Benchmark experiment on the passage of 14-MeV neutrons in models of an iron and iron-water shielding with slits

Conclusions  The experimental investigations and calculations of slit configurations in iron and iron—water shielding enable the following conclusions to be drawn.

Measurement of neutron spectra through composed material block bombarded with D-T neutrons

A 2-dimensional composed material assembly made of the iron and hydric block has been established. The neutron spectra from the assembly bombarded with 14-MeV neutrons at neutron generator have been obtained using the proton recoil technique with a stillbene detector. The detector positions were selected at the 60°, 120°, 180° on the surface of the iron spherical shell. The background neutron spectra consisted of background and room return radiation were subtracted with combination of methods of experimental shielding and MCNP calculation. The uncertainty of results was 6.3-7.4%. The experiment results were analyzed and simulated by MCNP code and two data library. The difference is integral neutron flux (background neutron subtracted) of measured results greater than calculations with maximum of 21.2% in the range of 1-16 MeV.     

Shielding design and analyses of KIPT neutron source facility

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of a neutron source facility. An electron accelerator drives a sub-critical facility (ADS) is used for generating the neutron source. The facility will be utilized for performing basic and applied nuclear researches, producing medical isotopes, and training young nuclear specialists. Monte Carlo code MCNPX has been utilized as the major design tool for the design, due to its capability to transport electrons, photons, and neutrons at high energies. However the ADS shielding calculations with MCNPX need enormous computational resources and the small neutron yield per electron makes sampling difficulty for the Monte Carlo calculations. The high energy electrons (E > 100 MeV) generate very high energy neutrons and these neutrons dominant the total radiation dose outside the shield. The radiation dose caused by high energy neutrons is ∼3-4 orders of magnitude higher than that of the photons. However, the high energy neutron fraction within the total generated neutrons is very small, which increases the sampling difficulty and the required computational time. To solve these difficulties, the user subroutines of MCNPX are utilized to generate a neutron source file, which record the generated neutrons from the photonuclear reactions caused by electrons. This neutron source file is utilized many times in the following MCNPX calculations for weight windows (importance function) generation and radiation dose calculations. In addition, the neutron source file can be sampled multiple times to improve the statistics of the calculated results. In this way the expensive electron transport calculations can be performed once with good statistics for the different ADS shielding problems. This paper presents the method of generating and utilizing the neutron source file by MCNPX for the ADS shielding calculation and similar accelerator facilities, and the accurate radiation dose analyses outside the shield using modest computational resources.     

Radiation Dosage Due to Neutrons Produced in a Vehicle by Protons in the Trapped Radiation Belt

The rate of production of neutrons by geomagnetically trapped protons incident on a vehicle was measured by a neutron counting system carried into the trapped radiation belt by a pod flow piggy back on an Atlas rocket on December 19, 1961. The flux of neutrons produced by radiation belt protons incident on the pod was determined to be at least 700 neutrons/(cm2 sec); the actual value depends somewhat on the energy spectrum of the neutrons. This flux was estimated to be equivalent to a dose rate in tissue of 0.10 rems/hr. On the basis of proton flux measurements made in the radiation belt by Freden and White, a calculation was made of the tissue dosage which would have been received in the same environment directly from protons. These calculations were made by obtaining a numerical integration of the dE/dx times RBE times flux product over the entire energy spectrum. The total dose calculated amounted to 2.78 rems/ hr. Further calculations were made to estimate the dose rates which would have been received by tissue in the same environment but with varying amounts of shielding around the vehicle. The proton dose is, of course, reduced by the shield but the neutron dose actually increases as the shielding thickness is increased. It is seen that the neutron dose equals the proton dose at .3 rems/hr. when aluminum shielding of 2.6" surrounds the vehicle and it exceeds the proton dose with thicker shielding.     

