Core Meltdown Research Articles

A comparative assessment of various single tray core catcher options for future sodium cooled fast reactors

Various single tray in-vessel core catcher options to handle whole core decay heat for a 1500 MWt capacity sodium cooled Indian fast reactor (SFR) have been analyzed towards their efficiency in rendering acceptable core catcher temperatures. For this purpose, a one dimensional transient multiphase heat transfer model that considers sodium boiling, fuel particles re-melting and slumping has been adopted. The boundary conditions for the one dimensional model are obtained from computational fluid dynamics (CFD) simulations of turbulent natural convective heat transfer within the lower sodium plenum of the SFR. It is found that the top surface of the debris bed dissipates more heat than the bottom surface of the core catcher. A single plate core catcher is seen to experience melting of the core catcher plate (CCP) for a short duration in case of CCP thickness greater than 80 mm. The heat shield plate (HSP) of the double plate core catcher is observed to undergo complete melting due to vapor locking between the HSP and CCP. The phenomena of melting and slumping of debris particles is seen in double plate core catchers with a refractory lining. However, safe temperature levels of HSP and CCP were observed in the case of double plate core catchers with refractory lining. This study suggests that a double plate core catcher with a delay bed is the potential option for future sodium cooled fast reactors, which are to be designed to handle whole core meltdown scenarios.

Mechanical and thermal properties of Zr-B and Fe-B alloys

ABSTRACT After the core meltdown in the Fukushima Daiichi (1F) nuclear power plant, various compounds in metal boride systems could potentially form as a result of the reactions between B4C (control material) and Fe, Zr alloys (control-blade sheaths, zircaloy claddings, channel box, etc.). Some previous studies have focused on the properties of intermetallic compounds of the Zr-B and Fe-B systems, such as ZrB2, FeB, and Fe2B. However, during the solidification of fuel debris, composites of these intermetallics rather than large chunks of single-phase intermetallics will form. This situation makes the understanding of the composition dependence of the mechanical and thermal properties of these metal borides a major task before debris retrieval takes place. In this study, we first investigated the temperature-dependent thermal conductivity of Fe-B and Zr-B eutectics from room temperature up to 900°C. We were then able to evaluate the composition-dependent indentation hardness of these metal borides at room temperature. Based on our experimental data, we concluded that the hardness and thermal conductivity of the Fe-B, Zr-B composites can be well estimated using the properties of the composites’ corresponding components, with Rule of Mixtures and Effective Medium Theory (EMT) calculations, respectively.

Optimization design method of PRHRS air heat exchanger based on natural circulation characteristics

Passive residual heat removal system (PRHRS) was the key equipment to ensure the inherent safety of the fourth generation of nuclear reactors, which can effectively reduce the probability of core meltdown in the event of an accident. A multi-physics coupling model of sodium cooled fast reactor PRHRS was established in this paper, which based on the natural circulation characteristics of PRHRS. The model was analytically calculated according to the flow balance and energy balance, and finally the model was dynamically balanced. The influence of the tube spacing of the serpentine finned tube and the opening area of the air door on the thermal behavior of the system was analyzed through PRHRS model, and the optimal design method of the air heat exchanger (AHX) of PRHRS was mastered. The research results provide a theoretical basis for the rational design of the air heat exchanger and the safe operation of the sodium cooled fast reactor PRHRS.

Design of core catchers for sodium cooled FBRs – Challenges

The whole core meltdown scenario in sodium cooled fast reactors is considered under design extension criteria and has a very low probability of occurrence. To mitigate such a hypothetical severe accident in fast breeder reactors, a core catcher has been provided to accommodate the core debris within the primary containment boundary. In sodium cooled fast breeder reactor (FBR), an in-vessel core catcher, is provided to receive and disperse the fuel debris arising out of core meltdown during Hypothetical Core Disruptive Accident (HCDA). The core catcher prevents hot debris from reaching and damaging the main vessel. It also enables adequate heat transfer by natural convection to keep debris in stable conditions. The present study focuses on design of core catcher with sacrificial barriers and techniques to improve natural cooling of debris on the core catcher for FBRs. Studies are carried out towards the selection of suitable sacrificial material for core catcher, numerical analysis on multilayer and thermal-hydraulic analysis on multi jets core catcher concepts for selection of best suitable technique for debris bed coolability. The core catcher concept with sacrificial barrier and cooling pipes is found to possess enhanced coolability of debris bed. Experimental studies on compatibility of sacrificial ceramic material with sodium indicate strong dependence on microstructure. It has also been found that the multilayer core catcher with MgO as delay bed of 50 mm thickness is adequate to reduce the bottom temperature of core catcher below the design safety limit. Judicious mix of all these findings is expected to have unique design to accommodate whole core meltdown in future fast reactors.

