l ^ ^ ' ^ ' ^

S I F O B T Но 117О/ПА/2Е

M£A0UH£l£gI3TS 0? THE SK-10 TYPE

BSiCTOB FUEL WHS Ш ША-Л COSE

J . Aleksandrowiez

M» Csernievekl

This rep,.?t baa been reproduced direotly from the beetavailable copy

ЕНФОРВииЩОННЫЙ ЦЕНТР ПО ЯДЕРНОЙ ЭНКРГШУполяомочввввг® Праввгвльетва ПНР

QO Иовзз&аовашш Ядержой Энергии

Д®ер«щ Культура и НаукиО ó i t я i

Available frost

fTOCLBAR SHXRGY Hf?ORHA¥ION CSNTER

of the P o l i s h OaveraE»nt Cozaajisaioner for Use

of Huelsar Energy

Palas© of Culture and Sc ience

Warsawg Poland

1 rosprowadami

L?K£MACJI 0 EHESQIIPvi2as»ooallse Ssądu do Spraw WjkorsyatAnift Bnergii Jądrowe

Wydaje

Instytut ЪайвЛ Jądrowych

Nakład 4 <? ega.. Objętość ark. wyd.Ark. druk 5,5 , Data złoteniaprz«z autora 30.11970г. , Oddano do druku

5 Ж 1990r D\-uk ukoAc гопоSP-09/SO/66 Zaia. nr 135/W

INSTITUTE OF NUCLEAR RESEARCH

TEMPERATURE USASUBKMENTS OF THE ВД-10 TYPE

REACTOR FUEL RODS IN SVA-4 COB!

POMIARY TEMPERATURY PECTÓW FALBOTTCH TYFU EK-1O

V RDZENI J REAEPOBA

ИЗМЕРЕНИЕ ТЕМПЕРАТЛ^Ы ТЕПЛОВЫДЕЛЯЮЩИХ СТЕРЖНЕЙ

ТИПА ЭК-IO В АКТИВНОЙ ЗОНЕ РЕАКТОРА

J. Aleksandrowi сz

М. Czarniewski

Abstract

The measurements programme of temperature of the

EK-1Q type of reactor fuel rods In steady and tr<*n3ient

states of heat transfer conditions has been performed

with EWA-2 and ВЭА-4 cores» Resulta of measurements

performed with Щ - 2 core were reported previously in

tho IBJ Report Mo 879/3Q/H. In thle report: the B«aeureiaant

reeuifi obtained in flxperiment performed with ISSA- core

are presented. The Bt'A— coro conaieted of fuel assemblies

equipped, with variable croee-eection tubes inside of

which the fuel ro&o Ład bees placed*

Two instrumented fuel ro&B war© applied for measure-

menta : one equipped with 5 surface thermocouples installed

in the fuel cladding , and second equipped with thermocouple

for fuel rod cor© temperature aeaeuremaut.

In steady states, the KK-10 fuel temperatur© was

measured for different cooling conditions and different

reactor power levels. Transient tests ware Initiated by

cooling water flow stopping azsd by гашр reactivity addition

into the core.

StresacaenJe

Prog raiB pomiarów tensperjitury prętó® paliwowych typu

EK-10 w ustalanych i przejściowych stanach wymiany ciepła

został wykoaajay w rdzeniach E»A-2 i ША-4, Rezultaty po~

miar6w wykonanych w rdseniu ША~2 były opublikowane wcześniej

w raporcie-IBJ Жо 879/Х1/Н.

Rapoi-t ten obejmuje wyniki pomiai-ó^ uayalcane w rdge—

niu Й£А«4. Rdzeń BSA-4 składał się a sekcji pallwosrych

wyposażonych w zwyżki9 wewnątra których, uiaieszczone były

pręty paliwowe. Jeden a nich został aryposażony w 5» termopar

Łainstalowanych w koszulce elementu paliwowego. Drugi

został wyposażony w termoparę mierzącą teispereturf wewnątras

paliwa. Pomiary w stanach ustalonycn byty przeprowadzone

dla rożnych warunków chłodzenia i dla różnych poziomów

шосу reaktora» Pomiary в stanach przejściowych obejmowały

przypadki zatrzymania przepływu wody w pierwszym obiegu

chłodzenia reaktor* oraz wymuszenie wzrostu mocy przez

szybkie liniowe wprowadzenie reaktywności,

ii

Ажнотац&я

Программ гзеервяий ¥»мпературы твевов твпа ЗЯ-Юв етадюнйргзоЕ и дерехвдноаз режимах бьша проведана iактявявх зонах реактора КВА-2 г ЙВА-4. Результаты ш&пе-peiifl в зон! ЕВА-2 был! опубдоковады раньш в докладаШШ » 879Д1/Р.

