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N 341 72-08

PAPERS PRESENTED BY JAPANAT THE MEETING ON VOID COEFFICIENTOF HEAVY WATE RMODERATED BOILINGLIGHT WATER COOLED REACTORS, PARIS.18-20 SEPT. 1972

KINETICS OF PROTOTYPE HEA`VYWATER REACTOR (FUGEN)

September, 1972

Power Reactor and Nuclear Fuel

Development Corporation, Tokyo, Japan

C 0 N T E N T S

I. Instrumentation and Control

I.1 Reactivity Control

1.2 Reactor Power Control

1.3 Reactor Protection

I.4 Nuclear Instrumentation

II. Analysis of Plant Dynamics

List of Figure

Fig. 1 Core configuration

Fig. 2 Safety circuit

Fig. 3 Schematic diagram of scram circuit

Fig. 4 Schematic diagram of. nuclear instrumentation

Fig. 5 Reactor control scheme

Fig. 6 Response for recirculation pumps trip

Fig. 7 Response for turbine trip (By-pass valve does not work)

Fig. 8 Response for reactivity disturbance at rated power

(corresponding to 10% power reduction)

Fig. 9 Response for reactivity disturbance at half power

Fig. 10 Response for reactivity disturbance with region power

control

Fig. 11 Response for steam drum level set point change

(0.1 m change of setting water level)

Fig. 12 Response for initial pressure set point change

(1 atm change of pressure setting)

I. Instrumentation and Control

In the FUGEN plant, the main instrumentation and control systems are

concentrated at a central control room. These system are designed to be

as testable as possible in operating conditions. The reactor power is

controlled manually over the entire power range and also automatically

in that over 40 % power.-

The steam drum pressure .in operating condition is maintained constant

by an initial pressure regulator (IPR).

I.1 Reactivity Control

Control rods and liquid poison in the moderator are prepared for

reactivity control. The moderator poison is used to suppress the excess

reactivity in the initial core and to compensate such slow reactivity

change as burnup.

I;2 Reactor Power Control

49 control rods. are distributed throughout the core as shown in Pig. 1.

4 control rods located in the center of each region are used for power

regulation. These four rods are called regulating rods. The position of

each regulating rod is automatically controlled by a signal of deviation

between averaged regional power and the setting power. The other 45 control

rods are used for reactor. shutdown with sufficient margin (one most ef-

fective rod stuck is considered), and some are also used for flattening

the power distribution.

I.3 Reactor Protection

The reactor protection system has two functions. The first is to give

an alarm, to block control rod withdrawal, to scram and to dump moderator

in off-normal operating conditions. The second is to close reactor

isolation valves, to operate emergency core cooling system; high pressure

core injection system (EPCI), low pressure core injection system (LPCI),

accumulated pressure core injection system (APCI), etc., in accident con-

ditions. Figure 2 shows the diagram of safety protection. A diagram of

the reactor scram logic is shown in Fig. 3. This system consists of a

dual one out of two circuits.

I.4 Nuclear Instrumentation

The reactor power is measured by -the in-core neutron monitoring systems

from the neutron source level to the power operation level through about

nine decades.

- I -

In order to obtain sufficient sensitivity to start up the reactor

and to secure detailed information of the neutron flux throughout the core,

all the neutron detectors are installed in the core. (see Fig. 1)

In-core neutron monitoring systems consist of the source range

monitor (SRM), the intermediate range monitor (11M) and the power range

monitor (PRII). The measurable ranges of these monitors are partially

overlapped and fission counters and chambers are used as neutron detectors

for these monitors.

I.4.1 Source Range Monitor

In order to obtain information on the neutron flux at start-up we

have four SRM. The detectors are movable fission counters. With SRM, the

neutron flux and reactor period are monitored.

I.4.2 Intermediate Range Monitor

In order to obtain information on the neutron flux in the intermediate

range, we have six SRM. Each consists of a fission chamber and a mean

square circuit etc. The measurable range of IR1I is changed adequately with

a range switch. If the indication of IRM exceeds over setting point in

the range, the reactor is scrammed for -protecting against an undesirable

power increase. IRM and SR11 are drawn out of the core in full power range.

1.4.3 Power Range Monitor

In power range, 64 local power range monitors (LPRM), 8 region average

power range monitors (R.APIM), 2 total region average power range monitors

(T.APRM) and the rod block monitor (RB4) are prepared to obtain detailed

information on the neutron flux throughout the core. Besides these, in

order to calibrate these monitors and to measure axial distribution of

neutron flux, we have the traverse incore probe (TIP).

LPRM are installed at 16 channels as shown in Fig. 1. Each channel

has four independent detectors at regular intervals vertically as illus-

trated in Fig. 4. Total number of LPRI is 64 (16 x 4). The detectors are

fission chambers (6 imm). Each detector assembly has a guide tube for TIP.

In this tube, a micro fission chamber is traversed to correct LPRI4 and

measure axial distribution of neutron flux.

