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ORIGINAL ARTICLE
Year : 2014  |  Volume : 37  |  Issue : 2  |  Page : 71-76  

Neutron spectral and dose distribution studies during fast reactor fuel fabrication


Health Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai, Maharashtra, India

Date of Web Publication18-Dec-2014

Correspondence Address:
Kousiki Ghosh
Health Physics Division, 1/AF, Bidhannagar, Kolkata - 700 064
India
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Source of Support: None, Conflict of Interest: None


DOI: 10.4103/0972-0464.147277

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  Abstract 

In mixed oxide (MOX) fuel fabrication facilities, the fabrication of MOX fuel involves various metallurgical operations such as mixing and milling of weighed quantities of uranium and plutonium oxides, precompaction, granulation of precompacts, final compaction of granules, sintering of green pellets, followed by fuel pin fabrication. All these operations are carried out in glove boxes, which have complex geometries due to housing of various equipments. Plutonium, a source of neutrons, is handled in large quantities in various forms such as powder, granules, pellets and pellets encapsulated in pins during these operations. It is obvious that extensive knowledge on the neutron spectral distributions in workplace is required from radiation protection and shielding point of view. In this paper, a brief introduction to the source of neutrons in MOX fuel handling facilities, studies that include experimental measurements of neutron spectra of various forms of MOX, contribution of neutron fluence in various energy groups and its dose equivalent, establishment of a simulation procedure for glove boxes handling Pu using FLUKA Monte Carlo codes, comparison of simulation results with the actual experimental measurements are presented. The results indicate that bare PuO 2 and MOX sources require 3 mm of lead shield to eliminate gamma interference in MICROSPEC with N-probe neutron spectrometer. Furthermore, the studies reveal that dose equivalent contribution from MOX pellets is very significant in the energy group of 1.0-1.5 MeV unlike PuO 2 powder and fuel pins, which exhibit significant contribution in the energy groups of both 1.0-1.5 MeV and 2.0-3.0 MeV. Also, the fuel pins show high neutron fluence in the energy group of 0.0-0.01 MeV, but they do not contribute significantly to dose equivalent. A good agreement between experimentally measured data and FLUKA simulated results has been observed.

Keywords: FLUKA Monte Carlo code, glove box, mixed oxide fuel, neutron spectra, PuO 2


How to cite this article:
Ghosh K, Ganesh G, Sunil C, Biju K, Purohit R G, Tripathi R M. Neutron spectral and dose distribution studies during fast reactor fuel fabrication . Radiat Prot Environ 2014;37:71-6

How to cite this URL:
Ghosh K, Ganesh G, Sunil C, Biju K, Purohit R G, Tripathi R M. Neutron spectral and dose distribution studies during fast reactor fuel fabrication . Radiat Prot Environ [serial online] 2014 [cited 2022 May 23];37:71-6. Available from: https://www.rpe.org.in/text.asp?2014/37/2/71/147277


  Introduction Top


India's three-stage nuclear power program involves utilization of mixed oxide (MOX) [1],[2],[3] fuel (MOX fuel containing uranium and plutonium oxides) in fast breeder reactors. In MOX fuel fabrication facilities, the fabrication involves various metallurgical operations in which Plutonium is handled in various forms such as powder, granules, pellets and pellets encapsulated in pins. All these operations are carried out in leak-tight glove boxes, which have complex geometries due to housing of various equipments. Plutonium obtained from reprocessing of irradiated spent fuel containing 238 Pu, 239 Pu 240 Pu and 242 Pu isotopes is a source of neutrons due to spontaneous fission as well as (α, n) reaction. Radiation monitoring in such workplaces involves measurements of the neutron field associated with the gamma field. Hence, the characterization of neutron spectra and field measurements are very important in understanding its fluence distribution in various energy groups and its impact on dose equivalent. It is imperative that extensive knowledge on the neutron spectral distributions in workplace is essential.


  Materials and methods Top


Experimental

Neutron energy spectra were measured with MICROSPEC and N-probe [4] in various locations of Plutonium handling facility. It is widely used in neutron spectroscopy for its better accuracy than the other commonly used portable neutron spectrometer and easy handling methods. [5] Neutron dose rate is obtained by multiplying fluence with energy dependent neutron fluence to dose conversion factors.

