Safety assessment methodology and analysis of production during plutonium recycling at the State Science Center of the Russian Federation—Scientific-Research Institute of Nuclear Reactors

The methodology and criteria for safety assessment of nuclear fuel cycle technological processes are proposed, substantiated, and checked in large-scale recycling of plutonium (500 kg). The results of comprehensive investigations of the radiation-ecological conditions during the experimental production of mixed uranium-plutonium fuel and fuel assemblies at the State Science Center of the Russian Federation— Scientific-Research Institute of Nuclear Reactors are presented. A methodology and an experimental data bank can be used for safety assessment of commercial recycling of plutonium and Np, Am, and Cm in the nuclear fuel cycle. 4 figures, 3 tables, and 13 references. State Science Center of the Russian Federation—Scientific-Research Institute of Nuclear Reactors. Translated from Atomnaya énergiya, Vol. 87, No. 4, pp. 266–275, October, 1999.     

Design and Calculation of Buncher in Pulsing Beam System

The buncher system is the one of the main parts in the pulsing beam generation system for 10 MeV central region of model （CRM） compact cyclotron and it will modulate the beam in the longitudinal direction. The beam with special phase width will be compressed into very short pulse in the time with buncher. With the modulation of a buncher, the charged particles in the continuous beam segments will come into being very short pulse after a long drift, and this pulse length should be less than the RF phase acceptance of the CRM cyclotron, that is, the pulsed beam can be captured by the RF system of the cyclotron completely. Because the velocities of the charged particles will be modulated with the buncher,     

Design optimisation and measuring techniques for the neutronics experiment on a HCLL–TBM mock-up

A neutronics experiment on a mock-up of the EU Test Blanket Module (TBM), helium cooled lithium lead concept, is in preparation with the objective to validate the capability of the neutronics codes and nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. Three independent measurements of the TPR will be performed using Li₂CO₃ pellets. Other measurement techniques have been developed using thermo-luminescence detectors, and diamond detectors covered with ⁶LiF. Neutron flux spectra will also be measured from fast energies down to thermal energies, relevant for TPR. Comparison of measured quantities (E) with the same calculated quantities (C) will be provided, together with the related uncertainties.The paper presents the results of development of the measurement techniques and their relevance for tritium measurements in TBM in ITER. It presents also the pre-analyses conducted to optimise the mock-up configuration so that the neutron spectra are as similar as possible to those in the TBM in ITER. Sensitivity/uncertainty assessments of the TPR show that the calculation uncertainty due to the uncertainties of the neutron cross sections amounts to a few %, depending on position. The largest uncertainties are due to the elastic scattering (n,2n), and (n,3n) reactions on Pb.     

Design and Calculation of Buncher in Pulsing Beam System

1 Basic theory of the buncher The buncher system is the one of the main parts in the pulsing beam generation system for 10 MeV central region of model (CRM) compact cyclotron and it will modulate the beam in the longitudinal direction. The beam with special phase width will be compressed into very short pulse in the time with     

Design and experimental activities in view of Water Detritiation–Isotopic Separation Systems combination in TRENTA facility

To support the research activities needed to characterize the performance of various components for the Water Detritiation System (WDS) and the Isotope Separation System (ISS) processes needed for the ITER design, an experimental facility called TRENTA, simulating for the ITER WDS and ISS protium separation column, has been commissioned at Tritium Laboratory Karlsruhe (TLK). The TRENTA facility has been conceived to allow operation in a closed loop with respect to tritium inventory and to allow investigations of key design and operation issues of combined CECE (Combined Electrolytic Catalytic Exchange) and CD (Cryogenic Distillation) processes in similar conditions as envisaged for the ITER WDS–ISS. Activities combining CECE and CD processes are on going at TLK. For the CD system, dedicated heat-exchangers have been designed and manufactured to make the combination possible. The system of heat-exchangers has to provide a double barrier to avoid tritium contamination of the helium stream. The design of the heat-exchangers for feeding of the CD column, equilibrator loop and condenser will be presented. In addition, the ongoing experimental activities for to the investigation of different CD fillings will be presented.     

