Neutron Multiplicity Counting Moments for Fissile Mass Estimation in ScatterBased Neutron Detection Systems
Abstract
Neutron multiplicity counting (NMC) techniques are widely used for nuclear materials accountability and international safeguards applications to quantitatively evaluate characteristic properties pertaining to fissile material. Mathematical models for NMC moments have been previously derived for systems that use capturebased detectors; however, these models are not applicable when scatterbased detectors are used because of “neutron cross talk.” Neutron cross talk occurs when a single neutron scatters and deposits energy above threshold into multiple detectors causing spurious increase in multiplicity counts; this, in turn, has caused fissile mass to be overestimated when not treated. In this paper, we propose new mathematical models derived from point kinetics to correct for neutron crosstalk effects up to any arbitrary order N, where N denotes the maximum number of counts a single neutron can cause. The new models were used to estimate the fissile mass of plutonium metal and oxide samples with effective ^{240}Pu mass ranging from 2.5 to 250 g. The adequacy of the models was confirmed using simulations of a conceptual scatterbased neutron multiplicity counter (e.g., organic scintillators) using MCNPX v2.7e with the PoliMi fission event generating extension. The fissile mass estimates with no correction for neutron crosstalk events yielded an average relative deviationmore »
 Authors:

 Univ. of Michigan, Ann Arbor, MI (United States). Dept. of Nuclear Engineering & Radiological Sciences
 Idaho National Lab. (INL), Idaho Falls, ID (United States)
 Univ. of Michigan, Ann Arbor, MI (United States). Dept. of Nuclear Engineering & Radiological Sciences; Chalmers Univ. of Technology, Göteborg, (Sweden). Division of Subatomic and Plasma Physics
 Publication Date:
 Research Org.:
 Univ. of Michigan, Ann Arbor, MI (United States)
 Sponsoring Org.:
 USDOE NA Office of Nonproliferation and Verification Research and Development (NA22); USDOE National Nuclear Security Administration (NNSA)
 OSTI Identifier:
 1454799
 Grant/Contract Number:
 NA0002534
 Resource Type:
 Accepted Manuscript
 Journal Name:
 Nuclear Science and Engineering
 Additional Journal Information:
 Journal Volume: 188; Journal Issue: 3; Journal ID: ISSN 00295639
 Publisher:
 American Nuclear Society  Taylor & Francis
 Country of Publication:
 United States
 Language:
 English
 Subject:
 46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY; Neutron multiplicity counting; neutron cross talk; fissile mass estimation
Citation Formats
Shin, Tony H., Hua, Michael Y., Marcath, Matthew J., Chichester, David L., Pazsit, Imre, Di Fulvio, Angela, Clarke, Shaun D., and Pozzi, Sara A. Neutron Multiplicity Counting Moments for Fissile Mass Estimation in ScatterBased Neutron Detection Systems. United States: N. p., 2017.
Web. doi:10.1080/00295639.2017.1354591.
Shin, Tony H., Hua, Michael Y., Marcath, Matthew J., Chichester, David L., Pazsit, Imre, Di Fulvio, Angela, Clarke, Shaun D., & Pozzi, Sara A. Neutron Multiplicity Counting Moments for Fissile Mass Estimation in ScatterBased Neutron Detection Systems. United States. doi:10.1080/00295639.2017.1354591.
Shin, Tony H., Hua, Michael Y., Marcath, Matthew J., Chichester, David L., Pazsit, Imre, Di Fulvio, Angela, Clarke, Shaun D., and Pozzi, Sara A. Fri .
"Neutron Multiplicity Counting Moments for Fissile Mass Estimation in ScatterBased Neutron Detection Systems". United States. doi:10.1080/00295639.2017.1354591. https://www.osti.gov/servlets/purl/1454799.
@article{osti_1454799,
title = {Neutron Multiplicity Counting Moments for Fissile Mass Estimation in ScatterBased Neutron Detection Systems},
author = {Shin, Tony H. and Hua, Michael Y. and Marcath, Matthew J. and Chichester, David L. and Pazsit, Imre and Di Fulvio, Angela and Clarke, Shaun D. and Pozzi, Sara A.},
abstractNote = {Neutron multiplicity counting (NMC) techniques are widely used for nuclear materials accountability and international safeguards applications to quantitatively evaluate characteristic properties pertaining to fissile material. Mathematical models for NMC moments have been previously derived for systems that use capturebased detectors; however, these models are not applicable when scatterbased detectors are used because of “neutron cross talk.” Neutron cross talk occurs when a single neutron scatters and deposits energy above threshold into multiple detectors causing spurious increase in multiplicity counts; this, in turn, has caused fissile mass to be overestimated when not treated. In this paper, we propose new mathematical models derived from point kinetics to correct for neutron crosstalk effects up to any arbitrary order N, where N denotes the maximum number of counts a single neutron can cause. The new models were used to estimate the fissile mass of plutonium metal and oxide samples with effective 240Pu mass ranging from 2.5 to 250 g. The adequacy of the models was confirmed using simulations of a conceptual scatterbased neutron multiplicity counter (e.g., organic scintillators) using MCNPX v2.7e with the PoliMi fission event generating extension. The fissile mass estimates with no correction for neutron crosstalk events yielded an average relative deviation from the true 240Pueff mass of 55.94% and 84.56% for metal and oxide samples, respectively. When neutron crosstalk events of order N = 2 are included in the model, the fissile mass estimates yielded an average relative deviation of 11.89% for metal and 13.21% for oxide samples. Accounting for neutron crosstalk events of order N = 3 resulted in fissile mass estimates with an average relative deviation of 9.58% and 10.51% for metal and oxide samples, respectively. These mass estimates were compared to a reference case (i.e., no neutron crosstalk effects) that yielded an average relative deviation of 6.81% and 4.77% for metal and oxide samples, respectively. In conclusion, the discrepancy between the estimates from the proposed model and the reference case is attributed to the assumed value of N, which sets a finite upper bound on the order of crosstalk events the model treats (i.e., the model for N = 3 assumes that a neutron will never cause more than three counts).},
doi = {10.1080/00295639.2017.1354591},
journal = {Nuclear Science and Engineering},
number = 3,
volume = 188,
place = {United States},
year = {2017},
month = {8}
}
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