Настошцй доклад содержит рввудь^аты измвреашй в эоееЕВА~Л4 которая соссояле ES тежнолоретескнз: c e s n ^ $ сеэб-генных профЕОрованныш грубкамз8 окрушштшш каждый шэ?8шювцделявдюс aitMesfOB., ДЛИ шгшербНЕМ аршиевнлгсь дга

снабженные ^ермоааршж, Ossa шп низ содержа 5зпресованйшж в оболочку тв#жч, во втором - кз-

нерялась feanepatypa ввутрм fB©.ie. йзыареаал в стационар-ном режим© прсводмдмсь прш {жзличны!; условию, охлавдсакаpeascopii i при разных уровнях иоцеосха» В оереходнш: р@-кюаах мссдедовалвсь случаш аожпой ocTasossis г^аркулядвн ви£ тажже увелжчанжя ЙОЩВОС М за сч©т ducfporo

i i i

1. INTRODUCTION

Measurements presented in this report were

carried out i-a October, 1966, ав а part of the prograame

of the BE-10 fuel rod surface temperature investigations

at the various heat transfer conditions<» Measurements

performed in ЕИА-2 reactor core are described in Eef.1.

In this report the fuel rod temperature measurement results

performed with BiA-4 core are contained. Also have been

included the results of measurements of the fuel rod core

temperature performed in July, 1967. These results allow

to estimate the main ЕЙА-4 core thermal properties in the

steady and transient states.

The detailed description of the ЭвА reactor and the

data concerning the EK-10 fuel rods can be found in Refs«2

and 3» In EWA-4 core the fuel assemblies equipped with

variable cross-section /Venturi/ tubes are applied - Fig.1.

Thia type of fuel assembly had been constructed as a

result of investigation described in Ref,4. The reactor

hot channel temperature decrease in effect of Venturi

tubes application is a result of water velocity increase,

and fitting the water velocity changes along the core

to the heat sources distribution. On the basis of preli-

minary results of the fuel surface temperature measurements

/Ref ,5/ the nominal power of EWA reactor had been increased

from u W up to 4 W for the same value of cooling water

flow rate.

The 1Я7А-4- core configuration saown on Pig.2, consist

of 50 fuel assemblies containing 751 fuel rode, There

were two positions of instrumented fuel rod chosen as near

as possible to the centrum of the core. These positions

mark.ad by I and II in Fig*2, correspond to the measurements

performed in 1966 and 1967 years, respectively. Insertion

of the instrumented fuel rods in exactly the same place

of the core could not be realized for technical reasons

of reactor operation,

2. STATIC MEASUREMENTS

In this section, the results of the fuel rod tempe-

rature measurements for different values of reactor power

and cooling water flow rate at the steady operating cond.it101

resulting tŁe steady state of heat transfer?have been

presented, and data concerning neutron flux distribution*

control rods calibration and temperature effects on

reactivity have been included. too«

2.1. Instrumentation

She instrumented fuel rods with marked positioning

of thexmouples are Jhown on ]?ig.3. Fuel cladding tempera-

ture was measured by means of Thermocoax 2AB Ac 05 TJ -

type theraocouplee with factory mada Inslutated hot

junctions.

2

These thermocouples were inserted and then pressed intc

the grooves milled in alluminium cladding.

Distance between hot Junctions onxthe rod equipped with

surface thermocouples only, was 90 аш. Thermocouples are

denoted by the t » tpi t,t t t % according to the counting

succeeion from a bottom core-plate - upwards. Thermocouple

denoted by the. t, was placed in the middle of fuel length.

Inlet water temperature was measured by Thermocoax 2ABAC 10

thermocouple (tg) with hot junction positioned about 100 mm

above the upper ends of the fuel elementv, Fuel г0.1 repar^

for core temperature measurements wus equipp-ia with thei-

mocouple of lABAc O 35 type with encapsulated and insulared

hot junction, Thia thermocouple ( t^ ) was mounted tusló^

of drilled hole in the fuel rod .-n the night where, for

E3?A-4 core, the maximum surface temperature was observed.

Second thermocouple saj installed in the fuel cladding

on the same bight. The position of this thermocouple (t'4)

corresponded to the position of tne t^ thermocouple on the

fuel rod equipped with surface thermocouples.

All thermocouples installed on the rods were led in the

cladding end then went through the elluminium guiding

tube festened to the rod. On the upper and of the guiding

tube the connector was placed. From this connector the

compensation cables are fed to the measuring apparatus

through the reactor covering penetration holes..

Thermocouple voltage signals were recorded using the

Hartman-Braun Photосошрепваtor DTA - Universal with six

neaaurixig channels. Reactor power was estimated on two

ways: by aeane of the ioni*stion chaabers (from reactor

control equipment) and by the N-16 activity meter (with

gama detector positioned on the tube of primary cooling

circuit) . Both aethods were calibrated by heat balance

using the values of water flow rate and temperature Increase

of core cooling water - Beasured by instruments contained

1л the stationary reactor control equipment.

2.2. Neutron Flux Distribution

Vertical distribution of the thermal neutron flux

in the place cf instrumented fuel rod ( In water ) - position

I - is shown in Fig.4.

On the graph thermocouple positions were marked by

vertical dashed lines with thermocouple numbers denoted.

Neutron flux distribution was measured by activation of

«be copper wires.

2.3* Control Rod Worths

The automatic control rod (AS) was calibrated by

aeans of period method. Remaining part .of the control rods

was calibrated by reference ( on low power level} to the

AR - worth in linear part its integral characteristic.

DisplaceBent of the control aad &afety rods in BVA-4 oore,

латке Pig.2.

