A Very High Temperature Reactor (VHTR) is one of the next generation nuclear systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. We measured the outlet flow rate by hot-wire anemometer which is an omnidirectional spherical probe of diameter 2.5mm. The experiment has been carried out under the condition that a copper wire with a scourer model and a cubic lattice model were inserting into the channel.

Quenching is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident (LOCA) in a nuclear reactor. In the present study, an experimental apparatus is designed and constructed with the purpose of conducting high-temperature transient pool boiling quenching experiments for zirconium (Zr-4) cylindrical test samples. Three thermocouples are inserted in the test sample to investigate the effect of axial distance on the minimum film boiling temperature. The Zr-4 rodlet is heated up to a temperature well above the minimum film boiling temperature (up to 600°C), and then plunged vertically in a quiescent pool of subcooled water. A data acquisition system is used to record the temperature of the embedded thermocouples with time. Data reduction is performed by an inverse heat conduction code to calculate the surface temperature and corresponding surface heat flux. A visualization study with a high-speed camera is conducted to record the quenching behavior on the test sample. It is found that the minimum film boiling temperature decreases with the axial distance, while the CHF temperature is relatively insensitive to the axial distance. The film boiling heat transfer coefficient decreases with surface temperature, and seems to be independent of axial distance. The quench front is observed to originate from the bottom and move upwards. It is found that the quench front velocity remains nearly constant in the lower region of the test sample, and significantly increases in the upper region.

A detailed safety assessment of innovative Generation IV reactor designs with heavy-liquid metal coolants, such as lead and lead-bismuth eutectic (LBE), requires an evaluation of the maximum core temperature in several postulated scenarios. Considering the complex geometry of fuel assemblies (FAs), and the low Prandtl number of the coolants, this flow scenario is challenging for the models used in numerical simulations, e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data are needed for validation. In recent years, a series of comprehensive heat transfer tests in fuel pin bundle simulators was performed at ENEA (Italy) and KIT (Germany) in the framework of the European collaborative projects THINS and SEARCH. Both grid and wire spacer geometries are considered, in a wide range of operating conditions (temperature, flow velocity and power density) representative of the ALFRED and MYRRHA fuel assemblies, in natural and forced-convective flow regimes. Although different experimental approaches were followed by each group (e.g. thermocouple position and average data treatment), there is a relatively good agreement on results in the overlapping regions. These experimental studies indicate that the mean Nusselt number is in well agreement with the predictions of empirical correlations developed for sodium systems. In particular, for wire-spaced FA, heat transfer results show values close to the Kazimi-Carelli correlation both for low and high flow rates at ENEA and KIT respectively. For grid-spaced FA, results are more in agreement with Ushakov correlation. Furthermore, large temperature differences are measured by thermo-couples installed at selected rods and sub-channels. A discussion on the influences of the spacer design and bundle size is included. This wide comparison allows an overview of the research on the HLM cooled fuel assembly in Europe.

An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle simulator, cooled by Heavy Liquid Metal and installed in the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The experimental tests concerned mass flow rate transition of the primary coolant from forced to natural circulation, with fuel pin bundle simulator characterized by non-uniform power distribution. The main objective of the performed experimental campaign was to perform integral system and local thermal-hydraulic analysis, in particular to investigate the flow in different flow regimes and specifically the transition from forced to natural circulation flow and, more specifically, analyze the behavior of the 19-pins wire-spaced fuel pin simulator (FPS) during such transient. Indeed, the performed test were characterized by non-uniform heating of the bundle (i.e. just some pins switched on), so the effects of this non-uniformity on the local temperatures and on the overall system behavior was evaluated. A deep investigation on the local temperature distribution was performed thanks to the accurate instrumentation provided in the bundle (67 thermocouples). For instance, in some cases, the wall temperatures relative to pins switched off remained below the relative sub-channel temperature, depending on the heating distribution. The obtained experimental data provided useful information for the characterization of the bundle and the computation of the heat transfer coefficient. Moreover, the collected system data can be helpful for STH codes validation, whereas the local fuel bundle data, especially the ones from dissymmetric tests can be useful for the qualification and benchmarking of CFD codes and coupled STH/CFD methods for HLM systems.

