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MCNP optimization of fast neutron beam thermalization device based on D-T neutron generator
Fusion Engineering and Design 151 (2020) 111385
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MCNP optimization of fast neutron beam thermalization device based on DT neutron generator
T
Cong Li, Shiwei Jing*, Yadong Gao, Wenjuan Zhang School of physics, Northeast Normal University, Changchun, 130024, Jilin, PR China
A R T I C LE I N FO
A B S T R A C T
Keywords: D-T neutron generator Thermalization MCNP5 Multiplier Collimator
The 14 MeV neutrons generated by Deuterium-Tritium (D-T) fusion reaction usually need to be moderated for Prompt Gamma Neutron Activation Analysis (PGNAA) applications. A fast neutron beam thermalization device was developed in Northeast Normal University based on a D-T neutron generator (model NG-9). A simple experimental system was set up firstly to validate the feasibility of simulation. The materials and dimensions of moderator and the size of multiplier, reflector and collimator were then optimized through MCNP5 to improve the efficiency of the thermalization device. The thermal, epithermal, fast and total neutron flux across the output surface in final system were increased for a factor of 9.65, 8.70, 5.84 and 6.94 times compared to the initial experimental system, respectively. And the optimized thermalization efficiency is 13.28 times higher than that of the original model of experimental system.
1. Introduction Prompt Gamma Neutron Activation Analysis (PGNAA) is a fast, online, precise, non-destructive method capable of entire elemental analysis. It is widely applied in many fields such as industries [1], therapy [2], environmental [3], security [4], radiography [5] and biophysical [6]. The prompt gamma rays can be induced by (n, n＇γ) and (nth, γ) interactions. The element type and content can be determined by analyzing the energy and the characteristic peak area of the prompt gamma rays [7]. The intensity of characteristic gamma rays emitted from (nth, γ) interaction will be significantly improved by increasing the thermal neutron flux of PGNAA device. Currently, the neutron sources used in PGNAA applications mainly include Americium-Beryllium (Am-Be) [8], 252Cf [9], Deuterium-Deuterium (D-D) and D-T [10,11]. D-T neutron generator has many advantages and is widely used for PGNAA applications. It is convenient to be turned off which are inherently safer for operators. The neutrons from D-T reaction can penetrate bulk samples and the materials composition can be identified through the characteristic gamma rays produced by both (n, n＇γ) and (nth, γ) interactions. In addition, the detection time can be decreased and the neutron spread will be increased because of high neutron flux. The environmental background will be reduced due to the source itself doesn’t emit gamma rays. In practical manufacturing process, there are some insulating materials around the D-T neutron tube such as Kapton to isolate tens of kV high voltage of
⁎
accelerator from the surroundings. However, the D-T neutron tube is usually treated as a 14 MeV point source in Monte Carlo simulation without considering the moderation effect of the insulating materials. The neutron energy emitting from the neutron generator ranges from 14 MeV mono-chromaticity to a broad band, which is a mixed neutron spectra of thermal, epithermal and fast neutrons. Recently, a simple thermal neutron-capture based PGNAA experimental device was designed and tested at Northeast Normal University. We subsequently optimized this thermalization device by changing the moderator, multiplier, reflector and collimator using MCNP5 to find the most suitable geometrical size, which will increase the thermal neutron flux and eventually improve the performance of PGNAA device. In this article, the thermal neutron flux of the thermalization device’s output surface was optimized. The optimized objects included three aspects: (1) to increase the flux of thermal neutrons, (2) to increase the total neutron flux and (3) to increase the thermalization efficiency compared with the initial experimental system. 2. PGNAA experiment and verification A basic experiment was carried out at the Institute of Radiation Technology in Northeast Normal University to verify the agreement between simulation and experiment (as shown in Fig. 1(a)). A D-T neutron generator, a gamma-ray spectrometer with bismuth germanium oxide (BGO) scintillators, and paraffin (8 × 25 × 30 cm) as
Corresponding author at: Northeast Normal University, No. 5268, Renmin Street, Nanguan District, Changchun, Jilin, PR China. E-mail address: [email protected] (S. Jing).
https://doi.org/10.1016/j.fusengdes.2019.111385 Received 11 June 2019; Received in revised form 24 October 2019; Accepted 27 October 2019 0920-3796/ © 2019 Elsevier B.V. All rights reserved.
