ISSN:0029-5450    

DOI:10.13182/NT87-A34014

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractA remote maintenance robot system for use in a hazardous environment is presented. The system consists of a programmable turntable, a robot, and hoist subsystems, which operate under the control of a supervisory computer to perform coordinated programmed maintenance operations on a pulsed nuclear reactor.

ISSN:0029-5450    

DOI:10.13182/NT87-A34015

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractThe development of an improved model to determine the dynamic response of the primary heat transport pump during severe transients in a Canada deuterium uranium reactor is presented. A lumped parameter model is proposed. A control volume formulation is employed for centrifugal pumps. The mathematical formulation of this model is based on mass, momentum, and energy balances as well as on Euler’s Theory of Turbomachines. Several constitutive relationships are adopted in the model to describe three-dimensional effects. In addition, the proposed model includes the consequent effect of different flow regimes and the slip between the two phases. Numerical results indicated that the proposed model favorably predicted the pump response and compared well with other pump-related models (Aerojet Nuclear Company) in the literature as well as in the experimental data.

ISSN:0029-5450    

DOI:10.13182/NT87-A34016

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractThe time series modeling approach is introduced to estimate the prompt-neutron decay constant. Neutron flux noise data of three fuel cycles of a high flux isotope reactor are analyzed. The noise data detected from an ionization chamber outside the reactor core surrounded by a beryllium reflector were recorded at full-power operation. The decay constant corresponding to a rounded-off corner break frequency can be estimated from the characteristic roots of adequate autoregressive moving average models. This implicit characteristic identification is one of the advantages of off-line modeling analysis. The estimated neutron lifetime in the beginning of fuel cycle is 38µs (expected value = 35µs). The estimated lifetime near the end of cycle is 66µs (expected value = 70µs).

ISSN:0029-5450    

DOI:10.13182/NT87-A34017

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractA real-time expert system (DIAREX) has been developed to diagnose plant failure and to offer a corrective operational guide for boiling water reactor (BWR) power plants. The failure diagnosis model used in DIAREX was systematically developed, based mainly on deep knowledge, to cover heuristics. Complex paradigms for knowledge representation were adopted, i.e., the process representation language and the failure propagation tree. The system is composed of a knowledge base, knowledge base editor, preprocessor, diagnosis processor, and display processor. The DIAREX simulation test has been carried out for many transient scenarios, including multiple failures, using a real-time full-scope simulator modeled after the 1100-MW(electric) BWR power plant. Test results showed that DIAREX was capable of diagnosing a plant failure quickly and of providing a corrective operational guide with a response time fast enough to offer valuable information to plant operators.

ISSN:0029-5450    

DOI:10.13182/NT87-A34018

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractAn operational concern in natural-convection-cooled research reactors is pool-top16N activity (PTNA). The conventional technique for reducing PTNA is to disperse the water plume rising above the core by a planar water jet and thus increase the transit time of16N nuclei to the pool top. The extension in transit time is a function of pool dynamics under dispersion. Ideally, a sufficiently deep stagnant water layer is formed below the pool top to confine16N activity to lower pool regions. The effects of changes in pool configuration and disperser design parameters on pool dynamics are not well known. These effects are important in determining the feasibility of a power upgrade without major facility modifications. Due to the complexity of pool geometry, pool dynamics under dispersion cannot be described by simple flow models. The COMMIX-1A code is used to simulate the pool dynamics of a typical natural-convection-cooled research reactor with plate-type elements as a function of pool configuration and disperser design parameters. The pool is partly described as continuum and partly as porous medium. All the major pool components are explicitly modeled. The differences between the shapes of some pool structures and computational cells are accounted for using the concept of directional surface permeability. The importance of local turbulence effects and cross-flow friction losses at the guide tubes above the core are also investigated. The results show the following:1. The conventional technique for reducing PTNA is effective in the power upgrade of natural-convection-cooled research reactors with large pools.2. If pool dimensions are small, a more feasible way to reduce PTNA is to place pool outlet suction points close to the top of the core.3. Containing the core outflow within a shroud increases the pool heat removal system inlet temperature, as well as reducing PTNA.