S&ch of the 1ER and 2ER systems has two neutron absorbing

rods coupled with one driving mechanism.

AR and 3HE - systems have only single rode . The calibration

results are contained in the Table 1.

Full range of the control rods displacement is 60 cm

(counted from the aero point above the core) « The value

of 30 cm corresponds to the middle of fuel length.

Table 1

Control

rod

system

AH

1RR

1HB

2RR

5RR

Total worth

0.3=8 0.436

Ć.53 Ьш>.

- -

3,00 4.00

:._i 1.61

Range

linear

cm

30-50

26-43

30-40

25-41

26-45

of j Slope of

part в

%

0

0

0

г

0

/cm

а010

.071

.060

.089

.036

linear

art

$

0.

0,

•') .

0.

0.

/on

01;

^9;

087

114

046

-.euiarks

at 2ER on35 cm level

Table 1. Control rods calibration data

2.4. Temperature Sffeots on Reactivity.

The homogeneous temperature effect i.e. the influence

of the core temperature on reactivity was measured at

very low reactor power level. Therefore practically no

temperaturę gradients were in the core I.e. fuel temperature

was equal to the core water resperature. The reactivity

change ae a function of average water temperature ID the

core wae measured in the temeprature range between в С and*

51°C. The results are plotted in ?ig.5«

Power effect on reactivity i.e. reactivity decreaee as

a function of power level for different cooling water flow

rates was measured at the constant value of inlet water

temperature equal to 22°C £ 1°C. The results are plotted

in Fig.6» In n«asuremant at convection flow the condition

of constant inlet water temperature was not fulfilled.

Toe water temperature during this measurement rlsed from

22°C up to 35°G ( it was no possibility to keep it constant),

So the curve for Q « 0 in Fig.6 illustrates also the tem-

perature effects on reactivity caused by the increase of

the inlet water temperature»

2.5. Temperature Distribution along the Fuel Rod for

Different Reactor Power Levels at the Constant

Cooling Water Flow Rate.

At this measurement the inlet water tempearture t&

was maintenend constant on the value 29°C ± 1°C. Water

flow rate Q waa equal to 9Ю nr/h . Control rods were од

following positions : 1ЕЙ - 33*6 cmj 2BE - 33e6 cm; Afi -

5HE - 30 cm. Power effect on reactivity was compensated

by the 1RS and 2RR systems ( changing their positions

from 3J«6 cm to 32.1 cm ) .

Results of measurements have been collected in the Table- 2

and plotted in ?ig*7.

Table 2

p

1.075

2.250

3.380

4.500

5.480

*1

°C

40.0

52.6

64.1

76.4

66.6

T 2

°0

40.0

52.6

64.0

75.0

80.O

ъ°G

40.0

51.4

63.4

7^.0

85.0

4

°G

41.0

67.0

79.0

90.0

H°c

56.6

45.0

52.6

60.0

67.0

28.3 I

28.3 |

29.5

29.5

30.0

Table 2. Temperature distributions along the fuel rod

for different reactor power levels. Q=91O mr/h=consi,

2.6. Tvsmperature Distributions along the Fuel Rod for

Different Cooling Water Flow Bates.

During this measurement reactor power was equal to

4300 kff and the inlet water temperature was kept constant

on the level 31°C ± 1°C. Control rods were on following

positions : 1ES - 33.5 cm e SBB - 33»5 cm, 3RB - 30 cm,

7

. 30 ca. Results are contained In Table 3 and graphically

shown on Fig.8.

Table 3

Q

m5/h

560

700

800

910

a

°c

89.

81.

76.

73.

Ъ

4

4

0

*2

°0

87.

81.

76.

73.

5

4

4

0

s°G

87.6

80.0

76.4

73.0

4°C

95.4

86.1

81.4

77.0

°C

70

65

62

60

.0

• 0

.0

.0

4°c

30.

30.

30.

31.

8

8

8

0

Table 3. Temperature distributions alone the fuel rod

for different cooling water flow rates.

P s 4300 ИГ ш const.

2.7. Teapsrature Distributions along the Fuel Rod for

Different Reactor Power Levels at a Natural Convection

Cooling,

In this aeasureffient the inlet water tenperature ( at

the bottom of the core ) could not be Measured. Its value

considerably rieed during aeasurement. Eo, these results

have only the qualitative meaning. Thermocouple tg measured

the instantaneous values of the water temperature on the

level of 10 ca above the core edge. In Table 4 the mean

8

value» of t^ signal are given.

Control rod position* «ere following!

1BH - 29И си» 2Ш - 35-0 ca, Аи - 30 cm.

Temperature effect» on reactivity were compensated by 3ER

scntrol rod changing its position from 45 cia to 35,7 cm.

Results are contained in Table 4 and shown on Fig.9*

Table 4

p

50

100

150

200

°c34.0

42.6

51 Л

57.6

*2

°C

42.0

55.0

67.6

77,6

H°C

47.0

62,6

76.4

87.6

4

°c

50.0

66.4

81.4

91.4

ъ°C

50.0

66.0

81.4

91.4

*6

°C

24.0

25.0

30.0

35.0

On F i g . 9 :curve mar-ked by

!I ;

I I

I I I

IV

Table 4. Temperatute distributions along the fuel rod

at a natural convection cooling for different

reactor power levels.

