Reactivity induced transient analysis when the occurrence of leakage in the dry irradiation channels of the Dalat Nuclear Research Reactor

 irradiation channels 
of the Dalat Nuclear Research Reactor 
Dau Duc Tu, Nguyen Minh Tuan, Le Vinh Vinh, Huynh Ton Nghiem, 
Nguyen Kien Cuong, Tran Quoc Duong, Bui Phuong Nam 
Reactor Center, Dalat Nuclear Research Institute, VINATOM 
01 Nguyen Tu Luc Str., Dalat, Lamdong, Vietnam. 
E-mail: tudd.re@dnri.vn 
(Received 06 December 2018, accepted 28 December 2018) 
Abstract: The leakage from the reactor pool back into the dry irradiation channels due to corrosion or 
mechanics based reason is a postulated event that could occur under operating conditions of the Dalat 
nuclear research reactor (DNRR), especially the channel 7-1 which has been installed more than 30 
years. When it occurs, the air space in these channels will be occupied by the water, subsequently a 
water column will appear in fuel region. The appearance of water column considerably enhances 
medium of neutron moderation for its surrounding fuel assemblies. As a result, a positive reactivity is 
inserted in the core and this event is classified as an insertion of excess reactivity. This event needs to 
be addressed by analysis and assessment from safety point of view and the results of analysis are also 
important for updating the reactor operating procedures. This paper presents assumptions, computer 
models and the results of analysis for such event in the DNRR by using MCNP5 code (code for 
neutronics analysis) and EUREKA-2/RR code (code for transient analysis). The calculation results 
include value of reactivity insertion, change in power of reactor, as well as surface temperature of the 
hottest fuel assembly. This research contributes to updating the reactor operating procedure. 
Keywords: Dalat nuclear research reactor (DNRR), MCNP5, EUREKA-2/RR, reactivity 
insertion, transient analysis. 
I. INTRODUCTION 
A. General information 
The Dalat reactor, which is a swimming 
pool type reactor and uses light water as 
moderator and coolant, has 500 kW thermal 
power and put into operation in March, 1984. It 
was reconstructed and upgraded from the USA 
made 250 kW TRIGA reactor. A number of 
structures from the original TRIGA reactor such 
as the aluminum tank with the surrounding 
concrete shield, four beam-ports, thermal 
column and the graphite reflector were 
remained. 
The reactor core is currently loaded 92 
low enriched fuel (LEU) assemblies (19.75%), 
3 irradiation channels (2 dry channels and 1 
wet channel), 12 beryllium rods and a neutron 
trap [1]. The fuel assemblies are VVR-M2 
tubular fuel assemblies designed 
and manufactured in Russia. Each fuel 
assembly consists of three coaxial annular 
tubes as well as a header and a tail. The 
outermost fuel element has a hexagonal shape 
of 32 mm in width across parallel sides and the 
other two inner ones have a circular shape of 
22 mm and 11 mm in outer diameter, 
respectively. The beryllium rods of the reactor 
have a hexagonal shape of 32 mm in width 
across its parallel sides [1]. 
B. Description of dry irradiation channels 
In the reactor core, there are two dry 
irradiation channels located in cells 7-1 and 13-
REACTIVITY INDUCED TRANSIENT ANALYSIS WHEN THE OCCURRENCE OF  
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2 and designed for neutron activation analysis 
(NAA). Their characteristics are described in 
Table I. They are vertical channels and made 
of watertight aluminum tubes. Samples for 
NAA are packaged in small plastic bags and 
then placed in a high purity polyethylene 
irradiation capsule (so-called rabbit). Finally, 
the capsules are sent to and returned from the 
channels by a pneumatic transfer system 
(PTS). At these two channels, samples are 
irradiated for a short time from 5 seconds to 5 
minutes to analyze elements by measuring 
emission ray gamma spectrum of the short-
lived radionuclides. The PTS of channel 7-1 is 
a remote-control system located nearby the 
reactor building. The PTS of channel 13-2 is 
mainly used for fast analysis of elements. Its 
terminal station and measuring equipments are 
located in the reactor building [2]. These 
channels enter the reactor from the top surface 
of the reactor pool, pass through the pool, 
penetrate into cells 7-1 and 13-2 of the reactor 
core and terminate 25 cm from the bottom of 
the reactor core as shown in Figures 1 and 2. 
