NUREG/IA-0105

International

Agreement Report

Assessment of RELAP5/MOD3

Version 5m5 Using Inadvertent

Safety Injection Incident

Data of Kori Unit 3 Plant

Prepared by

K. T. Kim, B. D. Chung, I. G. Kim, H. J. Kim

Korea Institute of Nuclear Safety

P. 0. Box 16

Daeduk-Danji

Taejon, Korea, 305-606

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

May 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

r-
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-0105

International

Agreement Report

Assessment of RELAP5/MOD3

Version 5m5 Using Inadvertent

Safety Injection Incident

Data of Kori Unit 3 Plant

Prepared by

K. T. Kim, B. D. Chung, I. G. Kim, H. J. Kim

Korea Institute of Nuclear Safety

P. 0. Box 16

Daeduk-Danji

Taejon, Korea, 305-606

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

May 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 is based on work performed under the sponsorship of The Korea

Advanced Energy Institute of Korea. The information in this report has been

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

between the United States and Korea (Agreement on Thermal-Hydraulic Research

between the United States Nuclear Regulatory Commission and The Korea Advanced

Energy Research Institute, May 1, 1986). Korea 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 Korea 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.
Abstract

An inadvertent safety injection incident occurred at Kori Unit 3 in

September 6, 1990 was analyzed using the RELAP5/MOD3 code. The event

was initiated by a closure of main feedwater control valve of one of

three steam generators. High pressure safety injection system was

actuated by the low pressure signal of main steam line.

The actual sequence of plant transient with the proper estimations

of operator actions was investigated in the present calculation. The

asymmetric loop behaviors of the plant was also considered by

nodalizing the loops of the plant into three.

The calculational results are compared with the plant transient

data. It is shown that the overall plant transient depends strongly

on the auxiliary feedwater flowrate controlled by the operator and

that the code gives an acceptable prediction of the plant behavior

with the proper assumptions of the operator actions. The results also

show that the solidification of pressurizer is not occurred and the

liquid-vapor mixture does not flow out through pressurizer PORV. The

behavior of primary pressure during pressurizer PORV actuation is

poorly predicted because the actual behavior of pressurizer PORV could

not be modelled in the present simulation.

iii

Executive Summary

The plant transients following an inadvertent safety injection

signal have a relatively high probability of occurrence in the

current Westinghouse design. It is well known that the severity of

the above event is not so great, however the effect of core

over-cooling and pressurizer solidification may become serious in

conjunction with the operator misoperation.

There are some experiences of events described above in the

Westinghouse 3-loop power plant in Korea. One of these event occurred

at Kori Unit 3 in September 6, 1990, which is an Westinghouse 3-loop

PWR rated at 900 NW(e) and was commissioned in 1985. The event was

initiated by a closure of main feed water control valve of one of the

steam generators (SO-B) while the reactor power was maintained to 83 X

full power. A SG-B low-low level signal generated by the deduction of

feedwater flow tripped the reactor. Although the turbine trip signal

was also generated by the reactor trip signal, reactor-turbine intertrip valve was stuck and consequently actual turbine trip was delayed.

During this period, the safety injection system was introduced by the

low pressure signal of main steam line and increased the primary water

inventory.

The event was simulated using the best estimated computer code,

RELAP5/MOD3 Version 5m5, with the proper estimations of operator

actions to identify the important thermal-hydraulic phenomena of

inadvertent safety injection and to demonstrate the code applicability

V
to the full scale plant analysis. There are three types of digital

records from plant computer in Kori Unit 3 ; the computer daily

logging sheet, pre-post trip review record, and sequence record of

events. However most plant data were lost due to the computer

malfunction during the transient, therefore calculational results were

compared with the plant analog strip chart record.

It was found that the code predicted well the plant behaviors,

with sufficient accuracy, indicating the code's capability to this

type of transients. The analysis shows that the core over-cooling was

prevented by terminating the safety injection, and the overpressurization due to the solidification of pressurizer was prevented

with the actuation of pressurizer PORV and injection of auxiliary

feedwater. The analysis shows also that the residual heat can be

removed by the safety injection and establishment of the natural

circulation. However there are some differences between the

calculational results and the plant data, particularly in the cold leg

temperature. It may be considered that this discrepancy was mainly

due to the temperature measurement errors in the low flow rate.

vi
Table of Contents

Abstract iii

Executive Summary v

Table of Contents vii

List of Tables viii

List of Figures ix

1. Introduction 1

2. Description of Plant and Accident Sequence 3

3. Code and Input Description 5

3.1. Code Description 5

3.2. Steady-State Calculation 5

3.3. Transient Calculation 7

4. Comparisons of Results and Discussions 10

4.1. Steam Generator Level 10

4.2. Steam Generator Pressure and Temperature 11

4.3. Behavior of Primary Side 13

5. Run Statistics 16

6. Conclusions 17

7. References 18

vii
List of Tables

Table 1. Design Specifications of Kori Unit 3 19

Table 2. Operating Parameters before Occation of the Accident 20

Table 3. Scram Table 20

Table 4. Loss Coefficients of Junctions in Steam Generator 21

Table 5. Comparisons of Initial Conditions 21

Table 6. Trip Set Point 22

Table 7a. Uncertainties of Indicators and/or Recorders

(Conditions II, II1, and IV) 23

Table 7b. Uncertainties of Indicators and/or Recorders

(Normal Operation) 23

Table 8. Comparisons of Sequence of Events 24

Table 9. Run Statistics Table 24

viii
List of Figures

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

FIG.

1.

2.

3.

4.

5.

6.

7a

7b.