Fusion Engineering and Design 151 (2020) 111385
C. Li, et al.
2.2. Experiment and verification According to the real materials and geometry of the thermalization device as shown in Fig. 1, MCNP5 is edited to describe the experiment device. The neutron generator is usually treated as a point source with monochromatic energy 14 MeV at the target in Monte Carlo simulation. In our work, the neutron source was simulated as a 4π- direction point source located at the target which is 10 cm distance from the surface of neutron tube (as shown in Fig. 4). The neutron tube is bounded by 1.7 cm thick Kapton and 0.1 cm thick Kovar alloy layer, respectively. The Kapton has a density of 1.40 g/cm3, and weight fraction of 11.57% O, 5.42% H, 77.96% C and 5.03% N. It is a kind of polymer containing imide ring in the main chain, which has high thermal stability and heat resistance. It is used to isolate the high-voltage from the surroundings in the neutron emitter. The Kovar alloy has a density of 8.30 g/cm3, and weight fraction of 53.70% Fe, 17.30% Co and 29.00% Ni. The three cylinder layers construct the actual neutron emitter. After passing through the Kapton and Kovar alloys, 21.1% of the neutron energy ranges from 0 to 13 MeV, and 78.9% of the neutron energy ranges from 13 to 14 MeV. This shows that the moderation effect of Kapton can’t be neglected. Subsequently, the input file is read by MCNP code and the results per source particle are listed in the output file [13]. The gamma spectra from simulations and experiments for different thickness of paraffin were compared to verify the agreement of MCNP5 simulation and experiment. The experimental gamma spectra were obtained by a multichannel analyzer (http://www.risehood.com/) with 60 s measuring time. F8 tally was used in simulation to obtain the photon pulse height spectra in the BGO detector with the corresponding E tally to divide the energy range between 0 and 14 MeV into 100 channels. And the Gaussian Energy Broadening function (GEB) was used to simulate the BGO responses. The gamma-ray cross sections are from the ENDF/B-VI library, and then the gamma spectra of experiment and simulation were analyzed by Origin software [14]. The characteristic energy peaks can be determined using the “find peaks” function in Origin. Subsequently, all counts in the channel scope were added together with background subtracted. As seen in Fig. 2, the obviously peak at 2.23 MeV is from gamma rays produced by the neutron capture reactions with hydrogen in paraffin. And the peak at 2.62, 4.44 and 6.13 MeV is generated by neutron inelastic scattering with lead, carbon and oxygen, respectively. The trend of gamma spectra between experiment and simulation shows apparently agreement. When the thickness of paraffin is 8 cm, the gamma counts of hydrogen peak reaches the maximum. For further verify the feasibility of simulation, the hydrogen peak counts were compared between simulation and experiment. The thermal neutron flux through the paraffin under different thickness of 4 cm, 8 cm, 12 cm and 16 cm were simulated with F2 card, which reflects an average flux in the measuring surface. Due to the difficulty in describing the accurate elemental composition and distributions of the background in the experiments and the differences between simulated and actual neutron tubes, the results between the simulations and the experiments can't be compared directly. Therefore, MCNP5 simulations are normalized to experiments by the ratio of experimental to simulation’s hydrogen peak areas [3]. After normalization, the characteristic peaks of hydrogen rise with increasing thermal neutron flux to some extent (as seen in Fig. 3). The shape of gamma spectra and the trend of hydrogen peak counts of experiment are in good agreement with the simulation in MCNP5, which verify the feasibility of simulation. The geometrical size was subsequently optimized to produce maximum thermal neutron flux based on the experimental device. The neutron flux across the output surface (as shown in Fig. 4) was simulated using F2 tally (surface average flux card). The energy ranges 0–10−5, 10−5–10−1 and 10−1–14 MeV represent thermal, epithermal and fast neutron, respectively. The particle histories are 2.65 × 107 as mentioned earlier, and the relative errors in all tallies are less than 2%. To assess the performance of the thermalization setup, two criteria of P
Fig. 1. Experimental and schematic diagram of the initial thermalization device.
neutron moderator were the main components of the thermalization device. The BGO detector (3 in. × 3 in.) was surrounded by a layer of lead with 3 cm thickness in order to prevent background gamma rays entering the detector. The paraffin was set as a sample to obtain the photon spectrum as well. Two identical lead blocks (8 × 18 × 30 cm) were set on each side of the neutron emitter along the X axis, as shown in Fig. 1(b). The one between the paraffin and neutron emitter is served as multiplier and the other is served as reflector. The multiplier lead will increase the neutron flux through (n, 2n) reaction [12]. The reflector lead will reflect back the neutrons in the direction towards the detector. The paraffin was placed between the lead multiplier and BGO detector to slow down neutrons. In addition, a 3He neutron detector with 2.5 cm in diameter and 14.5 cm length was placed above the BGO detector to monitor the neutron yield.
2.1. NG-9 neutron generator A model NG-9 D-T neutron generator manufactured by Northeast Normal University was used in the thermalization device. It is composed of a neutron emitter, a control box and correspondingly electronics system. The overall neutron emitter is a cylinder with 4.3 cm in radius and 89 cm length. A Penning ion source, an accelerator electrode, a pressure adjustment system and a target are enclosed within a neutron tube. The neutron tube is a cylinder with 2.5 cm in radius and 10 cm length. The Penning iron source in the neutron tube is a low gas pressure, cold cathode ion source with crossed electric and magnetic fields and it is used to generate deuterium and tritium ions. Both the ion beam and target contain 50% deuterium and 50% tritium. The ionized ions will pass through the accelerating electrode and strike a titanium target coated with aluminum oxide, where occurred the DeuteriumTritium reactions producing 14 MeV neutrons and 3.5 MeV α particles. The control console allows the operator to adjust the operating parameters of the neutron tube. In our work, the neutron source was operated in continuous mode with 75 kV acceleration voltage and 102 μA ion source current, producing about 2.65 × 107 neutrons per second. And this intensity will be used in MCNP simulation.
