Design Considerations for a Migma Advanced Fuel Fusion Reactor

The migma concept is being pursued at Fusion Energy Corporation as a means of achieving controlled fusion.1-4 The features which distinguish this concept from other controlled fusion concepts may be summarized as: 1. High energy 2. Ordered motion 3. Use of advanced fuels 4. Small physical size Beams of ions are injected into the field of a superconducting magnet at MeV energies. The resulting motions of trapped ions have a high degree of order in phase space compared with a thermalized gas. At MeV energies the two major ion loss mechanisms, charge transfer and multiple Coulomb scattering, are greatly suppressed compared with thermonuclear energies (1-100 keV), because the cross section for multiple Coulomb scattering falls off as T1-5 and that for charge transfer approximately as T-5. Because ions are injected at nearly the average energy of the migma, it may also be said that, as a practical matter, the use of ordered motions facilitates the attainment of colliding energies in the MeV range. The ion motion is essentially that of precessing orbits which all intersect within a central core that is small compared with a gyrodiameter. Motion along the magnetic field lines is confined by a non-adiabatic focusing. The high collision energies obtainable enable the use of what are called "Advanced Fuels," that is, fuels other than the deuteriumtritium (D-T) mixture planned for, e.g., the tokamak fusion reactor. These fuels require higher collision energies for useful reaction rates.     

Lattice Parameter Measurements on Cluster-Type Fuel for Advanced Thermal Reactor

Abstract

The behavior of neutrons in a highly heterogeneous unit cell consisting of D₂O moderator, H₂O coolant and a 28-pin fuel cluster contained in a pressure tube has been studied through lattice parameter measurements covering three different ²³⁵U enrichments, four coolant void fractions and two lattice pitches. A single-region core configuration was adopted, with which measurements were made to determine—in relation to coolant void fraction—the critical D₂O level, as well as various lattice parameters

A strong dependence on coolant void was observed for the critical level and the lattice parameters, in the case of the smaller 22.5 cm pitch lattice, due to the positive effect on core reactivity exerted by the slowing-down faculty of H₂O in the epithermal energy region. With the larger 25.0 cm pitch, however, no meaningful dependence on void fraction was shown by any of the measured values, and this was ascribed to a compensating negative effect due to enhanced thermal neutron self-shielding in the fuel region produced by the H₂O coolant.

The results of cell calculations obtained by means of the METHUSELAH-II code showed generally good agreement with experimentally determined data, for both critical D₂O levels and lattice parameters, in the case of coolant-filled lattices (0, 30 and 70% void fractions). For lattices devoid of coolant (100% void fraction), however, discrepancies in lattice parameters—particularly in p ²⁸—produced corresponding deviations in core reactivity amounting to 1% in excess of those incurred with other void fractions.     

聚变驱动乏燃料焚烧堆(FDS-SFB)燃料循环动态分析

针对聚变驱动乏燃料焚烧堆FDS-SFB燃料循环系统与一次通过燃料循环系统,利用系统动力学软件Vensim分别建立了这两种循环系统的动态分析模型,并根据假设的三种核电发展情景,分别计算了这两种燃料循环系统的资源需求、乏燃料累积量、钚累积量及次锕系元素累积量。初步计算结果表明:与一次通过式燃料循环系统相比,FDS-SFB燃料循环系统可减少天然铀需求量与乏燃料累积量,减少的程度与核电发展规模相关。     

高温氧化挥发法——一种先进乏燃料后处理的首端工艺技术

高温氧化挥发处理技术是乏燃料后处理的干法首端过程,其目的是在乏燃料后处理分离工艺前实现包壳与燃料芯块分离,燃料氧化和裂变产物3 H、85 Kr/Xe、14 C、129I、Cs的去除。此过程既有利于乏燃料元件的溶解,又有利于在乏燃料元件进入溶解工艺之前实现氚碘等裂变元素去除,是实现整个乏燃料后处理流程过程废液最小化和氚碘等裂变产物集中管理的最有效方法之一。本文针对氧化挥发技术在乏燃料后处理首端中的应用特点以及氧化温度、气氛等关键影响因素进行了综合分析和阐述。     

工艺参数对燃料板热轧过程影响的模拟研究

利用显示动力有限元法对U_(10)Mo/Al弥散燃料板的轧制过程进行模拟,研究了热轧工艺参数对弥散燃料板芯体轧制变形和接触压力的影响。有限元模拟计算结果表明:轧制速度对燃料板芯体的应变速率有明显影响;轧制过程中燃料板芯体所受的表面接触应力和变形均随压下率的增加而增加;随着摩擦系数的增加,芯体宽展率逐渐降低,而其表面接触压力相应增加。     

