An integral Pressurized Water Reactor (iPWR) type SMR was studied featuring Passive Safety Systems (PSSs). Different from active systems, PSSs are easily influenced by system parameters referred to as phenomenological factors such as heat loss, flow friction, oxidation, non-condensable gases and void fraction due to the low driving force of natural circulation. The system parameters also contribute to the uncertainty and dependency of PSS leading to the system unreliability. Thus, efforts are made to improve the reliability of PSS. A classical Probabilistic Risk Assessment (PRA) model describing active systems does not consider time evolution nor event ordering for PSS that dynamic PRA can accommodate. Here we developed and realized coupling between LabVIEW and CAFTA. Isolation Condenser System (ICS) is taken as the benchmark system due to the simple design in single phase without phase change phenomena in order to mainly remove decay heat and secondarily depressurize the reactor pressure vessel (RPV). A classical PRA model of ICS using CAFTA is coupled with real-time simulation of primary loop and ICS in LabVIEW, leading to a dynamic simulation result. The difference in failure probability using dynamic versus classical PRA revealed that for one there are more component demands with different event ordering, such that improved PSS reliability in the iPWR-type SMR designs is possible.

After the Fukushima Daiichi accident, predicting lower flammability limits (LFL) as a part of hydrogen risk analysis has become an ever important task. Although many experimental studies have been conducted extensively, the LFL results for mixtures abided by the severe accident conditions are still lacking. The objective of this study is to develop a calculated non-adiabatic flame temperature (CNFT) model, which facilitates to predict the LFL of hydrogen mixtures. This model considers heat loss due to radiative heat transfer from flame to ambient environment during flame propagation. The model shows better agreement with experimental results for various mixtures than previous model, which predicts the LFL through a calculated adiabatic flame temperature. Especially, prediction accuracy for H 2 -air-steam mixture and mixtures at elevated initial temperature is improved substantially. Thus it is worth to evaluate the applicability of the CNFT model in the hydrogen risk analysis during severe accident. The postulated hydrogen risk in the current Optimized Power Reactor 1000 MWe (OPR1000) under Station Blackout (SBO) scenario was investigated with MELCOR 1.8.6 code. As a result, it was observed that uncertainty of hydrogen risk calculated with the MELCOR default model can be reduced by the CNFT model. This study suggests that the developed CNFT model can enhance reliability of severe accident analysis related to the flammability of hydrogen mixtures.

The aim of the paper is to investigate the thermal conditions (temperature distribution, heat losses) in the support system of the Vapour Suppression Tank (VST) of the Vacuum Vessel Pressure Suppression System (VVPSS), a safety important system of ITER fusion reactor, protecting the Vacuum Vessel (VV) against overpressures. The VVPSS includes four VSTs of identical volume and mounted as two stacked assemblies. The study focuses on the optimization of the design of the thermal insulation of the bottom part of the VST, connecting each two-tank stack to the basement, and also on the identification of the thermal loads at the interface between the tank support and the tank pressure boundary. A Computational Fluid Dynamics (CFD) analysis of the VST has been performed for four different insulation configurations and considering both steady state and transient loads following accidental conditions. The results of the analysis are used to provide recommendation on the optimum configuration of the thermal insulation. Measures for minimization of the thermal gradient in the critical area of the joint between the tank hemispherical head and support skirt to limit the thermal fatigue on the welds are also suggested.