2
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squared ψth to ψ0 will ensure a higher possible value of the thermal neutron flux. And the criteria K represents the neutron increment at the output surface of the thermalization device. 3. Optimization The P value of 2.97 × 10−6 cm−2 s−1 can be obtained by the simulation of the initial experimental system (as shown in Fig. 1(a)). The simulation results showed in Fig. 3 illustrate that thermal neutron flux has great influence on the characteristic gamma-rays counts of thermal neutron capture reaction. The intensity of thermal neutrons determines the efficiency of PGNAA thermalization device, and the 14 MeV fast neutrons need to be moderated before interacting with the sample [9]. Neutrons emitted from target will be moderated and the total neutron flux will be decreased after passing through the Kapton and Kovar alloys. A proper multiplier was used subsequently to increase the number of neutrons. The neutrons still need to be moderated due to the energy of the neutrons from the multiplier is still too high for PGNAA. However, the thermal neutron flux will decrease after passing through the beam shaping assembly in the device. The thermal neutron flux can be increased by backward reflection of a fraction of neutrons through selecting lead as reflector around the neutron emitter. In addition, the neutrons need to be guided to the output surface with a collimator to improve the efficiency of the device. In our work, we aimed to maximize the thermalization efficiency of the device through optimizing materials and geometric configurations around the neutron emitter for PGNAA experiments. The optimization model is shown in Fig. 4 using the VISED software [16]. In order to improve the P value of this thermalization device, system characteristics (L1: moderator thickness, L2: moderator width, L3: multiplier thickness, L4: reflector width, L5: reflector height, L6: collimator height) were optimized to increase the thermal neutron flux of the output surface.
Fig. 2. Gamma spectra: (a) experiment; (b) simulation when the thickness of paraffin is 4, 8, 12, 16 cm, respectively.
3.1. Moderator In a thermal neutron-capture-based PGNAA thermalization device, the moderator is generally used to slow down the fast neutrons. The thermal neutron flux after passing through the moderator is determined by the incident neutron energy and the moderator parameters such as material and geometry [17]. The material of moderator should have the characteristic of scattering neutrons, in other words, it is capable of reducing the energy of fast neutron. Meanwhile, it shouldn’t have high absorption cross section avoiding losing neutrons. Four materials including High Density Polyethylene (HDPE) with a density of 0.95 g/ cm3, paraffin (ρ = 0.93 g/cm3), graphite (ρ = 2.25 g/cm3) and heavy water (D2O) with a density of 1.11 g/cm3 [9,12,18] were selected as moderator. The moderator was defined as a truncated cone with fixed radius of the upper surface (5 cm) and the lower surface (10 cm) situated above the neutron emitter, as shown in Fig. 4. The material and optimization thickness L1 of moderator were simulated. The P values rise with the increase of the thickness of HDPE, paraffin, and D2O at first (as shown in Fig. 5). In this case, the significance of neutron elastic scattering is stronger than neutron absorption. The P values begin to decrease after reaching the maximum because of the process of thermal neutron absorption is stronger than neutron elastic scattering. For graphite, the P values nearly keep a constant as the thickness increase indicating that the hydrogen existing in HDPE, paraffin, and D2O plays an important role in moderating neutrons. When the thickness of HDPE is 8 cm, the P value reaches the maximum. Subsequently, the width of moderator was determined by P value. Different lower surface radius (L2/2) of truncated cone moderator were simulated with the radius of upper surface (5 cm) fixed (as shown in Fig. 4). The relationship between P values and the moderator width for HDPE is shown in Fig. 6. By increasing the width of moderator, the P value firstly increases until it reaches to a peak value and it keeps the
Fig. 3. The relationship between thermal neutron flux and hydrogen peak count.
and K representing thermalization efficiency and the increasing rate of flux, respectively, are specified by
P = ψth2 ψT
K=
ψT − ψ0 ψ0
(1)
(2)
where ψth and ψT represent the thermal neutron and total neutron flux after optimization, respectively [15]. ψ0 is the total neutron flux of the initial experimental PGNAA thermalization device. The criteria P represents the ability of the device to moderate fast neutrons. The ratio of 3
Fusion Engineering and Design 151 (2020) 111385
C. Li, et al.
Fig. 4. A schematic diagram of an optimization thermalization device, not to scale.
scattering because of the high threshold. In addition, the neutron capture cross section is approximately inversely proportional to the square root of the neutron energy in the low energy region. Therefore, the nuclear reaction rate will improve with the decreasing of neutron energy. The fast neutron can be moderated and the neutron flux will be increased by setting up multiplier between the neutron emitter and the moderator. And then materials containing hydrogen will be used to moderate neutrons through elastic scattering. A layer of lead (ρ = 11.34 g/cm3) was chosen as multiplier in our device. The lead layer thickness L3 varied from 0 cm to 8 cm, 0.5 cm by step. As shown in Fig. 6, the criterion P increases with the increasing thickness firstly achieving maximum when the multiplier thickness L3 is 4 cm. When the lead thickness is over 4 cm, P values begin to decrease for the process of neutron absorption is greater than multiplication. Therefore, the optimized thickness of multiplied lead is determined as 4 cm with P = 1.41 × 10−4 cm−2 s−1.
