ICAP-SPAIN national program includes agreements

between CSN and each of the following organizations:

- Unidad Elictrica (UNESA)

- Uni6n Iberoamericana de Tecnologia Elictrica (UITESA)

- Empresa Nacional del Uranio (ENUSA)

- TECNATOM

- LOFT-ESPANA

The program is executed by 12 working groups and a generic code review group and is coordinated by the "Comiti de

Coordinaci6n". This committee has approved the distribution

of this document for ICAP purposes.

ix

D-1/90-MPNV

1. INTRODUCTION

In 1986, the Asociaci6n Nuclear Asc6 (AM) created a group for plant and

core thermal-hydraulic analysis. The objectives of the group are as

follows:

1. Create and update core and plant thermal-hydraulic models based on

best-estimate criteria.

2. Provide off-line engineering support to the different technical

branches of ANA (i.e., technical services, reactor operation):

a. Analyze operating events that result in event reports.

b. Assess plant systems and/or equiprent modifications as well as

plant operating procedures and emergency instructions.

c. Analyze plant behavior under incident or accident conditions in

the abovementioned cases.

d. Scenarios and core damage evaluation for probabilistic risk

assessment.

3. Review final safety analysis report transients and accidents based on

best-estimate criteria.

4. In the future and if appropriate, participate in Asc6 individual

plant examination.

The plant analysis activities developed so far include the following:

1. Implementation of RELAP5/MID2 (Ref.1) cycle 36.05 in its IBM version

in an IBM 4381 and 3090 and cycle 36.04 and its Control Data

Corporation version in a Cyber 180/830. The results of both versions

for the scenarios analyzed are in close agreement.

2. Thermal-hydraulic model of both the primary and secondary systems. /2/

-I-
3. Kinetic model specifically adapted to Asc6

4. Simulation of control and protection systems /3/.

5. Revision and detailed study of all start-up tests and every transient

that has occurred in either unit. A total of 60 cases were studied.

Because of the influence of plant dynamics and the quality and

availability of plant data, six cases were selected to validate the

complete plant model:

a. Black-out

b. Faulty pressurizer spray valve opening.

c. Turbine trip without steam dump and secondary relief valves

available.

d. Loss of feedwater (LOEW)

e. Turbine trip with all systems available

f. Turbine power step

6. Simulation of the above six transients

parameters /4/, /5/, /6/.

7. Participation in the Internacional Code

Program with two cases.

and adjustment of control

Assessment and Application

8. Analysis of transients such as small-break-loss-of-coolant accident

(SBLOCA), anticipated transient without scram (A'IWS), and others for

PRA studies.

The adjustment and qualification process is the first and rost important

part of plant analysis. Sufficiently accurate predictions with

meaningful sets of measured data provide validation of both the model

-2-
and the procedures to be used in the future to analyze various transient

and accident scenarios of general interest such as SBLOCA and AtWS.

-3-
2. PLANT AND TRANSDET DESCRIPTION

2.1 Plant Description

Asc6 Nuclear Station is a nuclear power plant with two 930-Mwe

Westinghouse Pressurized Water Reactors (PWR). The first

criticalities were reached in June 1983 and September 1985,

respectively, for units I an II. Today, both units are in their

sixth and fourth cycles of normal operation.

The irain characteristics of both units are given in Table 1. The

core contains 157 fuel assemblies of (17x17 -25) fuel rods and the

steam generators are typical ones with U-tubes and preheaters

(model D-3). All other major ccmponents are standard Westinghouse

components.

2.2 Plant Data Acquisition System

The plant data used in each assessment calculation is that

produced by the plant process computer on the post-trip report. It

types a value of each pre-selected variable every 10 seconds. The

post-trip report of this particular transient is given in Annex I.

2.3 Transient description

The transient being analyzed is divided into two parts. The first

one is an actual transient that took place on October, 14; 1987

and is used to assess REIAP5/Mod2 and the plant model

capabilities. The second part of the transient can be considered

as an aplication of the model since a hypothetical situation is

analyzed.

The assessment part is described below: After an electrical power

gird perturbation, the turbine and reactor tripped. Thirty-seven

seconds later, a black-out took (loss of offsite power) place and

the reactor coolant pumps (RCPs) tripped. The normal actuation of

the

-4-
errergency diesel generators allowed the correct safeguards

sequence.

In Table 2 a description of the main events is presented.

-5-
3. YrDDE DESCRIPTIN

Figure 1 shows the model used to simulate the plant. It consits of 134

volumes, 146 junctions, 32 heat structures, and 259 control variables.

The model includes the vessel, the three primary loops, the pressurizer,

the three steam generator, the three secundary loops, and the steam

lines. The turbine, condenser, and feedwater tank are modeled as timedependent volumes.

