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.

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 Approaches to Safety Justification for Loading of VSC-VVER Containers in ZNPP DSFSF

2019 ◽  
pp. 82-87
Author(s):  
Ya. Kostiushko ◽  
O. Dudka ◽  
Yu. Kovbasenko ◽  
A. Shepitchak
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Spent Fuel ◽  
Fuel Assemblies ◽  
Fuel Storage ◽  
The Government

The introduction of new fuel for nuclear power plants in Ukraine is related to obtaining a relevant license from the regulatory authority for nuclear and radiation safety of Ukraine. The same approach is used for spent nuclear fuel (SNF) management system. The dry spent fuel storage facility (DSFSF) is the first nuclear facility created for intermediate dry storage of SNF in Ukraine. According to the design based on dry ventilated container storage technology by Sierra Nuclear Corporation and Duke Engineering and Services, ventilated storage containers (VSC-VVER) filled with SNF of VVER-1000 are used, which are located on a special open concrete site. Containers VSC-VVER are modernized VSC-24 containers customized for hexagonal VVER-1000 spent fuel assemblies. The storage safety assessment methodology was created and improved directly during the licensing process. In addition, in accordance with the Energy Strategy of Ukraine up to 2035, one of the key task is the further diversification of nuclear fuel suppliers. Within the framework of the Executive Agreement between the Government of Ukraine and the U.S. Government, activities have been underway since 2000 on the introduction of Westinghouse fuel. The purpose of this project is to develop, supply and qualify alternative nuclear fuel compatible with fuel produced in Russia for Ukrainian NPPs. In addition, a supplementary approach to safety analysis report is being developed to justify feasibility of loading new fuel into the DSFSF containers. The stated results should demonstrate the fulfillment of design criteria under normal operating conditions, abnormal conditions and design-basis accidents of DSFSF components.  Thus, the paper highlights both the main problems of DSFSF licensing and obtaining permission for placing new fuel types in DSFSF.

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 Minimizing the Standard Deviation of the Thermal Load in the Spent Nuclear Fuel Cask Loading Problem

Energies ◽  
2020 ◽  
Vol 13 (18) ◽  
pp. 4869
Author(s):  
Joaquín Bautista-Valhondo ◽  
Lluís Batet ◽  
Manuel Mateo
Keyword(s):  
Standard Deviation ◽  
Nuclear Fuel ◽  
Thermal Load ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Search Algorithm ◽  
Deterministic Algorithm ◽  
Mixed Integer ◽  
Spent Fuel ◽  
Fuel Assemblies

The paper assumes that, at the end of the operational period of a Spanish nuclear power plant, an Independent Spent Fuel Storage Installation will be used for long-term storage. Spent fuel assemblies are selected and transferred to casks for dry storage, with a series of imposed restrictions (e.g., limiting the thermal load). In this context, we present a variant of the problem of spent nuclear fuel cask loading in one stage (i.e., the fuel is completely transferred from the spent fuel pool to the casks at once), offering a multi-start metaheuristic of three phases. (1) A mixed integer linear programming (MILP-1) model is used to minimize the cost of the casks required. (2) A deterministic algorithm (A1) assigns the spent fuel assemblies to a specific region of a specific cask based on an MILP-1 solution. (3) Starting from the A1 solutions, a local search algorithm (A2) minimizes the standard deviation of the thermal load among casks. Instances with 1200 fuel assemblies (and six intervals for the decay heat) are optimally solved by MILP-1 plus A1 in less than one second. Additionally, A2 gets a Pearson’s coefficient of variation lower than 0.75% in less than 260s CPU (1000 iterations).

Sorption of Actinides in Well-Defined Oxidation States on Geologic Media

1981 ◽  
Vol 11 ◽  
Author(s):  
B. Allard ◽  
U. Olofsson ◽  
B. Torstenfelt ◽  
H. Kipatsi ◽  
K. Andersson
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Oxidation States ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Term Storage ◽  
Geologic Media

The long-lived actinides and their daughter products largely dominate the biological hazards from spent nuclear fuel already from some 300 years after the discharge from the reactor and onwards . Therefore it is essential to make reliable assessments of the geochemistry of these elements in any concept for long-term storage of spent fuel or reprocessing waste, etc.

