Transmutation Of Actinides Research Articles

Use of Thorium for Transmutation of Plutonium and Minor Actinides in PWRs

An assessment is made of the potential for Th-based fuel to minimize Pu and minor actinide (MA) production in pressurized water reactors (PWRs). Destruction rates and residual amounts of Pu and MA in the fuel used for transmutation are examined. In particular, sensitivity of these two parameters to the fuel lattice hydrogen to heavy metal (H/HM) ratio and to the fuel composition was systematically investigated. All burnup calculations were performed using CASMO4, the fuel assembly burnup code. The results indicate that up to 1000 kg of reactor-grade Pu can be burned in Th-based fuel assemblies per gigawatt (electric) year. Up to 75% of initial Pu can be destroyed per passage through reactor core. Addition of MA to the fuel mixture degrades the burning efficiency. The theoretically achievable limit for total transuranium (TRU) destruction per passage through the core is 50%. Efficient MA and Pu destruction in Th-based fuel generally requires a higher degree of neutron moderation and, therefore, higher fuel lattice H/HM ratio than typically used in the current generation of PWRs. Reactivity coefficients evaluation demonstrated the feasibility of designing a Th-Pu-MA fueled core with negative Doppler and moderator temperature coefficients. Introduction of TRU-containing fuels to a PWR core inevitably leads to lower control material worths and smaller delayed-neutron yields than with conventional UO2 cores. Therefore, a major challenge associated with the introduction of Th-TRU fuels to PWRs will be the design of the whole core and reactor control features to ensure safe reactor operation.

Use of Thoria-Urania Fuels in PWRs: A General Review of a NERI Project to Assess Feasible Core Designs, Economics, Fabrication Methods, In-Pile Thermal/Mechanical Behavior, and Waste Form Characteristics

This paper provides an introduction to and a summary of the remaining papers in this issue of Nuclear Technology. The papers in this issue present the important results from a U.S. Department of Energy–sponsored Nuclear Engineering Research Initiative (NERI) project to study the efficacy of the thorium-uranium dioxide (ThO2-UO2) once-through fuel cycle in current light water reactors. The project addressed fuel cycle neutronics and economics; ThO2-UO2 fuel manufacturing; the in-pile thermal/mechanical behavior of ThO2-UO2 fuel during normal, off-normal, and accident conditions; and the long-term stability of ThO2-UO2 waste. Results from this work show that a small-scale separation of the uranium and thorium will enhance the fuel reactivity and achievable burnup from uranium-thorium dioxide fuels. Under conditions that meet the thermal requirements in present pressurized water reactors (PWRs), a properly designed microheterogeneous fuel will have more reactivity than all-uranium fuel, and the overall production of plutonium is significantly reduced. The use of thorium as a host for actinide fuels when PWRs are used for actinide transmutation was also explored. It was also determined that there were no fundamental obstacles to converting the current plants that manufacture uranium oxide–only fuel to a mixed ThO2-UO2 fuel. Also, the in-service and transient thermal and mechanical performance of homogeneous ThO2-UO2–based fuels with respect to safety is generally equal to or better than that of all-uranium fuel. Furthermore, a mixed thorium-uranium dioxide spent fuel appears to be a much more stable waste form than uranium oxide spent fuel.

Minor Actinide Transmutation Potential of Modified PROMETHEUS Fusion Reactor

This study presents the investigation of the burning and/or transmutation (B/T) of minor actinides (MAs) in the modified PROMETHEUS-H fusion reactor. The calculations were performed for an operation period (OP) of up to 10 years by 75% plant factor (η]) under a neutron wall load (P) of 4.7 MW/m 2 . In order to incinerate and transmute the MAs effectively, the transmutation zone (TZ) containing the mixture of MA nuclides, discharged from the pressured water reactor (PWR)-MOX spent fuel, was located in the modified blanket of the PROMETHEUS-H fusion reactor. Two different blanket modifications were considered, (the Models A and B). The MA mixture was spherically prepared and cladded with SiC to prevent fission products from contaminating coolant and the MAs from contacting coolant. Helium was selected for the nuclear heat transfer in the TZ. The effect of the MA volume fraction in the TZ on the B/T was also investigated. The results bring out that the MAs are converted by the transmutation rather than the incineration. Both models can reduce significantly the effective half-lives of the MA nuclides by burning and/or transmuting these nuclides, and at the same time produce substantial electricity in situ.