The Quench-06 experiment (ISP-45) has been used as a benchmark and training aid for Innovative Systems Software (ISS) and our users/students since it was completed in the early 2000s. The experiment was first analyzed by several international organizations using RELAP/SCDAPSIM/MOD3.2. These results were submitted to the “blind” and “open” phases of the ISP. The experiment was subsequently used for basic user training for experimental analysis by our RELAP/SCDAPSIM/MOD3.4 and MOD3.5 users. It is also used extensively in our university support and training internships. This paper describes an integrated uncertainty analysis of the QUENCH-06 electrically heated experiment, looking at the influence of uncertainties in experimental conditions and important models/correlations. The QUENCH calculations demonstrated the use of the new IUA, “Integrated Uncertainty Analysis”, option introduced into RELAP/SCDAPSIM/MOD3.4 in the summer of 2017 and MOD3.5 in the fall of 2017. The input models and results from both versions are discussed in the paper. The MOD3.4 results are based on the original input model developed for MOD3.2 and refined in the open phase of the ISP. The MOD3.5 results are based upon two base input models. The first was developed specifically to test the impact of MOD3.5 modeling improvements for the Quench electrically heated fuel rod simulator and was used in an early paper presented at this meeting in 2014. The second input model has been refined as part of the university support and training internship program and was used originally in 2016 to look at the influence of different approaches in modeling the insulated shroud used to minimize radial heat losses. The uncertainty analysis provided in this paper looks at the influence of uncertainties in (a) the parabolic equations for Zircaloy oxidation, (b) the tungsten heater element resistances, (c) the convective heat transfer coefficients, (d) the contact resistance of the heater elements, and (e) the thermal conductivity of the porous zirconia used in the shroud. The uncertainty analysis demonstrated very clearly that a bias was introduced into the 2014 MOD3.5 input model. This bias was subsequently determined to be primarily associated to the modeling of the shroud and associated radial heat losses. This bias was reduced in the 2016 version of the MOD3.5 input model and will be further refined as our training activities on the modeling of experiments continue.

A FLiNaK high temperature test loop, which was designed to support the Thorium Molten Salt Reactor (TMSR) program, was constructed in 2012 and is the largest engineering-scale fluoride loop in the world. The loop is built of Hastelloy C276 and is capable of operating at the flow rate up to 25m 3 /h and at the temperature up to 650°C. It consists of an overhung impeller sump-type centrifugal pump, an electric heater, a heat exchanger, a freeze valve and a mechanical one, a storage tank, etc. Salt purification was conducted in batch mode before it was transferred to and then stored in the storage tank. The facility was upgraded in three ways last year, with aims of testing a 30kW electric heater and supporting the heat transfer experiment in heat exchanger. Firstly, an original 100kW electric heater was replaced with a 335kW one to compensate the overlarge heat loss in the radiator. A pressure transmitter was subsequently installed in the inlet pipe of this updated heater. Finally, a new 30kW electric heater was installed between the pump and radiator, the purpose of which was to verify the core’s convective heat transfer behavior of a simulator design of TMSR. Immediately after these above works, shakedown test of the loop was carried out step by step. At first the storage tank was gradually preheated to 500°C so as to melt the frozen salt. Afterwards, in order to make the operation of transferring salt from storage tank to loop achievable, the loop system was also preheated to a relatively higher temperature 530°C. Since the nickel-base alloy can be severely corroded by the FLiNaK salt once the moisture and oxygen concentration is high, vacuum pumping and argon purging of the entire system were alternatively performed throughout the preheating process, with the effect of controlling them to be lower than 100ppm. Once the salt was transferred into the loop, the pump was immediately put into service. At the very beginning of operation process, it was found that flow rate in the main piping could not be precisely measured by the ultrasonic flow meter. Ten days later, the pump’s dry running gas seal was out of order. As a result, the loop had to be closed down to resolve these issues.

The Reactor Cavity Cooling System (RCCS) is an important passive safety system that is being incorporated in a number of high temperature reactor design concepts. The Natural convection Shutdown heat removal Test Facility (NSTF), located at Argonne National Laboratory, is an experiment with the objective of investigating the flow and thermal behavior of a particular air-cooled RCCS design. It consists of 12 ducts surrounded by a cavity with a heated wall, through which air flows via natural convection before exiting through two chimneys. The NSTF is a ½-scale facility, and is well instrumented in order to provide data for code validation, including Computational Fluid Dynamics (CFD)-grade data in a number of locations. Instrumentation includes fiber-optic Distributed Temperature Sensors (DTS) throughout one of the riser ducts and in the upper plenum. In conjunction with the experimental tests, CFD simulations were performed to support the design and optimization of these natural convection systems. The CFD simulations were performed using the “as-tested” geometry of the NSTF. All CFD simulations were steady-state. Both a full natural convection model and a smaller forced primary flow model were tested. The influence of boundary conditions, notably at the cavity walls, was tested. Initial simulations assumed adiabatic walls but these were later adapted to simulate heat losses, aided by thermal images taken of the exterior NSTF surfaces during testing. Simulations were run for tests at two different power levels. A number of turbulence models were compared to test their influence. Simulation results were compared with experimental data. Convergence was generally good for both models. It was found that the natural convection model was indeed beneficial for correctly estimating local temperatures in a number of areas, particularly near the top of the riser ducts and from DTS measurements along the flow path. Flow in the heated cavity was complex. In general, the experimental trends were predicted well by CFD, although magnitudes could be improved in some areas. The turbulence models tested had a relatively small effect on the shape of the temperature profile in the ducts and on heated surface temperatures. Results from the simulations have been of direct use in improving test procedures and choosing locations for more accurate instrumentation. In future work, full natural convection simulations of more tests will be performed. After this has been completed, best practices can be established for accurately simulating these general types of natural convection systems across a wide range of operating conditions.

