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AbstractAbstract
[en] The application of multi-concentration group mathematical modeling to the leaching of radionuclide waste-forms which have continuous porous matrix such as cemented waste form is described. The modeling is illustrated analyzing a hypothetical of some transport mechanisms such as molecular diffusion, ionic migration and convective flow for release of interest radionuclide from a solidified waste form which contains discrete particles of radioactive Sr-85 nuclides into the aqueous environment. The group parameters are derived from the classical electrochemistry concept of ion transport in dilute electrolytic solution. The numerical analysis is based on the Crank-Nicolson Implicit Methods which assures the stability of the solution at a practical time step. The results show that, for a short-time period of leaching in demineralized water the leaching behavior follows the predominating diffusion mechanism. After this point, the role of other processes apparent and continue until all radionuclides in the cement waste are leached out when compared to the Semi-Infinite Diffusion model which is based on pure diffusion mechanism. The multi-concentration group model can also be applied to long-term prediction of complicated release mechanisms of the radionuclides in the waste form of a particular disposal environment, unless the variables of interest such as the corrosion rate, the chemical reaction, erosion rate and etc. are determined by experiment or theoretical hypothesis. The appropriate differential equation then can be solved by the same general numerical approach
Primary Subject
Source
1987; 225 p; University Microfilms Order No. 87-12,902; Thesis (Ph. D.).
Record Type
Report
Literature Type
Thesis/Dissertation
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Nimnual, S.
Department of Nuclear Technology, Chulalongkorn Univ., Bangkok (Thailand)1975
Department of Nuclear Technology, Chulalongkorn Univ., Bangkok (Thailand)1975
AbstractAbstract
[en] An experiment was made to measure the dose from a short burst of X-rays in the order of 1 second or less by means of the cadmium sulphide photoconductive cell. If protected from light, the CdS cell has a very high resistance such that it does not discharge a capacitor appreciably. But during irradiation, the resistance decreases temporarily and an amount of charge will leak from the capacitor through the Ca S cell. The result to this experiment shows that the principle works very well but it is necessary to add another fixed high resistance of about 107 ohms into the circuit in order to get results independent of the dose-rate. The equipment used in this experiment can measure a dose as low as 6 m R
Source
1975; 26 p; Chulalongkorn University; Bangkok (Thailand); Available from Graduate School, Chulalongkorn Univ., Bangkok (TH); Thesis (Master Eng.)
Record Type
Miscellaneous
Literature Type
Thesis/Dissertation; Numerical Data
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Yang, J.W.; Nimnual, S.
Brookhaven National Lab., Upton, NY (USA). Funding organisation: Nuclear Regulatory Commission, Washington, DC (USA)1991
Brookhaven National Lab., Upton, NY (USA). Funding organisation: Nuclear Regulatory Commission, Washington, DC (USA)1991
AbstractAbstract
[en] Hydrogen transport and combustion in a PWR dry containment are analyzed using the CONTAIN code for a multi-compartment model of the Zion plant. The analysis includes consideration of both degraded core and full core meltdown accidents initiated by a small break LOCA. The importance of intercell flow mixing on distributions of gas composition and temperature in various compartments are evaluated. Thermal stratification and combustion behavior are discussed. 4 refs., 8 figs., 2 tabs
Primary Subject
Source
1991; 14 p; CONTRACT AC02-76CH00016; OSTI as DE91010977; NTIS; INIS; US Govt. Printing Office Dep
Record Type
Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
[en] The NRC is in the preliminary phase of evaluating the CANDU-3 reactor design. Brookhaven National Laboratory is supplying support in the analysis of the positive void feedback effect known to be present in the CANDU-3 design. This paper presents some results from the WIMSD code that was used to study a representative lattice cell under a voided condition
Primary Subject
Source
Annual meeting of the American Nuclear Society (ANS); Philadelphia, PA (United States); 25-29 Jun 1995; CONF-950601--
Record Type
Journal Article
Literature Type
Conference
Journal
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Yang, J.W.; Musicki, Z.; Nimnual, S.
Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution; Brookhaven National Lab., Upton, NY (United States). Funding organisation: Nuclear Regulatory Commission, Washington, DC (United States)1991
Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution; Brookhaven National Lab., Upton, NY (United States). Funding organisation: Nuclear Regulatory Commission, Washington, DC (United States)1991
AbstractAbstract
[en] Hydrogen issues applicable to PWRs with dry containment designs are reviewed based on existing information from the NRC's severe accident research program. Additional calculations were performed using the CONTAIN code for a multi-compartment model of the Zion plant. The review includes in-vessel and ex-vessel hydrogen generation, time and modes of hydrogen release, hydrogen mixing and transport in the containment, hydrogen combustion mechanism, hydrogen control methods and the equipment survivability. A cost-benefit analysis of the hydrogen ignition system was performed for the Zion and Surry plants. Potential for hydrogen detonation in these plants was evaluated. 47 figs., 36 tabs
Primary Subject
Source
Jun 1991; 199 p; BNL-NUREG--52271; CONTRACT AC02-76CH00016; OSTI as TI91015352; NTIS; INIS; GPO
Record Type
Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Nimnual, S.; Colombo, P.; Wacks, M.E.
