Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN’s Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X)

CARDOSO, Domingos D´Oliveira

FERREIRA, Francisco José de Oliveira

VOI, Dante Luiz

XAVIER, Larissa R. P.

Carpe dIEN

2017 International Nuclear Atlantic Conference - INAC 2017 
Belo Horizonte, MG, Brazil, October 22-27, 2017 
ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR – ABEN 
  
 
Proposal for a radiation shielding study aiming the implantation of 
neutrons beam shutter in the J-9 radiation channel of the Argonauta 
reactor of the Nuclear Engineering Institute. 
 
Larissa R. P. Xavier1, Domingos D’Oliveira Cardoso1, Francisco José de Oliveira 
Ferreira2 and Dante Luiz Voi2 
 
1 Instituto Militar de Engenharia (IME) 
Praça General Tibúrcio, 80 
22291-270 Rio de Janeiro, RJ 
larissa.xavier@cnen.gov.br 
domingosoliveiralvr71@gmail.com 
 
2Instituto de Engenharia Nuclear (IEN) 
Rua Hélio de Almeida, 75 
21941-614 Ilha do Fundão, RJ 
fferreira@ien.gov.br 
dante@ien.gov.br 
 
 
ABSTRACT 
 
Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear 
Engineering (IEN), regularly serves a network of users focused on research and development, and also provides 
its infrastructure for experimental classes and completion work course. Due to increasing demand for non-
destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or 
devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the 
irradiation channels without the reactor being turned off and with a shielding configuration that limits the 
occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the 
present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN’s 
Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor 
using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the 
spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or 
change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is 
involving the neutron flux simulation of the reactor and the various shields with different boundary conditions 
using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X). 
 
 
1. INTRODUCTION 
 
Radiological Protection is the set of measures that aims to protect man, his descendants and the 
environment from possible harmful effects of ionizing radiation, thus allowing the activities 
that make use of this type of radiation [2,4]. 
 
The basic principle of radiological protection known as ALARA (As Low As Reasonably 
Achievable) states that all exposures should be kept as low as reasonably achievable. This is 
because low-dose radiobiological studies have shown that there is no real dose threshold for 
stochastic effects. Thus, the exposure of any tissue involves a carcinogenic risk, depending on 
the radiosensitivity of that tissue. Therefore, the application of this principle requires the 
optimization of radiation protection in all situations where they can be controlled by protective 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
measures, particularly in the selection, planning of equipment, operations and protection 
systems [9, 13, 15]. 
 
One of the activities of radiological protection, which corroborates this principle, is the 
specification and design of materials that can be used as protection barriers, in order to 
minimize the effects of doses of ionizing radiation on workers and individuals in the public. 
 
The installation of a beam shutter in the J-9 irradiation channel will be an important factor for 
radiological safety, minimizing the chronic exposure of occupationally exposed individuals and 
other users of the reactor. Also from an operational point of view, it will allow access to the 
experimental arrangements without the reactor being shut down, as is the case today. Thus, 
there will be the improvement and expansion of the research activities with the reactor, as it 
will make possible the implementation of new experimental techniques, as well as optimize the 
attendance of the increasing demand related to the rendering of services using the reactor. This 
study will contribute not only to the attendance of the current neutron flux, but also to the 
possibility of increasing this flow, in order to be made available in the future demand of the 
activities of the Argonauta. 
 
 
2. METHODOLOGY 
 
The reactor in question is a low-power thermal reactor, designed to achieve up to 5 kW of 
power, although its usual operation is at 340 W. 
 
 
 
 
Figure 1:  Argonauta reactor located at the Nuclear Engineering Institute. 
 
 
The core of the Argonauta consists of two concentric aluminum cylinders forming a ring where 
the fuel elements formed by plates are placed and separated by graphite wedges. Between the 
plates circulates water for refrigeration and moderation of the neutrons. The current fuel 
configuration in the core is eight symmetrically distributed fuel elements, each consisting of 
seventeen uranium oxide (U3O8) fuel plates fixed by two aluminum pins. 
 
The outer thermal column is located on the front of the reactor, and contains thirteen removable 
graphite blocks for conducting experiments. The main channel, which is used in this work, 
called J-9, has the largest neutron flux since it reaches the outer aluminum tank at half height 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
of the fuel elements. The depth of the other channels reaches only the external reflector. At the 
exit of the J-9 channel is installed the crystal spectrometer, where is the BF3 detector used in 
the experiments [10, 11,14]. 
 
