The concept and neutronic characteristics of a multipurpose fast reactor (MPFR)

The paper describes a concept and its neutronic characteristics of a long-life multipurpose nuclear reactor with self-sustained liquid metallic fuel, which is proposed to meet the requirements for the energy production in the future. The application of the liquid plutonium–uranium metallic alloys as a nuclear fuel demonstrates high potential to reach excellent conversion characteristics and high burn-up values with relatively small reactivity swings. Considerations about the capability of in-vessel separation of the main fission products to prolong the core lifetime without fuel reloading or shuffling are also discussed from the viewpoint of the influence on core burn-up characteristics.     

Safety characteristics of the multipurpose fast reactor (MPFR)

The safety characteristics of a long-life multipurpose nuclear reactor (MPFR) with self-sustained liquid metallic fuel and lead coolant, which is proposed to meet the requirements for the energy production in the future, were investigated. The application of liquid plutonium–uranium metallic alloys used as a nuclear fuel demonstrated high potential to reach excellent reactor shutdown characteristics against anticipated transients without scram such as unprotected loss-of-flow and unprotected transient overpower. The calculations indicated that the thermal expansion of liquid fuels would cause the negative reactivity insertion that would be larger in magnitude than any other thermally induced reactivity changes. This created the reactivity balance for the passive shutdown and power stabilization capabilities of the MPFR core. It was found that MPFR satisfies such design characteristics to be a potential candidate providing the replacement of fossil fuels by alternative energy sources in the next century.     

Long life multipurpose small size fast reactor with liquid metallic-fueled core

A concept of a long life multipurpose nuclear reactor with self-sustained liquid metallic fuel is proposed to meet the requirements for the future energy production. The conceptual design is described and the core neutronic characteristics are obtained based on two-dimensional cylindrical diffusion reactor model. The influence of fission products separation in the liquid-fueled system on the core burnup capability is discussed. The burnup analysis shows a feasibility of the long life refueling-free core concept.     

商用压水堆核燃料研发进展与应用展望

压水堆（PWR）是目前核电厂反应堆的主力堆型,而核燃料是反应堆的能量源泉和放射性裂变物质的主要来源,关乎核电厂的经济性和安全性。本文对当前国际上面向商用PWR应用研发的掺杂UO₂燃料、高铀密度燃料、微封装燃料和金属燃料的性能特点、技术状态及前景进行了归纳和评价。在掺杂UO₂燃料中,大晶粒燃料具有较高的技术成熟度,将在PWR实现大规模商用;高铀密度燃料和金属燃料在高温水腐蚀氧化问题以及事故下的行为仍待研究解决;具有极致安全的微封装燃料更适合特殊用途的小型反应堆。应协同开展先进燃料组件设计、建立设计准则以及研发高保真的性能分析技术等,以充分发挥新型燃料的可靠性及高燃耗优势。      

Fission gas production in reactor fuels including the effects of ternary fission

An understanding of gas bubble formation and migration in nuclear fuel and its impacts on fuel and cladding materials requires knowledge of the isotopic composition of the gases and their generation rates. In this paper, we present results of simulations for the production of the dominant noble gases (helium, xenon, krypton) in nuclear fuels for different reactor core configurations and fuel compositions. The calculations were performed using detailed nuclear burn simulations with Monte Carlo nuclear transport, and included ternary fission to ensure an accurate treatment of helium production. For all reactor designs and fuels considered xenon was found to be the most dominant gas produced. Variation in the composition of fission gases is quantified for: (1) the burn time, (2) the composition of the fuel, and (3) the neutron energy spectrum. These three factors determine the relative fraction of each gas and its transmutation into or from stable gas by subsequent neutron capture.     

长寿命小型自然循环铅基快堆燃料选型

针对铅基快堆长寿命、小型化、自然循环的设计目标，构建铅基快堆堆芯模型并开展燃料选型研究，选取U-Pu、Th-U循环燃料及氧化物、氮化物、碳化物、金属燃料，分析比较了不同燃料的物性参数、在不同能谱条件下的堆芯物理特性。结果表明：在偏软能谱中，Th基燃料堆芯增殖能力更强，反应性系数负值更大，热工安全裕量更大、裂变产物容留能力更强；PuN-ThN燃料堆芯燃耗特性最佳，可在较疏松栅格条件下获得较强增殖能力，减少燃料装载量，确保固有安全性，兼顾堆芯长寿命、小型化、自然循环设计要求；但堆芯有效缓发中子份额较小，不利于反应性控制。     

