Main Results of the French Program on Partitioning of Minor Actinides,a Significant Improvement Towards Nuclear Waste Reduction

In the framework of the successive 1991 and 2006 Waste Management Act, French government supported a very significant R&D program on partitioning and transmutation of minor actinides (MA). This program aims to study potential solutions for still minimizing the quantity and the hazardousness of final waste, by MA recycling. Indeed, MA recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long-life of the waste. Within this framework, this paper aims to present the most recent progress obtained in CEA on the development of innovative actinide partitioning hydrometallurgical processes in support of their recycling, either in an homogeneous mode (MA are recycled at low concentration in all the standard reactor fuel) or in an heterogeneous mode (MA are recycled at higher concentration in specific targets, at the periphery of the reactor core). Recovery performances obtained on recent tests in high active conditions of the so-called GANEX and DIAMEX-SANEX process will be presented and discussed in light of the potential P&T scenarios. Finally, recent developments regarding the recycling of the sole Am will be presented as well as the results obtained on highly active solutions for this so-called EXAM process. This set of results gives to the French government a portfolio of potential recycling processes which could be separately and progressively implemented if decided.     

The Nuclear Fuel Cycle Back-End after Fukushima

The impact of the Fukushima accident on the nuclear fuel cycle back-end is obvious. In the present paper the various back-end options and their impact on the environment will be presented in view of this new perspective. The partitioning and transmutation (P&T) concept and the direct disposal of nuclear fuel including a very long intermediate storage will certainly be revisited with respect to safety considerations; here the instant release fraction (IRF) and the long-term stability of the fuel matrix under real repository conditions are highly relevant. Furthermore the impact of released radionuclides to the environment will have a higher attention.     

Recycling the Actinides,The Cornerstone of Any Sustainable Nuclear Fuel Cycles

The sustainability of the current nuclear fuel cycles is not completely achieved since they do not optimise the consumption of natural resource (only a very small part of uranium is burnt) and they do not ensure a complete and efficient recycling of the potential energetic material like the actinides. Promoting nuclear energy as a future energy source requires proposing new nuclear systems that could meet the criteria of sustainability in terms of durability, bearability and liveability. In particular, it requires shifting towards more efficient fuel cycles, in which natural resources are saved, nuclear waste are minimised, efficiently confined and safely disposed of, in which safety and proliferation-resistance are more than ever ensured. Such evolution will require (i) as a mandatory step, evolutionary recycling of the major actinides U and Pu up to their optimized use as energetic materials using fast neutron spectra, (ii) as an optional step, the implementation of the recycling of minor actinides which are the main contributors to the long term heat power and radiotoxicity of nuclear waste. Both options will require fast neutrons reactors to ensure an efficient consumption of actinides. In such a context, the back-end of the fuel cycle will be significantly modified: implementation of advanced treatment/recycling processes, minor-actinides recovery and transmutation, production of lighter final waste requiring lower repository space. In view of the 2012 French milestones in the framework of the 2006 Waste Management Act, this paper will depict the current state of development with regards with these perspectives and will enlighten the consequences for the subsequent nuclear waste management.     

Pyrochemical treatment of spent nitride fuels for MA transmutation

Nitride fuels have several advantages including high thermal conductivity and high metal density(like metallic fuels) and high melting point and isotropic crystal structure(like oxide fuels). Since the late 1990 s, the partitioning and transmutation of minor actinides(MA) has been studied to decrease the long-term radio-toxicity of high-level waste and to mitigate the burden of final disposal. Japan Atomic Energy Agency(JAEA) has proposed a dedicated transmutation cycle using an accelerator-driven system(ADS) with nitride fuels containing MA. The nitride fuel cycle we have developed includes a pyrochemical process. Our focus is on the electrolysis of nitride fuels and their refabrication from the recovered actinides; other processes are similar to the technology for metal fuel treatment and have been studied elsewhere. Here, we summarize our activity on the development of the pyrochemical treatment of spent nitride fuels.     

Radiometric detector options to aid in DOE high activity waste tank in situ characterization efforts

Radioactive waste cleanup and subsequent closure of waste storage tanks is currently underway at the Savannah river site, prompting the need to characterize the residual contents (heels) of the tanks. Occasionally, results from laboratory analyses indicate alternative sub-sampling strategies are needed, resulting in repetitive efforts to sample and analyze tank bottoms. The development of a system for in situ tank analyses using a radiometric probe, which could be lowered into a waste tank, could aid in identifying waste structures on tank bottoms requiring further sampling and characterization. Ideally, the probe would provide information for determining which structures were higher in concentrations of actinides and fission products characteristic of DOE high level waste (HLW) heels. Although only a limited set of isotopes can be measured directly without extensive radiochemical separations, the low-energy photon spectra of HLW do offer some intriguing possibilities for characterization using a radiometric probe. One possibility for obtaining a low-energy photon spectrum in the presence of high levels of interfering radiation would be to design a probe primarily based upon recently developed technology from Amptek Inc. Such a detector would be relatively insensitive to the high photon background, which would paralyze conventional gamma probes (i.e. sodium iodide) subjected to the same radiological conditions. The prototype detector is capable of successfully obtaining high resolution measurements at very high count rates (in excess of 500,000 counts per second). An overview of measurements obtained from various HLW samples using the prototype Amptek detector, as well as some additional detector technologies, which could enhance this prototype, will be discussed.     

