Design of the thermal neutron detection system for CJPL-Ⅱ

A low background thermal neutron flux detection system has been designed to measure the ambient thermal neutron flux of the second phase of the China Jinping Underground Laboratory(CJPL-Ⅱ),right after completion of the rock bolting work.A ~3He proportional counter tube combined with an identical ~4He proportional counter tube was employed as the thermal neutron detector,which has been optimised in energy resolution,wall effect and radioactivity of construction materials for low background performance.The readout electronics were specially designed for long-term stable operation and easy maintenance in an underground laboratory under construction.The system was installed in Lab Hall No.3 of CJPL-Ⅱ and accumulated data for about 80 days.The ambient thermal neutron flux was determined under the assumption that the neutron field is fully thermalized,uniform and isotropic at the measurement position.     

用于BNCT-SPECT的帕维亚TRIGA MARK II反应堆热柱优化设计

针对坐落于意大利帕维亚大学的TRIGA Mark II反应堆热柱结构进行优化设计,从而满足面向硼中子俘获治疗(BNCT)的单光子发射计算机断层成像(SPECT)研究要求。为提高计算效率并减小统计误差,对比分析使用SSW/SSR方法与直接使用反应堆为源项时热柱内照射位置处中子能谱,其结果基本一致,从而验证了SSW/SSR方法的可靠性。为在该反应堆开展BNCT中SPECT实验,热柱中子束需准直为笔形束。对比分析四种热柱优化方案下束流口处及探测器处热中子和光子通量:40 cm长石墨(射束口5 cm3 cm);0.5 cm厚硼包裹40 cm长石墨(射束口5 cm3 cm);30 cm长天然锂聚乙烯(射束口直径4 cm);30 cm长天然锂聚乙烯(20 cm长射束口直径5 cm,5 cm长射束口直径4 cm,5 cm长射束口直径2 cm)。结果显示,射束口处热中子通量分别为1.05108,2.52107,6.08107和5.10 107 #/(cm2s)。综合考虑中子准直效果及光子污染,方案三具有最优性能。为后续进行BNCT-SPECT理论和实验研究提供了基础,从而有效促进BNCT剂量准确评估方法的研究进程。     

Assessment of fast and thermal neutron ambient dose equivalents around the KFUPM neutron source storage area using nuclear track detectors

A set of five ²⁴¹Am–Be neutron sources are utilized in research and teaching at King Fahd University of Petroleum and Minerals (KFUPM). Three of these sources have an activity of 16 Ci each and the other two are of 5 Ci each. A well-shielded storage area was designed for these sources. The aim of the study is to check the effectiveness of shielding of the KFUPM neutron source storage area. Poly allyl diglycol carbonate (PADC) Nuclear track detectors (NTDs) based fast and thermal neutron area passive dosimeters have been utilized side by side for 33 days to assess accumulated low ambient dose equivalents of fast and thermal neutrons at 30 different locations around the source storage area and adjacent rooms. Fast neutron measurements have been carried out using bare NTDs, which register fast neutrons through recoils of protons, in the detector material. NTDs were mounted with lithium tetra borate (Li₂B₄O₇) converters on their surfaces for thermal neutron detection via and nuclear reactions. The calibration factors of NTD both for fast and thermal neutron area passive dosimeters were determined using thermoluminescent dosimeters (TLD) with and without a polyethylene moderator. The calibration factors for fast and thermal neutron area passive dosimeters were found to be 1.33 proton tracks and 31.5 alpha tracks , respectively. The results show variations of accumulated dose with the locations around the storage area. The fast neutron dose equivalents rates varied from as low as up to whereas those for thermal neutron ranged from as low as up to . The study indicates that the area passive neutron dosimeter was able to detect dose rates as low as 7 and from accumulated dose for thermal and fast neutrons, respectively, which were not possible to detect with the available active neutron dosimeters.     

Passive neutron area monitor with pairs of TLDs as neutron detector

A passive neutron area monitor has been designed using Monte Carlo methods; the monitor is a polyethylene cylinder with pairs of thermoluminescent dosimeters (TLD600 and TLD700) as thermal neutron detector. The monitor was calibrated with a bare and a thermalzed ²⁴¹AmBe neutron sources and its performance was evaluated measuring the ambient dose equivalent due to photoneutrons produced by a 15 MV linear accelerator for radiotherapy and the neutrons in the output of a TRIGA Mark III radial beam port.     

