Shutdown and Cooldown of SEE-THRU Nuclear Power Plant for Student Performance. MP-SEE-THRU-02 Rev. 004

Size: px
Start display at page:

Download "Shutdown and Cooldown of SEE-THRU Nuclear Power Plant for Student Performance. MP-SEE-THRU-02 Rev. 004"

Transcription

1 Student Operating Procedure Millstone Station Shutdown and Cooldown of SEE-THRU Nuclear Power Plant for Student Performance Approval Date: 10/4/2007 Effective Date: 10/4/2007

2 TABLE OF CONTENTS 1. PURPOSE Objective Discussion Applicability PREREQUISITES General PRECAUTIONS Safety Precautions General Precautions Operating Limits INSTRUCTIONS Securing the Secondary Plant Reactor Shutdown Plant Cooldown and System Shutdown of 11

3 1. PURPOSE 1.1 Objective This procedure provides instructions for the shutdown and cooldown of the SEE-THRU Nuclear Power Plant. Students under the instruction of a qualified instructor may use this procedure. 1.2 Discussion This procedure provides instructions to allow students to perform a shutdown and cooldown of the SEE-THRU Nuclear Power Plant. 1.3 Applicability This procedure is applicable for student operation of the SEE-THRU Nuclear Power Plant. 3 of 11

4 2. PREREQUISITES 2.1 General The instructor has conducted a pre-job brief with all students for this activity Students have reviewed and understand the safety precautions, general precautions and SEE-THRU Nuclear Power Plant operating limits Students have reviewed the SEE-THRU Nuclear Power Plant layout and are cognizant of the locations of the major plant components, operating controls, local valves, pumps and instrumentation indicators. 4 of 11

5 3. PRECAUTIONS 3.1 Safety Precautions Always operate the SEE-THRU Nuclear Power Plant with the containment plexiglass (cylinder and dome) and component plexiglass covers in place to prevent scalding from hot water or steam in the event of a primary or secondary plant leak or rupture. The containment and component covers must also be in place to prevent serious lacerations or other injuries from broken/flying glass should a glass component break. 3.2 General Precautions When selecting RUN for any pump, select STOP immediately if there is an abnormal noise or if an abnormal noise develops during operation Do NOT run any pump at its shutoff head for an extended period of time Do NOT allow the condenser to pump empty because this will cause air binding of Condensate Pump - P4 and/or Feed Pump - P When the reactor coolant system is solid (no steam bubble in pressurizer), a small change in temperature will result in a significant change in system pressure When operating Feedwater Regulating Valves FRV-FW1 and FRV-FW2, Reactor Vessel Vent Valve RC-15, and Pressurizer PORV Isolation Valve RC-22 in the closed direction, do NOT run them hard into their seat in order to prevent damage to the valves. 3.3 Operating Limits Ensure that the following operating limits are not exceeded: a. Reactor coolant system pressure is maintained less than 23 psig b. Reactor temperature is maintained less than 275 F c. Pressurizer temperature is maintained less than 275 F d. Condenser water level is maintained greater than 1/3 full e. Containment temperature is maintained less than 180 F 5 of 11

6 4. INSTRUCTIONS 4.1 Securing the Secondary Plant If feedwater heater is in operation, perform the following steps to shutdown feedwater heater: a. CLOSE valve Feedwater Supply MS5. b. CLOSE manual isolation valve MS-9. c. CLOSE valve Feedwater Heater Drain MS If vacuum pump is in operation, perform the following steps to shutdown vacuum pump: a. STOP Vacuum Pump P5. b. OPEN manual isolation valve MS-13. c. CLOSE manual isolation valve MS End of Section of 11

7 4.2 Reactor Shutdown LOWER Reactor Power to 0 amperes When Turbine Speed lowers to 500 RPM, perform the following steps to shutdown the secondary plant: a. TURN the Turbine Generator Exciter control knob fully counterclockwise to its minimum position. b. SWITCH the Turbine Generator Exciter switch to OFF. c. OPEN valve Bypass-MS3 (Steam Dump to the Condenser). d. CLOSE valve TSV-MS4 (Steam to the Main Turbine). e. VERIFY Turbine Trip annunciator light is lit PRESS the Reactor Trip push button VERIFY Reactor Trip button red light is lit AND Reactor Trip alarm is lit. 7 of 11

