Comparative study of ADS for minor actinides transmutation

The methodology of the efficiency comparison of the different ADS-burners is discussed. ADS with lead–bismuth coolant (fast neutron spectrum), molten salt ADS (intermediate spectrum) and heavy water ADS (thermal spectrum) are chosen as representatives for the comparison. The first results of the suggested approach are discussed.     

Wave-driver options for low aspect ratio steady-state tokamak reactors

Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A=2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value.     

利用超长寿命快堆嬗变亚锕元素的特性研究

对利用超长寿命快堆（ＵＬＬＦＢＲ）嬗变高放核废物亚锕元素（ＡＭｓ）的堆芯物理特性进行了初步分析,得出了在ＵＬＬＦＢＲ中适当布置ＡＭｓ,既可以嬗变ＭＡｓ,又可改善超长寿命反应堆的物理特性这一结论。     

All-metal transformer core for a low aspect ratio tokamak

A novel concept for incorporating an iron core transformer within a axisymmetric toroidal plasma containment device with a high neutron flux is described. This design enables conceptual design of low aspect ratio devices which employ standard transformer-driven plasma startup by using all-metal high resistance inserts between the toroidal field windings. This design avoids the inherent problems of a multi-turn air core transformer which will inevitably suffer from strong neutron bombardment and hence lose the integrity of its insulation, both through long term material degradation and short term neutron induced conductivity. A full 3-dimensional model of the concept has been developed within the MAXWELL program and the resultant loop voltage calculated. The utility of the result is found to be dependent on the resistivity of the high resistance inserts. Useful loop voltage time histories have been obtained using expected resistivities.     

Neutronics analysis of minor actinides transmutation in a fusion-driven subcritical system

The minor actinides (MAs) transmutation in a fusion-driven subcritical system is analyzed in this paper. The subcritical reactor is driven by a tokamak D-T fusion device with relatively easily achieved plasma parameters and tokamak technologies. The MAs discharged from the light water reactor (LWR) are loaded in transmutation zone. Sodium is used as the coolant. The mass percentage of the reprocessed plutonium (Pu) in the fuel is raised from 0 to 48% and stepped by 12% to determine its effect on the MAs transmutation. The lesser the Pu is loaded, the larger the MAs transmutation rate is, but the smaller the energy multiplication factor is. The neutronics analysis of two loading patterns is performed and compared. The loading pattern where the mass percentage of Pu in two regions is 15% and 32.9% respectively is conducive to the improvement of the transmutation fraction within the limits of burn-up. The final transmutation fraction of MAs can reach 17.8% after five years of irradiation. The multiple recycling is investigated. The transmutation fraction of MAs can reach about 61.8% after six times of recycling, and goes up to about 86.5% after 25.     

Effect of minor actinides doping on plutonium produced in large-scale FBR blanket

The present study focuses on the effect of minor actinides (MAs) addition into the FBR blanket as ways of increasing fraction of even-mass-number plutonium isotopes, especially ²³⁸Pu, aiming at enhancing the proliferation resistance of plutonium produced in the blanket. The MA loading potential to enhance the proliferation resistance of plutonium is investigated, with considering actual design constraints on the fuel decay heat from the fuel handling and fabrication points of view, as MAs considerably generate decay heat. It reveals that depending on doping quantity of MAs, it is possible to denature produced plutonium by MA transmutation. MA addition in the blanket gives a significant increment in ²³⁸Pu fraction of generated plutonium but less effect on other even-mass-number plutonium isotopes. However, it is important that MA compositions should be adequately controlled to satisfy both the proliferation resistance requirements and the decay heat constraints for fuel handling.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

钠冷氧化物燃料快堆嬗变MA研究

研究了次量锕系核素(MA)在钠冷氧化物燃料快堆中嬗变的基本物理特性。结果表明,MA核素加入堆芯燃料中后对堆芯动态参数和反应性反馈会产生显著的影响,如:βeff会有所减小、多普勒负反馈会显著减弱以及钠空泡反应性正反馈会显著增强。添加MA所带来的收益是燃耗反应性损失减小,且一定量的MA被嬗变掉,同时MA裂变也有相应的能量产出。MA嬗变的本质在于MA的焚毁,MA的焚毁比消耗与其所占全堆的裂变份额(包括由其转换的238Pu的裂变)成正比,为此相同MA裂变份额下的堆芯安全参数成为MA嬗变快堆设计的关键点。研究表明,堆芯小型化能够有效地减小堆芯的钠空泡反应性正反馈,同时对MA的焚毁比消耗影响较小。     

