Core Melt Solidification Characteristics in PRV Lower Head-Experimental Results From LIVE Tests

2010 ◽  
Vol 132 (10) ◽  
Author(s):  
Xiaoyang Gaus-Liu ◽  
Alexei Miassoedov ◽  
Thomas Cron ◽  
Jerzy Foit ◽  
Thomas Wenz ◽  
...  
Keyword(s):  
Vessel Wall ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Water Cooling ◽  
Melt Pool ◽  
External Cooling ◽  
Time Period ◽  
Crust Layer ◽  
Lower Head ◽  
Melt Solidification

Core melt solidification phenomena in the lower plenum of pressurized reactor vessel during external reactor vessel cooling is investigated in late in-vessel phase experiment tests under different external cooling conditions and melt pouring positions. The melt solidification behavior, which has not yet been given sufficient attention, is an important issue since it influences not only the transient but also the steady state of melt pool thermal hydraulics. A noneutectic melt (80 mol %KNO3–20 mol %NaNO3) was used to simulate the core melt. It has been found out that when the vessel is cooled with water during the whole test period (water cooling), the cooling is more effective than the case that the vessel lower head is first cooled with air and flooded by water (air/water cooling). Water cooling at the beginning leads to faster buildup of crust layer on the vessel inner wall and lower crust thermal conductivity compared with air/water cooling. In comparison with the air/water cooling, the water cooling also achieves shorter time period of crust growth. During the solidification period in all tests, the constitutional supercooling condition is fulfilled. Pouring position near the vessel wall results in considerable asymmetry in the heat flux distribution through the vessel wall.

Core Melt Solidification Characteristics in RPV Lower Head: Experimental Results From LIVE-Tests

2009 ◽  
Author(s):  
Xiaoyang Gaus-Liu ◽  
Alexei Miassoedov ◽  
Thomas Cron ◽  
Jerzy Foit ◽  
Thomas Wenz ◽  
...  
Keyword(s):  
Heat Flux ◽  
Vessel Wall ◽  
Growth Period ◽  
Cooling Condition ◽  
Water Cooling ◽  
Initial Water ◽  
External Cooling ◽  
Crust Layer ◽  
The One ◽  
Melt Solidification

Core melt solidification phenomena during external reactor vessel cooling is investigated in LIVE tests with different external cooling conditions and melt pouring positions. A non-eutectic simulant melt (80-20 mole% KNO3-NaNO3) is used in the LIVE tests. It is found out that when the vessel is cooled with water at the beginning of the melt pouring, the cooling is more effective than in the case of delayed water cooling condition, in which the vessel is first cooled with air and then flooded by water. The initial water cooling leads to a faster growth of crust layer, lower crust thermal conductivity and thinner crust layer than those under the delayed water cooling condition. The initial water cooling leads also to higher heat flux through the vessel wall during the steady state and shorter crust growth period in comparison with the delayed water cooling condition. The solidification of the melt is probably under supercooling condition. The pouring position near the vessel wall results in considerable asymmetric heat flux distribution at one latitude. The heat flux at the position of melt pouring is higher than the one at other locations.

Effect of CRGT Cooling on Modes of Global Vessel Failure of a BWR Lower Head

2012 ◽  
Author(s):  
Walter Villanueva ◽  
Chi-Thanh Tran ◽  
Pavel Kudinov
Keyword(s):  
Heat Transfer ◽  
Vessel Wall ◽  
Reactor Vessel ◽  
Creep Model ◽  
Guide Tube ◽  
Mechanical Loads ◽  
Control Rod ◽  
Melt Pool ◽  
Creep Analysis ◽  
Lower Head