8 s CAUTION s 1. When operating the Feedwater Regulating Valves FRV-FW1 and FRV- FW2, in the closed direction, be careful not to run them hard into the seat to prevent damaging the valves. 2. Proper water levels (approximately ½ to1 inch above the tube bundles) must be maintained while boiling is occurring in the secondary side of the steam generators 3. Do NOT allow the condenser to pump empty because this will cause air binding of Condensate Pump - P4 and/or Feed Pump - P When boiling stops in the steam generator, perform the following: a. CLOSE Feedwater Regulating Valves FRV-FW1 for A S/G and FRV- FW2 for B S/G. b. STOP Feed Pump P3. c. STOP Condensate Pump P4. s CAUTION s Steps through must be performed quickly to prevent an automatic actuation of the Safety Injection System Perform the following to de-energize the pressurizer heaters: a. CLOSE valve Pressurizer Spray RC4 AND ALLOW pressurizer pressure to stabilize between 19 and 21 psig. b. PRESS F1 then F5 on the Master Display/Controller to display Pressurizer Control. c. PRESS F7 switch from Automatic to Manual pressure control mode. d. VERIFY Pressurizer Control is in MANUAL. e. PRESS F6 On the Master Display/Controller f. ENTER 0% (TYPE in 0 and depress the enter key) g. VERIFY 0% output is displayed. h. PLACE the Pressurizer Heater switch in the OFF position. i. PRESS F1 then F2 on the Master Display/Controller to display pressurizer water and reactor temperatures. 8 of 11

9 s CAUTION s When the reactor coolant system is solid (no steam bubble in pressurizer), a small change in temperature will result in a significant change in system pressure Perform the following steps to fill the reactor coolant system to a water solid condition: a. CLOSE manual valve RC-22 b. OPEN valve RC-5, Pressurizer PORV c. START Safety Inj Pump - P1. d. PLACE the Injection SI1 key switch to Bypass (key in horizontal position - ¼ turn clockwise). e. OPEN valve Injection SI1. NOTE Pressurizer pressure may lower to 8 to 10 psig and then stabilize between 10 and 15 psig when the reactor coolant system is water solid and valve RC-22 is closed. f. SLOWLY OPEN manual valve RC-22 and allow the pressurizer to gradually fill. - End of Section of 11

10 4.3 Plant Cooldown and System Shutdown NOTE 1. Operation of valve RC-22 will require constant attention during the plant cooldown. 2. A slower cooldown rate will cause less stress on the plant components. 3. Opening manual valve RC-22 will provide a flowpath from the pressurizer to the quench tank/rwst THROTTLE manual valve RC-22 to MAINTAIN Pressurizer Pressure between 10 to 13 psig When reactor temperature reaches 150 F, perform the following: a. THROTTLE manual valve RC-22 to MAINTAIN Pressurizer Pressure at 5 psig. b. CLOSE valve Injection SI1. c. MONITOR Passive Accumulator Pressure gauge for lowering pressure. d. When Passive Accumulator pressure lowers to 3 psig, OPEN valve Injection SI1. e. THROTTLE manual valve RC-22 to maintain Pressurizer Pressure between 10 to 13 psig for the remainder of the cooldown WHEN pressurizer and reactor water temperatures are < 120 F, PERFORM the following: a. CLOSE valves Pressurizer PORV RC-5 and Injection SI1 simultaneously. b. FULLY OPEN manual valve RC-22 AND VERIFY Pressurizer Pressure remains stable. 10 of 11

11 4.3.4 Perform the following steps to stop reactor coolant system flow: a. STOP Reactor Coolant Pump P7. b. STOP Reactor Coolant Pump P6. c. STOP Safety Inj Pump - P1. d. CLOSE valve Flood SI ADJUST water levels by using the associated fill and drain valves and pumps to achieve the following: Reactor - SOLID. Pressurizer - SOLID. Passive Accumulator - 2/3 FULL. Passive Accumulator - 0 psig. Condenser coils - COVERED. Steam Generator water levels - above FEED RING VERIFY the status of the following components: Pressurizer SURGE-RC3 = OPEN. Main Steam Isolation Valves MSIV-MS1 and MSIV-MS2 = OPEN. BYPASS-MS3 (Steam Dump to the Condenser) = OPEN. ALL OTHER main control board valves = CLOSE (green status lights) STOP the Containment Air Recirculation Cooling (CARC) fan by turning the CARC control switch to the OFF position CLOSE valve Circ Water Reg CW CLOSE the CIRC WATER THROTTLE VALVE to condenser TRIP the sump pump GFCI CLOSE the CIRCULATING WATER supply line VALVE on wall DEPRESS the Reactor, PZR, S/G, RWST, T/G, and COND push buttons to de-energize the display lighting TURN OFF main power on console TURN OFF main power wall switches. - End of Section of 11

Westinghouse Small Modular Reactor. Passive Safety System Response to Postulated Events

Westinghouse Small Modular Reactor. Passive Safety System Response to Postulated Events Westinghouse Small Modular Reactor Passive Safety System Response to Postulated Events Matthew C. Smith Dr. Richard F. Wright Westinghouse Electric Company Westinghouse Electric Company 600 Cranberry Woods

More information

EPR: Steam Generator Tube Rupture analysis in Finland and in France

EPR: Steam Generator Tube Rupture analysis in Finland and in France EPR: Steam Generator Tube Rupture analysis in Finland and in France S. ISRAEL Institut de Radioprotection et de Sureté Nucléaire BP 17 92262 Fontenay-aux-Roses Cedex, France Abstract: Different requirements