2.8. Maximum Measured Temperature Diffei-ence between

Fuel Bod Cladding and Cooling Water At д^^ц. - tg

asa Function of Jater Plow Rate in Primary Circuit.

The water flow rate in primary cooling circuit for

constant reactor power level equal to 4.3 M* wae gradually

decreased. Control rods were on following p o s i t i o n s

1R-. - 33. b cm, 2RR - 55,5 cm, 3RR - 50 cm, AK - 50 ca.

The results show Table 5 auci Fig. 10.

The dashed curve i i Fig.10 was obtained fi'om

u in ś.) for reactor power equal to d

Table 5.

1

t, = tЦ- max

ч - ч

°G

°G

P C

120 U

С .

50.

39.

0

0

0

Q1O

77

31

46

.0

. о •

.0

800

81.-г

30.8

50.6

700

Só.1

30.5

55.3

bOO

95

3^

61

.0

.8

.2

;,60

Table b» Maxisiini aeasured temperature difference between

fuel rod cladding and cooling water as a function

of 'A-ater flow rate - Q. P = ^.3 Ш/ * const.

2.^. Vaxiflium Measured Temcerature Difference between r'uel

Rod Cladding ana Cooling .Vater. ^ t m a x = ^ ~ ^A a a

a Function of Reactor Power»

Inlet water temperature t^ was maintenend constant

с tiic level 2^°C + ?.C-J by the flow rate regulation in

secondary cooling circuit. During measurements following

coj.trcl го:з -лere Xept on constant level : 1RH - 30 cm;

-• • - .-''-•' '-Л ; AH - 30 cm. Potver effect on reactivity was

• vc:.; '.i:d rjy 5HFb control rod which cnanged its position

to 3a,5 c:a for a-,6 №9.

10

.и for J,1

Bteulte of measurements are collected In Table & and

graphically illustrated tn Fig.11»

Table 6

Q

B3/h

300

500

300

300

300

615

615

615

615

910

910

910

P

0.107

0.520

0,645

1.29

1.72

0,535

1.075

2.15

3.20

1,14

3.55

4,65

°C

24.4

52.0

43.0

62.6

76.4

30.8

39.5

57.0

73.3

35.162,0

74.5

* 6

°C

20.6

20.6

20.6

20.6

20.6

22.6

22.6

23.3

23.9

20.8

22.0

23.3

3.8

11.4 I

22.4

42.0

55.8

8.2

16.9

33.749.4

14.3

40.0

51.2

Ееазагкв ji

j1•urve I on

Fig.11 !

i

Curve II

on Fig.11, „

Curve IIT

on Pig .1 '

Table 6. Цят1 "И"» measured teoperature difference between

fuel rod cladding and cooling water ae a functi n

of reactor power.

11

. Maximum Measured Temperature Difference between

Fuel Hod and Cooling later ^tj^^ * ^

a Function of Inlet Water Temperature -

- t 6 as

This measurement allow to estimate the infleunce

of the water physical propertiea - changes /aainly

due to viscosity decrease with temperature/ on the heat

transfer coefficient. The average water temperature in

primary cooling circuit was gradually riaed at constant

reactor power level P = 4.3 US and water flow rate Q = 910

m^/h /. The temperature rise was achieved by water flow

rate regulation in the secondary cooling circuit. Power

level was maintenend constant with respect to the N-16

activity meter - indications. Control rods were kept on

the following positions :

1RR - 33.5 cm; 2HK - 33.5 ощ 3RR - 30 cm; AR - 30 cm.

Measurement results contains Table 7 and are shown on

Fig.12.

Table 7.

ч-*6

«na,

max °С

°С

6 ° С

71

2348

.5

еО

.5

77

31

45

. 0

.5

.5

86

45

41

.5

.0

;5

92

52

39

. 4

.9

.5

101

6 3 .

38.

0

0

Table 7. Uaxt mum measured temperature difference between

fuel cladding and cooling water ^ t ^ ^ * *4~^б

as a function of inlet water temperature - tg.

P = 4.3 » t Q. - 910

2.11. Maximum Measured Temperature Difference between

Fuel Rod Core Temperature and Fuel Cladding as

a Function of Beactor Power.

Temperature drop between fuel core thermocouple

measuring junction - t^ and fuel cladding junction - t

was measured in the EWA-4- core place,shifted a little

further along the core radius in comparison to the core

place I /Pig.2 position II/.

The influence of neutron flux change on the temperature

values can be neglected /within the range of temperature

error 4 2°C/e Results of measurements combined together

with the results of cladding temperature measurements in

ECTA-4 and BWA-2 cores are collected in Table 8 and plotted

in Fig.12.

The dependence betweer fuel core - fuel cladding temperature

drop /t^ - t^/ and reactor power is linear in the range

of measured temperatures and amounts about 38°С/Ш /assumed

neutron flux averaging factor is equal to 1.25Л All

presented temperature difference - power plots are normalized

with respect to the constant inlet water tempera-tore value

equal to 30°C. Correction was made using the dependences

plotted in Fig. 12 for БИА-4 core and in Fig.4 in Eef.1

for ША-2 core.

