Gateway for:

Member Countries

Fuel Materials with Non-Uranium Diluent

#1193


Development of Fabrication Technology and Studies into the Properties of Fuel Materials Containing Skeleton-Type Inert Diluent (Without U-238) to Utilize Weapon Grade and Civil Pu as well as Minor Actinides (Np, Am, Cm, etc.) in Power Reactors With Improv

Tech Area / Field

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

Status
3 Approved without Funding

Registration date
14.01.1998

Leading Institute
All-Russian Scientific Research Institute of Non-Organic Materials named after A. Bochvar, Russia, Moscow

Supporting institutes

  • FEI (IPPE), Russia, Kaluga reg., Obninsk\nSiberian Chemical Combine, Russia, Tomsk reg., Seversk

Collaborators

  • European Commission / Joint Research Center / Institute for Transuranium Elements, Germany, Karlsruhe\nCEA / DRN / DER / CEN Cadarache, France, Cadarache

Project summary

Fast reactors with specialized cores seem promising for transmutation of long-lived high activity level waste containing minor actinides (Np, Am, Cm) and for utilization of Pu the weapon's grade one included. Reactors of this type are distinguished from the traditional ones by the new fuel type not containing uranium-238, which is the main raw material for generating Pu. This places specific requirements on the fuel material compositions, the fissionable components of which are to be disposed in a carrier-material, the so-called inert diluent that weakly affects the physics of the reactor core.

One of the options of this fuel may be a composite material based on a porous skeleton made of refractory high temperature compounds, specifically, Zr carbide (nitride) or its solid solutions with heavy nuclides incorporated into pores.

Prior to filling with a fissionable substance it is advisable to work the porous Zr carbide (nitride) skeleton (the porosity in the range of 20-80%) to reach the specified diameter and length and impregnate with solutions that in the process of thermal destruction form oxides of heavy metals. This procedure results within core pores in the formation of mixed oxides of nuclides that do not form a solution in a melt (e.g., Np+Am).

The attention is attracted by the feasibility of profiling the content of a fissionable constituent over the radius and along the core length; in this case the accurate and fine dosage of a fuel constituent may be provided by the process parameters.

The availability of a high melting high thermal conductivity Zr carbide (nitride) skeleton increases the reactor safety at an emergency temperature rise within a fuel assembly.

The choice of the ZrC (ZrN) skeleton porosity and the extent of filling pores with fissionable isotopes are dictated by the reactor core specificity conditioned by the safety parameters such as sodium void effect of reactivity (SVER) related to a loss of sodium from a core as a result of an accident. With such a loss of a coolant the reactor is to change to a subcritical condition which necessitates detailed physics studies of reactors having fuel of this type.

To determine the sizes of fuel cores and their specific thermal conductivity the thermo-physical properties of porous materials are to be determined in the greater details as well as the behavior of fissionable nuclide oxides in a porous matrix is to be investigated at different temperatures.

The final objectives of the suggested project are:


- determine the main thermophysical properties of the composite fuel material under development to select the process parameters and fabricate specimens of these materials to be experimentally irradiated in BOR-60 and BN-600 reactors (in - pile and PIE are scheduled for the next project);
- define the initial data for subsequent fuel cores development based on the studied composite fuel to burn out Pu and minor actinides;
- design studies the feasibility of improving the fuel core performance via the profiled disposition of PuO2 within the core volume;
- complete design studies to optimize and validate the reactor cores loaded with fuels not containing U-238 but having the chosen porous matrix;
- generalize the acquired calculated, experimental and scientific data, issue the final scientific and technical report and recommendations to further continue the work.


Back