Transmutation Of Actinides Research Articles

Scenarios for the transmutation of actinides in CANDU reactors

With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past ( Boczar et al., 1996; Chan et al., 1997; Hyland and Dyck, 2007; Hyland et al., 2009). The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100–1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2− x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of “best estimates” provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

Fissile fuel breeding and minor actinide transmutation in the life engine

Progress on The National Ignition Facility (NIF) brings fusion a viable energy source in foreseeable future. Energy multiplication in a fusion–fission (hybrid) reactor could lead earlier market penetration of fusion energy for commercial utilization. Originally, scientists at the Lawrence Livermore National Laboratory (LLNL) have worked out a hybrid reactor design concept; the so-called Laser Inertial Confinement Fusion–Fission Energy (LIFE) engine, which has consisted of a spherical fusion chamber of ∼5 m diameter, surrounded by a multi-layered blanket with a beryllium multiplier zone after the first wall. However, earlier work had indicated extreme power peaks at immediate vicinity of the first wall of a hybrid assembly, if a beryllium multiplier is used. Hence, in the current work, the beryllium multiplier zone has been removed in order to mitigate fission power peaks at the vicinity of the first wall as a result of neutron moderation on beryllium. Furthermore, minor actinides (MA) will cause significant neutron multiplication under fusion neutron irradiation so that an extra beryllium multiplier will not be needed. Present work has made following modifications on the LLNL design of the original (LIFE) engine: • Omission of beryllium multiplier. • TRISO fuel has been suspended as micro-size particles in Flibe coolant in lieu of being dissolved in uranium salt or imbedded carbon matrix in macro-size pebbles. • Carbide fuel is used. • Fissionable fuel charge is kept lower than in the LLNL (LIFE) engine. The modified (LIFE) engine is kept similar to the LLNL design to a great degree in order to allow mutual feedback between two geographically separated teams towards a more advanced and improved design under consideration of totally independent views. The first wall is made of ODS (2 cm) and followed by a Li 17Pb 83 zone (2 cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2 cm) from the Flibe molten salt zone (50 cm), containing MA as fissionable fuel. A 3rd ODS layer (2 cm) separates the molten salt zone on the right side from the graphite reflector (30 cm). Calculations have been conducted for a fusion driver power of 500 MW th in S 8-P 3 approximation using 238-neutron groups. Minor actinides (MA) out of the nuclear waste of LWRs are used as fissile carbide fuel in TRISO particles with volume fractions of 0, 2, 3, 4 and 5% have been dispersed homogenously in the Flibe coolant. For these cases, tritium breeding at startup is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. In the course of plant operation, TBR and fissile neutron multiplication factor decrease gradually. For a self-sustained reactor, TBR > 1.05 can be kept for all cases over 8 years. Higher fissionable fuel content in the molten salt leads also to higher blanket energy multiplication, namely M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5% TRISO volume fraction at start up, respectively. For all investigated cases, fissile burn up exceeds 400 000 MW D/MT. Major damage mechanisms have been calculated as DPA = 50 and He = 176 appm per year. This implies a replacement of the first wall every 3 years.

Vaporisation of candidate nuclear fuels and targets for transmutation of minor actinides

The thermal stability and high temperature behaviour of candidate fuels and targets for transmutation of minor actinides has been investigated. Zirconia-based solid solution, MgO-based CERCER and molybdenum-based CERMET fuels containing Am and/or Pu in various concentrations were heated up to 2700 K in a Knudsen cell coupled with a quadrupole mass spectrometer, to measure their vapour pressure and vapour composition. The results reveal that the vaporisation of the actinides from the samples is not only determined by the thermodynamics of the system but is also related to the dynamic evolution of multi-component mixtures with complex composition or microstructure.

State of the Art of Pyroprocessing Technology in Japan

Abstract Minor actinide recycling in a fuel cycle is a potential technology to minimize the environmental radioactivity burden in waste disposal in the fast reactor era after 2050. Pyroprocessing technology with metal electrorefining expects that no additional process is required to separate minor actinides and short- cooled fuels can be treated due to no-use of organic solvent that degrades by radiation. Pyroprocessing has been explored for metal fuel cycle and nitride fuel cycle in Japan. Metal fuel fast reactor, which can achieve a high breeding ratio over 1.3, and its fuel cycle is a compact system by integrating pyroprocessing. Oxide fuel can be also treated by converting to metals by reduction. Separation of transuranium elements from high level liquid waste originating from aqueous reprocessing has been challenged. Verification of the process and development of an engineering scale device are the current interests for study. In addition, irradiation study of metal fuel with minor actinides currently much advances from the point of fundamental investigation. Accelerator-driven system (ADS) for transmutation of minor actinides is integrated with pyroprocessing for recycling system. The denitriding at anode and azotizing at cathode together with electrorefining have been fundamentally studied by use of plutonium.

