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ARTICLE
Year : 2011  |  Volume : 34  |  Issue : 4  |  Page : 262-266  

Analysis of neutron streaming through the trenches at linac based neutron generator facility, IGCAR


1 Radiological Safety Division, IGCAR, Kalpakkam, India
2 AERB-Safety Research Institute, Kalpakkam, India

Date of Web Publication17-Jan-2013

Correspondence Address:
Usha Pujala
Radiological Safety Division, IGCAR, Kalpakkam
India
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Source of Support: None, Conflict of Interest: None


DOI: 10.4103/0972-0464.106194

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  Abstract 

Shielded LINAC hall has been built to accommodate a radio frequency quadrupole (RFQ) linear accelerator (LINAC) based pulsed neutron generator at RSD, IGCAR. The concrete neutron shield wall thickness for the LINAC hall is finalized by adopting NCRP-51 methodology. The dimensions of the trenches and labyrinth are decided based on the neutron generator installation requirements. As per the AERB safety criteria, the radiation dose levels in the occupancy area should be less than 1μSv/h. However, the refined dose estimate with the inclusion of actual trench and labyrinth dimensions indicates an increased dose rate of ~3.10 μSv/h in LINAC control room. Hence, the control room is declared as controlled area. Additional shielding has been proposed for the trenches and labyrinth to make the control room as full occupancy area. For designing the additional shielding, experimental and theoretical analyses are needed to estimate the radiation streaming through the trenches and labyrinth. To start with, the neutron streaming through trenches have been studied using 185 GBq (5Ci) 241 Am-Be neutron source for qualifying the trenches. The 241 Am-Be source position is selected at the trench entrance such that it contributes the same neutron flux as that of neutron generator. Neutron dose rate and spectral measurements have been carried out at five locations along the trench from the entrance (LINAC hall side) to the exit (Control room side) of the trench.The experimental results are validated with the theoretical calculations using Monte Carlo N particle (MCNP) code. The analysis shows that the trenches are having a dose reduction factor better than 800 with respect to that of entrance dose. The observed dose rate at the trench exit is found to be less than 450nSv/h. In this paper, both the theoretical and experimental neutron streaming analyses through the trenches of LINAC hall are presented.

Keywords: Capture gamma, linear accelerator, Monte Carlo N particle, neutron streaming, radio frequency quadrupole


How to cite this article:
Pujala U, Thilagam L, Selvakumaran T S, Mohapatra D K, Raja E A, Subbaiah K V, Baskaran R. Analysis of neutron streaming through the trenches at linac based neutron generator facility, IGCAR. Radiat Prot Environ 2011;34:262-6

How to cite this URL:
Pujala U, Thilagam L, Selvakumaran T S, Mohapatra D K, Raja E A, Subbaiah K V, Baskaran R. Analysis of neutron streaming through the trenches at linac based neutron generator facility, IGCAR. Radiat Prot Environ [serial online] 2011 [cited 2020 Jun 2];34:262-6. Available from: http://www.rpe.org.in/text.asp?2011/34/4/262/106194


  1. Introduction Top


A compact RFQ LINAC based pulsed neutron generator is being procured from M/s D.V Effremov institute of electro physical apparatus (NIIEFA), Russia (IGC-Report, 2012). [1] In this generator, deuterium ions produced from an ion source are accelerated to the energy of 1 MeV by the RFQ LINAC and bombarded with beryllium target. The neutrons are produced from the 9 Be(d,n) 10 B nuclear reaction at the target. The expected neuron spectrum from the neutron generator (INADA et al., 1968) [2] is shown in [Figure 1]. The yield of neutron generator is 10 9 n/s/4π with 2.2 MeV average energy (Ē).
Figure 1: Neutron spectral distribution from a thick Be target for D+ ion beam of 0.9 MeV (INADA et al., 1968)

