NUREG/IA-O 121

ICSP-AS-SPR-R

International

Agreement Report

Assessment of a Pressurizer Spray

Valve Faulty Opening Transient 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

0817 - Barcelona

Spain

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

December 1993

Prepared as part of

Thc Agreement on Rcsearch Participation and Technical Exchange

under the International Thermal-Hydraulic Code Asscssment

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

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U.S. Government Printing Office

Mail Stop SSOP

Washington, DC 20402-9328

and

National Technical Information Service

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NUREG/IA-O 121

ICSP-AS-SPR-R

International

Agreement Report

Assessment of a Pressurizer Spray

Valve Faulty Opening Transient at

Asco Nuclear Power Plant

with RELAP5/MOD2

Prepared by

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

Asociacion Nuclear Asco

Crrres Torres, 7

0817 - Barcelona

Spain

Office or Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Wahshington, DC 20555-0001

December 1993

Prepared as part of

The Agreemcnt on Research Participation and lbchnical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program ([CAP)

Published by

U.S. Nuclear Regulatory Commission

ABSTACr

MThe Asociacibn Nuclear Asc6 has prepared a mocdel of Asc6 NPP using

RELAP5/MOD2. This mrodel, wh~ich include thenzalhydraulics, kinetics and

protection and controls, has been qualified in previous calculations of

several actual plant transients.

one of the transients of the qualification process is a "Pressurizer spray

valve faulty opening" presented in this report. it consists in a primaery

coolant depressurization that causes the reactor trip by overtarrerature and

l~ater on the actuation of' the safety injecin

T0he results are in close agreement with plant data.

iii

TABLE OF CNi'~lr

ABs'rRA=r.............. ....... .. .. .. .. ..

DEtIIV StI'ARY.o....................... .................. o Vi

1. I1NrIT~COUION............ o..... o ...... . o ........ o..... o......1I

2.PI.AN AND TANSI=1 DESCRIPTION ....o....................... o. ........4

2.1 PLA~NT DESCRIPTION.......o.....o............... ...o...........o....4

2.2 DA.TA ACQUISITION~ AND ANALYSIS SYSTM4 DESCRIPTION . ... o .... 4

2 .3 TRANSIE~vr DESCRIPTION . .o.o ..... o ... ............... 4

3. MOD1 DESCRIPTION. ....... o.o..o... o... oooos-o.........o 5

3. 1 TI-271'AL-HYDRAtL',IC MOE .. o........................ o..... 5

3.2 K=~~IC MODL..... ..................... . . 7.. 

3. 3 CON'TL AN P'R=CTI0 SYSrEMS MOELo....o......... o.... o.........7

4. STEADY STATE CALCUL~ATION.........o....o..... o............ oesso11 ...

5. TRANSIM'.r CAI.uIATCt AND CCMPARISON VERSUS XMfM 2 DAIA ....... o..... 12

8. PIEDE .. o...............***.oooooewoooo******o************~ ****e* 17

nMEX OF MABLES .. oooooo. o ........ *o.... o ..... o....... o.*.**..e*... oo 19

3Mxor Fi~u1w ... *.oo*o*******oo oo** **o**a*************e***oos** 30

v

E)t7CIVE SLVMAM

Asc6 Nuclear Powr Plant is a nuclear station with two PWR Qf -930 Mie of

Westinghouse design.

The The~nalhydraulic analysis gru of the Asociaci6n Nuclear Asc6 (ANA) has

prepared a model of the plant using REAP5fYDD2. This model includes

thermaihydraulics, kinetics and protection and controls.

AN-' s =matrm~ent 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 of a

Pressurizer Spray valve faulty opening".

T1his transient has been chosen because of two main reasons:

- Plant instrumentaticn data are reasonably accurate for the entire

transient.

- The assessment qualifies the behaviour of different system and

ca, nents; under abnormal conditions (spray valve, safety injection,

Reactor protection system).

The rra.in conclusions of the analysis are the following:

- Close agreement between results and data.

- 1Relap5/nrod2 Asc6 mordel is a valuable tool to analyze plant transients.

Vii

FO0R E WOR D

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 thermal hydraulic activities between the Consejo

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

Nuclear Regulatory Commission (USWRC) 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 El~ctrica (UITESA)

- Empresa Nacional del Uranio (ENUSA)

- TECNATOM

- LOFT-ESPARA

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.

