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.

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.

Safety Analyses for Preventing PCV Damage by FCI in Kashiwazaki-Kariwa 6,7

2017 ◽  
Author(s):  
Akinori Hayakawa ◽  
Shouichi Suehiro ◽  
Satoshi Mizuno ◽  
Yoshihiro Oyama ◽  
Shinichi Kawamura
Keyword(s):  
Pressure Vessel ◽  
Steam Explosion ◽  
Reactor Core ◽  
Cooling Water ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Response Analysis ◽  
Maximum Pressure ◽  
Containment Vessel ◽  
Reactor Pressure

The containment vessel failure mode, “molten fuel-coolant interaction outside the reactor pressure vessel” (“ex-vessel FCI”) is a phenomenon of rapid increase in the pressure inside the containment vessel or steam explosion caused by the contact between the molten reactor core and cooling water outside the reactor pressure vessel after the reactor core is damaged. To evaluate the viability of keeping confinement function of the containment vessels of Units 6 and 7 (ABWRs) of Kashiwazaki-Kariwa Nuclear Power Station against “ex-vessel FCI,” we conducted a code-based event progression analysis. For evaluation of the rapid increase, we employed the severe accident analysis code, MAAP, after organizing the critical phenomena of the event. In addition, assuming a case of steam explosion occurrence, we conducted an analysis with employing the steam explosion analysis code, JASMINE, and the structural response analysis code, AUTODYN-2D. As a result of the evaluation, the maximum pressure and temperature inside the containment vessels were lower than their limits. Moreover, the maximum stress applied to the lower part of the containment vessels was lower than the yield stress on support structure of the lower part of the containment vessels. Therefore, we could confirm that the containment vessels can keep their integrity.

scholarly journals Comparative Analysis of Severe Accident at WWER-1000 NPP with MELCOR 1.8.5 and 2.1 Code Versions

2020 ◽  
pp. 30-40
Author(s):  
O. Kotsuba ◽  
Yu. Vorobyov ◽  
O. Zhabin ◽  
D. Gumenyuk
Keyword(s):  
Comparative Analysis ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Computer Code ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Bottom Part ◽  
Control Volumes ◽  
Vessel Bottom ◽  
Reactor Pressure

An overview of the main improvements in updated version 2.1 of MELCOR computer code related to more representative mathematical modeling of complex thermohydraulic severe accident processes of core degradation, transfer of molten fragments to the bottom of the reactor, heating and failure of the bottom of the reactor pressure vessel is presented. The elements of WWER-1000 NPP computer model for the MELCOR 1.8.5 (control volumes, thermal structures and structures of the reactor core) that are reproduced for a reactor with the primary side, the secondary side and the containment are described. The changes implemented in WWER-1000 NPP model for MELCOR 1.8.5 to convert it to MELCOR 2.1 version that are mainly related to more detailed modeling of the reactor core and reactor pressure vessel bottom are provided. The paper presents the results of comparative analysis of severe accident scenario of total station blackout at WWER-1000 NPP with MELCOR 1.8.5 and 2.1. The comparison demonstrates good agreement between the main parameters’ results (pressure and temperature in hydraulic elements of the primary, secondary sides and the containment, temperature of core elements, the mass of the generated non-condensed gases and their concentration in the containment) obtained with these code versions for severe accident in-vessel phase. The identified differences in the time of core structures degradation and reactor vessel bottom failure are insignificantly affected by the behavior of the parameters in the primary side and the containment in the in-vessel phase of the severe accident and are related to more detailed modelling of the reactor core and bottom part of the reactor in MELCOR 2.1.

Nodalization of NEK in APROS and Steady State Simulation

2014 ◽  
Author(s):  
Tadeja Polach ◽  
Klemen Debelak ◽  
Ivica Bašić ◽  
Luka Štrubelj
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Building Blocks ◽  
Reactor Pressure Vessel ◽  
Primary Circuit ◽  
Physical Parameters ◽  
Detailed Model ◽  
Reactor Pressure

