Katsuyuki Kawashima

Concept of hydride fuel target subassemblies in a fast reactor core for effective transmutation of MA

Abstract U–Th–Zr alloys with four different compositions were hydrogenated and examined for their hydrogen holding capacities, microstructural and hardness changes with irradiation and thermal diffusivities. Considerably high hydrogen capacity was confirmed up to about 1173 K. A certain degree of irradiation stability was observed, and relatively high thermal diffusivity was ascertained for these hydrogenated ternary alloys. Based on these results, a new concept for effective transmutations of MA was proposed, where target assemblies containing Np and Am in hydrogenated form are loaded in a fast reactor core. This concept has a great potential to achieve the best transmutation of MA with improvement of safety characteristics in a fast reactor core.

Fast Reactor Core Concepts for Minor Actinide Transmutation Using Hydride Fuel Targets

Fast reactor core concepts are studied which reduce long-term radiotoxicity of nuclear waste by using minor actinides (MAs) in the form of zirconium-hydride fuel targets. A systematic parameter survey is carried out to investigate the fundamental characteristics of MA transmutation and the core safety parameters such as sodium void reactivity in a 1,000 MWe-class fast reactor core. Two core concepts are proposed, using 36 target assemblies, by adjusting the composition of hydride fuels. One is the MA burner core to transmute a large amount of MAs in a short time combined with Pu multi-recycling in fast reactors, whereby the MAs produced in about 13 LWRs can be transmuted every year with a 58% MA transmutation rate in discharged targets. The other is the MA once-through core to incinerate a small amount of MAs by fission, whereby the MAs produced in about 2 LWRs can be incinerated every year with a 64% MA incineration rate (a 93% transmutation rate) in discharged targets. This study shows these concepts have great potential to achieve good transmutation characteristics of MAs while providing the improved safety characteristics of a fast reactor core.

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

As part of an international experimental program REBUS, core physics experiments have been implemented on a UO2 core, which consists of 3.3 and 4.0 wt% UO2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 × 7 UO2 fuel rods in the center of the UO2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k effs of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.

Applicability evaluation to a MOX fueled fast breeder reactor for a self-consistent nuclear energy system

Abstract The potential for a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in a self-consistent nuclear energy system (SCNES). The FBRs annual production amounts of selected long-lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a two layer radial blanket region without a significant impact on core nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 10 2 years is as small as that of a typical uranium ore. Regarding self-controllability in the systems safety, the proposed FBR core concept has an inherent negative reactivity feedback with a gas expansion module, sodium plenum above the core and burnup reactivity compensation module. So sodium boiling and fuel melting will be avoided in anticipated transient without scram events. Regarding self-terminability, even if the MOX fuel melting should cause a core compaction process, re-criticality of the core can be avoided by a fuel dilution and relocation module.

Utilization of fast reactor excess neutrons for burning long lived fission products

Abstract An evaluation is made on a large MOX fuel fast reactors capability of burning long lived fission product Tc99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction.The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs.

Effect of Pu-Rich Agglomerate in MOX Fuel on a Lattice Calculation

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17 × 17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small.

Design Studies of a Low Sodium Void Reactivity Core Able to Accommodate Degraded TRU Fuel

As part of the Fast Reactor Cycle Technology Development (FaCT) Project, JSFR (Japan Sodium-Cooled Fast Reactor) core design efforts have been made to cope with the transuranic (TRU) fuel compositions expected during the light water reactor (LWR)–to–fast breeder reactor transition period, during which various kinds of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and their recycling method. The sodium void reactivity, which is one of the major core safety parameters, is considerably influenced by TRU fuel compositions. The criteria assigned to the JSFR core include a void reactivity effect limited to ∼6

Control rod worth and related nuclear characteristics of an axially heterogeneous liquid-metal fast breeder reactor core

; therefore, designing a core with reduced sodium void reactivity will offer a greater margin for the core to host TRU fuel. To this end, a new core concept called BUMPY is proposed. This homogeneous core exhibits a low sodium void reactivity, due to partial-length fuels with an upper sodium plenum interspersed within the core, among other standard fuel assemblies. This core configuration enhances the upward and lateral neutron leakage from the core fuel region toward the sodium plenum when voiding to reduce void reactivity. The BUMPY core is applied to the 750-MW(electric) JSFR core design. The core can meet the design target by adjusting the loading fraction of the partial-length fuels and the height of the step in fuel lengths. The calculated void reactivity of the selected BUMPY core is 2.5

Fast reactor core concepts to improve transmutation efficiency

(25% loading fraction, 30-cm step height), which is considerably reduced from the 5.3

Analysis of a Partial-Refueling Ultra-Long-Life Core Using Metallic Fuel for 1000-MW(electric) Liquid-Metal Fast Breeder Reactors

value of the reference core. This allows the BUMPY core to accommodate 5% to 9% more minor actinides in the core compared to the reference core.