Metallic insulation is commonly used in reactor vessel because of its resistance to radiation and corrosion. Since the main mode of heat loss of reactor vessel is thermal radiation, the ability to prevent radiation heat transfer is important for metallic insulation. But the thermal conductivity of metallic insulation is difficult to calculate owing to their complex geometry. This article uses FLUENT 14.0 to obtain the important parameter “view factor”, and then develops a computational model of effective conductivity of metallic insulation. Heat transfer test of metallic insulation was done, and the numerical simulation of metallic insulation was also performed. Based on results of test and simulation, the computational model is modified. The modified model can fit the test result better. Based on the modified model, the effective conductivity of metallic insulation increases with the increase of temperature of hot side and cold side, among which the temperature of hot side influences more. And when the temperature is high, the effective conductivity increases much faster.

Uncertainty evaluation constitutes a key feature of BEPU (Best Estimate Plus Uncertainty) process. The uncertainty can be the result of a Monte Carlo type analysis involving input uncertainty parameters or the outcome of a process involving the use of experimental data and connected code calculations. Those uncertainty methods are discussed in several papers and guidelines (IAEA-SRS-52, OECD/NEA BEMUSE reports). The present paper aims at discussing the role and the depth of the analysis required for merging from one side suitable experimental data and on the other side qualified code calculation results. This aspect is mostly connected with the second approach for uncertainty mentioned above, but it can be used also in the framework of the first approach. Namely, the paper discusses the features and structure of the database that includes the following kinds of documents: 1. The” RDS-facility” (Reference Data Set for the selected facility): this includes the description of the facility, the geometrical characterization of any component of the facility, the instrumentations, the data acquisition system, the evaluation of pressure losses, the physical properties of the material and the characterization of pumps, valves and heat losses; 2. The “RDS-test” (Reference Data Set for the selected test of the facility): this includes the description of the main phenomena investigated during the test, the configuration of the facility for the selected test (possible new evaluation of pressure and heat losses if needed) and the specific boundary and initial conditions; 3. The “QP” (Qualification Report) of the code calculation results: this includes the description of the nodalization developed following a set of homogeneous techniques, the achievement of the steady state conditions and the qualitative and quantitative analysis of the transient with the characterization of the Relevant Thermal-Hydraulics Aspects (RTA); 4. The EH (Engineering Handbook) of the input nodalization: this includes the rationale adopted for each part of the nodalization, the user choices, and the systematic derivation and justification of any value present in the code input respect to the values as indicated in the RDS-facility and in the RDS-test.