7c.

8a.

8b.

8c.

9a.

9b.

9c.

10.

11.

12a

12b

12c

13a

13b

13c

Sequence of Events

Kori Unit 3 Nodalization

Reactor Power (Calculated)

Main Feed Water Flow Rate

Main Steam Line Flow Rate

Aux. Feed Water Flow Rate

(Calculated)

(Calculated)

(Calculated)

I'

I.

Comparison

Comparison

Comparison

Comparison

Comparison

Comparison

Comparison

Comparison

Comparison

Comparison

Comparison

* Comparison

* Comparison

Comparison

* Comparison

Comparison

of

of

of

of

of

of

of

of

of

of

of

SG-A Level with Plant Data

SG-B Level with Plant Data

SG-C Level with Plant Data

SG-A Pressure with Plant Data

SG-B Pressure with Plant Data

SG-C Pressure with Plant Data

SG-A Temp. with Data

SG-B Temp. with Data

SG-C Temp. with Data

PZR Pressure with Data

PZR Level with Data

25

26

27

28

29

30

31

32

33

34

35

36

37

38

39

40

41

42

43

44

45

46

47

of

of

of

of

of

T

T

T

T

T

T

Hot (A) with Data

Hot (B) with Data

Hot (C) with Data

Cold (A) with Data

Cold (B) with Data

0 Comparison of Cold (C) with Data

ix
FIG. 14a. Comparison of T Cold (A) and SG-A Temp. 48

FIG. 14b. Comparison of T Cold (B) and SG-B Temp. 49

FIG. 14c. Comparison of T Cold (C) and SG-C Temp. 50

FIG. 15. Integrated Flow of PZR PORV (Calculated) 51

FIG. 16. Safety Injection Flow Rate (Calculated) 52

FIG. 17. Total Required CPU Time 53

x
1. Introduction

The RELAP5 code has been assessed with a great number of data of

the separate and integral effect test facilities, but the assessments

using the data of full scale power plant were scarce. The resulting

data from real power plants can be used to eliminate the scaling

problems in small scale test facilities. However the plant data are

scarcely available and have great uncertainties compared with the well

instrumented test facility data.

The plant transients following an inadvertent safety injection have

a relative high probability of occurrence in the current Westinghouse

design and the above incidents are considered to be American Nuclear

Society (ANS) Condition II events. We have some experiences of these

events in the Westinghouse 3-loop nuclear power plants in Korea. One

of these events occurred at September 6, 1990 in Kori Unit 3, which is

a Westinghouse 3-loop PWR rated at 900 NW(e) and was commissioned in

1985. The event was initiated by a closure of main feed water control

valve of one of the steam generators while the reactor power was

maintained at 83% full power. A SG-B low-low level signal generated

by the deduction of feedwater flow tripped the reactor. Although the

turbine trip signal was also generated by the a reactor trip signal,

reactor-turbine inter trip valve was stuck and thus actual turbine

trip was delayed. During this period the safety injection was

actuated by the low pressure of main steam line and increased the

primary water inventory.

-1-
Such incident result the core over-cooling and pressurizer

solidification, thereby the operator should determined if the safety

injection signal should be blocked. For a spurious occurrence, the

operator would stop the safety injection and maintain the plant in the

hot shutdown condition. It is well known that the severity of the

above event is not so great, however the effect of core over-cooling

and pressurizer solidification may become serious in conjunction with

the operator misoperation.

According to the actual sequence of plant transient, the event was

simulated using the best estimated computer code, RELAP5/m1D3 [1]

Version 5m5, with the proper estimations of operator actions to

identify the important thermal-hydraulic phenomena of inadvertent

safety injection and check a applicability of the code for the

full-scale plant analysis.

Description of the Kori Unit 3 plant and sequence of events are

presented in Chapter 2. The plant simulation model and input

description is presented in Chapter 3. Results and discussions are

presented in Chapter 4. Conclusions are presented in Chapter 6.

-2-
2. Descriptions of Plant and Accident Sequence

Kori Unit 3 nuclear power plant in which an inadvertent safety

injection event occurred is located on Yangsan, Kyoungnam, Korea.

Kori Unit 3 is a 900 MW(e) 3 loops Westinghouse PWR.

Kori unit 3 has been in commercial operation since Sep. 30, 1985.

The major design specifications are summarized in Table 1.

Accident happened at 15:35 Sep. 5, 1990. Before the accident

occurred, Kori unit 3 was operated in 83% of full power and 800 1IW(e)

of turbine generator to adjust the time of refueling. The major

operating parameters are presented in Table 2.

Sequence of events and major plant data were recorded by the main

computer. However a large portion of digital data was lost due to the

malfunction of computer. From the available records[2, 3, 4] of the

sequence of event, the event can be summarized as Figure 1.

Based on the limited operation records and strip charts of plant

parameters, the accident sequence is reconstructed as follows. While

operated in 83% of full reactor power, loss of power for 7300 PCB

Group 3 Frame 2 happened due to destruction by overheat. This led to

loss of control for SG-B main feedwater control valve, which is an

fail-close valve. Level of SG-B began to decrease. When the narrow

range level reached to 17%, SG lo-lo level signal caused to trip the

reactor.

Although turbine trip signal was generated by reactor trip signal,

reactor-turbine inter-trip valve was stuck and thus turbine trip was

-3-
delayed. During this period, steam in SGs was flowing out to turbine.

SGs were depressurized until main steam isolation valves (MSIVs) were

closed. Depressurization of steam line actuated safety injection

signal and main steam isolation signal.

