Neutron Radiation Characteristics of Plutonium Dioxide Fuel

Taherzadeh, M.; Gingo, P. J.

doi:10.13182/nt72-a16037

Cited by 17 publications

(21 citation statements)

References 8 publications

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“…The normalized distribution of the neutron spectrum χ αn (E) was obtained through approximation of experimental (EXFOR − Experimental Nuclear Reaction Data) and evaluated (ENDSF − Evaluated Nuclear Structure Data File) data (Gorshkov 1962, Herold 1968, Anderson and Neff 1969, Taherzadeh 1971, Taherzadeh and Gingo Peter 1972, Arkhipov et al 1972, Bair and Butler 1973, Anderson 1980 by the function of the type XXX…”

Section: Computational Modelmentioning

confidence: 99%

“…where the constants а and b depend on the type of the nuclei i and j. The investigations conducted in (Gorshkov 1962, Herold 1968, Anderson and Neff 1969, Taherzadeh 1971, Taherzadeh and Gingo Peter 1972, Arkhipov et al 1972, Bair and Butler 1973, Anderson 1980 had the purpose of obtaining highly accurate values of the neutron yields and spectra for 238 PuO 2 . The measurement error does not exceed 5%, and the compilation and the evaluation were performed in (Vlaskin et al 2015, Bulanenko 1980.…”

Section: Computational Modelmentioning

confidence: 99%

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Solution of neutron-transport multigroup equations system in subcritical systems

Shamanin

Bedenko

Нестеров

et al. 2018

NUCET

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Citation: Shamanin IV, Bedenko SV, Nesterov VN, Lutsik IO, Prets AA (2018) Solution of neutron-transport multigroup equations system in subcritical systems. Nuclear Energy and Technology 4(1): 79-85. https://doi. AbstractAn iteration method has been implemented to solve a neutron transport equation in a multigroup diffusion approximation. A thermoelectric generator containing plutonium dioxide, used as a source of thermal and electric power in spacecraft, was studied.Neutron yield and multigroup diffusion approximation data was used to obtain a continuous and group distribution of neutron flux density spectra in a subcritical multiplying system. Numerical multigroup approaches were employed using BNAB-78, a system of group constants, and other available evaluated nuclear data libraries (ROSFOND, BROND, BNAB, EXFOR and ENDSF).The functions of neutron distribution in the zero iteration for the system of multigroup equations were obtained by approximating an extensive list of calculated and experimental data offered by the EXFOR and ENDSF nuclear data libraries. The required neutronic functionals were obtained by solving a neutron transport equation in a 28-group diffusion approximation. The calculated data was verified. The approach used is more efficient in terms of computational efforts (the values of the neutron flux density fractions converge in the third iteration). The implemented technique can be used in nuclear and radiation safety problems. KeywordsSubcritical system, neutron distribution function, neutron transport, multigroup diffusion approximation. Research statusNeutron transport calculation methods are used in reactor physics, for process monitoring during nuclear fuel fabrication and reprocessing, to determine the irradiated fuel burnup, in biological shielding design, and to improve the radiation measurement procedures in the nuclear material accounting and control system.There are a lot of dedicated programs and codes that implement the Monte Carlo method based on selection of neutron interaction probabilities. The internationally recognized program packages include MMKKENO and MMKC (IPPE, Russia), MCU (Kurchatov Institute, Russia), MCNP (USA), KENO-3D (USA), MONK (Great Britain) and others. These programs make it possible to obtain highly accurate results thanks to a 3D geometry

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Section: Computational Modelmentioning

confidence: 99%

Section: Computational Modelmentioning

confidence: 99%

Solution of neutron-transport multigroup equations system in subcritical systems

Shamanin

Bedenko

Нестеров

et al. 2018

NUCET

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“…Candidates are small (very low-multiplication) standards composed of very pure oxide or oxide with very low (or very well defined) impurity concentrations. Upper limits on the allowable concentrations of ten impurity elements were calculated using basic nuclear and atomic data with a simple stopping power model [3]. Again, each limit represents the element content that contributes 1% of the spontaneous-fission neutron specific yield for pure weapons-grade plutonium in oxide form.…”

Section: A Impurity Limitsmentioning

confidence: 99%

“…Table III gives the target mass, diameter, and estimated thickness of PuO 2 powder for each of the three oxide standard sizes. The oxide density of 1.5 g/cm 3 determines the required height of each of the three container sizes conservatively, to assure ease of fabrication in a glovebox environment. …”

Section: B Multiplicationmentioning

confidence: 99%

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Preparation of Small Well Characterized Plutonium Oxide Reference Materials and Demonstration of the Usefulness of Such Materials for Nondestructive Analysis

Guillen¹,

Hsue²,

Huang³

et al. 2003

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Neutron background environment measured by the Mars Science Laboratory's Dynamic Albedo of Neutrons instrument during the first 100 sols

et al. 2013

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[1] The Dynamic Albedo of Neutrons (DAN) instrument on board Mars Science Laboratory has been operating successfully since the landing and has been making measurements regularly along Curiosity's traverse at the surface. DAN measures thermal (E < 0.4 eV) and epithermal neutrons (0.4 eV < E <~1 keV) while operating in two different modes: active and passive. The active mode uses a pulsed neutron generator (PNG) to study the geological characteristics of the subsurface. In the passive mode, DAN measures the background neutron environment. This paper presents results of measurements in the passive mode from landing through to sol 100 and provides an interpretation of the data based on extensive Monte Carlo simulations. The main observations are summarized as follows: (1) the thermal neutron counts vary strongly along the rover traverse while the epithermal counts do not show much variation; (2) the neutrons from the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) are a larger contributor to the DAN passive data than the Galactic Cosmic Ray (GCR)-induced neutrons; (3) for the MMRTG neutrons, both the thermal and the epithermal counts increase as a function of the subsurface water content; (4) on the other hand, for the GCR-induced neutrons, the thermal counts increase but the epithermal counts decrease as a function of the subsurface water content; and (5) relative contributions by the MMRTG and GCR to the DAN thermal neutron counts at the Rocknest site, where the rover was stationed from sol 59 to sol 100, are estimated to be~60% and~40%, respectively.

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Neutron Radiation Characteristics of Plutonium Dioxide Fuel

Cited by 17 publications

References 8 publications

Solution of neutron-transport multigroup equations system in subcritical systems

Solution of neutron-transport multigroup equations system in subcritical systems

Preparation of Small Well Characterized Plutonium Oxide Reference Materials and Demonstration of the Usefulness of Such Materials for Nondestructive Analysis

Neutron background environment measured by the Mars Science Laboratory's Dynamic Albedo of Neutrons instrument during the first 100 sols

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