Graphite Isotope Ratio Method
Verification of Graphite Isotope Ratio Method to Restore Neutron Local Fluence in Reactor Core with Graphite Moderator by the Example of the Reactor of the First A-Plant in Obninsk to Control Non-Proliferation of Nuclear Fissile Materials
Tech Area / Field
- FIR-ISO/Isotopes/Fission Reactors
- FIR-MOD/Modelling/Fission Reactors
- FIR-NOT/Nuclear and Other Technical Data/Fission Reactors
3 Approved without Funding
VNIIEF, Russia, N. Novgorod reg., Sarov
- FEI (IPPE), Russia, Kaluga reg., Obninsk
- Pacific Northwest National Laboratory, USA, WA, Richland
Project summaryNuclear power industry is being dynamically developed all over the world. Increasing number of countries are being involved in activities related to nuclear power plant (NPP) development. Nuclear power is closely connected with additional accumulation of fissile materials. Over the lifetime of a nuclear reactor, neutrons from the fission process do not only convert U-238 into fissile materials but also bring about changes in the elements of the structural materials of the reactor core components. A reactor total fissile material production is directly related to total neutron fluence and neutron energy distribution. Determination of the reactor physical model and lifetime neutron fluence in each cell of the reactor core containing fuel could provide an accurate estimate of a reactor total fissile material production.
Clearly, only items that remain in place throughout the reactor lifetime and that are placed sufficiently near fuel cells in a reactor core will contain the most reliable information about lifetime neutron fluence. Hence it is obvious that the most reasonable candidate is the graphite moderator that can provide measuring of neutron fluence over the lifetime. In the known operating experiences no significant portions of the graphite moderator were replaced in large production reactors. Partial replacement of the graphite moderator can be identified and considered in the estimation. Carbon can not be used as “indicator” element, that is, element that undergo transmutation in a predictable manner due to neutron irradiation, because other sources of carbon are present in the reactor core. For instance nitrogen is converted to carbon when it absorbs a neutron and some reactors use carbon dioxide as a coolant. This left only the graphite impurities as candidate indicator elements. Impurity concentrations and composition in graphite is not a constant value and vary from feedstock and manufacture’s technologies. Impurity concentrations can vary even within one graphite piece. Therefore changes in initial concentration can’t be used to measure neutron fluence. It became obvious that an accurate value of total lifetime neutron fluence could be obtained by measuring the change in the isotopic ratio of certain impurity elements only.
The initial isotopic ratios of each element are fixed by nature. In order to estimate neutron fluence at any location of the reactor core it is necessary to get a graphite sample from the place adjacent to this location and measure isotopic ratios of the indicator element. The main requirement is that the indicator element should be available in measurable quantity in the graphite sample obtained from the reactor core and its isotopic ratio could be accurately estimated. Restoring of local fluence by measured isotopic ratios requires information about neutron energy distribution and thermal fields in the reactor.
The proposed project is intended to develop computing-theoretical method of isotopic transmutation of indicator elements in graphite under neutron irradiation depending on neutron fluence and energy distribution, and to carry out experimental testing of GIRM technology to restore local neutron fluence in uranium-graphite AM reactor core at the First NPP in Obninsk. In the course of the project research will be carried out in order to identify initial composition and content of impurities in AM reactor graphite. It is necessary to select indicator elements that can be used for measuring isotopic ratios and estimation of neutron local fluence in AM reactor. The next step is to develop technology of graphite sampling and to take the required number of samples (~ 50) of reactor graphite from various locations of the core with different fluence values. It is supposed to develop method of indicator elements separation from the reactor graphite and to measure the elements isotopic ratios. It is also the task of the project to develop and/or adapt computing methods and the software for neutron fluence estimation based on the indicator elements isotopic ratios.
The project also includes a separate task to restore neutron fluence in the taken samples using detailed operation data on changes of the reactor power throughout its lifetime as well as power and composition of fuel assemblies located in the cells the graphite samples are taken from. The accuracy of such computing-experimental method of neutron fluence estimation in the samples taking into account all measuring and estimated errors will be not worse than 10%.
The final stage of the project will be concentrated on comparison of neutron local fluence in the graphite samples taken from various locations in the AM reactor core estimated using the GIRM method with that obtained using the detailed operation data. This will enable us to evaluate the GIRM method accuracy and its potential application to control fissile materials production in graphite reactors.
It is the objective of the project to test all technological chain of the GIRM method including identification of impurities in the reactor core graphite, selection of the indicator elements, measurements of isotopic ratios of the indicator elements in irradiated samples of the reactor graphite, development of physical models to restore neutron fluence based on measured isotopic ratios values.
The project will result in improvement of the methods and mastering technologies of measuring isotopic ratios of indicator elements in the reactor graphite with required accuracy. Our knowledge of types and quantity of impurities in the reactor graphite will be enhanced. Methods to estimate neutron local fluence in the uranium-graphite reactor core will be developed.
The Project results can be used to validate fissile material production in reactors with graphite moderator and can provide potential technological monitoring for fissile materials production.
Cooperation of experts from RFNC-VNIIEF developing up-to-date methods of computing of nuclear systems and isotopic analysis of chemical element micro quantities and specialist from SRS RF-PEI, who have a 50-years experience in the field of graphite reactor operation at the First NPP and conducting detailed reactor calculations will be a reliable base for successful development of the GIRM technology.
The Project will involve scientists and specialists directly related to nuclear weapons development. Thus this Project meets the ISTC objectives and goals.
At each stage of the project there will be an active exchange of information with the Partner and the work results will be discussed at joint meetings and workshops.