Investigation of core meltdown phenomena and radioactive materials release in VVER‐1000/V446 nuclear reactor at severe accident condition due to LBLOCA along SBO

In this study, VVER-1000/V446 nuclear reactor modeling using RELAP5/SCDAP3.4 code to investigate the reactor core behavior during severe accident conditions in a LBLOCA scenario along with station-black-out (SBO) is carried out. The analyses are performed in two stages, before and after the core heat up, without considering operator’s action on the accident management. Fuel assemblies in the core are grouped into five based on average power peaking factors and are modeled in SCDAP code. For each group a corresponding channel is modeled in RELAP5 code plus a bypass channel. In the first stage, the study of ECCS and KWU tanks efficiency to keep the reactor core in the safe condition, the calculation of the elapsed time before the reactor core heat up and the estimation of available time for operator’s action to avoid core degradation, are investigated. In the second part the results of Hydrogen production rate, cladding oxide thickness, cladding damage level, release of fission products into the coolant are studied. Analysis of the scenario by the code shows the production of around 350kg Hydrogen with the maximum rate of about 1kg/s and releasing a large amount of FPs in the order of 10kg. The results also demonstrate that the operators have ∼3 h before the fuel rod cladding rupture and ∼2.5 h before the inception of exothermic steam-zirconium reaction. Finally, using a geometric mesh for the lower plenum and applying the COUPLE code, the results show that the core slumping into the lower plenum and the lower plenum rupture occur at 17561 and 18370 seconds after the onset of accident, respectively.

Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

Sensitivity Analysis of Unsteady Corium Solidification in an Initially Emptied Horizontal Turbulent Pipe Flow

Abstract In a reactor core meltdown under postulated severe accidents, the molten material called corium could be ejected or relocated through existing vessel penetrations. There exists, however, a potential for plugging of melt flow due to its complete solidification providing the availability of an adequate heat sink. Simulations of the melt flow in a horizontal tube were carried out to conduct a sensitivity study on the effect of key parameters on the melt penetration distance and bulk temperature distribution of the corium. The Reynolds number was varied from 10,000 to 20,000, inlet temperature was varied from 2600 K to 3000 K, the corium thermal conductivity was varied from 10 W/m·K to 20 W/m·K, and the pipe diameter was varied from 0.0095 m to 0.019 m. In addition, a comparison was made with an analytical model based on a modified Epstein's model and a previous numerical model. The study provided insight into the lower bound, which was found to be 98 mm, and the upper bound was 258 mm when predicting the potential penetration length of corium in horizontal pipes.

Experimental research on heat transfer behavior of large scale two-layer salt melt pool based on COPRA facility

During a core meltdown accident, the significant decay heat within the corium pool will continually impose thermal load on the vessel wall and challenge the integrity of reactor pressure vessel (RPV). External Reactor Vessel Cooling (ERVC) is considered to be an effective strategy to reduce the potential of vessel failure. After long-term cooling, the initially uniform melt pool can reach a final steady state with stable two-layer or three-layer configuration. Based on the COPRA facility, a series of tests were performed to study the natural convection heat transfer phenomenon of stratified corium pools. The test section was a full-scale 2D slice, simulating the hemispherical lower head of the HPR1000 reactor design. Direct electrical heating was applied in the lower part of vessel to simulate the homogenously heated ceramic pool, while no heat was generated in the upper part to model the metallic layer. A binary mixture of 20 mol% NaNO3–80 mol% KNO3 was selected as simulant material due to the similar phase-change characteristics with the prototypical melt. During the experiment, effects of heating power, top boundary condition, corium pool height, decay heat distribution and simulant material on corium pool heat transfer behavior were analyzed. The results implied that the top cooling and decay heat distribution slightly influenced the temperature and heat flux of ceramic pool, but strongly impacted the characteristics of metallic layer. The comparison between the dimensionless parameters of water tests and salt tests also demonstrated the key role of simulant material. Furthermore, the downward heat transfer Nudn and the upward heat transfer Nuup obtained from COPRA two-layer experiment were in accordance with the previous data or empirical correlations. Note that using the same simulant material for oxide layer and metal layer fails to repeat the boundary conditions in the reactor case, however, the current study pays more attention to the qualitative analysis of different influence on natural convection heat transfer behavior. More valuable information about the crust characteristics and stratification effect can be obtained by the following CFD simulation of these tests.

