NUREG/IA-0092

International

Agreement Report

Assessment of RELAP5/MOD2

Computer Code Against the

Net Load Trip Test Data

From Yong-Gwang, Unit 2

Prepared by

Namnsung Arne, Sungjae ChoIKEPC

Sang Hoon Lee/KINS

Korea Electric Power Co.

17-15, Yongjeon-Dong, Dong-GLI

Taejeon, Korea

Korea Institute of Nuclear Safety

P.O. Box 16, Daeduck Danji

Taejeon, Korea

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-0092

International

Agreement Report

Assessment of RELAP5/MOD2

Computer Code Against the

Net Load Trip Test Data

From Yong-Gwang, Unit 2

Prepared by

Namnsung Arne, Sungjae ChoIKEPC

Sang Hoon LeeJKINS

Korea Electric Power Co.

17-15 Yongjeon-Dong, Dong-Gu

Taejeon, Korea

Korea Institute of Nuclear Safety

P.O. Box 16, Daeduck Danji

Taejeon, Korea

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 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, 1985). 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

The results of the RELAP5/IIOD2 computer code simulation for the 100 %

Net Load Trip Test in Yoný-Gwang Unit 2 are analyzed here and compared

with the plant operation data. The control systems for the control

rod, feedwater, steam generator level, steam dump, pressurizer level

and pressure are modeled to be functioned automatically until the

power level decreases below 30 % nuclear power. A sensitivity study

on control rod worth was carried out and it was found that variable

rod worth should be used-to achieve good prediction of neutron power.

The results obtained from RELAP5/IIOD2 simulation agree well with the

plant operating data and it can be concluded that this code has the

capability in analyzing the transient of this type in a best estimate

means.

ii i

TABLE OF CONTENTS

ABSTRACT ------------------------------------- ii

LIST OF TABLES ----------------------------------- vi

LIST OF FIGURES ---------------------------------- vii

CHAPTER 1. INTRODUCTION ---------------------------- 1

CHAPTER 2. PLANT DESCRIPTION------------------------ 2

CHAPTER 3. PLANT TRANSIENT DESCRIPTION--------------- 17

CHAPTER 4. NODALIZAT10N AND INITIALIZATION------------21

CHAPTER 5. CALCULATION RESULTS AND ANALYSIS-----------29

CHAPTER 6. SENSITIVITY STUDY ----------------------- 48

CHAPTER 7. NODALIZATION STUDY ------------------------ 52

CHAPTER 8. RUN STATISTICS -------------------------- 58

CHAPTER 9. CONCLUSIONS --------------------------- 61

REFERENCE

V
LIST OF TABLE

Table 1. Pressurizer Pressure Control System Setpoint------ 9

Table 2. Rod Control System Setpoint -------------------- 11

Table 3. Steam Generator Level Control System Setpoint----13

Table 4. Steam Dump Control System Setpoint -------------- 16

Table 5. Acceptance Criteria---------------------------- 20

Table 6. Initialization Data --------------------------- 27

Table 7. Boundary Condition at Gov. V/V------------------ 28

Table 8. Reactivity Worth of Control Rod----------------- 51

Vi
LIST OF FIGURE

Figure

Figure

Figure

2.1

2.2

2.3

Figure 2.4

Figure 2. 5

Figure 2. 6

Figure 2.7

PWR schematic diagram ---------------------- 6

PZR equipment Arrangement-------------------- 7

Functional Diagram of the

PZR pressure control system------------------- 8

Functional Diagram of

the Rod Control System --------------------- 10

Functional Diagram of the

S/G Level Control System -------------------- 12

Schematic Representation of the

Steam Dump Control System -------------------- 14

Functional Diagram of the

Steam Dump Control System -------------------- 15

Nodalization------------------------- 26

Reactor Power ----------- 37

Tavg - - - - - - --38

DEL. T -- - - - - - - -39

PZR. Pressure --------- 40

PZR. Level --------- 41

Spray Flow --------- 42

S/G Pressure --------- 43

S/G Level ----- ---- 44

Feed Water Flow --------- 45

Figure

Figure

Figure

Figure

Figure

Figure

Figure

Figure

Figure

Figure

3.1

5.1

5.2

5.3

5.4

5.5

5.6

5.7

5.8

5.9

Vii
Figure

Figure

Figure

Figure

Figure

Figure

Figure

Figure

Figure

5.10 Steam Flow

5.11 Steam Dump Flow

6.1 Sensitivity Study

7.1 Nodalization Study

7.1 Nodalization Study (Steam Dump

7.2 Nodalization Study (PZR Pres.)

7.3 Nodalization (S/G Pres.)

8.1 CPU(sec)

8.2 Time Step Size

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

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

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

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

Flow) -------------

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

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

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

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

46

47

50

54

55

56

57

59

60

viii
1.0 Introduction

The objective of this work was to perform the best estimate transient

analysis using the REJAP5/llodZ/36.04 computer code to demonstrate that

the Net Load Trip Test, hereinafter named NLTT, of Yong-Gwang Unit 2,

performed during its start up test period, can be simulated by this

best estimate computer code.

The purpose of 100 % NLTT is to prove the ability of the plant to

sustain a net load, which is required as a minimum house load when the

transmission line fault occurs, without reactor and turbine trip and

to evaluate the interaction between the control system and the system

response to the transient [7].

With the plant at 100 steady state conditions and all control

systems in automatic mode, NLTT was initiated by manually placing the

main transformer high side breaker in the tripped position. During

the transient, the major plant operating data were collected by using

the on-line computer data logging system and seven strip charts.

The plant geometry and the plant transient are described in chapter 2

and 3. In chapter 4, the plant simulation model and nodalization are

presented, while illustrating the initialization methodology.. The

base calculation results are analyzed in chapter 5 and the analysis of

sensitivity study and a run statics are included in chapter 7 and 8

respectively. Conclusions are added in chapter 9

- 1-
2.0 Plant description

This section describes the plant components and control systems which

are considered important to understand the Yong-Gwang Unit 2 NLTT

simulation. Yong-Gwang Unit 2 ,which is located on the southwestern

coast of the republic of Korea ,is a Westinghouse 3 loop PWR rated at

996.8 Nwe and Becthel was the AlE.' A schematic representation of the

Yong-Gwang Unit 2 is presented in Figure 2.1 [3].

