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- Chernikova, Dina, 1982, et al.
(author)
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Analysis of 235U, 239Pu and 241Pu content in a spent fuel assembly using Lead Slowing Down Spectrometer and time intervals matrix
- 2012
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In: JNMM, Journal of the Institute of Nuclear Materials Management. - 0893-6188. ; 40:2, s. 9-18
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Journal article (peer-reviewed)abstract
- Nowadays knowledge of the physical parameters of irradiated nuclear fuel is going to be a key issue for the continued and future use of nuclear energy. One of the major characteristics of spent fuel which plays an important role in international nuclear materials Safeguards is the quantity of plutonium (Pu) in wastes. It can be obtained through using of various techniques, one of which is the non-destructive assay (NDA) method of slowing-down time spectrometry in lead where the energy spectrum of neutrons can be represented as being monoenergetic with minor deviation from the peak value in each time moment after a fast neutron pulse. This fact was successfully used in developing several methods of Pu mass determination and confirmed the potential of the Lead Slowing Down Spectrometer (LSDS) to get detailed information about spent fuel [1-2]. A method, which we presented earlier [3], was based on a matrix of time intervals where large differences in the number of fissions of 235U and 239Pu are observed. This technique allows increasing precision in the Pu evaluation by decreasing the self-shielding effect significantly. As opposed to homogeneous-volume approximations used in our previous research, we describe the detailed Monte Carlo models of real fuel assemblies, as well as the effects of the influence of the scintillation detector to the system in question. Although the proposed method for characterization of spent fuel assemblies has only been studied using Monte Carlo simulations, it was possible to demonstrate the determination of 239Pu using a DT pulsed neutron source, a Lead Slowing Down Spectrometer, and fast timing scintillator that is sensitive to both photons and neutrons. Additional information about the system can be obtained from n-γ pulse shape discrimination.
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| 2. |
- Chernikova, Dina, 1982, et al.
(author)
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Development of the Neutron-Gamma-Neutron (NGN) approach for the fresh and spent fuel assay
- 2012
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In: Proceedings of The 53nd Annual Meeting of the Institute of Nuclear Materials Management.
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Conference paper (other academic/artistic)abstract
- In connection with current safety and complexity limitations for installations which usekeV neutrons for irradiation in the fresh and spent fuel assay, in medicine, geophysicaland detection fields, there is a long felt need of effective, light, inexpensive systems withlonger lifetimes and the possibility to ”switch off” the source during transportation andwork break. The present paper discusses an opportunity of creating a novel technique basedon using a pulsed neutron generator, hydrogen moderator and beryllium, which will enablecreation of a compact and inexpensive facility capable to satisfy all requirements. The mainidea of the proposed method consists in using photonuclear reaction in beryllium due togamma irradiation originating from (n,gamma) reaction in a hydrogen containing moderator.After the neutron pulse of a modern compact DD neutron generator in hydrogen-containingmoderator, high-energy neutrons (approximately 2.5 MeV) get slowed down mainly by elasticscattering to near thermal energies. At thermal energies, the neutrons diffuse through thematerial until they undergo thermal capture (capture is dominated by hydrogen neutronabsorbers). When a hydrogen atom captures a thermal neutron, it turns into deuteriumwith the release of a large component of 2.23 MeV gamma-rays. Beryllium is one of thefew elements in nature that undergoes a photonuclear reaction with this gamma energyrange (e.g. deuterium has a low energy 2.225 MeV photonuclear threshold). Thus, as theberyllium has lower neutron-binding energy, 1.667 MeV (photonuclear reaction threshold),a photonuclear reaction will take place with emission of neutrons with energy defined by thekinematic equation. Thus, this method allows to obtain a high quality epithermal neutronbeam without using a complex system with particle accelerator or radioisotope sources withlimited decay time. The proposed method has been studied using Monte Carlo simulations,which made it possible to define the theoretical limits of the use of the NGN approach inindustrial applications. The paper will presents the results of these investigations.
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