Use of Noise Analysis Methods in Process Monitoring of Future Fuel Cycles

Conference · Sun Sep 12 00:00:00 EDT 1993

OSTI ID:10183449

Mihalczo, J. T. ^[1]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

The ²⁵²Cf-source-driven noise analysis method has been used in measurements for subcritical configurations of fissile systems for a variety of applications including a dynamic measurement that was the first step in applying this method to monitor fuel in a future reprocessing plant These measurements showed that this method has potential to monitor many dynamic situations in processing plants such as in a continuous dissolver or batch dissolvers either to obtain the neutron multiplication factor k or to be used as a signature to verify that various processes are occurring in a repeatable or bounded way. This verification of normal operation would be by comparing noise analysis signatures for the process with reference signatures. Abnormal operation could be ascertained if the signature deviates by some specified amount from the reference signatures from normal operation. The deviations from normal could be specified by measurement or by Monte Carlo neutron transport theory methods directly calculating the measured parameters for the processing plant applications.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10183449
• Report Number(s): CONF-930913-24; ON: DE93018725
• Resource Relation: Conference: Global '93: Future Nuclear Systems: Emerging Fuel Cycles and Waste Disposal Options, Seattle, WA (United States), 12-17 Sep 1993; Other Information: PBD: [1993]
• Country of Publication: United States
• Language: English

Related Subjects

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
FUEL REPROCESSING PLANTS
ACOUSTIC MONITORING
FISSILE MATERIALS
NEUTRON DETECTION
NEUTRON DETECTORS
SPENT FUELS
MEASURING METHODS
Nuclear Criticality Safety Program (NCSP)
Fissile Systems
Monte Carlo Neutron Transport Theory Methods
Fuel Loading
Reactor Fuel Elements
Cross-Power Spectral Densities (CPSDs)
Autopower Spectral Density (APSD)
Light-Water Reactor (LWR) Spent Fuel
High Flux Isotope Reactor (HFIR) Fuel Elements
KENO-V.a
050800
440101
SPENT FUELS REPROCESSING
GENERAL DETECTORS OR MONITORS AND RADIOMETRIC INSTRUMENTS