Cross Sections of Threshold Reactions for Fission Neutrons: Nickel as a Fast Flux Monitor

The neutron flux trap effect was experimentally studied in the sub-critical assembly of the Atomic and Nuclear Physics Laboratory of the Aristotle University of Thessaloniki, using delayed gamma neutron activation analysis (DGNAA). Measurements were taken within the fuel grid, in vertical levels symmetrical to the Am-Be neutron source, before and after the removal of fuel elements, also permitting a basic study of the vertical flux profile. Three identical flux traps of diamond shape and an area of 96 cm2 were created by removing four fuel rods for each one. Two (n,γ) reactions and one (n,p) threshold reaction were selected for thermal, epithermal and fast flux study. For the thermal and epithermal flux, results obtained through the 197Au(n,γ)198Au, and 186W(n,γ)187W reactions were used, with and without Cd covers, to differentiate between the two flux regions. For the fast flux, the 58Ni(n,p)58Co reaction was used.All measurements were taken in a HPGe detector of 42% relative efficiency, with a resolution of 1.8 keV at 1332 keV and analyzed in the SPECTRW software package, developed at NCSR Demokritos.An interpolation technique based on local procedures is used to fit the cross sections and the flux spectra.End results show a thermal flux increase of 105% at the source level, and 90% across all levels, pointing to a high potential to increase the available thermal flux for future experiments. Furthermore, the vertical flux profile was found to be slightly asymmetric, with higher flux values at the top part of the assembly.

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Measurement of cross sections for flux monitor reactions using quasi-monoenergetic neutrons

The European Physical Journal Plus ◽

10.1140/epjp/s13360-021-01673-9 ◽

2021 ◽

Vol 136 (7) ◽

Author(s):

Vibhuti Vashi ◽

Rajnikant Makwana ◽

S. Mukherjee ◽

B. K. Soni ◽

M. H. Mehta ◽

...

Keyword(s):

Cross Sections ◽

Flux Monitor

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Cross sections of the (n, 2n) reaction of 59Co, 58Ni, 70Ge, 90Zr and 203Tl with fission neutrons

Journal of Inorganic and Nuclear Chemistry ◽

10.1016/0022-1902(81)80312-0 ◽

1981 ◽

Vol 43 (7) ◽

pp. 1427-1431 ◽

Cited By ~ 4

Author(s):

Toshiaki Sekine ◽

Hiroshi Baba

Keyword(s):

Cross Sections ◽

Fission Neutrons

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An Analytical Solution for the Two-Group Diffusion Theory Analysis of Multiregion Source-Driven Systems

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4042021 ◽

2019 ◽

Vol 5 (3) ◽

Author(s):

Huseyin Atilla Ozgener ◽

Ceyhun Yavuz

Keyword(s):

Analytical Solution ◽

Cross Sections ◽

Diffusion Theory ◽

Spherical Geometry ◽

Multiplication Factor ◽

Source Regions ◽

Design Variables ◽

Effective Multiplication Factor ◽

Fast Flux ◽

Effective Multiplication

An analytical solution of the two-group diffusion equations is derived for multiregion source-driven subcritical systems in spherical geometry. An analytical formulation for the calculation of the effective multiplication factor is also presented. Using typical two-group cross sections characterizing source, buffer, and blanket regions, parameters such as neutron amplification, source efficiency, and blanket fast flux peaking factor are calculated. The criticality solution is utilized to calculate the effective multiplication factor and the neutron source efficiency. The dependency of the calculated global parameters on design variables like the source, buffer, blanket thicknesses, and subcriticality level is studied. Thin source regions result in very high neutron amplification, at the expense of high blanket fast flux peaking factors. If a buffer region is put between the source and the blanket regions, flux peaking could be reduced at the expense of reduced neutron amplification. If the subcriticality level can be reduced without jeopardizing safety, the neutron amplification increases and the fast flux peaking is reduced.