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The layout of shielded LINAC hall constructed to house the neutron generator; control room and cooling room are shown in [Figure 2]. LINAC hall is surrounded by 110 cm thick ordinary concrete (density 2.35 g/cm 3 )shield wall. Towards the control and cooling room side, the concrete wall is followed by 32 cm brick wall. A gap of 7 cm exists between the concrete and brick walls and it is filled with thermocole sheets. In the western side of the LINAC hall, an entrance (2.5 m × 3 m) with movable shield door (having 45 cm thick paraffin followed by 5 cm thick lead) is provided for material movement. The dimensions of the concrete shielding wall and shield door are optimized based on NCRP-51 methodology (NCRP-51, 1977) [3] in such a way that the dose rates in the surrounding areas of the LINAC hall are not exceeding the value of 1 μSv/h. For personnel entrance to the LINAC hall from the control room a labyrinth is provided to reduce the radiation level in the control room. Two trenches are provided between the LINAC hall to the control room for the passage of power and control cables. Another trench is provided between LINAC hall and cooling room for the cooling channels passage. All the three trenches are equi-dimensional. Dimensions of the Labyrinth and trenches are decided based on the requirement of installation. The revised dose estimate using NCRP-51 methodology with the inclusion of the trench and labyrinth shows an elevated dose rate of ~3.1 μSv/h in the control room and the dose rate at the cooling room remains less than1 μSv/h. The higher dose rate in the control room may be attributed to the neutron and associated capture gamma streaming through the trenches and labyrinth. Hence, the control room is declared as controlled area at present.

Additional shielding has been proposed for the trenches and labyrinth to make the control room as full occupancy area. Before designing the additional shield, experimental and theoretical studies using MCNP code are carried out to quantify the radiation streaming through each of the structures. The study would be helpful in design optimization of the additional shielding. In the present study, the experimental and theoretical analyses of neutron streaming through the trenches are carried out and the results are presented.
Figure 2: Layout of the linear accelerator hall

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  2. Experimental Method and Theoretical Simulation Top


2.1. Neutron streaming experiments

The experimental set up for measuring the neutron streaming through the control cable trench is shown in [Figure 3]. A standard 185 GBq (5Ci) 241 Am-Be neutron source is positioned at the trench entrance such that it contributes the same neutron flux as that of neutron generator. The yield of the 185 GBq (5Ci) 241 Am-Be source is 1.1E + 07 n/s/4π with the neutrons energy distributed from thermal to 11 MeV with the average neutron energy,(Ē) = 4.4 MeV (Arthur Eleftherakis, 2011). [4] Neutron dose rate and spectral measurements have been carried out at five positions along the trench from the entrance (LINAC hall side) to the exit (Control room side) of the trench.
Figure 3: Experimental arrangement for measuring the Neutron streaming through the trench

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A neutron monitor (Model-BDKN-03, Make-ATOMTEX, Republic of Belarus) [5] and a Neutron REM counter (Model-2222A, Make-Wedholm, Sweden) [6] are used for dose measurements. Both are having response to the neutron radiation in the energy range from 0.025 eV to 14 MeV. Neutron monitor can measure the equivalent dose rate in the range from 0.1 μSv/h to 10 mSv/h and dose in the range from 0.1 μSv to 10 Sv. Whereas REM counter can measure the dose rates in the range from 1 μSv/h to 999.9 mSv/h. For neutron spectral measurement, dual detector ( 3 He gas counter and NE213 detector) based neutron spectrometer (Model-BTI MICROSPEC-2, Make-Bubble Technology Industries, Chalk River, Canada) is used [7] In addition, dose rate measurements are also carried out using this neutron spectrometer to inter-compare its performance with neutron monitor and REM counter. In the neutron spectrometer, neutron energy region from thermal to 0.8 MeV is covered by the 3 He counter and from 0.8 MeV to 20 MeV is covered by NE213 scintillator. The neutron spectrum of 241 Am-Be source measured by using Neutron spectrometer is shown in [Figure 4]. The neutron spectrum measured is found to be comparable with that of ISO 8529-2 recommended. 241 Am-Be neutron spectrum (Lebreton et al., 2007; CONRAD, 2006) [8],[9] shown in [Figure 5]. Though the spectral characteristics of the 241 Am-Be neutron source are different from that of LINAC based neutron generator, the present work provides a conservative estimate because of the higher average neutron energy of 241 Am-Besource. The neutron monitor, REM counter and spectrometer are factory calibrated and the calibration is cross checked by the measurements using standard 185 GBq (5Ci) 241 Am-Be source.
Figure 4: Measured 241Am-Be neutron spectrum with Neutron spectrometer BTI MICROSPEC-2

