In the wake of nuclear accidents such as Three Mile Island Unit 2 and Fukushima, the nuclear power industry’s safety record is scrutinized. Today the main concerns lies with hydrogen production in a nuclear reactor core when the zirconium fuel cladding reacts with the water coolant during an accident, creating...
Absorption of methyl iodide by aqueous hydrazine solutions
within spray chambers was studied theoretically and experimentally.
The objective of this work was to evaluate the absorption
rate of airborne methyl iodide by sprays of hydrazine solutions
under conditions expected to exist in post-accident nuclear
reactor containment atmospheres.
Experimental effort included...
Antineutrino detectors could provide a valuable addition to current safeguards
regimes. Antineutrinos are an attractive emission to monitor due to their low
interaction cross-section that prevents them from being shielded and the dependence
of their spectrum on the power level and isotopic content of a reactor core. While
there are...
At Oregon State University the Multi-Application Small Light Water Reactor (MASLWR) integral effects testing facility is being prepared for safety analysis matrix testing in support of the NuScale Power Inc. (NSP) design certification progress. The facility will be used to simulate design basis accident performance of the reactor's safety systems....
The 1979 incident at Three Mile Island drew attention to weaknesses in the capability to obtain and analyze reactor coolant
samples in an environment where the reactor core incurs significant
damage. As a result, the U.S. Nuclear Regulatory Commission promulgated regulations requiring that reactor coolant analytical capabilities be upgraded so...
There is renewed interest in the reliability and safety of nuclear power plants following the Fukushima Daiichi nuclear accident followed by 8.9 magnitude earthquake and Tsunami with the height of 15 m on March 11, 2011. Small Modular Reactors (SMRs) have been developed to improve safety systems by utilizing passive...
A limited scope Probabilistic Risk Assessment (PRA) of a vented nuclear fuel system for a Generation IV Gas-Cooled Fast Reactor Plant was performed. The goal of the study was to better understand the safety and licensing implications of vented fuel technology. A Level 1 PRA was performed to determine the...
This study focuses on investigation of the RETRAN modeling
concepts in the analysis of a safety valve discharge transient.
Three safety valve tests conducted at the Combustion Engineering
test facility were chosen for evaluation with RETRAN. It was
essential to apply the critical flow boundary condition at the
downstream nodes...
A transient code (TFETC) for calculating the temperature
distribution throughout the radial and axial positions of a
thermionic fuel element (TFE) has been successfully developed.
It accommodates the variations of temperatures, thermal power,
electrical power, voltage, and current density throughout the
TFE as a function of time as well as...