- Email: [email protected]
Detector sensitivity and calibration for yield measurements in fusion neutron sources
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Contents lists available at ScienceDirect
Nuclear Inst. and Methods in Physics Research, A journal homepage: www.elsevier.com/locate/nima
Detector sensitivity and calibration for yield measurements in fusion neutron sources Sanjay Andola a,b ,∗, A.M. Rawool a , T.C. Kaushik a,b a b
Applied Physics Division, Bhabha Atomic Research Centre, Mumbai-400 085, India Homi Bhabha National Institute, Mumbai-400094, India
ABSTRACT In this study, numerical simulations have been carried out to evaluate the effect of moderator parameters (material/dimensions) for standard radio-isotopic/fission (Pu-Be and 252 Cf) and fusion sources (D-D). These sources are used for calibrating thermal neutron detectors, which are often employed for determining absolute neutron yield in fusion-based sources. Simulations have been carried out using the Monte Carlo based ‘‘FLUKA’’ code for typical moderator materials and 3 He/BF3 proportional counter based thermal neutron detectors. Results have been validated against experimentally measured values for an available radio-isotopic source. A notable difference in the detector efficiency was observed due to the source energy term. The detection efficiency ratio for Pu-Be:D-D:252 Cf sources with the same moderator thickness (optimized for 252 Cf) was found to be 1 : 1.16±0.02 : 1.40±0.03. For these three types of sources, optimum acrylic (Perspex) and polypropylene moderator thicknesses have been observed to be 100,100,80 mm and 80,80,60 mm respectively with the corresponding efficiency ratio of 1 : 1.15±0.02 : 1.34±0.008.
1. Introduction The neutrons generated in fusion devices such as plasma focus, zpinch array of deuterated fibres are typically pulsed in nature and practically mono-energetic [1–3]. Various types of neutron detectors such as scintillators, activation detectors, and proportional counters are often employed for the fusion yield measurement. The first one works in the fast energy regime while the latter two are better suited for thermal neutron or as threshold energy neutron detectors. Plastic scintillation detectors have a disadvantage of being sensitive to Xrays that are inevitable products in such experiments [4]. Neutron detectors working in the thermal energy region have relatively better sensitivity due to the comparatively large cross-section of interaction among others. Thus, for the experiment where information only about neutron yield (without temporal profile) is required, thermal neutron detectors appear to be a better option. To quantify the yield, such detectors need to be generally calibrated with fission/radio-isotopic sources such as 252 Cf, Pu-Be, and Am-Be which in turn are calibrated for their absolute strength in various ways i.e., manganese bath system (MBS)and energy independent long counters [5,6]. The neutron yield in high yield fusion devices can be determined using foil activation detectors but for low yield devices, proportional counter detectors appear to be a superior option owing to their significantly higher sensitivity. The activation type detection system uses two approaches namely first principle and the ‘F’ factor approach for fusion yield measurement [7,8]. The first principle approach deals with the fundamental quantities such as the activation reaction cross section and ∗
detection efficiencies of the system while the F factor method is based on the entire system calibration methodology. The fraction of thermal neutrons in the spectrum of sources whether fission or fusion, even after moderation is quite small (up to a few percent) [9]. Performance of thermal neutron detectors can be maximized by selection of appropriate moderating material and optimizing their thickness. For thermal neutron detection, 3 He and BF3 based proportional counters are the suitable options due to their large cross section and good gamma discrimination [10,11]. To enhance the sensitivity of the detection system especially for low yield sources, these detectors can even be assembled as a bank [12]. However, irrespective of the actual detector, the fraction of thermal neutrons after moderation would not be same for fusion (to be monitored) and fission (used for calibration) sources. In addition to the thermal region, the other regions of the neutron spectra would also contribute to the overall signal. Therefore, evaluation of the effect of moderator and its thickness with respect to the signal and efficiency of detection in fusion or fission sources is necessary [13]. This aspect is the theme of the present investigation. In the present work, Monte Carlo simulations have been carried out using the FLUKA code to optimize moderating thickness of perspex and polyethylene sheets for mono-energetic (D-D fusion based) and wide energy radio-isotopic (fission) sources [14]. We have not considered the D-T fusion source due to its significantly higher neutron energy (14.06 MeV) as compared to wide energy radio-isotopic sources. The responses of 3 He proportional detectors have been simulated to investigate the effect of the moderator. Furthermore, the response of BF3
Corresponding author at: Applied Physics Division, Bhabha Atomic Research Centre, Mumbai-400 085, India. E-mail address: [email protected] (S. Andola).
https://doi.org/10.1016/j.nima.2019.162580 Received 12 April 2019; Received in revised form 6 August 2019; Accepted 11 August 2019 Available online 16 August 2019 0168-9002/© 2019 Elsevier B.V. All rights reserved.
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Fig. 1. Simulation geometries for the moderation thickness calculation (a) 3 He Single detector tube (b) BF3 Detector bank (c) 3 He Detector bank with concrete wall around the set up.
Fig. 2. Output of RESINUCLEi card (residual nuclei) at optimum thickness of perspex (100,100,80 mm) and polyethylene (80,80,60 mm) moderator in 3 He single detector with sources (a) perspex (b) Polyethylene.
detectors was also investigated since these detectors are generally being used as an alternative to 3 He detectors [15]. The radio-isotopic sources considered are Pu-Be (𝛼-n and SF) and 252 Cf (SF) and D-D as a fusion source. To evaluate detector response, the aforementioned sources are
encapsulated in a cylinder of 4 cm diameter and 10 cm in height. For the experimental validation, we have investigated 3 He and BF3 detector banks in addition to single detectors using a standard Pu-Be neutron source available in the laboratory. 2
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Fig. 3. Absolute detection efficiency (counts/source neutron) Pu-Be, D-D fusion neutron and 252 Cf sources with varying moderator thicknesses (a) 3 He in open-space with perspex as moderator, (b) 3 He detector in a closed-room of 6 m × 8 m × 6 m with 20 cm thick concrete walls, (c) 3 He detector with polyethylene moderator(d) BF3 with perspex moderator (e) BF3 with polyethylene moderator.
