Some main results of commissioning of the Dalat research reactor with low enriched fuel

ut by Lu metal foils 
neutron activation. A number of positions in 
the reactor core were chosen to measure 
thermal neutron flux distribution including 
neutron trap, irradiation channels 1-4 and 13-2, 
and 10 FAs at the cells: 1-1, 2-2, 2-3, 2-7, 3-3, 
3-4, 4-5, 6-4, 12-2 and 12-7. Figs 6 to 9 present 
the measured results of axial and radial neutron 
flux distributions of the reactor core. 
From the measured results, it can be 
seen that the maximum peaking factor of 
1.49 is achieved at outer corner of hexagonal 
tube of the fuel assembly in cell 6-4. Neutron 
distribution of working core has large 
deviation from North (thermal column) to 
South (thermalizing column). Neutron flux 
in southern region of the core (cell 12-1 and 
12-7) is about 28 % smaller than those in 
Northern region (cell 2-1 and 2-7). The 
asymmetry of the reactor core has reason 
from the not identical reflector that was noted 
from the former HEU fuel core. 
Position (mm) 
R
ea
ct
iv
it
y
 (
$
) 
Position (mm) 
R
ea
ct
iv
it
y
 (
$
) 
Fig. 4. Integral characteristics of regulating rod Fig. 5. Integral characteristics of 4 shim rods 
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1
1.1
-35 -30 -25 -20 -15 -10 -5 0 5 10 15 20 25 30 35
R
e
la
ti
v
e
 U
n
it
Position from the bottom to the top (cm)
FA cell 2-3
FA cell 3-3
FA cell 4-5
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1
1.1
0 5 10 15 20 25 30 35 40 45 50 55 60
R
e
la
ti
v
e
 U
n
it
Position from the bottom to the top (cm)
Fig. 6. Axial thermal neutron flux distribution in 
the fuel assemblies 
Fig. 7. Axial thermal neutron flux distribution in the 
neutron trap 
SOME MAIN RESULTS OF COMMISSIONING OF  
40 
Determination of effective worth of FAs, 
beryllium rods and void effect 
The measurements of effective worth of 
FAs, beryllium rods and void effect (by 
inserting an empty aluminum tube with 
diameter of 30 mm) were also performed. 
These are important parameters related to 
safety of the reactor. Positions for 
measurement of effective reactivity of FAs, 
Be rods and void effect were chosen to 
examine the distribution, symmetry of the 
core and the interference effects at some 
special positions. Effective reactivity of FAs, 
beryllium rods and void effect were 
determined by comparing position change of 
control rods before and after withdrawing FA 
or beryllium rod or before and after inserting 
watertight aluminum tube. Reactivity worth 
values were obtained using integral 
characteristics curves of control rods. 
Figs 10÷12 show the measured results 
of effective worth of 14 FAs in the reactor 
core at different positions; effective worth of 
beryllium rods around neutron trap and a 
new beryllium rod at irradiation channel 1-4; 
void effect at neutron trap, irradiation 
channel 1-4 and cell 6-3, which surrounded 
by other FAs. 
Fig. 8. Thermal neutron flux distribution of FAs 
and irradiation positions in comparison with 
neutron trap. 
Fig. 9. Thermal neutron flux distribution of the 
FA’s corners in comparison 
with neutron trap 
Fig. 10. Effective worth of FAs in the reactor core 
Fig. 11. Effective worth of Be rods in the reactor core 
LUONG BA VIEN et al. 
41 
. 
The most effective worth of fuel 
assembly measured at cell 4-5 is 0.53 $. 
Measured results of effective reactivity of fuel 
assemblies and Be rods show a quite large 
tilting of reactor power from North to South 
direction. Void effect has negative value in the 
reactor core (cell 1-4 and 6-3) while positive in 
the neutron trap. Void effect in neutron trap 
has positive value because almost neutrons 
coming in neutron trap are thermalized, that is 
absorption effect of water in neutron trap is 
dominant compared to moderation effect. The 
replacement of water by air or decreasing of 
water density when increasing steadily of 
temperature introduces a positive reactivity. 
With the core using HEU fuel also has positive 
reactivity of void in neutron trap. 
Determination of temperature coefficient 
of moderator 
Temperature coefficient of moderator is 
the most important parameter, demonstrating 
inherent safety of reactor. To carry out 
experiment, the temperature inside reactor pool 
was raised about 10
0
C by operating primary 
cooling pump without secondary cooling 
pump. To measure temperature coefficient of 
moderator, criticality of the reactor was 
established after each increased step of pool 
water temperature about 2.5
0
C. Basing on the 
change of regulating rod position (due to 
change of temperature in the reactor core) the 
temperature coefficient of moderator was 
determined. 
Heating process of water in reactor pool 
by operating primary cooling pump took long 
time so water in neutron trap also heated up 
and inserted positive reactivity (as explanation 
in measurement of void effect), as opposed to 
temperature effect in the reactor core. So, a 
hollow stainless steel tube 60 mm diameter 
was inserted in neutron trap to eliminate 
positive temperature effect of neutron trap. 
