Modular next generation fast neutron detector for portal monitoring

Modular next generation fast neutron detector for portal monitoring 1 2 corresp first-author eaboud@physics.tamu.edu eaboud@physics.tamu.edu eaboud@physics.tamu.edu https://orcid.org/0000-0003-1465-151X 1 1 2 3 1 2 1 2 1 1 1 2 Cyclotron Institute, Texas A&M University, College Station, Texas 77843, USA Cyclotron Institute, Texas A&M University, College Station, Texas 77843, USA Department of Physics and Astronomy, Texas A&M University, College Station, Texas 77843, USA Department of Physics and Astronomy, Texas A&M University, College Station, Texas 77843, USA Department of Statistics, Texas A&M University, College Station, Texas 77843, USA Department of Statistics, Texas A&M University, College Station, Texas 77843, USA Nuclear nonproliferation is of critical importance for global security. Dangerous fissile materials including highly enriched uranium and weapons-grade plutonium are especially important to detect. Active interrogation techniques may result in much better sensitivity but are difficult with conventional portal monitors that rely on detecting thermal neutrons. Also, most conventional portal monitoring systems rely on 3 He, which has a finite and continually decreasing supply. By designing a highly-segmented array of organic scintillators, we posit that we can accurately and quickly identify fissile materials, including weapons-grade plutonium and highly enriched uranium, being smuggled. We propose a new design for a fast-neutron detector that overcomes the limitations of the current generation of portal monitors. MCNP6 simulations have been performed in conjunction with the UMPBT statistical model to determine the sensitivity limitations of the proposed detector. Results suggest that the proposed detector may be ∼ 10 times more efficient than current-generation thermal neutron detectors and may be able to positively identify a ∼ 81 mg 235 U source in as little as 192 seconds utilizing active interrogation techniques. Fast Neutron Detection Portal Monitoring Non-Proliferation Introduction 1 Monitoring and halting the movement of unauthorized fissile materials becomes more important every day, specifically the monitoring of special nuclear materials (SNM). SNMs include 239 Pu, which is the primary component of weapons-grade plutonium (WGPu), and 235 U, which is the primary component of highly enriched uranium (HEU). These are radioactive materials and can be identified using γ and neutron detectors. Neutron detection is typically used together with γ -ray monitoring and has an advantage of lower natural background and background induced by radioactive materials used in industry and for medical applications. While WGPu emits significant amount of neutrons and is easier to detect, HEU is particularly challenging because of the low neutron emission rate and due to the fact that low energy γ -rays produced by HEU are relatively easy to shield. Portal monitoring systems often combine γ -ray detectors and neutron detectors. Due to naturally occurring radioactive materials and electronic noise or failures, in radiation portal monitors the false alarm rate for γ -rays is about 1 in 10 2 and for neutrons about 1 in 10 4 [ 1 1 ]. Current neutron portal monitoring techniques rely heavily on thermal neutron detectors, many of which require 3 He. 3 He has applications in many different fields, which has dwindled the increasingly small and finite terrestrial supply [ 2 2 ]. Therefore, 3 He is becoming increasingly expensive and hard to acquire [ 3 3 , 4 4 ]. Efforts to move away from 3 He detectors have been an area of interest for more than a decade now. Generally, SNMs of interest emit neutrons with average energies around 1 MeV – the fast-neutron region. Thermal neutron detectors rely on moderating fast neutrons to detect them, making them indistinguishable from ambient thermal neutrons. These detectors use total counting techniques, which are not position sensitive, to determine an increase in total neutron flux. Detecting only the total neutron flux doesn't allow them to be sensitive to low-emission fissile materials [ 5 5 - 7 7 5-7 ]. These techniques may also be prone to a high false alarm rate. Active interrogation techniques can be used to enhance the sensitivity to SNMs such as HEU and WGPu. Fission can be induced in these materials by using a low-energy neutron generator, which utilizes reactions such as 7 Li(p,n) 7 Be to produce neutrons. A beam of neutrons with energy around few tens of keV may induce fission in 235 U and 239 Pu, while not inducing fission in other isotopes such as 238 U and 232 Th. The 60-keV neutron beam produces a small dose and can penetrate bulk materials nearly as well as higher energy neutrons. These low-energy neutrons also lose their energy primarily from elastic collisions rather