2.1 Primary system

The primary system consists of a pressurized water reactor, reactor

coolant system and the associated auxiliary systems. The reactor

coolant system is arranged as three reactor coolant loops connected in

parallel to the reactor vessel, each containing a reactor coolant pump

and a steam generator. As shown in Figure 2.2, an electrically heated

pressurizer is connected to the hot leg of one reactor coolant loop.

The core is of the three region type and consists of 157 fuel

assemblies with 264 fuel rods per assembly. The nominal core power is

2775.5 Nwt. RCCA( Rod Cluster Control Assemblies) are used for

reactor rod control and consist of clusters of cylindrical absorber

rods. The reactor coolant pumps are Westinghouse vertical, single

stage and centrifugal pumps of the shaft seal type. The steam

generators are Westinghouse vertical u-tube- units which contain

inconel tube [5].

-2 -
The pressurizer pressure control system are making a Rey role in the

primary control system together with the control rod control system.

The pressurizer pressure control system whose functional block diagram

and setpoints are shown in Figure 2.3 and table 1 maintains pressure

at the setpoint value with the help of spray valves (158.95

Rgf/cm2), relief valves (164.23 1cgf/cm2),safety valves (175

kgf/cm2), proportional heaters and backup heaters [1J. The heaters,

spray valve and relief valves maintain the pressure at the setpoint

value and prevent reactor trip as a result of pressure variations

caused by plant transients.

As shown in Figure 2.4, the automatic control rod control system is

designed to maintain a programmed average temperature in the reactor

coolant by regulating the reactivity within the core. This control

system is designed to control the reactor power automatically in the

power range between 15 7% and 100 7. of rated power for the design

transients. Table 2 shows the setpoints of the control rod control

system.

2.2 Secondary system

The secondary system conveys steam from the steam generators to the

turbine generator system. The system consists of main steam piping,

power operated relief valves, safety valves, main steam isolation

valves, atmosphere and condenser dump valves. The steam generator

level control system 6nd the steam dump control system ,which are

shown in Figure 2.5 and 2.7, respectively, are the major control

logics in the secondary system.

- 3 -
The setpoints for these systems are shown in table 3 and 4. Each

steam generator is equipped with a three element feedwater

controller (feedwater flow, steam flow and water level) which

maintains a programmed water level (50 Z on the secondary side of

the steam generator during normal plant operation. This

controller continuously compares the measured feedwater flow with the

steam flow and a compensated steam generator downcomer water level

signal with a water level setpoint to regulate the main feedwater

valve opening. The steam dump control system whose schematic

representation is shown in Figure 2.6, comprises 16 valves which

can bypass steam to the condenser and to the atmosphere. Tavg

and steam pressure (Tref) activate the dump system, which is

interlocked with plant output to enhance overall control response.

If the difference between the required temperature set point of

the reactore coolant system and the actual temperature exceeds a

predetermined amount (4 c) which is called dead band, a signal

will actuate the steam dump to maintain the reactor coolant system

temperature within control range until a equilibrium condition is

reached. The required number of steam dump valves stroke fully open

or modulate, depending upon the magnitude of the temperature error

signal resulting from load reduction.. The dump valves can be

modulated closed after they are full cpen by the reactor coolant

average temperature mismatch signal. The steam dump systez. is

comprised of four banks which can bypass steam to condenser and

atmosphere , i.e. one' single bypass valve with setpoint of 6.48 F

one double bypass valve with 11.44 F, one single dump valve 14.24 F,
and one double dump valve with 19.2 F. These valves have a total

capacity of 64 Z of full load turbine steam flow [ 1].

2.3 Plant data acquisition

There are three type of digital recording from plant computer, the

computer daily logging sheet, pre-post trip review record and sequence

record of event. Besides of these digital recordings, there are three

analog strip chart recordings of which speed is 2 mnl/min. There, of

course, exists the uncertainty of the collected data which are due to

the reading and instrument error. According to the Ref.[8], the

reading and instrument error is 1.5 7., The plant operating data

are collected every two' seconds, and the simulation data is

calculated every one second.

-5 -
THROTTLE 7ALVE

HNE STEAM I

'1 ~ FEEDWATER

~.1- VALVE / -~----FEEENATER FLCW

CO LD LEG

SPRAY LINE

Figure 2-1. Pressurized Water Reactor Schematic Dia~ram
SA~FETY VALVES

RELIEF VALVES

T

TO PR~ESSURIZER

RELIEF TANK

I

Figure 2- 2,. Eressaii~er Pressure and Level Associated Equipment Arrangement
REFERENCE

PRESSURE

PRESSURE

SIGNAL

SPRAY CONTROL

PROPORT I O.NL

HEATJER CONMIL

BACKUP

HEATER CONTOL I

cc

I

POWER RELIEF

VALVE fl CZiTROL

PO(ER 'RELIEF

VALVE #2 CC'XTROL

POW(ER RELIEF

VALVE #3 CONTROL

Figure 2-3 .Functional Block Diagram of the Pressurizer Pressure Control System
Table 1. PRESSUR.ZER PRESSURE CONTROL SYSTEM SETPOINTS

PlO Controller

K21 1 psi/psi

'21 900o sec

'22 0 sec

P ref 2,235 psig

Spray controlle'r

Spray Initiation pressure 2,260 psjgta)

Prprinlgain 2%/si

-Proportional heater controller

Proportional gain -3.33%/psi

Setpoint for full-on proportional heaters 2.220 psig(a)

B fackup heater controller

Backup heaters turned on 2,210 ps1g~a)

.Power relief valve no. 1 2,335 pslgla)

Power relief valve no. 2 and no. 3 2.335 psig

9
NUCL[AR POWER +

%IRBINE LOAD REPRESENTED BY

-uPULSI CH1AMBER PESSURE

0

w-t

ROD HOTION

iN STEPS

Figure 2-A Functional Block Diagram o~f the Rod Conroi System
Table 2. ROD COUIROL SYSTEMI SEIPOINTS

Ref gain (BT re f/BQT) For a full load Tag 0.3lY*F/%

of 588.5

80 se c

3

4 10 sec

5 5 sec

No load temperature 5576F

Power mismatch controller

40 sec

flollnea~r galn (K1)