In order to conform to the new regulatory standard in Japan, seawater is regarded as the alternative water source both for BWRs (Boiling water reactors) and PWRs (Pressurized water reactors). For preventing further accident evolutions occurred in Fukushima Daiichi nuclear power plants, seawater was injected into the reactors for more than one week. With long-term seawater injection, sea salt compositions are condensed and many of them will precipitate when saturated concentrations are exceeded. Unlike corrosion issues, impacts of sea salt precipitation on the heat removal has not been studied widely in the past because it has not been regarded as the alternative water source before Fukushima Daiichi nuclear power plants accident. The existent knowledge base of boric acid precipitation under LOCA conditions was studied. Based on the existent study on impacts of boric acid precipitation under LOCA conditions, the experimental project of seawater consisting of four experiments was proposed. The existent database of physical properties, such as viscosity, of seawater is rather poor under severe accident conditions. In addition, it is likely that boric acid will be injected with water or seawater to prevent re-criticality. It is known that physical properties of boric acid vary widely under high temperature conditions. Measurement of viscosity of the seawater-boric acid mixture was conducted in high temperature using a rotational viscometer. Under conditions equivalent to the estimated bulk coolant conditions under a long-term cooling phase of severe accidents. In this range, no obvious change of viscosity is expected. Then a detailed structure of seawater precipitates was observed using a mockup fuel bundle with 5×5 in the square lattice and 500 mm in the length. Images of precipitates were taken using the X-ray CT. The water level, the concentration of sea salt and the heat flux are employed as experimental parameters. The heat flux, bubble stirring in downstream of spacers and heat loss by a non-heated channel box were identified as influential factors to local and overall precipitates in fuel bundles.

An important accident management measure for controlling severe accident transients in Light Water Reactors is the injection of water to cool the degrading core. Flooding of the overheated core, which causes quenching of the fuel rods, is considered a worst-case scenario regarding hydrogen generation rates which should not exceed safety-relevant critical values. Within the frame of the QUENCH test-program the loss of coolant accidents with the following flooding of overheated core in Light Water Reactors is analysed using an experimental facility. The modelling of QUENCH-03 and QUENCH-06 experiments was performed with RELAP/SCDAPSIM computer code. The observed calculation results showed that thermal properties of shroud materials (heat losses through the shroud) and electrical power of fuel simulators are the main source of uncertainty in the calculations. The main idea of this article is modification of input parameters to receive the best agreement with the measurements for the selected QUENCH test. Modified input parameters are used in the input deck for another QUENCH test. The good agreement between calculation results and measurements of both QUENCH tests demonstrated the correctness of modified parameters and legitimacy with the real physical processes.

Accident scenarios in spent fuel pools (SFP) were not a main topic in nuclear safety studies so far. Heat fluxes are low and the timeframe for counteractions in case of a loss of cooling is long. However, because of their huge activity inventory safety of spent fuel pools should be analyzed with great intensity. Beside some analytical approximations from the U.S. Nuclear Regulatory Commission (NRC) there is only few experimental data about loss of coolant accidents in a SFP [1, 3, 4, 5, 6]. After the reactor accident in Fukushima Daiichi NPP this issue became omnipresent in public perception. The presented paper investigates boil-off scenario of a boiling water reactor (BWR) SFP after a station blackout with focus on experimental findings. Since 2006 experiments for SFP boil-off scenarios were conducted at the Technische Universität Dresden [7]. For a detailed investigation under realistic geometrical fuel assembly conditions the test facility ADELA-II was built in 2011. It consists of two channels. The inner channel simulates a quarter of a BWR fuel element with a 24 rod bundle. In the outer channel 8 more heating rods were assembled to reduce heat losses from the inner channel and to simulate the surrounding. With a heated length of 3760 mm and a width of 120 mm it was designed in an axial and radial scale of 1:1. The heating profile is based on the heat flux profile of a fuel rod with low burn-up. To measure the detailed axial and radial temperature profiles 113 thermocouples are mounted. With these data conclusions for heat and mass transfer can be made with special regard to convection inside the assembly. Boil-off experiments at the ADELA-II facility with a supplied heat of 20 Watts per rod lead to rod surface peak temperatures of about 479 °C. Due to the high temperature and narrowed geometry the axial heat transport is harmed and radial heat conduction had a great influence on the test results. Accordingly, quasi adiabatic boil-off could not be verified. Thus, coolability of a fuel assembly strongly depends on the (radial) heat removal from the fuel assembly box to the surrounding. Further investigations have to be done for a better consideration of neighbored fuel elements and global thermohydraulic effects in the spent fuel pool.

