On a 14-MeV neutron source based on MCF for fusion materials research

Recently it has been proposed to use the 14-MeV neutrons produced in the muon-catalyzed dt fusion reaction for the thermonuclear reactors first wall materials testing. The results of the Monte Carlo computations of some of the most interesting conceptual schemes of such a source are presented. It was shown that a neutron flux of about 10¹⁴ n/cm² s, can be achieved in the case of a scheme based on a mirror trap with strong magnetic fields, as proposed in this work.     

Maximal design basis accident of fusion neutron source DEMO-TIN

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.     

Current-carrying element based on second-generation high-temperature superconductor for the magnet system of a fusion neutron source

Application of current-carrying elements (CCEs) made of second-generation high-temperature superconductor (2G HTS) in magnet systems of a fusion neutron source (FNS) and other fusion devices will allow their magnetic field and thermodynamic stability to be increased substantially in comparison with those of low-temperature superconductor (LTS) magnets. For a toroidal magnet of the FNS, a design of a helical (partially transposed) CCE made of 2G HTS is under development with forced-flow cooling by helium gas, a current of 20–30 kA, an operating temperature of 10–20 K, and a magnetic field on the winding of 12–15 T (prospectively ~20 T). Short-sized samples of the helical flexible heavy-current CCE are being fabricated and investigated; a pilot-line unit for production of long-sized CCE pieces is under construction. The applied fabrication technique allows the CCE to be produced which combines a high operating current, thermal and mechanical stability, manufacturability, and low losses in the alternating modes. The possibility of fabricating the CCE with the outer dimensions and values of the operating parameter required for the FNS (and with a significant margin) using already available serial 2G HTS tapes is substantiated. The maximum field of toroidal magnets with CCEs made of 2G HTS will be limited only by mechanical properties of the magnet’s casing and structure, while the thermal stability will be approximately two orders of magnitude higher than that of toroidal magnets with LTS-based CCEs. The helical CCE made of 2G HTS is very promising for fusion and hybrid electric power plants, and its design and technologies of production, as well as the prototype coils made of it for the FNS and other tokamaks, are worth developing now.     

LiB neutron converter for a neutrino source

The possibility of enhancing the antineutrino flux from an intense source based on a lithium converter with the use of the beta-decaying ⁸Li isotope (T _1/2 = 0.84 s) is considered. The ⁸Li nuclide is produced in the lithium converter during the (n,γ) activation of the ⁷Li isotope, and high-energy \(\tilde \nu _e \) with E _ν of up to 13 MeV are then emitted. An intense antineutrino source can be constructed on the basis of various high-intensity neutron sources enclosed within a lithium blanket. It is possible to use reactions that enhance the ⁸Li yield (e.g., ¹¹B(n,α)⁸Li) for fast neutrons. One possible design for a converter with the ⁷Li and ¹¹B isotopes is analyzed. The production of ⁸Li is determined using the MAMONT software for Monte Carlo neutron transport calculations.     

聚变中子能谱测量系统脉冲中子灵敏度的实验研究

为多种复杂环境下的稳态和脉冲DT聚变中子能谱测量建立了一种灵敏度优化反冲质子磁谱仪. 使用成像板和同位素α源测量了谱仪的反冲质子能量-位置投影关系. 利用稳态加速器中子源平台、通过单粒子计数方法结合三维带电粒子输运程序模拟,研究了谱仪脉冲中子灵敏度能量响应. 通过高探测效率参数设置使谱仪对DT中子的探测效率达到2×10^-5 cm²水平,从而在较弱中子源上获得了较高统计精度实验数据. 程序模拟结果与谱仪α粒子刻度和DT中子标定实验结果取得了良好的一致性,可由此发展精细解谱技术,以提高脉冲中子能谱测量的灵敏度和能量分辨. 关键词： 聚变中子能谱 磁反冲质子 脉冲中子灵敏度 粒子输运     

Spectrometry and dosimetry of a neutron source

Using the Monte Carlo methods, the spectrum, dose equivalent and ambient dose equivalent of a ²³⁹PuBe have been determined at several distances. The spectrum and both doses were determined to 100 cm with a Bonner sphere spectrometer. These quantities were obtained by unfolding the spectrometer count rates using artificial neural networks. The dose equivalent was measured with the area neutron dosemeter Eberline model NRD, to 100, 200 and 300 cm. All the measurements were carried out in an open space to avoid the room-return. With these results, it was found that this source has a yield of 8.41E(6) n/s.     

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction ^232+233+234U and ²³¹Pa are formulated. (1) The fuel cycle would shift from fissile ²³⁵U to ²³³U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of ²³¹Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.     

A femtosecond neutron source

The possibility of using the ultrashort ion bunches produced by circularly polarized laser pulses to drive a source of fusion neutrons with sub-optical cycle duration is discussed. A two-sided irradiation of a deuterated thin foil target produces two counter-moving ion bunches, whose collision produces an ultrashort neutron burst. Using particle-in-cell simulations and analytical modeling, it is calculated that, for intensities of a few 10¹⁹ W cm^-2, more than 10³ neutrons per Joule may be producedwithin a time shorter than one femtosecond. Another scheme based on a layered deuterium-tritium target is outlined. PACS 24.90.+d; 29.25.Dz; 52.38.ph; 52.50.Jm     

Proposal of a novel high intense neutron source for radiation therapy

When using the¹H(t, n)³He reaction instead of the³H(d, n)⁴He reaction for a high intense neutron source, the power dissipated in the target which limits the maximum obtainable intensity is a factor of 70 less. Therefore higher fluxes than withd-T sources can be obtained with nearly no low energy background which is important for radiation therapy. The relevant properties of this source are discussed and its practicability is investigated.     