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Figure 5: ISO 8529-2 recommended 241Am-Be neutron spectrum (Lebreton et al., 2007; CONRAD, 2006)

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2.2. Neutron streaming calculations using Monte Carlo method

Theoretical simulations are carried out using general-purpose, continuous energy, Monte Carlo N-particle transport code, MCNP (Briesmeister et al., 1997) [10] with ENDF/B-VI cross-section data. The exact experimental geometry as simulated in Monte Carlo N Particle code is shown in [Figure 6]. The compositions of concrete (NBS ordinary, density 2.35 g/cm 3 ) and brick (density 1.8 g/cm 3 ) given in MCNP Criticality Primer III (Roger Brewer, 2009) [11] have been used in all MCNP simulations. Other details used in MCNP simulations are given in [Table 1] (Arthur Eleftherakis and Martin Kocan, 2011; CONRAD, 2006; ICRP-21, 1971; NCRP-38; 1973; ICRP-74, 1997). [4],[9],[12],[13],[14]

Each simulation is performed for 2 billion neutron histories with the 'importance biasing' as variance reduction technique to reduce the statistical uncertainties less than 10% for F4 tallies. ISO 8529-2 recommended 241 Am-Be neutron spectrum [Figure 5] is used in all MCNP simulations.
Figure 6: Cross-sectional view of the trench with the five detectors locations

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Table 1: Monte Carlo N particle input details

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  3. Results and Discussions Top


Streaming neutron dose rates are measured at five different locations along the trench and are compared with theoretical MCNP simulations. The results of the analyses are presented in [Table 2] and plotted in [Figure 7] to show the dose rate reduction along the trench towards the control room side from LINAC hall side.
Figure 7: Calculated and measured streaming neutron dose rates at five locations

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Table 2: Calculated and measured streaming neutron dose rates

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The streaming neutron spectra measured using neutron spectrometer (BTI MICROSPEC-2) and simulated using MCNP are shown in [Figure 8] and [Figure 9] respectively for all the detector locations. They are found to be in agreement. Comparison of experimental measurements and theoretical analysis using MCNP calculations show that:

  1. The measured and calculated streaming neutron dose ratesusing ICRP-74 FDCF are found to be in good agreement with the percentage deviations within ±36% for all locations.
  2. The neutron dose rate at the control cables trench exit (D5) is observed to be ~800 times less than that of first location for the (Ē) = 4.4 MeV ( 241 Am-Bespectrum) neutrons as the input. Therefore for (Ē) = 2.2 MeV [ 9 Be (d, n) 10 B spectrum], the attenuation factor is expected to be more than 800. Since the dimensions of the trenches are same, this reduction factor is expected to be equal for all the other trenches. As the power cables trench is farther away from the neutron generator target as shown in [Figure 2], the neutron dose rate at its exit in the control room side will be less compared to that of the control cable trench.
  3. Dose rate at the exit of the control cables trench (D4 location) in the control room from streaming neutron is about 450 nSv/h for 185 GBq (5Ci) 241 Am-Besource. Therefore, the present study shows that the dose rate due to neutron streaming through the trenches to the control room will be less than 450 nSv/h during the operation of neutron generator.
Figure 8: Measured streaming neutron spectra using BTI MICROSPEC-2