Fig. 5. Criterion of P as a function of the moderator material and thickness L1.
3.3. Reflector The suitable material for reflector should reenter more neutrons to the device. Lead was selected as the reflector [12,17,19,20] owing to its high elastic scattering cross section and low absorption cross section. Meanwhile, the thermal neutron flux will obviously increase because of backward refection of neutrons. The width L4 and height L5 of reflector were simulated to obtain the maximum P value. The results are shown in Fig. 7. When the width L4 and height L5 are over 20 cm and 25 cm, the values of P keep constant illustrating that thermal neutrons cannot transmit to a farther distance. These results indicate that the width of 20 cm and the height of 25 cm for reflector are large enough for the setup. Fig. 6. Criterion of P as a function of the moderator width L2 and the multiplier thickness L3, respectively.
horizontal later. According to the maximum value of the criterion P = 1.40 × 10−6 cm−2 s−1 (as shown in Fig. 6), the optimized moderator of the thermalization device is determined as HDPE with the thickness L1 = 8 cm and the width L2 = 20 cm. 3.2. Multiplier As is known that when the neutron energy is above 7 MeV, the cross section of the 208Pb (n, 2n) 207Pb reaction will increase, which is capable of compensating neutron losses [12,15]. The inelastic scattering reaction cross section of medium and heavy materials is larger than light elements such as hydrogen for fast neutrons. Setting the high density metal can effectively slow down fast neutrons through inelastic
Fig. 7. The relationship between the criterion P and reflector width L4, height L5 and the collimator height L6, respectively. 4
Fusion Engineering and Design 151 (2020) 111385
C. Li, et al.
Table 1 P-values and neutron flux comparisons after optimization (unit: cm−2 s−1). thermal Initial system multiplied reflected collimated
epithermal −5
1.38 × 10 1.41 × 10−4 1.46 × 10−4 1.47 × 10−4
−6
6.07 × 10 5.80 × 10−5 5.94 × 10−5 5.89 × 10−5
fast
total −5
4.43 × 10 3.07 × 10−4 3.10 × 10−4 3.03 × 10−4
P −5
6.41 × 10 5.06 × 10−4 5.15 × 10−4 5.09 × 10−4
3.4. Collimator
for their financial support.
The direction of neutron towards the output surface is diffuse, which will decrease the neutron flux reaching to the sample. Nickel [15] (ρ = 8.90 g/cm3) was selected as collimator to return escaping neutrons back to the device. Because of its high elastic scattering cross sections and low absorption cross sections, the thermal neutron flux can be increased by using Ni as collimator. Different height of collimator L6 was simulated. As shown in Fig. 7, the maximum value of the criterion P = 4.24 × 10−5 cm−2 s−1 is achieved when the height is 3.5 cm. When the height is over 3.5 cm, P values begin to decrease duo to the process of thermal neutron absorption is stronger than neutron elastic scattering. When the height is over 5 cm, the P value tends to increase, probably due to the decrease of total neutron flux caused by the replacement of part of lead by Ni. As is shown in Table 1, both the thermal neutron flux and the total neutron flux of the final optimized system (after collimated) are higher than that of the initial experimental system. Meanwhile, the thermalization efficiency P is increased, which will improve the performance of PGNAA device.
References
K −6
2.97 × 10 3.91 × 10−5 4.12 × 10−5 4.24 × 10−5
– 6.88 7.03 6.94
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4. Conclusions A simple thermalization device was developed with a NG-9 neutron generator and a 3 in. × 3 in. BGO detector for PGNAA. The feasibility of the MCNP5 model is verified by comparing the results of experiment with simulation. Take the actual structure of neutron emitter into consideration, the device was optimized based on the initial device to obtain the maximum thermalization efficiency. The main process of optimization includes: (1) optimization of the moderator material, thickness L1 and width L2, (2) optimization of the multiplier thickness L3, (3) optimization of the reflector width L4 and height L5, (4) optimization of the collimator height L6. The final thermalization device was characterized by the following dimensions: L1 = 8 cm; L2 = 20 cm; L3 = 4 cm; L4 = 20 cm; L5 = 25 cm; L6 = 3.5 cm, respectively. The neutron flux going across the output surface in the final system are increased for a factor of 9.65, 8.70, 5.84 and 6.94 times for thermal, epithermal, fast and total neutrons compared with the initial thermalization device, respectively. And the criterion P = 4.24 × 10−5 cm−2 s−1 is about 13.28 times higher than that of the original experimental system for the geometry shown in Fig. 1, which will eventually improve the performance of a thermal neutron-capture based PGNAA device. Declaration of Competing Interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper. Acknowledgements The authors wish to acknowledge Department of Science and Technology of Jilin Province, China (20190303101SF), Education Department Project of Sichuan Province, China (16ZA0325) and Key Laboratory of Criminal Prosecutions in Sichuan Universities of China 5
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MCNP optimization of fast neutron beam thermalization device based on DT neutron generator
T
Cong Li, Shiwei Jing*, Yadong Gao, Wenjuan Zhang School of physics, Northeast Normal University, Changchun, 130024, Jilin, PR China
A R T I C LE I N FO
A B S T R A C T
Keywords: D-T neutron generator Thermalization MCNP5 Multiplier Collimator
The 14 MeV neutrons generated by Deuterium-Tritium (D-T) fusion reaction usually need to be moderated for Prompt Gamma Neutron Activation Analysis (PGNAA) applications. A fast neutron beam thermalization device was developed in Northeast Normal University based on a D-T neutron generator (model NG-9). A simple experimental system was set up firstly to validate the feasibility of simulation. The materials and dimensions of moderator and the size of multiplier, reflector and collimator were then optimized through MCNP5 to improve the efficiency of the thermalization device. The thermal, epithermal, fast and total neutron flux across the output surface in final system were increased for a factor of 9.65, 8.70, 5.84 and 6.94 times compared to the initial experimental system, respectively. And the optimized thermalization efficiency is 13.28 times higher than that of the original model of experimental system.