热中子分析探雷的中子源慢化装置设计与优化

与传统的地雷探测技术相比,热中子分析(Thermal Neutron Analysis,简称TNA) 探雷技术具有准确率高、虚警率低和对环境适应性强的特点,但探测速度较慢,制约了其广泛应用。为了提高地雷位置处的慢热中子通量,缩短探测时间,提出了一种基于252Cf 的中子源慢化装置设计构型,主要包含中子慢化层、中子反射层、本底 屏蔽层和侧向中子吸收层4 个部分。采用数值模拟的方法比较了4种常用中子慢化(反射) 材料的性能,优选高密度聚乙烯作为慢化材料,石墨作为反射材料。同时,为了满足辐射安全要求,对屏蔽材料的结构进行了优化计算。按照设计构型搭建了TNA 探雷实验平台。在104 n/s 中子源强下优化了慢化层和反射层的厚度,测试了装置慢化效能,在107 n/s 中子源强下评估了装置辐射安全性能。结果表明,采用该装置可使地雷位置处的慢热中子通量提升11 倍以上,并能有效保障辐射安全。Compared with the traditional landmine detection methods, Thermal Neutron Analysis (TNA) landmine detection has the advantages of high accuracy, low false alarm rate and strong adaptability to the environmental change.But the long detection time restrict the wide application of this technology. In order to shorten the detection time, one possible design of neutron moderation device based on 252Cf neutron source is proposed to enhance the moderated neutronflux in mine position. The device consists of four parts, the neutron moderator, the neutron reflector, the background shield and the useless neutron absorbing layer. Then, the performance of four widely used materials in neutronics was compared with MCNP5 code, and HDPE was chosen as the neutron moderator material, graphite as the neutron reflector material. The thickness of the useless neutron absorbing layer was optimized at the same time. Finally, an experimental platform of 252Cf neutron moderation device was assembled on the basis of simulation results, and a series of experiments were carried out to optimize the geometric dimensions and evaluate the dose equivalent with two different strengths neutron source, 104 and 107 n/s. The results indicate that this device can effectively enhance the thermal neutron flux at mine position by more than 11 times and ensure the radiation safety.     

Study of the improvement of TLD cards for personal neutron dosimetry

In this work, personal thermoluminescence dosimeter (TLD) cards type of GN-6770 (holder type 8806) from Harshaw were used for personal neutron dosimetry. The response of the dosimeters has been determined in terms of the personal absorbed dose and personal dose equivalent for different neutron energy components, based on the recommendations of ICRP-60 and ICRU-49. Neutron irradiation was performed using a 5 mCi Am–Be neutron source. The TLD reader, type Harshaw 6600, was installed and calibrated for accurate neutron doses equivalent to gamma-ray doses. It was found that fast neutron doses measured by TLD (badges or cards) are in agreement with those measured by neutron TE (tissue equivalent gas) ionization chambers and neutron monitors. Thermal neutron doses measured by TLD cards were overestimated when compared with those measured by neutron monitors. Additional Cd was used to reduce thermal neutron doses to be in agreement with actual thermal doses. Other configurations for TLD crystals are also suggested for accurate thermal neutron dose measurements.     

氧化镁中子过滤器研究

核电厂反应堆乏燃料水池格架材料在生产和使用过程中需要对其中子吸收性能进行监测和检测,针对这两方面需求,研制了乏燃料水池格架B₄C_Al材料的中子吸收性能检测设备。为了降低检测过程中超热中子本底的影响,考虑采用氧化镁超热中子过滤器滤除超热中子。对10 和5 cm氧化镁单晶的中子透射率与宏观总截面进行了理论计算,对慢化体表面中子成分进行蒙特卡罗模拟计算并开展实验测量。实验结果表明,10 cm氧化镁对采用8 cm聚乙烯慢化后的252Cf中子源的中子透射率为60.16%,相对镉比值比未加10 cm氧化镁时提高了93.85%,证明常温下采用氧化镁单晶做B₄C_Al检测装置的超热中子过滤器是可行的。     

Thermal neutron calibration method using an intense neutron beam from JRR-3M

We have developed a thermal neutron calibration method using a reactor produced neutron beam in JRR-3M of the Japan Atomic Energy Agency. Neutron-induced prompt gamma ray analysis has usually been performed in this beam line. Neutron energy distributions with negligible contributions from epithermal neutrons were measured by a time-of-flight method with a chopper made of ⁶LiF powder. The thermal neutron flux was determined by a gold foil activation method. We found that the thermal neutron beam in JRR-3M was well suited for calibration, neutron detector development or neutron dosimetry.     

