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Analysis of Nuclear Spent Fuel

#3769


Methods for Non-Destructive Analysis of Fuel in Chemical Reprocessing of Nuclear Power-Plant Spent Fuel

Tech Area / Field

  • FIR-FUC/Fuel Cycle/Fission Reactors
  • FIR-FUE/Reactor Fuels and Fuel Engineering/Fission Reactors
  • FIR-INS/Nuclear Instrumentation/Fission Reactors
  • FIR-NSS/Nuclear Safety and Safeguarding/Fission Reactors

Status
3 Approved without Funding

Registration date
24.05.2007

Leading Institute
Khlopin Radium Institute, Russia, St Petersburg

Collaborators

  • Idaho State University / Idaho Accelerator Center, USA, ID, Pocatello\nUniversity of Nevada Las Vegas / Department of Mechanical Engineering, USA, NM, Las Vegas

Project summary

Project aim: development and demonstration of the real-time method for estimation of content of uranium and plutonium isotopes in imitators of dissolved irradiated fuel (or spent fuel) of NPP.

High importance of the work follows from the problem of non-proliferation. Components of irradiated fuel are potentially very attractive for those, who intend to create a nuclear explosive. Thus, during the recycling process of irradiated fuel these materials should be under strict control to prevent their leakage. However, existing technical means cannot provide real-time control. The adequate non-destructive control devices should estimate plutonium quantity accurate within 1%. Yet such devices do not exist now.

This problem is important as well considering necessity of providing criticality-safe chemical processing of chemical recycling of irradiated fuel.

At present there are no non-destructive methods of direct measurement of plutonium and uranium isotopes in irradiated fuel. Existing non-destructive methods estimate content of plutonium or uranium isotopes via secondary characteristics which lead to ambiguity and unreliability of results. Precise measurement of isotope concentration in solution is necessary to define correctly the criticality safety of devices, which are used in chemical recycling of irradiated fuel. Mistakes while defining criticality can cause serious accidents.

It is very important to control uranium and plutonium content in solution of irradiated fuel in real-time mode, in particular. Only such control will give the opportunity to avoid plutonium leakage and to prevent possible criticality accidents due to incorrect measurement of fissionable material amount in devices, which are used during the chemical recycling process.

To fulfill these tasks we offer to use direct method to measure concentration of components in dissolved irradiated fuel using neutron “slowing-down time” spectrometer.

The main idea of this method is that next to the dissolved irradiated fuel we place a lead or graphite moderator of hundreds of liters in volume. Moderator is irradiated by short-term (about 10 ns) homoenergetic flux of fast neutrons, produced by pulsed neutron generator. During tens of microseconds neutrons gradually slow down, but still stay homoenergetic. Moderated neutrons reach the irradiated fuel solution causing fission of it's isotopes. Threshold detector based on fission chambers detects fission neutrons. Since the neutron flux slows down to the energy lower than one of the threshold (about 1 MeV), fission chambers become insensitive to initial neutron flux and, thus, detect only fission neutrons. Dependences of fission cross-sections of different isotopes on energy vary. Accordingly, dependence of counting rate of fission chambers on slowing-down time of neutrons (on condition that in every point of time they are homoenergetic), or in other words response of the device to each isotope will be unique. Response of the device to dissolved irradiated fuel material is, in fact, combination of responses to isotopes in the solution, that makes possible to separate total response into components, that will correspond with different isotopes and will give the opportunity to define their absolute contents in the sample of dissolved irradiated fuel.

Specialists of V.G. Khlopin Radium Institute have necessary professional skills and experience in development of systems for identification of radioactive materials using active neutron methods. Earlier such methods were used in the framework of several projects: IAEA ¹ 12600, NATO ¹ CP NR SFR 981003. Chemical and technological equipment of the Institute allow carrying out tests of the proposed method using imitators of irradiated fuel solutions. Specialists of V.G. Khlopin Radium Institute are experienced in calculations with MCNP4C2, MCNP5, MCNP-PoliMi to accomplish measurements of radioactive materials using active neutron methods.

In the framework of the present project we are going to develop new method and to implement a device consisting of neutron generator, moderator, system of neutron detectors based on fission chambers, electronic control system, electronic data acquisition system. This method will be laboratory tested on imitators of irradiated fuel solutions that will be prepared at Khlopin Radium Institute.

The following work will be carried out in the framework of the project:

  1. Mathematical modeling of the device to provide optimization of measurements, modeling of responses to main isotopes, development of algorithms for accurate estimation of isotopes concentration using variants of least-squares method;
  2. Production of neutron detectors based on fission chambers and neutron moderator;
  3. Development and production of control and data acquisition systems;
  4. Assembling and testing of the device together with neutron generator, moderator, neutron detectors, control system and data acquisition system;
  5. Laboratory tests of the method for control of absolute concentration of plutonium and uranium isotopes in imitators of dissolved irradiated fuel.

In accordance with the results of development and laboratory testing of the method the proposal will be worked out considering its probable application at fuel reprocessing facilities.


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