than reactions or inelastic scattering, creating an easily identifiable neutron background. Previous studies have shown positive results in the identification of small quantaties of 235 U, even when shielded by various materials [ 8 8 - 10 10 8-10 ]. Recently, there have been attempts to combat the limitations of monitoring systems based on thermal neutron detection [ 11 11 - 13 13 11-13 ]. Hausladen et al. [ 13 13 ] uses a segmented array of liquid scintillators with pinhole apertures. PMT after-pulsing caused the liquid scintillator setup to perform worse than expected for neutron and γ -ray pulse shape discrimination. Source identification tested well for high-rate sources, but may have difficulty with lower-rate sources. The approach of Rose et al. [ 11 11 ] uses high-energy γ -ray ( E γ > 4 MeV) and fast-neutron active interrogation for spatial imaging. Both the high-energy γ rays and the fast neutrons can induce fission in SNM. However, the high-intensity background inhibits the ability for direct detection and forces the reliance on secondary reactions, including β -decay measurements. Hamel et al. proposed a dual-particle imager (DPI), which can localize both neutrons and γ rays. A neutron generator can be used to induce fission and a veto can be used so that events while the generator is on are discarded. Image reconstruction was done for both neutrons and γ rays using the stochastic origin ensemble statistical technique with the simple backprojection method [ 14 14 ]. Image reconstruction can localize the source, but does not reliably identify the source. After localization, source identification is done by analyzing the intensity decay time while the neutron generator is off. The DPI method can localize and identify SNM, but it requires a multi-step analysis and can take a significant amount of time to identify the source [ 12 12 ]. In order to resolve the existing limitations, we developed a new direct technique for fast-neutron portal monitoring that overcomes the sensitivity limitations of fast-neutron thermalization and allows for the use of active-interrogation techniques. By using small hydrocarbon ( p -terphenyl) scintillators in a large array, we are able to detect fast neutrons to localize fissile materials, including HEU and WGPu. Our novel design, combined with Bayesian statistical analysis can perform with an efficiency that is a factor of 12 better than the standard technique based on 3 He-tubes with a moderator, even for passive interrogation. Insensitivity of this detector to low-energy neutrons makes it ideal for applications that involve active interrogation. The results described in this paper are intended as proof of viability calculations. A prototype detector is being built and will become available in the near future. The pulse shape discrimination thresholds and time resolution that are used in the simulations presented below were adopted from the already existing basic units of this detector. Simulated Detector Design 1 Our proposed design took inspiration from the Gamma-Ray Burst Monitor [ 15 15 ]. Similar to their design of identifying an astrophysical event, we utilize small modular crystals to localize a radioactive source. Our conceptual device consists of an array of 2.5 × 2.5 × 2.5cm 3 p -terphenyl crystals. The detector size is approximately 50×50cm 2 in cross section and 25-cm deep — 4000 crystals in total. The cross section size was chosen for prototyping purposes and the depth was chosen to be approximately two mean free paths of a 1 MeV neutron in p -terphenyl ( Fig. 1 Fig. 1 ). The scintillators are grouped into layers of one crystal in height ( Fig. 2 Fig. 2 ). Each of the layers are spaced 20 mm apart. The detector is simulated in an idealized geometry where the crystals are packed tightly. p -terphenyl 2 P -terphenyl ［C 18 H 14 ］ is an organic crystalline scintillator. P -terphenyl has been of increased interest in fast neutron detection during the past decade due to its large light output (about 20,000 ph/MeV [ 18 18 ]), excellent time resolution (477(12) ps [ 19 19 ]), fast decay time (14.1 ns [ 18 18 ]), and exceptional pulse shape discrimination (PSD), which gives us confidence for positive neutron identification with a good detection efficiency and fast response times. PSD is necessary to filter out γ -ray events from neutron events in the detector. γ rays from ambient background and from various reactions will swamp the detector, but with definitive PSD those events can be filtered out. Prior studies have shown that PSD in p -terphenyl is resolvable down to about the 100 keVee level [ 20 20 , 21 21 ]. Current testing with a pseudo-bar configuration has shown accurate position resolution down to 300 keVee and PSD down to about 150 keVee [ 22 22 ]. In line with the previous studies the PSD threshold used in the simulations was