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Transient 3D simulation for heating and melting process of PWR core after SBO
Abstract
In the heat transfer between the core and coolant in a reactor pressure vessel (RPV), a bounding structure such as a shroud has an important influence on the heating and melting process of the core. To study the 3D transient heat transfer and melting process of the Hualong 1 PWR using an IVR strategy, a numerical analysis model has been established based on the 2D transient heat conduction of each rod, which is coupled considering the residual heat of the core, the convective and radiation heat transfer inside the core, and natural convection heat transfer of the outer water chamber. A transient 3D numerical simulation code, 3DTMCOR, has been developed to analyze the melting process of an IVR reactor core with high precision. The effects of heat generation from a zirconium–water reaction, radiation heat transfer, and the adoption of an IVR strategy on the heating and melting process of the reactor core have been investigated. It was found that the first melting time for the same 100 MW PWR from the 3DTMCOR code is about 2500 s later than that of MAAP. The core temperature will first decrease with a decrease in core power within 100 s, then increase as the water level drops down with a zirconium–water reaction of 100 s to 1000 s, then decrease again with a decrease in core power from 1000 s to 4300 s, and increase a second time when water in the RPV is evaporated after 4300 s. At 4860 s, a continuous melting area first appears in the core top. Radiation heat exchange is very important for transferring heat from the fuel rods to structures such as the control rods and bounding structures, which is also the main cooling mode in the core after water is evaporated. The first melting time with the precise radiation exchange model will occur 200 s later than in the results with the empirical model. The water chamber outside the RPV does not have a remarkable effect on the heating and melting process of the core.
... In most cases, pressurized water is used as the coolant in these types of reactors, and the classical convective heat transfer correlations can be applied for calculations [2]. However, the literature review results [2][3][4][5][6][7][8][9][10][11][12][13][14][15][16] show that there is increased interest in nanofluid application for heat transfer enhancement in PWRs. Section 3 reports a review of convective heat transfer correlations valid for BWRs. ...
... Zhang et al. [9] developed a theoretical model (see Equations (6) and (7) in Table 1) to simulate the 3D transient heating and melting progress of the reactor core following a station blackout accident, based on a 2D heat transfer analysis of each rod. This study examined the Hualong 1 reactor, a 100 MW PWR that utilizes an in-vessel retention (IVR) design. ...
... The resulting equations were derived for a smooth tube with a diameter of 26 mm and a length of 2 m, inclined at 20 degrees to the horizontal. Nux and Rex are evaluated at temperature tx: [7,[9][10][11] Heat transfer to water and steam at variable specific heat at near-critical region (turbulent, 1D, single-phase, experimental, stationary). [44] Heat transfer at supercritical region in flow of carbon dioxide and water in tubes, (turbulent, 1D, single-phase, experimental, stationary). ...
Nuclear reactors are very complex units in which many physical processes occur simultaneously. Efficient heat removal from the reactor core is the most important of these processes. Heat is removed from the reactor core via heat conduction, radiation, and convection. Thus, convective heat transfer and its conditions play a crucial role in the operation and safety of nuclear reactors. Convective heat transfer in nuclear reactors is a very complex process, which is dependent on many conditions and is usually described by different correlations which combine together the most important criteria numbers, such as the Nusselt, Reynolds, and Prandtl numbers. The applicability of different correlations is limited by the conditions of heat transfer in nuclear reactors. The selection of the proper correlation is very important from the reactor design accuracy and safety points of view. The objective of this novel review is to conduct a comprehensive analysis of the models and correlations which may be applied for convective heat transfer description and modeling in various types of nuclear reactors. The authors review the most important research papers related to convective heat transfer correlations which were obtained by experimental or numerical research and applied calculations and heat transfer modeling in nuclear reactors. Special focus is placed on pressurized water reactors (PWRs), boiling water reactors (BWRs), CANDU reactors, small modular reactors (SMRs), and molten salt reactors (MSRs). For each type of studied reactor, the correlations are grouped and presented in tables with their application ranges and limitations. The review results give insights into the main research directions related to convective heat transfer in nuclear reactors and set a compendium of the correlations that can be applied by engineers and scientists focused on heat transfer in nuclear reactors. Prospective research directions are also identified and suggested to address the ongoing challenges in the heat transfer modeling of present and next-generation nuclear reactors.
