神光Ⅱ与神光Ⅲ原型上的辐射驱动小收缩比内爆靶设计

本文建立辐射驱动小收缩比(5至10倍)的内爆靶丸的设计方法,并利用辐射流体力学程序Multi-1d对神光Ⅱ和神光Ⅲ原型装置的辐射驱动小收缩比内爆靶丸进行模拟计算,设计出小收缩比内爆靶丸的构形,计算结果给出压缩比、中子产额、压缩面密度随燃料压力变化等内爆物理参量.神光Ⅱ装置上的内爆靶燃料分别采用DD气体和DT气体两种;神光Ⅲ原型装置的内爆靶燃料采用DD气体.神光Ⅱ上小收缩比内爆实验结果与模拟结果基本一致,表明Multi-1d用于设计此类实验的可靠性. 关键词： 辐射驱动 小收缩比 内爆 面密度     

用于中子能谱测量的反冲质子磁谱仪

根据神光-Ⅲ原型内爆实验条件下初级中子能谱诊断的要求,提出了反冲质子磁谱仪原型的设计方案,并通过蒙特卡罗程序Geant4对设计方案进行了仿真模拟。该质谱仪的测量能区在6~30 MeV,测量产额范围大于1012,能够测量内爆实验的全中子能谱。模拟结果表明:设计的谱仪对于14 MeV的中子探测效率大于10-11,能量分辨达到1%,信噪比大于10,满足神光-Ⅲ原型装置内爆实验离子温度诊断的要求。     

神光Ⅲ主机装置首次间接驱动内爆集成实验

基于新建成的神光Ⅲ主机装置开展了首次激光间接驱动内爆集成实验。国内首次采用多环脉冲整形激光注入黑腔产生X光辐射驱动内爆,通过优化激光打靶参数控制驱动不对称性,演示了以惯性压缩为主、收缩比约15倍的DT靶丸内爆实验能力, 实现了准一维的高静产额(YOC)和高中子产额的物理指标;其中,真空黑腔DT靶丸最高中子产额为1.91012,YOC达到60%;充气黑腔DT靶丸最高中子产额为2.41012,YOC大约70%。该实验为未来开展多台阶整形辐射驱动、更高倍数收缩比的高压缩内爆综合实验、验证点火靶物理设计和关键调控措施有效性奠定了基础。     

氘氘中子产额铟活化诊断方法

提出了活化法测量DD中子产额的实验方法,该方法可提高DD中子产额测量的精度。方法基于铟同位素115In与DD中子的非弹性散射反应,活化反应释放的射线被HPGe探测器记录,根据活化系统标定灵敏度推算出中子产额。分析了探测器记录的活化射线数与中子产额间的关系。介绍了一套活化测量的系统设计。通过蒙特卡罗方法模拟了活化样品出射的射线数与样品厚度的关系,模拟结果表明:样品厚度取为1 cm可兼顾活化效率和测量精度。在加速器上对铟活化样品进行了标定实验,实验结果表明:在聚变中子产额大于2109的实验中可使用铟活化诊断方法,中子产额测量的相对标准误差在10%以内。随着聚变中子产额的不断提高,铟活化测量中子产额的精度可进一步提高。     

ICF中子发射时间诊断

 对ICF中子发射时间的诊断技术进行了研究,研制了基于快闪烁体和微通道板式光电倍增管的中子发射时间探测器。在某大型激光原型装置上进行了中子发射时间的实验测量,成功获得多发实验的中子发射时间与打靶激光脉冲的时间及中子发射时间之间的关系。实验结果表明：中子发射时间探测器对DT中子和DD中子都能够响应,中子产额测量下限达到10⁷,时间测量不确定度小于20 ps;CH烧蚀层越厚,中子发射时间越长。     

