Reactor Coolant System Research Articles

Operator actions evaluation for MSGTR accident mitigation

The safety assessment of nuclear power plants for multiple failure accidents such as Multiple Steam Generator Tube Rupture (MSGTR) and the verification of the suitability of accident management measures have gained prominence in the nuclear engineering field after the Fukushima accident. Compared to Steam Generator Tube Rupture (SGTR), the MSGTR accident progresses rapidly due to the high break flow rate, and thus appropriate operator actions are essential to prevent the release of radioactive materials into the environment. It is crucial for the operators to identify the MSGTR early and perform each operator action within the allotted time in order to mitigate the accident successfully. In this study, a reference analysis model for MSGTR is developed, differences from SGTR are analyzed, and the required time for major operator actions are derived through various sensitivity analyses. In conclusion, the operator can confirm that there is a high possibility of MSGTR occurring through the changes in the subcooling degree at the hot-leg and the Pressurizer (PZR) water level. The appropriate time to be initiated for major operator actions are as follows: (1) Reactor Coolant Pump (RCP) manual trip within 10 min after the reactor trip, (2) Reactor Coolant System (RCS) temporary-cooldown operation through Main Steam Isolation Bypass Valve (MSIBV) and Steam Bypass Control System (SBCS) within 15 min after the reactor trip, (3) PZR spray operation – RCS temporary-cooldown operation in progress, and (4) RCS controlled-cooldown operation using Atmospheric Dump Valve (ADV) within 15 min after the end of the RCS temporary-cooldown operation. Additional possible action is the restart of RCP, but the execution of this action depends on the operator’s judgment. The results of this investigation have significant implications which can be utilized for improving the accident mitigation strategies and the operator training for MSGTR.

Compensated Feed water Pump Control Analysis of Effect of Base-Load Electricity Demand Reduction on Nuclear Steam Supply System

The effect of compensated feedwater (FW) pump control on a nuclear steam supply system with a significant reduction of baseload electricity demand as a common-cause failure could result in temperature elevation of the reactor coolant system and corresponding pressure increases in the pressurizer and steam generators above the set points. The shutting and opening of the pressure relief valve causes the fluid flow rate to transition from laminar to turbulence flow, where a sudden burst, chaotic movement, and inertial forces and weight of the fluid have the potential to cause a break in pipelines leading to a loss-of-coolant accident. This study employs the Fourier transform to simulate the impact of force as the power spectral density (in dBm/Hz) measured in 1 to 99 label harmonics over a specified time window using MATLAB/Simulink library tools. The experimental results show that compensated FW pump control could significantly reduce the effect of turbulence and reveal a perturbation settlement state prior to steady-state laminar flow.

Comprehensive oxidative process for the removal of antimony from the surfaces of nuclear reactor coolant systems

An oxidative process using hydrogen peroxide for the dissolution of metallic antimony, stripping of the dissolved antimony ions, and the compatibility of the oxidative process with the structural materials present in the coolant systems of nuclear reactors have been investigated. This process has direct application for the removal of antimony activity and thereby reducing the radiation exposure hazard to the operating personnel. The rate of dissolution of metallic antimony increased in the presence of hydrogen peroxide. The re-deposition of dissolved antimony on the oxide coated carbon steel coupons was ≤0.02 mg/cm2. Three substrates investigated, namely, strong base anion exchange resin and two chitosan-TiO2 based substrates, for the removal of dissolved antimony were equally efficient. The process could dissolve antimony without any significant re-deposition on the oxide coated carbon steel surface and could dissolve antimony incorporated in the oxide film of zircaloy. Corrosion compatibility of relevant alloys, namely, carbon steel, incoloy-800, Zr2·5Nb and monel-400, at pH 10.2 and 9.0, showed that only monel-400 underwent pitting type of attack at pH 10.2.

