Monte Carlo simulations of response functions for gas filled and scintillator detectors with MCNPX code

Monte Carlo simulation of a detector response function presents a very challenging problem. The detector response functions have been calculated for different neutron and gamma detectors: ³He gas filled proportional counter, NE213 organic scintillator, BrillanCe 350 or LaCl₃(Tl), and an ionization chamber with mixed gas composition. MCNPX code was used for simulations. The simulations were done with different neutron and gamma energies. The effects of neutron scattering, wall effects, recoil continua and contribution from charged particles have been included. The detector response function for the NE213 organic scintillator was obtained with consideration of light output curves of different products of neutron reactions with materials of the scintillator. The simulated data has been compared with experiments.     

直径30 cm的等高圆柱形载钆液闪探测性能

载钆液闪探测器是高能物理及核物理实验中重要的粒子探测工具。通过研制得到了一台大体积的直径为30 cm等高圆柱形载钆液闪探测器,载钆液闪溶液的载钆量为0.5%wt;利用252Cf中子源进行了中子与γ分辨性能实验测试,结果表明,直径30 cm等高圆柱形载钆液闪的中子与γ分辨性能较差;利用飞行时间技术通过符合测量的方法,分别测量了中子与伽马分辨谱中的中子与γ信号的时间分布,两者峰位之间的时间差为2 ns;利用252Cf裂变电离室的裂变碎片信号作为开门信号,通过符合测量的方法,获得了直径30 cm等高圆柱形载钆液闪的中子俘获时间分布实验数据,中子俘获平均时间为11μs。对于较大体积条件下,载钆液闪的中子与γ分辨性能较差的物理现象,通过实验给出了合理解释和分析。     

A comparison of traditional and hybrid radiation detector dead-time models and detector behavior

High intensity radiation measurements are confounded by detector dead-time and pulse pile-up problems. A computational method was used to compare the traditional dead-time models with recently proposed hybrid dead-time models. A computational algorithm based on a decay source method was used to study the behavior of various dead-time models. Validation of the code was performed for the hybrid models by confirming that the predictions lie between the two ideal dead-time models; the paralyzing and the non-paralyzing model. It was interesting to note that two seemingly similar hybrid dead-time models produced significantly different results. Lee and Gardner's model based on two dead-times and Patil and Usman paralysis factor based model are inherently different in their logic as well as results. For Lee and Gardner's model altering the orders of dead-times produced significantly different response. These hybrid models should be studied further to investigate both the dependence and the variation of model parameters on detector design and operating conditions. It is well accepted that one dead-time does not apply to all detectors and even for the same detector applicability of the same model under all operating condition is questionable. Therefore, dead-time model should be chosen carefully for the specific detector, operating conditions and radiation to be measured to correctly represent the physical measurement phenomenon.     

Characterization of the neutron flux distribution at the Athens Tandem Accelerator NCSR “Demokritos”

Quasi-monoenergetic neutron beams, in the energy range 7-11.5 MeV, produced via the ²H(d,n) reaction, have been used at the 5.5 MV tandem T11/25 Accelerator Laboratory of NCSR “Demokritos”. The flux variation of the neutron beam is monitored with a BF₃ detector, while the absolute flux is obtained with respect to reference reactions. An investigation of the energy dependence of the neutron fluence has been carried out using two independent techniques: by a liquid scintillator BC501A detector and deconvolution of its recoil energy spectra performed by means of the DIFBAS code, as well as via the multiple foil activation technique in combination with the SULSA unfolding code. The neutron facility has also been characterized by means of Monte Carlo simulations with MCNP5.     

Comparison of photoneutron yield from beryllium irradiated with bremsstrahlung radiation of different peak energy

In an effort to set up microtron based photoneutron source and in order to optimize the neutron yield, photoneutron production from beryllium has been studied for different volumes of beryllium irradiated by different peak energy bremsstrahlung radiation. The theoretical estimation of neutron yield has been carried out using the MCNP simulation for 8.75, 8.15 and 7.58 MeV peak energy of bremsstrahlung radiation. The experimental measurements were carried out using two types of detectors: SSNTD CR-39 and custom designed Silver wrapped GM detector. The neutron yield corresponding to beryllium of volume 381.70 cm³ are found to be 2.13E+09, 2.00E+09 and 1.74E09 n/s (MCNP calculation values) for electrons of energy 8.75, 8.17 and 7.48 MeV, respectively. The experimental results are compared with the MCNP simulated results and are good agreement.     