Thermophysical properties of molten Zr1−xOx (x=0.1, 0.2) measured by electrostatic levitation

During the core meltdown accident at the Fukushima Daiichi (1F) nuclear power plant, molten Zr based alloys and oxides formed at elevated temperatures as the fuel cladding is made up of zircaloys. As decommission is taking place currently, thermophysical data of the Zr–O system are important for both understanding the details of the accident and for the ongoing decommission process. Given the high melting temperature of zirconium oxides and their high reactivity, there have been limited reports on their properties. Here, we used a contactless electrostatic levitation (ESL) technique to measure the density, surface tension, and viscosity of molten Zr1−xOx (x = 0.1, 0.2) alloys at elevated temperatures.

Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

Oxidation of molten zirconium droplets in water

Zirconium, which is used as the cladding material of nuclear fuel rods in LWRs, can react with steam in the case of a core meltdown accident. This results in the release of hydrogen which poses a significant risk of hydrogen explosion. Oxidation of Zr occurs either during the core degradation when the steam flows over the hot fuel rod surfaces or during an FCI when molten corium falls into a water pool (e.g. in the lower head). An experimental study was performed at the MISTEE-OX facility at KTH to observe and quantify the oxidation of molten zirconium droplets in a water pool. During the experimental runs, single droplets of molten zirconium were discharged into a subcooled water pool and the dynamic events were recorded using a high-speed camera. The bubble dynamics indicate a process of cyclic oxidation and hydrogen release from the rear periphery of a droplet while it is quenched in the water. The debris (solidified state of the droplet) after each run was collected for compositional and microstructural analysis via SEM/EDS. The obtained data were employed to estimate the oxidation fractions of the droplets and the results revealed several interesting insights into the oxidation phenomenon of the Zr melt. The water subcooling was observed to have a significant influence on the oxidation: the degree of oxidation decreased with increase in the water subcooling. Furthermore, the degree of oxidation was also influenced by the depth into the debris, forming compounds whose oxygen content decreases from the outer surface towards the core of the debris. Therefore, the qualitative and quantitative results presented in this paper are important in the context of developing a phenomenological understanding of the oxidation behaviour of zirconium melt during the FCI as well as to improve and validate the currently available models implemented in the state-of-art steam explosion codes.

Experimental Investigation of Melt Coolability and Ablation Behavior of Oxidic Sacrificial Material at Prototypic Conditions in Scaled Down Core Catcher

To minimize the potential risk of design extension conditions (DEC) with core meltdown, some advanced reactors employ ex-vessel core catchers which stabilize and cool the corium for prolonged period by strategically flooding it. This paper describes the coolability of the melt pool and ablation process in a scaled down ex-vessel core catcher employing sacrificial material which reduces the specific volumetric heat, temperature, and density of the melt pool. To understand these phenomena, a simulated experiment was carried out. The experiment was performed by melting about 500 kg of corium simulant using thermite reaction at about 2500 °C. The bricks of oxidic sacrificial material were arranged in the core catcher vessel which was surrounded by a tank filled with water up to a certain level. After the time required for melt inversion, water was introduced to flood the test section from the top. The melt pool temperatures were monitored at various locations using “K” and “C” type thermocouples to obtain ablation depth at different elevations with time. The results show that the coolability of the molten pool in the presence of water for the present geometry is achievable with outside vessel temperatures not exceeding 100 °C. A ceramic stable crust was observed at the top surface of the melt pool, which prevented water ingression into the molten corium. The ablation rate was found to be maximum at the lower corners of the brick arrangement with the maximum value being 0.75 mm/s. An average rate of about 0.18 mm/s was obtained in the brick matrix.

COMPASS - New modeling and simulation approach to PWR in-vessel accident progression

Abstract The objective of this paper is to discuss the modeling principles of phenomena governing core degradation/melting and in-vessel melt relocation during severe accidents in light water reactors. The proposed modeling approach has been applied in the development of a new accident simulation package, COMPASS (COre Meltdown Progression Accident Simulation Software). COMPASS can be used either as a stand-alone tool to simulate in-vessel meltdown progression up to and including RPV failure, or as a component of an integrated simulation package being developed in Korea for the APR1400 reactor. Interestingly, since the emphasis in the development of COMPASS modeling framework has been on capturing generic mechanistic aspects of accident progression in light water reactors, several parts of the overall model should be useful for future accident studies of other reactor designs, both PWRs and BWRs. The issues discussed in the paper include the overall structure of the model, the rationale behind the formulation of the governing equations and the associated simplifying assumptions, as well as the methodology used to verify both the physical and numerical consistencies of the overall solver. Furthermore, the results of COMPASS validation against two experimental data sets (CORA and PHEBUS) are shown, as well as of the predicted accident progression at TMI-2 reactor.

