Electrochemical synthesis of boron-containing coatings on Mg alloy for thermal neutron shielding

The work provides the results of the one-step formation of boron-containing coatings on an Mg–Mn–Ce alloy by plasma electrolytic oxidation. The results of studies of the composition, structure and morphology of heteroxide coatings are presented. It was established that the boron is contained in the coating mainly in the form of B or B_(2)O_(3). The introduction of B changes the color of coatings, and also helps to increase their porosity. The method of determining the full cross section of the interaction of thermal neutron absorption efficiency by samples material using the installation of neutron-activation analysis based on ^(252)Cf was developed. It was shown that the introduction of boron into the formed coatings allows to increase the macroscopic cross-section of the interaction of samples with thermal neutrons by 3.8 times. This effect opens the potential for the use of synthesized material in the field of nuclear technologies and aerospace industry.

Static Structural Analysis for a Neutron Shielding Block in ITER

The ITER neutron shielding blocks are located between the outer shell and the inner shell of the vacuum vessel to provide neutron shielding. Considering the combined loads acting on the shielding blocks during ITER plasma operation, the structure of the shielding blocks must be evaluated. Using the finite element method with ANSYS analysis software, static structural analysis is performed, including elastic analysis and limit analysis for one typical shielding block. The evaluated results based on RCC-MR code show that the structure of this shielding block can meet the design requirement.

Limit Analysis for the Mechanical Structure of the ITER Neutron Shielding Block

The ITER neutron shielding blocks are located between the inner shell and the outer shell of the vacuum vessel （VV） with the main function of providing neutron shielding. Conskicring the combined loads of the shielding blocks during the plasma operation of the ITER, limit analysis for one typical ferromagnetic （FM） shielding block has been performed and the structural design has bccn evaluated based on the American Society of Mechanical Engineers （ASME） criterion and European standards. Results show that the collapse load of this shielding block is three times the specified load, which is much higher than the design requirement of 1.25. The structure of this neutron shielding block has a sufficient safety margin.

Numerical Analysis on Neutron Shielding Structure of ITER Vacuum Vessel

The neutron shielding component of ITER （International Thermonuclear Experimental Reactor） vacuum vessel is a kind of structure resembling a wall in appearance. A FE （finite element） model is set up by using ANSYS code in terms of its structural features. Static analysis, thermal expansion analysis and dynamic analysis are performed. The static results show that the stress and displacement distribution are allowable, but the high stress appears in the junction between the upper and lower parts. The modal analysis indicates that the biggest deformation exists in the port area. Through modal superposition, the single-point response has been found with the lower rank frequency of the acceleration seismic response spectrum. But the deformation and the stress values are within the permissible limit. The analysis results would benefit the work in the next step and provide some reference for the implementation of the engineering plan in the future.

B4C/NRL flexible films for thermal neutron shielding

Boron carbide/natural rubber latex(B_4 C/NRL)flexible films were prepared via dip-molding with B_4 C content in the range of 5–55 wt% for thermal neutron(0.0253 e V) shielding. B_4 C was well dispersed in NRL according to microscopic observation. Both the inside and outside surfaces of the film were smooth. For B_4 C/NRL flexible films, the minimum elongation at break was greater than 600%, the minimum tensile strength was greater than 12 MPa, and the hardness was in the range of 35–55 HA,which were suitable for preparing flexible wearable products. The attenuation efficiencies of the B_4 C/NRL flexible films for thermal neutrons were also calculated. The B_4 C/NRL flexible films exhibit good attenuation effect for thermal neutrons.

Calculation of photon shielding properties for some neutron shielding materials

The objective of the present study is to calculate photon shielding parameters for seven polyethylene-based neutron shielding materials. The parameters include the effective atomic number(Z_(eff)), the effective electron density(N_(eff)) for photon interaction and photon energy absorption,and gamma-ray kerma coefficient(kc). The calculations of Z_(eff)are presented as a single-valued and are energy dependent. While Z_(eff)values were calculated via simplistic powerlaw method, the energy-dependent Z_(eff)for photon interaction(Z_(PI-eff)) and photon energy absorption(Z_(PEA-eff)) are obtained via the direct method for energy ranges of 1 keV–100 GeV and 1 keV–20 Me V, respectively. The kccoefficients are calculated by summing the contributions of the major partial photon interactions for energy range of 1 keV–100 MeV. In most cases, data are presented relative to pure polyethylene to allow direct comparison over a range of energy. The results show that combination of polyethylene with other elements such as lithium and aluminum leads to neutron shielding material with more ability to absorb neutron and crays. Also, the kerma coefficient first increases with Z of the additive element at low photon energies and then converges with pure polyethylene at energies greater than 100 keV.

