Reactor Pressure Vessel Lower Head Research Articles

A scoping study on remelting process of a debris bed in the lower head of reactor pressure vessel

Coolability and retention of core melt (corium) in the lower head of reactor pressure vessel (RPV) has been accepted as a severe accident management strategy to maintain the reactor pressure vessel (RPV) integrity of a light water reactor. To qualify the in-vessel melt retention strategy, lots of studies have been performed to investigate the natural convection heat transfer of a melt pool in the lower head. However, little work has been attributed to the precursory phase of the melt pool, i.e., the remelting process of a debris bed which is formed in the lower head at the very beginning of corium relocation from the core the lower head. The present study is motivated to conduct an experimental study on the debris remelting process. For this purpose, a dedicated test facility named COREM (COrium REMelting) is conceived and constructed, which features internal heating of electromagnetic induction and visualization of debris remelting dynamics. The test section is a semicircular vessel representing a slice of scaled-down RPV lower head, whose front and back walls are made of transparent tempered glass which facilitate visualization and induction heating of the debris bed. Fiber probes with multiple optical temperature sensors are mounted in the semicircular wall of the test section to measure its temperature distribution. In the scoping test, n-octanol and Wood's metal are selected as the simulant materials of metallic and oxidic components of corium, respectively, and their particles ware loaded in the test section to form a debris bed. This paper presents the first two scoping tests which have been performed with the COREM facility so far, under different mixing ratios of two debris materials. The measured data include the photography of debris remelting processes and the temperatures in the debris bed and the semicircular wall. Based on the temperature distributions, heat flux profile along the semicircular vessel wall is also estimated. The scoping tests well reproduce the dynamic process of debris remelting, with two distinct stages of fusion of n-octanol and Wood’s metal, i.e., melting successively from low to high melting-point debris particles. During the remelting process, the vessel wall temperatures increase with the polar angle firstly and then decrease gradually, leading to the highest temperature appearing in the middle of the lower layer of the stratified molten pool which is finally formed.

The Experimental and Simulation Results of LIVE-J2 Test—Investigation on Heat Transfer in a Solid–Liquid Mixture Pool

Since the reactor pressure vessel (RPV) lower head failure determines the subsequent ex-vessel accident progression, it is a key issue to understanding the accident progression of the Fukushima Daiichi Nuclear Power Station (1F). The RPV failure is largely affected by thermal loads on the vessel wall, and thus, it is inevitable that the thermal behavior of the molten metallic pool with the co-existence of solid oxide fuel debris must be understood. In past decades, numerous experiments have been conducted to investigate homogeneous molten pool behavior. Few experiments, however, address the melting and heat transfer process of the debris bed consisting of materials with different melting temperatures. The LIVE-J2 experiment aims to provide experimental data on a solid-liquid mixture pool in a simulated RPV lower head under various conditions. The experiment was performed in the LIVE-3D facility at the Karlsruhe Institute of Technology. The LIVE-J2 experiment started from the end state of the previous LIVE-J1 experiment where a eutectic binary mixture of KNO3-NaNO3 (nitrate) was solidified and filled the gap of the ceramic beads inside the LIVE-vessel. The information obtained in the LIVE-J2 experiment includes transient and steady-state melting temperature and vessel wall temperature distributions. The extensive measurements of the melting temperature indicate the heat transfer regimes in a solid-liquid mixture pool. The test results showed that the conductive heat transfer is dominant during steady state along the vessel wall boundary and that convective heat transfer takes place inside the mixture pool. After the addition of liquid nitrate on top of the mixture pool, different behavior was observed in each layer. In the upper pure-liquid nitrate layer, convective heat transfer was well developed, resulting in a homogeneous temperature, while within the lower solid/liquid debris mixture zone a large temperature gradient was observed, suggesting that conductive heat transfer was dominant. Besides the experimental performance, the test case was numerically simulated using Ansys Fluent. The simulation results generally agree with the measured experimental data. The flow regime and transient melt evolution were able to be estimated by the calculated velocity field and the crust thickness, respectively.