BNCT中子源用RFQ加速器

 分析了加速器作为硼中子俘获治疗(BNCT)中子源的优势,提出以射频四极场（RFQ）加速器作为BNCT用中子源的首选机型。对该RFQ的参数进行了选择,利用质子轰击锂靶近阈反应产生的前冲中子束能散低、散角小的优势,设定RFQ最终能量为1.9 MeV。采用“匹配均温”设计方法进行了此强流质子RFQ的束流动力学设计,并对设计方案进行了传输模拟,在入口归一化均方根发射度为0.25 mm·mrad、流强为100 mA时,束流传输效率为99.3％。选择合适厚度的锂靶,经过整形即可得到满足BNCT治疗需要的中子束。     

Dosimetry and spectrometry at accelerator based neutron source for boron neutron capture therapy

An innovative accelerator-based neutron source for boron neutron capture therapy has started operation at the Budker Institute of Nuclear Physics, Novosibirsk. This facility is based on a compact vacuum insulation tandem accelerator designed to produce proton current up to 10 mA. Epithermal neutrons are proposed to be generated by 1.915 MeV protons bombarding a lithium target using ⁷Li(p,n)⁷Be threshold reaction.In the article, techniques to detect neutron and gamma-rays at the facility are described. Gamma radiation is measured with NaI and BGO gamma-spectrometers. The total yield of neutrons is determined by measuring the 477 keV γ-quanta from beryllium decay. For the rough analysis of the generated neutron spectrum we used bubble detectors. As the epithermal neutrons are of interest for neutron capture therapy the NaI detector is used as activation detector. We plan to use a time-of-flight technique for neutron spectra measurement. To realize this technique a new solution of short time neutron generation is proposed.     

Fuel breeding in a hybrid muon catalyzed fusion reactor

We discuss the processes of nuclear fuel burnup and plutonium breeding in the uranium blanket of a hybrid mesocatalytic reactor. The time dependence of the nuclear fuel isotopic concentrations was calculated by the BURNFL code. Using a three-dimensional Monte Carlo MORSE-H code the plutonium and tritium breeding coefficients, the fission rates of uranium and plutonium and a specific power distribution in the blanket were computed. The total fusion energy multiplication factor was obtained as a function of the fuel residence time using results of a detailed calculation of the mesocatalytic channel and estimations of the electronuclear channel.     

Feasibility study of a strong ultra cold neutron source

The heat that would be produced in a superthermal ultra-cold neutron source proposed by Golub and Pendlebury is evaluated when it is placed in a cold neutron field. The source strength achievable in the present technology is also discussed.Paper based on the KEK Internal Report 83-7 (Dee), p 223, and 84-2 (Dec), p 139, by H.Y. and M.U.     

散裂中子源靶站谱仪的物理设计

中子散射广泛地应用于凝聚态物质研究和应用的众多学科领域,是研究物质微观结构和动态的理想工具之一.散裂中子源能是新一代的加速器基脉冲中子源,能为中子散射提供高通量的脉冲中子.文章重点介绍了散裂中子源项目CSNS中靶站和谱仪的建设内容和设计工作的进展.     

Cooling of Akmal-Pandharipande-Ravenhall neutron stars with a rotochemical heating source

Employing phenomenological density-dependent critical temperatures of strong singlet-state proton pairing and of moderate triplet-state neutron pairing, we investigate the effects of rotochemical heating on the thermal evolution of superfluid neutron stars whose cores consist of npe matter with the Akmal-Pandharipande-Ravenhall equation of state. Since the star is not quite in the weak interaction equilibrium state during spin-down, the departure from the chemical equilibrium leads to the rotochemical heating in a rotating NS which will increase the stellar＇s temperature. Our calculations show that the rotochemical heating delays the cooling of superfluid neutron stars considerably and makes the previous classification of NS cooling ambiguous. What＇s more, our model is currently consistent with all the observational data, and in particular some middle-aged and cold NSs （PRS J0205＋6449 in 3C 58, PRS J1357-6429, RX J007.0＋7303 in CTA 1, Vela） can be better explained when taking into account rotochemical heating.     

Cooling of Akmal-Pandharipande-Ravenhall neutron stars with a rotochemical heating source

Employing phenomenological density-dependent critical temperatures of strong singlet-state proton pairing and of moderate triplet-state neutron pairing, we investigate the effects of rotochemical heating on the thermal evolution of superfluid neutron stars whose cores consist of npe matter with the Akmal-Pandharipande-Ravenhall equation of state. Since the star is not quite in the weak interaction equilibrium state during spin-down, the departure from the chemical equilibrium leads to the rotochemical heating in a rotating NS which will increase the stellar's temperature. Our calculations show that the rotochemical heating delays the cooling of superfluid neutron stars considerably and makes the previous classification of NS cooling ambiguous. What's more, our model is currently consistent with all the observational data, and in particular some middle-aged and cold NSs (PRS J0205+6449 in 3C 58, PRS J1357-6429, RX J007.0+7303 in CTA 1, Vela) can be better explained when taking into account rotochemical heating.     

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根据第一壁后聚变中子通量和能谱的空间分布和 ITER 可能输出的聚变功率、聚变中子产额、第一壁(产氚包层)之后的中子通量的蒙塔-卡罗中子输运(MCNP)计算,对氦冷却固体增殖剂中子学和产氚包层模块(NT-TBM-HCSB)的聚变中子诊断提出概念性设计,本设计提出用固定位置微型裂变室探测器和可移动封装薄箔活化分析系统测量中子倍增器之后、固体氚增殖剂中及其后的中子通量;用可移动天然金刚石探测器的紧凑型能谱仪测量中子能谱。