D-1/90-MPNV ii x

1. NTROD0CtON

In 1986, the Asociaci~n Nuclear Asc6 (ANA) -created a croup for plant and

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

follows:

1. Create and update core and plant thezna 1-hydraulic mrodels based on

best-estinrate criteria.

2. Provide off-lin~e engineering support to the different technical

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

a. Analyze operating events that result in event reports.

b. Assess plant systems and/or equipmrent mocdifications as weill as

plant operating procedures and emergency instructions.

c. Analyze plant behavior under incident or accident conditions in

the abovvrentiorned 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 Aso6 individual

plant examination.

The plant analysis activities developed so far include the following:

1. Imrplement~ation of RELAP5/lMt02 (Pef.l) cycle 36.05 in its IBM version

in an MA 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 agreemnt.

2. Thermal-hydraulic model of both the pr~imay and secondary systems. /2/

-1 -
3. Kinetic --cdel specifically adapted to Aso5

4. Si.rm~lation of control and- protection systems /3/.

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

th-at has occurred in either unit. A total of 60 cases were studied.

Because of the influence of plant dynmuics and the quality and

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

ccuplete plant mod~el:

a. Station blackout

b. Faulty pressurizer spray valve openi~ng.

c. Turbine trip without steam dump~ and secondary relief valves

available.

d. Loss of feedwater MLOMW

e. Tumrbine trip with all systers available

f. Turbine powr step

6. Sinxilation of the above six transients and adjustrrent of control

paramreters /4/, /5/, /6/.

7. Participation in the Internacional Code Assessrent, and Application

Program with two cases.

-2-
8. Analysis of transients such as s-nall-break-loss-of-coolant accident

(SBIJX-A) , anticipated transient without scrxn (A`TWS) , and others for

PA studies.

The adjustment and qualification process is the first and nost inrortant

part of plant analysis. Sufficiently accurate predictions with

rrea~r.ingful sets of rreasu~red data provide validation of both the rrodel

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

and accident scenarios of general interest such as SBLC' and AnS

-3-.
2. PLAINr AND TRANS=~T DESCRIMTON

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 ncrmal operation.

The mirn characteristics of both units are given in Table I. The

core contains 157 fuel asserblies of (17x17 -25) fuel rods arnd the

steam generators are typical ones with U-tubes and preheaters

(rrodel D-3). All other rmajor cat nents are standard Westinghouse

components.

2.2 Plant Data Acqu~isition System

*t.e 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 started fran 100% rated conditions because of an

unnoticed failure on the primrary pressure control system.

The- failure turned off all the pressurizer heaters arid opened the

spray valve to its maxixrmn mass flow rate. Innediately U-4 primary

pressure started a continuous decrease. TIhe reactor scramed 117

seconds after the failure of the pressure control systemi by the

overta.eprature protection system. The reactor trip further

decreases the pressure causing the injection of safety water 10

secnds after reactor trip.

-4-
3. MODEL DESCRIPTION

Figure 1 shows the mocdel used to simulate the plant. It consists 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 tin-.-

dependent volumes.

The usual practice of izrplenenting FPELAPS by homogenizing the multiple

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

not followd in this model because of the following reasons:

a) non - synr~mtric distribution of auxiliary feed-water am~ong the three

steam generators

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

c) Different number of plugged tubes in each steam gener ator.

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

break, small break in the primary circuit, one reactor colant pump

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

boundary conditions, in each loop.

e) Ccaputer 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 mocdeled by volume 120

and the proper heat structures. Volume 130 simulates the

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

Volume 1.40 mo~del the upper plenumn and volumes 150 and 160 the

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

-5-
pressurizer through volumre 510. The pressurizer is divided

Linto two voltzms.

The locer one (volum~ 520) i~s divided into five nodes. Heat

structures, sixrulating the pressurizer actual heaters, are

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

order to mod~el the junctions connecting the pressurizer with

the safety and relief valves and with the spray system.

