Cold Leg Research Articles

Characteristics of vertically falling liquid film affected by cross gas flow in a narrow flat channel

Gas crossflow on a falling liquid film occurs in many practical cases in industries, particularly, in the downcomer of the APR1400, when accidents such as damage to the cold legs occur, hot steam crossflows over the vertical falling water film formed by the water impingement on the wall through a direct-vessel injection nozzle. Such crossflow may degrade the heat transfer ability of the liquid film. To investigate the film characteristics of downward flowing water, across which air was blowing horizontally, an experimental setup was designed and assembled. Experiments were conducted under different water flow rates (2, 3, 4, and 5 L/min) and different air velocities (0, 3.4, 4.9, 6.9, 8.4, and 9.8 m/s) to determine their effect on the film structure. A confocal chromatic sensor was employed to detect film thickness, film width, and wave amplitude. It was found that the base film thickness and wave amplitude do not vary when superficial air velocity is not larger than 4.9 m/s. As air velocity increases, the base film thickness exhibits a decreasing trend and wave amplitude exhibits an increasing trend. In particular, wave amplitude shows a different tendency from that found in annular flow. Moreover, local base film thickness and wave amplitude decreased and increased along the film flowing direction, respectively. Furthermore, the average film thickness was found constant throughout the whole air velocity range. In addition, film width expanded nonlinearly as the water flow rate increased, and the slope of the film centerline under different superficial gas velocities was presented for the first time.

Experimental measurements of flow mixing in cold leg of a pressurized water reactor

The main objective of OECD/NEA Cold Leg Mixing CFD-UQ Benchmark exercise is to evaluate the use of CFD simulations for safety issues related to mixing phenomena in the cold leg and downcomer of a Pressurized Water Reactor. An experimental facility has been designed and constructed to collect measurements of density-driven the flow mixing. Special techniques have been applied to simulate high density difference fluids while maintaining optical properties required to apply laser-based flow measurements techniques. Particle Image Velocimetry (PIV) was applied to determine the velocity field, combined with Laser Induced Fluorescence (LIF) to simultaneously capture the concentration of species. Pressure drop, cold front speed, and physical properties of the solutions have been presented. Results and measurement uncertainty are discussed. This article provides a unique set of data to support the advancement and the assessment of CFD codes in predicting mixing in buoyancy-driven flows in nuclear and non-nuclear applications.

Effect on hydrogen risk of total amount of inert gas during post-inerting in AP1000 NPP

To study the hydrogen risk at severe accident of LB-LOCA of cold leg in 1# SG compartment with gravity injection failure and hydrogen igniters without power in AP1000 NPP, we investigate the effect of total amount of inert gas on reducing hydrogen risk during AP1000 post-inerting. The total amount of inert gas is 42 tons, 56 tons and 70 tons respectively. We analyze the average volume fraction of hydrogen, flame acceleration factor and deflagration to detonation transition factor in 1# SG compartment and in the upper space of containment, and the pressure in containment. The results show that, post-inerting can reduce hydrogen risk in containment at severe accident, the more the total amount of inert gas is injected, the lower the possibility of FA and DDT is. Moreover, we must ensure that containment filtration and discharge system is functioning normally at accident conditions to prevent containment from overpressure.

Compatibility of FeCrAlMo with flowing PbLi at 500°-650 °C

To explore the upper temperature limit of a fusion blanket utilizing eutectic Pb-17at.%Li for tritium breeding, a series of three mono-metallic thermal convection loops have been fabricated from alloy APMT (Fe–21Cr–5Al–3Mo) and operated with commercial PbLi and increasing peak temperatures of 550°, 600° and 650 °C, flow rates of 0.4–0.7 cm/s and temperature gradients of 85–115 °C. Chains of APMT specimens including 25 mm long tensile specimens were hung in the hot and cold legs of the loops to evaluate post-exposure room temperature tensile properties after 1000 h exposures in each experiment. For specimens exposed to flowing PbLi above 500 °C, only minor changes in tensile properties were observed after exposure. Most APMT specimens were pre-oxidized to form alumina prior to exposure and exhibited small mass losses as the alumina was partially transformed to α-LiAlO2 that tended to spall after exposure. Specimens without pre-oxidation exhibited higher mass losses and formed γ-LiAlO2 but with less spallation. While promising for increasing blanket temperatures, the small mass losses suggest that the surface oxide is not completely stable. However, even at the highest temperatures, there was no indication of classic dissolution/precipitation behavior suggesting that higher temperature compatibility may be possible.