A short description of the code is provided in the next section followed by details on set up considered for simulation and experimental validation.
library in FLUKA has data for 250 materials, typically used in physics, dosimetry and accelerator engineering. The Neutron interaction crosssection of different isotopes is taken into account using the LOW-MAT card. The scorecard used for sensitivity calculations in the proportional counters was RESINUCLEi card. This card calculates residual nuclei formed in the gas region of the detector. In 3 He and BF3 filled detectors, residual nuclei 1 H, 3 H and 4 He, 7 Li produced by the interaction of neutrons in the active gas volume indicate absolute efficiency of the detector.
2. Simulation code FLUKA is a Monte Carlo code available on an open platform and widely used in calculations of particle transport or interaction with matter [14]. It has been widely used in accelerators, shielding designs, calorimetry, activation of materials, accelerator driven systems, cosmic rays, neutrino physics, and radiotherapy. We have carried out Monte Carlo simulations using this code for neutron transport in moderating media, detector gas, and surrounding regions. The default card used in FLUKA were ‘‘neutron’’ and ‘‘precision’’. Both the cards transport low energy neutrons with energy ranging from 10−5 eV to 20 MeV in 260 groups, in addition to this the latter one also deals with other types of radiations. The cross-section
3. Set-up for numerical simulation As noted above 3 He and BF3 proportional counters have been considered for the detection of neutrons originating from the fission or fusion sources. The 3 He detector is one-metre-long filled with 3 He and Kr gases to pressure of 3 bars and 1.5 bars respectively [16]. 3
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Table 1 Moderator thickness with corresponding efficiency ratios of 3 He and BF3 single detectors and detector banks for Pu-Be, D-D fusion and sources. Detector
3
He 3 He BF3 3 He BF3
single Detector (open-space) single Detector single detector Detector Bank detector Bank
Perspex moderator
252 Cf
Polyethylene moderator
Thickness (Pu-Be,D-D,CF) mm
Efficiency
100,100,80 100,100,80 100,100,80 80,80,60 60,60,40
1:1.15:1.31 1:1.13:1.35 1:1.14:1.34 1:1.14:1.31 1:1.17:1.32
Thickness (Pu-Be,D-D,CF) mm
Efficiency
80,80,60 80,80,60
1:1.15:1.35 1:1.19:1.34
sum mode and embedded in a grooved perspex moderator. The size of the grooved perspex sheet is 5 cm (t) × 32 cm (w) × 96 cm (l). The BF3 detector bank consists of five single proportional counters embedded in a grooved perspex moderator of 7 cm thickness. The size of this grooved perspex sheet is 7 cm (t) × 36 cm (w) × 100 cm (l) for BF3 detector bank. The neutron source is encapsulated in an aluminium cylinder having 4 cm diameter and 10 cm height. The source was placed at 100 cm from the surface of the detectors with an isotropic spatial source distribution being assumed for all the cases. The energy spectrum of the Pu-Be source was taken from a continuous spectrum by digitizing it in 46 energy bins [17]. The 252 Cf source energy spectrum was sampled according to a Watt spectrum with probability density being taken 𝐸 1 as [18]𝑃 (𝐸) = 𝐶𝑒− 𝑎 𝑆𝑖𝑛ℎ(𝑏𝐸) 2 where C, a, and b are constants which depend upon the type of isotope [19] 𝐶=(
Fig. 4. Comparison of efficiencies with variation in Perspex thickness for (a) Detector bank (b) BF3 detector bank using different neutron sources.
𝜋𝑎𝑏 1∕2 𝑎𝑏4 ) 𝑒 𝑎 4
The constants a, b for 252 Cf isotope are 1.025 and 2.926 respectively. A combinational geometry was designed in FLUKA for the detection set up. A total 5 cycles of 106 histories were run for each configuration. LOW-Mat card was used for hydrogen in CH2 bonds, 3 He and 10 B and 11 B in BF , which sets the correspondence between FLUKA materials 3 and low-energy neutron cross sections. For simulations, we have considered two geometries of the surroundings; a closed room and completely open-space. The latter is a hypothetical case that has been simulated for only one detector configuration (3 He single detector) to realize the build-up in neutron population due to wall reflections that contribute to the absolute efficiency of a detector. In closed-room (a real situation) geometry, the detector has been placed in a room of size 6 × 8 × 6 m3 as shown in Fig. 1. The thickness of walls, ceiling, as well as the floor each is considered as 20 cm while for the open-space geometry, the detector has been placed in an assumed open space (no walls, ceiling or floor). The density and composition of concrete, polyethylene and perspex was taken from standard FLUKA material library.
3 He
4. Experimental setup The active length of the detector is taken as 920 mm with 38 mm diameter. The BF3 detector is 5 cm in diameter and one-metre long with active length of 920 mm and gas pressure of 0.73 bar. The Moderator under consideration was placed in front of the detector at touching position. The RESINUCLEi card, which simulates the number of residual nuclei formed by interaction of neutrons in the 3 He/BF3 gas, has been used to estimate the efficiency of the detector. In the experimental scenario both the residual nuclei further ionize the detector gas and produce a voltage pulse, which is amplified and discriminated against electronic noise and background gamma pulses, however in simulations the number of residual nuclei have been considered for calculation of absolute detector efficiency (counts/source-neutron or c/n). The absolute efficiency of the detector was calculated for moderator thicknesses varying from 0 to 200 mm in steps of 20 mm. As mentioned earlier, 3 He and BF3 detector banks have also been simulated (in addition to single detectors). The 3 He detector bank consists of six proportional counters (described above) connected in
Simulations have been validated against experimental results for the cases of detectors within walls (a realistic physical situation similar to closed-room geometry described above) using a Pu-Be source. A Pu-Be source of intensity 1 ± 0.05 × 105 neutron/s was placed at 100 cm from the surface of the detector to minimize the effect of source size on the detection efficiency. The counting was carried out for 10, 50 and 50 s for detector bank, single detector and background respectively. The net counts were corrected for background in every case. 5. Results and discussions The response of above-mentioned sources in single as well as in the form of banks of 3 He and BF3 detectors has been numerically simulated for varying thickness of moderator. The simulation geometry of the detection system is shown in Fig. 1. The output of the RESINUCLEi card at optimum thickness of perspex and polyethylene in single detectors is shown in Fig. 2. As described in later segment of this para the optimum 4
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Fig. 5. Comparison of experimental and Monte-Carlo simulation results with perspex and Pu-Be neutron source for (a) 3 He detector (b) 3 He detector bank (c) BF3 detector (d) BF3 detector bank. Table 2 Ratio of efficiencies for three sources with same moderator thickness (optimized for Configuration
3
He 3 He BF3 3 He BF3
detector (open-space) single detector single detector detector bank detector bank
Perspex
252
Cf). Polyethylene
Thickness mm
Efficiency (Pu-Be:D-D:
80 80 80 60 40
1:1.15:1.42 1:1.15:1.42 1:1.16:1.41 1:1.15:1.32 1:1.16:1.43
