Vectorization and parallelization of a production reactor assembly code

In order to efficiently use new features of supercomputers, production codes, usually written 10 – 20 years ago, must be tailored for modern computer architectures. We have chosen to optimize the CPM-2 code, a production reactor assembly code based on the collision probability transport method. Substantional speedups in the execution times were obtained with the parallel/vector version of the CPM-2 code. In addition, we have developed a new transfer probability method, which removes some of the modelling limitations of the collision probability method encoded in the CPM-2 code, and can fully utilize parallel/vector architecture of a multiprocessor IBM 3090.     

Optimizing fuel assembly placement in a research reactor

Institute for Nuclear Research, Academy of Sciences of the Ukrainian SSR. Translated from Atomnaya Energiya, Vol. 70, No. 4, pp. 257–259, April, 1991.     

Optimizing fuel-pin assembly location in a research reactor

Nuclear Research Institute, Ukrainian Academy of Sciences. Translated from Atomnaya Énergiya, Vol. 72, No. 6, pp. 624-626, June, 1992.     

Development of a steady state analysis code for a molten salt reactor

The molten salt reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for transmutation, and production of electricity, hydrogen and fissile fuels. In this study, a single-liquid-fueled MSR is designed for conceptual research, in which no solid material is present in the core as moderator, except for the external reflector. The fuel salt flow makes the MSR neutronics different from that of conventional reactors using solid fuels, and couples the flow and heat transfer strongly. Therefore, it is necessary to study the core characteristics with due attention to the coupling among flow, heat transfer and neutronics. The standard turbulent model is adopted to establish the flow and heat transfer model, while the diffusion theory is used for the neutronics model, which consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six groups of delayed neutron precursors. These two models which are coupled through the temperature and heat source are coded in a microcomputer program. The distributions of the velocity, temperature, neutron fluxes, and delayed neutron precursors under the rated condition are obtained. In addition, the effects of the inflow temperature, inflow velocity, and the fuel salt residence time out of the core are discussed in detail. The results provide some valuable information for the research and design of the new generation molten salt reactors.     

Selection of a toroidal fusion reactor concept for a magnetic fusion production reactor

The basic fusion driver requirements of a toroidal materials production reactor are considered. The tokamak, stellarator, bumpy torus, and reversed-field pinch are compared with regard to their demonstrated performance, probable near-term development, and potential advantages and disadvantages if used as reactors for materials production. Of the candidate fusion drivers, the tokamak is determined to be the most viable for a near-term production reactor. Four tokamak reactor concepts (TORFA/FED-R, AFTR/ZEPHYR, Riggatron, and Superconducting Coil) of approximately 500-MW fusion power are compared with regard to their demands on plasma performance, required fusion technology development, and blanket configuration characteristics. Because of its relatively moderate requirements on fusion plasma physics and technology development, as well as its superior configuration of production blankets, the TORFA/FED-R type of reactor operating with a fusion power gain of about 3 is found to be the most suitable tokamak candidate for implementation as a near-term production reactor.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing from its 1982 version (except for Tables II and III and Fig. 1), explaining the fact that some of the material is dated.     

Verification of a self-developed CFD-based multi-physics coupled code MPC-LBE for LBE-cooled reactor

To perform an integral simulation of a pool-type reactor using CFD code, a multi-physics coupled code MPC-LBE for an LBE-cooled reactor was proposed by integrat...     

Criticality qualification of a new Monte Carlo code for reactor core analysis

In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal–hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the “Accelerator part” of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the “Reactor part” of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.     

Fuel assembly design study for a reactor with supercritical water

The European concept of the High Performance Light Water Reactor (HPLWR) differs from current light water reactors in a higher system pressure beyond the critical point of water, as well as a higher heat-up of the coolant within the core and thus higher core outlet temperatures, leading to a significant increase in turbine power and thermal efficiency of the power plant. The motivation to develop a novel fuel assembly for the HPLWR is caused by the high variation of coolant density in the core by more than a factor of seven. A systematic design study shows that a square fuel assembly with two rows of fuel rods and a central moderator box is best to minimize the structural material, to optimize the moderator to fuel ratio and to reduce differences of fuel rod power. Using neutronic and thermal-hydraulic analyses, a detailed mechanical design of a fuel assembly of the HPLWR has been worked out. Moreover, concepts for the head piece, the foot piece, the steam plenum and the lower mixing plenum, including the lower core plate, have been developed to account for the individual flow paths of this reactor. These allow a leak-tight counter current flow of moderator water and coolant as well as uniform mixing of different mass flows. The assembly design concept can be used as a general key component for any advanced core design of this reactor.     