A few locations of the fuel fabrication facility were selected for the present study. It includes cylindrical stainless steel container containing PuO 2 in powder form, MOX powder, MOX pellets kept inside the glove boxes [Figure 1] and the fuel pin storage tray [Figure 2] where fuel pins are stored temporarily and transferred from one location to another inside the facility. As the nature of neutron spectra depends on the sample matrix and room geometry, [6] this comparison study was performed. The glove boxes provided with a shield offers gamma rejection capability for the instrument to eliminate gamma interference in the measurements. Whereas, the measurements carried out for a bare source and a fuel pin tray containing a significant number of fuel pins require adequate lead shielding for the detector to eliminate gamma contribution. The various experimental set up is explained as follows:

  • PuO 2 in powder form stored inside a stainless steel container which is further sealed with double PVC sheet and kept inside a larger stainless steel container was used for carrying out measurements. The measurements were carried out by keeping the detector 5 cm away from the surface of the cylindrical stainless steel container. The measurements were also carried out by placing lead shield of various thicknesses while keeping the detector in the same position to study gamma contribution in the measurements
  • The neutron spectral measurements were carried out when a cylindrical container containing PuO 2 was stored inside the glove box, representing actual workplace conditions. The detector was kept at 1 m height from the floor. This is also the position of neutron personal dosimeter during work. As the glove box is already sufficiently shielded, gamma interference is not significant in the neutron dose measurements. Hence, additional shield was not required for this measurement. Detector was kept at 3 cm away from the shielded part of the glove box from PuO 2 container
  • Two cylindrical containers with MOX pellets kept inside the glove box were considered for measurements. Measurements were taken from outside the lead shielded portion of the glove box to eliminate gamma contribution
  • Fuel pins encapsulated with MOX fuel pellets were considered for the study. These fuel pins are stored in a trolley that is used for transporting fuel pins from one location to another within the facility. Neutron spectra measurements representing actual working conditions were carried out for the pins stored in the tray. The detector was placed 3 cm away from the tray containing pins. Measurements were taken without shield as well as by placing one 3 mm lead shield between the tray containing pins and the detector to eliminate gamma interference in the instruments.
Figure 1: Glove boxes

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Figure 2: Fuel pins

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Monte Carlo simulation

Detailed simulations were performed using Monte Carlo Transport code FLUKA by almost exact modeling of the actual experimental conditions. FLUKA [7] is a well-known Monte Carlo particle transport code. Scientific Linux computation platform is required to run the code.

While performing simulations, the modeled workplace was kept similar to the actual measurement geometries as much as possible for better matching. In FLUKA simulations two cases were considered for the cylindrical PuO 2 container, that is, the neutron source (i) bare source inside the room and (ii) source inside the shielded glove box. The simulated diagrams for these two situations have been shown in [Figure 3] and [Figure 4], respectively. For the bare source wall and floor of the room were also considered. Several right parallelepiped, right circular cylinder and spehrical (SPH) bodies were used to build the geometry representing the actual measurement conditions. The source is 1 kg PuO 2 powder in a cylindrical container. The neutron spectra of spontaneous fission and (α, n), are made to emit/sampled within the cylindrical volume of the PuO 2 powder isotropically. The major inputs are the energy distribution of neutrons and probabilities of emission, direction of emission of the neutrons, position of the emission in the source volume. A source subroutine representing the cylindrical source of neutron was written and compiled along with the source card. In this subroutine 80% contribution from spontaneous fission and 20% contribution from (α, n) reaction were considered. Initial guess spectrum for (α, n) reaction of PuO 2 was taken from literature [8] and spontaneous fission spectrum was calculated using Watt fission spectrum of 240 Pu. USRTRACK card was used to estimate neutron fluence distribution, USRTRACK along with AUXSCORE card (for folding the neutron fluence with "AMB74" conversion coefficients) were used to estimate ambient dose equivalent. 10 7 neutron histories had been given for each simulation.
Figure 3: Cross sectional full view along X axis at (0, 0, 0) considering wall and PVC floor of the room

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Figure 4: Cross section view of the FLUKA simulated geometry along X-axis at (26.5, 0, 0). This is the plane showing four gauntlets (green), viewing window (yellow) and stainless steel wall (brown). Origin is taken as the geometrical center of the glove box

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  Results and discussions Top


Measured neutron spectra by MICROSPEC with N-probe showing gamma interference of the instrument in mixed field environment

It is observed that for bare source gamma dose rate = 3.0 mGy/h (measured by GM survey meter) and neutron dose rate = 182 μSv/h (measured by He 3 based REM counter).For shielded source gamma dose rate = 0 . 025 mGy/h (measured by GM survey meter) and neutron dose rate = 178 μSv/h (measured by He 3 based REM counter).