1. Introduction Prompt Gamma Neutron Activation Analysis (PGNAA) is a fast, online, precise, non-destructive method capable of entire elemental analysis. It is widely applied in many fields such as industries [1], therapy [2], environmental [3], security [4], radiography [5] and biophysical [6]. The prompt gamma rays can be induced by (n, n＇γ) and (nth, γ) interactions. The element type and content can be determined by analyzing the energy and the characteristic peak area of the prompt gamma rays [7]. The intensity of characteristic gamma rays emitted from (nth, γ) interaction will be significantly improved by increasing the thermal neutron flux of PGNAA device. Currently, the neutron sources used in PGNAA applications mainly include Americium-Beryllium (Am-Be) [8], 252Cf [9], Deuterium-Deuterium (D-D) and D-T [10,11]. D-T neutron generator has many advantages and is widely used for PGNAA applications. It is convenient to be turned off which are inherently safer for operators. The neutrons from D-T reaction can penetrate bulk samples and the materials composition can be identified through the characteristic gamma rays produced by both (n, n＇γ) and (nth, γ) interactions. In addition, the detection time can be decreased and the neutron spread will be increased because of high neutron flux. The environmental background will be reduced due to the source itself doesn’t emit gamma rays. In practical manufacturing process, there are some insulating materials around the D-T neutron tube such as Kapton to isolate tens of kV high voltage of
⁎
accelerator from the surroundings. However, the D-T neutron tube is usually treated as a 14 MeV point source in Monte Carlo simulation without considering the moderation effect of the insulating materials. The neutron energy emitting from the neutron generator ranges from 14 MeV mono-chromaticity to a broad band, which is a mixed neutron spectra of thermal, epithermal and fast neutrons. Recently, a simple thermal neutron-capture based PGNAA experimental device was designed and tested at Northeast Normal University. We subsequently optimized this thermalization device by changing the moderator, multiplier, reflector and collimator using MCNP5 to find the most suitable geometrical size, which will increase the thermal neutron flux and eventually improve the performance of PGNAA device. In this article, the thermal neutron flux of the thermalization device’s output surface was optimized. The optimized objects included three aspects: (1) to increase the flux of thermal neutrons, (2) to increase the total neutron flux and (3) to increase the thermalization efficiency compared with the initial experimental system. 2. PGNAA experiment and verification A basic experiment was carried out at the Institute of Radiation Technology in Northeast Normal University to verify the agreement between simulation and experiment (as shown in Fig. 1(a)). A D-T neutron generator, a gamma-ray spectrometer with bismuth germanium oxide (BGO) scintillators, and paraffin (8 × 25 × 30 cm) as
Corresponding author at: Northeast Normal University, No. 5268, Renmin Street, Nanguan District, Changchun, Jilin, PR China. E-mail address: [email protected] (S. Jing).
https://doi.org/10.1016/j.fusengdes.2019.111385 Received 11 June 2019; Received in revised form 24 October 2019; Accepted 27 October 2019 0920-3796/ © 2019 Elsevier B.V. All rights reserved.