Sensitivity analysis of in-vessel accident progression behavior in Fukushima Daiichi Nuclear Power Plant Unit 3

The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th 2011 put the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions resulting from loss of offsite power and AC power. It is believed that the Station Blackout (SBO) and loss of heat sink led to core meltdown in all Units 1–3. Despite past research efforts on the severe accident progression in Fukushima NPP Units 1–3, there are still knowledge gaps and uncertainties existing in understanding of the severe accident scenarios and consequences.Hence, this study aims at identifying the modeling uncertainties and addressing the sensitivity parameters in Fukushima NPP Unit 3. A more detailed Control Volume (CV) division model of the reactor core region has been developed to better simulate the thermal-hydraulic behavior of liquid water and steam, which is considered to be crucial in simulating the core uncovery and degradation process. The boundary conditions such as the water injection rates by the Reactor Core Isolation Cooling (RCIC) system, the High Pressure Core Injection (HPCI) system and Alternative Water Injection (AWI) to the reactor core were determined based on the available reactor water level and pressure measurement data. The current study suggested that the local vapor heatup behavior could influence the core melting and relocation behavior, which can lead to different core degradation scenarios. With the current modeling assumptions in MELCOR, the best estimate conditions for RPV pressure history of Unit 3 suggested that 6 SRVs could have remained open when the major core slumping took place at ca. 45:20 h (ca. 12:00, March 13) with 50 to 80 tons of water inventory in the lower plenum. The current analysis also suggested that the efficiency of the AWI to the reactor core could have been only 15% as of reported by TEPCO with the current modeling conditions if debris dryout was assumed to have occurred at around ca. 54.0 h (20:40 h, March 13th). As for lower head failure, there is still large uncertainty in predicting lower head failure time with Larson-Miller creep rupture model in the current MELCOR modeling.

Experimental study on natural convection heat transfer in a two-layer corium pool based on COPRA facility

During core meltdown accidents, the heat flux distribution along the wall of lower plenum was closely related to the potential of vessel failure. Concerning with the natural convection heat transfer behavior in stratified corium pools, several two-layer tests were performed on COPRA facility recently. The test vessel (a 1/4 circular slice with an inner radius of 2.2 m and a width of 0.2 m) is divided into two parts by a plate made of 304 stainless steel. The lower part, heated by 18 electrical heating rods, is applied to simulate the oxide pool. The upper part without heat generation is assumed to be a metallic layer. During the experiment, different heating powers and boundary conditions were employed to obtain the temperature distributions of oxide pool and metal layer, as well as the heat flux through the vessel wall. Prototypic Ra′ range, varied from 1.88 × 1016 to 5.39 × 1016, was successfully achieved in the oxide pool. Despite the smaller downward Nudn numbers, the upward Nuup numbers within this study were well in accordance with the reported data. It should be mentioned that the results of water are unrepresentative for the reactor situation since no crust was formed in the bottom and top of oxide pool. However, the current study will still attract interest by the qualitative analysis of top cooling influence on natural convection phenomenon and provide valuable information to other researchers who concern with the crust formation and focusing effect on stratified corium pools behavior.

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.

Neutronic modeling of debris beds for a criticality evaluation

After the Fukushima accident, the interest of the scientific community in severe accident research has been renewed. One of the severe accident research issues that needs to be further investigated is the recriticality potential of the debris bed that is formed after the core meltdown. The uncertainty regarding fuel debris conditions is very high. Consequently, one of the challenges of the criticality evaluation is the neutronic modeling of the debris itself. Conservative assumptions and simplifications have to be performed to overcome the uncertainties and to achieve a computationally feasible debris bed model. This paper presents a suitability analysis of several debris bed models for a criticality evaluation. The objective is to identify the most suitable model, i.e. the model with the best compromise between accuracy and simplicity, from a set of candidates. For that, a detailed near-to-reality model was developed whose higher fidelity results have been used as a reference to evaluate the adequacy of various simplifications. This work focuses on the modeling of the porous internal structure of the debris, concretely on the fuel particles, which are characterized by their shape, size and spatial distribution. The final proposed neutronic model represents the debris bed as a regular 3D arrangement of monosized spherical particles. This is a very suitable model delivering accurate results as long as an appropriate equivalent diameter deq is chosen. However, it was proved that not always the same deq value is adequate for representing the debris in the whole porosity range. The Monte Carlo codes MCNP6.1 and Serpent 2.1 were used to construct the debris bed models and to compute the infinite multiplication factor(k∞).