The GDT Based Neutron Source as a Driver in a Sub-Critical Burner of Radioactive Wastes

Transmutation of long-lived radioactive nuclear waste, including plutonium, minor actinides and fission products, represents a highly important problem of fission reactor technology and is presently studied worldwide in large-scale. Sub-critical systems seem to be a promising option for efficiently burning plutonium and minor actinides provided a sufficiently high-intense neutron source is available. For a number of years the Budker Institute of Nuclear Physics (Russia) in collaboration with the Russian and European organizations developed the project of a 14 MeV neutron source for fusion material irradiation and other applications. The projected plasma type neutron source is based on the Gas Dynamic Trap (GDT) which is a special magnetic mirror system for the plasma confinement. This poster presents different version of the GDT-based neutron source for hybrid fusion-fission sub-critical system for the transmutation of the long-live radioactive waste in spent nuclear fuel.

ADS Fuel Developments in Europe: Results from the EUROTRANS Integrated Project.

Fuels to be used in Accelerator Driven Systems dedicated to Minor Actinides transmutation can be described as highly innovative in comparison with those used in critical cores. Indeed, ADS fuels are not fertile, so as to improve the transmutation performance and they contain high volumetric concentrations (~50%) of minor actinides and plutonium compounds. This unusual fuel composition results in high gamma and neutron emissions during its fabrication, as well as degraded performances under irradiation. Ceramic-Ceramic and Ceramic Metallic composite fuels consisting of particles of (Pu,MA)O2 phases dispersed in a magnesia or molybdenum matrix were investigated within the European Research programme for Transmutation, as driver fuels for a prospective 400MWth transmuter: the European Facility for Industrial Transmutation. Fuel performances and safety of preliminary core designs were evaluated to support the project. Out -of-pile as well as in-pile experiments were carried out to gain essential knowledge on properties and behaviour under irradiation of these types of fuel. This paper gives an overview of experimental results within the project. © 2010 Published by Elsevier Ltd

Minor Actinide Bearing Fuels: Fabrication and Irradiation Experience in Europe

Minor actinides constitute a significant contribution to the long term radiotoxicity of spent fuel. Their separation from spent fuel, fabrication and irradiation in homogeneous or heterogeneous reactor recycling modes provides a means to dramatically reduce their impact. Several challenges must be overcome in the separation and fabrication steps. Advanced robust fabrication processes are required to minimise dust and to produce fuels specifically designed to optimise their destruction through neutron irradiation. At the Institute for Transuranium Elements fabrication routes based on sol gel and infiltration have been developed to meet these demands. Dedicated fuels and targets have been produced and their irradiation behaviour is being tested. Fuels based on MOX and UO2 are being tested with respect to Gen IV fast reactor initiatives. Furthermore, dedicated minor actinide transmutation fuels based on the inert matrix concept (e.g. (Zr,Y,Am)O2 and Mo based (Pu,Am)O2 composite fuels) have been produced for application in the double strata concept, whereby the minor actinides should be transmuted in an ADS. Achievements and current status of these programmesare presented and discussed in this paper.

Predicting thermo-mechanical behaviour of high minor actinide content composite oxide fuel in a dedicated transmutation facility

The European Facility for Industrial Transmutation (EFIT) of the minor actinides (MA), from LWR spent fuel is being developed in the integrated project EUROTRANS within the 6th Framework Program of EURATOM. Two composite uranium-free fuel systems, containing a large fraction of MA, are proposed as the main candidates: a CERCER with magnesia matrix hosting (Pu,MA)O 2− x particles, and a CERMET with metallic molybdenum matrix. The long-term thermal and mechanical behaviour of the fuel under the expected EFIT operating conditions is one of the critical issues in the core design. To make a reliable prediction of long-term thermo-mechanical behaviour of the hottest fuel rods in the lead-cooled version of EFIT with thermal power of 400 MW, different fuel performance codes have been used. This study describes the main results of modelling the thermo-mechanical behaviour of the hottest CERCER fuel rods with the fuel performance code MACROS which indicate that the CERCER fuel residence time can safely reach at least 4–5 effective full power years.

Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2− x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW −1 h −1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

Impact of Partitioning and Transmutation on High-Level Waste Disposal for the Fast Breeder Reactor Fuel Cycle

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.

