Minor actinide transmutation in fast reactor metal fuels irradiated for 120 and 360 equivalent full-power days

An irradiation experiment on uranium–plutonium–zirconium (U–Pu–Zr) alloys containing 5 wt% or less minor actinides (MAs) and rare earths was carried out in the Phénix fast reactor. The isotope compositions of the fuel alloys irradiated for 120 and 360 equivalent full-power days (EFPDs) were chemically analyzed by inductively coupled plasma–mass spectrometry after 3.3–5.3 years of cooling. The results of chemical analysis indicated that the discharged burnups of the fuel alloys irradiated for 120 and 360 EFPDs were 2.1–2.5 and 5.3–6.4 at%, respectively. The changes in the isotopic abundances of plutonium, americium, and curium during the irradiation experiment were assessed to discuss the transmutation performance of MA nuclides added to U–Pu–Zr alloy fuel. Multigroup three-dimensional diffusion and burnup calculations accurately predicted the changes in these isotopic abundances after fuel fabrication. An evaluation of the MA transmutation ratio based on the results of chemical analysis revealed that the quantity of MA elements in the U–19Pu–10Zr–5MA (wt%) alloy decreased by about 20% during the irradiation experiment for 360 EFPDs.     

Short-term irradiation behavior of minor actinide doped uranium plutonium mixed oxide fuels irradiated in an experimental fast reactor

A fuel irradiation program is being conducted using the experimental fast reactor ‘Joyo’. Two short-term irradiation tests in the program were completed in 2006 using a uranium and plutonium mixed oxide fuel which contains minor actinides (MA-MOX fuel). The objective of the tests is the investigation of early thermal behavior of MA-MOX fuel such as fuel restructuring and redistribution of minor actinides. Three fuel pins which contained MA-MOX: 2% neptunium and 2% americium doped uranium plutonium mixed oxide (Am,Pu,Np,U)O_2−x fuel were supplied for testing. The first test was conducted with high-linear heating rate of approximately 430 W cm⁻¹ for only 10 min. After the first test, one fuel pin was removed for examinations. Then the second test was conducted with the remaining two pins at nearly the same linear power for 24 h. In these tests, two oxygen-to-metal molar ratios were used for fuel pellets as a test parameter. Non-destructive and destructive post-irradiation examinations results are discussed with early on the behavior of the fuel during irradiation.     

Management of actinide waste inventories in nuclear phase-out scenarios

The improvement of the “radiological cleanliness” of nuclear energy is a primary goal in the development of advanced reactors and fuel cycles. The multiple recycling of actinides in advanced nuclear systems with fast neutron spectra represents a key option for reducing the potential hazard from high-level waste, especially when the fuel cycle is fully closed. Such strategies, however, involve large inventories of radiotoxic, transuranic (TRU) nuclides in the nuclear park, both in-pile and out-of-pile. The management of these inventories with the help of actinide burners is likely to become an important issue, if nuclear energy systems are eventually phased out, i.e. replaced by other types of energy systems.     

The kinetics of actinide redistribution by vapor migration in mixed oxide fuels (II). By pores

The migration velocity of a closed lenticular pore in hyperstoichiometric mixed oxide fuel has been calculated, and the U/Pu ratio of the solid near the moving pore has been determined. Pore migration in mixed oxides differs from the analogous process in pure stoichiometric urania in that the composition as well as the temperature of the two faces of the pore are different. It was found that pore migration is not as effective a mechanism of actinide redistribution as vapor transport along cracks. The velocity of the pores in (U_0,8Pu_0,2) O_{2 + x} is $\text{～3}\text{1}\text{2}$ times faster than that in pure UO₂.     

The kinetics of actinide redistribution by vapor migration in mixed oxide fuels (I). By cracks

The kinetics of vapor migration-driven restructuring and actinide redistribution along continuous cracks was studied analytically. The fissured fuel was modeled by a network of rectangular cracks with the temperature gradient along the major axis. Redistribution of fuel constituents was considered to occur by molecular diffusion in the gas filling the cracks. Hyperstoichiometric U-Pu oxide was treated so that only a single diffusing species (UO₃) needed to be considered, at least in the early stages of the process. The dependence of the kinetics upon oxygen-to-metal ratio, temperature gradient and crack dimensions was investigated for fuel containing 20% plutonium.     

Some aspects of the use of ZrN as an inert matrix for actinide fuels

The sodium compatibility and the nitric acid dissolution of (Zr_0.80U_0.20)N, prepared by carbothermic reduction of the oxide, were determined. No interaction with liquid sodium (T=823 K) was observed. The material readily dissolved in nitric acid (T=378–383 K). From these results it is concluded that ZrN is an attractive inert matrix in fast reactor fuels for the incineration of plutonium and minor actinides.     

Plutonium management with thorium-based fuels and inert matrix fuels in thermal reactor systems

The plutonium that is produced by light water reactors worldwide is currently re-used to a limited extent. In the last century, the expected introduction of fast reactors and the associated need for large amounts of plutonium did not take place. The result is that worldwide a stockpile of excess plutonium has formed, which is the dominant contributor to the radiotoxicity of spent nuclear fuel for storage times from 10² to 10⁵ years. One option to reduce or stabilize the plutonium stockpile is to utilize this plutonium in advanced fuel types, such as thorium-based and inert matrix fuels. Because these fuels do not contain uranium, the plutonium consumption rate is very high. In this paper, the status of the fuel research and some recent developments are given.     

