In-vessel Retention Strategy Research Articles

Critical heat flux enhancement by porous coatings and prediction model development under External Reactor Vessel Cooling conditions

Critical heat flux (CHF) is the important limitation to estimate the effectiveness of the In-vessel retention (IVR) strategy for the severe accident mitigation. However, as the reactor power increases in the new-generation nuclear power plant, it is necessary to ensure the sufficient safety margin by taking some measures for the CHF enhancement. Porous coatings are recognized as an effective technique to improve the CHF. In this paper, the applicability of the porous coatings has been evaluated. The porous coatings consist of a dense layer and a porous layer. CHF experiments were constructed at the REPEC-III facility with 1:1 height ratio to the prototypic External Reactor Vessel Cooling (ERVC) environment. CHF on the porous surface is enhanced 32.4%–48.5 % under the ERVC conditions. Based on the experimental data, a mechanistic model considering the porous surface characteristics to predict CHF of the flow boiling under the ERVC natural circulation is developed and has a good predictive ability. The predicted CHF is compared with the experimental data of large-scale ERVC experiments. The mechanistic model can predict CHF within ±15 % error. Through the full-height experimental and theoretical analyses, porous coatings are an effective method to improve CHF and has a good application perspective for the IVR strategy in the nuclear power plant.

Experimental investigation on creep behaviors and life prediction across phase-transformation of thermal aged 16MND5 steel

Reactor pressure vessel (RPV) is one of the key equipment for nuclear power plant, and long-term exposure to high temperature is an important factor causing its performance deterioration. Under core meltdown conditions, RPV steel experiences high temperature creep due to the In-Vessel Retention (IVR) mitigation measures. This paper aims to investigate high-temperature mechanical properties, especially for the creep behaviors of 16MND5 steel after thermal aging at 650 °C for 2000 h. The results showed that the yield stress and ultimate tensile strength of the 16MND5 steel are decreased, and the creep rupture time is rapidly shortened after thermal aging. The microstructure was analyzed by scanning and transmission electron microscopy (SEM, TEM). 16MND5 steel showed a typical bainitic phase before thermal aging. After 2000 h of thermal aging, the micro-grains were coarsened, and finally changed into a ferritic microstructure. Based on Larson-Miller (LM), Orr-Sherby-Dorn (OSD) parameters methods and Kachanov-Rabotnov (K–R) model to predict uniaxial creep life after thermal aging, the prediction results are satisfactory, and all within the double tolerance band. This work reveals the creep damage mechanism of 16MND5 steel after thermal aging, and provides basic data for the implementation of IVR strategy in nuclear power plants after long-term service.

Critical heat flux model for heated horizontally oriented cylindrical vessel and its application to PHWR calandria under severe accident scenario

In-vessel retention (IVR) of molten corium is a key severe accident management strategy adopted in pressurized heavy water reactors (PHWRs) and is recognised as the only option to manage a complete core melt scenario. Since the PHWR calandria vessel is completely surrounded by a large pool of water, the vessel itself is expected to act as a ‘core catcher’. The success of IVR strategy depends up on whether the actual heat flux imparted by molten corium is less than the critical heat flux (CHF). In the present work, a hydrodynamics model is developed and applied to obtain the variation of CHF along outer surface of 220 and 700 MWe Indian PHWR calandria vessel and tube. The thermal margin available for success of IVR strategy is evaluated. The governing equations for external buoyancy driven boundary layer flow are derived and solved in non-dimensional form. The details of the mathematical model, benchmarking exercises, validation aspects and application of the model to PHWR calandria vessel and tube are discussed. Parametric studies are presented to bring out the influence of diameter, system pressure, subcooling of water, depth of submergence and slip ratio on the variation of CHF along the cylindrical vessel surface.