Two-phase pressurized thermal shock analysis with CFD including the effects of free-surface condensation

Pressurized thermal shock (PTS) is the stress experienced by a reactor pressure vessel (RPV) due to a rapid temperature and pressure change, or large spatial variations in temperature because of plumes and stripes of cold water surrounded by hotter water. PTS resulting from a loss-of-coolant accident was investigated for the nuclear Power Plant Gösgen-Däniken, which is a German-type three-loop pressurized-water reactor. The most critical regions in the RPV during such PTS transients are the most irradiated regions in the downcomer wall under the cold leg in the vicinity of the first weld of the first ring of the RPV. Cooling plumes or stripes develop at the RPV wall or core barrel wall depending on parameters like the water level in the downcomer, the mass flow rate of the injected cooling water and condensation effects.For fracture-mechanics analysis of the RPV, temperature profiles and heat-transfer coefficients corresponding to either cooling plumes or stripes at the wall (here mainly at the inner RPV wall) are needed. Various methodologies can be followed for generating the temperature profiles and heat-transfer coefficients resulting from the mixing of cold injection water with the hot RPV water. One is the calculation of these data by analytical fluid-mixing codes validated with experiments, such as KWU-MIX. Alternatively, computational fluid dynamics (CFD) tools can be used after suitable validation. One advantage of CFD is the possibility to show attachment of plumes and stripes to the inner wall of the RPV or detachment towards the core barrel.A free-surface condensation model was implemented in ANSYS CFX for simulating two-phase transients when the water level is below the cold-leg nozzle. For the purpose of validating free-surface condensation and the maximum mass flow rate for the stripe of liquid water before detachment from the RPV wall, experimental data from the full-scale Upper Plenum Test Facility (UPTF) were used.The comparison of CFD and KWU-MIX analyses of the limiting transient for the core weld obtained from the selection process with KWU-MIX shows that the inherent conservatism in the standard KWU-MIX approach leads to higher thermal loading than the best-estimate CFD approach and demonstrates the benefits of a CFD tool to quantify potential margins.

Flow Boiling in Minigap in the Reversed Two-Phase Thermosiphon Loop

The paper presents the results of experimental investigations of a model of a heat exchanger featuring a minigap, which is perceived as an evaporator for an inverted thermosiphon. The system works with a single component test fluid. The tested evaporator generates pumping power in the test loop in a way similar to the mammoth pump. The tests regarded a module of the heat exchanger, consisting of a hot leg and a cold leg with the width by the length of 0.1 × 0.2 m, heated by a uniform heat flux. In the tests, the minigaps of 1, 2 and 3 mm were formed. Two fluids, namely, distilled water and ethanol, were tested in the facility. Two-phase flow structures for both working fluids and various operational parameters, together with comprehensive visualization material, are presented. The specifics of pressure changes and its influence on operating parameters and flow structure are discussed.

Assessment of VVER 1000 core degradation for bounding cases with ASTEC 2.1.1.0

This paper presents an assessment and comparison of the core degradation progression in four Station Blackout (SBO) scenarios analysed by the ASTEC computer code. Two types of SBO scenarios namely with high- and low-pressure conditions are considered. The low-pressure conditions in the SBO scenarios were simulated by introducing a small break loss of coolant accident in the cold leg with an equivalent internal diameter of 80 mm. The selection of the break size was based on the idea to have a significantly faster primary side pressure reduction for simulation of low-pressure conditions and a significantly longer coolant injection from the hydro accumulators (passive safety system).For both types of selected scenarios, the plant behaviour was analysed without operator actions, which allows assessing the times to reach the important set points during the accident progression. The set points include; a) the prediction of the dryout of the steam generators (SG) (or loss of SG effectiveness), b) the loss of natural circulation leading to core uncovery and heat up, c) the beginning of hydrogen generation, d) different stages in core degradation, e) the actuation of the passive safety system etc. The analyses were performed until reactor vessel failure takes place in both of the investigated scenarios. The purpose of these analyses is to study the reactor core behaviour parameters and to estimate the time available to perform operator actions.In addition, this investigation is focused on the assessment of the effectiveness of operator actions as prescribed in the Severe Accident Management Guidlines (SAMGs) for the investigated reactor type. The referenced NPP is KNPP equipped with two VVER 1000 V320 reactors. The ASTECv2.1.1.0 computer code has been used for the investigation. The aim of these analyses is to assess the possibility preserving the reactor core from damage during a severe accident and to assess the hydrogen generation that occurs as a result of the overheated core reflooding at high- and low-pressure initial conditions. The injection of a coolant by an active system will start at the same core exit temperature for each scenario, but at different pressure.

