Some main results of commissioning of the Dalat Nuclear Research Reactor with low enriched fuel
Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 36-45
Some Main Results of Commissioning
of The Dalat Research Reactor with Low Enriched Fuel
Nguyen Nhi Dien, Luong Ba Vien, Pham Van Lam, Le Vinh Vinh, Huynh Ton Nghiem,
Nguyen Kien Cuong, Nguyen Minh Tuan, Nguyen Manh Hung, Pham Quang Huy,
Tran Quoc Duong, Vo Doan Hai Dang, Trang Cao Su, Tran Tri Vien
Nuclear Research Institute –Vietnam Atomic Energy Institute
01-Nguyen Tu Luc, Dalat, Vietnam
(Received 5 March 2014, accepted 13 April 2014)
Abstract: After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for
conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the
commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried
out from 24 November 2011 to 13 January 2012. The experimental results obtained during the
implementation of commissioning programme showed a good agreement with design calculations and
affirmed that the DNRR with LEU core have met all safety and exploiting requirements.
Keywords: HEU, LEU, physics start up, energy start up, effective worth, Xenon poisoning, Iodine pit.
I. INTRODUCTION
hours testing operation without loading at
nominal power (December, 13rd, 2011).
Physics and energy start-up of the Dalat
Nuclear Research Reactor (DNRR) for full
core conversion to low enriched uranium
(LEU) fuel were performed from November
24th, 2011 until January 13th, 2012 according to
an approved program by Vietnam Atomic
Energy Institute (VINATOM). The program
provides specific instructions for manipulating
fuel assemblies (FAs) loading in the reactor
core and denotes about procedures for carrying
out measurements and experiments during
physics and energy start-up stages to guarantee
that loaded LEU FAs in the reactor core are in
accordance with calculated loading diagram
and implementation necessary measurements
to ensure for safety operation of DNRR.
II. PHYSICS START UP
Physics startup of reactor is the first
phase of carrying out experiments to confirm
the accuracy of design calculated results,
important physical parameters of the reactor
core to meet safety requirements. Physics
startup includes fuel loading gradually until to
approach criticality, loading for working core
and implementing experiments to measure
parameters of the core at low power such as
control rods worth, shutdown margin,
temperature effect,…
A. Fuel loading to approach criticality
The loading of LEU FAs to the reactor
core was started on November, 24th, 2011
following a predetermined order in which each
step loaded one or group LEU FAs to the
reactor core. After each step, the ratio of
Main content of the report is a brief
presentation of performed works and achieved
results in the physics and energy start up stages
for DNRR using LEU fuel assemblies, that is
from starting loading LEU fuel to the reactor
core (November, 24th, 2011) until finishing 72
N0
(N0 is initial number of neutron count rate,
Ni
©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
LUONG BA VIEN et al.
Ni is that to be obtained after step ith) was
for working core, effective worth of loaded
fuel assemblies and shutdown margin were
preliminarily evaluated to ensure shutdown
margin limit not be violated. Fig. 3 shows the
current working core of DNRR, including 92
LEU FAs (80 fresh LEU FAs and 12 partial
burnt LEU FAs, the burn up about 1.5 to 3.5
%) and neutron trap at the center. Total mass of
U-235 that was loaded to the reactor core is
about 4246.26 g. Shutdown margin (or
subcriticality when 2 safety rods are fully
withdrawn) is 2.5 $ (about 2% k/k), smaller
than calculated value (3.65 $) but still
completely satisfy the requirement >1% for
the DNRR. Excess reactivity of the core
configuration is about 9.5 $, higher than
calculated value (8.29 $), ensuring operation
time of the reactor more than 10 years with
recent exploiting condition. So, it can be said
that the current working core meets not only
safety requirements but reactor utilization
also (ensure about shutdown margin and
sufficient excess reactivity for reactor
operation and utilization).
evaluated to estimate critical mass. At 15h35
on November, 30th, 2011 the reactor reached
critical status with core configuration including
72 LEU FAs and neutron trap in center (see
Fig. 1 and 2).
Established critical core configuration
with 72 LEU FAs having neutron trap is in
good agreement with design calculated
results. With 72 LEU FAs, by changing
position of some fuel assemblies, all new
criticality conditions were achieved with
lesser inserting position of regulating rod. It is
concluded that the above critical configuration
(Fig. 1) is the minimum one among
established configurations. The critical mass
of Uranium is 15964.12 g in which Uranium-
235 is 3156.04 g.
B. Fuel loading for the working core
After completion of fuel loading to
approach criticality, fuel loading for working
core was carried out from December, 6th, 2011
to December, 14th, 2011. During fuel loading
Fig. 1. Critical core configuration and
order of loaded fuel assemblies
Fig. 2. N0/Ni ratio versus number of FAs
loading to the core
37
SOME MAIN RESULTS OF COMMISSIONING OF …
LEU Fuel
Wet channel
Dry channel
Empty cell
Berrylium
Aluminum
Neutron trap
Fig. 3. Working core configuratiom with 92 LEU FAs
C. Performed experiments in the working core
configuration
working core in configuration with 82 fresh
LEU FAs and 92 LEU FAs.
Determination of control rod worth
Control rods worths and integral
characteristics in core configuration with 92
LEU FAs are presented in Table I, Fig. 4
and 5. Measured results were smaller than
design calculated results about 12% in
average.