Development and Validation of Thermal-Mechanical Creep Failure Module for Reactor Pressure Vessel Lower Head

For severe accidents, in-vessel retention (IVR) is a very effective and crucial severe accident mitigation measure. The lower head of the reactor pressure vessel plays a vital role in the IVR strategy. The failure of the lower head may lead to the release of radioactive substances into the environment. During the implementation of IVR, the lower head is in a high-temperature environment, and its main failure form is creep failure. Therefore, to ensure the successful implementation of the IVR strategy and prevent radioactive material leakage, it is necessary to conduct an in-depth analysis of the lower head. In this paper, the lower head thermal-mechanical creep failure (LHTCF) module is developed based on the theory of plate and shell and Norton-type constructive creep laws. Through the mechanical analysis of the lower head, seven failure criteria are used to evaluate the integrity of the lower head. Finally, the LHTCF module is integrated into the integrated severe accident analysis (ISAA) program, and the accuracy of the module is validated by numerical calculation of the Organisation for Economic Co-operation and Development Lower Head Failure (OLHF) experiment. Through the comprehensive judgment of different failure criteria, the final simulation results are in good agreement with the experimental data. The results show that the wall thickness at the crack decreases sharply before failure due to the effect of creep, and the stress increases abruptly at the failure time. The LHTCF module developed in this paper can accurately predict the creep behavior of the lower head, and the calculated failure time, position, and thickness distribution agree well with the experimental results.

Numerical Analysis of Heating Technique in Corium Melt Pool Convection Flow Field & Thermal Interaction in a Volumetrically Heated Molten Pool

In-Vessel Retention (IVR) is one of the existing strategies of severe accident management of LWR, which intends to stabilize and isolate corium & fission products inside the reactor pressure vessel (RPV) and primary containment structure. Since it has become an important safety objective for nuclear reactors, it is therefore needed to model and evaluate relevant phenomena of IVR strategy in assessing safety of nuclear power reactors. One of the relevant phenomena during accident progression in the oxidic pool is non-uniform high heat generation occurring at large scale. Consequently, direct experimental studies at these scales are not possible. The role computer codes and models are therefore important in order to transpose experimental results to reactor safety applications. In this paper, the state-of-the-art ANSYS FLUENT CFD code is used to simulate Non-uniform heat generation in the lower plenum by the application of Cartridge heating under severe accident conditions to derive the basic accident scenario. However, very few studies have been performed to simulate non-uniform decay heat generation by Cartridge heaters in a pool corresponding lower plenum of power reactor. The current investigation focuses on non-uniform heating in the fluid domain by Cartridge heaters, which has been done using ANSYS FLUENT CFD code by K-epsilon model. The computed results are based on qualitative assessment in the form of temperature and velocity contour and quantitative assessment in terms of temperature and heat flux distribution to assess the impact of heating method on natural convective fluid flow and heat transfer.

Thermodynamic analysis for molten corium stratification with U-Zr-O-Fe database

Aided by CALPHAD (CALculation of PHAse Diagram) method, a self-consistent thermodynamic corium database of U-Zr-O-Fe system is established based on reoptimization of key binary and ternary subsystems. According to molten corium stratification thermodynamic analysis for STFM-Fe tests in MASCA project, the corium would separate into metallic liquid phase and oxidic liquid phase due to the existence of miscibility gap and their mass fraction and composition can be well predicted. Further, the density of the oxide phase is accurately calculated by a linear fitting relationship obtained from the measured oxide density, and the density of the metal phase is calculated by the empirical formula. Thus the molten corium would exhibit a two-layer structure with a higher density metal phase below and a lower density oxide phase above. The good agreement with the STFM-Fe tests indicates the reliability of the database to analyze the molten corium stratification which tends to improve the effectiveness of in-vessel retention strategy by providing the thermodynamic prediction.