A source efficient ADS for minor actinides burning

Taking advantage of the good neutron economy of nitride fuel, a compact accelerator-driven system (ADS) for burning of minor actinide fuels has been designed, based on the fuel assembly geometry developed for the European Facility for Industrial Transmutation (EFIT) within the EUROTRANS project. The small core size of the new design permits reduction of the size of the spallation target region, which enhances proton source efficiency by about 80% compared to the reference oxide version of EFIT. Additionally, adoption of the austenitic steel 15/15Ti as clad material allows to safely reduce the fuel pin pitch, which leads to an increase of fuel volume fraction and therefore makes the neutron energy spectrum faster, consequently increasing minor actinides fission probabilities. Our calculations show that one can dramatically increase neutron source efficiency up to 0.95 without a significant loss of neutron source intensity, i.e. having high proton source efficiency. Consequently, the accelerator current required for operation of the ADS with a fission power of 201 MWth and a burn-up of 27 GW d/t per year (365 EFPD) is reduced by 67%.     

Temperature distribution in mixed ThO2–UO2 fuel rods located in blanket of an inertial fusion energy breeder

Investigations of neutronic analysis and temperature distribution in fuel rods located in a blanket driven ICF (Inertial Confinement Fusion) have been performed for various mixed fuels and coolants under a first wall load of 5 MW/m². The fuel rods containing ThO₂ and UO₂ mixed by various mixing methods for achieving a flat fission power density are replaced in the blanket and cooled with different coolants; natural lithium, flibe, eutectic lithium and helium for the nuclear heat transfer. It is assumed that surface temperature of the fuel rod increases linearly from 500 °C (at top) to 700 °C (at bottom) during cooling fuel zone. Neutronic and temperature distribution calculations have been performed by MCNP4B Code and HEATING7, respectively. In the blanket fueled with pure UO₂ and cooled with helium, M (fusion energy multiplication ratio) increases to 3.9 due to uranium having higher fission cross-section than thorium. The high fission energy released in this blanket, therefore, causes proportionally increasing of temperature in the fuel rods to 823 °C. However, the M is 2.00 in the blanket fueled with pure ThO₂ and cooled with eutectic lithium because of more capture reaction than fission reaction. Maximum and minumum values of TBR (tritium breeding ratio) being one of main neutronic paremeters for a fusion reactor are 1.07 and 1.45 in the helium and the natural lithium coolant blanket, respectively. These consequences bring out that the investigated reactor can produce substantial electricity in situ during breeding fissile fuel and can be self-sufficient in the tritium required for the DT fusion driver in all cases of mixed fuels and coolant types. Quasi-constant fission power density profiles in FFB (fissile fuel breeding) zone are obtained by parabolically increasing mixture fraction of UO₂ in radial and axial directions for all coolant types. Such as, in the helium coolant blanket and the case of PMF (parabolically mixed fuel), Γ (peek-to-average fission power density ratio) of the blanket is reduced to 1.1, and the maximum temperatures of the fuel rods in radial direction of the FFB zone are also quasi-constant. At the same time, in the case of PMF, for all coolant types, the temperature profiles in the radial direction of the fuel rods rise proportionally with surface temperature from the top to the bottom of fuel rods in the axial direction. In other words, for each radial temperature profile in the axial direction, temperature differences between centerline and surface of the fuel rods are quasi-constant. According to the coolant types, these temperature diffences vary between 30 and 45 °C.     

加速器驱动先进核能系统的乏燃料循环再生研究北大核心CSCD

乏燃料后处理是核燃料循环的关键环节,制约核电的可持续发展。借助于加速器驱动先进核能系统(ADANES)提供的高通量、硬能谱的外源中子,其乏燃料后处理只需除去乏燃料中的挥发性裂变产物和影响次锕系元素嬗变的中子毒物,长寿命的次锕系元素Np、Am、Cm可与二氧化铀一起转化为新的燃料元件在加速器驱动燃烧器中燃烧、嬗变、增殖和产能。基于此,本课题组提出了加速器驱动的乏燃料后处理及再生制备的技术路线,包括高温氧化粉化与挥发、选择性溶解分离和燃料再生制备。本文主要介绍了近几年本课题组在这三方面所取得的一些成就,希望能为加速器驱动先进核能系统的乏燃料后处理提供基础数据。     

Gallium-cooled liquid metallic-fueled fast reactor

We have examined the effects on core characteristics of using two different types of Pu-based metallic alloy fuels in the gallium-cooled fast reactor core. In the proposed concept, the liquid metal fast nuclear reactor uses metallic fuel in the liquid phase and gallium coolant at high temperature (inlet 1700K, outlet 1900K). The liquid fuel is continuously supplied to the reactor during operation at full reactor power. The reactor power is controlled by rotational control drums with absorber material. The aim was to evaluate reactor core neutronics and safety characteristics demonstrating a feasibility of the reactor system. Although gallium has large absorption cross section in the high neutron energy region, we can design the core with rather good neutronics performances. The large negative reactivity feedback induced by the thermal expansion of liquid metallic fuel ensures the core's inherent safety against the unprotected loss-of-flow transient.     