scholarly journals Recycling versus Long-Term Storage of Nuclear Fuel: Economic Factors

2013 ◽  
Vol 2013 ◽  
pp. 1-7 ◽  
Author(s):  
B. Yolanda Moratilla Soria ◽  
Maria Uris Mas ◽  
Mathilde Estadieu ◽  
Ainhoa Villar Lejarreta ◽  
David Echevarria-López
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Economic Factors ◽  
International Studies ◽  
Closed Cycle ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Open Cycle ◽  
Term Storage

The objective of the present study is to compare the associated costs of long-term storage of spent nuclear fuel—open cycle strategy—with the associated cost of reprocessing and recycling strategy of spent fuel—closed cycle strategy—based on the current international studies. The analysis presents cost trends for both strategies. Also, to point out the fact that the total cost of spent nuclear fuel management (open cycle) is impossible to establish at present, while the related costs of the closed cycle are stable and known, averting uncertainties.

Preparing the Spent Nuclear Fuel for Long-Term Storage and Final Disposal by Encapsulating It Into Glass-Ceramic Waste Forms

2001 ◽  
Author(s):  
Sergey Yu. Sayenko ◽  
G. A. Kholomeyev ◽  
B. A. Shilyaev ◽  
A. V. Pilipenko ◽  
E. P. Shevyakova ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Glass Ceramic ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Final Disposal ◽  
Ceramic Composition ◽  
Long Term Storage ◽  
Term Storage

Abstract This paper describes the research work carried out at the NSC KIPT to develop and apply a final waste form in the shape of a monolithic solid block for the containment of spent nuclear fuel. To prepare radioactive waste for long-term storage and final deep geological disposal, investigations into the development of methods of immobilizing HLW simulators in protective solid matrices are being conducted at the NSC KIPT. For RBMK spent nuclear fuel it is proposed and justified to encapsulate the spent fuel bundles into monolithic protective blocks, produced with the help of hot isostatic pressing (HIP) of powder materials. In accordance with this approach, as a material for the protective block made up of the glass-ceramic composition prepared by sintering at isostatic pressure, the powder mixture of such natural rocks as granite and clay has been chosen. Concept approach and characterization of waste form, technological operations of manufacturing and performance assessment are presented. The container with spent fuel for long-term storage and final disposal presents a three barrier protective system: ceramic fuel UO2 in cladding tube, material of the glass-ceramic block, material of the sealed metal capsule. Investigations showed that the produced glass-ceramic material is characterized by high stability of chemical and phase compositions, high resistance in water medium, low porosity (compared with the porosity of natural basalt). With the help of mathematical calculations it was shown that the absorbed dose of immobilizing material by RBMK spent fuel irradiation for 1000 years of storage in the geological disposal after 10 years of preliminary cooling will be ∼ 3.108 Gy, that is 2–3 orders of magnitude less than the values corresponding to preserving radiation resistance and functional parameters of glasses and ceramics. The average value of velocity of linear corrosion in water medium of the protective layer made up of the glass-ceramic composition determined experimentally makes up ∼ 15 mm per year. This allows to use glass-ceramic compositions effectively as an engineering barrier in the system of spent fuel geological disposal and to increase the lifetime of the waste container, in particular, up to 3000 years with the layer thickness ∼ 40 mm. The possible release of radionuclides from the waste container during its interim storage in the open air (near-surface storage) is estimated. The calculations are made by taking into account the possible increase of coefficients of radionuclide diffusion from 10−16 to 10−14 m2/c as a result of spent fuel radiation affecting the protective layer. The obtained results showed that the protective barrier (about 40 mm) at the base of the glass-ceramic composition, ensures reliable isolation from the environment against the release of radionuclides from the controlled near-surface long-term storage far up to 1000 years. The relatively limited release of radionuclides will make up about 1% for the period of more than 400 years, and 10% - in 1000 years. During this period of time, the radionuclides 90Sr and 137Cs will completely turn into stable 90Zr and 137Ba and the decay of many transuranium elements will occur. The results from laboratory scale experiments, tests and calculations carried out so far, show that the proposed glass-ceramic materials may be used as basic materials for manufacturing the monolithic protective block in which the spent fuel elements will be embedded with the aim of further long-term storage or final disposal.