More information

RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING

RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING Science and Technology Journal of BgNS, Vol. 8, 1, September 2003, ISSN 1310-8727 RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING Pavlin P. Groudev, Rositsa V. Gencheva,

More information

NPP Simulators Workshop for Education - Passive PWR NPP & Simulator Overview

NPP Simulators Workshop for Education - Passive PWR NPP & Simulator Overview NPP Simulators Workshop for Education - Passive PWR NPP & Simulator Overview Wilson Lam (wilson@cti-simulation.com) CTI Simulation International Corp. www.cti-simulation.com Sponsored by IAEA Modified

More information

Joint ICTP-IAEA Essential Knowledge Workshop on Deterministic Safety Analysis and Engineering Aspects Important to Safety. Trieste,12-23 October 2015

Joint ICTP-IAEA Essential Knowledge Workshop on Deterministic Safety Analysis and Engineering Aspects Important to Safety. Trieste,12-23 October 2015 Joint ICTP- Essential Knowledge Workshop on Deterministic Safety Analysis and Engineering Aspects Important to Safety Trieste,12-23 October 2015 Safety classification of structures, systems and components

More information

EXPERIMENTS ON THE PERFORMANCE SENSITIVITY OF THE PASSIVE RESIDUAL HEAT REMOVAL SYSTEM OF AN ADVANCED INTEGRAL TYPE REACTOR

EXPERIMENTS ON THE PERFORMANCE SENSITIVITY OF THE PASSIVE RESIDUAL HEAT REMOVAL SYSTEM OF AN ADVANCED INTEGRAL TYPE REACTOR EXPERIMENTS ON THE PERFORMANCE SENSITIVITY OF THE PASSIVE RESIDUAL HEAT REMOVAL SYSTEM OF AN ADVANCED INTEGRAL TYPE REACTOR HYUN-SIK PARK *, KI-YONG CHOI, SEOK CHO, SUNG-JAE YI, CHOON-KYUNG PARK and MOON-KI

More information

CANDU Safety #6 - Heat Removal Dr. V.G. Snell Director Safety & Licensing

CANDU Safety #6 - Heat Removal Dr. V.G. Snell Director Safety & Licensing CANDU Safety #6 - Heat Removal Dr. V.G. Snell Director Safety & Licensing 24/05/01 CANDU Safety - #6 - Heat Removal.ppt Rev. 0 vgs 1 Overview the steam and feedwater system is similar in most respects

More information

Risks Associated with Shutdown in PWRs

Risks Associated with Shutdown in PWRs International Conference Nuclear Option in Countries with Small and Mi Opatija, Croatia, 1996 Risks Associated with Shutdown in PWRs Igor Grlicarev Slovenian Nuclear Safety Administration Vojkova 59, 1113

More information

Main Steam & T/G Systems, Safety

Main Steam & T/G Systems, Safety Main Steam & T/G Systems, Safety Page 1 This steam generator, built for the Wolsong station in Korea, was manufactured in Canada by the Babcock and Wilcox company. In Wolsong 2,3, and 4 a joint venture

More information

THREE MILE ISLAND ACCIDENT

THREE MILE ISLAND ACCIDENT THREE MILE ISLAND ACCIDENT M. Ragheb 12/4/2015 1. INTRODUCTION The Three Mile Island (TMI) Accident at Harrisburg, Pennsylvania in the USA is a severe and expensive incident that has seriously affected,

More information

Small Modular Reactors: A Call for Action

Small Modular Reactors: A Call for Action Small Modular Reactors: A Call for Action Overview of Five SMR Designs by Dr. Regis A. Matzie Executive Consultant Adapted May 2015 for the Hoover Institution's Reinventing Nuclear Power project from a

More information

RELAP 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-07

RELAP 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-07 Fifth International Seminar on Horizontal Steam Generators 22 March 21, Lappeenranta, Finland. 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-7 József Bánáti Lappeenranta University of

More information

INVESTIGATION OF CRITICAL SAFETY FUNCTION INTEGRITY IN CASE OF STEAM LINE BREAK ACCIDENT FOR VVER 1000/V320

INVESTIGATION OF CRITICAL SAFETY FUNCTION INTEGRITY IN CASE OF STEAM LINE BREAK ACCIDENT FOR VVER 1000/V320 International Conference 12th Symposium of AER, Sunny Beach, pp.99-105, 22-28 September, 2002. INVESTIGATION OF CRITICAL SAFETY FUNCTION INTEGRITY IN CASE OF STEAM LINE BREAK ACCIDENT FOR VVER 1000/V320

More information

Verification of the MELCOR Code Against SCDAP/RELAP5 for Severe Accident Analysis

Verification of the MELCOR Code Against SCDAP/RELAP5 for Severe Accident Analysis Verification of the Code Against SCDAP/RELAP5 for Severe Accident Analysis Jennifer Johnson COLBERT 1* and Karen VIEROW 2 1 School of Nuclear Engineering, Purdue University, West Lafayette, Indiana 47907-2017,