Table I. Characteristics of the dry vertical irradiation channels 
Characteristics Channel 7-1 Channel 13-2 
Thermal neutron flux, n/cm
2
/s (3.8 ÷ 4.4) 1012 (3.8 ÷ 4.2) 1012 
Epithermal neutron flux, n/cm
2
/s (4.0 ÷ 4.5) 1011 (3.7 ÷ 4.1) 1011 
Fast neutron flux, n/cm
2
/s (3.0 ÷ 5.0) 1012 (3.0 ÷ 6.0) 1012 
Loading capacity, gram 20 4 
Irradiation time (min.), sec 45 5 
Irradiation time (max.), min. 20 30 
Sample moving speed, m/s 10 20 
Fig. 1. Technical drawing of pneumatic transfer channel 13-2 [2] 
DAU DUC TU et al. 
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Fig. 2. The layout of pneumatic transfer systems (PTS) [2] 
C. The leakage and assumptions for the 
transient analysis 
Under normal operating conditions, the 
leakage from the reactor pool back into the 
pneumatic irradiation channels due to 
corrosion or mechanics based reasons is a 
postulated event that could occur at the DNRR. 
When it occurs, the air space in these channels 
will be occupied by the water, subsequently a 
water column will appear in fuel region. The 
appearance of water column considerably 
enhances medium of neutron moderation for its 
surrounding fuel assemblies. As a result, a 
positive reactivity is inserted in the core that 
leads to an unwanted increase in fission rate, 
reactor power and fuel temperature. The value 
and speed of reactivity insertion mainly depend 
on the leakage rate, volume and location of the 
channels, power level and coolant temperature, 
etc. This event is classified as an event of 
insertion of excess reactivity. The events of 
insertion of excess reactivity that has the most 
potential consequences as a shim rod (SR) or 
the automatic regulating rod (AR) of the 
reactor can be withdrawn with the maximum 
speed of 3.4 mm/s and 20 mm/s, respectively, 
has been analyzed and described in the safety 
analysis report (SAR) for the DNRR [2]. 
However, the event presented here has not 
been analyzed in detail so far and the result of 
this analysis is important to develop and update 
the DNRR operating procedures after 35 years 
of operation. 
The main assumption selected for 
transient analysis is a small break occurring at 
one of two dry irradiation channels (channel 7-
1 or 13-2) with diameter of 2 mm (Ø = 2 mm) 
when the reactor is operating at the nominal 
power (500 kW). This assumption was made in 
order to simulate the above mentioned leakage. 
The other physics parameters needed for the 
transient analysis are given in Table II. 
MCNP5 Monte Carlo code and ENDF/B-VII.0 
nuclear data library were used to analyze 
reactivity insertion and EUREKA-2/RR code 
was used to study the transient of core power 
and fuel surface temperature. In fact, the 
possibility of simultaneous leakage at the 2 dry 
channels is extremely low; therefore it was not 
included in this investigation. 
II. CALCULATION MODELS 
In the investigation, the reactor core and 
its components such as fuel assemblies, control 
rods, beryllium blocks, graphite reflector, beam 
ports, thermal column, etc. were modeled 
almost as exactly as they are. Figures 3 and 4 
show in detail MCNP5 model for the core 
configuration with 92 fuel assemblies of low 
enriched fuel (LEU) and the irradiation 
channels of the DNRR [3]. 