Relation between material configuration and radiation shielding capability - application of radiation shielding characteristic functions

For the purpose of finding a principle for material configuration which an ideal radiation shielding in slab geometry should obey, radiation energy dependence of material configuration is studied. In the course of study, radiation shielding capability for each system of different material configuration is evaluated by using radiation shielding characteristic functions defined as dose rates of transmitted radiations in response to isotropic incidence of radiations to the slab shield with pulse-like narrow energy distributions.In shielding neutrons by steel and water layers, recommendable material configuration depends on energy distribution of incident neutrons; all steel layers should be located in the source side of all water layers, if incident neutron energies are above 5 MeV: either homogeneous array of steel and water layers or above mentioned material configuration is recommendable, if incident neutron energies are between 2 MeV and 5 MeV: all water layers should be located in the source side of all steel layers, if incident neutron energies are below 2 MeV or incident neutrons have energy spectrum of fission neutrons.Above recommendation can be understood well by considering both energy dependence of neutron cross sections of each material and the maximum amount of energy degradation at elastic scattering in each material.In designing a neutron shield, shielding of secondary gamma rays is important as well as neutron shielding. This importance is demonstrated for several types of actual cask walls which are composed of many material layers by using the characteristic functions of neutrons and gamma rays for cask walls.     

可控中子源中子-伽马测井仪中屏蔽结构的优化

本文设计了一种使用遗传算法调用蒙特卡罗计算软件MCNP的方案,用以优化设计中子-伽马测井仪中的屏蔽结构。以D-D聚变中子源和BGO探测器为研究对象,以最小化探测器内的辐照本底为优化目标,设计出了3种不同厚度的屏蔽结构。模拟结果表明,这些屏蔽结构具有优异的屏蔽性能,可有效地降低探测器中的辐射本底。     

The neutron background at the surface of the earth

Conclusions The calculations that were made have shown that the production of neutrons by cosmic rays in the upper layers of the earth's crust can exert a strong influence on the slow neutron flux at sea level.In order for us to compare calculation with experiment, it would have been necessary to compute the slow neutron flux at surface of the ground which arises from neutrons produced in the atmosphere. For such an estimate, it is necessary to know the intensity and spectrum of atmospheric neutrons and the possible distortions of them at the surface of the earth. We did not have such data available, and therefore such an estimate was not made in this paper.However, it is clear that the contribution of the neutrons which are produced in granite by cosmic rays to the total neutron flux above its surface is considerably greater than in the case of water. Therefore, one can suppose that the difference in slow neutron fluxes above water and soil surfaces are caused to a considerable degree by the cosmic ray production of neutrons in the upper layers of the earth's crust.Translated from Atomnaya Énergiya, Vol. 17, No. 6, pp. 492–496, December, 1964     

Neutron damage calculations in Cu,Nb and Au to 32 MeV: Application to sputtering and deuteron-breakup neutron sources

Primary recoil distributions and specific damage energies have been computed for high energy deuteron-breakup neutrons in Cu, Nb and Au. The calculations are based on theoretical neutron cross sections and consider in particular a d-Be spectrum broadly peaked at 15 MeV with some neutrons above 30 MeV. The theoretical results are similar to corresponding calculations for monoenergetic 15-MeV neutrons and are in good agreement with range measurements of (n, 2n) recoils generated by high energy d-Be neutrons in Nb and Au. The calculations are also consistent with recent d-Be neutron sputtering experiments in Nb and Au and demonstrate the usefulness of deuteron-breakup neutron sources for simulating fusion neutron effects.     

Use of concretes for high-temperature shielding of nuclear reactors

The authors have studied the possibility of using chromite and chomotte heat-resistant concretes for the thermal shields of reactors. They observe neutron fluxes of various intensities (up to 10¹³ neutrons/cm²·sec, with spectrum similar to fission spectrum), absorbed by shields of these materials. They compute the transmission of neutrons and of fluxes of gamma quanta and the heat emission in the shielding. They calculate the temperatures in the shielding for various neutron fluxes, concrete thicknesses and cooling conditions. They perform a statistical calculation of the temperature stresses for shielding constructed of heat-resistant ferroconcrete.It was established that nuclear reactor shields can be made from heat-resistant ferroconcrete when the neutron fluxes on the concrete are up to 10¹³ neutrons/cm²·sec, for temperatures up to 1000–1100° C and temperature differences of up to 900° C.Translated from Atomnaya Énergiya, Vol. 19, No. 6, pp. 524–529, December, 1965Report read by G. I. Budker at the International Conference on High-Energy Accelerators (Frascati, Italy).     