An in-vessel stage of a severe core melt accident in a Nordic type Boiling Water Reactor (BWR) is considered wherein a decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. This process induces creep leading to a mechanical failure of the reactor vessel wall. The focus of this study is to investigate the effect of Control Rod Guide Tube (CRGT) and top cooling on the modes of global vessel failure of the lower head. A coupled thermo-mechanical creep analysis of the lower head is performed and cases with and without CRGT and top cooling are compared. The debris bed heat-up, re-melting, melt pool formation, and heat transfer are calculated using the Phase-change Effective Convectivity Model and transient heat transfer characteristics are provided for thermo-mechanical strength calculations. The creep analysis is performed with the modified time hardening creep model and both thermal and integral mechanical loads on the reactor vessel wall are taken into account. Known material properties of the reactor vessel as a function of temperature, including the creep curves, are used as an input data for the creep analysis. It is found that a global vessel failure is imminent regardless of activation of CRGT and top cooling. However, if CRGT and top cooling is activated, the mode and timing of failure is different compared to the case with no CRGT and top cooling. More specifically, with CRGT and top cooling, there are two modes of global vessel failure depending on the size of the melt pool: (a) ‘ballooning’ of the vessel bottom for smaller pools, and (b) ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool for larger pools. Without CRGT and top cooling, only a ballooning mode of global vessel failure is observed. Furthermore, a considerable delay (about 1.4 h) on the global vessel failure is observed for the roughly 30-ton debris case if CRGT and top cooling is implemented. For a much larger pool (roughly 200-ton debris), no significant delay on the global vessel failure is observed when CRGT and top cooling is implemented, however, the liquid melt fraction and melt superheat are considerably higher in non-cooling case.

Structural Integrity of a Reactor Vessel Lower Head Under In-Vessel Steam Explosion Loads

2013 ◽  
Author(s):  
Guohong Xue ◽  
Yinbiao He ◽  
Ming Cao ◽  
Hao Yu ◽  
Yongjian Gao
Keyword(s):  
Steam Explosion ◽  
Power Plants ◽  
Structural Integrity ◽  
Equivalent Plastic Strain ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Conservation Of Energy ◽  
Lower Head ◽  
Pressure Pulses

Passive nuclear power plants emphasize the “In vessel retention” idea such that, after a postulated severe accident event, the reactor vessel wall, flooded with emergency cooling water, will maintain its structural integrity and consequently keep the molten core inside the reactor vessel. However, steam explosion may still occur when the melting core or molten metal is mixed with cooling water. The huge pressure pulses from the steam explosion may be a threat to the structural integrity of the reactor vessel lower head and the potential failure may make the situation difficult to control. This paper presents a detailed analysis on the structural integrity of a reactor vessel lower head. First, a mathematical model is built to relate the equivalent plastic strain in the lower head under explosive loads based on the law of conservation of energy. Then a finite element model, using the computer code ABAQUS, is built and the material’s yield strength as a function of strain rate was simulated using the Bodner-Symonds methodology. With this model, the dynamic response and the structural integrity of the reactor vessel lower head is studied, considering the effect of the magnitude, the shape and the duration of the pressure pulses. The method used in this paper is believed to be applicable to other types of devices containing potential explosive materials and thus could provide guiding significance to similar problems.

scholarly journals Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
Alejandro Nuñez-Carrera ◽  
Raúl Camargo-Camargo ◽  
Gilberto Espinosa-Paredes ◽  
Adrián López-García
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Core Material ◽  
Detailed Model ◽  
Lower Head ◽  
Reactor Pressure

The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

Micro-Porous Coatings and Enhanced CHF for Downward Facing Boiling During Passive Emergency Reactor Cooling

2016 ◽  
Author(s):  
Albert E. Segall ◽  
Faruk A. Sohag ◽  
Faith R. Beck ◽  
Lokanath Mohanta ◽  
Fan-Bill Cheung ◽  
...  
Keyword(s):  
Cold Spray ◽  
Heat Removal ◽  
Porous Coating ◽  
Porous Coatings ◽  
Spray Coatings ◽  
Spray Process ◽  
External Cooling ◽  
Loss Of Coolant Accident ◽  
Lower Head ◽  
Viable Approach

