[en] The proceedings of this conference are divided into 5 volumes. Volume one is subdivided into Parts A and B. Here in Part A, the following subjects are covered: fundamental thermal hydraulics; structural analysis and design of nuclear facilities; numerical and experimental two-phase flow developments; advances in reactor pressure vessel technology; advances in computational thermal hydraulics; thermal hydraulics of nuclear safety and reliability; advanced reactors thermal hydraulic safety--natural circulation; advances in numerical simulation of steam explosions; computational and experimental two-phase flow developments; two-phase flow and heat transfer fundamentals; advances in stability analysis; and material issues in nuclear applications. Separate abstracts were prepared for most papers in this volume

[en] The proceedings of this conference are divided into five volumes. This volume is divided into the following sections: operation and maintenance; accident simulation and analysis; stability analysis; Korea Project -- safety analysis and radiation release; severe accident analysis; severe accident features; severe accident management; emerging risk-based PRA applications; PSA/PRA applications; PSA -- applications and procedures; PSA -- procedure and policy; dynamic PSA applications; and PSA -- policy and miscellaneous. Separate abstract were prepared for all papers in this volume

[en] The proceedings of this conference are divided into five volumes. This volume covers the following topics: operational experience; maintenance methods and programs; organization, management, QA, and human factors; performance monitoring; instrument and control systems; analytical methods; risk methodology; plant and system improvements; fuel cycle technologies; fuel cycle related technologies--component applications; enrichment and processing; economics; technical issues in capacity and expansion in Russia; nuclear energy policy advances; energy mix--the role of nuclear energy; latest international developments in non-proliferation and safeguards; plutonium disposition and mixed oxide fuels; aspects of plutonium disposition; and plutonium burning and disposition reactors and accelerators. Separate abstracts were prepared for most papers in this volume

[en] The proceedings from this conference are divided into five volumes. This volume is divided into the following sections: spent fuel and nuclear waste processing; transportation and characterization of nuclear wastes; spent fuel and high level waste storage and disposal; D and D (decontamination and decommissioning) of reactors; D and D of nonreactor facilities; D and D related techniques and technology; assessment and management of aging, degradation, and damage; aging effects on failures; license extension and regulatory issues; assessment of aging effects; global advances in nuclear codes and standards; steam generator reliability; and reliability of reactor vessels and internals. Separate abstracts were prepared for most papers in this volume

[en] The proceedings for this conference are contained in 5 volumes. This volume is divided into the following areas: advanced reactor requirements; advanced reactor design and analysis; arrangement and construction; specific reactor designs; demonstration testing; safety systems and analysis; component demonstration testing; advanced reactor containment design; licensing topics and updates; accelerator applications and spallation sources; and advanced reactor development. Separate abstracts were prepared for most papers in this volume

[en] The proceedings of this conference are divided into 5 volumes. Volume one is divided into Parts A and B. Here in Part B, the following subjects are covered: advances in thermal hydraulic computer codes for reactor analysis; human health effects of low level radiation; advances in liquid metal reactor technology; computational and experimental developments; two-phase flow and heat transfer fundamentals; advances in flow field analysis and testing; advanced reactor thermal hydraulic safety; thermal hydraulics of nuclear safety and reliability; advances in numerical simulation of steam explosions; and computational and experimental two-phase flow developments. Separate abstracts were prepared for most papers in this volume

[en] Experimental study has been performed to study periodical-boiling instability in parallel channels, which occurs at low heat fluxes and is of concern in the start-up process of an SBWR. Three channels, made of glass tubes to facilitate visual observation, are placed in bulk water and are heated with electric heaters inserted in the channels. Two types of channels of different length are used to investigate the effect of unheated risers or chimney on periodical boiling. Relationship of bubble development and flow instability is also studied. It is found that slug bubbles are usually followed by rapid boiling, which results in high fluctuation of the flow. Refilling of the subcooled water from the top causes bubble condensation. Sometimes rapid condensation occurs and results in crack sounds. The longer channel does not necessarily have better coolability and stability characteristics; the average flow rate is actually lower and fluctuation amplitude is higher compared to those for the shorter channel

[en] Integral experiments simulating the postulated accidents in the Westinghouse AP600 reactor have been conducted using the ROSA-V Large Scale Test Facility (LSTF). These experiments allowed the N₂ gas for the pressurization of accumulator tanks to enter the primary system after the depletion of the tank water inventory. The gas migrated into the Passive Residual Heat Removal (PRHR) system heat exchanger tubes and into the Core Makeup Tanks (CMTs), and influenced the performance of these components which are unique to the AP600 reactor. Specifically, the PRHR was disabled soon after the N₂ gas discharge in most of the experiments, although the core decay power was removed well by the steam discharge through the Automatic Depressurization System (ADS) after the PRHR was disabled. The N₂ gas ingress into the CMTs occurred in the experiments with relatively large breaks (≥ 2 inch in equivalent diameter), and contributed to a smooth draindown of the CMT inventory into the primary system

[en] Nuclear power plants designed in India in the last twenty years have been designed for earthquake loading using the current licensing practices. Designers and equipment suppliers have therefore been required to consider seismic loading as a major load case. In India, the nuclear power plants have been seismically qualified using state-of-the-art techniques involving both seismic analysis and testing to ensure that the power plant is capable of safely surviving an earthquake that the plant is likely to experience during their operating life. Guidelines and criteria for meeting the qualification requirements are followed as given in various AERB (Indian Atomic Energy Regulatory Board), NRC, IAEA guides, ASME codes and IEEE standards. In this paper various methods available for qualification of structures, systems, mechanical and electrical equipment are explained. The approach and guidelines used within Indian nuclear industry which are evolved from simple analytical requirements to the more elaborate current requirements involving complex analysis and testing on shake table are also summarized

[en] In a breeder-type nuclear power plant, liquid metal is used as a coolant due to the high heat capacity factor. Also, some proposals for fusion reactor blanket design include liquid metal as a possible coolant. In both cases the understanding of natural circulation of liquid-metal flow behavior is an integral part of the thermal hydraulic analysis, especially under two-phase flow conditions. Experimental investigations have been conducted to study a liquid metal two-phase natural circulation flow system. A lead-bismuth (PbBi) eutectic mixture is used as a working fluid in a heated metal walled natural circulation loop. Gas injection induces natural circulation through the gas-lift mechanism. A real-time neutron radiography system is used to visualize the two-phase mixture, specifically the interface and the flow regime. Measurements of void fraction, void fluctuation and bubble propagation are performed