After safety injection signal, the operators tripped turbine

manually and opened PCB-7372 and PCB-7300 to protect generator.

Because of opening of PCB-7372 and PCB-7300, automatic power transfer

to off-site power following generator trip was failed. Thus off-site

power was lost for 3 minutes until operator manually transferred the

switch to off-site power. At this time the reactor coolant pumps

(RCPs) were failed due to a trouble of 480 V Non-lE diesel generator.

Safety injection was supplied for about 12 minutes. Aux. feedwater

was also supplied to SGs sufficiently. Without RCPs, heat transfer to

SGs maintained by a natural circulation.

Restarting RCP-A at about 38 minutes after accident, the plant was

begun to be recovered to hot shutdown condition.

-4-
3. Code and Input Description

3.1. Code Description

The RELAP5/MOD3 Version 5m5 is used for the simulation of the

event. There are no model change of the code, however a minor

correction is done to overcome an input processing failure with tue 3

pump controllers for self initialization control. The incorrect

coding on the line 139 of subroutine 'rssi.F', 13a(7) = 0, is replaced

with 13a(7) = 2.

3.2. Steady-State Calculation

Steady-state calculation is conducted to obtain the major operating

parameters which matches to the pre-accident conditions. Since SI

signal is actuated by secondary side, the calculation is aimed at the

obtaining of the secondary side conditions, especially, steam

generator level.

Nodalization of the plant is shown in Figure 2. In nodalization,

the whole system was divided into 227 volumes, 274 junctions, and 289

heat structures. The components are described in the following

paragraphs.

The plant is nodalized in 3 loops in order to simulate asymmetric

effects between loops. 3 loops are connected to a lower plenum

component through 3 downcomer components, which interconnected with

-5-
cross flow junctions. Lower plenum component is connected to upper

plenum component through core component and core bypass channel

component.

There are two heat structures in the core component to simulate the

fuel rods; one is for the average fuel rods, the other for the hottest

rod. Heat generation from the two heat structures is 83% of full

power. Scram table is used as shown in Table 3. ANSI/ANS-5.1-1979

decay heat model[3) is used in this calculation.

Pressurizer (PZR) component is connected to loop-B. In the

steady-state calculation, PZR component is also connected to

time-dependent volume of 15.51 KPa in order to obtain the steady-state

system pressure. This volume is disconnected in the transient

calculation. Pressurizer heaters and spray are not modeled. PZR PORV

(Power Operated Relief Valve) component is modelled and described in

the following chapter.

Upper plenum component is connected to each hot leg of three loops,

and to upper head component through downcomer upper annulus component

and guide thimble component.

Pump components are used. The characteristic curves of RCP(Reactor

Coolant Pump) are given for four regions. Pump controllers determine

speeds of each pump components with the parameters; 336.85 of gain,

1.0 of proportional time constant, 10.0 of integral time constant, and

4582.16 kg/sec of the desired flow rate.

Feedwater of 7.07 IPa and 226.7 C is-supplied to steam generators.

The feedwater controllers determine each feedwater flow rates. The

-6-
given control parameters are as the followings; 1.0 of proportional

time constant, 5.0 of integral time constant, 0.5 of the desired

level, 0.05 of scale factor Sk, and 10.0 of scale factor Sm.

Flowing the riser component from downcomer component to separator

component, feedwater is vaporized. In separator component, separated

water flows downward to downcomer component and steam flows upward to

steam plenum component. Table 4 shows loss coefficients of several

junctions in steam generator, which results in the recirculation

ratio, 3.95.

Steam within steam plenum component flows to common header

component through steam dryer component, dome component, main steam

isolation valve component, and steam line component. Common header

component connects turbine component and three steam line components.

Because of the characteristic of the feedwater controller,

feedwater flow rate has matched with steam flow rate after a

considerably long calculation. Moreover, 0.2 sec is given as the

maximum time step because a numerical oscillation in separator

component was occurred at the lager time step size.

The results of steady-state calculation are summarized and compared

with the plant data in Table 5.

3.3. Transient Calculation

At 10 sec of transient calculation, accident is initiated by

closing the main feedwater control valve of loop-B. When the SG

-7-
narrow range level decreases to 17%, reactor trip signal is generated

to scram the reactor after 2 sec of signal delay.

Safety injection signal is generated when one of three steam lines

is depressurized below 4.13 fPa. In this case, steam line pressures

are compensated with lead/lag controller (50/5). Time delays of 2 sec

for SI signal and 10 sec for SI pump start are considered.

Although the safety injection flow rate is calculated by the pump

characteristic curve, it is adjusted by multiplying 2/3 based on the

operator's records. It is injected for 795 sec until the operator

terminates.

HFIV and ISIV are closed in 5 sec after signal. Time delays of 5

sec and 10 sec after SI signal are considered for NFIV and ISIV,

respectively.

Pressurizer PORV is opened when pressure is increased to 100 psi

(689 kPa) above the reference pressure and closed when decreased to 80

psi (552 kPa) above the reference pressure. Pressurizer pressure is

compensated by a proportional-integral-differential controller with

the parameters; 2249.7 psia (15.5 KPa) of the reference pressure, 5.0

of proportional gain, and 900.0 of reset time constant.

Motor-driven auxiliary feedwater pumps are actuated by one SG

low-low level signal. 60 sec of response time and 10 sec of pump

delay is considered. Pumps perform their maximum capacity in 10 sec

from start. A turbine-driven auxiliary feedwater pump is actuated by

two SGs low-low level signal. 60 sec of response time and 2 sec of

pump delay are also considered. Pump performs its maximum capacity in

-8-
10 sec from start. 7.31 kg/sec of steam is extracted from SG-A and B

to simulate more actually the plant.

