Depressorization accident analysis for the FTTR by the TAC-NC

036&5442/91 53.00+0.lm copyright (0 1991pictvmoll Rru pk

,%cqyVol. 16.No. I/2. pp.471-480.1991 ~ntaiinGratBritia.AUri~u-cd

DEPRESSURIZATION ACCIDENTANALYSIS FOR THE HTTR BY THE TAC-NC

K. KUNITOMI, I.

NISHIGUCHI, H.WADA. T.TAKEDA,

M. HISHIDA, Y. SUDO, T. TANAKA,and S. SAID3

Department Japan Atomic

Energy Research

Oarai-machi,

Ibaraki-ken,

- The two-dimensional

Abstract modified

from

temperature accidents

the

transients of

in

special

function

of

of

Increases

the

TAC-NC code results

transients

fuel

temperature, temperature

of

forced

cooling

Test

Reactor)

includes

by natural

pressure

a

convection

vessel

during

results

rapidly

however, since

thermal

due

remains

most of

to

capacity

of

of

the

simulated

core

results. accident

the HTTR.

The maximum fuel

the

reactor

scram

decay

heat.

below

the core

by

ones of

a depressurltation after

that

out

the experimental

with experimental

during

after

was carried

with

by the TAC-NC code for

slightly

large

loss

engineering

transfer

in

Analytical

decreases

fuel the

the

test.

evaluated

temperature

heat

were in good agreement

Temperature were

of

TAC-NC is to calculate

The TAC-NC code

regions

the analytical

ingress

temperature

case

code

in order

accident.

Verification an air

the

to calculate

the depressurlzatlon comparison

analysis

accident.

and colder

Japan

code TAC-2D

the HlTR (High Temperature

hotter

Institute,

319-11,

thermal

analytical

such as a depressuriration

between

of HTTR Project,

the

decay

The

initial

and

maximum maximum

heat

is

absorbed

in

the

core

codes

for

the H’l7’R.l

graphite

in and

reflector.

1. INTRODUCTION The JAERI has been developing licensing verification u)Y

of

the of

the

HTTR, codes

the

the safety JAERI has

as well

as

analysis

been

for

IYb2--cF

471

carried

establishment

out

the

necessary

and upgrading

For

the

R&D for of

HTGR

K. Kmmxa

472

technology

2.3

basis.

evaluated

by these

In the residual

case

heat

of

is

VCSs (reactor

the

pressure

primary

the temperature

Vessel

pipe

limit

although a break

inherent

safety

features

of

the depressurization

area

accident

postulated

accidents

(depressuriration

of

Systems)

and fuel

the

pressure that

reactor

were

temperature

cooling

there

is

accident),

pressure

decreases

and forced

the HTTR is

according

stopped.

no forced

by

does not exceed

rapidly is

vessel

as

One of

cooling

system

the

during

accident.

The TAC-NC code was developed during

for

surface

coolant

through

of

rupture

Cooling

leaks

analyses

codes.

removed from the outer

coolant

core

safety

In addition,

newly developed

et al

the depressurization

to calculate accident

the

temperature

and to confirm

transients

the inherent

in

safety

the

feature

the HTTR.4 This

by the

paper

presents

comparison

depressurization

the analytical

with

the

accident

of

model and verification

experimental

results

results

and analytical

of

the TAC-NC

results

of

the

the HTTR.

2. ANALYTICALMODELOF TAC-NC The TAC-NC code composed

of

Fig.

the

of

core

through

conduction, 1 shows

accident

radiation,

the

the HTTR.

by primary the stacked

radiation

was developed

and finally

thermal

to calculate forced

and natural

and hydraulic

Under normal condition, forced

fuel

coolant,

block

to

and the

the thermal

1 to

removed from the outer

during heat

2 4; of

generated vessel

of

the depressurization

is mainly

pressure surface

heat

Fig. 1

‘.

Ther-ma1 and hydraulic

the reactor

behavior

during

depressuritation

is

accident

from

transfered

pressure

.. .

.

removed

by conduction

by VCSs.