Multi-dimensional finite element analyses of OECD lower head failure tests

For severe accident assessment of reactor pressure vessel (RPV), it is important to develop an accurate model that can predict transient thermo-mechanical behavior of the RPV lower head under the given condition. The present study revisits the lower head failure with two- and three-dimensional finite element models. In particular, we aim to give clear insight regarding the effect of the three-dimensionality present in the distribution of the thickness and thermal load of the lower head. For a rigorous validation of the result, both the OLHF-1 and the OLHF-2 tests are considered in this study. The result suggests that the three-dimensional effect is not negligible as far as the failure location is concerned. The non-uniformity of the thickness distribution is found to affect the failure location and time. The thermal load, which may not be axisymmetric in general, has the most significant effect on the failure assessment. We also observe that the creep property can affect the global deformation of the lower head, depending on the applied mechanical load.

An investigation of characteristics of flow and heat transfer in the two-layer molten pool under swing motion condition

In-Vessel Retention (IVR) of molten corium is an important strategy for management of severe accidents adopted by some active service generation II plus reactors and advanced generation III reactors like AP1000, Hualong one and APR1400. One of the most important factors contributing to IVR analysis can be attributed to the characteristics of heat transfer in the molten pool in the lower head of reactor pressure vessel (RPV), which directly determines the heat load imposed on the vessel wall of the RPV lower head. Besides, IVR is also an important strategy for management of severe accidents of Ocean Floating Reactor (OFR), whose flow and heat transfer characteristics are influenced by the ocean motion conditions. However, since all previous molten pool experiments were conducted under static condition, characteristics of the heat transfer in the molten pool thus obtained may be different from those under ocean motion conditions, thus not applicable to the latter. To address this issue, in this study, an experimental system was established, which includes a two-layer molten pool test facility and a swing motion platform. Based upon the system, the effects of the swing angle and swing period on the characteristics of heat transfer in the two-layer melt pool were investigated. In addition, a scaling method was proposed to select simulant materials for the two-layer corium pool. Results of the experiment show that the swing period and angle have a significant effect on the temperature field of the molten pool and the rate of heat transfer. We thus suggest that, in future studies, new heat transfer correlations, models and IVR analysis method for ocean floating reactors should be researched and developed to evaluate the IVR design of ocean floating reactors, which is critical for management of severe accidents of the ocean floating reactors.

Vessel Failure Analysis of a Boiling Water Reactor During a Severe Accident

In a postulated severe accident, the thermo-mechanical loads from the corium debris that has relocated to the lower head of the reactor pressure vessel (RPV) can pose a credible threat to the RPV’s structural integrity. In case of a vessel breach, it is vital to predict the mode and timing of the vessel failure. This affects the ex-vessel accident progression and plays a critical role in the development of mitigation strategies. We propose a methodology to assess RPV failure based on MELCOR and ANSYS Mechanical APDL simulations. A Nordic-type boiling water reactor (BWR) is considered with two severe accident scenarios: i) SBO (Station Blackout) and ii) SBO + LOCA (Loss of Coolant Accident). In addition, the approach considers the dynamic ablation of the vessel wall due to a high-temperature debris bed with the use of the element kill function in ANSYS. The results indicate that the stress failure mechanism is the major cause of the RPV failure, compared to the strain failure mechanism. Moreover, the axial normal stress and circumferential normal stress make the dominant contributions to the equivalent stress σ at the lower head of RPVs. As expected, the region with high ablation is most likely the failure location in both SBO and SBO + LOCA. In addition, comparisons of the failure mode and timing between SBO and SBO + LOCA are described in detail. A short discussion on RPV failure between ANSYS and MELCOR is also presented.