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive heat removal system proposed for the Advanced High-Temperature Reactor (AHTR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops relying completely on buoyancy as the driving force. A fluidic diode has been proposed in the DRACS primary loop to maintain the passive feature. Fluidic diodes are passive flow control devices with low flow resistance in one direction and high flow resistance in the opposite direction. The fluidic diode is orientated such that during reactor normal operation the primary salt flow in the DRACS is restricted, thus preventing excessive heat loss from the reactor to the DRACS. However, when the DRACS is functioning during reactor accidents, the primary salt flow is in the forward flow direction of the diode that features low flow resistance. To investigate the reliability and thermal performance of the DRACS, a high-temperature DRACS test facility (HTDF) is being designed and constructed at The Ohio State University (OSU). In this HTDF, a conventional vortex diode has been proposed. In this paper, a detailed design process of the vortex diode for the HTDF is presented. Design parameters, such as the desired flow rates in and pressure drops across the fluidic diode, were first determined for both the forward and reverse flow directions, following which was the parametric CFD study of multiple vortex diodes with variant nozzle size, chamber size, and inlet flow rates. Flow structures inside the diode, and the effects of the nozzle size, chamber size, and Reynolds number on the Euler number were examined for both flow directions. Correlations of the forward and reverse Euler numbers and the diodicity were developed and used to develop a vortex diode design that would be applicable to the HTDF.

High-temperature gas-cooled reactor-pebble bed module (HTR-PM) of China, based on the technology and experience of the 10MW high-temperature gas-cooled reactor (HTR-10), is currently in the design and experiment phase. The helium blower is the key equipment in the first loop of the HTR-PM. Active magnetic bearings (AMBs) are replacing ordinary bearings as the perfect sustaining assembly for the helium blower because they have several advantages: they are free of contact, do not require lubrication, are not subject to the contamination of wear, have endurance, and control performance very well. So the AMB is the appropriate supporting assembly. The rotor’s length of the helium blower is about 3.3 m, its weight is about 4000 kg and the rotating speed is 4000 r/min. The maximal radial load of the upper AMB or the lower AMB is about 1950kg, and the maximal axial load is about 4500kg. So the axial and radial AMBs must have enough carrying capacity to support the rotor. It is difficult to design the AMB in limited space. Many factors may restrict the AMB design, for example, clearance between AMB and rotor, electric current, and environment temperature, etc. In this paper, design of magnetic field will be analyzed. Heat loss of AMB will be discussed. The designing scheme of magnetic field will offer the important theoretical base for AMB design of the helium blower of HTR-PM.

It is necessary to access the pool top area for loading and unloading irradiation test pieces by a required irradiation period during a normal operation of an open-pool-type research reactor installed on the bottom of reactor pool with a depth of about 13m. However, when the reactor pool top radiation level exceeds the limit of radiation level by the rising of reactor coolant contaminated by radioactivity due to a natural convection of the pool water, access to the reactor pool top area is denied. In the case of HANARO, a hot-water layer (HWL, hereinafter) is maintained below a depth of 1.2m from the top of the reactor pool in order to reduce the radiation level of the reactor pool top area by suppressing natural convection of the pool water. After normal operation of the HWL, the pool top radiation level is safely maintained below the limit of the pool top radiation level. For studying the characteristics of the HWL under downward flow pattern of the reactor coolant (hereafter as DFP), the heat loss of the HWL is calculated based on the model for HANARO HWL. The heat loss characteristics of the HWL were reviewed by reactor power level increment. This paper suggests a HWL heat loss relation formula by an increment of the reactor power level. It was confirmed that the total heat loss of the HWL under DFP of reactor coolant linearly increased larger than that of upward flow pattern (hereafter, UFP). The reason is that the bottom convection loss of DFP increased 4.5 times than that of UFP by the 10 times increment of the core bypass flow rate under DFP than that under the UFP.

The ANL Plant Dynamics Code (PDC) for the analysis of supercritical carbon dioxide (S-CO 2 ) Brayton cycle power converters has been under development at Argonne National Laboratory for several years. In previous years, limited validation of the PDC models on an individual basis was carried out using experimental data obtained from facilities directed at individual components. Recently, experimental data from the SNL/BNI small-scale S-CO 2 Brayton cycle demonstration that is being assembled in a staged fashion has been provided to ANL. The loop configuration with a single turbo-alternator-compressor (TAC) was modeled with the Plant Dynamics Code and the performance prediction of individual components and the entire loop obtained from the PDC was compared with the experimental data. Overall, reasonably good agreement is obtained for steady state conditions around the loop when a value is inferred for the heat loss downstream of the turbine such that a steady state loop energy balance can be established.