set to be at about the 100 keVee level. MCNP6 Simulations 1 Monte Carlo N Particle (MCNP6) [ 23 23 ] simulations were performed to estimate the performance of our detector. The simulations generated neutrons from an isotropic 235 U+n Watt fission source, which peaks at roughly 1 MeV in the neutron energy spectrum, and an ambient neutron background. Note that the ambient neutron background was produced by default functionality [ 24 24 ]. For the continental United States, there are 39 grid points where the ambient neutron background was sampled. Due to the proximity, the data point at ground level near New Orleans, Louisiana was chosen. At this data point the ambient neutron flux is 0.9907×10 -2 n/cm 2 /s. This results in approximately 100 neutrons/s entering our detector from background sources. Our analysis of the neutron tracks included recoil energy calculations [ 25 25 ] and the PSD threshold explained in section 2.1. All events below the PSD threshold were discarded. Data Analysis 1 Utilization of Double- and Single-Scattering Events 2 The location of the neutron source can be found using segmented neutron detector through observation of neutron double-scattering events (see for example Hamel et al. [ 14 14 ]). Application of this technique, in which a neutron needs to produce at least two signals in two separate crystals, in principle, is possible with the proposed detector system. The time of flight (TOF) for a fast neutron between two crystals may be small, but with the time resolution of around 600 ps for the pseudo-bar [ 22 22 ] most double-scattering events are resolvable. Figure 3 Figure 3 shows the simulated statistics of time difference between the first and second scattering events of the same neutron from the 235 U+n source. It is clear that significant fraction of double-scattering events can be resolved. While double-scattering technique offers an advantage of source localization, sole reliance on double-scattering events would reduce efficiency dramatically. Therefore, we will focus on single-scattering events in further discussion. Our design of a highly-segmented crystal array can utilize the density of scatterings in each of the crystals to detect the presence of the source. Crystals closest to the source will see more events than crystals further from the source due to the attenuation in the detector itself and this effect is the key for identification of the localized source that increases neutron flux from one side of the detector. Scattering Distribution of Directional Neutrons and Ambient Neutron Background 2 The proposed detector is specifically designed to be approximately two mean free paths for a 1-MeV neutron in p -terphenyl ( Fig. 1 Fig. 1 ). As shown in Fig. 4(a) Fig. 4(a) , when we have a source located in front of the detector we observe a large number of neutrons scattering in the first half. The number of source neutrons dramatically decreases with respect to the depth of the detector ( Fig. 5 Fig. 5 ). Figure 4(b) Figure 4(b) shows the relatively homogeneous distribution for hits from background neutrons. This background can be well characterized by periodic sampling of the background rate in each of the crystals. Uniformly Most Powerful Bayesian Tests 1 Uniformly most powerful Bayes tests (UMPBTs) [ 26 26 ] were used to define tests for positive identification. If a positive identification requires that the Bayes factor (i.e., likelihood ratio) for a test exceeds γ , then maximizing the probability of a positive identification can be accomplished by assuming the rate through the detector, λ 1 ［n/s］, which is chosen to maximize the function f (λ 1 ), which corresponds to the number of counts in the detector, given by <math display="block" id="M1"><mrow><mi>f</mi><mo stretchy="false">(</mo><msub><mi>λ</mi><mn>1</mn></msub><mo stretchy="false">)</mo><mo>=</mo><mfrac><mrow><mi>ln</mi><mo stretchy="false">(</mo><mi>γ</mi><mo stretchy="false">)</mo><mo>+</mo><mstyle displaystyle="true"><munder><mo>∑</mo><mi>j</mi></munder></mstyle><mi>g</mi><mo stretchy="false">(</mo><mi>j</mi><mo stretchy="false">)</mo><mo stretchy="false">(</mo><msub><mi>λ</mi><mn>1</mn></msub><mo>−</mo><msub><mi>λ</mi><mn>0</mn></msub><mo stretchy="false">)</mo><mi>Δ</mi><mi>t</mi></mrow><mrow><mi>ln</mi><mo stretchy="false">(</mo><mfrac><mrow><msub><mi>λ</mi><mn>1</mn></msub></mrow><mrow><msub><mi>λ</mi><mn>0</mn></msub></mrow></mfrac><mo stretchy="false">)</mo></mrow></mfrac><mn>.