Effects due to a Pu-C source on a HPGe detector and the corresponding neutron shielding

A gamma spectrum of a Pu-C source is measured using a p-type HPGe detector, whose three peaks （full energy, single-escape and double-escape peak） can be used as a calibration source for the beam energy measurement system of BEPCII. The effect of fast neutron damage on the energy resolution of the HPGe detector is studied, which indicates that the energy resolution begins to deteriorate when the detector is subject to 2×107 n/cm^2 fast neutrons. The neutron damage mechanism and detector repair methods are reviewed. The Monte Carlo simulation technique is utilized to study the shielding of the HPGe detector from the fast neutron radiation damage, which is of great significance for the future commissioning of the beam energy measurement system.

Effect of Gd on neutron absorption properties and electrochemical corrosion behavior of Zr-Gd alloy in boiling concentrated HNO3

A Zr-Gd alloy with neutron poisoning properties and resistance to boiling concentrated HNO3 corrosion was developed based on a corrosion-resistant Zr-702 alloy to meet the demand for neutron shielding in the closed-loop treatment of spent fuel and the nuclear chemical industry.In this study,1 wt.%,3 wt.%,5 wt.%,7 wt.%,and 9 wt.%Zr-Gd alloys were designed and fabricated with Zr-702 as the control element.The electrochemical behavior of the Zr-Gd alloys in boiling concentrated HNO3 was investigated,and the neutron shielding effect on plate thickness and Gd content was simulated.The experimental results demonstrate that the corrosion resistance of the alloy decreased slightly before~7-9 wt.%with increasing Gd content;this is the inflection point of its corrosion resistance.The alloy uniformly dissolved the Gd content that could not be dissolved in the Zr lattice,resulting in numerous micropores on the passivation coating,which deteriorated and accelerated the corrosion rate.The MCNP simulation demonstrated that when the Gd content was increased to 5 wt.%,a 2-mm-thick plate can shield 99.9%neutrons;an alloy with a Gd content≥7 wt.%required only a 1-mm-thick plate,thereby showing that the addition of Gd provides an excellent neutron poisoning effect.Thus,the corrosion resistance and neutron shielding performance of the Zr-Gd alloy can meet the harsh service requirements of the nuclear industry.

Study on the gamma rays and neutrons energy response optimization of a scintillating fiber detector for EAST with Geant4

A new scintillating fiber detector inside magnetic shielding tube was designed and assembled for use in the next round of fusion experiments in the experimental advanced superconducting tokamak to provide D–T neutron yield with time resolution.In this study,Geant4 simulations were used to obtain the pulse height spectra for ideal signals produced when detecting neutrons and gamma rays of multiple energies.One of the main sources of interference was found to be low-energy neutrons below 10–5 MeV,which can generate numerous secondary particles in the detector components,such as the magnetic shielding tube,leading to high-amplitude output signals.To address this issue,a compact thermal neutron shield containing a 1-mm Cd layer outside the magnetic shielding tube and a 5-mm inner Pb layer was specifically designed.Adverse effects on the measurement of fast neutrons and the shielding effect on gamma rays were considered.This can suppress the height of the signals caused by thermal neutrons to a level below the height corresponding to neutrons above 4 MeV because the yield of the latter is used for detector calibration.In addition,the detector has relatively flat sensitivity curves in the fast neutron region,with the intrinsic detection efficiencies(IDEs)of approximately 40%.For gamma rays with energies that are not too high(<8 MeV),the IDEs of the detector are only approximately 20%,whereas for gamma rays below 1 MeV,the response curve cuts off earlier in the low-energy region,which is beneficial for avoiding counting saturation and signal accumulation.

Design of a shielding collimator device for a small-angle monoenergetic neutron source

To obtain a small-angle monoenergetic neutron source,a shielding collimator device is designed for the neutron source generated by a neutron tube.The device is divided into the collimator and the capture cave.The collimator is made of three layers of stainless steel and borated polyethylene and is used to constrain neutrons in a small angle.The capture cave is used to increase the number of times neutron inelastic scattering occurs in the opposite direction of the radiation field,thereby reducing the proportion of scattered neutrons in the radiation field.Material thickness,aperture size,and the optimum structure of the capture cave were simulated using MCNP.The design features a neutron emission angle within a range of 3° and neutron fluxes in the radiation field,which are higher by two orders of magnitude than those outside the radiation field.This research has practical value for the generation of monoenergetic small-angle neutron sources and neutron applications.