Experimental investigation of PWR accident scenarios at the PKL test facility

PKL is the only test facility in Europe that replicates the entire primary side and the most important parts of the secondary side of western-type Pressurized Water Reactors (PWR) in the scale of 1:1 in heights. It is also worldwide the only test facility with 4 identical reactor coolant loops arranged symmetrically around the Reactor Pressure Vessel (RPV) for simulation of nonsymmetrical boundary conditions between the reactor loops. Thermal-hydraulic phenomena observed in PWRs are simulated in the PKL test facility for over 40 years. The analyses carried out in these years encompass a large spectrum of accident scenario simulations and corresponding cool-down procedures. The overall goal of the PKL experiments is to show that under accident conditions - even for extreme and highly unlikely assumptions as additional loss of safety systems - the core cooling can be maintained or re-established by automatic or operator- performed procedures and that a severe accident e.g. a core melt-down can be avoided under all circumstances. Another goal of the tests performed in the PKL facility is the provision of data for validation of thermal-hydraulic system codes. This paper presents recent modifications of the PKL facility, applied in order to adapt the facility to the latest western-type designs currently built in the world. The paper discusses also important results obtained in the last years.

Thermal-hydraulic modeling for deterministic safety analysis of portable equipment application in the VVER-1000 nuclear reactor during loss of ultimate heat sink accident using RELAP5/MOD3.2 code

Thermal-hydraulic model is an important tool in deterministic safety analysis of the Nuclear Power Plant (NPP) which would confirm the adequacy and efficiency of provisions in the defense in depth concept to cope with challenges of the plant safety. In this paper, for the first time, a validated thermal-hydraulic model is presented for Deterministic Safety Analysis of the portable equipment to be applied in VVER-1000 nuclear reactor during severe accident. Therefore, the VVER-1000 as a case study reactor model is assessed and evaluated through RELAP5/MOD3.2 thermal hydraulic code during the loss of ultimate heat sink accident. Since LUHS accident is absent in the Final Safety Analysis Report of the nuclear power plant, this accident scenario is analyzed for the first time in this study. The simulation is done for two situations of with and without safety modification. This modification is related to portable equipment application which would maintain the primary or secondary circuit cooling in long term. The results of these simulations indicate that LUHS would lead to fuel failure and finally core meltdown. But, the modified LUHS simulation shows that this designed portable equipment can remove the decay heat of the core or transfer the heat into the ultimate heat sink in a continuous manner until the safety systems recovery become possible. Indeed, the most important achievements of this paper consist of: preparation of emergency instructions, accident management at the time of loss of ultimate heat sink and the training of NPP's personnel.

The Influence of Crust Layer on Reactor Pressure Vessel Failure Under Pressurized Core Meltdown Accident

The so-called in-vessel retention (IVR) was considered as a severe accident management strategy and had been certified by Nuclear Regulatory Commission (NRC) in U.S. as a standard measure for severe accident management since 1996. In the core meltdown accident, the reactor pressure vessel (RPV) integrity should be ensured during the prescribed time of 72 h. However, in traditional concept of IVR, several factors that affect the RPV failure were not considered in the structural safety assessment, including the effect of corium crust on the RPV failure. Actually, the crust strength is of significant importance in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the lower head (LH) of the RPV. Consequently, the RPV integrity is significantly influenced by the crust. A strong, coherent crust anchored to the RPV walls could allow the yet-molten corium to fall away from the crust as it erodes the RPV, therefore thermally decoupling the melt pool from the coolant and sharply reducing the cooling rate. Due to the thermal resistance of the crust layer, it somewhat prevents further attack of melt pool from the RPV. In the present study, the effect of crust on RPV structural behaviors was examined under multilayered crust formation conditions with consideration of detailed thermal characteristics, such as high-temperature gradient across the wall thickness. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative RPV to figure out the possibility of high temperature induced failures with the effect of crust layer.