聚变-裂变混合堆水冷包层中子物理性能研究

研究直接应用国际热核聚变实验堆（ITER）规模的聚变堆作为中子驱动源，采用天然铀为初装核燃料，并采用现有压水堆核电厂成熟的轻水慢化和冷却技术，设计聚变-裂变混合堆裂变及产氚包层的技术可行性。应用MCNP与Origen2相耦合的程序进行计算分析，研究不同核燃料对包层有效增殖系数、氚增殖比、能量放大系数和外中子源效率等中子物理性能的影响。计算分析结果显示，现有核电厂广泛使用的UO₂核燃料以及下一代裂变堆推荐采用的UC、UN和U₉₀Zr₁₀等高性能陶瓷及合金核燃料作为水冷包层的核燃料，都能满足以产能发电为设计目标的新型聚变 裂变混合堆能量放大倍数的设计要求，但只有UC和U₉₀Zr₁₀燃料同时满足聚变燃料氚的生产与消耗自持的要求。研究结果对进一步研发满足未来核能可持续发展的新型聚变-裂变混合堆技术具有潜在参考价值。     

Molten salts and nuclear energy production

Molten salts (fluorides or chlorides) were considered near the beginning of research into nuclear energy production. This was initially due to their advantageous physical and chemical properties: good heat transfer capacity, radiation insensitivity, high boiling point, wide range solubility for actinides. In addition it was realised that molten salts could be used in numerous situations: high temperature heat transfer, core coolants with solid fuels, liquid fuel in a molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing and coolant and tritium production in the case of fusion. Molten salt reactors, one of the six innovative concepts chosen by the Generation IV international forum, are particularly interesting for use as either waste incinerators or thorium cycle systems. As the neutron balance in the thorium cycle is very tight, the possibility to perform online extraction of some fission product poisons from the salt is very attractive. In this article the most important questions that must be addressed to demonstrate the feasibility of molten salt reactor will be reviewed.     

South Africa's opportunity to maximise the role of nuclear power in a global hydrogen economy

Global concern for increased energy demand, increased cost of natural gas and petroleum, energy security and environmental degradation are leading to heightened interest in using nuclear energy and hydrogen to leverage existing hydrocarbon reserves. The wasteful use of hydrocarbons can be minimised by using nuclear as a source of energy and water as a source of hydrogen. Virtually all hydrogen today is produced from fossil fuels, which give rise to CO₂ emissions. Hydrogen can be cleanly produced from water (without CO₂ pollution) by using nuclear energy to generate the required electricity and/or process heat to split the water molecule. Once the clean hydrogen has been produced, it can be used as feedstock to fuel cell technologies, or in the nearer term as feedstock to a coal-to-liquids process to produce cleaner synthetic liquid fuels. Clean liquid fuels from coal - using hydrogen generated from nuclear energy - is an intermediate step for using hydrogen to reduce pollution in the transport sector; simultaneously addressing energy security concerns. Several promising water-splitting technologies have been identified. Thermo-chemical water-splitting and high-temperature steam electrolysis technologies require process temperatures in the range of 850 °C and higher for the efficient production of hydrogen. The pebble bed modular reactor (PBMR), under development in South Africa, is ideally suited to generate both high-temperature process heat and electricity for the production of hydrogen. This paper will discuss South Africa's opportunity to maximise the use of its nuclear technology and national resources in a global hydrogen economy.     

Innovations in the ENHS reactor design and fuel cycle

A number of approaches were explored for improving characteristics of the encapsulated nuclear heat source (ENHS) reactor and its fuel cycle, including: increasing the ENHS module power, power density and the specific power, making the core design insensitive to the actinides composition variation with number of fuel recycling and reducing the positive void coefficient of reactivity. Design innovations examined for power increase include intermediate heat exchanger (IHX) design optimization, riser diameter optimization, introducing a flow partition inside the riser, increasing the cooling time of the LWR discharged TRU, increasing the minor actinides' concentration in the loaded fuel and split-enrichment for power flattening. Another design innovation described utilizes a unique synergism between the use of MA and the design of reduced power ENHS cores.

