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.

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.

scholarly journals Analysis of Some Computer Codes Used for Evaluation of Spent Nuclear Fuel Radiation Parameters

2019 ◽  
pp. 44-50
Author(s):  
A. Smaizys ◽  
E. Narkunas ◽  
V. Rudychev ◽  
Y. Rudychev
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Input Data ◽  
Spent Nuclear Fuel ◽  
Operation Time ◽  
Computer Code ◽  
Chemical Compositions ◽  
Spent Fuel ◽  
Management Options ◽  
Computer Codes

The radiation parameters such as radionuclide content and activities, fluxes and energy spectrum of gamma and neutrons of spent nuclear fuel are essential when planning further spent fuel management options – interim wet or dry storage or disposal into a geological repository. Radiation parameters determine the design of a storage or disposal facility, what materials, structures and thicknesses of structures should be used to provide adequate biological shielding. Experimental measurements of spent fuel radiation parameters are rather complicated and expensive, therefore numerical methods are widely used. Various computer codes (APOLLO, BOXER, CASMO, FISPACT, ORIGEN-S, WIMS, etc.) have been developed to simulate the irradiation processes of nuclear fuel and to obtain resulting radiation parameters. Irrespective of the used computer code, the input data firstly must be entered into that code. When simulating nuclear fuel irradiation and burn-up in the reactor core, the geometrical parameters of the fuel assembly, materials’ data (chemical compositions, densities), the operating parameters of the reactor (power, operation time, coolant parameters, etc.) shall be entered into the program as initial data. Fairly often approximations of the input data are performed, for example, fuel rods in a fuel assembly are homogenized and geometrically described as a solid cylinder, the reactor operation time is assumed as continuous and at constant power. The particularity of the input data and accepted assumptions depend on what initial information is available and on the capabilities of the computer code. The modelled spent fuel radiation parameters depend not only on the input data and assumptions, but also on the cross-section databases that are used in computer codes. Computer codes TRITON, ORIGEN-S and FISPACT have been used to model the concentration of actinides and fission products in the spent fuel from the RBMK-1000 reactor. The obtained results are compared and possible reasons for the differences in the modelling results are discussed.

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.

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 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.

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 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.

Effects of matrix composition and sample preparation on instant release fractions from high burnup nuclear fuel

2014 ◽  
Vol 1665 ◽  
pp. 261-266 ◽  
Author(s):  
O. Roth ◽  
J. Low ◽  
K. Spahiu
Keyword(s):  
Sample Preparation ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Matrix Composition ◽  
Spent Fuel ◽  
Preparation Methods ◽  
Rapid Release ◽  
Leaching Solution ◽  
First Contact

ABSTRACTThe rapid release of fission products segregated either to the gap between the fuel and the cladding or to the UO2 grain boundaries from spent nuclear fuel in contact with water (often referred to as the instant release fraction - IRF) is of interest for the safety assessment of geological repositories for spent fuel due to the potential dose contribution. In September 2012 a study was initiated with the aim of comparing the instant release behavior of fuels with and without additives/dopants. Preliminary results from this (ongoing) study indicate that the release of uranium during the first contact periods was higher than during the tests with fuel segments, even though the fuel was cut open recently [1]. This could be due to the sample preparation method which included axial cutting of the cladding in order to remove the fuel fragments used in the study. In the present work, leaching data from both studies are presented and the releases are discussed comparing the two sample preparation methods and considering the effect of matrix composition. The leaching studies have been performed in air using 10 mM NaCl + 2 mM NaHCO3 as leaching solution.

Effects of matrix composition on instant release fractions from high burn-up nuclear fuel

2013 ◽  
Vol 1518 ◽  
pp. 145-150 ◽  
Author(s):  
Olivia Roth ◽  
Jeanett Low ◽  
Michael Granfors ◽  
Kastriot Spahiu
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Matrix Composition ◽  
Initial Contact ◽  
Gas Release ◽  
Spent Fuel ◽  
Rapid Release ◽  
Fission Gas ◽  
Fission Gas Release

ABSTRACTThe release of radionuclides from spent nuclear fuel in contact with water is controlled by two processes – the dissolution of the UO2 grains and the rapid release of fission products segregated either to the gap between the fuel and the cladding or to the UO2 grain boundaries. The rapid release is often referred to as the Instant Release Fraction (IRF) and is of interest for the safety assessment of geological repositories for spent fuel due to the potential dose contribution.Previous studies have shown that the instant release fraction can be correlated to the fission gas release (FGR) from the spent fuel. Studies comparing results from samples in the form of pellets, fragments, powders and a fuel rodlet have shown that the sample preparation has a significant impact on the instant release, indicating that the differentiation between gap release and grain boundary release should be further explored.Today, there are trends towards power uprates, longer fuel cycles and increasing burn-up putting additional requirements on the nuclear fuel. These requirements are met by the development of new fuel types, such as UO2 fuels containing dopants or additives. The additives and dopants affect fuel properties such as grain size and fission gas release. In the present study we have performed experimental leaching studies using two high burnup fuels with and without additives/dopants and compared the fuel types with respect to their instant release behavior. The results of the leaching of the samples for the 3 initial contact periods; 1, 7 and 23 days are reported here.

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