Neutronic Analysis for Transmutation of Minor Actinides and Long-Lived Fission Products in a Fusion-Driven Transmuter (FDT)

This study presents the transmutations of both the minor actinides (MAs: ²³⁷Np, ²⁴¹Am, ²⁴³Am and ²⁴⁴Cm) and the long-lived fission products (LLFPs: ⁹⁹Tc, ¹²⁹I and ¹³⁵Cs), discharged from high burn-up PWR-MOX spent fuel, in a fusion-driven transmuter (FDT) and the effects of the MA and LLFP volume fractions on their transmutations. The blanket configuration of the FDT is improved by analyzing various sample blanket design combinations with different radial thicknesses. Two different transmutation zones (TZ_MA and TZ_FP which contain the MA and LLFP nuclides, respectively) are located separately from each other. The volume fractions of the MA and the LLFP are raised from 10 to 20% stepped by 2% and from 10 to 80% stepped by 5%, respectively. The calculations are performed to estimate neutronic parameters and transmutation characteristics per D–T fusion neutron. The conversion ratios (CRs) for the whole of all MAs are about 65–70%. The transmutation rates of the LLFP nuclides increase linearly with the increase of volume fractions of the MA, and the ⁹⁹Tc nuclide among them has the highest transmutation rate. The variations of their transmutation rate per unit volume in the radial direction are quasi-concave parabolic.     

Multilateral Assessment of the Fast Reactor System as a Component of the Future Sustainable Nuclear Energy and Paths for the System Deployment

The paper reviews main findings of the Joint Assessment Study on a Nuclear Energy System (NES) based on a Closed Nuclear Fuel Cycle with Fast Reactors (CNFC-FRs) that was performed within the IAEA project INPRO.     

Behavior of Fission Products Zirconium and Barium in Fast Reactor Fuel Irradiated to High Burnup

Barium and Zr generated in nuclear fuels can precipitate as multi-component oxide with some other fission products. In addition, the solubility of Ba in the fuel depends on the oxygen potential and the temperature and Zr can easily dissolve into the fuel matrix. Therefore, the behavior of the Ba-Zr oxide inclusions during irradiation is rather complex. In this work, the composition of multi-component oxides and the distributions of Ba and Zr as a function of relative radius were evaluated with X-ray microanalysis. As results, the oxide inclusions containing both Ba and Zr and containing only Ba were observed in the fuel irradiated to the burnup of 13.3 and 10.6 at%, respectively. These results were discussed in terms of the solubility of Ba and Zr in the fuel and in terms of the rO₂–UO₂ phase diagram, together with the radial distributions of Ba and Zr in fuel matrix.     

Fusion-Driven Transmutation of Fission Product Cesium in its Elemental Form

Assuming fission reaction as a dominant energy source for a long-term perspective, the goal of transmutation of fission products is to cut their increasing accumulation and to keep their inventories at easily manageable level. Opposite to relatively short-lived ¹³⁷Cs (T_1/2=30yr) whose natural decay converge equilibrium mass to the level of order of 11 per GW of fission energy, an approach to similar equilibrium inventory for long- lived ¹³⁵Cs (T_1/2=2.3×10⁶ yr) requires artificial transmutation that preassumes its isotopic separation in the most studies. The present paper addresses cesium transmutation without preliminary isotope separation that means an approach to equilibrium for all the isotopes including stable ¹³³Cs. A high-flux blanket driven by Fusion Neutron Source with ITER-like parameters is proposed to transmute cesium in the elemental form. Transmutation efficiency is estimated in terms of equilibrium inventory and characteristic time to reach equilibrium both governed by the mean life-time of nuclides in transmuter. The analytical results show that the mean life-time of the target isotope ¹³⁵ Cs is as short as 21 yr which is in more than order of magnitude shorter than achieved in advanced fission reactors. It reveals that one Fusion Neutron Source with ITER-like parameters could transmute elemental cesium from 23 PWRs, the fraction of power associated with transmutation being as small as 3%.     

先进核燃料循环体系研究进展

概述了先进核燃料循环体系的概念 ,论述了目前后处理与分离 嬗变领域的研究进展和技术发展趋势     

Research and Development of Minor Actinide-containing Fuel and Target in a Future Integrated Closed Cycle System

Research and development of minor actinide-containing fuels and targets, i.e., (Pu,Am)O₂–MgO, (Pu,Np)O₂–MgO, (U,Pu,Np)O₂, (U,Pu,Np)N and (Pu,Np,Zr)N, for use in a future integrated closed cycle system that includes fast reactor and accelerator driven sub-critical system is underway. The present statuses of fabrication test and property measurements are given. Design concept of the oxide target is described in detail together with a screening of the support material. A new apparatus for the measurement of mechanical properties at the elevated temperature is installed for use in evaluating the fuel-cladding mechanical interaction. Development histories with future prospects of two types of Np-containing fuels for the fast reactor are mentioned. Preliminary test results for a new nitride target for the accelerator driven sub-critical system are given. Finally, an irradiation test plan in the experimental fast reactor JOYO is briefly described.     

