In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3D©

2002 ◽  
Author(s):  
D. L. Knudson ◽  
J. L. Rempe
Keyword(s):  
Heat Transfer ◽  
Molten Pool ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Complex Processes ◽  
Lower Head ◽  
Environmental Laboratory ◽  
Molten Materials ◽  
Core Materials

Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D© has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D© relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D© are outlined.

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.

Numerical Simulation of Liquid Jet Behavior in Shallow Pool by Interface Tracking Method

2020 ◽  
Author(s):  
Takayuki Suzuki ◽  
Hiroyuki Yoshida ◽  
Naoki Horiguchi ◽  
Sota Yamamura ◽  
Yutaka Abe
Keyword(s):  
Numerical Simulation ◽  
Hydrodynamic Interaction ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Water Pool ◽  
Molten Material ◽  
Jet Breakup ◽  
Jet Behavior ◽  
Lattice Resolution ◽  
Core Materials

Abstract In the severe accident (SA) of nuclear reactors, fuel and components melt, and melted materials fall to a lower part of a reactor vessel. In the lower part of a reactor vessel, in some sections of the SAs, it is considered that there is a water pool. Then, the melted core materials fall into a water pool in the lower plenum as a jet. The molten material jet is broken up, and heat transfer between molten material and coolant may occur. This process is called a fuel-coolant interaction (FCI). FCI is one of the important phenomena to consider the coolability and distribution of core materials. In this study, the numerical simulation of jet breakup phenomena with a shallow pool was performed by using the developed method (TPFIT). We try to understand the hydrodynamic interaction under various, such as penetration, reach to the bottom, spread, accumulation of the molten material jet. Also, we evaluated a detailed jet spread behavior and examined the influence of lattice resolution and the contact angle. Furthermore, the diameters of atomized droplets were evaluated by using numerical simulation data.

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.

Experimental Study on the Improved In-Vessel Corium Retention Concepts for the Severe Accident Management

2002 ◽  
Author(s):  
K. H. Kang ◽  
R. J. Park ◽  
K. M. Koo ◽  
S. B. Kim ◽  
H. D. Kim
Keyword(s):  
Heat Removal ◽  
Elevated Pressure ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Severe Accident ◽  
Experimental Facility ◽  
Dual Strategy ◽  
Accident Management ◽  
Lower Head ◽  
Inner Diameter

Feasibility experiments were performed for the assessment of improved In-Vessel Corium Retention (IVR) concepts using an internal engineered gap device and also a dual strategy of In/Ex-vessel cooling using the LAVA experimental facility. The internal engineered gap device made of carbon steel was installed inside the LAVA lower head vessel and it made a uniform gap with the vessel by 10 mm. In/Ex-vessel cooling in the dual strategy experiment was performed installing an external guide vessel outside the LAVA lower head vessel at a uniform gap of 25 mm. The LAVA lower head vessel was a hemispherical test vessel simulated with a 1/8 linear scale mock-up of the reactor vessel lower plenum with an inner diameter of 500 mm and thickness of 25 mm. In both of the tests, Al2O3 melt was delivered into about 50K subcooled water inside the lower head vessel under the elevated pressure. Temperatures of the internal engineered gap device and the lower head vessel were measured by K-type thermocouples embedded radially in the 3mm depth of the lower head vessel outer surface and in the 4mm depth of the internal engineered gap device, respectively. In the dual strategy experiment, the Ex-vessel cooling featured pool boiling in the gap between the lower head vessel and the external guide vessel. It could be found from the experimental results that the internal engineered gap device was intact and so the vessel experienced little thermal and mechanical attacks in the internal engineered gap device experiment. And also the vessel was effectively cooled via mutual boiling heat removal in- and ex-vessel in the dual strategy experiment. Compared with the previous LAVA experimental results performed for the investigation of the inherent in-vessel gap cooling, it could be confirmed that the Ex-vessel cooling measure was dominant over the In-vessel cooling measure in this study. It is concluded that the improved cooling measures using a internal engineered gap device and a dual strategy promote the cooling characteristics of the lower head vessel and so enhance the integrity of the vessel in the end.

Experimental Investigation on Critical Heat Flux From Downward-Facing Flat Plate for Different Orientation Angles

2018 ◽  
Author(s):  
K. H. Deng ◽  
Y. Zhang ◽  
C. L. Wang ◽  
Y. P. Zhang ◽  
W. X. Tian ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Heating Surface ◽  
Severe Accident ◽  
Width Ratio ◽  
Lower Head ◽  
Length Width ◽  
Core Materials

After the severe accident inside a nuclear reactor, the IVR (In-vessel retention) management strategy is an effective way to keep the integrity of pressure vessel and reduce risk of radioactive leakage by holding the damaged core materials through External Reactor Vessel Cooling (ERVS). The damaged core materials aggregate in the lower head of pressure vessel and releasing heat to the lower head. Therefore, it is very important to remove heat timely to keep the integrity of pressure vessel by ERVS. The shape of lower head is hemispherical and the local Critical Heat Flux (CHF) of different parts changed with latitude. In this paper, influence of orientation angles, area and length-width ratio on CHF of plate heating surface for saturated pool boiling is investigate experimentally. The results show that CHF increases with increasing orientation angles and decreasing area, meanwhile, length-width ratio has a significantly effect on CHF.