Fig. 13 shows measured results of 
temperature coefficient of moderator with 
initial temperature of 17.7 
o
C. Based on these 
results, the temperature coefficient of 
moderator is determined about -9.1 10-3 $/oC. 
Measured result without steel pipe containing 
air at neutron trap was about -5.2 10-3 $/oC. 
Thus, temperature coefficient of moderator 
including neutron trap still has negative value. 
Temperature coefficient of moderator of the 
core loaded with 88 HEU FAs measured in 
1984 was -8.0 10-3 $/oC. 
Fig. 12. Measured results of void effect at 
some positions in the reactor core 
Fig. 13. Negative reactivity insertion 
dependent on pool water temperature 
temperature 
Temperature (
0
C) 
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$
) 
SOME MAIN RESULTS OF COMMISSIONING OF  
42 
III. ENERGY START-UP 
A. Power ascension test 
On January 6
th
, 2012 reactor power has 
been increased at levels of 0.5% nominal 
power, 10% nominal power and 20% nominal 
power. At each power level, thermal neutron 
flux in neutron trap, irradiation channels 1-4, 
13-2 and rotary specimen was measured by 
using Au foil activation method. Also, on 
January 17
th
, 2012 thermal neutron flux of 
positions mentioned above was measured at 
power level 100%. Measured results of thermal 
neutron flux at several irradiation positions in 
the reactor core with different power levels are 
presented in Table II. 
Table II. Measured results of thermal neutron flux at several irradiation positions at different reactor power levels 
Irradiation 
positions 
Power (% Nominal power) 
0,5 10 20 100 
Neutron trap 1.143E+11 2.063E+12 4.174E+12 2.122E+13 
Channel 1-4 5.288E+10 9.719E+11 1.965E+12 8.967E+12 
Channel 13-2 4.749E+10 8.542E+11 1.682E+12 N/A 
Rotary Specimen N/A N/A N/A 4.225E+12 
Based on the reactor power determined 
by thermal neutron flux measurements at low 
power levels, on January 9
th
, 2012 the reactor 
was ascended power: 0.5%, 20%, 50%, 80% 
and then operated at 80% nominal power 
during 5 hours for determination of thermal 
power and examination of technological 
parameters and gamma dose before raising the 
reactor power to nominal level. 
Thermal power of the reactor 
corresponding to 80% nominal power level 
(based on indication of control system) after 5 
hours calculated based on primary cooling 
system parameters was about 372 kW. This 
value enables us to raise the reactor power to 
full power level. 15h32 on January, 9
th
, 2012 the 
reactor was raised to 100% nominal power and 
maintained at this power about 65 hours before 
decreasing to 0.5% nominal power to measure 
Xenon poisoning transient. Table III presents 
the values of thermal power of the reactor 
during the first 8 hours after the reactor reached 
100% nominal power. The data indicate that 
thermal power is just only about 460 kW, lower 
than design nominal power about 10%. 
Table III. Thermal power of the reactor with operation time after the reactor reached 100% nominal power 
Time 
Tin (1) 
[
o
C] 
Tout (1) 
[
o
C] 
GI 
[m
3
/h] 
PI 
[kW] 
15h30 29,2 22,4 49,4 390 
16h00 30,3 22,9 49,3 423 
17h00 31,0 23,1 49,8 456 
18h00 31,0 23,0 49,8 462 
1h00 30,9 22,9 49,8 462 
20h00 30,8 22,9 49,6 454 
21h00 30,8 22,9 49,8 456 
22h00 30,7 22,8 50,5 457 
23h00 30,6 22,7 50,1 459 
24h00 30,5 22,6 49,6 455 
LUONG BA VIEN et al. 
43 
B. Xenon poisoning transient and Iodine hole 
The experiment to determine the curve 
built up of Xenon poisoning and then 
calculating its equilibrium poisoning was 
conducted from January 9
th
, 2012 to January 
12
th
, 2012 when the reactor was in 100% 
nominal power (indicating of control system 
without adjusting power) . Next, Iodine hole 
was also determined from 12 to January 13
th
, 
2012 after reducing power of the reactor from 
100% to 0.5% nominal power by monitoring 
the shift position of regulating rod. 
Fig. 14 presents measured results of 
Xenon poisoning curve and Iodine pit of the 
above experiment. Xenon equilibrium 
poisoning and other effects is totally about -1.1 
eff and the maximum depth of Iodine pit 
determined about -0.15 eff after 3.5 hours 
since the reactor was down to 0.5% nominal 
power. After adjusting thermal power up to 
500 kW, during the long operation from 
March, 12-16, 2012, after the reactor was 
operated 68 hours at nominal power, total 
value of poisoning and temperature effects is 
about -1.32 eff. 