thickness of perspex and polyethylene for the three sources (252 Cf, PuBe and D-D) is 100, 100, 80 mm and 80, 80, 60 mm respectively. The residual nuclei (3 H and 1 H in 3 He; and 7 Li and 4 He in BF3 ) formed due to neutron interactions indicate the absolute efficiency (reactions per source-neutron) for the given detector configuration. It can be seen (Fig. 2) that the polyethylene provides better moderation with smaller thickness as compared to perspex. The absolute efficiency of single 3 He and BF3 detectors with varying perspex and polyethylene thickness for each of the three types of sources is shown in Fig. 3. It was observed that as the moderator thickness is increased; the detection efficiency reaches to a certain optimum value and then decreases. The increase in the efficiency is quite expected due to enhanced thermalization of source neutrons by the moderator. Further increase in moderator thickness beyond a certain value causes absorption of the thermalized neutrons by the hydrogenous media due to its finite absorption cross section in hydrogen [20], thereby leading to a lowering in the efficiency of detection. Simulations of the single 3 He detector with perspex moderator have been carried out for the setup located in open-space and for a closed-room, as mentioned above (Section 3). The maximum efficiency for a single detector, with the Pu-Be source, in open-space and in a closed-room are observed to be 1.20×10−4 and 2.94×10−4 respectively. Significant enhancement (2.4 times) in the detector efficiency is observed for the closed-room configuration compared to that in openspace. This is essentially due to the reflection of neutrons from the
252
Cf)
Thickness mm
Efficiency (Pu-Be:D-D:
60 60
1:1.18:1.44 1:1.16:1.34
252
Cf)
concrete walls. Simulations have also been carried out to investigate the effect of individual surfaces of closed-room on the detector efficiency for a single 3 He detector with Pu-Be neutron source. The floor (being the closest) and ceiling provide an enhancement in the efficiency of 33% and 4% as compared to open-space case. In comparison to the case of the geometry consisting detector, floor, and ceiling, the geometry consisting surface behind the detector and surface behind the source enhances the efficiency to 32% and 5% respectively, however, both the side walls provide an enhancement of only 2%. The combined effect of all surfaces provides 2.4 times enhancement in detector efficiency as compared to set-up in the open-space. The effect of moderation material has been studied with perspex and polyethylene in both types of single detectors. The optimum moderator thickness of Pu-Be:D-D: 252 Cf sources for perspex and polyethylene were found out to be 100, 100, 80 mm and 80, 80, 60 mm respectively. The difference in efficiency with these two moderators is expected due to hydrogen content; however, the efficiency at optimum moderator thickness seems to be almost the same for both. The ratio of efficiency of the 3 He detector in open space with Pu-Be:D-D:252 Cf source was 1:1.15:1.31 at 100,100,80 mm perspex moderator thickness, while the same for the detector configured within the walls, the ratio of efficiency comes out to be 1:1.13:1.35. The slight difference in the ratio of these two could be due to statistical error (±1%–2%) in the simulations. For a realistic approach, all other simulations were carried out within the walls. In BF3 detectors the 5
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Fig. 6. Fluence of neutron after the moderator while source is at 100 cm from the surface of detector (a) Fluence after moderator with varying thickness with D-D source (b) The neutron fluence after 60 mm moderator thickness. (Inset: fluence of thermal neutrons).
efficiency ratio at optimum moderator thickness for Pu-Be:D-D: 252 Cf source is 1:1.14:1.34, while with polyethylene moderator the ratio is slightly different (1:1.19:1.34). Comparative data of the detection efficiency for all the three sources at optimum moderator thickness is listed in Table 1. The average efficiency ratio for these sources in 3 He and BF3 detector is 1 : 1.13±0.006 : 1.35±0.007 and 1 : 1.16±0.02 : 1.34±0.008 for perspex and polyethylene respectively. Fig. 4 illustrates the variation of detection efficiency with moderator thickness. Compared to the two other sources, the highest efficiency was observed for the 252 Cf source that could be due to the relatively softer energy spectrum of 252 Cf. The counts with D-D sources are found to be relatively higher in large moderation thickness than in case of 252 Cf and Pu-Be sources. Due to the continuous energy spectrum for Pu-Be and 252 Cf, excess moderation not only thermalizes the fast neutrons but also absorbs the lower energy ones. The results of simulation of the aforementioned detectors for the PuBe source is validated through experimental measurements as shown in Fig. 5. Due to electronic processing, (described in Section 3), the low height neutron pulses could be lost in pulse discrimination, that reduces the registered count rate, so the experimental values are generally found to be less than simulated results. It was observed that experimental values of efficiency are higher than the simulated ones for the case of detector banks (Fig. 5(b) and (d)). This may be attributed to reflection of neutrons from various other materials present in the laboratory, but in case of single detectors, the effect of reflection from the surrounding laboratory material has not been observed due to the small surface
area of the detector. In BF3 detectors, the pulse discrimination is better owing to its higher full energy peak (764 keV for 3 He against 2.3 MeV for BF3 ), hence the effect of reflection appears to be higher in the BF3 as compared to 3 He detector banks. The trends of variation in efficiency with moderator thickness are seen in agreement with each other. The optimum moderation thickness for single detector and the bank are also found to be different. This can be attributed to additional moderation provided by the grooved perspex sheets inside detector banks. To check the difference in counts due to grooved perspex we have also simulated the detection efficiency of both the detector banks without grooved perspex. It is seen that embedding by perspex sheets provides additional moderation to the source neutrons that results up to a 25% enhancement in the detection efficiency in the 3 He Detector bank. In the BF3 detector bank enhancement in detection efficiency is more than factor of two. The enhancement is maximum for 252 Cf (23% in 3 He and 150% in BF3 ) source compared to the other two (5 and 8% in 3 He and 106 and 109% in BF3 for Pu-Be and D-D source respectively). The ratio of efficiencies for the three sources have been observed to be 1:1.12:1.25 and 1:1.15:1.35 for the 3 He and BF3 detector bank respectively. The enhancement in the case of BF3 is expected because of the larger width of available moderator surrounding the detector. The average neutron fluence without moderator at the detector location is 1.3±0.06×10−5 n/cm2 . The fluence after the moderator near its optimum thickness for D-D fusion neutron source is plotted in Fig. 6. The fluence can be plotted in /cm2 /energy-bin/source-neutron, 6
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
which is always less than 1. The energy bins in the FLUKA library are non-uniform, so the fluence has been plotted here in /cm2 /MeV/sourceneutron, which is higher than 1 for lower energy range. It is clearly seen that the neutron fluence is increasing in the lower energy region up to a certain moderator thickness. Beyond 120 mm of moderator thickness, the thermal fluence starts decreasing; leading to lowering of detection efficiency. The fluence after 60 mm Perspex moderator clearly shows that neutron flux after the moderator is higher for 252 Cf source compared to the other two sources that impart to higher sensitivity. The Ratio of detection efficiency of the detectors for the same thickness (optimized for 252 Cf) of polyethylene moderator for all the three sources is shown in Table 2. For such a case, the ratio of detection efficiency for these sources can be taken as 1 : 1.16±0.02 : 1.4±0.03. Similarly, the optimum moderation thickness plays a crucial role in the correct determination of yield as seen with three types of sources (PuBe, D-D and 252 Cf), which is 100, 100 and 80 mm for the perspex and 80, 80, 60 mm for polyethylene. It suggests that for the same detector, if the moderator thickness is optimized for calibrating source, there is a difference in the detection efficiencies while monitoring fusion device due to the energy spectrum that needs to be accounted. This also implies that the calibration factor for the type of source needs to be correctly taken into account for the yield measurement in either type of source.