The usual practice of iplementing REAP5 by homogenizing the nultiple

steam generator loops into two loops (one including the pressurizer) is

not followed in this model because of the following reasons:

a) non - symmetric distribution of auxiliary feed-water among the three

steam generators

b) Different length of the steam lines of each loop.

c) Different number of plugged tubes in each steam generator.

d) Non - symmetric transients (loss of feed water, steam generator tube

break, small break in the primary circuit, one reactor coolant pum.

trip, and so on) that require modeling of different actuations or

boundary conditions, in each loop.

e) Computer availability

3.1 Thermal - Hydraulic Model

a.- Primary System.

The model /2/ of the primary system includes the main

components of the plant. The core is modeled by volume 120

and the proper heat structures. Volume 130 simulates the

by-pass region between the core baffle and the core barrel.

Volume 140 model the upper plenrum and volumes 150 and 160 the

vessel upper head. The three hot lines depart fran the core

-6-
upper plenum. Loop 3 (volume 410) is connected with the

pressurizer through volume 510. The pressurizer is divided

into two volumes.

The lower one (volume 520) is divided into five nodes. Heat

structures, simulating the pressurizer actual heaters, are

attached to the first two nodes. Volume 525 is a branch in

order to model the junctions connecting the pressurizer with

the safety and relief valves and with the spray system.

Volume 420 models the remaining of the hot leg. Volumes 430

and 440 simulate the water boxes of the steam generator. The

primary side of the steam generator is modeled by volume 431

divided into nine nodes. Volumes 450, 465, 466 and 470

represent the cold leg. Volume 460 model the primary coolant

puup, proper homologous curves, given by the vendor, have

been used for this purpouse. Volume 468 models the Safety

Injection System.

b.- Secondary System

The model of the secondary system starts with Tine Dependent

Volume 870 (Loop 3) that represent the feed-water going to

the steam generator. Volume 871 models the Auxiliary Feed

Water. The downcomer is simulated by means of volumes 800,

801, 822 and 825. The steam generator preheater is modeled by

volumes 806 and 807, and the reaning of the tubes zone by

volumes 808, 806, 809 and 810. The steam separator with

volume 820. Volume 830, 840 and 850 model the steam dryer and

the domo of the steam generator. The steam line starts at

volume 880. Safety and relief valves (ccmponents 886 and 884)

are connected to volume 881. Cmponent Valve 885 models the

isolation valve. Time Dependent Volumes 994 and 999 represent

the free atmosphere. The steam is conducted throughout volume

883 to the steam-collector, volume 900. Finally Valves 906,

903 and 907 model the by-pass to condenser valve, and the

turbine stop and control valves, respectively.

-7-
Proper heat structures are used to connect thermally the

.primary side of the steam generator with the secondary side.

Actual values are used for all the variables except for the

hydraulic diameter, heat transfer surface, and thermal

conductivity of -the tubes material where some changes were

introduced in order to achieve the actual heat transfer rate

without any change in primary average temperature and

secondary pressure.

The data used to model volumes and junctions as well as heat

structures were taken form plant design information /7/.

3.2 Kinetic Model

The kinetic model /3/ was prepared using the RELAP5/lod2 spaceindependent reactor kinetics option with data from the AM Nuclear Analysis Group. The model includes a scram table of

reactivity versus time. The total control rod drop tine is the

actual value measured at plant. This table is activated by reactor

trip.

The control model supplies the reactivity of the C and D control

rod banks.

This control reactivity is added to the feedback reactivities

calculated by the kinetic model from the data supplied for the

specific burn-up condition of each transient.

3.3 Control and Protection Systems Model

The protection and control systens were modeled using REEAP5/MDD2

control blocks and following specific setpoint studies, logical

diagrams and technical specifications of the plant /3/, /8/, /9/. The

model includes the following systems:
a. - Reactor Trip System

The reactor can be tripped in the FELAP5/MOD2 model because of

the following effects:

- Low primary pressure.

- High primary pressure.

- Low speed at any pump.

- High pressurizer level.

- High reactor power.

- Low level at any steam generator.

- Overtemperature.

- Overpower.

- Turbine trip.

- Safety injection.

In Figure 2 the logic of the reactor protection system is

presented.

b.- High pressure Injection System

Using the following signals:

- Very low Average Tenperature.

- Low steam generator pressure.

- High steam mass flow rate.

- Low primary pressure.

- Large pressure difference between S.G.

the logic of the safety injection system was reproduced. The

massflow rate injected is modeled by means of the punms

characteristic curves.

c.- Turbine Trip and Control System

The position of the turbine control valve is controlled as a

function of the difference between the Pequired Power and

-9-
Actual Power, with the proper control block to model the actual logic of the plant.

The Turbine Trip (closure of the turbine stop valve) is

modeled. The signals of Safety Injection, very high steam

generator level and Reactor trip, are used to trip the turbine.

Turbine run-back has not yet been modeled.

d.- Feed Water Control System

The feed water control system has been modeled as shown in

Figure 3. The massflow rate calculated by the control system

is injected by means of a tine dependent junction.