A Very High Temperature Reactor (VHTR) is one of the next generation nuclear reactor systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. The flow velocity distribution was obtained by a PIV method. In this paper, we discuss the heat transfer and fluid flow characteristics of the proposed channel. From the results obtained in the experiment, it was found that the amount of removed heat decreased with increasing of temperature of gas when a copper wire was inserted into the channel with high porosity. This is because the mass flow rate decreased with increasing of viscosity of gas. Since it is expected that the porosity of a porous material will have an optimum value, further studies will be needed.

In nuclear plant piping, when high-temperature water penetrates from main pipe, a thermal stratification may be formed in a branch pipe with a closed end. The penetrating flow of high-temperature water is called as a cavity flow. If a thermal stratification interface is formed in an elbow of vertical-horizontal branch pipes, thermal fatigue may occur due to periodic temperature fluctuations. Because high cycle thermal fatigue may have significant influence on structure integrity of pipe, it is important to evaluate the position which is called penetration depth of a cavity flow in which a thermal stratification interface is formed. In the present evaluation guideline which was formulated by the Japan Society of Mechanical Engineers, applicable range is limited only to 50 mm diameter of vertical-horizontal branch pipe. Therefore, it needs to expand the applicable diameter range of vertical-horizontal branch pipe. In this research, 200 mm diameter piping test is conducted to confirm characteristic phenomena, and to build evaluation method of penetration depth in large diameter piping. This paper presents experimental results and consideration of thermal hydraulics phenomena about thermal stratification. The horizontal piping test and the vertical-horizontal branch piping test are conducted. Pipes are made of acrylic resin for visualization of water flow. The water temperature is less than 60°C. The horizontal piping test reproduces natural circulation flow by installing a heater which imitated heating from cavity flow, and a chiller which imitated heat dissipation from a blockage valve. Temperature profiles are obtained by using thermocouples and optical fibers. Moreover, water flow in piping is visualized by injected ink. From the visualization and the temperature measuring result in the horizontal piping test, flow characteristic of the natural circulation region of large diameter piping is clarified The vertical-horizontal branch piping test reproduces cavity flow penetration from main pipe. Temperature profiles are obtained at cross-section direction and axial direction. Moreover, the amount of mass transfer at the thermal stratification which is formed between a cavity flow and natural circulation is evaluated experimentally by using a lithium tracer method. The second paper[Ref.2] explains modeling of evaluation method for cavity flow penetration depth in large diameter piping.

In the third generation pressurized water reactor AP1000 plant, the Automatic Depressurization System (ADS) is one of the most important passive safety system. As the steam continues to discharge into the IRWST, the high-temperature and high-pressure steam condenses in the IRWST intensely and rapidly. In the present work, multi-hole spargers DCC experimental bench has been built to study the condensation and mixing phenomena in the accidental depressurization events in AP1000. The key parameters including the temperature, velocity, and flow patterns are monitored by the thermocouples, PIV technique, and high speed camera, etc. Based on the experimental results, the heat transfer characteristics of the steam-jet Direct Contact Condensation (DCC) in the ADS depressurization process were analyzed. The experimental comparison shows that the spraying steam behaviors are closely related to the steam mass flux, which determined the ejection patterns. The spraying steam is condensed rapidly in the core region with the flow regime locating in chugging, transitional chugging, and stable ejection stages. The 88-hole experimental results fit well with Fukuda’s condensation oscillation correlations, whereas the 2-hole experimental results are slightly lower than the values predicted by traditional single hole correlation.

