The code RELAP5-3D is used for the analysis of a 0.5% cold-leg Small-Break Loss-of-Coolant Accident (SBLOCA) experiment PL-12 conducted at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility with the passive core cooling injection. The code predictions include the analysis for primary system pressure, loop flow rate, loop and CMT temperatures, coolant inventory distribution in pressurizer, Accumulator (ACC), and Core Makeup Tank (CMT), mass flow rate in Passive Residual Heat Removal (PRHR) system, and other core thermal-hydraulics. A comparison between the calculated results and the experiment data shows (a) a good match with the prediction of the primary system pressure, CMT-1 upper portion temperature, and loop-3 hot-leg fluid temperature, (b) underprediction of the loop-3 mass flow rate and CMT-1 lower portion temperature, (c) overprediction of the pressurizer liquid level, CMTs liquid level and PRHR system mass flow rate.

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In liquid-metal-cooled fast reactors (LMFRs), the occurrence of hypothetical core disruptive accidents (HCDAs) has been considered. Although the probability of occurrence of HCDAs is quite low, it is indispensable to evaluate of the potential risk of HCDAs. To reduce further the potential risk of HCDAs, the core concept that molten fuel will be discharged from the core region before occurrence of an energetic re-criticality is desirable. As a basic study to establish the core concept, we have previously reported a mechanism on breakup of molten metal jet penetrating a sodium pool at instantaneous contact interface temperatures below its freezing point. In the present study, we carried out a series of experiments to confirm a mechanism of thermal fragmentation of a single molten drop penetrating a sodium pool, which is important to understand the fragmentation of molten metal drops after breakup of molten metal jet. A single molten copper drop from 5g to 0.25g was dropped into the sodium pool through the argon gas atmosphere. Thermal fragmentation originating inside the molten copper drop with a thin solid crust at its lower surface was clearly observed at all runs of 5.0g molten drop and 1.2g molten, drop. At runs of 0.5g molten drop and 0.25g molten drop with almost its freezing point, a single debris which was a shell structure of irregular shape with a large open mouth, was obtained even at low sodium temperatures less than 433K. In this way, even the initial temperatures are almost its freezing point, the molten drop fragments. In the present experimental condition, latent heat contributes to fragmentation. Thermal fragmentation of a single molten drop is caused by boiling of sodium entrapped into the molten drop and absorbed latent heat. According to our previous study, this sodium entrapment occurs because of sodium micro jet driven into the upper surface of the molten drop. In order to confirm this action, an experiment to measure impact of the micro jet driven into the upper surface of the molten drop penetrating a water pool was carried out using a strain gauge. The instantaneous pressure rise from 0.03 atm to 0.04 atm by the micro jet driven into the upper surface of the drop was detected. The dissolution of argon gas into molten copper drop was also investigated. The cross section of the frozen drops without fragmentation showed no trace of argon gas bubbles. Therefore, the dissolution of argon gas essentially does not affect fragmentation observed.

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The steam injector, which attains higher discharge pressure than the supply steam pressure, is a passive jet pump that has no movable part and drives feedwater by supersonic jet. Applying the multistage steam injector to a feedwater heating system of a nuclear power plant could be of great advantage for simplification of the feedwater heating system and reduction of the volume of the turbine building. It is remarkably important to improve the thermal-hydraulic performance of the first-stage steam injector to enhance the thermal efficiency of the plant, since it utilizes low-pressure steam of 0.05MPa, extracted from the turbines. The analytical simulation for the four-stage steam injector system by CFD could realize the enhancement of the thermal-hydraulic performance of the multistage steam injector by its design improvement. Further issues exist in the application of reduced-scale experimental and analytical results for the design of equipment as large as an actual plant. A-fifth-scaled test facility, consisting of the three-stage steam injectors and the last stage steam injector with the jet deaerator, enlarges by two times in a flow rate as large as a-seventh-scaled test facility. The test results and analyses exhibited the similarity in the performance of both facilities. With this in view, we developed a four-stage simplified feedwater system using the steam injectors as shown below and proved its viability by a-seventh-scaled and a-fifth-scaled model tests in a bid to significantly rationalize the BOP for the turbine feedwater and condensation system. The CFD code was applied to the multistage SI system to analyze the whole performance from the first-stage SI through the third-stage SI. A-fifth-scaled test facility, consisting of the three-stage steam injectors and the last stage SI with the jet deaerator, enlarges by two times in a flow rate as large as a-seventh-scaled test facility. The test results and analyses exhibited the similarity in the performance of both facilities. The predicted discharge pressure and temperature of the SI by the CFD analyses for a full-size model showed almost same values as those of scaled models. The similarity could be valid in the large scaled-up multistage SI system.