.

behavior

convection.

behavior generated

reactor

and hydraulic

and vessel

473

Safety and licensingissues

In

the

case

immediately

of

the

by detecting

secondary

coolant.

stopped

decrease

After

and residual

radiation

turns

downwards

of

the

heat

accident,

differential

reactor

is

Coolant

by VCSs.

the core,

depressuriration

or air

pressure

scram,

removed

from

the

ingressed

surface

inner

of

forced

the

between

is

the

scrammed

primary

by primary

by natural

and

coolant

convection

is and

the break area flows upward in

the reactor

the

reactor

cooling

core

through

downward under the top head of

along

the

reactor

pressure

vessel

pressure

and flows

vessel

by natural

convection. The TAC-NC can calculate reactor code

core

and pressure

which

was developed

dimensional

the

of

network

calculation

each

by GA company

convection

The flow

equations

steady-state

under these

and

transient

condition.

temperatures

It was based

to calculate

transient

on

in the

the

TAC-ZD

temperatures

in one

problem.5

Natural model.

the

vessel

of

passage

core

is

which consists

the

to be solved

the

natural

are

simulated

of

flow passages

circulation

(see

the steady-state,

and continuity

equation,

by one-dimensional and plenums

Fig.

4).

of

state

is

The basic

one-dimensional

equation

flow

network

assumed flow

network

momentum equation

and energy

equation

for

for

for each

plenum. The momentum equation

has the fallowing

form:

x.

ap

1

DUi

Dt---

for each passage

- ( 1

Lpcg

t ci

I+

IUJ

ui

di

where c :inlet

resistance

d :hydraulic P :fluid

diameter

P :fluid

flow

the flow

passage

passage

velocity

over

the cross

section

density

X :friction

factor

passage

z :coordinate

D Dt-

of

in the flow

pressure

u :average

i :flow

coefficient

atu at

number

of

flow

direction

k-

3. TAC-NC CODE VERIFICATION Verification analytical ingress pipe

results test

rupture

was

of

the

TAC-NC

with

the

experimental

carried

out

to evaluate

accidents.

code

was conducted ones transient

of

the

by the air

behavior

comparison

ingress of

the air

test.

of

the

The air

in the case of

414

K. KUN~TOMI et al

3.1.

Test

Apparatus

Fig.2

shows

apparatus. which

and Procedure6 a vertical

The test

consists

temperature

apparatus

of

a reactor

plenum,

simulated

bottom covers a primary

of

and

a about

core

cross

l/IO

pressure

simulation

a plenum plenums

vessel,

the HTTR and measuring

sectional

scale

simulator,

top and bottom

the reactor

pipe of

is

horizontal

inlet

view

of

the

test

model

of

the

HTTR

simulating

simply

the high

to

top

and

corresponding

to

corresponding

and outlet

pipes

the

system.

t WATS

11 d;yJ

TOP Cx?

\

LWER

FENUM

BOTTOM COVER

- jb1NLE-r

PIPE

f Fig.2

Vertical

The reactor between 1200 air

core

heater 12.7

mm,

simulator

pipes.

cross

sectional

consists

of

inner

The length,

view of

43 heater

diameter

test

pipes

apparatus

and

and power of

The heater

mm and 36 W respectively.

pipes

raised

fiber

insulations

a heater

pipe

are

the temperature

of

up to 45O’C. Thermocouples

temperature between flow

and horizontal

transient

hotter

The

test

core

the

is

nozzles

pipes pipe

region

at the inlet

procedure

and outlet

(A) (B) Simulated

after

and colder

meter installed Inlet

on the heater

of

and other

structures

rupture.

Flow

the test

vessel

are

rate

of

installed natural

was measured

to measure convection

by the ultrasonic

pipe.

as follows: are closed

temperature

is

with raised

blankflanges. up to

the

maximum 450°C

by the heater

pipes.

(f-3 After state

the

temperature

condition,

ingressed

into

distribution

the blankflanges the test

vessel

in the test are

vessel

is

attained

removed from the nozzle

by diffusion

and natural

in a steady-

and ambient

convection.

air

is

Safety and licensingissues

3.2.

Analytical

Condition

The experimental (A) Total

heater

(C)

Inlet

conditions

temperature

Figs. Natural

along

compared with

of the heater of

the cooling

flow

the test

the analysis

are as follows.

200 W

3 and 4 show

circulation

passages

and Model

power is

(B) Maximum temperature

475

the

pipe

is

3OO’C.

water

is

2S’C.

analytical

channels vessel,

pipes

_lilz __ ___.__ __._ .._ . _ _. __ _ _ .-_ _.~~

model

consist

of

and natural simulated

convection core

flow

flow

network.

channels,

flow

and plenums.

c

L

PLENUM DRESIMULATOR ‘LOW Rwv%Es

/o

,FLow PASSAGE

FLOiV DIRECTION

,LOWER PLENUM AHJTLET PIPE

ALORG TEST

VESSEL

_OWER PLENUM 1

.BOTTOM COVER /INLET

PIPE

II INLET

._ ._ liLlHl

Fig.3

Analytical

3.3.