CINEMA-SMART development and its application to SBLOCA analysis for Korean small integral reactor

The Korean small integral reactor of SMART (System-integrated Modular Advanced ReacTor) was developed, which contained core, reactor coolant pumps, steam generators and pressurizer within a single reactor pressure vessel. In the code development of severe accident sequence analysis for the SMART, a CINEMA (Code for INtegrated severe accidEnt Management Analysis)-SMART has been developing from an initiation event to a containment failure including fission product behavior in the vessel and the containment. The CINEMA-SMART is composed of CSPACE, SACAP (Severe Accident Containment Analysis Package), and SIRIUS (SImulation of Radioactive nuclide Interaction Under Severe accident), which are capable of core melt progression with thermal hydraulic analysis of the RCS (Reactor Coolant System), severe accident analysis of the containment, and fission product analysis, respectively. The main sequence of a SBLOCA (Small Break Loss Of Coolant Accident) for the SMART has been analyzed using CINEMA-SMART computer code. This analysis is focused on an evaluation of core melt progression and containment thermal hydraulic behavior during a severe accident. The CINEMA results have shown that the integrity of the containment can be maintained during the SBLOCA accident in the SMART.

A cost-informed component maintenance index and its applications

All systems and components are unreliable in the sense that they will fail. While a failed component in a system is being repaired (i.e., corrective maintenance), preventive maintenance (PM) may be conducted on the other components to improve the reliability of the system. The selection of different components for PM may result in a variety of maintenance policies with different cost implications. It is therefore necessary to develop appropriate tools such as importance measures to guide engineers in selecting components for PM in order to minimize relevant costs. There is little research, nevertheless, that jointly minimizes the total expected cost of maintenance and maximizes the number of components for PM. To fill in this gap, this paper proposes an importance index, Cost-Informed Component Maintenance Index (CICMI). It then derives some propositions of the proposed index and different maintenance policies, respectively. A method to optimize the number of components for PM subject to cost constraints is then proposed. A case study on a reactor coolant system is performed to illustrate the applicability of the proposed methods.

Pressurizer system dynamic model for transient control in PWR

Abstract The pressurizer system of pressurized water reactor (PWR) maintains the reactor coolant system pressure during steady-state operation and limits pressure changes during transients. The in/out surge transients will cause pressure variations and they are controlled by either the spray system or the heater system. The spray system is actuated when the pressure exceeds a preset value. The heater system is initiated when the pressure falls below a preset value. The fundamental understanding and a reliable modeling of the pressurizer system behavior under steady state and transient conditions are needed to simulate overall nuclear power plant behavior. In the present study, an algorithm using Python 3.7 is developed to represent the dynamic behavior of the pressurizer system under steady-state and during in/out surge transients. Moreover, RELAP5 code is used to simulate the pressurizer system during the prescribed transients. The analysis and assessment results demonstrate satisfactory control performance during the in/out surge transients that guarantee the safety of PWR during operation. Also, the comparison between Python algorithm and RELAP5 model illustrates the capability and effectiveness of the Python algorithm for dynamic simulation and control.

Analysis of depressurization ability for IRIS containment during SBLOCAs using RELAP5 and CONTEMPT

As an integral and modular pressurized water reactor (PWR), the international reactor innovative and secure (IRIS) eliminates the threat of large break LOCAs (LBLOCAs) by integrating all major reactor coolant system components into the reactor vessel and provides compact steel containment with a suppression pool. Unlike large reactor containment, IRIS containment has modest volume and compact internal structure. The analysis of traditional dry containment cannot be applied to the IRIS reactor. Despite the disappearance of LBLOCAs (exceeding 4 in), small steel containment still faces the threat of high–frequency small break LOCAs (SBLOCAs) (e.g., the largest possible 4 in chemical and volume control system (CVCS) line break and the lowest elevation 2 in direct vessel injection (DVI) line break). To identify the impact of these characteristics on the integrity of IRIS reactor containment, a safety assessment framework that combines deterministic and probabilistic methods is proposed for the pressure response of containment under SBLOCAs of IRIS reactor. First in the modeling phase, primary and secondary loop models have been built considering the design of IRIS, and accident consequences under typical SBLOCAs were evaluated using RELAP5/MODE4. A simplified model of containment was presented herein based on the primary structure of IRIS containment. Second in the input parameters identification and uncertainty quantification stage, the responses of containment were simulated using CONTEMPT–LT. Thus, the main factors affecting the depressurization capacity of the containment were identified, and the reasonable probability density distributions were determined. Third, in the uncertainty propagation stage, Monte Carlo method was used to map the uncertainty of input parameters to the output variable. Fourth in the sensitivity analysis and uncertainty analysis phase, the contribution of each input parameter was measured using sensitivity analysis, and their interactions were investigated. The range of deviations of the containment pressure response was measured using uncertainty analysis, and finally, the calculated results were combined with the containment fragility curve to evaluate containment failure probability.

