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.

The Development and Innovation of Spent Fuel Reprocessing in Fuel Cycle

2010 ◽  
Author(s):  
Guang Jun Chen ◽  
Yu Lin Cui ◽  
Guo Guo Zhang ◽  
Hong Jun Yao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Fuel Cycle ◽  
Fission Products ◽  
Purex Process ◽  
Spent Fuel ◽  
Spent Fuel Reprocessing ◽  
Long Life ◽  
Fuel Reprocessing

With an increased population and an increasing demand for power, nuclear power has attracted an increasing attention and mass nuclear power plant have been built in different countries in the past several decades. At present, about ten thousands ton spent fuels are discharged from nuclear power plant every year and the estimated capacity will approximately add up to 5×105 ton. Therefore, spent fuel reprocessing, by which the co-extraction and separation as well as purification of Uranium and Plutonium could be realized and ensure the recycle of uranium resources and the management of nuclear waste, is a vital step in nuclear fuel cycle including two major strategies, i.e. once-through cycle and closed fuel cycle. It is worth noting that the utilization of MOX fuel made by plutonium mixed with uranium has been successfully achieved in thermal reactor. Fortunately, the middle experiment plant of china spent fuel reprocessing has been being debugged and will be operated completely in future two years. Various reprocessing schemes have been proposed for the extraction of actinides from fission products and other elements presented in spent nuclear fuel. However, after numerous studies of alternate reprocessing methods and intensive searches for better solvents, the PUREX process remains the prime reprocessing method for spent nuclear fuels throughout the world. High burning and strong radioactive spent fuel resulting from the evolution of various reactors drive the development of the advanced PUREX technology, which emphasizes the separation of neptunium and technetium besides the separation of the Uranium and Plutonium from the majority of highly active fission products. In addition, through Partitioning and Transmutation method, some benefits such as segregating the actinides and long life fission products from the high level waste can be obtained. The GANEX process exploited by CEA, which roots in COEX process belonged to advanced PUREX process, considers the separation of the actinides and long life fission products. The study on the pyro-chemical processing such as the method of electro-deposition from molten salts has still not replaced the traditional PUREX process due to various reasons. In conclusion, the future PUREX process will focus on the modified process including predigesting the technical flowsheets and reducing reprocessing costs and using salt-less reagent in order to minimize the waste production.

Atmospheric release of volatile fission products from operation of nuclear power reactors and spent fuel reprocessing facilities and prospects for extracting the products

1978 ◽  
Vol 44 (2) ◽  
pp. 153-157
Author(s):  
B. Ya. Galkin ◽  
L. I. Gedeonov ◽  
N. N. Demidovich ◽  
R. I. Lyubtsev ◽  
I. V. Petryanov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Fission Products ◽  
Spent Fuel ◽  
Spent Fuel Reprocessing ◽  
Fuel Reprocessing ◽  
Atmospheric Release

Reprocessing of Spent Nuclear Fuel: A Policy Analysis

2006 ◽  
Author(s):  
Todd P. Lagus
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Energy Content ◽  
The United States ◽  
Spent Fuel ◽  
Cycle System ◽  
Spent Fuel Reprocessing ◽  
The Cost ◽  
Fuel Reprocessing

Nuclear spent fuel reprocessing has lately reemerged as a subject of debate in the energy policy world. Since a 1977 Presidential Directive which deferred reprocessing of spent nuclear fuel (SNF), the United States has utilized a “once through” or “open cycle” system of nuclear fuel processing, which leaves most of the energy content in uranium unused. Current reprocessing technology increases the cost of nuclear electricity while only offering limited storage benefits. Advanced technologies have the potential to increase proliferation resistance, the need for more geologic repositories, and allow the United States to regain an international presence in reprocessing technology. The United States should not immediately engage in spent fuel reprocessing, but should begin aggressive research and development for new reprocessing technologies.

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.

Centralization of Canada’s Spent Nuclear Fuel

2013 ◽  
Author(s):  
Krista Nicholson ◽  
John McDonald ◽  
Shona Draper ◽  
Brian M. Ikeda ◽  
Igor Pioro
Keyword(s):  
Nuclear Fuel ◽  
Site Selection ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Selection Process ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Candu Reactors ◽  
High Level

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel. The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method. The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.

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.

The VR Simulation for Spent Fuel Reprocessing

2008 ◽  
Author(s):  
Bo Zhu ◽  
Yanhua Yang
Keyword(s):  
Nuclear Power ◽  
Spent Fuel ◽  
Nuclear Power Reactor ◽  
Operation Process ◽  
Fast Development ◽  
Spent Fuel Reprocessing ◽  
Visualization Technology ◽  
Near Future ◽  
And Training ◽  
Fuel Reprocessing

The spent fuel disposal is a very important step during the recycling of nuclear fuel and the operation process is very complex and dangerous. With the fast development of nuclear power reactor in china, more and more spent fuel will occur and should be treated in the near future. For researching the spent fuel management and training the operators, we have been developing the virtual reality (VR) Simulation System for spent fuel reprocessing (VRSS) aimed at demonstrating and analyzing the treatment process. The VRSS is based the VR and visualization technology. Combined with the regulation of radiation protection, the integrated virtual operate environment is developed to enable user navigation the working place, and obtain target information “it can demonstrate the structure and component of equipment, it also can illustrate the disassembly craft transform principle and the operation way of extraction facility. It even can prompt the user how to operate when spent fuel handling accident occur.

Measurement of Atmospheric Sea Salt Concentration in the Dry Storage Facility of the Spent Nuclear Fuel

2006 ◽  
Author(s):  
Masumi Wataru ◽  
Hisashi Kato ◽  
Satoshi Kudo ◽  
Naoko Oshima ◽  
Koji Wada ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Sea Salt ◽  
Storage Facility ◽  
Japan Society ◽  
Spent Fuel ◽  
Dry Storage ◽  
Storage Methods ◽  
Interim Storage

Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing. There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack. Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking (SCC) of the canister is not sufficient. Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility. Air including the sea salt particles goes into the dry storage facility (Figure 1). Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding. The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building. For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack.

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).

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