| 1. |
- Chernikova, Dina, 1982, et al.
(author)
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Experimental and numerical investigations of radiation characteristics of Russian portable/compact pulsed neutron generators: ING-031 ING-07, ING-06 and ING-10-20-120
- 2014
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In: Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment. - : Elsevier BV. - 0168-9002. ; 746, s. 74-86
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Journal article (peer-reviewed)abstract
- The present paper discusses results of full-scale experimental and numerical investigations of influence of construction materials of portable pulsed neutron generators ING-031, ING-07, ING-06 and ING-10-20-120 (VNIIA, Russia) to their radiation characteristics formed during and after an operation (shutdown period). In particular, it is shown that an original monoenergetic isotropic angular distribution of neutrons emitted by TiT target changes into the significantly anisotropic angular distribution with a broad energy spectrum stretching to the thermal region. Along with the low energetic neutron part, a significant amount of photons appears during the operation of generators. In the pulse mode of operation of neutron generator, a presence of the construction materials leads to the "tailing" of the original neutron pulse and the appearance of an accompanying photon pulse at similar to 3 ns after the instant neutron pulse. In addition to that, reactions of neutron capture and inelastic scattering lead to the creation of radioactive nuclides, such as Co-58, Cu-62, Cu-64 and F-18, which form the so-called activation radiation. Thus, the selection of a portable neutron generator for a particular type of application has to be clone considering radiation characteristics of the generator itself. This paper will be of interest to users of neutron generators, providing them with valuable information about limitations of a specific generator and with recommendations for improving the design and performance of the generator as a whole. (c) 2014 Elsevier B.V. All rights reserved.
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| 2. |
- Batyaev, V.F., et al.
(author)
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Monitoring fissile and matrix materials in closed containers by means of pulsed neutron sources
- 2013
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In: Atomic Energy. - : Springer Science and Business Media LLC. - 1573-8205 .- 1063-4258. ; 115:2, s. 116-122
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Journal article (peer-reviewed)abstract
- Computational and experimental studies of the possibility of determining the fissile and matrix materials by means of differential neutron attenuation are presented. The time response of 235U fission under the action of 14 MeV neutrons from an ING-07T pulsed neutron generator, fabricated by VNIIA, on a 70-liter steel container holding uranium in graphite, iron and polyethylene matrices is analyzed. It is shown that milligram quantities of 235U can be detected and the matrix type and density as well as the location of fissile material inside a container can be determined.©2013 Springer Science+Business Media New York.
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| 3. |
- Chernikova, Dina, 1982, et al.
(author)
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Time intervals matrix analysis of 235U and 239Pu content in a spent fuel assembly using Lead Slowing down Spectrometer
- 2011
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In: Proceedings: Special Session on Determination of Pu Mass in Spent Fuel with NDA of the 52nd Annual Meeting of the Institute of Nuclear Materials Management.
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Conference paper (other academic/artistic)abstract
- A key motivation for developing the technology capable of quantifying plutonium (Pu) in spent fuel assemblies with nondestructive assay (NDA) techniques is knowledge of the physical parameters of irradiated nuclear fuel which is important both for nuclear safeguards and nuclear fuel management. One of the most attractive NDA approaches applied for the determination of the total amount of plutonium using a pulsed neutron source is the 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 is based on a matrix of time intervals where large differences in the number of fissions of 235U and 239Pu are observed [3]. 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 in this work 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 spent fuel assemblies characterization has only been studied using Monte Carlo simulations, it was possible to demonstrate the 239Pu determination using a DT pulsed neutron source, Lead Slowing Down Spectrometer and fast timing scintillatior which is sensitive to both photons and neutrons, and n-γ pulse shape discrimination allows to get additional information about the system.
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| 4. |
- Romodanov, V. L., et al.
(author)
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Sensitivity of nonstationary neutron spectra in lead to nuclear data variations
- 2012
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In: Atomic Energy. - : Springer Science and Business Media LLC. - 1573-8205 .- 1063-4258. ; 112:4, s. 281-287
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Journal article (peer-reviewed)abstract
- The computed variations of the parameters of nonstationary neutron spectra in lead moderator due to the use of different libraries of the constants characterizing the sensitivity of these parameters to nuclear data variations are presented. The testing of the nuclear constants is based on the varying neutron spectra formed in a lead sphere at different times after a neutron pulse from a source. The computational studies showed considerable sensitivity of the differential and integral time characteristics of the neutron field in lead to the nuclear data library and showed that the constant component of the computational error in the neutron spectra in lead moderator must be taken into account.
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| 5. |
- Chernikova, Dina, 1982, et al.
(author)
-
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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| 6. |
- Romodanov, V. L., et al.
(author)
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A method for nondestructive assay of nuclear materials in facilities with a pulse neutron generator and a digital technology for discrimination between neutrons and photons
- 2013
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In: Instruments and Experimental Techniques. - 1608-3180 .- 0020-4412. ; 56:6, s. 620-627
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Journal article (peer-reviewed)abstract
- A method for determining the U-235 concentration in fuel assemblies of a high-power channel-type PBMK reactor is described. The measure of U-235 content of an analyzed sample is the number of neutrons from thermal-neutron fission of U-235, normalized to the number of gamma quanta produced in thermal neutron capture by hydrogen nuclei in the scintillator or by B-10 in the glass of a photomultiplier tube. A pulse neutron generator based on DT reaction is the neutron source, and an organic scintillator with the pulse shape discrimination between neutrons and gamma rays with the aid of the digital technology is a detector. The scintillator is also used as a neutron moderator. Simulation of the method shows that the U-235 content of the analyzed sample can be determined for 1 min with an accuracy of 1% or better. The efficiency of the method has been confirmed by experimental investigations on a model of the setup.
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