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.     

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of ²³⁷Np, ²³⁸Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of ²³⁷Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO₂ or UO₂–Gd₂O₃), which was similar to the case of PWR fuel. The differences in the number densities of ²³⁵U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu, ¹⁴⁹Sm, and ¹⁵¹Sm have a large impact on k_eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.     

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes.     

Comparison of accelerator-based with reactor-based nuclear waste transmutation schemes

An overview of the most significant studies in the last 35 years of partitioning and transmutation of commercial light water reactor spent fuel is given. Recent Accelerator-based Transmutation of Waste (ATW) systems are compared with liquid-fuel thermal reactor systems that accomplish the same objectives. If no long-lived fission products (e.g., ⁹⁹Tc and ¹²⁹I) are to be burned, under ideal circumstances the neutron balance in an ATW system becomes identical to that for a thermal reactor system. However, such a reactor would need extraordinarily rapid removal of internally-generated fission products to remain critical at equilibrium without enriched feed. The accelerator beam thus has two main purposes (1) the burning of long-lived fission products that could not be burned in a comparable reactor's margin (2) a relaxing of on-line chemical processing requirements without which a reactor-based system cannot maintain criticality. Fast systems would require a parallel, thermal ATW system for long-lived fission product transmutation. The actinide-burning part of a thermal ATW system is compared with the Advanced Liquid Metal Reactor (ALMR) using the well-known Pigford-Choi model. It is shown that the ATW produces superior inventory reduction factors for any near-term time scale.     

Interaction of uranium-molybdenum fuel with an aluminum matrix with deep burnup

A thermodynamic analysis is used to calculate the phase and component composition of uraniummolybdenum fuel with burnup 200 GW·days/ton. The equilibrium composition of the gas phase, consisting mainly of gaseous cesium whose pressure reaches 30 kPa, is determined more accurately. The quantitative composition of the phase of solid solutions of tellurides, whose formation degrades the structure of a fuel granule, is presented. Thermal tests of the fuel composition (U–Mo)–Al were performed. The investigation was performed in the presence of simulators of the chemically active fission products of cesium and iodine at different temperature. The interaction zone of (U–Mo)–Al is investigated by means of metallography and scanning electron microscopy. The data obtained on the composition of the indicated zone made it possible to conjecture the character of interaction between the fuel material and the aluminum matrix.     

Metallic fuel for Na-cooled reactors with an extended burnup cycle

In recent years, various reactors and fuel-cycle concepts have been proposed as alternatives to the (Pu-U)O₂ mixed-oxide fuel cycle. This interest has been stimulated by the need to utilize the U resources and also to contribute to the solution of the proliferation problem. To date, essentially all combinations of fuel-cycle mixes have been considered, except the denatured FBR operating on an extended burnup cycle. The basic feature of the proposed concept is a ²³³U²³⁸U LMFBR using metallic fuel, enriched at the beginning of life to about 6 at. % cooled with Na, and designed to operate in such a way that, once the reactor is built, it only needs natural or depleted U as feed for the rest of the life of the reactor. The denatured breeder simply enriches the U to the level necessary to maintain criticality. Calculations show that the reactivity swing over each refueling interval, the fuel-pin performance and some safety parameters are all within current technology constraints.     

基于FCM燃料的商业压水堆中子学分析

全陶瓷微封装(Fully Ceramic Microencapsulated,FCM)燃料是一种将三结构同向性型(Tri-structural isotropic,TRISO)燃料颗粒弥散于SiC基质的先进燃料,具有良好的包容裂变产物的能力,能有效地改善核燃料在严重事故下保持结构完整性的能力,有利于降低核电站发生大量放射性物质泄漏的风险,是耐事故燃料(Accident Tolerant Fuel,ATF)的主要研究方向之一。与传统UO_2陶瓷芯块燃料相比,FCM燃料的U装量较少,且燃料基体采用SiC,慢化能力较好,可能导致FCM燃料应用于商业压水堆时寿期初慢化剂温度系数为正,不能满足堆芯的固有安全性。本文以标准AFA3G 17×17栅格形式的UO_2-Zr合金燃料组件为参照对象,采用中核集团自主研发的NESTOR软件,分析了17×17和13×13两种栅格形式的FCM燃料(UN核芯)组件的中子学特性,评价了由13×13栅格形式的FCM燃料(UN核芯)组成反应堆堆芯的总体物理特性。研究表明:含钆可燃毒物的13×13栅格形式的FCM燃料(UN核芯)组件可满足欠慢化要求,13×13栅格形式的FCM燃料(UN核芯)用于大型商业压水堆堆芯的慢化剂温度系数可以为负,首循环堆芯可达到与参照堆芯接近的燃耗深度与循环长度,能初步满足商业压水堆堆芯的固有安全性和经济性的要求。     

Neutronics analysis for MSR cell with different fuel salt channel geometries

The neutronic properties of molten salt reactors (MSRs) differ from those of traditional solid fuel reactors owing to their nuclear fuel particularity. Based on...     

Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc…). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called “BUCAL1”. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k_∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.     