</mn></mrow></math> In this equation, background neutron rate is denoted by λ 0 and a total neutron rate is denoted by λ 1 . A UMPBT threshold of γ is required to declare a positive identification; that is, γ =10 6 means that the ratio of likelihoods evaluated at λ 1 and λ 0 , is required to exceed 10 6 for detection. The function g ( j ) is the average fraction of neutron scatterings per crystal j relative to the ambient neutron background simulated with MCNP6, and Δ t is the real world time that the simulation encompasses. For the purpose of the following calculations, g ( j ) was averaged over 50 simulations of the ambient neutron background, with 10 million events each. The distribution of the number of counts in each crystal was fitted with Poisson function, providing the average, which was then normalized to the total ambient neutron flux. Naturally, the sum of all g ( j ) is less than unity (∑ g ( j )=0.244) and specific values depend on the location of the crystal in the detector. In the real life settings these numbers can be directly measured periodically when there is no sample and corrected automatically (if necessary) in real time. The function f (λ 1 ) is minimized with respect to λ 1 for a fixed time (Δ t ). This would give the minimum source intensity for which identification can be made at the required significance ( γ ) within a fixed time interval. Conversely, one can determine the minimum amount of time required to identify a source with a given intensity at the required level of significance. The limit is set by: <math display="block" id="M2"><mrow><mi>Σ</mi><msub><mi>y</mi><mrow><mi>i</mi><mi>j</mi></mrow></msub><mo>></mo><mi>f</mi><mo stretchy="false">(</mo><msub><mi>λ</mi><mn>1</mn></msub><mo stretchy="false">)</mo><mn>,</mn></mrow></math> where ∑ y ij is the sum of events in the detector as a whole. The limit occurs when ∑ y ij =f(λ 1 ), inferring that a source can be detected if it emits enough neutrons to satisfy Eq. (2). The minimization of f( λ 1 ) is done by varying the value of λ 1 . When f( λ 1 ) and λ 1 are plotted against one another, it is clear where the minimum lies ( Fig. 6 Fig. 6 ) for any given time interval of the measurements. The trend is asymptotic at λ 1 = λ 0 , as expected. By taking into account the geometric efficiency, the source neutron rate can be derived. The estimated number of source neutrons emitted per second is simply: <math display="block" id="M3"><mrow><msub><mi>I</mi><mrow><mtext>src,det</mtext></mrow></msub><mo>=</mo><mo stretchy="false">(</mo><msub><mi>λ</mi><mn>1</mn></msub><mo>−</mo><msub><mi>λ</mi><mn>0</mn></msub><mo stretchy="false">)</mo><mn>,</mn></mrow></math> <math display="block" id="M4"><mrow><msub><mi>I</mi><mrow><mtext>src</mtext></mrow></msub><mo>=</mo><msub><mi>I</mi><mrow><mtext>src,det</mtext></mrow></msub><mo>×</mo><mfrac><mrow><mn>4</mn><mi>π</mi></mrow><mrow><msub><mi>Ω</mi><mrow><mtext>sim</mtext></mrow></msub></mrow></mfrac><mn>,</mn></mrow></math> with I src,det being the number of neutrons emitted per second by the source within the solid angle covered by the detector active area. An intrinsic efficiency of the detector is given by: <math display="block" id="M5"><mrow><msub><mi>ϵ</mi><mrow><mtext>sim</mtext></mrow></msub><mo>=</mo><mfrac><mrow><msub><mi>N</mi><mrow><mtext>det,sim</mtext></mrow></msub></mrow><mrow><msub><mi>N</mi><mrow><mtext>tot,sim</mtext></mrow></msub></mrow></mfrac><mo>×</mo><mfrac><mrow><mn>4</mn><mi>π</mi></mrow><mrow><msub><mi>Ω</mi><mrow><mtext>sim</mtext></mrow></msub></mrow></mfrac><mn>,</mn></mrow></math> where N det,sim is the number of neutrons from the source that produce a signal above the threshold in the detector, N tot,sim is the total number of source neutrons in the simulation, and Ω sim is the solid angle encompassed by the detector with respect to the point source in the simulation. The intrinsic efficiency is of course a function of energy threshold, which was set to a realistic value as explained in Sect. 3 in order to have clean neutron identification. The time it takes for the detector to identify the source is then: <math display="block" id="M6"><mrow><mi>t</mi><mo>=</mo><mfrac><mrow><mi>f</mi><mo stretchy="false">(</mo><msub><mi>λ</mi><mn>1</mn></msub><mo stretchy="false">)</mo><mo>/</mo><msub><mi>ϵ</mi><mrow><mtext>sim</mtext></mrow></msub></mrow><mrow><msub><mi>I</mi><mrow><mtext>src</mtext></mrow></msub><mo>×</mo><mfrac><mi>Ω</mi><mrow><mn>4</mn><mi>π</mi></mrow></mfrac></mrow></mfrac><mn>,</mn></mrow></math> where Ω is the solid angle from the source position being calculated. As one may expect, the identification time scales with the solid angle and detector efficiency. Comparison to industry standards 2 The National Committee on Radiation Instrumentation (NCRI) sets the requirements for radiation portal monitors. A Pacific Northwest National Laboratory (PNNL) study (Kouzes et al.) in 2010 describes the requirements for alternative neutron portal monitors for the current 3 He monitors [ 27 27 ]. Following the parameters discussed in this study, the fulfilment of requirements is tested. The requirements set are in terms of 252 Cf spontaneous fission, which has a notable conversion of 1 ng to 2.1×10 3 n/s. The efficiency requirement for an alternative