Uncertainty evaluation constitutes a key feature of BEPU (Best Estimate Plus Uncertainty) process. The uncertainty can be the result of a Monte Carlo type analysis involving input uncertainty parameters or the outcome of a process involving the use of experimental data and connected code calculations. Those uncertainty methods are discussed in several papers and guidelines (IAEA-SRS-52, OECD/NEA BEMUSE reports). The present paper aims at discussing the role and the depth of the analysis required for merging from one side suitable experimental data and on the other side qualified code calculation results. This aspect is mostly connected with the second approach for uncertainty mentioned above, but it can be used also in the framework of the first approach. Namely, the paper discusses the features and structure of the database that includes the following kinds of documents: 1. The “RDS-facility” (Reference Data Set for the selected facility): this includes the description of the facility, the geometrical characterization of any component of the facility, the instrumentations, the data acquisition system, the evaluation of pressure losses, the physical properties of the material and the characterization of pumps, valves and heat losses; 2. The “RDS-test” (Reference Data Set for the selected test of the facility): this includes the description of the main phenomena investigated during the test, the configuration of the facility for the selected test (possible new evaluation of pressure and heat losses if needed) and the specific boundary and initial conditions; 3. The “QP” (Qualification Report) of the code calculation results: this includes the description of the nodalization developed following a set of homogeneous techniques, the achievement of the steady state conditions and the qualitative and quantitative analysis of the transient with the characterization of the Relevant Thermal-Hydraulics Aspects (RTA); 4. The EH (Engineering Handbook) of the input nodalization: this includes the rationale adopted for each part of the nodalization, the user choices, and the systematic derivation and justification of any value present in the code input respect to the values as indicated in the RDS-facility and in the RDS-test.

The objective of this paper is to calculate heat losses from a CANDU-6 fuel-channel while modifying it according to the specified operating pressure and temperature conditions of SuperCritical Water-cooled Reactors (SCWRs). Heat losses from the coolant to the moderator are significant in a SCWR because of high operating temperatures (i.e., 350–625°C). This has adverse effects on the overall thermal efficiency of the Nuclear Power Plant (NPP), so it is necessary to determine the amount of heat losses from fuel-channels proposed for SCWRs. Inconel-718 was chosen as a pressure tube (PT) material and PT minimum required thickness was calculated in accordance with the coolant’s maximum operating pressure and temperature. The heat losses from the fuel-channel were calculated along the heated length of the fuel-channel. Steady-state one-dimensional heat-transfer analysis was conducted, and programming in MATLAB was performed. The fuel-channel was divided into small segments and for each segment thermal resistances of the fuel-channel components were analyzed. Further, the thermophysical properties of the coolant, annulus gas, and moderator were retrieved from the NIST REFPROP software. The analysis outcome resulted in a total heat loss of 29.3 kW per fuel-channel when the pressure of the annulus gas was 0.3 MPa.

This paper focuses on thermal-design options of a new pressure channel for SuperCritical Water-cooled Reactors (SCWRs). The objectives of this paper are to estimate heat losses from the coolant to the moderator for a preliminary fuel-channel design and to investigate effects of the insulator thickness and moderator pressure on the overall heat losses. In order to fulfill the objectives, the heat losses for an existing reactor were calculated and compared with available values from open literature. These calculations became the basis for calculation of the heat loss for the chosen new fuel-channel design. MATLAB, and NIST REFPROP software were utilized for programming and calculation of thermo-physical properties as needed, respectively. Heat losses for different thicknesses of the ceramic insulator were calculated. These calculations showed that the heat losses for the optimum thickness of insulator, which was calculated to be 7 mm, were about 31 MW. In current CANDU reactors the operating pressure of the moderator is close to the atmospheric pressure; higher operating pressures will allow operation of the moderator at higher temperature while preventing occurrence of boiling in the calandria vessel. Higher moderator temperatures will results in a lower temperature difference between the coolant and the moderator, hence reducing the heat sink from the coolant to the moderator. The effect of the moderator pressure on the heat loss was investigated, which showed that the heat loss can be reduced by increasing the operating pressure of the moderator by approximately 1 MW per 0.1 MPa increase in pressure.