神光Ⅲ主机双束组激光间接驱动内爆实验研究

在神光Ⅲ主机装置上,利用已经建成的两个激光束组,开展了激光间接驱动内爆物理磨合实验,是神光Ⅲ主机装置首次出中子实验。实验采用1400μm×2100μm黑腔,500μm的塑料靶丸充1 MPa的DD燃料,激光从黑腔两端55°注入。实验获得的最高中子产额为9.7×108。实验结果表明,实验黑腔的耦合效率约为50%;使用的黑腔偏长,靶丸被压缩为"薄饼形";中子产额和激光能量正相关;中子发射峰值时刻主要依赖于烧蚀层厚度。     

小收缩比内爆实验初步研究

在神光III原型装置上,利用8路激光间接驱动充高气压DT靶丸,开展小收缩比内爆实验.实验中测量得到中子产额、离子温度、聚变反应速率峰值时刻(bangtime)等关键物理量,以及它们随烧蚀层厚度变化的规律.从定性和定量两个方面对实验结果进行了分析讨论,推测流体力学不稳定性和内爆不对称性是导致实验结果与一维辐射流体计算结果不一致的原因.     

激光间接驱动惯性约束聚变内爆物理实验研究

激光间接驱动惯性约束聚变利用辐射烧蚀驱动靶丸球形内爆,在减速阶段将内爆动能转化成热斑内能,同时压缩燃料,达到点火条件,实现聚变点火。根据目前认识,影响内爆压缩过程的主要因素包括内爆对称性、燃料熵增因子、内爆速度和混合。内爆物理实验研究的目的是发展对上述影响因素的实验表征方法,获取这些影响因素随靶设计参数的变化规律,建立相应的实验调控能力,最终达到不断提升内爆性能的目的。为此,在内爆对称性方面,开展了Bi球自发光实验,用于研究点火脉冲前2ns驱动不对称性;在内爆速度方面,开展了球面弯晶单能流线实验,测量得到内爆速度和剩余质量随时间的变化;在混合方面,开展了内壳层示踪涂层内爆混合实验,测量得到环形发光图像。为考察综合内爆性能,在神光Ⅱ和神光Ⅲ原型装置上开展了DT内爆实验,获得了中子产额随初始靶参数的变化规律。     

基于纯冲击波聚心的准一维氘氘内爆物理实验

在神光Ⅲ原型装置上利用八路6400J/1ns激光注入1100μm×1850μm的黑腔内产生210eV的高温辐射场,均匀辐照填充氘氘燃料的靶丸实现内爆。实验中选择高气压薄壳靶丸实现纯冲击波聚心内爆。通过闪烁体探测器、中子条纹相机等多套诊断设备获取了中子产额、聚变反应时刻等关键内爆参数。结合一维数值模拟表明,实验测量的中子产额与干净一维数值模拟计算的中子产额之比达到90%;同时通过人为破坏内爆对称性等方式表明,该设计下内爆中子产生机制集中于冲击波聚心,其内爆性能受到高维因素影响极低,从而实现了准一维内爆。     

塑料闪烁探测器测量辐射驱动内爆的中子产额

本文介绍了在ＬＦ－１２装置上，用塑料闪烁探测器诊断辐射驱动内爆低产额中子的过程，并讨论了中子产额与靶型及靶参数的关系。结果表明，在ＬＦ－１２装置上，对不同的靶，激光辐射驱动内爆的中子产额为１０＾３－１０＾５。     

用于神光Ⅲ燃料面密度诊断的下散射中子探测系统设计

根据辐射屏蔽后5.5m测点处的辐射场情况,分别设计了电流型探测器系统和成像型探测器系统。通过Geant4数值模拟分析可得:在面密度达到10mg/cm2、初级中子产额为1012时,电流型探测器系统满足测量的信噪比,信噪比达到40∶1;在面密度达到10mg/cm2、初级中子产额为1011时,成像型探测器系统满足测量的信噪比,信噪比好于10∶1;面密度增大时,信噪比有所改善;但是当初级中子产额达到1012时,出现中子信号重叠现象,可通过缩短曝光时间或者减小塑料闪烁体厚度来降低中子重叠率。     

反冲质子磁分析技术用于氘氚中子能谱测量研究

介绍了一种基于反冲质子法和磁分析技术的氘氚聚变诊断方法, 适用于稳态及脉冲条件下的等离子体温度、燃料密度和中子产额的精确诊断. 设计了小型的原理性装置, 磁分析器使用高性能钕铁硼二极永磁铁, 焦平面上使用CR-39固体径迹探测器或PIN探测器测量质子位置分布. 使用²³⁹Pu α 源对磁分析器进行了实验标定, 建立了配套的模拟程序. 利用蒙特卡罗方法模拟分析了装置整体性能, 并在K-400加速器上进行了中子实验研究.     