observed that for the measurements of fusion neutrons, calibration of detectors using radio-isotopic sources could yield an error even with the optimum thickness of moderator which could be up to 40% using pure fission sources. Acknowledgement It is a pleasure to acknowledge technical support by our colleague Mr A V Patil during the experiments in this work. References [1] D. Klir, J. Kravarik, P. Kubes, K. Rezac, S.S. Ananev, Y.L. Bakshaev, P.I. Blinov, A.S. Chernenko, E.D. Kazakov, V.D. Korolev, Phys. Plasmas 15 (2008) 032701. [2] J.W. Mather, Phys. Fluids 8 (1965) 366. [3] C.L. Ruiz, G.W. Cooper, S.A. Slutz, J.E. Bailey, G.A. Chandler, T.J. Nash, T.A. Mehlhorn, R.J. Leeper, D. Fehl, A.J. Nelson, Phys. Rev. Lett. 93 (2004) 015001. [4] G.H.V. Bertrand, M. Hamel, S.p. Normand, F. Sguerra, Nucl. Instrum. Methods Phys. Res. A 776 (2015) 114. [5] L.V. East, R.B. Walton, Nucl. Instrum. Methods 72 (1969) 161. [6] E.J. Axton, P. Cross, J.C. Robertson, J. Nucl. Energy Parts A/B. React. Sci. Technol. 19 (1965) 409. [7] C.W. Barnes, A.R. Larson, G. LeMunyan, M.J. Loughlin, Rev. Sci. Instrum. 66 (1995) 888. [8] G.W. Cooper, C.L. Ruiz, Rev. Sci. Instrum. 72 (2001) 814. [9] L.B. Rees, J.B. Czirr, Nucl. Instrum. Methods Phys. Res. A 691 (2012) 72. [10] D.H. Beddingfield, N.H. Johnson, H.O. Menlove, Nucl. Instrum. Methods Phys. Res. A 455 (2000) 670. [11] Z.M. Zeng, H. Gong, Q. Yue, J.M. Li, Nucl. Instrum. Methods Phys. Res. A 866 (2017) 242. [12] A. Lintereur, K. Conlin, J. Ely, L. Erikson, R. Kouzes, E. Siciliano, D. Stromswold, M. Woodring, Nucl. Instrum. Methods Phys. Res. A 652 (2011) 347. [13] S. Pszona, Nucl. Instrum. Methods Phys. Res. A 402 (1998) 139. [14] A. Ferrari, P.R. Sala, A. Fasso, J. Ranft, FLUKA: A multi-particle transport code (Program version 2005), 2005. [15] R.T. Kouzes, J.H. Ely, L.E. Erikson, W.J. Kernan, A.T. Lintereur, E.R. Siciliano, D.L. Stephens, D.C. Stromswold, R.M. Van Ginhoven, M.L. Woodring, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 623, 1035. [16] T.C. Kaushik, L.V. Kulkarni, R.K. Rout, A.M. Rawool, S.C. Gupta, A.M. Shaikh, Neutron detectors for wide range monitoring in pulsed electromagnetic environment, in: Proceedings of DAE-BRNS National Symposium on Compact Nuclear Instruments and Radiation Detector-2005, 2005. [17] M.E. Anderson, R.A. Neff, Nucl. Instrum. Methods 99 (1972) 231. [18] C.J. Everett, E.D. Cashwell, A third Monte Carlo sampler, Los Alamos Report LA-9721-MS, 1983. [19] MCNP-A General Monte Carlo N-Particle Transport Code, Version 5, Los Alamos National Laboratory Los Alamos, NM, 2003. [20] B. Hamermesh, G.R. Ringo, S. Wexler, Phys. Rev. 90 (1953) 603.