Monte Carlo simulations of a portable prompt gamma system for nondestructive determination of chloride in reinforced concrete

In this work, the MCNP code was used to perform Monte Carlo simulations of the operation of a portable prompt gamma neutron activation (PGNA) system for chloride detection in reinforced concrete. The system consists of a moderated ²⁵²Cf neutron source, a high purity germanium (HPGe) gamma ray detector and a portable multichannel analyzer. The system maximum weight is 23 kg with a largest dimension of 31 cm. The simulations utilized a hybrid approach, which consisted of using MCNP simulations to model neutron transport and ray tracing for gamma ray transport, which considerably reduces computation time in comparison to a fully coupled neutron/photon Monte Carlo simulations. The simulations have shown that the current moderator design effectively thermalizes the neutron energy spectrum. At low to moderate chloride concentrations, the hybrid simulation model of the PGNA chloride detector shows very good agreement with experimental data. The MCNP computations predicted that for a standard error of 10% in counting statistics, the detection of a 2000 ppm chloride concentration (the corrosion threshold) in reinforced concrete can be achieved in a seven minute counting period. This represents a significant improvement over the current standard destructive method of measuring chlorides in concrete. Over the range of water to cement (w/c) ratios normally found in concrete mixes (0.38-0.55), the chloride signal strength shows very little variation especially at the lower chloride concentrations. Thus for all practical purposes the chloride signal remains insensitive to the w/c ratio. Similarly, the chloride signal strength does not vary significantly if limestone coarse or fine aggregate is used in place of quartz.     

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF₃ detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample’s surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.     

一种新型中子高效率探测器技术研究

根据现有的中子探测器多采用气体探测材料,普遍存在体积大,效率不高和响应时间长的缺点,结合中子探测的最新发展,用蒙特卡罗方法通用软件MCNP(4B版本)对中子探测材料和结构进行了研究,并在此基础上设计出了一种新型中子探测器模型,对它的探测性能进行了数值仿真。该新型中子探测器的设计不论是对民用还是对军用都将有着十分重要的意义。     

Design of moderator and multiplier systems for D–T neutron source in BNCT

Boron Neutron Capture Therapy (BNCT) is an outstanding way to treat Glioblastoma Multiforme. Epithermal neutrons with energy from 1 eV to 10 keV represent the most effective range for brain tumor therapy. In this research we have focused on ³H(d, n)⁴He reaction as a neutron source using Cock Craft Walton accelerator. High neutron yield with 14.1 MeV energy can be generated via accelerating a deuteron beam with 110 keV energy.A Monte Carlo simulation code (MCNP4C) was used to design the D–T source. Pb and ²³⁸U are suggested as neutron multipliers; AlF₃ and BeO as a moderator and reflector, respectively. An Al layer is used for decreasing the ratio of fast to total neutron fluxes. Epithermal neutron flux in the suggested system is 10⁸ n/cm² s and is a suitable flux for BNCT applications. Finally the suggested configuration is compared to the most recent works and it is shown that the proposed configuration works better.     

Dead time corrections on the Feynman-Y curve using the backward extrapolation method

Dead time losses in neutron detection, caused by both the detector and the electronics dead time, is a highly nonlinear effect, known to create high biasing in physical experiments as the power grows over a certain threshold. Analytic modeling of the dead time losses is a highly complicated task due to the different nature of the dead time in the different components of the monitoring system (paralyzing vs. non-paralyzing), and the stochastic nature of the fission chains. The most basic analytic models for a paralyzing dead time correction assume a non-correlated source, resulting in an exponential model for the dead time correction. While this model is often used and very useful for correcting the average count rate in low count rates, it is totally impractical in noise experiments and the so-called Feynman-α experiments. In the present study, a new technique is introduced for dead time corrections, based on backward extrapolation of the losses, created by imposing increasing artificial dead time on the data, back to zero. The method is implemented on neutron noise measurements carried out in the MINERVE reactor, demonstrating high accuracy in restoring the corrected values of the Feynman-Y variance-to-mean-ratio.     