C. Power adjustment 
In the process of gradually raising power 
in energy start-up, although power indication 
on control system was 100% but calculated 
thermal power of the reactor through flow rate 
of primary cooling system and difference 
between inlet and outlet temperatures of the 
heat exchanger was only 460 kW, smaller than 
nominal power about 10%. The reason was 
mainly due to power density of the core using 
92 LEU FAs were higher than the mixed core 
using 104 FAs before. The adjustment to 
increase thermal power of the reactor was 
performed by changing the coefficients on the 
control panel. After adjusting, the reactor was 
operated to determine thermal power at power 
setting 100%. The results of thermal power 
obtained from the next long operation was 
about 510.5 kW. This value includes 500 kW 
thermal power of the reactor and about 10 kW 
generated by primary cooling pump. 
D. Measurement of neutron flux and neutron 
spectrum after power adjustment 
After carrying out reactor power 
adjustment, thermal neutron flux at some 
Xe 
poisonning 
Iodine Pit 
PpppPithole 
Time (hour) 
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eg
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e 
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ti
v
it
y
 (
$
) 
Fig. 14. Negative reactivity insertion by Xenon poisoning with operation time and Iodine pit 
SOME MAIN RESULTS OF COMMISSIONING OF  
44 
irradiation positions in the reactor core and 
neutron spectrum in neutron trap were 
measured again by neutron activation foils. 
Measured maximum neutron flux at neutron 
trap was 2.23 1013 n/cm2.s (compared with 
calculated result was 2.14÷2.22 1013 n/cm2, 
depending on shim rods position). Those in 
channel 1-4 and 13-2 were 1.07 1013 n/cm2.s 
and 8.61 1012 n/cm2.s, respectively. The 
experimental error of neutron flux was 
estimated about 7%. 
From reaction rate measured by foils 
irradiation method in neutron trap, neutron 
spectrum obtained by SAND-BP computer 
code. Obtained results of neutron spectrum in 
neutron trap (Fig. 15) showed that comparing 
with mixed-core HEU-LEU fuel, when 
neutron trap having thinner Beryllium layer, 
thermal neutron flux increased while epi-
thermal and fast neutron flux decreased with a 
significant percentage. 
IV. CONCLUSIONS 
After completing design calculation and 
preparation, start up of DNRR with entire LEU 
FAs core was implemented following a 
detailed plan. As a result, physics and energy 
start up were carried out successfully. DNRR 
was reached criticality at 15:35 on November, 
30
th
, 2011 with 72 LEU FAs, consistent with 
calculated results. Then, the working core with 
92 LEU FAs has been operating 72 hours for 
testing at nominal power during from January, 
9
th
, 2012 to January, 13
th
, 2012. 
Experimental results of physical and 
thermal hydraulics parameters of the reactor 
during start up stages and long operation cycles 
at nominal power showed very good agreement 
with calculated results. On the other hand, 
experimental results of parameters related to 
safety such as peaking factor, axial and radial 
neutron flux distribution of reactor core, 
negative temperature coefficient, temperature 
of the reactor tank, temperature at inlet/outlet 
of primary cooling system and secondary 
cooling system,it could be confirmed that 
current core configuration with 92 LEU FAs 
meets the safety and exploiting requirements. 
Measured neutron flux at irradiation 
positions and actual utilization of the 
reactor after full core conversion also 
showed that the reactor core using LEU fuel 
is not much different than previous core 
using HEU fuel. 
Fig. 15. Measured neutron spectrum in neutron trap before and after conversion 
N
eu
tr
o
n
 f
lu
x
/L
et
h
ag
y
, 
n
/c
m
2
.s
ec
LUONG BA VIEN et al. 
45 
ACKNOWLEDGMENTS 
The NRI’s staffs that performed start up 
work of DNRR with entire LEU fuel core 
would like to express sincere gratitude to the 
leadership of Ministry of Science and 
Technology, Vietnam Atomic Energy Institute, 
Vietnam Agency for Radiation and Nuclear 
Safety, who have regularly regarded, guided 
and created the best condition for us to 
implement our works. We also express our 
thanks to Argonne National Laboratory and 
experts from RERTR program (Reduced 
Enrichment for Research and Test Reactors) 
and specialists, professionals in program 
RRRFR (Russian Research Reactor Fuel 
Return) has supported in finance as well as 
useful discussions during design calculation of 
full core conversion, upgrading equipments 
and carrying out start up of DNRR. 
REFERENCE 
[1] P. V. Lam, N. N. Dien, L. V. Vinh, H. T. 
Nghiem, L. B. Vien, N. K. Cuong, “Neutronics 
and Thermal Hydraulics Calculation for Full 
Core Conversion from HEU to LEU of the 
Dalat Nuclear Research Reactor”, RERTR Int’l 
Meeting, Lisbon, Portugal, 2010. 
[2] L. B. Vien, L. V. Vinh, H. T. Nghiem, N. K. 
Cuong, “Transient Analyses for Full Core 
Conversion from HEU to LEU of the Dalat 
Nuclear Research Reactor”, RERTR Int’l 
Meeting, Lisbon, Portugal, 2010. 
[3] “Process of physics and energy start up for full 
core conversion using LEU fuel of the Dalat 
Nuclear Research Reactor”, Nuclear Research 
Institute, 2011. 
[4] “Operation logbook of DNRR”, 2011-2012