6. Summary and conclusions In the present work, a comparative study has been carried out to investigate the effect of source energy term on moderator thickness for three different types of sources including mono energetic (fusion) and broad energy spectrum (radio-isotopic and fission). For fusion sources of unknown yields, it is necessary to calibrate the detectors using standard fission/radio-isotopic sources such as Pu-Be, 252 Cf or Am-Be etc. Monte Carlo simulations have been carried out to optimize the moderator thickness in commonly used proportional counters i.e. 3 He and BF3 gas filled neutron detectors. The simulation results have been validated through experiments using a Pu-Be source. The optimum moderator thicknesses are in agreement with the simulation, although the efficiencies are found to be higher in experiments possibly due to reflection of neutrons from various materials in the laboratory. The results are significant and useful for applications where neutron yields from sources of unknown strength need to be characterized based on calibration of detectors using standard fission sources. Moreover, it is
7
Contents lists available at ScienceDirect
Nuclear Inst. and Methods in Physics Research, A journal homepage: www.elsevier.com/locate/nima
Detector sensitivity and calibration for yield measurements in fusion neutron sources Sanjay Andola a,b ,∗, A.M. Rawool a , T.C. Kaushik a,b a b
Applied Physics Division, Bhabha Atomic Research Centre, Mumbai-400 085, India Homi Bhabha National Institute, Mumbai-400094, India
ABSTRACT In this study, numerical simulations have been carried out to evaluate the effect of moderator parameters (material/dimensions) for standard radio-isotopic/fission (Pu-Be and 252 Cf) and fusion sources (D-D). These sources are used for calibrating thermal neutron detectors, which are often employed for determining absolute neutron yield in fusion-based sources. Simulations have been carried out using the Monte Carlo based ‘‘FLUKA’’ code for typical moderator materials and 3 He/BF3 proportional counter based thermal neutron detectors. Results have been validated against experimentally measured values for an available radio-isotopic source. A notable difference in the detector efficiency was observed due to the source energy term. The detection efficiency ratio for Pu-Be:D-D:252 Cf sources with the same moderator thickness (optimized for 252 Cf) was found to be 1 : 1.16±0.02 : 1.40±0.03. For these three types of sources, optimum acrylic (Perspex) and polypropylene moderator thicknesses have been observed to be 100,100,80 mm and 80,80,60 mm respectively with the corresponding efficiency ratio of 1 : 1.15±0.02 : 1.34±0.008.
1. Introduction The neutrons generated in fusion devices such as plasma focus, zpinch array of deuterated fibres are typically pulsed in nature and practically mono-energetic [1–3]. Various types of neutron detectors such as scintillators, activation detectors, and proportional counters are often employed for the fusion yield measurement. The first one works in the fast energy regime while the latter two are better suited for thermal neutron or as threshold energy neutron detectors. Plastic scintillation detectors have a disadvantage of being sensitive to Xrays that are inevitable products in such experiments [4]. Neutron detectors working in the thermal energy region have relatively better sensitivity due to the comparatively large cross-section of interaction among others. Thus, for the experiment where information only about neutron yield (without temporal profile) is required, thermal neutron detectors appear to be a better option. To quantify the yield, such detectors need to be generally calibrated with fission/radio-isotopic sources such as 252 Cf, Pu-Be, and Am-Be which in turn are calibrated for their absolute strength in various ways i.e., manganese bath system (MBS)and energy independent long counters [5,6]. The neutron yield in high yield fusion devices can be determined using foil activation detectors but for low yield devices, proportional counter detectors appear to be a superior option owing to their significantly higher sensitivity. The activation type detection system uses two approaches namely first principle and the ‘F’ factor approach for fusion yield measurement [7,8]. The first principle approach deals with the fundamental quantities such as the activation reaction cross section and ∗
detection efficiencies of the system while the F factor method is based on the entire system calibration methodology. The fraction of thermal neutrons in the spectrum of sources whether fission or fusion, even after moderation is quite small (up to a few percent) [9]. Performance of thermal neutron detectors can be maximized by selection of appropriate moderating material and optimizing their thickness. For thermal neutron detection, 3 He and BF3 based proportional counters are the suitable options due to their large cross section and good gamma discrimination [10,11]. To enhance the sensitivity of the detection system especially for low yield sources, these detectors can even be assembled as a bank [12]. However, irrespective of the actual detector, the fraction of thermal neutrons after moderation would not be same for fusion (to be monitored) and fission (used for calibration) sources. In addition to the thermal region, the other regions of the neutron spectra would also contribute to the overall signal. Therefore, evaluation of the effect of moderator and its thickness with respect to the signal and efficiency of detection in fusion or fission sources is necessary [13]. This aspect is the theme of the present investigation. In the present work, Monte Carlo simulations have been carried out using the FLUKA code to optimize moderating thickness of perspex and polyethylene sheets for mono-energetic (D-D fusion based) and wide energy radio-isotopic (fission) sources [14]. We have not considered the D-T fusion source due to its significantly higher neutron energy (14.06 MeV) as compared to wide energy radio-isotopic sources. The responses of 3 He proportional detectors have been simulated to investigate the effect of the moderator. Furthermore, the response of BF3
Corresponding author at: Applied Physics Division, Bhabha Atomic Research Centre, Mumbai-400 085, India. E-mail address: [email protected] (S. Andola).
https://doi.org/10.1016/j.nima.2019.162580 Received 12 April 2019; Received in revised form 6 August 2019; Accepted 11 August 2019 Available online 16 August 2019 0168-9002/© 2019 Elsevier B.V. All rights reserved.
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Fig. 1. Simulation geometries for the moderation thickness calculation (a) 3 He Single detector tube (b) BF3 Detector bank (c) 3 He Detector bank with concrete wall around the set up.
Fig. 2. Output of RESINUCLEi card (residual nuclei) at optimum thickness of perspex (100,100,80 mm) and polyethylene (80,80,60 mm) moderator in 3 He single detector with sources (a) perspex (b) Polyethylene.
detectors was also investigated since these detectors are generally being used as an alternative to 3 He detectors [15]. The radio-isotopic sources considered are Pu-Be (𝛼-n and SF) and 252 Cf (SF) and D-D as a fusion source. To evaluate detector response, the aforementioned sources are
encapsulated in a cylinder of 4 cm diameter and 10 cm in height. For the experimental validation, we have investigated 3 He and BF3 detector banks in addition to single detectors using a standard Pu-Be neutron source available in the laboratory. 2
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Fig. 3. Absolute detection efficiency (counts/source neutron) Pu-Be, D-D fusion neutron and 252 Cf sources with varying moderator thicknesses (a) 3 He in open-space with perspex as moderator, (b) 3 He detector in a closed-room of 6 m × 8 m × 6 m with 20 cm thick concrete walls, (c) 3 He detector with polyethylene moderator(d) BF3 with perspex moderator (e) BF3 with polyethylene moderator.
A short description of the code is provided in the next section followed by details on set up considered for simulation and experimental validation.
library in FLUKA has data for 250 materials, typically used in physics, dosimetry and accelerator engineering. The Neutron interaction crosssection of different isotopes is taken into account using the LOW-MAT card. The scorecard used for sensitivity calculations in the proportional counters was RESINUCLEi card. This card calculates residual nuclei formed in the gas region of the detector. In 3 He and BF3 filled detectors, residual nuclei 1 H, 3 H and 4 He, 7 Li produced by the interaction of neutrons in the active gas volume indicate absolute efficiency of the detector.
