m£a0uh£l£gi3ts 0? the sk-10 type bsictob fuel whs Ш ША …
TRANSCRIPT
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
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Available frost
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of the P o l i s h OaveraE»nt Cozaajisaioner for Use
of Huelsar Energy
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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
EWA-4fuel rods -
neutron sourceAft. /Я*. гЯЯ. 3R* - control mdsAZ - ьаМу rod
Pig, 2 БЯА-4- core configuration
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
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
£
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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.
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