Study on small long-life LBE cooled fast reactor with CANDLE burn-up – Part I: Steady state research

Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction.

The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead–Bismuth is used as coolant.

From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%.     

Nuclear burning wave in fast reactor with mixed Th-U fuel

The possibility of creating a self-sustained regime of a running nuclear burning wave in the critical fast reactor with the mixed Th-U fuel is demonstrated. The calculations were performed in the deterministic approach based on solving the non-stationary multi-group diffusion equation of neutron transport together with the set of equations of the fuel component burn-up and the nuclear kinetics of precursor nuclei of delayed neutrons. The presence of the constructional material Fe and the coolant (the Pb-Bi eutectic) in the reactor composition is taken into account. The calculation results of the space-time evolution of neutron flux and fuel component concentrations are presented for different values of the Th-U ratio in the fuel. The calculations show the remarkable stability of the nuclear burning wave regime against neutron flux distortions in the reactor, which is a result of the negative feedback on reactivity inherent to this regime. This is one of the most important features of the reactor of this type, which ensures its intrinsic safety.     

超高通量快中子研究堆核燃料概念设计研究

提高中子注量率是高通量研究堆的发展趋势,能够大幅加速反应堆材料研发进程。但若提高中子注量率至1016 cm?2·s?1将导致功率密度峰值相较于现有研究堆高数倍,对反应堆和核燃料设计带来许多挑战。为此,本文从中子学、传热、燃料材料堆内行为等方面半定量分析了提高中子注量率对核燃料性能的影响,并提出应对超高通量和功率密度挑战的设计措施,为发展超高通量快中子研究堆燃料设计提供指导。      

Study of a self-regulated nuclear burn wave regime in a fast reactor based on a thorium–uranium cycle

The possibility of a wave of slow nuclear burning in a fast reactor in thorium–uranium fuel cycle is investigated. The calculations were performed using a model based on the solution of a nonstationary nonlinear diffusion equation for a cylindrical homogeneous reactor using the concept of a radial geometric factor (buckling) and the effective multigroup approximation taking account of the nuclear kinetics of the precursors of delay neutrons and burnup and production of the main nuclides of the thorium–uranium fuel cycle. The calculations showed that the generation and propagation of a wave of nuclear burning traveling with velocity approximately 2 cm/yr are possible in a thorium–uranium medium. However, the addition of even small quantities of a construction material and coolant to the composition of the reactor makes it impossible to obtain the burn wave regime. A self-maintained nuclear burn regime is also established in this case and exists for a long time (∼5 yr), but the system does not transition into a regime with a nuclear burn wave propagating along the axis of the reactor.     

超高通量快中子研究堆需求分析

实现超高快中子通量是世界先进研究堆的重要发展方向,对于加快第四代先进核能系统燃料及材料创新发展具有重要意义。本文从先进核能堆内结构材料与核燃料的辐照考验、长反应链超钚元素生产等角度,初步分析了我国建设超高通量快中子研究堆的必要性。在此基础上,确定了超高通量快中子研究堆的堆芯最大中子注量率及其冷却剂,给出了反应堆主要参数及冷却剂流动方案。反应堆热功率为200 MW,冷却剂为铅铋合金,最大中子注量率大于1016 cm?2·s?1。      

Characterization of neutron field in the experimental fast reactor JOYO for fuel and structural material irradiation test