A collimator-moderator arrangement is used within the J-9 channel providing the neutron 
characteristics, which are presented below [3]. 
 
 
 
 
Figure 2:  Collimator-moderator arrangement on the J-9. 
 
 
Table 1:  Characteristics of the neutron flux 
 
Reactor power:170 W 
Thermal flux 2.23 106 n.cm-2s-1 
Epithermal flux 3.18 103 n.cm-2s-1 
Neutron/gamma ratio 3.00 106 n.cm-2mR-1 
L/D collimation rate 63.25 
 
 
To perform the neutron detection at the output of the J-9 were used an experimental array 
composed of: pre-amplifier, multichannel analyzer, which includes the 2000 V power supply, 
amplifier, and multichannel buffer, and associated electronic [1,6,7,8,10,19]. 
 
In the first stage of the work the alignment of the neutron beam of the J-9 channel with the BF3 
detector was regulated, in order to obtain the best position, that is, where the largest neutron 
flux was found. 
 
After this stage, different types of materials were selected by SEREA – Reactor Service – from 
Institute of Nuclear Engineering, in order to attenuate the neutron beam, and to provide a better 
radioprotection for the operators and users, by choosing the shield that meets simultaneously 
the above quoted items positioned at the exit of J-9 channel. The materials used and their 
characteristics may be observer below [5]. 
 
 
 
 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
Table 2:  Characteristics of the materials that served as shielding to the beam of J-9 
radiation channel  
 
Material Thickness (mm) Density (kg/m3) 
Stainless steel 8.5 8000 
Water 23.5 1000 
Aluminum 13.3 2697 
Bakelite (sample port) 2.0 1375 
Cadmium 0.5 8650 
Lead 10.7 11340 
Concrete (high density) 22.3 3600 
Gadolinium 0.1 7901 
Graphite 15.0 2160 
Iron 8.5 7874 
Borated Paraffin 12.5 * 
Paraffin 18.0 900 
                                    * In determination 
 
 
3. RESULTS 
 
Following the methodology, it was initiated by the alignment of the emerging beam of the J-9 
channel with the spectrometer, where the BF3 detector is located. The best position for the 
spectrometer would be the point where the largest direct beam was measured, as it would be 
possible to work with the largest neutron beam, and consequently, a better counting statistic. 
To obtain the proper positioning, the spectrometer was mounted on a railing system that allows 
the rotation of this equipment up to 360 degrees, with minimum variations of 1 degree. 
 
 
 
 
Figure 3:  Rotatory adjustment with the function of aligning the spectrometer to the 
irradiation beam of J-9 channel. 
 
 
For the first experiment, measurements were made between the positions of 354 to 0 and 0 to 
6 degrees, with an interval of 1 degree. For each grade three measurements were made (each 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
of 30 seconds), and the average was found. It was observed that the highest neutron flux was 
obtained at the 359 degree position, as shown in the Figure 4. In this condition, the relation 
between the peak width assigned to neutron detection and the peak width obtained with gamma 
detection was also checked. The width found for the neutron was 682 channels, and for gamma 
the 24 channels, thus the ratio of 28.4 was obtained. The work was carried out with the power 
of the reactor in 34 Watts. 
 
 
 
Figure 4:  Number of neutron counts per degree, in the direct flow. 
 
 
At this stage, keeping the spectrometer aligned at position 359, the second phase of the work 
was started. The power of the reactor was increased to 170 Watts and the relation between the 
neutron peak width and the gamma width was the 686 channels for 24 channels, obtaining a 
value of 28.6, thus maintaining the above-mentioned relation even with the increase in power, 
confirming that the detector saturation did not occur. In the same way, three measures (each 
300 seconds) were made for the direct (unshielded) beam, and then measurements were taken 
for different types of materials, which could be used as shielding. In the following figure, it is 
possible to observe the structure mounted with some of the barriers. 
 
 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
 
 
Figure 5: Structure assembled for the experiment, with some of the shields (borated 
paraffin, lead, concrete and aluminum) 
 
 
Starting with the direct beam the average of the counts found for the neutron fluxes was 
348014, with a standard deviation of 6%. One of the spectra obtained in the multichannel buffer 
is followed. 
 
 
 
 
Figure 6:  Spectrum of the direct irradiation beam of the J-9 channel. 
 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
The same procedure used for the direct beam was made for each of the materials separately. 
Below are the average counts and the standard deviation for the various shields. 
 