In-Vessel Corium Behavior and Steam Explosion Risks: A Review of Experimental Approaches

2017 ◽  
Author(s):  
Pei Shen ◽  
Wenzhong Zhou
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Radioactive Material ◽  
Severe Accident ◽  
Reactor Accident ◽  
Experimental Approaches

Although no one would like to see, a severe nuclear reactor accident may result in reactor core melting, the fuel melt dropping into water in the reactor vessel, and then interacting with coolant into steam explosion. Steam explosion is a result of very rapid and intense heat transfer and violent interaction between the high temperature melt and low temperature coolant. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. Steam explosion may endanger the reactor vessel and surrounding structures. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the containment integrity. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, fist, small to intermediate scale experiments related to premixing, triggering and propagation stages are reviewed and summarized in tables. Then the intermediate to large scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities’ work is also discussed. The studies on steam explosion are vital for reactor severe accident management, and will lead to improved reactor safety.

Thermal and Structural Analysis of Reactor Vessel Lower Head Considering Core Meltdown Accident

2018 ◽  
Author(s):  
Juan Luo ◽  
Jiacheng Luo ◽  
Lei Sun ◽  
Peng Tang
Keyword(s):  
Structural Analysis ◽  
Creep Deformation ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Creep Failure ◽  
The Core ◽  
Lower Head ◽  
Core Meltdown

In the core meltdown severe accident, in-vessel retention (IVR) of molten core debris by external reactor vessel cooling (ERVC) is an important mitigation strategy. During the IVR strategy, the core debris forming a melt pool in the reactor pressure vessel (RPV) lower head (LH) will produce extremely high thermal and mechanical loadings to the RPV, which may cause the failure of RPV due to over-deformation of plasticity or creep. Therefore, it is necessary to study the thermomechanical behavior of the reactor vessel LH during IVR condition. In this paper, under the assumption of IVR-ERVC, the thermal and structural analysis for the RPV lower head is completed by finite element method. The temperature field and stress field of the RPV wall, and the plastic deformation and creep deformation of the lower head are obtained by calculation. Plasticity and creep failure analysis is conducted as well. Results show that under the assumed conditions, the head will not fail due to excessive creep deformation within 200 hours. The results can provide basis for structural integrity analysis of pressure vessels.

Materials Interaction Tests to Identify Base and Coating Materials for an Enhanced In-Vessel Core Catcher Design

2004 ◽  
Author(s):  
J. L. Rempe ◽  
D. L. Knudson ◽  
K. G. Condie ◽  
W. D. Swank ◽  
K. Y. Suh ◽  
...  
Keyword(s):  
High Temperature ◽  
Base Material ◽  
Severe Accident ◽  
Oxide Material ◽  
Coating Materials ◽  
Engineering Research ◽  
The Core ◽  
Lower Head ◽  
Core Catcher ◽  
Core Materials

An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (U.S.)–Korean International Nuclear Engineering Research Initiative (INERI) investigating methods to insure In-Vessel Retention (IVR) of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulating oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Initial evaluations suggest that a thermally-sprayed oxide material is the most promising candidate insulator coating for a core catcher. As part of the effort to develop an in-vessel core catcher design, a series of high temperature materials interaction tests were conducted for thermal sprayed coatings and base materials with properties deemed most promising. This paper reports results from these materials interactions tests and efforts to optimize parameters for applying the thermal spray coatings.

LIVE Experiments on Melt Behavior in the RPV Lower Head

2008 ◽  
Author(s):  
Alexei Miassoedov ◽  
Thomas Cron ◽  
Jerzy Foit ◽  
Xiaoyang Gaus-Liu ◽  
Silke Schmidt-Stiefel ◽  
...  
Keyword(s):  
Heat Flux ◽  
Molten Pool ◽  
Vessel Wall ◽  
Flux Distribution ◽  
Severe Accident ◽  
Experimental Program ◽  
Strong Impact ◽  
Residual Water ◽  
Heat Flux Distribution ◽  
Lower Head

Behavior of the corium pool in the lower head is still a critical issue in understanding of PWR core meltdown accidents. One of the key parameter for assessing the vessel mechanical strength is the resulting heat flux at the pool-vessel interface. A number of studies [1]–[3] have already been performed to pursue the understanding of a severe accident with core melting, its course, major critical phases and timing and the influence of these processes on the accident progression. Uncertainties in modeling these phenomena and in the application to reactor scale will undoubtedly persist. These include e.g. formation and growth of the in-core melt pool, relocation of molten material after the failure of the surrounding crust, characteristics of corium arrival in residual water in the lower head, corium stratifications in the lower head after the debris re-melting [4]. These phenomena have a strong impact on a potential termination of a severe accident. The main objective of the LIVE program [5] at FZK is to study the core melt phenomena both experimentally in large-scale 3D geometry and in supporting separate-effects tests, and analytically using CFD codes in order to provide a reasonable estimate 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 non-eutectic melts (mixture 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 for the assessment of the correlations derived for the molten pool behavior. The information obtained from the LIVE experiments includes heat flux distribution through the reactor pressure vessel wall in transient and steady state conditions, crust growth velocity and dependence of the crust formation on the heat flux distribution through the vessel wall. Supporting post-test analysis contributes to characterization of solidification processes of binary non-eutectic melts. Complimentary to other international programs with real corium melts, the results of the LIVE activities provide data for a better understanding of in-core corium pool behavior. The experimental results are being used for development of mechanistic models to describe the in-core molten pool behavior and their implementation in the severe accident codes like ASTEC. The paper summarizes the objectives of the LIVE program and presents the main results obtained in the LIVE experiments up to now.

Large eddy simulation on turbulent heat transfer in reactor vessel lower head corium pools

2018 ◽  
Vol 111 ◽  
pp. 293-302 ◽  
Author(s):  
Luteng Zhang ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Wenxi Tian ◽  
G.H. Su ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Large Eddy Simulation ◽  
Reactor Vessel ◽  
Turbulent Heat Transfer ◽  
Turbulent Heat ◽  
Eddy Simulation ◽  
Large Eddy ◽  
Lower Head
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