Drop calorimetric measurements on a versatile monazite phase loaded with simulated radioactive waste

Monazite is one of the candidate ceramic matrices for the immobilization of high level radioactive waste (HLW) from the reprocessing of spent nuclear fuel. The monazite phase, Ce_0.8Ca_0.2PO₄, can accommodate cations of different valences due to the mixed valence state (+3 and +4) of Ce in this compound, by facilitating the oxidation and reduction of the Ce³⁺ and Ce⁴⁺ as required by the in-coming cation. This will assist in accommodating HLW of different compositions in the monazite crystal structure even if the average valence of the HLW elements is other than 3. Therefore, the monazite phase, Ce_0.8Ca_0.2PO₄, can be a versatile host for the immobilization of HLW. The enthalpy increment and heat capacity of this versatile monazite phase and a simulated waste form based on it with 20 mass% HLW oxides were measured by drop calorimetry in the temperature range from 373 to 873 K, and the results are compared with those measured for CePO₄.     

Hydrometallurgical Reprocessing of BREST-OD-300 Mixed Uranium-plutonium Nuclear Fuel

The duration of external fuel cycle of BREST-OD-300 reactor with mixed U-Pu nitride fuel (MNIT) including hydrometallurgical reprocessing should not exceed 3 years. An average burnup of the fuel should be 6% of heavy metal (HM) with the potential increase up to 10% HM. Therefore, the technology should provide the reprocessing of spent nuclear fuel (SNF) after less than 2 years cooling time and with fissile materials (FM) content of 10 – 15%. Pellets technology has been chosen for the MNIT fuel production. That means necessity to receive the recycled actinides oxides of high purification coefficient (∼ 10⁶). Currently on a laboratory scale, the following process stages have been tested on the real products: actinide oxides production and rare-earth and trans-plutonium elements separation. Moreover, on a pilot scale the process of high level radioactive waste (HLW) and intermediate level radioactive waste (ILW) concentration by evaporation has been tested, as well as the Am-Cm separation. In 2015, the design of the MNIT SNF reprocessing facility has been started, placed at the JSC Siberian Chemical Plant site as a part of the pilot demonstration power complex (PDPC) with BREST-OD-300 reactor. MNIT SNF reprocessing plant (RP) should be put in operation after 2020.     

Behaviour of ruthenium in nitric media (HLLW) in reprocessing plants: a review and some perspectives

Journal of Radioanalytical and Nuclear Chemistry - The reprocessing of spent nuclear fuel produces high level liquid waste (HLLW). Due to the decay heat, the concentrated nitric solutions...     

VHTR,ADS, and PWR Spent Nuclear Fuel Analysis

The aim of this study is to analyze and compare the discharged-spent fuel of three types of nuclear systems: a Very High-Temperature Gas Reactor (VHTR), a lead-cooled Accelerator-Driven System (ADS) and a standard Pressurized Water Reactor (PWR). The two first systems, VHTR, and ADS were designed to use reprocessed fuels. UREX+ and GANEX techniques were used for the reprocessing processes respectively. The fuel burnup simulated for the systems in other works have been used to obtain the final composition of the spent fuel discharged. After discharge, the radioactivity, the radiotoxicity, and the decay heat were evaluated through the ORIGEN 2.1 code until 10⁷ years and compared to the literature.     

Reduction and removal of neptunium from synthetic high level liquid waste by TRPO process

Valency control of neptunium is an important issue in the partitioning of high level liquid waste (HLLW) from power-reactor spent fuel treatment. The redox behavior of neptunium in HLLW is quite different from that in nitric acid because of the effect of the large amount of ions in HLLW. In order to remove neptunium from HLLW, we studied the reduction of neptunium in synthetic HLLW (SHLLW) to maintain its valency at IV so that it can be extracted by TRPO extractant in the well developed Chinese TRPO process. Five different reductants were tested and the reduction behavior was investigated. The influence of some active elements in SHLLW was studied. The mechanism that the reductants react with neptunium through Fe element was supposed and proved by experiments. The reduction rate of Np(V) was highly enhanced by Fe element. Finally, a hybrid reductant was suggested and good reduction efficiency was obtained.     

