Investigation of the pressure vessel lower head potential failure under IVR-ERVC condition during a severe accident scenario in APR1400 reactors

2021 ◽  
Vol 376 ◽  
pp. 111107
Author(s):  
Muritala Alade Amidu ◽  
Yacine Addad ◽  
Jeong Ik Lee ◽  
Dong Hoon Kam ◽  
Yong Hoon Jeong
Keyword(s):  
Pressure Vessel ◽  
Severe Accident ◽  
Accident Scenario ◽  
Lower Head ◽  
Potential Failure

Test Design, Results, and Archived Database for the OECD Lower Head Failure Program Integral Experiments

2006 ◽  
Author(s):  
Larry L. Humphries ◽  
Tze Yao Chu ◽  
John H. Bentz
Keyword(s):  
Wall Temperature ◽  
Pressure Vessel ◽  
Severe Accident ◽  
Core Material ◽  
Material Strength ◽  
Test Design ◽  
Temperature Differential ◽  
Pressurized Water ◽  
Accident Management ◽  
Lower Head

In the event of a severe core meltdown accident, core material can relocate to the lower head of a pressurized water reactor (PWR) vessel resulting in significant thermal and pressure loads to the vessel. The potential for failure of the pressure vessel makes possible the release of core material to the containment. The objective of this experimental/analytical program is to characterize the mode, timing, and size of lower head failure (LHF) under severe accident conditions. The OECD Lower Head Failure (OLHF) project investigates lower head failure for conditions of low reactor coolant system (RCS) pressure (2–5 MPa) and prototypic through-wall temperature differential (ΔTW >200K). Low RCS pressure is motivated by the desire to use the data to develop models for assessing accident management strategies involving reactor pressure vessel (RPV) depressurization. Pressure transient is useful in assessing the effect of water injection as part of accident management strategy. Prototypic through-wall temperature differential, ΔTW, is of importance because of the need to provide data where stress redistribution in the vessel wall occurs (as a result of decreasing material strength with temperature). Test design and results for the four OLHF integral tests are reported and summarized in this paper. A short description of the test conduct and heating history is followed by a description of the vessel failure site, the vessel deformation, temperature profiles, stress state, and rupture dynamics for each test. Key observations and conclusions are summarized for each test. The ∼1/5 scale tests are extensively instrumented to provide temperature, pressure, and displacement data. The vessel surfaces are mapped both before and after the test to provide measurements of pre-test thickness, post-test thickness, and cumulative vessel deformation. Data has been assessed and qualified in data reports for each test. The data has been preserved in MSEXCEL™ spreadsheets with macro utilities to facilitate access and analysis of the data. As a result, there exists a well-archived, well-qualified database for model development and validation.

Study on the Accidental Rupture of Hot Leg or Surge Line in SBO Accident

2006 ◽  
Author(s):  
Kun Zhang ◽  
Xuewu Cao
Keyword(s):  
Pressure Vessel ◽  
Severe Accident ◽  
Base Case ◽  
Management Guidelines ◽  
Accident Management ◽  
Lower Head ◽  
Station Blackout ◽  
Surge Line ◽  
Reactor Pressure ◽  
Relap5 Code

The postulated total station blackout accident (SBO) of PWR NPP with 600 MWe in China is analyzed as the base case using SCDAP/RELAP5 code. Then the hot leg or surge line are assumed to rupture before the lower head of Reactor Pressure Vessel (RPV) ruptures, and the progressions are analyzed in detail comparing with the base case. The results show that the accidental rupture of hot leg or surge line will greatly influence the progression of accident. The probability of hot leg or surge line rupture in intentional depressurization is also studied in this paper, which provides a suggestion to the development of Severe Accident Management Guidelines (SAMG).

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.

Computational Simulation of Natural Convection of a Molten Core in Lower Head of a PWR Pressure Vessel

2010 ◽  
Author(s):  
Camila B. Vieira ◽  
Jian Su
Keyword(s):  
Natural Convection ◽  
Rayleigh Number ◽  
Excellent Agreement ◽  
Pressure Vessel ◽  
Flat Surface ◽  
Computational Simulation ◽  
Internal Heat ◽  
Severe Accident ◽  
Bottom Surface ◽  
Lower Head

The paper reports computational simulation of natural convection of a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel. Numerical simulations were carried out for a two-dimensional semi-circular slice geometry which represents the lower head of a PWR pressure vessel. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate is simulated by using a commercial computational fluid dynamics (CFD) software ANSYS CFX 12.0. The bottom circular surface and top flat surface are both kept at a same constant temperature. The Boussinesq model is used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-ω based SST (shear stress transport) turbulent model is used to model the turbulent eddy viscosity. Two Prandtl numbers are considered, 6.13 and 7.0. Five Rayleigh numbers were simulated for each Prandtl number: 109, 1010, 1011, 1012, and 1013. The average Nusselt numbers on the bottom surface calculated by the computational simulations are in excellent agreement with Mayinger et al. (1976) correlation, with the average Nusselt number on the top flat surface calculated by the computational simulation is in excellent agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations only at Ra = 109. Numerical results of the velocity and temperature fields show the α-phenomena at lower Rayleigh number and ν-phenomena at lower Rayleigh number.