核裂变产物对地表环境Ba同位素体系的影响

近年来,钡(Ba)同位素研究在地表生态环境领域的应用越来越广泛。然而,上世纪的核试验和核事故泄露在地表环境中释放了大量含有人工核素Ba的裂变产物,它们可能会改变地表自然样品的Ba同位素特征,从而不利于Ba同位素研究在该领域的发展和应用,因此需要评估其影响程度。本文根据Ba同位素最新研究进展和国际上已发表的数据,总结不同地表储库中Ba同位素的分布特征,利用核裂变理论计算核试验裂变产物的Ba同位素组成及裂变成因Ba的加入对环境样品Ba同位素组成的改变程度。结果显示,自然样品加入占其Ba含量0.01%的裂变成因Ba,能显著改变它的Ba同位素组成,因此需要引起重视。此外,基于计算结果发现,同时测量样品δ^137/134Ba和δ^138/134Ba值,根据其异常特征可以有效地排除受到沾染的样品。     

Prediction of Nongaseous Fission Products Behavior in the Primary Cooling System of High Temperature Gas-Cooled Reactor

In high temperature gas-cooled reactors (HTGRs), some amounts of fission products (FPs) are released mainly from fuel with failed coatings and are transported in the primary cooling system with the primary coolant during normal operation. In that case, condensable FPs plateout on the inner surface of components in the primary cooling system. On the other hand, since the HTGRs use helium gas as primary coolant, the primary coolant is not activated itself and very small amount of corrosion products is generated. Then, γ-ray emitted from the FPs becomes main source in shielding design of the HTGRs, and not only release amount from fuel but also plateout distributions of the FPs should be properly evaluated. Therefore, prediction of plateout behavior in the primary cooling system of HTGRs was carried out based on the calculation result of plateout distribution in High Temperature Engineering Test Reactor. Before the calculation, analytical model was verified by comparison with experimentally obtained plateout distributions and the applicability of the model to predict the plateout distributions in the primary cooling system of HTGR was certified.

This report describes the predicted result of plateout distribution in the primary cooling system of HTGR together with the verification result of the analytical model.     

Build-Up and Decay of Actinide Nuclides in Fuel Cycle of Nuclear Reactors

In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1,000MW. The following items were examined;

1. quantities of actinide nuclides build-up in the reactor

2. build-up and decay of activities of actinides in the spent fuel

3. build-up and decay of activities of actinides after reprocessing, and

4. variation of isotopie composition of plutonium with high burn-up.

It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

Sensitivities of Some Characteristics of Nuclear Equilibrium State to One-Group Constants

In the nuclear equilibrium state, the nuclide densities and h-value (a neutron balance index) can be estimated from one-group nuclear constants. One-group cross sections usually contain some uncertainties originated from nuclear data errors, analysis errors and design uncertainties. The method to obtain sensitivity coefficients of nuclide densities and h-value to one-group constants are derived by the use of perturbation method, and some numerical results are obtained for both fast and thermal reactors. The numerical results show several notable characteristics. It is appeared that many of these characteristics can be explained using only the main nuclide chain. The method developed in this paper, some derived characteristics and/or numerical results for typical reactors may be useful for future studies based on the nuclear equilibrium state.     

Equilibrium Distributions of Actinides and Fission Products in Pyrochemical Separation Systems, (II) LiCl-KCl/Cd System

Equilibrium distributions of actinides and fission products were determined in a LiCl-KCl/Cd system at 800–973 K. The redox potential of the system was controlled with the addition of reductant Li. Different distribution behaviors due to different group elements were observed. The group partitioning in this system was thus supported to be feasible. On reduction, however, the actinide and lanthanide elements were found to be less soluble in the Cd phase and to remain at the interface in the form of their solid intermetallics. A group partitioning process in which reductive extraction is combined with filtration was proposed on the basis of the present observations.     

