On feasibility of a closed nuclear power fuel cycle with minimum radioactivity

Practical implementation of a closed nuclear fuel cycle implies solution of two main tasks. The first task is creation of environmentally acceptable operating conditions of the nuclear fuel cycle considering, first of all, high radioactivity of the involved materials. The second task is creation of effective and economically appropriate conditions of involving fertile isotopes in the fuel cycle. Creation of technologies for management of the high-level radioactivity of spent fuel reliable in terms of radiological protection seems to be the hardest problem.     

Variants of closing the nuclear fuel cycle

Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.     

Possibility of utilizing the D-3He fuel cycle with 3He production in a spherical tokamak reactor

Possible operating regimes of a spherical tokamak reactor based on the D-³He fuel cycle with ³He production are considered. The parameters of the plasma and magnetic system are calculated for several versions corresponding to the high power efficiency (with a power gain factor in plasma of Q = 20) in a reactor with an aspect ratio of A = 1.5. According to calculations, for an axial magnetic field in vacuum of B ₀ = 2 T, a plasma radius of a = 3 m, an average 〈β〉 value of 0.53, and a plasma temperature of 〈T〉 = 48 keV, the reactor power can reach P _fus = 500 MW. In order to achieve a power of P _fus = 1500 MW in a reactor with a = 2 m, 〈β〉 = 0.36, and 〈T〉 = 40 keV, the magnetic field should be increased to B ₀ = 5 T.     

Swedish containers for disposal of spent nuclear fuel and radioactive waste

The purpose of a disposal is to isolate the radioactive waste from man and the environment. If the isolation is broken, the leakage and transport of radioactive substances must be retarded. The package is one of several barriers, used to achieve these two main functions. For short-lived, low and intermediate level waste four standard containers of steel and concrete are used. Spent fuel will be placed in a canister consisting of a pressure-bearing insert of cast nodular iron and an outer corrosion barrier of copper before it is deposited in a deep geological repository. In particular, the development of a high integrity copper canister for the isolation of spent nuclear fuel is described in this paper. To cite this article: T. Hedman et al., C. R. Physique 3 (2002) 903–913.     

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction ^232+233+234U and ²³¹Pa are formulated. (1) The fuel cycle would shift from fissile ²³⁵U to ²³³U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of ²³¹Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.     

On the possibility of nuclear spin polarization in fusion reactor plasmas

Individual nuclei may remain polarized for long times in fusion reactor plasmas. The ensemble of nucleiu is depolarized by precession of the spins in inhomogeneous fields. Reversible mixing may be expected to occur rapidly with respect to the fusion-time scale. Irreversible mixing can be shown to occur on times of the order of fusion reaction time due to additional perturbations of the particle orbit.     

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of ²³³U and mode of operation in self-sufficient cycle. For the mode of accumulation of ²³³U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for ²³³U production. Plutonium was placed in fuel channels, while ²³²Th was located in target channels. Maximum content of ²³³U in target channels was estimated to be ∼13 kg/t of ThO₂. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires ²³³U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO₂ and certain amount of ²³³U. It is shown that nonuniform distribution of ²³³U in fuel assembly is preferable.      

On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.     

Measure of low radioactive contamination and nuclear recoil quenching factor using cryogenic particle detectors

For double beta decay and dark matter experiments the determination of the radioactive contamination of materials used is of fundamental importance. The cryogenic particle detectors allow the realization of detectors with various types of materials and in this way the radioactivity is measured in a calorimetric approach with very high sensitivity. In particular a new complete analysis for different types of leads, modern and ancient roman, was realized and here presented. The detector sensitivity for a nuclear recoil event is a crucial point in the dark matter experiments. The nuclear recoil detection efficiency for cryogenic particle detectors was directly measured using two TeO₂ crystals facing one with other and performing a coincidence-anticoincidence analysis of the collected data. The determination of the quenching factor for low energy nuclear recoils in the TeO₂ cryogenic detectors shows that these devices are very sensitive to a possible dark matter particle interaction. In cryogenic detectors a nuclear recoil generates signals with the same amplitude of an electron with the same energy. The performed measure and the data analysis will be discussed in detail. Finally the experimental setup and the preliminary test of the new array of TeO₂ cryogenic detectors for double beta decay and dark matter search will be discussed in detail.     