Volum.r 420 mocdels the remaining of the hot leg. volumes 430

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

prirary 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 modxel the primary colant

pt~mp, proper hamlolous curves, given by the vendor, have

been used for this purp,.ise. Voltrnm 468 nrioels the Safety

Injection System.

b. - Secondary Systan

The mocdel of the secondary system staurts with Time Dependent

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

the steam generator. Volume 871 models the Auxiliary Feed

Water. The dcwncomer is simtlated by mreans of -omluaes 800,

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

volumres 806 and 807, and the remning -of the tubes zone by

volumres 808, 806, 809 and 810. The stearn separator with

volumre 820. Volumre 830, 840 and 850 nr:del the steam dryer and

the dczrc of the steam generator. Ste~ steam line starts at

volumre 880. Safety and relief valves (cmponents 886 and 884)

are connected to volume 881. Component Valve 885 moedels the

isolation valve. Tirie- Dependent Volumres 994 and 999 represent

the free atmrosphere. The steam is conducted throughout volumie

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

903 and 907 mo~del the by-pass to condenser valve, and the

turbine stop and control valves, respectively.

-6-
Proper heat struactures are used to connect thermally the

prirrary side of the steam generator wuth the secondlary si-de.

Actual values are used for all the var-iables except for the

hydraulic diamrter, heat transfer surface, and thenrml

conductivity of the tubes material where scae changes were

introduced in order to achieve the actual heat transfer rate

without any change in primary average terperature 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 mtodel /3/ was prepared using the RELAPS/Mod2 spaceindependent reactor kinetics option with data from the ANA Nu-

.clear Analysis Group. The mrodel includes a scrxn table of

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

actual value mreasured 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 mocdel from the data supplied for the

specific bur-up condition of each transient.

3.3 Control and Protection System Model

The protection and control systems were mrodeled using RELAPS/l4)D2

control blocks and following specific setpint studies, logical

diagrams~ and technical specifications of the plant /3/, /8/, 19/. The

mrodel includes the following systems:

-.7-
a. - Reactor Trip System

-1he reactor can be tripped in this ASCX7 model because of the

follow.ing effects:

- Lcw primaray pressure.

- High primrary pressure.

- Low speed at any pmp.-

- High pressurizer level.

- High reactor pc;'Aer.

- Low level at any steam generator.

- verte-a~xrat=re.

-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 Temerature.

- 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 reprodced. The

rnassflow' rate injected is mocdeled by mrans of the purnps

characteristic curves.

c .- Tu\rbine Trip and Control System

The position of the turbine control valve is controlled as a

function of the difference between the Required Powr and
Actual Power, with the proper control block to mocdel the act-ual logjic of the plant.

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

mocdeled. 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 mocdeled.

d. - Feed Water Control System

T~he feed water control systemn has been mrodeled as shown in

Figure 3. The rnassflo; rate calculated by the control system

is injected by m~ans of a tire dependent junction.

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

e .- Pressurizer level and pressure control system

The mocdel of the pressure control systen actuates upon

heaters, spray valve and charging pu.mp. Pressurizer safety and

relief valves and level control systemi are also simulated.

f.- Steam Dua control system

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

is represented.

g. - Average Tenprature Control System

The average torperature control system mode led with

PEWA5/MOD2 is shown in Figure 5. As can be observed this

system controls both the primary average temperature and the

prrimary-secorKdary power mistmach.

-9-
other svstc--s nrr'e led are:

h.- Stean-line Isolation logic.

i. - Main Feedwater Isolation logic.

j.- Safety and Relief valves of the secondary.

-10-
4.- STEADY STATE CALCULATION

Asteady state calculation was performed with the plant at 100% rated

condition. Tkhe objective is to obta~n a stable condition to start

transients.

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

given for the mrain plant variables. A catplete description of the plant

sensors and signals Is given in ref. 10.

-11-
5.- TRANSIENT CALCMlATION AND Ct4PARISON VERSUS ACTUAL DATA

The caniarison between model predictions (see table 5) against plant

data (see table 4) shows an overall grood agreement with some minor

disagreements that will be explained later.

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

As can be seen in figures .6 to 18 the transient which begun at second

100, starts with an unnoticed opening of the spray valves to its fully

open position (figure 13). This causes a continuous decrease in primary

pressure (figure 6), reactor power because of the correspondent decrease

in moderator density, primary average temperature (figure 7) and

secondary pressure (Figure 10). This event produces a decrease in the

overtemperature setpoint signal that (see Figure 17) reaches the actual

delta-tenperature and trips the reactor. The trip takes place in the

mo~del at second 216.3 (see Table 6) while in plant it was at seoond 217,

but in both cases for the sane reason: overtemperature. The low pressure

setpoint of the protection system was close to be reached but it is not

the signal that trips the plant.