neutron monitor is 2.5 cps/ng (counts per second per nanogram) for a source 2 m away from the detector. The source is also required to be moderated by 2.5 cm of polyethylene. The 252 Cf source used emitted a total of 950,000 neutrons over a span of 1.388 seconds. This source intensity corresponds to a source of 326 ng. The 252 Cf source and a 2.5-cm thick polyethylene wall were implemented in an MCNP6 simulation to reconstruct the tests of the 3 He detector from Kouzes et al. [ 27 27 ]. The specific geometry was duplicated according to previous MCNP simulations for the detector [ 1 1 ]. A single 3 He tube at 3 atm was used for this simulation. The simulations yielded a neutron rate of 597 n/s being detected, which corresponds to an efficiency of 1.83 cps/ng. The difference between this efficiency and the quoted 3.0 cps/ng is likely due to the undefined polyethylene thickness of the real 3 He detector. A comparative simulation was constructed with the proposed detector replacing the 3 He detector. The simulation yielded a neutron rate of 3203.22 n/s after a 2.5 cm thick polyethylene moderator. It should be noted, however, that the polyethylene moderator is not necessary for the proposed detector apparatus, but included in order to comply with the NCRI standards and make a direct comparison to previous studies. This rate corresponds to an efficiency of 9.83 cps/ng. The efficiency then needed to be normalized to the size of the 3 He detector. Normalization was done based on the ratio of the volumes of the two detectors, which provides only an estimate for the scaling of the efficiency. The total volume of the 3 He detector is 170,190 cm 3 (183 cm × 15 cm × 62 cm) [ 1 1 ] and the total volume of the proposed detector is 62,500 cm 3 . This leads to a detection efficiency of 26.8 cps/ng that is directly comparable to the previous study. Table 1 Table 1 shows the comparison of the calculated efficiency of the proposed detector to commercially-available detectors tested in Kouzes et al . [ 27 27 ]. Values marked with a † were taken from Kouzes et al. 1 1 Detector type 1 1 Efficiency (cps/ng) 1 1 Proposed Detector - Without UMPBT 1 1 26.8 1 1 3 He proportional detector (1 tube) 1 1 3.0(2) † 1 1 BF 3 proportional detector (3 tubes) 1 1 3.7(2) † 1 1 Boron-lined proportional detector 1 1 3.0(2) † 1 1 Lithium-loaded glass fibers 1 1 1.7(6) † 1 1 Coated non-scintillating plastic fibers 1 1 2.0(1) † In order to make a direct comparison to the 3 He detector, positive detection confidence levels were calculated for a 252 Cf source (rate of 3112 n/s) at a distance of 2 m from the detector and at various exposure times. A series of 50 MCNP simulations were performed for each time with and without a source present. A Poisson average was taken for the sets of 50 simulations to get the average number of neutrons detected from the source, C src , and from the background, C BG . For each exposure time, the error on the background was calculated as: <math display="block" id="M7"><mrow><mi>ϵ</mi><mo>=</mo><msqrt><mrow><msub><mi>C</mi><mrow><mtext>BG</mtext></mrow></msub></mrow></msqrt><mn>.</mn></mrow></math> With ϵ , the confidence interval was calculated with: <math display="block" id="M8"><mrow><mi>x</mi><mo>=</mo><mfrac><mrow><msub><mi>C</mi><mrow><mtext>Total</mtext></mrow></msub><mo>−</mo><msub><mi>C</mi><mrow><mtext>BG</mtext></mrow></msub></mrow><mi>ϵ</mi></mfrac><mn>,</mn></mrow></math> with C Total being the sum of C src and C BG . The confidence interval was then used to calculate the confidence level in the form 1 in γ , with: <math display="block" id="M9"><mrow><mi>γ</mi><mo>=</mo><mfrac><mn>1</mn><mrow><mn>1</mn><mo>−</mo><mi>E</mi><mi>r</mi><mi>f</mi><mo stretchy="false">(</mo><mfrac><mi>x</mi><mrow><msqrt><mn>2</mn></msqrt></mrow></mfrac><mo stretchy="false">)</mo></mrow></mfrac><mn>.</mn></mrow></math> Similarly, the confidence level for the UMPBT model was found using the same data with Eq.(1). Figure 7 Figure 7 shows the abundant sensitivity advantage that the proposed detector (with the UMPBT model) has over a 3 He detector. Active interrogation technique with the proposed detector 2 Since the proposed detector uses fast neutrons to detect the presence of fissile material, it becomes possible to use a beam of low-energy neutrons for active interrogation. It is one of the key advantages of the proposed detector system. In particular, it has been shown in Ref. [ 9 9 ] that 60 keV neutrons produced by 7 Li(p,n) 7 Be neutron generator can be utilized for this purpose. For the case of 235 U, the UMPBT model can be used to estimate the minimum time required to detect a fissile material under active interrogation. Plotting the two calculated values (Eqs. (3) and (6)) against one another provides a sensitivity limitation trend for our detector ( Figure 8 Figure 8 ). Table 2 Table 2 shows the minimum time it would take to detect various source strengths at various distances. 