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Erosive-corrosive wear in steam-extraction lines of power plantsVu, Hung Viet January 1982 (has links)
Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Mechanical Engineering, 1982. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND ENGINEERING. / Bibliography: leaf 42. / by Hung Viet Vu. / M.S.
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Population estimates and projections for nuclear power plant safety analyses and evacuation plans : the Shoreham nuclear power station methodologyDonnelly, Kathleen A January 2010 (has links)
Typescript (photocopy). / Digitized by Kansas Correctional Industries
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Experimental investigation of liquid entrainment in a reactor hot-leg with a vertical branchWelter, Kent B. 26 January 2001 (has links)
A literature review of current phase separation publications was conducted. Data
sets were collected and compiled into a Two-Phase Flow Separation Database.
Examination of this database indicating a need for further investigation into the liquid
entertainment phenomena for smaller hot-leg to branch diameters and intermittent flow
regimes. A detailed analysis to the prototypic phase separation process is presented and
the associated phenomena are identified. Appropriate scaling criteria were employed for
the design of a scaled test facility. Geometry and the flow conditions of the test facility
were determined accordingly to Wu et. al (1998).
A series of phase separation tests conducted at the Air-water Test Loop for
Advanced Thermal-hydraulic Studies (ATLATS) and Advanced Plan Experiment
(APEX) has been completed. Results show that the criteria developed by Smoglie (1984)
used in RELAP5, reasonably predicts the onset of liquid entrainment. However, the
steady-state entrainment correlation in RELAP5 significantly underpredicts primary
coolant removal rates. This discrepancy is due to the effects of downstream boundary
conditions and pool entrainment and carry-over from the reactor vessel. Due to pool
entrainment, entrainment through the branch continues when the reactor vessel mixture
level drops below the bottom of the hot-leg. This investigation shows that RELAP5 is
non-conservative when predicting coolant removal rates due to steady state liquid
entrainment in a horizontal mainline with a vertical branch for stratified, stratified-wavy,
transition, and stepped hot-leg flow regimes. / Graduation date: 2001
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A knowledge-based approach for monitoring and situation assessment at nuclear power plantsHeaberlin, Joan Oylear 21 July 1994 (has links)
An approach for developing a computer-based aid to
assist in monitoring and assessing nuclear power plant
status during situations requiring emergency response has
been developed. It is based on the representation of
regulatory requirements and plant-specific systems and
instrumentation in the form of hierarchical rules. Making
use of inferencing techniques from the field of artificial
intelligence, the rules are combined with dynamic state data
to determine appropriate emergency response actions.
In a joint project with Portland General Electric
Company, a prototype system, called EM-CLASS, was been
created to demonstrate the knowledge-based approach for use
at the Trojan Nuclear Power Plant. The knowledge domain
selected for implementation addresses the emergency
classification process chat is used to communicate the
severity of the emergency and the extent of response actions
required. EM-CLASS was developed using Personal Consultant
Plus (PCPlus), a knowledge-based system development shell
from Texas Instruments which runs on IBM-PC compatible
computers. The knowledge base in EM-CLASS contains over 200
rules.
The regulatory basis, as defined in 10 CFR 50, calls
for categorization of emergencies into four emergency action
level classes: (1) notification of unusual event, (2) alert,
(3) site area emergency, and (4) general emergency. Each
class is broadly defined by expected frequency and the
potential for release of radioactive materials to the
environment. In a functional sense, however, each class
must be ultimately defined by a complex combination of in-
plant conditions, plant instrumentation and sensors, and
radiation monitoring information from stations located both
on- and off-site. The complexity of this classification
process and the importance of accurate and timely
classification in emergency response make this particular
application amenable to an automated, knowledge-based
approach.
EM-CLASS has been tested with a simulation of a 1988
Trojan Nuclear Power Plant emergency exercise and was found
to produce accurate classification of the emergency using
manual entry of the data into the program. / Graduation date: 1997
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Development of systems analysis program for space reactor studiesLewis, Bryan R. 14 June 1993 (has links)
An overall systems design code was developed to model
an advanced in-core thermionic energy conversion based
nuclear reactor system for space applications at power
levels of 10 to 50 kWe. The purpose of this work was to
provide the overall shell for the systems code and to also
provide the detailed neutronic analysis section of the code.