ISSN:0029-5450    

DOI:10.13182/NT87-A34019

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractSelecting a nuclear power plant site for power production and water desalination is a very complex problem, especially in countries with moderate technology. Many interrelated factors affect the process, and professional judgments by various experts are involved. Four sites, all located on the West Coast of Saudi Arabia along the Red Sea, were chosen as potential sites for building such a plant. (All sites were in either the northern or southern section of the coast; the central part was excluded for pilgrims’safety.) The East Coast was completely eliminated in the initial screening process due to its strategic location, the existence of oil fields and refineries, and its proximity to other Arabian (Persian) Gulf countries (to minimize radioactive releases to these countries in case of an accident). A computer code based on Saaty’s eigenvalue technique and developed in a previous study was used in this analysis. Twenty-one main criteria were considered, and the sites were ranked to determine which was most desirable. Site 4 was found to be most suitable, followed by site 3.

ISSN:0029-5450    

DOI:10.13182/NT87-A34020

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractThe RIBD code results and calculated detector pulse-height distributions have been used to assess the capabilities of gamma-ray spectrometry as applied to the measurement of reactor coolant activity in the event of a severe fuel failure in a nuclear power plant. The associated interference effects of nearby photopeaks and the Compton continuum of higher energy gamma rays were considered in this assessment. Key radionuclides representative of differing degrees of fuel damage have been found to be measurable under severe accident conditions.

ISSN:0029-5450    

DOI:10.13182/NT87-A34021

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractMelt refining as a means of uranium decontamination of metallic wastes by electroslag refining was examined. Electroslag refining was selected because it is easy to scale up to the necessary industrial levels. Various thicknesses of iron and aluminum cylinders with uranium concentrations close to actual metallic wastes were melted by adding effective fluxes for decontamination. Thin-walled iron and aluminum cylinders with a fill ratio (electrode/mold cross-section ratio) of 0.05 could be melted, and the energy efficiency obtained was 16 to 25%. The ingot uranium concentration of the iron obtained was 0.01 to 0.015 ppm, which was close to the contamination level of the as-received specimen, while for aluminum it was 3 to 5 ppm, which was a few times higher than the as-received specimen contamination level of∼0.9ppm. To melt a thin aluminum cylinder in a steady state, with this fill ratio of 0.05, instantaneous electrode driving response control was desired. Electroslag refining gave better decontamination and energy economization results than by a resistance furnace.

ISSN:0029-5450    

DOI:10.13182/NT87-A34022

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor

摘要:

AbstractBased on available experimental data on the interaction of sodium bentonite and groundwater, a model has been developed that represents a means of extrapolation from laboratory data to the conditions in compacted bentonite. The basic reactions between sodium bentonite and groundwater are described by an ion exchange model for sodium, potassium, magnesium, and calcium. The model also assumes equilibrium with calcite and quartz. The calculations are carried out for two types of granitic groundwater: the Swiss reference groundwater (ionic strength I = 0.24 M) and the standard Swedish groundwater (I = 0.0044 M). It is calculated that the pore water of compacted sodium bentonite will have a pH of 9.7 and a carbonate activity of 8×10−4M if the dry bentonite is saturated with Swiss reference groundwater; it will have a pH near 10.2 and {} = 8×10−3M for standard Swedish groundwater. The long-term situation, which is important for nuclear waste disposal, is modeled by the assumption that the near field of a radioactive waste repository behaves like a mixing tank. It is calculated that sodium bentonite will be slowly converted to calcium bentonite over a long period. The model is used to calculate short- and long-term maximum solubilities of thorium, uranium, neptunium, plutonium, americium, and technetium in the near-field pore water of a potential Swiss nuclear waste repository. The redox potential in the near field is assumed to be controlled by the corrosion products of the iron canister. Using a conservative chemical thermodynamic data base, the maximum solubility of thorium is calculated to be between 2×10−10and 10−8M, that of uranium between 3×10−11and 3×10−8M, that of neptunium between 10−9and 10−5M, that of plutonium between 3×10−10and 4×10−5M, that of americium between 2×10−7and 5×10−5M, and that of technetium will not exceed 10−9M.

ISSN:0029-5450    

DOI:10.13182/NT87-A34023

出版商:Taylor&Francis    

年代:1987

数据来源: Taylor