激光间接驱动内爆靶丸的X光诊断

 报道了神光Ⅱ激光聚变实验中内爆燃料靶丸区电子温度、电子密度以及燃料面密度的X光诊断结果。在电子温度诊断中，采用X射线光谱学方法，根据聚变靶丸燃料区的Ar示踪元素的Ly-β线与He-β线的强度比推断出靶丸燃料区电子温度为(950±100) eV；在电子密度诊断中，利用靶丸燃料区Ar元素的He-β线Stark展宽确定聚变靶丸芯部的电子密度为(0.9±0.2)×10²⁴ cm^-3；在燃料区面密度诊断中，利用X光单能照相技术获得了内爆靶丸的燃料面密度为（3.2±0.5） mg/cm²。     

Radiation effect of neutrons produced by D–D side reactions on a D–3He fusion reactor

One of the most important characteristics in D–³He fusion reactors is neutron production via D–D side reactions. The neutrons can activate structural material, degrading them and ultimately converting them into high-level radioactive waste, while it is really costly and difficult to remove them. The neutrons from a fusion reactor could also be used to make weapons-grade nuclear material, rendering such types of fusion reactors a serious proliferation hazard. A related problem is the presence of radioactive elements such as tritium in D–³He plasma, either as fuel for or as products of the nuclear reactions; substantial quantities of radioactive elements would not only pose a general health risk, but tritium in particular would also be another proliferation hazard. The problems of neutron radiation and radioactive element production are especially interconnected because both would result from the D–D side reaction. Therefore, the presentation approach for reducing neutrons via D–D nuclear side reactions in a D–³He fusion reactor is very important. For doing this research, energy losses and neutron power fraction in D–³He fusion reactors are investigated. Calculations show neutrons produced by the D–D nuclear side reaction could be reduced by changing to a more ³He-rich fuel mixture, but then the bremsstrahlung power loss fraction would increase in the D–³He fusion reactor.     

Study of the Elution Profile of the 99mTc Gel Generator

The generator with gel matrix is considered a perspective type for obtaining ^99mTc. The elution yield of a generator and the course of the elution profile A_Vn = f(V_n) are affected by the properties and conditions of preparation of zirconium molybdate (generator matrix). Furthermore, it is influended by neutron flux density as well as irradiation time and temperature of cooling water during the irradiation. Zirconium molybdate, Mo: Zr = 1.72, in the form of a monodispersion (mean particle size x = 0.175 mm), was irradiated under the neutron flux density ?_n = (0.7 … 2.2) × 10¹³ cm^?2 s^?1. The elution profile and the elution yield of ^99mTc were examined in dependence on the temperature of drying the Zr–Mo matrix (40–100 °C). The kinetics of the passage of ^99mTc into the solution, in dependence on the temperature of drying the hydrogel, was investigated. Some predetermined substitutional functions were tested and the mechanism of the ^99mTc passage into solution was explained.     