.error signal at breakpoint of nonlinear gain + 1%

* Low gain 0.34F/%

* High gain 1.5"F/%

Variable gain (K.~ AT 100% 1.0

* Down to 50% Inversely

*proportional to

* turbine power

*: to~ 0% 2-.0

Control bank overlap 113 steps

...Start 0 with A.* 115 steps

Stop-A *.228 steps

Start C with B 230 steps

Stop B * 343.steps

* Start D with C 345 steps,

*...estop C.* 458 steps

-III-
FEEDWATER FLOW SIGNAL

C

STEAM FLOW SIGNAL

S-E AM

-MAIN FEEDWATER

VALVE POsITIONER

c'J

S-.AGE

F=ESSURE

NUCLEAR POWER SIGNAL BYPASS FEEDWATER,

VALVE POSITIONER

Figure 2-5 Functional Block Diagram of the Steam Generator Level Control System
Table 3. STEAM GENERATOR LEVEL CONTROL SYSTEM SETPOINTS

Level.Program

Prograrrner100% Span -... 128 Inches

s0

Percent. of

narrow range

level span

I.

100

PERCENT TURDBINE POWER

(FROM IMPULSE STAGE PRESSURE)

Level Controlle'r

~30

131

S32

136

K30

K34.

5 sec

200 sec

5 sec

200 sec

750 sec

3.3% full flow/% level-span

-1% valve lift/% fulf V'Elow

2% valve.lift/% full flow

K 4% valve lift/% nuclear power 35

13
walsAS lojiuoo dwun] cJI wa1~4l JO UO11elUasaidaU oilewaqoS 9..z ain!d

'-A

S3 A'lVA 9)

S3AIVA dwnfl

WV3V1S U3SN30N03

(SAlVA 00.

S3AIVA dVwil0

V4V31S OIU3HdSQVYUV

3AlVA

dOIS fI 3A1VA

(S3AIVA V) A13:IVS

S3AIVA

OIU3HdSOW.rV1
TREF -

AUCTIONEERED

TAVG _

TRIP BANK A

CONDENSER

VALVES

TRIP BANK B

CONDENSER

VALVES

TRIP BANK C

ATMOSPHERIC

VALVES

TRIP BANK D

ATMOSPHERIC

VALVES

TNO LOAD_

STEAM D.-MP j LOAD SELECTC= SWITCH

R LEJCTIC-%

-0;T MODULATE

TURBINE TRIP --- O-CPEN!CLOSE

DUMP VALVES

SET ST EAM DUMP HEADER

PRESSURE (IPRESSURE CONTROLLER

STEAM \F7 I 

HEADER - -IK 1 1 I +

PRESSURE ELj4fJ r 

In

j

_j

TURBINE

IMPULSE

CHAMBER

PRESSURE

______________ ALLOW

CONDENSER CONDENSER

PRESSURE INTER- STEAM DL?%'P

LOCK CIRCULATING - VALVES TO BE

WATER PUMPS I EITHER

INTERLOCK I TRIPPED OPEN

~ OR MODULATED

ALLOW HALF OF

TURBINýENOT 0.ONDENSER ST.EAM

TRIPP D DUMP VALVES TO

TRIPPED E EITHER T;:iPPED

INTERLOCK ENOR

+ M'ODULATED

CO'NDENSER AVAILABLE SIGNAL

TURBINE TRIP

Figure 2.9 Functio~nal Block Diagram of the Steam Dumlp Control System
Table 4. STEAMI DUMP CONiTROL SYSTEM SETPOIHTS

Load rejection controller

~ 1 120 sec

'112 10 sec

713 Ssec

Deadband 50F

Proportional gain in percent 3.8%/*F

of total dump capac'aty.per OFta]

Trip open temperatures 9.86F, l9.6or,

24.6-F, 31.3-F

Sudden load-loss setpoint (C-7A) 10% of full. load

Load-loss greater than 10% (C-7D) 50% or fu'.1 load

Turbine trip controller

112 10 sec

713 5 sec

Dead Band 0OF

Proportional gain in percent 1.79%/"F

of total dump capacity per Fa

Trip open temperatures 10.4.OF, 31.30F

Header pressure controller

Set pressure 1,092 psig

Proportional gain in percent of 1%/psi

total dump capacity per psi,[a] Kll

Reset time constant, -rl 180 sec

16
3. Plant transient description

With the plant at steady state 100 power conditions and all control

systems in automatic mode, NUTT is initiated by manually placing thle

main transformer high side breaker in the tripped position. Right

after the initiation of the transient, the decrease in the S/G steam

flow resulting from the large load reduction limits the amount of

energy which can be removed from the primary system. Thle secondary

side steam flow rate decrease is not matched by an equivalent drop in

the core power level. This results in increasing the primary system

pressure and temperature (Tavg). The automatic control rod control

system ,whose diagram is shown in Figure 2.4 ,is designed to control

the reactor in the power range between 15 7. of full power and 100 Z of

full power, while maintaining a programmed average temperature in the

reactor coolant system ~by regulating the reactivity within the core

[2]. The control rod is working according to the two temperature

error signals automatically. The error signals are composed of the

difference between Tavg and Tref which means the difference between

newtron power and turbine power. . ,

As shown in Figure 2.3, the pressurizer pressure and level control

system functions to maintain the proper water inventory in the reactor

coolant system. Due to the coolant density change during the

transient, the pressurizer level and pressure fluctuates as well and

the pressurizer functions to maintain the proper water inventory by

- 17 -
the help of spray and heaters. On a sudden turbine load reduction,

the steam generator level encounters initially a shrink effect due to

a collapse of void in the steam generator. Therefore the level error

deviation controller calls for an increase in feedwater flow to

recover the steam generator level. At the same time, the decrease in

the steam flow results in unequal flow at the feed/steam flow

deviation controller and this controller calls for a decrease in

feedwater flow. Since the level error produced by the shrink is a

transitory and short term effect, the output of the level deviation

controller is delayed by sen~ing it through a filtered unit which has

5 second lag time. This allows the flow deviation controller to take

the load in controlling the feedwater flow and will drop flow to match

the decrease in steam flow.