The auxiliary feed water system is also included in the model.

e.- Pressurizer level and pressure control system

The model of the pressure control system actuates upon

heaters, spray valve and charging pump. Pressurizer safety and

relief valves and level control system are also simulated.

f.- Steam Dump control system

In Figure 4 the model used for the steam dump control system

is represented.

g.- Average Temperature Control System

The average temperature control system modeled with

RELAP5/Yj3D2 is shown in Figure 5. As can be observed this

system controls both the primary average temperature and the

prnmary-secondary power mistmach.

-10-
Other systems modeled are:

h.- Steam line Isolation logic.

i.- Main Feedwater Isolation logic.

j.- Safety and Relief valves of the secondary.

-11-
4.- STEADY STATE CALCULATION

A steady state calculation was performed with the plant at 100% rated

condition. The objective is to obtain a stable condition to start

transients.

In Table 3 a comparison between the model results and the plant data is

given for the main plant variables. A complete description of the plant

sensors and signals is given in ref. 10.

-12-
5.- ASSESSM1T: TRANSIENT CA1CU=ATION AND COMPARISON VERSUS ACTUAL DATA

The plant transient took place on October, 14; 1987. As said before, the

plant data acquisition system available at that time was only able to

print-out a value of each variable preselected every 10 seconds. On the

other hand, for this transient, only four variables were printed with

enough accuracy to validate the RELAP5 nmdel (see Table 4 and Table 5).

Table 6 shows a comparison of the cronology of the main events.

These variables are:

- loop 3 mass flow rate.

- Primary Average Temperature.

- Vessel Delta-Tenperature.

- Primary pressure.

The comparison of model prediction versus actual values for these variables is given next:

a.- Primary Mass-Flow Rate

In Figure 6 the canarison between actual data and RELAP5/MDD2 is

given.

From second 40 to 100 steady state at 100% rated power is assumed;

however the plant was not stable during that period since the

electrical network perturbation cause oscillations in turbine and

reactor power. Since neither the actual cause that triggered the

transient for the actual situation of the plant are perfectly known,

due to lack of data, it was decided to trip the reactor and turbine

from an 100% stable condition. Also during the 37 seconds between

reactor and turbine trip and black-out, network perturbations caused

same primary pumps velocity changes as can be observed in Figure 6.

Once black-out took place the prediction of mass flow rate decay

shows a close agreement with actual data. To achieve this agreement

it was used the RELAP5/YFD2 Pump component adjusting accordingly the

-13-
friction coefficcients.. The camparison ends at 280 seconds (180

seconds after reactor trip) because of lack of plant data beyond

that point.

b.- Vessel Delta - Temperature

At Figure 7 the cacparison between actual data and RELAP5/MOD2

prediction of this variable with and without signal treatment is

given. As can be observed a slightly smaller decrease is obtain with

the model during the period between reactor trip and black-out (100

to 137 seconds). It is due to the primary pumps behaviour during

this time interval that induces an increase in primary mass flow

(see Figure 6). Once black-out takes place the small difference

between actual data and prediction is probably due to the difference

in primary mass flow rate, and reactor residual power because of the

grid perturbation that took place before reactor trip.

c.- Primary Average Temperature

Figure 8 shows actual data of this variable campared against code

prediction with and without signal proccessing. A good comparison is

obtained for most of the transient although at the latest part of it

a growing mistmatch can be observed. It is probably due to reactor

residual power and the actuation of the steam-dump at the early

stages of the transient. Since there is not data of the steam-dump

behaviour it was not possible to improve the agreement.

d.- Primary Pressure

Figure 9 shows the comparison between primary pressure data and

REEAPS/MDD2 prediction. An almost constant mismatch is observed

during the transient. The main reason, since a better agreement is

obtained in average temperature, could be due to the pressurizer

heaters actuation during the time from reactor trip to black-out. No

data is available on heaters behaviour and its proper abtuation is

assumed in the model.

-14-
6.- APPLICATION: TRANSIET DESCRIPTION

An interesting hypothetical case was suggested to study after the blackout transient was assessed.

The case basically consisted in re-starting a primary pump, once

electrical power in recovered (500 seconds after reactor trip is

assumed) and once if is at rated speed assume a new black-out. The

objective being to know how long it will take to restore natural

circulation throught-out the core since two-loop will have reverse mass

flow rate at the time of the second black-out.

Table 7 shows the cronology of the main events of the application

transient.

