Neutron yield measurement system of HL-2A tokamak

This research presents the development of HL-2A neutron yield measurement which includes ²³⁵U fission chamber and BF3 and ³He proportional counters. Equivalent noise formula of the radiation detection signal amplification system was derived to guide the development of the signal amplification system. Then all detectors were calibrated in situ by using the ²⁵²C_f neutron source. The neutron yield of the HL-2A during neutral beam heating was analyzed. These results indicate that the developed neutron flux diagnostic system can obtain neutron yield results under various experimental conditions of the HL-2A tokamak, and can provide information on neutron yield.     

Design study on an accelerator-based facility for BNCT and low energy neutron source

We have challenged to reduce an accelerator beam power for an accelerator-based BNCT facility. The required neutron source strength at the target has been estimated so as to make the epithermal neutron flux in the patient irradiation field exceed 1.7 × 10⁹ n/cm²s. The energy of the incident proton and the arrangement of the moderator assemblies are optimized. The beam current and the accelerating voltage are determined so that the accelerator power becomes minimum. The beam power required for the treatment in one hour is 62.5 kW. The proposed facility is equipped with a 2.5 MeV proton accelerator of 25 mA. a lithium target, and a heavy water moderator contained in an aluminum tank.     

CARR堆冷中子瞬发伽玛活化分析系统及实验研究

利用中国先进研究堆（CARR）在国内首次开展了冷中子瞬发伽玛活化分析（CNPGAA）实验,采用定制加长的电制冷高纯锗（HPGe）探测器和先进的数字多道谱仪DSPEC®-502进行测量,获得了NH₄Cl样品中元素冷中子瞬发伽玛谱和本底谱等数据,同时利用伽玛放射源¹⁵²Eu、¹³⁷Cs、⁶⁰Co以及NH₄Cl产生的瞬发伽玛射线对探测器在宽能区0.1~8 MeV进行能量刻度。为降低环境辐射本底,HPGe探测器外围采用环形锗酸铋（BGO）康普顿谱仪,10 cm铅以及含⁶Li和¹⁰B材料对中子束流准直屏蔽。此外,利用金片活化法测量了CARR堆运行功率为15 MW时有无冷源情况下冷中子导管B(CNGB)末端1 m处的中子注量率,结果显示有冷源时中子注量率可提高一个量级。     

Applicability of neutron noise technique for diagnostics of in-core components for heavy water reactors

Neutron flux signal is composed of a steady or mean component resulting from the flux produced by power operation of the reactor and a very small fluctuating component called ‘noise’ component. Analysis of neutron noise from suitably located sensors is a proven technique to monitor the in-core components of light water reactors (LWRs). However, the use of neutron noise has been rare, if any, for heavy water reactors (HWRs) as it was generally felt that the unfavourable transfer function characteristics of the reactors would limit its applicability. To assess the applicability of technique in pressurised heavy water reactors (PHWRs), experiments were carried out using in-core and out-of-core neutron sensors in a research reactor. This paper discusses the measurement details and results of the experiment. This paper concludes that the neutron noise technique can be effectively utilised for diagnostics/characterisation of the in-core components of heavy water reactors.     

用于核素在线瞬发伽马数据测量的热中子源设计

基于MCNP程序建立了西安脉冲堆热中子源设计的蒙特卡罗深穿透耦合屏蔽计算方法;采用MCNP临界源模型计算了热柱方腔前表面的中子、伽马平面源的参数,并与实验值进行了对比,给出了平面源的修正系数;基于中子、伽马等效平面源,采用新型硼铝复合材料以及铅、铋等材料,优化设计了热中子束流滤束装置,给出了热中子束流滤束装置的升级改造方案,得到热中子通量密度较原设计方案提高3倍、中子伽马通量密度比值大于10的平行热中子束,且束流外侧区域的中子、伽马本底剂量率接近0.025 mSv/h的辐射防护标准。     

Optimization study of some neutron imaging parameters in a reactor based neutron radiography facility

Neutron imaging technique can be used as a means of material Non-Destructive testing. One of common neutron sources for neutron radiography is nuclear research reactor. In this work, several neutron imaging parameters such as aperture distance and the radiography plane location from the neutron source as well as the aperture diameter have been computationally optimized to deliver a proposed neutron beam. According to the results, the aperture diameter of 3.5–4 cm which was located at 55–85 cm from the outer layer of reactor core and the position of image plane at 300–400 cm fulfills delivering of the suitable neutron flux and other required conditions. W, Fe and Pb walls with an identified length formed the convergent-divergent collimator and shielded the neutron and gamma out of beam path. Bi and Fluental filters with an optimal dimension were used to efficiently improve the neutron beam profile at a sample position.     