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.     

声核中子测量的屏蔽设计

通过测定声空化核效应实验室各测量点的中子注量率,了解实验室墙壁和地面对出射中子的散射,选定散射中子相对较弱的位置作为声核中子测量点。利用SHIELD程序模拟不同材料的中子屏蔽效果,选用4cm铁和20cm含硼石蜡组成屏蔽体,以降低中子本底。测定影屏蔽及影屏蔽结合BF₃正比计数管环绕屏蔽两种方式下的散射修正因子F_s,提出以统计显著性增量S.S.I≥3/[KF(]F_s[KF)]作为超声中子计数相对于非超声中子计数的增量ΔC是否具有统计意义的判据。     

Accelerator driven subcritical system as a future neutron source in Kyoto University Research Reactor Institute (KURRI) —Basic study on neutron multiplication in the accelerator driven subcritical reactor—

A basic study on the nuclear characteristics in the accelerator driven subcritical reactor (ADSR) was performed through a series of neutronics calculations in view of a future neutron source in Kyoto University Research Reactor Institute (KURRI) for the joint use program among researchers of Japanese universities. In this series of calculations, it was assumed that three kinds of monoenergetic neutrons were isotropically generated at the center of spherical and homogeneous cores with different moderator-to-fuel volume ratios in order to examine the spectrum mismatching effect between injected neutrons and fission neutrons born in the subcritical core. The results of calculations clearly showed the spectrum mismatching effect on the neutron multiplication in the ADSR.     

The design of “multiple energies from one source” in D-T neutron tube

To obtain multiple monoenergetic neutron sources and realize the on-site calibration of radiation monitoring equipment for nuclear-involved places,the structural characteristics and neutron source features of D-T neutron tube were analyzed;Monte Carlo method was adopted to simulate the effect of interaction between typical materials and different energy neutrons;multilayered shielding materials were combined and optimized to acquire the optimal scheme to shield the neutron sources from the neutron tube.On the base,a tapered alignment filtration construction was designed and Monte Carlo method was employed to simulate the effect of alignment construction.The result showed that the tapered alignment filtration construction can create monoenergetic neutrons including14.1 MeV,0.18 MeV and thermal neutrons and demonstrated good monochrome performance which provides multiple monoenergetic sources for the on-site calibration.     

Measurement of neutron tissue dose outside reactor shielding

RUS instruments, developed by the authors, are described that enable one to measure the flux and tissue dose rate of intermediate neutrons, which make a significant contribution to neutron tissue dose outside reactor shielding. The neutron dose composition was investigated in experiments at the IRT-1000 reactor, and it was shown that it depends essentially on shielding composition. It was established that the neutron tissue dose computed from readings taken with the RPN-1 instrument were actually too low by a factor amounting to one and one half outside water shielding and to five outside concrete shielding.Translated from Atomnaya Énergiya, Vol. 15, No. 5, pp. 386–393, November, 1963     

中子残余应力谱仪静态屏蔽体表面辐射剂量率初算

中子残余应力谱仪静态屏蔽体主要用于对谱仪装置的附加闸门、中子导管等组件的辐射剂量的屏蔽,使装置操作人员可以安全地在装置周围活动。通过MCNP5程序对谱仪装置静态屏蔽体的屏蔽能力进行了计算,可为该方案的改进、优化提供依据,以便最终制造出满足辐射剂量要求的屏蔽体。     