In figures 10 to 18 the main variables are given:

Fig.

it

it

it

10

11

12

13

14

15

16

17

18

19

PRfMARY VARIABLES

Reactor Power

Primary Mass Flow Rates

Loop 3 Mass Flow Rate

Primary Pressure

Primary Average Temperature

Vessel Delta-Temperature

Secondary Pressure

Secondary Mass Flow Rates

Secondary Mass Flow Rates

Steam generator Narrow Range Level

0I

to

of

If

If

-15-
The main conclusion of the analysis is that mass flow rates throught-out

the core goes back to the natural circulation values previous to the

PLump 2 transient without having negative mass-flow rate at any moment,

and that natural circulation is restored in all loops after flow

reversal in loop 1 and 3 (see Figures 11 and 12)

From Figure 15 it can be seen that the increase of temperature in the

vessel returns to the values of the first part of the transient after

the second black-out. The oscillation observed fram second 900 to the

end of the time analyzed is due to the small oscillation in mass flow

rate after the second black-out (see Fig. ii and 12)

Figures 14 and 13 show the evolution of primary average temperature and

pressure. As can be observed after the second black-out tenperature and

pressure get stable values above those reached after the first one. It

is due (see Figure 16 which represents secondary pressure) to the

decrease in secondary pressure after the first black-out because of the

actuation of the auxiliary feed water system which introduces cold (313

QK) water above the steam generator water level so in the vapor region,

causing condensation and then depressurization in the steam generator.

After the second black-out the steam generator level (see Fig. 19) is

such that no auxiliary feed-water is introduced into the system (see

Fig. 17 and 18). Consequently no secondary side depressurization takes

place resulting in a higher primary average temperature and pressure at

the end of this application transient than in the assessment one.

-16-
7.- RUN STATISTICS

Calculations were carried out on a CYBER 180/830 NOS 2.5 property of

Fundaci6n Leonardo Torres Quevedo located at Santander - Spain.

RELAP5/Mod. 2 cycle 36.04 was used in all the calculations.

In Table 8 a typical run statistics is presented.

-17-
8. ONCLUSIObNS

Both cases, the actual transient and the hypothetical one, have been

simulated with RMEUS/MDD2 Asc6 model.

The results of the actual transient are in agreement with plant data

although only a limited number of variables were recorded.The

nodalization used has allowed the correct simulation of the behaviour of

the plan.

This calculation, along with those of the rest of transients of the

qualification matrix, provides the validation of the model.

This transient provides an assessment of the RCP function that is valid

for future predictions.

The simulation of signal treatment seems the correct one.

The Steam-dump actuation, although in this transient is truncated by the

black-out, can be improved as long as predictions of primary temperature

and pressure are higher than actual data.

The results of the hypothetical transient provide the improvement of the

knowlegde of plant dynamics.

Inactive loops (1 and 3) get reverse flow after the re-start of punp 2.

Natural circulation is restored in all loops after the second black-out.

During all the transient, core mass-flow is always positive.

The hypothetical situation is as safe as the actual one.

The model of Asco using Relap5/Mod2 is a valuable tool to analyze plant

transients and to provide engineering support to plant operation.

-18-
9.- RERNS

1) V. H. RAMSON and R.J. WAGNER, "PELAP5/MOD2: Code Manual, "EGG-SAAM6377, EG&G Idaho, Inc. (Apr. 1984).

2) F. REVENTOS, J. SANMEZ-BAPTISTA, and P. MORENO, "Modelo

Termohidru3lico de la C.N. Asc6", Asociaci6n Nuclear Asc6 (June

1987).

3) F. REVENTOS, J. SACEZ-BAPTSTA, and P. MDRENO, "Miodelo

Tenmohidriulico y de protecci6n y control de C.N. Asc6," Asociaci6n

Nuclear Asc6 (May 1988).

4) F. REVENTOS, J. SANCHEZ-BAPTISTA, and P. MO1REO, "Anilisis de transitorios en A.N.A. con RELAP5/MOD2," presented at 14th Annual Mtg.

Sociedad Nuclear Espafiola, Marbella, Spain, October 26-28, 1988.

5) F. REVENTOS, J. SANCHEZ-BATIISTA, P. MORENO, and A. PEREZ NAVAS,

"Transient Analysis for ASOD Nuclear Power Plant Using RELAP5/MOD2,"

Proc. Ist Int. REIAP5 Users' Seminar, College Station Texas, January

31-February 2, 1989.

6) F. REVENTOS, J. SAN Z-BAPTISTA, A. PEREZ-NAVAS and P. MORENO.

"Transient Analysis in the Asc6 NPP using RELAP5/MDD2," Nuclear

Technology, Volume 90, Number 3, pp 294-307. June 1990.

7) "Final Safety Analysis Report-Asc6 I" (June 1983), "Final Safety

Analysis Report-Asc6 II" (Sep. 1985), and later revisions, Asociaci6n Nuclear Asc6.

8) F. BALLERINI, "Setpoint Study Asc6 Units I and 2," Westinghouse

Electric Corporation (Mar. 1976).

9) "Precautions, limitations and Setpoints of Asc6 Units I and 2".

E/PS/76/068. Rev. 11. Sep. 1988.

-19-
10) C. Simon, et al "Central Nuclear de Asc6 Units 1 and 2. Emergency

Recovery Guidelines Setpoint Values. Calculation and Methodology

Appendix D. Instrmrentation channels statistical calculation of

Uncertainties". WENX 88-08. Rev. 2. October 1989.

