scholarly journals ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

2018 ◽  
Vol 2018 ◽  
pp. 1-24
Author(s):  
J. C. de la Rosa Blul ◽  
S. Brumm ◽  
F. Mascari ◽  
S. J. Lee ◽  
L. Carenini
Keyword(s):  
Failure Time ◽  
Hydrogen Generation ◽  
Severe Accident ◽  
Primary System ◽  
Similar Amount ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Expansion Model ◽  
Gas Expansion ◽  
Comparison Exercise

A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.

Comparative Study of Loss-of-Coolant Accident Using MAAP4.03 and RELAP5/MOD3.2.2

2002 ◽  
Author(s):  
Chang Hwan Park ◽  
Doo Yong Lee ◽  
Ik Jeong ◽  
Un Chul Lee ◽  
Kune Y. Suh ◽  
...  
Keyword(s):  
Flow Model ◽  
Correction Term ◽  
Sensitivity Study ◽  
Severe Accident ◽  
Primary System ◽  
Two Phase ◽  
Flow Rates ◽  
Flow Models ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.

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.

scholarly journals Analysis of Containment Pressure and Temperature Changes Following Loss of Coolant Accident (LOCA)

2015 ◽  
Vol 5 (4) ◽  
pp. 1-8
Author(s):  
Van Thai Nguyen ◽  
Ngoc Dung Kieu
Keyword(s):  
High Pressure ◽  
Cooling System ◽  
Primary System ◽  
Similar Work ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Water ◽  
Main Steam Line ◽  
Main Steam ◽  
Work Done

This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

2017 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhang Zhijian ◽  
Tian Zhao-fei ◽  
A. M. Mustafa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Simulation ◽  
Primary System ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

Validation Status of MAAP5 Core Melt Progression Model

2014 ◽  
Author(s):  
Chan Y. Paik ◽  
Paul McMinn ◽  
Christopher Henry ◽  
Wison Luangdilok
Keyword(s):  
Hydrogen Generation ◽  
Computer Code ◽  
Severe Accident ◽  
Accident Analysis ◽  
Primary System ◽  
Analysis Program ◽  
Fuel Rod ◽  
Progression Model ◽  
Hydrogen Mass ◽  
Quench Test

The Modular Accident Analysis Program (MAAP) is a computer code that is used for integrated severe accident analysis. The latest MAAP5 version was validated against the PHEBUS-PF FPT0 and FPT1 tests performed at CEA/IPSN, PBF-SFD Test 1–4 performed at INEL, and QUENCH Test 06 performed at FZK. PHEBUS FPT0, PHEBUS FPT1, and PBF-SFD Test 1–4 are in-pile experiments where a test bundle was housed in the center of a reactor. Comparisons were focused on fuel and shroud temperature histories, and hydrogen generation histories. For the PHEBUS tests, primary system and containment responses were also compared. In general, fuel and shroud temperatures are well predicted by MAAP5. Overall hydrogen mass generated is also well predicted except that MAAP5 over-predicts the total hydrogen mass generated for the FPT1 test. The hydrogen generation at the time of the peak oxidation phase for the FPT0 test is under-predicted while the hydrogen generation for the FPT1 test is over-predicted. In general, MAAP tends to over-predict mass relocations from the upper part of the bundle due to fuel rod collapses by the end of separate effects test transients.

Numerical Simulation of the Behavior of Fission Products in the Primary Circuit of the VVER During the LOCA Severe Accident

2009 ◽  
Author(s):  
S. V. Tsaun ◽  
V. V. Bezlepkin ◽  
A. E. Kiselev ◽  
I. A. Potapov ◽  
V. F. Strizhov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Fission Products ◽  
Severe Accident ◽  
Primary Circuit ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Loss Of Coolant ◽  
Mass Activity