Potential reprocessing improvements in the Advanced Fuel Recycle System

From the aspects of economical competitiveness, proliferation resistance, and minimizing waste problems, PNC has proposed an improved recycle concept for the FBR fuel cycle, termed Advanced Fuel Recycle System. Reprocessing in this system is based on the well-known PUREX flowsheet and features a “single cycle Pu/U co-extraction flowsheet” with lower decontamination factor (DF) than that in the conventional process. This feature is practical because of the FBR's low neutronic sensitivity to impurities.

Such a simplified extraction process without purification cycles should substantially reduce not only the number of process components but also the quantities of liquid to be treated in other related processes, so it will lead to the proportional reduction in waste processing, waste itself, and all other related equipments and facilities. This should improve overall economics. One method being examined to further reduce the liquid throughputs and simplify the process is to apply the crystallization technique to dissolver solution.

Overall, with this proposed concept, proliferation resistance will be significantly improved because plutonium is always recovered as a mixture with the uranium and DF of the plutonium product is low.

Reprocessing and fabrication processes are integrated into one fuel cycle plant in this system further contributing to these improvements.     

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation’s energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.     

法杰马第二代先进燃料组件

本文介绍了法杰马所提供的先进燃料组件(AFA)设计所具有的特点,同时介绍了第二代先进燃料组件的技术改进。     

先进核燃料循环体系研究进展

概述了先进核燃料循环体系的概念 ,论述了目前后处理与分离 嬗变领域的研究进展和技术发展趋势     

法国先进核燃料组件在中国应用前景浅析

介绍了法国先进燃料组件（ＡＦＡ）系列核燃料的特点及其在中国的应用现状，阐述了广东核电集团有限公司核电发展战略和第三代先进燃料组件（ＡＦＡ－３Ｇ）在中国的应用前景，并从物理，热工水力和燃料组件的机械完整性等方面作了初步论证，对当前开展的有关工作进行了讨论。     

先进核燃料特环体系研究进展

概述了先进核燃料特环体系的概念，论述了目前后处理与分离－嬗变领域的研究进展和技术发展趋势。     

Reactivity Effect of Non-Uniformly Distributed Fuel in Fuel Solution Systems

A numerical method to determine the optimal fuel distribution for minimum critical mass, or maximum k-effective, is developed using the Maximum Principle in order to evaluate the maximum effect of non-uniformly distributed fuel on reactivity. This algorithm maximizes the Hamiltonian directly by an iterative method under a certain constraint—the maintenance of criticality or total fuel mass. It ultimately reaches the same optimal state of a flattened fuel importance distribution as another algorithm by Dam based on perturbation theory.

This method was applied to two kinds of spherical cores with water reflector in the simulating reprocessing facility. In the slightly-enriched uranyl nitrate solution core, the minimum critical mass decreased by less than 1% at the optimal moderation state. In the plutonium nitrate solution core, the k-effective increment amounted up to 4.3Δk within the range of present study.     

先进压水堆燃料组件计算不确定性分析

采用广义微扰理论,研究核数据不确定性对先进压水堆AP1000燃料组件宏观截面参数计算不确定性的贡献与影响机理。通过比较、分析不同因素对组件参数计算不确定性的贡献,给出组件宏观截面参数相关系数矩阵;采用敏感性分析方法及分步比较的思路研究在不同堆芯运行状态下核数据对AP1000燃料组件宏观参数计算不确定性贡献的机理。研究结果表明:核数据自身不确定性通过组件输运计算最终传递给宏观截面参数的不确定性是基本恒定的。其中,~(235)U平均裂变中子数反应、~(238)U辐射俘获反应、~(238)U共振非弹性散射反应及~1H共振弹性散射反应对组件宏观截面参数计算不确定性贡献尤为突出。同时,温度升高导致组件kinf及宏观截面参数计算不确定性增加;燃料富集度降低及可燃毒物的存在均使组件kinf计算不确定性增加;组件快群截面计算不确定性远大于热群截面计算不确定性。其中~(238)U辐射俘获反应、共振非弹性散射反应等截面信息应重点关注并且需要进一步评价和改进。     

钍基先进CANDU堆(TACR)钍-铀燃料功率影响研究

在CANDU堆燃料栅元物理的研究中,通常选择堆芯平均的燃料比功率对栅元进行计算模拟,而在TACR中,由于使用了钍燃料,比功率的不同就可能对核反应产生影响,并通过影响棒束栅元的基本截面参数而影响到全堆计算的结果.本文对不同定功率条件下,含全铀燃料和钍-铀燃料棒束的栅元截面参数随辐照值的变化以及钍燃料棒束中233Pa和233U的质量份额进行了计算分析,认为功率会对钍燃料的栅元宏观截面产生影响,在全堆计算中,栅元基本参数应尽量使用基于历史的局部参数法.     