The prediction of the pressurizer response during various transients plays an important role in determining what actions should be taken to control or mitigate the consequence of an accident. For this reason, a range of different experiments have been completed which aim to support a preliminary understanding of the phenomena which are important to the analysis of a PWR pressurizer. As so far, the researches on these subjects are focusing on the low pressure, there are rarely researches on the high pressure. Hence, the effects of a non-condensable gas on pressurizer insurge transients under high pressure are necessary and significant. The study concentrates on the pressure response and heat transfer during an insurge transient under the high pressure. These experiments were run in two cylindrical stainless steel tanks: the primary tank and the storage tank. In this research, the experiments looked at the effects of water level increase and the presence of a non-condensable gas with varying mass fraction. Nitrogen was chosen as the non-condensable gas. The storage tank was filled with water and pressurized with nitrogen. An insurge to the primary tank with hot walls was initiated by opening the high pressure solenoid valve between two tanks. The experiments assigned lots of thermocouples for monitoring the radial and axial temperature distribution. Experiments showed that the non-condensable gas has a substantial influence on the pressure response. And the magnitude of the increase in peak pressure is quite different with the change of experiment initial conditions.

An experimental setup was designed to find out behavior of deposition thickness of corrosion products. This experimental setup is designed on the bases of ITER (International Thermonuclear Experimental Reactor) loop. The experiment was done for different values of pH to study its effect on the deposition thickness. System pressure was kept at 0.5 MPa with maximum limit of temperature at 150°C. The pH value was kept in the range of 7.00–10.50. Artificial corrosion particles (Fe 3 O 4 ) were introduced manually in the loop, which was allowed to stabilize. Four test tubes were used to determine the thickness of deposition, and the temperature at these points were measured by thermocouples. It is also very important to determine the temperature distribution around such points, in order to establish its relationship with pH and deposition thickness. Results show that the change of temperature and pH (of coolant) have a strong effect on deposition thickness. The temperature distribution, found in the various components of the experimental loop, will be used to correlate its relation with the deposition thickness and the coolant pH in future studies.

In the framework of the European Commission LEADER project, an experimental campaign of seven tests was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the Spiral-Tube Steam Generator (STSG) of the European Lead Fast Reactor (ELFR). The LIFUS5/Mod2 facility is composed by a water tank of 15 L injecting subcooled water up to 200 bar into the reaction tank of 100 L (420 mm of diameter), which is connected by a 3 inch pipe to the dump tank of 2 m 3 . A dedicated test section was designed, assembled and implemented in the reaction tank. It is composed by 188 tubes, vertically disposed with triangular pitch inside a cylindrical support. This tube bundle is representative of a portion of the STSG of ELFR. The cylindrical support is closed at the lower and upper end by two tube plates and has a perforated lateral shell (300 mm of diameter and 400 mm high). The reaction tank is filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C up to the top tube plate, with an argon cover gas at about 2 bar. The water is injected at about 180 bar and 270°C through the central tube, at middle height of the bundle. The water-LBE interaction is characterised by high quality data acquisition system: 6 fast Pressure Transducers (PTs) working at 10 kHz for precisely characterize the first narrow injection peaks, 70 low constant time Thermocouples (TCs) to understand the vapour evolution path and 13 strain gages (SGGs) for measuring the strain of the bundle and main vessel. The overall LEADER experimental campaign is constituted by seven tests, divided in three series (B1, B2 and B3), characterized by different injection orifice diameters of 4, 8.9 and 12.6 mm, respectively. This paper presents the experimental results of the first two tests of series B2 (B2.1 and B2.2) having 8.9 mm of injection orifice. The first test analysed showed a first narrow pressure peak of about 32 bar, some milliseconds after the cap rupture instant. The following pressurization due to the evaporation of water entered into the reaction vessel was of an analogues magnitude for both the tests (about 50 bar) and lasted some tenths of second. The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 150°C. The outer rank was cooled down to about 300°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests and about 200 kg.