3. TRANSIENT TESTS

The main purpose of the performed transient test

series was to collect the data of the E»A-4 core behaviour

for most credibile incidents in reactor operation. Such

incidents are: stopping of the oooling water flow through

the core and reactor power rise as a result of the ramp

reactivity addition into the core* /Comparison of the

ЕЖА-2 and SWA-4 cores behaviour in transient states are

contained in Ref.10/.

3.1. Instrumentation for Measurements in Transient States

In transient states the requirements concerning the

transfer frequency band was a base for estimation of tho

properties of apparatus u^ed.

Below a aeasurlig Instrumentation applied for transient

signals of temperature, water flow rate euad reactor power

is described.

3.1*1. Temperature Measurements

The applied Thermocoax thermocouples of 2 ABA с 0*5

TJ-type according to SOOERN data have a response time

equal to 35 msec. /Response for step temperature change

in water/. The necessary condition for the measuring instru-

mentation frequency band was 4 cpe for tranafering and

recording of the temperature signals. It was fulfiled by

шзе of a double beem oscilloscope with attached film-caaera.

For amplification of the thermocouple signal the operational

amplifier waa used /Ref.5/. However% for most of the

temperattire transients the transfer frequency of aboux-

0.5 eps gave satisfactory accuracy of signal recording •

So, in such cases the phot ocomp ensat or - recorder H37.3-

type /USSR/ was applied.

3.1.2. Water Flow Measurements

Water flow decay after water pumps stopping was

recorded using the pressure signal from the flow

measuring orificed plate in primary cooling circuit. The

sensor consisting of two variable reluctance magnetic

typ© pressure transducers was connected to a bx idye e n -

suring system /Ref.6/. This system sensitive to the

pressure drop on the flow diafragm gives the output signal

proportional to the square root of water flow rate. The

voltage signal from this measuring system was iec into

oscilloscope and recorded on photo—type by attached

film-camera. To cut off the nich frequency signal coarponent

/noise from pumping system and pipe vibration/ the low

pass filter limiting the upper frequency band of the

measuring eet-up to about 100 cps — was applied.

ТаЫа 3. Juel rod temperature measured data collection.

p

IM

0.10

0.50

1.00

1.08

1.50

2.00

2.25

2.50

з.оо3.38

4.00

4.50

5.4в

JA

910

910

910

910

935

910

900

935

910

910

910

910

910

?uel neat temperatur* ( t f )measurement i s EWA-4 core

*f

°C

35.9

53.4

21.4

114.0

156.4

195.0

ч

°С

32.1

37.6

42.1

53.3

68.0

80.3

Ч

°С

30.9

32.6

30.9

31.9

з?.з

39,0

"ft°с

3.8

15.8

29.3

59.7

88.4

114.7

*ч-\°С

1.2

5.0

11.2

21.4

31.7

41.3

Cladding tempera-ture oeaeureaentIn 3jjJSA-4 core

4

41.0

54.5

67.0

79.0

90.0

«6

°C

28,3

28,3

29.;

29.i

32.C

o c

12.7

-

26.2

37.5

49.5

CO.Q

Gladding tempera-ture measurement/fief .1/1» m - 2 core

°c

39.1

49.5

59.8

70.1

79.5

88.9

*6

°c

28.9

28. о

28.9

28.5

26.-

28.9

°C

10,.2

20,6

30.9

41.6

51.0

60.0

Resarts •

for ВЯ1-2Q s 9.35 ur^/h

n

3.1«3* Reactor Power Measurement

The fast transients of the reactor power were recorded

using neutron sensitive compensated ion chamber of BWEJ-8

type /Ref.?/.

The chamber was calibrated with respect to the reactor

power by the heat balance method- The current signal from

the chamber was recorded on phytotype using oscilloscope

and film—camera. The frequency response of the chamber

together with the current measuring and recording apparatus

was well above 100 cps. /Ref .8/.

The ion chamber current signals at slow power tran-

sients in reactivity excitation tests were recorded using

NORMA Model 115 milivoltmeter - recorder.

3.2. Cooling Water Flow Failure Teats

The scheme of electrical connections applied in the

case of water flow rate stopping i-п primary cooling circuit

is shown in Fig. 14. Opening of the push button contactor

W-1 causes a break of voltage зирр1у to the water ршцрв

with the aid of the line breaker AHJ. The same voltage

signal switches on the photo—type driving nechanism in the

film-camera. In the same time a voltage supply to safety

amplifiers is switdied off and reactor scram occures.

Inherent time delay of the auxiliary delay system was

less than 5 m sec.

This case of reactor scram /short delay/ correspond to

17

the total electrical power supply breek.

The longest possible tine delay in reactor aafety

circuits occure*» «ben the acr&m signal comes from the

water flow rate meter as the water flow rate decreases

below 20% of its nominal value . That is artificial created

case of reactor scraa «hen the safety amplifiers are sup-

plied from the voltage source independent of the source

supplying the water pumps. The initial conditions of tem-

perature distribution along the fuel rod before scram can

be found in Fig.7.