Auxiliary feedwater system can be controlled by the operator in the

accident situation. And there might be several operations to control

auxiliary feedwater flow rate. Although those operations are not

recorded, those may be deduced based on the operating parameter

records as the followings.

(1) For SG-A, auxiliary feedwater flow rate is controlled above 55% of

SG level to match with steam flow rate to the turbine-driven auxiliary

feedwater flow rate.

(2) For SG-B, auxiliary feedwater flow rate is reduced to 8.85 kg/sec

above 30% of SG level.

(3) For SG-C, it is reduced to 5.44 kg/sec above 10% of SG level. The

above deductions are reflected in this calculation.

Trip setpoints described in this chapter are summarized in Table 6.

-9-
4. Comparisons of Results and Discussions

The calculation is conducted for 3600 sec of problem time. The

calculational results are compared with the plant data. However, the

plant data have large uncertainties as shown in Tables 7a and 7b[5].

Furthermore, the large time shifts exist in the records since the pens

are positioned on the recorders to be able to cross one another.

Therefore, the analysis was performed to access a code performance

based on the comparison of trends rather than comparison of absolute

values.

The major calculated events are summarized and compared with the

plant data in Table 8. The timings of reactor trip and safety

injection signal are earlier than those of the plant data. It seems

that those discrepancies come from the difference between the time

delay considered in the calculation and the actual time delay.

After reactor trip, reactor power rapidly decreases to decay power

level as shown in Figure 3. Main feedwater and main steam flow rates

rapidly decrease by closing the isolation valves and do not flow as

shown in Figures 4 and 5. Major parameters along with the accident

sequence are discussed in the following sections.

4-1. Steam Generator Level

As mentioned previously, it was inferred that the operator

controlled the auxiliary feedwater flow rate according to the SG

- 10 -
levels. Estimated auxiliary feedwater flow rate is shown in Figure 6

as a function of time. The calculational results for SG levels are

compared with the plant data in Figures 7a, 7b, and 7c. The direct

comparison of calculated values with the plant data may be meaningless

because there were many unrecorded operator actions for the recovery

of SG levels, and thus the resulted plant data was used for the

simulation to estimate the operator actions. The major difference

between the calculation and the plant data is that S6 levels do not

drop below '0' in the initial period of the plant data. It is guessed

that levels are not correctly recorded because of the dead band of

narrow range level gauge.

Although there are a little time shifts and discrepancies after RCP

re-running, the general behavior of water level is well agreed with

the plant trend. Therefore it is inferred that the operator actions

on the auxiliary feedwater control are reasonably considered.

4-2. Steam Generator Pressure and Temperature

The calculational results for SG pressure are shown in Figures 8a,

8b, and 8c. At the beginning of the event, a low-low level signal due

to a sudden decrease of main feedwater actuates reactor trip.

Normally the reactor trip causes the turbine trip, but in this

transient the turbine stop valve does not closed. The SG pressure

decreases following the excess outgoing of steam and reduction of void

generation due to the reactor trip. Although the actual pressure was

- 11 -
not reached to the safety injection set-point (Steam Line Low

Pressure; 4.14 Mpa), the pressure signals led by lead-lag controller

generate the safety injection signal which close the hain Steam

Isolation Valves (lSIVs).

SG pressures rapidly increase as 1SIVs close, and then slowly

decrease as the supply of the auxiliary feedwater reaches its capacity

so that the secondary heat removal begins to overcome the reactor

decay power. PORVs (Power Operated Relief Valves) are simulated to

open at 7.76 Npa. The anlysis shows that the supply of the auxiliary

feedwater alone can provide the sufficient heat removal capability

without the operation of the PORVs.

While safety injection water cools down the primary side and

auxiliary feedwater is supplied to cooldown the secondary side, SG

pressures gradually decrease. This decrease is continued until safety

injection is terminated and then SG pressures are remained constantly.

After about 1800 sec, SG pressures increase gradually because of the

step-wise reductions of auxiliary feedwater flow rate. After RCP-A

re-running at 2376 sec, reactor coolant flows through loop by the

forced circulation. The coolant flow leads to the increase of heat

transfer in SG. The rate of pressure change increases as shown in the

figures. The overall trend in the calculated secondary pressures is

in good agreement with the plant data. The discrepancy on the

magnitude may be resulted from an estimation of operator action and

assumption of the closing characteristics of isolation valves.

The change of temperatures are shown in Figures 9a, 9b, and 9c. It

- 12 -
shows that the steam in SGs is under a nearly saturated condition and

thus the temperature behavior is almost identical to the pressure

behavior.

4-3. Behavior of Primary Side

With the excess cooling of steam generators due to the failure of

turbine stop valve closure following the reactor trip, the pressure

and temperature in primary side decrease rapidly as shown in Figures

11, 1Z, 13, and 14. Immediately after the reactor trip, the hot leg

and cold leg temperature decrease rapidly. After the steam flow was

isolated by SI signal, pressure and temperature turn to increase. The

flow coastdown due to RCP trips and the decay heat increase the

hot-cold leg difference until the establishment of the natural

circulation in the loop. And then RCS temperature gradually decreases

because of being cooled down by safety injection and auxiliary

feedwater injection. Safety injection flow rate is determined by the

characteristic curve of safety injection pump, and thus varies

according to RCS pressure as shown in Figure 17.