Also described is a radically different ENHS reactor concept that has a solid core from which heat pipes transport the fission power to a coolant circulating around the reflector. Promising features of this design concept include enhanced decay heat removal capability; no positive void reactivity coefficient; no direct contact between the fuel clad and the coolant; a core that is more robust for transportation; higher coolant temperature potentially offering higher energy conversion efficiency and hydrogen production capability.     

兆瓦级空间气冷堆系统启停堆瞬态分析

本文针对兆瓦级高温气冷堆布雷顿循环系统,采用Fortran语言开发系统分析程序TASS,包括堆芯、透平-发电机-压气机、回热器、冷却器和热管式辐射散热器等模型。通过设计值与程序计算值对比对TASS进行验证,并利用TASS对系统启动、停堆瞬态工况进行数值模拟。结果显示,通过分两阶段、阶梯式引入正反应性和提高涡轮机械的转轴速度,堆芯流量和功率匹配良好,系统可在3.5 h内完成启动过程,达到反应堆功率3 406 kW、流量14.2 kg/s的稳态运行。系统停堆过程中,反应堆可依靠自身的非能动余热排出能力,确保芯块和包壳温度与熔点间存在较大安全裕量,实现安全停堆。     

Uranium-zirconium hydride fuel properties

Properties of the two-phase hydride U_0.3ZrH_1.6 pertinent to performance as a nuclear fuel for LWRs are reviewed. Much of the available data come from the Space Nuclear Auxiliary Power (SNAP) program of 4 decades ago and from the more restricted data base prepared for the TRIGA research reactors some 3 decades back. Transport, mechanical, thermal and chemical properties are summarized. A principal difference between oxide and hydride fuels is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium particles has been observed in hydride fuels. Avoidance of this source of swelling limits the maximum fuel temperature to ∼650 °C (the design limit recommended by the fuel developer is 750 °C). To satisfy this temperature limitation, the fuel-cladding gap needs to be bonded with a liquid metal instead of helium. Because the former has a thermal conductivity ∼100 times larger than the latter, there is no restriction on gap thickness as there is in helium-bonded fuel rods. This opens the possibility of initial gap sizes large enough to significantly delay the onset of pellet-cladding mechanical interaction (PCMI). The large fission-product swelling rate of hydride fuel (3× that of oxide fuel) requires an initial radial fuel-cladding gap of ∼300 m if PCMI is to be avoided. The liquid-metal bond permits operation of the fuel at current LWR linear-heat-generation rates without exceeding any design constraint. The behavior of hydrogen in the fuel is the source of phenomena during operation that are absent in oxide fuels. Because of the large heat of transport (thermal diffusivity) of H in ZrH_x, redistribution of hydrogen in the temperature gradient in the fuel pellet changes the initial H/Zr ratio of 1.6 to ∼1.45 at the center and ∼1.70 at the periphery. Because the density of the hydride decreases with increasing H/Zr ratio, the result of H redistribution is to subject the interior of the pellet to a tensile stress while the outside of the pellet is placed in compression. The resulting stress at the pellet periphery is sufficient to overcome the tensile stress due to thermal expansion in the temperature gradient and to prevent radial cracking that is a characteristic of oxide fuel. Several mechanisms for reduction of the H/Zr ratio during irradiation are identified. The first is transfer of impurity oxygen in the fuel from Zr to rare-earth oxide fission products. The second is the formation of metal hydrides by these same fission products. The third is by loss to the plenum as H₂.The review of the fabrication method for the hydride fuel suggests that its production, even on a large scale, may be significantly higher than the cost of oxide fuel fabrication.     

反应堆用SiC陶瓷基复合包壳材料研究进展

核燃料元件的包壳材料是反应堆安全的重要屏障。随着核动力反应堆向高燃耗、长燃料循环寿命、高安全性趋势的发展,传统Zr合金包壳材料因其铀燃耗极限(62 MW·d/kg)、高温腐蚀、氢脆、蠕变、辐照生长、芯/壳反应等缺陷,已不能满足未来第四代核能系统燃料元件对包壳材料的苛刻要求。Si C因其更小的中子吸收截面、低衰变热、高熔点及优异的辐照尺寸稳定性等优点,以Si C为基体的陶瓷基复合材料成为新一代包壳材料研究的热点。结合Si C的晶体结构、热物理特性,对其在第四代核反应堆包壳材料中的设计思路、中子辐照效应、热一力性能、与UO,的化学反应等进行了概述,对SiC基复合材料在未来核能领域的应用前景进行了展望。     

Core temperature transient in the early phase of core uncovering at TMI-2

This paper describes a best-estimate analysis of the initial core boil-down and heat-up transient at Three Mile Island Unit (2) on 28 March 1979. This transient began shortly after all reactor coolant pumps were secured (100 min after reactor trip) and was terminated by a period of sustained high pressure injection of emergency cooling water, starting at 202 min.