Waste management '87: Waste isolation in the US, technical programs, and public education1987
Waste management '87: Waste isolation in the US, technical programs, and public education1987
AbstractAbstract
[en] The application of multi-concentration diffusion (MCD) group mathematical modelling to the long-term leaching behavior of cement-based radioactive waste forms is described. The modelling is illustrated by analyzing hypothetical cases using transport mechanisms such as molecular diffusion, ionic migration and convective flow for release of SR-85 from a solidified waste form. The group parameters, which describe the transport mechanisms, are derived from the classical electrochemistry concept of ion transport in dilute electrolytic solutions. The numerical analysis is based on the Crank-Nicolson Implicit Method, which assures the stability of the equation at practical time intervals. The boundary conditions are based on the experimental results. The results show that, for a short time period of leaching in demineralized water, the leaching behavior follows the predominating diffusion mechanism. After this point, the role of other processes are apparent and continue until all radionuclides in the cement waste form are leached. When compared to the Semi-Infinite Diffusion model, which is based on a pure diffusion mechanism, the MCD model predicts higher leach rates
Primary Subject
Secondary Subject
Source
Post, R.G; vp; 1987; vp; University of Arizona; Tucson, AZ (USA); Waste management '87; Tucson, AZ (USA); 1-5 Mar 1987
Record Type
Book
Literature Type
Conference; Numerical Data
Country of publication
AQUEOUS SOLUTIONS, BOUNDARY CONDITIONS, CEMENTS, DIFFUSION, ELECTROCHEMISTRY, EQUATIONS, EXPERIMENTAL DATA, FEASIBILITY STUDIES, GROUTING, IONS, LEACHING, LOW-LEVEL RADIOACTIVE WASTES, MATHEMATICAL MODELS, NUMERICAL SOLUTION, RADIOACTIVE WASTE DISPOSAL, RADIONUCLIDE MIGRATION, SOLID WASTES, STRONTIUM 85
BETA DECAY RADIOISOTOPES, BUILDING MATERIALS, CHARGED PARTICLES, DATA, DAYS LIVING RADIOISOTOPES, DISPERSIONS, DISSOLUTION, ELECTRON CAPTURE RADIOISOTOPES, ENVIRONMENTAL TRANSPORT, EVEN-ODD NUCLEI, HOMOGENEOUS MIXTURES, HOURS LIVING RADIOISOTOPES, INFORMATION, INTERMEDIATE MASS NUCLEI, ISOMERIC TRANSITION ISOTOPES, ISOTOPES, MANAGEMENT, MASS TRANSFER, MATERIALS, MIXTURES, NUCLEI, NUMERICAL DATA, RADIOACTIVE MATERIALS, RADIOACTIVE WASTES, RADIOISOTOPES, SEPARATION PROCESSES, SOLUTIONS, STRONTIUM ISOTOPES, WASTE DISPOSAL, WASTE MANAGEMENT, WASTES
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Jo, J.; Lin, C.C.; Mufayi, V.; Neymotin, L.; Nimnual, S.
Brookhaven National Lab., Upton, NY (United States). Funding organisation: Nuclear Regulatory Commission, Washington, DC (United States)1992
Brookhaven National Lab., Upton, NY (United States). Funding organisation: Nuclear Regulatory Commission, Washington, DC (United States)1992
AbstractAbstract
[en] Studies suggest that the risk of severe accidents during low power operation and/or shutdown conditions could be a significant fraction of the risk at full power operation. The Nuclear Regulatory Commission has begun two risk studies to evaluate the progression of severe accidents during these conditions: one for the Surry plant, a pressurized water reactor (PWR), and the other for the Grand Gulf plant, a boiling water reactor (BWR). This paper summarizes the approach taken for the Level 2/3 analysis at Surry for one plant operating state (POS) during shutdown. The current efforts are focussed on evaluating the risk when the reactor is at mid-loop; this particular POS was selected because of the reduced water inventory and the possible isolation of the loops. The Level 2/3 analyses are conditional on core damage having occurred. Initial results indicate that the conditional consequences can indeed be significant; the defense-in-depth philosophy governing the safety of nuclear power plants is to some extent circumvented because the containment provides only a vapor barrier with no capability for pressure holding, during this POS at Surry. However, the natural decay of the radionuclide inventory provides some mitigation. There are essentially no predicted offsite prompt fatalities even for the most severe releases
Primary Subject
Source
1992; 13 p; 20. water reactor safety information meeting; Bethesda, MD (United States); 21-23 Oct 1992; CONF-921007--25; CONTRACT AC02-76CH00016; OSTI as DE93013342; NTIS; INIS; US Govt. Printing Office Dep
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Jo, J.; Lin, C.C.; Nimnual, S.; Mubayi, V.; Neymotin, L.