3.2.1. Flo» Failure Test at Short Delay Scram

In this test the scram signal from electronic

amplifiesrs in safety circuits appearss practically in

the moment of electrical power supply break. T&e reactor

power and water flow decrease after APU-line breaker

switching off is shown In Pig.15 within about 1 sec time

increment„

Tine equal to zero on this diagram corresponds to the

moment of connection of the APU passive contacts. Th®

reactor power and water flow start to decrease almost in

the баае tine of about 150 m sec after voltage break.

The dashed part of the curve 1A the plot of water flow

decay refere to the water flow rate values below 300 *r/h.

The pressure drop on the flow diafrag» in this range

18

is much lower than nominal value and consequently tfcu*

error is very high»

In Fig,, 16 the thermocouple transient signals within the

range of 1 mln have been shown. It is seen from this

Figure that the flow inversion took place about u-'J sec

after voltage break and then typical convective distributio

of temperature along the fuel rod-occurea. The fuel aeat

temperature - t- decreases Ггош 200°С to about 70°C within

about 2 вес - Fig.1?. Differences between shapes of the

t^ and t , - transients arose probaoly ae a results of

the differences iiz Venturi tube profiles and differences

in initial water temperature before всгвы.

3.2.2. Plow Failure Teal; at Long Delay Scraia

In Pig.18, have been shown the reactor power and

water flow transients when the pumpe voltage supply brsak

is not followed by reactor eafety system - operation.

In this case the reactor operates on full power level

during about 1 sec at a very low value of water flow rate.

As a result of that,temperature peaks in transient signals

of thermocouples-appear. It is seen in Fig,19. In The t f

transient /Fig.20/ a slightly marked peak /a few centi-

grade a on the level of 200°C/ was observed /not shown in

the Fig.20/.

19

The series of measureaonts of the t^ thermocouple tran-

sient signals after reactor scraa are shown In Fig.21,

when various values of the scram signal*time delay have

been applied.

3.3. Reactivity Addition Tests

Fuel cladding temperaturę and reactor power transients

at convection water flow and at a very low values of

forced flow rate were investigated during these type of

tests. The reactivity addition rate for all the testa was

about 0.32 t /sec . Reactivity wes inserted into the core

by use of the highest possible motor speed of 1RH - control

rod* The reactor power after reactivity insertion was

controlled only by the influence of the fuel and water

temperature rise on the multiplication factor. Magnitude

of these temperature effects illustrate Pigs.Ъ ajad 6.

3*5.1 • Reactivity Excitation Tests at Convection Cooling

At raap reactivity addition into the core up to total

values inserted of about 0.5 В , power and fuel surface

temperature transients have an asymptotic form. When total

reactivity exceedes this value the peak in power transients

appears. The asymptotic temperature and reactor power

values for quasi equilibrium state of heat transfer corre-

to the values measured in steady states /Tig.9/,

20

The typical transients of fuel surface temperature and

reactor power at total reactivity inserted of about 0.5 $

are shown in Fig.22.

5.2,2. Reactivity Brcitation Tests at Low Values of

Forced Water Plow Rate.

In the range of low values of forced water flow rates

through the core for total reactivities inserted above

certain value depending on the flow rate, In a part of

the core, a flow inversion occures from forced one /downwards/

to natural convection flow /upwards/.

The time moment when the chaage of water flow direction

occures is identified by the observed redistribution of

the temperature values along the fuel rod. In such a case

the fuel surface temperature in this region of the core

where flow inversion occured rises by considerable value.

This phenomena was investigated for total reactivities

inserted in the range of /0.4 - 0.6/ $ and corresponding

values of flow rates of about 30 nr/h. /The estimation

of water flow rate value based on the extrapolation of

the dependence between flow rate and pumps loading current

is very unaccurate/. After flow inversion the surface

temperature /t^ - transient record/ increased by about

12°C, The corresponding redistribution of temperature

along the fuel rod., was observed on the type - record of

21

the surface thermocouple signals measured by H * В photo-

coapensator — recorder.

In Fig.23 the flow inversion - case in EWA-4- core is

illustrated for total value of reef-ivity equal to 0.58 / .

4. RESULTS DISCUSSION

The comparison of the thermal ^rcpertiee of the

EWA-4 and SWA-2 cores on the base of steady state measure-

ments can be performed witn limited accuracy because the

spatial distributiona of the power generation in these

two cores have not been measured.

Two following conditions were roughly coxj;iraied by transient

test series results :

1. The ratio of power generated in the instrumented fuel

rod in E//A-2 and КЛА—^ cores is approximately equal

to the ratio of fuel rods number in t.-.e both согеь /for

the same reactor power level/; ч'Р: and 751»

2. Spatial distributions of the power in both cores are

approximately the same.

The ratio of temperature peak.s measured by thermocouple

t^ in 5WA-2 and EWA-4 cores at long delay flow stopping

scram is roughly equal to iha ratio of power per one fuel

element in both cores. This fact indicates also that

vertical and radial neutron i'lux averaging factors /estima-

ting core hot spots/ arv appro:;imately the sajae in both

cores.

The effect of the application ot the variable cross-

aection tubes around the iuel rode can De ee jjoaoed. on the

base oi the comparison of the average power and flow ratios

per one fu»l element in the SHA-2 and E*A-4 cores.

