Design analyses for full core conversion of the dalat nuclear research reactor

0.9
1.0
0 2 4 6 8 10 12 14 16 18 20
Time (s)
P
o
w
e
r 
(M
W
)
0
2
4
6
8
10
12
14
16
18
20
D
N
B
R
Period scram
Overpower scram with set point 110% FP
DNBR
40
50
60
70
80
90
100
110
120
0 2 4 6 8 10 12 14 16 18 20
Time (s)
T
e
m
p
e
ra
tu
re
 (
o
C
)
Period scram
Overpower scram with set point 110% FP
Fig. 7. Reactor power and cladding temperature transient of one shim rod withdrawal 
from a stable operation of 100%FP. 
Cooling pump failure 
In the event of in-service primary or 
secondary cooling pumps stopped working, the 
reactor is automatically shutdown by an 
abnormal technological signal on low flow rate 
(the setpoint is 40 m
3
/h for the primary flow, 
and 70 m
3
/h for the secondary flow). The 
residual heat after shutdown is about 6% FP 
(30 kW) in maximum and the natural 
convection process can itself assure the good 
cooling of the core. 
If the reactor is purposely maintained at 
full power operation, failure of cooling pumps 
leads to loss of heat removal from the pool 
water, and thus gradually increases of the pool 
water temperature. The results show that the 
clad temperature reaches the maximum 
allowable operating clad temperature of 103 
o
C 
at about 55 min; i.e. the reactor could continue 
its operation for 55 minutes within the envelope 
of the limiting conditions of operation. The 
results also show that even at the end of the 
simulation (7000 s) the clad temperature has 
been well below that of the acceptance criterion 
for anticipated operational occurrences. 
Earthquake 
The postulated event of an earthquake of 
intensity grade VI is assumed to occur while 
the reactor is at full power. Owing to the 
measures undertaken in design and 
construction, the removal of all control rods 
would not exceed 10 mm and insert a step 
positive reactivity estimated of 0.3$. With this 
reactivity insertion, the scram set-point of 
reactor overpower is attained almost 
instantaneously. If the reactor scram is 
initiated by overpower signal with a delay of 
0.16 sec, the fuel surface temperature 
increases slightly before decreases with the 
power, the residual heat after shutdown is 
sufficiently removed from the fuel by natural 
convection of pool water without considerable 
increase of the temperature. 
Fig. 8 shows the analyzing results of 
the earthquake event assuming the 
protection system fails to shutdown the 
reactor, and Because of the loss of offsite 
power due to the earthquake, the primary and 
secondary pumps stop operating. In this case, 
the reactor power increases to the max value of 
1.525 MW after 200 seconds from the 
initiation of this event. The reactor power then 
rapidly decreases because the significant 
increasing of core water temperature so that the 
positive reactivity insertion is overtaken by the 
negative reactivity feedback (about -0.44$). 
The reactor is then kept at subcritical state. The 
cladding temperature reaches a maximum 
value of 118.2
o
C, then decreases with no 
DESIGN ANALYSES FOR FULL CORE CONVERSION OF  
11 22 
significant overheating of the fuel. The 
maximum outlet water reaches 89
o
C and 
gradually decreases to a value at about 60
o
C, 
which is still far below the saturation 
temperature. The minimum DNBR of 4.79 is 
much higher than the acceptance value. 
In case the cooling pumps remain 
working after the earthquake event (very 
unlikely); the peak power reaches 1.57 MW 
within 300 seconds and decreases due to 
negative temperature feedback to a stable 
value of about 1.12 MW. The cladding 
temperature reaches to a maximum value of 
118.38
o
C then gradually decreases to a stable 
value of 115
o
C without nucleate boiling. The 
maximum temperature of outlet water is 89
o
C 
at the peak power then decreases and 
stabilizes at about 82
o
C, well below the 
saturation point. The minimum DNBR in this 
case estimated about 4.74 is still far from the 
acceptance criterion. 
Fig. 9. Power and Temperature responses to earthquake event while cooling pumps are stopped functioning. 
Fuel cladding failure (MHA) 
For the derivation of source term of this 
event, it is assumed that no core melting occurs 
but cladding rupture of one fuel assembly is 
involved. It is also assumed that the damaged 
fuel assembly is irradiated at the maximum 
neutron flux position in the core and the fuel 
damage occurs immediately at the end of 
operating cycle of 100 hrs with no decay. 
From the damaged fuel assembly, 100% 
of noble gases (Xe, Kr), 25% halogens (I), and 
1% of other radionuclides (Cs, Te) [21] are 
released directly to the reactor building with 
the assumption of no retention of volatile 
fission products in the pool water. During the 
accident evolution, the emergency ventilation 
system is not in place, the normal ventilation 
system V1 is in operation but HEPA filter with 
95% efficiency is not available, and there are 
no decay and deposition of radionuclides 
within the reactor building. 