Study of a nTHGEM-based thermal neutron detector

With new generation neutron sources, traditional neutron detectors cannot satisfy the demands of the applications, especially under high flux. Furthermore, facing the global crisis in ~3He gas supply, research on new types of neutron detector as an alternative to ~3He is a research hotspot in the field of particle detection. GEM(Gaseous Electron Multiplier) neutron detectors have high counting rate, good spatial and time resolution, and could be one future direction of the development of neutron detectors. In this paper, the physical process of neutron detection is simulated with Geant 4 code, studying the relations between thermal conversion efficiency, boron thickness and number of boron layers. Due to the special characteristics of neutron detection, we have developed a novel type of special ceramic n THGEM(neutron THick GEM) for neutron detection. The performance of the n THGEM working in different Ar/CO_2 mixtures is presented, including measurements of the gain and the count rate plateau using a copper target X-ray source. A detector with a single n THGEM has been tested for 2-D imaging using a ~(252)Cf neutron source. The key parameters of the performance of the n THGEM detector have been obtained, providing necessary experimental data as a reference for further research on this detector.     

铝基碳化硼材料中子屏蔽性能的蒙特卡罗模拟

采用蒙特卡罗方法, 运用MCNP4C程序研究了碳化硼含量20%–40%、中子能量200 eV–15 keV、材料厚度0.3–2 cm对B₄C/Al复合材料中子屏蔽性能的影响. 结果表明: 碳化硼含量与中子透射系数呈一次线性下降关系; 同含量的碳化硼, B₄C/Al材料的中子屏蔽效果要大大优于聚乙烯碳化硼材料; 在等厚度条件下, 模拟试样B20等的中子屏蔽效果要优于水、铜、混凝土等常规屏蔽材料; 材料厚度与中子透射系数呈指数下降关系, 且单位厚度的增加对中子透射系数改变很大; 含硼量对热中子透射系数影响很大; 在热中子能区, 中子每单位能量的变化对中子透射系数改变较大; 在慢中子能区, 中子每单位能量的变化对中子透射系数改变很小. 关键词： 中子透射系数 蒙特卡罗 铝基复合材料 碳化硼     

TMSR白光中子源本底屏蔽设计

为了满足钍基熔盐堆对核数据的需求，中国科学院上海应用物理研究所自行设计并建造了紧凑型的15 MeV电子加速器驱动的白光中子源。电子直线加速器、中子产生靶以及探测器系统处于同一个实验大厅，中子/伽马射线本底较高，原有屏蔽并不能满足在低能区进行热中子物理实验测量的低中子本底需求。为了降低热中子本底，提高在热区的测量能力，需要对中子源进行局部屏蔽。根据调试运行经验以及模拟计算结果，分析了中子伽马射线本底的来源，利用MCNP5模拟计算了混凝土、铅、含硼聚乙烯对中子/伽马射线的屏蔽效果，优化设计了局部屏蔽方案。模拟计算结果显示，该屏蔽方案可将热中子本底降低三个量级，伽马本底降低两个量级。屏蔽后的实验测量结果表明，探测器处的有效热中子与本底热中子的比值达到约100:1，屏蔽效果显著，为后续在热中子能区顺利开展中子物理测量实验奠定了基础。     