-20-
TABLES

TABLE 1. DESCRIPTION OF THE tMIN CHARACT!I•STICS OF ASCO I AND II

NUCLEAR STATION.

TABLE 2. MAIN EVETS THAT TOOK PLACE DURING THE BIACK-OUT.

TABLE 3. C2MPARISON BE7LEE RELAP5/M1D2 VALUES AND ACTUAL DATA FOR

STEADY STATE.

TABLE 4. DESCRIPTION OF PLANT DATA blEASUR11NTS.

TABLE 5. DESCRIPTION OF RELAP5/I'fD2 VARIABLES.

TABLE 6. COMPARISON OF THE CRONODLOGY OF THE MAIN EVENTS (ASSESSMENT

TRANSIENT).

TABLE 7. CRONOLOGY OF THE MAIN EVENTS OF THE APPLICATION TRANSIENT.

TABLE 8. RUN STATISTICS.

-21-
- Electrical power 930 Mw7e

- Thenral reactor power 2686 Mwth

- Fuel U02

- Number of assemblies 157

- Fuel rods per fuel assemrbly (17x17 - 25) = 264

- Active length of fuel rods 3.657 m

- Outside diameter of fuel rods 4.75 x 10-3 m

- Cladding tube material Zr -4

- Cladding tube wall thickness 0.655x10-3 m.

- Average linear heat generation rate 17.2 Kw/m.

- Absorber rods per control assembly 24

- Absorber material Ag - In - Cd.

- Number of coolant loops 3

- Reactor operating pressure (pressurizer) 15.51 Mpa.

- Coolant Average Temperature 581.3 QK

- Coolant flow rate 14287 Kg/s

Table I.- Description of the main characteristics of Asc6 I and II

Nuclear Station. (1 of 3)

-22-
Steam Generator

- Type

- Number

- Height

- Diameter (Upper shell)

- Tube material

- Average tube length

- Design pressure/temperature

(steam plant side)

- Inner diameter of tubes

- Outer diameter of tubes

Reactor coolant Punms

- Type

- Discharge head

- Design flow rate

-Speed

Westinghouse D-3

3

20.6 m

4.445 m

Inconel 600

15.94 m

8.17 MPa/589 QK

1. 687x10-2 m.

1.905x10-2 m.

Westinghouse 93-DS

86.25 m

5.928 m3 /s

155 rad/s

Table I.- Description of the main characteristics of Asc6 I and II

Nuclear Station. (2 of 3)

-23-
Pressurizer

- Height

- Diameter (inner)

- Voltme

- Operating saturation pressure

- Heating power of the heater rods

Steam/Power Conversion Plant

- Feed Water flow rate

- Main steam flow rate

- Steam moisture at steam

generator outlet

- Feedwater temperature

12.835 m

2.134 m

39.64 m3

15.51 MPa

1.40 Iwe

497.5 Kg/s/loop

1492 Kg/s

0.25%

497.05 QK

Table 1 - Description of the main characteristics of Asc6 I and II Nuclear

Station. (3 of 3)

-24-
Second Event

0.0 Turbine and Reactor Trip because of electrical power network

perturbation.

0.4 Steam-Dump starts to open.

2.0 Steam Dump fully open.

20.0 Steam Dump starts to close.

37.0 Black-out.

47.0 Start-up of the Diesels.

92.0 Auxiliary Feed water available.

Table 2 - Main events that took place during the black-out.

-25-
VARIABLE RELAP5/MOD2 PLANT DATA

PRIMARY MASS FLCW RATE (Kg/s) 14027 14287 *

CORE BY-PASS I'JASS FLOW RATE (%) 2.71 2.71 *

VESSEL DELTA-T (QK) 33.24 33.26

REACTOR POWER (H4) 2681 2686

PRIMARY PRESSURE (WMa) 15.50 15.51

PRfl'Y AVERAGE T•MERATURE (QK) 581.1 581.3

RECIRCUIATION RATIO 2.286 2.29 *

UP-STREAM FLOW RATIO

IN THE S.G. PREUETER .517 .520 *

STEA4 GENERATOR NARROW RANGE LEVEL .66 .66

SECONDARY PRESSURE (MPa) 6.808 6.821

STEAM COLICI)R PRESSURE (MPa) 6.705 6.724

STEAM MASS F'LOW RATE (Kg/s) 1478.5 1492.0

* DESING DATA

Table 3 - Coaparison between RELAP5/MDD2 values and actual date for steady

state.

-26-
EITrIFECA'OR

T0421A

T0424A

F0442A

P0499A

DESCRIP ON

Average Temperature neasured in loop 2.

Vessel Delta - Tenperature measured in loop 2.

Meassured Mass Flow Rate of loop 3.

Pressure at loop 2, hot leg.