As reactor coolant contracts due to the decrease of temperature,

pressurizer level is decreased. However, as the inventory of reactor

coolant increases due to safety injection, pressurizer level stops

decreasing and turns to increase according to the amount of safety

injection. At this time, pressure also increases to the PORV setpoint

according to the increase of the reactor coolant inventory, and PZR

- 13 -
PORV is opened. According to the plant data, it is found that PZR

PORV was in the cycling of open/close. PORV was sometimes opened at

higher pressure and closed at lower pressure than the setpoint of

open/close. It may be resulted from the valve degradation during the

process. In the simulation, the above valve degradation was not

modeled because of it's complexity.

The calculated RCS cold leg and SG temperatures are compared with

the plant data as shown in Figure 15. The calculated RCS cold leg

temperatures are almost the same as SG temperatures because the heat

is exchanged completely. But, in case of the plant data, RCS cold leg

temperatures do not match with SG temperatures, because the

temperature measurement in the loop of low flowrate is not reliable.

As safety injection signal is resetted and safety injection is

terminated at 861 sec, the increasing rates of RCS hot leg

temperatures are blunted and stabilized as shown in Figures 13a, 13b,

and 13c. This, in turn, influences to SG pressure to be stabilized.

As the RCS inventory increases no more, PZR level does not increase

and is stabilized. As pressure gradually decreases along with heat

transfer to the secondary side, PZR pressure decreases below the PORV

setpoint and then PORV is closed. There is no more flow through PORV

as shown in Figure 16.

As the auxiliary feedwater flow rates are controlled to decrease

step by step, RCS temperature increases gradually and reactor coolant

swells. Thus, in turn, PZR level increases again. According to the

increase of PZR level, PZR pressure increases to open PORV again.

- 14 -
Analysis shows that PZR did not become a solid state and the liquid

did not flow out through PORV. Following the restarting RCP at 2331

sec, the operators begin to recover the plant and keep the reactor at

the hot shutdown condition by normal charging and letdown. There was

no record of operator actions during the recovery phase and thus the

simulation was terminated at 3600 sec.

In this study, there are many assumptions or estimations of

operator actions and the overall plant trend depends strongly on the

operator actions. Moreover the plant digital data were lost during

the major events due to the malfunction of data logging computer.

However the calculated trends with the proper assumptions of the

operator actions are generally agreed with the analog strip chart for

the plant trend record, except the loop temperature.

- 15 -
5. Run Statistics

In order to compare the run times with other organization, it

should be mentioned that the computer type used is MIPS 32-bit

workstation and the operating system is one of UNIX System V, RICSos.

The time step sizes and other run statistics are summarized in Table

9. The requested maximum time step size is 0.5 sec in this work.

Because of Courant limit, time step sizes become smaller after RCP

rerun at 2376.5 sec than before. The total CPU time is shown in

Figure 17 as a function of transient time.

The total CPU time required for simulation of the whole transient

3600 sec. is 42092.4 sec. The input processing time is 36.75 sec. and

the number of volumes is 227. The number of time steps is 37882. Thus

the grind time is 4.89 milli seconds (the grind time = (total CPU time

- input processing time) / (number of time step x number of volumes) =

(42092.4 - 36.75) / (37882 x 227)).

- 16 -
6. Conclusions

An analysis of an inadvertent safety injection event for Kori Unit

3 was carried out using RELAP5/MOD3 Version 5m5 on HIPS RC-3240

workstation. The results are compared with the available plant data

to assess the code applicability. The major conclusions are as the

followings.

(1) The calculational trends well agree with the plant data, except

the loop temperature.

(2) There are some differences between the calculational result and

the plant data. It is due to the uncertainties of the instruments and

the estimations of the operator actions.

(3) The asymnetric effects of the plant can be simulated by nodalizing

the plant into three loops. The calculation predicts well the

asymmetric effects if the predictions of the operator actions are

proper.

(4) The characteristics of valve degradation such as pressurizer PORV

are recognized to be important in the transient analysis of pressure

behavior.

Generally speaking, RELAP5/MOD3 code has the capability of

simulating the safety injection incident following the operator

actions.

- 17 -
7. Reference

1. "RELAP5/MOD3 Code Manual (Draft)," NUREG/CR-5535, EGG-Z596, Aug.

1991.

2. Computed Daily Log Sheet of Kori Unit 3, on 6 September 1990.

3. Computed Post Trip Review Sheet of Kori Unit 3' on 6 September

1990.

4. Computer Sequence of Events Record of Kori Unit 3, on 6 September

1990.

5. "Final Safety Analysis Report, Kori Nuclear Power Plant Unit 3/4,"

Korea Electric Power Company, 1985.

6. "Emergency Operating Procedure, Kori Nuclear Power Plant Unit 3/4,"

Korea Electric Power Company, 1990.

- 18 -
Table 1. Design Specifications of Kori Unit 3

Items Specifications

1. RCS

- core power 2775.0 lWth - fuel assembly 17 X 17 standard - number of FA 157 assemblies

2. Loops

- number of loops 3 loops - pump power: 10 lWth / 3 RCP's

3. Pressurizer

- Pressurizer spray

- Heaters Backup Heaters/

Proportional Heaters - PORV 3 units - SRV 3 units

4. Steam Generator Model F Type - Steam line / SG 1 unit

Flow restrictor/SG 1 unit - Number of U-tubes 5626 - Swirl vane moisture 3 units

separator - Shevron type dryer 1 unit

- MSIV 1 unit per SG - MFIV 1 unit per SG

Items Specifications

5. CVCS

- Charging Pumps 3 units - PZR Level Control Automatic letdown

and charging control

6. ECCS

- Accumulator per Loop 1 unit - HPSI pumps 3 units

(Charging pump) - LPSI pumps 2 units

(RHR pumps)