Experimental research of heat transfer in heavy metallic layer under different top boundary conditions

The External Reactor Vessel Cooling (ERVC) is crucial to realizing In-Vessel Retention (IVR) strategy when reactor core degradation happens; consequently, fission products would be limited in Reactor Pressure Vessel (RPV). Because of the fission products in the lower head of RPV, vessel integrity is threatened. Due to the immiscibility of oxide and metal materials, the molten pool in the lower head of RPV would turn to a stratified configuration. Two-layer (light metallic layer atop of oxidic layer) and three-layer (light metallic layer, oxidic layer, heavy metallic layer) configurations have been proposed. The possible heavy metallic layer has a different shape, top boundary conditions, and decay heat compared with the other two layers. Thereby, the objective of the heavy metallic layer experiment is to study its heat transfer characteristics, and two possible top boundary conditions are performed. The heating cables simulating the decay heat are placed in a hemispheric test section (diameter = 2.4 m). Water is used as the experimental simulant with a height of 0.3 m. The experimental results demonstrate that increasing the top cooling rate, the ratio of top cooling power to sidewall cooling power, and the ratio of upward heat flux to sideward heat flux increase. In addition, they share almost the same trend. The central melt temperature rises with the height increment, and the maximum normalized temperature reaches about 1.3 for both top cooling and insulated boundary conditions. Thermal stratification is found in the experiment, and the possible natural convection is limited in the heavy metallic layer. Correlations of UCLA, Mini-ACOPO, and RASPLAV reasonably predict the heavy metallic layer's normalized sideward heat flux distribution.

Preliminary Analysis of an Aged RPV Subjected to Station Blackout

Today, 46% of operating Nuclear Power Plants (NPP) have a lifetime between 31 and 40 years, while 19% have been in operation for more than 40 years. Long Term Operation (LTO) is an urgent requirement for all of the nuclear industry. The aim of this study is to assess the performance of a reactor pressure vessel (RPV) subjected to a station blackout (SBO) event. Alterations suffered by the material properties and creep at elevated temperatures are considered. In this study, coupling between MELCOR and Finite Element Method (FEM) codes is carried out. In the Finite Element (FE) model, the combined effects of ageing and creep are implemented through degraded material properties and a viscoplastic model. The reliability of the model is validated by comparing the FOREVER/C1 experimental results. The results show that the RPV lower head bends downwards with a maximum radial expansion of about 260 mm and RPV thermomechanical properties are reduced by more than 50% at high temperatures. The effects of ageing, creep and long heat-up strongly affect the resistance of the RPV system until the point of compromising it in the absence of/delayed emergency intervention. Aged RPV at end-of-life may collapse earlier, and in less time, with the same accidental conditions.

Estimation of the fuel debris thermal energy at the time of the major core slumping of Fukushima Daiichi Nuclear Power Plant Unit-3 with MELCOR-2.2

To provide supportive information to understand the current debris status in Fukushima Daiichi Nuclear Power Plant Unit-3, sensitivity analyses have been carried out with MELCOR-2.2 for two scenarios, with/without direct leakage from the Reactor Pressure Vessel (RPV) to the Drywell (D/W). The particular focus of the analyses is the estimated debris thermal energy up to and at the time of the major core slumping event, which may be the key to determine the following debris cooling and failure mechanism of the lower head of the RPV. In the analysis, the debris relocation velocity from the core region to the RPV lower head (VFALL), the number of Safety Relief Valves (SRVs) opening, and the amount of core slumping at the major RPV pressure peak were tuned so that the plant data of RPV and PCV pressure from ca. 9:00 to 12:00 March 13th 2011could be reproduced in each scenario. Best-estimate case conditions are summarized for the with/without direct leakage from RPV to D/W scenarios. As a result, as a consideration common to both scenarios, MELCOR analysis tends to estimate significant core oxidation before Automatic Depressurization System (ADS) actuation, and concomitantly estimates little core oxidation from the time after ADS to the time of the major RPV pressure peak. In addition, although a remarkable difference was found in the amount of hydrogen generated in RPV during the major core slumping between the best-estimate with/without leakage cases, limited difference can be observed for the core oxidation heat generation during the major core slumping. This is because the early Zircaloy oxidation prior to ADS actuation in the current MELCOR modelling resulted that the hydrogen generation source during the major core slumping at 12:00 March 13th was primarily from oxidation of stainless steel, which was not as exothermic as that of Zr oxidation.