Brookhaven National Lab., Upton, NY (United States). Funding organisation: Nuclear Regulatory Commission, Washington, DC (United States)1992
Brookhaven National Lab., Upton, NY (United States). Funding organisation: Nuclear Regulatory Commission, Washington, DC (United States)1992
AbstractAbstract
[en] Studies and operating experience suggest that the risk of severe accidents during low power operation and/or shutdown (LP/S) conditions could be a significant fraction of the risk at full power operation. Two studies have begun at the Nuclear Regulatory Commission (NRC) to evaluate the severe accident progression from a risk perspective during these conditions: One at the Brookhaven National Laboratory for the Surry plant, a pressurized water reactor (PWR), and the other at the Sandia National Laboratories for the Grand Gulf plant, a boiling water reactor (BWR). Each of the studies consists of three linked, but distinct, components: a Level I probabilistic risk analysis (PRA) of the initiating events, systems analysis, and accident sequences leading to core damage; a Level 2/3 analysis of accident progression, fuel damage, releases, containment performance, source term and consequences-off-site and on-site; and a detailed Human Reliability Analysis (HRA) of actions relevant to plant conditions during LP/S operations. This paper summarizes the approach taken for the Level 2/3 analysis at Surry and provides preliminary results on the risk of releases and consequences for one plant operating state, mid-loop operation, during shutdown
Primary Subject
Secondary Subject
Source
Nov 1992; 12 p; Probabilistic safety assessment international topical meeting (PSA 93); Clearwater Beach, FL (United States); 27-29 Jan 1993; CONF-930116--40; CONTRACT AC02-76CH00016; OSTI as DE93008052; NTIS; INIS; US Govt. Printing Office Dep
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
[en] MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in nuclear power plants. It is being developed for the US Nuclear Regulatory Commission by Sandia National Laboratories and is designed to provide an improved severe-accident/source term analysis capability relative to the older Source Term Code Package. Early assessment efforts for MELCOR emphasized validation of containment modeling. Current assessment efforts have realized the need to focus more on in-vessel phenomenology, which is an area of major uncertainty in the assessment of severe-accident radiological consequences. In-vessel analyses reported to date using MELCOR have been the Power Burst Facility (PBF) SFD 1-1 test, the Three Mile Island Unit 2 standard problem, both using MELCOR Version 1.7.1, and the PBF SFD 1-4 test, which used MELCOR Version 1.8BC. The purpose of this paper is to describe a MELCOR (Version 1.8DN, the most recent) simulation of the National Research Universal (NRU) Full-Length High-Temperature 2 (FLHT-2) test and to compare results with test data and predictions from the US Nuclear Regulatory Commission's mechanistic code SCDAP
Primary Subject
Secondary Subject
Source
Annual meeting of the American Nuclear Society (ANS); Orlando, FL (United States); 2-6 Jun 1991; CONF-910603--
Record Type
Journal Article
Literature Type
Conference
Journal
Country of publication
BWR TYPE REACTORS, BYPASSES, FISSION PRODUCT RELEASE, FUEL ELEMENT FAILURE, HEAT TRANSFER, HYDRAULICS, HYDROGEN, LOSS OF COOLANT, M CODES, OXIDATION, PRODUCTION, PWR TYPE REACTORS, REACTOR CORE DISRUPTION, REACTOR SAFETY, REACTOR SHUTDOWN, RESEARCH PROGRAMS, S CODES, SANDIA LABORATORIES, SCRAM, SOURCE TERMS, TEST FACILITIES, TIME DEPENDENCE, US NRC
ACCIDENTS, CHEMICAL REACTIONS, COMPUTER CODES, ELEMENTS, ENERGY TRANSFER, ENRICHED URANIUM REACTORS, NATIONAL ORGANIZATIONS, NONMETALS, POWER REACTORS, REACTOR ACCIDENTS, REACTORS, SAFETY, SANDIA NATIONAL LABORATORIES, SHUTDOWN, THERMAL REACTORS, US AEC, US DOE, US ERDA, US ORGANIZATIONS, WATER COOLED REACTORS, WATER MODERATED REACTORS
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Madni, I.K.; Nimnual, S.; Foulds, R.
Brookhaven National Lab., Upton, NY (United States). Funding organisation: Nuclear Regulatory Commission, Washington, DC (United States)1993
Brookhaven National Lab., Upton, NY (United States). Funding organisation: Nuclear Regulatory Commission, Washington, DC (United States)1993
AbstractAbstract
[en] This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock ampersand Wilcox (B ampersand W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides
Primary Subject
Secondary Subject
Source
1993; 9 p; Probabilistic safety assessment international topical meeting (PSA 93); Clearwater Beach, FL (United States); 27-29 Jan 1993; CONF-930116--41; CONTRACT AC02-76CH00016; OSTI as DE93009539; NTIS; INIS; US Govt. Printing Office Dep
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
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