ROSA/LSTF test and RELAP5 code analyses on primary feed-and-bleed operation during total loss-of-feedwater transient of PWR

An experiment was conducted using the rig of safety assessment/large-scale test facility (ROSA/LSTF), which simulated the primary feed-and-bleed operation during a total loss-of-feedwater transient of a pressurized water reactor. The primary feed-and-bleed operation included coolant injection from high-pressure injection (HPI) system into cold legs and the full opening of a power-operated relief valve (PORV) of pressurizer (PZR). In the LSTF test, the primary and steam generator (SG) secondary-side pressures were maintained, respectively, at around 16 MPa and 8 MPa by cycle opening of the PORV and of the SG relief valve until the start of the PORV full opening. Coolant behavior in the PZR affected the primary depressurization. The RELAP5/MOD3.3 code predicted well the overall trends of the major thermal hydraulic responses observed in the LSTF test, and indicated insufficient predictions of primary pressure, primary coolant distribution, and accumulator flow rate. The sensitivity analysis results clarified the primary pressure, the core collapsed liquid level, and the cladding surface temperature are largely influenced by the combination of the HPI total flow rate, the onset timing of the PORV full opening, and the PORV flow area within the defined calculation conditions.

Improvement of horizontal stratification criteria in the SPACE code

The flow regime for two-phase flow plays an important role in nuclear safety analysis, because different physical models for two-phase flow are used depending on the flow regime. This paper deals with the flow regime criteria for horizontal stratification, which frequently occurs in the hot or cold leg during a loss-of-coolant accident (LOCA). Although there are several horizontal stratification criteria based on the Kelvin-Helmholtz theory, the condition for the onset of waves leading to a wavy-stratified flow or a non-stratified flow is still unclear. This study approached the horizontal stratification criteria from the perspective of viscous/inviscid flow instability theories. The criterion between smoothly stratified flow and transition region was adopted with viscous Kelvin-Helmholtz (VKH) analysis. The onset of ill-posedness for the governing equation set, which is related to the inviscid Kelvin-Helmholtz (IKH) analysis, was used to determine the transition criterion between the transition region and non-stratified flow. These criteria were incorporated into the SPACE thermal-hydraulic system code. Simulation results show good agreements with experimental data, and the predictability of horizontal stratification is thereby further improved.

Simulation of hydrogen distribution and effect of Engineering Safety Features (ESFs) on its mitigation in a WWER-1000 containment

In this study, thermal–hydraulic parameters inside the containment of a WWER-1000/v446 nuclear power plant are simulated in a double-ended cold leg accident for short and long times (by using CONTAIN 2.0 and MELCOR 1.8.6 codes), and the effect of the spray system as an engineering safety feature on parameters mitigation is analyzed with the former code. Along with the development of the accident from design basis accident to beyond design basis accident, the Zircaloy–steam reaction becomes the source of in-vessel hydrogen generation. Hydrogen distribution inside the containment is simulated for a long time (using CONTAIN and MELCOR), and the effect of recombiners on its mitigation is analyzed (using MELCOR). Thermal–hydraulic parameters and hydrogen distribution profiles are presented as the outcome of the investigation. By activating the spray system, the peak points of pressure and temperature occur in the short time and remain below the maximum design values along the accident time. It is also shown that recombiners have a reliable effect on reducing the hydrogen concentration below flame propagation limit in the accident localization area. The parameters predicted by CONTAIN and MELCOR are in good agreement with the final safety analysis report. The noted discrepancies are discussed and explained.

