The use of U3Si2 dispersed fuel in Low-Power Research Reactors

The use of U₃Si₂ as a Low Enriched Uranium (LEU) dispersed fuel in Low-Power Research Reactors is investigated in this paper. The fuel proves to be usable if some of the original fuel rods (HEU UAl₄–Al fuel) are still simultaneously employed (mixed core) without changing the structure of the actual core. About 3.5712 mk Initial Excess Reactivity (IER) is procured. Although the worths of both the control rod and the reactivity devices decrease, the safety of these reactors is higher in the case of the new LEU fuel. If the dimensions of the meat and/or the clad are allowed to change these reactors can be run with a meat 2.15 mm outer radius, and a clad 0.58 mm thickness. The IER will then be 4.1537 mk, and both the control rod (CR) worth and the safety margins decrease.     

燃料板失效机理研究与分析

U3Si2-Al燃料板是研究堆的新型燃料,其拉伸失效机理是燃料组件设计中需要考虑的重要因素.为此,对大小两种尺寸的燃料板试样,在不同的工艺条件下进行了拉伸试验.基于试验得到的数据结果以及试验过程中对断口拍摄的电镜照片,分析并得到了U3Si2-Al燃料板的一般拉伸失效机理.此外,还对芯体缺陷对拉伸失效的影响进行了讨论,以及拉伸试验中燃料板各部分的力学特性.最后,考虑燃料板在实际工作环境下的拉伸失效可能性,分析了燃料板的实际使用安全性.     

研究堆用U3Si2-Al燃料板的热稳定性试验

U3Si2-Al板状燃料组件是一种推广应用的新型燃料元件,在国内首次应用。燃料组件的各项性能,特别是热稳定性必须通过实验验证。通过对铀密度为3.02 g/cm3的U3Si2-Al燃料板的热稳定性试验,得到:热稳定性试验会使燃料板的体积略有增大;120℃及250℃的热循环下,燃料板无明显变形,表面无变化,400℃的热循环下,燃料板略有弯曲,个别芯体裸露的燃料板表面有起泡现象;循环温度越高,芯体中U3Si2颗粒开裂越严重等实验结论,为该燃料组件的结构设计、安全分析、加工工艺提供了关键参数,并为该组件的堆内运行提供了借鉴。     

U3Si2-Al板状燃料组件冲刷试验后的解体研究

通过对U3Si2-Al板状燃料组件的解体试验研究,得到了110 ℃水温、7 m/s流速下50 d的水力冲刷试验和18 m/s流速下的流致振动试验对该新型燃料组件各项参数的影响,为板状燃料组件的设计、选材、加工、应用提供了实验数据.     

Design and experience of HEU and LEU fuel for WWR-M reactors

Thin-walled WWR-M5 fuel elements were designed and manufactured and have been used successfully for 16 years; they contain twice as much uranium-235 as the WWR-M2 and WWR-M3 fuel elements. The fuel elements have been optimized with regard to their neutron physics and thermal-hydraulic parameters and fuel consumption has been minimized. The mean specific power in the core of the WWR-M reactor was raised to 230 kW l⁻¹, the measured maximum volume thermal specific power was 900±100 kW l⁻¹ and the surface specific power was 136±15 W cm⁻². The WWR-M5 fuel elements enable the power of the WWR-M pooltype reactor to be raised to 30 MW while simultaneously increasing the number of cells in the core available for experimentation by a factor of approximately two and reducing fuel element consumption. Reactor tests of WWR-M fuel elements with reduced fuel enrichment (36 and 21%) were carried out for a meat uranium density up to 2–3 g cm⁻³. Conversion of WWR-SM-type reactors to these fuel elements did not lead to a loss in reactivity and enabled their power to be increased to 20–30 MW.     

Reactivity temperature coefficients for the HEU and LEU fuel of the MNSR reactor

Analysis of the Reactivity Temperature Coefficients of the Miniature Neutron Source Reactor (MNSR) for normal and accidental conditions (above 45 °C) using HEU-UAl₄ and the LEU: U₃Si, U₃Si₂ and U9Mo fuel were carried out in this paper. The Fuel Temperature Coefficient (FTC), Moderator Temperature Coefficient (MTC), and Moderator Density Coefficient (MDC) were calculated using the GETERA code. The contribution of each isotope presented in the fuel cell was calculated for the temperature range of 20 °C–100 °C at the beginning of the core life. The average values of the FTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were found to be: −2.23E-03, −1.85E-02, −1.96E-02, −1.85E-02 mk/°C respectively. The average values of the MTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −8.91E-03, −1.24E-04, −4.70E-03, 2.10E-03 mk/°C respectively. Finally, the average values of the MDC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −2.06E-01, −2.03E-01, −2.04E-01, −2.03E-01 mk/°C respectively. It's found also that the dominant reactivity coefficient for all types of fuel is the MDC.     

