Characterization of FUTURIX-FTA Irradiated Nuclear Fuel Samples

2017 ◽  
Author(s):  
Concettina Andrello ◽  
Daniel Freis ◽  
Rosa Lo Frano ◽  
Dimitri Papaioannou ◽  
Fabienne Delage
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Heat Rate ◽  
Current Status ◽  
High Level Waste ◽  
Minor Actinide ◽  
Spent Fuel ◽  
Post Irradiation ◽  
Irradiation Experiment

The amount of spent fuel and high-level waste already available, and which will be produced by the future NPPs operation, calls for the evaluation of any possible technological solution that could minimize the burden of their disposal: reduction of Minor Actinide (MA) content, in addition to the radiotoxicity and radioactivity, and of the generated thermal load (decay heat). In this context, R&D efforts currently focus on the development of methodologies and technical solutions for Partitioning and Transmutation. MAs and long-lived fission products are in fact the main contributors to the long-term radiotoxicity of spent nuclear fuel, and their transmutation to short-lived fission products, in fast spectrum nuclear reactors, in transmuters or in Accelerator Driven Systems (ADS), by neutron irradiation of dedicated fuels/targets, is a promising and widely investigated option. In order to provide substantial input for the safety assessment of innovative nuclear fuels dedicated to MA transmutation, several irradiation tests are being carried out. In some options, the investigated fuels/targets are uranium free, or of low uranium content, to improve the transmutation performance and contain high concentrations of MA and plutonium compounds. Two molybdenum based CER-MET fuels, called ITU-5 & ITU-6, were prepared at JRC Karlsruhe for the irradiation experiment FUTURIX-FTA (FUel for Transmutation of transURanium elements in phenIX/Fortes Teneurs en Actinide). The experiment performed from 2007 to 2009 in the Phénix reactor, France, in cooperation with CEA. The experiment ended after 235 equivalent full power days (EFPD) at a Linear Heat Rate of circa 130 W/cm and reached burn-ups of 18 %FIHMA and 13 %FIHMA, respectively. Afterwards, the pins were transported to the Hot Cells of JRC Karlsruhe for Post Irradiation Examination. After a short summary describing the fuel preparation and irradiation conditions of the FUTURIX FTA irradiation experiment, the paper will give an overview on the current status and further planning of the Post Irradiation Examinations of ITU-5 & ITU-6 at JRC Karlsruhe. Finally, the results of the characterisations will be discussed and conclusions on the irradiation performance will be drawn. The results of this experiment will help to increase the knowledge and understanding of the irradiation behaviour of metal based transmutation targets and the qualification and validation of models developed to predict fuel safety performance.

scholarly journals German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

2013 ◽  
Vol 2013 ◽  
pp. 1-11 ◽  
Author(s):  
A. Schwenk-Ferrero
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
High Level Waste ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Cycle Simulation ◽  
High Level

Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content) and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104to 106years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

Distribution of Palladium During Spent Fuel Reprocessing

2003 ◽  
Author(s):  
Yu. Pokhitonov ◽  
V. Romanovski ◽  
P. Rance
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Platinum Group Metals ◽  
Spent Fuel ◽  
Platinum Metals ◽  
Spent Fuel Reprocessing ◽  
Platinum Group ◽  
Fuel Reprocessing

