Showing papers in "Nuclear Technology in 1989"
TL;DR: The Integral Fast Reactor (IFR) as discussed by the authors is an innovative liquid metal reactor concept being developed at Argonne National Laboratory, which aims to exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system.
Abstract: The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired.
237 citations
TL;DR: The primary objective of the U.S. Department of Energy Three Mile Island Unit 2 (TMI-2) research program, conducted by the EG&G Idaho TMI-1 Accident Evaluation Program, is to develop a comprehensiv...
Abstract: The primary objective of the U.S. Department of Energy Three Mile Island Unit 2 (TMI-2) research program, conducted by the EG&G Idaho TMI-2 Accident Evaluation Program, is to develop a comprehensiv...
141 citations
TL;DR: In this paper, a neutronic study of an accelerator-based neutron irradiation facility (ANIF) for boron neutron capture therapy (BNCT) was performed using three-dimensional Monte Carlo transport calculations.
Abstract: A neutronic study of an accelerator-based neutron irradiation facility (ANIF) for boron neutron capture therapy (BNCT) was performed using three-dimensional Monte Carlo transport calculations The major components of the ANIF are a radio-frequency quadrupole proton accelerator, a /sup 7/Li target, and a moderator assembly Neutrons are generated by bombarding the /sup 7/Li target with 25-MeV protons The neutrons emerging from the /sup 7/Li target are too energetic to be used for BNCT and must therefore be moderated Calculations show that, among all materials for the ANIF, beryllia (BeO) and heavy water (D/sub 2/O) are the best moderators Between them, beryllia provides better neutron spectra, but D/sub 2/O gives higher neutron intensities Adding alumina (Al/sub 2/O/sub 3/) to D/sub 2/O improves the neutron spectra, but it also increases gamma-ray contamination
96 citations
TL;DR: In this paper, the performance of particles at extremely high temperatures have been investigated to achieve an understanding of coating failure mechanisms and to establish the data base for safety and risk analyses of hypothetical accidents in large and medium-sized HTRs.
Abstract: Coated particles embedded in graphitic elements are the fuel for the High-Temperature Reactor (HTR) Experimental investigations of the performance of particles at extremely high temperatures have been conducted to achieve an understanding of coating failure mechanisms and to establish the data base for safety and risk analyses of hypothetical accidents in large- and medium-sized HTRs The primary mechanism for coating failure and fission product release in the 1900 to 2500/sup 0/C temperature range is thermal decomposition of silicon carbide (SiC) Heating tests have provided the activation energy of this process and the correlation of SiC decomposition with coating failure and subsequent fission product release
67 citations
TL;DR: In this article, a light water reactor fuel rod bundle containing Ag-In-Cd absorber rods or Al203/B4C burnable poison rods with increasing temperature up to the complete melt is presented.
Abstract: Chemical interactions that may occur in a light water reactor fuel rod bundle containing Ag-In-Cd absorber rods or Al203/B4C burnable poison rods with increasing temperature up to the complete melt...
62 citations
TL;DR: In this paper, the first phase in the investigation of the feasibility of storing light water reactor spent fuel in air, oxidation tests were performed on nonirradiated UO2 pellets over the temperature range of 150 to 345°C.
Abstract: As a first phase in the investigation of the feasibility of storing light water reactor spent fuel in air, oxidation tests were performed on nonirradiated UO2 pellets over the temperature range of 150 to 345°C. The objective of the tests was to determine the important independent variables that affect the oxidation behavior of fuel. Pellets tested at the high end of the temperature range (>230°C) oxidized very rapidly from the standpoint of projected storage periods in air. These results suggest that acceptable spent-fuel storage temperatures should be <230°C. The tests also revealed that the oxidation was initially retarded by the presence of a coating, probably a higher oxide, that formed on pellets during the period of air storage before they were tested. The oxide coating became increasingly semiprotective after longer storage periods. Other variables identified as important to oxidation behavior of fuel were temperature, radiolysis of a static air atmosphere, fuel microstructure, gadolinia content, a...
36 citations
TL;DR: In this paper, core debris samples obtained from different regions of the Three Mile Island Unit 2 (TMI-2) core were examined to characterize the interaction among core components and the coolant, to determine the peak temperatures at which the interactions occurred, and to evaluate core melt progression in TMI2.
Abstract: Core debris samples obtained from different regions of the Three Mile Island Unit 2 (TMI-2) core were examined to characterize the interaction among core components and the coolant, to determine the peak temperatures at which the interactions occurred, and to evaluate core melt progression in TMI-2. Estimates of peak temperatures were needed from these samples because of the strong influence that temperature has on core damage progression and fission product behavior. The peak temperatures can be bounded by comparing the observed microstructure and compositions with established phase diagrams. The microstructures were determined by optical metallography and scanning electron microscopy, and compositions were determined by energy and wavelength dispersive X-ray spectroscopy and scanning Auger spectroscopy.The material interactions among the core components are very complex and involve not only the interaction between the Zircaloy cladding and the UO2 fuel, but interactions with control rod material...