 
Table 3:  Average counts and standard deviation for the different materials 
 
Material Average counts Standard deviation 
Paraffin wax 18909 1% 
Lead 276259 6% 
Gadolinium 93186 3% 
Concrete 131429 5% 
Paraffin 6686 2% 
Iron 199702 6% 
Graphite 198425 1% 
Aluminum 282927 3% 
Cadmium 24992 1% 
Bakelite 211973 1% 
Stainless steel 169914 0% 
Sample port with water 2496 5% 
Sample port without water 151313 0% 
Water 148818 0% 
 
 
The neutron count for water was found by decreasing the value found for the sample port 
without water and that found for the sample port with water. 
 
For a better presentation of the obtained results, the spectrum obtained with the direct beam 
and with the studied block material will be arranged together, thus allowing the choice of the 
best experimental arrangement of neutron shielding for the output of the J-9 channel. 
 
 
 
 
Figure 7:  Direct beam x Attenuated beam by Borated Paraffin 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
 
 
Figure 8:  Direct beam x Attenuated beam by Lead 
 
 
 
 
Figure 9:  Direct beam x Attenuated beam by Gadolinium 
 
 
 
 
Figure 10:  Direct beam x Attenuated beam by Concrete 
 
 
 
 
Figure 11:  Direct beam x Attenuated beam by Paraffin 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
 
 
Figure 12:  Direct beam x Attenuated beam by Iron 
 
 
 
 
Figure 13:  Direct beam x Attenuated beam by Graphite 
 
 
 
 
Figure 14:  Direct beam x Attenuated beam by Aluminum 
 
 
 
 
Figure 15:  Direct beam x Attenuated beam by Cadmium 
 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
 
 
Figure 16:  Direct beam x Attenuated beam by Bakelite 
 
 
 
 
Figure 17:  Direct beam x Attenuated beam by Stainless steel 
 
 
 
 
Figure 18:  Direct beam x Attenuated beam by Sample pot with water 
 
 
 
 
Figure 19:  Direct beam x Attenuated beam by Sample pot without water 
 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
The counts and spectra obtained due to the different attenuations were observed, the best 
materials to be used for the shielding of the neutron flux under study were found. Below is 
presented the attenuation capacity achieved by each material and consequently its possible 
utility for the construction and implementation of an obturator for the neutron beam from the 
J-9 irradiation channel of the Argonauta reactor. 
 
 
Table 4:  Percentage of attenuation of the different materials 
 
Material Percentage of attenuation 
Paraffin 98,1% 
Bored paraffin 94,6% 
Cadmium 92,8% 
Gadolinium 73,2% 
Concrete 62,2% 
Water 57,2% 
Stainless steel 51,2% 
Graphite 43,0% 
Iron 42,6% 
Bakelite 39,1% 
Lead 20,6% 
Aluminum 18,7% 
 
 
The attenuation for the sample port with water and no water were only used to find the value 
of the attenuation for the water, so these materials were not included in the table above. 
 
 
4. CONCLUSIONS 
 
With the experimental data obtained in this work the best materials to be used as shielding for 
the neutron flux of the output of the J-9 channel were found. They are: paraffin, borated paraffin 
and cadmium. This work continues with the development of the reactor neutron flux simulation 
stage and the various shields with different conditions using the transport code, Monte Carlo 
N-Particle Extended (MCNP-X) [12,16,17,18]. After obtaining the simulation results, they will 
be compared with the experimental results presented here, in order to point out the best material 
composition, thicknesses and/or arrangements to be used in the implementation of a neutron 
beam shutter of the J-9 irradiation channel of the IEN’s Argonauta reactor. 
 
 
ACKNOWLEDGMENTS 
 
This work would not be possible without the knowledge and teachings of my dear advisers 
Domingos, Francisco and Dante, and the technical support of the entire Argonauta reactor team, 
especially Rogerio Chaffin, João Carlos Santos Pereira, André Luis Nunes Barbosa, João 
Alberto Gadelha and Flavio Guimarães Porto. Thanks also to Lilia Crissiuma Palhares who 
supported me and made it possible. To the Institute of Military Engineering (IME), to the 
Institute of Nuclear Engineering (IEN) and the National Commission of Nuclear Energy 
(CNEN). 
INAC 2017, Belo Horizonte, MG, Brazil. 
 
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Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute.

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Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute.

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