Simulation model for investigating safety margin of three-layer molten pool after late phase in-vessel water injection

The thermal focusing effect is an essential issue in the research of In-Vessel Retention (IVR) because it might cause vessel failure. Late phase in-vessel water injection is applied to the three-layer corium pool surface to alleviate the severe accident and strengthen the IVR strategy. Analyzing the heat transfer in this corium pool after in-vessel water injection is an important issue. A model named “Three-layeR configuration mOlten pool after the late phase in-vessel water Injection under ERVC-IVR Strategy (TROIS)” has been developed for analyzing the heat transfer issues. Its calculated results could reasonably analyze a two-layer case of AP600 and a three-layer experiment with both the top and sidewall cooling conditions. Furthermore, it is applied to analyze the hypothetical core degradation case of AP1000. The calculated results demonstrate that the safety margin of the two-layer configuration is larger than that of the three-layer configuration. And vessel failure would not occur in the three-layer configuration with or without in-vessel water injection, according to the comparison results of local heat flux and CHF. Late phase in-vessel water injection can alleviate this thermal focusing effect and enlarge the safety margin.

Experimental Research for CHF Sensitivity of Heat Flux Distribution under IVR Conditions

In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is one of the most effective severe accident mitigation measures in the nuclear power plants. The most influential issues on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the lower head, and the external cooling of reactor pressurized vessel (RPV). In the molten pool research, there are mainly two different molten pool configurations: two layers and three layers. Based on the different distributions of heat flux in molten pool configurations, a new problem was raised: whether the in-vessel heat flux distribution will affect the CHF on the outer wall of RPV and further affect the effectiveness of IVR measures? A full-height external reactor vessel cooling and natural circulating facility was conducted to study the CHF sensitivity of different heat flux distributions. The experimental results show that the characteristics of natural circulation are similar and the CHF of the RPV lower head external surface is not obviously affected under the different heat flux distributions. The varying heat flux distribution during severe accident process will not threaten significantly the success of IVR strategy.

Experimental study on the CHF enhancement effect of nanofluids on the oxidized low carbon steel surface

Studies about using nanofluids to enhance the Critical Heat Flux (CHF) of In-vessel Retention (IVR) strategy in the third-generation reactor have been conducted extensively and show a significant CHF enhancement effect. However, low carbon steel SA508 used in the reactor vessel is easy to oxidize and the oxidation can lead to changes in the surface properties which may affect the CHF enhancement effect of nanofluids. In this study, pool boiling CHF experiments with low carbon steel SA508 surfaces were conducted in distilled water and nanofluids under different boiling time to investigate the CHF enhancement effect of nanofluids under low carbon steel surface oxidization condition. CHF in distilled water increases rapidly with boiling time due to the rapid surface oxidation during the boiling and the increase ratio can be nearly 2 due to the surface oxidation. CHF in nanofluids is stable and independent of boiling time. The difference between CHF in nanofluids and CHF in distilled water decrease to 17% under the longest boiling time conditions due to the surface oxidation. The deposition layer of nanoparticles on the surface leads to the capillary wicking and decrease in the nucleation site and thus, CHF is enhanced. This study is of great significance for exploring the actual effect of nanofluids on the CHF enhancement of IVR strategy.

Heat transfer experiment with the prototypical Ra number of heavy metallic layer in IVR strategy

In-Vessel Retention (IVR) strategy is significant to limit molten corium in the Reactor Pressure Vessel (RPV) during severe accident scenarios. A heat transfer experiment has been established to study the thermal load on the cooled wall adjacent heavy metallic layer. The test section of the heavy metallic layer experiment is a hemisphere, and its diameter is 2.4 m; the height of simulant is set to 0.3 m. Heavy metallic layer experiment uses water as the simulant, and it uses the internal heating coils to represent the decay heat. Three independent sidewall coolant channels are used to cool the whole test section. The top boundary condition of the test section could be a cooling or insulated condition. The experimental results illustrate that the Mayinger heat transfer correlation could predict the heavy metallic layer experiment results, and it is recommended to predict the sideward heat flux. Applicating the heat transfer correlation and experimental qlocal/qmean distribution to the heavy metallic layer in core degradation situation, the heat transfer correlation calculated results demonstrate that it is unlikely for the local sideward heat flux to exceed the Critical Heat Flux (CHF). The possible top cooling boundary for the heavy metallic layer contributes to alleviating wall melting. Generally, it is unlikely that the vessel failure occurs at the cooled wall adjacent the heavy metallic layer.