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 Corrosion tests in water of fuel elements irradiated in the world’s first NPP reactor

2019 ◽  
Vol 5 (4) ◽  
pp. 337-343
Author(s):  
Sergey N. Ivanov ◽  
Sergey I. Porollo ◽  
Yury D. Baranaev ◽  
Vladimir F. Timofeev ◽  
Yury V. Kharizomenov
Keyword(s):  
Dose Rate ◽  
Water Samples ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Fuel Composition ◽  
Spent Fuel ◽  
Artificial Defect ◽  
Corrosion Tests ◽  
Long Term Storage ◽  
Fuel Elements

Spent nuclear fuel (SNF) storage in reactor spent fuel pools (SFP) is one of the crucial stages of SNF management technology: it requires special measures to ensure nuclear and radiation safety. During long-term storage in water-filled SFPs, leak-tight canisters in which SFAs are usually placed can become unsealed, which will result in the development of corrosion processes in the fuel element (FE) claddings. We studied fragments of spent fuel elements of the AM reactor of the World’s First NPP during their long exposure in the aqueous medium. The aim of the study was to obtain experimental data on the corrosion changes in the FE claddings and fuel composition during storage as well as on the release of radioactive fission products from them. For the study, a laboratory facility for exposing fuel elements in the water was developed and experimental fragments of fuel elements were made. The study was carried out in the hot chamber of the SSC RF-IPPE. The change in the activity of the water was estimated by the γ-dose rate from the selected water sample. The diameter measurements and metallographic studies were carried out in various sections of FE fragments. Corrosion tests were carried out on fragments of spent fuel elements of the AM reactor of the World’s First NPP that were stored for a long time (more than 50 years – FEs with U-Mo fuel and ~ 20 years – FEs with UO2 fuel) using standard technology – first in SFP canisters filled with water and then in dry canisters in the air. Placing the fuel elements in the water did not lead to through damage to the FE claddings and a significant change in the size (diameter) of the outer cladding. Metallographic studies of the FE fragments after the corrosion tests showed the presence of intergranular and local frontal corrosion on the surface of the claddings, the depth of which exceeded the depth of the cladding corrosion defects before testing. The rate of radionuclide release from the fuel composition was estimated by the γ-dose rate of water samples taken from the glasses with FE fragments. Throughout the test period, the dose rate of water samples from the glasses with defect-free FEs remained constant. The dose rate from water samples taken from the glasses with the FE fragments with an artificial defect grew during storage.

scholarly journals Problems of radiation safety calculations related to spent nuclear fuel transport casks

2020 ◽  
Vol 6 (1) ◽  
pp. 43-47
Author(s):  
Artem V. Sobolev ◽  
Pavel A. Danilov
Keyword(s):  
Gamma Radiation ◽  
Effective Dose ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Radiation Safety ◽  
Spent Fuel ◽  
Preliminary Assessment ◽  
Radiation Sources ◽  
Fuel Assemblies ◽  
Long Time

The paper discusses the stages of calculating the radiation safety of spent nuclear fuel (SNF) transport packages, in particular, transport casks and some related problems. The problem of describing the source of neutrons and gamma radiation of spent nuclear fuel is shown. For individual designs of fuel assemblies, data are given on isotopes that make the main contribution to the neutron source as well as on gamma rays in nuclear fuel material and structural materials. The authors emphasize the necessity of analyzing the influence of the initial spent fuel parameters on the formation of the radiation spectrum and, therefore, on the radiation situation around the transport casks. Consideration is given to the problem of assessing the attenuation of gamma radiation in calculating protection analytically and using software. Due to the ambiguity of the position of the zone with the highest effective dose value on the SNF transport cask surface, it is indicated that preliminary estimates are required to take into account all radiation sources and their nonuniformities. All the problems presented in the paper are currently being solved by means of rather complex and voluminous calculations that take a long time. In order to be able to conduct a preliminary assessment of the radiation situation around the transport casks, the authors propose to create a methodology that will determine the type of interrelations between the maximum effective dose and input parameters, such as fuel burnup, decay, fuel composition, protection material in the SNF transport cask, etc. This methodology will make it possible to improve the efficiency of the process of designing the SNF transport casks, avoid possible design errors and, in particular, when used as intended, resolve the issue of the SNF cask loading configuration.

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