In this study, a hybrid model for 3D natural convection combined with surface thermal radiation in a closed differentially heated cube was developed. Within this model, numerical procedure for fluid flow was considered in terms of the lattice Boltzmann method under Bhatnagar-Gross-Krook approximation with D3Q19 scheme. On the other hand, unsteady three-dimensional energy equation was solved by means of the finite difference technique. Developed hybrid approach was validated against experimental and up to date numerical data. Mathematical modelling was conducted for two-dimensional and three-dimensional problem formulations under variation of Rayleigh number, conduction-radiation number and surface emissivity. It was found that discrepancy between 2D and 3D results was increased with an enhancement in the radiation heat transfer rate. Computational performance of hybrid lattice Boltzmann method was several times higher than conventional CFD approach.
The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence. The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core. Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective. This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants.
Following a core melt, molten material may slump to the lower head of a reactor pressure vessel (RPV). In that case, some structures like lower parts of fuel elements and a core support plate would remain intact. Since the melt is at high temperature and there are no obstacles between the melt and the supporting plate, the plate is exposed to an intense radiation heating. The radiation heat exchange model of the lower head was developed and applied to a finite element code COUPLE which is a part of the detailed mechanistic code RELAP5/SCDAPSIM. The radiation enclosure consisted of three surfaces: the upper surface of the relocated material, the inner surface of the RPV wall above the relocated material and the lower surface of the core support plate. View factors were calculated for the enclosure geometry that is changing in time because of intermittent accumulation of molten material. The enclosure surfaces were covered by mesh of polygonal areas and view factors were calculated, for each pair of the element areas, by solving the definite integrals using the algorithms for adaptive integrations by means of Gaussian quadrature. Algebraic equations for radiosity and irradiation vectors were solved by LU decomposition and the radiation model was explicitly coupled with the heat conduction model. The results show that there is a possibility of the core support plate failure after being heated up due to radiation heat exchange with the melt.
* The work represents the very first deterministic analysis of TMI2 accident of such an extent in the world.
MELCOR has become the preferred code of the Swiss nuclear industry and of PSI for severe accident analysis, on account of its integrated systems-level approach and validation against experiments and more detailed codes, while MACCS is commonly used by safety authorities for independent assessment of off-site consequences, in particular health effects. The present work arises out of a programme to assess MELCOR independently using empirical data consistent with the recommendations of the OECD/CSNI validation matrix for core degradation codes. The MELCOR 1.8.5RD calculations are based on a model for phases 1 and 2 provided by the code developers but with a simplified thermal hydraulic noding in certain regions and the inclusion of a simple representation of the fission product release and transport pathways. The model has also been extended to simulate phases 3, 4, and the continuing initial period of core recovery and stabilisation. These calculations are a first attempt to demonstrate a MELCOR–MACCS capability to simulate the whole plant accident sequence beyond phase 4, including the containment response and off-site consequences arising from fission product release from the containment. Emphasis is placed on the overall accident evolution and whole plant response, rather than the detailed behaviour. Results are compared with observed and deduced data for the major accident signatures and rough estimates for exposure based on off-site monitoring. The results provide a good basis for the NPP analysis foreseen.
The reflooding, i.e. injecting water to the uncovered degraded core, is the most important accident management measure to terminate a severe accident and to stop the core from being melt in LWR. The reflooding of LWR core during severe accident may lead to the core cooling and cessation or to the temperature escalation and further development of the accident. That will depend on several important parameters, characterizing the core state and the way of the reflooding. Appropriate understanding of the complex core reflooding phenomena is necessary for the prediction of the system evolution.
