Calculation of Evaporation From Fukushima NPP Spent Fuel Pools

2018 ◽  
Author(s):  
Mirza M. Shah
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Spent Fuel ◽  
Decay Heat ◽  
Wide Range ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
Safety Systems ◽  
Good Agreement

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and post-accident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The post-accident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25 % lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

Calculation of Evaporation From Fukushima Nuclear Power Plant Spent Fuel Pools

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Mirza M. Shah
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Spent Fuel ◽  
Decay Heat ◽  
Wide Range ◽  
Spent Fuel Pool ◽  
Safety Systems

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and postaccident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The postaccident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25% lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

Analyses of Severe Accident Sequences in the Spent Fuel Pool of the VVER-1000 Type of Reactor With MELCOR 1.8.6

2014 ◽  
Author(s):  
Miroslav Kotouč
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Geometry ◽  
Steel Plates ◽  
Severe Accident ◽  
Spent Fuel ◽  
Concrete Walls ◽  
Nuclear Disaster ◽  
Spent Fuel Pool ◽  
Fukushima Nuclear Disaster

Since the Fukushima nuclear disaster in 2011, much attention has been paid to investigation of severe accidents (SA) progression in spent fuel pools (SFP) of various types of nuclear power plants (NPP). In Czech Republic, 4 VVER-440 and 2 VVER-1000 types of reactors (at the Dukovany and Temelin NPPs, respectively) are currently under operation. In order to enhance their safety, especially with respect to station black-out (SBO) events, numerical analyses have been carried out evaluating the risks associated with accidents occurring also in the SFP. The present paper deals with analyses of 2 postulated scenarios (loss of cooling and loss of coolant) and is mainly focused on the input deck preparation for the integral, lumped parameter (LP) code for SA analyses — MELCOR 1.8.6. Emphasis is put on description of correct implementation of the complex geometry of the SFP, consisting of 3 distinct pools separated by concrete walls (lined with steel plates) in which fuel assemblies (FA) are stored in an absorber grid (rack). In the description of the prepared numerical model, light is shed on the encountered modeling issues, and corresponding hints and solutions are proposed in order to provide guidance for preparing adequate models for various types of SFP designs. Finally, some of the most characteristic results are presented for both postulated scenarios.

Experimental Investigation Into Buckling Failure of Slender Metal Plates in Severe Accident Conditions

2014 ◽  
Author(s):  
Byeongnam Jo ◽  
Wataru Sagawa ◽  
Koji Okamoto
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Load ◽  
Severe Accident ◽  
Steel Columns ◽  
Wide Range ◽  
Buckling Failure ◽  
The Difference ◽  
Accident Conditions

Buckling failure load of stainless steel columns under compressive stress was experimentally measured in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling failure load defined as load which causes failure of the column (plastic collapse) was measured in a wide range of temperatures from 25 °C up to 1200 °C. The load values measured in this study were compared to numerical estimations by eigenvalue simulations (for an ideal column) and by nonlinear simulations (for a column with initial bending). Two different methods for measurement of the buckling failure load were employed to examine the effect of thermal history on buckling failure. Different load values were obtained from two methods in high temperature conditions over 800 °C. The difference in the buckling failure load between two methods increased with temperature, which was explained by the effect of creep at high temperatures. Moreover, the influence of asymmetric temperature profiles along a plate column was also explored with regard to the failure mode and the buckling failure load. In present study, all of the buckling processes were visualized by a high speed camera.

scholarly journals Temperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode

2021 ◽  
Vol 7 (1) ◽  
pp. 9-13
Author(s):  
David A. Hakobyan ◽  
Victor I. Slobodchuk
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Cooling Mode ◽  
Temperature Conditions ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK reactors have not been fully resolved so far. For this reason, nuclear power plants are forced to search for new options for the disposal of spent fuel, which can provide at least temporary SNF storage. One of the possible solutions to this problem is to switch to compacted SNF storage in reactor spent fuel pools (SFPs). As the number of spent fuel assemblies (SFAs) in SFPs increases, a greater amount of heat is released. In addition, no less important is the fact that a place for emergency FA discharging should be provided in SFPs. The paper presents the results of a numerical simulation of the temperature conditions in SFPs both for compacted SNF storage and for emergency FA discharging. Several types of disturbances in normal SFP cooling mode are considered, including partial loss of cooling water and exposure of SFAs. The simulation was performed using the ANSYS CFX software tool. Estimates were made of the time for heating water to the boiling point, as well as the time for heating the cladding of the fuel elements to a temperature of 650 °С. The most critical conditions are observed in the emergency FA discharging compartment. The results obtained make it possible to estimate the time that the personnel have to restore normal cooling mode of the spent fuel pool until the maximum temperature for water and spent fuel assemblies is reached.