To calibrate control rod worth, doubling
time method was applied for regulating rod
while reactivity compensation method was
used for shim rods and safety rods. The
calibration of control rods of DNRR were
implemented two time during fuel loading for
Table I. Effective worth of regulating rod, 4 shim rods and 2 safety rods
in core configuration with 92 LEU FAs.
Effective reactivity ($)
Control Rod
Measured value Calculated value
Regulating rod
Shim rod 1
Shim rod 2
Shim rod 3
Shim rod 4
Safety rod 1
Safety rod 2
0.495
2.966
3.219
2.817
2.531
2.487
2.195
0.545
3.237
3.263
3.473
3.086
2.744
2.795
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LUONG BA VIEN et al.
Position (mm)
Position (mm)
Fig. 5. Integral characteristics of 4 shim rods
Fig. 4. Integral characteristics of regulating rod
Thermal neutron flux distribution
measurement in 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.
Measurement of thermal neutron flux
distribution following axial and radial in the
reactor core was carried out 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.
1.1
1.1
1
1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
0.9
FA cell 2-3
0.8
FA cell 3-3
0.7
FA cell 4-5
0.6
0.5
0.4
0.3
0.2
0.1
0
0
5
10
15
20
25
30
35
40
45
50
55
60
-35
-30
-25
-20
-15
-10
-5
0
5
10
15
20
25
30
35
Position from the bottom to the top (cm)
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
39
SOME MAIN RESULTS OF COMMISSIONING OF …
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
Determination of effective worth of FAs,
beryllium rods and void effect
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.
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
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. 10. Effective worth ofFAs in the reactor core
Fig. 11. Effective worth of Be rods in the reactor core
40
LUONG BA VIEN et al.
Temperature (0C)
Fig. 13. Negative reactivity insertion
dependent on pool water temperature
Fig. 12. Measured results of void effect at
some positions in the reactor core
The most effective worth of fuel
established after each increased step of pool
water temperature about 2.50C. Basing on the
change of regulating rod position (due to
change of temperature in the reactor core) the
temperature coefficient of moderator was
determined.
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.
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
Determination of temperature coefficient
of moderator
o
initial temperature of 17.7 C. Based on these
results, the temperature coefficient of
moderator is determined about -9.110-3 $/oC.
Measured result without steel pipe containing
air at neutron trap was about -5.210-3 $/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 -8.010-3 $/oC.
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 100C by operating primary
cooling pump without secondary cooling
pump. To measure temperature coefficient of
moderator, criticality of the reactor was
41
SOME MAIN RESULTS OF COMMISSIONING OF …
III. ENERGY START-UP
A. Power ascension test
13-2 and rotary specimen was measured by
using Au foil activation method. Also, on
January 17th, 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.
On January 6th, 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,
Table II. Measured results of thermal neutron flux at several irradiation positions at different reactor power levels
Power (% Nominal power)
Irradiation
positions
0,5
10
20
100
Neutron trap
Channel 1-4
1.143E+11
2.063E+12
4.174E+12
2.122E+13
5.288E+10
4.749E+10
N/A
9.719E+11
8.542E+11
N/A
1.965E+12
1.682E+12
N/A
8.967E+12
N/A
Channel 13-2
Rotary Specimen
4.225E+12
Based on the reactor power determined
by thermal neutron flux measurements at low
power levels, on January 9th, 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.
system parameters was about 372 kW. This
value enables us to raise the reactor power to
full power level. 15h32 on January, 9th, 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%.
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
Table III. Thermal power of the reactor with operation time after the reactor reached 100% nominal power
Tin (1)
[oC]
Tout (1)
[oC]
GI
PI
[kW]
Time
[m3/h]
15h30
16h00
17h00
18h00
1h00
29,2
30,3
31,0
31,0
30,9
30,8
30,8
30,7
30,6
30,5
22,4
22,9
23,1
23,0
22,9
22,9
22,9
22,8
22,7
22,6
49,4
49,3
49,8
49,8
49,8
49,6
49,8
50,5
50,1
49,6
390
423
456
462
462
454
456
457
459
455
20h00
21h00
22h00
23h00
24h00
42
LUONG BA VIEN et al.
B. Xenon poisoning transient and Iodine hole
C. Power adjustment
In the process of gradually raising power
The experiment to determine the curve
built up of Xenon poisoning and then
calculating its equilibrium poisoning was
conducted from January 9th, 2012 to January
12th, 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 13th,
2012 after reducing power of the reactor from
100% to 0.5% nominal power by monitoring
the shift position of regulating rod.
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.
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.
D. Measurement of neutron flux and neutron
spectrum after power adjustment
After carrying out reactor power
adjustment, thermal neutron flux at some
Xe
Time (hour)
Fig. 14. Negative reactivity insertion by Xenon poisoning with operation time and Iodine pit
43
SOME MAIN RESULTS OF COMMISSIONING OF …
irradiation positions in the reactor core and
start up were carried out successfully. DNRR
was reached criticality at 15:35 on November,
30th, 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,
9th, 2012 to January, 13th, 2012.
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
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.
and 8.611012 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.
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.
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
Fig. 15. Measured neutron spectrum in neutron trap before and after conversion
44
LUONG BA VIEN et al.
ACKNOWLEDGMENTS
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.
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.
[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
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