We have studied nuclear power plant for district heating. Already some towns and villages in Hokkaido have requested small reactor for district heating. Using existing technology allows us to shorten development period and to keep a lid on development cost. We decided to develop new reactor based on “MUTSU” reactor technology because “MUTSU” had already proved its safety. And this reactor was boron free reactor. It allows plant system to reduce the chemical control system. And moderator temperature coefficient is deeply negative. It means to improve its operability and leads to dependability enhancement. We calculated burn-up calculation of erbium addition fuel. In the result, the core life became about 10 years. And we adapt the cassette type refueling during outage in order to maintain nonproliferation. In the district heating system, a double heat exchanger system enables to response to load change in season. To obtain the acceptance of public, this system has a leak prevention system of radioactive materials to public. And road heating system of low grade heat utilization from turbine condenser leads to improve the heat utilization efficiency. We carried out performance evaluation test of district heating pipeline. Then the heat loss of pipeline is estimated at about 0.440 °C/km. This result meets general condition, which is about 1 °C/km. This small plant has passive safety system. It is natural cooling of containment vessel. In case of loss of coolant accident, decay heat can remove by natural convection air cooling after 6 hours. Decay heat within 6 hours can remove by evaporative heat transfer of pool on containment vessel.

Reflooding of an uncovered and overheated core is one of the most expected measures during the early phase of a severe accident progression. The capability of analytical tools to model correctly processes during independent sub-phases of this accident progression is on different levels, and the most concentrated effort is focused on the study of fuel cladding quenching topic. The main objective of the Forschungszentrum Karlsruhe (Germany) Quench experimental program is to investigate the hydrogen generation during the reflooding of overheated core. The Nuclear Research Institute Rez (hereafter NRI) contributes to this program by the analytical simulation of the Quench bundle experiments with the MELCOR code mainly and ICARE2 code as well since 2000. The NRI analyses covered the Quench-01, Quench-03 and Quench-06 with version MELCOR 1.8.5 (including reflood model), and Quench-01 and Quench-11 tests with the latest version MELCOR 1.8.6. The tests Quench-01 and Quench-06 were characteristic of the lower reflooding onset temperature (1900–2050 K), and the tests Quench-03 and Quench-11 had high reflooding onset temperature (> 2350 K) with the fast heat-up phase before reflooding. The integral code MELCOR is capable to sufficiently predict the heat up and reflooding phases for the tests with the lower onset temperature, but the Quench-03 test with high onset temperature was not predicted correctly starting in the heat-up phase and also the hydrogen generation during bundle reflooding was significantly underpredicted. The interpretation of the Quench-03 test with ICARE2 code confirmed the MELCOR results, but its more detail description enabled to identify a cause. The underestimation of the bundle temperatures is influenced by the shroud insulation behaviour, which could be interpreted as the reduction of heat losses through shroud and simultaneously more intensive heat-up of a bundle. The phenomenon, which causes this behaviour, is not yet known, and is not important for plant applications, although it is relevant for the test interpretation. Also the interpretation of the Quench-11 showed an importance of correct modeling of the bundle shroud, which is in the MELCOR 1.8.6 simulated by newly implemented independent component. The Quench-11 test analysis showed simplified failure criteria for the cases when the shroud is not supported by the former components, which are assumed in the basic MELCOR model fundamentals. Concerning the plant application of the MELCOR code for the scenario with the overheated core reflooding, the user has to give attention on the correct treatment of heat losses through shroud and its degradation modeling, mainly if the shroud supporting formers absent in the core periphery design. Hydrogen generation during reflooding is predicted well for lower onset temperatures (below 2100 K) and slightly underpredicted for higher onset temperatures. It is important to take these code capabilities into account because a reflooding of the reactor core will consist of varied reflooding onset temperatures.