加速器驱动钠冷金属燃料快堆次锕系核素嬗变特性研究

分析加速器驱动系统(ADS)钠冷金属燃料快堆重金属燃料不同核素对堆芯有效增殖系数(Keff)的影响,给出了燃料成分的确定方法,详细分析次锕系核素(MA)嬗变特性.运用耦合了MCNP4c3与ORJGEN2的三维燃耗程序COUPLE对堆芯进行稳态与燃耗计算.结果分析表明,调节燃料中239Pu的质量比例并使其在燃耗过程中保持稳定是使Keff达到设计值并在燃耗过程中保持稳定的有效手段.散裂中子引起堆芯内区较外区更硬的中子能谱,有利于提高MA的裂变截面与裂变吸收比.全堆MA嬗变支持比为8.3,具有较好的嬗变效果.由于堆芯内区的高通量,堆芯内外区的嬗变率有明显差异,将MA集中布置于内区有利于减少装料量,改善总体嬗变效果.     

Monte Carlo calculations on transmutation of plutonium and minor actinides of pebble bed high temperature reactor

The paper aimed to maximize the fuel burnup performance of plutonium and minor actinides fueled pebble bed high temperature reactor (PBMR-400). The PBMR-400 was designed as a reference core. The neutronic calculations were performed by the code combination MCNP-ORIGEN-MONTEBURNS. In this study, neutronic performances of three different types of nuclear fuels (Reactor Grade Plutonium – RGPu; Weapon Grade Plutonium – WGPu and Minor Actinides – MAs) combined with natural uranium were conducted in a PBMR-400 full core. The neutronic performances were compared with the original uranium fuel designed for this reactor. Neutronic calculations showed that 9.6 wt % enriched uranium has a core effective multiplication factor (k_eff) of 1.2395. Corresponding to this k_eff values the natural UO₂/RG-PuO₂; natural UO₂/WG-PuO₂ and natural UO₂/MAO₂ mixture were found 70%/30%, 76%/24% and 63%/37%, respectively. The operation times were computed as ∼2000, ∼2500 and 1400 days whereas, the corresponding burnup values were obtained as ∼163 000, ∼194 000 and ∼116 000 MWD/T, respectively, for end of life k_eff set equal to 1.08.     

Influence of void fraction change on plutonium and minor actinides recycling in BWR with equilibrium burnup

A study on the influence of void fraction change on plutonium and minor actinides recycling in standard boiling water reactor (BWR) with equilibrium burnup model has been conducted. We considered the equilibrium burnup model since it is a simple time independent burnup method that can handle all possible produced nuclides in any nuclear system.

The uranium enrichment for the criticality of the reactor diminishes significantly for the plutonium and minor actinides recycling case compared to that of the once-through cycle of BWR case. This parameter decreases much lower with the increasing of the void fraction. A similar propensity was also shown in the required natural uranium per annum. The annual required natural uranium was calculated by assuming that the uranium concentration in the tail of the enrichment plant is 0.25 w%. The amount of loaded fuel reduces slightly with the increment of the void fraction for plutonium and minor recycling in BWR.     

Lower hybrid heating and current drive system for a tokamak reactor

A lower hybrid heating system has been designed for heating a tokamak reactor to ignition and for sustaining steady-state operation by driving the toroidal plasma current. The power spectrum from an active/passive waveguide grill is computed, and the resulting equilibrium current density profile is computed from a full electromagnetic WKB analysis of wave propagation in a cylinder. The corresponding toroidal current profile is a low-current equilibrium which is stable to various ideal modes at an economically acceptable beta. The electronic circuitry is designed to minimize the electric power required for current drive, and the resulting design appears to provide reliable operation in a reactor environment. The same system can drive current during reactor startup if some of the waveguides are modified slightly. A typical sequence of startup equilibria is calculated.     