Evolution of microstructure and second-phase particles in Zr–Sn–Nb–Fe alloy tube during Pilger processing

Evolution of microstructure and second-phase particles (SPPs) in Zr–Sn–Nb–Fe alloy tube were investigated during Pilger process using electron backscatter diffraction, secondary electron and transmission electron microscopy imaging techniques. Results show that the Pilger rolled tubes present heterogeneous structures with the C axes of less deformed grains mostly concentrated in the axial direction. During the Pilger rolling, the increase of deformation caused weakening of linear distribution of second-phase particles. The mean diameters of the precipitates are in the range of 70–100 nm in all specimens, and the growth mechanism of SPPs follows second-order kinetics. The grain growth is controlled by Zener pinning in the Pilger rolling–annealing specimens. Clusters containing the Zr(Nb,Fe)₂ and β_Nb precipitates formed in the Zr–1.0Sn–1.0Nb–0.12Fe alloy. Most of the particles located in grain boundaries are the Zr(Nb,Fe)₂ Laves phase with hexagonal structure, and stacking faults have been found in the Zr(Nb,Fe)₂ precipitates. The types, morphology and distribution of precipitates depend on the constituent and structural fluctuations of the nucleation area.     

In-depth analysis of international collaboration and inter-institutional collaboration in nuclear science and technology during 2006–2015

Nuclear science and technology (NST) is a promising but dangerous field, and the collaboration patterns were cautiously encouraged to promote this field. An in-depth analysis was conducted to reveal overall trends, collaboration patterns, citation impact, and collaboration networks in NST category of the Science Citation Index Expanded. Newly developed indicators of TC_year, CPP_year, SNI, and CPP_SNI were employed for the concrete analysis. Fruchterman Rheingold default layout algorithm in Gephi 0.9.1, a data visualization and manipulation software, was also used to visualize the network and relationships of collaboration among countries, and institutions. Results show that the leading institutions and countries involve more inter-institutional and international collaborations. The USA and European countries were important centers of global international collaboration network, while Iran was isolated from the international community. National Institute for Nuclear Physics from Italy and Japan Atom Energy Agency were two centers of inter-institutional collaboration network. International collaboration improved scientific impact of almost all countries in NST field. Moreover, a close relationship between one country's scientific papers and nuclear power stations was revealed. This study originally revealed the collaboration patterns and impact of NST during 2006–2015, and provided important information and prior collaborators for researchers and policy-makers.     

Shielding Design and Analysis of Fusion–Fission Hybrid Energy Reactor Blanket

A preliminary design of fusion–fission hybrid energy reactor (FFHER) has been proposed by Institute of Nuclear Physics and Chemistry based on current fusion science and well-developed fission technology. In FFHER, shield blocks provide nuclear shielding and thermal shielding for internal and external blanket components. The hybrid of fusion core and fission blanket makes the spectra rather complex. Therefore, it is necessary to make detail shielding design and carry out radiation analysis according to the blanket structure and material property. In this study, a shielding design of combining several different material shield blocks has been proposed. The shielding analysis is performed by Monte Carlo (MC) method. For the radiation deep-penetration problem, the flux and statistical relative error of forward MC estimate are applied to get an optimal weight window for global variance reduction (GVR). The spatial distribution of neutron and gamma flux have been assessed along the shield block depth. Power deposited and dose rate distributions have also been simulated and analysed. Neutron irradiation damage has been studied to evaluate the material damage. Based on the configuration analysis, nuclear analysis and GVR method, an optimal FFHER blanket shielding design has been obtained.     

Design and analysis of the IFMIF–EVEDA beam dump cooling system

The IFMIF–EVEDA beam dump is designed to stop a 9 MeV, 125 mA continuous wave deuteron beam that deposits along its surface a total of 1.125 MW. The beam dump design is based on a 2.5 m long copper cone whose inner surface absorbs the beam. This piece is cooled by water flowing at high velocity through the annular channel formed between it and a second piece (shroud) made of four truncated cones of slightly different slopes.In this paper the beam dump cooling system will be briefly described, and the relevant 1D and 3D results will be presented paying especial attention to the computational fluid dynamics results.     

Design and development of a plastic scintillator based whole body β/γcontamination monitoring system

Plastic scintillation detectors based whole bodyβ/γcontamination monitors are developed for use in radi-ation facilities. This microcontroller-based multi-detec...     