Using dependence for ЕЯА-4 /Fig.11/ and lor EIA-2 /Pig.3,

fief»1/ it can be found that the maximum temperature gra-

dient for ЕЯА-4 core is about JO°C lower than corresponding

gradient for ВЯА-2 core.

In power units it corresponds to the value of 2.5 Ш .

Also comparing directly the slopes of the dependences

in above mentioned figures for Q^ = 935 m^/h /EWA-2 core/

and for Q = 9 Ю nr/h /EWA-3 core/ it is to be seen that

the value for HffA-2 core is equal to about 20°С/Ш and for

EWA-4 core to about 11°C/MW.

It siiowc that Venturi tube effect allows approximately to

aciDle tne maximum reactor power for assumed constant fuel

cladding temperature and water flow rate.

On the base of measured maximum temperature gradient

between fuel centre and fuel cladding temperatures, At =

= tf-t^, the rough estimation of the fuel core conductivity

- X can be performed according to the formula

At я ^ y * applied in WWES reactor design. In this formula

r m is the radius of the fuel core equal to 35 nm and

^ is heat generation density in instrumented fuel rod,

which for neutron flux averaging factor 1.3 anri 751 fuel

rods amounts З И 5 - Ю kcai/nr h.

The obtained value of A. equal to 9 kcal/m Ь °0 is

significantly smaller than that given in reactor design

data /25 kcal/m h °0/.

It is worth to note that in transient test в results

at flow failure tests, the flow inversion occures in

almost the same time about 40 sec after break of the

supplying voltage. However, the typical convective distri-

bution of temperature on the fuel rod occures much

earlier in Ж&-4- core. This indicates smaller thermal

inert!on, of BWA-4 than ЗЖА-2 core.

It could be previously expected as a result of the effect

of higher fractional alluminium volume in the SWA-4

core.

In the reactivity excitation tests at low values of

forced flow rate in BfA-2 core the local effect of the

fuel cladding temperaturę increase, after flow inversion,

is followed by the corresponding global reactor power

decrease. This effect suggests that flow inversion had taken

place in the significant part of the core practically at

the same time. In HfA-4 core this effect has not been

observed, «hat Is probably- confined with a more local

effect of the flow inversion in the case of EWA-4 core.

On the base of the results of the pexfasa»-.! transient

tests following concludlon can be formulated:

1. In the case of water flow stopping in primary cooling

circuit the ютт1 mum possible tia*» delays In reactor

safety system operation do not cause any significant

fuel temperature increase in ВЙА-2 and ША-Ч- cores.

2. The reactivity excitation testa up to total reactivi-

ties inserted of about 0.5 Ж Indicated strong effect

of reactor power self-limitation to the level of

about 200 kSV. /at Initial water temperature of about

20°C/.

3. The reactivity excitation teats at low values of forced

flow rates indicated that the requirement concerning the

proper value of the forced water flow rate at reactor

operation on low power levels should be estimated

on the base of consideration of the conditions of the

flow inversion effect in the core» /Flow rate value

estimated from these conditions xs significantly

higher than that necessary for the removal of the

heat generated in the core/. In such cases usually

a more safe is reactor operation at convection cooling

of the core.

HEFERENCES

1. J.Alekeandrowlcs, UUCseroiewski, L.Labno

Cladding Temperature Measurements of the БК-10 Type

Reactor fuel Rods in HiA-2 Core

Report IBJ No 879/И/В 1968

2» I Geneva Conference paper 622

3. W.Butkowski, W.Szteke, M.WleoKorkowakl

SK-1O Dispersion Fuel Elements for Experimental

and University Reactors

Eeport IBJ Ко 585/ŁT7/B /Dec, 196V

4. W.Byszewaki, U.Slechta, J «Aleksandrowiez

Investigation on the Colling of the 1WES Reactor Fuel

Element 1л Cylindrical Channels in a Channel with

Variable Cross-Section in the Water Loop

Nuklsonika 8, 507 Л963/

5. W.Bysz-ewski » Private communication.

6- J.Podgórski,

Operationverataerker, Sntwicklung, Meseungen und

Bauveieae Inat i tut t for Atoaenergi, Ejeller, Norway

Innerreport E- 25 /Dec. 1965/

7. J.Dziedzic

Instrukcja obsługi miernika MPZ-25-8

Instytut Badań Jądrowych Z-d В-И /1966/

8. J.Jabłoński, A.Janikowski, J.Topa

Progress in fi&actor Detectors Design and Construction

Carried out in the Tears 1963 - 1965,

Nukleonika £, 34-9 /1966/

9* L.Gafliorowaki, L.Labno

Measurements of the Frequency Response of Neutron

Sensitive Current Ion Chambree,

Nukleonika ^ , 175 /1966/

26

10, ii.Cżernieweicl, L.Łabno

Fuel Temperature Transients In SWA-2 and EWA-4 Coras

of Flow Failure and Rang? Reactivity Tests

Report IBJ No 1090/łI/HA/PH /1968/

27

г

Juti

Ф15

\

4(4 \Core cenie

* *

! f

-eŁbIv o:

