Sensitivity of an (α,n) neutron source to (α,n) cross sections and stopping powers

Favorite, Jeffrey A.

doi:10.1016/j.anucene.2019.107154

Title: Sensitivity of an (α,n) neutron source to (α,n) cross sections and stopping powers

Abstract

The first-order sensitivities of the (α,n) neutron source rate density in a homogeneous material to isotopic (α,n) cross sections, elemental electronic stopping power data, and general coefficients for the nuclear stopping power are derived for the total neutron source and the energy-dependent neutron source. The physics bases for some of the equations in the SOURCES4C calculation of the stopping power are briefly discussed. The derivatives have been implemented in SOURCES4C. In simple PuO₂ and PuB₂ test problems, the computed derivatives compare extremely well with central differences. As a result, the first-order sensitivities are applied to show that improving the value of one of SOURCES4C’s stopping-power coefficients changes the (α,n) neutron source rate by 1.738 × 10^–6% in the PuO₂ problem.

Authors:

Favorite, Jeffrey A. ^[1]

Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

Publication Date:: Fri Dec 13 00:00:00 EST 2019

Research Org.:: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

Sponsoring Org.:: USDOE National Nuclear Security Administration (NNSA). Office of Defense Nuclear Nonproliferation R&D (NA-22); USDOE National Nuclear Security Administration (NNSA)

OSTI Identifier:: 1604039

Alternate Identifier(s):: OSTI ID: 2325475

Report Number(s):: LA-UR-19-27042
Journal ID: ISSN 0306-4549; TRN: US2104185

Grant/Contract Number:: 89233218CNA000001

Resource Type:: Accepted Manuscript

Journal Name:: Annals of Nuclear Energy (Oxford)

Additional Journal Information:: Journal Name: Annals of Nuclear Energy (Oxford); Journal Volume: 138; Journal Issue: C; Journal ID: ISSN 0306-4549

Publisher:: Elsevier

Country of Publication:: United States

Language:: English

Subject:: 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; (α,n) neutron source; Sensitivity; SOURCES4C; Stopping power; Plutonium oxide; Plutonium boride

Citation Formats


                    Favorite, Jeffrey A. Sensitivity of an (α,n) neutron source to (α,n) cross sections and stopping powers.  United States: N. p., 2019. 
Web.  doi:10.1016/j.anucene.2019.107154.

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                    Favorite, Jeffrey A. Sensitivity of an (α,n) neutron source to (α,n) cross sections and stopping powers.  United States.  https://doi.org/10.1016/j.anucene.2019.107154

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                    Favorite, Jeffrey A. Fri .  
"Sensitivity of an (α,n) neutron source to (α,n) cross sections and stopping powers".  United States.  https://doi.org/10.1016/j.anucene.2019.107154.  https://www.osti.gov/servlets/purl/1604039.

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@article{osti_1604039,

  title        = {Sensitivity of an (α,n) neutron source to (α,n) cross sections and stopping powers},

  author       = {Favorite, Jeffrey A.},

  abstractNote = {The first-order sensitivities of the (α,n) neutron source rate density in a homogeneous material to isotopic (α,n) cross sections, elemental electronic stopping power data, and general coefficients for the nuclear stopping power are derived for the total neutron source and the energy-dependent neutron source. The physics bases for some of the equations in the SOURCES4C calculation of the stopping power are briefly discussed. The derivatives have been implemented in SOURCES4C. In simple PuO2 and PuB2 test problems, the computed derivatives compare extremely well with central differences. As a result, the first-order sensitivities are applied to show that improving the value of one of SOURCES4C’s stopping-power coefficients changes the (α,n) neutron source rate by 1.738 × 10–6% in the PuO2 problem.},

  doi          = {10.1016/j.anucene.2019.107154},

  journal      = {Annals of Nuclear Energy (Oxford)},

  number       = C,

  volume       = 138,

  place        = {United States},

  year         = {Fri Dec 13 00:00:00 EST 2019},

  month        = {Fri Dec 13 00:00:00 EST 2019}

}

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Works referenced in this record:

XXXIX. On the α particles of radium, and their loss of range in passing through various atoms and molecules
journal, September 1905

Bragg, W. H.; Kleeman, R.
The London, Edinburgh, and Dublin Philosophical Magazine and Journal of Science, Vol. 10, Issue 57
DOI: 10.1080/14786440509463378

Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE
journal, May 2011

Gauld, Ian C.; Radulescu, Georgeta; Ilas, Germina
Nuclear Technology, Vol. 174, Issue 2
DOI: 10.13182/NT11-3

In-line (α,n) source sampling methodology for monte carlo radiation transport simulations
journal, September 2017

Griesheimer, David P.; Pavlou, Andrew T.; Thompson, Jason T.
Nuclear Engineering and Technology, Vol. 49, Issue 6
DOI: 10.1016/j.net.2017.08.004

A simple universal fit formula to experimental nuclear stopping power data
journal, September 1976

Kalbitzer, S.; Oetzmann, H.; Grahmann, H.
Zeitschrift f�r Physik A: Atoms and Nuclei, Vol. 278, Issue 3
DOI: 10.1007/BF01409171

Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix
journal, August 2016

Pigni, M. T.; Croft, S.; Gauld, I. C.
Progress in Nuclear Energy, Vol. 91
DOI: 10.1016/j.pnucene.2016.04.006

Reflection of energetic particles from atomic or ionic chains in single crystals
journal, March 1970

Varelas, C.; Biersack, J.
Nuclear Instruments and Methods, Vol. 79, Issue 2
DOI: 10.1016/0029-554X(70)90141-2

Sources: A code for calculating (alpha, n), spontaneous fission, and delayed neutron sources and spectra
journal, May 2009

Wilson, W. B.; Perry, R. T.; Charlton, W. S.
Progress in Nuclear Energy, Vol. 51, Issue 4-5
DOI: 10.1016/j.pnucene.2008.11.007

The electronic and nuclear stopping of energetic ions
journal, October 1977

Ziegler, J. F.
Applied Physics Letters, Vol. 31, Issue 8
DOI: 10.1063/1.89771

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Similar Records

Title: Sensitivity of an (α,n) neutron source to (α,n) cross sections and stopping powers

Abstract

Citation Formats

XXXIX. On the α particles of radium, and their loss of range in passing through various atoms and molecules journal, September 1905

Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE journal, May 2011

In-line (α,n) source sampling methodology for monte carlo radiation transport simulations journal, September 2017

A simple universal fit formula to experimental nuclear stopping power data journal, September 1976

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Sources: A code for calculating (alpha, n), spontaneous fission, and delayed neutron sources and spectra journal, May 2009

The electronic and nuclear stopping of energetic ions journal, October 1977

XXXIX. On the α particles of radium, and their loss of range in passing through various atoms and molecules
journal, September 1905

Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE
journal, May 2011

In-line (α,n) source sampling methodology for monte carlo radiation transport simulations
journal, September 2017

A simple universal fit formula to experimental nuclear stopping power data
journal, September 1976

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journal, August 2016

Reflection of energetic particles from atomic or ionic chains in single crystals
journal, March 1970

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journal, May 2009

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