Core Thermohydraulic Design with 20% LEU Fuel for Upgraded Research Reactor JRR-3

This paper presents the outline of the core thermohydraulic design and analysis of the research reactor JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% low enriched uranium (LEU) plate-type fuel. For the condition of normal operation, the upgraded JRR-3 core is planned to be cooled by two cooling modes of forced-convection at high power and natural-convection at low power. The major feature of core thermohydraulics is that at the forced-convection cooling mode the core flow is a downflow, under which fuel plates are exposed to a severer condition than an upflow in cases of operational transients and accidents. The core thermohydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margins both against the onset of nucleate boiling (ONR) not to allow the nucleate boiling anywhere in the core and against the departure from nucleate boiling (DNB). The safety margins against ONB and DNB were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the ONB, and the minimum DNB ratio (ratio of DNB heat flux to the maximum heat flux) was evaluated to be about 2.1, which gives a sufficient margin against the DNB. The core thermohydraulic characteristics were also clarified for the natural-convection cooling mode.     

Preliminary conceptual design of a new moderated reactor utilizing an LEU fuel for space nuclear thermal propulsion

Nuclear Thermal Rocket (NTR) propulsion is a viable and meritorious option for human exploration into deep-space because of its high thrust, improved specific impulse, well established technology, bimodal capability, and enhanced mission safety and reliability. The NTR technology has already been investigated and tested by the United States of America and Russia and the former Soviet Union. The representative Nuclear Engine for Rocket Vehicle Applications (NERVA) type reactors traditionally used Highly Enriched Uranium (HEU) fuels, shaped in hexagonal fuel element geometries because of the importance of making a high power reactor with a minimum size. Although the HEU-NTR designs are the best choice in terms of rocket performance and technical maturity, they inevitably provoke nuclear proliferation obstacles not only for all research and development activities by civilians and non-nuclear weapon states but also for potential commercialization. To overcome the security issues due to HEU, the non-proliferative, small-size NTR engine with low thrust levels of 41 kN–53 kN (9.2 klb_f ∼ 11.9 klb_f), Korea Advanced NUclear Thermal Engine Rocket utilizing a Low-Enriched Uranium fuel (KANUTER-LEU), is being designed for future generations. Its design goals are to make use of an LEU fuel for its fairly compact core, but to minimize the rocket performance sacrifice relative to the traditional HEU-NTRs. To achieve these goals, a new space propulsion reactor is conceptually designed with the key concepts of a high uranium density fuel with resistance against high heating and H₂ corrosion, a thermal neutron spectrum core, and a compact and integrated fuel element core design with protective cooling capability. In addition, a preliminary design study of neutronics and thermal-hydraulics was performed to explore the design space of the new LEU-NTR reactor concept. The result indicates that the innovative reactor concept has great potential, both to implement the use of an LEU fuel and to create comparable rocket performance, compared to the existing HEU-NTR designs.     

New fuel rod design based on LEU/Th fuel and an assessment on in-core fuel cycle length of the MHR core

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50-80% relative to the current design, with only a modest increase in the fissile material loading (15-20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once - through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10-15% over that of current MHR design.     

DFT + U as a reliable method for efficient ab initio calculations of nuclear materials

Density functional theory (DFT)-based ab initio methods become standard research tools in various research fields, including nuclear materials science. However, having strongly correlated f-electrons, lanthanide- and actinide-bearing nuclear materials are computationally challenging for DFT methods and straightforward DFT calculations of these materials can easily produce false results. In this contribution we benchmark the DFT + U method, with the Hubbard U parameter derived ab initio, for prediction of structural and thermochemical parameters of nuclear materials, including various actinide-bearing molecular complexes and lanthanide-bearing monazite- and xenotime-type prospective ceramic nuclear waste host forms. Our studies show that the applied DFT + U method improves significantly prediction of DFT by producing results with uncertainties similar to those of the higher order, but computationally unfeasible ab initio methods, and the experimental data, and thus allows for reliable and feasible ab initio computation of even chemically complex nuclear materials.     