R.APRT1 consists of the instrument which averages the signals of LPRM

in 1/4 core. 'When average neutron flux exceeds over 120 % of rated power,

R.APF14 gives scram signal.

- 2 -

T.APBM sums the signals of R.APRM.

RBE4 prevents the control rods being drawn out continuously by mis-

operation. When one of the 64 LPRM reaches the setting neutron flux, the

RBM works.

I : , I::

_ , _

I

CALANDRIA TANK

/ HEAVEY WATEI

/.~ 00 0 0

/ 00 0000

/ * 00 0 0 0 00/ 0.000 0000

+ ~ + +0 0+0 0+0 0+0 0

*0 0,0 0,0°0,0 0

I

0.0 0 ° 0

0 0+0 0+0 0 0

. 0+0 0o0 0 0

A A0.0000.00 00\

0 0,0 0+0 0+0 01

- . . .I1000 0 00o 0a0o0

0+00 0 0 0 0

\ *06 0+0 0-00 0-000 0 0000

0 00000

0 0'0 0'0 0'0 0'O 0+0 0+0 0 0 00 0 0 0 0 0 0 0

O'o00000*0 0-000 000 00

000000 /

0 FUEL ASSEMBLY 224 * LPRM 16 x 4+ CONTROL ROD 45 D r R M 6- REGULATING ROD 4 A S R M 4

@ I NEUTRON SOURCE I_

Fig. 1 CORE CONFIGURATION

- 4 -

I Mode switch dump I

Steam drum water level extremely low

I Container pressure high .

I He pressure high

-

Earthquake

Manual

Mode switch dump

Mode switch- shut down

Container pressure high

Steam drum pressure ah h

Steam drum water level low

Neutron flux hi gh

Neutron flux Indicate low

-

I Condeser vacuum low I

Turbin main stop valve close

[ Recirculation flow rate low

| He pressure high

[ Electric power loss

Nuclear instrumentation. -not work .

| D20 level low

Earthquoke

Manual

i Steam line radiation level high

Steam pipe rupture

Manual

Steam drum water level low

| Manua I --- }

Steam drum water level extremely low

Manual I. - f

|Container radiation level ~!high

Container pressure high

Steam drum water level extremely low

Manual

Air ejector radiation level high

I-

Steam isolation

valve close

0.N

E

* 0

*0

I.z

'0

0

Reactor core isolation

cooling operation

Container ventilation

Emergency core coolingsystem operation

IEjector outlet valve close

Fig. 2 Safety circuit

-5 -

Line A L in e B

Fig. 3 Schematic diagram of scram circuit

- 6 -

I

-1 0

Construction of power range -instrumentation

Each RPRM has twoindicator

I

Region averaged power. range monitor -

R..A PR M .

Same us left; -4 ; - . , I

.- Sme~as right

- . i

1 1 1 1 I-I

I4

1 .4

I* *oat tpw

\ Local powl

4

I I

or r(

I 4p 4I I

I 4 I

inge monitor (LPRM ) 16 x4 64)

-Power range monitor assembly ( T I P guide tube and 4 LPRM )

Region 2 Region 4Region i Region 3

Fig. 4 Schematic diagram of nuclear instrumentation

j *--

II. Analysis of Plant Dynamics

The analysis of the plant dynamics was studied using the ATM (Advanced

Thermal Reactor) plant dynamic analysis code, "TETRAC".

Some of the analysis is shown in Figs. 6-12. The following points are

noted. The plant has stable response against perturbation due to change of

setting pressure, setting drum water level, setting power level, change of

recirculation flow rate, and other perturbation in the normal operation

mode.

Stable reactor power control will be established against some fluctu-

ations in terms of the Doppler coefficient or the void coefficient from

their respective design values.

Fig. 5 shows the reactor control scheme. As mentioned above, reactor

power is controlled with control rods. Steam drum level is controlled

with three signals: drum level, steam flow and feed water flow.

Fig. 6 shows the response for recirculation pumps trip. In this

case, all of four recirculation pumps are tripped. After about one

sec.reactor is scrammed owing to low recirculation flor wate (of Fig. 2)

Fig. 7 shows the response for turbine trip. When the turbine is

tripped, the main stop valve is closed and the turbine bypass valve will

work. At present we assume that bypass valve does not work. The steam

drum pressure increases and reaches scram level (cf Fig. 2). Negative

void coefficient brings neutron flux increase till the reactor is scrammed.

Fig. 8 shows the response for reactivity disturbance at rated power.

Negative reactivity corresponding to 10% power-reduction is inserted by

step. The neutron flux undershoots but soon are stable.

Fig. 9 shows the response for reactivity disturbance at half power.

Positive reactivity corresponding to 10% power-up is inserted by step.

The neutron flux overshoots but soon are stable as described above.

Fig. 10 shows the response for reactivity disturbance with region

power control. Negative step reactivity is inserted into No.1 region.

The regulating rods in each region move to maintain the reactor power.

The reactor power soon recovered to the rated power.