Design and analysis of a uranium hexafluoride actinide transmutation reactor

The long-term (> 1000 years) hazards of high-level wastes (HLW) can be reduced substantially by practising waste-actinide partitioning-transmutation (P-T). This paper investigates the waste-actinide transmutation performance of a uranium hexafluoride actinide transmutation reactor (UHATR). Using mostly present-day and near-term technology, a preliminary UHATR design is established. Because of the gaseous nature of the fuel, very high neutron fluxes are obtained. Compared with an LWR, the average blanket thermal flux of this UHATR is about 10–30 times higher, leading to a 15-fold improvement in the percentage of actinides fissioned per year of irradiation.     

EPR机组反应堆厂房空气污染评估

为评估EPR机组功率运行和停堆期间反应堆厂房的空气污染水平,本文介绍了空气污染评估方法,并建立了H-3的空气浓度估算模型;估算了功率运行期间和停堆期间反应堆厂房设备间和工作间的空气污染水平。评估结果显示,在停堆期间,反应堆厂房空气污染的主要核素是H-3,大约为0.07 DAC,其导致的大修集体剂量可以达到56人·mSv。     

Neutronic potential of water cooled reactor with actinide closed fuel cycle

Neutronic potential of water-cooled reactor on the efficient use of Uranium by multiple recycling of all actinides have been examined after a brief-review of alternative coolants to sodium which are applicable for fast neutron reactors. The water-cooled reactor which was designed to have tight lattice and lower Hydrogen/Heavy-Metal number density ratio (H/HM<0.5) showed enough neutronic potential to make the fuel cycle closed for actinides and could be feasible to be operated not as a breeder but as a self-fuel sustainer. The recycle system requires no external fissile supply and generates no actinide as waste except for the inevitable recovery loss, therefore the Uranium resource could be efficiently used and sustained for a long period similar to the case of sodium cooled system.     

Spectrum shifting as a mechanism to improve performance of VHTRs with advanced actinide fuels

Reprocessing of spent LWR fuel is an intrinsic part of the closed fuel cycle. While current technologies treat recovered minor actinides as high level wastes, the primary objective of one of the U.S. DOE Nuclear Energy Research Initiative (NERI) projects is to assess the possibility, advantages and limitations of designing a single-batch (no-refueling) very high temperature reactor (VHTR) configuration that utilizes transuranic nuclides (TRU) as a fuel component. Since both VHTR core design concepts, pebble bed and prismatic block assembly, permit flexibility in component configuration, fuel utilization and management, it is possible to improve fissile properties by neutron spectrum shifting through configuration adjustments. The presented analysis is focused on the TRU-impact on the single-batch mode (no-refueling) VHTR core lifetime. As a result of the analysis, promising performance characteristics have been demonstrated. The TRU-core configurations are expected to be suitable for long-term autonomous operation without intermediate refueling.     

Fission gas production in reactor fuels including the effects of ternary fission

An understanding of gas bubble formation and migration in nuclear fuel and its impacts on fuel and cladding materials requires knowledge of the isotopic composition of the gases and their generation rates. In this paper, we present results of simulations for the production of the dominant noble gases (helium, xenon, krypton) in nuclear fuels for different reactor core configurations and fuel compositions. The calculations were performed using detailed nuclear burn simulations with Monte Carlo nuclear transport, and included ternary fission to ensure an accurate treatment of helium production. For all reactor designs and fuels considered xenon was found to be the most dominant gas produced. Variation in the composition of fission gases is quantified for: (1) the burn time, (2) the composition of the fuel, and (3) the neutron energy spectrum. These three factors determine the relative fraction of each gas and its transmutation into or from stable gas by subsequent neutron capture.     

Comparative study of minor actinide transmutation in sodium and lead-cooled fast reactor cores

This paper shows that lead-cooled and sodium-cooled fast reactors (LFRs and SFRs) can preferentially consume minor actinides without burning plutonium, both in homogeneous and in heterogeneous mode. The former approach consists of admixing about 5% of minor actinides (MAs) into uranium–plutonium fuels in the core and using a limited number of thermalising pins consisting of UZrH_1.6. These are needed to keep the negative Doppler feedback larger than the positive coolant reactivity coefficient. Our Monte Carlo burn-up calculations showed that a 600 MW_e LFR self-breeder without blankets can burn an average of around 67 kg annually of MAs with a reactivity swing of only about −0.7$ per year. The reactivity swing of a corresponding 600 MW_e SFR is more than three times larger due to the poorer breeding and half the critical mass in comparison to the LFR. However, when axial and radial blankets loaded with 10% MAs are added, the SFR burns 25% more MAs (131 kg/yr) and breeds 30% more Pu (150 kg/yr) than an equally sized LFR. When only the blankets are loaded with MAs, the SFR breeds 30% more Pu (198 kg/yr) and still burns about 60 kg a year of MAs. However, in terms of severe accident behaviour, the LFR, with its superior natural coolant circulation and larger heat capacity, has definite advantages.     

On estimation of critical temperatures of fast reactor fuels

The applicability of Watson's method to estimate critical temperatures T_c of fast reactor fuels using their normal boiling points T_b, liquid density at T_b and molecular weight is reviewed in the context of the peculiar situation arising from the dissociation of those polyatomic nuclear materials into different gas-like species in the vapour phase. A simpler empirical relation proposed by Gates and Thodos for evaluating T_c directly from T_b appears to be more appropriate for such materials, with its predictions comparing favourably with more sophisticated approaches for both oxide and carbide fuels. Based on the consistency between these independent approaches, the most probable values of T_c are recommended.