Feasibility evaluation on European capabilities for 238Pu based radioisotope power systems

Research outputpeer-review

Abstract

Production of 238 Pu proceeds via neutron irradiation of 237 Np, which is created as a by-product in nuclear fission reactors, with a typical production slightly less than 1000 g/tU. When reprocessing spent nuclear fuel as e.g. done in the ORANO plant at La Hague (France) via the plutonium uranium reduction extraction (PUREX) process, neptunium is partially co extracted with uranium from dissolved irradiated nuclear fuel but as of today, it is not further refined, but instead added to the high-level waste and vitrified. The PUREX process can in principle be modified for neptunium recovery, and reprocessing of civil spent fuel can thus provide an abundant source of 237 Np. Neutron irradiation of separated 237Np to produce 238pU is conceptually simple, but producing sizeable quantities of 238 Pu with acceptable isotopic purity, separating it from the host matrix in which it is generated, its further processing and encapsulation, poses formidable technological challenges. All plutonium isotopes of technological interest are extremely radiotoxic alpha emitters. The elevated specific activity of isotope 238 adds highly concentrated radiolysis issues in liquid phase processing, static charging problems during powder handling, and heat generation when forming solid samples to the normal challenges of handling highly radiotoxic materials. Also, the precursor material 237Np is a radiotoxic alpha emitter with a lesser specific activity compared to 238 Pu. Especially the very rich and weakly explored chemistry of Np is a challenge here. In the present contribution, options for the irradiation of 237Np containing transmutation targets in the BR-2 high flux reactor of SCK CEN (Mol, Belgium) are presented and boundary conditions for the production of such targets are discussed. The principal technology steps are: •Fabrication of 237Np targets for neutron irradiation •Production of 238 Pu by neutron irradiation of 237Np targets •Processing of irradiated targets and Pu/Np separation •Conversion of separated 238 Pu to solid PuO 2 pellets Several technology options exist in each of the steps listed above, and these are reviewed with specific attention paid to those options which have been brought to actual production stage in the past, or which are intensively pursued today. Two principal options stand out: mixed ceramic-metal (CERMET) based routes developed and implemented at the Savannah River Site between the early 1950's and late 1980's and ceramic based routes pursued a.o. by Oak Ridge National Laboratory that have recently gained more attention. Technology options chosen in the early days are of course not necessarily the ones which would today be preferred. The European nuclear technology also developed differently over the past decades than the US nuclear technology, particularly with respect to civil spent fuel reprocessing, Pu separation and (U,Pu)O 2 manufacturing for Light Water Reactor (LWR) application. The two front-end processes (CERMET and full-oxide) for Np target production have similarities with established industry-scale fuel manufacturing processes in Europe: CERMET processes are applied in Materials Test Reactor (MTR) fuel fabrication and full-oxide processes are the reference technology for power reactors. The full-oxide process is furthermore also implemented at industrial scale for mixed uranium-plutonium (MOX) oxide fuel manufacturing, which shares similar radiotoxicity concerns as for Np-targets. CERMET front-end processes are in Europe applied for uranium-based fuels. An assessment of 238 Pu production capabilities in the BR-2 reactor has shown that suitable core positions can be selected with sufficiently low by-production of unwanted 236 Pu. Further assessment will be needed to evaluate which is the flexibility regarding core positions that guarantee limited production of 236 Pu. Production rates of 238 Pu were evaluated for unperturbed flux conditions to deduce an upper boundary of possible production rates. Preliminary calculations under perturbed flux conditions were performed for a design option that yet has to be optimized to deduce a lower boundary. The unperturbed flux results showed a theoretical upper boundary for the transformation yield slightly below 20%, achieved after three cycles of 28 days each and 28 days downtime between each cycle. Two production campaigns (i.e. six cycles) can reasonably be foreseen per year. Prolonged irradiation reduces the Pu vector below acceptable quality. The perturbed flux calculations for an un-optimized target showed a lower boundary slightly above 5% under the same irradiation conditions. Actual production yields are expected to be closer to the lower boundary than the upper boundary. Assuming a loading of 3 kg of Np, one may thus expect 150 g 238 Pu for a single production campaign, or 300 g 238 Pu per year. Design optimizations are expected to improve this yield. Regarding the processing of irradiated targets, the principal concerns are waste-related. The dissolution stage for the NpO 2 Al CERMET targets is particularly problematic and from this perspective full-ceramic NpO 2 targets would be preferred. The anion exchange purification stage as historically applied at SRS could be replaced by solvent extraction with TBP, most probably in combination with additional purification of the products via ion exchange. It is believed that the TBP process will create less waste than an anion exchange process and therefore seems to be the preferred method. Experience with these processes is available in Europe. The decay of 238 Pu dominates the radiology of the plutonium generated from 237Np irradiation. Decay parameters of a typical Pu-vector issued by 237Np irradiation have been compared with those of LWR MOX for which much broader experience exists in Europe. Even compared to MOX, issued from high burnup UO2 and including the ingrowth of 241 Am equivalent to a time lapse of 2 1/2 years, the radiological parameters per unit mass of Pu issued from 237 Np irradiation are higher by about two orders of magnitude. This difference will have to be taken into account for the Pu conversion process and the PuO 2 production process.
Original languageEnglish
Title of host publication2023 13th European Space Power Conference, ESPC 2023
Place of PublicationElche, Spain
PublisherIEEE - Institute of Electrical and Electronics Engineers
Number of pages13
Edition2023
ISBN (Electronic)979-8-3503-2899-8
ISBN (Print)979-8-3503-2900-1
DOIs
StatePublished - 6 Nov 2023

Publication series

Name2023 13th European Space Power Conference, ESPC 2023

ASJC Scopus subject areas

  • Aerospace Engineering
  • Energy Engineering and Power Technology
  • Electrical and Electronic Engineering

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