Yongwei Yang

Prediction calculations and experiments for the first criticality of the 10 MW High Temperature Gas-cooled Reactor-Test Module

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is a pebble bed experimental reactor built by the Institute of Nuclear Energy Technology (INET), Tsinghua University. This paper introduces the first critical prediction calculations and the experiments for the HTR-10. The German VSOP neutronics code is used for the prediction calculations of the first loading. The characteristics of pebble-bed high temperature gas-cooled reactors are taken into account, including the double heterogeneity of the fuel element, the buckling feedback of the spectrum calculation, the effect of the mixture of fuel elements and graphite balls, and the correction of the diffusion coefficients in the upper cavity based on transport theory. Also considered are the effects of impurities in the fuel elements, in the graphite balls and in the reflector graphite on the reactivity. The number of fuel elements and graphite balls in the initial core is predicted to provide reference for the first criticality experiment. The critical experiment adopts a method of extrapolating to approach criticality. The first criticality was attained on December 1, 2000. The first criticality experiment shows that the predicted critical number of the fuel elements and graphite balls is in close agreement with the experimental results. Their relative error is less than 1.0%, implying the physical predictions and the results of the criticality experiment are much beyond expectations.

Study on fission blanket fuel cycling of a fusion–fission hybrid energy generation system

This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion?fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion?fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion?fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

Fuel management of the HTR-10 including the equilibrium state and the running-in phase

Abstract The mode of fuel management of the HTR-10 was studied, including the simulation of the fuel shuffling process and the measurement of the burnup of a fuel element. The prior consideration was the design of the equilibrium state. Based on this the fuel loading of the initial core and the fuel shuffling mode from the initial core through the running-in phase into the equilibrium state were studied. The code system VSOP was used for the physical layout of the HTR-10 at the equilibrium state and in the running-in phase. For the equilibrium state, in order to lessen the difference between the peak and the average burnup, 5-fuel-passage-through-the-core was chosen for the fuel management. The average burnup of the spent fuel for the equilibrium core is 80 000 MWd t −1 , and the peak value of it is less than 100 000 MWd t −1 when the burnup of the recycled fuel element is under 72 000 MWd t −1 . The mixture of fuel element and graphite element was used for the initial core loading, the volume fractions of the fuel and the graphite elements were 0.57 and 0.43, respectively. During the running-in phase, the volume fraction of graphite will decrease with the fresh fuel elements being loaded from the top of the core and the graphite elements discharged from the bottom of the core. The fuel shuffling mode is similar to that of the equilibrium state. The burnup limit of recycled fuel element is also 72 000 MWd t −1 and the peak burnup is less than 100 000 MWd t −1 . Finally the core will be full of fuel elements with a certain profile of burnup and reaches the equilibrium state. According to the characteristics of the pebble-bed high temperature gas-cooled reactor, a calibrating method of concentration of 137 Cs was proposed for the measurement of fuel burnup.

Application of Origen2.1 in the decay photon spectrum calculation of spallation products

Origen2.1 is a widely used computer code for calculating the burnup, decay, and processing of radioactive materials. However, the nuclide library of Origen2.1 is used for existing reactors like pressurized water reactors. To calculate the photon spectrum released by the decay of spallation products, we have made specific libraries for the ADS tungsten spallation target, based on the results given by the FLUKA Monte Carlo code. All the data used to make the Origen2.1 libraries are obtained from Nuclear structure & decay Data (NuDat2.6). The accumulated activity of spallation products and the contribution of nuclides to photon emission are given in this paper.

Burnup Analysis of Thorium-Uranium Based Molten Salt Blanket in a Fusion-Fission Hybrid Reactor

Abstract Progress on the fusion-fission hybrid reactor (FFHR) brings fusion a viable energy source in foreseeable future. Energy multiplication in a FFHR makes a much easier prerequisite for the fusion reaction than a fusion reactor. The molten salt reactor has advantages on heat transfer and post-processing of the spent fuels. A fission blanket made of molten salt was studied for the FFHR. The molten salt consists of F-Li-Be, with nuclear fuels dissolved in it. When thorium-uranium-plutonium fuels were added into a F-Li-Be molten salt zone with a component of 71% LiF -2% BeF2 -13.5% ThF4 -8.5% UF4 -5% PuF3, the appropriate blanket energy multiplication factor and TBR can be obtained. Two different molten salt models (Single molten salt zone model and multi molten salt zone model) were designed and compared in this study. The changes in blanket multiplication factor, M, and the tritium breeding ratio, TBR, during burnup life are investigated. The burnup analysis of the molten salt blanket was carried out by the COUPLE2 code. Through the burnup analysis, the breeding of the fissile fuel 233U and the transmutation of the minor actinides were also studied.

Sustainable Development of Nuclear Energy and Study on ADS in China

As is well known, the disposal of high level nuclear waste is one of the key issues restricting the sustainable development of nuclear energy in the world. The partitioning and transmutation strategy (P&T) has been considered as an alternative for the disposal of high level nuclear waste in future advanced nuclear fuel cycles. For the transmutation of high level nuclear waste, accelerator driven sub-critical system (ADS) has been regarded as the most feasible choice. In this paper, the main technical characteristics of ADS system are described, the world research status and development trends of ADS are introduced, and ADS study plans in China are discussed.