A Method of Isotopic Ratio Determination of Uranium by Ge(Li) Detector

The effect of the presence of a reentrant hole for extracting the neutron beam from within experimental systems of two different geometries is analyzed theoretically with use made of multi-group 2- dimensional discrete S_n method without resorting to bold assumptions for neutron transport nor drastic simplification of geometry. One of the two experimental systems is a rectangular light water prism 12 cm high of 40 × 40 cm² cross section, poisoned with Cd and/or In, and provided with a 1, 2 or 3 cm diameter reentrant hole. The other system is a 1″ thick natural uranium plate sandwiched between two layers of pure light water, each 4.6 cm thick, which also is provided with a 1cm diameter reentrant hole.

The following is concluded by comparing the angular neutron flux with and without the reentrant holes. With the first experimental system, perturbations of the order 10～25% is caused, which is particularly strong below about 0.3 eV, except when the hole diameter is 1cm. The perturbation effect increases as the reentrant hole becomes larger in diameter and shallower in depth. In the case of the second experimental system, the effect results in about 2% increase of the neutron flux at the bottom of the reentrant hole when the bottom is located in the natural uranium plate. On the other hand, if the bottom is in the light water region, the neutron flux is reduced by about 2～4% at the peak of the thermal neutron spectra.     

Target optimization for a 10-MeV E-beam neutron converter

Accelerator-based target design and optimization is an approach for neutron generation. The target plays an important role for a neutron source on an electron accelerator. For optimizing a neutron source using 10 MeV electron beams of Rhodotron-TT200, Pb, Ta, or W alloys with Be were calculated as photo-neutron converter. The neutron yield, flux and energy were simulated using the MCNPX code. The results indicate that a 10 MeV electron beam is capable of producing high-intensity neutron flux of 1013n·cm–2·s–1 with average energy of 0.8 MeV.     

中子三轴谱仪部件模拟与优化计算

根据三轴谱仪的安装位置及后续结构,利用蒙特卡罗方法模拟计算了中子三轴谱仪样品处的中子谱,获得了中子均匀性、注量率等参数。模拟结果表明:采用双聚焦单色器的中子注量率是采用平板单色器中子注量率的1.54倍,但中子分布均匀性变差,垂直发散度增大到约4°;聚焦单色器的聚焦长度对中子均匀性无影响,但聚焦长度越长其垂直发散度越小;入射单色中子波长λ0.4nm时,用热解石墨做过滤器比较合适,而当λ≥0.4nm时,用铍做过滤器更合适一些。     

Radiation Field for Fission Neutron Enhanced Boron Neutron Capture Therapy Employing Fast Neutron Source Reactor “YAYOI”

Boron Neutron Capture Therapy (BNCT) of a localized tumor needs a sufficient thermal neutron flux at the tumor. A surgical operation including ennucleation of the main part of tumor is required for the case of thermal neutron beam from a thermal reactor because of the rapid decrease of the neutron flux in the tissue. Intermediate neutrons with little fast neutron component are only produced by a specifically designed reactor which awaits to be build.

In the present paper, a positive use of fast neutron beams in addition to BNCT is proposed for treatment of some kind of localized tumors employing a fission fast neutrons from a fast neutron source reactor “YAYOI” of University of Tokyo which is licenced as such. Dose distributions in a water phantom located at a proposed position for two collimator cases were measured and its availability was confirmed as a possible port for therapy.     

Beam Hole Configuration and Analysis of Medical Reactor for Boron Neutron Capture Therapy

This report presents an investigation of beam holes to be provided in a medical reactor for Boron Neutron Capture Therapy. The principal requirement for the beam holes is to deliver the therapeutic doses of thermal and epithermal neutrons in a modest time (30 to 60min) with minimal fast neutron and γ-contaminants. Characteristics of the beam holes have been evaluated by 2-dim. n-γ coupling S _N transport calculations. Reexaminations and revisions of the beam hole design have brought improvements of the characteristics, especially an increase of the thermal neutron flux at the horizontal thermal neutron beam port and a decrease of the fast neutron flux at the vertical epithermal neutron beam port. The design objectives for the beam holes set up in this study may be achievable even if the thermal power of the reactor is reduced from 2 to 1MW.     

NPDMC中子和光子耦合输运孔道计算三维蒙特卡罗程序

ＮＰＤＭＣ是一个我国研制的多群中子和光子耦合输运的三维孔道计算蒙特卡罗程序。该程序配备了国内制作的８７群中子和２５群光子截面数据库,在程序设计中,采用了一些独特的方法和技巧,成功地解决了蒙特卡罗方法应有和到辐射屏蔽计算中所遇到的“深穿透”和“小概率”等难题。ＮＰＤＭＣ程序可以计算各种类型的研究性反应堆和动力堆中的管道,束孔和缝隙的中子能量,能谱,剂量率和γ光子通量及剂量率等。     