RELAP5 validation for core makeup tank analysis based on separate effect tests

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during accident and has been widely used in the field of nuclear thermal-hydraulics and safety analysis. The core make-up tank (CMT) is a crucial device of the passive core cooling system in the AP series pressurized water reactor nuclear power plants. Since the CMT not only provides full-pressure injection but also triggers auto depressurization system by its level signal during small-break loss of coolant accident. Therefore, it is necessary to carry out CMT separate effect tests to verify the feasibility and accuracy of RELAP5/MOD 3.4 to simulate CMT. The previous studies found that the lower version of the RELAP5 program has problems in simulating CMT, and the applicability of the latest RELAP5/MOD3.4 has not been evaluated. In this paper, a CMT separate effect test facility were presented and then RELAP5/MOD 3.4 was used to establish a CMT test bench model to evaluate the applicability, and obtain the general rules and methods of RELAP5/MOD 3.4 to simulate CMT. It’s found that RELAP5/MOD 3.4 can accurately simulate most of the CMT operating conditions. However, large uncertainty is expected in the simulation of delayed time and the flowrate of delayed drainage stage is not affected by the heat transfer coefficient of steam direct contact condensation. The simulation of delay time depends on experimental data and the CMT delayed drainage experiment cannot verify the accuracy of the RELAP5/MOD 3.4 direct contact condensation model.

Numerical Study of Natural Circulation Flow in Reactor Coolant System during a Severe Accident

The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.

Fault Diagnosis Using Bond Graphs in an Expert System

A fault diagnosis method using bond graphs in an expert system is proposed for a reactor coolant system. Firstly, the time causality graph and the variable relationship graph are derived from the bond graph. Secondly, the fault signature matrix is obtained by combining the change relationship of fault parameters. Finally, the fault signature matrix is used as the rule of the inference engine design in the expert system for fault diagnosis. In this paper, the key equipment of the reactor coolant system is used to verify the fault diagnosis method of the bond graph expert system, and the path reasoning relationship between alarms is obtained, which can accurately obtain the deep knowledge required by the operators. A new idea for fault diagnosis in a nuclear power plant’s expert system is provided by this method.

Minimum dose path planning during reactor coolant system maintenance with corrosion product activity

To facilitate the periodic maintenance of the reactor coolant system, the nuclear power plant must be shut down. However, maintenance work on the primary loop components could result in post-shutdown radiation exposure. Activated corrosion products are the main source of radiation inside the containment building during post-shutdown. Traditionally, radiation detectors are used to measure in-containment radiation to protect workers from radiation risk during maintenance. However, these devices cannot predict corrosion product activity and the subsequent radiation dose. This study presents a minimum dose path planning methodology for protecting workers during the reactor coolant system maintenance in a typical pressurized water reactor nuclear power plant. Specifically, the study uses the corrosion product release and activation (CoPRA) code to estimate the corrosion product activity from the primary loop’s three major post-shutdown source terms. Subsequently, the FLUKA Monte Carlo code is used to simulate dose rates from the corrosion product activity, and the Three-Degree of Vertex shortest path algorithm is implemented to find the minimum dose path. Comparison results show that the predicted dose is consistent with the measured dose from the nuclear power plant. The results also show the flexibility and adaptability of the proposed method in finding the shortest path with the minimum dose in different reactor operating conditions.