球环型产氚聚变堆中子学分析

对球环型产氚聚变堆概念设计中的中子学设计进行了计算分析。此设计利用了球形环的先进等离子体物理性能和紧凑的结构特征,并尽量利用真空室内的空间安置氚生产包层以减少氚泄漏而提高氚增殖率,达到年产氚量1kg的目标。2D中子学计算得到的氚增殖率高于1．68的设计是其它类似设计没有达到的,进一步体现出球环型产氚聚变堆的先进性。     

Improved analysis on multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle

An FBR closed fuel cycle involves recycling of the discharge fuel, after reprocessing and refabrication, to utilize the unburnt fuel remains and the freshly bred fissile material. Our previous study in this regard for the PFBR indicated a comfortable feasibility of multiple recycling with selfsufficiency. In the present work, more refined estimations are done using the most recent nuclear data, viz. ENDF/B-VII.0, and with the most recent specification of the fuel composition. Among others, this paper brings out the importance of taking into account the energy self-shielding effects in the cross-section averages used in the study. While self-shielded averages lead to realistic predictions, unshielded averages significantly overpredict breeding in the blankets and underpredict loss in the cores.     

Ultra-dense deuterium: A possible nuclear fuel for inertial confinement fusion (ICF)

The ejection of deuterons with kinetic energy release (KER) of 630 eV was proved recently by measuring the laser-induced ion time-of-flight (TOF-MS) with two different detectors at different distances [S. Badiei, P.U. Andersson, L. Holmlid, Int. J. Mass Spectrom. 282 (2009) 70]. Realizing that the only possible energy release mechanism is Coulomb explosions, the D-D distance in the ultra-dense deuterium was determined to be constant at 2.3 pm. Using a long TOF-MS path now gives improved resolution. We show the strong effect of collisions in the ultra-dense material, and demonstrate that the kinetic energy of the ions increases with laser pulse power but that the number of ions formed is independent of the laser pulse power. This indicates special properties of the material. We also show that the two forms of condensed deuterium D(1) and D(−1) can be observed simultaneously as well resolved mass spectra of different forms. No intermediate bond lengths are observed. The two forms of deuterium are stable and well separated in bond length. We suggest that they switch rapidly back and forth as predicted by theory. A loosely built form with planar clusters of D(1) is observed here to be related to D(−1) formation.     

The internal propagation of fusion flame with the strong shock of a laser driven plasma block for advanced nuclear fuel ignition

An accelerated skin layer may be used to ignite solid state fuels. Detailed analyses were clarified by solving the hydrodynamic equations for nonlinear force driven plasma block ignition. In this paper, the complementary mechanisms are included for the advanced fuel ignition: external factors such as lasers, compression, shock waves, and sparks. The other category is created within the plasma fusion as reheating of an alpha particle, the Bremsstrahlung absorption, expansion, conduction, and shock waves generated by explosions. With the new condition for the control of shock waves, the spherical deuterium-tritium fuel density should be increased to 75 times that of the solid state. The threshold ignition energy flux density for advanced fuel ignition may be obtained using temperature equations, including the ones for the density profile obtained through the continuity equation and the expansion velocity for the r ≠ 0 layers. These thresholds are significantly reduced in comparison with the ignition thresholds at x = 0 for solid advanced fuels. The quantum correction for the collision frequency is applied in the case of the delay in ion heating. Under the shock wave condition, the spherical proton-boron and proton-lithium fuel densities should be increased to densities 120 and 180 times that of the solid state. These plasma compressions are achieved through a longer duration laser pulse or X-ray.     

Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle

Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.     