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2.2. Experiment and verification According to the real materials and geometry of the thermalization device as shown in Fig. 1, MCNP5 is edited to describe the experiment device. The neutron generator is usually treated as a point source with monochromatic energy 14 MeV at the target in Monte Carlo simulation. In our work, the neutron source was simulated as a 4π- direction point source located at the target which is 10 cm distance from the surface of neutron tube (as shown in Fig. 4). The neutron tube is bounded by 1.7 cm thick Kapton and 0.1 cm thick Kovar alloy layer, respectively. The Kapton has a density of 1.40 g/cm3, and weight fraction of 11.57% O, 5.42% H, 77.96% C and 5.03% N. It is a kind of polymer containing imide ring in the main chain, which has high thermal stability and heat resistance. It is used to isolate the high-voltage from the surroundings in the neutron emitter. The Kovar alloy has a density of 8.30 g/cm3, and weight fraction of 53.70% Fe, 17.30% Co and 29.00% Ni. The three cylinder layers construct the actual neutron emitter. After passing through the Kapton and Kovar alloys, 21.1% of the neutron energy ranges from 0 to 13 MeV, and 78.9% of the neutron energy ranges from 13 to 14 MeV. This shows that the moderation effect of Kapton can’t be neglected. Subsequently, the input file is read by MCNP code and the results per source particle are listed in the output file [13]. The gamma spectra from simulations and experiments for different thickness of paraffin were compared to verify the agreement of MCNP5 simulation and experiment. The experimental gamma spectra were obtained by a multichannel analyzer (http://www.risehood.com/) with 60 s measuring time. F8 tally was used in simulation to obtain the photon pulse height spectra in the BGO detector with the corresponding E tally to divide the energy range between 0 and 14 MeV into 100 channels. And the Gaussian Energy Broadening function (GEB) was used to simulate the BGO responses. The gamma-ray cross sections are from the ENDF/B-VI library, and then the gamma spectra of experiment and simulation were analyzed by Origin software [14]. The characteristic energy peaks can be determined using the “find peaks” function in Origin. Subsequently, all counts in the channel scope were added together with background subtracted. As seen in Fig. 2, the obviously peak at 2.23 MeV is from gamma rays produced by the neutron capture reactions with hydrogen in paraffin. And the peak at 2.62, 4.44 and 6.13 MeV is generated by neutron inelastic scattering with lead, carbon and oxygen, respectively. The trend of gamma spectra between experiment and simulation shows apparently agreement. When the thickness of paraffin is 8 cm, the gamma counts of hydrogen peak reaches the maximum. For further verify the feasibility of simulation, the hydrogen peak counts were compared between simulation and experiment. The thermal neutron flux through the paraffin under different thickness of 4 cm, 8 cm, 12 cm and 16 cm were simulated with F2 card, which reflects an average flux in the measuring surface. Due to the difficulty in describing the accurate elemental composition and distributions of the background in the experiments and the differences between simulated and actual neutron tubes, the results between the simulations and the experiments can't be compared directly. Therefore, MCNP5 simulations are normalized to experiments by the ratio of experimental to simulation’s hydrogen peak areas [3]. After normalization, the characteristic peaks of hydrogen rise with increasing thermal neutron flux to some extent (as seen in Fig. 3). The shape of gamma spectra and the trend of hydrogen peak counts of experiment are in good agreement with the simulation in MCNP5, which verify the feasibility of simulation. The geometrical size was subsequently optimized to produce maximum thermal neutron flux based on the experimental device. The neutron flux across the output surface (as shown in Fig. 4) was simulated using F2 tally (surface average flux card). The energy ranges 0–10−5, 10−5–10−1 and 10−1–14 MeV represent thermal, epithermal and fast neutron, respectively. The particle histories are 2.65 × 107 as mentioned earlier, and the relative errors in all tallies are less than 2%. To assess the performance of the thermalization setup, two criteria of P
Fig. 1. Experimental and schematic diagram of the initial thermalization device.
neutron moderator were the main components of the thermalization device. The BGO detector (3 in. × 3 in.) was surrounded by a layer of lead with 3 cm thickness in order to prevent background gamma rays entering the detector. The paraffin was set as a sample to obtain the photon spectrum as well. Two identical lead blocks (8 × 18 × 30 cm) were set on each side of the neutron emitter along the X axis, as shown in Fig. 1(b). The one between the paraffin and neutron emitter is served as multiplier and the other is served as reflector. The multiplier lead will increase the neutron flux through (n, 2n) reaction [12]. The reflector lead will reflect back the neutrons in the direction towards the detector. The paraffin was placed between the lead multiplier and BGO detector to slow down neutrons. In addition, a 3He neutron detector with 2.5 cm in diameter and 14.5 cm length was placed above the BGO detector to monitor the neutron yield.
2.1. NG-9 neutron generator A model NG-9 D-T neutron generator manufactured by Northeast Normal University was used in the thermalization device. It is composed of a neutron emitter, a control box and correspondingly electronics system. The overall neutron emitter is a cylinder with 4.3 cm in radius and 89 cm length. A Penning ion source, an accelerator electrode, a pressure adjustment system and a target are enclosed within a neutron tube. The neutron tube is a cylinder with 2.5 cm in radius and 10 cm length. The Penning iron source in the neutron tube is a low gas pressure, cold cathode ion source with crossed electric and magnetic fields and it is used to generate deuterium and tritium ions. Both the ion beam and target contain 50% deuterium and 50% tritium. The ionized ions will pass through the accelerating electrode and strike a titanium target coated with aluminum oxide, where occurred the DeuteriumTritium reactions producing 14 MeV neutrons and 3.5 MeV α particles. The control console allows the operator to adjust the operating parameters of the neutron tube. In our work, the neutron source was operated in continuous mode with 75 kV acceleration voltage and 102 μA ion source current, producing about 2.65 × 107 neutrons per second. And this intensity will be used in MCNP simulation.