The excess steam resulting from the load decrease in the turbine is

bypassed to the condenser and to the atmosphere via the steam dump

system. The steam dump control system ,shown in Figure 2.7, reduces

the magnitude of nuclear power plant system response to the transient,

following a large turbine load reduction by dumping steam directly to

the main condenser and atmosphere. This system has a total capacity

of approximately 64 percent of the full load turbine steam flow at

full load steam pressure.

This test was conducted as one of the tests which should be done

during commissioning in Yong-Gwang Unit 2. The transient initiated by

opening the high side breaker , PCB 7400 and PCB 7471 and after 180

- 18 -
seconds the reactor power reached to 50 full power when the plant

control was changed to manual operation. After 240 seconds, the

reactor power and turbine power reached to 11.5 of full power and

27 riwe, respectively and after 10 min. synchronized to the grid again

and returned to the 100 7 of power level [7]. This test was

performed with satisfying the acceptance criteria specified in Table

5 . The result of this test turned out acceptable.

- 19 -
Table 5. Acceptaiicc Crit~oria for NlI1TT

- TI3N does not reach ovbrspeed trip setpoint

- Safety Injection is not initiated

- RX and TBN miust not trip

- No manual intervention should be necessary for control

rod, feedwater, steam dump, pressurizer pressure and

level until power level decreases below 30 % nuclear

power.

-SIG and PhR safety valves should not actuate.

- 20 -
4. Nodalization and Initialization

The used code version was RELAP5/110D2/CY 36.04 and was processed

on a CDC machine cyber 170-875. The input deck was prepared by

KEPCO. The nodalization philosophy is based on the guideline of

RELAP5 code manual and the detailed data for specific volumes and

junctions are based on the design and drawing values for

Yong-Gwang Unit 2.

The nodalization can be found in Figure 4.1, which was

previously developed for the LOCA analysis [6]. The whole system

is devided into 118 volumes, 122 Junctions and 79 heat slabs.

The reactor core is described with 6 volumes and core heat slab

is modeled by using the RELAP5/MOD2 point kinetics model. The

reactor vessel consists of 4 volumes, ie, the downcomer,the lower

plenum and core inlet, the upper plenum the upper head and

control rod guide tube. Steam generator is devided into the

primary and the secondary system with the boundary of U-tube and

the primary system is described with 10 volumes and the secondary

with 9 volumes. Two steam generators are lumped in order to

match with the RCS loop. Separator is depicted by using the

RELAP5 SEPARATOR model. The RCP and the accumulator are also

modeled in this simulation. RELAP5 ANS79 model is applied for

computing the decay heat and the negative reactivity of

temperature and doppler coefficient is considered.

- 21 -
The plant consists of 3 loop coolant system but for the

convenience of the nodalization, the reactor coolant system is

modeled with two loops, one is the loop which lumps the two loops

in which the pressurizer is not connected and the other is the

single loop with pressurizer. Each loop is depicted with the

RELAP5 pipe component and consider the fluid-metal interaction by

using the heat structure. The safety and relief valve nodes, of

steam generator and pressurizer are not included because it was

confirmed by the simple calculation that they did not open

during this transient.

The outlets of both steam generators are connected to a single

volume, steam header, which is connected to the turbine and the

condenser volume. One assumption is needed in order to model

this steam dump control system. The atmosphere volume is assumed

to have the same pressure as the condenser volume,thus the

condenser volume in this nodalization contains the atmosphere

volume to which the steam is designed to bypass through the steam

dump valve. The steam dump system comprised of 4 banks. So

the condenser volume is devided into 4 volumes and connected to

the steam header with the survo valve junction. These survo

valves behaves realistically in that the valve opening area

varies as a function of normalized valve area calculated by steam

dump system. For the area change option, the abrupt area change

option is selected for valve hydrodynamics in order to calculate

- 22 -
the exact steam dump flow rate. Turbine and condenser are

modelled as a time dependent volume.

For the steady state initialization, the RELAP5MOD2 computer

code requires the initial conditions which result in zero time

deviation for each of balance equation [4]. The initial

condition for the RELAP5 simulation should agree with the power

plant steady state condition at 100 %. full power. The simulated

steady state initial conditions along with the plant steady state

data for 100 % full power are presented in Table 6, which shows

an excellent agreement with the plant data. The steady state

condition is obtained in the condition that all primary and

secondary control systems are in auto mode.

In the RELAP5/?lod2 simulation, it is very important to get the

accurate Tref value because Tref value affects to actuate the

control rod and the steam dump system. In order to get the exact

Tref value, the steam flow rate at the governor valve is used as

the boundary condition because steam flow rate can be considered

to be proportional to the turbine power and Tref can be obtained

by using turbine power. It is also very important to compute the

steami flow at the governor valve exactly in order to calculate

the accurate turbine power. Since the curve of valve stem

position vs. time or the normalized valve flow area vs. time of

the governor valve from the test results is not available, one

assumption is made in this analysis. This assumption deals with

- 23 -
the steam flow rate calculation at the governor valve. Without

knowing the precise behavior of the governor valve, the steam

flow rate can not be accurately calculated. Therefore the steam

flow rate is intended to be imposed as a boundary condition by

applying the time dependent junction at the location of governor

valve and can be obtained from Tref value in the plant operating

data. The boundary condition is shown in Table 7.

The steam generator level control system is modelled to function

automatically with the three element method, ie, feedwater flow,

steam flow, and steam generator level. The feedwater valve is

working by controlling the feedwater pump speed in the power

plant. It is practically difficult to model the feedwater pump

speed control and in order to overcome the difficulty

encountered, one assumption is made. The feedwater flow rate

is determined as a boundary condition -by applying the time

dependent junction at the feedwater valve and this boundary

condition is determined by the normalized feedwater valve area

calculated by the feedwater control system. In order to get

the exact boundary condition, an approximate method is applied in

which the history of valve area is adjusted so that the desired

feedwater flow is obtained. This can be done by normalizing the

feedwatwer flow rate and the valve flow area response will follow

the same curve. At full power, the normalized valve area was

0.88 [51.

-24 -
In the NUjTI of the nuclear power plant, many plants have

experience to fail the NU2T due to the 01 t trip resulted from

the temperature change in the primary system. In the RELAP5

simulation, this OT T is modeled with the limitation in the

calculation of the nuclear calibration term for T trips. Since

the RELAP5/MOD2 code can not compute the nuclear calibration

term, the present modeling contains the OT T trip model without

the nuclear calibration term.

