[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

[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

[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

[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

[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

[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

[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

[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

[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

[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