2. Simulation code FLUKA is a Monte Carlo code available on an open platform and widely used in calculations of particle transport or interaction with matter [14]. It has been widely used in accelerators, shielding designs, calorimetry, activation of materials, accelerator driven systems, cosmic rays, neutrino physics, and radiotherapy. We have carried out Monte Carlo simulations using this code for neutron transport in moderating media, detector gas, and surrounding regions. The default card used in FLUKA were ‘‘neutron’’ and ‘‘precision’’. Both the cards transport low energy neutrons with energy ranging from 10−5 eV to 20 MeV in 260 groups, in addition to this the latter one also deals with other types of radiations. The cross-section
3. Set-up for numerical simulation As noted above 3 He and BF3 proportional counters have been considered for the detection of neutrons originating from the fission or fusion sources. The 3 He detector is one-metre-long filled with 3 He and Kr gases to pressure of 3 bars and 1.5 bars respectively [16]. 3
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Table 1 Moderator thickness with corresponding efficiency ratios of 3 He and BF3 single detectors and detector banks for Pu-Be, D-D fusion and sources. Detector
3
He 3 He BF3 3 He BF3
single Detector (open-space) single Detector single detector Detector Bank detector Bank
Perspex moderator
252 Cf
Polyethylene moderator
Thickness (Pu-Be,D-D,CF) mm
Efficiency
100,100,80 100,100,80 100,100,80 80,80,60 60,60,40
1:1.15:1.31 1:1.13:1.35 1:1.14:1.34 1:1.14:1.31 1:1.17:1.32
Thickness (Pu-Be,D-D,CF) mm
Efficiency
80,80,60 80,80,60
1:1.15:1.35 1:1.19:1.34
sum mode and embedded in a grooved perspex moderator. The size of the grooved perspex sheet is 5 cm (t) × 32 cm (w) × 96 cm (l). The BF3 detector bank consists of five single proportional counters embedded in a grooved perspex moderator of 7 cm thickness. The size of this grooved perspex sheet is 7 cm (t) × 36 cm (w) × 100 cm (l) for BF3 detector bank. The neutron source is encapsulated in an aluminium cylinder having 4 cm diameter and 10 cm height. The source was placed at 100 cm from the surface of the detectors with an isotropic spatial source distribution being assumed for all the cases. The energy spectrum of the Pu-Be source was taken from a continuous spectrum by digitizing it in 46 energy bins [17]. The 252 Cf source energy spectrum was sampled according to a Watt spectrum with probability density being taken 𝐸 1 as [18]𝑃 (𝐸) = 𝐶𝑒− 𝑎 𝑆𝑖𝑛ℎ(𝑏𝐸) 2 where C, a, and b are constants which depend upon the type of isotope [19] 𝐶=(
Fig. 4. Comparison of efficiencies with variation in Perspex thickness for (a) Detector bank (b) BF3 detector bank using different neutron sources.
𝜋𝑎𝑏 1∕2 𝑎𝑏4 ) 𝑒 𝑎 4
The constants a, b for 252 Cf isotope are 1.025 and 2.926 respectively. A combinational geometry was designed in FLUKA for the detection set up. A total 5 cycles of 106 histories were run for each configuration. LOW-Mat card was used for hydrogen in CH2 bonds, 3 He and 10 B and 11 B in BF , which sets the correspondence between FLUKA materials 3 and low-energy neutron cross sections. For simulations, we have considered two geometries of the surroundings; a closed room and completely open-space. The latter is a hypothetical case that has been simulated for only one detector configuration (3 He single detector) to realize the build-up in neutron population due to wall reflections that contribute to the absolute efficiency of a detector. In closed-room (a real situation) geometry, the detector has been placed in a room of size 6 × 8 × 6 m3 as shown in Fig. 1. The thickness of walls, ceiling, as well as the floor each is considered as 20 cm while for the open-space geometry, the detector has been placed in an assumed open space (no walls, ceiling or floor). The density and composition of concrete, polyethylene and perspex was taken from standard FLUKA material library.
3 He
4. Experimental setup The active length of the detector is taken as 920 mm with 38 mm diameter. The BF3 detector is 5 cm in diameter and one-metre long with active length of 920 mm and gas pressure of 0.73 bar. The Moderator under consideration was placed in front of the detector at touching position. The RESINUCLEi card, which simulates the number of residual nuclei formed by interaction of neutrons in the 3 He/BF3 gas, has been used to estimate the efficiency of the detector. In the experimental scenario both the residual nuclei further ionize the detector gas and produce a voltage pulse, which is amplified and discriminated against electronic noise and background gamma pulses, however in simulations the number of residual nuclei have been considered for calculation of absolute detector efficiency (counts/source-neutron or c/n). The absolute efficiency of the detector was calculated for moderator thicknesses varying from 0 to 200 mm in steps of 20 mm. As mentioned earlier, 3 He and BF3 detector banks have also been simulated (in addition to single detectors). The 3 He detector bank consists of six proportional counters (described above) connected in
Simulations have been validated against experimental results for the cases of detectors within walls (a realistic physical situation similar to closed-room geometry described above) using a Pu-Be source. A Pu-Be source of intensity 1 ± 0.05 × 105 neutron/s was placed at 100 cm from the surface of the detector to minimize the effect of source size on the detection efficiency. The counting was carried out for 10, 50 and 50 s for detector bank, single detector and background respectively. The net counts were corrected for background in every case. 5. Results and discussions The response of above-mentioned sources in single as well as in the form of banks of 3 He and BF3 detectors has been numerically simulated for varying thickness of moderator. The simulation geometry of the detection system is shown in Fig. 1. The output of the RESINUCLEi card at optimum thickness of perspex and polyethylene in single detectors is shown in Fig. 2. As described in later segment of this para the optimum 4
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Fig. 5. Comparison of experimental and Monte-Carlo simulation results with perspex and Pu-Be neutron source for (a) 3 He detector (b) 3 He detector bank (c) BF3 detector (d) BF3 detector bank. Table 2 Ratio of efficiencies for three sources with same moderator thickness (optimized for Configuration
3
He 3 He BF3 3 He BF3
detector (open-space) single detector single detector detector bank detector bank
Perspex
252
Cf). Polyethylene
Thickness mm
Efficiency (Pu-Be:D-D:
80 80 80 60 40
1:1.15:1.42 1:1.15:1.42 1:1.16:1.41 1:1.15:1.32 1:1.16:1.43
thickness of perspex and polyethylene for the three sources (252 Cf, PuBe and D-D) is 100, 100, 80 mm and 80, 80, 60 mm respectively. The residual nuclei (3 H and 1 H in 3 He; and 7 Li and 4 He in BF3 ) formed due to neutron interactions indicate the absolute efficiency (reactions per source-neutron) for the given detector configuration. It can be seen (Fig. 2) that the polyethylene provides better moderation with smaller thickness as compared to perspex. The absolute efficiency of single 3 He and BF3 detectors with varying perspex and polyethylene thickness for each of the three types of sources is shown in Fig. 3. It was observed that as the moderator thickness is increased; the detection efficiency reaches to a certain optimum value and then decreases. The increase in the efficiency is quite expected due to enhanced thermalization of source neutrons by the moderator. Further increase in moderator thickness beyond a certain value causes absorption of the thermalized neutrons by the hydrogenous media due to its finite absorption cross section in hydrogen [20], thereby leading to a lowering in the efficiency of detection. Simulations of the single 3 He detector with perspex moderator have been carried out for the setup located in open-space and for a closed-room, as mentioned above (Section 3). The maximum efficiency for a single detector, with the Pu-Be source, in open-space and in a closed-room are observed to be 1.20×10−4 and 2.94×10−4 respectively. Significant enhancement (2.4 times) in the detector efficiency is observed for the closed-room configuration compared to that in openspace. This is essentially due to the reflection of neutrons from the