Neutronic simulation of water-based nanofluids as a coolant in VVER-1000 reactor

The main goal of this study is to perform the neutronic simulation of nanofluids application to reactor core. The variation of the Bushehr VVER-1000 reactor primary neutronics parameters is investigated with using different nanofluids as coolant. In the present neutronic simulation, water-based nanofluids containing various volume fractions of Al₂O₃, Si, Zr, TiO₂, CuO, Ti, Cu and Ag nanoparticles are investigated. Optimization of type and volume fraction of nanoparticles affects the reactor neutronic characteristics. The results achieved by using WIMS and CITATION codes, show that below 0.1 percent volume fraction of Al₂O₃ is the optimum nanoparticle for normal operation and Ag/water nanofluid is suggested to use as a reactor safety enhancement tool.     

Irradiation behavior of ground U(Mo) fuel with and without Si added to the matrix

In the framework of the IRIS-TUM irradiation program, several full size, flat dispersion fuel plates containing ground U(Mo) fuel kernels in an aluminum matrix, with and without addition of silicon (2.1 wt.%), have been irradiated in the OSIRIS reactor. The highest irradiated fuel plate (with an Al-Si matrix) reached a local maximum burnup of 88.3% ²³⁵U LEU-equivalent and showed a maximum thickness increase of 323 μm (66%) but remained intact. This paper reports the post irradiation examination results obtained on four IRIS-TUM plates. The evolution of the fission gas behavior in this fuel type from homogeneously dispersed nanobubbles to the eventual formation of large but apparently stable fission gas bubbles at the interface of the interaction layer and the fuel kernel is illustrated. It is also shown that the observed moderate, but positive effect of Si as inhibitor for the U(Mo)-Al interaction is related to the dispersion of this element in the interaction layer, although its concentration is very inhomogeneous and appears to be too low to fully inhibit interaction layer growth.     

U3Si2-Al燃料元件板力学性能试验研究

对研究堆使用的不同规格的U3 Si2 Al弥散型复合燃料板元件的力学性能参数进行了试验研究和分析 ,同时结合板型燃料组件的结构特点对燃料板在组件运行期间可能受到的拉、压、弯的承载能力进行了测试研究。研究结果表明 :国内生产的板状元件在其拉伸性能、结构抗力等方面基本达到了国外类似燃料的水平 ,满足了研究堆的设计要求。     

The use of WIMS and CITATION codes in fuel loading required for the conversion of HEU MNSR core to LEU

The Nigerian Research Reactor-1 (NIRR-1) falls in the category of Miniature Neutron Source Reactors (MNSR) using a fuel of 90% HEU. It is therefore desirable to convert it from this enrichment to LEU (less than 20%) in conformity with the new global trend of making research reactor fuel as unattractive as possible to groups that may be interested in using such highly enriched cores for non-peaceful purposes. In this work, we have developed a computational scheme based on WIMS and CITATION that would theoretically achieve this objective as easily as possible. The scheme systematically reduces the enrichment from 90% (or any other initial values) to less than 20% in steps of 5% or any desired percentage variation. Two fuel types (UAl₄ and UO₂) are considered in here, while maintaining the size and geometry of the core as well as the excess reactivity (between 3.5 and 4 mk). Our results show that the U-235 loading increases sharply as enrichment decreases. It has also been noticed that at 5% enrichment the fuel loading for both types is 2505 g. However, at 90% enrichment, the loading drops sharply to 998 g for UAl₄ fuel and 946 g for UO₂ fuel. Below the enrichment of 5%, the operation of NIRR-1 with both fuel types can be considered unrealistic as this requires structural adjustment which the work tries to maintain constant.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Neutronic optimization for a blanket of the laser fusion reactor,SENRI-I

Neutronic calculations were performed to optimize the SENRI blanket in terms of energy multiplication as well as tritium breeding ratio. The blanket employs a thick ( 64-cm) Li layer as breeder/coolant. Three approaches were taken here to achieve the goal: (1) reduction of⁶Li in the lithium, (ii) replacement of the Li layer by a molten-salt (flibe) layer, and (iii) shipment of excess tritium to a nonbreeding blanket. It was found that the excess tritium produced in the SENRI blanket could be used effectively to obtain additional power by fueling a nonbreeding D-T reactor.