New Methods for HTR Fuel Waste Management

Abstract

Considering the need to reduce waste production and greenhouse emissions by still keeping high energy efficiency, various 4 th generation nuclear energy systems have been proposed. As far as graphite moderated reactors are concerned, one of the key issues is the large volumes of irradiated graphite encountered (1770 m 3 for fuel elements and 840 m 3 for reflector elements during the lifetime (60 years) of a single reactor module [1]). With the objective to reduce volume of waste in the HTR concept, it is very important to be able to separate the fuel from low level activity graphite. This requires to separate TRISO particles from the graphite matrix with the sine qua non condition to not break TRISO particles in case of future embedding of particles in a matrix for disposal.According to National Regulatory Systems, in case of limited graphite waste production or of short duration HTR projects (e.g. in Germany), direct disposal without separation is acceptable. Nevertheless, in case of large scale deployment of HTR technology, such approach is not economical and sustainable. Previous attempts in graphite management (furnace, fluidised bed and laser incinerations and encapsulation matrices) dealt with graphite matrix only. These are the reasons why we studied the management of irradiated compact-type fuel element. We simulated the presence of fuel in the particles by using ZrO 2 kernels. Compacts with ZrO 2 TRISO particles were manufactured by AREVA NP.Two original methods have been studied. First, we tested high pressure jet to erode graphite and clean TRISO particles. Best erosion rate reached about 0.18 kg/h for a single nose ending. Examination of treated graphite showed a mixture of undamaged TRISO particles, particles that have lost the outer pyrolytic carbon layer and ZrO 2 kernels. Secondly, we studied the thermal shock method by immerging successively graphite into liquid nitrogen and hot water to cause fracturing of the compact. This produced particles and graphite fragments with diameter ranging from several centimetres to less than 500 µm. This relatively simple and economic method may potentially be considered as a pre-treatment step and be coupled with other method(s) before reprocessing and recycling for example.

CHARACTERIZATION OF DIRECT ADDITIVE MANUFACTURED U3Si2 SURROGATE TO PREDICT U3Si2 MICROSTRUCTURES

“…Surrogate TRISO fuel was produced for irradiation testing at the High Flux Isotope Reactor (HFIR) at Oakridge National Laboratory [8]. Other studies on new methods for High Temperature Reactor (HTR) fuel waste management also simulated the presence of fuel in TRISO particles by using ZrO2 kernels [9]. The previous research proved the suitability in employing zirconium as a surrogate material for uranium.…”

“…Zirconium was down selected as a surrogate material due to its chemical and crystallographic properties, which are comparable to those of uranium. Research studies have proven the feasabiltiy of employing ZrO2 as surrogate material for uranium fuel in TRISOcoated particles [8] [72]. Zirconium and silicon powders were mixed in the proper stoichiometric proportions to undergo a laser synthesis process and ultimately produce Zr3Si2.…”

The chemical durability of glass and graphite–glass composite doped with cesium oxide

“…In our research, the process of intercalation of TRISO particles from its graphite matrix caused damage to the outer coating of TRISO particles and released fission products from the kernel. An intercalation with a high speed water jet, applied to a graphite matrix in parallel and then in perpendicular direction, generated respectively 45.3% and 5.8% of TRISO particles without coating [4].…”

“…Although graphite is impermeable and chemically stable during an operation of a high temperature reactor, the porous structures of graphite facilitates release radionuclides to the biosphere [1]. Furthermore, it is expected that around 1770 m 3 of irradiated graphite would be generated from only the fuel elements during 60 years of operation of a High Temperature Reactor (HTR), making waste minimisation an important target in the development of this technology [2]. In order to reduce waste volume, fuel compacts are generally pulverized to separate the low level waste graphite from high level waste fuel particles.…”