252
Cf)
Thickness mm
Efficiency (Pu-Be:D-D:
60 60
1:1.18:1.44 1:1.16:1.34
252
Cf)
concrete walls. Simulations have also been carried out to investigate the effect of individual surfaces of closed-room on the detector efficiency for a single 3 He detector with Pu-Be neutron source. The floor (being the closest) and ceiling provide an enhancement in the efficiency of 33% and 4% as compared to open-space case. In comparison to the case of the geometry consisting detector, floor, and ceiling, the geometry consisting surface behind the detector and surface behind the source enhances the efficiency to 32% and 5% respectively, however, both the side walls provide an enhancement of only 2%. The combined effect of all surfaces provides 2.4 times enhancement in detector efficiency as compared to set-up in the open-space. The effect of moderation material has been studied with perspex and polyethylene in both types of single detectors. The optimum moderator thickness of Pu-Be:D-D: 252 Cf sources for perspex and polyethylene were found out to be 100, 100, 80 mm and 80, 80, 60 mm respectively. The difference in efficiency with these two moderators is expected due to hydrogen content; however, the efficiency at optimum moderator thickness seems to be almost the same for both. The ratio of efficiency of the 3 He detector in open space with Pu-Be:D-D:252 Cf source was 1:1.15:1.31 at 100,100,80 mm perspex moderator thickness, while the same for the detector configured within the walls, the ratio of efficiency comes out to be 1:1.13:1.35. The slight difference in the ratio of these two could be due to statistical error (±1%–2%) in the simulations. For a realistic approach, all other simulations were carried out within the walls. In BF3 detectors the 5
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
Fig. 6. Fluence of neutron after the moderator while source is at 100 cm from the surface of detector (a) Fluence after moderator with varying thickness with D-D source (b) The neutron fluence after 60 mm moderator thickness. (Inset: fluence of thermal neutrons).
efficiency ratio at optimum moderator thickness for Pu-Be:D-D: 252 Cf source is 1:1.14:1.34, while with polyethylene moderator the ratio is slightly different (1:1.19:1.34). Comparative data of the detection efficiency for all the three sources at optimum moderator thickness is listed in Table 1. The average efficiency ratio for these sources in 3 He and BF3 detector is 1 : 1.13±0.006 : 1.35±0.007 and 1 : 1.16±0.02 : 1.34±0.008 for perspex and polyethylene respectively. Fig. 4 illustrates the variation of detection efficiency with moderator thickness. Compared to the two other sources, the highest efficiency was observed for the 252 Cf source that could be due to the relatively softer energy spectrum of 252 Cf. The counts with D-D sources are found to be relatively higher in large moderation thickness than in case of 252 Cf and Pu-Be sources. Due to the continuous energy spectrum for Pu-Be and 252 Cf, excess moderation not only thermalizes the fast neutrons but also absorbs the lower energy ones. The results of simulation of the aforementioned detectors for the PuBe source is validated through experimental measurements as shown in Fig. 5. Due to electronic processing, (described in Section 3), the low height neutron pulses could be lost in pulse discrimination, that reduces the registered count rate, so the experimental values are generally found to be less than simulated results. It was observed that experimental values of efficiency are higher than the simulated ones for the case of detector banks (Fig. 5(b) and (d)). This may be attributed to reflection of neutrons from various other materials present in the laboratory, but in case of single detectors, the effect of reflection from the surrounding laboratory material has not been observed due to the small surface
area of the detector. In BF3 detectors, the pulse discrimination is better owing to its higher full energy peak (764 keV for 3 He against 2.3 MeV for BF3 ), hence the effect of reflection appears to be higher in the BF3 as compared to 3 He detector banks. The trends of variation in efficiency with moderator thickness are seen in agreement with each other. The optimum moderation thickness for single detector and the bank are also found to be different. This can be attributed to additional moderation provided by the grooved perspex sheets inside detector banks. To check the difference in counts due to grooved perspex we have also simulated the detection efficiency of both the detector banks without grooved perspex. It is seen that embedding by perspex sheets provides additional moderation to the source neutrons that results up to a 25% enhancement in the detection efficiency in the 3 He Detector bank. In the BF3 detector bank enhancement in detection efficiency is more than factor of two. The enhancement is maximum for 252 Cf (23% in 3 He and 150% in BF3 ) source compared to the other two (5 and 8% in 3 He and 106 and 109% in BF3 for Pu-Be and D-D source respectively). The ratio of efficiencies for the three sources have been observed to be 1:1.12:1.25 and 1:1.15:1.35 for the 3 He and BF3 detector bank respectively. The enhancement in the case of BF3 is expected because of the larger width of available moderator surrounding the detector. The average neutron fluence without moderator at the detector location is 1.3±0.06×10−5 n/cm2 . The fluence after the moderator near its optimum thickness for D-D fusion neutron source is plotted in Fig. 6. The fluence can be plotted in /cm2 /energy-bin/source-neutron, 6
S. Andola, A.M. Rawool and T.C. Kaushik
Nuclear Inst. and Methods in Physics Research, A 944 (2019) 162580
which is always less than 1. The energy bins in the FLUKA library are non-uniform, so the fluence has been plotted here in /cm2 /MeV/sourceneutron, which is higher than 1 for lower energy range. It is clearly seen that the neutron fluence is increasing in the lower energy region up to a certain moderator thickness. Beyond 120 mm of moderator thickness, the thermal fluence starts decreasing; leading to lowering of detection efficiency. The fluence after 60 mm Perspex moderator clearly shows that neutron flux after the moderator is higher for 252 Cf source compared to the other two sources that impart to higher sensitivity. The Ratio of detection efficiency of the detectors for the same thickness (optimized for 252 Cf) of polyethylene moderator for all the three sources is shown in Table 2. For such a case, the ratio of detection efficiency for these sources can be taken as 1 : 1.16±0.02 : 1.4±0.03. Similarly, the optimum moderation thickness plays a crucial role in the correct determination of yield as seen with three types of sources (PuBe, D-D and 252 Cf), which is 100, 100 and 80 mm for the perspex and 80, 80, 60 mm for polyethylene. It suggests that for the same detector, if the moderator thickness is optimized for calibrating source, there is a difference in the detection efficiencies while monitoring fusion device due to the energy spectrum that needs to be accounted. This also implies that the calibration factor for the type of source needs to be correctly taken into account for the yield measurement in either type of source.