- 25 -
100

Ll*x VýJ %:.v ,

NO I

Fig 4.0 Nodalization

-'X, Cross nlow junction
Table 2. Plant Initialization Data

Plant Data RELAP5/MOD2 Data

1.0 Core

Inlet

Temperature 291.7 (c) 292.62 (c)

Outlet

Temperature 326.6 (c) 327.08 (c)

Average

Temperature 309.2 (c) 309.85 Cc)

Core Power 2775.5 Mwt 2775.5 Mwt

2.

Pressurizer

Level 0.58 0.58

Pressure 157 (kgf/cm2) 157.4 (kgf/cm2)

3.0

Steam

Generator

Pressure 66. 8(kgf/cm2) 67. 2(kgf/cm2)

Water Level 0.5 0.5

Steam Flow 511.75(kg/sec) 511.55 (kglsec)

Feed Water

Flow 513.88(kg/sec) 512.26 (kg/sec)

Feed Water

Temperature 226.5 ( c) 226.7 ( c)

-27-
Table 3. Boundary Condition for Steam Flow at Governor Valve

Time (sec) Steam Flow (ib/sec)

0.0 3381.0

1.0 2028.6

2.0 1196.7

3.0 983.87

4.0 770.87

5.0 561.24

10.0 135.24

200.0 135.24

-28-
5.0 Calculation results and discussions

5.1 Primary system

The core power is controlled by the control rod with the two. error

signals. Right after the initiation of transient, the control rod

begin inserting into the core in order to decrease Tavg and the

neutron power. In the RELAP5/110D2 simulation, the greatest effect of

the control rod control system is shown between 20 seconds and 80

seconds. The control rod decreases the reactor power with the rate of

0.33 X of full power per second(Fig 5.1). The reactor power begins

to increase again up to 79% of full power for 20 seconds due .-to the

effect of the negative moderator temperature coefficient which appears

near 80 seconds. After 100 seconds, the core power continues

decreasing toward 30% of full power at which the power plant.-control

mode was changed to the manual control mode. The actual power plant

responds differently. It continues to decreuse down to 50 % of full

power with a reduction rate of 0.28 % of full power per second. The

effect of IITC(moderator temperatur e coefficient) can be found around

100 seconds for 30 seconds. This,*later effect of IIIC in. the power

plant operating data causes a little lower value after 85 seconds than

the simulation data. The main reason of this later effect of MTfC may

result from the lack of the accurate reactivity information in the

core during the test and more details will be added in the sensitivity

'29 -
study. ilowever it can be concluded that the core power calculated by

the REIAP5/Mod2 computer code agrees well with the plant operating

data.

Following the transient, Tavg increases because of the reduced heat

removal capacity of the secondary system. With the help of the

sliding Tavg program, reactor power is adjusted to maintain a

programmed decreasing Tavg as the turbine load is decreased.

Decreasing the turbine load mak~es Tavg and steam pressure increase.

The control rod control system inserted the control rod into the core

in order to decrease Tavg equal to the programmed Tavg. The simulated

Tavg value in Fig 5.2 shows two peaks of temperature around 10 second

and 40 second, respectively. These two peaks can also be found in the

pressurizer liquid level, the pressurizer pressure and the spray flow

which are shown in Figure 5.4, 5.5, and 5.6, respectively even though

these peaks are very small. The timing of these two peak values

approximately agreed well with the pressurizer liquid level and

pressure but a little latter-around 7 seconds than the timing of the

spray flow rate. As shown in Fig. 5.2, The simulated Tavg value

follows the plant operating data quite accurately up to about 80

second but after that time, it indicates a little higher value which

s hows the similar trend to core power behavior. The difference

between the outlet and inlet temperature in the RELAP5/IIOD2 simulation

(Figure 5.3) reduecd from 34.46 c to 22.9 c for 180 seconds, in which

overall agreement was satisfactory. Right after the initiation of

- 30 -
the transient, The cold leg temperature increases very fast due to the

sudden reduction of heat removal of the secondary system ,meanwhile

the hot leg temperature is constant and this makes the temperature

difference between outlet and inlet temperature drop very fast, this

reduction continues until 6 second. After 6 second, the hot leg

temperature increases due to the constant core power generation until

10 second. After that, it continues to decrease.

The pressurizer level control system functions to maintain the proper

water inventory in the reactor coolant system. This level is

maintained by controlling the balance between water leaving the

system, via the letdown flow path to the CVCS, and water entering the

system from charging pump. Since letdown is a fixed amount, the

balance is maintained by varing the changing flow. Like this kind of

simulation, it is judged that the changing charging flow does not

affect much to control the pressurizer level ,so this is modelled as a

constant flow. The simulated pressurizer level shows a little lower

trend than the plant operating data until 80 second and a little

higher after that time. Even though the plant operating data shows an

oscillation , the trend of the simulated pressurizer level is similar

to that of the plant operating data. The pressurizer pressure is

controlled by using the heaters and spray. The decrease of the

secondary heat removal induces the Tavg increase and RCS volume

increases with yielding the large volume water enters into the

pressurizer. During this transient, the pressurizer is insurged.

- 31 -
After insurging, vapor bubble in the pressurizer is compressed and

pressure initially increased. This makes the pressurizer spray

actuated to reduce pressurizer pressure by condensing steam from the

spray. The initial peak pressure in the simulation is shown around 5

seconds which is similar to the plant operating data. The inherent

pressure control due to the subcooled water makes the PZR pressure

decreases, after that another peak pressure can be obserbed due to the

reduction of heat removal of secondary system at 40 second. After

peak, it continues to decrease. Generally, the pressurizer pressure

trend in the REIAP/Mod2 simulation agrees well with that of plant

operating data except a little higher value after 80 second.

5.2 Secondary system

The secondary system cosists of two loops like the reactor coolant

system. The calculation results for both loops shows the exactly same

trend for the level and for the flow rates, the lumped loop shows

double flow rate. For the comparison with the plant operating data,

single loop data was used.

Each steam generator is equipped with three element feedwater

controllers (feedwater flow, steam__flow and steam generator water

level)., This controller continuously compares the measured feedwater

flow with the steam flow and also compares a compensated signal of

water level in the steam generator downcommer with a setpoint of water

level to regulate the main feedwater valve opening.. The RELAP5/IIOD?

