Abstract
When performing research at a reactor facility, experimenters often need to determine the neutron fluence achieved during an operation. Facilities typically provide guidance in the form of neutron fluence per megajoule (MJ) or through passive dosimetry results. After experiment completion, there is sometimes a delay of several days (or weeks) before the passive dosimetry results are available. In the interim, an experimenter does not have confirmation that the desired irradiation levels were reached. Active dosimetry may provide an estimate of neutron fluxes, but few active detectors are available that have been calibrated to measure neutron fluxes obtained inside the Annular Core Research Reactor (ACRR) central cavity environment. For past experiments at the ACRR, the neutron fluence was calculated by integrating the response of a fission chamber rate detection signal and then normalizing this integral to fluence determined from passive dosimetry. An alternative method of directly measuring neutron flux is desired; the new methodology described provides a complete neutron flux profile after a reactor pulse, utilizing fission chamber physics in combination with a compensating ion chamber to extract and convert a current signal to neutron flux as a function of time.
Highlights
The fielding experiment of the fission chamber in the lead-boron 36” bucket (LB36) Annular Core Research Reactor (ACRR) environment consisted of four pulse runs and three steady-state runs, each at varying energy levels
Current signals were collected from the fission chamber, LND ion chamber, and the compensating ion chamber (CIC)
The values for F are nearly identical for the nickel and sulfur dosimetry data; the MCNP data appears to overestimate the value for F, likely due to an underestimation in the number of fissions that occur within the fission chamber
Summary
To the fission rate, the current is proportional to the incident neutron flux [1] The magnitude of this proportionality is largely dependent on two factors: (1) the spectrum-averaged fission cross section for the fissile materials for the incident neutrons and (2) the characteristic energy required to ionize atoms of the fill gas. The method outlined attempts to utilize only the current produced by neutrons inducing fission in the fission chamber, but the fission chamber is sensitive to ionization caused by gamma radiation present in the reactor environment. If fielded in a characterized reactor environment where the neutron spectrum is well established, this compensated method employing the fission chamber will successfully provide a time-dependent neutron flux profile from the detector signal
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