钍基先进CANDU堆燃料循环方式研究

在钍基先进CANDU堆的概念设计中，钍燃料的循环利用方式是一重要问题。文章采用中心两圈为钍燃料、外面两圈为稍加浓缩铀燃料的CANFLEX燃料棒束，通过对燃料棒束栅元物理特性的研究，提出了一套切实可行的直接自身再循环的燃料棒束循环方案。     

乏燃料后处理萃取过程中的界面污物

在乏燃料后处理萃取工艺工程中，萃取剂和溶剂的辐解以及料液中不溶性固体微粒的存在导致产生界面污物。界面污物严重影响萃取柱的正常操作。文章就有关界面污物的研究状况进行概要评述。目前，普遍认为，在Purex流程萃取过程中，尤其是在一循环中，界面污物的产生与Zr和TBP降解产物HDBP、HzMBP、H3PO4形成的沉淀以及料液中存在的不溶性RuO2、Pd等微粒的表面化学现象有关。沉淀是否产生以及形成的界面污物类型与HDBP／Zr摩尔比和水相条件密切相关。此外，煤油等稀释剂的降解产物也是形成界面污物的一个不可忽略的因素。从萃取设备中排出的界面污物可用Na2CO3或草酸进行处理。在界面污物模拟实验中，需同时考虑Zr与TBP降解产物HDBP、H2MBP、H3PO4形成沉淀和不溶性微粒RuO2、Pd的影响等多种因素。     

真空环境下燃料组件压紧弹簧疲劳应力松弛研究

反应堆燃料组件压紧弹簧需要长期在高温和压应力下服役,其材料的正确选型对燃料组件的设计具有重要意义。针对反应堆运行过程中可能因机械振动和流致振动等引起的压紧弹簧疲劳破坏现象,本文对三种备选奥氏体不锈钢压紧弹簧(GH2132、632和GH4169)在450℃条件下进行了10万次疲劳循环的真空疲劳试验。试验结果表明:所有弹簧均保持形态完整,没有出现断裂失效现象。GH4169弹簧在试验后没有发现明显的析出物。同时其处于低层错能状态,位错运动形式更多的是单滑移,变形过程中塑性的累积困难,表现出较好的抗应力松弛性能。最终优选GH4169作为反应堆组件压紧弹簧备选材料,为反应堆燃料组件设计与制造提供一定参考。     

真空环境下燃料组件压紧弹簧疲劳应力松弛研究

反应堆燃料组件压紧弹簧需要长期在高温和压应力下服役,其材料的正确选型对燃料组件的设计具有重要意义。针对反应堆运行过程中可能因机械振动和流致振动等引起的压紧弹簧疲劳破坏现象,本文对三种备选奥氏体不锈钢压紧弹簧(GH 2132、632和GH4169)在450℃条件下进行了10万次疲劳循环的真空疲劳试验。试验结果表明:所有弹簧均保持形态完整,没有出现断裂失效现象。GH4169弹簧在试验后没有发现明显的析出物。同时其处于低层错能状态,位错运动形式更多的是单滑移,变形过程中塑性的累积困难,表现出较好的抗应力松弛性能。最终优选GH4169作为反应堆组件压紧弹簧备选材料,为反应堆燃料组件设计与制造提供一定参考。     

AFA 3G燃料组件骨架导向管与格架的压力电阻焊工艺研究

骨架点焊是压水堆燃料组件制造过程中的重要工序。对AFA 3G骨架点焊焊接技术，即格架与变径导向管之间的压力电阻点焊，在我国尚属一个崭新的技术课题。在大量实验的基础上，就不同的焊接规范对焊点质量的影响进行了详细的讨论与研究，包括焊点的剪切力、焊点熔核尺寸以及焊点腐蚀性能，从而获得了比较理想的焊接规范。