基于加速器的多终端硼中子俘获治疗装置的束流整形组件设计

硼中子俘获治疗(Boron Neutron Capture Therapy, BNCT)是一种新型的精准放射治疗方法,束流整形组件(Beam Shaping Assembly, BSA)作为硼中子俘获治疗装置的重要组成部分,对于产生适用于BNCT的中子束至关重要。通过BSA可以将快中子慢化到适当的能量范围,并且减少其他不需要的束流成分,进而满足BNCT用于治疗的中子束要求。本文利用蒙特卡罗模拟软件MCNP,基于2.5 MeV/30 mA的质子加速器,设计出可将质子打Li靶产生的快中子束慢化到热中子能量范围(<0.5 eV)和超热中子能量范围(0.5 eV~10 keV)的多终端BSA方案。提出的热中子BSA方案使用D₂O作为慢化体,BeO作为反射体,Bi作为$\gamma $过滤体,而超热中子BSA方案使用MgF₂作为慢化体,Pb作为反射体,6LiF作为热中子过滤体。热/超热中子方案在BSA出口处的束流参数,均满足国际原子能机构(IAEA)对BNCT治疗提出的束流指标要求。     

Investigation of stability of ITER candidate Hall sensors under neutron irradiation

Hall sensors offer an attractive true non-inductive method of magnetic field measurements for fusion reactors. Their use for steady state magnetic diagnostics of ITER is presently limited by their questionable radiation and thermal stability. Issues of stable and reliable operation in ITER like radiation and thermal environment are addressed by the contribution. Recently, novel Hall sensors, compatible with temperatures up to 200°C, were developed and their radiation stability was tested at LVR-15 experimental fission reactor. Overview of the experimental set-up on LVR-15 reactor is given. Degradation of the sensor’s sensitivity by several tens of percents was observed after neutron irradiation by the total neutron fluence of 2 × 10¹⁷ n/cm² in LVR-15. This level of neutron fluence is comparable to that expected to occur over the whole ITER life time for a sensor location just outside the ITER vessel. The in-situ recalibration techniques are expected to handle the observed degree of Hall sensors performance degradation in ITER environment.     

Neutron moderation in a bulk sample and its effects on PGNAA setup geometry

In a prompt gamma ray neutron activation analysis (PGNAA) setup, the neutron moderation in the bulk sample also plays a key role. This can even dominate the thermalization effects of the external moderator in some cases. In order to study the neutron moderation effect in the bulk sample, moderators with two different sizes of the sample were tested at the King Fahd University of Petroleum & Minerals (KFUPM) PGNAA facility. In these tests, the thermal neutron relative intensity and prompt gamma ray yield from the two moderators were measured using nuclear track detectors (NTDs) and NaI detector, respectively. As predicted by Monte Carlo simulations, the measured intensity of thermal neutron inside the large sample cavity due to the external moderator was smaller than that from the smaller sample cavity. Due to its larger size, additional thermalization of neutrons will take place in the larger sample. In spite of smaller thermal neutron yield from the external moderator at the large sample location, higher yield of the prompt gamma ray was observed as compared to that from the smaller sample. This confirms the significance of neutron moderation effects in the bulk sample and can thereby affect the PGNAA geometry size. This allows larger samples in conjunction with smaller moderators in the PGNAA setup.     

Gamma rays from thermal neutron capture in gold

The thermal column of the 10 kW University of Maryland pool reactor was used as the neutron source for this nuclear spectroscopy study. A 40 cm³ lithium drifted germanium crystal was the gamma (γ) ray detector used to measure the high energy prompt photon emissions from thermal neutron capture. The energy region was from 2.6–7.0 MeV. New capture gamma ray lines were observed.     

Analysis of neutron spectra and fluxes obtained with cold and thermal moderators at IBR-2 reactor: Experimental and computer-modeling studies

The results of experimental and computer-modeling investigations of neutron spectra and fluxes obtained with cold and thermal moderators at the IBR-2 reactor (Joint Institute for Nuclear Research (JINR), Dubna) are presented. These studies are for the YuMO small-angle neutron scattering (SANS) spectrometer (IBR-2 beamline 4). The neutron spectra have been measured for two methane cold moderators for the standard configuration of the SANS instrument. The data from both moderators under different conditions of their operation are compared. The ratio of experimentally determined neutron fluxes of cold and thermal moderators is shown at different wavelengths. Monte Carlo simulations have been carried out to determine the spectra for cold-methane and thermal moderators. The results of calculations of the ratio of neutron fluxes of cold and thermal moderators at different wavelengths are demonstrated. In addition, the absorption of neutrons in the air gaps on the way from the moderator to the investigated sample is presented. SANS with the protein apoferritin was done with both cold methane and a thermal moderator and the data were compared. The prospects for the use of a cold moderator for a SANS spectrometer at IBR-2 are discussed. The advantages of using the YuMO spectrometer with a thermal moderator with respect to the tested cold moderator are shown.     

