Target fuels for plutonium and minor actinide transmutation in pressurized water reactors (original) (raw)
Related papers
Use of Thorium for Transmutation of Plutonium and Minor Actinides in PWRs
The objective of this work was to assess the potential of thorium based fuel to minimise Pu and MA production in Pressurised Water Reactors (PWRs). The assessment was carried out by examining destruction rates and residual amounts of Pu and MA in the fuel used for transmutation. 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 burn-up calculations were performed using CASMO4 -the fuel assembly burn-up code. The results indicate that up to 1 000 kg of reactor grade Pu can potentially be burned in thorium based fuel assemblies per GW e Year. Up to 75% of initial Pu can be destroyed per path. Addition of MA to the fuel mixture degrades the burning efficiency. The theoretically achievable limit for total TRU destruction per path is 50%. Efficient MA and Pu destruction in thorium 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 fuel with negative Doppler and moderator temperature coefficients.
Characterization of metallic fuel for minor actinides transmutation in fast reactor
Progress in Nuclear Energy, 2016
The METAPHIX programme is a collaboration between the Central Research Institute of Electric Power Industry (CRIEPI, Japan) and the Joint Research Centre-Institute for Transuranium Elements (JRC-ITU) of the European Commission dedicated to investigate the safety and effectiveness of a closed nuclear fuel cycle based on Minor Actinides (MA: Np, Am, Cm) separation from spent fuel, incorporation in metal alloy fuel and transmutation in fast reactor. Nine Na-bonded experimental pins of metal alloy fuel were prepared at ITU and irradiated at the Phenix reactor (CEA, France) achieving 2.5 at.%, 7 at.% and 10 at.% burn-up. Four metal alloy compositions were irradiated: U-Pu-Zr used as fuel reference, U-Pu-Zr þ 5 wt.% MA, U-Pu-Zr þ 2 wt.% MA þ 2 wt.% Rare Earths (RE: Nd, Y, Ce, Gd), and þ5 wt.% MA þ 5 wt.% RE, respectively. RE reproduce the expected output of a pyrometallurgical reprocessing facility. Post Irradiation Examination is performed using several techniques, covering properties ranging from the macroscopic morphology of the fuel matrix to the microanalysis of phases and elemental redistribution/segregation. The irradiated fuel is characterized by many phases occurring along the fuel radius. The fuel underwent large redistribution of the fuel constituents (U, Pu, Zr) and many secondary phases are present with a variety of compositions. The distribution of phases in the irradiated fuel containing minor actinides and rare earths is essentially similar to that observed in the basic ternary alloy fuel.
Nuclear Energy and Technology, 2022
In terms of nuclear raw materials, the issue of involving thorium in the fuel cycle is hardly very relevant. However, in view of the large-scale nuclear power development, the use of thorium seems to be quite natural and reasonable. The substitution of traditional uranium-plutonium fuel for uranium-thorium fuel in fast neutron reactors will significantly reduce the production of minor actinides, which will make it attractive for the transmutation of long-lived radioactive isotopes of americium, curium and neptunium that have already been and are still being accumulated. Due to the absence of uranium-233 in nature, the use of thorium in the nuclear power industry requires a closed fuel cycle. At the initial stage of developing the uranium-thorium cycle, it is proposed to use uranium-235 instead of uranium-233 as nuclear fuel. Studies have been carried out on the transmutation of minor actinides in a fast neutron reactor in which the uranium-thorium cycle is implemented. Several optio...
Multi-group analysis of Minor Actinides transmutation in a Fusion Hybrid Reactor
EPJ Nuclear Science and Technology, 2023
New nuclear technologies are currently being study to face High Level Waste treatment and disposal issues. Generally, GEN-IV fission Fast Reactors (FR) are considered the waste-burners of the future. In fact, a fast flux turns out to be the best choice for actinides irradiation in critical reactors because of favorable cross section conditions. Differently, Fusion Fission Hybrid Reactors (FFHR) are futuristic devices based on the combination of fusion and fission systems and could represent an alternative to FRs. In such systems, the choice spectrum of the neutron flux that irradiates HLW may be non-obvious due to some operational constraints which have to be considered. To design and optimize these systems as waste-burners, one should fully understand the transmutation dynamics occurring into the fission region. A multi-energy-group analysis by FISPACT-II code has been set to analyze the conversion processes in scenarios characterized by different neutron energy spectra and fluences. The results of this study show that, despite fast fluxes are characterized by better behaviors in terms of radiotoxicity treatment, the difficulties of reaching high reaction yields may require solutions involving moderators or broadened neutron fluxes to increase the reactions probabilities and, consequently, actinides mass conversion yield.
LWR spent fuel transmutation with fusion-fission hybrid reactors
Progress in Nuclear Energy, 2013
In this paper the transmutation of light water reactors (LWR) spent fuel is analyzed. The system used for this study is the fusion-fission transmutation system (FFTS). It uses a high energy neutron source produced with deuterium-tritium fusion reactions, located in the center of the system, which is surrounded by a fission region composed of nuclear fuel where the fissions take place. In this study, the fuel of the fission region is obtained from the recycling of LWR spent fuel. The MCNPX Monte Carlo code was used to setup a model of the FFTS. Two fuel types were analyzed for the fissile region: the mixed oxide fuel (MOX), and the inert matrix fuel (IMF). Results show that in the case of the MOX fuel, an important Pu-239 breeding is achieved, which can be interesting from the point of view of maximal uranium utilization. On the contrary, in the case of the IMF fuel, high consumption of Pu-239 and Pu-241 is observed, which can be interesting from the point of view of non-proliferation issues. A combination of MOX and IMF fuels was also studied, which shows that the equilibrium of actinides production and consumption can be achieved. These results demonstrate the versatility of the fusion-fission hybrid systems for the transmutation of LWR spent fuel.
2019
The Russian Federation is developing a number of technologies within the «Proryv» project for closing the nuclear fuel cycle utilizing mixed (U-PuMA) nitride fuel. Key objectives of the project include improving fast reactor nuclear safety by minimizing reactivity changes during fuel operating period and improving radiological and environmental fuel cycle safety through Pu multi-recycling and MA transmutation. This advanced technology is expected to allow operating the reactor in an equilibrium cycle with a breeding ratio equaling approximately 1 with stable reactivity and fuel isotopic composition. Nevertheless, to reach this state the reactor must still operate in an initial transient state for a lengthy period (over 10 years) of time, which requires implementing special measures concerning reactivity control. The results obtained from calculations show the possibility of achieving a synergetic effect from combining two objectives. Using MA reprocessed from thermal reactor spent fuel in initial fuel loads in FR ensures a minimal reactivity margin during the entire fast reactor fuel operating period, comparable to the levels achieved in equilibrium state with any kind of relevant Pu isotopic composition. This should be combined with using reactivity compensators in the first fuel micro-campaigns. In the paper presented are the results of simulation of the overall life cycle of a 1200 MWe fast reactor, reaching equilibrium fuel composition, and respective changes in spent fuel nuclide and isotopic composition. It is shown that MA from thermal and fast reactors spent fuel can be completely utilized in the new generation FRs without using special actinide burners.
The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6th Framework Program, is mainly aimed at providing additional key elements for the utilisation and transmutation of plutonium and minor actinides in contemporary and future (high temperature) gas-cooled reactor design, which are promising tools for improving the sustainability of the nuclear fuel cycle, hereby also contributing to the reduction of Pu and MA stockpiles, and to the development of safe and sustainable reactors for CO2-free energy generation. The project runs from
Evaluation on transmutation performance of minor actinides with high-flux BWR
Annals of Nuclear Energy, 2001
The performance of high-¯ux BWR (HFBWR) for burning and/or transmutation (B/T) treatment of minor actinides (MA) and long-lived ®ssion products (LLFP) was discussed herein for estimating an advanced waste disposal with partitioning and transmutation (P&T). The concept of high-¯ux B/T reactor was based on a current 33 GWt-BWR, to transmute the mass of long-lived transuranium (TRU) to short-lived ®ssion products (SLFP). The nuclide selected for B/T treatment was MA (Np-237, Am-241, and Am-243) included in the discharged fuel of LWR. The performance of B/T treatment of MA was evaluated by a new function, i.e. [F/T ratio], de®ned by the ratio of the ®ssion rate to the transmutation rate in the core, at an arbitrary burn-up, due to all MA nuclides. According to the results, HFBWR could burn and/or transmute MA nuclides with higher ®ssion rate than BWR, but the ®ssion rate did not increase proportionally to the¯ux increment, due to the higher rate of neutron adsorption. The higher B/T fraction of MA would result in the higher B/T capacity, and will reduce the units of HFBWR needed for the treatment of a constant mass of MA. In addition, HFBWR had a merit of higher mass transmutation compared to the reference BWR, under the same mass loading of MA.
Malaysian Journal of Fundamental and Applied Sciences
The evaluation of RSG-GAS research reactor for transmutation reactor was proposed to study its effectiveness to transmute minor actinides (MA), specifically Am-241, to support geologic storage/disposal. The Am-241 radionuclide was assumed to be discharged from 1000MWe PWR’s spent fuel. The mass of Am-241 discharged from within a year operation of 1000MWe PWR was 1.65E+03 gram, while the optimum Am-241 mass which can be transmuted in RSG-GAS - and still meet the safety requirements of reactivity - was 8.0E+03 gram. This was equivalent to about cumulative Am-241 discharged from 5 units of 1000MWe PWR. In 10 cycles of RSG-GAS operation (about 2 years), the remaining of Am-241 is only about 100 grams. The ratio of Am-241 transmuted (8.0E+03 gram) and Am-241 produced in the RSG-GAS core (1.98E-02 gram) within 1-year operation shows the effectiveness of RSG-GAS as a transmutation reactor.
Nuclear Engineering and Design, 2011
A study was conducted to evaluate the feasibility of minor actinide (MA) transmutation in light water reactors (LWR). The purpose of this work was to provide a guide for future investigations into MA transmutation in LWR. This work considered the effects of various Am/Cm separation efficiencies as well as homogeneous and heterogeneous MA bearing fuel assemblies. The MA content was introduced into the reactor as mixed oxide plus minor actinide (MOX + MA) fuel. Three Am/Cm separation efficiencies were independently considered: 99.9%, 99.0%, and 90.0%. In order to evaluate the feasibility of MA transmutation, the fuel performance of the various assemblies and core designs, as well as their respective safety related parameters, were calculated. The reduction of the burden of high level waste (HLW) motivated the investigation of MA transmutation. It was found that the MA bearing fuel assemblies and their subsequent core designs were able to perform within the safety limits required as well as achieving similar burnups to a UO 2 core. The Am transmutation rates were ∼40% for the homogeneous assemblies and up to 68% for the MA targets in the heterogeneous assemblies after the described burnup, however, there was a significant amount of Cm produced during burnup. This Cm production was due to the more favorable neutron capture reaction over fission for Am in the thermal spectrum. Future work should examine the benefits of Am transmutation at the expense of large Cm production rates.