Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 °C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 °C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions.     

用钴活化法测定反应堆中热中子积分通量

本文叙述了用钴活化法测定高通量堆中热中子积分通量的方法。测得的热中子积分通量值与计算值作了比较。本法适于测定在高通量堆中长期辐照的较高热中子积分通量。     

Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling

Fuel rod design for high power density supercritical water-cooled fast reactor was conducted with mixed-oxide (MOX) fuel and stainless steel (SUS304) cladding under the limiting cladding surface temperature of 650 °C. Fuel and cladding integrities, and flow-induced vibration were taken into account as design criteria. Designed fuel rod has the diameter of 7.6 mm and is arranged in the fuel assembly with pitch-to-diameter ratio of 1.14. New core arrangement for negative void reactivity is proposed by three-dimensional tri-z core calculation fully coupled with thermal hydraulic calculation, where ZrH layer concept is used for negative void reactivity. The core has high power density of 156 W/cm³ and its equivalent diameter is only 2.7 m for 1000 MWe class reactor core. High average core outlet temperature of 500 °C is achieved by introducing radial fuel enrichment zoning and downward flow in seed assembly. Small pressure vessel size and simplified direct steam cycle with higher thermal efficiency give an economical potential in aspect of capital and operating cost.     

Numerical examination of temperature control in helium-cooled high flux test module of IFMIF

For long term irradiation of the International Fusion Materials Irradiation Facility (IFMIF), test specimens are needed to retain constant temperature to avoid change of its irradiation characteristics. The constant temperatures control is one of the most challenging issues for the IFMIF test facilities. We have proposed a new concept of test module which is capable of precisely measuring temperature, keeping uniform temperature with enhanced cooling performance. In the system according to the new design, cooling performances and temperature distributions of specimens were examined numerically under diverse conditions. Some transient behaviors corresponding to the prescribed temperature control mode were perseveringly simulated. It was confirmed that the thermal characteristics of the new design satisfied the severe requirement of IFMIF.     

Conceptual design and safety characteristics of a new multi-mission high flux research reactor

Research reactors with neutron fluxes higher than 10¹⁴ n cm^-2 s^-1 are widely used in nuclear fuel and material irradiation,neutron-based scientific research, and medical and industrial isotope production. Such high flux research reactors are not only important scientific research facilities for the development of nuclear energy but also represent the national comprehensive technical capability. China has several high flux research reactors that do not satisfy the...     

Liquid salt cooled flexible conversion ratio fast reactor: Neutronic design

This paper presents the neutronic design of a liquid salt cooled fast reactor with flexible conversion ratio. The main objective of the design is to accommodate interchangeably within the same reactor core alternative transuranic actinides management strategies ranging from pure burning to self-sustainable breeding. Two, the most limiting, core design options with unity and zero conversion ratios are described. Ternary, NaCl-KCl-MgCl₂ salt was chosen as a coolant after a rigorous screening process, due to a combination of favourable neutronic and heat transport properties. Large positive coolant temperature reactivity coefficient was identified as the most significant design challenge. A wide range of strategies aiming at the reduction of the coolant temperature coefficient to assure self-controllability of the core in the most limiting unprotected accidents were explored. However, none of the strategies resulted in sufficient reduction of the coolant temperature coefficient without significantly compromising the core performance characteristics such as power density or cycle length. Therefore, reactivity control devices known as lithium thermal expansion modules were employed instead. This allowed achieving all the design goals for both zero and unity conversion ratio cores. The neutronic feasibility of both designs was demonstrated through calculation of reactivity control and fuel loading requirements, fluence limits, power peaking factors, and reactivity feedback coefficients.     

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.     

Design of highly moderated pressurized water reactor based on critical heat flux considerations

The difficulty of applying the existing critical heat flux correlations to the design of a modified 100% mixed plutonium uranium oxide fuelled assembly are outlined. A core with increased moderating ratio (RMA) was designed. The purpose of this design modification is to permit the consumption of large amounts of plutonium which cannot be burned in a conventional pressurized water reactor (PWR) owing to safety considerations. The design criteria required that the minimum departure from nucleate boiling ratio (DNBR) in the modified fuel does not exceed the value of the standard N4 PWR, which served the basis for the design. The desired conditions were achieved by reducing the fuel pin diameter to increase the moderating ratio and increasing the number of grids with mixing vanes to improve the minimum DNBR in the modified assembly. The design methodology and some of the proposed design options are presented.     

Neutronic design of a sub-critical, source driven reactor at the RA-8 facility

We present a LEU-ADS design based on an existing Argentine experimental facility, the RA-8 pool type zero power reactor. The versatility of this reactor allows measurement of different core configurations using different fuel enrichment, burnable poison rods, water perturbations, different control rods types in critical or subcritical configurations with an external source.To assess the feasibility of the LEU-ADS, multiplication factors, kinetic parameters, spectra, and time flux evolution were computed. Two external sources were considered: an isotopic source, and a D-D pulsed neutron source.Parameters for different core configurations were calculated, and the feasibility of using continuous and pulsed neutron sources was verified.     