36 citations
TL;DR: In this article, the authors examined the control rod behavior in severe reactor accidents with a goal of improving the methodology used to estimate reactor accident source terms, which was used to improve the estimation of the source terms.
Abstract: Silver-indium-cadmium (Ag-In-Cd) control rod behavior in severe reactor accidents is examined with a goal of improving the methodology used to estimate reactor accident source terms. Control rod be...
35 citations
TL;DR: The modular high-temperature gas-cooled (MHTGR) as discussed by the authors was designed primarily to provide passive safety that will prevent fuel damage over a wide spectrum of accidents.
Abstract: The modular high-temperature gas-cooled reactor (MHTGR) is modularized primarily to provide the passive safety that will prevent fuel damage over a wide spectrum of accidents. Specifically, this ra...
28 citations
TL;DR: Based on the Helmholtz instability at the microlayer/vapor interface as a trigger condition for microlayer dryout, Lee and Mudawwar as discussed by the authors developed a mechanistic critical heat flux (CHF) model for subcooperative dryout.
Abstract: Based on the Helmholtz instability at the microlayer/vapor interface as a trigger condition for microlayer dryout, Lee and Mudawwar developed a mechanistic critical heat flux (CHF) model for subcoo...
28 citations
TL;DR: In this paper, the effects on durability of composition variations in West Valley Nuclear Services Company preliminary waste glass composition WV205 are discussed and the general trends are discussed in terms of the structural roles of the components.
Abstract: In this article the effects on durability of composition variations in West Valley Nuclear Services Company preliminary waste glass composition WV205 are discussed. MCC-3 results at times from 7 to 180 days are presented for 50 glass compositions. The results are suggestive of a large plateau region where durability is good and weakly dependent on composition, adjoining a region in which durability is a much steeper function of composition. The same effect is observed when the redox state of the iron, which comprises --12wt% of the glass, is varied. The general trends are discussed in terms of the structural roles of the components. The effects of the alkalies and alkaline earths correlate quite well with the field strengths of these ions.
TL;DR: A method to expel radioiodine from a spent-fuel solution is important for iodine control in reprocessing plants as mentioned in this paper. But this method has been investigated without considering the influence of external factors.
Abstract: A method to expel radioiodine from a spent-fuel solution is important for iodine control in reprocessing plants. Many authors have investigated the procedure without considering the influence of ot...
TL;DR: Samples of the bores obtained from the melted core of the Three Mile Island Unit 2 (TMI-2) reactor were investigated as part of the TMI2 accident evaluation program.
Abstract: Samples of the bores obtained from the melted core of the Three Mile Island Unit 2 (TMI-2) reactor were investigated as part of the TMI-2 accident evaluation program. The samples included fuel rod ...
TL;DR: In this paper, uncertainty in the estimation of parameters for common-cause failure models arise not only because of the small number of common cause failure events but also because recorded events may not be r...
Abstract: Uncertainties in the estimation of parameters for common-cause failure models arise not only because of the small number of common-cause failure events but also because recorded events may not be r...
TL;DR: In this paper, the collapse of the upper debris bed was the main cause of core failure and core material relocation to the lower vessel plenum during the Three Mile Island Unit 2 (TMI-2) acc...
Abstract: It is postulated that the collapse of the upper debris bed was the main cause of core failure and core material relocation to the lower vessel plenum during the Three Mile Island Unit 2 (TMI-2) acc...
TL;DR: The results of the bulk material examinations performed on samples from the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel (RPV) are summarized and the materials chemistry that resulted i... as discussed by the authors.
Abstract: The results of the bulk material examinations performed on samples from the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel (RPV) are summarized and the materials chemistry that resulted i...
TL;DR: In this paper, a reliability analysis using the GO-FLOW methodology is given for the emergency core cooling system (ECCS) of a marine reactor experiencing either a collision or a grounding accident.
Abstract: A reliability analysis using the GO-FLOW methodology is given for the emergency core cooling system (ECCS) of a marine reactor experiencing either a collision or a grounding accident. The analysis ...
TL;DR: In this article, oxidation tests were conducted with light water reactor spent fuel, and the initial rate of weight gain for spent fuel was up to 50 times greater than the initial rat...
Abstract: To support dry storage technology, oxidation tests were conducted with light water reactor spent fuel. The initial rate of weight gain for spent fuel was up to 50 times greater than the initial rat...
TL;DR: In the short time that plant-specific, full-scope probabilistic risk assessments (PRAs) have been performed, extensive progress has been made in understanding and managing risk.
