球形托卡马克聚变嬗变堆中子学设计

基于对球形托卡马克ST聚变堆的研究,提出了ST聚变嬗变堆的设计概念。对堆芯参数作了初步选择,确定了一组适合于嬗变包层的堆芯参数供中子学计算和结构设计参考,给出了在以嬗变次锕系元素（MA）核废物为目标的一维中子学计算结果。     

次临界能源堆物理设计进展

聚变-裂变混合能源堆包括聚变中子源和次临界能源堆,主要目标是生产电能。回顾了国内外混合堆的发展历史,给出混合能源堆设计的边界条件和约束条件,说明次临界能源堆以铀锆合金为燃料、水为冷却剂的设计思想。利用输运燃耗耦合程序MCORGS计算了混合能源的燃耗,给出了中子有效增殖因数、能量放大倍数和氚增殖比等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。论述了混合堆的热工设计并进行了安全分析。对于燃耗数值模拟程序,通过多家对算,保证其计算结果的可信性。针对次临界能源堆的特点,利用贫铀球壳建立了贫铀聚乙烯装置和贫铀LiH装置,并且专门设计加工了天然铀装置,开展铀裂变率、造钚率、产氚率等中子学积分实验,验证了数值模拟的可靠性。     

聚变实验增殖堆FEB-E放射性废物处置指标的计算

应用中子输运程序ＢＩＳＯＮ３．０、增殖堆放射性计算程序ＦＤＫＲ、剂量率计算程序ＤＯＳＥ完成了聚变实验增殖堆ＦＥＢ－Ｅ的放射性、核废物特性及废物处置额定容量（ＷＤＲ）的计算。结果表明,在停堆以后几周内,ＦＥＢ－Ｅ设计的经一壁和包层结构材料满足１０ＣＦＲ６１Ｃ级核废物处置额定容量的要求。对包层中的重要锕系元素＾２３２Ｕ、＾２３７Ｎｐ的含量也作了计算分析。     

行波堆深燃耗的计算特点

行波堆属于新概念堆型,卸料燃耗深度可达400 GWd/tHM,是现有快堆的3~4倍、压水堆的6~8倍,较高的卸料燃耗深度对堆芯物理分析工具计算正确性提出挑战。基于此,以KYLIN-1程序为基础,从能谱、裂变产物核素重要性、燃耗计算误差累积等方面探究行波堆深燃耗计算特点。对典型行波堆六角形组件分析结果表明:低富集度铀组件寿期初、末能谱差别较大,采用单一权重谱制备的多群截面库用于其燃耗计算时,无限增殖系数偏差较大;为保证行波堆深燃耗计算的正确性,燃耗链应包含重要的70种裂变产物核素;行波堆深燃耗计算时,由于燃耗步增多累积的误差较小,无限增殖系数偏差每燃耗步约为0.001%。     

冯开明先进燃料靶惯性约束聚变的燃耗和增益

计算了球形均匀冯开明先进燃料靶惯性约束聚变(ICF)的燃耗和增益。讨论了这种堆系统的能量平衡。设计了一种新型的由毛细管阵列组成具有抗辐射损伤、可自动更新的液悉金属锂自由表面多孔漫璧。用它取出聚变能。同时与D-T热核燃料靶系统的燃耗和增益及它们不同的堆工程特性作了比较。     

D-3He先进燃料靶惯性约束聚变的燃耗和增益

计算了球形均匀D-3He先进燃料靶惯性约束聚变(ICF)的燃耗和增益。讨论了这种堆系统的能量平 衡。设计了一种新型的由毛细管阵列组成具有抗辐射损伤、可自动更新的液态金属锂自由表面多孔湿壁,用它取 出聚变能。同时与D-T热核燃料靶系统的燃耗和增益及它们不同的堆工程特性作了比较。     

球形托卡马克堆嬗变中子学计算的比较研究

基于对球形托卡马克（ST）聚变堆的研究，提出了ST聚变－嬗变堆的设计概念。运用一维输运燃耗计算程序BISON3.0进行了优化设计，确定了适合于嬗变少额锕系MA核素的堆芯等离子体参数、包层结构及合适的换料周期。在一维计算的基础上，运用二维中子学程序TWODANT进行了二维中子输运计算；结合TWODANT给出的中子通量，运用一维放射性计算程序FDKR进行了燃耗计算，并给出了有关的计算结果。     