1 1 Source distance (m) 1 1 I  src,det  (n/s) 1 1 I  src  (n/s) 1 1 Time (sec) 1 1 1 1 1 1 1 1 50.3 1 1 16703 1 1   1 1 10 1 1 502.7 1 1 1477 1 1   1 1 100 1 1 5027 1 1 192 1 1   1 1 200 1 1 10053 1 1 120 1 1   1 1 800 1 1 40212 1 1 60 1 1 2 1 1 50 1 1 10053 1 1 331 1 1   1 1 200 1 1 40212 1 1 120 1 1 3 1 1 22 1 1 9953 1 1 686 1 1   1 1 89 1 1 40263 1 1 209 1 1 5 1 1 8 1 1 10053 1 1 1851 1 1   1 1 32 1 1 40212 1 1 487 1 1 10 1 1 2 1 1 10053 1 1 7966 1 1   1 1 8 1 1 40212 1 1 1851 The source intensity was converted into mass to determine how much of the material must be present in order to be detectable. We assume that neutron flux for a 60 keV beam is ϕ =5 × 10 6 cm -2 s -1 , as in Ref. [ 9 9 ]. The fission cross section, <math id="M10"><mrow><mi>σ</mi><mo>≈</mo><mn>2</mn></mrow></math> barns [ 28 28 ] at 60 keV, and the prompt neutron multiplicity, ν p =2.42 [ 28 28 ], of the 235 U+n reaction were used. Using equation (10), Avagadro's number, N A , and the mass M 235 U =235 g, the conversion was performed to estimate the mass for a range of source neutron intensities ( Table 3 Table 3 ). A similar plot to Fig. 8 Fig. 8 was then made with respect to mass ( Fig. 9 Fig. 9 ). <math display="block" id="M11"><mrow><mi>M</mi><mo>=</mo><mfrac><mrow><msub><mi>I</mi><mrow><mtext>src</mtext></mrow></msub></mrow><mrow><msub><mi>ν</mi><mtext>p</mtext></msub><msub><mi>N</mi><mtext>A</mtext></msub><mi>σ</mi><mi>ϕ</mi></mrow></mfrac><msub><mi>M</mi><mrow><msup><mrow/><mrow><mn>235</mn></mrow></msup><mtext>U</mtext></mrow></msub></mrow></math> It is clear that the proposed detector system becomes extremely sensitive to even very small amounts of fissile material under active interrogation. Figures 8 Figures 8 and 9 9 indicate that 1 g of 235 U can be detected in less than a minute if modest flux of 5×10 6 cm -2 s -1 of 60 keV neutrons is used for active interrogation. Note that it is not possible to use systems based on thermal neutron detection in this case because of the background which will be unavoidably caused by the interrogating neutron beam. The detection times found for the various masses at one meter from the detector are also displayed. The confidence level used is γ =10 6 . 1 1 I  src,det  × 10 3 (n/s) 1 1 Mass (g) 1 1 Time (sec) 1 1 0.01 1 1 0.0081 1 1 1477 1 1 0.1 1 1 0.0807 1 1 192 1 1 0.2 1 1 0.161 1 1 120 1 1 0.8 1 1 0.646 1 1 59.8 1 1 1.5 1 1 1.21 1 1 47.4 1 1 2 1 1 1.61 1 1 43.1 1 1 3 1 1 2.42 1 1 38.2 1 1 4 1 1 3.23 1 1 35.3 1 1 5 1 1 4.03 1 1 33.4 1 1 10 1 1 8.07 1 1 28.4 1 1 25 1 1 20.2 1 1 23.7 Summary and Conclusion 1 Special nuclear materials should be carefully monitored and controlled. The monitoring of these materials is of critical importance for global security, especially weapons-grade plutonium and highly enriched uranium. Detection of these materials can be facilitated by active interrogation with low-energy ( < 100 keV) neutrons. Even without active interrogation, our proposed detector design and analysis method overcomes the limitations set by thermal-neutron detectors as well as the higher false alarm rate of γ -ray detectors. Using MCNP6 simulations and the uniformly most powerful Bayesian tests statistical model, the limits of the proposed detector were probed. For a fissile material undergoing fission one meter away from the detector, without any moderation, it would take about 25 minutes to identify a source emitting ≈ 500 n/s, and similarly 192 seconds for ≈ 5000 n/s. For the specific case for 235 U+n, with the specific parameters discussed in Sect. 5.2, these limits translate to 8.1 mg and 80.7 mg of 235 U, respectively. Comparisons to commercially-available neutron detectors, studied in Kouzes et al. [ 27 27 ], were made ( Table 1 Table 1 ). The requirements set by the National Committee on Radiation Instrumentation at the time of the previous study required a minimum neutron detection efficiency of 2.5 cps/ng. The detector proposed in the current work surpasses that requirements set. The proposed detector performs with an efficiency of 26.8 cps/ng. With the UMPBT model, a comparison of confidence levels for a particular source at various exposure times revealed a significant increase in sensitivity compared to the 3 He detector ( Fig. 7 Fig. 7 ). With respect to the previously-studied detectors described in Kouzes et al., the proposed neutron detection has a significant advantage in efficiency and sensitivity. Our method can be applied to the development of a monitoring device and the times produced scale linearly with the solid angle of the detector with respect to the source. Increasing the number of the proposed detectors in various positions around the source will increase the sensitivity and reduce the required detection time of the detector significantly. References R.T. Kouzes , E.R. Siciliano , J.H. Ely et al. , Passive neutron detection for interdiction of nuclear material at borders . Nucl. Instrum. Meth. A 584 , 383 - 400 ( 2008 ) doi: 10.1016/j.nima.2007.10.026 http://doi.org/10.1016/j.nima.2007.10.026 Passive neutron detection for interdiction of nuclear material at borders J.L. Lacy , A. Anthanasiades , C.S. Martin et al. , Straw-based portal monitor 3 He replacement detector with expanded capabilities . IEEE Nucl. Sci. Conf. R 4865 - 4868 ( 2011 ) doi: 10.1109/NSSMIC.2011.6152486 http://doi.org/10.1109/NSSMIC.2011.6152486 Straw-based portal monitor 3He replacement detector with expanded capabilities R.T. Kouzes and J.H. Ely , PNNL-19360: Status summary of 3 He and neutron detection alternatives for homeland security ( 2010 ). PNNL-19360: Status summary of 3He and neutron detection alternatives for homeland security D.A. Shea and L. Daniel , The helium-3 shortage: Supply, demand, and options for congress. , Congressional Research Service Washington, DC ( 2010 ). The helium-3 shortage: Supply, demand, and options for congress. P.E. Vanier , L. Forman , R. Daren , Thermal neutron imaging in an active interrogation environment . AIP Conf. Proc. 1099 , 583 ( 2009 ) doi: 10.1063/1.3120104 http://doi.org/10.1063/1.3120104 Thermal neutron imaging in an active interrogation environment C.E. Moss , C.L. Hollas , G.W. McKinnety et al. , Comparison of active interrogation techniques, IEEE Nucl . Sci. Conf. R 2005 , 329 - 332 ( 2005 ) doi: 10.1109/NSSMIC.2005.1596264 http://doi.org/10.1109/NSSMIC.2005.1596264 Comparison of active interrogation techniques, IEEE Nucl R.C. Runkle , D.L. Chichester , and S.L. Thompson , Rattling nucleons: New developments in active interrogation of special nuclear material . Nucl. Instrum. Meth. A 663 , 75 - 95 ( 2011 ) doi: 10.1016/j.nima.2011.09.052 http://doi.org/10.1016/j.nima.2011.09.052 Rattling nucleons: New developments in active interrogation of special nuclear material D. Dietrich , C. Hagmann , P. Kerr et al. , A kinematically beamed, low energy pulsed neutron source for active interrogation . Nucl. Instrum. Meth. B 241 , 826 - 830 ( 2005 ) doi: 10.1016/j.nimb.2005.07.139 http://doi.org/10.1016/j.nimb.2005.07.139 A kinematically beamed, low energy pulsed neutron source for active interrogation C.A. Hagmann , D.D. Dietrich , J.M. Hall et al. , Active detection of shielded SNM with 60-keV neutrons . IEEE T Nucl. Sci. 56 , 1215 - 1217 ( 2009 ) doi: 10.1109/TNS.2009.2012859 http://doi.org/10.1109/TNS.2009.2012859 Active detection of shielded SNM with 60-keV neutrons P. Kerr , M. Rowland , D. Dietrich et al. , Active detection of small quantities of shielded highly-enriched uranium using low-dose 60-kev neutron interrogation . Nucl. Instrum. Meth. B 261 , 347 - 350 ( 2007 ) doi: 10.1016/j.nimb.2007.04.190 http://doi.org/10.1016/j.nimb.2007.04.190 Active detection of small quantities of shielded highly-enriched uranium using low-dose 60-kev neutron interrogation P.B. Rose , A.S. Erickson , M. Mayer et al. , Uncovering Special Nuclear Materials by Low-energy Nuclear Reaction Imaging . Scientific Reports 6 , 24388 ( 2016 ) doi: 10.1038/srep24388 http://doi.org/10.1038/srep24388 Uncovering Special Nuclear Materials by Low-energy Nuclear Reaction Imaging M.C. Hamel , J.K. Polack , M.L. Ruch et al. , Active neutron and gamma-ray imaging of highly enriched uranium for treaty verification OPEN . Scientific Reports 7 , 7997 ( 2017 ) doi: 10.1038/s41598-017-08253-x http://doi.org/10.1038/s41598-017-08253-x Active neutron and gamma-ray imaging of highly enriched uranium for treaty verification OPEN P. Hausladen , M. Blackston , E. Brubaker et al. , Fast neutron coded-aperture imaging of special nuclear material configurations , 53rd Annual Meeting of the INMM ( 2012 ) Fast neutron coded-aperture imaging of special nuclear material configurations M.C. Hamel , J.K. Polack , A. Poitrasson-Rivière et al. , Stochastic image reconstruction for a dual-particle imaging system . Nucl. Instrum. Meth. A 810 , 120 - 131 ( 2016 ). doi: 10.1016/j.nima.2015.12.002 http://doi.org/10.1016/j.nima.2015.12.002 Stochastic image reconstruction for a dual-particle imaging system C. Meegan , G. Lichti , P.N. Bhat et al. , The Fermi gamma-ray burst monitor . Astrophy. J. 702 , 791 - 804 ( 2009 ). doi: 10.1088/0004-637X/702/1/791 http://doi.org/10.1088/0004-637X/702/1/791 The Fermi gamma-ray burst monitor D.A. Brown , M.B. Chadwick , R. Capote et al. , ENDF/B-VIII.0: The 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data . Nucl. Data Sheets 148 , 1 - 142 ( 2018 ). doi: 10.1016/j.nds.2018.02.001 http://doi.org/10.1016/j.nds.2018.02.001 ENDF/B-VIII.0: The 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data M.B. Chadwick , M. Herman , P. Obložinský et al. , ENDF/B-VII.1 nuclear data for science and technology: Cross sections, covariances, fission product