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Figure 9: Calculated streaming neutron spectra using Monte Carlo N particle

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  4. Conclusion Top


Experimental and theoretical analysis of neutron streaming through the trenches of the LINAC hall shows that the neutron dose rate in the control room is about 450 nSv/h for 185 GBq (5Ci) 241 Am-Besource with average neutron energy (Ē) 4.4 MeV. Hence, it will be less than 450nSv/h during the operation of neutron generator with average neutron energy (Ē) 2.2 MeV. The dimensions of trench with the concrete boundary are found to be sufficiently large to moderate the neutrons and their absorption. Also, the study validates the present MCNP methodology for neutron streaming analysis. Capture gamma analysis associated with neutron streaming through the trenches is in progress. The requirement of the shield cover over the trench will be decided based on the total dose rate due to neutron and capture gamma.

 
  References Top

1.IGC-Report, Safety Analysis Report for Installation and Testing of LINAC based Neutron Source, 2012; IGC/RSEG/RSD/RTAS/92615/SR/3003/REV-D.  Back to cited text no. 1
    
2.Inada T, Kawachi K. Neutrons from Thick Target Beryllium (d,n)Reactions at1.0MeVto3.0MeV. Nucl Sci Technol 1968;5:22-9.  Back to cited text no. 2
    
3.NCRP Report No 51. Radiation protection design guidelines for 0.1-100 MeV particle accelerator facility; 1977.  Back to cited text no. 3
    
4.Eleftherakis A, Kocan M. Report on Results from preliminary checks on Am-Be Neutron source #71, Australian Government, Department of Defence, Human Protection and Performance Division, Defence Science and Technology Organization; 2011.  Back to cited text no. 4
    
5.ATOMTEX Radiation Monitor AT1117M users manual.  Back to cited text no. 5
    
6.WEDHOLM MEDICAL Neutron monitor 2222A users manual.  Back to cited text no. 6
    
7.BTI Spectroscopic Survey System MICROSPEC-2 users manual.  Back to cited text no. 7
    
8.Lebreton L, Zimbal A, Thomas D. Experimental comparison of 241 Am-Be neutron fluence energy distributions. Radiat Prot Dosimetry 2007;126:3-7.  Back to cited text no. 8
[PUBMED]    
9.CONRAD-European Union Co-ordinated Network for Radiation Dosimetry, "UNCERTAINTY ASSESSMENT IN COMPUTATIONAL DOSIMETRY", June 2006.  Back to cited text no. 9
    
10.Briesmeister JF."MCNP- A General Monte Carlo N-Particle Transport Code, Version4B," LA-12625-M; 1997.  Back to cited text no. 10
    
11.Brewer R. "MCNP Criticality Primer III". New Mexico: Los Alamos National Laboratory; 2009.  Back to cited text no. 11
    
12.ICRP-74, International Committee on Radiological Protection, Conversion Coefficients for use in Radiological Protection against External Radiation, Report 74, Annals of the ICRP, Vol. 26. Oxford: Pergamon Press; 1997.  Back to cited text no. 12
    
13.NCRP-38, National Council on Radiation Protection and Measurements, Protection Against Neutron Radiation, Report 38, Washington D.C: NCRP Publications; 1973.  Back to cited text no. 13
    
14.ICRP-21, International Commission on Radiological Protection, ICRP Committee 3 Task Group, In: Grande P, O'Riordan MC, chair-men, Data for Protection Against Ionizing Radiation from External Sources: Supplement to ICRP Publication 15," Oxford: Pergamon Press; April 1971.  Back to cited text no. 14
    


    Figures

  [Figure 1], [Figure 2], [Figure 3], [Figure 4], [Figure 5], [Figure 6], [Figure 7], [Figure 8], [Figure 9]
 
 
    Tables

  [Table 1], [Table 2]


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