Numerical Study on the Effect of Jet Cross Section Shape on the Corium Jet Breakup Behavior With Lattice Boltzmann Method

In a core meltdown accident in light water reactors, molten corium may drop into the lower plenum of the pressure vessel and interact with water, which is called fuel–coolant interaction (FCI). The behavior of the corium jet breakup in water during FCIs is important for the in-vessel retention strategy and has been extensively studied. While in previous studies, the jet cross-section shapes are naturally assumed to be circular, which is actually not always the case, in this study, the breakup processes of the corium jets with four different elliptical cross-section shapes and three different penetration velocities are simulated with color-gradient lattice Boltzmann method. The effect of the cross-section shape on the hydrodynamic breakup behavior of the corium jet is analyzed in detail. It is found that the effect of the cross-section shape on the jet penetration depth is very limited. With the increase in the aspect ratio under the same penetration velocity, the jet breakup length decreases gradually. In general, the dimensionless corium surface area increases with the increase in the aspect ratio for the jets under the same penetration velocity.

Investigation of water injection on molten Zirconium-Stainless steel pool in IVR strategy

Water injection on the surface of the molten pool could enhance the effectiveness of In-Vessel Retention (IVR) strategy and mitigate the thermal load on the vessel wall. The wATer injectiOn on molten Zirconium-Stainless steel Metallic pool experiment (ATOM) has been established to investigate the effect of water injection on the molten pool for IVR strategy. The mass percentage of Zr is 59.5% for the simulant. The mass flow rate of the water injection is 0.10 kg/s, and the maximum heat flux removed by water is 1.18 MW/m2. The experimental results indicate that water injection on the surface of the light metallic layer can significantly reduce the heat flux transferred to the sidewall and mitigate the thermal focusing effect on the Reactor Pressure Vessel (RPV). Hydrogen is produced during the water injection. Neither the vapor explosion nor the melt ejection is found during the whole experiment.

Investigation on the effect of low aspect ratio on heat transfer characteristics for light metallic layer

The nuclear safety and the integrity of the Reactor Pressure Vessel (RPV) could be ensured by the In-Vessel Retention (IVR) strategy during the severe nuclear accident. The HEat transfer behavior of Light Metallic layer experiment for the Low aspect Ratio (HELM-LR) has been established to investigate the heat transfer characteristics and the concentration factor of the light metallic layer in a three-dimensional molten pool. The melt simulant is water. The test section is heated from below and can be cooled from both top and side. The experimental results demonstrate that the melt temperature increases along the vertical direction and decreases along the radial direction. The possible natural convection in the light metallic layer is formed, due to the temperature difference between the melt near the center and the melt near the sidewall. However, the effect of the natural convection would be limited with the low aspect ratio (H/D). Increasing the top cooling could lead to reducing the concentration factor (qside/qinput). When the bottom heat flux is 1 MW/m2 and the concentration factor is below 204%, the heat flux transferred to the AP1000 sidewall would not exceed the Critical Heat Flux (CHF) and the effect of IVR strategy would be enhanced. Consequently, it could avoid the vessel failure during the nuclear severe accident.