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At present, there are two possible approaches which are used in practice for the thermal-hydraulic analysis of the reactor core in simulations with relevant crossflow effects are : the representation of the core by a system of separated parallel flow subchannels with provision of a forced crossflow mixing by implementation of additional formulations and the porous body approach, where the core geometry is replaced by a structure of homogenized zones of porous media. If crossflow is comparable with flow along the bundles, the porous body approach is the only one suitable method to perform steady-state and transient flow calculations for safety analyses. The CFD - code CFX-4 offers the porous region model for the modeling of the core geometry as a homogenized medium. This model is characterized by a set of properties (volume porosity, resistance to flow and so on). The Core Crossflow Experimental Facility (CCEF) was built with the objective to obtain experimental data for the flow in a rod bundle under the conditions of forced crossflow with relatively low Reynolds numbers and variable flow angle. The test section of CCEF (a plexi-glass model) contains the test rod bundle. The test rod bundle is based on the geometry of a typical PWR with a pitch/diameter ratio of 1.33 and was scaled with a factor 1.5. The crossflow was induced by an asymmetrical outlet condition for the bundle. The test bundle consists of the 100 plexiglass rods, arranged in 4 rows. Additional internals, which are typical for a reactor core geometry, were not installed in the test section. The working fluid in the experiments was water. For the velocity measurements the one-component laser-Doppler anemometer (LDA) was used. Calculations for the comparison with experimental data were performed in two different approaches. The detailed geometry was created for the calculation of flow fields with κ-ε turbulence model and a relative simple geometry of the test section was built by means of the CFX-4 PreProcessing for the flow simulations in the porous region approach. The results of the experimental investigations on the CCEF and the comparison with the calculations, performed with CFX-4 Code in the geometry of the test section are presented in this paper.

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If cooling is inadequate during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 accident. In such a case, concerns about containment failure and associated risks can be eliminated if it is possible to ensure that the lower head remains intact so that relocated core materials are retained within the vessel. Accordingly, in-vessel retention (IVR) of core melt as a key severe accident management strategy has been adopted by some operating nuclear power plants and planned for some advanced light water reactors. However, it is not clear that currently proposed external reactor vessel colling (ERVC) without additional enhancements can provide sufficient heat removal for high power reactors (i.e., reactors with power levels above 1000 MWe). Consequently, a joint United States/Korean International Nuclear Energy Research Initiative (I-NERI) has been launched to develop recommendations to improve the margin for IVR in high power reactors. This program is initially focussed on the Korean Advanced Power Reactor-1400 MWe (APR1400) design. However, recommendations will be developed that can be applied to a wide range of existing and advanced reactor designs. The recommendations will focus on modifications to enhance ERVC (improved data, vessel coatings to enhance heat removal, and an enhanced vessel/insulation configuration to facilitate water ingress and steam venting) and modifications to enhance in-vessel debris coolability (enhanced in-vessel core catcher configuration, thickness, and material). In this paper, late-phase melt conditions affecting the potential for IVR of core melt in APR1400 were established as a basis for developing the I-NERI recommendations. The selection of 'bounding' reactor accidents, simulation of those accidents using the SCDAP/RELAP5-3D^[○!C]code, and resulting latephase melt conditions are presented.

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An enhanced in-vessel core catcher is being designed and evaluated as part of a joint U.S. -Korean International Nuclear Engineering Research Initiative (INERI) investigating methods to insure in-vessel retention of materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. For reactor designs with penetrations, the core catcher is manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary) : a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optinal coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Key properties of possible core catcher base and coating materials were reviewed; and a set of candidate materials was identified based on cost, material properties (melting temperature, ultimate strength, thermal conductivity, resistance to thermal shock, coefficients of linear expansion), and the potential for chemical interactions. Scoping thermal and structural analyses were completed to obtain additional insights about the thickness and type of material that should be selected for each layer of the core catcher. Scoping flow analyses were also completed to determine the impact of this core catcher on reactor vessel coolant flow. Last, scoping materials interaction tests were completed to determine if iron oxide forms in the presence of more promising candidate oxide coating materials, the temperature at which such iron oxide forms, and the ability of iron oxide formation to liquefy candidate oxide coatings. Results from these scoping analyses suggest that thermally-sprayed magnesium oxide is the most promising candidate for a core catcher coating. SCDAP/RELAP5-3D^[○!C]thermal analysis results also indicate that the performance of the core catcher is not significantly impacted by the type of steel (SA304 or SA533B1) selected for the base material, the thickness of the base or coating material, or the porosity of the coating material. However, structural analyses results suggest that the core catcher's base material should be at least 2 cm thick to support the loads associated with relocated materials during a severe accident. Last, analyses indicate that the impact of the core catcher on reactor vessel coolant flow is negligible.