In the framework of Gen IV Heavy Liquid Metal Fast Reactors (HLMFRs) development, research activities are carried out for improving the Advanced Lead Fast Reactor European Demonstrator (ALFRED) project. In particular, the functioning and safety of the Steam Generator Bayonet Tube (SGBT) of ALFRED reactor (developed by ANSALDO Nucleare) needs to be experimentally investigated and numerical tools have to be verified and validated against reliable high quality data of SGBT stationary and accidental transient conditions. At the ENEA CR Brasimone, a full scale portion of the SGBT of ALFRED, composed by seven bayonet tubes arranged as triangular pitch inside a hexagonal shroud, was designed and assembled. It is called Heavy liquid mEtal-pRessurized water cOoled tube (HERO) and it was implemented as heat exchanger in the Integral Circulation Experiment (ICE) test section. The ICE-HERO test section is presently ready to be set in the large pool CIRCE facility for carrying out integral tests in Lead Bismuth Eutectic alloy (LBE) pool at about 400°C. ICE test section is equipped with a fuel pin simulator (FPS) of 37 electrically heated pin for an overall maximum thermal source of 1 MW and an argon injector above FPS for performing enhanced circulation (gas lift). The hot LBE (about 480°C) exiting from FPS moves upwards passing through the thermally insulated riser pipe and reaches an upper small tank, isolated from the outer colder LBE pool, called separator, from which descends shell side of the HERO heat exchanger and returns to the LBE pool, closing the integral circulation. The secondary side of HERO component (7 bayonet tubes) is fed by about 0.35 kg/s of water at 180 bar and 335°C. Each bayonet tube is composed by three concentric tubes. In the first (inner) tube, liquid water flows downwards in subcooled conditions and thermally insulated from outer parts. Between the first and second tube (annular cross-sectional area) water moves upwards and it is heated up to superheated steam condition. The annular gap between the second and third tube is filled by helium at 5 bar and steel powder for monitoring the possible rupture of the third or second tube and do not reduce excessively the heat transfer coefficient. The HERO component is highly instrumented for characterising the heat exchanger capabilities of the bayonet tubes in stationary and transient conditions. An overall number of 65 thermocouples are implemented, acquiring water, LBE and helium temperatures. Two pressure transducers monitor the feed water and steam line pressure. Each bayonet tube has four differential pressure transducers for measuring the pressure drops on the descending, ascending, and total water path and computing the water mass flow rate of each tube by a calibrated orifice. A Coriolis flowmeter is adopted to precisely determine the feed water mass flow rate. Moreover the signals of thermocouples, pressure transducers and bubble tubes implemented in the ICE test section will be acquired for an exhaustive analysis of the tests. A sensitivity analysis performed by RELAP5 code was carried out for defining the geometry and material to be adopted for the bayonet tube construction. The materials foreseen to be implemented have undergone modifications, AISI powder instead of SiC, air gap in place of the thermal insulating paint and AISI pipe instead of T91. The RELAP5 thermodynamic analysis predicted that the bayonet tube in updated configuration is able to provide superheated steam at about 400°C. The expected performance of HERO-TS is retained acceptable for the application to a prototypic unit and can be scaled to ALFRED.

