A Method for Online Calibration of MAAP Simulations in a Severe Accident Management System Database

2017 ◽  
Author(s):  
Weifeng Xu ◽  
Fangqing Yang ◽  
Peng Chen ◽  
Yehong Liao
Keyword(s):  
Management System ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Outlet Temperature ◽  
Online Calibration ◽  
Support Plate ◽  
Initial Event ◽  
Accident Management ◽  
Accident Simulation ◽  
Plate Failure

During a nuclear plant accident, five accident events are usually considered, including core uncovery, core outlet temperature arrived at 650 °C, core support plate failure, reactor vessel failure and containment failure. In accident emergency aspect, when an accident happens, the initial event can be utilized in the severe accident management system which is based on MAAP to simulate the long process of the accident, so as to provide support for operators to take actions. However, in MAAP, many sensitivity parameters exist, which reflect phenomenological uncertainty or models uncertainty and will influence the happening time of the five accident events above. In this paper, based on MAAP5 and LOCAs, the CPR1000 is simulated to analyze the influences of MAAP5’s sensitivity parameters reflecting phenomenological uncertainty on the accident process, which is aimed to find out the sensitivity parameters associated to the five important accident events and build the database between these sensitivity parameters and five accident events’ happening time. Then, based on the research above, a preliminary approach to optimize the MAAP5’s accidents simulation is introduced, which is realized by adjusting sensitivity parameters. Finally, the application of this research will be showed in a severe accident management system developed by us. The research results offer great reference significance for the severe accident simulation and prediction in MAAP5.

Introduction of Three Methods Used for the Nuclear Accident Diagnosis

2017 ◽  
Author(s):  
Peng Chen ◽  
Weifeng Xu ◽  
Fangqing Yang ◽  
Yehong Liao
Keyword(s):  
Management System ◽  
Nuclear Power ◽  
Nuclear Accident ◽  
Severe Accident ◽  
Severe Accidents ◽  
Accident Management ◽  
Accident Simulation ◽  
Diagnosis Method ◽  
Nuclear Severe Accidents ◽  
Diagnosis Technique

In order to deal with the nuclear severe accidents, the severe accident management systems are popularly considered and developed at home and abroad recently. A severe accident management system usually includes these functional parts: accident monitor, accident diagnosis, accident simulation, accident prognosis and SAMG support. Here, the accident diagnosis part is mainly concerned, and three nuclear accident diagnosis methods are introduced here, including BP neural network method, SDG expert diagnosis technique and artificial diagnosis method, which are also applied to a severe accident management system developed by us. In this paper, firstly, the severe accident management system developed by us will be introduced briefly. Then, three accident diagnosis methods for nuclear power plant (NPP) are showed and described in detail. At last, two cases including LOCA and SGTR accidents are used for the verification of these accident diagnosis methods and some analyzing results and conclusions are given. The results show that the three diagnosis methods are very useful for the accident diagnosis of NPP, which can diagnose the accident type accurately and offer much information or support to the severe accident management system and operators. The paper offers some reference significance for the research of accident diagnosis methods and the development of severe accident management system.

External Cooling: The SWR 1000 Severe Accident Management Strategy

2004 ◽  
Author(s):  
Nikolay Ivanov Kolev
Keyword(s):  
Management Strategy ◽  
External Pressure ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Core Flooding ◽  
External Cooling ◽  
Structural Resistance ◽  
Safety Margins ◽  
Accident Management ◽  
Management Concept

This paper provides the description of the basics behind design features for the severe accident management strategy of the SWR 1000. The hydrogen detonation/deflagration problem is avoided by containment inertization. In-vessel retention of molten core debris via water cooling of the external surface of the reactor vessel is the severe accident management concept of the SWR 1000 passive plant. During postulated bounding severe accidents, the accident management strategy is to flood the reactor cavity with Core Flooding Pool water and to submerge the reactor vessel, thus preventing vessel failure in the SWR 1000. Considerable safety margins have been determined by using state of the art experiment and analysis: regarding (a) strength of the vessel during the melt relocation and its interaction with water; (b) the heat flux at the external vessel wall; (c) the structural resistance of the hot structures during the long term period. Ex-vessel events are prevented by preserving the integrity of the vessel and its penetrations and by assuring positive external pressure at the predominant part of the external vessel in the region of the molten corium pool.

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.

