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.

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.

Simulation of the TMI-2 Accident With ATHLET-CD and Analysis of the Late-Phase Modeling

2012 ◽  
Author(s):  
Mathias Hoffmann ◽  
Marco K. Koch
Keyword(s):  
Hydrogen Generation ◽  
Late Phase ◽  
Current Version ◽  
Molten Material ◽  
Plant Behavior ◽  
Accident Data ◽  
Core Components ◽  
Molten Materials ◽  
Two Phases ◽  
Reactor Pressure

This paper provides the results of a simulation of the TMI-2 accident with the current version of ATHLET-CD Mod. 2.2A as part of code validation activities at Ruhr-Universität Bochum (RUB). The calculated plant behavior during the first four phases of the accident is discussed and analyzed in comparison to available post-accident data and measurements. The calculation captures the plant response in terms of the thermal-hydraulics very well during the first two phases. However, during the reflooding of the degraded core some discrepancies between the calculation and TMI-2 data are identified. The code basically underestimates the hydrogen generation in this phase. Moreover, the debris bed and molten pool behavior during this phase cannot be simulated yet. An essential limitation of the current capabilities of the code in terms of the late-phase is the lack of models addressing the relocation of molten materials to the lower plenum of the reactor pressure vessel. Based on this analysis, the next steps needed to model the relocation of molten core components to the lower plenum are identified. These are the lateral leveling of accumulated molten material inside molten pools as well as the slumping to the lower plenum via different paths.

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 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.

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.

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