Local Failure Modes of a Nuclear Reactor Pressure Vessel Nozzle Under Severe Accident Conditions

2002 ◽  
Author(s):  
Young J. Oh ◽  
Kwang J. Jeong ◽  
Byung G. Park ◽  
Il S. Hwang
Keyword(s):  
Weld Metal ◽  
Pressure Vessel ◽  
Failure Modes ◽  
Nuclear Reactor ◽  
Creep Rupture ◽  
Local Failure ◽  
Severe Accident ◽  
Lower Head ◽  
Accident Conditions ◽  
Reactor Pressure

Most past studies for the creep rupture of a nuclear reactor pressure vessel (RPV) lower head under severe accident conditions, have focused on global deformation and rupture modes. Limited efforts were made on local failure modes associated with penetration nozzles as a part of TMI-2 Vessel Investigation Project (TMI-2 VIP) in 1990’s. However, it was based on an excessively simplified shear deformation model. In the present study, the mode of nozzle failures is investigated using data and nozzle materials from Sandia National Laboratory’s Lower Head Failure Experiment (SNL-LHF). Crack-like separations were revealed at the nozzle weld metal to RPV interfaces indicating the importance of normal stress component rather than the shear stress in the creep rupture. Creep rupture tests were conducted for nozzle and weld metal materials, respectively, at various temperature and stress levels. Stress distribution in the nozzle region is calculated using elastic-viscoplastic finite element analysis (FEA) using the measured properties. Calculation results are compared with earlier results based on the pure shear model of TMI-2 VIE It has been concluded from both LHF-4 nozzle examination and FEA that normal stress at the nozzle/lower head interface is the dominant driving force for the local failure with its likelihood significantly greater than previously assumed.

Simplified Modeling of a PWR Reactor Pressure Vessel Lower Head Failure in the Case of a Severe Accident

2002 ◽  
Author(s):  
V. Koundy ◽  
M. Durin ◽  
L. Nicolas ◽  
A. Combescure
Keyword(s):  
Pressure Vessel ◽  
High Pressures ◽  
Numerical Models ◽  
Numerical Results ◽  
Reactor Pressure Vessel ◽  
The Other ◽  
Severe Accident ◽  
Core Material ◽  
Lower Head ◽  
Reactor Pressure

In order to characterize the timing, mode and size of a possible lower head failure (LHF) of the reactor pressure vessel (RPV) in the event of a core meltdown accident, several large-scale LHF experiments were performed under the USNRC/SNL LHF program. The experiments examined lower head failure at high pressures (10 MPa in most cases) and with small throughwall temperature differentials. Another recent USNRC/SNL LHF program, called the OLHF program, has been undertaken in the framework of an OECD project. This was an extension of the first program and dealt with low and moderate pressures (2 MPa to 5 MPa) but with large throughwall temperature differentials. These experiments should lead to a better understanding of the mechanical behavior of the reactor vessel lower head, which is of importance both in severe accident assessment and the definition of accident mitigation strategies. A well-characterized failure of the lower head is of prime importance for the evaluation of the quantity of core material that can escape into the containment, since this defines the initial conditions for all external-vessel events. The large quantity of escaping corium may lead to direct heating of the containment. This is an important severe accident issue because of its potential to cause early containment failure. The experiments also provide data for model development and validation. For our part, as one of the program partners, numerical modeling was performed to simulate these experiments. This paper presents a detailed description of three of our numerical models used for the simulation. The first model is a simplified semi-analytical approach based on the theory of a spherical shell subjected to internal pressure. The two other methods deal with 2D finite element (2D-FE) modeling: one combines the Norton-Bailey creep law with a damage model proposed by Lemaitre-Chaboche while the other uses only a creep failure criterion but takes into account thermo-metallurgical phase transformations. The numerical results are consistent with the experimental measurements. The effect on the numerical results of the multiphase transformation of the shell material and of the two failure criteria used, one involving necking (Conside`re’s criterion) and the other involving creep damage (Lemaitre-Chaboche), is discussed.

scholarly journals Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-14 ◽  
Author(s):  
Mindaugas Valinčius ◽  
Tadas Kaliatka ◽  
Algirdas Kaliatka ◽  
Eugenijus Ušpuras
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Pressure Vessel ◽  
Heat Fluxes ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Focusing Effect ◽  
Accident Management ◽  
Lower Head ◽  
Reactor Pressure

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

scholarly journals Study of tearing behaviour of a PWR reactor pressure vessel lower head under severe accident loadings

2008 ◽  
Vol 238 (9) ◽  
pp. 2411-2419 ◽  
Author(s):  
Vincent Koundy ◽  
Cataldo Caroli ◽  
Laetitia Nicolas ◽  
Philippe Matheron ◽  
Jean-Marie Gentzbittel ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Lower Head ◽  
Reactor Pressure

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.

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