Thermal-hydraulic modeling and analyses of the water-cooled EU DEMO using RELAP5 system code

The conceptual design of the Primary Heat Transfer System (PHTS) of the water-cooled European (EU) DEMO foresees two independent cooling circuits, the breeding zone PHTS and the first wall PHTS. During the pulse time (120 min) the first delivers thermal power to the turbine, the latter delivers thermal power to the Intermediate Heat Transfer System (IHTS) equipped by an Energy Storage System (ESS). The IHTS delivers partially thermal power to the turbine in pulse so that to accumulate a suitable amount of energy in ESS to operate the turbine during the dwell time (10 min) at almost constant load, despite the EU DEMO pulsation of the generated thermal power. A dynamic model of the primary systems of water-cooled EU DEMO is developed using RELAP5/Mod3.3 system code to verify components sizing and to investigate code predictive capabilities. The model includes the primary and secondary side of the breeding zone and first wall, and in particular: in-vessel (i.e. breeding blanket) and ex-vessel components (i.e. main collectors, hot and cold legs, heat exchangers, steam generators, pumps). Preliminary assessments of the nodalization have been carried out, in particular checking pressure drops along the systems and heat exchanger performances.

Scaling and integral effect test on transition from depressurization to long-term gravity injection

The passive safety commercial pressurized water reactors (PWRs), AP1000 and CAP1400 are equipped with the automatic depressurization system (ADS). The fourth stage of ADS (ADS-4) has the largest discharging capacity among the all ADS stages. During a small break loss-of-coolant accident (SBLOCA), only when the system is fully depressurized by the ADS-4 to near the containment atmospheric pressure, the in-containment refueling water storage tank (IRWST) can deliver the large coolant inventory for the long-term core cooling by its water head. Before the IRWST injection, the core reaches its minimum inventory. So the transition from the ADS-4 depressurization to the IRWST injection is a very challenging period in a SBLOCA. Therefore, to evaluate the passive system performance by the integral effect test (IET), the transition must be simulated properly. This work conducted the scaling analysis for the ADS-4 depressurization and the IRWST injection for designing the related components of the ACME IET facility. A uniform scaling law based on the existing T-branch liquid entrainment models was proposed. For scaling the onset of the IRWST injection, the requirement for reproducing the synergetic effect in an IET was introduced in order to preserve the important event/process occurrence time with the proper conditions, and the scaling laws for the full pressure IRWST injection were obtained. For investigating the transition thermal hydraulic behaviors and the ADS-4 configuration influences on the core safety, one cold leg break test and three double-ended direct-vessel-injection (DEDVI) line break tests with the different ADS-4 failure modes were selected from the ACME SBLOCA test matrix, and the main test data related to the core safety were analyzed and compared. It was found that the ADS-4 actuation significantly accelerated the system depressurization but worsened the core inventory depletion at the same time. In addition, the break condition and the ADS-4 failure mode could differently affect the system depressurization and the core level. The ACC injection during the ADS-4 depressurization in DEDVI was critical for the core level makeup. The insufficient ADS-4 venting capacity delayed IRWST injection, but also mitigated the core inventory depletion. It suggests the ADS-4 design can be potentially optimized in the passive safety system design.

Emergency condition analysis for MBLOCA along with SBO initiated severe accident using MELCOR

Much interest in severe accident research has been attracted since Fukushima accident, which indicates that severe accident could possibly happen although the probability is extremely low. Although many researches have been conducted in severe accident progression simulation and emergency condition analysis, it is still necessary to do some research on emergency conditions of severe accident that have not been sufficiently analysed. In this paper, a 250mm medium break loss of coolant accident (MBLOCA) in cold leg along with station blackout (SBO) initiated severe accident was simulated using MELCOR code. Before the calculation, steady state was simulated in order to check the model. Main parameters of the model were compared with operating parameters of nuclear power plant (NPP), and relative error was acceptable, which indicated that the model could be used to simulate the accident. Emergency condition progression of the accident was obtained, including start of core uncover, start of zirconium-water reaction, failure of fuel cladding and failure of the lower head. Thermal-hydraulics response of reactor was analysed. The calculation ended with failure of lower head, and the response of the containment was not studied. Results show that a 250mm break is relative a large one among MBLOCA, and the progression is similar with that of large break loss of coolant accident (LBLOCA). The accident progresses quite fast. The core starts to melt at 526s, and lower head fails at 2314s. Results of the paper could be used to provide basis for decision making in case of emergency and severe accident mitigation.