ADAM: An Accident Diagnostic, Analysis and Management System — Applications to Severe Accident Simulation and Management

2002 ◽  
Author(s):  
M. J. Zavisca ◽  
M. Khatib-Rahbar ◽  
H. Esmaili ◽  
R. Schulz
Keyword(s):  
Real Time ◽  
Water Injection ◽  
Management Training ◽  
Computer Applications ◽  
Severe Accident ◽  
Diagnostic Analysis ◽  
Gas Generation ◽  
Accident Management ◽  
Accident Simulation ◽  
On Line

The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM’s severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of “alarms” are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management.

scholarly journals BWR/5 Pressure-Suppression Pool Response during an SBO

2013 ◽  
Vol 2013 ◽  
pp. 1-8
Author(s):  
Javier Ortiz-Villafuerte ◽  
Andrés Rodríguez-Hernández ◽  
Enrique Araiza-Martínez ◽  
Luis Fuentes-Márquez ◽  
Jorge Viais-Juárez
Keyword(s):  
Pressure Rise ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Power Station ◽  
The Core ◽  
Initial Event ◽  
Station Blackout ◽  
Suppression Pool ◽  
Failure Temperature ◽  
Design Pressure

RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

In-Vessel Retention of Molten Core Debris for CAP1400

2010 ◽  
Author(s):  
Junrong Wang ◽  
Huajian Chang ◽  
Wenxiang Zheng ◽  
Zhiwei Zhou
Keyword(s):  
Management Strategy ◽  
Reactor Vessel ◽  
Chinese Version ◽  
Severe Accident ◽  
Paper Briefly ◽  
Accident Management ◽  
Containment Integrity

In-vessel retention (IVR) of core melt through external reactor vessel cooling (ERVC) is a key severe accident management strategy to ensure that the vessel head remains intact and eliminate consequent major threats to containment integrity. To maintain the margin against the failure of the reactor vessel and make sure of the feasibility of IVR for its role to confine the molten corium with 1400MW power, CAP1400, Chinese version of large passive PWR, systematic investigations including experimental and analytical researches of IVR are very important to the development of CAP1400. This paper briefly reviews the progress and tasks of a four-year project which was planned to analyze, evaluate, improve and validate the effectiveness of IVR employed in the CAP1400.

A Novel Point Estimate Procedure for IVR Calculations in Core-Molten Severe Accident

2010 ◽  
Author(s):  
Yapei Zhang ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Wenxi Tian
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Safety Margin ◽  
Thermal Response ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Point Estimate ◽  
Accident Management ◽  
Water Reactors ◽  
Estimate Procedure

In-Vessel Retention (IVR) of core melt is a key severe accident management strategy adopted by operating nuclear power plants and advanced light water reactors (ALWRs), AP600, AP1000 etc. External Reactor Vessel Cooling (ERVC), which involves flooding the reactor cavity to submerge the reactor vessel in an attempt to cool core debris relocated to the vessel low head, is a novel severe accident management for IVR analysis. In present study, IVR analysis code in severe accident (IVRASA) has been proposed to evaluate the safety margin of IVR in AP600 with anticipative depressurization and reactor cavity flooding in severe accident. For, IVRASA, the point estimate procedure has been developed for modeling the steady-state endpoint of core melt configurations. Furthermore, IVRASA was developed in a more general fashion so that it is applicable to compute various molten configurations such as UCSB FInal Bounding State (FIBS) etc. The results by IVRASA were consistent with those of the UCSB and INEEL. Benchmark calculations of UCSB-assumed FIBS indicate the applicability and accuracy of IVRASA and it could be applied to predict the thermal response of various molten configurations.

MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

2014 ◽  
Author(s):  
Pavlin P. Groudev ◽  
Antoaneta E. Stefanova ◽  
Petya I. Vryashkova
Keyword(s):  
Parametric Study ◽  
Cold Water ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Fuel Cladding ◽  
Accident Management ◽  
Management Guidance ◽  
Fuel Behavior ◽  
Accident Conditions

This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed. The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident. The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.

Preliminary Analysis of Effect of the Intentional Depressurization on Fission Product Behavior During TMLB’ Severe Accident

2009 ◽  
Author(s):  
Gaofeng Huang ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Fission Product ◽  
Reactor Vessel ◽  
Inert Gas ◽  
Severe Accident ◽  
Base Case ◽  
Reactor Coolant ◽  
Accident Management ◽  
Direct Containment Heating ◽  
Accident Sequences ◽  
Coolant System

It has been identified that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its effects. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB’ base case and opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomena of revaporization is strong in the RCS.

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