During a Reaction Initiated Accident (RIA) or Loss of Coolant Accident (LOCA), passive external-cooling of the reactor lower head is a viable approach for the in-vessel retention of Corium; while this concept can certainly be applied to new constructions, it may also be viable for operational systems with existing cavities below the reactor. However, a boiling crisis will inevitably develop on the reactor lower head owing to the occurrence of Critical Heat Flux or CHF that could reduce the decay heat removal capability as the vapor phase impedes continuous boiling. Fortunately, this effect can be minimized for both new and existing reactors through the use of a Cold-Spray delivered, micro-porous coating that facilitates the formation of vapor micro-jets from the reactor surface. The micro-porous coatings were created by first spraying a binary mixture with the sacrificial material then removed via etching. Subsequent quenching experiments on uncoated and coated hemispherical surfaces showed that local CHF values for the coated vessel were consistently higher relative to the bare surface. Moreover, it was observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit varied appreciably along the outer surface. Nevertheless, the results of this intriguing study clearly show that the use of Cold Spray coatings could enhance the local CHF limit for downward facing boiling by more than 88%. Moreover, the Cold-Spray process is amenable to coating the lower heads of operating reactors.

Heat Transfer Coefficient Measurement for Downward-Facing Heat Transfer on Curved Rectangular Water Channel

2016 ◽  
Author(s):  
Jun Yeong Jung ◽  
Yong Hoon Jeong
Keyword(s):  
Heat Transfer ◽  
Test Section ◽  
Flow Boiling ◽  
Water Channel ◽  
Outer Wall ◽  
Reactor Vessel ◽  
Working Fluid ◽  
Heater Surface ◽  
Ir Camera ◽  
Lower Head

In-Vessel Retention by External Reactor Vessel Cooling (IVR-ERVC) is method of removing the decay heat by cooling reactor vessel after corium relocation, and is also one of severe accident management strategies. Estimating heat transfer coefficients (HTCs) is important to evaluate heat transfer capability of the ERVC. In this study, the HTCs of outer wall of reactor vessel lower head were experimentally measured under the IVR-ERVC situation of Large Loss of Coolant Accident (LLOCA) condition. Experimental equipment was designed to simulate flow boiling condition of ERVC natural circulation, and based on APR+ design. This study focused on effects of real reactor vessel geometry (2.5 m of radius curvature) and material (SA508) for the HTCs. Curved rectangular water channel (test section) was design to simulate water channel which is between the reactor vessel lower head outer wall and thermal insulator. Radius curvature, length, width and gap size of the test section were respectively 2.5 m, 1 m, 0.07 m and 0.15 m. Two connection parts were connected at inlet and outlet of the test section to maintain fluid flow condition, and its cross section geometry was same with one of test section. To simulate vessel lower head outer wall, thin SA508 plate was used as main heater, and test section supported the main heater. Thickness, width, length and radius curvature of the main heater were 1.2 mm, 0.07 m, 1 m and 2.5 m respectively. The main heater was heated by DC rectifier, and applied heat flux was under CHF value. The test section was changed for each experiment. The HTCs of whole reactor vessel lower head (bottom: 0 ° and top: 90 °) were measured by inclining the test section, and experiments were conducted at four angular ranges; 0–22.5, 22.5–45, 45–67.5 and 67.5–90 °. DI water was used as working fluid in this experiment, and all experiments were conducted at 400 kg/m2s of constant mass flux with atmospheric pressure. The working fluid temperatures were measured at two point of water loop by K-type thermocouple. The main heater surface temperatures were measured by IR camera. The main heater was coated by carbon spray to make uniform surface emissivity, and the IR camera emissivity calibration was also conducted with the coated main heater. The HTCs were calculated by measured main heater surface temperature. In this research, the HTC results of 10, 30, 60 and 90 ° inclination angle were presented, and were plotted with wall super heat.

Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

2005 ◽  
Vol 152 (2) ◽  
pp. 162-169 ◽  
Author(s):  
Yong Hoon Jeong ◽  
Soon Heung Chang ◽  
Won-Pil Baek
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Test Section ◽  
Vessel Wall ◽  
Reactor Vessel

scholarly journals Experimental study of water cooling effect on heat transfer to increase output power of 180 watt peak photovoltaic module

2018 ◽  
Vol 67 ◽  
pp. 01009
Author(s):  
Arrad Ghani Safitra ◽  
Fifi Hesty Sholihah ◽  
Erik Tridianto ◽  
Ikhsan Baihaqi ◽  
Ni Nyoman Ayu Indah T.
Keyword(s):  
Heat Transfer ◽  
Experimental Study ◽  
Surface Temperature ◽  
Output Power ◽  
Cooling Water ◽  
Photovoltaic Module ◽  
Water Cooling ◽  
Water Replacement ◽  
Pv Modules ◽  
Water Height

Photovoltaic (PV) modules require solar radiation to generate electricity. This study aims to determine the effect of water cooling PV modules on heat transfer, output power, and electrical efficiency of PV modules. The experiments carried out in this study were to vary the heights of flooded water (with and without cooling water replacement control) and cooling water flow. Variations in the height of flooded water are 0,5 cm, 1 cm, 2 cm, and 4 cm. While the flow rate variations are 2 L/min, 4 L/min, and 8 L/min. The flooded water replacement control will be active when the PV surface temperature reached 45°C. When the temperature dropped to 35°C, the cooler is disabled to let more photon to reach PV surface. The results showed that the lowest heat transfer occurred in the variation of 4 cm flooded water height without water replacement control, i.e. 28.53 Watt, with an average PV surface temperature of 32.92°C. The highest average electric efficiency occurred in the variation of 0,5 cm flooded water height with water replacement control, i.e. 13.12%. The use of cooling water replacement control is better due to being able to skip more photons reach PV surface with low PV temperature.

Evaluations of the Coolability Through the Inherent In-Vessel Gap Cooling in the LAVA Experiments

2002 ◽  
Author(s):  
K. H. Kang ◽  
R. J. Park ◽  
J. T. Kim ◽  
S. B. Kim ◽  
H. D. Kim
Keyword(s):  
Current Flow ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Gap Formation ◽  
Lower Head ◽  
Melt Composition ◽  
Counter Current ◽  
Vessel Bottom ◽  
Cooling Model

The analysis of the LAVA (Lower-plenum Arrested Vessel Attack) experimental results focused on gap formation and in-vessel gap cooling characteristics have been performed. In the LAVA experiment, Al2O3/Fe thermite melt (or Al2O3 only) was used as a corium simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation in case there was an internal pressure load across the vessel. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were mainly determined by the possibilities of the water ingression into the gap. The possibility of heat removal through the gap in the LAVA experiments was confirmed from that the vessel cooled down with the conduction heat flux through the vessel by 70 to 470 kW/m2. Also the quantitative evaluations of the in-vessel coolability using gap cooling model based on counter current flow limits (CCFL) have been performed for the LAVA experiments in parallel. It could be inferred from the analysis for the LAVA experiments that the vessel could effectively cooldown via heat removal through the gap cooling even if 2mm thick gap should form between the interface of the melt and the vessel in the 30 kg of Al2O3 melt tests. In the case of large melt mass of 70 kg of Al2O3 melt, however, the infinite possibility of heat removal through a small size gap such as 1 to 2 mm thick couldn’t be guaranteed due to the difficulties of water ingression through the gap into the lower head vessel bottom induced by the CCFL. Synthesized the experimental results and the analytical evaluations using the CCFL model, it could be found that the coolability through gap cooling was affected mainly by the melt composition and mass and also the gap thickness.

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