Engineering topics

Harvey, Taylor Joseph

Publications and source records attributed to Harvey, Taylor Joseph.

4 records

DOE OSTI2023

MCNP Simulation of the SFRC and Minimum Detection Threshold Determination (Rev. 1)

In this report the Minimum Detection Threshold (MDT) for the Spent Fuel Rod Counter (SFRC) was determined using Monte Carlo N-Particle transport computer code (MCNP). A model of the SFRC was simulated in addition to a fuel transport pipe and spent Magnox fuel at descending positions down the pipe. The neutron MDT for Magnox fuel free-falling along a discharge pipe was calculated to be equal to 4080 n/s. The gamma MDT for the same free falling fuel as measured by the SFRC’s ionization chamber was determined to be 455 γ/s. These values were determined to be a conservative value to use when the instrument is deployed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

SFRC Measurement Control Summary - FY23 Q3/Q4

Measurements with the Spent Fuel Rod Counter (SFRC) to be performed bi annually include 3 He detector efficiency checks (in both modules) and ion chamber channel tests with a pocket pulser. These measurements will all be performed at TA-35, building 27 where the SFRC is stored. Refer to LA-UR-22-31737 and LA-CP-22-20079 SFRC measurement control documents for a description of additional tests. Note that during this measurement period, counts could not be confirmed for the Si detectors, as an analog pulse to TTL box necessary for the DAQs to communicate with the detectors was still under construction. For this reason, typical measurements at the Calibrated Dose Facility were not performed for the second half of FY23. When these components are finished by the NEN electrical shop, the Si detectors will be implemented into these measurement control checks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Automated Fissile Mass Estimation Python Script for the Advanced Experimental Fuel Counter (AEFC)

The Advanced Experimental Fuel Counter (AEFC) is a non-destructive assay (NDA) instrument developed to quantify the residual fissile mass in research reactor spent fuel assemblies. The system is installed in a spent fuel pool and measurements are acquired underwater. A calibration curve relates the totals and coincidence counting signal from active interrogation of the spent fuel assemblies to the fissile mass remaining in the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Plutonium Basket Counter Measurement Control Q3-Q4 2023

The Plutonium Basket Counter (PBC) is a spent fuel nuclear safeguards instrument that was developed to measure spent-fuel elements from MAGNOX-type research reactors and determine their plutonium content. The instrument is designed to measure fuel elements in spent fuel cooling pools through underwater operation, but it is also able to measure radiation sources in air. The PBC determines fuel plutonium content by detecting neutrons with an array of Helium-3 detectors. The electronics make use of a JSR-15 shift register for data collection and IAEA Neutron Coincidence Counting (INCC) software for data analysis. The PBC is used to determine the 240 Pu content in spent MAGNOX fuel elements grouped into basket-like bundles, thus the instrument’s name. Reactor burnup calculations can be used in conjunction with the data from the PBC to estimate the total plutonium content in the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