加氚对D-3He系统聚变燃烧过程的影响

对加少量氚的D-³He聚变系统的点火燃烧过程进行了数值模拟研究,得到了有关的物理图象和一些主要计算结果。研究结果表明,加少量氚可以解决D-³He聚变系统的点火问题和加速其燃烧过程,从而提高燃耗。     

In-Zinerater液态包层输运燃耗数值模拟

Z-Pinch惯性约束聚变是未来一种有竞争力的能源候选方案。Z-Pinch驱动的聚变裂变混合堆可高效地嬗变反应堆乏燃料中分离出的超铀元素。对美国Sandia国家实验室提出的In-Zinerater混合堆概念进行了中子学分析和数值模拟。在三维输运燃耗耦合程序MCORGS中增加了处理在线添加燃料与去除裂变产物的功能,实现了对液态燃料燃耗过程的模拟。增加6Li丰度和燃料初装量保持寿期初反应性不变,可以减缓寿期内反应性下降趋势。逐步增加包层内超铀元素装量,可以控制整个寿期内反应性基本恒定。聚变功率取20 MW,通过反应性控制,5年内包层能量放大倍数在160～180之间,氚增殖比在1.5～1.7之间,优于In-Zinerater基准设计方案。     

核能与聚变裂变混合能源堆

未来20年将是核能发展的一个关键时期.2035年左右,快堆有望投入商用;磁约束聚变、激光聚变、Z箍缩聚变也都有演示堆计划.聚变演示堆存在纯聚变与聚变裂变混合能源堆两种可能,而后者可降低聚变功率,缓解高能中子对材料的辐照损伤.另外,氘氚聚变供能时间有限.文章介绍了混合能源堆的概念.能源堆可充分利用铀资源,且后处理不涉及铀钚分离,有很好的防扩散性能.裂变堆、聚变堆、能源堆共同发展,可望使核能在不太长的时间内获得大规模应用,并可为人类提供千年以上的能源供应.     

The radiation characteristics of the transport packages with vitrified high-level waste

The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.     

Axial Neutron Flux Evaluation in a Tokamak System: a Possible Transmutation Blanket Position for a Fusion–Fission Transmutation System

A sub-critical advanced reactor based on Tokamak technology with a D–T fusion neutron source is an innovative type of nuclear system. Due to the large number of neutrons produced by fusion reactions, such a system could be useful in the transmutation process of transuranic elements (Pu and minor actinides (MAs)). However, to enhance the MA transmutation efficiency, it is necessary to have a large neutron wall loading (high neutron fluence) with a broad energy spectrum in the fast neutron energy region. Therefore, it is necessary to know and define the neutron fluence along the radial axis and its characteristics. In this work, the neutron flux and the interaction frequency along the radial axis are evaluated for various materials used to build the first wall. W alloy, beryllium, and the combination of both were studied, and the regions more suitable to transmutation were determined. The results demonstrated that the best zone in which to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements of W alloy/W alloy and W alloy/beryllium would be able to meet the requirements of the high fluence and hard spectrum that are needed for transuranic transmutation. The system was simulated using the MCNP code, data from the ITER Final Design Report, 2001, and the Fusion Evaluated Nuclear Data Library/MC-2.1 nuclear data library.     

The effect of metallic environment and low temperature on the 253Es α decay rate

In recent publications it has been pointed out that the α decay of transuranic elements in nuclear waste can be considerably speeded up by putting them into metals. The proposed mechanism is based on the effect of electron screening of radioactive nuclei (according to the Debye electron plasma model), which grows enhanced as temperature decreases. To verify the predicted phenomenon, half-lives of ²⁵³Es nuclei implanted in a metallic iron foil were measured at the temperature from 4 K to 50 mK. The results agree with the room-temperature data reported in the literature; no temperature dependence of the half-life was found within the error of ≈2%.     

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.     

Measurement of 248Cm/252Cf Cf Spontaneous Prompt Fission Neutron Spectrum

In order to treat the waste material of nuclear power and develop new type of clean nuclear power,it is necessary to measure the neutron adta of long half life nuclei existed in the waste material.The prompt spontaneous neutron spectrum is one of the most important unclear data for new type nuclear power facilities as well as for understanding the mechanism of fission neutron emission.The measurements of ²⁴⁸Cm/²⁵²Cf spontaneous prompt fission neutron spectrum in the neutron energy range form 200keV to 12MeV wer performed by using TOF method.A micro-ionization chamber aws used as fission fragment detector and stibene crystal as neutron detector.The flying paths of neutrons for the measurements were 30cm,50cm and 100cm respectively.The spontaneous prompt fission neutron spectrum of ²⁴⁸Cm was fitted by the Maxwellian distribution and the temperature was determined as (1.401±0.006)MeV in the corresponding neutron energy range.     