A capture-gated fast neutron detection method

To address the problem of the shortage of neutron detectors used in radiation portal monitors(RPMs),caused by the ~3He supply crisis, research on a cadmium-based capture-gated fast neutron detector is presented in this paper. The detector is composed of many 1 cm × 1 cm × 20 cm plastic scintillator cuboids covered by 0.1 mm thick film of cadmium. The detector uses cadmium to absorb thermal neutrons and produce capture γ-rays to indicate the detection of neutrons, and uses plastic scintillator to moderate neutrons and register γ-rays. This design removes the volume competing relationship in traditional ~3He counter-based fast neutron detectors, which hinders enhancement of the neutron detection efficiency. Detection efficiency of 21.66% ± 1.22% has been achieved with a 40.4 cm × 40.4cm × 20 cm overall detector volume. This detector can measure both neutrons and γ-rays simultaneously. A small detector(20.2 cm × 20.2 cm × 20 cm) demonstrated a 3.3 % false alarm rate for a ~(252)Cf source with a neutron yield of 1841 n/s from 50 cm away within 15 s measurement time. It also demonstrated a very low(0.06%) false alarm rate for a 3.21 × 10~5 Bq ~(137)Cs source. This detector offers a potential single-detector replacement for both neutron and the γ-ray detectors in RPM systems.     

A conceptual high flux reactor design with scope for use in ADS applications

A 100 MWt reactor design has been conceived to support flux level of the order of 10¹⁵ n/cm²/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D₂O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium-aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of 8 × 10¹⁴ n/cm²/s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.      

点火靶高熵内爆的数值模拟

使用一维多群输运程序RDMG与二维少群扩散程序LARED-S对点火靶高脚与低脚内爆进行数值模拟.相对于低熵内爆,高脚高熵内爆通过提高预脉冲的辐射温度使得烧蚀面与物质界面的流体稳定性得到明显的改善,能够抑制流体不稳定的增长与热斑混合的发展.同时,高熵设计导致燃料的压缩变差,阻滞时刻燃料的压缩密度与面密度相应降低,中子产额降低.因此,高脚高熵内爆是通过牺牲燃料的高压缩,来换取靶丸内爆流体稳定性能的改善.     

Cumulative fission yield measurements with 14.7 MeV neutrons on 238U

The fission yield data in the 14 MeV energy neutron induced fission of ²³⁸U play an important role in decay heat calculations and generation-IV reactor designs. In order to accurately measure fission product yields (FPYs) of ²³⁸U induced by 14 MeV neutrons, the cumulative yields of fission products ranging from ⁹²Sr to ¹⁴⁷Nd in the ²³⁸U(n, f) reaction with a 14.7 MeV neutron were determined using an off-line γ-ray spectrometric technique. The 14.7 MeV quasi-monoenergetic neutron beam was provided by the K-400 D-T neutron generator at China Academy of Engineering Physics (CAEP). Fission products were measured by a low background high purity germanium gamma spectrometer. The neutron flux was obtained from the ⁹³Nb (n, 2n)^92mNb reaction, and the mean neutron energy was calculated using the cross-section ratios for the ⁹⁰Zr(n, 2n)⁸⁹Zr and ⁹³Nb(n, 2n)^92mNb reactions. With a series of corrections, high precision cumulative yields of 20 fission products were obtained. Our FPYs for the ²³⁸U(n, f) reaction at 14.7 MeV were compared with the existing experimental nuclear reaction data and evaluated nuclear data, respectively. The results will be helpful in the design of a generation-IV reactor and the construction of evaluated fission yield databases.     

Absence of voids in fast neutron irradiated zirconium by small angle neutron scattering

Abstract

The aim of this research was to resolve a difference of opinion in the literature on the presence of voids in fast neutron irradiated zirconium. There is a great interest in the study of zirconium, since zirconium and its alloys are used extensively in modern power reactors, for example in the fuel rods as a containment material for enriched uranium. A polycrystalline sample of zirconium was irradiated in the HERALD reactor at 40°C with 10²⁰ fast neutrons per cm^?2. The neutron scattering from irradiated and unirradiated standard samples was studied over a wide Q range from 0.001 to 1.12 Å^?1 on a D11 Spectrometer at the ILL (France). The defect cross-section (the difference between the scattering of the standard zirconium crystal and irradiated crystal) was nearly flat as a function of Q (momentum transfer vector) with an average value of 8.5 mb/Str/atom. This indicates a point defect concentration of about 1.8%. Thus the absence of any small angle (Q dependent) defect scattering indicates that large damage regions (e.g. voids) are not produced in zirconium by fast neutron irradiation.