7. Aux. Feed Water Pump - Motor-Driven 2 units - Turbine-Driven 1 unit

- 19 -
Table 2. Operating Parameters before Occation of the Accident

Parameters Conditions

Reactor Power 83% of Full Power

Turbine-Generator Power 800 IW(e)

Levels of Steam Generator 50%, 50%, 50%

Boron Concentration 258 ppm

Control Rod Position D-Bank 172 Steps

RCS Temperature 306.2 C

PCS Pressure 157 kg/cm2

Table 3. Scram Table

Trip Time (sec) Reactivity ($)

0.0 0.0

0.2 -0.09667

0.4 -0.3664

0.6 -1.2818

0.8 -6.00514

1.0 -7.7259

1.2 -8.5833

1.4 -9.0077

1.6 -9.9426

1.8 -9.3966

2.0 -9.5202

2.2 -9.5703

2.4 -9.667

2.0e20 -9.667

- 20 -
Table 4. Loss Coefficients of Junctions in Steam Generator

Loss Coefficient

Junction # From Vol. To Vol. Foreward Reverse

315 310 320 10.0 10.0

330 330 340 0.45 0.45

330 330 305 0.1 0.1

330 320 330 0.1 0.1

Table 5. Comparisons of Initial Conditions

Parameters Plant Data Simulated Remark*

Reactor Power 83% of Full Power 83% of Full Power

PZR Pressure 157. kgf/cmZ 158.19 kgf/cm2 800-08

(2250.1 psia)

PZR Level 53.0 % 54.95 %

T(hot) 320.7 C 321.5 C 200-01

T(cold) 291.7 C 292.1 C 240-02

S/G Pressure -70. kgf/cm2 66.58 kgf/cm2 330-01

(947.01 psia)

Feed/Steam Flow -1.6e6 kg/hr 1.54e6 kg/hr

(979.8 lb/sec) (941.13 lb/sec)

S/6 Level 0.50 0.50

Note: * volume number of calculated value

- 21 -
Table 6. Trip Set Point

Trip Parameter Set Point

MFCV-B Close Event Initiation 10 sec

Rx Trip 1/3 SG lo-lo Level 17%

SI 1/3 Steam Line lo-P 599.7 psia

(50/5 lead-lag)

SI Reset Manual 805 sec after SI

MFIV, MSIV SI Signal

Aux Feed

Motor-Driven 1/3 SG lo-lo Level or SI 17%

Turbine-Driven 2/3 SG lo-lo Level 17%

RCP Manual 51 sec

after Rx Trip

PORV (open) PZR Pressure 100 psid (PID)

(close) PZR Pressure 80 psid (PID)

(Pref = 2235psig)

Charging Manual 2324 sec

& Letdown after Rx Trip

- 22 -
Table 7a. Uncertainties of Indicators and/or Recorders

(Conditions 11, 111, and IV)

Indicators/Recorders Range Accuracy

T hot (wide range) -17.8 - 371.1 C ±4%

T cold (wide range) -17.8 - 371.1 C ±4%

PZR Level entire distance +35%

between taps

Primary System 0.0 - 20.7 MPa 8%

Pressure

Steam Line Pressure 0.0 - 8.96 KPa 14%

Steam Generator 0.0 - 100.0 of Span ±35%

Water Level

(wide range)

Aux. Feedwater 0.0 - 144.2 m3/hr ±5%

Flow Rate

Table 7b. Uncertainties of Indicators and/or Recorders

(Normal Operation)