The design code that was developed is to be used to evaluate
a reactor system based upon a single cell thermionic fuel
element which uses advanced technology to enhance the
performance of single cell thermionic fuel elements.
A literature survey provided information concerning how
other organizations performed system studies on similar
space reactor designs. / Graduation date: 1994
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System modeling and reactor design study of an advanced incore thermionic space reactorLee, Hsing Hui 12 October 1992 (has links)
Incore thermionic space reactor design concepts which operate at a
nominal power output range of 20 to 50 kWe are described. Details of the
neutronic, thermionic, thermal hydraulics and shielding performance are
presented. Due to the strong absorption of thermal neutrons by natural
tungsten, and the large amount of that material within the reactor core,
two designs are considered.
An overall system design code has been developed at Oregon State
University to model advanced incore thermionic energy conversion based
nuclear reactor systems for space applications. The code modules include
neutronics and core criticality, a thermionic fuel element performance
module with integral thermal hydraulics calculation capability, a
radiation shielding module, and a module for the waste heat rejection.
The results show that the driverless single cell ATI configuration,
which does not have driver rods, proved to be more efficient than the
driven core, which has driver rods. It also shows that the inclusion of
the true axial and radial power distribution decrease the overall
conversion efficiency. The flattening of the radial power distribution by
three different methods would lead to a higher efficiency. The results
show that only one thermionic fuel element (TFE) works at the optimum
emitter temperature; all other TFEs are off the optimum performance and
result in 40 % decrease of the efficiency of the overall system. / Graduation date: 1993
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Process analysis and aspen plus simulation of nuclear-based hydrogen production with a copper-chlorine cycleChukwu, Cletus 01 August 2008 (has links)
Thermochemical processes for hydrogen production driven by nuclear energy are promising alternatives to existing technologies for large-scale commercial production of hydrogen, without dependence on fossil fuels. In the Copper-Chlorine (Cu-Cl) cycle, water is decomposed in a sequence of intermediate processes with a net input of water and heat, while hydrogen and oxygen gases are generated as the products. The Super Critical Water-cooled Reactor (SCWR) has been identified as a promising source of heat for these processes. In this thesis, the process analysis and simulation models are developed using the Aspen PlusTM chemical process simulation package, based on experimental work conducted at the Argonne National Laboratory (ANL) and Atomic Energy of Canada Limited (AECL). A successful simulation is performed with an Electrolyte Non Random Two Liquid (ElecNRTL) model of Aspen Plus. The efficiency of the cycle based on three and four step process routes is examined in this thesis. The thermal efficiency of the four step thermochemical process is calculated as 45%, while the three step hybrid thermochemical cycle is 42%, based on the lower heating value (LHV) of hydrogen. Sensitivity analyses are performed to study the effects of various operating parameters on the efficiency, yield, and thermodynamic properties. Possible efficiency improvements are discussed. The results will assist the development of a lab-scale cycle which is currently being conducted at the University of Ontario Institute of Technology (UOIT), in collaboration with its partners. / UOIT
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Nonintrusive intelligent monitoring for nuclear power plant emergency classificationGreene, Kenneth R. (Kenneth Ray), 1958- 13 May 1991 (has links)
A prototype real-time process monitoring emergency
classification expert system, RT/EM-CLASS, has been developed
for use at the Trojan Nuclear Power Plant. This knowledge-based
system features the integration of electronically sensed plant
data with the menu selection data representation of its
predecessor, EM-CLASS. This prototype demonstrates the
techniques required to acquire plant process data from another
computer and use that data in an expert system to determine the
proper Emergency Action Level.
Several benefits are realized by the RT/EM-CLASS application.
These include:
The resources required to make a classification are
reduced thereby freeing the responsible person to devote
time to other important tasks.
The classification may be completed more often and with
better data than the current system allows.
The human user is less likely to make an erroneous
Emergency Action Level classification.
Prototype implementation required resolution of an efficiency
problem of relating plant process data to the expert system data
forms. This was achieved through the development of multi-conditional
rules that significantly reduce the size of the rule set.