.32 -
simulation indicates that the steam generator level drops to 46% o f

span level at 4 second (Figure 5.8). This first level drop caused

from the fact that on a sudden turbine load reduction, the steam

generator level encounters initially a shrink effect due to a collapse

of void in the steam generator but because the response of the feed

steam flow deviation controller functions much faster than the

response from the actual level/programmed level, the level deviation

controller calls for an increase in feedwater flow. So the steam

generator level begins to increase in order to be equal to the

progranmed level of 50 % of span level. It recovers up to 49 % level

of span about 10 second, and continues the constant level until 110

second, then it begins to decrease. The plant operating data shows a

different trend. It drops down to 32% level of span, recovers around

10 second, continue to keep a constant level until 70 second. When it

agrees well with the simulation data after 70 second it begins to show

oscillation. The general steam generator level trend by RELAP5/110D2

simulation shows a good agreement with that by the plant opeiating

data except some oscillation after 70 second.

The steam generator pressure increases due to the turbine valve

closing and reaches to the first peak pressure at 3 second, then it

begins to decrease due to the steam dump valve opening (Figure 5.7).

Around 8 second, It begins to increase again because the effect of

turbine valve closing is higher than that of steam dump valve opening,

and it increase gradually in order to reach the second peak pressure

- 33
about 30 second. The plant operating data reveals a little higher

response (2kgf/cm2) until 80 second. With showing the similar

general trend to the simulation pressure, except the oscillation over

the range of 6 kgf/cm2 with the trend of gradual increase. But it

shows similar two peak pressures which can be explained the same way

as above.

The three element valve controller functions to maintain the

necessary main feedwater valve position so that sufficient feedwater

flows into the steam generator to maintain the level at the programmed

value. The three signals that determines the valve position are the

level error signal , the steam flow rate signal, and the feedwater

flow rate signal. The result of RELAiP5/tlodZ simulation (Figure 5.9)

shows that the feedwater drops to 450 kg/sec as soon as the transient

is initiated (3 sec) due to the effect of the sudden steam flow

reduction. The plant operating data shows that the feedwater flow

drops to 370 kg/sec at 2 second. This can be explained as follows.

On the load reduction, the steam generator level encounters initially

a shrink effect due to a collapse of void in the steam generator and

the level deviation calls for an increase in feedwater flow. At the

same time, the decrease in steam flow developes unequal flow at the

feed/steam flow deviation controller and this controller calls for a

decrease in feedwater flow. Since the level error produced by the

shrink is a transitory , short term effect, the output of level

deviation is delayed by sensing it through a filter unit with the 5

- 34 -
second lag time. This allows the flow deviation controller to take

the load in controlling the feedwater flow and will drop flow to match

the decrease in steam flow. But soon the effect of steam generator

level deviation induces the feedwater flow to increase. The simulated

feedwater flow shows the peak flow rate of 510 kg/sec but the plant

operating data shows 520 kg/sec at 10 sec and 12 sec, respectively.

Even though the plant operating data shows a big oscillation with the

decrease trend but the simulated feedwater flow shows the continuous

decrease after peak flow.

The simulated steam flow (Figure 5.10) shows a sharp reduction down

to 340 kg/sec due to a TBN valve closing, meanwhile the plant

operating data indicates the same sharp reduction with the value of

250 kg/sec. Following this sharp reduction, it begins to increase

with the help of steam dump valve opening. After the peak flow rate,

the simulated steam flow decrease gradually with the less sharp slope

than the plant operating data. Sharp change of feedwater flow is not

seen when compared with steam flow due to the fact that the feedwater

controller senses both the error and the accumulation of error. This

integration factor is the main reason why the feedwater is less sharp

than the steam flow. The plant operating data for the steam generator

level, steam flow and feedwater flow show the oscillation but the

simulated data reveals that it has some limitation to simulate in

details enough to describe the oscillation. In spite of this

limitation, it can be concluded that the result of simulation agrees

well with the plant o~erating data.

- 35 -
On beginning the transient, the steam dump valves trip open

immediately and remain full open for approximately 35 seconds in the

simulation data. After that ,it begins to decrease continuously. Tile

plant operating data is not available, so it is impossible to compare

these data with the plant operating data. The Figure 5.11 shows the

steam dump flow rate calculated by the RELAP5/MOD2 code. the results

of calculation in the primary and secondary system illustrates that

the dump system works very accurately because the steam dump flow rate

determines the thermal-hydraulic parameters of primary and secondary

system.

36 -
1.00

Rfta' 1*va PUANT

0.00

0.00 50.00 100.00 150'.00 200.00

time(sec)

Ficiure 5.1 REACTOR POWER
340.00

~-300.00

-T

-RELAP

SPLANT

260.0C -1- !I'

0.00 50.00 100.00 150.00 200.00

time(sec)

Figure 5.2 Tavg
50.00

*3 0.00

______RELAP

***IC*PLANT

a San a

a a

owIV)

10.00 111r

0.00 50.00

II I II I I II I II I II IIII 1 I 11 I11I11I 1 'II

100.00 150.00 200.00

time(sec)

Figure 5.3 DEL T
2*00.00l

E

U)

~1050.00

0.

0

200.00

time(sec)

Figure 5.4 PRESSURIZER PRESSURE
1 .00

Lii

N.

0~0

-4

200.00

time(sec)

Figure 5.5 PZR LEVEL
0.10 - -RELAP

-000

'1,i

S0.05 /

0.005.0 10015.0 200

time(sec)

Figure 5.6 PRESSURIZER SPRAY FLOW (kg/sec)
Ld

CL

100.00

80.00-

60.00-

- RELAP ~ PLANT

* an 'a

~~=anan fit. ~a1r fin.

a. a. fi '~fl fig fit. a. ft.

*9gt .f

I

CV)

I.

40.i~:00 -I11

0.00 50.00 100.00 150.00 200.00

time(sec)

Figure 5.7 STEAM GENERATOR PRESSURE
1 .00

RELAP

P PLNT

-J

Li.