Extensively released hydrogen due to zirconium-water reaction during severe accidents in containment of pressurized water reactor raises explosion crisis. Since the containment is the last barrier for fatal irradiation species, efficient measures should be implemented to restrain the hydrogen. Hence, hydrogen elimination and monitor devices are wildly applied to address this issue. Detection of hydrogen once has been conducted by a traditional hydrogen concentration measuring system with sampling devices and hydrogen sensors that located outside the containment. This arrangement, however, is a compromise between actual requirements for hydrogen measuring device and absence of favorable hydrogen sensors which could be applied in the harsh environment under severe accidents. Most recently, R&D of hydrogen concentration measuring system with in-situ hydrogen sensor has attracted increasing attention. Mitsubishi Heavy Industries, Ltd is focusing their job on an electrochemical hydrogen sensor based on solid state electrolyte. Besides, AREVA has developed a system depending on thermal conductivity detector associated with catalytic combustion sensor which requires external power supply to heat the assembly. In PERIC, we have developed a hydrogen concentration measuring system with in-situ hydrogen sensor which can be set in accident confident area. The hydrogen sensor is originally based on catalytic recombination of hydrogen and oxygen. Generally, catalyst prepared using noble metals such as platinum and palladium is scientifically loaded in the hydrogen sensor to serve as hydrogen sensitive material. On the event of severe accidents, mixture of hydrogen and air can spontaneously diffuse into the hydrogen sensor, where, part of the mixture is involved in a chemical exothermic reaction on the catalyst to generate water and heat. Generally, an increased concentration of hydrogen will raise relatively higher reaction temperature of the hydrogen sensor. The hydrogen concentration related temperature of the hydrogen sensor is detected using steel armored thermocouple. Besides, environmental temperature and pressure in the containment are also acquired to assist calculation. All the data are transferred to a signal processing cabinet, which, performs the calculation and indication functions using programmable logic controller and digital display device, respectively. There is no organic material, mechanical moving and power consumption part in the hydrogen sensor and thermocouple. The system indicated reliable performance in simulated containment under condition of high temperature, pressure, steam, and etc. The hydrogen concentration measuring system illustrated excellent endurance to poisoning species such as iodine and aerosol. Furthermore, the hydrogen sensor also suggested high resistance to irradiation. The system can survive a severe earthquake, and its seismic certification toward to safety shutdown earthquake is class I. Over 80 systems so far have be applied in pressurized water reactor in China and or Pakistan. The latest model is designed according to requirements of CAP1400.

A primary pipe rupture accident is one of the design-basis accidents of a Very-High-Temperature Reactor (VHTR). When a primary pipe rupture accident occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, it is important to understand the mixing processes of different kinds of gases in the stable and unstable stratified fluid layers. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Therefore, in order to predict or analyze the air ingress phenomena during a pipe rupture accident, it is important to develop a method for the prevention of air ingress during an accident. We carried out experiments to obtain the mixing process of two-component gases and flow characteristics of localized natural convection. This study also investigated a control method for the natural circulation of air through the injection of helium gas. An experiment has been carried out to investigate a control method of natural circulation of air by injection of helium gas. The experimental apparatus consists of a reverse U-sharped vertical slot and a storage tank. One side-slot consists of the heated and cooled walls. The other side-slot consists of the two cooled walls. The dimensions of the vertical slots are 598 mm in height, 208 mm in depth, and 70 mm in width. Each two vertical slots were connected and were a reverse U-shaped passage. The dimensions of the connecting passage were 16 mm in height, 106 mm in depth, and 210 mm in length. The storage tank was connected to the lower part of the reverse U-shaped passage. The dimensions of the storage tank were 398 mm in length, 398 mm in depth, and 548 mm in width. The reverse U-shaped passage and the storage tank were separated by a partition plate. The wall and gas temperature were measured by a K-type thermocouple. Experimental results regarding mixing process of two component gases in vertical fluid layer were as follows. The heavy gas was transported to the slot by the molecular diffusion and natural convection. As time elapses, natural circulation of heavy gas suddenly occurs through the reverse U-shaped slot. As a result of experiments, the onset time of natural circulation is affected by not only molecular diffusion coefficient but also the strength of natural convection. When the helium gas is injected into the channel, it is possible to control the natural circulation of air. The onset time of the reproduction of the natural circulation can be varied by changing the injection rate of the helium gas.

