scholarly journals Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

2012 ◽  
Vol 2012 ◽  
pp. 1-11 ◽  
Author(s):  
Gilberto Espinosa-Paredes ◽  
Raúl Camargo-Camargo ◽  
Alejandro Nuñez-Carrera
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water Injection ◽  
Reactor Core ◽  
Cooling Water ◽  
Fission Products ◽  
Severe Accident ◽  
Core Damage ◽  
Laguna Verde

The loss-of-coolant accident (LOCA) simulation in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP) sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

scholarly journals Antineutrino Flux from the Laguna Verde Nuclear Power Plant

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
Marisol Chavez-Estrada ◽  
Alexis A. Aguilar-Arevalo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Plastic Scintillator ◽  
Reactor Core ◽  
Simulation Code ◽  
Fission Fragments ◽  
Antineutrino Flux ◽  
Laguna Verde ◽  
Reactor Cores

We present a calculation of the antineutrino flux produced by the reactors at the Laguna Verde Nuclear Power Plant in México, based on the antineutrino spectra produced in the decay chains of the fission fragments of the main isotopes in the reactor core, and their fission rates, which have been calculated using the DRAGON simulation code. We also present an estimate of the number of expected events in a detector made of plastic scintillator with a mass of 1 ton, at 100 m from the reactor cores.

Study on the Application of CDAG Methodology in EPR Nuclear Power Plant

2017 ◽  
Author(s):  
Yin Yuhao ◽  
Huang Yichao ◽  
Zhao Feng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fission Product ◽  
Damage Assessment ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Damage ◽  
Calculation Results ◽  
Design Characteristics ◽  
European Nuclear

The Westinghouse Owners Group Core Damage Assessment Guidance (CDAG), which has been authorized by the NRC staffs, is now used by licensee emergency response organization staff for estimating the extent of core damage that may have occurred during an accident at a Westinghouse nuclear power plant. On the other hand, EPR is a 3rd generation nuclear power plant, which applies the advanced European nuclear power technology. This paper introduced Core Damage Assessment Guidance methodology in detail. The CDAG methodology is then attempted to apply to the EPR nuclear power plant. Detailed calculations have been performed for the setpoints of containment radiation monitors (CRM) and core exit thermocouples (CETs) with EPR design characteristics, which are the two main methods for estimation core damage amount. This paper also focuses the discussion on the reasons of difference of core damage estimating results between CRM method and CETs method; based on the discussion, several advices are provided when the two methods show a reasonable discrepancy in conclusions. Several conclusions can be made from the discussions in this article that 1)the Westinghouse Owners Group CDAG methodology proved to be reasonable when applied to EPR power plant for core damage assessment under severe accident; 2) the CDAG methodology which reflect the latest understanding of fission product behavior, is very simple and timely for core damage assessment based on NPP (nuclear power plant) real-time parameters; 3) conservative calculation results of setpoints on CRM and CETs based on EPR design show a reasonable trend and range; 4) it is concluded that several factors such as the releasing way, RCS fission product retention, fuel burnups might have great impact on the estimating results, when the results from two main indications (CRM and CETs) show an unexpected response.

Development, Application, and Implementation of RAMCAP to Characterize Nuclear Power Plant Risk From Terrorism

2006 ◽  
Author(s):  
John P. Gaertner ◽  
Grant A. Teagarden
Keyword(s):  
Decision Making ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Reactor Core ◽  
Severe Accident ◽  
Spent Fuel ◽  
Financial Loss ◽  
Worst Case ◽  
Management Procedures

In response to increased interest in risk-informed decision making regarding terrorism, EPRI and ERIN Engineering were selected by U.S. DHS and ASME to develop and demonstrate the RAMCAP method for nuclear power plant (NPP) risk assessment. The objective is to characterize plant-specific NPP risk for risk management opportunities and to provide consistent information for DHS decision making. This paper is an update of this project presented at the American Nuclear Society (ANS) International Topical Meeting on Probabilistic Safety Analysis (PSA05) in September, 2005. The method uses a characterization of risk as a function of Consequence, Vulnerability, and Threat. For each site, worst case scenarios are developed for each of sixteen benchmark threats. Nuclear RAMCAP hypothesizes that the intent of the perpetrator is to cause offsite radiological consequences. Specific targets are the reactor core, the spent fuel pool, and nuclear spent fuel in a dry storage facility (ISFSI). Results for each scenario are presented as conditional risk for financial loss, early fatalities and early injuries. Expected consequences for each scenario are quantified, while vulnerability is estimated on a relative likelihood scale. Insights for other societal risks are provided. Although threat frequencies are not provided, target attractiveness and threat deterrence are estimated. To assure efficiency, completeness, and consistency; results are documented using standard RAMCAP Evaluator software. Trial applications were successfully performed at four plant sites. Implementation at all other U.S. commercial sites is underway, supported by the Nuclear Sector Coordinating Council (NSCC). Insights from RAMCAP results at 23 U.S. plants completed to date have been compiled and presented to the NSCC. Results are site-specific. Physical security barriers, an armed security force, preparedness for design-basis threats, rugged design against natural hazards, multiple barriers between fuel and environment, accident mitigation capability, severe accident management procedures, and offsite emergency plans are risk-beneficial against all threat types.

scholarly journals Online Simulation of Nuclear Power Plant Primary Systems

2020 ◽  
Vol 2020 ◽  
pp. 1-9
Author(s):  
Hongyun Xie ◽  
Haixia Gu ◽  
Chao Lu ◽  
Jialin Ping
Keyword(s):  
Information System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Real Time ◽  
Nuclear Power ◽  
Measurement Data ◽  
Reactor Core ◽  
Power Industry ◽  
Nuclear Power Industry ◽  
Online Simulation

Real-time Simulation (RTS) has long been used in the nuclear power industry for operator training and engineering purposes. And, online simulation (OLS) is based on RTS and with connection to the plant information system to acquire the measurement data in real time for calibrating the simulation models and following plant operation, for the purpose of analyzing plant events and providing indicative signs of malfunctioning. OLS has been applied in certain industries to improve safety and efficiency. However, it is new to the nuclear power industry. A research project was initiated to implement OLS to assist operators in certain critical nuclear power plant (NPP) operations to avoid faulty conditions. OLS models were developed to simulate the reactor core physics and reactor/steam generator thermal hydraulics in real time, with boundary conditions acquired from plant information system, synchronized in real time. The OLS models then were running in parallel with recorded plant events to validate the models, and the results are presented.

Electrochemical decontamination of irradiated nuclear graphite from corrosion and fission products using molten salt

2021 ◽  
Author(s):  
Tatiana Grebennikova ◽  
Abbie N Jones ◽  
Clint Alan Sharrad
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Waste Management ◽  
Molten Salt ◽  
Nuclear Power ◽  
Fission Products ◽  
Nuclear Graphite ◽  
Irradiated Graphite ◽  
The World ◽  
The Uk

Irradiated graphite waste management is one of the major challenges of nuclear power-plant decommissioning throughout the world and significantly in the UK, France and Russia where over 85 reactors employed...

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

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