[en] Complete text of publication follows. Tectonic processes affect the creation, destruction, and mobilization of magnetic materials within the lithosphere of the Earth. The scale of these processes dictates the appropriate observation platform. In the near surface, volcanism and related igneous processes such as dike emplacement, rifting, and faulting act to modify pre-existing magnetic signatures, thus providing critical details of the processes involved. Within the mantle and deep crust, subduction processes produce serpentinite by the dewatering of subducting slabs. This produces highly magnetic but light material, and affects not only magnetic and gravity signatures, but also seismic behavior and properties of the slab and overlying materials. This has important implications for seismic risk assessment. The thickness of the terrestrial magnetic crust can also be related to tectonic processes. On the largest scale, diffuse plate boundary zones within continents are seen to have thinner crust than continental regions away from these zones. On intermediate scales, thickness variations are associated with rifting, and regions with enhanced heat flows.

[en] The interest in the application of heat pipes for heat transfer system is well known in industrial fields. Heat pipe uses the working fluid in containment as cylindrical shape tube. Vaporization occurs in evaporation section due to the heat input and vapor is transferred to condensation section. At the condensation area, the working fluid is condensed and immersed in the wick structure, which has highly porous media. The condensed working fluid returns to evaporator section by capillary wicking of wick structure. The driving force for working fluid is affected by capillary and gravitational force. The heat pipes for nuclear systems have been suggested as horizontal loop heat pipes for reactor core cooling system or vertical heat pipes for passive cooling for spent fuel. In the present research, preliminary tests of horizontal and vertical heat pipe were studied for its heat transfer performance. The main purpose of the research was the analysis of heat transfer behavior of heat pipe and the performance of heat transfer. The thermal performances of horizontal and vertical heat pipe were measured experimentally. Vertical heat pipe showed better performance compared to horizontal one, at high heat input region. The heat transfer coefficients of horizontal heat pipe were lower than vertical one because of gravitational force. Overall heat transfer coefficient of vertical heat pipes were enhanced to 28.5 % compared to the horizontal heat pipes. The horizontal heat pipes revealed high thermal resistance up to 54.3 % compared to vertical heat pipes. Therefore, vertical heat pipes analyzed better heat transfer performance than horizontal heat pipe

[en] A separate effect test facility consisting of hot-leg, surge line and pressurizer is installed to investigate the gas flow into the pressurizer and the entrained water holdup in the pressurizer. The collapsed and mixture levels are measured with changes of gas flow rate during the liquid holdup process. Onset of liquid collapse, CCFL and frictional drop in the surge line are examined during the water collapse process. Scaling analysis is performed to have scale similarities between test facility and real plant. CCFL and velocity similitude are applied to geometric scale parameters in the test facility. Scale similarity for the collapsed and mixture levels are examined. The collapsed level has a similarity from the present scaling methodologies. The mixture level also has a similarity in case that the void fraction is preserved. Preliminary experimental results are obtained for the liquid holdup process. The collapsed level becomes a control parameter instead of the water level in hot-leg together with the gas flow rate

[en] The objective of the present report is to document the design of our first experiment to measure generic flow phenomena expected to occur in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In the process, fabrication sketches are provided for the use of CFD (computational fluid dynamics) analysts wishing to employ the data for assessment of their proposed codes. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. One aspect of the complex flow in a prismatic VHTR is being addressed: flow and thermal mixing in the lower plenum (''hot streaking'' issue). Current prismatic VHTR concepts were examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses were applied to determine key non-dimensional parameters and their magnitudes over this operating range. The flow in the lower plenum can locally be considered to be a situation of multiple jets into a confined crossflow--with obstructions. Flow is expected to be turbulent with momentum-dominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls--with near stagnant surroundings at one extreme and significant crossflow at the other