The methods and models for the analysis of the radiological consequences of the design basis and severe accidents in a Nuclear Power Plant (NPP) are presented in this paper when using the system code SOCRAT. The system code SOCRAT/V3 was elaborated for a realistic analysis of radiological consequences of severe accidents in a NPP. The following models of the fission products (FP) behavior are included into the code SOCRAT/V3: (i) the condensation and the evaporation of the FP in the gaseous phase and (ii) the sedimentation, the evaporation, and the coagulation of the aerosol-shape FP. The latter processes are governed by gravity, Brownian and turbulent diffusion, thermophoresis, turbophoresis and so forth. The behavior of the FP during the loss-of-coolant accidents (LOCA) is presented to demonstrate the possibilities of the code SOCRAT/V3. The main stages of the accident (the core dryout, the core reflooding, the core degradation, the hydrogen generation, the FP release, etc.) are described. Corresponding estimations of the mass, activity, and decay heat of the suspended, settled and released into containment the FP (Xe, Te, Cs, CsI, Cs2MoO4, and so forth) are represent as well.

The Development and Application of SCDAP-3D©

2002 ◽  
Author(s):  
E. W. Coryell ◽  
E. A. Harvego ◽  
L. J. Siefken
Keyword(s):  
Computer Code ◽  
Alternative Fuel ◽  
Severe Accident ◽  
Initial Development ◽  
Standard Problem ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Environmental Laboratory ◽  
Reactor Accidents ◽  
Thermal Hydraulic Analysis

The SCDAP-3D© computer code (Coryell 2001) has been developed at the Idaho National Engineering & Environmental Laboratory (INEEL) for the analysis of severe reactor accidents. A prominent feature of SCDAP-3D© relative to other versions of the code is its linkage to the state-of-the-art thermal/hydraulic analysis capabilities of RELAP5-3D©. Enhancements to the severe accident models include the ability to simulate high burnup and alternative fuel, as well as modifications to support advanced reactor analyses, such as those described by the Department of Energy’s Generation IV (GenIV) initiative. Initial development of SCDAP-3D© is complete and two widely varying but successful applications of the code are summarized. The first application is to large break loss of coolant accident analysis performed for a reactor with alternative fuel, and the second is a calculation of International Standard Problem 45 (ISP-45) or the QUENCH 6 experiment.

scholarly journals Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept

2015 ◽  
Vol 2015 ◽  
pp. 1-9
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Severe Accident ◽  
Fusion Reactors ◽  
Transfer System ◽  
Light Water Reactors ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Peak ◽  
Water Reactors ◽  
Heat Transfer System ◽  
Do So

As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters.

Analysis of Possibility of Pressure Tube Cold Pressurization During ECCS Injection for a Small Break LOCA

2002 ◽  
Vol 124 (4) ◽  
pp. 483-486 ◽  
Author(s):  
D. Mukhopadhyay ◽  
S. K. Gupta ◽  
V. Venkat Raj
Keyword(s):  
Control System ◽  
System Design ◽  
Heavy Water ◽  
Pressure Control ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Control System ◽  
Safe Limits ◽  
Water Reactors

ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.

scholarly journals Study on Airborne Radionuclide Dispersion in Floating Nuclear Power Plant under the Loss-of-Coolant Accident

2021 ◽  
Vol 2021 ◽  
pp. 1-8
Author(s):  
Shuliang Zou ◽  
Na Liu ◽  
Binhai Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Effective Dose ◽  
Nuclear Power ◽  
Severe Accident ◽  
Emergency Plan ◽  
Dose Limit ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Dynamics Method

Floating nuclear power plant is a kind of nuclear power plant on a barge moored specifically in an area of the sea. In order to study the factors influencing airborne radionuclide dispersion induced by the loss-of-coolant accident in floating nuclear power plant, the floating nuclear power plant platform was taken as the research object, and the dispersion of airborne radionuclide under combined conditions of platform positions, wind directions, and break directions (north, south, west, and east) was simulated by the CFD (computational fluid dynamics) method. The results show that northern and southern breaks have less dangerous island area than western and eastern ones but have more platform dangerous area than the western and eastern ones. The risk of the southern break is the greatest, and that of the western break is the least. Rotating the floating nuclear power plant platform in a certain angle can reduce the damage of loss-of-coolant accident. The effects of the dose received by the personnel under the condition of the severe accident were evaluated based on previous research, showing that the inhalation effective dose and the effective dose of plume immersion exposure were less than the radiation dose limit of 0.25 Sv within two hours in the accident. The results of the study can provide reference for the design of floating nuclear power plant platform and the formulation of emergency plan.

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