observed that for the measurements of fusion neutrons, calibration of detectors using radio-isotopic sources could yield an error even with the optimum thickness of moderator which could be up to 40% using pure fission sources. Acknowledgement It is a pleasure to acknowledge technical support by our colleague Mr A V Patil during the experiments in this work. References [1] D. Klir, J. Kravarik, P. Kubes, K. Rezac, S.S. Ananev, Y.L. Bakshaev, P.I. Blinov, A.S. Chernenko, E.D. Kazakov, V.D. Korolev, Phys. Plasmas 15 (2008) 032701. [2] J.W. Mather, Phys. Fluids 8 (1965) 366. [3] C.L. Ruiz, G.W. Cooper, S.A. Slutz, J.E. Bailey, G.A. Chandler, T.J. Nash, T.A. Mehlhorn, R.J. Leeper, D. Fehl, A.J. Nelson, Phys. Rev. Lett. 93 (2004) 015001. [4] G.H.V. Bertrand, M. Hamel, S.p. Normand, F. Sguerra, Nucl. Instrum. Methods Phys. Res. A 776 (2015) 114. [5] L.V. East, R.B. Walton, Nucl. Instrum. Methods 72 (1969) 161. [6] E.J. Axton, P. Cross, J.C. Robertson, J. Nucl. Energy Parts A/B. React. Sci. Technol. 19 (1965) 409. [7] C.W. Barnes, A.R. Larson, G. LeMunyan, M.J. Loughlin, Rev. Sci. Instrum. 66 (1995) 888. [8] G.W. Cooper, C.L. Ruiz, Rev. Sci. Instrum. 72 (2001) 814. [9] L.B. Rees, J.B. Czirr, Nucl. Instrum. Methods Phys. Res. A 691 (2012) 72. [10] D.H. Beddingfield, N.H. Johnson, H.O. Menlove, Nucl. Instrum. Methods Phys. Res. A 455 (2000) 670. [11] Z.M. Zeng, H. Gong, Q. Yue, J.M. Li, Nucl. Instrum. Methods Phys. Res. A 866 (2017) 242. [12] A. Lintereur, K. Conlin, J. Ely, L. Erikson, R. Kouzes, E. Siciliano, D. Stromswold, M. Woodring, Nucl. Instrum. Methods Phys. Res. A 652 (2011) 347. [13] S. Pszona, Nucl. Instrum. Methods Phys. Res. A 402 (1998) 139. [14] A. Ferrari, P.R. Sala, A. Fasso, J. Ranft, FLUKA: A multi-particle transport code (Program version 2005), 2005. [15] R.T. Kouzes, J.H. Ely, L.E. Erikson, W.J. Kernan, A.T. Lintereur, E.R. Siciliano, D.L. Stephens, D.C. Stromswold, R.M. Van Ginhoven, M.L. Woodring, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 623, 1035. [16] T.C. Kaushik, L.V. Kulkarni, R.K. Rout, A.M. Rawool, S.C. Gupta, A.M. Shaikh, Neutron detectors for wide range monitoring in pulsed electromagnetic environment, in: Proceedings of DAE-BRNS National Symposium on Compact Nuclear Instruments and Radiation Detector-2005, 2005. [17] M.E. Anderson, R.A. Neff, Nucl. Instrum. Methods 99 (1972) 231. [18] C.J. Everett, E.D. Cashwell, A third Monte Carlo sampler, Los Alamos Report LA-9721-MS, 1983. [19] MCNP-A General Monte Carlo N-Particle Transport Code, Version 5, Los Alamos National Laboratory Los Alamos, NM, 2003. [20] B. Hamermesh, G.R. Ringo, S. Wexler, Phys. Rev. 90 (1953) 603.
6. Summary and conclusions In the present work, a comparative study has been carried out to investigate the effect of source energy term on moderator thickness for three different types of sources including mono energetic (fusion) and broad energy spectrum (radio-isotopic and fission). For fusion sources of unknown yields, it is necessary to calibrate the detectors using standard fission/radio-isotopic sources such as Pu-Be, 252 Cf or Am-Be etc. Monte Carlo simulations have been carried out to optimize the moderator thickness in commonly used proportional counters i.e. 3 He and BF3 gas filled neutron detectors. The simulation results have been validated through experiments using a Pu-Be source. The optimum moderator thicknesses are in agreement with the simulation, although the efficiencies are found to be higher in experiments possibly due to reflection of neutrons from various materials in the laboratory. The results are significant and useful for applications where neutron yields from sources of unknown strength need to be characterized based on calibration of detectors using standard fission sources. Moreover, it is
7