Analysis of the thermal-hydraulic parameters of VVER-1200 due to loss of coolant accident concurrent with loss of offsite power

This paper aims to evaluate severe accident in VVER-1200. The consequences of severe accident depend greatly on containment safety features and its performance in retaining radioactive material. The specific type of LOCA includes hot leg or cold leg break in the primary coolant circuit. For both simulations a loss of offsite power is also initiated after 500 seconds. It shows that the Reactor Cooling System (RCS) pressure drops from 155 bars to below 5 bars in both the cases to reduce RCS leakage. Reactor Core Thermal Power, Nuclear Flux Power and Turbine Load fall significantly after 23.5 seconds because of the occurrence of Reactor Trip. Both HA-1 and HPI are activated to prevent Core damage. The flow of HPI drops to zero after the loss of offsite power. The temperature of the fuel falls below 200 ºC within 150-200 seconds of the malfunction in both the cases. The corresponding pressure, temperature, void fraction of reactor etc. are analysed in this paper. The results obtained satisfactorily meet all PSAR (Preliminary Safety Assessment Report) data regarding Loss of Coolant Accident (LOCA). The results provide useful information about safety features of VVER-1200 in case of Loss of Coolant Accident (LOCA) concurrent with loss of offsite power.

Evaluation of atmospheric dispersion of radioactive materials in a severe accident of the BNPP based on Gaussian model

Regarding the importance of radioactive materials dispersion in the environment and their effect on living organism's well-being, this article examines the atmospheric dispersion of these materials around the Bushehr nuclear power plant (BNPP) in a hypothetical severe accident. One of practical dispersion models is AERMOD which is based on Gaussian distribution. This model which is an extension of Gaussian model is normally used for predicting the gaseous pollution. The aim of this study is to investigate the dispersion of radioactive materials in BNPP's surroundings due to a specific type of Large Break LOCA (LBLOCA) accident. In this study, the wind direction, wind speed, and the topological conditions of the power plant's environment are considered. In order to execute this model, the meteorology and topological data of the region have been provided. Besides, using the CONTAIN code, the data about released radioactive materials from the containment have been provided. The double ended cold leg break (DECL) accident in containment has been modeled and its validity has been checked with FSAR. Extracting all required data, the simulation is done in AERMOD and the volumetric concentration and activity of fission products are calculated. The volumetric activity of fission products in Tangestan, Bandar-E-Bushehr and Delvar cities are evaluated as 7.5658E+07, 7.5136E+07 and 2.9655E+06 (Bq/m3), respectively. From results it could be seen that the most dispersion of the radioactive materials are in direction of the non-residential regions, although the pollution of the cities around the power plant is inevitable.

Development of MELCOR thermal hydraulic model of AP1000 and its verification for a DECL break

The AP1000 is currently an advanced pressurized water reactor with emergency core cooling systems (ECCS) completely dependent on natural circulation. It is under construction in the USA and has been commissioned in China and is being considered by several countries to use this advanced technology for a power source. With this objective in mind, a representative MELCOR model (for the AP1000) has been developed to better understand the thermal hydraulics involved and to assess the performance of the safety systems for a design basis accident such as large break LOCA (LBLOCA) with availability of the safety systems. The results from these calculations were compared to those obtained using WCOBRA-TRAC, developed by Westinghouse Electric Company, which are reported in the AP1000 European Design Control Document.MELCOR was selected for the design assessment activities specifically because of its capabilities to model severe accident scenarios and features such as containment behaviour. It is one of the system thermal hydraulic codes, which are recommended for severe accident analysis in light water reactors. The aim of the current study is to assess the thermal hydraulic models and correlations in MELCOR against a specific code developed for a design basis accident such as a LBLOCA. This exercise will validate the input model for a LBLOCA and then use the input model for station blackout studies and the impact on the containment.A double ended cold leg (DELC) break in the loop containing the Core Makeup Tanks (CMT) is considered for this study. The calculated break flow, peak cladding temperature, accumulator, and CMT injection flow rates are compared to those reported. These calculated results are found to be satisfactory according to ECCS acceptance criteria for the temperature and hydrogen generation. This study has been further extended to analyse the containment behaviour during the accident. The calculations for the transient were extended to observe the behaviour of the Automatic Depressurization System (ADS) and IRWST injection capabilities which can effectively avoid core uncovery for a longer duration of the time. These systems can only work when the reactor system is depressurized. MELCOR calculated a PCT value (occurs just after the break) as 1242 K while the total hydrogen produced was calculated to be 2.216 kg which is only 0.61% of the total potential hydrogen production from oxidation.

