A PC based on-line advanced plant simulator (OLAPS)
for high quality simulations of Portland General
Electric's Trojan Nuclear Facility is presented. OLAPS
is designed to simulate the thermal-hydraulics of the
primary system including core, steam generators, pumps,
piping and pressurizer. The simulations are based on a
five equation model that...
Condensation Induced Water-Hammer, CIWH, has been an
historical problem for the nuclear power industry over the
past 2 decades. It has caused damage to plant systems, and
considerable anguish to plant operators. This thesis has
embarked on an attempt to characterize the fluid motion, heat
transfer, mixing, and stability of...
Steam interfacial condensation in a core makeup tank was simulated using the
code RELAP5/MOD3 version 8.0 to predict the violent pressure oscillation phenomena
in a core makeup tank. Six base cases were carried out to study the effects of back
pressure and of vacuum conditions produced in the core makeup...
The flooding phenomenon can be defined as the maximum attainable flow
condition beyond which the well defined countercurrent flow pattern can no longer
exist. Thus the countercurrent flow limit (CCFL) or the flooding limit may be thought
of as the flow condition at which the strong interaction between the two...
A series of small break loss of coolant accident (SBLOCA) analyses in nuclear plant shutdown operations was simulated using the code RELAP5A,MOD3 version 8.0 to predict the SBLOCA phenomena in the Zion-l nuclear power plant The first objective is to study the impact of SBLOCA (1" and 2" breaks) on...
The topic of this thesis is to discuss an
investigation made for a fluid velocity measuring device.
The OSU thermal hydraulic testing facility is interested
in using a small probe to measure fluid velocities in the
annular region of a model nuclear reactor vessel. A
heater with a thermocouple embedded...
The purpose of this paper is to describe the flow oscillations which occur in the
AP600 long term cooling test facility at Oregon State University. The AP600 system is
an advanced pressurized water reactor design utilizing passive emergency cooling
systems.
A few hours after the initiation of a cold leg...
The Department of Nuclear Engineering at Oregon State University has performed a series of confirmatory tests for the United States Nuclear Regulatory Commission (USNRC). These tests have been conducted in the Advanced Plant Experiment (APEX) Facility which is a one quarter height scaled simulation of the Westinghouse Advanced Passive 600...
Counter Current Flow Limitation (CCFL) was observed in the pressurizer surge line of the Oregon State University APEX facility during test NRC-10. This test simulated a one-inch diameter cold leg break with a failure of three of four of the fourth-stage Automatic Depressurization System (ADS) valves. The result was a...
The purposes of this paper are to present the results of an experimental effort to measure the level swell in an air/water system and to generate a methodology for determining the volume-averaged void fraction within the Oregon State University (OSU) Advanced Plant Experiment (APEX) Test Facility. The results were then...