More information

Passive Cooldown Performance of Integral Pressurized Water Reactor

Passive Cooldown Performance of Integral Pressurized Water Reactor Energy and Power Engineering, 2013, 5, 505-509 doi:10.4236/epe.2013.54b097 Published Online July 2013 (http://www.scirp.org/journal/epe) Passive Cooldown Performance of Integral Pressurized Water Reactor

More information

Module 05 WWER/ VVER (Russian designed Pressurized Water Reactors)

Module 05 WWER/ VVER (Russian designed Pressurized Water Reactors) Module 05 WWER/ VVER (Russian designed Pressurized Water Reactors) 1.3.2016 Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at

More information

INTEGRAL EFFECT NON-LOCA TEST RESULTS FOR THE INTEGRAL TYPE REACTOR SMPART-P USING THE VISTA FACILITY

INTEGRAL EFFECT NON-LOCA TEST RESULTS FOR THE INTEGRAL TYPE REACTOR SMPART-P USING THE VISTA FACILITY HEFAT7 5 th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics 1- July 7, Sun City, South Africa CK INTEGRAL EFFECT NON-LOCA TEST RESULTS FOR THE INTEGRAL TYPE REACTOR SMPART-P

More information

Implementation of Lessons Learned from Fukushima Accident in CANDU Technology

Implementation of Lessons Learned from Fukushima Accident in CANDU Technology e-doc 4395709 Implementation of Lessons Learned from Fukushima Accident in CANDU Technology Greg Rzentkowski Director General, Power Reactor Regulation Canadian Nuclear Safety Commission on behalf of CANDU

More information

Design of Traditional and Advanced CANDU Plants. Artur J. Faya Systems Engineering Division November 2003

Design of Traditional and Advanced CANDU Plants. Artur J. Faya Systems Engineering Division November 2003 Design of Traditional and Advanced CANDU Plants Artur J. Faya Systems Engineering Division November 2003 Overview Canadian Plants The CANDU Reactor CANDU 600 and ACR-700 Nuclear Steam Supply Systems Fuel

More information

A DYNAMIC ASSESSMENT OF AUXILIARY BUILDING CONTAMINATION AND FAILURE DUE TO A CYBER-INDUCED INTERFACING SYSTEM LOSS OF COOLANT ACCIDENT

A DYNAMIC ASSESSMENT OF AUXILIARY BUILDING CONTAMINATION AND FAILURE DUE TO A CYBER-INDUCED INTERFACING SYSTEM LOSS OF COOLANT ACCIDENT A DYNAMIC ASSESSMENT OF AUXILIARY BUILDING CONTAMINATION AND FAILURE DUE TO A CYBER-INDUCED INTERFACING SYSTEM LOSS OF COOLANT ACCIDENT Z.K. Jankovsky The Ohio State University Columbus, USA Email: jankovsky.3@osu.edu

More information

PALO VERDE NUCLEAR GENERATING STATION

PALO VERDE NUCLEAR GENERATING STATION PALO VERDE NUCLEAR GENERATING STATION Introduction to Plant Systems Lesson Plan Engineering Training (Program) Date: May 17, 2012 LP Number: NGT95C000102 Rev Author: Ben Acosta Title: Introduction to Plant

More information

B. System Design and Performance Requirements

B. System Design and Performance Requirements 15625 Water Chillers This document provides design standards only, and is not intended for use, in whole or in part, as a specification. Do not copy this information verbatim in specifications or in notes

More information

Nuclear Power A Journey of Continuous Improvement

Nuclear Power A Journey of Continuous Improvement Nuclear Power A Journey of Continuous Improvement Westinghouse Non Proprietary Class 3 Our Place in Nuclear History Innovation 1886 and forever Implementation & Improvement 1957 through Today Renaissance

More information

The ESBWR an advanced Passive LWR

The ESBWR an advanced Passive LWR 1 IAEA PC-Based Simulators Workshop Politecnico di Milano, 3-14 October 2011 The ES an advanced Passive LWR Prof. George Yadigaroglu, em. ETH-Zurich and ASCOMP yadi@ethz.ch 2 Removal of decay heat from

More information

Secondary Systems: Steam System

Secondary Systems: Steam System Secondary Systems: Steam System K.S. Rajan Professor, School of Chemical & Biotechnology SASTRA University Joint Initiative of IITs and IISc Funded by MHRD Page 1 of 10 Table of Contents 1 SECONDARY SYSTEM

More information

DESIGN AND SAFETY PRINCIPLES LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION

DESIGN AND SAFETY PRINCIPLES LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION DESIGN AND SAFETY PRINCIPLES LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION VIENNA OKTOBER 3-6, 2016 1 ANPP * ANPP is located in the western part of Ararat valley 30 km west of Yerevan close to