Economic analysis of a magnetic fusion production reactor

The magnetic fusion reactor for the production of nuclear weapon materials, based on a tandem mirror design, is estimated to have a capital cost of $1.5 billion and to produce 10 kg of tritium/year for $22,000/g or 940 kg/year of plutonium in the plutonium mode for $250/g plus heavy metal processing. A tokamak-based design is estimated to cost $1.5 billion and to produce 10 kg of tritium/year for $29 thousand/g. For comparison, a commercially sized tandern mirror fusion breeder selling excess electricity and fissile material to commercial markets is estimated to cost $3.6 billion and to produce tritium for $2.6 thousand/g and plutonium for $34/g plus heavy metal processing.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing of its 1982 version, explaining the fact that some of the material is dated.     

Mechanical design of a magnetic fusion production reactor

The mechanical aspects of tandem mirror and tokamak concepts for the tritium production mission are compared and a proposed breeding blanket configuration for each type of reactor is presented in detail, along with a design outline of the complete fusion reactor system. In both cases, the reactor design is developed sufficiently to permit preliminary cost estimates of all components. A qualitative comparison is drawn between both concepts from the view of mechanical design and serviceability, and suggestions are made for technology proof tests on unique mechanical features. Detailed cost breakdowns indicate less than 10% difference in the overall costs of the two reactors.This paper represents Work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing of its 1982 version, explaining the fact that some of the material is dated.     

HYSYS软件在反应堆系统仿真中的应用探索

系统仿真软件可以模拟运行工况变化对系统整体运行带来的影响,在系统瞬态分析和安全研究中起着重要的作用。Aspen HYSYS软件是世界知名的油气过程仿真和优化的系统软件,具有强大的二次开发功能,可以用于反应堆系统仿真。在植入熔盐物性、修改熔盐换热模型的基础上,建立并调用点堆模型的动态链接库,尝试将HYSYS与点堆耦合起来,弥补HYSYS无法对熔盐堆等反应堆进行仿真的缺憾。在此基础上,对中国科学院上海应用物理研究所的熔盐堆设计进行了系统仿真,给出了熔盐堆在不同的运行工况下的系统响应分析结果,并与RELAP5仿真结果进行比较。结果表明,耦合程序有较高的可用性,能够达到预期的效果。     

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.     

Nuclear design and analysis of a magnetic fusion production reactor

Tandem-mirror- and tokamak-based magnetic fusion production reactors are predicted to have tritium breeding ratios of 1.67 and 1.49, respectively. The latter value replaces one (1.56) that is used elsewhere in the sequence of papers in this issue. Blanket energy multiplication for both is predicted to be about 1.3. With the tandem mirror operating in the plutonium production mode, the net plutonium-plus-tritiurn breeding ratio is 1.74. Blanket energy multiplication for the plutonium mode is predicted to be 2.4 at a plutonium-uranium ratio of 0.7% and a uranium volume fraction of 3%.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing of its 1982 version, explaining the fact that some of the material is dated.     

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.     

Fusion technology for a magnetic fusion production reactor

The tandem mirror and tokamak are being considered as candidate fusion drivers for a materials production reactor that could be implemented in the 1990s. This report considers, in detail, the required performance characteristics of the fusion plasma and the major technological subsystems for each fusion driver. These performance characteristics are compared with the present state of the art, corresponding development needs are identified, and technology program requirements, in addition to those now being supported by the Department of Energy, are pointed out. The tandem mirror and tokamak fusion drivers are also compared with regard to their required advancements in plasma performance and technology development.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing of its 1982 version, explaining the fact that some of the material is dated.     

CFD code application: calculating heat transfer in a reactor with supercritical parameters

Thermohydraulic calculations of isolated and communicating cells of a rod bundle were performed by the channel method for CANDU-X fuel assemblies and by a three-dimensional method. It was established that in solving the problem for the tightest cell in the case q = const the azimuthal nonuniformity of the temperature was found to decrease by 77°C but it too was inadmissibly large. The temperature distribution along the surface of a fuel element in the case q = const was found to be different from the solution of the adjoint problem. A region with elevated coolant temperature, impeding heat exchange between two neighboring cells, was found between two adjoining cells. It was found that to evaluate computational reliability an experimental study must be performed on rod assemblies with supercritical coolant parameters.