Feasibility of Reprocessed Uranium in LWR Fuel Cycle for Protected Plutonium Production

Protected plutonium production (PPP) is an intrinsic measure to enhance the proliferation resistance of Pu by raising the ²³⁸Pu isotopic concentration, which denatures Pu on account of the high spontaneous fission neutron (SFN) rate and large decay heat (DH). This study is aimed at examining the feasibility of reprocessed uranium (RepU) with or without the addition of minor actinide (MA) in LWR fuel cycle for PPP and to make a tentative economic assessment of RepU possessing the PPP feature. It was analytically clarified that RepU enriched to 5% ²³⁵U by centrifugation produced denatured Pu at higher burnup than about 40GWd/t. By the addition of more than 0.5% MA to RepU and natural uranium both enriched to 5%, Pu generated in the uranium fuel with MA added could be denatured up to 40 GWd/t at least. A diagram designed with functions of SFN rate and DH explicated the PPP features of re-enriched RepU and enriched natural uranium with or without MA addition. The economic assessment indicated that the cost of fuel cycle applying re-enriched RepU would be comparable to that of the conventional fuel cycle, if the cost of the source RepU is low. In addition, the LWR fuel cycle applying RepU for PPP was discussed.     

Comparison of Calculated Values with Measured Values on the Amount of TRU and FP Nuclides Accumulated in Gadolinium Bearing PWR Spent Fuels

Destructive analyses for five spent fuel samples taken from a Gd bearing fuel assembly were done. The measured amounts of actinides of ^234-238U, ²³⁷Np, ^238-242p_u 241,242m,243Am ^242,244Cm, and fission products of ¹³⁴Cs and ¹⁵⁴Eu were used for evaluating the accuracy of calculation made by CASMO-MICBURN and ORIGEN-2 codes. The effect of Gd on the neutron spectrum was taken into account in the CASMO-MICBURN calculation.

The amounts of ²³⁵U, ²³⁹Pu and ²⁴¹Pu calculated by CASMO-MICBURN agreed well with the observed values within about 3%. On the other hand, the amounts obtained from ORIGEN-2 calculation showed lower values than those observed, especially by —12% in average in ²³⁵U for Gd₂0₃U0₂ fuel. The main cause of this large difference may be attributed to the effect of Gd on the neutron spectrum. The amounts of the other actinides by both calculation codes revealed no significant difference in nearly 10% except for ^242mAm, in which a large fluctuation among the samples was observed. About 10% difference between the measured values and the calculated values was also observed for ¹³⁴Cs, but the calculated values for ¹⁵⁴Eu showed a significant difference from measured values.     

Parametric Survey for Benefit of Partitioning and Transmutation Technology in Terms of High-level Radioactive Waste Disposal

Benefit of implementing Partitioning and Transmutation (P&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. First, the amount and characteristics of HLWs caused from UO₂ and MOX spent fuels of light-water reactors (LWR) were evaluated for various reprocessing schemes and cooling periods. The emplacement area in the repository site required for the disposal of these HLWs was then estimated with considering the temperature constrain in the repository. The results showed that, by recycling minor actinides (MA), the emplacement area could be reduced by 17–29% in the case of UO₂-LWR and by 63–85% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of 241Am which was a long-term heat source. Further 70–80% reduction of the emplacement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types. To achieve this benefit of P&T, however, it is necessary to confirm the engineering feasibility of these unconventional disposal concepts.     

Simplified Method for Calculation of Neutron Capture Transformation Effects of Fission Products on Decay Power

A simplified method is proposed for the calculation of the effects of neutron capture transformations of fission products (FPs) on the decay power of FPs. The decay power of FPs after shutdown changes by the neutron capture transformations of FP nuclides during reactor operation. It is proposed to calculate the neutron capture transformation effects considering the production of the following 7 nuclides ¹⁰³Ru, ¹³⁴Cs, ¹³⁶Cs, ^148mPm, ¹⁴⁸Pm, ¹⁵⁴Eu and ¹⁵⁶Eu by the neutron capture reaction of the direct mother nuclide alone giving a cumulative fission yield for the mother nuclide. The present method was assessed by com-paring the calculation results with the rigorous calculation results for the thermal-neutron fission of ²³⁵U irradiated between 1 and 5 yr in a light water reactor with thermal-nentron flux between 3 x 10¹³ and 6 x 10¹³ n/cm²·s and for the fast-neutron fission of ²³⁹Pu irradiated between 1 and 5 yr in a fast breeder reactor with total neutron flux between 3 x 10¹⁵ and 6 x 10¹⁵ n/cm²·s. It has been clarified that the present method can calculate the neutron capture transformation effects within the accuracy of ±1% of the decay power for the irradiation of 1yr and cooling time less than 10⁹s irrespective of fission type and neutron flux. The accuracy varies little with neutron flux but considerably with irradiation time. For a irradiation of 5 yr the present method can calculate the capture effect within the accuracy of +1% and -5% of the decay power. The accuracy can be improved to ±1% of the decay power with the simple correction factors.