Gateway for:

Member Countries

Nitride Fuel for Fast Neutron Nuclear Reactors

#K-1479


Development of Technology for Producing High-Effective Nitride Fuel UN with Controlled Microstructure for Advanced Fast Neutron Nuclear Reactors

Tech Area / Field

  • FIR-FUE/Reactor Fuels and Fuel Engineering/Fission Reactors

Status
3 Approved without Funding

Registration date
12.12.2006

Leading Institute
Joint Stock Company "Ulba Metallurgical Plant", Kazakstan, Ust Kamenogorsk

Collaborators

  • Lawrence Livermore National Laboratory, USA, CA, Livermore

Project summary

This project is aimed at development of technology for producing high-effective nitride fuel UN with controlled microstructure for advanced fast neutron nuclear reactors. Developed technology will be used for producing nitride fuel UN, use of which as the main component of fuel elements in advanced energy reactors of fourth generation will allow main advantages of new reactors to be effectively realized.

At the beginning of the XXI century many countries took programs on accelerated development of nuclear power engineering based on two main courses: intensification of functioning atomic power reactors usage and construction of nuclear reactors of fourth generation named breeders that enable to increase nuclear fuel breeding and burn off long-lived isotopes (actinides) obtained out of irradiated fuel of thermal reactors. U-238 and Th-232 can be involved in fuel cycle since their natural stores are much higher than the U-235 stores (the main fuel for thermal reactors). Besides, dump uranium remained after nuclear fuel enrichment on U-235 can be used, and the problem of Plutonium utilization, which is the main component of nuclear weapons, due to its burning out in the core of breeder can be solved. Exploitation of fast neutron reactors allows both ecological problems and problems of power substitution at the expense of growth of electric power portion derivable from NPP to be solved under existing conditions of decline of gas and oil production. It will be possible if the average depth of nuclear fuel burning out in fast neutron reactors is 100÷150 MWd/kgU, i.e. it will be 2,5-3 times higher than in thermal reactors. In this connection, special requirements to physical and technical properties of nuclear fuel will be demanded to solve the given tasks. Nuclear fuel for new generation reactors must be of high radiation resistance, geometrics stability, high level of plasticity and thermal conductivity since heat-and-power engineering characteristics of fourth generation reactors will be much higher than characteristics of functioning today power reactors. Since such qualities do not inhere to uranium dioxide ceramic nuclear fuel for thermal neutron reactors thus nitride fuel UN can be the most feasible fuel for new reactors. Nitride fuel UN has a number of considerable advantages such as firm and steady bonding with melting temperature of 26300C, dissociation pressure at 17000C does not exceed 1,33∙10-4 Рa., theoretical density is 14.32 g/cm3 and high thermal conductivity is 20 W/m∙K.

There is interest to dense fuels, including nitride fuel as the most promising, for new reactors. Dense nitride fuel UN is one of the less studied among the other types of nuclear fuel. It is important to make a choice of nitride fuel production process method to obtain necessary characteristics and fuel modification at macro- and micro level during its production. Uranium Production Department of Joint Stock Company “Ulba Metallurgical Plant” (UP JSC “UMP”) produces different types of uranium products more than 30 years. Today, these products include uranium dioxide ceramic powders of nuclear grade, U3O8, ceramic nuclear fuel pellets for VVER and RBMK reactors. As a part of JSC “UMP” Tantalum Production Department (TP) also has unique highly scientific technologies enables to produce high capacity tantalum powder, and tantalum and niobium production with high level of purity. Technology of magnesium thermal reduction of heavy metals is used at the plant. Such technology can be used during production process of pure metal uranium out of uranium containing materials including uranium tetrafluoride, one of the main uranium containing material used during nitride fuel production process. In the 70th years of the XX century a scheme of producing uranium tetrafluoride by combined method has been developed and implemented at JSC “UMP”. This method includes combination of uranium tetrafluoride precipitation out of mixed salt-fluoride solution with semi-wet method of producing uranium tetrafluoride. UP JSC “UMP” has a wide experience on optimization of physical and technical properties of nuclear fuel that provides for the highest effectiveness of its operation in reactors. First of all it concerns fuel investigations at micro level, which are carried out for regulation and optimisation of grain size, open and inside porosity, uniformity of microstructure and plastic properties of fuel. JSC “UMP” has a wide experience and scientific and technical potential for development of economically efficient technology during production process of high effective nitride fuel UN. Guarantor of project implementation is high quality of JSC “UMP” production based on certified quality assurance system, availability of universal fluid-extraction technology of raw material and technical wastes recycling followed by production process of uranium containing product with high level of nuclear purity with the given characteristics, technological flexibility enables to re-form quickly towards new production output. Furthermore, there is scientific potential that allows investigations, development and implementation of new technology at high level to be carried out.

This project is aimed at development of economically feasible process for producing effective nitride fuel UN for advanced nuclear reactors of fourth generation. Nitride fuel UN obtained will have all necessary physical and technical characteristics in compliance with condition of exploitation in advanced fast neutron nuclear reactors. Nitride fuel will be of high level of plasticity and geometrics stability during irradiation, high tendency to retention of gaseous products of fission due to optimisation of size grain at the level of 25-30 μm. The most effective methods and modes of producing uranium tetrafluoride, the main raw material for UN fuel, will be realized by suggested technology. There will be considered the most effective and economically feasible methods of producing metal uranium with high level of nuclear purity such as: magnesium thermal reduction of uranium tetrafluoride up to metal uranium; carrying out metal uranium hydrogenation and uranium hydride nitration during one production operation; producing fuel pellets (fuel density is not less than 95% from theoretical density) out of uranium nitride by powder metallurgy methods. At the first stage of project implementation experimental technological site for realization of the main procedures of worked out technology will be created. The first task of the project will be the solution of mentioned above problems. The second task will be concerned with regulation and optimization of microstructure of fuel pellets out of uranium nitride. At this stage the following steps will be realized: correction of developed technology modes and searching of additional steps enable to regulate characteristics of nitride fuel microstructure purposefully for its optimisation; guarantee microstructure with high level of pore and grain homogeneity, minimization of open and inside porosity, geometrics stability of pellets during high temperature annealing, producing grain size at the level of 25÷30 m, required for decreasing possibility of gaseous products of fission penetration under fuel element shell. At the third stage of project implementation method of regulation of nitride fuel plastic properties will be developed. Invented procedure will allow nitride fuel plasticity to be maximum increased that is necessary in condition of its high burning out. Given task will be aimed at activation of relaxation possibilities of fuel matrix under irradiation at the expense of its microalloying and correction of pore and grain fuel microstructure.

At all stages of project implementation the close interaction with Collaborator is implied. It includes information exchange, cross checking of obtained results, testing of technology flow sheet developed during project implementation, help in visiting international meetings on project theme, conduction of mutual meetings and work seminars.


Back

The International Science and Technology Center (ISTC) is an intergovernmental organization connecting scientists from Kazakhstan, Armenia, Tajikistan, Kyrgyzstan, and Georgia with their peers and research organizations in the EU, Japan, Republic of Korea, Norway and the United States.

 

ISTC facilitates international science projects and assists the global scientific and business community to source and engage with CIS and Georgian institutes that develop or possess an excellence of scientific know-how.

Promotional Material

Значимы проект

See ISTC's new Promotional video view