贫化铀用于核动力堆防护的屏蔽性能与放射性毒性计算分析

针对贫化铀的γ射线屏蔽进行了实验与模拟计算验证。构建了核动力压水堆屏蔽模型,模拟输出的屏蔽层内中子能谱与实际能谱分布较为一致。采用蒙特卡罗程序与燃耗计算程序相耦合的方法,模拟计算了贫化铀在不同位置处中子、γ混合辐射场中的综合屏蔽性能,并与铅作为屏蔽材料进行了对比分析。模拟计算了屏蔽层中子辐照贫化铀40 a后的活化和裂变产物,分析了材料辐照前后年摄入量限值（ALI）定义下的放射性毒性,结果表明,新增二次产物对放射性毒性影响不大。      

Preliminary spectrum shifter design for intermediate energy nuclear data benchmark experiments with DT neutrons

Integral benchmark experiments with DT neutrons are not always sufficient for nuclear data benchmarking in the MeV region, below 10 MeV. A neutron spectrum shifter, which will be placed between a sample and a DT neutron source, is effective to moderate DT neutrons incident to the sample. In order to estimate effects of the spectrum shifter, the ratio of the contribution of 14 MeV neutrons in the leakage neutron and gamma-ray spectra was calculated with MCNP-4C for an experimental configuration at FNS of JAEA, Japan. The calculations were carried out for a Li₂TiO₃ sample with a Be, D₂O, or ⁷LiD spectrum shifter. It was found out that the Be shifter was superior to others and the Be shifter was effective to decrease the contribution of 14 MeV neutrons especially for secondary gamma-ray spectrum measurements.     

NXcom – A program for calculating attenuation coefficients of fast neutrons and gamma-rays

This work is concerned with a construction and use of NXcom computer program for calculating the removal and attenuation coefficients of transmitted fast neutrons and γ-rays, respectively, through mixtures, composites, concretes and compounds. The program uses only one input data file for neutrons and γ-rays calculations. For γ-ray attenuation, the program predictions were tested by comparing them with the well-known WinXcom program results and an excellent agreement was noticed. Also, it has been used for calculating the values of macroscopic effective removal cross-sections Σ_R (cm⁻¹) for five new published polyamide and anhydride composites designed for shielding mixed neutron and γ-rays. The obtained values for Σ_R using the program and the reported attenuation thicknesses which were based on the Monte Carlo N-Particle (MCNP) code showed the same trend. The NXcom program can be used as a preliminary effective tool for testing the shielding material against fast neutrons and γ-rays.     

Spectra of fast reactor neutrons after passing through graphite,lead, and iron

By means of a fast neutron scintillation spectrometer with one hydrogen-containing detector, the spectra of fast reactor neutrons after passing through various thicknesses of lead, graphite, and iron were measured in the range 0.7–11 MeV. The measurements were carried out in a water-moderated water-cooled experimental reactor in barrier geometry. The results of the experiments enabled us to determine the deformation of the neutron spectrum in relation to the penetration through the layers of material, and to calculate the relaxation lengths and the removal Cross sections. These quantities were punished earlier for fission spectrum neutrons in the energy range 0.7^–3 MeV.Translated from Atomnaya Énergiya, Vol. 16, No. 1, pp. 32–40, January, 1964     

基于SuperMC对结构缝隙影响中子屏蔽的研究

本文基于Monte Carlo粒子输运计算程序SuperMC,计算了四种含硼聚乙烯(B-PE)结构缝隙对两种谱中子的衰减倍数。为了便于比较不同结构缝隙对中子屏蔽性能的影响,统一与相同厚度无缝隙材料相比得到中子衰减倍数相对减小量,并在相同条件下对计算结果进行了实验验证。结果表明:对于厚度6 cm的B-PE材料,斜缝结构的快中子衰减倍数相对减小量为直缝结构的1/8,斜缝结构的慢化中子衰减倍数相对减小量为直缝结构的1/3,斜缝结构对中子屏蔽产生的负面影响最小。