[en] Research and development for the safety of liquid metal fast breeder reactor with metallic fuel have been carried out for many years. They are mainly related to rapid decrease of fission reaction due to core inherent safety during ATWS events and to feasible decay heat removal way in a passive manner. The inherent safety of metallic fuel was proved by the EBR-II test in April, 1986; however, the core inherent safety and the performance and reliability of passive safety system were not verified yet with respect to commercial reactor design. Since no advanced liquid metal reactors adopting the concept of core inherent safety and passive safety system have been built in the world, the information about safety analysis bases and safety acceptance criteria is deficient. KAERI (Korea Atomic Energy Research Institute) has developed the conceptual design of a prototype reactor of KALIMER (Korea Advanced LIquid Metal Reactor). Since the KALIMER design is under development, performance bases may significantly change as design goes on; however, the safety related design bases will not change much. In this report, preliminary safety analysis bases applicable to the KALIMER conceptual design have been drawn from the experience of design and licensing of abroad liquid metal reactors. Containment accidents and hypothetical core disruptive accidents are out of scope of this report

[en] This paper presents safety aspects of the reactor SVBR-75/100. Information and technical drawings are given on design features, main performances, reactor arrangement, refuelling, multi-fuel capacity,safe reactivity control algorithm, autonomous heat removal, SG localizing system, passive heat sink, core blockades, reactivity incidents and prospects. (A.L.B.)

[en] The American National Standard for Decay Heat Power in Light Water Resources (ANSI/ANS-5.1) was issued in 1979 and has been used extensively for the past decade. Since the standard was issued, there have been new decay-heat measurements, and new and improved calculational capabilities have been developed. In this article the standard is compared with these new data and with calculations based on improved methods. Three foreign standards or proposed standards have also been developed, and their contents are compared with the present American standard. Proposals for improving the standard are presented and discussed. 27 refs., 9 figs., 3 tabs

[en] Residual Heat Removal (RHR) pumps installed in pressurized water reactor power plants are used to provide the removal of decay heat from the reactor and to provide low head safety injection in the event of loss of coolant in the reactor coolant system. These pumps are subjected to rather severe temperature and pressure transients, therefore, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. RHR pumps have traditionally been a significant maintenance item for many utilities. The close-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. The casing separation requires the loosening of numerous highly torqued studs. Once the casing is separated, the impeller is dropped from the motor shaft to allow removal of the mechanical seal and casing cover from the motor shaft. Galling of the impeller to the motor shaft is not uncommon. The RHR pump internals are radioactive and the separation of the pump casing to perform routine maintenance exposes the maintenance personnel to high radiation levels. The handling of the impeller also exposes the maintenance personnel to high radiation levels. This paper introduces a design modification developed to convert the close-coupled RHR pumps to a coupled configuration

[en] Conclusions: • The demonstration of the «Non credibility» of the situation related to the long - term loss of DHR function is organized through: – Probabilistic assessment approach; – Demonstration of negligible risk; – Probabilistic goals. • Results show the inadequacy of design measures to meet the safety requirement of 10^-7/reactor year: – System redundancies and configuration. • Results subject to the assumptions taken in the analysis: – Lack of statistically reliable data for LMRs; – Level of definition of the systems, which are not yet established; – Conservative value of the frequency of the initiator, corresponding to the normal shutdown. • Results show the relevance of CCFs; • Other provisions that could justify the “practical elimination”: – Diversification of components to cope with CCFs; – DHR function through vault cooling

[en] The station blackout accident is one of the most concerning scenarios for the KALIMER design. It is a limiting scenario in a loss of heat removal, and the scenario also provides a measure for the effectiveness of the passive safety system which ultimately secures the KALIMER safety. When the accident occurs, all the pumps in both the primary and intermediate systems are tripped, and thereby the normal heat removal path through the IHX is not available. The path for the heat removal, therefore, is made only through the PDRC (Passive Decay Heat Removal Circuit), while the core cooling becomes solely dependent on the natural circulation in the primary system. Since the PDRC does not come into normal operation before sodium begins over-flowing from the hot pool into the cold poor, the safety analysis in the early stage before the PDRC activation is of great concern in order to assess the timing of the PDRC engagement. The most dominant factor affecting the transient is natural circulation, and the core flow is determined from the friction and heat transfer of the core wire-wrapped rods. For this reason, they should be evaluation objects in this study. The safety margins obtained in the study will provide a quantitative basis for the KALIMER safety assurance