0.50

V)

999*~~ *9 * * 99 99 9. 9* 9

-- *9* * * 39 -- -~ 3* *39* *9 *9 9 * 3* 9* 9 9 * * 3 9 9 9 * * C

* *9* *

3 9

99

* 9 9

I

0.00 -jIIII I111

0.00 50.00 100.00 150.00 200.00

tirne(sec)

Figure 5.8 STEAM GENERATOR LEVEL
900.00

RELAP

SPLANT

0

a,

In

~' 500.00

0

-J

LL~

UWOK Sn'.

* ft* 0 an.

.: RE

ft .4 .4 .4 S Rift ft ft U a a

9'

-S.

I

Ln

1 00.00 - .0.00 50.00

1ia1 1 11 1 1 ii1ii1 iII I I II I I III I I

100.00 150.00 2)0 0.0 0

time(sec)

Figure 5.9 FEED WATER FLOW(kg/sec)
800.00

400.00

V

G)

C',)

0

-J

IU)

RELAR

SPLANT

'.L

0.00 11 1

0.00 50.00 100.00 150.00 200.00

time("sec)

Figure 5.10 STEAM FLOW (kg/sec)
2ý000.00

R ELAP

0

1000.00

0.00

0.00 50.00 100.00 150.00 200.00

time(sec)

Figure 5.11 STEAM DUMP FLOW
6.0 sensitivity study

The sensitivity study has been performed for the control rod control

system. The variable reactivity worth for the control rod was used to

achieve good prediction of neutron power and also to check whether thle

control rod control system is working properly. Figure 6.1 shows the

result of simulation which indicates the -reactor power with tile

different control rod reactivity worth. DATA2 has a double reactivity

worth compared to the DATAl, which has the reactivity worth with

minor modification in the reactivity worth obtained from the nuclear

design report. One thing should be pointed out here that it is very

difficult to get the accurate reactivity worth in the core during the

test. The various reactivity worth was used to achieve the good

prediction of neutron power. Because the main purpose of this

sensitivity study is to know the effect of the control rod reactivity

worth, the other parameter except the reactivity worth was not changed

at all. DATA2, which has the double reactivity worth, shows the

earlier feedback effect of the moderator temperature coefficient. It

reveals this effect around 50 second with the peak power rise up to 80 X

full power but DATAl, having the original reactivity worth,

indicates the effect of moderator temperature coefficient at 80 second

with a less peak power up to 7O7.full power. After this feedback

effect, DATA2 shows the sharper reduction of the reactor power than

DATAl. Around 180 second, DATA2 shows l0higher reactor power than

- 48 -
UDiTA1. The effect on the other thermal-hydraulic parameters of primary and

secondary system, i.e. coolant temperature, pressure and steam generator

pressure, was shown the sane as reactor power.

- 49 -
1.00-

0O.50-

0

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

DATAW

DATA2

,~~*~PLUNT

9

9 4 9. 99 9*

9. *~"V

*9 9 *999 99 9* 9

9

I

0.C00 jF

0.00

w i I I I I I I I I I I I I 1 1 I I I I I1 1 1 I I I I I I I 1 1

50.00 100.00 150.00 200.00

time(sec)

Fig. 6.1 * SENSITIVITY STUDY ( RX POWER)
Table 4. Reactivity Worth of Control Rod

-51-
7.0 Nodalization Study

The steam dump system is selected for the nodalizat ion study

because the dump steam flow rate via steam dump valves has an

important effect on deciding the behaviors of the thermalhydraulic

parameters in the primary and secondary system. It is of great

interest to find out the effect of nodalization in the

steam dump system on thermalhydraulic parameters. As shown in

Figure 7.1, two cases of nodalization are selected for this

nodalization study. Case 1 illustrates the boundary condition

model which uses two volumes( turbine and condenser) and one time

dependent junction (steam dump valve) where the steam dump flow

rate can be treated as the boundary condition in the steam dump

valves. The steam dump flow rate of this boundary condition is

determined by the value specified in the P,IJ,S

(precautionjlimitation and setpoints) which is a function of

temperature error (Tavg-Tref) [1).

CASE 2 describes the steam dump system in detail. The steam dump

system is composed of 4 banks with 16 steam dump valves. In order to

describe this system, the dump system modeling in the RELAP5/tNOD2

simulation consists of 4 volumes for condenser and atmosphere and 4

junctions for valves where the survo valve model was used in RELAP5

and the valve is working according to the 'normalized flow area

calculated by the control system.

- 52 -
The simulations for these two different cases (case 1 and case 2)

were performed to evaluate the effect of the different nodalization in

the steam dump system .The steam dump flow rates for both cases are

shown in Figure 7.2. The steam dump flow rate in Case 1 is around 120

kg/sec lower than that in Case 2 for the first 70 seconds and then

drops faster than Case 2. Figure 7.3 and Figure 7.4 show the

comparisons of pressurizer pressure and steam generator pressure

between two cases. As shown in Fig. 7.2, Case 2 computed the -steam

dump flow rate higher than Case 1 after 70 seconds and this induced

that the pressurizer and steam generator pressure, shown in. Figure

7.3 and Figure 7.4, in Case 1 reveals a little higher trend than that

in case 2. The results of the comparison indicates that valve

junction model is more effective to describe the steam dump system

than the boundary condition model.

- 53 -
1.0 BOUNDARY CONDITION NODALIZATION

2.0 VALVE JUNCTION NODALIZATION

Fig 7.1 Nodalization study ( two cases)

-54-
200/0.00

S1000.00

IL

02

DCase

Case IT

9. \

'I'..

********************. N~ -

'ON -

9

OIn

LO)

0.C0)0 -j-i0.00 50.00 100.00 150.00 200.00

time(sec)

Fig. 7.2 Nodalization study (steam dump flow)
180.00

-N 1 60.00

E

w 140.00

a_

2

0. Case k.

Case .r

S* PLANT -

. .... ....