Optimum ions for a nuclear reactor with neutron illumination

The main advantage of fission reactors with neutron illumination is that, owing to the neutron illumination, they operate in the subcritical mode and, hence, are safe. Neutrons required for the illumination are generated during nuclear fragmentation in a target irradiated by 1–2 GeV protons. A substantial disadvantage of this method for energy generation is the severe restriction imposed on the power of a plant. This restriction is due to the restriction imposed on the intensity of an accelerated-proton beam by the tolerances for the activation of accelerator structural elements. The substitution of heavier nuclei (from carbon to argon) for protons is shown to substantially increase the intensity of an accelerated-ion beam and to provide a commercially reasonable thermal power of ∼4–6 GW.     

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.     

Systematic impact of spent nuclear fuel on θ13 sensitivity at reactor neutrino experiment

Reactor neutrino oscillation experiments, such as Daya Bay, Double Chooz and RENO are designed to determine the neutrino mixing angle θ13 with a sensitivity of 0.01--0.03 in sin2 2θ13 at 90% confidence level, an improvement over the current limit by more than one order of magnitude. The control of systematic uncertainties is critical to achieving the sin2 2θ13 sensitivity goal of these experiments. Antineutrinos emitted from spent nuclear fuel (SNF) would distort the soft part of energy spectrum and may introduce a non-negligible systematic uncertainty. In this article, a detailed calculation of SNF neutrinos is performed taking account of the operation of a typical reactor and the event rate in the detector is obtained. A further estimation shows that the event rate contribution of SNF neutrinos is less than 0.2% relative to the reactor neutrino signals. A global χ2 analysis shows that this uncertainty will degrade the θ13 sensitivity at a negligible level.     

Odd-isotope enrichment studies of Gd by double resonance laser-ionization for the production of burnable nuclear reactor poison

 Because of its large neutron capture cross-section, gadolinium (Gd) is widely used as burnable poison in nuclear reactors. Only two of its naturally occurring seven isotopes are strong neutron absorbers; hence, it would be desirable to enrich Gd in those isotopes. We demonstrate experimentally that significant enrichment can be achieved by two-step laser-ionization by relatively broad-band (bandwidth 1–2 GHz) lasers and by utilizing the isotope shifts only. Our results indicate that the content of the strong absorbers can be raised from the natural 30% to nearly 70%. We also compare the performance of this partially enriched Gd to natural Gd and to pure ¹⁵⁷Gd, the most strongly absorbing isotope, as burnable poison. Received: 3 April 1996/Revised version: 14 October 1996     

不同类型核燃料对热管冷却反应堆燃耗性能的影响

热管冷却反应堆采用固态反应堆设计理念,具有功率密度高、结构紧凑、固有安全性高等特点,在深空探索、深海勘探、偏远地区等场景中具有广阔的应用前景。核燃料作为热管冷却反应堆的重要组成部分,不同类型核燃料在堆芯燃耗分析时会呈现不同的中子学性能。基于美国爱达荷国家实验室(INL)提出的热管冷却反应堆INL Design A,利用清华大学蒙特卡罗中子输运程序RMC (Reactor Monte Carlo code)建立堆芯物理模型,选取UO₂,(U_0.9Pu_0.1)O₂,U-10Zr,U-8Pu-10Zr,UN,UC这6种核燃料开展燃耗计算,分析了不同核燃料、不同功率水平对热管冷却反应堆堆芯燃耗性能的影响。计算结果表明:在堆芯燃耗深度相同情况下(20.8 GW·d·t?1),装载U-8Pu-10Zr燃料的堆芯所需235U富集度最低(9.8%),具有较好的U-Pu增殖性能。堆芯功率处于5 MW的热管冷却反应堆,燃料中241Pu的存在不仅没起到增大堆芯燃耗深度的作用,反而导致堆芯剩余反应性和堆芯寿期末次锕系核素(MAs)的产量增大,影响反应堆的安全性与经济性。因此,对于装载含有Pu燃料的小功率长寿期热管冷却反应堆,需重点关注241Pu对堆芯燃耗性能的影响。