v. J

EWA-4fuel rods -

neutron sourceAft. /Я*. гЯЯ. 3R* - control mdsAZ - ьаМу rod

Pig, 2 БЯА-4- core configuration

1

- I

USuribc*

coupkt

s1

Pig.J lastromented fuel rods

30

-\

I 4

center

ofH* t'4anś if

Neutron flux distribution in the core position I /in water/

31

8

-00/,

-QP&

-QD6-

-Ш0-

•QJ2-

-QJ4-

-QJ6-

EWA-A

Ю 20 30 ^0

Average water temperature # °C50

Fig.Ъ Homogenous temperature effect on reactivity

ЕЕ

5f arcscscj «3 *

Pig.? Temperature distributions along the fuel rod for different reactorpower levels. Q * 910 m5/n ш const. I -P«1O?5 Ш; II -P«2.25 Mf;III -P =5.3S » ; IV -P=4.5 «i V 6 Ш

Fig, 8 Temperature distributions along the fuel rod for different cooling water

flow rates. P=4.3 Ш =const. I -Q=91O nP/h; II -Q=800 m- /h; III -Q=7OO17 -QrSGO m3/h •

9»

•то-V

iX бО<

Ul HH *Mt

"5Г

Fig.9 Temperature distributions aloruj the fuel rod for different reactor power levels at

natural convection coolin,r» I -r=pOO Ш-, II -P=1OO kW; III -P=15O icW; 17 -Pa200 k».

I

I

I

/

в т

I

£

о•J

шS3

-i

оо

ои

3 •*

a •Ф II

Xl M

0) >О Фa *->a? xй •••СП

*-" 3VH О

S a1 5 CO

<D Vi

g, Оa' i:

о3

•.5

Fig.11 Maxijaujm temperature difference between fuel rc»i n

I -Q=300 rn^/h ; II - * = 6 1 ; m /hj III -Q=91O аУ

.ir.;; ani water из и function of reactor power.

Ч>

iл

$*]

45 S» ff* «О

Fig. 12 Maxima tempeiature difference between duel ród cladding and cooling water as a functionof inlet water temperature. P=4.3 MW=conat. q=9io m5/h=cons.

v

aо•w•POO)

Т

йф

оо.

оfjоЯ)

оя

нг»

®

11д.

аО)

•rł

г

J> Ж-i

SafetyЫ UA

Op m

L _ I

Fig. 14 Schematic diagram of the electrical connections

of water flow failure tests.

Q

0,1 02 02 0.4 0,5 0,6 Q7 QB Q9 1p 1.1

Time after power supply break ; sec

Fig. 15 Reactor рокег and wtf--r flc* decay after electrical power supply breaJc. Sbort delay асгая.

Core portion-IH

JO 40 SO 80Tim* ofttr рея**- supply bnot,s*e ^

70

fuel cla^ciinr *с>г.регы,аг ti-iboi-ncs at short, delay ьспг,

Core position-i

о ю го зо 40Time offer power supply breok, tee

-to йг

FiK. 1? fuel core and lu,l claddib.* t^peraturf trar^ientt at short

delay scrajn.

Q2 0,6 ф 1,0 1,2 1/Ь 1,6 1,8 2р

Time after power supply break , sec

Fig. 18 Reactor power and water flow decay after electrical power supply ureak.

Long delay зсгаш.

tWA-4 Core posit ion -1(P), « 4MW

(tt)0 гол

Pig- 19 Fuel cladding temperature transients at lone ле!а> bci1.*.-

EtHA-4 Core position

if*)* -Ąmt

( <99'C

G tO Ю 30Time afttr po**tr suppkj

SO 70sac

Fuel core and fu«l cladding tempersturt u 'delay _ .raffi.

- . -- t '! -'^ -

Tjme after power supply break , sec

18 Eeactor power a.ad water flow decay after ulectrical power supply break.

Long delay acram.

tNA-4 Core position-14M#20-C

(P)(tt)o

Pig. 19 Fuel cladaing temperature tranrients at Ion; ;elav

Ь-

ШЛ-Ą Core position • i

(te)e =37*C

(a)l-

k, sec.sow го so

Time after power iuppiy

Fir. ^0 Fuel core ала luel cla.ldinr temperature t r a Łdelay L.rani,

TO

u . ac lo:.-

- 4 солеCoat «OJ / т/о* ±

•vmr twnr , •»*

?ig. 21 Fuel cladding tejc>erature transients at varioue valueeof tlae dela^ of the reactor всгеш signal.

-г--

ЮО

90-

SO-

t40Ą

я-

о

£łVA -4 Core position - п4

Z 3Tum eft» arbitrary

•250

-200

iЮО

v so

Fig. 22 Temperature and power trans i ел/с s after ramp reactivity addition.

Total value of reactivity inserted - 0.51 $ at the rate 0.52 /sec

-4. Core position I0.56 #

*/

J

400

300

[200

J2 3 4

Time after arbtraty г em.

б

Fig. 23 Temperature an- :wer transients at i'lo-л invex-sion. Total reac-civity

inserted - C.-..с g, а с thv rate O..-J2 $ /sec. Flow rate about j>0 m /fa.