强流氘氚聚变中子源HINEG设计研究

强流氘氚聚变中子源HINEG(High Intensity D-T Fusion Neutron Generator)研发分两期:HINEG-Ⅰ为直流脉冲双模式,已成功产生中子强度1.1×10~(12)n/s的氘氚聚变中子,并实现连续稳定运行;HINEG-Ⅱ中子强度设计指标为10~(14)~10~(15)n/s量级,重点突破强流离子源和高载热氚靶技术。HNEG中子源可开展中子学方法程序与核数据、辐射屏蔽与防护、材料活化与辐照损伤机理和部件中子学性能等核能与核安全研究,同时也可在核医学与放射治疗、中子照相等领域拓展核技术应用研究。本文简要介绍HINEG总体设计方案与关键技术研究进展。     

中国先进研究堆高分辨中子粉末衍射谱仪模拟研究

应用蒙特卡罗方法对中国先进研究堆 (CARR)上的高分辨中子粉末衍射谱仪 (HRPD)的物理设计进行了模拟研究。给出了谱仪的分辨率曲线、样品处中子通量密度等主要参数的模拟计算结果 ,讨论了高分辨模式和高强度模式下的分辨率和强度变化 ,表明设计方案是可行的     

Evaluation of Reactivity Worth of Air Filled Beam Tubes in a Small Neutron Source Reactor by 3-D Transport Calculations

Four air filled beam tubes are present in the small pool type ²³³U fuelled, BeO and water reflected neutron source reactor KAMINI under construction in IGCAR, Kalpakkam. The positive reactivity insertion possible due to the accidental ingress of surrounding water into these empty beam tubes was estimated using the 3-D transport code TRITAC. To save computer time and also to highlight the effects caused by filling of air gaps with water, a similar but symmetrical core with more beam tubes was also studied in detail. Microscopic cross sections were derived from WIMS library and assembly homogenization was performed using a computer code SMAXY. Condensed eight group cross sections were used in TRITAC.

During the study reported here, need was found for the use of tight flux convergence criterion in TRITAC for getting reliable estimates of air to water worths. From the study of the symmetrical core, it is found that the presence of outer water reflector significantly reduces the air to water worth of these beam tubes. Further, the air to water worth of these beam tubes is quite small when the region where water replaces air is, beyond 15 cm from core boundary. Thus it is inferred that the total air to water worth of all beam tubes in the KAMINI reactor will be small compared to the total delayed neutron fraction of ²³³U.     

70MeV质子加速器产生的白光中子源强度和能谱的计算

用７０ＭｅＶ质子束轰击厚靶可产生白光中子源。选用金属钨作厚靶材料，应用适用于几十ＭｅＶ能区的核反应模型程序计算了中子能谱和角分布。结果表明：７０ｍｅＶ质子束在钨厚靶中终止前有４．６％的质子发生了反应，２００μＡ的７０ＭｅＶ质子束流产生的中子总强度为１．０１×１０１４Ｓ－１，中子平均能量为４．２ＭｅＶ，１０ＭｅＶ以上的中子强度为１．１５×１０１３Ｓ－１，高能中子主要在朝前的小角度区发射。     

Neutron Transport Benchmark Calculations, (I)

Five benchmark problems were selected as standards for the evaluation of calculations on neutron penetration. For three of these problems there were available reliable experimental results obtained in absolute intensities of the neutron angular spectrum in graphite and in water and of the neutron reaction rate in a two-dimensional water shield. The two remaining problems were based on PALLAS transport calculations in graphite and in water in plane geometry. For the first three problems, comparisons were made between the results of PALLAS calculations using ENDF/B-I cross sections and those obtained by experiment. It is concluded that the PALLAS calculations result in fairly good agreement with experiment, except at certain spatial meshes in two-dimensional geometry, due to ray effect. Comparisons are also made of the angular flux spectrum in graphite between calculations using ENDF/B-I and B-III data, as well of the scalar flux spectrum in water between those by PALLAS and by ANISN codes.     

Evaluation of gamma-ray and neutron energy for area monitoring system in the IFMIF/EVEDA accelerator building

The engineering validation of the IFMIF/EVEDA prototype accelerator, up to 9 MeV by supplying the deuteron beam of 125 mA, will be performed at the BA site in Rokkasho. A design of this area monitoring system, comprising of Si semiconductors and ionization chambers for covering wide energy spectrum of gamma-rays and ³He counters for neutrons, is now in progress. To establish an applicability of this monitoring system, photon and neutron energies have to be suppressed to the detector ranges of 1.5 MeV and 15 MeV, respectively. For this purpose, the reduction of neutron and photon energies throughout shield of water in a beam dump and concrete layer is evaluated by PHITS code, using the experimental data of neutron source spectra. In this article, a similar model using the beam dump structure and the position with a degree of leaning for concrete wall in the accelerator vault is used, and their energy reduction including the air is evaluated. It is found that the neutron and photon flux are decreased by 10⁴-order by employing the local shields using concrete and polyethylene around beam dump, and the photon energy can be suppressed in the low energy.