On monitoring the tritium breeder in a lead–lithium cooled ceramic breeder (LLCB) module of the ITER

The operation of a tritium breeder is a most process among engineering problems of DEMO. In this study, a design for monitoring tritium-breeding in the reactor is discussed. Additionally, a system for the experimental estimation of the tritium-breeding ratio (TBR) and the tritium-breeding dynamics in a lead–lithium cooled ceramic breeder (LLCB) test module used in the ITER is proposed. The systems are based on tritium and neutron-flux measurements under the ITER plasma D–T experiments and the use of lithium ortho-silicate and lithium carbonate samples and neutron detectors. Different lithum-6 and lithium-7 isotope contents in the samples are used to measure neutron spectrum. The samples and detectors are delivered in containers to the test breeder module (TBM) on a monitor channel connecting the TBM to an operating zone of the ITER. The tritium content in the samples is measured in a laboratory by the liquid scintillation method.Pneumatic control is used to deliver the samples to the TBM and to extract the samples using the channel during plasma-operational pauses. Neutron calculation is performed to estimate the tritium content in the samples and the heat distribution in the materials of the channel under reactor irradiation. A measurement accuracy of the tritium content in the carbonate and orthosilicate samples can attain a level of 7% and 10%, respectively. The results of the channel-cooling calculation performed under the nominal operating conditions of the TBM (a plasma pulse) are presented in the paper.     

Progress on the development of H-concentration probes in eutectic lead–lithium: Synthesis and characterization of electrochemical sensor materials

Dynamic tritium concentration measurement in lithium–lead eutectic (17% Li–83% Pb) is of major interest for a reliable tritium testing program in ITER TBM and for an experimental proof of tritium self-sufficiency in liquid metal breeding systems. Potentiometric hydrogen sensors for molten lithium–lead eutectic have been designed at the Electrochemical Methods Lab at Institut Quimic de Sarria (IQS) at Barcelona and are under development and qualification. The probes are based on the use of solid state electrolytes and works as Proton Exchange Membranes (PEM).In this work, the following compounds have been synthesized in order to be tested as PEM H-probes: BaCeO₃, BaCe_0.9Y_0.1O_3?δ, SrCe_0.9Y_0.1O_3?δ and Sr(Ce_0.9–Zr_0.1)_0.95Yb_0.05O_3?δ. Potentiometric measurements of the synthesized ceramic elements have been performed at different hydrogen concentrations at 500 °C. In this campaign, a fixed and known hydrogen pressure has been used in the reference electrode. The sensors constructed using the proton conductor elements BaCeO₃, SrCe_0.9Y_0.1O_3?δ and Sr(Ce_0.9–Zr_0.1)_0.95Yb_0.05O_3?δ exhibited quite stable output potential and its value was quite close to the theoretical value calculated with the Nernst equation (deviation less than 100 mV). Unstable measurement was obtained using BaCe_0.9Y_0.1O_3?δ as a solid state electrolyte in the sensor.     

Radiotoxicity and Residual Energy Release of Spent Uranium–Plutonium and Thorium Fuel during Prolonged Storage

A computational comparison is made of the radiotoxicity and residual energy release of spent uranium–plutonium and thorium–uranium nuclear fuel during long-term storage. The contribution of the most significant nuclides, whose primary extraction and transmutation make it possible to decrease energy release and radiotoxicity of the remaining stored wastes, is determined. 2 figures, 4 references.     

Radiation Characteristics of Fuel and Wastes in the Uranium–Plutonium and Thorium–Uranium Fuel Cycle

The radiation characteristics of fuel cycles of various reactors – replacement candidates in the future nuclear power – are compared. Proceeding from the basic requirements (safety, fuel supply, and nonproliferation of fissioning materials), inherently safe fast reactors of the BREST type can be used as the basis for large-scale nuclear power. Thermal reactors, which can burn enriched uranium, thorium–uranium fuel, or mixed uranium–plutonium fuel with makeup with fissioning materials from fast reactors, will operate for a long time simultaneously with fast reactors in the future nuclear power. VVÉR-1000 and CANDU reactors are examined as representatives of thermal reactors; for each of these reactors the operation in various variants of the fuel cycle is simulated. It is shown that with respect to radiation characteristics of the fuel and wastes the thorium–uranium fuel cycle has no great advantages over the uranium–plutonium cycle.     

Amorphization of the interaction products in U–Mo/Al dispersion fuel during irradiation

The microstructures of the product resulting from interaction between U–Mo fuel particles and the Al matrix in U–Mo/Al dispersion fuel are discussed. We analyzed the available characterization results for the Al matrix dispersion fuels from both the out-of-pile and in-pile tests and examined the difference between these results. The morphology of pores that form in the interaction products during irradiation is similar to the porosity previously observed in irradiation-induced amorphized uranium compounds. The available diffraction studies for the interaction products formed in both the out-of-pile and in-pile tests are analyzed. We have concluded that the interaction products in the U–Mo/Al dispersion fuel are formed as an amorphous state or become amorphous during irradiation, depending on the irradiation conditions.     

Design and analysis on tritium system of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

As early application of fusion technology, the fusion–fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion–fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.