Tenkte dere kanskje kunne være litt interessert (ihvertfall noen av dere?) i hva jeg egentlig
holder på med på Blindern;) Så special for you: om en mulighet med thorium i kjernekraftverk - på en måte slik at vi ikke trenger å utvikle en ny reaktor (som koster MYE, og er flere tiår inn i fremtiden, dessverre).
Bare for å understreke: jeg mener ikke at vi ikke skal forske på å utvikle ny reaktorteknologi, altså, men jeg syns det er spennende å se hva vi kan få til per i dag, med den teknologi som vi faktisk har. På samme måte som at jeg ikke mener vi ikke skal forske på solceller - men solcellene er heller ikke gode nok i dag til å konkurrere med den energien vi kan produsere med kjernekraft. Ja, takk! Begge deler!
The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR) has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U) is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that waste radio-toxicities and decay heats are up to a factor of 20 lower after 103 years. Two innovative fuel types named S90 and S20, ThO2 mixed with 90% and 20% enriched UO2 respectively, are compared as an alternative to standard uranium oxide (UOX) and uranium/plutonium mixed oxide (MOX) fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted.
Thorium is a fertile material that can easily be transformed into fissile 233U by thermal neutron capture. 233U does not exist in nature but is an excellent fissile material which has a higher η (∼2.3), the number of neutrons emitted per neutron absorbed, than the analogous fissile plutonium isotopes 239, 241Pu (∼1.7, ∼2.1) from the U/Pu cycle. This implies that the yield of new fissile material in fuels containing fertile 232Th will always be greater than for those containing fertile 238U, assuming a thermal neutron spectrum. Furthermore, the neutron capture/fission ratio of 233
sup>U (∼0.11) is lower than that of 239Pu (∼0.55) and five successive neutron captures on fertile 232Th are required to produce a trans-uranium element (TRU), while only one is necessary for fertile 238U. Therefore, production rates for the TRU’s, the major long-lived nuclear waste components will be much lower if 238U is replaced with 232Th.
We focus here on the EPR reactor and examine the consequences of removing significant quantities of fertile 238U from conventional UOX fuel and replacing it with fertile 232Th. This is achieved by blending thorium with uranium enriched to higher content of fissile 235U than the typical 4% needed for PWRs. We examine two new types of innovative fuel for the EPR, S20, a mixture of 20% enriched UO2 and ThO2, and S90, a mixture of 90% enriched UO2 and ThO2, and compare with conventional UOX fuel and MOX fuel fabricated from the plutonium recovered in spent UOX reprocessing.
The EPR is a Generation III + PWR concept, which is an evolutionary design rather than revolutionary and based on proven pressurized water technology - currently the most widely used worldwide.
In theory, light water reactors (LWRs) with a thermal neutron spectrum using 233U/232Th fuels can breed self-sustaining amounts of new fissile material since 2.3 neutrons are available per neutron absorbed, and indeed breeding was proven in the Shippingport Light Water Breeder Reactor . However, for a light water power reactor with realistically long cycle lengths there will be significant losses of neutrons to captures on the hydrogen in the water coolant, the fission products, the control poisons, the 233Pa intermediate thorium cycle nucleus and the structural materials, and thus the breeding ratio will be less than unity and breeding will be impossible. For breeding ratios of less than one, the thorium cycle will not be self-sustaining and the reactor will be dependent on a supply of some fissile material from the uranium cycle – either new fissile 235U or 239,241Pu.
dedicated exclusively as thorium cycle breeder reactors have been suggested, e.g. the Thorium Molten Salt Reactor  and the use of thorium in sodium critical reactors , or in subcritical accelerator driven systems [4-6]. However, these technologies are very different from current power reactor designs and cannot use existing cycle infrastructure. A large amount of research and development is needed for these systems and thus commercial power production could be decades into the future. Since LWRs comprise around 80% of the world fleet of operating plants, the fastest way to explore the potential of thorium is through the use of innovative solid fuels in the existing infrastructure of operating plants.
The once through cycle (OTC) is the cheapest fuel cycle in the short term. However, the potential energy content of the residual fissile and fertile isotopes is lost, and the OTC also gives the largest possible volume of high-level waste. The potential energy content of the spent fuel provides an incentive to recover the fissile isotopes. Fuel recycling also reduces the mass of high-level waste and the time of high radio-toxicity of waste, thus reducing the requirements for both the number of repositories and the duration of geological storage. The goal of the present work is to examine thoriated fuels as an alternative to present MOX recycled fuels that should provide greater economy of uranium resources and lower waste inventories.
2 EPR simulations with the MURE code
An EPR assembly based on a 17 × 17 lattice was modeled using MURE (MCNP Utilities for Reactor Evolution), based on the Monte-Carlo neutron transport code MCNP5 . MURE is a precision research code that has been developed jointly at the Institut de Physique Nucléaire d’Orsay (IPNO) and the Laboratoire de Physique Subatomique et Cosmologie (LPSC) of Grenoble . MURE performs calculations of the time-evolution of reactor fuels with full 3D geometry. Consecutive MCNP calculations are performed to determine reaction rates and deduce core material evolution over time at a constant reactor power by solving the set of coupled differential equations for the production and destruction of isotopes.
Fig. 1. Cross section of the simulated EPR assembly