The experimental fast reactor JOYO has been operated as an irradiation test facility for fast reactor fuel and structural material since 1983 with its MK-II core. During this time, an extensive study was conducted to characterize the neutron field in order to assure the accuracy and reliability of neutron fluence. Neutron flux for a given irradiation test was calculated using a core management code system based on three-dimensional diffusion theory. It was then corrected with the adjusted neutron spectrum by means of the multiple foil activation method. The neutron fluence calculation accuracy in the fuel region was evaluated within a 5% error by comparing the burn-up of spent fuel with the measured values, which had been obtained from their post-irradiation examination. At positions away from the fuel region, the neutron flux distribution was calculated using a two-dimensional transport code. A Monte Carlo code was also used to analyze the detailed neutron flux distribution within an irradiation test subassembly that had a heterogeneous internal structure. With the neutron flux results various irradiation parameters, such as displacement per atom (dpa) and helium production, could be evaluated. A helium accumulation fluence monitor has been developed to measure not only neutron fluence but also helium production. Neutron flux and fluence obtained from the core management calculations were compiled as a database for users’ convenience together with related irradiation information and fuel subassembly material compositions. These data are expected to be widely used in the post-irradiation analysis of fuel and structural material.     

Study for 228Th reduction in thermal reactor with Th-U fuel cycls

By using computercode WIMS/CENDL, the effects of some parameters, core configuration such as fuel element structure, neutron flux and burn-up, are discussed in thispaper. It is shown that high neutron flux, small fuel rod diameter,large volume ratio of coolant to fuel, seed-blank heterogeneous corearrangement and ²³¹Pa chemical separation are necessary for reducing ²²⁸Th production in reactor.     

Study for 228Th reduction in thermal reactor with Th—U fuel cycls

By using computer code WIMS/CENDL,the effects of some parameters,core configuration such as fuel element structure,neutron flox and burn-up,are discussed in this paper.It is shown that high neutron flux,small fuel rod diameter,large volume ratio of coolant to fuel,seed-blank heterogeneous core arrangement and 231 Pa chemical separation are necessary for reducing 228 Th production in reactor.     

Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead–bismuth, Pb–Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.     

Design research of small long life CANDLE fast reactor

The feasibility of a small long life fast reactor with CANDLE burn-up concept was investigated. It was found that a core with 1.0 m radius and 2.0 m length can bring about CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead–Bismuth is used as coolant. From equilibrium analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year, which easily permits a long core life design. The averaged core discharged fuel burn-up is about 40%. For better understanding of the effect of the coolant to fuel volume ratio, comparison was made among five cases. In these cases the coolant channel radii were different from one case to another, while fuel pin pitch was fixed. Comparisons were also made with a fixed coolant channel radius and different fuel pin pitches. A simulation of core operation is implemented and the results show that the present design can establish the long time steady CANDLE burn-up successfully without a burn-up control mechanism.     

Design concept of a small nuclear reactor for large-diameter NTD-Si using a conventional PWR full-length fuel assembly

A design concept for a small nuclear reactor dedicated to large-diameter neutron transmutation doping silicon (NTD-Si) is proposed. Conventional PWR (Pressurized Water Reactor) full-length fuel assembly is used to assure stable and reliable supply of fuel. Criticality, neutron transportation, and core burn-up calculations are performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the proposed reactor can be critical over 18 years, and excess reactivity can be suppressed by a combination of Gd₂O₃ burnable poison and soluble boron. Preliminary steady-state single-channel thermal hydraulic analysis showed that heat removal from core is possible under 1 atm operating pressure. Si ingots up to 30 cm in diameter can be irradiated in the reactor irradiation channels, and the uniform irradiation condition can be achieved for a large-diameter Si ingot.     

超高通量快中子试验堆堆芯初步概念设计

针对先进核能系统发展需要,提出了超高通量堆的堆芯概念设计。本文采用板型燃料、正方形燃料组件设计,设置宽流道保证堆芯冷却剂占有较高的体积份额。堆芯采用52盒燃料组件,设置8盒控制棒组件和较厚的反射层。通过堆芯概念设计方案评价,结果表明堆芯循环长度可达100EFPD(等效满功率天),所提出的超高通量堆的最大中子注量率可达到1.08×10¹⁶ cm^-2·s^-1。     