The Use of Molybdenum-Based Ceramic-Metal (CerMet) Fuel for the Actinide Management in LWRs

The technical and economic aspects of the use of molybdenum depleted in the isotope 95Mo (DepMo) for the transmutation of actinides in a light water reactor are discussed. DepMo has a low neutron absorption cross section and good physical and chemical properties. Therefore, DepMo is expected to be a good inert matrix in ceramic-metal fuel. The costs of the use of DepMo have been assessed, and it was concluded that these costs can be justified for the transmutation of the actinides neptunium, americium, and plutonium.

Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor

Significant amounts of nuclear wastes consisting of plutonium, minor actinides and long lived fission products are produced during the operation of commercial nuclear power plants. Therefore, the destruction of these wastes is very important with respect to public health, environment and also the future of nuclear energy. In this study, transmutation of minor actinides (MAs) discharged from LMFBR spent fuel in a high power density fusion reactor has been investigated under a neutron wall load of 10 MW/m2 for an operation period of 10 years. Also, the effect of MA percentage on the transmutation has been examined. The fuel zone, containing MAs as spheres cladded with W-5Re, has been located behind the first wall to utilize the high neutron flux for transmutation effectively. Helium at 40 atm has been used as an energy carrier. At the end of the operation period, the total burning and transmutation are greater than the total buildups in all investigated cases, and very high burnups (420–470 GWd/tHM) are reached, depending on the MA content. The total transmutation rate values are 906 and 979 kg/GWth year at startup and decrease to 140 and 178 kg/GWth year at the end of the operation for fuel with 10% and 20% MA, respectively. Over an operation period of 10 years, the effective half lives decrease from 2.38, 2.21 and 3.08 years to 1.95, 1.80 and 2.59 years for 237Np, 241Am and 243Am, respectively. Total atomic densities decrease exponentially during the operation period. The reductions in the total atomic densities with respect to the initial ones are 79%, 81%, 82%, 83%, 85% and 86% for 10%, 12%, 14%, 16%, 18% and 20% MAs, respectively.

04/00751 Thermophysical properties of inert matrix fuels for actinide transmutation: Ronchi, C. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 54–65

04/00751 Thermophysical properties of inert matrix fuels for actinide transmutation: Ronchi, C. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 54–65

Efficiency of preliminary transmutation of actinides before ultimate storage

The concept of preliminary transmutation of minor actinides before placement to the long-term storage is considered. The purpose of such preliminary transmutation before ultimate storage is to incinerate a part of actinides and to transform another part into new actinides providing low level of radiotoxicity accumulated in the storage. Modes of transmutation in reactors of PWR, CANDU, and Superphenix types are compared. Among power reactors, heavy-water PHWR (CANDU) type reactor is most acceptable for preliminary transmutation.

Measurement of Effective Capture Cross Section of 238Np for Thermal Neutrons

The effective capture cross section (σ) of 238Np for thermal neutrons has been measured as data for the burn-up analysis of irradiated fuels and for the study of nuclear transmutation of minor actinides. A sample of 237Np was irradiated for 10 hours at the Kyoto University Reactor. Neptunium-238 and -239 were produced simultaneously via the double neutron capture reactions 237Np(n, γ)238Np(n, γ)239Np. The neutron flux at the irradiation position was monitored using Au and Co wires. The epithermal index r√T/T 0 in Westcott's convention was deduced as 0.03. To measure the activities of 238Np and 239Np, a high-purity Ge detector with a BGO (Bi4Ge3O12) anti-coincidence shield has been developed. The σ of 238Np was deduced from the ratio of these activities. The result obtained is 479± 24 b for the σ of 238Np. The result is compared with data in the evaluated nuclear data libraries, ENDL-86, ENDF/B-VI, and JENDL-3.3.

Measurement of Effective Capture Cross Section of 238Np for Thermal Neutrons

The effective capture cross section (σ) of 238Np for thermal neutrons has been measured as data for the burn-up analysis of irradiated fuels and for the study of nuclear transmutation of minor actinides. A sample of 237Np was irradiated for 10 hours at the Kyoto University Reactor. Neptunium-238 and -239 were produced simultaneously via the double neutron capture reactions 237Np(n, γ)238Np(n, γ)239Np. The neutron flux at the irradiation position was monitored using Au and Co wires. The epithermal index r√T/T 0 in Westcott's convention was deduced as 0.03. To measure the activities of 238Np and 239Np, a high-purity Ge detector with a BGO (Bi4Ge3O12) anti-coincidence shield has been developed. The σ of 238Np was deduced from the ratio of these activities. The result obtained is 479± 24 b for the σ of 238Np. The result is compared with data in the evaluated nuclear data libraries, ENDL-86, ENDF/B-VI, and JENDL-3.3.