2
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squared ψth to ψ0 will ensure a higher possible value of the thermal neutron flux. And the criteria K represents the neutron increment at the output surface of the thermalization device. 3. Optimization The P value of 2.97 × 10−6 cm−2 s−1 can be obtained by the simulation of the initial experimental system (as shown in Fig. 1(a)). The simulation results showed in Fig. 3 illustrate that thermal neutron flux has great influence on the characteristic gamma-rays counts of thermal neutron capture reaction. The intensity of thermal neutrons determines the efficiency of PGNAA thermalization device, and the 14 MeV fast neutrons need to be moderated before interacting with the sample [9]. Neutrons emitted from target will be moderated and the total neutron flux will be decreased after passing through the Kapton and Kovar alloys. A proper multiplier was used subsequently to increase the number of neutrons. The neutrons still need to be moderated due to the energy of the neutrons from the multiplier is still too high for PGNAA. However, the thermal neutron flux will decrease after passing through the beam shaping assembly in the device. The thermal neutron flux can be increased by backward reflection of a fraction of neutrons through selecting lead as reflector around the neutron emitter. In addition, the neutrons need to be guided to the output surface with a collimator to improve the efficiency of the device. In our work, we aimed to maximize the thermalization efficiency of the device through optimizing materials and geometric configurations around the neutron emitter for PGNAA experiments. The optimization model is shown in Fig. 4 using the VISED software [16]. In order to improve the P value of this thermalization device, system characteristics (L1: moderator thickness, L2: moderator width, L3: multiplier thickness, L4: reflector width, L5: reflector height, L6: collimator height) were optimized to increase the thermal neutron flux of the output surface.
Fig. 2. Gamma spectra: (a) experiment; (b) simulation when the thickness of paraffin is 4, 8, 12, 16 cm, respectively.
3.1. Moderator In a thermal neutron-capture-based PGNAA thermalization device, the moderator is generally used to slow down the fast neutrons. The thermal neutron flux after passing through the moderator is determined by the incident neutron energy and the moderator parameters such as material and geometry [17]. The material of moderator should have the characteristic of scattering neutrons, in other words, it is capable of reducing the energy of fast neutron. Meanwhile, it shouldn’t have high absorption cross section avoiding losing neutrons. Four materials including High Density Polyethylene (HDPE) with a density of 0.95 g/ cm3, paraffin (ρ = 0.93 g/cm3), graphite (ρ = 2.25 g/cm3) and heavy water (D2O) with a density of 1.11 g/cm3 [9,12,18] were selected as moderator. The moderator was defined as a truncated cone with fixed radius of the upper surface (5 cm) and the lower surface (10 cm) situated above the neutron emitter, as shown in Fig. 4. The material and optimization thickness L1 of moderator were simulated. The P values rise with the increase of the thickness of HDPE, paraffin, and D2O at first (as shown in Fig. 5). In this case, the significance of neutron elastic scattering is stronger than neutron absorption. The P values begin to decrease after reaching the maximum because of the process of thermal neutron absorption is stronger than neutron elastic scattering. For graphite, the P values nearly keep a constant as the thickness increase indicating that the hydrogen existing in HDPE, paraffin, and D2O plays an important role in moderating neutrons. When the thickness of HDPE is 8 cm, the P value reaches the maximum. Subsequently, the width of moderator was determined by P value. Different lower surface radius (L2/2) of truncated cone moderator were simulated with the radius of upper surface (5 cm) fixed (as shown in Fig. 4). The relationship between P values and the moderator width for HDPE is shown in Fig. 6. By increasing the width of moderator, the P value firstly increases until it reaches to a peak value and it keeps the
Fig. 3. The relationship between thermal neutron flux and hydrogen peak count.
and K representing thermalization efficiency and the increasing rate of flux, respectively, are specified by
P = ψth2 ψT
K=
ψT − ψ0 ψ0
(1)
(2)
where ψth and ψT represent the thermal neutron and total neutron flux after optimization, respectively [15]. ψ0 is the total neutron flux of the initial experimental PGNAA thermalization device. The criteria P represents the ability of the device to moderate fast neutrons. The ratio of 3
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Fig. 4. A schematic diagram of an optimization thermalization device, not to scale.
scattering because of the high threshold. In addition, the neutron capture cross section is approximately inversely proportional to the square root of the neutron energy in the low energy region. Therefore, the nuclear reaction rate will improve with the decreasing of neutron energy. The fast neutron can be moderated and the neutron flux will be increased by setting up multiplier between the neutron emitter and the moderator. And then materials containing hydrogen will be used to moderate neutrons through elastic scattering. A layer of lead (ρ = 11.34 g/cm3) was chosen as multiplier in our device. The lead layer thickness L3 varied from 0 cm to 8 cm, 0.5 cm by step. As shown in Fig. 6, the criterion P increases with the increasing thickness firstly achieving maximum when the multiplier thickness L3 is 4 cm. When the lead thickness is over 4 cm, P values begin to decrease for the process of neutron absorption is greater than multiplication. Therefore, the optimized thickness of multiplied lead is determined as 4 cm with P = 1.41 × 10−4 cm−2 s−1.