Passive safety systems represent one field of research concerning the safety-related enhancement of nuclear power plants. Passive safety systems can ensure the safe removal of decay heat without an input of electrical or mechanical energy for commissioning or operation. The heat removal chain is guaranteed on the basis of the physical principles condensation, heat conduction, boiling and natural circulation. The thermal hydraulic processes in passive safety systems disagree with the plant-specific thermal hydraulics because of different operating conditions. Since the established system codes are validated for the plant-specific conditions, the operational behavior of passive safety systems is currently not sufficiently predictable. On this account, the German Federal Ministry of Education and Research initiated the joint project PANAS to investigate the decay heat removal by passive safety systems on the basis of experimental analyses, modelling and validation. Object is the heat removal chain in advanced boiling water reactors consisting of emergency condensers (EC; heat transfer from reactor core to core flooding pools) and containment cooling condensers (CCC; heat transfer from the containment to the shielding/storage pool). At Technische Universität Dresden, the test facility GENEVA was constructed for the experimental investigation of the operational behavior of the CCC. GENEVA models the CCC concerning the original thermal hydraulic conditions of the heat source and heat sink as well as the tube geometry for the heat transfer. In this way, the comparability of the thermal hydraulic phenomena is given. Previous experiments focused on the stability analysis of the natural circulation in the test facility. The focus of PANAS is on the condensation process of saturated steam at the outside of the slightly inclined tubes and the convection respectively boiling of both a stable and an unstable two-phase flow inside these tubes. For a detailed analysis, condensation rates at the outside as well as the flow structure inside have to be investigated experimentally. Therefore, the instrumentation in the heat transfer section of GENEVA is considerably enhanced. This enhancement comprises an optical measuring system for the film thickness or droplet size of the condensate, a tipping scale for the condensate mass flow, void probes for the steam void fraction and more than 100 thermocouples outside and inside the tubes for temperature profiles in axial, radial and azimuthal direction. By reference to these parameters, it is possible to examine the thermal hydraulic models for the heat transfer. The paper outlines the available models in system codes regarding condensation and boiling concerning the operating conditions of the CCC. Since dropwise condensation could be observed in previous experiments and the condensation models in system codes focus on film condensation, the review is extended beyond native models. A sensitivity analysis of the reviewed models regarding condensation shows huge differences concerning the value of the heat transfer coefficient. Furthermore, the courses of the condensation models present different dependencies regarding the heat transfer coefficient and the wall temperature. Due to this, the necessity of the experimental investigation and later the revision of the condensation models in system codes is confirmed. The comparison of the reviewed models with first experimental results outlines the tendency for the numerical description of the condensation process. Based on the investigation and validation of models concerning the heat transfer processes in the CCC, the operational behavior will be accurately predictable by established system codes, which enhances the safety investigation and the licensing. Although the conception of this investigation is founded on the CCC, the adapted models will be able to characterize the heat transfer processes boiling and condensation for saturation conditions at a relatively low pressure (maximum 4 bar) and for natural convection in general.

The effective thermal diffusivity and conductivity of pebble bed in the High Temperature Gas-cooled Reactor (HTGR) are two vital parameters to determine the operating temperature and power in varisized reactors with the restriction of inherent safety. A high-temperature heat transfer test facility and its inverse method for processing experimental data are presented in this work. The effective thermal diffusivity of pebble bed will be measured at temperature up to 1600 °C in the under-construction facility with the full-scale in radius. The inverse method presents a global optimal relationship between thermal diffusivity and temperature through those thermocouple values in the pebble bed facility. Furthermore, the robustness and uncertainty analyses are also set forth here to illustrate the validity of the algorithm and the corresponding experiment.

The Westinghouse Owners Group Core Damage Assessment Guidance (CDAG), which has been authorized by the NRC staffs, is now used by licensee emergency response organization staff for estimating the extent of core damage that may have occurred during an accident at a Westinghouse nuclear power plant. On the other hand, EPR is a 3 rd generation nuclear power plant, which applies the advanced European nuclear power technology. This paper introduced Core Damage Assessment Guidance methodology in detail. The CDAG methodology is then attempted to apply to the EPR nuclear power plant. Detailed calculations have been performed for the setpoints of containment radiation monitors (CRM) and core exit thermocouples (CETs) with EPR design characteristics, which are the two main methods for estimation core damage amount. This paper also focuses the discussion on the reasons of difference of core damage estimating results between CRM method and CETs method; based on the discussion, several advices are provided when the two methods show a reasonable discrepancy in conclusions. Several conclusions can be made from the discussions in this article that 1)the Westinghouse Owners Group CDAG methodology proved to be reasonable when applied to EPR power plant for core damage assessment under severe accident; 2) the CDAG methodology which reflect the latest understanding of fission product behavior, is very simple and timely for core damage assessment based on NPP (nuclear power plant) real-time parameters; 3) conservative calculation results of setpoints on CRM and CETs based on EPR design show a reasonable trend and range; 4) it is concluded that several factors such as the releasing way, RCS fission product retention, fuel burnups might have great impact on the estimating results, when the results from two main indications (CRM and CETs) show an unexpected response.