Figure 13 shows the pressurizer spray mass flow rate accordingly with

delta-P between cold leg and pressurizer (see figure 18).

After the trip a further decrease in plant pressure occur, because of

the power reduction and sate 10 seconds later, at second 236.*4, the high

Pressure Injection System becamre active injecting abouat 25 Kg/s into the

primary system (Figure 14) until the end of the transient.

Because of turbine 'trip the steam-dump valves open, controlled by

primra~ry average temperature, (see Figures 10 and 16) lowering secondary

pressure and in same seconds steam dumrp valves close. Afterwards the

continous decrease in secondary pressure is caused by steam extractions.

Control system causes main feed water trip (see Figure 15) and auxiliary

feed-water is injected only to steamn generators 1 and 3 (because of loop

2 disfunction).

-12-
The ca~parison in the behaviour of the main variables is explained below.

a.- Primary Pressure.

In figure 6 pressure at the hot leg is presented. As can be seen the

cariparison is good but the pressure decrease predicted by PREAP because

of the spray actuation (second 100 to plant trip), has a lower slope

than the actual one. It may be caused by a mo~re effective vapor

condensation in the plant, caused by the liquid dropplet, than in the

model1.

b. - Primary Temerature

In figure 7 mo~del primary average te-mperature with and without signal

proccesing is compared versus the plant value. In figure 8 model hot leg

temrperature of loop 1, without signal proccesing is ccarar~ed against

plant data.

It can be observed that when signal procces sing is introduced the

comarion is fairly good.

c.- Pressurizer level.

In figure 9 the caTparison between mo~del prediction and plant data is

given. The overall carparison is fair although a larger increase in

plant level is observed both before the reactor trip and when level

recovery starts. It may be due to limitations to the actual simrulation

of the surge line in the model.

d. - Secondary Pressure

In figure 10 the ccnparison of this variable versus actual data is

given. It can be observed that the short term behaviour of the steam

dump valves is not well reproduced in the model and a faster decrease of

secondary pressure than the actual one is obtained. The long term

behaviour is correctly predicted. This problem is being studied also

-13-
with other transients. Except for this point a good corrparison is

obtained.

e. - Steam Generators level.

In figure 11 the carparison between steam generator 1 narrow range level

(steam generator 3 behaviour is totally similar since auxiliary feed

water is injected at both steam generators) and actual plant data is

given.

Figure 12 shows the catparison for steam generator 2 which does not

receive auxiliary feed-water in this transient.

In both level predictions saue oscillations can be observed when na~rrow

range level decrease under about 30%. Sam diferent. nodalizaticns have

been tested before chosing the best one in order to minimize these

oscillations. The overall conarisons shoed in figures 11 and 12 ar

good, except for steam generator 2 during the last 40 seconds, where

plant data show a quicker decrease that has not yet an explanation.

-14-
6.- RUN STATISTICS

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

Fundacicn Leonardo Torres Quevedo located at Santander - Spain.

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

In Table 8 a Typical run statistics is presented.

-15-
7.- .)NCLUSIONS

The transient has been simulated with Relap5/zrod2 Asc6 model.

The results are in close agreement with plant data.

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

qualification matrix, provides the validation of the model.

The function of the pressurizer spray valve is correctly predicted by

the model.

The transient provides the assessment of the high pressure Safety

Injection, as well as the reactor protection system.

Although in long term, the actuation of the steam-di-mp seems to be

correctly predicted, in short term this actuation can be =nproved.

Some diferent steam generator nodalizations; have been tested before

chosing the best one in order to min~imize level oscilations. The level

predictions are, any way in fair agreement with actual data.

The mocdel of Asc6 using Relapsfnod.2 is a valuable tool to analyze plant

transients.

-16-
8. - __________

1) V. -H. RAYSON and R.J. WAZR "PELAPS/MC)D2: Cod~e Manual, "EGGSAAM-6377, E)G&G Idaho, Inc. (Apr. 2984).

2) F. REVENI1OS, J. S7ANCREZ-BAPTISTA, and P. MORENO, "Modelo

Te~rrhidr5Lulico de la C.N. Ascb", Asociaci6n Nuclear Asc6 (June

1987).