Sensitivity of fuel failure detection by the delayed neutron measurement in the primary cooling circuit of HANARO

The sensitivity of the fuel failure detection system based on the delayed neutron measurement in the primary cooling circuit of a research reactor, HANARO is investigated. The neutrons around the primary cooling pipe during normal operation of HANARO are measured with BF₃ detector, and their count rate is 900 cps. They are regarded as photoneutrons due to the high energy gamma-rays from N-16 and delayed neutrons from the fission of the uranium contaminated on the fuel surface. The contribution of each neutron source is analyzed by measuring the changes of the neutron counts before and after the abrupt shutdown of reactor. In order to estimate the sensitivity of the fuel failure detection, the neutron count rate of BF₃ detector is predicted by Monte Carlo calculation. The generation, transportation and detection of the photoneutrons and the delayed neutrons are simulated for the geometry similar to the experiments. From the calculations and experiments, it is ascertained that the photoneutron contribution to the total count rate is about 20–30%, and that the delayed neutron count rate is expected to about 720 cps. The fission rate in the flow tube of the reactor core by the surface contamination is obtained from the deduced delayed neutron count rate, and it is estimated to 1.66 × 10⁵ fissions/cm³ s. From the MCNP calculation, it is confirmed that this fission rate can originate from the contaminated uranium of 120 μg, which is about 13% of the maximum allowable surface contamination on the fuel surface. The sensitivity of U-235 mass detection by the delayed neutron measurement can be concluded to about 0.2 μg-U²³⁵/cps. Thus, it is confirmed that the delayed neutron detection is sensitive enough to monitor the fuel failure, and that the neutron count rate is high enough for stable signal with short counting time.     

Measurements and model calculations of activation cross sections for 232Th(n,2n)231Th reaction between 13.57 and 14.83 MeV neutrons

In this study, activation cross sections were measured for the reaction of ²³²Th(n,2n)²³¹Th (T_1/2 = 25.5 h) by using neutron activation technique at six different neutron energies from 13.57 and 14.83 MeV. Neutrons were produced via the ³H(²H,n)⁴He reaction using SAMES T-400 neutron generator. Irradiated and activated high purity Thorium foils were measured by a high-resolution γ-ray spectrometer with a high-purity Germanium (HpGe) detector. In cross section measurements, the corrections were made for the effects of γ-ray self-absorption in the foils, dead-time, coincidence summing, fluctuation of neutron flux, low energy neutrons. For this reaction, statistical model calculation, which the pre-equilibrium emission effects were taken into consideration, were also performed between 13.57 and 14.83 MeV energy range. The cross sections were compared with previous works in literature, with model calculation results, and with evaluation data bases (ENDF/B-VII, ENDF/B-VI, JEFF-3.1, JENDL-4.0, JENDL-3.3, and ROSFOND-2010).     

用于闪烁薄膜探测器灵敏度标定的中子屏蔽体设计

为标定薄膜厚度大于0.1mm的闪烁薄膜探测器灵敏度,采用MCNP程序建模优化设计了适用于在中国原子能科学研究院放射性计量测试部的5SDH-2串列加速器上进行实验的中子屏蔽体。实验表明,该屏蔽体可将偏离通道的中子注量减弱到通道中子注量的十分之一以下,将本底信号抑制在较光电倍增管暗电流略低的水平上,对于薄膜厚度大于0.1mm的探测器,可使其信噪比大于1∶1。计算表明,准直孔的散射对探测器测量灵敏度的影响不超过5％。     

Development of large-area composite stilbene scintillator for fast neutron detection