Characterization of high level waste for minor actinides by chemical separation and alpha spectrometry

The high level waste (HLW) generated from the reprocessing of the spent fuel of pressurized heavy water reactor has been characterized for the minor actinides. The radiation dose of the waste solution was reduced by radiochemical separation of cesium from HLW by solvent extraction with chlorinated cobalt dicarbollide dissolved in 20% nitrobenzene in xylene. Minor actinides (Np, Pu, Am, Cm) in the high level waste were assayed by alpha spectrometry following radiochemical separation. The gross alpha activity determined by liquid scintillation agrees well (within 10%) with the cumulative quantities of actinides determined by alpha spectrometry.     

An Improved Method for Testing Weak VOCs in Dry Plants with a Purge and Trap Concentrator Coupled to GC-MS

The purge and trap (P&T) technique was improved for measuring the release of organic compounds with weak volatility (weak VOCs) from dry plant materials. Using distilled water as a dispersant, the plant tissues were mulled and placed in the purge tube of a P&T concentrator. Then the sample-containing purge tube was heated to 80 °C with helium as the carrier gas, and the purged volatiles were preconcentrated in the trap prior to analysis with GC-MS. The VOCs in Chinese herbal medicinal plants Swertia tetraptera, Saussurea involucrate and S. lacostei, which had been stored dry for 1–2.5 years were assayed with this improved method and conventional P&T techniques. Our results show this new P&T method had great promise for determining the VOCs in dry plant materials. Using this new technique, we identified 38 weak VOCs with a large peak area from the dry samples. In contrast, less than five VOCs were detected by the conventional P&T method. So the improved heat-purge and trap system showed to be more efficient for measuring the release of the weak VOCs from dry plant materials.     

Chemical Pre-Treatment of used Fuel for Long–Term Storagea

Repository programs throughout the world have been slowed by the need for increased local public involvement in the siting and licensing process. The result has been an increase in the dry storage of used fuel at reactor sites and the potential that such storage may be extended for many decades, even centuries. While there are sound technical reasons to believe that dry storage can be conducted safely, there are increasing concerns that the ultimate transfer to either a future repository or a centralized separations plant may result in fractured cladding and serious handling issues, including criticality concerns. These concerns would be increased for higher burn-up fuels. Currently, various chemical pre-treatment processes under R&D for application to commercial used oxide fuel have been investigated at the laboratory scale as methods to simplify and increase the safety of the remaining stages of conventional solvent extraction processing. This includes advanced decladding methods and various oxidation/reduction processes designed to release volatile and semi-volatile fission products, produce finely divided uranium oxide powder, and ameliorate the subsequent nitric acid dissolution step. The paper examines the potential for combining several chemical and physical pre-treatment steps to minimize long-term concerns about safe transport of used fuel, possibly providing another option for future nuclear waste management. Laboratory data from both cold and hot testing will provide the basis for the evaluation. An example of a potential pre-treatment process includes shearing, advanced voloxidation and off-gas treatment, the possible mixing of the resulting uranium oxide with a secondary oxide, and densification and recanning in nitric acid-soluble storage containers for extended time periods. Chemical decladding may be feasible to replace shearing. Zirconium recycle may also be feasible, significantly reducing high level waste quantities. Both analytic and experimental data will be applied to the examination of this potential fuel cycle option.     

An advanced partitioning process for key elements separation from high level liquid waste

To separate MA(Am,Cm) and some fission product elements(FPs) such as Tc,Pd,Cs and Sr from high level liquid waste(HLLW) systematically,we have been studying an advanced aqueous partitioning process,which uses selective adsorption as the separation method.For this process,we prepared several novel adsorbents which were immobilized in a porous silica/polymer composite support(SiO 2-P).Adsorption and separation behavior of various elements was studied experimentally in detail.Small scale separation tests using simulated HLLW solutions were carried out.Pd(II) was strongly adsorbed by the AR-01 anion exchanger and effectively eluted off by using thiourea.Successful separation of Pd(II) from simulated HLLW was achieved.Tc(VII) also exhibited strong adsorption on AR-01 and could be eluted off by using U(IV) as a reductive eluent.Am(III) presented significantly high adsorbability and selectivity onto R-BTP/SiO 2-P adsorbents over various FPs including Ln(III).The R-BTP adsorbents were fairly stable in 3 M HNO 3,but instable against-irradiation-3M HNO 3.An advanced partitioning process consisting of three separation columns for the target elements separation from HLLW was proposed and the obtained experiment results indicated that the proposed process is essentially feasible.     