In-Vessel Retention by External Reactor Vessel Cooling (IVR-ERVC) is method of removing the decay heat by cooling reactor vessel after corium relocation, and is also one of severe accident management strategies. Estimating heat transfer coefficients (HTCs) is important to evaluate heat transfer capability of the ERVC. In this study, the HTCs of outer wall of reactor vessel lower head were experimentally measured under the IVR-ERVC situation of Large Loss of Coolant Accident (LLOCA) condition. Experimental equipment was designed to simulate flow boiling condition of ERVC natural circulation, and based on APR+ design. This study focused on effects of real reactor vessel geometry (2.5 m of radius curvature) and material (SA508) for the HTCs. Curved rectangular water channel (test section) was design to simulate water channel which is between the reactor vessel lower head outer wall and thermal insulator. Radius curvature, length, width and gap size of the test section were respectively 2.5 m, 1 m, 0.07 m and 0.15 m. Two connection parts were connected at inlet and outlet of the test section to maintain fluid flow condition, and its cross section geometry was same with one of test section. To simulate vessel lower head outer wall, thin SA508 plate was used as main heater, and test section supported the main heater. Thickness, width, length and radius curvature of the main heater were 1.2 mm, 0.07 m, 1 m and 2.5 m respectively. The main heater was heated by DC rectifier, and applied heat flux was under CHF value. The test section was changed for each experiment. The HTCs of whole reactor vessel lower head (bottom: 0 ° and top: 90 °) were measured by inclining the test section, and experiments were conducted at four angular ranges; 0–22.5, 22.5–45, 45–67.5 and 67.5–90 °. DI water was used as working fluid in this experiment, and all experiments were conducted at 400 kg/m 2 s of constant mass flux with atmospheric pressure. The working fluid temperatures were measured at two point of water loop by K-type thermocouple. The main heater surface temperatures were measured by IR camera. The main heater was coated by carbon spray to make uniform surface emissivity, and the IR camera emissivity calibration was also conducted with the coated main heater. The HTCs were calculated by measured main heater surface temperature. In this research, the HTC results of 10, 30, 60 and 90 ° inclination angle were presented, and were plotted with wall super heat.

In AP1000 plant, the automatic depressurization system (ADS) works to discharge the high-temperature and high-pressure steam from the Reactor Coolant System (RCS) primary side to the In-containment Refueling Water Storage Tank (IRWST) in the LOCA conditions. In particular, for the AP1000 plant, both the IRWST and ADS spargers are specially designed, and the ADS spargers are located in one corner of the IRWST. All the factors lead to the special and complicated thermal and flow behavior in the IRWST, which in turn have great influences on the condensation effects of the ejected steam. In the present work, an overall scaled-down IRWST and ADS sparger models are built to study the condensation and mixing phenomena in the accidental depressurization events in AP1000. Thermocouples matrix with more than 200 T-type sheathed thermocouples are utilized to measure the three dimensional temperature in the large tank. The Particle Image Velocimetry (PIV) is employed for the measurement of the natural convection flow velocity. Based on the experimental data, the local spraying flow patterns, flow behavior, and thermal stratification phenomena in IRWST etc. are analyzed. The results indicate that the spraying steam condensation flow patterns are closely related to the subcooling degree in the IRWST. In addition, the stratification number is developed to evaluate the thermal stratification extent in the IRWST, which indicates that only part of the fluid are used efficiently for condensing the spraying steam directly.