Research on debris in-vessel cooling and retention behavior for the small modular reactor ACP100

In this paper, the debris in-vessel cooling and retention behavior of the small modular reactor ACP100 design in severe accident is studied using MELCOR 2.1 code. A complete MELCOR input deck of ACP100 has been built. In ACP100, in-vessel retention strategy is essential for safety enhancement, and is realized by passive water injection into the flow channel between reactor pressure vessel outer surface and reactor cavity, namely cavity injection system, when the core exit temperature is higher than 650 °C. The severe accident is conservatively assumed to be triggered by a double-ended rupture of one direct vessel injection line. The collapse behavior of the fuel assembly is estimated by MELCOR fuel rod degradation model, and the failure behavior of the lower core plate is predicted by the ANSYS software. The results show that the core center partially collapse and melt to form a core molten pool, while most fuel assemblies surrounding the core molten pool remain intact. The lower core plate is still able to support the core molten pool and uncollapsed fuel assemblies throughout the simulation. The heat transfer processes associated with the core barrel, reactor pressure vessel, and lower core plate are also discussed. The decay heat can be effectively removed by the cavity injection system, so that debris retention in the reactor pressure vessel is feasible and even the appearance of molten pool in the lower head can be avoidable.

Studies on Key Effect Factors of Natural Circulation Characteristics for Advanced PWR Reactor Cavity Flooding System

In order to enhance the ability of severe accident mitigation for Pressurised Water Reactor (PWR), different kinds of severe accident mitigation strategies have been proposed. In-Vessel Retention (IVR) is one of the important severe accident management means by External Reactor Vessel Cooling. Reactor cavity would be submerged to cool the molten corium when a severe accident happens. The success criterion of IVR strategy is that the heat flux which transfers from the corium pool must be lower than the local critical heat flux (CHF) of the reactor pressure vessel (RPV) outside wall and the residual thickness of the RPV wall can maintain the integrity. The residual thickness of RPV is determined by the heat flux transfer from the corium pool and the cooling capability of outer wall of the RPV. There are various factors which would influence the CHF and the cooling capability of outer wall of the RPV. In order to verify the optimized design which is beneficial to the heat transfer and the natural circulation outside the actual reactor vessel, a large-scale Reactor Vessel External Cooling Test (REVECT) facility has been built. A large number of sensitivity tests were carried out, to study how these sensitivity factors affect CHF value and natural circulation. Based on the test results, the structure of the test section flow channel has an obvious effect on the CHF distribution. The flow channel optimized can effectively enhance the CHF value, especially to enhance the CHF value near the “heat focus” region of the molten pool. The water level in the reactor pit has also a great impact on the natural circulation flow. Although natural circulation can be maintained with a low water level, it will lead to a decrease of the cooling capacity. Meanwhile, some noteworthy test phenomena have been found, which are also essential for the design of the reactor pit flooding system.

Modeling of the corium crust of a stratified corium pool during severe accidents in light water reactors

During Severe Accidents (SAs) in Light Water Reactors (LWRs), the core heat up due to the lack of fuel cooling in the Reactor Pressure Vessel (RPV) may lead to the formation of a so called corium pool (molten core oxidic and metallic liquid materials) relocating in the lower head of the RPV. Within the framework of SA, one strategy to minimize risks of containment failure is the In Vessel Retention (IVR) strategy. To demonstrate the feasibility of such a strategy and to ensure the vessel integrity during the accident, one of the key points is to correctly evaluate the thermal load applied on the RPV wall by the lower head corium pool. Among all the complex coupled transient phenomena taking place in the RPV, the possible solidification of the corium pool, forming the so-called corium crust, has been shown by experimental and numerical results to be a key phenomenon for the IVR strategy. However, in most SA codes, the corium crust is only taken into account as a stationary model playing the role of a thermal resistance between the corium pool and the RPV wall. In this paper, a transient corium crust model is presented which explicitly takes into account the mass, temperature, composition and thermodynamic properties of the crust and pool material. Numerical results, in the context of the IVR strategy, highlight the impact of the transient corium crust modeling in comparison to a stationary modeling. The proposed model opens new perspective of modeling such as the dissolution of the crust and/or its thermo-mechanical behavior.