Based on the reflood flow and heat transfer characteristics, a thermo-hydraulic analysis model for the reflood phase was developed and presented in this paper. Based on this model, a code module was developed for the thermo-hydraulic analysis of reflood. The calculation results for the different conditions are compared with the FLECHT data and the QUENCH data, respectively. The results showed that the calculation results and the experiment data are in substantial agreement. Then the influences of the system pressure, the subcooling of coolant and the wall temperature on reflood characteristics were studied, too. The quench velocity falls when the wall temperature went up. The quench velocity increases when the system pressure increases. The quench velocity drops with the increasing of water temperature. The paper also provides a theoretical basis for safety analysis of fuel element cladding during reflood phase.
The Fukushima accident has drawn attention even more to the importance of external events and loss of energy supply on safety analysis. Since 2011, several Station Blackout (SBO) analyses have been done for all type of reactors. The most post-Fukushima studies analyze a pure and straight SBO transient, but the Fukushima accident was more complex than a standard SBO. At Fukushima accident, the SBO was a consequence of an external flooding from the tsunami and occurred 40 min after an emergency shutdown (SCRAM) caused by the earthquake. The first objective of this paper is to assume the consequences of an external flooding accident in a PWR site caused by a river flood, a dam break or a tsunami, where all the plant is damaged, not only the diesel generators. The second objective is to analyze possible actions to be performed in the time between the earthquake event (that causes a SCRAM) and the external flooding arrival, which could be applicable to accidents such as dam failures or river flooding in order to avoid more severe consequences, delay the core damage and improve the accident management. The results reveal how the actuation of the different systems and equipments affect the core damage time and how some actions could delay the core damage time enough to increase the possibility of AC power recovery.
Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding.
During a severe accident such as a LOCA, fuel rods are exposed to steam. Rods are heated by fission-product decay and by steam oxidation of the cladding. The energy balance on the rod includes heat losses by convection to the surrounding steam and by thermal radiation to the surrounding fuel rods and to the ultimate heat sink, the reactor pressure vessel. The present note analyzes the latter problem by simulating the fuel rods in the core as concentric thin shells that emit radiant energy. The result is an estimate of the configuration factor that appears in the conventional radiation heat-flux equation.
The model describing massive melt blockage (slug) relocation and physico-chemical interactions with steam and surrounding fuel rods of a bundle is developed on the base of the observations in the CORA tests. Mass exchange owing to slug oxidation and fuel rods dissolution is described by the previously developed 2D model for the molten pool oxidation. Heat fluxes in oxidising melt along with the oxidation heat effect at the melt relocation front are counterbalanced by the heat losses in the surrounding media and the fusion heat effect of the Zr claddings attacked by the melt. As a result, the slug relocation velocity is calculated from the heat flux matches at the melt propagation front (Stefan problem). A numerical module simulating the slug behaviour is developed by tight coupling of the heat and mass exchange modules. The new model demonstrates a reasonable capability to simulate the main features of the massive slug behaviour observed in the CORA-W1 test.
The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the risk oriented accident analysis methodology (ROAAM) and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow for the delineation of the failure boundaries. The technical treatment in this assessment includes: (a) new data on energy flow from either volumetrically heated pools or non-heated layers on top, boiling and critical heat flux in inverted, curved geometries, emissivity of molten (superheated) samples of steel, and chemical reactivity proof tests; (b) a simple but accurate mathematical formulation that allows prediction of thermal loads by means of convenient hand calculations; (c) a detailed model programmed on the computer to sample input parameters over the uncertainty ranges, and to produce probability distributions of thermal loads and margins for departure from nucleate boiling at each angular position on the lower head; and (d) detailed structural evaluations that demonstrate that departure from nucleate boiling is a necessary and sufficient criterion for failure. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is ‘physically unreasonable’ Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings.