Transmutation Investigation of a Typical VVER-1000 Reactor Burnup Products to less Toxicity Isotopes in a Fusion-Fission Hybrid Reactor (original) (raw)
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Transmutation considerations of LWR and RBMK spent nuclear fuel by the fusion–fission hybrid system
Nuclear Engineering and Design, 2018
The performance of the fusion-fission hybrid system based on the molten salt (flibe) blanket, driven by a plasma based fusion device, was analyzed by comparing transmutation scenarios of actinides extracted from the LWR (Sweden) and RBMK (Lithuania) spent nuclear fuel in the scope of the EURATOM project BRILLIANT. The IAEA nuclear fuel cycle simulation system (NFCSS) has been applied for the estimation of the approximate amount of heavy metals of the spent nuclear fuel in Sweden reactors and the SCALE 6 code package has been used for the determination of the RBMK-1500 spent nuclear fuel composition. The total amount of trans-uranium elements has been estimated in both countries by 2015. Major parameters of the hybrid system performance (e.g., k scr , k eff , Φ n (E), equilibrium conditions, etc.) have been investigated for LWR and RBMK trans-uranium transmutation cases. Detailed burn-up calculations with continuous feeding to replenish the incinerated trans-uranium material and partial treatment of fission products were done using the Monteburns (MCNP + ORIGEN) code system. About 1.1 tons of spent fuel trans-uranium elements could be burned annually with an output of the 3 GW th fission power, but the equilibrium stage is reached differently depending on the initial trans-uranium composition. The radiotoxicity of the remaining LWR and RBMK transmuted waste after the hybrid system operation time has been estimated.
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.
Burn-up and Neutronic Analysis of VVER-1000 Nuclear Reactor
In this paper, burn-up analysis of the Bushehr Nuclear Power Plant (BNPP), VVER-1000 , performed by ORIGEN2 code is presented. Neutronic calculation at nominal conditions is carried out during its first operational cycle using the coupled ORIGEN2-MCNP4C. Modeling of all rods, channels, FAs and as a result, the core, is carried out using the MCNP4C code. By modeling the core, the new fuel and material composition obtained from ORIGIN-2, are fed into MCNP4C data card to determine criticality and power distribution. The calculated fuel inventory, criticality and fuel assembly power distribution as a function of time are comparable with the plant's PSAR.
Some safety studies for conceptual fusion--fission hybrid reactors. Final report
1978
EPRI Project Manager Noel A. Amherd Fossil Fuel and Advanced Systems Division liioTRICUTIGN OF THIS iJ0CU:viE21T ^S UN.LJ^l'TJ-^D LEGAL NOTICE This report was prepared by the University of California at Los Angeles (UCLA) as an account of work sponsored by the Electric Power Research Institute, Inc. (EPRI). Neither EPRI, members of EPRI, UCLA, nor any person acting on behalf of either (a) makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed m this report may not infringe privately owned rights, or (b) assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.
LWR spent fuel transmutation in a high power density fusion reactor
Annals of Nuclear Energy, 2004
The prospect of light water reactor (LWR) spent fuel incineration in a high power density fusion reactor has been investigated. The neutron wall load is taken at 10 MW/m 2 and a refractory alloy (W-5Re) is used in the first wall. Neutron transport calculations are conducted over an operation period of 48 months on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S 8-P 3 approximation. In the neutron rich environment, the tritium breeding ratio remains > 1.05 so that the tritium self-sufficiency is maintained for the fusion reactor. The presence of fissionable nuclear waste fuel in the investigated blanket causes significant energy amplification. The energy multiplication factor is 4atstartupanditincreasessteadilyupto5.55duringpowerplantoperationsothatevenamodestfusionreactorcansupplyasignificantquantityofelectricity.Inthecourseofnuclearwasteincineration,mostofthefissionablefuelisburntinsitu.Inadditiontothat,excessfissilefuelproductionenhancesthenuclearqualityofthenuclearfuel.Startingwithaninitialcumulativefissilefuelenrichment(CFFE)valueofthespentfuelof2.1724 at startup and it increases steadily up to 5.55 during power plant operation so that even a modest fusion reactor can supply a significant quantity of electricity. In the course of nuclear waste incineration, most of the fissionable fuel is burnt in situ. In addition to that, excess fissile fuel production enhances the nuclear quality of the nuclear fuel. Starting with an initial cumulative fissile fuel enrichment (CFFE) value of the spent fuel of 2.172%, CFFE can reach 4% after an irradiation period of 4atstartupanditincreasessteadilyupto5.55duringpowerplantoperationsothatevenamodestfusionreactorcansupplyasignificantquantityofelectricity.Inthecourseofnuclearwasteincineration,mostofthefissionablefuelisburntinsitu.Inadditiontothat,excessfissilefuelproductionenhancesthenuclearqualityofthenuclearfuel.Startingwithaninitialcumulativefissilefuelenrichment(CFFE)valueofthespentfuelof2.17212 months. Then the spent fuel becomes suitable for a new recharge in an LWR as a regenerated fuel. Further residence in the fusion blanket continues to upgrade the nuclear waste so that after 48 months, CFFE can reach such a high level (9%) that it becomes qualified to be used in a new type of the advanced high temperature reactors for the Generation-IV.
On the safety of conceptual fusion-fission hybrid reactors
Nuclear Engineering and Design, 1979
A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (e) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundant decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered.
Applied Radiation and Isotopes, 1995
A deep geological repository for safe long-term storage of long-lived radioactive materials (waste) arising from nuclear fuel irradiation in reactors is a need generally accepted, whatever the strategy envisaged for further use of the irradiated fuel (e.g. : reprocessing and re-use of uranium and plutonium ; no reprocessing and final disposal).
Reprocessing Free Nuclear Fuel Production Fusion Fission Hybrids in Action
2011
Fusion fission hybrid, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. The principle and detailed workings of a totally reprocessing-free (ReFree) Th-U conversion cycle are presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional water-cooled reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U fuel does not throttl...
Annals of Nuclear Energy, 2011
The Georgia Institute of Technology has developed several design concepts of tokamak based fusion-fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a "fusion transmutation of waste reactor". It was designed for a criticality k eff <0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR cycle =1.9 and a first wall lifetime of 12-16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead-bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k eff <0.97 and produces less fission power of 1.5 GW th. Its main performance parameters turn out substantially better.