The analysis is primarily directed to understanding the progression of core damage, rather than providing a detailed characterization of the core end-state condition. The latter objective can be achieved only after vessel head removal and visual examination.

The thrust of the present effort has been to: (1) develop a core coolant mixture level (dry-out level) calculation which satisfies the boundary conditions implied by various instrument responses and system operational characteristics; (2) couple the level calculation with a core heat-up modelto simulate the accumulation of thermal damage in the exposed, upper regions of the core; (3) compare calculated gross damage to the core with measurements of hydrogen and fission product releases subsequent to the accident.

Results indicate that:

1. (i) Observed containment hydrogen levels were due to Zircaloy/stainless steel corrosion that occurred during the period of core uncovering between the de-activation of the loop A reactor coolant pump (100 min after trip) and sustained operation of the high pressure injection system 100 min later. Appreciable zircaloy oxidation probably commenced at 150 min after trip, and continued at a high rate until the sustained high pressure injection at 202 min caused a major core quench.

2. (ii) There was some potential for fuel liquefaction. Calculations imply that peak fuel temperatures did not exceed the UO₂ pellet melting temperature, but 30% of the fuel was exposed to temperatures where liquid U---Zr---O alloys could have formed.

3. (iii) A substantial fission product release was obtained from fuel over-heating; however, an apparent disparity between the expected fission product release by calculation and the high range of fission product estimates obtained from plant measurements suggests that a significant release fraction may have originated from powdered or rubbilized fuel during cooldown. Additional gas releases may have developed from hot spots which persisted after core quench.

4. (iv) Steam temperatures in the upper plenum, at the outlet nozzle elevation, were generally below 900°C (1650°F) although this value was probably exceeded for a few min during the partial fuel quench caused by activation of the loop 2B reactor coolant pump, at 174 min after trip. The metal-work in the upper plenum, above the upper tieplate did not experience appreciable heating.

Thermal damage to the fuel and consequential weakening and mechanical disruption of the core was essentially complete 230 min after turbine trip.     

Numerical Study on Effects of Fuel Mixture Fraction and Li-6 Enrichment on Neutronic Parameters of a Fusion–Fission Hybrid Reactor

This study presents the effects of mixture fractions of nuclear fuels (mixture of fissile–fertile fuels and mixture of two different fertile fuels) and ⁶Li enrichment on the neutronic parameters (the tritium breeding ratio, TBR, the fission rate, FR, the energy multiplication ratio, M, the fissile breeding rate, FBR, the neutron leakage out of blanket, L, and the peak-to-average fission power density ratio, Γ) of a deuterium–tritium (D–T) fusion neutron-driven hybrid blanket. Three different fertile fuels (²³²Th, ²³⁸U and ²⁴⁴Cm), and one fissile fuel (²³⁵U) were selected as the nuclear fuel. Two different coolants (pressurized helium and natural lithium) were used for the nuclear heat transfer out of the fuel zone (FZ). The Boltzmann transport equation was solved numerically for obtaining the neutronic parameters with the help of the neutron transport code XSDRNPM/SCALE4.4a. In addition, these calculations were performed by also using the MCNP4B code. The sub-limits of the mixture fractions and ⁶Li enrichment were determined for the tritium self-sufficiency. The considered hybrid reactor can be operated in a self-sufficiency mode in the cases with the fuel mixtures mixed with a fraction of equal to or greater than these sub-limits. Furthermore, the numerical results show that the fissile fuel breeding and fission potentials of the blankets with the helium coolant are higher than with the lithium coolant.     

A new paradigm for core design aimed at the sustainability of nuclear energy: The solution of the extended equilibrium state

The future expansion of nuclear energy, a technology identified as one of the main candidates for reducing the world’s dependence on fossil fuels, requires a thorough analysis of the sustainability of this energy source for long-term supply. Generation-IV nuclear systems could represent a turning point for energy production by minimizing the environmental footprint of the fuel cycle. A new paradigm is thus required for reactor design, focusing, at the core design level, on both the closure of the fuel cycle and the effective utilization of natural resources.