Fast neutron applications have gained popularity with the growth of fast neutron production facilities. Covering a larger area and/or wider angle can be one of the advantages of a fast neutron detector. In the present study, a large-area composite stilbene scintillator with the dimensions of 200 mm (D) × 20 mm (H) was fabricated to examine its scintillation properties and to evaluate its applicability to fast neutron detection. The detector response of small- and large-area composite stilbene scintillators for neutrons and gamma rays was measured and compared with that of commercial and small single-crystal stilbene scintillators. To this end, the response of each scintillator was measured for radioisotopes as well as mono-energetic neutrons generated by a Tandem accelerator. The neutron–gamma separation performance of the large-area composite stilbene scintillator was evaluated in terms of figure-of-merit (FoM) using the digital pulse shape discrimination method. The composite stilbene scintillator showed good energy linearity, as determined from its recoil proton spectra, with reasonable n–γ separation capability. The results indicated that the composite stilbene scintillator could be applied to the field of fast neutron detection, especially when a large area and/or a wide angle is to be covered and could be a good alternative to liquid scintillators.     

Modeling, analysis and prediction of neutron emission spectra from acoustic cavitation bubble fusion experiments

Self-nucleated and external neutron nucleated acoustic (bubble fusion) cavitation experiments have been modeled and analyzed for neutron spectral characteristics at the detector locations for all separate successful published bubble fusion studies. Our predictive approach was first calibrated and validated against the measured neutron spectrum emitted from a spontaneous fission source (²⁵²Cf), from a Pu–Be source and from an accelerator-based monoenergetic 14.1 MeV neutrons, respectively. Three-dimensional Monte-Carlo neutron transport calculations of 2.45 MeV neutrons from imploding bubbles were conducted, using the well-known MCNP5 transport code, for the published original experimental studies of Taleyarkhan et al. [Taleyarkhan, et al., 2002. Science 295, 1868; Taleyarkhan, et al., 2004. Phys. Rev. E 69, 036109; Taleyarkhan, et al., 2006a. PRL 96, 034301; Taleyarkhan, et al., 2006b. PRL 97, 149404] as also the successful confirmation studies of Xu et al. [Xu, Y., et al., 2005. Nuclear Eng. Des. 235, 1317–1324], Forringer et al. [Forringer, E., et al., 2006a. Transaction on American Nuclear Society Conference, vol. 95, Albuquerque, NM, USA, November 15, 2006, p. 736; Forringer, E., et al., 2006b. Proceedings of the International Conference on Fusion Energy, Albuquerque, NM, USA, November 14, 2006] and Bugg [Bugg, W., 2006. Report on Activities on June 2006 Visit, Report to Purdue University, June 9, 2006]. NE-213 liquid scintillation (LS) detector response was calculated using the SCINFUL code. These were cross-checked using a separate independent approach involving weighting and convoluting MCNP5 predictions with published experimentally measured NE-213 detector neutron response curves for monoenergetic neutrons at various energies. The impact of neutron pulse-pileup during bubble fusion was verified and estimated with pulsed neutron generator based experiments and first-principle calculations. Results of modeling-cum-experimentation were found to be consistent with published experimentally observed neutron spectra for 2.45 MeV neutron emissions during acoustic cavitation (bubble) fusion experimental conditions with and without ice-pack (thermal) shielding. Calculated neutron spectra with the inclusion of ice-pack shielding are consistent with the published spectra from experiments of Taleyarkhan et al. [Taleyarkhan, et al., 2006a. PRL 96, 034301] and Xu et al. [Xu, Y., et al., 2005. Nuclear Eng. Des. 235, 1317–1324] where ice-pack shielding was present, whereas without ice-pack shielding the calculated neutron spectrum is consistent with the experimentally observed neutron spectra of Taleyarkhan et al. [Taleyarkhan, et al., 2002. Science 295, 1868; Taleyarkhan, et al., 2004. Phys. Rev. E 69, 036109] and Forringer et al. [Forringer, E., et al., 2006a. Transaction on American Nuclear Society Conference, vol. 95, Albuquerque, NM, USA, November 15, 2006, p. 736; Forringer, E., et al., 2006b. Proceedings of the International Conference on Fusion Energy, Albuquerque, NM, USA, November 14, 2006] and also that from GEANT computer code [Agostinelli, S., et al., 2003. Nuclear Instrum. Methods Phys. Res. A 506, 250–303] predictions [Naranjo, B., 2006. PRL 97 (October), 149403] in which ice shielding was also absent.The results of this archive confirm for the record that the confusion and controversies caused from past reports [Reich, E., 2006. Nature (March) 060306. news@nature.com; Naranjo, B., 2006. PRL, 97 (October) 149403] have resulted from their neglect of important details of bubble fusion experiments. Results from this paper demonstrate that ice-pack shielding between the detector and the fusion neutron source, gamma photon leakage and neutron pulse-pileup due to picosecond duration neutron pulse emission effects play important roles in affecting the spectra of neutrons from acoustic inertial confinement thermonuclear fusion experiments.     