Indicators/Recorders Range Accuracy

T hot (wide range) -17.8 - 371.1 C ±4%

T cold (wide range) -17.8 - 371.1 C ±4%

PZR Level entire distance ±3.5%

(at 15.5 MPa) between taps

Primary System 0.0 - 20.7 MPa ±4%

Pressure

Steam Line Pressure 0.0 - 8.96 MPa ±4%

Steam Generator +7 to -5 ft from ±4%

Water Level nominal full load

(wide range) level

Aux. Feedwater 11.4 - 181.7 m3/hr ±3%

Flow Rate

- 23 -
Table 8. Comparisons of Sequence of Events

Event Plant Data Simulated

1IFCV-B Close 0:10 0:10

Aux Feed (MiD) Actuation Signal 0:51 0:42.8

Rx Trip 0:52 0:44.8

SIS 1:03 0:56.2

Aux Feed (TD) Actuation Signal 1:09 1:05.5

RCP Trip 1:43 1:31.8

Aux Feed (MD) (60+10 sec Delay) 1:54.8

Aux Feed (TD) (60+2 sec Delay) 2:07.5

PORV Open

SI Reset 14:28 14:21.25

Aux Feed-C Closed 15:20.5

Aux Feed-A Partially Closed 22:15.25

Aux Feed-B Partially Closed 28:31.55

Charging & Letdown 39:36 39:29

Aux Feed All Closed 39:39 39:32

RCP Run 39:43 39:36

Table 9. Run Statistics Table

Transient CPU Time Time # of Primary Secondary

Time(sec) (sec) Step Cycle Mass Error Mass Error

Size

(sec) (kg) Fraction (kg) Fraction

0.0 36.8 0.02 0 0.0 0.0 0.0 0.0

150.0 1958.9 0.5 1848 236.4 1.36e-3 67.8 4.02e-4

650.0 6673.9 0.25 6097 221.4 1.l1e-3 498.1 2.20e-3

1150.0 12033.5 0.25 10728 224.6 1.15e-3 984.0 3.86e-3

1650.0 15783.4 0.25 14009 220.9 1.13e-3 1588.3 5.75e-3

2150.0 19068.7 0.125 16991 230.7 1.18e-3 2972.1 1.05e-2

2200.0 19349.0 0.25 17244 230.4 1.18e-3 3202.1 1.12e-2

2300.0 19996.3 0.5 17834 231.3 1.19e-3 3544.7 1.24e-2

2376.5 20504.5 0.25 18297 229.8 1.18e-3 3853.1 1.34e-2

2876.5 29319.0 0.0625 26306 4.3 2.26e-5 79.8 2.78e-4

3376.5 38149.6 0.0625 34306 9.4 4.99e-5 157.2 5.49e-4

3600.0 42092.4 0.0625 37882 0.9 4.97e-6 176.3 6.15e-4

-24-
00:34 TBN manual trip 00:47

01:35 Generator manual trip 01:36

zation

n

I13:36

38:43

. 38:47

38:50 * Times (min:sec).are from

the reactor trip.

Figure 1. Sequence of Events

- 25 -
I , - F

Eli I in 13VO fi

//7 Heat Structure • .l

-~-Cross Flow

FIG. 2. Kori Unit a3 Nodlization
KOR 1 -3 INADVERTENT SI ACCIDENT (RELAP5/MOD3) 3.0

2.5

2.0

LU

09.

C)

1.5

I

--4

i 1.0

0.500

0 I~ 1i~~ Ii

0 500 1000 1500 2000 2500 3000 3500 4000

T IME (SEC)

FIG. 3. Reactor Power (Calculated)
KOR 1 -3 INADVERTENT SI ACCIDENT (RELAP5/MOD3) 600

500

cnf

LLI

0'

400

300

200

100

0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 4. Main Feed Watern Flow Rate (Calculated)
KORI-3 INADVERTENT SI ACCIDENT (RELAP5/MOD3) 500

400

Lii

LUV.0

0 1

L LI t--

0

_J

L300

200 1

CI1

100

0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 5. Main Steam Line Flow Rate (Ca I cu I ated)
KOR 1 -3 I NADVERTENT SI ACCIDENT (RELAP5/MOD3) 50

40

UI

w

e-n

LU

F0

_I

L30

20

10

0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 6. Aux. Feed Water Flow Rate (Cal cul a ted)
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON)

0

LU

I1.0

0.900

0.800

0.700

0.600

0.500

0.400

0.300

0.200

0.100

0

I

CA

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 7a. Comparison oF SG-A Level with Plant Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON)

L)

LL

LLJ

LLU

- .1

1.0

0.900

0.800

0.700

0.600

0.500

0.400

0.300

0.200

0. 100

0

I~

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 7b. Comparison of SG-B Level with Plant Data
KOR]-3 INADVERTENT SI ACCIDENT (COMPARISON) 1.0

C

0

C-)

L

LL

Ii -J

._J

0.900

0.800

0./00

0.600

0.500

0.400

0.300

0.200

0.100

1

0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 7c. Comparison of SG-C Level with Plant Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON)

LU

Of

LU

0-

8.0

7.0

6.0

5.0

4.0

3.0

2.0

1.0

0

CA,

0 500 1000 1500 2000 2500 3000 3500 4000

T IME (SEC)

FIG. 8a. Comparison of SG-A Pressure with Plant Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON)

CLL

cf)

LLU

CL

8.0

7.0

6.0

5.0

4.0

3.0

2.0

1.0

0

co) I

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 8b. Comparison of SG-B Pressure with Plant Data
KORI1-3 INADVERTENT SI ACCIDENT (COMPARISON)

LLU

:D

Uc)

00

w.Ofc

n~

8.0

7.0

6.0

5.0

4.0

3.0

2.0

1.0

0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 8c. Comparison oF SG-C Pressure with Plant Data
KOR I-3 INADVERTENT SI ACCIDENT (COMPARISON) GOO

0j)

-U

-u

LLI

I---

.<r

or"

LLI

C)

L2: LLI

F--.

580

560

520

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 9a. Comparison of SG-A Temp. w i th Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 600

LIJ

F580

560

540 00

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 9b. Comparison oF SG-B Temp. I th Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 600

0)

Ld

HLLJ

0-

M:

580

560

540

520

!-.

(I

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 9c. Comparison of SI-C Temp. w i th Data
KOR I-3 INADVERTENT SI ACCIDENT (COMPARISON) 20

18

16

14

12

- RELAP5/MO3 .....Plant Data

no

0r

LLU (': 12-

f

10

8.0

6.0

4.0

2.0

0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 10. Comparison oF PZR Pressure with Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 1.0

c- 0

Cdu

L.L

LU

LU

- j

0.900

0.800

0.700

0.600

0.500

0.400

0.300

0.200

0.100

0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 11. Comparison of PZR Level with Data
KOR 1 -3 INADVERTENT SI ACCIDENT (COMPARISON) 600

- RELAP5/MOD3 ..... Plant Data

0;)

a)

H--

H580

560

540

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 12a. Comparison of T Hot (A) with Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 600

CD

H-E

H580

560

540 CI

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 12b. Comparison of T Hot (B) w i th Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 600

- RELAP5/MOD3 ..... Plant Data

0)

Of

U)

-o

iii

a_

HLH

580

560

540

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 12c. Comparison of T Hot (C) w i th Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 600

0j)

-0

LUJ

HLUJ

F580

560

C., 540

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 13a. Comparison of T Cold (A) w i th Data
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 600 - RELAP5/MO03

Plant Data

0)

Iii

LU

H--

LU

LU

0-

580

560

I

540

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 13b. Comparison of T Cold (B) with Data
KOR 1 -3 INADVERTENT SI ACCIDENT (COMPARISON) 600

I

0)

-0

w

Hc--

580

560

540 -3

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

T IME (SEC)

FIG, 13c. Comparison of T Cold (C) with Data
KOR 1-3 INADVERTENT SI ACCIDENT (COMPARISON) 600

RELAP5/MOD3 (T Cold)

Plant Data (T Cold)

RELAP5/MOD3 (SG Temp.)