Probabilistic Assessment of Countermeasures for Loss of Ultimate Heat Sink to Spent Fuel Pool

2017 ◽  
Author(s):  
Bumpei Fujioka ◽  
Naoki Hirokawa ◽  
Daisuke Taniguchi
Keyword(s):  
Heat Sink ◽  
Nuclear Power ◽  
Water System ◽  
Cooling Water ◽  
Severe Accident ◽  
Probabilistic Assessment ◽  
Spent Fuel ◽  
The Past ◽  
Plant Safety ◽  
Spent Fuel Pool

In the Fukushima Dai-ichi nuclear power station, Loss of Ultimate Heat Sink (LUHS) was caused by the great east japan earthquake and the subsequent tsunami [1]. It resulted in severe accident in three units. In that time, fuel damage in Spent Fuel Pool (SFP) were prevented by the various countermeasures such as makeup by pump truck and recovery of injection systems /cooling water system. In the past, Probabilistic Safety Assessment (PSA) has been developed with a focus on the reactor. After the accident, it has been acknowledged that SFP PSA is important to enhance the plant safety. In this study, probabilistic assessment is performed to suggest countermeasures for LUHS to SFP.

Self-Propelling Cooling Systems: Back-Fitting Passive Cooling Functions to Existing Nuclear Power Plants

2012 ◽  
Author(s):  
Dominik von Lavante ◽  
Dietmar Kuhn ◽  
Ernst von Lavante
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Electrical Power ◽  
Reactor Core ◽  
Passive Cooling ◽  
Spent Fuel ◽  
Cooling Systems ◽  
Spent Fuel Pool

The present paper describes a back-fit solution proposed by RWE Technology GmbH for adding passive cooling functions to existing nuclear power plants. The Fukushima accidents have high-lighted the need for managing station black-out events and coping with the complete loss of the ultimate heat sink for long time durations, combined with the unavailability of adequate off-site supplies and adequate emergency personnel for days. In an ideal world, a nuclear power plant should be able to sustain its essential cooling functions, i.e. preventing degradation of core and spent fuel pool inventories, following a reactor trip in complete autarchy for a nearly indefinite amount of time. RWE Technology is currently investigating a back-fit solution involving “self-propelling” cooling systems that deliver exactly this long term autarchy. The cooling system utilizes the temperature difference between the hotter reactor core or spent fuel pond with the surrounding ultimate heat sink (ambient air) to drive its coolant like a classical heat machine. The cooling loop itself is the heat machine, but its sole purpose is to merely achieve sufficient thermal efficiency to drive itself and to establish convective cooling (∼2% thermal efficiency). This is realized by the use of a Joule/Brayton Cycle employing supercritical CO2. The special properties of supercritical CO2 are essential for this system to be practicable. Above a temperature of 30.97°C and a pressure of 73.7bar CO2 becomes a super dense gas with densities similar to that of a typical liquid (∼400kg/m3), viscosities similar tothat of a gas (∼3×105Pas) and gas like compressibility. This allows for an extremely compact cooling system that can drive itself on very small temperature differences. The presented parametric studies show that a back-fitable system for long-term spent fuel pool cooling is viable to deliver excess electrical power for emergency systems of approximately 100kW. In temperate climates with peak air temperatures of up to 35°C, the system can power itself and its air coolers at spent fuel pool temperatures of 85°C, although with little excess electrical power left. Different back-fit strategies for PWR and BWR reactor core decay heat removal are discussed and the size of piping, heat exchangers and turbo-machinery are briefly evaluated. It was found that depending on the strategy, a cooling system capable of removing all decay heat from a reactor core would employ piping diameters between 100–150mm and the investigated compact and sealed turbine-alternator-compressor unit would be sufficiently small to be integrated into the piping.

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Sumit V. Prasad ◽  
A. K. Nayak
Keyword(s):  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Public Domain ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Critical Issues

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Experimental Investigation on Heat Transfer for Two-Phase Flow Under Natural Convection

2012 ◽  
Author(s):  
Milan Amižić ◽  
Estelle Guyez ◽  
Jean-Marie Seiler
Keyword(s):  
Heat Transfer ◽  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Liquid Pool ◽  
Severe Accident ◽  
Bottom Plate ◽  
Interfacial Region ◽  
Two Phase ◽  
Wide Range

In the frame of severe accident research for the second and the third generation of nuclear power plants, some aspects of the concrete cavity ablation during the molten corium–concrete interaction are still remaining issues. The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat melt-through. For the purpose of experimental investigation of thermal-hydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched jointly by CEA, EDF, IRSN, GDF-Suez and SARNET. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the smallest pool configuration (50 cm × 25 cm × 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s. This paper presents some preliminary conclusions deduced from the experiments which involve a liquid pool with the gas injection only from the bottom plate. A comparison with existing models for the assessment of heat transfer has also been carried out.

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