More information

ACR-1000: ENHANCED RESPONSE TO SEVERE ACCIDENTS

ACR-1000: ENHANCED RESPONSE TO SEVERE ACCIDENTS ACR-1000: ENHANCED RESPONSE TO SEVERE ACCIDENTS Popov, N.K., Santamaura, P., Shapiro, H. and Snell, V.G Atomic Energy of Canada Limited 2251 Speakman Drive, Mississauga, Ontario, Canada L5K 1B2 1. INTRODUCTION

More information

Nuclear Service Valves

Nuclear Service Valves GE Energy Consolidated* Pressure Relief Valves Nuclear Service Valves Pressure relief solutions for the nuclear industry Safety Relief Valves Safety Valves Advanced Technology and Equipment Manufactured

More information

Preliminary Lessons Learned from the Fukushima Daiichi Accident for Advanced Nuclear Power Plant Technology Development

Preliminary Lessons Learned from the Fukushima Daiichi Accident for Advanced Nuclear Power Plant Technology Development Preliminary Lessons Learned from the Fukushima Daiichi Accident for Advanced Nuclear Power Plant Technology Development A. Introduction The IAEA Report on Reactor and Spent Fuel Safety in the Light of

More information

ES Fluid & Thermal Systems Page 1 of 6 STEAM TURBINE LABORATORY

ES Fluid & Thermal Systems Page 1 of 6 STEAM TURBINE LABORATORY ES 202 - Fluid & Thermal Systems Page 1 of 6 STEAM TURBINE LABORATORY Objective The objective of this laboratory experience is to demonstrate how mechanical power can be generated using a steam turbine

More information

Nuclear Power Plant Safety Basics. Construction Principles and Safety Features on the Nuclear Power Plant Level

Nuclear Power Plant Safety Basics. Construction Principles and Safety Features on the Nuclear Power Plant Level Nuclear Power Plant Safety Basics Construction Principles and Safety Features on the Nuclear Power Plant Level Safety of Nuclear Power Plants Overview of the Nuclear Safety Features on the Power Plant

More information

Application for Permission to Extend the Operating Period and Application for Approval of Construction Plans of Unit 3 at Mihama Nuclear Power Station

Application for Permission to Extend the Operating Period and Application for Approval of Construction Plans of Unit 3 at Mihama Nuclear Power Station November 26, 2015 The Kansai Electric Power Co., Inc. Application for Permission to Extend the Operating Period and Application for Approval of Construction Plans of Unit 3 at Mihama Nuclear Power Station

More information

SMR/1848-T03. Course on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors June 2007

SMR/1848-T03. Course on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors June 2007 SMR/1848-T03 Course on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors 25-29 June 2007 Applications of Natural Circulation Systems N. Aksan Paul Scherrer Institut (PSI), Villingen,

More information

CONDENSATION INDUCED WATER HAMMER

CONDENSATION INDUCED WATER HAMMER CONDENSATION INDUCED WATER HAMMER RISKS AND ISSUES ASSOCIATED WITH SAGD OPERATIONS Jeff Dancey, P. Eng. Baker Engineering and Risk Consultants, Inc. OUTLINE CONDENSATION INDUCED WATER HAMMER (CIWH) - EXPLANATION

More information

Nuclear Power Volume II - Nuclear Power Plants

Nuclear Power Volume II - Nuclear Power Plants PDHonline Course E338 (5 PDH) Nuclear Power Volume II - Nuclear Power Plants Instructor: Lee Layton, PE 2012 PDH Online PDH Center 5272 Meadow Estates Drive Fairfax, VA 22030-6658 Phone & Fax: 703-988-0088

More information

CAREM: AN INNOVATIVE-INTEGRATED PWR

CAREM: AN INNOVATIVE-INTEGRATED PWR 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18) Beijing, China, August 7-12, 2005 SMiRT18-S01-2 CAREM: AN INNOVATIVE-INTEGRATED PWR Rubén MAZZI INVAP Nuclear Projects

More information

Status report 85 - VVER-1500 (V-448) (VVER-1500 (V-448))

Status report 85 - VVER-1500 (V-448) (VVER-1500 (V-448)) Status report 85 - VVER-1500 (V-448) (VVER-1500 (V-448)) Overview Full name Acronym Reactor type Coolant Moderator Neutron spectrum Thermal capacity Gross Electrical capacity Design status Designers VVER-1500

More information

NUCLEAR TRAINING CENTRE COURSE 134 FOR ONTARIO HYDRO USE ONLY

NUCLEAR TRAINING CENTRE COURSE 134 FOR ONTARIO HYDRO USE ONLY NUCLEAR TRAINING CENTRE COURSE 134 FOR ONTARIO HYDRO USE ONLY NUCLEAR TRAINING COURSE COURSE 134 1 - Level 3 - Equipment & System Principles 4 - TURBINE, GENERATOR & AUXILIARIES Index 134.00-0 Objectives

More information

Analysis of a Station Black-Out transient in SMR by using the TRACE and RELAP5 code