Plant Data (SG Temp.)

0)

LU

Ha--

LU

F--

580

580

003 540

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 14a. Comparison of T Cold (A) and SG-A Temp.
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 600

RELAP5/MOD3 (T Cold)

Plant Data (T Cold)

RELAP5/MOD3 (SG Temp.)

Plant Data (SG Temp.)

0T)

Ir

Of

-D

LU

IiLLd

580

560

540 I.

520

500 I I . I I

0 500 1000 1500

I 1 1 1

2000 2500 3000 3500 4000

TIME (SEC)

FIG. 14b. Comparison of T Cold (B) and SG-B Temp.
KORI-3 INADVERTENT SI ACCIDENT (COMPARISON) 500

RELAP5/MOD3 (T Cold)

Plant Data (T Cold)

RELAP5/MOD3 (SG Temp.)

Plant Data (SG Temp.)

0)

cr

-i

w

aLLJ H--

580

560

u, 540

520

500

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 14c. Comparison oF T Cold (C) and SG-C Temp.
KOR 1 -3 INADVERTENT S 1 ACCIDENT (RELAP5/MO03)

C-9

\Z

U)

8000

7000

O0O0

5000

4000

3000

2000

1000

0

I

c--n

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 15. InLegrated Flow of PZR PORV (Ca I cu I a ted)
KORI-3 INADVERTENT SI ACCIDENT (RELAP5/MOD3) 15

12

UwU

cnf

C-0

Ir

9.0

6.0 C-,

3.0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 16. SaFety Injection Flow Rate (Calculated)
KORI-3 INADVERTENT SI ACCIDENT (RELAP5/MOD3) 50

x

U

LU

c!)

U

40

30

20 I

Ca

10

0

0 500 1000 1500 2000 2500 3000 3500 4000

TIME (SEC)

FIG. 17. Total Required CPU Time

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

t2-_89) IAolmned by NRC. Add Vol., Iupp., Rev.. NRCM 12. aind Addendum Nurmber. It iny.)

3o,.IM12 BIBLIOGRAPHIC DATA SHEET NUREG/IA-0105

ISee inStruction$ on the reversel

2. TITLE AND SUBTITLE

Assessment of RELAP5/MOD3 Version 5m5 Using Inadvertent Safety Injection 3. DATE REPORT PUBLISHED

Incident Data of Kori Unit 3 Plant MONTH V 'EAR

May 1993

4. FIN OR GRANT NUMBER

L2245

5. AUTHOR(S) 6. TYPE OF REPORT

K. T. KTechnical Report

K.T.Kim, B. D. Chung, I. G. Kim, H. J. Kim 7. PERIOD COVERED liclus, Damt)

8..PR..... N............AE.NDAD.SS....C.. .... D...n......rR.io... N deReultryCo nluon ndmaligadr~~l onrcr,~po ic

8. PERFORMING ORGANIZATION -- NAME AND ADDRESS (it NRC. provide Division. Office of Replan. U.1 Nucloa Refiulroray Co imsio, and mallino addresvl ontractor, Provide nae mod .Mkvag addm".)

Korea Institute of Nuclear Safety

P. 0. Box 16

Daeduk-Danji

Taejon, Korea, 305-606

9. SPONSORING ORGANIZATION- NAME AND ADDRESS (II NRC. tvp. '3mw as above°"; tcontractor. rovide NRC Division. Office or Region. U.1. NWOW..' Regulaory Commission.

&,W maili addrom)

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT 1200 words orv)

An inadvertent safety injection incident occurred at Kori Unit 3 in September 6, 1990 was analyzed using the

RELAP5/MOD3 code. The event was initiated by a closure of main feedwater control valve of one of three steam

generators. High pressure safety injection system was actuated by the low pressure signal of main steam line.

The actual sequence of plant transient with the proper estimations of operator actions was investigated in the present

calculation. The asymmetric loop behaviors of the plant was also considered by nodalizing the loops of the plant

into three.

The calculational results are compared with the plant transient data. It is shown that the overall plant transient

depends strongly on the auxiliary feedwater flowrate controlled by the operator and that the code gives an

acceptable prediction of the plant behavior with the proper assumptions of the operator actions. The results also

show that the solidification of pressurizer is not occurred and the liquid-vapor mixture does not flow out through

pressurizer PORV. The behavior of primary pressure during pressurizer PORV actuation is poorly predicted because

the actual behavior of pressurizer PORV could not be modelled in the present simulation.

12. KEY WORDS/OESCR!PTORS ILn twor orphssres that will assist nwasthom in locating the report.1 13. AVAILABI LItV STATEMENT

ICAP Program Unlimited

RELAP5/MOD3 14. sECURTY CLASSIFICATION

Kori Unit 3 ,r,,I Pag, 

Unclassified

1 gt RAener

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 33S (2491


Federal Recycling Program

NUREGIIA-0105 ASSESSMENT OF RELAPS/MOD3 VERSION SmS USING INADVERTENTI'

SAFETY INJECTION INCIDENT DATA OF KORI UNIT 3 PLANT

MAY 1993

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

, i