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.

Numerical and experimental analysis of a rectangular single-phase natural circulation loop with asymmetric heater position

A simple rectangular natural circulation loop (NCL) consists of a heater, a cooler, cold leg and hot leg. The main aim of an NCL is to reject heat from heat sources to heat sinks with the use of a mechanical pump. NCL is used in various energy systems, such as solar heaters, nuclear reactors, geothermal power production and engine and computer cooling. In this study, a simple rectangular single-phase NCL with an asymmetric heater is numerically analyzed with respect to tested experimental data. First governing equations, i.e., mass, momentum, and energy balance equations are derived. Then, these equations are written in dimensionless form with proper definitions of the dimensionless parameters. The boundary conditions of NCL are heater power and cooler temperature that are modeled by modified Grashof number and modified Stanton number, respectively. A continuous friction law is used for numerical solution of governing equations that are applicable for the laminar, transition and turbulent flow regimes. The governing equations are solved for both the steady state and transient. The equations of the transient state are solved with both linear and nonlinear methods. In most of the earlier literature, a stability map has been reported that is only applicable for laminar or turbulent flows. But in the present work, a stability map is obtained, which is valid for all the three regions (laminar, transition, and turbulent). Finally, the experimental data are compared with the numerical results.

Large Eddy Simulation of the injection of cold ECC water into the cold leg of a pressurized water reactor

The aging of the reactor pressure vessel (RPV) can be a limiting factor for the lifetime extension of pressurized water reactors. In case of loss of coolant accidents, cold emergency core cooling (ECC) water will be injected under high-pressure thermal-hydraulic conditions into the cold leg that is filled with hot coolant. The hot RPV wall can be exposed to this cold fluid, suffering important thermal stresses. This process is called pressurized thermal shock (PTS). The efficiency of the PTS depends on the mixing between cold ECC water and hot coolant in the cold leg and the upper downcomer. A single-phase PTS scenario was investigated experimentally in the TOPFLOW facility under representative thermal-hydraulic conditions. A transient of 390 s of this experiment is analyzed here numerically with Large Eddy Simulations and two different hypothesis to account for thermal effects: incompressible fluid and dilatable fluid (low Mach number compressible fluid). Detailed information is given on the temporal development of both thermal stratification and flow behavior in the cold leg. Calculated temperature profiles in the cold leg are compared to detailed temperature measurements. The dilatable approach reproduces well the experimental results and outperforms the incompressible fluid approach.

Numerical analysis of the ROCOM boron dilution benchmark experiment using the CUPID code

Abstract The Rossendorf Coolant Mixing Model (ROCOM) boron dilution benchmark experiment was assessed using the CUPID code in the framework of a Coordinated Research Project by the International Atomic Energy Agency, namely, “Application of Computational Fluid Dynamics Codes for Nuclear Power Plant Design.” Two-equation Reynolds-Averaged Navier–Stokes (RANS) turbulence models, such as the standard k-e model, Re-Normalization Group (RNG) k-e model, and Shear Stress Transport (SST) k-ω model, were tested, and predictions of boron concentration were compared to measurements at the downcomer and core inlet in the reactor pressure vessel. The analysis result showed that the de-borated water slug injected from one cold leg was diffused by the turbulence mixing effect as it flowed downward in the downcomer and passes through a perforated drum in the lower plenum. The applied turbulence model affected the mixing patterns of the de-borated water, which resulted in a local distribution of boron concentration at the core inlet. This was the main concern of the boron dilution phenomena from the viewpoint of safety analysis. The result showed combining the CUPID code with the RANS turbulence models could reasonably predict the boron dilution phenomena in the reactor vessel when the proper y-plus wall treatment was assured, considering the characteristics of the turbulence models.