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.

Experimental investigation on saturated pool boiling CHF for downward facing heating surface with different sizes and aspect ratio

IVR-ERVC (In-Vessel Retention – External Reactor Vessel Cooling) is a key severe accident management for keeping the integrity of the reactor pressure vessel (RPV) lower head. One of most important topics is the characteristics of the CHF (Critical Heat Flux) on the outside wall of RPV lower head. Some experiments had been conducted to investigate the effect of the orientation on characteristics of the boiling heat transfer and CHF for downward facing heating plate. While the size of the heating plate used may have an appreciable effect on the characteristics of the CHF. In this work, experiments were carried out to investigate the effects of the heater surface area and length-width ratios on saturated pool boiling and CHF for downward facing heating surface. Experimental results indicated that the inclined angle has little effect on the saturated pool boiling heat transfer, and the wall superheat of triggering of CHF was greater for the larger inclined angle. For the same aspect ratio of the heating surface, the CHF decreased with the increase of the heating surface area for the same surface orientation. It also indicated that the CHF was influenced by the bubble escaping length and wetted perimeter complicatedly, and the short edge effect had an appreciable influence on CHF based on present experimental work. Meanwhile, a modified El-Genk and Guo CHF correlation was proposed based on CHF data with the same aspect ratio of the heating surface, which could consider the effect of heating surface area on pool boiling CHF.

Analysis of hemispherical vessel ablation failure involving natural convection by MPS method with corrective matrix

In a severe accident of a light water reactor, the reactor pressure vessel (RPV) lower head may fail due to ablation at the vessel wall boundary involving natural convection of molten core materials. Accurate prediction of RPV lower head failure is essential for assessing severe accident progression and improving accident management because it greatly influences the subsequent ex-vessel accident progressions. However, there have been still large uncertainties about RPV lower head failure mode in the Fukushima Daiichi Nuclear Accident in 2011. The Lagrangian based MPS (moving particle semi-implicit) method has advantage of analyzing such phenomena involving complex interfaces and liquid-solid phase changes over other Eulerian mesh-based method. In the preceding study, small-scale Pb–Bi hemisphere vessel ablation experiment, with silicone oil as simulated molten core, was reproduced qualitatively by original MPS method. However, ablation mechanism associated with natural convection of the high temperature liquid could not be discussed because of significant influence of numerical discretizing error. In this study, the improved MPS method coupling corrective matrix in the particle interaction model which largely suppress the numerical fluctuation was adopted to analyze the experiment. The results show that the ablated metal relocation may enhance convective heat transfer in the downstream. As a result, ablation of the vessel wall extends from the level, close to the silicone oil surface down to the bottom of the vessel rather than previously simulated localized ablation near the silicone oil surface.

In- and ex-vessel coupled analysis for in-vessel retention

In-vessel retention (IVR) in the manner of external reactor vessel cooling (ERVC) is an important severe accident mitigation strategy, which has been applied to some advanced light water reactors, e.g. AP1000. This main assessment method on the effectiveness of IVR is the lumped parameter (LP) model, which has been applied to assess the safety margin of IVR-ERVC for decades. However, this model deals with the safety margin analysis with several isolated processes, without considering the strong coupled relations about the heat transfer processes in the internal and external reactor pressure vessel (RPV). This paper addresses this coupled issue based on modified lumped parameter model using an iterative method, which focuses on the relations between the inner wall surface temperature of the reactor pressure vessel lower head and the heat flux through the RPV inner wall with a modified heat transfer method. The method has been applied to analyze the molten pool behaviour of LIVE-L4 tests. Comparison between the experimental data and the calculation results is performed to validate the accuracy of the coupled analysis method. High attention is paid to some important parameters, e.g. the heat flux through vessel, the inner and outer wall surface temperatures of the RVP lower head and the crust thickness. In addition, the paper analyzes the effectiveness of IVR for AP600 based on the modified method and compares with the UCSB assumed Final Bounding State (FIBS) benchmark calculation results to verify the modified coupled method. This method is applied to conduct the sensitivity analysis for some design parameters of AP600.