Numerical Study on the Corium Pool Heat Transfer With OpenFOAM

In order to ensure the successful implementation of in-vessel retention strategy to terminate molten corium pool evolution, it is necessary to evaluate the heat flux distribution on the lower head wall of reactor pressure vessel. In the present study, the models of internal heating and natural convention buoyancy, as well as the models of WMLES turbulence and phase changing were applied in the open source CFD software OpenFOAM to perform numerical simulations for the COPRA single-layer molten pool experiment. The distributions of temperature, heat flux and crust thickness were obtained. The simulation results were in good comparison with COPRA experimental data, proving the validity of the developed model for corium pool heat transfer characteristics. The simulation method and results could be applied to further in-depth study of thermal behavior in the corium pool.

Effect of metal honeycomb structure on enhancing CHF in saturated downward-facing flow boiling

There is a strong need to find a feasible approach to increase the Critical Heat Flux in downward-facing flow boiling effectively for the In-Vessel Retention strategy in advanced nuclear power plants. In this study, we designed metal porous honeycomb plates with different structures and used them in downward-facing flow boiling experiment to find a honeycomb plate that can enhance CHF in the downward-facing flow boiling to the largest extent and investigate the enhancement mechanism. The effect of the honeycomb structure and water supply through porous media on CHF are investigated. It is found that the reduction in boiling area has big effect on CHF and water supply through the porous media has significant effect on increasing CHF for some kinds of honeycomb structures.

Creep Analysis of Reactor Pressure Vessel Lower Head under Core Meltdown Severe Accident

During the core meltdown severe accident in the nuclear power plants, the strategy of In-Vessel Retention (IVR) to contain the melt in the Reactor Pressure Vessel (RPV) is a key mitigation measure. During the implementation of IVR strategy, the RPV lower head is likely to fail due to excessive creep deformation under the combined action of extremely high temperature loads and mechanical loads. Therefore, it is necessary to perform the analysis of the creep deformation of the RPV lower head, to ensure the structural integrity of the RPV under the condition of melt retention. In this paper, under the assumption of IVR, the finite element method is used to perform the thermal-structural coupling analysis of the RPV lower head. The temperature and stress fields of the vessel wall, and the plasticity and creep deformation of the lower head are calculated. Combining the plasticity and creep rupture criteria, the failure was analyzed. Results show that the deformation of the structure will be greatly increased when creep is considered. During the IVR strategy under severe accident, the main failure mode of the RPV lower head is creep failure instead of plastic failure. Further analysis implies that the internal pressure has a significant influence on the creep deformation and failure time. This paper provides the method for the creep and failure analysis of the RPV lower head under severe accident.

Assessment of different turbulence models on the large scale internal heated water pool natural convection simulation

In vessel retention (IVR) is treated as one of the strategies in the severe accident management aspect for the safety of nuclear reactor. The investigation of internal heated melt pool natural convection behavior is important for providing the reference for the IVR strategy. Various researchers conducted internal heated melt pool simulation with focusing on the different aspects and conditions. Based on the previous research, the large scale internal heated melt pool flow behavior located in the turbulence flow region and turbulence models used could affect the simulation result of the internal heated melt pool. Therefore, it is important to conduct the simulation of comprehensive assessment of different turbulence models on the melt pool simulation. In this work, the simulation of large scale internal heated water pool, which is one of the COPRA (COrium Pool Research Apparatus) water tests, is conducted by using ANSYS FLUENT 18.2. The large scale internal heated water pool transient behavior is investigated. The effect of various turbulence models on the internal heated water pool simulation is investigated. This work could provide better understanding of temperature distribution and velocity distribution of large scale internal heated water pool natural convection behavior from the numerical aspect. Based on the current simulation, Realizable k-epsilon turbulence model with the consideration of full buoyancy effects, RNG k-epsilon turbulence model with the consideration of full buoyancy effects, DES (Spalart-Allmaras) turbulence model and DES (SST k-omega) turbulence model are recommended for the simulation of large scale internal heated water pool. This work could bring the benefits for the further study on the numerical investigation of large scale internal heated water pool natural convection behavior.