Sludge Removal at Krško NPP Steam Generators during Outage

Sludge removal is performed on two steam generators (SG’s) at the Krško Nuclear Power Plant (NEK) during every outage. SG’s are a meeting point of four major plant systems: Reactor Coolant System (RC) on the primary side and three systems on the secondary side – Auxiliary Feedwater System (AF), Main Feedwater System (FW) and Main Steam System (MS). Sludge removal activities take place on the secondary side of the SG’s on the top of the tube sheet. It consists of classical Sludge Lancing (SL) which is done by spraying water at different angles (30°, 90°, 150°) between the tube gaps in the steam generator tube bundle with a pressure around 220 bars. Another method is Inner Bundle Lancing (IBL) which means spraying water at a much higher pressure (NEK’s contractor reached a pressure of approximately 590 bars). Such water is sprayed directly on the top of the tube sheet with a robot guided lance which is placed inside a steam generator. The robot is controlled by an operator and at times fully autonomous to provide the highest protection measures possible. After these activities, a televisual inspection (TVI) of the top of the tube sheet is performed to access the hard sludge area and to search for potential foreign objects in the SG’s. If an object is found, an attempt to retrieve it would usually take place. Other methods of sludge removal as upper bundle flushing or chemical cleaning have not been implemented in NEK thus far. Since the power plant uprate in May 2000, NEK conducted SL on both SG’s every outage with IBL in 2013 and 2015 and the same method was used in the 2018 outage. The purpose of these activities is mainly to extend the full load operation of the plant, prevent denting processes in the SG’s from occurring, stop the buildup of hard sludge area to increase/sustain efficiency and remove foreign objects found in the SG’s. SG’s U-tubes are a barrier between the primary side coolant and the secondary side of NEK and the environment. Therefore, it is crucial to keep the highest level of integrity of the U-tubes because any leak could potentially mean a release of radioactive material to the atmosphere. This paper describes the purpose and workflow of sludge removal in NEK.

Krško NPP Experience with RTD Bypass Elimination

This paper describes Krško experience and problems that had to be resolved with RTD Bypass Elimination project (RTDBE). Following RTD bypass manifold isolation valve leak in 2008, Krško decided to perform RTDBE modification on reactor coolant system narrow range temperature measurement system. The installation was performed during Outage 2013. Soon after the plant returned to power, newly configured measurement channels showed that OPΔT reactor trip was oversensitive to spikes caused by auxiliary relay operation in the cabinets nearby. The solution was to reconfigure OPΔT trip filtering constants to filter out short-term spikes in the signal. After almost full operating cycle of trouble-free operation, RTD failures started occurring on reactor coolant system cold leg, which was caused by the high frequency vibrations (3-5 kHz) induced by reactor coolant pumps. To resolve RTD failures, Krško ordered re-design of the RTDs to add robustness and specific qualification in high-frequency vibration operating environment. Improved RTDs were installed in Outage 2016 and were operating one full cycle with minor deviations.

RELAP5/MOD3.3 Analyses of Core Heatup Prevention Strategy during Extended Station Blackout in PWR

