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Title: Preliminary Investigation of 252Cf-Driven Neutron Noise Analysis for Subcritical Fuel Solution Systems

Abstract

A method for determining the reactivity of highly subcritical systems of fissile material, using neutron-noise power spectral densities in conjunction with a 252Cf source, had previously been tested in two fast reactor critical assemblies (a mockup of the Fast Flux Test Facility reactor and unreflected enriched uranium metal assemblies) and one thermal reactor (a light-water moderated and reflected lattice of Oak Ridge Research Reactor fuel elements). The last-mentioned test demonstrated the effectiveness of the method in water-moderated systems and thereby prompted the present study of its application to facilities for fuel preparation, reprocessing, and storage. To investigate the applicability of this method to facilities for fuel preparation, reprocessing, and storage, limited experiments were performed with a uranyl fluoride solution. The Los Alamos National Laboratory SHEBA facility, an unreflected cylindrical tank (56 cm diam), was partially filled with a solution containing 5 wt % 235U-enriched uranium (rho (U) approx. 1 g/cc, H/U atomic ratio approx. 550). The tank had a central 5-cm-diam axial hole in which the 252Cf source was placed at the midpoint of the solution. Measurements were performed at various solution heights from 60% of the critical height to near-critical height (approx. 36 cm).

Authors:
 [1];  [1];  [1]
  1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Publication Date:
Research Org.:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
OSTI Identifier:
6489679
Report Number(s):
CONF-810606-41
ON: DE81023217; TRN: 81-011105
DOE Contract Number:  
W-7405-ENG-26
Resource Type:
Conference
Resource Relation:
Conference: American Nuclear Society Annual Meeting, Miami Beach, FL (United States), 7 Jun 1981
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; CALIFORNIUM 252; NEUTRON SOURCES; NUCLEAR FUELS; REACTIVITY; SPENT FUEL STORAGE; SUBCRITICAL ASSEMBLIES; SPENT FUELS; ACTINIDE ISOTOPES; ACTINIDE NUCLEI; ALPHA DECAY RADIOISOTOPES; CALIFORNIUM ISOTOPES; ENERGY SOURCES; EVEN-EVEN NUCLEI; EXPERIMENTAL REACTORS; FUELS; HEAVY NUCLEI; ISOTOPES; MATERIALS; NUCLEI; PARTICLE SOURCES; RADIATION SOURCES; RADIOISOTOPES; REACTOR MATERIALS; REACTORS; RESEARCH AND TEST REACTORS; STORAGE; YEARS LIVING RADIOISOTOPES; Nuclear Criticality Safety Program (NCSP); Light-Water Moderated; Reflected Lattice; Oak Ridge Research Reactor; Fuel Elements; Fast Flux Test Facility Reactor; Los Alamos National Laboratory SHEBA Facility; Fissile Material; Fuel Solution Tanks; 050700* - Nuclear Fuels- Fuels Production & Properties; 050800 - Nuclear Fuels- Spent Fuels Reprocessing; 220600 - Nuclear Reactor Technology- Research, Test & Experimental Reactors

Citation Formats

Mihalczo, J. T., Kryter, R. C., and King, W. T. Preliminary Investigation of 252Cf-Driven Neutron Noise Analysis for Subcritical Fuel Solution Systems. United States: N. p., 1981. Web.
Mihalczo, J. T., Kryter, R. C., & King, W. T. Preliminary Investigation of 252Cf-Driven Neutron Noise Analysis for Subcritical Fuel Solution Systems. United States.
Mihalczo, J. T., Kryter, R. C., and King, W. T. 1981. "Preliminary Investigation of 252Cf-Driven Neutron Noise Analysis for Subcritical Fuel Solution Systems". United States. https://www.osti.gov/servlets/purl/6489679.
@article{osti_6489679,
title = {Preliminary Investigation of 252Cf-Driven Neutron Noise Analysis for Subcritical Fuel Solution Systems},
author = {Mihalczo, J. T. and Kryter, R. C. and King, W. T.},
abstractNote = {A method for determining the reactivity of highly subcritical systems of fissile material, using neutron-noise power spectral densities in conjunction with a 252Cf source, had previously been tested in two fast reactor critical assemblies (a mockup of the Fast Flux Test Facility reactor and unreflected enriched uranium metal assemblies) and one thermal reactor (a light-water moderated and reflected lattice of Oak Ridge Research Reactor fuel elements). The last-mentioned test demonstrated the effectiveness of the method in water-moderated systems and thereby prompted the present study of its application to facilities for fuel preparation, reprocessing, and storage. To investigate the applicability of this method to facilities for fuel preparation, reprocessing, and storage, limited experiments were performed with a uranyl fluoride solution. The Los Alamos National Laboratory SHEBA facility, an unreflected cylindrical tank (56 cm diam), was partially filled with a solution containing 5 wt % 235U-enriched uranium (rho (U) approx. 1 g/cc, H/U atomic ratio approx. 550). The tank had a central 5-cm-diam axial hole in which the 252Cf source was placed at the midpoint of the solution. Measurements were performed at various solution heights from 60% of the critical height to near-critical height (approx. 36 cm).},
doi = {},
url = {https://www.osti.gov/biblio/6489679}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Jan 01 00:00:00 EST 1981},
month = {Thu Jan 01 00:00:00 EST 1981}
}

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