The principal purpose of spent fuel reprocessing consists in the recovery of the uranium and plutonium and the separation of fission products so as to allow re-use of fissile and fertile isotopes and facilitate disposal of waste elements. Amongst the fission products present in spent nuclear fuel of Nuclear Power Plants (NPPs,) there are considerable quantities of platinum group metals (PGMs): ruthenium, rhodium and palladium. Given current predictions for nuclear power generation, it is predicted that the quantities of palladium to be accumulated by the middle of this century will be comparable with those of the natural sources, and the quantities of rhodium in spent nuclear fuel may even exceed those in natural sources. These facts allow one to consider spent nuclear fuel generated by NPPs as a potential source for creation of a strategic stock of platinum group metals. Despite of a rather strong prediction of growth of palladium consumption, demand for “reactor” palladium in industry should not be expected because it contains a long-lived radioactive isotope 107Pd (half-life 6,5·105 years) and will thus be radioactive for a very considerable period, which, naturally, restricts its possible applications. It is presently difficult to predict all the areas for potential use of “reactor” palladium in the future, but one can envisage that the use of palladium in radwaste reprocessing technology (e.g. immobilization of iodine-129 and trans-plutonium elements) and in the hydrogen energy cycle may play a decisive role in developing the demand for this metal. Realization of platinum metals recovery operation before HLW vitrification will also have one further benefit, namely to simplify the vitrification process, because platinum group metals may in certain circumstances have adverse effects on the vitrification process. The paper will report data on platinum metals (PGM) distribution in spent fuel reprocessing products and the different alternatives of palladium separation flowsheets from HLW are presented. It is shown, that spent fuel dissolution conditions can affect the palladium distribution between solution and insoluble precipitates. The most important factors, which determine the composition and the yield of residues resulting from fuel dissolution, are the temperature and acid concentration. Apparently, a careful selection of fuel dissolution process parameters would make it possible to direct the main part of palladium to the 1st cycle raffinate together with the other fission products. In the authors’ opinion, the development of an efficient technology for palladium recovery requires the conception of a suitable flow-sheet and the choice of optimal regimes of “reactor” palladium recovery concurrently with the resolution of the problem of HLW partitioning when using the same facilities.

Integrated Model of Korean Spent Fuel and High Level Waste Disposal Options

2009 ◽  
Author(s):  
Yongsoo Hwang ◽  
Ian Miller
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Near Field ◽  
Integrated Model ◽  
High Level Waste ◽  
Spent Fuel ◽  
Parameter Uncertainties ◽  
Design Concepts ◽  
High Level

This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21’st century. The model addresses alternative design concepts for disposal of SNF of different types (CANDU, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model’s results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses.

scholarly journals The Limitations of an Air-Oxidation Breakaway Model to Predict a Zirconium Fire in a Spent Nuclear Fuel Pool Accident

2019 ◽  
Vol 11 (22) ◽  
pp. 6364
Author(s):  
Sanggil Park ◽  
Min Bum Park
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Critical Phenomenon ◽  
Fission Products ◽  
Model Parameters ◽  
Spent Fuel ◽  
Air Oxidation ◽  
Loss Of Coolant Accident ◽  
Spent Nuclear Fuel Pool ◽  
Spent Fuel Pool

The OECD/NEA Spent Fuel Pool (SFP) project was conducted to investigate consequences of spent nuclear fuel pool accident scenarios. From the project, it was observed that cladding temperature could abruptly increase at a certain point and the cladding was completely oxidized. This phenomenon was called a “zirconium fire”. This zirconium fire is one of the crucial concerns for spent fuel pool safety under a postulated loss of coolant accident scenario, since it would lead to an uncontrolled mass release of fission products into the environment. To capture this critical phenomenon, an air-oxidation breakaway model has been implemented in the MELCOR code. This study examines this air-oxidation breakaway model by comparing the SFP project test data with a series of MELCOR code sensitivity calculation results. The air-oxidation model parameters are slightly altered to investigate their sensitivities on the occurrence of the zirconium fire. Through such sensitivity analysis, limitations of the air-oxidation breakaway model are identified, and needs for model improvement is recommended.

scholarly journals Natural Clay Minerals as a Starting Material for Matrices for the Immobilization of Radioactive Waste from Pyrochemical Processing of SNF

2021 ◽  
Vol 13 (19) ◽  
pp. 10780
Author(s):  
Anna V. Matveenko ◽  
Andrey P. Varlakov ◽  
Alexander A. Zherebtsov ◽  
Alexander V. Germanov ◽  
Ivan V. Mikheev ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Storage Time ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Bentonite Clay ◽  
Ceramic Materials ◽  
High Level Waste ◽  
Natural Clay ◽  
High Level

