scholarly journals New neutron cross section and fission yield data for SNManalysis

2003 ◽  
Author(s):  
R.B. Firestone ◽  
G.L. Molnar ◽  
Zs. Revay ◽  
T. Belgya
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Fission Yield ◽  
Yield Data

THE RELATIVE ABUNDANCES OF NEODYMIUM AND SAMARIUM ISOTOPES IN THE THERMAL NEUTRON FISSION OF U235 AND U233

1955 ◽  
Vol 33 (5) ◽  
pp. 830-837 ◽  
Author(s):  
E. A. Melaika ◽  
M. J. Parker ◽  
J. A. Petruska ◽  
R. H. Tomlinson
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Neutron Capture ◽  
Natural Uranium ◽  
Fission Yield ◽  
Yield Data ◽  
Neutron Fission ◽  
Thermal Neutron Fission ◽  
The Cross ◽  
Fission Yields

The relative fission yields of neodymium and samarium isotopes have been measured with a mass spectrometer for samples of natural uranium and U233 that had been irradiated with moderated neutrons. The cross sections for neutron capture by Sm149 and Sm151 have been determined to be 66,200 ± 2500 barns and 12,000 barns respectively, relative to the cross section of a B10 monitor. The half-lives of Pm147 and Sm151 have been evaluated to be 2.52 ± 0.08 yr. and ~93 yr., respectively, from samarium fission yield data for samples differing in age by seven years.

scholarly journals Propagation of Neutron Cross Section, Fission Yield, and Decay Data Uncertainties in Depletion Calculations

2014 ◽  
Vol 118 ◽  
pp. 480-483 ◽  
Author(s):  
J.S. Martinez ◽  
W. Zwermann ◽  
L. Gallner ◽  
F. Puente-Espel ◽  
O. Cabellos ◽  
...  
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Fission Yield ◽  
Decay Data

Investigation of the total neutron cross section for natural hafnium on the filtred neutron beam with energy 59 keV

2011 ◽  
Vol 30 (0) ◽  
pp. 142-147
Author(s):  
О. О. Грицай ◽  
А. К. Гримало ◽  
В. В. Колотий ◽  
В. М. Венедиктов ◽  
С. П. Волковецький ◽  
...  
Keyword(s):  
Cross Section ◽  
Neutron Beam ◽  
Neutron Cross Section ◽  
Total Neutron ◽  
Total Neutron Cross Section

Measurement of the Thermal Neutron Cross Section of the90Sr(n, γ)91Sr Reaction

1994 ◽  
Vol 31 (3) ◽  
pp. 173-179 ◽  
Author(s):  
Hideo HARADA ◽  
Toshiaki SEKINE ◽  
Yuichi HATSUKAWA ◽  
Noriko SHIGETA ◽  
Katsutoshi KOBAYASHI ◽  
...  
Keyword(s):  
Thermal Neutron ◽  
Cross Section ◽  
Neutron Cross Section ◽  
Thermal Neutron Cross Section

A Bent Crystal Neutron Spectrometer and Its Application to Neutron Cross-Section Measurements

1947 ◽  
Vol 72 (2) ◽  
pp. 109-115 ◽  
Author(s):  
R. B. Sawyer ◽  
E. O. Wollan ◽  
S. Bernstein ◽  
K. C. Peterson
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Neutron Spectrometer ◽  
Bent Crystal

Resolution of Bremsstrahlung Experiments

1961 ◽  
Vol 14 (1) ◽  
pp. 174 ◽  
Author(s):  
HH Thies
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Curve Fitting ◽  
Optical Resolution ◽  
Yield Data ◽  
Statistical Errors ◽  
Practical Evaluation ◽  
Bremsstrahlung Yield

This paper investigates how mathematical approximations and statistical errors are transmitted into computed cross sections in the analysis of experimental bremsstrahlung yield data. The resolution of bremsstrahlung experiments is defined in analogy with optical resolution and an expression for the practical evaluation of resolution is derived. Methods of cross-section computation, and smoothing and curve-fitting are discussed.

Total Neutron Cross Section of Deuterium below 1000 keV

1972 ◽  
Vol 29 (26) ◽  
pp. 1745-1748 ◽  
Author(s):  
P. Stoler ◽  
N. N. Kaushal ◽  
F. Green ◽  
Edward Harms ◽  
Luciano Laroze
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Total Neutron ◽  
Total Neutron Cross Section

Cross Section and Fission Yields Induced Uncertainty in the Vver-440 Burnup Calculation

2021 ◽  
Author(s):  
Branislav Vrban ◽  
Stefan Cerba ◽  
Jakub Luley ◽  
Filip Osuský ◽  
Vladimir Necas
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Reactor Operation ◽  
Precise Estimation ◽  
Stochastic Techniques ◽  
Core Characteristics ◽  
Fission Yields ◽  
Operational Data ◽  
Burnup Calculation

Abstract The properties of nuclear fuel depend on the actual isotopic composition which develops during a reactor operation. In practice, the prediction accuracy of burnup calculations serves as the basis for the future precise estimation of a core lifetime and other safety-based core characteristics. The present study quantifies nuclear data induced uncertainties of nuclide concentrations and multiplication factors in VVER-440 fuel depletion analysis. The well-known SCALE system and the TRITON sequence are used with the NEWT deterministic solver in the SAMPLER module that implements stochastic techniques to assess the uncertainty in computed results. The propagation of uncertainties in neutron cross section and fission yields is studied through the depletion calculation of 2D heterogeneous VVER-440 fuel assembly with an average enrichment of 4.87 wt % of 235U and six gadolinium rods with 3.35 % of Gd2O3. In the paper, fixed nominal depletion conditions are based on the real operational data of the Slovak NPP Bohunice unit 4 during cycle 30. In total 250 cases with uncertain parameters are computed and the results are evaluated by an auxiliary tool.

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