3) F. REVRMIS, J. SAHEZ-BA?1'ISTA, and P. MOlRENO, "Modelo

Termzhidraulico y de proteccibn y control de C.N. Asc6," Asociaci~x Nuclear Asc6 (May 1988).

4) F. REVENTOS, J. SANCEZ-BAPTISTA, and P. MORENO, "Analisis de

transitorios en A.N.A. con IELAP5/MOD2," presented at 14th Annual

Mtg. Sociedad Nuclear Espa~fiola, Marbella, Spain, October 26-28,

1988.

5) F. REVENIPOS, J. SA~NCHEZ-BAPTISTA~, P. 14)RENZ), and A. PEREZ NAVAS,

"Transient Analysis for ASCO Nuclear Power Plant Using

RE.AP5 /MDD2,1" Proc. Ist Int. F.AP5 Users' Seminar, College

Station Texas, January 31-February 2, 1989.

6) F. REvmOMS, J. SANCHEZ-BAPTISTA, A. PEREZ-NAVAS and P. MOJREN.

"Transient Analysis in the AsoB NPP using RELAPS/Mt'D2," Nuclear

Technology, voltme 90, Niunrer 3, pp 294-307. June 1990.

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

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

8) F. BA12ERINI, "Setpoint Study Asc6 Units I and 2," Westingt~se

Electric Corporation (m~ar. 1976).

9) "precautions, Lin-itations and Setpoints of Asc6 Units 1 and 2".

E/PS176/068. Rev. 11. Sep. 1988.

-17-
10) C. Simo~n, et al "Central Nuiclear de Asc6 Units 1 and 2. Emrgency

Recovery Guidelines Setpoint Values. Calculation and Methodology

Appendi.x D. Instrumentation channels statistical calculation of

Uncertainties". N=C 88-08. Rev. 2. October 1989.

-18-
TABLES

TABLE 1. DESCRIPTION~ OF THE MAIN CHARACTERISTICS OF AS00 I AND II

NUCLEAR STATION.

TABIBE 2. MAIN EVUMTIS THA TOOK PLACE DURIN~G TH TRZANSI=N.

TABLE 3. Cl1tlARISCN BETWEE FMEAP5IMZD2 VALUJES AND ACIIJAIJ DATA FOR

STEADY STATE.

TABLE 4. DESCRIPTION OF PLANT DATA MEASURIEN'T2TS.

TABLE 5.* DESCRIPTION OF FEAP5 fMOD2 VARLIABES.

TABLE 6. COMPARISON OF TkX CRONOIb3 OF THE MAIN EVENTS.

TABLE 7. RUN SEATISTICS.

-19-
- Electrical power 930 Mw

- Thez~al reactor pow.er 2686 Yh,7th

- Fuel U02

- Number of assemblies 157

- Fuel rod~s per fuel asserbly (17x17 - 25) = 264

- Active length of fuel rods 3.657 m,

- Ouatside diareter of fuel rods 4.75 x 10- i

- Cladding tube imaterial Zr -4

- Cladd~ing tube w-all thickniess 0.655x10-3 M.

- Average linear heat generation rate 17.2 Xw/m.

- Absorber rods per control assembly 24

- AZbsorber material Ag- in -Cd.

- Ntrrber of coolant loops 3

- Reactor operating pressure (pressurizer) 15.51 Npa.

- Coolant Average Terrerature 581.3 QK

- Coolant flow rate 14287 Kg/s

Table 1.- Description of the main c-haracteristics of Asc6 I and II

Nuclear Station. (1 of 3)

-20-
Steam Generator

- Numnber

- Height

- Diaxmter (Upper shell)

- Tube mrterial

- Average tube length

- Design pressure/tenperature

(steam plant side)

- Inner diameter of tubes

- outer dliazreter of tubes

Reactor coolant Punms

- Discharge head

- Design flow rate

- Speed

Westinghouse D-3

3

20.6 m

4.445 m

Inconel 600

15.94 m

8.17 1.2a/589 QK

l.687x10- M.

1.905x102 M.