Fig. 5. Criterion of P as a function of the moderator material and thickness L1.
3.3. Reflector The suitable material for reflector should reenter more neutrons to the device. Lead was selected as the reflector [12,17,19,20] owing to its high elastic scattering cross section and low absorption cross section. Meanwhile, the thermal neutron flux will obviously increase because of backward refection of neutrons. The width L4 and height L5 of reflector were simulated to obtain the maximum P value. The results are shown in Fig. 7. When the width L4 and height L5 are over 20 cm and 25 cm, the values of P keep constant illustrating that thermal neutrons cannot transmit to a farther distance. These results indicate that the width of 20 cm and the height of 25 cm for reflector are large enough for the setup. Fig. 6. Criterion of P as a function of the moderator width L2 and the multiplier thickness L3, respectively.
horizontal later. According to the maximum value of the criterion P = 1.40 × 10−6 cm−2 s−1 (as shown in Fig. 6), the optimized moderator of the thermalization device is determined as HDPE with the thickness L1 = 8 cm and the width L2 = 20 cm. 3.2. Multiplier As is known that when the neutron energy is above 7 MeV, the cross section of the 208Pb (n, 2n) 207Pb reaction will increase, which is capable of compensating neutron losses [12,15]. The inelastic scattering reaction cross section of medium and heavy materials is larger than light elements such as hydrogen for fast neutrons. Setting the high density metal can effectively slow down fast neutrons through inelastic
Fig. 7. The relationship between the criterion P and reflector width L4, height L5 and the collimator height L6, respectively. 4
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Table 1 P-values and neutron flux comparisons after optimization (unit: cm−2 s−1). thermal Initial system multiplied reflected collimated
epithermal −5
1.38 × 10 1.41 × 10−4 1.46 × 10−4 1.47 × 10−4
−6
6.07 × 10 5.80 × 10−5 5.94 × 10−5 5.89 × 10−5
fast
total −5
4.43 × 10 3.07 × 10−4 3.10 × 10−4 3.03 × 10−4
P −5
6.41 × 10 5.06 × 10−4 5.15 × 10−4 5.09 × 10−4
3.4. Collimator
for their financial support.
The direction of neutron towards the output surface is diffuse, which will decrease the neutron flux reaching to the sample. Nickel [15] (ρ = 8.90 g/cm3) was selected as collimator to return escaping neutrons back to the device. Because of its high elastic scattering cross sections and low absorption cross sections, the thermal neutron flux can be increased by using Ni as collimator. Different height of collimator L6 was simulated. As shown in Fig. 7, the maximum value of the criterion P = 4.24 × 10−5 cm−2 s−1 is achieved when the height is 3.5 cm. When the height is over 3.5 cm, P values begin to decrease duo to the process of thermal neutron absorption is stronger than neutron elastic scattering. When the height is over 5 cm, the P value tends to increase, probably due to the decrease of total neutron flux caused by the replacement of part of lead by Ni. As is shown in Table 1, both the thermal neutron flux and the total neutron flux of the final optimized system (after collimated) are higher than that of the initial experimental system. Meanwhile, the thermalization efficiency P is increased, which will improve the performance of PGNAA device.
References
K −6
2.97 × 10 3.91 × 10−5 4.12 × 10−5 4.24 × 10−5
– 6.88 7.03 6.94
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4. Conclusions A simple thermalization device was developed with a NG-9 neutron generator and a 3 in. × 3 in. BGO detector for PGNAA. The feasibility of the MCNP5 model is verified by comparing the results of experiment with simulation. Take the actual structure of neutron emitter into consideration, the device was optimized based on the initial device to obtain the maximum thermalization efficiency. The main process of optimization includes: (1) optimization of the moderator material, thickness L1 and width L2, (2) optimization of the multiplier thickness L3, (3) optimization of the reflector width L4 and height L5, (4) optimization of the collimator height L6. The final thermalization device was characterized by the following dimensions: L1 = 8 cm; L2 = 20 cm; L3 = 4 cm; L4 = 20 cm; L5 = 25 cm; L6 = 3.5 cm, respectively. The neutron flux going across the output surface in the final system are increased for a factor of 9.65, 8.70, 5.84 and 6.94 times for thermal, epithermal, fast and total neutrons compared with the initial thermalization device, respectively. And the criterion P = 4.24 × 10−5 cm−2 s−1 is about 13.28 times higher than that of the original experimental system for the geometry shown in Fig. 1, which will eventually improve the performance of a thermal neutron-capture based PGNAA device. Declaration of Competing Interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper. Acknowledgements The authors wish to acknowledge Department of Science and Technology of Jilin Province, China (20190303101SF), Education Department Project of Sichuan Province, China (16ZA0325) and Key Laboratory of Criminal Prosecutions in Sichuan Universities of China 5