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A PRACTICAL EFFECT OF THE RECENT O16 CROSS SECTION AND MUBAR UNCERTAINTIES

EPJ Web of Conferences ◽

10.1051/epjconf/202124715010 ◽

2021 ◽

Vol 247 ◽

pp. 15010

Author(s):

D. Kent Parsons ◽

Scott A. Turner ◽

Peter J. Jaegers

Keyword(s):

Elastic Scattering ◽

Cross Section ◽

Cross Sections ◽

Neutron Transport ◽

Surprising Result ◽

Typical Application ◽

Scattering Cross ◽

Practical Effect ◽

Fission Neutrons ◽

Scattering Cross Sections

With the recent release of ENDF/B VIII.0 data, additional covariance data was provided for many isotopes, including O16. The detail of elastic scattering and mubar covariance data for O16 increased dramatically between ENDF/B VII.1 and ENDF/B VIII.0. This new covariance data has been processed with NJOY2016 and investigation has begun on the effects of these new uncertainty data. The uncertainties are applied to multi-group scattering cross sections and P1 Legendre components in deterministic neutron transport. A simple but typical application of shielding fission neutrons with concrete has been used to assess the practical effects of the new covariance data for O16. A somewhat surprising result is that the mubar uncertainty can have a significant effect on the calculated shielding and criticality results.

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THE S32 (n,p) P32 REACTION AS A FAST-NEUTRON FLUX MONITOR

Canadian Journal of Chemistry ◽

10.1139/v63-016 ◽

1963 ◽

Vol 41 (1) ◽

pp. 123-133 ◽

Cited By ~ 17

Author(s):

D. C. Santry ◽

J. P. Butler

Keyword(s):

Neutron Flux ◽

Fast Neutron ◽

Cross Sections ◽

Extensive Study ◽

Neutron Production ◽

Flux Monitor ◽

Fast Neutron Flux ◽

Neutron Flux Monitor

An extensive study is made of the use of sulphur as a monitor for neutrons with energies from 2 to 20 Mev. Irradiated disks of pressed sulphur are burnt before counting to increase the efficiency and reproducibility of determining the P32 activity. It was established that on burning irradiated sulphur 95.5 ± 0.5% of the P32 activity is retained with a reproducibility of ± 1.0%. Measurements of relative cross sections for the S32 (n,p) P32 reaction have been extended to 20.3 Mev using the T (d,n) He4 and D (d,n) He3 reactions to produce monoenergetic neutrons. Neutron production using Zr−T targets is examined as is the effect of extraneous neutrons from the D (d,np) D process on the S32 (n,p) reaction.

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A study on improving the performance of a research reactor's equilibrium core

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp1304362m ◽

2013 ◽

Vol 28 (4) ◽

pp. 362-369

Author(s):

Atta Muhammad ◽

Masood Iqbal ◽

Tayyab Mahmood

Keyword(s):

Cross Sections ◽

Group Structure ◽

Energy Group ◽

Time Cycle ◽

Computer Codes ◽

Fission Neutrons ◽

Improved Performance ◽

Enriched Uranium ◽

Fuel Elements ◽

Uranium Silicide

Utilizing low enriched uranium silicide fuel (U3Si2-Al) of existing uranium density (3.285 g/cm3), different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of ?unit flux time cycle length per 235U mass per cycle?. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3) without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

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Cross-observables and cross-isotopes correlations in nuclear data from integral constraints

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2018011 ◽

2018 ◽

Vol 4 ◽

pp. 35 ◽

Cited By ~ 2

Author(s):

Eric Bauge ◽

Dimitri A. Rochman

Keyword(s):

Cross Sections ◽

Nuclear Data ◽

Fission Neutron ◽

Wide Range ◽

Data Files ◽

Fission Neutrons ◽

Prompt Fission ◽

Benchmark Evaluation ◽

Prompt Fission Neutron ◽

Good Agreement

Most recent evaluated nuclear data files exhibit excellent integral performance, as shown by the very good agreement between experimental and calculated keff values over a wide range of benchmark integral experiments. However, the propagation of the uncertainties associated with those nuclear data to integral observables, generally produces calculated distribution which are much (3–5 times) wider than the experimental uncertainties. Reducing the variances of the evaluated data to achieve consistency at the integral level would lead to unreasonably narrow variances in the light of differential experimental data. One way of solving that paradox could be to allow, for different observables like fission cross-sections (σf), the prompt fission neutron spectra (χ), and the average multiplicity of fission neutrons ([see formula in PDF]) to be correlated in a Bayesian-like, Total Monte-Carlo approach, under constraints from integral experiments from the ICSBEP (International International Criticality Safety Benchmark Evaluation Project) benchmark compilation. Future developments will be highlighted and restrictions imposed by the current formatting of nuclear data will be discussed.

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