Table 4 - Description of plant data measurements.

-27-
VARIABLE

P220010000

MIWJ 100030000

CNTIRLVAR 5

CaYrMVAR 26

CL'RLVAR 6

CNTILVAR 25

MFIOWJ 105

MFJ10001

MFJ10002

P681010000

MFJ906

MFJ675

MFJ672

MFJ772

CNTILVAR 16

DESCRIPTION4

Pressure at the Hot leg of loop 2.

Mass Flow rate of loop 3, cold leg.

Average Teqperature.

Proccessed Average Terrperature.

Vessel Delta-T (Hot minus cold leg Temperature).

Processed Vessel Delta-T.

Nuclear Reactor Power.

Mass Flow Rate at the inlet of the lower Plenum.

Mass Flow Rate of loop 1, cold leg.

Mass Flow Rate of loop 2, cold leg.

Secondary pressure at the steam line of loop 1.

Steam-Dump Mass Flow Rate.

Main Feed Water Mass Flow Rate of Loop 1.

Auxiliary Feed Water Mass Flow Rate of Loop 1.

Auxiliary Feed Water Mass Flow Rate of Loop 2.

Steam Generator I narrow range water level.

Table 5 - Description of REIAPS/M)D2 variables.

-28-
Event

Turbine and Reactor Trip because of

electrical power network perturbation.

Steam-Dump starts to open.

Back-up heaters actuation.

Steam Dump fully open.

Steam Dump starts to close.

Low average temperature signal.

Black-out.

Low primary flow.

Start-up of the Diesels.

Auxiliary Feed water motor pumps available.

(Loops 1 and 3)

TiIre (s)

Plant

100.0

100.4

102.0

120.0

137.0

147.0

192.0

Relap

100.0

101.0

101.4

104.0

113.0

118.3

137.0

138.2

147.0

192.0

Table 6 - Ccrparison of the cronology of the main events.

-29-
Event

Re-start pump 2

Pump 2 reaches naoinal speed

Re-trip pum 2

Loop 3 recovers positive flow

Secondary refief valves start cycling

Time (s)

600. (*)

610.

700. (*)

780.

790.

(*M boundary conditions

Table 7 - Cronology of the main events of the application transient.

-30-
CMTR ASSESSMENT

TRANSIEN Tf'M (s)

C (total number of actives volumes in the model)

DT (Total number of time steps)

CPU x 1000

C x DT

CPTIME/TRANSIENT TIME

Table 8 - Run Statistics.

CYBER 180/830

180

14241

173

4359

18.884

79.11

-31-
FIGURES

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

NODALIZATION DIAGRAM OF ASCO NUCLEAR POWER PLANT.

LOGIC OF THE REACZOR PR ION SYSTEM.

F WTER CONTROL SYSEM.

STEAM-DtP CONTROL SYSTEM.

PRIMARY AVERAGE T•EPATURE ,NT SYSTEM.

PRIPMAR MASS FLOW RATE. (ASS)

VESSEL DELTA TAZIERATURE. (ASS)

PRIMARY AVERAGE TEMPEPATURE. (ASS)

PRIMARY PRESSURE. (ASS)

REACMR POWER.

MASS FILO RATES.

LOOP 3, MASS FLOW RATE.

PRIMARY PRESSURE.

PRfIMARY AVERAGE T24ERATURE.

VESSEL DELTA TEMPERATURE.

SEONDARY PRESSURE.

SMASS FILE RATES.

SECONDARY MSS FLOW RATES.

STEAM GENERATOR MARROW RANGE LEVEL.

FIGURE 11.

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

12.

13.

14.

15.

16.

17.

18.

FIGURE 19.

FR/ac/irb/jv

IST00281

27.12.90

-32-
NODALIZATION DIAGRAM OF I

ASCO NUCLEAR POWER PLAI• N

JULY 1990

CONDENSER

~ PRESSUIZUER

.l • E IJIIl l

SI ;- E4

m14

N R~H!~BM

t

" lSTEAM

TT ?4IRATCR I

T I

-L F

, STEAM FI.14IRATOH 3 E

cc

-

-T

l mp

Ef1

. . . .

aB~Ji~~ F.J

H

~1[L

r.-"-l

fH::TlEEj -r.---i Ptn tp-I - [IVU, -2 --1-- ---. . 'qg' 1 w•,'c: :: 5 Eli, "' I . ! 4 Is l -R~I
PRESSURIZER LEVEL

REACTOR POWER SG LO EE U SG. LOW1P

E LESAFETY

SG3 LEVEL INJECTION

I RCP

FIGURE 2. -LOGIC OF THE REACTOR PROTECTION SYSTEM
REACTOR

POWER

-TO FWI

MODULATION

STEAM FLOW LOOP 1

STEAM FLOW LOOP 2

STEAM FLOW LOOP 3

FEED-WATER TRIP SIGNAL

FIGURE 3. FEED-UATER CONTROL SYSTEM
LOW-LOW AVERAGE TEMPERATURE

STEAM-DUMP AVAILABLE

TURBINE TRIPPED

-TO

-TO STEAM-DUMP

VALVES

TURBINE POWER

-Po

FIGURE. 4 STEAM-DUMP CONTROL SYSTEM
AVERAGE

TEMPERATURE

LAG LEAD

LAG

FIGURE. 5 - AVERAGE TEMPERATURE CONTROL SYSTEM
AIA

BLACK-OUT ASSESSMENT

6

, I

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

4 -

0

. '

I I

i .