Pyrochemistry is a promising technology that can provide benefits for the safe reprocessing of relatively fresh spent nuclear fuel with a short storage time (3–5 years). The radioactive waste emanating from this process is an electrolyte (LiCl–KCl) mixture with fission products included. Such wastes are rarely immobilized through common matrices such as cement and glass. In this study, samples of ceramic materials, based on natural bentonite clay, were studied as matrices for radioactive waste in the form of LiCl–KCl eutectic. The phase composition of the samples, and their mechanical, hydrolytic, and radiation resistance were characterized. The possibility of using bentonite clay as a material for immobilizing high-level waste arising from pyrochemical processing of spent nuclear fuel is further discussed in this paper.

A Trade Study for Waste Concepts to Minimize HLW Volume

2008 ◽  
Vol 1124 ◽  
Author(s):  
Dirk Gombert ◽  
Joe Carter ◽  
Bill Ebert ◽  
Steve Piet ◽  
Tim Trickel ◽  
...  
Keyword(s):  
Waste Management ◽  
Nuclear Fuel ◽  
Heat Dissipation ◽  
Spent Nuclear Fuel ◽  
Management Strategies ◽  
Radioactive Decay ◽  
Fission Products ◽  
Purex Process ◽  
High Level Waste ◽  
Glass Waste

AbstractAdvanced nuclear fuel reprocessing can partition wastes into groups of common chemistry. This enables new waste management strategies not possible with the plutonium, uranium extraction (PUREX) process alone. Combining all of the metallic fission products in an alloy and the balance as oxides in glass minimizes high level waste (HLW) volume. Implementing a waste management strategy using state-of-the-art combined waste forms and storage to allow radioactive decay and heat dissipation prior to placement in a repository makes it possible to place almost 10x the HLW equivalent of spent nuclear fuel (SNF) in the same repository space. However, using generic costs based on preliminary studies for waste stabilization facilities and separations modules, this analysis shows that combining the non-actinide wastes and using only one glass waste form is the most cost-effective.

Synthesis of Uranium Phosphate Phases and Potential Retardation Effects on Spent Fuel Radionuclide Migration

2002 ◽  
Vol 757 ◽  
Author(s):  
A. S. Turner ◽  
D. J. Wronkiewicz
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Potassium Phosphate ◽  
Fission Products ◽  
Chemical Compositions ◽  
Spent Fuel ◽  
Variable Chemical ◽  
Calcium Phosphorus ◽  
Uranyl Phosphates ◽  
Uranium Phosphate

ABSTRACTThe UO2 in spent nuclear fuel is unstable in the oxidizing conditions within the volcanic tuffs at the proposed nuclear repository at Yucca Mountain, Nevada. Over time, the UO2 will oxidize and corrode, releasing actinides and fission products to the surrounding environment. However, uranyl (U6+) phosphates (autunite, phurcalite, sodium autunite, etc.) are stable in such an oxidizing environment. The mobility of released radionuclides may be greatly retarded if they can be incorporated into these naturally stable phosphate phases, while the complex structures, variable chemical compositions, and natural analogue occurrences of the uranyl phosphates suggests such a process is favorable. Current tests have focused on synthesizing such phases by reacting uranium oxynitrate or UO3 with a calcium, sodium, or potassium phosphate and a base (if necessary) in a Teflon reaction vessel. Excess water is added, and the solution is heated at 90°C for 7, 35, or 182 days. SEM analyses have confirmed that various uranium phosphate crystalline solids have formed. XRD results indicate that tests using two different calcium phosphorus source materials, Ca2P2O7 and Ca10(OH)2(PO4)6, have both created synthetic phosphuranylite, Ca(UO2)[(UO2)3(OH)2(PO4)2] 2*12H2O. The formation of this phase is appears to be kinetically favored over other similar phases. Results utilizing sodium phosphates, NaH2(PO4)*H2O and Na2HPO4, have produced sodium autunite (NaUO2PO4*2H2O), but other phases are probably also present. Test results utilizing potassium phosphate, K3PO4, were inconclusive. Experiments using surrogate radionuclides are currently being performed in order to determine whether radionuclides, such as 239Pu, 137Cs, and 99Tc, released from corroded spent nuclear fuel can become incorporated into the crystalline structure of specific uranium phosphate phases, effectively limiting any further migration.