In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and one of the most important and realistic accident for Lead Fast Reactors (LFR) fuel assembly. The blockage in a fast reactor Fuel Assembly (FA) may have serious effects on the safety of the plant leading to the FA damaging or melting. The temperature of the coolant leaving the FA is considered an important indicator of the health of the FA (i.e. the effective heat removal) and is usually monitored via a dedicated, safety-related system (e.g. thermocouple). The external or internal blockage of the FA may impair the correct cooling of the fuel pins, be the root cause of anomalous heating of the cladding and of the wrapper and potentially impact also fuel pins not directly located above or around the blocked area. In order to model the temperature and velocity field inside a wrapped FA under unblocked and blocked conditions, detailed experimental campaign as well as 3D thermal hydraulic analyses of the FA is required. The present paper is focused on the CFD pre-test analysis and design of the new experimental facility ‘Blocked’ Fuel Pin bundle Simulator (BFPS) that will be installed into the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The BFPS test section will be installed into the NACIE-UP loop facility aiming to carry out suitable experiments to fully investigate different flow blockage regimes in a 19 fuel pin bundle providing experimental data in support of the development of the ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) LFR DEMO. In particular, the ‘Blocked’ Fuel Pin bundle Simulator (BFPS) cooled by lead bismuth eutectic (LBE), was conceived with a thermal power of about 250 kW and a uniform wall heat flux up to 0.7 MW/m 2 , relevant values for a LFR. It consists of 19 electrical pins placed on a hexagonal lattice with a pitch to diameter ratio of 1.4 and a diameter of 10 mm. The geometrical domain of the fuel pin bundle simulator was designed to reproduce the geometrical features of ALFRED, e.g. the external wrapper in the active region and the spacer grids. Pre-tests calculations were carried out by applying accurate boundary conditions; the conjugate heat transfer in the clad is also considered. The numerical simulation test matrix covered the envisioned experimental range in terms of mass flow rate; the wall heat flux was imposed in order to have a fixed temperature difference across the BFPS in unblocked conditions. The blockages investigated are internal blockages of different extensions and in different locations (central subchannel blockage, corner sub-channel blockage, edge subchannel blockage, one sector blockage, two sector blockage). High resolution RANS simulations were carried out adopting the ANSYS CFX V15 commercial code with the laminar sublayer resolved by the mesh resolution. The loci of the peak temperatures and their width as predicted by the CFD simulations are used for determining the location of the pin bundle instrumentation. The CFD pre-test analysis allowed also investigating the temperature distribution in the clad to operate the test section safely.

An experimental campaign investigating the postulated Steam Generator Tube Rupture (SGTR) event, in relevant configurations for Heavy Liquid Metal Reactors (HLMRs), was carried out in the separate-effect facility LIFU5/Mod2, at ENEA CR Brasimone. Ten tests were performed injecting pressurized subcooled water into the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with a cover gas of argon at about 2 bar. Fast pressure transducers, thermocouples and strain gages provided high-quality measurement data for improving the phenomena understanding and supporting the development and validation phase of computer codes for SGTR numerical simulation. The experimental campaign is composed by two series of tests, characterized by different water pressure: 40 and 16 bar. The first two tests belonging to the low pressure experiments are presented, highlighting the pressurization time trends of the water injection tank, injection line and reaction vessel. The injected water mass flow rate and temperature trends in the reaction vessel were measured. The former test is the reference one and the latter was carried out for investigating the injection of water with higher sub-cooling. A post-test analysis of the two mentioned tests was carried out by SIMMER-III code. The pressure profile in the water injection tank was set as boundary condition of the calculation. The numerical analysis provided injection line and reaction tank pressurization in agreement with the experimental data. The lower water temperature test provided a better accordance with the measured data, due to the lower evaporation along the injection line. The SIMMER-III analysis also studied the water-LBE interaction from the volume fraction point of view and the energy released in the total reaction tank and in its cover gas.