Neutronic Study of an Innovative Thorium-Uranium–Based Fusion-Fission Hybrid Energy Reactor with 233U Breeding Enhancement by Using Dual-Coolant System

Abstract In this technical note, an innovative thorium-uranium–fueled fusion-fission hybrid reactor (FFHR) design that employs a dual-coolant system to enhance 233U breeding and is based on a three-dimensional engineering model is presented. The reactor consists of two kinds of modules: a water-cooled, thermal spectrum power generation natural uranium–fueled module and helium-cooled, fast spectrum fissile-breeding natural thorium–fueled modules, which are arranged alternately in the poloidal direction of the blanket. An interesting and important neutronic characteristic of the FFHR is found in this technical note: Energy multiplication is primarily determined by the uranium module parameters and is almost independent of the thorium module parameter. Uranium module design should first consider improving energy production. The 232Th neutron capture rate is primarily determined by the thorium module parameters. The uranium module parameter has almost no influence on the 232Th neutron capture rate in the thorium module. The uranium and thorium modules have weak coupling in neutronic behavior. However, with the fixed design parameters of the uranium and thorium modules, the most important influencing factor on energy multiplication factor M (the ratio of total blanket energy output and the fusion energy) and the 233U breeding rate is the fraction of the external fusion neutron source irradiated on the uranium or thorium module or the blanket coverage rate of the uranium or thorium modules. Based on this characteristic, an innovative hybrid reactor design that employs a dual-coolant system is proposed in this technical note. Uranium modules still use water as the coolant to maintain a high energy multiplication factor, whereas helium is used as the coolant for the thorium module to obtain a fast neutron spectrum to enhance the 233U breeding. The simulation results show that the helium-cooled thorium module is 2.5 times more efficient in 233U breeding compared to the original water-cooled thorium module design. Approximately 10 tons of 233U is produced after 20 years of operation for the helium-cooled thorium module design.

Thermal Analysis for the Dense Granular Target of CIADS

For the China Initiative Accelerator Driven System (CIADS), the energy of the protons is 250 MeV, and the current intensity will reach 10 milliamperes. A new concept of a dense granular spallation target is proposed for which the tungsten granules are chosen as the target material. After being bombarded with the accelerated protons from the accelerator, the tungsten granules with high-temperature flow out of the subcritical reactor and the heat is removed by the heat exchanger. One key issue of the target is to remove the 2.5 MW heat deposition safely. Another one is the heat exchange between the target and the subcritical reactor. Based on the model of effective thermal conductivity, a new thermal code is developed in Matlab. The new code is used to calculate the temperature field of the target area near active zone and it is partly verified by commercial CFD code Fluent. The result shows that the peak temperature of the target zone is nearly 740°C and the reactor and the target are proved to be uncoupled in thermal process.

Physical Analysis of LBE Spallation Target Coupled With the Reactor for CIADS

For the project of the Chinese Initiative Accelerator Driven Sub-critical system (CIADS), the Lead-Bismuth-Eutectic (LBE) spallation target is one of the two alternatives, which has high good thermal performance, mature technology, and other advantages. The physical design of the spallation target determines the neutron yield and the utilization of the neuron source, as well as the performance of the sub-critical reactor and other key issues. Based on the Monte Carlo program MCNPX, we did the preliminary design of spallation target coupled with the reactor with a keff about 0.95. The energy deposition density distribution of the target and the window were calculated. In the mean time, the neutron flux density, the neutron energy spectrum, and the power amplification factors were calculated. By changing the positions of the target, the radii of the beam pipe and the thickness of target, we studied the variation of the neutronic parameters mainly mentioned above. The energy deposition density distribution was used as the heat source of the thermal-hydraulics analyses. From the neutronic parameters, we found that to get the maximum power amplification factor, the target window should be put at the positions 11.4 cm above the center of the core. Actually, when the target was put above the center of the core, from 0cm to 22cm, the maximum differences of the power amplification factor is less than 4.0%, which means the position will have little influences in this range. When the target window was put at the center, increasing of the window’s thickness will lead the decreasing of the power amplification factor. The enlargement of the beam pipe radii will decrease the maximum that the amplification factors can reach. Meanwhile, the increasing of the beam radii will enlarge the power amplification factor slightly. The physics analysis of the LBE target coupled with the reactor can give more information to the optimization of the target structure and the sub-critical reactor for CIADS.Copyright

Research on Nuclear Heat Deposition Behavior in the Spallation Target of an Accelerator Driven Subcritical System

The power density distribution behavior of the ADS spallation target, which is a key factor in the thermal-hydraulic and mechanical design of the high-power ADS target, was investigated under different proton incident energy. A Chinese ADS conceptual design of spallation target was proposed in this paper. The deposition heat in the spallation target was calculated by MCNPX code. From the results, it was found that the Bragg peak phenomenon weakens as the proton source incident energy increases. Large Bragg peak was observed for proton incident energy below 500MeV, however for the proton source energy above 900MeV, Bragg peak phenomena was not obvious. Analysis on the nuclear reactions behavior and ionization process induced by source proton in the target was carried out to address this issue. Meanwhile, the results show that the proton leakage rate from the target, which is an important factor in proton radiation shielding, greatly depends on the proton range (penetration depth) under different incident energy. In order to stop or keep the protons in the target, the minimum axial length (in the incoming particles direction) of the target at a given proton source incident energy should be determined according to the corresponding proton range. The results of this paper will be useful to guide the design of spallation target of a reference ADS.Copyright