04/00206 JET first wall and divertor protection system: Cenedese, A. et al. Fusion Engineering and Design, 2003, 66–68, 785–790

A feasibility study on natural uranium spallation target in accelerator-driven system (ADS) for minor actinide (MA) transmutation was performed. As a result of comparative study of uranium and lead-bismuth (PbBi) targets in the bare case without blanket surrounding, it was found that uranium target had better neutron generation performance, but limited by the geometrical size due to high neutron absorption in 238U. In ADS for MA transmutation, uranium used as target instead of PbBi also absorbs neutrons passing the target area.More realistic concept of pin type uranium spallation target cooled by liquid PbBi was considered aiming at enhancing spallation target performance in terms of neutron generation efficiency and operation temperature. The uranium pin target design had nothing better effects on neutron balance of such system than a conventional PbBi target in ADS and it was concluded that uranium target was not suitable for the full-scale ADS.

The concept of the volumetric neutron source on the basis of the JUST-T tokamak for minor actinides transmutation

By concept development of the compact volumetric neutron source on the spherical tokamak JUST basis for minor actinides transmutation with aspect ratio A = 2, some key plasma physics problems are arising: start of discharge; plasma current maintenance in stationary stage; appropriate neutron fluence for transmutation. On the basis of accepted physical and technical preconditions of the concept the combined scenario of current start and ramp-up, its stationary maintenance due to bootstrap effect and drive by neutral particle injection are considered. The plasma current is proposed to be initiated inductively and then bootstrap current will generate with using additional heating by injection of neutral beams. Necessary neutron flux for effective transmutation (Γn ≈ 0.4 MW/m2) will be reached by both plasma-beam reactions and thermal plasma fusion reactions. By the way of preliminary consideration engineering studies of vacuum chamber, toroidal magnetic system, divertor, and blanket with energy multiplicatio...

A comparative physics study of alternative long-term strategies for closure of the nuclear fuel cycle

The appropriate management of radioactive waste arising from the nuclear fuel cycle is considered to be a key issue in the development of future, more sustainable nuclear energy systems. In this context, the partitioning and transmutation of actinides could play an important role through the achievement of very significant reductions in the actinide content and radiotoxicity of the high-level waste requiring geological disposal. The current paper reports on the results of a detailed physics study carried out to compare the pros and cons of alternative strategies for closure of the nuclear fuel cycle. Different long-term “steady-state” scenarios have been considered, involving the deployment, to varying degrees, of light water reactors (LWRs) and advanced fast-spectrum systems. The same nuclear data and calculation methods have been used throughout, so that a consistent and reliable comparison of the relative performance of the three basic fuel cycle options (once-through, plutonium recycle, and recycling of all actinides) has been made possible. In addition, with transmutation having been considered employing both critical and accelerator-driven fast-spectrum systems, the study has provided an evaluation of the advantages and disadvantages of these two different advanced system types.

Transmutation of Long-Lived Actinides in Power Reactors

Heavy- and light-water power reactors can be used for partial transmutation of nuclear wastes, thereby making it possible to limit the accumulation of long-lived radionuclides in storage sites for spent nuclear fuel to relatively low levels. During the operation of PHWR-880 in the self-service regime, the equilibrium radiotoxicity in long-term storage and the time when it is reached are 4–5 times less than in the analogous regime for a VVER-1000 reactor.

Inelastic neutron scattering study of the vibration frequencies of hydrogen in calcium dihydride

We report incoherent inelastic neutron scattering measurements on commercially available standard grade CaH 2 (∼95% pure). This compound is of interest because of a current research programme on the transmutation of long-lived nuclear wastes. CaH 2 is one possible material that could be used for localised moderation of the neutron spectrum in the PHENIX Fast Reactor, to optimise the core for the transmutation of higher actinides using dedicated moderated targets. The particular advantage of CaH 2 is that it is believed to be relatively stable in liquid sodium. For the case of ZrH 2, modelling existing experimental core data requires the use of the MCNP Monte Carlo code using the detailed IINS cross-sections for ZrH 2 from the ENDF-B/VI file. No such data exist for CaH 2, hence the need for the present experiment. The measurements used the IN1 Be-Filter instrument at the ILL going from 15 to 200 meV, at both 20 °C and 10 K. The data at 20 °C show evidence of acoustic modes at about 20 meV, a strong feature at 40 meV due to Ca(OH) 2 impurities and two strong bands between 70 and 100 meV and between 110 meV and 140 meV—each split into three distinguishable peaks. This is consistent with the crystal structure, which has two H sites, one a distorted octahedral site (lower band) and the other a distorted tetrahedral (upper).