Analysis of a Station Black-Out transient in SMR by using the TRACE and RELAP5 code Journal of Physics: Conference Series OPEN ACCESS Analysis of a Station Black-Out transient in SMR by using the TRACE and RELAP5 code To cite this article: F De Rosa et al 2014 J. Phys.: Conf. Ser. 547

More information

Lesson 10: Overall Unit Control Module 1: Boiler Pressure Control

Lesson 10: Overall Unit Control Module 1: Boiler Pressure Control IAEA - CANDU I & C Module 1: Boiler Pressure Control Lesson 10: OVERALL UNIT CONTROL Module 1: Boiler Pressure Control MODULE 1: BOILER PRESSURE CONTROL (BPC) MODULE OBJECTIVES: At the end of this module,

More information

OPERATING EXPERIENCE REGARDING THERMAL FATIGUE OF UNISOLABLE PIPING CONNECTED TO PWR REACTOR COOLANT SYSTEMS

OPERATING EXPERIENCE REGARDING THERMAL FATIGUE OF UNISOLABLE PIPING CONNECTED TO PWR REACTOR COOLANT SYSTEMS OPERATING EXPERIENCE REGARDING THERMAL FATIGUE OF UNISOLABLE PIPING CONNECTED TO PWR REACTOR COOLANT SYSTEMS ABSTRACT Paul Hirschberg John Carey Arthur F. Deardorff EPRI Project Manager Structural Integrity

More information

The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000 TM. Roger Schène Director,Engineering Services

The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000 TM. Roger Schène Director,Engineering Services The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000 TM Roger Schène Director,Engineering Services 1 Background Late 80: USA Utilities under direction of EPRI and endorsed by NRC : Advanced

More information

20/06/2011 Seminar on Geothermal Exploitation Santiago de Chile

20/06/2011 Seminar on Geothermal Exploitation Santiago de Chile Contents Power Plants Steam Power plants Binary Power plants Geothermal Power Plants Single flash systems Binary systems 1 Equipment Well head Gathering piping system Steam separators and moisture separators

More information

Safety Provisions for the KLT-40S Reactor Plant

Safety Provisions for the KLT-40S Reactor Plant 6th INPRO Dialogue Forum on Global Nuclear Energy Sustainability: Licensing and Safety Issues for Small and Medium-sized Nuclear Power Reactors (SMRs) 29 July - 2 August 2013 IAEA Headquarters, Vienna,

More information

SCHNEIDER-KESSEL BERLIN

SCHNEIDER-KESSEL BERLIN SCHNEIDER-KESSEL BERLIN STEAM BOILER and HOT WATER BOILER WASTE HEAT RECOVERY BOILER - Series AHK GENERAL Waste Heat Recovery boilers of Series AHK are steam generators in special smoke tube design to

More information

NUCLEAR POWER NEW NUCLEAR POWER PLANTS IN 2012

NUCLEAR POWER NEW NUCLEAR POWER PLANTS IN 2012 NUCLEAR POWER NEW NUCLEAR POWER PLANTS IN 2012 AP1000 IN FEBRUARY 2012, THE FIRST NUCLEAR POWER PLANTS IN THE US IN 35 YEARS WERE LICENSCED TO BEGIN CONSTRUCTION. TWO WESTINGHOUSE AP1000 NUCEAR REACTOR

More information

National Maritime Center

National Maritime Center National Maritime Center Providing Credentials to Mariners U.S.C.G. Merchant Marine Exam (Sample Examination) Page 1 of 22 Choose the best answer to the following Multiple Choice Questions. 1. Excessive

More information

OPG Proprietary Report

OPG Proprietary Report N/A R001 2 of 114 Table of Contents Page List of Tables and Figures... 5 Revision Summary... 6 Executive Summary... 7 1.0 INTRODUCTION... 9 1.1 Objectives... 10 1.2 Scope... 10 1.3 Organization of Summary...

More information

Combined-Cycle Units and Block Design Data (Voluntary Reporting)

Combined-Cycle Units and Block Design Data (Voluntary Reporting) Unit Design Data Combined-Cycle Units and Block Design Data (Voluntary Reporting) (Note: The NERC Board of Trustees approved the GADS Task Force report (dated July 20, 2011 here i which states that design

More information

In April 1986, unit 4 of the Chernobyl nuclear

In April 1986, unit 4 of the Chernobyl nuclear Safety of RBMK reactors: Setting the technical framework The IAEA's co-operative programme is consolidating the technical basis for further upgrading the safety of Chernobyl-type reactors by Luis Lederman

More information

Summary. LOCA incidents: Gas and liquid metal cooled reactors. List of LOCA incidents: 3-4

Summary. LOCA incidents: Gas and liquid metal cooled reactors. List of LOCA incidents: 3-4 Summary NTEC Module: Water Reactor Performance and Safety Lecture 13: Severe Accidents II Examples of Severe Accidents G. F. Hewitt Imperial college London List of LOCA incidents: 3-4 Water cooled reactors