Effect of X-ray energies on induced photo-neutron doses

Photoneutrons induced by two high energies range from the Elekta medical linear accelerator (10 and 18 MV) were measured by nuclear track detectors (NTDs). CR-39 NTD in contact with converter screen slide films, natural boron of thickness 40 μm coated on the polyester film (BN1). Detectors were exposed at 100 cm SSD with field size 20×20 on the patient table, with chest phantom and with build-up Perspex used for high-energy exposure. CR-39 registers the thermal neutron by the (n–α) reaction with the thin layer of boron and the fast neutron was measured through the (n–p) elastic scattering with the H₂ molecules in the CR-39 constituents.It was found that the total neutron dose (thermal and fast) from the 18 MV X-ray is higher than that of 10 MV. The measured thermal neutron dose is relatively smaller than the fast neutron dose in the case of direct exposure at the two X-ray energies. On the other hand, in the case of measurements on phantom and upon the use of build-up Perspex sheets, the ratio of fast to that of thermal is less than that of direct exposure.     

SiC基中子探测器对热中子的响应

以SiC二极管和中子转换材料6LiF为基础,研制了SiC基中子探测器,并用241Am源与临界装置分别研究了SiC基中子探测器的粒子响应、热中子响应。结果表明:SiC基中子探测器能够满足241Am源粒子的计数测量,但由于SiC二极管灵敏区薄,故不能用于5.48 MeV的粒子能谱测量;SiC基中子探测器对热中子响应良好,不同功率下,脉冲幅度谱中可清晰看到由6Li(n,)3H的反应产物粒子、3H粒子形成的双峰;脉冲幅度甄别法可将射线及电子学噪声甄别掉;在直接测量与符合测量中,SiC基中子探测器的计数率均与临界装置功率成线性关系,且直接测量的线性度比符合测量的线性度好,最好可达0.999 97。研究表明:降低肖特基接触金属的厚度、增加其外延层厚度、提高其外延层品质,可将SiC二极管用于带电粒子能谱测量。     

Modelisation and distribution of neutron flux in radium–beryllium source (<Superscript>226</Superscript>Ra–Be)

Using the Monte Carlo N-Particle code (MCNP-6), to analyze the thermal, epithermal and fast neutron fluxes, of 3 millicuries of radium–beryllium, for determine the qualitative and quantitative of many materials, using method of neutron activation analysis. Radium–beryllium source of neutron is established to practical work and research in nuclear field. The main objective of this work was to enable us harness the profile flux of radium–beryllium irradiation, this theoretical study permits to discuss the design of the optimal irradiation and performance for increased the facility research and education of nuclear physics.     

The dose comparison between the THOR and HFR epithermal neutron beams

This study is a part of the beam comparison campaign, inter-center dose comparison, between boron neutron capture therapy facilities at the Tsing Hua Open-pool Reactor and the High Flux Reactor. The clinical information exchange can improve the dosimetry uncertainty for medical physics in a mixed field. The method of paired Mg(Ar) and TE(TE) ionization chambers was used to determine the gamma-ray and neutron dose rates. Furthermore, activation foils, including gold, copper, and manganese, were employed to estimate the thermal and epithermal neutron fluxes. Measurements were performed free in air and also in a PMMA phantom. All the chambers were calibrated using a ⁶⁰Co primary standard source at the Institute of Nuclear Energy Research, Taiwan. Spectrum dependent neutron sensitivity of TE(TE) chamber is one of the important parameters to evaluate dose components. The requested neutron spectra were calculated by the Monte Carlo code MCNP. The measured thermal neutron fluxes, gamma-ray and neutron dose rates of the THOR beam in the phantom were 2.6, 2.2, and 2.1 times of the HFR beam at 2.5-cm depth, respectively. The higher thermal neutron flux and neutron and gamma-ray dose rates are due to the higher epithermal neutron beam intensity of the THOR.