skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Computer Models for IRIS Control System Transient Analysis

Abstract

This report presents results of the Westinghouse work performed under Task 3 of this Financial Assistance Award and it satisfies a Level 2 Milestone for the project. Task 3 of the collaborative effort between ORNL, Brazil and Westinghouse for the International Nuclear Energy Research Initiative entitled “Development of Advanced Instrumentation and Control for an Integrated Primary System Reactor” focuses on developing computer models for transient analysis. This report summarizes the work performed under Task 3 on developing control system models. The present state of the IRIS plant design – such as the lack of a detailed secondary system or I&C system designs – makes finalizing models impossible at this time. However, this did not prevent making considerable progress. Westinghouse has several working models in use to further the IRIS design. We expect to continue modifying the models to incorporate the latest design information until the final IRIS unit becomes operational. Section 1.2 outlines the scope of this report. Section 2 describes the approaches we are using for non-safety transient models. It describes the need for non-safety transient analysis and the model characteristics needed to support those analyses. Section 3 presents the RELAP5 model. This is the highest-fidelity model used formore » benchmark evaluations. However, it is prohibitively slow for routine evaluations and additional lower-fidelity models have been developed. Section 4 discusses the current Matlab/Simulink model. This is a low-fidelity, high-speed model used to quickly evaluate and compare competing control and protection concepts. Section 5 describes the Modelica models developed by POLIMI and Westinghouse. The object-oriented Modelica language provides convenient mechanisms for developing models at several levels of detail. We have used this to develop a high-fidelity model for detailed analyses and a faster-running simplified model to help speed the I&C development process. Section 6 describes an ACSL model that Westinghouse started but suspended developing for the moment. ACSL is an old simulation language that Westinghouse used on many projects. It may (or may not) offer some advantages during the later stages of detailed plant design and analysis, but supporting the ACSL model does not appear to be necessary at this time. Section 7 summarizes our expectations for future development.« less

Authors:
; ;
Publication Date:
Research Org.:
Westinghouse Electric Company LLC
Sponsoring Org.:
USDOE - Office of Nuclear Energy, Science and Technology (NE)
OSTI Identifier:
933157
Report Number(s):
STD-AR-06-04 (rev. 2)
TRN: US0804272
DOE Contract Number:
FG07-05ID14690
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; BENCHMARKS; COMPUTERS; CONTROL SYSTEMS; DESIGN; FINANCING; NUCLEAR ENERGY; ORNL; SIMULATION; TRANSIENTS; VELOCITY; IRIS; Computer Models; Transient Analysis; Operating Transients; Safety Analysis

Citation Formats

Gary D. Storrick, Bojan Petrovic, and Luca Oriani. Computer Models for IRIS Control System Transient Analysis. United States: N. p., 2007. Web. doi:10.2172/933157.
Gary D. Storrick, Bojan Petrovic, & Luca Oriani. Computer Models for IRIS Control System Transient Analysis. United States. doi:10.2172/933157.
Gary D. Storrick, Bojan Petrovic, and Luca Oriani. Wed . "Computer Models for IRIS Control System Transient Analysis". United States. doi:10.2172/933157. https://www.osti.gov/servlets/purl/933157.
@article{osti_933157,
title = {Computer Models for IRIS Control System Transient Analysis},
author = {Gary D. Storrick and Bojan Petrovic and Luca Oriani},
abstractNote = {This report presents results of the Westinghouse work performed under Task 3 of this Financial Assistance Award and it satisfies a Level 2 Milestone for the project. Task 3 of the collaborative effort between ORNL, Brazil and Westinghouse for the International Nuclear Energy Research Initiative entitled “Development of Advanced Instrumentation and Control for an Integrated Primary System Reactor” focuses on developing computer models for transient analysis. This report summarizes the work performed under Task 3 on developing control system models. The present state of the IRIS plant design – such as the lack of a detailed secondary system or I&C system designs – makes finalizing models impossible at this time. However, this did not prevent making considerable progress. Westinghouse has several working models in use to further the IRIS design. We expect to continue modifying the models to incorporate the latest design information until the final IRIS unit becomes operational. Section 1.2 outlines the scope of this report. Section 2 describes the approaches we are using for non-safety transient models. It describes the need for non-safety transient analysis and the model characteristics needed to support those analyses. Section 3 presents the RELAP5 model. This is the highest-fidelity model used for benchmark evaluations. However, it is prohibitively slow for routine evaluations and additional lower-fidelity models have been developed. Section 4 discusses the current Matlab/Simulink model. This is a low-fidelity, high-speed model used to quickly evaluate and compare competing control and protection concepts. Section 5 describes the Modelica models developed by POLIMI and Westinghouse. The object-oriented Modelica language provides convenient mechanisms for developing models at several levels of detail. We have used this to develop a high-fidelity model for detailed analyses and a faster-running simplified model to help speed the I&C development process. Section 6 describes an ACSL model that Westinghouse started but suspended developing for the moment. ACSL is an old simulation language that Westinghouse used on many projects. It may (or may not) offer some advantages during the later stages of detailed plant design and analysis, but supporting the ACSL model does not appear to be necessary at this time. Section 7 summarizes our expectations for future development.},
doi = {10.2172/933157},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Wed Jan 31 00:00:00 EST 2007},
month = {Wed Jan 31 00:00:00 EST 2007}
}

Technical Report:

Save / Share:
  • This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyzemore » a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.« less
  • The MARVEL (Multi-Loop Analysis of Pressurized Water Reactor System Transient) digital computer code was developed to calculate multiloop detailed transient behavior of pressurized water reactor systems. The program simulates two reactor coolant loops including two steam generators and associated systems. It also simulates reactor kinetics, reactor control and protection system, safeguards system, and other subsystems. The code can also be used for three- or four-loop plants by assuming that the other loops are operated in the same way as either of the two loops. The program can be utilized as a tool for various types of accident analyses and controlmore » studies, including startup of an inactive reactor coolant loop, loss of reactor coolant flow, reactivity insertion incidents, steam line break accident, steam generator tube rupture accident, and others.« less
  • TRAAP is a computer program that records, analyzes, and plots data gathered by a transient recorder. It controls data acquisition by allowing the user to check the transient recorder before taking data and offering the choice of using or not using a low-pass filter. It transfers data to and from floppy disks for storage and playback. It analyzes data in a variety of ways (including compressing, smoothing, simple mathematical operations, differentiating, integrating) and provides four options for graphic output. TRAAP is a flexible program the capabilities of which can readily be increased; several new options are currently being developed. 9more » figures.« less