MFILW10 + F

o u ......................- .

40 80 20 16 20 4 8

FIUR 6 TME(sc - ,FOJ00 + 42 ,0 
ANA

BLACK-OUT ASSESSMENT

40

35

30

0

AJ

w

in

(n

wi

25

20

15

10

5

0

40 80 120 160 200 240 280

FIGURE 7 TIME (sec)

- CNTRLVAR 6 CNTRLVAR 25
ANA

BLACK-OUT ASSESSMENT

590

w 57 I -i .- ...i '.. -i ... i- .L ... i i Z. . . ..--. .......... .. ........ . ......

586. . ..

Id574 --- ......... .... .. . ...... . . . . ..... ......-................. ..... ~ ..... .-.. .. . ..... 5

IL In5 8 .. . . . .. . .. . . . .. .... . . .... . -... . ... .. . .... ..... .... ..... ...... .. . .. . . . . . . . . . . . .•. . . .

2....................................... -... ~

tw 576•"

NL5574T • 2

(t 572

< 568

564 -- ,...

40 80 120 160 200 240 280

FIGURE 8 TIME (sec) - CNTRLVAR 5 CNTRLVAR 26
ANA

BLACK-OUT ASSESSMENT

18

17 --

1

!

!

i

16 -1-

a_

n_

15 -1-

14- ...

I-

+

13 -F-

+ t

* I

+ + + t + t + + + T * p I I I

* I

SI.-

* I

* I

* I

I I ~L -~-J

..

12 I

W I I I ! !

240

!

40 80 120 160 200 280

FIGURE 9

P220010000

TIME (sec)+ P0499A
AiA

BLACK-OUT APPLICATION

3

- '1I "J' 1 -r F- -~-~ -. 

-

020

2.8 

-

2.6

2.4-

2.2-

2-

1.8-

1.6-

1.4

1.2-

1-

0.8-

0.6-

-J

*

* .1

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

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

.

* 

I

.

.

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

. ... .. 

. .. ". 

.

.

I...- 

- .I 

-

I iI *.

.

I

.. 

.. -.. - .!" 

. -. __...... ... ...

_ _ I _

II I II

. .... 2|... .--. 

.

.

i ..--

I ~ j 

. , 

*1*

...

.

..

-.

...

. , 

.

.

0.4 

-

0.2 -

0

I

I I L I I

0

0.2

I

0

0.4

I

0

0.6

I

I

0 0.8

1 1.2

(Thousands)

FIGURE 10 TIME (sec)

-- RKTPOW 

0
ANA

BLACK-OUT APPLICATION

Ed0

ELL-

(n

in

15

14

13

12

11

109876543210

-1

I I ' ' 

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

I 

.

- -1 ~ -... I--...- **--~~--**-...___

I I I' :--i~ 

-

I 

I

* 

I

I I 

.

. I . .

* j 

I S1-.-.

I 

. .i.

I 

I --- I---- . - .. 1 F

I __ 

I

___

l

I ______

F-41... 11 J I 0 0.2 0.4 0.6

(Thousands)

FIGURE 11 TIME (sec)

MFJ10001

0.8

1

1.2

I

- MFLOWJ 105

- MI.C0 

I
ANA

BLACK-OUT APPLICATION

5

4

0)

Q0

<C

a~.

0

U)

3

2

1

0

-1

0 0.2 0.4 0.6 0.8 1

(Thousands)

FIGURE 12 TIME (sec)

MFLOWJ 100030000

1.2
ANA

BLACK-OUT APPLICATION

18 -

I I

1 7 -. . ... .. .. .... ....... .i . .. .. .. ....i . .. . . .... I . ... . ..... .. . ... ... 

I I . ,i I I[

o I I I

ft.

1 6. . .... .... . .......... .. ....- .. ........ iIL ..... ..... ...

I- • I .

U15

U14

13. ..4

II I

I i I ,i .. . . ...

12-I j-- v , 

0 0.2 0.4 0.6 0.8 1 1.2

(Thousands)

FIGURE 13 TIME (sec)

P220010000
IdH

a: w

0~

IdH

Id0

'I:

590

588

586

584

582

580

578

576

574

572

570

568

566

564

562

560

II

- T ~T 

-

ANA

BLACK-OUT APPLICATION

.. .. .. 

.

... ... .

...

... .i.. 