环形元件超高通量堆堆芯初步概念设计

基于环形燃料元件,提出了一种超高通量堆(UFR)堆芯概念设计。UFR燃料组件设计采用61个燃料元件构成的六角形组件,堆芯采用52盒燃料组件、9盒控制棒组件和厚反射层设计。通过开展堆芯概念设计方案评价,给出了堆芯循环长度、中子注量率、中子能谱、中子空间分布等关键参数。结果表明,在当前的总体参数下所提出的UFR的最大中子注量率可达到1.0×10¹⁶ cm^-2·s^-1。     

ECRIX-H experiment: Synthesis of post-irradiation examinations and simulations

The purpose of the ECRIX-H experiment is to study the behaviour of a composite ceramic target made of AmO_1.62 microdispersed in an MgO matrix irradiated for 318 EFPD in the Phenix sodium-cooled fast reactor (SFR), in a specific carrier sub-assembly equipped with annular blocks of CaH_x acting as a neutron moderator. Results indicate that magnesia-based inert matrix targets display satisfactory behaviour and moderate swelling under irradiation, even for significant quantities of helium produced and a high burn-up. On this basis, the design of transmutation fuel pins for recycling of minor actinides (MA) in accelerator-driven systems (ADS) or in fast neutron reactors (FR) could be optimised so as to increase their performance level (initial MA content, burn-up, etc.).The measured Am fission rate (25 at.%) was found to be lower than that predicted by neutronic simulations probably due to the inaccuracies linked to the complexity of neutron modelling and the uncertainties on nuclear data related to moderated neutron spectrum. In addition, as most of the initial Am transmuted into Pu under irradiation, a PuO_x-type phase was created within the initial AmO_1.62 particles, leading to the incomplete dissolution of the irradiated targets under standard reprocessing conditions. This issue will have to be considered and investigated in greater detail for all transmutation fuels and targets devoted to the multi-recycling of MA.     

Core concept of a passive-safety fast reactor “METAL-KAMADO” and reactivity coefficients

New concept of a passive-safety simple fast reactor “METAL-KAMADO” with metallic fuels is presented, which has same concept as a passive-safety thermal reactor “KAMADO”. A fuel element of the “METAL-KAMADO” consists of metallic fuel (U–10%Zr) and cooling holes of He gas flow. These fuel elements are located in a reactor water pool of atmospheric pressure (0.1 MPa) and low temperature (<60 °C). In case of LOF, decay heats of fuel elements are removed by natural heat transfer from surfaces of the fuel elements to the reactor water pool.

Preliminary neutronic calculations of the “METAL-KAMADO” show possibility of high burn-up of more than 120 GWd/t with 10% enriched U–Zr fuel. Reactivity coefficients of the core are also discussed.     

计算核反应堆堆芯功率的热平衡试验

反应堆功率运行时，燃耗变化会引起堆外中子通量密度变化，造成RPN核功率测量系统测得的反应堆功率与实际功率出现偏差。为了保证PRN反应堆堆芯功率测量的正确，大亚湾核电站利用热平衡的方法，即利用能量平衡原理计算反应堆堆芯的功率，然后对RPN测得的反应堆堆芯功率数据进行校核。本文主要对热平衡测量核反应堆堆芯功率的方法，计算原理进行全面的描述。     

Characteristics of Two-Reactor Electronuclear Systems

The Monte Carlo method is used to simulate electronuclear systems consisting of two-cascade subcritical zones: a liquid-metal fast reactor, which is used as a booster, and a thermal reactor, where most of the energy is released. Reactors of the type VVÉR-1000, MSBR-1000, and CANDU-6 are considered. The systems considered, functioning in the safe regime (k _eff = 0.94–0.98), possess much higher maximum power in the entire range of k _eff than similar systems without an intermediate fast booster reactor. At the same time, for high thermal neutron flux and with both fast and thermal zones present nuclear wastes can be efficiently transmuted in them, decreasing the required proton current in the beam by approximately a factor of 10. This is especially important when the liquid-salt thermal breeding reactors are considered as the basic electricity producing zone.