The accident at the Fukushima Dai-ichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power for several days, so called extended station blackout (SBO). A set of measures have been proposed and implemented in response of the accident at the Fukushima Dai-ichi nuclear power plant. The purpose of the study was to investigate the application of the deterministic safety analysis for core heatup prevention strategy of the extended SBO in pressurized water reactor, lasting 72 h. The prevention strategy selected was water injection into steam generators using turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump. Method for assessment of the necessary pump injection flowrate is developed and presented. The necessary injection flowrate to the steam generators is determined from the calculated cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed method allows assessment of the necessary injection flowrates of pump, TD-AFW or portable, for different plant configurations and number of flowrate changes. The RELAP5/MOD3.3 Patch04 computer code and input model of a two-loop pressurized water reactor is used for analyses, assuming different injection start times, flowrates and reactor coolant system losses. Three different reactor coolant system (RCS) coolant loss pathways, with corresponding leakage rate, can be expected in the pressurized water reactor (PWR) during the extended SBO: normal system leakage, reactor coolant pump seal leakage, and RCS coolant loss through letdown relief valve unless automatically isolated or until isolation is procedurally directed. Depressurization of RCS was also considered. In total, six types of RCS coolant loss scenarios were considered. Two cases were defined regarding the operation of the emergency diesel generators. Different delays of the pump injection start following the station blackout were assumed and analysed. For each scenario, two kinds of SBO calculations for two-loop PWR were performed, base and verification. Base calculations were needed to determine necessary minimum flowrate for steam generators feeding in such a way that they are not overfilled or emptied. Namely, it was assumed that instrumentation is not available during extended SBO. The verification calculations have been then performed to verify if the determined minimum flowrates are sufficient to prevent the core heatup. The calculated results show effectiveness of the proposed extended SBO prevention strategy provided that the water injection is available in the first two hours after SBO occurring at full power. If diesel generator is running after loss of offsite power for some time, e.g. one hour, the available time for steam generator water injection is significantly longer. The obtained results demonstrate the need for assessment of the pump injection flowrates before the utilization of the pump for mitigation of the event. The applicability of the developed method for assessment of the required pump injection flowrate has been validated on PWR.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Lines Due to Outflow Activities

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. Electric Power Research Institute (EPRI) has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicates that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from other thermal transients, especially those resulted from outflow activities, may also contribute to the failure of reactor coolant system (RCS) branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME section III class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients, particularly the transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

On the validation of ATHLET 3-D features for the simulation of multidimensional flows in horizontal geometries under single-phase subcooled conditions

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

Research on Diagnosis of Loss of Flow Accidents in Nuclear Reactor Coolant System

The reactor coolant system is one of the most important systems in a nuclear power plant. The safe and stable operation of the reactor coolant system is related to the safety of the nuclear power plant. With the rapid development of nuclear power technology, it is particularly important to ensure the safety of nuclear power plants. This paper takes the third-generation reactor AP1000 as the research object, firstly analyzes the typical accidents of the coolant system, and establishes a three-layer fault model, and then introduces the counterfactual reasoning method to take the Loss of Flow Accident as an example to identify the accident, and compares it with the traditional data-driven The Bayesian network diagnosis method was compared with the Bayesian network diagnosis method, and finally the accident was diagnosed by the simulation software Pctran. The results show that the method can accurately identify the fault and is better than the Bayesian method, which can provide decision support for accident handling and ensure the safety of the reactor.

Numerical simulation and experimental validation of flow characteristics for a butterfly check valve in small modular reactor

The butterfly check valve is widely used in small modular reactors to prevent pumps from being damaged by backflow during normal operating conditions. However, the large flow resistance of the valve has an important impact on the flow characteristics of the reactor coolant system during natural circulation state. The present study aims to perform and validate a series of simulations of the flow characteristics for a butterfly check valve in a small modular reactor employing the computational fluid dynamics (CFD) method. Steady state simulations were first performed to obtain the relationship between the valve loss coefficient and the valve opening angle. Based on dynamic mesh techniques, transient state simulations with five different turbulence models, standard k-ε model, RNG k-ε model, realizable k-ε model, SST k-ω model, and BSL-RSM model, were applied to obtain valve loss coefficient and valve flow coefficient respectively. Furthermore, experiments were carried out and compared with the results of transient numerical simulations. It must be pointed out that compared to other turbulent models, the simulation results using the standard k-ε model agree better with the experiment results. The experimental and numerical results show that the numerical simulation methodology established in this paper can be used to obtain the valve flow coefficient when a small modular reactor is under natural circulation conditions, and the coefficient can be used to accurately analyze the natural circulation characteristics of small modular reactors.