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Fluent and RELAP5-3D^[○!C]/ATHENA are coupled using a technique that permits implicit inter-actions between them using an Executive Program (Weaver, Tomlinson, & Aumiller, 2002). Hence, if necessary, the Executive will allow Fluent and RELAP5-3D^[○!C]/ATHENA to move forward in calculation space on a time-step-by-time step basis. In addition, the point to three-dimension neutronics/reactor kinetics subroutine in RELAP5-3D^[○!C]/ATHENA (based on NESTLE) can be used, by itself, together with Fluent so a three-dimensional fluids model can be coupled with a three-dimensional neutronics model while the balance of the RELAP5-3D^[○!C]/ATHENA subroutines are used to model the system piping and other system components. Fluent and RELAP5-3D[○!C]/ATHENA were linked using an Executive Program (see Figure 1) that (a) monitors the calculational progression in each code, (b) determines when each code has converged, (c) governs the information interchanges between the codes, and (d) issues permission to allow each code to progress to the next time step. The Executive Program was interfaced with Fluent and RELAP5-3D^[○!C]/ATHENA using user-defined functions. User-defined functions were also used to ensure the fluid properties used by Fluent and RELAP5-3D^[○!C]/ATHENA are equivalent. The Executive Program uses the Parallel Virtual Machine (PVM) as the control medium. The Executive Program interacts with Fluent and RELAP5-3D^[○!C]/ATHENA and governs the interactions between Fluent and RELAP5-3D^[○!C]/ATHENA since the two codes are each independent domains. As noted in Weaver, Tomlinson, and Aumiller, 2001^* : "..volume 1 is adjacent to and connected to volume I, and volume 2 is adjacent to and connected to volume II. The boundary volumes in one of the domains (i.e., 1 and 2) represent normal volumes in the interior of the other computational domain (i.e., I and II). Information about these volumes must be passed between the domains at the coupling boundary to achieve an integrated solution." Using the above approach, the domains can be coupled explicitly or semi-implicitly depending on the problem type.

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An experimental study of the effect of flow obstacles on critical heat flux (CHF) has been conducted recently in horizontal and vertical tubes cooled with water and the refrigerant R-134a. The objective of this study is to compare the CHF experimental data obtained in horizontal and vertical tubes (with upward flow) cooled with water and R-134a, in order to assess the effects of orientation and fluid type on CHF enhancement. The investigated ranges of flow parameters in water (the R-134a equivalent values are given in brackets and have been transformed using CHF fluid-to-fluid modeling relationships) are an outlet pressure of 10 (1.67)MPa, mass fluxes from 1000 (705) to 6515 (4605)kg/m^2s, and critical qualities from-0.1 to +0.4. A comparison of the CHF enhancement data for a horizontal tube with those for a vertical tube shows a strong orientation effect on CHF enhancement for R-134a within the investigated range. In general, at equivalent flow conditions, CHF enhancement in water and R-134a flowing in vertical tubes is about the same at high mass fluxes. However, at lower mass fluxes, CHF enhancement in water is lower than that in R-134a. On the other hand, in horizontal tubes and at equivalent flow conditions, CHF enhancement in water is noticeably higher than that in R-134a at high mass fluxes. However, at lower mass fluxes, this trend changes to the opposite (i.e., at higher positive qualities, CHF enhancement in R-134a can be higher than that in water). Also, the limiting critical quality region can be observed in a R-134a-cooled horizontal CHF-enhanced tube, while it is not observed in the enhanced tube cooled by water.

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Systems analysis codes, such as RELAP5-3D^[○!C], are generally aimed at modeling the behavior of an entire system such as a pressurized water reactor plant. While such codes generally have the capability to model multi-dimensional effects, their capacity to produce widely accepted analyses of multi-dimensional behavior is limited by the assumptions and capabilities that stem from their field equation formulations. Historically RELAP5-3D^[○!C]was developed first to analyze the behavior of two-phase systems that could be modeled in one-dimension. Because of the need to analyze two-phase flow, the assumptions used to define the field equations resulted in a simplification of the viscous stress terms and the use of many empirical relationships that cannot be traced to first-principles, e.g., flow regime transitions and the models describing the interactions between phases. The RELAP5-3D^[○!C]field equation set was later extended to analyze two- and three-dimensions. However, the assumptions inherent to the one-dimensional equation set were retained. The starting point and the needs that led to the development of codes such as Fluent and RELAP5-3D^[○!C]led to different products with different capabilities, limitations, strengths, and weaknesses. In summary, the fundamental strengths and weaknesses of the Fluent and RELAP5-3D^[○!C]codes, from an analysis perspective, are given in Table 1. The CFD codes are without peer when analyzing the complex flow behavior of single-phase systems in two- or three-dimensions, for either steady-state or transient behavior. The systems analysis codes, such as RELAP5-3D^[○!C], are without peer for analysis of two-phase systems in one-, two-, or three- dimensions.

