Improvement of the RELAP5-3D Model of Condensation in the Presence of Noncondensables

Vapor Density ◽

Pressurized Water ◽

Plant Behavior ◽

University Of Wisconsin ◽

Accident Scenarios ◽

Saturation Vapor

Condensation of steam on the primary side of a steam generator in a pressurized water reactor (PWR) is one means of removing decay heat during some accident scenarios, including small break loss of coolant accident (SBLOCA). However, when noncondensable gasses mix with steam, it impairs condensation. To correctly predict plant behavior, nuclear power plant (NPP) safety analysis codes such as RELAP5-3D must model the effect of condensation in the presence of noncondensables properly. A potential error in the RELAP5-3D code was reported in the condensation model in the presence of noncondensables by the University of Wisconsin[1]. The report indicated that the calculated condensation heat flux was under-predicted due to the modeling of the mass transfer in the gas mixture. The original documentation describing the implementation of the model was reviewed and compared with alternative formulations. An alternative that uses saturation vapor density at temperature of total pressure instead of the saturation vapor density at vapor partial pressure for calculating vapor mass flux was implemented. Comparison of the alternative method with the original against experimental data for several test cases showed improvement for most test cases.

Improvement of the RELAP5-3D Condensation Heat Transfer Model in the Presence of Noncondensable Gases

Journal of Heat Transfer ◽

10.1115/1.4047048 ◽

2020 ◽

Vol 142 (8) ◽

Author(s):

Nolan Anderson ◽

Piyush Sabharwall

Keyword(s):

Heat Removal ◽

Correct Prediction ◽

Transfer Model ◽

Pressurized Water ◽

Plant Behavior ◽

Decay Heat ◽

Noncondensable Gases ◽

Accident Scenarios

Abstract Condensation of steam on the primary side of steam generator in a pressurized water reactor is one of the means of removing decay heat during accident scenarios such as a small break loss of coolant accident. With the presence of noncondensable gases, the rate of removal of decay heat reduces, affecting the ability of the nuclear plant to remove heat in accident scenarios. Therefore, correct prediction of heat removal capability is very significant to predict the plant behavior. In this study, an analytical model is compared with a numerical solution with the use of experiments performed at University of California, Berkeley and at MIT. A modified correlation is proposed and compared to experimental observation for various noncondensable gases.

Simulation of Loss of Coolant Accident in an Integral Pressurized Water Reactor (IPWR)

Advanced Materials Research ◽

10.4028/www.scientific.net/amr.230-232.410 ◽

2011 ◽

Vol 230-232 ◽

pp. 410-414

Author(s):

Salah Ud Din Khan ◽

Min Jun Peng ◽

Muhammad Zubair

Keyword(s):

Nuclear Power ◽

Zirconium Hydride ◽

Nuclear Power Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Sequence Of Events ◽

Loss Of Coolant ◽

Safety And Reliability

In this paper research has been carried out on the Loss of Coolant Accident (LOCA) in an Integral Pressurized Water Reactor(IPWR) by using thermal hydraulic system code Relap5/Mod3.4.The designing of Integrated Pressurized Water Reactor (IPWR) incorporates the safety and reliability of the reactor to withstand under accidental vulnerabilities. In this study, the reactor under consideration is Uranium Zirconium Hydride Nuclear Power Reactor INSURE-100 with the power output of 100MW.In the current research, the reactor has been described in detail according to the requirement for the simulation of LOCA using Relap5 code with the possibility of occurrence of the time sequence of events. The graphs obtained shows good agreement for the safe operation of IPWR under LOCA.

Operating Procedures: Do they Reduce Operator Errors?

