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Title: Validation of the Fast-Neutron Multiplicity Expressions for Fissile Mass Estimation

Abstract

Neutron multiplicity analysis is a non-destructive assay technique that involves measurement of the multiplets of neutrons that are emitted from fission of fissile material. Typically, multiplicity analysis is conducted by means of a neutron multiplicity counting (NMC) system that uses capture-based thermal-neutron detectors (e.g., 3He-gas detectors). To relieve the reliance on such detectors, we have developed a fast-neutron multiplicity counting (FNMC) system that uses scatter-based, organic scintillators to directly detect the fast fission neutrons emitted from the sample. One challenge with FNMC systems, however, is that they are prone to effects of neutron cross-talk, where a single neutron may scatter and cause more than one detection event. This leads to an overestimate of the fission rate (F) and inaccurate estimation of other sample parameters such as the leakage multiplication (M) and the alpha-ratio (a) when using the traditional capture-based NMC equations. We have formulated fully generalized multiplicity expressions that account for cross-talk effects for any arbitrary order N. The validity of these expressions were verified with MCNPX-PoliMi simulations and experimental measurements; this paper will present the validation of the proposed generalized multiplicity expressions. Californium-252 and Pu metal plates were measured using the FNMC system and the data was used formore » neutron multiplicity analysis to estimate sample parameters F, M, and a. We demonstrate that accounting for neutron cross-talk effects of order N = 2 and 3 improves the accuracy of the estimated sample parameters.« less

Authors:
; ; ; ; ; ;
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1407411
Report Number(s):
INL/CON-17-41102
DOE Contract Number:  
DE-AC07-05ID14517
Resource Type:
Conference
Resource Relation:
Conference: INMM 58th Annual Meeting, Indian Wells, California, USA, July 16–20, 2017
Country of Publication:
United States
Language:
English
Subject:
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY; Neutron Multiplicity; Safeguards; Well Counter

Citation Formats

Shin, T. H., Hua, M. Y., Di Fulvio, A., Marcath, M. J., Chichester, D. L., Clarke, S. D., and Pozzi, S. A. Validation of the Fast-Neutron Multiplicity Expressions for Fissile Mass Estimation. United States: N. p., 2017. Web.
Shin, T. H., Hua, M. Y., Di Fulvio, A., Marcath, M. J., Chichester, D. L., Clarke, S. D., & Pozzi, S. A. Validation of the Fast-Neutron Multiplicity Expressions for Fissile Mass Estimation. United States.
Shin, T. H., Hua, M. Y., Di Fulvio, A., Marcath, M. J., Chichester, D. L., Clarke, S. D., and Pozzi, S. A. Thu . "Validation of the Fast-Neutron Multiplicity Expressions for Fissile Mass Estimation". United States. https://www.osti.gov/servlets/purl/1407411.
@article{osti_1407411,
title = {Validation of the Fast-Neutron Multiplicity Expressions for Fissile Mass Estimation},
author = {Shin, T. H. and Hua, M. Y. and Di Fulvio, A. and Marcath, M. J. and Chichester, D. L. and Clarke, S. D. and Pozzi, S. A.},
abstractNote = {Neutron multiplicity analysis is a non-destructive assay technique that involves measurement of the multiplets of neutrons that are emitted from fission of fissile material. Typically, multiplicity analysis is conducted by means of a neutron multiplicity counting (NMC) system that uses capture-based thermal-neutron detectors (e.g., 3He-gas detectors). To relieve the reliance on such detectors, we have developed a fast-neutron multiplicity counting (FNMC) system that uses scatter-based, organic scintillators to directly detect the fast fission neutrons emitted from the sample. One challenge with FNMC systems, however, is that they are prone to effects of neutron cross-talk, where a single neutron may scatter and cause more than one detection event. This leads to an overestimate of the fission rate (F) and inaccurate estimation of other sample parameters such as the leakage multiplication (M) and the alpha-ratio (a) when using the traditional capture-based NMC equations. We have formulated fully generalized multiplicity expressions that account for cross-talk effects for any arbitrary order N. The validity of these expressions were verified with MCNPX-PoliMi simulations and experimental measurements; this paper will present the validation of the proposed generalized multiplicity expressions. Californium-252 and Pu metal plates were measured using the FNMC system and the data was used for neutron multiplicity analysis to estimate sample parameters F, M, and a. We demonstrate that accounting for neutron cross-talk effects of order N = 2 and 3 improves the accuracy of the estimated sample parameters.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2017},
month = {6}
}

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