Neutronic Evaluation of a MHR System to Transmutation of Minor Actinides

The goal is to simulate the modular helium reactor (MHR) core to analyze the neutronic parameters behavior due the insertion of Pu isotopes and minor actinides (MAs) using shuffling scheme without compromising the safety parameters. Initially the core is filled with driver fuel (DF). After the burn-up, these fuels are then reprocessed and used to produce the transmutation fuel (TF). Some cycles after, the core is filled with DF and TF fuels. DF fuel is composed of Pu and Np while TF fuel is a mixture of Pu and MAs. The shuffling scheme was evaluated after each cycle. It was verified that neutronic parameters and isotopic composition reach equilibrium when this scheme is used. The WIMS code was used to perform the simulations and the following neutronic parameters were evaluated: infinite multiplication factor, spectrum hardening, and reactivity temperature coefficients.

Thermodynamics of Dissolution for Bis(triazine)−Bipyridine-Class Ligands in Different Diluents and Its Reflection on Extraction

Hydrochemical separation processes are one of the methods used for the treatment of spent nuclear fuel. Solvent extraction is also used in many other non-nuclear applications like the mining industry. In the nuclear case, hydrochemical separation processes are already employed in the world today for the recovery of uranium and plutonium. The method is however also considered for future separation systems for use in combination with the transmutation of the minor actinides. In a hydrochemical separation process the two phases are the pregnant (usually) aqueous feed and the organic phase comprising a diluent together with one or more extractants. One such class of extractants developed for partitioning and transmutation purposes is the bis(triazine)−bipyridine-type (BTBP) molecules. When assessing the feasibility and loading properties of such an extraction system, the solubility of the ligands is of the outmost importance. The understanding of whether the dissolution is enthalpically or entropically driven will also help the understanding of the differences in extraction observed between various diluents and temperatures. In this paper the enthalpy and entropy of dissolution of the BTBP-class ligands have been determined for different diluents. It has also been shown that it is possible to predict the extraction behavior of these molecules in the selected diluent once the solubility is known.

Hydrogen production potential of APEX fusion transmuter fueled minor actinide fluoride

Main aim of this study is to investigate hydrogen production potential of Advanced Power EXtraction (APEX) fusion reactor cooled with the molten-salt mixtures, as well as its neutronic performance to transmute minor actinides (MAs). In the original APEX reactor concept, fusion power ( P f) is quite high (4000 MW), and the FLiBe molten-salt flows as molten-salt wall. The FLiBe molten-salt is mixed with molten minor actinide tetra fluoride salt (MAF 4) to transmute minor actinides, and at the same time, to increase the energy multiplication. In addition to this mixture of FLiBe and MAF 4, FLiNaBe, LiF and Eutectic Lithium instead of FLiBe are mixed individually with MAF 4, and are used as the molten-salt coolant. Furthermore, two different compositions of MA nuclides are considered as follows: (i) The MA nuclides discharged from the pressured water reactor (PWR)-MOX spent fuel and (ii) The MA nuclides discharged from the PWR-UO 2 spent fuel. The neutronic analyses have been performed for these eight different molten-salt mixture cases and for both one and three-dimensional geometry models by using the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B code, respectively. In order to produce hydrogen in large-scale, Steam-Methane Reforming (SMR) combined with the Mineral CO 2 Sequestration (MCS) is selected. Furthermore, the sulfur–iodine (S–I) thermochemical water splitting and high-temperature electrolysis (HTE) cycles, which are the most promising water-splitting cycles, are analyzed. The results of calculations bring out that the considered fusion reactor has a good neutronic performance, and it can produce in a considerable amount of the hydrogen production (up to 426 kg/s), as well as the minor actinide transmutation (up to 4.849 t/yr).

Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity—Irradiation Tests of Np and Np-U Samples in the Experimental Fast Reactor JOYO (JAEA) and the Advanced Test Reactor at INL—

A project on Protected Plutonium Production (P3) was proposed by Tokyo Institute of Technology as part of a nonproliferation research program for plutonium (Pu) utilization in nuclear reactors. The project is aimed at the production of inherently protected Pu by the addition of 237Np to uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. In the first, a determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor JOYO was done to evaluate 238Pu production from 237Np under fast neutron spectra. The fast reactor can undertake P3, which can be better performed in the reflector region. In the test, the percentage of the total amount of 238Pu atoms transmuted from 237Np atoms by irradiation was around 90%. In the second test, 2, 5, and 10% Np-containing U samples were irradiated in the Advanced Test Reactor at the Idaho National Laboratory (INL) to evaluate 238Pu production in the thermal neutron region. The fuel specimens were removed from the core at 100, 200, and 300 effective full power days (EFPDs), and then a postirradiation examination was completed at an analytical laboratory in the Materials & Fuels Complex (MFC) at the INL. For the samples after irradiation for 300 EFPDs, Np depletions were about 60% for 2% neptunium (Np)-U samples and about 50% for 5 and 10% Np-U samples. The 238Pu-to-Pu ratios were about 20, 30, and 45% for 2, 5, and 10% Np-U samples, respectively.

The role sol–gel process for nuclear fuels-an overview

The paper reviews the sol–gel methods used for the preparation of nuclear fuel materials in the form of microspheres. It also discusses how these microspheres can be fabricated into nuclear fuels for reactors such as High Temperature Gas Cooled Reactors and Fast Reactors. The performance of these microsphere-based fuels is reviewed. More recent applications, such as the transmutation of minor actinides, (Np, Am and Cm) and hydrogen production, are also briefly covered.

Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

Abstract Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R&D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic–ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic–metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg HM discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW (th) loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life. Presented results were used for testing newly-developed models installed in the TRANSURANUS code to deal with such innovative fuels and T91 steel cladding. Agreement among codes predictions was satisfactory for fuel and cladding temperatures, pellet-cladding gap and mechanical stresses.

Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity — Irradiation Tests of Np and Np-U Samples in the Experimental Fast Reactor JOYO (JAEA) and the Advanced Test Reactor at INL —

A project on Protected Plutonium Production (P3) was proposed by Tokyo Institute of Technology as part of a nonproliferation research program for plutonium (Pu) utilization in nuclear reactors. The project is aimed at the production of inherently protected Pu by the addition of 237Np to uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. In the first, a determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor JOYO was done to evaluate 238Pu production from 237Np under fast neutron spectra. The fast reactor can undertake P3, which can be better performed in the reflector region. In the test, the percentage of the total amount of 238Pu atoms transmuted from 237Np atoms by irradiation was around 90%. In the second test, 2, 5, and 10% Np-containing U samples were irradiated in the Advanced Test Reactor at the Idaho National Laboratory (INL) to evaluate 238Pu production in the thermal neutron region. The fuel specimens were removed from the core at 100, 200, and 300 effective full power days (EFPDs), and then a postirradiation examination was completed at an analytical laboratory in the Materials & Fuels Complex (MFC) at the INL. For the samples after irradiation for 300 EFPDs, Np depletions were about 60% for 2% neptunium (Np)-U samples and about 50% for 5 and 10% Np-U samples. The 238Pu-to-Pu ratios were about 20, 30, and 45% for 2, 5, and 10% Np-U samples, respectively.

Neutron-induced capture cross sections via the surrogate reaction method

The surrogate reaction method is an indirect way of determining cross sections for nuclear reactions that proceed through a compound nucleus. In this method, the compound nucleus is produced via an alternative (surrogate) reaction and its decay (by fission, gamma or neutron emission) is measured in coincidence with the outgoing appropriate charged particle. This technique has enabled neutron-induced cross sections to be extracted for nuclear reactions on short-lived nuclei that otherwise could not be measured. The CENBG collaboration has successfully applied this technique to determine the neutron-induced fission cross sections of several short-lived nuclei such as 233Pa, 242,243Cm and 241Am. These data are very important for the development of the Th/U cycle and for minor actinide transmutation. We currently investigate whether this powerful technique can also be used to determine the neutron-induced capture cross sections. For this purpose we will use the surrogate reaction 174Yb(3He,pγ)176Lu to infer the well known 175Lu(n,γ) cross section and compare the results with the directly measured neutron-induced data. The experimental set-up and the first results will be presented. We will also discuss our future plans to use the surrogate method for extracting actinides (n,γ) cross sections.

Applicability of dynamic programming to the accelerator-driven system (ADS) fuel cycle shuffling scheme for minor actinide (MA) transmutation

The accelerator-driven system (ADS) is under development in several countries to reduce the burden for conditioning and disposal of the high-level radioactive waste (HLW) by transmuting minor actinide (MA). It is expected that the fuel shuffling can make the power distribution flat and transmute MA effectively by using only one kind of fuel composition. But the total number of calculation cases becomes huge for finding the globally optimum case. In order to find the best shuffling scheme for MA transmutation by ADS, we attempt to develop a calculation code within an acceptable time by employing dynamic programming. It is used successfully for a cylindrical core with three fuel regions with 20 times of fuel shuffling.