Analytical

(1)

Steady-state Fig.

OUTLET PIPE

model of

test

temperature

of

the

Figs. the heater heater

the inlet

about

temperature

and outlet

Flow network model of

test

apparatus

are

in

in the test

pipes

good

agreement

the temperature

2 hours.

in

the

are opened.

test

vessel

before

The analytical

the

results

agree

vessel and analytical

temperature

and analytical

30°C after

vessel

ones.

Analytical

pipes. to predict

experimental

Fig.4

temperature

6 and 7 show the experimental

pipes

difficult

in the test

steady-state

with the experimental Transient

apparatus

Results

5 shows

blankflanges well

PIPE

RAMAL MRECTION

t -

(2)

UPPER

AORE SIMULATOR

with

the

and middle

experimental

at the top position

temperature This

temperature

at the bottom

temperature

at

the

increase

top

of of

at the surface position It

one. the

heater

the heater

was attributed

of

is

somewhat

pipe.

pipes to heat

of the Both

increased transfer

K. KUNITOMIet al

476

from

the

bottom

temperature

and middle

however is about

position

to

top

by natural

15°C as low as experimental

convection. one.

INSULATION

CORE SIMULATOR 500

OO1

0.05 I

4

I 0.15 II

0.I1

RADIAL

0.2 111

1 mJ

DISTANCE

sits

CORE SIMULATU3

SUS

500

0

0,2

BOTTOM Fig.5

Steady-state

0.4 AXIAL

0.6 DISTANCE

temperature

(ml

TOP

in test

apparatus

*

EXPERIMENT(MID&E ..__.

-

EXPERIMENTITOP);

-

EXPfRIMENTiBOTTOM

---- ANAlYTICALRESULTS

CROSS-SECTION

0' 0 Fig.6

Relationship of heater

I

5

I5 TIME

between analytical pipe

temperature

I

I

I

I

10 ELAPSED

20

25

(h 1 and experimental

( center

pipe

)

results

Analytical

477

Safety and licensingissues

Fig.7

Relationship

equivalent

thermal

average

of

depends

on both

the

insulations. the

in

which

however,

Hence,

for

conductivity,

the

and vessel

Flow rate Fig.

result

of natural

of

flow

might

due

result.

is considered

to be lower of

equivalent

pipes

of

conductivity of

HTTR,

The value in

the

convection

in

of

thermal

change

such

as

to be conservative

to be higher

the

core

conductivity

on

in the insulation.

properties

are considered

be evaluated

natural

The

volume

from the

permeability

the local

dependence

the

reason.

and insulations. and

the permeability of

following

is calculated

thermal

considers

)

the

simulator factor

results

than real

thermal so

that

temperature.

circulation

experimental

the calculation

the

pipe

to

simulator

the dominant

analyses

the flow

rate

primary

core

and emissivity

temperature

( peripheral

due

not consider

capacity

8 shows

the

analysis

safety

heat

and experimental

such as the heater

is

does

EXfWlMENT (BOTTOM1 ANpcMlCAL RESULTS

the core

of

affects

temperature,

is

of

simulator

The present

(3)

temperature

Permeability

insulation

* ---

temperature

temperature

simulator.

fuel

pipe

conductivity

core

EXPERIMENT lM1DDl.E) EXPERIMENT (TOP1

between analytical

of heater

The difference

-o---01

rate to

This

of

natural

natural is

convection

convection

because

is

analytical

than the experimental

the graphite

in the test

oxidation

of

a little

pressure

loss

The results

one.

vessel.

Analytical

higher of

than

the

the simulated

core

are conservative

for

the core.

