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Boiling Water Reactor Simulator with Passive Safety Systems

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Boiling Water Reactor Simulator with Passive Safety Systems Boiling Water Reactor Simulator with Passive Safety Systems User Manual October 2009 INTERNATIONAL ATOMIC ENERGY AGENCY, 2009 The originating Section of this publication in the IAEA was: Nuclear Power Technology Development Section International Atomic Energy Agency Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria 2 3 Foreword Given the renewed worldwide interest in nuclear technology, there has been a growing demand for qualified nuclear professionals, which in turn has resulted in the creation of new nuclear science and technology education programs and in the growth of existing ones. Of course, this increase in the number of students pursuing nuclear degrees, has also contributed to a large need for qualified faculty and for comprehensive and up-to-date curricula. The International Atomic Energy Agency (IAEA) has established a programme in PC-based Nuclear Power Plant (NPP) simulators to assist Member States in their education and training endeavors. The objective of this programme is to provide, for a variety of nuclear reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and their associated training materials, sponsors training courses and workshops, and distributes documentation and computer programs. The simulators operate on personal computers and are provided for a broad audience of technical and non- technical personnel as an introductory educational tool. The preferred audience, however, are faculty members interested in developing nuclear engineering courses with the support of these very effective hands-on educational tools. It is important to remember, however, that the application of these PC-based simulators is limited to providing general response characteristics of selected types of power reactor systems and that they are not intended to be used for plant-specific purposes such as design, safety evaluation, licensing or operator training. The IAEA simulator collection currently includes the following simulators: • A WWER-1000 simulator provided to the IAEA by the Moscow Engineering and Physics Institute in Russia. • The IAEA generic Pressurized Water Reactor (PWR) simulator has been developed by Micro- Simulation Technology of USA using the PCTRAN software. This simulator is a 600 MWe generic two-loop PWR with inverted U-bend steam generators and dry containment system that could be a Westinghouse, Framatome or KWU design. • The IAEA advanced PWR simulator has been developed by Cassiopeia Technologies Inc. (CTI) of Canada, and is largely based on a 600 MWe PWR design with passive safety systems, similar to the Westinghouse AP-600. • The IAEA generic Boiling Water Reactor (BWR) simulator has also been developed by CTI and represents a typical 1300 MWe BWR with internal recirculation pumps and fine motion control rod drives. This simulator underwent a major enhancement effort in 2008 when a containment model based on the ABWR was added. • The IAEA Pressurized Heavy Water Reactor (PHWR) simulator is also a CTI product and is largely based on the 900 MWe CANDU-9 system. • The IAEA advanced PHWR simulator by CTI from Canada, which represents the ACR-700 system. • The IAEA advanced BWR, which largely represents the GE ESBWR passive BWR design and was also created by CTI. The simulator development is a concerted effort from the developer, the Agency’s staff, and from Dr. Bharat Shiralkar, a thermal-hydraulics expert. This activity was initiated under the leadership of Mr. R. B. Lyon. Subsequently, Mr. J. C. Cleveland and later Ms. S. Bilbao y León and Mr. S.D. Jo from the Division of Nuclear Power became the IAEA responsible officers. More information about the IAEA simulators and the associated training is available at http://www.iaea.org/NuclearPower/Education/Simulators/ 4 EDITORIAL NOTE The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA. 5 Table of Content 1. INTRODUCTION..............................................................................................9 1.1 PURPOSE ............................................................................................................9 1.2 HISTORICAL BACKGROUND INTRODUCTION ......................................................10 1.3 EVOLUTION OF BWR FROM BWR/1 THROUGH ESBWR .....................10 1.3.1 General Progression of BWR Designs ........................................................12 1.3.2 Reactor System Design ................................................................................13 1.3.3 Safety System Design...................................................................................18 1.3.4 Containment Design ....................................................................................21 2. ESBWR MAJOR SYSTEMS .............................................................................25 2.1 NATURAL CIRCULATION DESIGN........................................................................25 2.2 ESBWR PRESSURE VESSEL AND INTERNALS .....................................................28 2.3 ESBWR CORE..................................................................................................31 2.3.1 Nuclear Fuel Assemblies ..............................................................................31 2.3.2 Control Rod Drive System ...........................................................................32 2.3.3 Rod Sequence Control System (RSCS) ........................................................35 2.3.4 Neutron Monitoring System.........................................................................36 2.3.5 Reactor Protection System (RPS)................................................................37 2.3.6 Core Nuclear reactivity Characteristics .....................................................40 2.3.7 Reactivity Control........................................................................................40 2.3.8 Fuel Management........................................................................................41 2.4 ESBWR OPERATION IN NATURAL CIRCULATION.................................................42 2.5 ESBWR NUCLEAR BOILER SYSTEM ...................................................................44 2.6 MAIN STEAM LINE ISOLATION VALVES (MSIV).................................................44 2.7 SAFETY-RELIEF VALVES .....................................................................................45 2.8 AUTOMATIC DEPRESSURIZATION SYSTEM ..........................................................45 2.9 MAIN TURBINE AND GENERATOR.......................................................................47 2.9.1 Main Condenser and Turbine Bypass Valves...............................................47 2.9.2 Circulating Water System............................................................................47 2.10 REACTOR FEEDWATER SYSTEM........................................................................47 2.10.1 Feedwater Lines (FWLs) ............................................................................48 2.10.2 Feedwater Control System.........................................................................48 2.11 PASSIVE SAFETY FEATURES.............................................................................49 2.11.1 Gravity Driven Core Cooling System.......................................................50 2.11.2 Passive Containment Cooling System ......................................................52 2.11.3 Isolation Condenser System .....................................................................54 2.12 STANDBY LIQUID CONTROL SYSTEM ................................................................57 2.13 REACTOR WATER CLEANUP/SHUTDOWN COOLING SYSTEM.............................60 2.14 CONTAINMENT SYSTEM ....................................................................................60 2.14.1 Containment Inerting System ....................................................................63 2.14.2 Containment Monitoring System ................................................................64 3. PASSIVE BWR NPP SIMULATOR – GENERAL..........................................65 3.1 SIMULATOR STARTUP...............................................................................69 3.2 SIMULATOR INITIALIZATION..................................................................69 3.3 LIST OF BWR SIMULATOR DISPLAY SCREENS ....................................69 3.4 PASSIVE BWR SIMULATOR DISPLAY COMMON FEATURES ...............70 4. PASSIVE BWR SIMULATOR DISPLAY SCREENS ....................................72 6 4.1 PASSIVE BWR PLANT OVERVIEW SCREEN..............................................72 4.2 PASSIVE BWR CONTROL LOOPS SCREEN ................................................75 4.3 PASSIVE BWR POWER/FLOW MAP & CONTROLS...................................78 4.4 PASSIVE BWR REACTIVITY & CONTROLS SCREEN............................83 4.5 PASSIVE BWR SCRAM PARAMETERS SCREEN ....................................85 4.6 PASSIVE BWR TURBINE GENERATOR SCREEN...................................86
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