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Determining the moderator temperature coefficient in a PWR by stochastic fluctuation analysisHerr, John D. 11 May 2006 (has links)
An on-line, passive method for monitoring the moderator temperature coefficient (MTC) in a pressurized water reactor is described and compared to an analytical model. The method correlates fluctuations in the neutron flux with fluctuations in the coolant temperature. Results show that the frequency response function between the in-core neutron flux and thermocouple signals is proportional to the MTC and can be used to monitor changes in the MTC throughout the fuel cycle. An analytical model is also formulated and used to investigate how the neutron flux and coolant temperature fluctuations are related and change in the axial core direction. The analytical model results agree reasonably well with the experimental results. / Ph. D.
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Thermal-hydraulic analysis of a spectral shift reactor coreDamiano, Brian January 1982 (has links)
A computer program has been developed which performs a thermal-hydraulic analysis of a mechanical spectral shift reactor fuel assembly. In this reactor type, the fuel to water volume ratio is changed to shift the neutron energy spectrum and control reactivity. The ability to handle different fuel to water ratios is the unique feature of this program.
The required input parameters are the coolant inlet pressure and temperature, the mass flow rates, and the power density. The cladding temperature and the coolant pressure and temperature are calculated throughout the fuel assembly.
Three test cases were performed using a conceptual reactor core design. The fuel to water ratio of this design can be varied from 1.25 to 0.75 by removing blank rods from the core. Each fuel assembly is enclosed in a hex can. The fuel assembly flowrates can be independently selected.
The test case results indicate that proper flowrate selection can produce equal fuel assembly enthalpy gains. The maximum allowable pressure drop determines the maximum coolant flowrate and the maximum power density variation between fuel assemblies. The coolant exit properties are relatively unaffected by changing the fuel to water ratio. This indicates that little heat is transferred into the moderator regions that are formed when void rods are removed.
The most useful program modification would be the addition of a routine that defines the coolant channel parameters for any fuel assembly. / Master of Science
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Liquid entrainment at an upward oriented vertical branch line from a horizontal pipeWelter, Kent B. 25 September 2002 (has links)
Under simulated accident conditions, tees in the primary coolant loop of a
Pressurized Water Reactor (PWR) can deviate from their original design purpose
and become separators that effectively remove core heat sink capacity. This method
of primary coolant removal is a phenomelogical subset of phase separation known
as liquid entrainment, whereby liquid is forced from its original path by the inertia
of the gas. A comprehensive literature review revealed common deficiencies in
previous studies. The Westinghouse AP600 advanced reactor design was chosen to
assess the validity of entrainment models. Following a systematic scaling analysis
of the prototypic design a model separate effects test was proposed and constructed
at Oregon State University. Just under 100 tests were run to fill the deficiencies
found in the literature review. New data from the Air-water Test Loop for
Advanced Thermal-hydraulic Studies (ATLATS) could not be predicted by
published correlations. A new theoretical model for predicting liquid entrainment
onset and steady state entrainment was developed. Comparison with all available
data shows a marked improvement for predicting the mass flow rate out the vertical
branch. / Graduation date: 2003
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Critical heat flux estimation for annular channel geometryPagh, Richard T. 26 April 2001 (has links)
Critical Heat Flux (CHF) is an important safety parameter for the design of nuclear
reactors. The most commonly used predictive tool for determination of CHF is a look-up
table developed using tube data with an average hydraulic test diameter of 8 mm. There
exist in the world today nuclear reactors whose geometry is annular, not tubular, and
whose hydraulic diameter is significantly smaller than 8 mm. In addition, any sub-channel
thermal hydraulic model of fuel assemblies is annular and not tubular.
Comparisons were made between this predictive tool and annular correlations developed
from test data. These comparisons showed the look-up table over-predicts the CHF
values for annular channels, thus questioning its ability to perform correct safety
evaluations.
Since no better tool exists to predict CHF for annular geometry, an effort was
undertaken to produce one. A database of open literature annular CHF values was
created as a basis for this new tool. By compiling information from eighteen sources and
requiring that the data be inner wall, unilaterally, uniformly heated with no spacers or
heat transfer enhancement devices, a database of 1630 experimental values was
produced.
After a review of the data in the database, a new look-up table was created. A look-up
table provides localized control of the prediction to overcome sparseness of data.
Using Shepard's Method as the extrapolation technique, a regular mesh look-up table was
produced using four main variables: pressure, quality, mass flux, and hydraulic diameter.
The root mean square error of this look-up table was found to be 0.8267. However, by
fixing the hydraulic diameter locations to the database values, the root mean square error
was further reduced to 0.2816. This look-up table can now predict CHF values for
annular channels over a wide range of fluid conditions. / Graduation date: 2001
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Study of interfacial condensation in a nuclear reactor core makeup tankMa, Chang Chun 13 December 1993 (has links)
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 tank by rapid steam
condensation. The effect of varying the liquid conduction thermal layer thickness was
studied. In addition, the code's ability to predict condensation heat transfer was
evaluated.
Violent pressure oscillations were found in the early period of a transient. The
violent pressure oscillations had no effect on the total amount of injection water from
core makeup tank. The conduction thermal layer thickness was found to only effect
the liquid temperature history. The current version of RELAP5/MOD3 was found to
be incapable of dealing with the condensation heat transfer problem in which the
volume liquid temperature is lower than the temperature of the heat structure which is
connected to that hydraulic volume. / Graduation date: 1994
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Development of a simulation model for PWR reactor coolant system陳炳林, Chan, Ping-lam. January 1989 (has links)
published_or_final_version / Mechanical Engineering / Master / Master of Philosophy
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Mechanisms and quantities of tritium and krypton-85 production and release in a conceptual thorium-232/uranium-235 pressurized light-water prebreeder reactorOstenak, Carl Andrew January 1978 (has links)
No description available.
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Simulacao do modelo termodinamico de pressurizador tipico de PWR em regime transiente por meio do programa CSMPWOISKI, EMANUEL R. 09 October 2014 (has links)
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Estudo de acidente de perda de refrigerante por grande ruptura na usina nuclear Angra-1BORGES, EDUARDO M. 09 October 2014 (has links)
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Analise termo-hidraulica e neutronica de reatores a agua pressurizada (PWR)ALVES, CARLOS H. 09 October 2014 (has links)
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