Characterization of a new shielding rubber for use in neutron–gamma mixed fields

A variety of formulations was investigated for the fabrication of an appropriate shielding rubber to be used in neutron–gamma mixed fields. Having considered the required mechanical properties together with tungsten as the gamma-ray absorbing element, calculations with MCNPX 2.6 code confirmed that the incorporation of 5 weight percentage(wt%) of boron carbide exhibited the best performance as a thermal neutron absorber. A series of both experimental and simulation results are provided for comparison.

Rem-meter correction factor for measuring high energyneutrons outside concrete shielding

Correction factors of both Rem-meters, the 10 inch diameter single-sphere Remmeter and the standard A-B Rem-meter, were estimated for measuring high energy neutron dose equivalent outside a concrete shielding wall and the effects that the emitted neutron spectra become remarkably "harder" penetrated through a concrete shielding wall, and the energy response of the Rem-meter were taken in account. The estimated results could be applied in the measurement of neutron dose equivalent for the intermediate energy heavy ion reactions to avoid the difficulty induced by the energy response of the Rem-meters.

环保型钆硼柔性热中子吸收材料特性研究

本文以聚苯乙烯-乙烯-丁二烯-苯乙烯(SEBS)为基体、以Gd_(2)O_(3)、B_(4)C为功能填料,研制了一种柔性热中子吸收材料。SEM测试结果表明,不同含量的材料中Gd_(2)O_(3)和B_(4)C微米颗粒分布相对均匀;XRD和FT-IR测试结果表明,Gd_(2)O_(3)、B_(4)C与基体SEBS未发生化学反应,属于物理性混合。通过实验和蒙特卡罗模拟进行热中子屏蔽性能验证,对照组材料的实验透射率为31.97%~35.35%,实验组材料的实验和模拟透射率分别为32.11%~36.54%和26.26%~31.31%。对实验组进行了热中子面透射率均匀性测试,结果表明,(10%Gd_(2)O_(3)+40%B_(4)C)/SEBS和(30%Gd_(2)O_(3)+40%B_(4)C)/SEBS材料的平均透射率分别为34.34%和31.60%,所有采样点的绝对偏差在±0.5%以内,标准差为0.33%和0.26%,离散系数为0.0096和0.0082。该柔性材料有效弥补了传统刚性射线屏蔽材料的不足,在核设施异形复杂结构表面包覆防护和可穿戴辐射防护服领域具有潜在应用价值。

主动式多重性方法对黄饼中铀定量的模拟研究

黄饼材料中铀的准确定量是后续处理工艺选择的关键,文中在主动式多重性方法的基础上,提出了一种通过记录分析中子源诱发238U裂变信息,进行铀定量的方法。但由于黄饼材料自身存在中子自屏蔽效应以及含水量的差异,导致定量结果存在偏差。为了进一步提高定量准确性,使用MCNP(Monte Carlo N-Particle Transport)结合MATLAB程序优化选择了241Am-Be源作为激发源;另外,通过对不同质量及含水量系列化样品的模拟发现:铀定量误差主要来自于泄漏增殖因子ML与增殖因子M差距的不匹配。通过MCNP模拟获取M随铀质量变化规律的曲线后,根据样品净含量选择合适的增殖因子M,再根据二重计数率D进行定量计算,获得铀定量的相对误差小于5%;含水量的变化带来的中子自屏蔽效应对多重计数率影响较大,通过S0/Si与D0/Di的关系对二重计数率D进行修正后再进行计算,铀定量的相对误差能够控制在10%左右;该研究对中子多重性方法在黄饼生产与测量中的应用推广具有重要的参考价值。

几种屏蔽材料的中子屏蔽性能实验研究

先进固态模块化可移动式反应堆核电系统的发展和新型固态燃料的应用,对中子屏蔽材料类型、设备质量和设计紧凑程度都提出了更高的要求。针对设备轻量化的需求,定制特殊尺寸的屏蔽材料装填筒,以粉末松散填充的方式制备了碳化硼、氢化钛和氢化锆三种屏蔽测试材料,置于单管中子射线检测系统中进行屏蔽实验和数据采集。基于蒙特卡洛的粒子输运模拟程序(RMC)模拟^(241)Am-Be中子源系统下的中子屏蔽,并对其进行数据拟合。实验结果表明,碳化硼材料在屏蔽实验中性能显著优于氢化钛和氢化锆,蒙特卡洛模拟得到的数据与单管中子射线检测系统得到的实验数据拟合程度较高。该研究建立的计算、实验及对比验证方法能够用于指导新型中子屏蔽材料的性能研究。