Ablation and thermal stress analysis of RPV vessel under heating by core melt

After a reactor core melts and collapses suddenly, the melting core accumulates on the base surface of the bottom head of a reactor pressure vessel (RPV), causing severe thermal ablation and thermal stress, endangering the safety of the RPV bottom head. In this study, a 1000-MW pressurized water reactor is considered an example to study the heat transfer ablation and thermal stress of an RPV lower head after a core collapses, by performing numerical simulations. A two-dimensional (2D) heat transfer model is used to analyze the coupling heat transfer between the wall surface of the RPV, two-layer melting corium pool, and outer water chamber. The transient 2D temperature and ablation of the lower head wall surface are calculated. The thermal stress of the RPV lower head and deformation are also investigated using ANSYS software. The results show that (1) The upper crust is the least thick, with a thickness of approximately 0.01 m, whereas the side crust is the thickest at approximately 0.12 m at the base of the lower head. (2) The minimum thickness of the bottom wall decreases linearly with time, starting from the collapse time of the core to 2500 s, when it becomes 0.04 m. It does not change thereafter, but the melting zone is further expanded. (3) The lower head wall starts to melt from 200 s after the core collapses. The melting mass first increases sharply, and subsequently, increases slightly with time. The total melted material is 3000 kg at 5000 s. (4) The heat flux at the inner and outer surfaces of the lower head immersed in the uranium melt layer increases with the azimuth angle, reaching a maximum value of 750 kw/m2 and 250 kw/m2, respectively, at the interface with the metal melt. The heat flux at the RPV inner wall covered by the metal melt is approximately constant at 400 kw/m2, whereas that at the outer surface decreases with the azimuth angle. (5) The stress at the lower head is concentrated at the inner surface, with a maximum value of 625.65 MPa. The radial deformation increases with time only until 2200 s, with a maximum deformation of 28.39 mm occurring in the lower part of the RPV bottom.

Features of heat and deformation behavior of a VVER-600 reactor pressure vessel under conditions of inverse stratification of corium pool and worsened external vessel cooling during the severe accident. Part 2. Creep deformation and failure of the reactor pressure vessel

In the present manuscript the features of deformation of medium-power VVER-600 reactor pressure vessel (RPV) during a hypothetical severe accident (SA) in the case of worsened regimes of external vessel cooling are considered and analyzed. The analysis of thermal and deformation behavior of the RPV was performed for thermal loads on the reactor vessel that were defined before. The thermal loads acting on the reactor vessel from the corium melt were defined for two types of corium melt structures. A two-layer stratified melt, when a less dense layer of molten steel is located over the oxide heat-generating phase of the corium, was considered as the first structure of the molten pool. The inverse three-layer corium melt was considered as the second type of the melt structure. The numerical simulation of heating and creep deformation processes of the reactor vessel was performed with allowance for creep effect and failure of the RPV while varying both the value of in-vessel overpressure (from 0.2 to 0.8 MPa) and the conditions of external cooling of the RPV in the SA. The external cooling of the RPV were simulated by giving corresponding values of heat transfer coefficients (HTC) on the external surface of the vessel wall. The values of HTCs on the external surfaces of both cylindrical part of the vessel and the vessel bottom varied from 350 to 900 W/(m2K). The performed analysis resulted in the fact that under the conditions of worsened external cooling of the RPV in the SA, significant deformations of the RPV structure are observed. Particularly, during the SA the vertical displacement of the RPV lower head (LH) until the moment of its failure may attain 500 mm and more. Such considerable creep deformations of the reactor vessel structure are observed in the case of forming the inverse structure of the corium pool. Here, the LH deformations make up the bulk of total RPV deformation. Such a model of the vessel deformation may lead to partial or complete blockage of cooling gaps used for external cooling of the reactor vessel in the SA. In turn, the disturbance of the regime of external reactor vessel cooling may cause its overheating and premature RPV failure. The dependences of RPV failure time and the value of vertical displacement of the RPV structure on the value of in-vessel overpressure were obtained for the considered group of SA scenarios.