Mitigation of long-term station blackout accident of advanced boiling water reactors

Taiwan Power Company proposed ultimate response guidelines (URGs) to mitigate the so-called ‘Fukushima type accident’ of nuclear power plants. As stipulated in URGs, operators will depressurize the reactor coolant system if necessary to bring in low-pressure water. Containment venting is then used to dissipate the heat generated from the decay. These collective actions are called DIVing (depressurizing, injecting and venting) in URGs. The actions are designed to avoid large-scale evacuation of the public around the damaged plant during an accident. In this study, the effectiveness of DIVing in preventing radioactive emission to the environment is evaluated using RELAP5 codes. The plant analyzed is Lungmen Nuclear Power Station, which has two 1350-MWe advanced boiling water reactors (ABWR) designed by General Electric. In the Standard Operating Procedure of DIVing in URGs, controlled depressurization is executed to bring the pressure to the lowest operating pressure of Reactor Core Isolation Cooling (RCIC) system. The core water level is kept high while RCIC is in operation.. Emergency depressurization is carried out when RCIC fails. The executing time of DIVing as specified in URGs is too conservative. The study identifies the acceptable initial states for operators to carry out DIVing actions. The success criterion is to keep the peak cladding temperature below 1088 K (1500°F) during the transient, which is the temperature of cladding breach and the initiation of gap release of radionuclide from fuel rod.In this study, depressurization transients in the station blackout accident of ABWR are simulated using RELAP5/3D. The results show that the peak cladding temperature obtained during a depressurization transient depends on the time of depressurization, initial pressure, initial water level and injection rate of low-pressure water. When the water level and injection rate are fixed, a region of successful depressurization on the depressurization time vs pressure plot is delineated based on the results of RELAP5 simulations. Regions of successfully depressurization for typical combinations of injection rate and water level are identified, and acceptable region’s boundary is linear.A computation aid, Guidance for Emergency Depressurization (GfED), was developed based on the simulation results of the RELAP5/3D code to assist operators in determining whether the emergency depressurization under given conditions can be performed successfully. GfED uses low-pressure injection flow rate, water level and initial pressure upon emergency depressurization as input parameters to calculate the latest time that is acceptable for emergency depressurization. The technique of linear interpolation is adopted. By comparing the result and value input by the operators, GfED can predict whether it is safe to depressurize.GfED also guides operators to determine the initial condition of a successful depressurization. Three of the four input parameters are fixed in the guidance, and the limiting value of the fourth parameter of a successful depressurization is calculated. These values are informative for operators to manoeuvre the system while the reactor core isolation cooling system is still running.

Experimental investigation on thermal hydraulic interaction of RCS (reactor coolant system) and containment for intermediate break loss-of-coolant accident (IBLOCA) scenario in ATLAS-CUBE test facility

In order to experimentally investigate the thermal hydraulic interaction between the reactor coolant system (RCS) and the containment, the integral effect tests were conducted in a test facility, ATLAS-CUBE. Three tests for an intermediate-break loss-of-coolant accident (IBLOCA) were conducted for simulating the break of a cold leg or a direct vessel injection (DVI) nozzle. The effect of the mass/energy (M/E) supply from the break and the asymmetric phenomena inside the containment according to the break simulation position were investigated with considering its impact on the pressure/temperature (P/T) behavior inside the containment. As the test results, both of the RCS and the containment were effectively cooled down during the transient, by activation of the safety injection system and the containment spray system, respectively. The M/E supply from the RCS increased a pressure and a steam-gas mixture temperature of the containment, while a wall condensation on the passive heat sink and a spray injection contributed to cool down the steam-gas mixture. The two-phase flow characteristics of the break flow such as the quality or the subcooling of the liquid phase could highly affect the P/T build-up of the steam-gas mixture in the containment. The ATLAS-CUBE test data in this study will contribute to evaluate the thermal hydraulic safety analysis codes for the RCS and the containment.