改性高浓度硼酸溶液对蛇纹石屏蔽混凝土力学性能与中子屏蔽性能的影响

蛇纹石屏蔽混凝土是一种长期应用于高温条件下仍能保持高结晶水含量的中子屏蔽材料,但其密实度低,缺乏有效吸收热中子的硼元素。作为硼元素的来源,硼酸是生产屏蔽混凝土的有效材料,但在以水泥为主的水化体系中,硼酸的掺入会抑制水泥水化过程,导致材料强度及耐久性劣化。本研究通过改性高浓度硼酸溶液,诱导水泥浆体中游离硼酸盐生成高硼含量与结晶水含量的戈硼钙石,增加体系中硼元素含量的同时解除硼酸对水泥水化的抑制作用,研究改性高浓度硼酸溶液对蛇纹石屏蔽混凝土工作性能、力学性能、高温残余抗压强度、微观结构与屏蔽性能的影响。结果表明:掺加改性高浓度硼酸溶液对蛇纹石屏蔽混凝土的凝结时间、常温力学性能、高温残余抗压强度均无不良影响。掺加改性高浓度硼酸溶液后,蛇纹石屏蔽混凝土在400℃高温残余抗压强度提高了43.5%,常温下半值层降低了31.7%,350℃保温至恒重后半值层降低了21.4%,对中子射线的屏蔽能力增强。

控制棒对核动力航天器辐射屏蔽特性的影响

针对控制棒及其导向管贯穿阴影式屏蔽体的情况,本文进行了核动力航天器全场辐射屏蔽特性的分析,并提出局部屏蔽的优化方案。利用蒙特卡罗方法在核动力航天器全场范围内对中子与光子进行输运计算,求解得到控制棒吸收体处于不同位置时的中子注量率与光子剂量分布情况,分析了控制棒及其导向管对阴影屏蔽体的屏蔽特性及航天器全场辐射特性的影响,并针对辐射薄弱点提出了局部屏蔽优化方案,对优化后的方案进行了计算验证。结果表明:控制棒导向管垂直贯穿屏蔽体会导致仪器平台辐射剂量超限值,极端工况下光子剂量达到限值的10倍,中子注量达到了限值的1.8倍。利用局部屏蔽体可以有效消除控制棒导向管带来的辐射剂量超限问题,能在控制质量增幅的情况下保证仪器平台所处区域的辐射安全。

碳化硼/铅/聚乙烯复合材料的热稳定性及屏蔽性能

制备了质量分数3%、4%、5%及6%碳化硼以和不同厚度的碳化硼/铅/聚乙烯复合材料,并对碳化硼/铅/聚乙烯复合材料的热稳定性和屏蔽性能进行了研究。经测试可知:当不同试样处于同一厚度时,添加质量分数6%碳化硼复合材料的高温基础性能较好;热重测试表明,该试样的质量损失较低,有效保持了其基础性能。同时,添加质量分数为6%碳化硼的复合材料的应用效果最佳,具有较好的粒子屏蔽性能。该结果为碳化硼/铅/聚乙烯复合材料的制备和性能评估提供了可靠的依据。

百千瓦级空间核反应堆屏蔽优化研究

屏蔽体尺寸和重量对空间核反应堆和核动力航天器性能有着重要影响,因而屏蔽设计优化是空间核动力系统设计的关键。本文以JIMO项目反应堆为对象,在铍-碳化硼-钨-氢化锂分层组合屏蔽方案的基础上,考虑到辐照剂量的径向分布,采用蒙特卡罗方法计算了负载处辐照剂量和氢化锂中子剂量,分析了屏蔽设计原理,并提出了分步优化方法以实现屏蔽优化。根据结果分析,调整了铍和碳化硼的厚度比例、钨半径及布置位置,获得了优化的屏蔽方案,在满足屏蔽要求的基础上质量减少了98.41 kg。提出的屏蔽方案及设计流程可为空间核电源屏蔽设计优化提供参考。