Physical model features and validation status of three-dimensional simulation model for BWR in-vessel core degradation

After the severe accident at Fukushima Daiichi Units 1–3 in 2011, Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R) initiated the development of a Boiling Water Reactor (BWR) in-vessel detailed core degradation model (called “Multifunction Model”) for decreasing computational uncertainties of severe accidents as a fundamental research activity. The objective of this study is to present the validation status of the Model against in-vessel core degradation phases, focusing on the description of important physical models and validation scheme. The Model focuses on in-vessel core degradation from cladding temperature rise to molten material flow to a lower part of reactor vessel, since there are a multitude of large uncertainties in simulating in-vessel core degradation. To simulate in-vessel core degradation more realistically, physical models for all the dominant phenomena, which are (a) multi-phase, multi-component and multi-velocity field, (b) three-dimensional configuration of internal components in a Reactor Pressure Vessel (RPV), (c) multiple melting points and chemical interactions, (d) candling phenomenon, (e) melt blockage on spacers, and (f) molten corium interaction with water pool in an RPV lower head, have been developed at S/NRA/R. The Model of in-vessel core degradation phenomena was assessed against the DF-4 experiment at Sandia National Laboratories (SNL) and the QUNECH-06 experiment at Karlsruhe Institute of Technology (KIT) in terms of estimating the cladding temperature escalation and hydrogen generation, and the CORA-18 experiment at KIT in terms of estimating axial- and lateral-directional motion of molten corium. Additionally, the Model of metallic melt relocation around a BWR core support plate, and molten corium breakup in the lower head was validated by the XR-2 at SNL, and the FARO L-19 and the KROTOS K-37 at Joint Research Centre (JRC) facilities. The Model demonstrated a reasonable capability to simulate the main features of in-vessel core degradation phases from a core zone to a lower head zone. In the next stage, sensitivity analyses of the Model will be carried out for future practical application from 2016 to 2017.

Evaluation of VVER-1200/V-491 Reactor Pressure Vessel integrity during large break LOCA along with SBO using MELCOR 1.8.6

After Fukushima accident and stress test recommended by IAEA for existing reactors, higher safety requirements are enforced upon nuclear power plants during design extension and severe accident conditions. Based on those arguments, Vietnam Government requests a lot of effective safety solutions, in designs proposed for the nuclear power plants in Ninh Thuan province of Vietnam, which can prevent the accident progression toward severe accidents and mitigate severe accident consequences. One of safety requirements is related to delay time of core melt during design extension condition. Especially, if the worst case of accidents occurs, the reactor vessel integrity must be maintained at least 24 hours from the beginning of the accident. With the aim at investigation of Reactor Pressure Vessel (RPV) integrity, in this study, MELCOR 1.8.6 code is used to evaluate the integrity of RPV lower head for VVER-1200/V-491 reactor during a Large Break Loss of Coolant Accident (LBLOCA) in combination with Station Blackout (SBO) event. The study figures out several parameters related to melt down progress such as: rupture position and rupture timing, the amount of hydrogen generated. Availability of the second stage hydro-accumulators (HA2) in the VVER-1200/V-491 is assumed as an additional improvement to delay the timing of core melt as well as to maintain the vessel integrity for long-term.