用于在线测量反应堆内热中子相对通量密度分布的闪烁体光纤探测器研究

闪烁体光纤探测器采用双探头甄别中子信号,利用252 Cf裂变源对探测器系统进行了测试,并与3 He计数管的计数进行了对比。在启明星1#上进行了热中子相对通量密度分布的测量,结合Geant4得到的不同能量段的中子转化率及MCNPX模拟得到的反应堆中子能谱,对探测器进行了相对效率刻度,测试结果与固体核径迹探测器测得的裂变率分布进行了对比。测量结果表明,闪烁体光纤探测器对于252 Cf中子源的响应基本符合点源的衰减趋势,与3 He计数管的测量结果符合较好。在启明星1#热区测得的热中子相对通量密度分布与固体核径迹探测器测量到的结果一致,快区测得的热中子相对通量密度分布与3 He计数管的测量结果及MCNPX的模拟结果符合较好。测量结果为闪烁体光纤探测器的研究提供了较好的实验依据。     

Neutron interrogation of high-enriched uranium by a 4 MeV linac

For revealing unauthorized transport (illicit trafficking) of nuclear materials, a non-destructive method reported earlier, utilizing a 4 MeV linear accelerator for photoneutron interrogation, was further developed. The linac served as a pulsed neutron source for assay of highly enriched uranium. Produced in beryllium or heavy water by bremsstrahlung, neutrons subsequently induced fission in the samples. Delayed neutrons were detected by a newly designed neutron collar built up of 14 ³He counters embedded in a polyethylene moderator. A PC controlled multiscaler served as a time analyzer, triggering the detector startup by the beam pulse. Significant progress was achieved in enhancing the detector response, hence the sensitivity for revealing illicit material. A lower sensitivity limit of the order of 10 mg ²³⁵U was determined in a 20 s measurement time with a reasonable amount of beryllium (170 g) or of heavy water (100 g) and a mean electron current of 10 μA. Sensitivity can be further enhanced by increasing the measurement time.     

Accurate Estimation of Subcriticality Using Indirect Bias Estimation Method, (II)

Subcriticalities were estimated by applying the Indirect Bias Estimation Method to subcritical experiments on a light-water moderated/reflected low-enriched UO₂ lattice cores. Two measurable values, prompt neutron time-decay constant and spatial-decay constant were calculated using MCNP 4A and JENDL-3.2. With these calculation errors, the biases in calculated reactivity were derived from the Indirect Bias Estimation Method. The differences between the calculated and measured spatial-decay constants were more or less at the same extent of experimental errors. These results show that the accuracy of subcriticality estimation of MCNP 4A and JENDL-3.2 ranges within the uncertainty which can be achieved by the exponential experiment. The differences between calculated and measured prompt neutron decay constants derive significant biases in calculated reactivity. The subcriticalities were estimated by using the effective multiplication factors adjusted based on these biases in calculated reactivity.     

Design of epicadmium-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate an epicadmium-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one epicadmium covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model which has an epicadmium-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the epicadmium-shielded channel was made. The final k_eff of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new epicadmium designed model was recorded as 1.00332. Also, a final prompt neutron lifetime of 1.5237 × 10⁻⁴ s was recorded for the new epicadmium designed model while a value of 1.5571 × 10⁻⁷ s was recorded for the original MCNP design of the GHARR-1. The neutron energy causing fission for the original MCNP design of the GHARR-1 was 1.3533 × 10⁻² MeV while that of the new epicadmium designed model was 1.3513 × 10⁻² MeV.