Westinghouse 93-DS

86.25 m

5.928 m 3 s

155 rad/s

Table 1.- Description of the main characteristics of Asc6 I and ii

Nuclear Station. (2 of 3)

-21-
Pressurizer

- Height

- Diameter (inner)

- Volume

- 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 mn

39.64 Mn3

15.51 MPa

1. 40 Mwe

497.5 Kg/s/loop

1492 Kg/s

0.25%

497.05 OK

Table 1 - Description of the ma~in characteristics of Asc6 I and II Nuclear

Station. (3 of 3)

-22-
Second Event

0.0 Pressure control system failure.

117.0 Reactor Trip.

117.1 Low pressure signal.

127.0 Safety injection.

Table 2 - main events that took place during the transient.

-23-
VARIABLE

PRIMARY 1H.SS FLOWq RATE (Kgfs)

CORE BY-PASS MA.SS FLCW RATE(%

VESSEL DELTA-T (QK)

REACTOR POWER. (nq)

PRIMARY PRESSURE (MPa.)

PRIJ-7Y AVERAGE T~ERAýTtRE (QK)

RECII UIA TION RATIO

UP-STREA2,M FLOW RATIO

IN THE S.G. PRSHATER.

STEAM GEERATOR NARROW RANGE LEVEL

SBTO\NDARY PRESSURE (MPa)

STEAMv' OJLLECR PRESSURE 02~a)

STEN-M FASS FLCW RATE (Kgfs)

RELAP5/14OD2

14027.

2.71

33.24

2681.

15.50

581.1

2.286

PLANT DAT'A

14287.

2.71

33.26

2686.

15.51

581.3

2.29

.517

.66

6.808

6.705

1478.5

.520

.66

6.821

6.724

1492.0

* DESING DATA

Table 3 - Conparison between RELP5/rflD2 values and actual date for steady

state.

-24-
IDENTIFIC'IOR DESCRIPTION

T04 21A Average Temperature me~asured in loop 2.

T0424A Vessel Delta - Tepe-rature measured in loop 2.

F0442A Meassured Mass Flow Rate of loop 3.

P0499A Pressure at loop 2, hot leg.

P0484A Pressurizer pressure.

P0419A Teuperature at loop 2, hot leg.

N0482A Pressurizer level.

P0400A Secondary pressure, loop 1.

N0400A Level of the steam generator 1.

N0420A Level of the seteamn generator 2.

N0440A Level of the steam generator 3.

Table 4 - Description of plant data umeasurements.

-25-
VARIABIE VARIABLEDESCRIPTION

P220010000

T220010000

P520050000

1vk7M~J 100030000

CNTRLVAR 5

CNTRLVAR 26

CNTRLVAR 6

a.TERLVAR 25

C=RVAR 13

oRTPo1 0

MF10WJ 105

YFJ10001

MFJ10002

P681010000

MFThJ90 6

Pressure at the Hot leg of loop 2.

Temperature at the hot leg of loop 2.

Pressurizer pressure.

Mass Flow rate of loop 3, cold leg.

Average Tezmperature.

Proccessed Average Temperature.

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

Processed'Vessel Delta-T.

-Pressurizer level.

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-Dtxrp Mass Flow Rate.

Table 5 - Description of RELAP5/I4?D2 variables. (1 of 2)

-26-
MFThJ675 Main Feed Water Mass Flow Rate of Loop 1.

MFJ672 Auxiliary Feed Water Mass Flow Rate of Loop 1.

MFJ3772 Auxiliary Feed Water Mass Flow Rate of Loop 2.

CNTRLVAR 16 Steamn Generator 1 narrow range water level.

CNT~RLVAR 17 Steam Generator 2 narrow range water level.

Q'NTRLVAR 18 Steam Generator 3 narrow range water level.

YSW535 Pressurizer Spray Mass Flow Rate.

MFL7367 Safety Injection Mass Flow Rate.

CNTRLVAR 147 Mlain Feed Water Mass Flow Rate of Loop 1.

CNTRLVAR 36 Overtemperature Setpoint.

CNTRLVAR 47 Processed Vessel Delta-T (Reactor Protection System)

P36501 Cold leg Pressure

Table 5 - Description of RELAP5/DMD2 variables. (2 of 2)

-27-
Event

Pressure control system failure

Reactor trip (overtemperature)

Steam-Dump starts to open

Low pressure signal

Steam-Dump~ fully open

Safety injection

Low average temperature signal

Tinre (s)

Plant Relap

100.0 100.0

217.0 216.3

- 216.5

217.1 218.0

- 223.5

227.0 236.3

- 236.7

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

-28-
CCMPWtr~ ASSESSMEW

TRANSIENT TflE (s)

CP~rLT (s)

C (total niumiber of actives voluTies in the rodel)

DT (Total numb~er of timre steps)

CPU x 1000

C x DT

CPhMI/TRANSIENT TflE

Table 7 - Ruin Statistics.