Fig. 11 shows the response for steam drum level set point change.

The setting point of drum level is changed by 0.1 m. The level controller

works and after about one minute water level becomes stable.

- 8-

Fig. 12 shows the response for initial pressure set point change.

The setting point of steam pressure is changed by 1 atm. Steam pressure

controller works and soon all phenomena become stable.

- 9 -

Pressure setting

I DemandI I

GeneratorL_0n\ I Turbin by-pass

'Fe d water pump l

I 0 15) \Reactor

Recircu lotion pump

Fg 5 ReDZO leve scemcnrole or

Fi g. 5 Reactor controlI scheme,.

CO I= (D°I-

0

W > ¢, \ Maximum fuel temperature/

00 = N RN ; \ 2Minimum critical heat flux ratio0_

o E. -._

._03iO(x°S F j ~e~ flow rat

= 8 0 Surac heatX\< $>\~E _E

- 0 -O.

No o 262

E3 E -

*. . .- .0

E E-

C 04 8 12 1 2 0

Time after AC power loss (secl

Fig. 6 Response recirculation pumps trip

0T20:IrI-

0If)0

LC)~

EoC%

0CLi T

0)

01) O

C )_ _

I CM3E Ci_ I

-o

E0

0

0

-o0)

-0

- o

E

Cj

- 0-

0

.Lfl

0)

2

level

CL

Time after main steam valve close (sec

Fig. 7 Response for turbine trip ( By-passvalve does not work)

11 I

co 0 r CD - ° _ _ _ _ _

. I . '

co - ! -- 1 - Ico I

- . urface heat flux i Recirculation flow

- o 0

xEo E .-o __ _______________ --- ___________ _____

t .flux1I Id

. CSteam drum pressure V Steam drum water level Minimum ir~tem drm ;heat03

- - - - -- - - - - -- T__- - - -

E X Steam drt. oE, E.. ° -Minimum critical heat flux ratio

E e ~l n sI . I, I ,I.o ' . .

CD __ _ _ _ _ _j _ _ _ _ _ 60___ _ _ J _ __0 10 20 30 40 50 60 70 80

Fig. 8 Response for reactivity disturbance at rated power(corresponding to iO % power reduction)

90 100lime Nsec)

* P

A0DE

C-j

21-

(0

Ne

0

Q)

U)

LDi

.4-

C0

E:

0'

90 0ooTime (sec)

Fig. 9 Response for reactivity disturbance at half power

Sr-0

dQ -Li - -J._ _ _ _ _ - _

ID IS

-Steam drum water level

lI

II

, ..

.

0UI'

0

0onj

\3n

01-

E0

0.

0

C0U-

I,-

Cj

0

0

CS3

* If

E I0

LQ

I

-o- I0

a)

3:

a)

0* Lfl

-Steam drum pressure

coW

0N'

0N

NCMI,-

Ilr------r----------Il

! ~I!1-

0 o! 0

0

0kr

21

00

a) oM

tnina)CM.

E

-o0

C,,

0

or

I

x

0a)0

J= Rf

a), .

0

cn -

0

C00. _

* 0.

00. .oU

*. o

Q

.-

C0

z0

o~

0 o

.'.-* * I-

Surface heat flux |Neutron flux t) -, )

I----

. ~No.l.region control rod displacement

No.E.lZ region control rod displacement

I -- _-_____

II

I

1.------ _________.

~~~~~~~~~~I__--_----___-____

I

Ii

i

I

0) NP

I-No. .region control rod displacement

I I . I

I I

csC 10 20 30 40 50 60 70 80 90 100

- 'Time in sec

Fig. 10 Response for reactivity disturbance with region power control

ocO~ r -

o I coo -_ _ov ¢,

_ OSurface heat flux Recirculation flowo 0 e ron 0 o -- - - - - - _, - - - - - - - -ta rmwae ee

o E _ >

co ---- - -0_ I------LO O We 1

00

Surfacee (e se lo

FSteam drumxatio leveln( LO LO c c sl

E I -

= 0

LOC -. I ta drumr Ste eamdrmprssr00

E) FC) -

('a (0

0D 10 20 30 40 50 60 70 80 90 100Time (sec)

Fig. II Response f or steam drum level set point change(0.1 m change of setting water level)

-I

CQ

LOco 2o

co

I

I

a

a

0.

E2

E

C,,

>C.

=3 Cuj.4-

C)

a,

08

CnI-

_I_

3:

S0

a,Ct)

-

0-

.4

-0

C.)

~4

Surface heat flux i

I I I '

I_ ____ _ __-

Neutron: flux

o' 1- ! _ _

\-Recirculation, flow

co ----- --- --oeaw flow

o I I I

.,Steamd low pr ssur- - -- - - - - I- - - - - - - -

I _ _ _ _ _ _ II_ _ _ _ _ _

01

0o

0 10 20 30 40 50 60 70 80 90 100Time (sec)

Fig. 12 Response for initial pressure set point change ( latmi change of pressure setting )