yields and decay data . Nucl. Data Sheets 112 , 2887 - 2996 ( 2011 ). doi: 10.1016/j.nds.2011.11.002 http://doi.org/10.1016/j.nds.2011.11.002 ENDF/B-VII.1 nuclear data for science and technology: Cross sections, covariances, fission product yields and decay data T. Yanagida , K. Watanabe , Y. Fujimoto , Comparative study of neutron and gamma-ray pulse shape discrimination of anthracene, stilbene, and p-terphenyl . Nucl. Instrum. Meth. A 784 , 111 - 114 ( 2015 ) doi: 10.1016/j.nima.2014.12.031 http://doi.org/10.1016/j.nima.2014.12.031 Comparative study of neutron and gamma-ray pulse shape discrimination of anthracene, stilbene, and p-terphenyl A. Sardet , C. Varignon , B. Laurent et al. , p-Terphenyl: An alternative to liquid scintillators for neutron detection . Nucl. Instrum. Meth. A 792 , 74 - 80 ( 2015 ) doi: 10.1016/j.nima.2015.04.038 http://doi.org/10.1016/j.nima.2015.04.038 p-Terphenyl: An alternative to liquid scintillators for neutron detection J. Iwanowska , L. Swiderski , M. Moszynski et al. , Neutron/gamma discrimination properties of composite scintillation detectors . J. Instrum. 6 , P07007 ( 2011 ) doi: 10.1088/1748-0221/6/07/P07007 http://doi.org/10.1088/1748-0221/6/07/P07007 Neutron/gamma discrimination properties of composite scintillation detectors M.L. Ruch , M. Flaska , S.A. Pozzi , Pulse shape discrimination performance of stilbene coupled to low-noise silicon photomultipliers . Nucl. Instrum. Meth. A 793 , 1 - 5 ( 2015 ) doi: 10.1016/j.nima.2015.04.053 http://doi.org/10.1016/j.nima.2015.04.053 Pulse shape discrimination performance of stilbene coupled to low-noise silicon photomultipliers D.P. Scriven , G. Christian , G.V. Rogachev et al. , A position and pulse shape discriminant p-terphenyl detector module . Nucl. Instrum. Meth. A 1010 , 165492 ( 2021 ) doi: 10.1016/j.nima.2021.165492 http://doi.org/10.1016/j.nima.2021.165492 A position and pulse shape discriminant p-terphenyl detector module T. Goorley , M. James , T. Booth et al. , Initial MCNP6 release overview . Nucl. Technol. 180 , 298 - 315 ( 2012 ) doi: 10.13182/NT11-135 http://doi.org/10.13182/NT11-135 Initial MCNP6 release overview G.E. McMath , G.W. McKinney , and T. Wilcox , MCNP6 Cosmic & Terrestrial Background Particle Fluxes–Release 4 ( 2015 ) MCNP6 Cosmic & Terrestrial Background Particle Fluxes–Release C. Matei , F.J. Hambsch , S. Oberstedt , Proton light output function and neutron efficiency of a p-terphenyl detector using a 252Cf source . Nucl. Instrum. Meth. A 676 , 135 - 139 ( 2012 ) doi: 10.1016/j.nima.2011.11.076 http://doi.org/10.1016/j.nima.2011.11.076 Proton light output function and neutron efficiency of a p-terphenyl detector using a 252Cf source V.E. Johnson , Uniformly most powerful Bayesian tests . Ann. Stat. 41 , 1716 - 1741 ( 2013 ). doi: 10.1214/13-AOS1123 http://doi.org/10.1214/13-AOS1123 Uniformly most powerful Bayesian tests R.T. Kouzes , J.H. Ely , L.E. Erikson et al. , Alternative neutron detection testing summary , PNNL ( 2010 ) Alternative neutron detection testing summary K. Shibata , O. Iwamoto , T. Nakagawa et al. , JENDL-4.0: A New Library for Nuclear Science and Engineering . J. Nucl. Sci. Technol. 48 , 1 - 30 ( 2011 ) doi: 10.1080/18811248.2011.9711675 http://doi.org/10.1080/18811248.2011.9711675 JENDL-4.0: A New Library for Nuclear Science and Engineering All authors contributed to the design of the detector and the methodology of the study. Simulations and analyses were performed by Eric Aboud with assistance by Sunghoon Ahn. The UMPBT model was provided by Valen E. Johnson. The first draft of there manuscript was written by Eric Aboud and all authors provided feedback on drafts of the manuscript. All authors read and approved the final manuscript. 10.1007/s41365-022-00990-7 This material is based upon work was supported by the National Nuclear Security Administration (NNSA) through the Center for Excellence in Nuclear Training and University Based Research (CENTAUR) under Award No. DE-NA0003841, by the U.S.Department of Energy, Office of Science, Office of Nuclear Physics, under Award No. DE-FG02-93ER40773, and the College of Science at Texas A&M University through Strategic Transformative Research Program (CoS STRP). E. Aboud acknowledges funding through the Nuclear Solutions Institute's Graduate Student Merit Fellowship. 2021-10-12 2021-12-04 2021-12-11 2022-01-29 2022-02-16 © China Science Publishing & Media Ltd. (Science Press), Shanghai Institute of Applied Physics, the Chinese Academy of Sciences, ChineseNuclear Society and Springer Nature Singapore Pte Ltd. 2022 This is a post-peer-review, pre-copyedit version of an article published in Nuclear Science and Techniques. The final authenticated version is available online at: http://dx.doi.org/10.1007/s41365-022-00990-7 http://dx.doi.org/10.1007/s41365-022-00990-7 . 2022 Nuclear Science and Techniques 01 33