The evaluation of dose to a member of 
the public is calculated by code MACCS2 
version 1.13.1, using the following 
assumptions: (1) The radionuclides are 
released to the environment through the 40 m 
stack; (2) The Gaussian plume model is used to 
calculate air concentration of radioactivity; (3) 
Tadmor and Gur parameterization is used for 
this analysis; (4) No building in the vicinity (an 
open area release), plume rise mechanics only 
due to momentum rise (non-buoyant plume) 
and no wet deposition are assumed; (5) The dry 
deposition velocity is assumed to be 0.01 m/s, 
which corresponds to a particle with an 
aerodynamic equivalent diameter of 2 m to 4 
m (for unfiltered particulate releases) [15]; (6) 
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000
Time (s)
P
o
w
e
r 
(M
W
)
0
2
4
6
8
10
12
14
16
D
N
B
R
DNBR
Power
30
40
50
60
70
80
90
100
110
120
0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000
Time (s)
T
e
m
p
e
ra
tu
te
 (
o
C
)
Max. Cladding Temperature
Max. Water Temperature at Outlet
Water Temperature at Inlet
LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG 
11 23 
Surface roughness length is specified as 50 cm; 
(7) Mixing layer height is assumed to be 500 m 
(see Table 36 in Appendix VII of Ref. 21); (8) 
The breathing rate is 3.3x10
-4
 m
3
/s; (9) No 
shielding and sheltering are assumed; (10) 
Doses at each downwind distance are 
calculated for one year after the arrival of the 
plume (11). The environmental release is 
assumed to begin at the start of the weather 
conditions: Pasquill class D2.0 (most frequent 
stability class and most frequent wind speed). 
The effective equivalent doses, including 
cloudshine dose, inhalation dose and 
groundshine dose, as a function of the distance 
from the source are shown in Table XII and 
Fig. 10. It is seen that radiation exposure to the 
general public with the maximum effective 
dose of 0.64 mSv/year at distance from 400 m 
to 500 m from the stack. This value is lower 
than the annual dose limit of 1.0 mSv specified 
for the public [22]. 
Table XII. The annual effective dose to the public vs distance for the MHA. 
Distance 
(m) 
Effective Dose 
(mSv) 
Distance 
(m) 
Effective Dose 
(mSv) 
50 4.80E-02 1100 3.18E-01 
150 1.43E-01 1300 2.59E-01 
250 4.95E-01 1500 2.16E-01 
350 6.42E-01 1700 1.83E-01 
450 6.44E-01 1900 1.57E-01 
550 5.94E-01 2250 1.23E-01 
650 5.33E-01 2750 9.14E-02 
750 4.74E-01 3250 7.08E-02 
850 4.21E-01 3750 5.66E-02 
950 3.75E-01 4250 4.64E-02 
Fig. 10. The annual effective dose to the public in MHA event within 5 km. 
0.00E+00
1.00E-01
2.00E-01
3.00E-01
4.00E-01
5.00E-01
6.00E-01
7.00E-01
0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000
Downwind Distance, m
E
ff
ec
ti
v
e 
E
q
u
iv
a
le
n
t 
D
o
se
, 
m
S
v
DESIGN ANALYSES FOR FULL CORE CONVERSION OF  
24 
IV. CONCLUSIONS 
Neutronics, steady-state thermal-
hydraulic and transient/accidents analyses for 
Dalat Nuclear Research Reactor show that with 
a slight change in arrangement of Be rods, the 
main features of 92 LEU WWR-M2 FA cores 
are equivalent to those of HEU and current 
mixed fuel cores. 
The negative values of reactivity feedback 
coefficients show the inherent safety feature and 
shutdown margin of both candidate cores meets 
the safety required value of -1% k/k. The 
working core with 92 fresh LEU fuel assemblies 
can be operated for 600FPDs or about 11 years 
based on the current operating schedule without 
shuffling. The neutron fluxes at the irradiation 
positions are not much different from those of the 
current mixed fuel core. 
In thermal hydraulics aspect, the 
requirement of thermal-hydraulic safety margin 
for two candidate cores in normal operational 
condition is satisfied. The calculated maximum 
cladding temperature in operational condition 
is below the permissible value of 103
o
C. 
In transient/accidents aspect, some 
postulated initiating events and accident related 
to the conversion of the DNRR to full LEU 
core were selected and analyzed. Based on the 
calculated results, conclusions might be 
withdrawn as following: 
- The excess reactivity insertions when 
inadvertent withdrawals of control rod from 
start-up or nominal power operation are 
prevented by safety settings to initiate the 
reactor scram at overpower and fast period. 
None of these initiators would lead to the ONB 
and DNB, ensuring the integrity of the fuel 
cladding. The residual heat after shutdown is 
sufficiently removed from the fuel by natural 
convection of pool water. 
- If one of the cooling pumps stopped 
working, the reactor is automatically shutdown 
by a scram signal on low flow rate. The decay 
heat is removed from the fuel by natural 
convection of pool water. In this event, if the 
reactor was purposely maintained at full power, 
it could be safely operated for 55 minutes when 
maximum cladding temperature is still lower 
than the permissible value of 103
o
C. 
- The postulated earthquake event of 
MSK intensity grade VI would cause a step 
reactivity insertion of 0.3$. Even if the reactor 
fails to be scrammed, this positive reactivity 
can be covered by negative temperature 
feedback if the cooling pumps are stopped 
simultaneously, keeping the reactor sub-critical. 
In case the cooling pumps continue operating 
after earthquake event, the negative temperature 
feedbacks act to bring the reactor power to a 
stable level of about 1.12 MW without nucleate 
boiling. The minimum DNBR is much higher 
than the acceptance criterion of 1.5. 
- The maximum hypothetical accident 
assumes 100% of noble gases (Xe, Kr), 25% 
halogens (I), and 1% of other radio-nuclides (Cs, 
Te) in a most power fuel assembly after a long 
run are released into the environment through 
40m high stack. This event is considered to be 
very unlikely to occur for the DNRR. Even so, it 
would not cause undue radiological risk to the 
environment or the public. 
ACKNOWLEADGMENTS 
The authors would like to express their 
gratitude to experts from the Reduced 
Enrichment for Research and Test Reactors 
(RERTR) program of Argonne National 
Laboratory for financial support as well as very 
useful discussions during design calculation of 
full core conversion for the Dalat Nuclear 
Research Reactor. 
LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG 
11 25 
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