[Figure 5]a and b represent neutron spectra measurements taken for bare source and source inside the glove box along with shield, respectively.

From [Figure 5]a and b, it is observed that 3 mm lead shield is required to eliminate gamma contribution. No extra shield is required when the source is inside glove box as its own shielding is sufficient to eliminate the gamma contribution. Though N-probe is having 1000:1 gamma rejection ratio but due to higher gamma contribution from 241 Am in MOX fuel and also due to pile-up effect, a complete elimination of gamma interference may not be possible.
Figure 5: (a) Measured neutron spectra with MICROSPEC and N-probe showing gamma interference of the instrument in bare PuO2 powder source. (b) Measured neutron spectra with MICROSPEC and N-probe for source inside the glove box with and without 3 mm extra Pb shield between glove box and detector

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The reduction in neutron fluence due to keeping lead shield in between the source outside the glove box and detector is negligible. This is confirmed theoretically using FLUKA. The results are given in [Table 1].
Table 1: Neutron fluence rate obtained from FLUKA by keeping various thicknesses of lead shield in between source and detector


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Experimental results and measured neutron spectra by MICROSPEC with N-probe at different locations of the workplace

As shown in [Figure 6] neutron spectra at different locations of the workplace having plutonium in a different form is presented graphically. It can be observed that maximum neutron fluence is obtained from the neutrons of energy 1.5 MeV and 3 MeV irrespective of the workplace position, though the contribution ratios differ for various forms depending on scattering and slowing down of neutrons by different forms of sample matrix.
Figure 6: Experimental measured neutron spectra at different locations of the workplace

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The neutron spectrum obtained from PuO 2 powder shows equal contribution of neutron fluence at 1.5 MeV and 3 Mev, whereas there is a very little contribution of neutron fluence in lower energy side of the spectrum. The spectrum obtained from MOX pellets as a source shows that the significant contribution of neutrons comes from 1.5 MeV in comparison to higher energy (3 MeV) region, which is possibly due to slowing down of the higher energy component by the compact fuel matrix itself. On the other hand for fuel pins the contribution of neutrons of energy <1 MeV is significantly higher than that of higher neutron energy groups. This may be attributed to slowing down of neutrons by other fuel pins placed side by side in the same tray and also a stainless steel clad of the fuel pin.

[Table 2] presents data on a comparison of neutron fluence of various energy groups and its percentage contribution in dose equivalent for various forms of Pu. From [Table 2], it is found that maximum contribution (27.94%) in the neutron dose is obtained for 2-3 MeV energy range when the PuO 2 is handled in powder form. On the other hand for MOX pellets and fuel pins the maximum contribution in neutron dose (MOX pellet-34.24% and fuel pins-27.34%) is coming from the neutron of the energy range of 1-1.5 MeV. As the density of pellets is higher than that of the powder form the process of slowing down of neutrons by the matrix itself probably causes this type of spectra. However in the case of fuel pins higher neutron fluence is obtained at lower energy range of 0-0.01 MeV and the corresponding dose contribution is not significant due to lower fluence to dose conversion coefficients in this energy range.

Comparison between neutron spectra obtained from FLUKA simulation and MICROSPEC N-probe data

Simulations have been carried out using FLUKA Monte Carlo code maintaining the same experimental conditions of the workplace. One glove box had been chosen for the simulation. It was modeled accurately for comparison with the actual experimental measurement. This glove box did not contain any other material except the source cylinder at the time of experimental measurement. The results are presented in [Table 3] and graphically represented in [Figure 7].
Table 2: Comparison of neutron fluence and dose rate per gram of Pu in various forms


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Table 3: Comparison of dose rate from experimental measurement and simulation


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Figure 7: Source inside the glove box

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In [Figure 7] measured and simulated neutron spectra have been plotted in the same graph. In this case, a cylinder containing PuO 2 is kept inside the glove box during measurements. Spectrum obtained from FLUKA simulation is almost similar to the experimental spectrum. Neutron scattering from surrounding materials due to complex geometry explains some mismatch in the spectra due to simulation.