I.

ko

120-OC -J

0.00 50.00 100.00 150.00 200.00

time(sec)

Fig. 7.3 Nodalization study (pressurizer pres.)
1 00.00 -r-i-r--r--r aaaaaaa 11 1 11 1 aar i r Ii III 

'' Case I

Case

S* PLANT

S80.00

EL ...... LO

S60.00

40.00 aa

0.00 50.00 100.00 150.00 2-00.00

timne(sec)

Fig. 7.4 Nodalization study (S,/C pres.)
8.0 Run statistics

The computer type used in this analysis is CDC CYBER 170-875 which

has one unified extended memory unit and the operating system is the

NOS 2.6.1 level 700. The Figure 8.1 and 8.2 show a plot of CPU vs RT

and a plot of DT vs RT. The total CPU time fot the simulation of the

whole transient 180 seconds is 705.44 second. The calculated grind

time is around 1.66 second and the requested maximum time step size is

0.1 second. Figure 8.1 indicates that the CPU time is proportional to

the real transient time with the slope of 3.9. The time step

size,shown in figure 8.2, shows a constant value(O.05) . This result

reveals that this transient is not so severe that the advancement of

calculation was succesful with the given maximum time step.

-58-
1000.00-

500.00-

0.00 -

0.

0LO

100.00

time(sec)

00

Figure 8.1 CPU
0.10

ILl

N

U)

nr~

Lu ~ ILi,

Lii

0

'.0

n n f')

0.00 50.00 ~100.00 150.00 200.00

tirne(sec)

Figure 8.2 TIME STEP SIZE
9.0 conclusions

The best estimate transient analysis for Yong-Gwang unit 2 NUTT

was performed by using the RELAP5/MOD2 computer code. Based on

the comparisons between the calculated and measured results, the

following conclusions can be drawn.

For the NU1 T, comparisons of the plant major primary system T/H

parameters, such as reactor power, Tavg, PZR level and pressure,

indicates that REIJAP5 could give the accurate results to the

plant operating data. However, the calculated primary

thermal-hydraulic parameters show slightly higher than the plant

operating data for the latter part of transient( after 70

second). This is due to model uncertainties in RTD sensing line

and lower steam generator steam pressure. It is also observed

that sucessful prediction of the PZR level and pressure is

largely dependent upon the accurate prediction of the transient

behavior of the RCS Tavg.

The calculated behaviors of secondary side parameters, such as

steam generator water level, steam pressure, feedwater flow and

steam flow rates are also agreed well with the plant operating

data. If the steam header pressure is prescribed as the

boundary condition, the calculated behaviors of the secondary

side parameters may be agreed better, but for the auto control

of steam dump system, it is not adapted.

- 61 -
The sensitivity study on the control rod worth is carried out

for the 100 %. NUTT. It was found that various rod worth should

be used to achieve good prediction of neutron power. Hence,

additiofial sensitivity studies regarding control rod modeling are

required in the future.

The nodalization study for the two cases, boundary condition

model and valve junction model, was performed and comparison was

made. It is observed that the valve junction model is more

effective to describe the behavior of this NUTT transient than

the boudary condition model.

- 62 -
Reference

1. Precation,Limitation and Setpoints for NSSS, January,1986

prepared by Westinghouse

Z. Setpoint Study for PKNU 7 ,October 1984,by Westinghouse

3. Young Gwang Unit 2 F.S.A.R by KEPCO in 1983

4. RELAP5/MOD2 Code Manual Volume 2 by EG4CG

5. Yang Gwang Unit 2 System Manual by KEPCO in 1980

6. Developement of Thernial-ilydraulic Safety Analysis Code

for Nuclear Power Plant by KAERI in 1989

7. Test Procedure for Net Load Trip Test in Yang Gwang Unit

2 prepared by Westinghouse

8. Instrument Accuracy by KEPCO in 1987

- 63 -

NRC FORM 335 U.S. NUCLEAR REGULATORY CO' SION I. REPORT NUMBER 12491(Aulgtwd bv NRIC. Add Vol.. Su"o.. Rev..

Nq~tA 102. W Addendum Nvnbert. It sanyj n9C/l112. BIBLIOGRAPHIC DATA SHEET NUREG/IA-0092

IS"e instructions on the reverse)

2. TITLE AND SUBTITLE

Assessment of RELAP5/MOD2 Computer Code Against the Net 3. DATE REPORT PUBLISHED

Load Trip Test Data from Yong-Gwang Unit 2 MONTH 'tEAR

June 1993

4. FIN~ OR GRANT NUMBER

L2245

S. AUTHOR(IS) 6. TYPE OF REPORT

Namnsung Arne, Sungjae Cho/KEPC

Sang Hoon LeeIKINS Technical'

7. PERIOD COVER ED finciissive Dam),

a. PERFORMING ORGANIZATION - NAME AND ADDRESS (ii NRC. provide Division. Office or Region. U.1. Nude.,a.'gpuiarosv COMnilto. endmawifing ddreu.-itcontwtoro' povidv now* and meJEng eddre.&

Korea Electric Power Co.

17-15 Yongieon-Dong,.Dong-Gu Taejeon, Korea

Korea Institute of Nuclear Safety P.O. Box 16, Daeduck Danji Taejeon, Korea

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (II NRC. Wye me~ as abo ve";. iicontrawror. Provide NRC Diviuion. OffLieo e~ egion. UA.1 NuclweerRfuatsorvCommission.

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D'.C. 20555

[10. SUPPLEMENTARY.NOTESI

11. ABSTRACT 0f0 words oeless)

The results of the REILAP5/MOD2 computer code simulation for the 100% Net Load TripTest in Yong-Gwang Unit

2 are analyzed here and compared with the plant operation data. The control systems for the control rod, feedwater,

steam generator level, steam dump, pressurizer level and pressure are modeled to be functioned automatically until the power level decreases below 30% nuclear power. A sensitivity study on control rod worth was carried out and it

was found that variable rod worth should be used to achieve good prediction of neutron power. The results obtained

from RELAP5MQD2 simulation agree wiell with the plant operating data and it can be concluded that this code has

the capability in analyzing the transient of this type in a best estimate means.

12. KEY WORDS/DESCRIPTORS IL ist words or phreses that wilt anit Mweeshen; In locadtin the APPovt.1 13. AVAILABILITY STATEMENT

RELAP5/MOD 2 Unlimited

ICAP Program 14.-SECUAITY CLASSIFICATIONz

Yong-Gwang Unit 2 (This Peg.)

1rhis Reroono

Unclassified

15. NUMBER OF PAGES

16. PR ICE

.%RC FORM 33542-891


Federal Recycling Program

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juviL t77

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

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NLN 353 (4)