Modelling of RBMK Spent Nuclear Fuel Characteristics

2014 ◽  
Author(s):  
Arturas Smaizys ◽  
Povilas Poskas ◽  
Ernestas Narkunas
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Experimental Investigations ◽  
Spent Fuel ◽  
Radionuclide Content ◽  
Fuel Assemblies ◽  
Fuel Characteristics ◽  
Final Shutdown ◽  
Rbmk Fuel

After the final shutdown of Ignalina NPP, total amount of spent nuclear fuel is approximately 22 thousands of fuel assemblies. Radionuclide content and its characteristics in spent fuel are initial data for analysis of various safety related areas such as shielding, thermal analysis, radioactive releases and other processes. Experimental investigations of radionuclide content and characteristics in spent nuclear fuel are complicated and expensive, therefore numerical evaluation methods are widely used. Numerical modelling of spent RBMK fuel characteristics was performed using TRITON code from SCALE 6.1 system. Activities of fission products and actinides, gamma and neutron sources, decay heat obtained with TRITON code are compared with previous modelling results obtained using SAS2H sequence from the former SCALE 4.3 version. Some evaluated parameters are compared with published experimental data for RBMK spent nuclear fuel.

scholarly journals New Methods for Evaluation of Spent Fuel Condition during Long-Term Storage in Slovakia

2009 ◽  
Vol 2009 ◽  
pp. 1-5
Author(s):  
M. Mikloš ◽  
V. Kršjak
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Detection System ◽  
Fission Products ◽  
Spent Fuel ◽  
Limit Values ◽  
Long Term Storage ◽  
Fuel Assemblies ◽  
Leak Tightness ◽  
Evaluation And Monitoring

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system “Sipping in pool” delivered by Framatomeanp with external heating for the precise defects determination. In 2006, a new inspection stand “SVYP-440” for monitoring of spent nuclear fuel condition was inserted. This stand has the possibility to open WWER-440 fuel assemblies and examine fuel elements. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. With appropriate approach of conservativeness, new factor for specifying spent fuel leak tightness is introduced in the paper. By using computer simulations (based on SCALE 4.4a code) for fission products creation and measurements by system “Sipping in pool,” the limit values of leak tightness were established.

Materials Characteristics and Dissolution Behavior of Spent Nuclear Fuel

MRS Bulletin ◽  
1994 ◽  
Vol 19 (12) ◽  
pp. 24-27 ◽  
Author(s):  
L.H. Johnson ◽  
L.O. Werme
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Waste Form ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Power Stations ◽  
Geologic Disposal

The geologic disposal of spent nuclear fuel is currently under consideration in many countries. Most of this fuel is in the form of assemblies of zirconium-alloy-clad rods containing enriched (1–4% 235U) or natural (0.71% 235U) uranium oxide pellets. Approximately 135,000 Mg are presently in temporary storage facilities throughout the world in nations with commercial nuclear power stations.Safe geologic disposal of nuclear waste could be achieved using a combination of a natural barrier (the host rock of the repository) and engineered barriers, which would include a low-solubility waste form, long-lived containers, and clay- and cement-based barriers surrounding the waste containers and sealing the excavations.A requirement in evaluating the safety of disposal of nuclear waste is a knowledge of the kinetics and mechanism of dissolution of the waste form in groundwater and the solubility of the waste form constituents. In the case of spent nuclear fuel, this means developing an understanding of fuel microstructure, its impact on release of contained fission products, and the dissolution behavior of spent fuel and of UO2, the principal constituent of the fuel.

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