[Table 3] shows a close match between the measured data from MICROSPEC with N-probe and the FLUKA simulated result.


  Conclusion Top


The measurements indicate that when the source is inside the shielded glove box gamma component does not interfere significantly. But for a bare source outside the glove box, gamma interference in the neutron dose measurement is significant. Hence, a bare PuO 2 , MOX pellet sources, and MOX fuel pins require 3 mm of lead shield to eliminate gamma interference in measurements using MICROSPEC with N-probe instrument. The results indicate that the dose equivalent contribution from MOX pellets is very significant in the energy group of 1.0-1.5 unlike bare PuO 2 and fuel pins which exhibit significant contribution in the energy groups of both 1.0-1.5 MeV and 2.0-3.0 MeV. Also, it is observed that the major contribution in ambient dose equivalent for MOX fuel pins (containing pellets) comes from energy group 1.0-1.5 MeV. Even though the fuel pins show high neutron fluence in the energy group of 0.0-0.01 MeV, they do not contribute significantly to dose equivalent. Also, the results indicate that the pellets show high neutron fluence in the low energy region compared with powder that has a lesser density than the pellets. Due to significant variation in composition and densities of PuO 2 powder and MOX pellets, variation in fluence of different energy groups has been observed. Fuel pins contain MOX pellet inside it. However, the difference between these two is stainless steel clad of fuel pin. Hence, it is concluded that the change in the neutron spectrum coming out of the powder, MOX pellet, and the fuel pin is attributed due to the scattering and slowing down of neutrons within the fuel matrix. The simulations using the FLUKA code shows a good agreement with the measured data. Hence in a situation where experimental measurements cannot be carried out due to practical difficulties, simulation using FLUKA code can be used to obtain information on the radiation field of that area.


  Acknowledgments Top


The authors are thankful to Shri Mohd. Afzal, OS, BARC and Shri. P. G. Behere, Scientific Officer, BARC for their encouragement and support. Also, we wish to thank Dr. P. K. Sarkar, Ex. Head, HPD, BARC for his guidance and help in carrying out these studies.

 
  References Top

1.
Baldev R, Kamath HS, Natarajan R, VasudevaRao PR. A perspective on fast reactor fuel cycle in India. Prog Nucl Energy 2005;47:369-79.  Back to cited text no. 1
    
2.
Kamath HS. Recycle Fuel Fabrication for Closed Fuel Cycle in India. Energy Procedia 2011;7:110-9.  Back to cited text no. 2
    
3.
Somayajulu PS, Kumar A, Panakkal JP, Kamath HS. PFBR fuel pellet fabrication for experimental irradiation. Characterisation and Quality Control of Nuclear Fuels. 2004. p. 318.  Back to cited text no. 3
    
4.
Ing H, Djeffal S, Clifford T, Machrafi R, Noulty R. Portable spectroscopic neutron probe. Radiat Prot Dosimetry 2007;126:238-43.  Back to cited text no. 4
    
5.
Verbinski VV, Burns WR, Love TA, Zobel W, Hill NW. Calibration of an organic scintillator for neutron spectroscopy. Nucl Instrum Methods 1968;65:8-25.  Back to cited text no. 5
    
6.
Tsujimura N, Yoshida T, Sagawa N, Shoji S. Neutron spectra measurements at JAEA MOX fuel facility. Prog Nucl Sci Technol 2011;1:154-7.  Back to cited text no. 6
    
7.
Fasso A, Ferrari A, Ranft J, Sala PR. FLUKA: A Multi-Particle Transport Code, CERN-2005-10, INFN/TC_05/11, SLAC-R-773; 2005.  Back to cited text no. 7
    
8.
Craig RA, Bliss M. Predicted Performance of Neutron Spectrometers Using Scintillating Fibers. Pacific Northwest National Laboratory Report No. PNNL-13111; 2000.  Back to cited text no. 8
    


    Figures

  [Figure 1], [Figure 2], [Figure 3], [Figure 4], [Figure 5], [Figure 6], [Figure 7]
 
 
    Tables

  [Table 1], [Table 2], [Table 3]



 

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