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Minor Actinide Transmutation in Nitrides Fuel Environment

#2680


MATINE - Study of Minor Actinide Transmutation in Nitrides: Modelling and Measurements of Out-of-pile Properties

Tech Area / Field

  • FIR-MAT/Materials and Materials Conversion/Fission Reactors
  • FIR-NSS/Nuclear Safety and Safeguarding/Fission Reactors

Status
8 Project completed

Registration date
27.01.2003

Completion date
27.03.2007

Senior Project Manager
Tocheny L V

Leading Institute
FEI (IPPE), Russia, Kaluga reg., Obninsk

Supporting institutes

  • NIIAR (Atomic Reactors), Russia, Ulianovsk reg., Dimitrovgrad\nAll-Russian Scientific Research Institute of Non-Organic Materials named after A. Bochvar, Russia, Moscow

Collaborators

  • Kungl Tekniska Hogskolan / Stockholm Physics Centre, Sweden, Stockholm\nCEA / Direction de l'Energie Nucleaire (DEN), France, Saclay

Project summary

Management of radioactive waste (RW) in an environmentally safe manner is an important issue being addressed by all the countries developing a nuclear industry. In many countries it has become a serious political issue attracting intense critical attention of general public.

Neutron transmutation of long lived radioactive minor actinides (MA – neptunium, americium, curium) by the fission process, producing energy and simultaneously turning them into shorter lived nuclides, is being intensely analyzed and discussed as a possible solution of this problem. Several possibilities for the transmutation of long lived nuclides by nuclear reactions have been suggested. Recently there has been a renewed interest in the accelerator driven transmutation schemes, which seems to show good promise. Any type of reactor can be arranged to be in a subcritical state. Different fuel and fuel cycle concepts have been investigated. It has been well recognized that the minor actinides Np, Am, Cm can be effectively transmuted in a fast neutron flux, since the ratio of their fission to capture cross-section is quite large for fast neutrons. Since the minor actinide fission efficiently mostly in a fast neutrons spectrum, the present scheme in many countries is based on generating an intense fast neutron flux, with the options to consider sodium, lead-bismuth or helium coolants with a solid fuel (based on sodium-cooled fast reactor technology).

Transmutation and incineration are innovative options in the management and disposal of fission products and actinides. In order to improve the efficiency of these processes, materials inert to neutron activation are being considered as alternatives to UO2 as a support material, as the latter generates actinides during irradiation. These inert matrices are selected on the basis of their thermal conductivity, melting point, compatibility with the reactor coolant and their resistance to damage by neutrons, fission products and -decay. ZrN is one material, which meets these criteria.

Inert matrix nitride fuels (An,Zr)N, An - Pu, Am, Cm, Np, are under investigation now in Japan, Europe, USA, Russia. The work is devoted to the development of fabrication processes, materials properties determination and implementation of fuel irradiation into experimental reactors.

In the European CONFIRM project, computer simulation of uranium free nitride fuel irradiation up to about 20% burn-up will be made to optimize pin and pellet designs. Other computations will be performed especially concerning the safety evaluation of nitride fuel. Plutonium zirconium nitride (Pu,Zr)N and americium zirconium nitride pellets will be fabricated and their conductivity and stability at high temperature will be measured in the frame of CONFIRM. (Pu,Zr)N pins of optimized design will be fabricated and irradiated in the Studsvik reactor. Fabrication of nitrides (Pu,Zr)N, (Pu,Y)N, (Am,Zr)N, (Am,Y)N was carried out by JAERI. Solid solutions of nitrides were obtained. The irradiation of pins with (Pu,Zr)N is under preparation for the JMTR reactor. ANL plans to fabricate, characterize and irradiate at the ATR reactor several fuel types, including (Pu,Np,Am,Zr)N. In the frame of Russian-French collaborative experiment BORA-BORA the irradiation of two fuel pins with (Pu,Zr)N is under way now at the BOR-60 reactor.

Nevertheless up to now there is little experience on uranium free nitride fuel as well as on in-reactor behavior of MA fuels. Hence there is the necessity and importance of international collaboration in this field.

The objective of the Project is to measure some basic properties of (Pu,Zr)N fuel and to carry out the comprehensive calculation modelling of in-reactor behavior of uranium free nitride fuel in the fast spectrum of ADS in order to identify the optimum fuel pin design that would perform well under irradiation in a fast neutron spectrum.

The Project will pide into four tasks. Management of these tasks will be performed by the IPPE, which will be the Institution responsible for monitoring progress of the tasks, ensuring adequate documentation and acting as task integrator. Besides, IPPE will be in evolved in implementing of two tasks.

In the first task compilation of literature data from existing nitride fuel investigations in – and outside of Russia will be done in addition with compilation of data on MA fuels irradiation behavior. The objective of this task is to perform input data for KONDOR and VIKOND codes. IPPE, RIAR and VNIINM will perform this task.

The second task will cover fabrication of (Pu,Zr)N pellets and measurement of its thermal conductivity, creep rates and high temperature stability. The task will be performed by VNIINM.

The third task will involve modelling of the performance of (Pu,Am,Zr)N fuel under irradiation up to high burn-up in fast neutron spectrum. The fuel form would be either pellets or vibropacked granulates/microspheres. In the case of pellets, modelling is to be made for both helium, sodium and lead-bismuth bonded pins to compare their relative performance and to select the optimum fuel pin design. Modelling will be carried out by IPPE using KONDOR code. For vibropacked fuel modelling is to be made for helium-bonded pins only. RIAR will perform the calculation using VIKOND code.

The forth task will be devoted to technical – economical assessments of the feasibility of fabricating nitride fuels with atomic fraction of ZrN inert matrix up to 6010 %, containing up to 10 atomic percent curium, at RIAR site. RIAR will perform the task.

Similar work is being performed also by countries of European Union (CONFIRM Program), which are the foreign collaborators of the Project. The proposed work in Russia is likely to greatly benefit from a comparison of results obtained at EU, verification of codes, and accumulation of representative experimental data on properties of uranium free nitrides.

CEA (France), PSI (Switzerland) and KTH (Sweden) will participate in the analysis of the comprehensive research results to compare with own results and to make a decision on best solution for ADS experimental fuel pin to be tested in fast reactors.


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