Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Hiroshi Madokoro ◽  
Alexei Miassoedov ◽  
Thomas Schulenberg
Keyword(s):  
Structural Analysis ◽  
Message Passing ◽  
Message Passing Interface ◽  
Severe Accident ◽  
Coupled Analysis ◽  
Accident Analysis ◽  
Analysis Tool ◽  
Lower Head ◽  
Calculation Results ◽  
Reactor Pressure

Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

The Development of Candling Module Code in Module In-vessel Degraded Analysis Code MIDAC and the Relevant Calculation for CPR1000 During Large-Break LOCA

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Jun Wang ◽  
Yuqiao Fan ◽  
Yapei Zhang ◽  
Xinghe Ni ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Mass Distribution ◽  
Pressure Vessel ◽  
Hydrogen Generation ◽  
Peak Temperature ◽  
Severe Accident ◽  
Molten Material ◽  
Loss Of Coolant ◽  
Calculation Results ◽  
Relevant Calculation ◽  
Reactor Pressure

The occurrence of Fukushima has increased the focus on the development of severe accident codes and their applications. As a part of Chinese “National Major Projects,” a module in-vessel degraded analysis code (MIDAC) is currently being developed at Xi’an Jiaotong University. The developing situation of a candling module and relevant calculation for CPR1000 for large break loss of coolant analysis (LOCA) are presented in this paper. The candling module focuses on the melting, moving, and relocation of the melting core materials and necessary thermal hydraulic information. MIDAC’s LOCA accident calculation results of Chinese pressure reactor 1000 (CPR1000) cover the melting mass distribution, peak temperature, and hydrogen generation. The results have been compared with MAAP. Through comparison, the candling module of MIDAC proved to be able to predict the moving trend of the molten material mass relocation in the reactor pressure vessel.

Preliminary Severe Accident Analysis of Fusion-Fission Hybrid Reactor Using the MELCOR Code

2014 ◽  
Author(s):  
Zhang Dabin ◽  
Zhiwei Zhou ◽  
Heng Xie ◽  
Tang Yang
Keyword(s):  
Cooling System ◽  
Safety Performance ◽  
Severe Accident ◽  
Hybrid Reactor ◽  
Accident Analysis ◽  
Analysis Model ◽  
Decay Heat ◽  
Calculation Results ◽  
Core Meltdown ◽  
Dry Out

The fusion-fission hybrid conceptual reactor is a proposed means of generating power, which adopts a water cooled fission blanket based on ITER. Due to the water cooled fission blanket, safety performance of the hybrid reactor should be considered, including decay heat remove, core uncovered, core meltdown, core degradation, radioactivity releases, etc., similar with the PWRs. The main goal of this work is to develop the fission blanket model by using MELCOR code, and to evaluate the safety performance under severe accidents preliminarily. Based on MELCOR 1.8.5, some modification is made for the severe accident analysis of fission blanket. Using modified MELCOR code, an analysis model is developed for the fission blanket and the cooling loop. The strategy of the In-Vessel Retention (IVR) using the ex-vessel cooling method is evaluated during a large break LOCA. The calculation results describes the main phenomena during the severe accident progression, including core dry out, meltdown, relocation, etc.. Simulation result is shown that the decay heat in the fission zone can be removed out by the ex-vessel cooling system merely, and the fuel max temperature will not reach the melting point.

CORBA AND MPI-BASED “BACKBONE” FOR COUPLING ADVANCED SIMULATION TOOLS

2014 ◽  
Vol 3 (2) ◽  
pp. 83-90 ◽  
Author(s):  
M. Seydaliev ◽  
D. Caswell
Keyword(s):  
Fission Product ◽  
Computer Simulations ◽  
Message Passing ◽  
Message Passing Interface ◽  
Nuclear Reactor ◽  
Atmospheric Dispersion ◽  
Safety Analysis ◽  
Severe Accident ◽  
Nuclear Reactor Safety ◽  
Exchange Data

