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Boiling Water Reactor Simulator TRAINING COURSE SERIES No. 23 Boiling water reactor simulator Workshop material INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 2003 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 BOILING WATER REACTOR SIMULATOR IAEA, VIENNA, 2003 IAEA-TCS-23 ISSN 1018–5518 © IAEA, 2003 Printed by the IAEA in Austria September 2003 FOREWORD The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced 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 development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, “Reactor Simulator Development” (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 “WWER-1000 Reactor Simulator” (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 “Pressurized Water Reactor Simulator” (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. W.K. Lam, of Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA. The IAEA officer responsible for this publication was R.B. Lyon from the Division of Nuclear Power. 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. CONTENTS 1. INTRODUCTION ................................................................................................1 1.1 Purpose........................................................................................................1 1.2 Historical background .................................................................................1 1.3 Prominent characteristics of the BWR ........................................................3 2. 1300 MW(e) BOILING WATER REACTOR NPP SIMULATOR.....................5 2.1 Simulator startup.........................................................................................7 2.2 Simulator initialization................................................................................7 2.3 List of BWR simulator display screens.......................................................8 2.4 Generic BWR simulator display common features.....................................8 3. SIMULATOR DISPLAY SCREENS.................................................................10 3.1 BWR plant overview screen......................................................................10 3.2 BWR control loops screen.........................................................................12 3.3 BWR power/flow map & controls.............................................................14 3.4 BWR reactivity & controls screen.............................................................18 3.5 BWR scram parameters screen..................................................................20 3.6 BWR turbine generator screen ..................................................................21 3.7 BWR feedwater and extraction steam screen............................................22 4. SIMULATOR EXERCISES...............................................................................24 4.1 Introductory exercises ...............................................................................24 4.1.1 Power maneuver: 10% power reduction and return to full power.........................................................................................24 4.1.2 Reduction to 0% full power and back to 100% full power ..............24 4.1.3 Turbine trip and recovery.................................................................25 4.1.4 Reactor scram and recovery .............................................................25 4.2 Malfunction exercises................................................................................26 4.2.1 Loss of feedwater — Both FW pumps trip. .....................................26 4.2.2 Increasing core flow due to flow control failure ..............................28 4.2.3 Decreasing core flow due to flow control failure.............................29 4.2.4 Decreasing steam flow from dome due to pressure control failure ....................................................................30 4.2.5 Increasing steam flow from dome due to pressure control failure ....................................................................30 4.2.6 Turbine throttle PT fails low ............................................................32 4.2.7 Safety relief valve (SRV) on one main steam line fails open ..........33 4.2.8 Feedwater level control valve fails open..........................................34 4.2.9 Turbine trip with bypass valve failed closed....................................35 4.2.10 Inadvertent withdrawal of one bank of rods...................................36 4.2.11 Inadvertent insertion of one bank of rods.......................................37 4.2.12 Inadvertent reactor isolation...........................................................38 4.2.13 Loss of feedwater heating...............................................................39 4.2.14 Power loss to three reactor internal pumps (RIPs) .........................41 4.2.15 Steam line break inside drywell .....................................................42 4.2.16 Feedwater line break inside drywell...............................................43 4.2.17 Reactor vessel bottom break — 3000 kg/sec LOCA .....................45 4.2.18 Load rejection.................................................................................47 5. STEADY STATE MODEL................................................................................48 5.1 Purpose......................................................................................................48 5.2 Boiling water reactor mass and energy balance ........................................48 5.3 Boiling water reactor spreadsheet model ..................................................51 5.3.1 Procedures for spreadsheet model....................................................53 5.3.2 Steady state model solutions ............................................................55 6. DYNAMIC MODEL DESCRIPTION ...............................................................56 6.1 Reactor model............................................................................................57 6.2 Fuel heat transfer.......................................................................................59 6.3 Decay heat model ......................................................................................60 6.4 Coolant heat transfer .................................................................................61 6.5 Core hydraulics and heat transfer..............................................................63 6.5.1 Incompressible flow for non-boiling region.....................................63 6.5.2 Compressible flow for boiling region ..............................................65 6.5.3 Boiling boundary..............................................................................66 6.5.4 Summary of multi-nodal approach for simulating core hydraulics and heat transfer .............................................................66 6.6 Saturated enthalpy, saturated liquid density..............................................67 6.7 Core exit enthalpy, core quality, void fraction..........................................68 6.8 Dome mass balance and energy balance ...................................................68 6.9 Saturated steam density and dome pressure ..............................................69 6.10 Driving pressure in boiling core................................................................70 6.11 Recirculation flow & pressure losses ........................................................71 6.12 Coolant recirculation pumps .....................................................................72 6.13 Feedwater flow..........................................................................................72 6.14 Main steam system ....................................................................................73 6.15 Control and protection systems .................................................................74 6.15.1 Control rods control system ...........................................................75 6.15.2 Recirculation flow control..............................................................76
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