Proceedings of the Human Factors and Ergonomics Society Annual Meeting ◽

10.1177/154193129403800404 ◽

1994 ◽

Vol 38 (4) ◽

pp. 205-209 ◽

Cited By ~ 2

Author(s):

Sharolyn A. Converse

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Human Error ◽

Error Rates ◽

Pressurized Water ◽

Computerized Procedures ◽

Accident Scenarios ◽

Accident Conditions ◽

Subjective Estimates

Computerized operating procedures have been suggested as a mechanism for reducing human error in nuclear power plants. The Computerized Procedures Manual (COPMA-II) is an electronic procedure system that can be used to execute procedures, to track progress through plant procedures, and to automatically monitor plant parameters. To evaluate the effectiveness of COPMA-II, eight teams of two licensed reactor operators operated a scaled pressurized water reactor under normal and accident conditions, using both COPMA-II and traditional paper procedures. Error rates, times to initiate procedures, times to complete procedures, and subjective estimates of workload were collected for each scenario. The most interesting finding of the study was that, for one accident scenario, performance with COPMA-II was twice as accurate as performance with paper procedures. However, operators initiated responses to both accident scenarios fastest with paper procedures. Procedure type did not moderate time to complete procedures.

Functional Information of System Components Influenced by Counteractions on Computer-Based Procedure

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4040366 ◽

2018 ◽

Vol 4 (4) ◽

Author(s):

Tulis Jojok Suryono ◽

Akio Gofuku

Keyword(s):

Nuclear Power ◽

Radioactive Material ◽

Functional Information ◽

Pressurized Water ◽

Plant Behavior ◽

Computer Based ◽

The Impact ◽

Multilevel Flow Modeling ◽

Emergency Operating

In an emergency condition of nuclear power plant, operators have to mitigate the accident in order to remove the decay heat and to prevent the release of radioactive material to the environment following the emergency operating procedures (EOPs). The action of operators on a component, for example, changing the parameter level of a component, which is described in a procedure step, will impact other components of the plant and the plant behavior. Nowadays, the advanced main control rooms have been equipped with the computer-based procedures (CBPs) which provide some features and benefits which are not available in paper-based procedures. However, most of CBPs do not provide information of the impact of the counteractions on each procedure step (components influenced and future plant behavior) although it is useful for operators to understand the purpose of the procedure steps before making decisions and taking the actions. This paper discusses the functional information and the method to generate the information using multilevel flow modeling (MFM) model of operator actions on some procedure steps of a simplified EOP of pressurized water reactor (PWR) plant, as an example.

Analyses of loads on reactor pressure vessel internals in a pressurized water reactor due to a loss-of-coolant accident considering fluid-structure interaction

Kerntechnik ◽

10.3139/124.110123 ◽

2010 ◽

Vol 75 (6) ◽

pp. 311-315

Author(s):

P. Akimov ◽

L. Obereisenbuchner

Keyword(s):

Pressure Vessel ◽

Fluid Structure Interaction ◽

Reactor Pressure Vessel ◽

Pressurized Water ◽

Water Reactor ◽

Structure Interaction ◽

Loss Of Coolant ◽

Reactor Pressure

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

Progress in Nuclear Energy ◽

10.1016/j.pnucene.2021.103648 ◽

2021 ◽

Vol 134 ◽

pp. 103648

Author(s):

Katarzyna Skolik ◽

Chris Allison ◽

Judith Hohorst ◽

Mateusz Malicki ◽

Marina Perez-Ferragut ◽

...

Keyword(s):

Pressurized Water ◽

Water Reactor ◽

Loss Of Coolant

Volume 5: Fuel Cycle, Radioactive Waste Management and Decommissioning; Reactor Physics and Transport Theory; Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls; Fusion Engineering ◽

The Failure Analysis and Processing of Digital Reactor Protection System

10.1115/icone21-15423 ◽

2013 ◽

Author(s):

Jun Zhao ◽

Xing Zhou ◽

Jin Hu ◽

Yanling Yu

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Phase 1 ◽

Protection System ◽

Communication Failure ◽

Pressurized Water ◽

Power Failure ◽

Self Test ◽

Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.