Abstract: In the short time that plant-specific, full-scope probabilistic risk assessments (PRAs) have been performed, extensive progress has been made in understanding and managing risk. This paper discusses the various lessons learned during this time. Twenty-one specific PRAs are evaluated.
TL;DR: The behavior of melts in severe accident sequences affects the nature (composition and fission product inventory) of the debris released from the vessel upon lower head failure in unmitigated accid...
Abstract: The behavior of melts in severe accident sequences affects the nature (composition and fission product inventory) of the debris released from the vessel upon lower head failure in unmitigated accid...
Journal Article•
TL;DR: In this paper, Colloid formation of uranium, thorium, radium, lead, polonium, strontium, rubidium, and cesium in briny groundwaters is studied to predict their capability as vectors for transporting radionuclides.
Abstract: Colloid formation of uranium, thorium, radium, lead, polonium, strontium, rubidium, and cesium in briny (high ionic strength) groundwaters is studied to predict their capability as vectors for transporting radionuclides. This knowledge is essential in developing models to infer the transport of radionuclides from the source region to the surrounding environment. Except polonium, based on the experimental results, colloid formation of uranium, thorium, radium, lead, strontium, rubidium, and cesium is unlikely in brines with compositions similar to the synthetic Palo Duro Basin brine. This observation of no colloid formation is explained by electrokinetic theory and inorganic solution chemistry.
TL;DR: In this article, a theoretical critical heat flux (CHF) model based on microlayer dryout and Helmholtz instability for subcooled tube flow under pressurized water reactor operation conditions is first extended to t
Abstract: A theoretical critical heat flux (CHF) model based on microlayer dryout and Helmholtz instability for subcooled tube flow under pressurized water reactor operation conditions is first extended to t...
TL;DR: Colloid formation of uranium, thorium, radium, lead, polonium, strontium, rubidium, and cesium in briny groundwaters is studied in this article.
Abstract: Colloid formation of uranium, thorium, radium, lead, polonium, strontium, rubidium, and cesium in briny (high ionic strength) groundwaters is studied to predict their capability as vectors for tran...
TL;DR: Three samples from the ceramic melt, the lower crust, and the lower plenum of the previously molten part of the Three Mile Island Unit 2 core have been analyzed by X-ray diffraction to determine th...
Abstract: Three samples from the ceramic melt, the lower crust, and the lower plenum of the previously molten part of the Three Mile Island Unit 2 core have been analyzed by X-ray diffraction to determine th...
TL;DR: In this article, the authors describe how to recognize immediately periods of continuously decreasing levels of fuel rod integrity in order to prevent complications in routine power plant maintenance as well as accident situations caused by more severe fuel rod degradation.
Abstract: Periods of continuously decreasing levels of fuel rod integrity due to debris-induced cladding damage, vibration-induced fretting wear of the cladding, etc cause difficulties in the assessment of fuel rod performance from coolant activity data The calculational models currently in use for this purpose in nuclear power plants are not sufficiently capable of indicating cases in which they are invalid This can mislead reactor operators by misinterpretation of the coolant activity data, especially in situations where fast reactions are necessary A quick test of validity is suggested to check the applicability of the currently available calculational models for estimating the number and average size of fuel rod defects This paper describes how to recognize immediately periods of continuously decreasing levels of fuel rod integrity in order to prevent complications in routine power plant maintenance as well as accident situations caused by more severe fuel rod degradation
TL;DR: The development and demonstration of a method to establish in-service inspection priorities through the use of probabilistic risk assessment (PRA) results is demonstrated to be a useful tool for identifying systems and piping sections or welds that need to be inspected.
Abstract: Some of the goals of the Nondestructive Evaluation Reliability Program are to assess current inspection requirements for all pressure boundary systems and components, to determine whether improvements to the requirements are needed, and, if necessary, to develop recommendations for revising the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and regulatory requirements Part of the work performed in addressing this goal was the development and demonstration of a method to establish in-service inspection priorities through the use of probabilistic risk assessment (PRA) results The Oconee-3 PRA and the observed weld failure data of the nuclear plants operating in the United States are used to identify and prioritize the most risk-important systems for inspection Failure modes and effects analysis methodology is then used to identify and prioritize the most risk important piping sections of the Oconee-3 emergency feedwater system Based on the results of this study, this method is demonstrated to be a useful tool for identifying systems and piping sections or welds that need to be inspected
TL;DR: In this article, detailed microstructural and microchemical examinations of samples of debris extracted from the lower plenum region of the Three Mile Island Unit 2 reactor were performed using optical and electron microscopes.
Abstract: Detailed microstructural and microchemical examinations of samples of debris extracted from the lower plenum region of the Three Mile Island Unit 2 reactor were performed using optical and electron...