Gateway for:

Member Countries

Safety use of Dispersive Fuel

#3119


Design-Theoretical and Experimental Research on Justification of Safe Use of Dispersive Fuel in a Nuclear Reactor Core

Tech Area / Field

  • FIR-NSS/Nuclear Safety and Safeguarding/Fission Reactors

Status
8 Project completed

Registration date
07.10.2004

Completion date
15.02.2012

Senior Project Manager
Tocheny L V

Leading Institute
VNIIEF, Russia, N. Novgorod reg., Sarov

Collaborators

  • General Atomics, USA, CA, San Diego\nAtomic Energy Canada Limited, Canada, ON, Chalk River

Project summary

One of directions of development of nuclear power engineering is bound up with the development of a high-temperature gas-cooled reactor (HTGR) with fuel based on micro fuel elements pressed in a graphite matrix.

In the given project it is offered to develop procedure for testing of various types of fuel based on micro fuel elements and to define boundaries of safe use of this fuel in a nuclear installation core.

Development of the high-temperature helium-cooled reactor technology caused development of micro fuel elements - spheres of a small diameter made from uranium dioxide - kernels with multilayer ceramic protective coating. Such fuel is capable to hold fission products in its volume under temperature up to 1600оС for a long term and to provide high burnup up to 10-15 % of heavy nuclei. This technology now is the most ecologically safe and the most economic in comparison with other types of nuclear fuel.

Rules of nuclear safety of atomic power station require that in a detail design it will be shown for design accidents bound up with fast increase of reactivity that specific energy released in fuel elements in each moment of campaign will not exceed the threshold value causing fuel element destruction and that fuel melting will be excluded, and for hypothetical accidents requirements at which excess of threshold energy of fuel element destruction is possible must be given. Now there is no such standard characteristic for HTGR fuel elements.

Specific enthalpy in HTGR fuel elements considerably exceeds the proved limit of enthalpy for fuel elements used in water-moderated tank reactors. It is necessary to note, that investigations of limiting enthalpy definition for unirradiated HTGR fuel elements with reduced-enrichment uranium fuel have been carried out on pulse reactors. Analysis of these investigations has not been completed, but nevertheless preliminary results have shown that limiting enthalpy of fuel element destruction was (2.5 - 3) kJ/g at reactor pulse duration near 0.1 s and і30 kJ/g at pulse duration more than 2 s.

In the given project it is supposed to explore boundaries of safe operation of fuel in nuclear reactor core made from micro fuel elements by carrying out of design-theoretical and experimental researches of boundaries of destruction of micro fuel element coating during irradiation of samples of micro fuel elements on a pulse nuclear reactor.

During activity under the project it is supposed to solve the following problems:

· To explore proofness of samples of micro fuel elements to nuclear heating under irradiation on the pulse nuclear reactor with different energy-releases and pulse durations from 3 up to 10 ms and from 0.5 up to 50 s and variable size of inserted energy.

· To explore degree of kernel carbidizing (quantity of uranium dioxide of the kernel reacted with carbon of the coating) varying the specified parameters. To determine quantity of formed carbon oxides.

· To determine low temperature of start of interaction of uranium dioxide with graphite under nuclear heating. To carry out comparison of found low temperature with temperature of interaction start under stove heating.

· To determine for different irradiation durations the minimal quantity of energy inserted during irradiation at which coating destruction occurs.

· To estimate by design-theoretical methods temperature of the kernel during and after a pulse, time dependence of local temperature on radius of micro fuel elements. Observationally during an irradiation. To measure temperature of a micro fuel element surface during irradiation.

· In experiments it is supposed to use additional samples made from homogeneous mixture of a fine-dyspersated powder of uranium dioxide with graphite in order to estimate influence of a granular size of uranium dioxide in a mix with graphite on reaction rate of components interaction under irradiation on the pulse nuclear reactor.

· The irradiation complex and testing techniques which can be used for research of other types of micro fuel elements will be modernized during activity under the project.

The main result of project carrying out will be obtaining of the value of maximal enthalpy of a micro fuel element average on volume. This value characterizes boundaries of safe operation of dispersion fuel in a nuclear reactor core.

This project is assumed to be the first step of work in this direction. The goal of the next stage, which goes beyond this Project, will be testing of fresh and burnt-up fuel based on micro fuel elements containing plutonium.


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