More information

Equipment Design. Detailed Plant Conceptual Design. Version 9.0

Equipment Design.  Detailed Plant Conceptual Design. Version 9.0 Equipment Design Version 9.0 Detailed Plant Conceptual Design SOAPP CT sizes all major plant equipment, based on your Project Input, the process configuration derived from this input, and the results of

More information

CAREM Prototype Construction and Licensing Status

CAREM Prototype Construction and Licensing Status IAEA-CN-164-5S01 CAREM Prototype Construction and Licensing Status H. Boado Magan a, D. F. Delmastro b, M. Markiewicz b, E. Lopasso b, F. Diez, M. Giménez b, A. Rauschert b, S. Halpert a, M. Chocrón c,

More information

1/58 Components of solar systems

1/58 Components of solar systems 1/58 Components of solar systems storage heat exchangers safety and protection devices air vents, check valve control & measurement Thermosiphon circulation system 2/58 circulation induced by buoyancy

More information

CLASSIFICATION OF SYSTEMS, STRUCTURES AND COMPONENTS OF A NUCLEAR FACILITY

CLASSIFICATION OF SYSTEMS, STRUCTURES AND COMPONENTS OF A NUCLEAR FACILITY CLASSIFICATION OF SYSTEMS, STRUCTURES AND COMPONENTS OF A NUCLEAR FACILITY 1 Introduction 3 2 Scope of application 3 3 Classification requirements 3 3.1 Principles of safety classification 3 3.2 Classification

More information

APR1400 Safe, Reliable Technology

APR1400 Safe, Reliable Technology APR1400 Safe, Reliable Technology OECD/NEA Workshop on Innovations in Water-cooled Reactor Technology Paris, Feb 11 12, 2015 Presented by Shin Whan Kim Contents 1. Introduction 2. Major Safety Design Characteristics

More information

HYGIENIC DESIGN ASPECTS OF PASTEURIZER TO PREVENT CROSS CONTAMINATION OF PASTEURIZED MILK

HYGIENIC DESIGN ASPECTS OF PASTEURIZER TO PREVENT CROSS CONTAMINATION OF PASTEURIZED MILK Review paper UDC 637.133.3 HYGIENIC DESIGN ASPECTS OF PASTEURIZER TO PREVENT CROSS CONTAMINATION OF PASTEURIZED MILK Prabhakar Kanade 1, Asaithambi Subramani 1* 1 Mother Dairy Fruit & Vegetable Private

More information

Review Article Analyses of the OSU-MASLWR Experimental Test Facility

Review Article Analyses of the OSU-MASLWR Experimental Test Facility Hindawi Publishing Corporation Science and Technology of Nuclear Installations Volume 212, Article ID 528241, 19 pages doi:1.1155/212/528241 Review Article Analyses of the OSU-MASLWR Experimental Test

More information

System & Equipment Checkout Procedure Client Name Plant Location

System & Equipment Checkout Procedure Client Name Plant Location System & Equipment Checkout Procedure Client Name Plant Location Procedure Title: Boiler Feed Water System Prepared By: Applied Performance Strategies 1 of 7 Approved By: Date: 01/11/08 Date: Purpose:

More information

AP1000 European 19. Probabilistic Risk Assessment Design Control Document

AP1000 European 19. Probabilistic Risk Assessment Design Control Document 19.39 In-Vessel Retention of Molten Core Debris 19.39.1 Introduction In-vessel retention of molten core debris through water cooling of the external surface of the reactor vessel is a severe accident management

More information

PSP Series Water Purification Systems

PSP Series Water Purification Systems Operations and Maintenance Manual for PSP Series Water Purification Systems PSP-1000 PSP-1600 PSP-2700 Rev. 2 / 08/26/14 Compact Design, Self-Contained Water Processing Unit for Point-Of-Use or Point-Of-Entry

More information

The 2011 Tohoku Pacific Earthquake and Current Status of Nuclear Power Stations

The 2011 Tohoku Pacific Earthquake and Current Status of Nuclear Power Stations The 2011 Tohoku Pacific Earthquake and Current Status of Nuclear Power Stations March 31, 2011 Tokyo Electric Power Company Tohoku Pacific Ocean Earthquake Time: 2:46 pm on Fri, March 11, 2011. Place:

More information

UKEPR Issue 04

UKEPR Issue 04 Title: PCSR Sub-chapter 10.4 Other features of steam and power conversion systems Total number of pages: 29 Page No.: I / III Chapter Pilot: M. LACHAISE Name/Initials Date 25-06-2012 Approved for EDF by:

More information

The Westinghouse AP1000 Advanced Nuclear Plant Plant description

The Westinghouse AP1000 Advanced Nuclear Plant Plant description The Westinghouse AP1000 Advanced Nuclear Plant Plant description Copyright 2003, Westinghouse Electric Co., LLC. All rights reserved. Table of Contents 1 Introduction 1 2 Description of the nuclear systems

More information

BfS SAFETY CODES AND GUIDES - TRANSLATIONS. Edition 08/97. Contents. Bundesamt für Strahlenschutz Salzgitter

BfS SAFETY CODES AND GUIDES - TRANSLATIONS. Edition 08/97. Contents. Bundesamt für Strahlenschutz Salzgitter BfS SAFETY CODES AND GUIDES - TRANSLATIONS Edition 08/97 Contents Guides for the Periodic Safety Review of Nuclear Power Plants Basics of the Periodic Safety Review Safety Status Analysis Probabilistic

More information

Chapter 8. Vapor Power Systems

Chapter 8. Vapor Power Systems Chapter 8 Vapor Power Systems Introducing Power Generation To meet our national power needs there are challenges related to Declining economically recoverable supplies of nonrenewable energy resources.