Activation detection using indium foils for simultaneous monitoring neutron and photon intensities in a reactor core

In this work, the time-sequential changes in neutron and photon intensities in the core of a research reactor were monitored for 330 days following the time of shutdown. Indium foils, used as activation detectors, were irradiated during different time periods in the core. The activated nuclides ^116m1In and ^115mIn were measured and their activities were used as indices of the changes in neutron and photon intensities during the period of interest. Photon and neutron sources that may have been responsible for the activation reactions with indium foils were identified. The activities of both ^116m1In and ^115mIn rapidly decreased in the first 20 h (Phase I), and subsequently their activity levels were proportionally varied with ¹⁴⁰La activity between 20 and 3000 h (Phases II and III). After 3000 h (Phase IV), neutron fluence rate slowly decreased to levels of less than 10 cm^?2 s^?1. Then, the main neutron sources were identified as the neutron-emitting transuranic nuclides and the most important photon source that activated indium to form ^115mIn was identified as ⁶⁰Co. The activation detectors based on indium foils were found to be effective for simultaneously monitoring the variations of neutron and photon levels in reactor cores under sub-critical conditions.     

Peculiarities of the modern neutron spectrometry

Neutron spectrometry provides many branches of science and technology with the necessary data. Usually the main part of the data is supplied by powerful neutron time-of-flight spectrometers. Nevertheless there are many other very effective but simpler and cheaper neutron spectroscopy methods on accelerators, suitable for solution of plenty of scientific and applied problems (for example, in astrophysics and radioactive waste transmutation). The methods of slowing-down spectrometry in lead and graphite, generating of neutron spectra, characteristic for nucleosynthesis in the stars, and neutron spectrometry by means of primary γ-transition shift are discussed in the report.     

Neutron measurements around a TN85-type storage cask with high-active waste

Several types of casks have been deposited in the German interim storage facility for spent fuel assemblies and vitrified high-active waste (HAW) at Gorleben since 1995, most of them of the CASTOR^® type. In 2008 a delivery of 11 TN85-type casks arrived. They belong to the Transnuclear/Areva cask family and, compared to the flasks of the German (GNS) CASTOR^® type, they differ in the neutron shielding design.Generally, radiation exposure of personnel during transportation and storage of casks containing spent fuel and vitrified waste is caused by mixed photon/neutron fields. Frequently, especially at casks for vitrified waste from reprocessing, neutrons are the major component of radiation exposure.Spectrometric and dosimetric investigations were made around a cask of the TN85-type. Neutron fluence spectra and reference values of the ambient dose equivalent H*(10) were measured by means of a Bonner sphere spectrometer (BSS) at several locations on the cask surface and in its environment. Moreover, commercial area dosemeters, LB6411 neutron monitors and conventional AD 6-type photon dosemeters were used. In addition, the responses of two electronic personal dosemeters for mixed fields (EPD-N2, DMC 2000GN) and a TLD albedo dosemeter were investigated.The neutron spectra obtained from the BSS are presented and compared with former measurements at CASTOR^® type casks. The relative responses of the LB6411 survey meter and the individual dosemeters are discussed. The LB6411 monitor indicates H*(10) around the TN85 cask with tolerable measuring uncertainties. The personal dosemeters provide acceptable results for photons but overestimate the neutron dose considerably.     

Prospects for transmutation of nuclear waste and associated proton-accelerator technology

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, a major concern or objection is the long-term burden that is constituted by the radiotoxic waste from the spent fuel. The concept of Partitioning & Transmutation, a scientific and technological answer, is therefore of high interest. Its deployment may use dedicated “Transmuter” or “Burner” reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a solution of high interest. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at an introduction into the field in order to focus, in its later part, on the development of the required accelerator technology.     

The applications of nuclear track detectors in transuranic contamination problems

Nuclear Track Methodology (NTM) is very well known for its possibilities in applications in a wide variety of fields. The appearance of the polycarbonate as alpha radiation detector material, established a very confident monitoring surface alpha contamination, soil and water alpha activity. The purpose of the study is to investigate the utility of the CR-39 (Allyl Diglycol Carbonate) in fast semi-quantitative transuranic contaminant evaluation, including the distribution in underground contaminated soils, hot spots and transuranic material accumulations.