4. TEMPERATURE TRANSIENTSDURINGDEPRESSURIZATIONACCIDENTIN HTTR BY TAC-NC The major design residual large

heat

removal

heat capacity

feature after

of

the HTTR is

faults

due

and low power density

that

there

to inherent of

the core.

is

safety

no forced

cooling

characteristics

such

for as

K. KUNITOMet al

478

EXPERIMENDU RESULT

ELAPSED Fig.8

Relationship

between analytical

of

of natural

flow

The HTTR cooling safety the

rate

system

system

stand-by

residual

from

core

around

the

reactor

in the case transients

the TAC-NC in order

to

of

gas temperature

the

a single

behaviors

the properties

temperature that

the

the maximum fuel

to

decay

The

of

safety

the

to

Both

MCS. in

order

serve

in

remove VCSs are

to

to cool

accident

cool

the

the reactor

are evaluated

characteristics

before

power

and the one of

the graphite

of

the accident

for

the

with HTTR.

the analysis

the VCSs are assumed to be failed

heat.

in the accident

It,

after

fuel

remains

the

increases

beginning

up to

the fuel

conductivity

change

the relevant

as

is

of

core

life,

life.

at

about fuel

530°C

at

about

temperature

and the maximum slightly

30 hours temperature

30 hours of

so

9 shows

The maximum fuel

below the initial

under the allowable

up

in high

Fig.

scram and then increases 138O’C

burn

properties

structure.

at the end of core

the reactor remains

uses

of

in each evaluated

temperature

however,

temperature It also

at

rapidly

The peak

The analysis

was obtained

temperature

decreases

depend on the magnitude

such as the thermal

exposure.

temperature

was started.

vessel

gradually.

inherent

auxiliary

The ACS is

accident.

the depressurization

the accident

temperature

vessel

temperature

after

and irradiation highest

pressure

accident

in

operation

and they

vessel,

an

(VCSs).

and operated

a trouble normal

(MCS),

system

operation

the inherent

failure.

Their since

is

the

and outlet

are 95O’C and 30 MWrespectively

results

considering

systems

the depressurization

during

confirm

simply

cooling

reactor

when there during

Temperature

designed

vessel

normal

rate

shield

outlet

the

and experimental

of a main cooling

100% flow

and the core

Initial

therefore,

during

( h)

convection

and two reactor

condition

at each

biological vessel

(ACS),

heat

operated

is,

It consists

characteristics.

cooling

TIME

550°C.

after

due the

1495OC.

and decreases

Safety and liising

I

479

issues

t END Cf LIFE f

UAXIMUM PRESSURE VESSEL TEhi%iii% 1 BEGINNING OF LIFE I

200-

Oo In

20

40

11

3

60 8

L

ELARSED TIME Fig.9

Analytical during

results

of

fuel

depressurization

80 n

1

1

10

(h)

and pressure

vessel

temperature

accident

5. CONCLUDING REMARKS

(1)

Verification results

of

of

temperature thermal flow ones.

the TAC-NC code

the air

ingress

due to

was properly

natural

The results

were

out through

It was confirmed

were in good agreement

conductivity

rate

was carried

test.

with

for

results

the calculation.

was a little

conservative

comparison

the analytical

the experimental

used for

convection

the

that

than

of

the of

when equivalent

Analytical

higher

the evaluation

with results

the

result

of

experimental

the graphite

oxidation

in the core.

(2)

The maximum fuel does

not exceed

the core

1. S.Saito,

decay

Conference

Status

al.,

Test

Reactor

et al.,

Safety

of

the

temperature

absorbed

HTTR during in this

by the large

of HTGR Development

“Design

technology

3. S.Maruyama,

is

on HTGR, Dimitrovgrad et

Engineering Reactor

the intial heat

“Present

2. S.Saito,

temperature

(HTTR)“,

of

Dimitrovgrad

accident

is because

of

most

of

the core.

11th International

1989.

Consideration

In-core

This

capacity

Program in Japan”,

IAEA Technical

and siting,

“Verification

analysis. thermal

USSR,June 19-20,

and Safety

the depressurization

in

the

High

Committee Meeting USSR, June 21-23,

Thermal and Hydraulic

Temperature

on Gas-cooled 1989. Analysis

Code

K.KUNIYUMI etal

480

FLOWNET/TRLJMPfor the High TemperatureEngineeringTest Reactor at JAERI", to be published in NDRETH-4,October1989. 4. K.Kunitomi, et al., "Two-dimensional Thermal Analysis Code TAC-NC for High TemperatureEngineeringTest Reactor and its Verification",JAERI-M 89-001, 1989. 5. S.S.Clark and J.F.Petersen, "TAC-2D A General Purpose Two-Dimensional Heat Transfer Computer Code", GA-9292, September 1969. 6. M.Hishida

et al., "Studies on the Primary Pipe Rupture Accident of a High

TemperatureGas Cooled Reactor", to be published in NURETH-4, October 1989.