Safety Analysis of Fast Reactor Core with Uranium-Free Fuel for Actinide Transmutation

The solving of ecological problems of future nuclear power is connected with the solving of long-lived radioactive waste utilization problems. This concerns primarily plutonium and minor actinides (neptunium, americium, and curium), accumulated in the spent fuel of nuclear reactors. One of the ways this can be solved is to use a fast reactor with uranium-free fuel. The physics of this type of reactor was widely investigated during the last year for the BN-800 reactor. The solutions of the most important problems were (a) a decrease in nonuniformity of the power distribution and (b) an increase of the Doppler effect. The next stage of such core investigations is an evaluation of self-protection to beyond-design accidents. Preliminary results show a high safety level of the BN-800 reactor with uranium-free fuel in unprotected loss-of-flow and unprotected transient overpower events.

Transmutation of transuranic actinides in a spherical torus tokamak fusion reactor

In this paper, a concept for a fusion-driven waste transmutation reactor (FDTR) based on a spherical torus (ST) is put forward. A set of plasma parameters suitable for the transmutation blanket was chosen. The two-dimensional neutron transport code TWODANT, the three-dimensional Monte Carlo code MCNP/4B, the one-dimensional neutron transport and burn-up calculation code BISON3.0 and their associated data libraries were used to calculate the transmutation rate, the energy multiplication factor and the tritium-breeding ratio of the transmutation blanket. The calculation results for the system parameters and the actinide series isotopes for different operation times are presented. The engineering feasibility of the centre-post of FDTR has been investigated and the results provided. A preliminary neutronics calculation based on an ST transmutation blanket shows that the proposed system has a high transmutation capability for minor actinide wastes.

The chemistry and physics of modelling nitride fuels for transmutation

Against the background of increased interest in nitride fuel for transmutation of minor actinides, this paper presents the current status of UK modelling of nitride fuel pins. A new equilibrium chemistry model for nitride fuel, suitable for inclusion in fuel pin modelling codes, is described. High-temperature experiments on (U,Zr)N in a sealed capsule indicate that actinide nitrides in a nitrogen atmosphere are stable up to their melting point. The paper also recommends a set of physical and chemical properties for advanced nitride fuels, highlighting the areas where data are presently missing.

Thermophysical properties of inert matrix fuels for actinide transmutation

The thermal transport properties of a set of ‘inert matrix fuels’ (IMFs) for actinide transmutation were investigated and compared. Heat capacity, thermal diffusivity and conductivity were measured in the usable temperature range for MgO- and ZrO 2-based IMFs containing uranium, plutonium and americium, and sintered from co-precipitated or blended oxide powders.

Post irradiation examination of irradiated americium oxide and uranium dioxide in magnesium aluminate spinel

To study MgAl 2O 4 spinel as inert matrix material for the transmutation of minor actinides, two capsules were irradiated at the high flux reactor in Petten, containing 12.5 wt% micro-dispersed 241AmO x in spinel and 25 wt% micro-dispersed enriched UO 2 in spinel. During irradiation, the initially present 241Am was converted for 99.8% to fission products (50%), plutonium (30%), curium (16%) and 243Am (4%). The UO 2 spinel target experienced a burn-up of 32% fission per initial metal atom. The post irradiation examination of the AmO x inert matrix target showed swelling of 27 vol.%, and a gas release of 48% for He and 16% for Xe and Kr. The UO 2 inert matrix target also showed a large volumetric swelling of 11%, directed mainly radially. Ceramography on the UO 2 inert matrix target revealed a complete restructuring of the spinel grains upon irradiation and the absence of porosity, suggesting that amorphisation is the main cause of the swelling.

Fabrication of nitride fuels for transmutation of minor actinides

At the Japan Atomic Energy Research Institute, the concept of the transmutation of minor actinides (MA: Np, Am and Cm) with accelerator-driven systems is being studied. The MA nitride fuel has been chosen as a candidate because of the possible mutual solubility among the actinide mononitrides and excellent thermal properties besides supporting hard neutron spectrum. MA nitrides of NpN, (Np, Pu)N, (Np, U)N, AmN, (Am, Y)N, (Am, Zr)N and (Cm, Pu)N were prepared from the oxides by the carbothermic reduction method. The prepared MA nitrides were examined by X-ray diffraction and the contents of impurities of oxygen and carbon were measured. The fabrication conditions for MA nitrides were improved so as to reduce the impurity contents. For an irradiation test of U-free nitride fuels, pellets of (Pu, Zr)N and PuN + TiN were prepared and a He-bonded fuel pin was fabricated. The irradiation test started in May 2002 and will go on for two years in the Japan Materials Testing Reactor.