C"YMM 180/830

290

37460

173

7600

28. 491

129.17

-29-
FIGURES

FIGURE 1. NODAIJIZATION DIAGRAM OF ASCO NUCLEAR POWER PULNr.

FIGURE 2. LOGIC OF THE REACTOR PROTECTION SYSTEM4.

FIGURE 3. FEED@.TER CONTROL SYS=E~.

FIGURE 4. STEAI4-DUMP CON~l!RL SYSTEMY.

FIGURE 5. PPfl'2R AVERAGE TEMERATURE CONTROL SYSTEM.

FIGURE 6. PRESSURIZER PRESSURE.

FIGURE 7. PRIMARY AVERAGE TEMPERATURE.

FIGURE 8. HOT LEG T2ERATURE.

FIGURE 9. PRESSURIZER LEVEL.

FIGURE 10. SECONDARY PRESSURE.

FIGURE 11. STEAM GENERATOR 1 NARRO RANGE LEVEL.

FIGURE 12. STEAM GENERATEOR 2 NARROW RANGE LEVEL.

FIGURE 13. SPRAY MASS FLCM RATE.

FIGURE 14. SAFETY WATER MASS FLW~ RATE.

FIGURE 15. FEE-WATER MASS FLOW RATE.

FIGURE 16. STEAM-DUMP MASS FLOW7 RATE.

FIGURE 17. VESSEL DELTA T PERATURE.

FIGURE 18. PRIMA~RY PRESSURE.

FR/ac/mb/jv

IST00282

27.12.90

-30-
NODALIZATION DIAGRAM OF

A3CO NUCLEAR POWER PLANT

JULY 1990

::4ZZ~j-'UZ~

"s on

CONDENSER

PPMUsInZrER

i-El

405
IHIGH PRESSURIZER LEVEL

PRESSURIZER LEVEL

NIGH RP

iREACTOR TRIP

SG2 LEVEL

SG3 LEVEL

SAFETY INJECTION

I RCP

FIGURE 2. -LOGIC, OF THE REACTOR PROTECTION SYSTEM4
REACTOR

POWER

STEAM4 FLOW LOOP 1

-TO FUI

MODULATION

-TO FV2

MODULATION

STEAM4 FLOW LOOP 2

kTION

STEAM FLOW LOOP 3

FEED-WATER TRIP SIGNAL

FIGURE 3. FEED-WATER CONTROL SYSTEM
-TO STEAM-DUMP

VALVES

FIGURE. 4 STEAM-DUM4P CONTROL SYSTEM
REACTOR

POWER

AVERAGE

TEMPERATURE

LAG

FIGURE. 5 - AVERAGE TEMPERATURE CONTROL SYSTEM
ANA

PRESS. SPRAY VALVE OPENING

18

17

w

(I)

w

N

U)

'in

a.

16

15

14

13

12

11

10

0 100 200 300 400

FIGURE 6

P520050000

TIME (sec)

+ P0484A
ANA

PRESS. SPRAY VALVE OPENING

600 T r

w

w

w

w

CL

590-

580 -

I-- & -I I

4- -+ - ~ 4 -1-

570

560

550

0 100 200 300 400

FIGURE 7 TIME (sec)

- CNTRLVAR 5 CNTLVA 5NTR0VAR2 0 T0421A
ANA

PRESS. SPRAY VALVE OPENING

620

610

600

ci

wd

2:,

I-.

590

580

570

560

550

540

0 100 200 300 400

FIGURE 8

TEMPF 22001

TIME (sec)

+ T0419A
ANA

PRESS. SPRAY VALVE OPENING

N

Fr

D,

;v)

(n

1

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0

0100 200 300

FIGURE 9 TIME (sec)

400

- CNTRI-VAR 13 N0482A
It

Cf,

Id

0

0

Id(n)

8

7.9

7.8

7.7

7.6

7.5

7.4

7.3

7.2

7.1

7

6.9

6.8

63.7

6.6

6.5

6.4

6.3

6.2

6.1

6

ANA

PRESS. SPRAY VALVE OPENING

++

___________________~~~ ___________________ _________ ~~~~ +_________ +_________ 

0 100 200

TIME (sec)