A MELCOR model of Fukushima Daiichi Unit 1 accident

A MELCOR model of Fukushima Daiichi Unit 1 accident was developed. The model is based on publicly available information, and the MELCOR input file is published as Electronic Supplementary data with this paper. In order to reproduce the measured containment pressure, it was necessary to model a leak from the reactor coolant system. Recirculation pump seal leak, starting 5h after the earthquake, was assumed in this study. The reactor pressure vessel lower head was calculated to fail, and all fuel was discharged to the containment. Almost all of the radioactive noble gases and about 0.05% of the cesium inventory were released to the environment, according to this calculation. Calculated release rates from Units 1, 2, and 3 were compared with measured radiation dose rate in the plant area.

A MELCOR model of Fukushima Daiichi Unit 3 accident

A MELCOR model of the Fukushima Daiichi Unit 3 accident was developed. The model is based on publicly available information, and the MELCOR input file is published as electronic supplementary data with this paper. According to the calculation, the reactor pressure vessel lower head failed about 53h after the earthquake. At the end of the calculation, 30% of the fuel was still inside the reactor and 70% had been discharged to the containment. Almost all of the radioactive noble gases and 0.95% of the cesium inventory were released to the environment during the accident.

Analysis of PWR RPV lower head SBLOCA scenarios with the failure of high-pressure injection system using MAAP5

Through the use of MAAP5 code we carried out an analysis of the severe accident scenarios initiated by an unusual small-break loss-of-coolant accident (SBLOCA) located at the reactor pressure vessel (RPV) lower head with the failure of high-pressure injection system (HPIS) in a PWR power plant. The cases of 0.4-inch-break LOCA with and without manual depressurization of the reactor coolant system (RCS), and 1.0-inch-break LOCA were studied. The results showed that without manually depressurizing the RCS, reactor core underwent slow heatup and completely melted and eventually failed the RPV lower head at the primary system pressure of 6.87 MPa for 0.4-inch-break LOCA and 2.71 MPa for 1.0-inch-break LOCA. There was a strong interaction between the discharged melt and collected pool water in the reactor cavity at the time of RPV lower head failure by creep. A 0.472 MPa containment peak pressure was caused by hydrogen combustion for 0.4-inch-break LOCA. For 1.0-inch-break LOCA the pressure peaked twice, the first one (0.396 MPa) by intense steam generation and the second one (0.331 MPa) by hydrogen combustion. These pressure spikes were lower than the containment design pressure of 0.52 MPa. It was not likely to stop the cavity from being eroded by molten corium through continuously pouring water down into the cavity. Thus there were constant reactor cavity ablation and hydrogen generation, leading to dangerous erosion depths and elevated hydrogen volume fraction in the presence of containment spray system. On the other hand, timely manually opening the pressurizer (PZR) safety valves (SVs) was an effective mitigation measure to recover the core coolabilty by cold-leg accumulator injection system (AIS) and low-pressure injection system (LPIS). Besides, reasonably manually opening the steam generators (SGs) SVs while keeping the auxiliary feedwater on also helped to depressurize the RCS and prevent the severe accident. Both of the two mitigation measures successfully prevented the core from complete melt, but the latter one is preferable to the former one provided no steam generator tube rupture takes place.

LIVE Experiments on Melt Behavior in the Reactor Pressure Vessel Lower Head

The main objective of the LIVE program at Karlsruhe Institute of Technology is to study the core melt phenomena both experimentally in large-scale three-dimensional (3D) geometry and in supporting separate-effects tests, and analytically using computational fluid dynamics (CFD) codes in order to provide a reasonable estimation of the remaining uncertainty band under the aspect of safety assessment. Within the LIVE experimental program several tests have been performed with water and with noneutectic and eutectic melts (mixtures of KNO3 and NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used to assess the correlations derived for the molten pool behavior.