REACTIVITY INDUCED TRANSIENT ANALYSIS WHEN THE OCCURRENCE OF 
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Table II. Summary of the reactor and safety parameters needed for the transient analysis 
Parameters Values 
Power, kW 500 
Coolant inlet temperature, 
o
C 32 
Reactor kinetics parameters 
- Prompt neutron life, s 8.925 10-5 
- Delayed neutron fraction (1$) 7.551 10-3 
Temperature reactivity coefficients 
- Moderator, (ΔK/K); (293 - 400oK) - 1.264 10-2 
- Fuel, (ΔK/K)/oC; 
 (293 - 400
o
K) - 1.86 10-3 
 (400 - 500
o
K) - 1.92 10-3 
 (500 - 600
o
K) - 1.56 10-3 
- Void, ΔK/K % of void 
 (0 - 5%) -0.2432 
 (5 - 10%) -0.2731 
 (10 - 20%) -0.3097 
Reactivity control 
- Shutdown worth, % (2 safety rods) 4.98 
- Maximum withdrawal speed of one shim rod, 
mm/s 
3.4 
 and of the regulating rod, mm/s 20 
Reactor protection characteristics 
- Response time to overpower scram, s 0.16 
- Response time to fast period scram, s 
 Start-up range 9.1 
 Working range 6.7 
- Drop time of control rods, s 0.67 
Fig. 3. Cross-sectional and vertical views of the reactor modeled by MCNP5 
Channel No.2 
Channel No.1 
Channel No.3 
Channel No.4 
DAU DUC TU et al. 
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Fig. 4. Core configuration with 92 LEU fuel assemblies and the irradiation channels modeled by MCNP5 
The EUREKA-2/RR model for the 
DNRR is shown in Figure 5. The reactor core 
contains 92 fuel assemblies and they were 
divided into 5 distinct channels. Each fuel 
channel can consist of one or more fuel 
assemblies, in which fuel channel No. 1 was 
chosen as the hottest channel containing the 
hottest fuel assembly with maximum FR (radial 
power factor). The number of fuel assembly in 
each channel and FR factors taken from 
MCNP5 calculations are given in Table III. 
The axial power distribution was also taken 
from MCNP5 calculations. 
In the present model, each fuel channel 
was divided more into 10 heat slabs along with 
10 nodes as shown in Figure 5. The model uses 
a total of 54 nodes, 50 heat slabs and 59 
junctions. The junction No. 59 is declared as a 
fill junction to simulate the primary coolant 
flow (50 m
3
/h) passing through the reactor core 
and by pass volume as shown in Figure 5. 
Table III. The number of fuel assemblies in each channel and radial power factors 
Fuel channel No. The number of fuel assemblies Radial power factor (FR) 
1 1 1.421 
2 17 1.256 
3 18 0.932 
4 28 0.887 
5 28 0.985 
With the scenario of leaking the 
irradiation channel in reactor core, the 
equation (1) was used to calculate the water 
column height (h) depending on the time (t) 
and the parameters L, S1, S2. 
2
2
1
2
2
 t
S
gS
LLh (1) 
In which: 
- L: Height from the center of the 
reactor core to the surface of the reactor pool 
surface 
- h: Height of water column in the 
irradiation channel (depending on the time 
since the accident started) 
- S1: The section of breaking 
- S2: The section of irradiation channel 
- g: The gravity acceleration 
Channel No.7-1 
Channel No.13-2 
Channel position 
Core center 
REACTIVITY INDUCED TRANSIENT ANALYSIS WHEN THE OCCURRENCE OF 
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Fig. 5. Schematic diagram of the model prepared for EUREKA-2/RR analysis [4] 
III. RESULTS AND DISCUSSION 
The MCNP5 calculations for the DNRR 
were performed with the critical calculation 
(KCODE) and the ENDF/B-VII.0 nuclear data 
library. The number of simulated events is 260 
cycles and each of 10
6
 particles. The calculation 
results of reactivity insertion depending on time 
are shown in Table IV and Figures 6 and 7. 