300 400

FIGURE 10

P681 010000 P0400A
ANA

PRESS. SPRAY VALVE OPENING

in

1

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0

0 100 200 .300 400

FIGURE 11

QNTRLVAR 16

TIME (sec)

+ N0400A
ANA

PRESS. SPRAY VALVE OPENING

C14

1

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0

0 100 200 300

FIGURE 12 TIME (sec)

400

- CNTRLVAR 17 N042-OA
ANA

PRESS. SPRAY VALVE OPENING

50

40

0

-J

LL

(n)

in

30

20

10

0

0 100 200 300 400

FIGURE 13 TIME (sec)

.MFLOWJ535
w

0

Li.

9

8

7

6

5

4

3

2

1

0

ANA

PRESS. SPRAY VALVE OPENING

0 100 200 300 400

FIGURE 14

MFLOWJ

TIME (sec)

36700
600

0~

0

0

w

0

-J

LL

U) in

w

Id

Li500

400

300

ANA

PRESS. SPRAY VALVE OPENING

200

100

0

0 100 200 300 400

FIGURE 15

CNTRLVAR 147

TIME (sec) - MFLOWJ 672
ANA

PRESS. SPRAY VALVE OPENING

600

500

0

K

0

-J

1±.

(I)

U)

0~

0

U)

400

300

2-0-0

100

0

.10 100 200 300 400

TIME (sec)

MFLOWJ 906

FIGURE. 16
ANA

PRESS. SPRAY VALVE OPENING

60

50

40

w

w

30

20

10

0

-10

0 100 200 .300 400

FIGURE 17

CNTRL-VAR 36

TIME (sec)

CNTRLVAR 47
(n

(L

15

14.8

14.6-

14.4

14.2

14

13.8

13.6

13.4

13.2

13

12.8

12.6

12.4

ANA

PRESS. SPRAY VALVE OPENING

160 180 200 220 240 260

FIGURE 18

-P52005

TIME (sec)

-P36501
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-891 IAuslgnod by NRC. Add Vol., Supp.. Rev.. NRCM 1102, and Addendum Numbers, if env.)I

3201, 3202 BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/IA-Ol 21

2. TITLE AND SUBTITLE ICSP-AS-SPR-R Assessment of a Pressurizer Spray Valve Faulty Opening Transient at ASCO Nuclear Power Plant with RELAP5/MOD2 3. DATE REPORT PUBLISHED

December 1993

4. FIN OR GRANT NUMBER

L2245

5. AUTHOR(S) 6. TYPE OF REPORT

F. Reventos, J.S. Baptista, A.P. Navas, P. Mor eno

7. PERIOD COVERED (Inclusive Dares)

8. PERFORMING ORGAN IZATION - NAME AND ADDRESS (IfNRC. providelDivision. Office or Region. U.S. Nuclear Regulatory Commission, and mailing address: If contractor, provide

JN11 milngadroucl ear Asco

C/Tres Torres, 7

0817 - Barcelona

Spain

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

and mailing eddressj

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

1D. SUPPLEMENTARY NOTES

11. ABSTRACT 1200 words or les)

The Asociacion Nuclear Asco has prepared a model of Asco Nuclear Power Plant using

RELAP *5/MOD2. This model, which includes thermaihydraulics, kinetics and protection

and controls, has been qualified in previous calculations of several actual plant

transients. One of the transients of the qualification proce~ss is a "pressurized

spray valve faulty opening" presented in this report. It consists in a primary

coolant depressurization that causes the reactor trip by overtemperature and later

on the actuation of the safety injection. The results ara in close agreement with

plant data.

12. KEY WORDS/DESCR!PTORS (List words or phtrises that will &asist resarcher in locating the report.) 13. AVAILABILITY STATEMENT

ICAP, ASCO, RELAP5/MOD2, Spray Valve unlimited

14. SECURITY CLASSIFICATION

(This Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-891
Federal Recycling Program

A3Z)1,ZI~LfN 4- UP A f'l{t~L11SUItE SP1HAY VALVE PAUIJY OJPENING

NUREC/IA-0121 TRANSIENT AT ASCO NUCLEAR POWER PLANT WITH RELAIP5/iNOD2

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

DECEMBER 1993

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

I