Advanced-ORIENT Cycle Project: Summary of Phase I Fundamental Studies

The Advanced Optimization by Recycling Instructive Elements (Adv.-ORIENT) Cycle strategy has been proposed as a basic concept for trinitarian research on the separation, transmutation, and utilization (S, T, & U) of nuclides and elements on the basis of the FBR fuel cycle. Working in this direction, validation of the principal separation methods and related safety research were performed from 2006 through 2011 as the first phase. The separation scheme was composed of four ion exchange (IXC) steps and one catalytic electrolytic extraction (CEE) step. The fundamental technological aspects are summarized as the Phase I program.     

Analysis of high-level waste for tetraphenylborate using HPLC

Summary One waste remediation process used at the Savannah River Site was the in-tank precipitation of the beta-emitting ¹³⁷Cs from high-level waste (HLW) using sodium tetraphenylborate (NaTPB) followed by processing the resulting decontaminated filtrate into grout at the Saltstone Production Facility (SPF). A simple method was developed for the monitoring of tetraphenylborate (TPB) in high-level waste (HLW) containing up to 0.38 Ci/gal of ¹³⁷Cs. Separation was achieved by extraction of the high sodium-bearing waste with acetonitrile followed by analysis using reversed-phase high performance liquid chromatography (HPLC). The sample preparation method allowed for the handling of an organic extraction layer that had 94% less acitivity than the HLW sample. The subsequent HPLC analysis of the extraction layer determined the TPB concentration in HLW waste to 0.8 mg/l with a %rsd of 8.     

HEU measurements of holdup and recovered residue in the deactivation and decommission activities of the 321-M Reactor Fuel Fabrication Facility at the Savannah River Site

Summary This paper contains a summary of the holdup and material control and accountability (MC&A) assays conducted for the determination of highly enriched uranium (HEU) in the deactivation and decommissioning (D&D) of the Reactor Fuel Fabrication Facility at the Savannah River Site (SRS). The facility was used to fabricate HEU fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the SRS production reactors. The facility operated for more than 35 years. During this time thousands of uranium-aluminum alloy (U-Al) production reactor fuel tubes were produced. After the facility ceased operations in 1995, all of the easily accessible U-Al was removed from the building, and only residual amounts remained. The bulk of this residue was located in the equipment that generated and handled small U-Al particles and in the exhaust systems for this equipment (e.g., chip compactor, casting furnaces, log saw, lathes A & B, cyclone separator, Freon?cart, riser crusher, …, etc). The D&D project is likely to represent an important example for D&D activities across SRS and across the Department of Energy weapons complex. The Savannah River National Laboratory was tasked to conduct holdup assays to quantify the amount of HEU on all components removed from the facility prior to placing in solid waste containers. The ²³⁵U holdup in any single component of process equipment must not exceed 50 g in order to meet the container limit. This limit was imposed to meet criticality requirements of the low level solid waste storage vaults. Thus, the holdup measurements were used as guidance to determine if further decontamination of equipment was needed to ensure that the quantity of ²³⁵U did not exceed the 50 g limit and to ensure that the waste met the Waste Acceptance Criteria (WAC) of the solid waste storage vaults. Since HEU is an accountable nuclear material, the holdupassays and assays of recovered residue were also important for material control and accountability purposes. In summary, the results of the holdup assays were essential for determining compliance with the Waste Acceptance Criteria, Material Control & Accountability, and to ensure that administrative criticality safety controls were not exceeded. This paper discusses theg-ray assay measurements conducted and the modeling of the acquired data to obtain measured holdup in process equipment, exhaust components, and fixed geometry scrap cans. It also presents development work required to model new acquisition configurations and to adapt available instrumentation to perform the assays.     

Preliminary study of the pyrolysis of steam classified municipal solid waste

Steam classified municipal solid waste (MSW) has been studied for use as a combustion fuel, feedstock for composting, and cellulytic enzyme hydrolysis. A preliminary study has been conducted using a prototype plasma arc pyrolysis system (in cooperation with Plasma Energy Applied Technology Inc., Huntsville, AL) to convert the steam classified MSW into a pyrolysis gas and vitrified material. Using a feed rate of 50 lbs/h, 300 lbs of the material was pyrolysized. The major components of this pyrolysis gas were H₂, CO, and CO₂. A detailed presentation of the emission data along with details on the system used will be presented.     

Study on the adsorption behavior of silica gel for zirconium,plutonium and other fission products in high-level liquid waste

The adsorption behavior of silica gel for Zr, Pu and fission products (Fe, Mo, Nd, etc.) in high-level liquid waste (HLLW) have been studied. The dynamic desorption for Zr and Pu are also investigated. Silica gel is found to have high selectivity for Zr to other elements in HLLW and its adsorption capacity can be enhanced by H₂C₂O₄ elution and multiple reuse. However, the adsorption for Pu(IV) is also found. This may cause spent silica gel to be alpha waste. Further study should be carried out on the adsorption behavior for Pu(IV).