Table IV. Time-dependent reactivity insertion for the channels 7-1 and 13-2 when air occupies by water 
Time 
 (second) 
Reactivity of channel 7-1 
(cents) 
Reactivity of channel 13-2 
(cents) 
1 1.62 0.02 0.35 0.03 
2 5.07 0.05 1.96 0.05 
3 7.09 0.04 2.90 0.09 
4 8.52 0.06 3.57 0.09 
5 9.63 0.05 4.09 0.08 
DAU DUC TU et al. 
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6 10.54 0.07 4.51 0.06 
7 11.31 0.07 4.87 0.05 
8 11.97 0.06 5.18 0.06 
9 12.55 0.04 5.46 0.08 
The calculation results shown in 
Figures 6 and 7 revealed that about 9.0 
seconds after the leakage occurs, the 
channels 7-1 and 13-2 will be flooded with 
water to the corresponding height on the top 
surface of the reactor core and the 
maximum reactivity insertions are 12.55 and 
5.46 cents, respectively. The reactivity 
insertion at the channel 7-1 is greater than 
the channel 13-2. This can be explained by 
the fact that the channel 7-1 is located 
closer to the center of the reactor core and 
its volume is larger than the volume of the 
channel 13-2. 
Fig. 6. Reactivity insertion versus time when water occupies the air space in the channel 7-1 
Fig. 7. Reactivity insertion versus time when water occupies the air space in the channel 13-2 
For the next step, the calculation results 
of reactivity depending on the time by MCNP5 
were introduced into the transient calculations 
using EUREKA-2/RR. The calculation results 
of the transition of power and fuel surface 
temperature for the reactor at the hottest 
channel for the cases of channel 7-1 and 13-2 
are shown in Figures 8 and 9, respectively. 
REACTIVITY INDUCED TRANSIENT ANALYSIS WHEN THE OCCURRENCE OF 
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The transient calculations (see Figure. 
8) show that as soon as the event occurs, the 
time to reach the power scram set point are 3 
and 8 seconds (power scram point of the 
reactor was set at 550 kW). From Figure 9, it 
is seen that after 3.4 and 8.5 seconds, the fuel 
surface temperature increases to reach to the 
maximal values (90.8 
o
C and 92.2 
o
C) for the 
channels 7-1 and 13-2, respectively. These 
temperature values are much lower than the 
design limit imposed for the transient 
conditions (the design limit of fuel surface 
temperature is 400 
o
C). Therefore, this type of 
transient does not cause safety concern for the 
DNRR if the reactor protection system fulfills 
its safety functions. 
Fig. 8. The change in reactor power 
Fig. 9. The fuel surface temperature 
IV. CONCLUSIONS 
This report presented the investigation 
of reactivity insertion and transient analysis 
in power and fuel surface temperature for 
the DNRR when the leakage occurs in the 
dry irradiation channels 7-1 and 13-2. The 
results showed that approximately 9.0 
seconds after starting the event, the 
channels 7-1 and 13-2 will be flood with 
water to the height corresponding to the top 
of reactor core. The maximal values of 
DAU DUC TU et al. 
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reactivity insertion determined by MCNP5 
for the channels 7-1 and 13-2 were 
12.55 0.04 and 5.46 0.08 cents, 
respectively. The transient analysis by 
EUREKA-2/RR showed that the fuel surface 
temperature does not exceed 92.2 
o
C that is 
very low compared to design limit (400 
o
C). 
 The results of this investigation are 
considered to update the operating procedures 
of the DNRR after 35 years of operation. 
REFERENCE 
[1]. Nuclear Research Institute, Regulations of 
operation of Dalat nuclear reactor, March 2014. 
[2]. Nuclear Research Institute, Safety analysis 
report for the Dalat nuclear reactor, 2012. 
[3]. MCNP - A General Monte Carlo N-Particle 
Transport Code, Version 5, April, 2003. 
[4]. EUREKA-2/RR: A Computer Code for the 
Reactivity Accident Analyses in Research 
Reactors, Private Communication.