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Minor Actinide Transmutation in Inert Matrices Fuels

#3608


MATINE 2- Study of Minor Actinide Transmutation in INErt Matrices Fuels: Modeling, Fabrication and Measurements of Out-of-Pile Properties

Tech Area / Field

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

Status
3 Approved without Funding

Registration date
29.08.2006

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

Supporting institutes

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

Collaborators

  • Nexia Solutions Ltd, UK, Preston\nCEA / Direction de l'Energie Nucleaire (DEN), France, Saclay\nRoyal Institute of Technology / Nuclear and Reactor Physics, Sweden, Stockholm

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. 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
α
-decay.

Inert matrix fuels 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. Fabrication of nitrides (Pu,Zr)N, (Pu,Y)N, (Am,Zr)N, (Am,Y)N was carried out by JAERI. The PIE of pins with (Pu,Zr)N are carried out after its irradiation in the JMTR reactor. The FUTURIX-FTA (ЕU-USA-Japan) Program is currently under preparation. It covers the irradiation of different innovative fuels with MA in the PHENIX reactor, including inert matrices fuels. In the frame of FUTUR&CONFIRM EC Programs the investigations of (Pu,Zr)N, PuAmO2-MgO, (PuAmZr)O2 are carried out. In the frame of Russian-French collaborative BORA-BORA Program the irradiation of four fuel pins with ZrN and MgO matrices in the BOR-60 reactor have been completed, PIE are under way. In the frame of ISTC Project #2680 (MATINE) the calculation modeling of (Pu,Am,Cm,Zr)N fuel performance under irradiation in the fast neutron spectrum of ADS (EC Program) up to high burn-up is done. Two fuel types – vipac and pelleted – have been considered. The measurements of thermo-physical properties of (Pu,Zr)N laboratory samples are carried out (thermal conductivity, high temperature creep, high thermal stability). Besides the technical – economical assessments of the feasibility of fabricating nitride fuels containing up to 10 atomic percent curium are performed.

The results of the calculations of nitrides in-reactor behavior, received in the ISTC MATINE Project (#2680), show that possible problems may be with fuel performance at high-burn-ups in ADS core, particularly for He-bonded fuel pins, due to fuel-cladding mechanical interaction (FCMI). Since helium bonding is preferable in the case of aqueous reprocessing options, alternative fuel compositions are of interest. One of the possible options is the composite material based on porous matrix made of refractory compounds, Zr carbide (nitride) or its solid solutions, with heavy nuclides incorporated into the pores. Prior to the filling by fission fraction the porous Zr carbide (nitride) matrix of required diameter and length is fabricated. Matrix porosity is in the range of 20-80%. It is impregnated by the solutions that under the thermal destruction process form the heavy metals oxides. The oxide layer covers the pore surface. This procedure results in the formation of nuclides mixed oxides even of those that do not form solid solution in a melt (e.g., Np+Am).

The objective of the Project is to identify the optimum fuel pin design that would perform well under irradiation in ADS fast neutron spectrum. To accomplish the objective the fabrication and properties measurements of PuO2-ZrC samples with porous matrices will be done, as well as optimization of neutronics and thermo-hydraulics parameters of pin irradiation with (PuAmCmO2 + ZrC) fuel with ZrС<40 w%, in ADS core and the calculation modeling of in-reactor behavior of such fuel pins. In addition, on the base of the investigations carried out in the frame of ISTC Project #2680 (MATINE), the fabrication and examination of (PuAmZr)N and CmN vibropacked samples will be done.

The Project will cover 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 frame of the first task the neutronic and thermo-hydraulic calculations of ADS core parameters will be carried out aiming the optimization of irradiation conditions of (PuAmCmO2 + ZrC) fuel in ADS core for providing of fuel performance and efficient actinides utilization. The calculations will be done by IPPE.

The second task will involve modeling of (Pu,Am,Cm)O2 - ZrC fuel performance under irradiation up to high burn-up in ADS core. The calculations will be carried out for helium-bonded and sodium-bismuth-bonded pins in order to compare its performance and give recommendations on the optimized option. The work will be done by IPPE.

In the third task the fabrication of PuO2-ZrC samples with ZrС<40 w%, measurement of its thermal conductivity, creep rates and mechanical properties will be done. The task will be performed by VNIINM.

The forth task will include the fabrication of the samples of vibropacked (PuAmZr)N with ZrN<40 w% and CmN fuels, its characterization and measurements of high temperature stability. RIAR will perform the task.

Similar work is being performed also by countries of EC that are the foreign collaborators of the Project. The proposed work in Russia is likely to greatly benefit from a comparison of results obtained abroad, verification of codes, and accumulation of representative experimental data on properties of uranium free nitrides.

Foreign collaborators will participate in the analysis of the results obtained in order with to make a decision on best solution for ADS experimental fuel pin for further possible test in the BOR-60 reactors. The results of the investigations carried out in the frame of Project #2680 will be taken into account as well.


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