There is a growing international interest in using coupled, multidisciplinary computer simulations for a variety of purposes, including nuclear reactor safety analysis. Reactor behaviour can be modeled using a suite of computer programs simulating phenomena or predicting parameters that can be categorized into disciplines such as Thermalhydraulics, Neutronics, Fuel, Fuel Channels, Fission Product Release and Transport, Containment and Atmospheric Dispersion, and Severe Accident Analysis. Traditionally, simulations used for safety analysis individually addressed only the behaviour within a single discipline, based upon static input data from other simulation programs. The limitation of using a suite of stand-alone simulations is that phenomenological interdependencies or temporal feedback between the parameters calculated within individual simulations cannot be adequately captured. To remove this shortcoming, multiple computer simulations for different disciplines must exchange data during runtime to address these interdependencies. This article describes the concept of a new framework, which we refer to as the “Backbone,” to provide the necessary runtime exchange of data. The Backbone, currently under development at AECL for a preliminary feasibility study, is a hybrid design using features taken from the Common Object Request Broker Architecture (CORBA), a standard defined by the Object Management Group, and the Message Passing Interface (MPI), a standard developed by a group of researchers from academia and industry. Both have well-tested and efficient implementations, including some that are freely available under the GNU public licenses. The CORBA component enables individual programs written in different languages and running on different platforms within a network to exchange data with each other, thus behaving like a single application. MPI provides the process-to-process intercommunication between these programs. This paper outlines the different CORBA and MPI configurations examined to date, as well as the preliminary configuration selected for coupling 2 existing safety analysis programs used for modeling thermal–mechanical fuel behavior and fission product behavior respectively. In addition, preliminary work in hosting both the Backbone and the associated safety analysis programs in a cluster environment are discussed.

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).

scholarly journals A thermal structural analysis tool for RPV lower head behavior during severe accidents with core melt

2018 ◽  
Vol 4 (0) ◽  
pp. 18-00038-18-00038 ◽  
Author(s):  
Hiroshi MADOKORO ◽  
Alexei MIASSOEDOV ◽  
Thomas SCHULENBERG
Keyword(s):  
Structural Analysis ◽  
Analysis Tool ◽  
Severe Accidents ◽  
Lower Head

Comparison of CORA and MELCOR Core Degradation Simulation and MELCOR Oxidation Model

2014 ◽  
Author(s):  
Jun Wang ◽  
Michael L. Corradini ◽  
Wen Fu ◽  
Troy Haskin ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Test Data ◽  
Flow Rate ◽  
Hydrogen Generation ◽  
Severe Accident ◽  
Generation Rate ◽  
Accident Analysis ◽  
University Of Wisconsin ◽  
Calculation Results ◽  
Oxidation Model ◽  
The University

MELCOR is widely used and sufficiently trusted for severe accident analysis. However, the occurrence of Fukushima has increased the focus on severe accident codes and their use. A MELCOR core degradation calculation was conducted at the University of Wisconsin – Madison. The calculation results were checked by comparing with a past CORA experiment. MELCOR calculation results included the flow rate of argon and steam, the generation rate of hydrogen. Through this work, the performance of MELCOR COR package was reviewed in detail. This paper compares the hydrogen generation rates predicted by MELCOR to the CORA test data. While agreement is reasonable it could be improved. Additionally, the MELCOR zirconium oxidation model was analyzed.

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.

Creep Evaluation for a PWR Reactor Pressure Vessel Lower Head Under Severe Accident Conditions Considering Sustained Internal Pressure

2018 ◽  
Author(s):  
YongJian Gao ◽  
Ming Cao ◽  
YinBiao He
Keyword(s):  
Internal Pressure ◽  
Failure Modes ◽  
Structural Integrity ◽  
Creep Damage ◽  
Pressure Vessels ◽  
Mitigation Strategies ◽  
Field Analysis ◽  
Severe Accident ◽  
Lower Head ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of appropriate severe accident mitigation strategies for AP1000 Nuclear Power Plant (NPP), and assurance of prevention against to thermal failure and structural failure of Reactor Pressure Vessels (RPV) is the prerequisite of IVR. A Finite Element Model fora RPV considering lower head melting was established, the creep calculation was carried out after the temperature field analysis, and the stress-strain responses for different times were obtained. By means of choosing representative evaluation sections and applying the Accumulative Damage Theory based on Larson-Miller Parameter, the Creep Damage calculations and evaluations were conducted. The results showed that the failure modes associated with creep rupture would not happen under IVR condition when a certain amount of internal pressure sustained. The approaches employed in this paper could be utilized in structural integrity evaluation of RPV under IVR for other new type NPPs.

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.

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