Extension of core life-time in fast breeder reactor by introducing inner blanket and doping minor actinides

In order to achieve a longer-life Fast Breeder Reactor (FBR) compared with conventional one, the feasibility study based on proto-type large scale sodium cooled FBR has been performed by utilizing a characteristic of a fertile material of minor actinides (MA) and an inner blanket arranged radially at the center of the core. The analytical results showed that the long-life core without the inner blanket could be achieved by doping MA into an active core because ²³⁸Pu transmuted from MA worked as the fissile material. In case of the core with the inner blanket, it was found that if MA is doped into the inner blanket, the longer-life core also could be achieved by shifting of the main fission reaction zone geometrically from the active core to the inner core due to producing of ²³⁸Pu in the inner blanket. It was also found that if MA is doped into both the inner blanket and the active core, the core life can be extended further. As for the safety characteristics, it has been confirmed that the sodium loss reactivity is improved in case of introducing the inner blanket due to the enhancement of neutron leakage. It has also been confirmed that the sodium loss reactivity is largely affected if the region of high neutron flux, that is the region of main fission reaction is voided.     

Actinide destruction and power peaking analysis in a 1000 MWt advanced burner reactor using moderated heterogeneous target assemblies

The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.     

MOX燃料模块快堆的嬗变研究

介绍了模块快堆的堆芯概念方案和模块快堆的嬗变特性 ,并论述了在适当时候引入模块快堆对削减我国次锕系核素积累量增长的作用     

Oclator: The one coil low aspect toroidal reactor

A new approach to construct a tokamak-type reactor(s) is presented here. Basically, the return conductors of toroidal field coils are eliminated and the toroidal field coil is replaced by one single large coil, around which will be placed several tokamaks or other toroidal devices. The elimination of return conductors should, in addition to other advantages, improve the accessibility and maintainability of the tokamaks, and offer a possible alternative to the search for special materials to withstand large neutron wall loading, as the frequency of changeover would be increased due to minimum downtime. It also makes it possible to have a low aspect ratio tokamak, which should improve the limit, so that a low toroidal magnetic field strength might be acceptable, meaning that NbTi superconducting wire could be used. This system is named OCLATOR (One Coil Low Aspect Toroidal Reactor).     

Bootstrap current fraction scaling for a tokamak reactor design study

We have derived new bootstrap current fraction scalings for systems codes by solving the Hirshman–Sigmar model, which is valid for arbitrary aspect ratios and collision conditions. The bootstrap current density calculation module in the ACCOME code was used with the matrix inversion method without the large aspect ratio assumption. Nine self-consistent MHD equilibria, which cover conventional, advanced and spherical tokamaks with normal or reversed shear, were constructed using numerical calculations in order to compare the bootstrap current fraction values with those of the new model and all six existing models. The Wilson formula successfully predicted the bootstrap current fraction, but it requires current density profile index for the calculation. The new scaling formulas and IPDG accurately estimated the bootstrap current fraction for the normal and weakly reversed shear tokamaks, regardless of the aspect ratio. However, none of the existing models except the Wilson formula can accurately estimate the bootstrap current fraction for the reversed shear tokamaks, which is promising for the advanced tokamak operation mode.     

Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy.     

Design and optimization of an equilibrium reload with MOX fuel with minor actinides

In this work the design and optimization of an equilibrium core for a boiling water reactor (BWR), loaded with fuel composed of plutonium and minor actinides (Np, Am and Cm), is presented. The plutonium and minor actinides are obtained from the recycling of the spent fuel of a BWR, and are mixed with depleted uranium obtained from the enrichment tails. The design and optimization of the equilibrium reload is achieved in two steps. In the first step, the fuel assembly is adjusted and the reload pattern is designed, in order to obtain the target cycle length. In order to improve the shutdown margin, two actions were taken: to increase the boron-10 content in the control rods, and to add a burnable absorber (gadolinia) in some fuel rods. In the second step, the reload pattern, obtained in the first step, is optimized to maximize the energy, under the thermal and reactivity margins constraints; a system based on Genetic Algorithms was used in the optimization process. Results show that 5% more energy was obtained with the optimized reload.