Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10¹⁷, 6.1 × 10¹⁶ (E < 0.55 eV), 7.6 × 10¹⁶ (E > 0.1 MeV) and 4.0 × 10¹⁶ (E > 1 MeV) m⁻² s⁻¹. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.     

D-T快中子照相准直屏蔽体设计及中子束特性的模拟研究

设计一个用于氘氚(D-T)快中子照相的准直屏蔽体系统,对D-T中子发生器快中子在准直屏蔽体材料中输运的MCNP模拟研究,给出准直中子束的中子能谱、注量率及均匀性、γ射线能谱和γ射线注量率等重要参数.模拟结果显示,用D-T中子发生器中子源和合理的准直屏蔽体系统可得到快中子照相所需的准直快中子束.     

Formation of beams for neutron capture therapy using A tangential reactor channel

Institute of General Nuclear Physics, Russian Scientific Center of the Kurachatov Institute. Translated from Atomnaya Énergiya, Vol. 74, No. 5, pp. 394–400, May, 1993.     

基于氘氚中子源硼中子俘获治疗的中子慢化整形研究

硼中子俘获治疗（Boron Neutron Capture Therapy,BNCT）是一种具有广阔前景的癌症治疗方法。氘氚中子源是未来可供选择的BNCT中子源之一,由于氘氚中子源产生的中子能量为14.1 MeV,不能直接用于BNCT,需要进行束流慢化整形。使用蒙特卡罗模拟程序MCNP5设计了相应的束流整形组件（Beam Shaping Assembly,BSA）,模拟验证了用半径为14 cm的天然铀球做中子倍增层的优越性,计算结果表明：采用50 cm厚的BiF3和10 cm厚的TiF3组合慢化层,17 cm厚的AlF3补充慢化层,0.2 mm厚的Cd热中子吸收层,3.5 cm厚的Pb作为γ屏蔽层,以及10 cm厚的Pb反射层,获得了较为理想的治疗中子束,输出中子束的空气端参数满足国际原子能机构（International Atomic Energy Agency,IAEA）的建议值。     

Study of neutron production and moderation for sulfur neutron capture therapy

Neutron capture therapy with Sulfur-33, similar to boron neutron capture therapy with Boron-10, is effective in treating some types of tumors including ocular melanoma. The key point in sulfur neutron capture therapy is whether the neutron beam flux and the resonance capture cross section of ~(33)S(n;α)~(30) Si reaction at 13.5 keV can achieve the requirements of radiotherapy. In this research,the authors investigated the production of 13.5 keV neutron production and moderation based on an accelerator neutron source. A lithium glass detector was used to measure the neutron flux produced via near threshold~7 Li(p,n)~7 Be reaction using the time-of-flight method. Furthermore, the moderation effects of different kinds of materials were investigated using Monte Carlo simulation.     

Design of the thermal neutron beam for neutron radiography at the Syrian MNSR

Neutron beam design was studied at the Syrian reactor (MNSR, 30 kW) with a view to generating thermal neutron beam in the vertical irradiation sites for neutron radiography. The design of the neutron collimator was performed using MCNP4C and the ENDF/B-V cross-section library. Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, >10 keV, respectively. To produce a good neutron beam quality, bismuth was used as photon filter. In this design, the L/D ratio of this facility had the value of 125. The thermal neutron flux at the beam exit was about 2.548 × 10⁵ n/cm² s. If such neutron beam were built into the Syrian MNSR many scientific applications would be available using the neutron radiography.     

Accelerator-based source of epithermal neutrons for neutron capture therapy

Computational studies are performed for choosing an optimal material and dimensions of a moderator for forming a beam of epithermal neutrons for boron-neutron-capture therapy based on a proton accelerator and the reaction 7 Li(p, n)7 Be as the neutron source. It is shown that the best material for this is magnesium fluoride. An optimal configuration is proposed for a combined moderator consisting of magnesium fluoride and teflon. The computational results are compared with the experimental data.Translated from Atomnaya Énergiya, Vol. 97, No. 3, pp. 195–200, September, 2004.     

Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed.In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.     

基于D-D中子源的硼中子俘获治疗慢化体设计

为了探讨利用D-D中子源评估硼中子俘获治疗(Boron Neutron Capture Therapy,BNCT)中子通量探测器性能的可能性,本文利用蒙特卡罗模拟程序MCNP5(Monte Carlo N Particle Transport Code, version 5)设计了基于D-D中子源的BNCT慢化体,并最终给出了一种"5 cm聚乙烯(Polyethylene,PE)+24 cm氟化钛(TiF3)+22 cm氟化镁(MgF2)"的组合作为慢化层、20 cm的镍(Ni)作为反射层以及0.03 cm的镉(Cd)作为热中子吸收层的慢化体设计方案。模拟计算结果表明:D-D中子源经设计的慢化体慢化后形成的中子场可以用于BNCT中子通量探测器性能的实验测试。     

小型源热中子照相装置准直慢化系统设计

小型中子源中子照相技术具有便携性强,应用范围广的优点,在检测一些较大或难以移动的样品时较固定式(反应堆中子源)中子照相系统具有优势.采用MCNP软件对一小型中子源中子照相装置的热中子准直屏蔽系统进行了理论设计,确定中子慢化体由238U和聚乙烯构成,辅以石墨反射层和硼聚乙烯吸收层,经优化计算,预计成像处热注量率达104 ...     

Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.     

Neutron and capture γ-ray accumulation factors in engineering calculations of reactor shielding

Translated from Atomnaya Énergiya, Vol. 66, No. 5, pp. 321–324, May, 1989.     

含硼卟啉在硼中子捕获治疗中的研究进展

硼中子捕获治疗(boron neutron capture therapy,BNCT)是利用10B(n,α)7Li反应产生的高能α粒子和反冲7Li原子进行治疗的高传能线密度辐射治疗方式,含硼化合物是硼中子捕获治疗的重要方面,其中含硼卟啉是上世纪90年代起广受关注的含硼化合物.介绍了硼中子捕获治疗及含硼卟啉的特点,阐述了...     

Design of an epi-thermal neutron flux intensity monitor with GaN wafer for boron neutron capture therapy

Boron neutron capture therapy (BNCT) is a promising cancer therapy. Epi-thermal neutron (0.5 eV < E_n < 10 keV) flux intensity is one of the basic characteristics for modern BNCT. In this work, based on the ⁷¹Ga(n,γ)⁷²Ga reaction, a new simple monitor with gallium nitride (GaN) wafer as activation material was designed by Monte Carlo simulations to precisely measure the absolute integral flux intensity of epi-thermal neutrons especially for practical BNCT. In the monitor, a GaN wafer was positioned in the center of a polyethylene sphere as neutron moderator covered with cadmium (Cd) layer as thermal neutron absorber outside. The simulation results and related analysis indicated that the epi-thermal neutron flux intensity could be precisely measured by the presently designed monitor.     

Preliminary design of hybrid energy reactor and integral neutron experiments

We propose a preliminary design for a fusion–fission hybrid energy reactor (FFHER), based on current fusion science and technology (with some extrapolations forward from ITER) and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. We study the interaction between the fusion neutron and the uranium fuel with the aim of to achieving greater energy multiplication and tritium sustainability. We also review other concept hybrid reactor designs. We design integral neutron experiments in order to verify the credibility of our proposed physical design. The combination of this program of research with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology foundation for the future development of the FFHER concept in China.     

混合能谱超临界水堆堆芯设计分析

提出了一种新型的超临界水堆概念设计:混合能谱超临界水堆,它包括慢谱区和快谱区两部分.其慢谱区燃料组件采用双排燃料组件,快谱区采用简单的正方形栅元燃料组件.慢谱区与快谱区的燃料组件都采用同向流动方式来简化堆芯设计.慢谱区的冷却剂出口温度远低于整个堆芯的出口温度,这大大降低了慢谱区包壳的温度峰值.此外,由于快谱区冷却剂密度很小,流速很高,故可采用较大的栅元结构,这有效地降低了包壳周向局部传热不均匀性.所以混合堆在充分继承慢谱、快谱堆芯优点的基础上,弥补两者的不足.     

D-T快中子治疗准直屏蔽体设计及中子特性模拟

设计了一个用于D-T快中子治疗的准直屏蔽体,通过D-T中子在准直屏蔽体中的MCNP模拟,计算了屏蔽体外透射中子和透射光子在水中的吸收剂量,由此评价了准直屏蔽体的屏蔽效果。利用MCNP程序,模拟了准直中子束及中子束中的γ射线在源皮距(SSD)100cm处的能谱,计算了γ射线与中子束在水中吸收剂量的比值,对准直中子束中γ射线的污染水平进行了评价。完成了准直中子束在人体组织等效水箱中输运的MCNP模拟,给出了吸收剂量深度分布、吸收剂量横向分布和吸收剂量等剂量曲线。     

Prioritization of components important for safety and sensitive to ageing. Application for TRIGA reactor

Due to potential impact on the SSC performances, the identification of ageing effects and implementation of appropriate methods for mitigation of these effects represents an important preoccupation of many organizations and has received even more attention in the last years in the perspective of Long Term Operation. The efforts used for developing time-dependent reliability models for all SSC could be considerable and may outweigh the benefits, but they are not always necessary, mainly because in some cases, the maintenance, test, inspection and surveillance methods are good provision to mitigate the ageing effects. To efficiently use the limited resources, it is necessary to identify and prioritize the SSC that need time-dependent reliability models by using specific criteria. This paper contributes to the effort of performing an efficient process of evaluation of ageing effects using probabilistic safety assessment models. The paper presents the approach developed for SSC selection, and the results of its application carried out for two particular systems of the TRIGA research reactor.