A model of the primary circuit and part of the secondary circuit of the Slovenian Krško NPP – NEK was built using APROS - Advanced PROcess Simulation environment. The data used to describe the properties of the system modelled in APROS, were the data describing Krško NPP and its operational properties after the uprating and the introduction of the 18-month cycle. Basis for data collections, nodalization, structure and simplifications was NEK RELAP5\MOD3.3 Engineering Handbook and the 23rd cycle. In order to build a model describing all the important parameters, the available elements in APROS environment were used as building blocks for each system. The goal was to create a detailed model nodalization, which would give accurate results and would run on reasonable processing power. Each submodel was checked to verify that the partial results are within the allowable limits and that the description of the physical parameters is consistent with the real components. The model includes reactor pressure vessel, reactor coolant pumps and primary piping, steam generator, part of main steam, part of feedwater, pressurizer and reactor core kinetics. The regulation of pressurizer level and pressure, steam generator level and control rod is also modelled. The model consists of more than 400 thermal hydraulic volumes. The aim of building this model was a through thermal hydraulic analysis of the PWR systems present in the NPP Krško. Several simulations of the steady states at different power levels were performed. The resulting data describing the flow rates in steam generator feedwater, reactor pressure vessel, including bypass flows, heat transfer in reactor core and steam generator, thermal losses to containment, liquid level in pressurizer and steam generator, pressure drops in primary circuit and other parameters were then compared to the results of different types of calculation and to the testing data obtained from Krško NPP. The next step was to identify variations in results and determine whether they are consequence of wrong parameters, measurement deviation or numerical error. In that manner the model was verified and validated (in the sense of comparison with available system surveillance plant test results) to ensure the correct setup, initial and boundary conditions were applied in order to get reliable steady state results.

RELAP Simulation on Boiling and Flow Phenomena Under External Reactor Vessel Cooling Condition

2010 ◽  
Author(s):  
Yongchun Li ◽  
Weihua Zhou ◽  
Yanhua Yang ◽  
Bo Kuang ◽  
Xu Cheng
Keyword(s):  
Pressure Vessel ◽  
Flow Boiling ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Study Program ◽  
External Cooling ◽  
Flow Phenomena ◽  
Key Issues ◽  
Reactor Pressure

External reactor vessel cooling (ERVC) of the In-vessel retention (IVR) system is widely accepted as a feasible way to remove decay heat from the lower head of the reactor pressure vessel (RPV) under severe accident (SA) conditions. However, some issues relating to ERVC still need to be evaluated before its application, such as boiling and flow phenomena and CHF prediction, etc. To study these key issues, an experimental study program named REPEC (Reactor Pressure Vessel External Cooling) is performed at Shanghai Jiao Tong University. Steady state experiments focusing on flow boiling phenomena investigation are carried out with comprehensive measurements, including temperature distribution, pressure drop and mass flow rate. As a part of studies on boiling mechanism and flow phenomena between RPV and the insulation, the experiment is analyzed and simulated with RELAP code. The code simulation covers most of the experimental cases, and a comparison between simulation results and experimental data are presented and 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.

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

2018 ◽  
Vol 4 (4) ◽  
Author(s):  
Jianfeng Mao ◽  
Shiyi Bao ◽  
Zhiming Lu ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Core Material ◽  
Melt Pool ◽  
Crust Layer ◽  
Accident Management ◽  
Core Meltdown ◽  
Reactor Pressure

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.

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

Severe Accident Improvements for CAREM-25 to Arrest Reactor Vessel Meltdown Sequences

2002 ◽  
Author(s):  
L. E. Pomier Ba´ez ◽  
J. E. Nun˜ez Mac Leod ◽  
J. H. Baro´n
Keyword(s):  
Power Plant ◽  
Pressure Vessel ◽  
Transient Behavior ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Element Analysis ◽  
Management Actions ◽  
Core Catcher ◽  
Reactor Pressure

Advanced nuclear reactor designs, such as the CAREM reactor, include several improvements related to safety issues either enhancing the passive safety functions or allowing plant operators more time to undertake different management actions against radioactive releases to the environment. In the development of the nuclear power plant CAREM, the possibility of including a passive metallic in-vessel container in its design is being considered, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and thereof prevent its failure. The paper comprises the analyses, via numerical simulation, for the conceptual design of such a container type. Simulation model characteristics helping to establish geometrical dimensions, materials and container compatibility with power plant engineering features is addressed. The paper also presents the first model developed to analyze the complex relocation phenomena in the core of CAREM during a severe accident sequence caused by a loss of coolant. The PC version of MELCOR 1.8.4 code has been used to predict the transient behavior of core parameters. The finite element analysis (FEA) system ALGOR has been used to evaluate the thermal regime of the reactor pressure vessel wall, when the in-vessel metallic core catcher is present and when it is not present. Two different scenarios have been considered for heat transfer outside the reactor vessel, a pessimistic (dry) and optimistic (wet) conditions in the reactor cavity. This paper presents reactor variables behavior during the first hours of the event being studied, giving preliminary conclusions about the use and capability of a metallic in-vessel core catcher.

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