1-

I 

'

_ _ 

I

-.________ 1. 

-- 

"

I

I

* 

I

* 

I

* -..-4 I I

* ..-...J..~..-.......4

4--- 

I I ___________________ 1

IF......-.. 

*

*1- I 

I

I

0 0.2 0.4 0.6

(Thousands)

FIGURE 14 TIME (sec)

CNTRLVAR 5

0.8

1 1.2
ANA

BLACK-OUT APPLICATION

40

30

in

w

20

10

0

-10

0 0.2 0.4 0.6 0.8 1

(Thousands)

FIGURE 15 TIME (sec)

CNTRLVAR 6

1.2
ANA

BLACK-OUT APPLICATION

w

U)

U)

a. 0z

En

9

8.8

8.6

8.4

8.28

7.8

7.6

7.4

7.27

6.8

6.6

6.4

6.26

I

I

T f r 

-

-v -i-. i- I I .1. 1 1. J. I 

I

0 0.2

I I 1 I

0.4 0.6

(Thousands)

FIGURE 16 TIME (sec)

P 681010000

0.8

1 1.2
AN.A

BLACK-OUT APPLICATION

(n

in

700 -

600 -

500 -

400 -

300 -

200 -

-. ------1--~---F---

K

100 i .- I

0 ________ t 9 4 4-~I + 4 + I

0 0.2 0.4

FIGURE 17 - MFLOWJ906

0.6 0.8

(Thousands)

TIME (sec)

- MFLOWJ672

1 1.2
700

600

500

0

-J

Ii.

(n

U)

400

ANA

BLACK-OUT APPLICATION

I I

I _ _ _ _ I _ _ _ _ _ _ _ _ _ _ _

300

200

100

0

0 0.2 0.4

FIGURE 18

0.6 0.8

(Thousands)

TIME (sec)

MFLOWJ772

1 1.2

MFLOWJ675
ANA

BLACK-OUT APPLICATION

-J

Id0z6

ui

1 _

0.9 -

0.8 -

0.7

0.6

0.5

0.4-

0.3

0.2 

-

0.1 -

0-

K/7

-1- ..-

I!oa

Ii

"tr i...,4.4

Ia

i

!

!

!

|e

-- II

Ii

ii I I I I .4 1 I 1 4 I

0 0.2

I I 1 

1

0.4 0.6

(Thousands)

FIGURE 19 TIME (sec)

CNTRLVAR 16

0.81

i

1.2

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION I. REPORT NUMBER

12-89) IAslgened by NRC. Add Vol.. Supo.. Rev.,

NRCM 1102. a Addendum Numberi, It ay.1

3201,3202 BIBLIOGRAPHIC DATA SHEET NUREG/IA-0119

(See insrructions on the aeversel

2. TITLE AND SUBTITLE ICSP-AS-B0UT-R

3. DATE REPORT PUBLISHED

Assessment and Application of Blackout Transients at MONTH I EAR

Asco Nuclear Power Plant with RELAP5/MOD2 June 1993

4. FIN OR GRANT NUMBER

L2245

5. AUTHORIS) 6. TYPE OF REPORT

Francesc Reventos, Pablo Moreno, Jose Sanchez, and Alberto Perez Technical Report 7. PERIOD COVERED fncslovr Oaetest

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (II NRC. Provide Oividtnof Office or fie;on. U.S. Nuclear R*Vwtlroroy CommijuisA, •'•d maili, g eddmu; ii contrstorprOviOW mmo, and mailint addmis.I

Asociacion Nuclear Asco

C/ Tres Torres, 7

08017 - Barcelona

Spain

g. SPONSORING ORGANIZATION - NAME AND ADDRESS (It NRC. type "Same as dbove". ifcontramcor. orovide NRC Oivision. Office or Region. U.S. Nuclear fegufatory Commission.

andotmallnj addrtssj.

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT 1200 woegs oir les)

The Asociacion Nuclear Asc6 has prepared a model of Asc6 NPP using RELAP5/MOD2. This model, which

include thermalhyraulics, kinetics and protection and controls, has been qualified in previous calculations of several

actual plant transients.

The first part of the transient presented in this report is an actual black-out and one of the transients of the

qualification process. The results are in agreement with plant data.

The second part of the transient is a hypothetical case. It consists in re-starting a primary pump and assume a new

black-out.

The phenomenology prediction of this second part has been useful from the operation and safety point of view.

12. KEY WORDS/DESCRIPTORS (List ivr orphose s moit will uslit mooarr** In loctinF rle oo'tn.l 13. AVAILABILITY STATEMENT

ICAP Program Unlimited

NPP with RELAP5/MOD2 14. SECURITY CLASSIFICArION

ASCO ri•,, Page,

Blackout Unclassified

IThki Reoonl

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 42491
Federal Recycling Program

ASCO NUCLEAR POWER PLANT WITH RELAPS/MOD2

JU1nRt ivyY3

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300