Gas-cooled reactor thermal hydraulic analyses with MELCOR

Pursuant to the Energy Policy Act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the reference design for the Next Generation Nuclear Plant (NGNP). Stemming from a U.S. Nuclear Regulatory Commission (NRC) HTGR research initiative, a need was identified for validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform licensing analyses. Because the NRC has used MELCOR for light water reactors (LWR) in the past and because MELCOR was recently updated to include gas-cooled reactor (GCR) physics models, MELCOR is among the system codes of interest to the NRC. This paper describes MELCOR modeling of the General Atomics' Modular High Temperature Gas-Cooled Reactor (MHTGR). The MHGTR is a suitable design for demonstration of MELCOR GCR modeling competency for two reasons: 1) the MHTGR is a predecessor to the more advanced General Atomics’ Gas-Turbine Modular High Temperature Reactor (GTMHR), and 2) experimental data useful for benchmark calculations may soon become available. Using the most complete literature references available for the MHTGR design, researchers at Texas A&M University (TAMU) constructed a MELCOR input deck for the MHTGR to partially validate MELCOR GCR modeling capabilities. Normal and off-normal system operating conditions were modeled with appropriate boundary and initial conditions. MELCOR predictions of system response were obtained for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) scenarios. Code results were checked against nominal MHTGR design parameters, physical intuition, and anticipated GCR thermal hydraulic response. No inherent deficiencies in MELCOR modeling capability were observed, suggesting that the newly-implemented GCR models are adequate for systems-level analysis. If and when experimental benchmark data becomes available, further validation activities may proceed given the modeling efforts discussed herein.     

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.     

多功能聚变裂变混合实验堆FDS-MFX氦冷包层三维中子学初步设计与分析

FDS-MFX(Multi-Functional eXperimental fusion-fission hybrid reactor)是一个基于现实可行技术的多功能聚变裂变混合实验堆概念,分3个阶段相继开展实验研究,分别采用纯氚增殖包层、铀燃料包层和乏燃料包层.本文重点对其中铀燃料包层后期阶段中高浓缩铀模块的摆放方式...     

GPU加速的球床式氟盐冷却高温堆堆芯热工水力程序研发

球床式氟盐冷却高温堆(Pebble Bed Fluoride-salt Cooled High Temperature Reactor,PB-FHR)是一种先进的第四代反应堆。三维堆芯热工水力程序能够模拟具有复杂空间效应的工况,但计算耗时较高。图形处理器(Graphics Processing Unit,GPU)具有大量计算单元,可有效提高程序的计算速度。本文研发了GPU加速的PB-FHR堆芯热工水力程序(GPU-accelerated Thermal Hydraulic Code,GATH),采用非热平衡多孔介质模型建立堆芯物理模型,研究并实现了GPU高速求解算法。对PB-FHR的堆芯模型进行了热工水力分析,与商用计算流体力学软件ANSYS CFX的计算结果进行了对比,验证了程序的正确性。GPU加速性能分析的结果表明,程序整体的加速比率可达8.39倍,证明所研发的GPU求解算法能有效提升堆芯热工水力分析的计算效率。     

Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% ²³⁵U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required ⁹⁹Mo/^99mTc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.     

池式快堆三维热工流体力学设计研究

随着计算机软硬件技术的发展,三维数值分析技术已经成为池式快堆堆芯和钠池热工设计和计算分析的重要组成部分,并在其中发挥着不可替代的作用.通过对池式快堆几个典型热工现象的分析,展示了我国第一座池式快堆(中国实验快堆)热工设计和安全分析中所拥有的设计手段和工具,总结了三维数值分析技术在快堆工程中的应用,并指出了其对今后快堆热工设计的重要意义.     

Steady thermal hydraulic analysis for a molten salt reactor

The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.     

Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.     

Neutronic design of hydride fueled BWRs

This paper summarizes the neutronic part of a study of the feasibility of designing BWR cores to have enhanced power density and simplified fuel bundle by using hydride instead of oxide fuel. A 3D fuel bundle neutronic analysis is performed for a limited number of geometries to determine attainable discharge burnup, pin-by-pin power distribution, axial power distribution, reactivity coefficients, reactivity worth of control elements and burnable absorber effects. It is found that hydride fuel bundle design can be simplified by eliminating water rods and partial length fuel rods and by reducing the volume of water in-between the fuel bundles. Both an ideal and more practical bundle designs are examined. A companion study of the thermal-hydraulic and vibration characteristics of BWR cores predicts that the increase in the number of fuel rods per given core volume enables increasing the BWR power density by up to ∼30% relative to oxide fuelled core design. The net outcome is expected to be improved BWR economics even though hydride fuel requires higher uranium enrichment to compensate for its reduced uranium loading.