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The Japan Atomic Energy Research Institute (JAERI) and the High Energy Accelerator Research Organization (KEK) are promoting a plan to construct a neutron scattering facility at the Tokai Research Establishment, JAERI, under the High-Intensity Proton Accelerator Project (the JAERI/KEK Joint Project). In the facility, 1 MW pulsed proton beam from a high-intensity proton accelerator will be injected into a mercury target in order to produce high-intensity neutrons for use in the fields of life and material sciences. The spallation mercury target system is designed for aiming to start its construction from FY2003. A spallation neutron source is proposed using a high-energy (3.0GeV) proton beam with a current of 0.333mA (1MW) and operating at 25Hz with a pulse duration less than 1μs. A cross flow type (CFT) mercury target has been designed in order to distribute mercury flow according to an axial heat generation distribution caused by spallation reaction. The inner structure arrangement of the mercury target vessel was determined in order to realize appropriate mercury flow distribution according to axial heat generation distributin, based on the thermal hydraulic analytical results of 3GeV, 1MW proton beam injection by using the STAR-CD code. This paper presents the final design of the CFT target using blade distributors, CFD analytical results of the CFT target. The flow distribution in the CFT target is one of key feasibility issues in order to effectively produce the CFT target design. The general computational fluid dynamics (CFD) code STAR-CD was used to analyze the time-averaged thermal hydraulics and flow distribution in the CFT target. In the analysis, an inlet temperature of 50℃ and an inlet mercury velocity of 1.0m/s were assumed. As results, a maximum velocity of 2.48m/s was observed near the front end of the outlet plenum and a maximum of 125.5℃ was observed near the beam window where the volumetric heat generation rate was relatively large. The maximum temperature of 125.5℃ is far below the mercury saturation temperature of 356℃ under atmospheric pressure. This result satisfied the thermal-hydraulic design criteria of " Maximum mercury temperature in the target shall be less than 300℃". A maximum outer surface temperature was obtained at the beam window and was calculated to be 207℃. On the other hand, a maximum inner surface temperature was calculated to be 132℃. A double walled safety hull with water jackets of heavy water is going to cover the mercury target in order to ensure the safety and to collect mercury in a case of mercury leakage caused by the target beam window failure.

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Lead bismuth is to be used as spallation neutron sources and coolant in the sub-critical reactor at the accelerator driven transmutation system (ADS), for reducing the amount of the minor actinide (MA) and the long-lived fission products (LLFP) in the high level radioactive waste. As lead bismuth is used at elevated temperatures, technical issue is to compromise with material performance under flowing lead bismuth and/or with radiation damage by high-energy protons and neutrons. This paper presents a recent activity of target design and lead bismuth technology, which covers flowing tests of lead bismuth, stagnant corrosion tests, oxygen sensor and cleaning techniques. Materials used are SS316.

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My particular view about Russian nuclear energetic perspectives is presented. The nearest and the further perspectives are considered. The arguments are adduced that the most probable scenario of nuclear energetic development is its stabilization in the near future. For further development the arguments of supporters and opponents of nuclear energetic are analyzed. Three points of view are considered. The first point of view that there is not alternative for nuclear energetic. My notes are the following ones. a) I express a skeptic opinion about a statement of quick exhaustion of fossil organic fuel recourses and corresponding estimations are presented. b) It is expressed skeptic opinion about the statement that nuclear energetic can have a visual influence on "team effect". c) I agree that nuclear energetic is the most ecological technology for normal work but however we can't disregard possibilities of catastrophic accidents. The second point of view that the use of nuclear energetic can't have the justification. I adduce the arguments contrary to this statement. The third point of view that nuclear energetic is a usual technology and the only criteria for discussions about what dimension and where one ought develop it is total cost of its unit. Expressed an opinion that the deceived for the choose of a way the skill of the estimate correctly and optimized so named the external parts of the unit energy costs for different energy technologies.

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