VVER nuclear fuel burnup with different absorbers

A computational study is performed of the fuel burnup in VVER-1000 using different absorbers in open and closed fuel cycles. It is shown that mixtures of plutonium isotopes (energy and others) can give the same effect as gadolinium, which is currently used. Fuel burnup increases. When neptunium, americium, and curium isotopes are used as a consumable absorber in a closed fuel cycle, the accompanying effect is elimination of long-lived α-emitting radionuclides which have accumulated in long-term repositories.     

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

铅铋堆嬗变燃料初步选型与分析

分离嬗变目前看来是次锕系核素回收和处理的比较可行路径,嬗变燃料研究也是第四代核能系统和先进核燃料循环技术的研究热点。本文简要归纳了嬗变燃料研究的主要特点以及当前国内外的研究现状,建议我国大力发展相关技术,特别是快堆嬗变燃料。此外,通过分析铅铋快堆/加速器驱动次临界系统(Accelerator Driven Sub-critical System,ADS)的特点和传统快堆燃料性质,得出氮化物嬗变燃料是目前铅铋堆嬗变燃料优选方案的结论。对中国科学院正在建设的ADS嬗变系统,期望未来考虑开展氮化物燃料的辐照和嬗变性能测试。     

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power histories and specifications of the fuel rods DH and DK of IFA-519.9 irradiated in Halden reactor were input, and calculated rod internal pressures were compared with experimental data for the range of 25–93 MWd kg⁻¹ UO₂, and factors affecting pellet temperature were discussed. Also some sensitivity studies were conducted with respect to the effect of swelling rate and grain growth. As a result, it is found that the prediction is sensitive to the models of thermal conductivity and swelling rate of fuel, and FEMAXI–V analytical system proved to give a reasonable prediction even in the high burnup region.     

Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor–corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP–ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems’ k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.     

Neutronics physics analysis of a large fluoride-salt-cooled solid-fuel fast reactor with Th-based fuel

Fast reactors based on thorium fuel have enhanced inherent safety. Fluoride salt performs well as a coolant in high-temperature nuclear systems. In this paper,we present a reference core for a large fluoride-salt-cooled solid-fuel fast reactor(LSFR) using thorium–uranium fuel cycle. Neutronics physics of the LSFR reference core is investigated with 2D and 3D in-core fuel management strategy. The design parameters analyzed include the fuel volume fraction, power density level and continuous removal of fission products with 3D fuel shuffling that obtains better equilibrium core performance than 2D shuffling. A self-sustained core is achieved for all cases,and the core of 60% fuel volume fraction at 50 MW/m~3 power density is of the best breeding performance(average breeding ratio 1.134). The LSFR core based on thorium fuel is advantageous in its high discharge burn-up of 20–30% fissions per initial heavy metal atom, small reactivity swing over the whole lifetime(to simplify the reactivity control system), the negative reactivity temperature coefficient(intrinsically safe for all cases) and accepted cladding peak radiation damage. The LSFR reactor is a good alternative option for the deployment of a self-sustained thorium-based nuclear system.     

Neutronics physics analysis of a large fluoride-salt-cooled solid-fuel fast reactor with Th-based fuel

A synchrotron radiation source called TURKAY is proposed as a sub-project of the Turkish Accelerator Center Project. The storage ring of TURKAY is a low emittance synchrotron and the radiation ranges between 0.01 and 60 keV can be generated from the insertion devices and bending magnets placed on it. The injector system of the facility will mainly consist of a 150 MeV linac and full energy booster. In this study, we present design considerations and beam dynamics studies of the pre-injector linac and booster ring for TURKAY.     

LWR spent fuel transmutation with fusion-fission hybrid reactors

In this paper the transmutation of light water reactors (LWR) spent fuel is analyzed. The system used for this study is the fusion-fission transmutation system (FFTS). It uses a high energy neutron source produced with deuterium-tritium fusion reactions, located in the center of the system, which is surrounded by a fission region composed of nuclear fuel where the fissions take place. In this study, the fuel of the fission region is obtained from the recycling of LWR spent fuel. The MCNPX Monte Carlo code was used to setup a model of the FFTS. Two fuel types were analyzed for the fissile region: the mixed oxide fuel (MOX), and the inert matrix fuel (IMF). Results show that in the case of the MOX fuel, an important Pu-239 breeding is achieved, which can be interesting from the point of view of maximal uranium utilization. On the contrary, in the case of the IMF fuel, high consumption of Pu-239 and Pu-241 is observed, which can be interesting from the point of view of non-proliferation issues. A combination of MOX and IMF fuels was also studied, which shows that the equilibrium of actinides production and consumption can be achieved. These results demonstrate the versatility of the fusion-fission hybrid systems for the transmutation of LWR spent fuel.     

聚变驱动乏燃料焚烧堆氦冷包层中子学设计与分析

采用VisualBUS程序和HENDL数据库,对聚变驱动乏燃料焚烧堆氦冷包层开展了中子学设计与分析工作,设计目标是在满足Keff小于0.95,功率密度小于100 MW·m-3和氚自持的前提下,获得至少1 GWe的能量输出和最大增殖、嬗变能力,且系统能够连续稳定运行.文中通过对包层中的乏燃料成分和是否装载贫铀开展优化分析并给出了优化方案,该方案能够很好地满足设计目标.