The development of RT/EM-CLASS presents a methodology
for building knowledge based applications that perform nonintrusive
real-time monitoring of dynamic systems. This
methodology features
Use of existing analytical and Al tools where possible
Monitoring of a dynamic system
Non-intrusive acquisition of data from the system
This technology might be applied to portions of the nuclear
engineering design process (control rod programming in Boiling
Water Reactors, for example) to emulate the guidance and
activities of an expert. A substantial portion of the effort by the
expert engineer involves preparation of the code input, running the
computer code, analyzing the results, and based on the results,
deciding what case to submit next. A suitably designed expert
system could act in the place of the engineer in this dynamic
design process.
This methodology has been tested against the 1988 emergency
exercise at the Trojan Nuclear Power Plant. / Graduation date: 1992
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Nuclear design analysis of low-power (1-30 KWe) space nuclear reactor systemsGedeon, Stephen R. 23 November 1993 (has links)
Preliminary nuclear design studies have been completed on ten
configurations of nuclear reactors for low power (1-30 kWe) space
applications utilizing thermionic energy conversion. Additional design
studies have been conducted on the TRICE multimegawatt in-core
thermionic reactor configuration. In each of the cases, a reactor
configuration has been determined which has the potential for operating
7 years with sufficient reactivity margin. Additional safety
evaluations have been conducted on these configurations including the
determination of sufficient shutdown reactivity, and consideration of
water immersion, water flooding, sand burial, and reactor compaction
accident scenarios. It has been found, within the analysis conducted
using the MCNP Monte Carlo neutron transport code, that there are
configurations which are feasible and deserve further analysis. It has
also been found that solid core reactors which rely solely on conduction
for heat removal as well as pin type cores immersed in a liquid metal
bath have merit. The solid cores look attractive when flooding and
compaction accident scenarios are considered as there is little chance
for water to enter the core and cause significant neutron moderation. A
fuel volume fraction effect has also been found in the consideration of
the sand burial cases for the SP-100 derived configurations. / Graduation date: 1994
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Stochastic Modeling of Deterioration in Nuclear Power Plant ComponentsYuan, Xianxun January 2007 (has links)
The risk-based life-cycle management of engineering systems in a nuclear power
plant is intended to ensure safe and economically efficient operation of
energy generation infrastructure over its entire service life. An important
element of life-cycle management is to understand, model and forecast the
effect of various degradation mechanisms affecting the performance of
engineering systems, structures and components.
The modeling of degradation in nuclear plant components is confounded by large
sampling and temporal uncertainties. The reason is that nuclear systems are
not readily accessible for inspections due to high level of radiation and
large costs associated with remote data collection methods. The models of
degradation used by industry are largely derived from ordinary linear
regression methods.
The main objective of this thesis is to develop more advanced techniques based
on stochastic process theory to model deterioration in engineering components
with the purpose of providing more scientific basis to life-cycle management
of aging nuclear power plants. This thesis proposes a stochastic gamma process
(GP) model for deterioration and develops a suite of statistical techniques
for calibrating the model parameters. The gamma process is a versatile and
mathematically tractable stochastic model for a wide variety of degradation
phenomena, and another desirable property is its nonnegative, monotonically
increasing sample paths. In the thesis, the GP model is extended by including
additional covariates and also modeling for random effects. The optimization
of age-based replacement and condition-based maintenance strategies is also presented.
The thesis also investigates improved regression techniques for modeling
deterioration. A linear mixed-effects (LME) regression model is presented to
resolve an inconsistency of the traditional regression models. The proposed
LME model assumes that the randomness in deterioration is decomposed into two
parts: the unobserved heterogeneity of individual units and additive
measurement errors.
Another common way to model deterioration in civil engineering is to treat the
rate of deterioration as a random variable. In the context of condition-based
maintenance, the thesis shows that the random variable rate (RV) model is
inadequate to incorporate temporal variability, because the deterioration
along a specific sample path becomes deterministic. This distinction between
the RV and GP models has profound implications to the optimization of
maintenance strategies.
The thesis presents detailed practical applications of the proposed models to
feeder pipe systems and fuel channels in CANDU nuclear reactors.
In summary, a careful consideration of the nature of uncertainties associated
with deterioration is important for credible life-cycle management of
engineering systems. If the deterioration process is affected by temporal
uncertainty, it is important to model it as a stochastic process.
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