More information

Downsizing a Claus Sulfur Recovery Unit

Downsizing a Claus Sulfur Recovery Unit INFRASTRUCTURE MINING & METALS NUCLEAR, SECURITY & ENVIRONMENTAL Downsizing a Claus Sulfur Recovery Unit OIL, GAS & CHEMICALS By Charles L. Kimtantas and Martin A. Taylor ckimtant@bechtel.com & mataylo1@bechtel.com

More information

University of Houston Master Construction Specifications Insert Project Name SECTION WATER TREATMENT SYSTEMS PART 1 - GENERAL

University of Houston Master Construction Specifications Insert Project Name SECTION WATER TREATMENT SYSTEMS PART 1 - GENERAL SECTION 23 25 00 - WATER TREATMENT SYSTEMS PART 1 - GENERAL 1.1 RELATED DOCUMENTS: A. The Conditions of the Contract and applicable requirements of Division 1, "General Requirements", and Section 23 01

More information

What Nuclear Reactor Companies Need

What Nuclear Reactor Companies Need September 6, 2017 What Nuclear Reactor Companies Need Scott Bailey Vice President, Supply Chain NuScale Nonproprietary Copyright 2017 NuScale Power, LLC Acknowledgement & Disclaimer This material is based

More information

Development and use of SAMGs in the Krško NPP

Development and use of SAMGs in the Krško NPP REPUBLIC OF SLOVENIA Development and use of SAMGs in the Krško NPP Tomaž Nemec Slovenian Nuclear Safety Administration tomaz.nemec@gov.si IAEA TM on the Verification and Validation of SAMGs, Vienna, 12

More information

Heller Indirect Dry Cooling System: Dry cooling solution from the polar circle to hot deserts

Heller Indirect Dry Cooling System: Dry cooling solution from the polar circle to hot deserts Heller Indirect Dry Cooling System: Dry cooling solution from the polar circle to hot deserts In the developing Russian power industry, projects across the country set an example of the wide application

More information

Reduction Systems and Equipment at Moss Landing Power Plant

Reduction Systems and Equipment at Moss Landing Power Plant B&W s NO x Reduction Systems and Equipment at Moss Landing Power Plant Presented to: ICAC NO x Forum March 23-24, 2000 Washington D.C., U.S.A. Bill Becker Don Tonn Babcock & Wilcox Barberton, Ohio, U.S.A.

More information

Guidance page for practical work 15: modeling of the secondary circuit of a PWR

Guidance page for practical work 15: modeling of the secondary circuit of a PWR Guidance page for practical work 15: modeling of the secondary circuit of a PWR 1) Objectives of the practical work The aim is to investigate the potential of Thermoptim in modeling and calculation of

More information

Research Article Extended Station Blackout Coping Capabilities of APR1400

Research Article Extended Station Blackout Coping Capabilities of APR1400 Science and Technology of Nuclear Installations, Article ID 941, 1 pages http://dx.doi.org/1.1155/214/941 Research Article Extended Station Blackout Coping Capabilities of APR14 Sang-Won Lee, Tae Hyub

More information

Evaluation of AP1000 Containment Hydrogen Control Strategies for Post- Fukushima Lessons Learned

Evaluation of AP1000 Containment Hydrogen Control Strategies for Post- Fukushima Lessons Learned Evaluation of AP1000 Containment Hydrogen Control Strategies for Post- Fukushima Lessons Learned James H. Scobel and Hong Xu Westinghouse Electric Company, EEC 1000 Westinghouse Dr. Cranberry Township,

More information

OPG Proprietary Report

OPG Proprietary Report N/A R000 2 of 101 Table of Contents Page List of Tables and Figures... 5 Revision Summary... 6 Executive Summary... 7 1.0 INTRODUCTION... 9 1.1 Objectives... 10 1.2 Scope... 10 1.3 Organization of Summary...

More information

BARC BARC PASSIVE SYSTEMS RELIABILITY ANALYSIS USING THE METHODOLOGY APSRA. A.K. Nayak, PhD

BARC BARC PASSIVE SYSTEMS RELIABILITY ANALYSIS USING THE METHODOLOGY APSRA. A.K. Nayak, PhD BARC PASSIVE SYSTEMS RELIABILITY ANALYSIS USING THE METHODOLOGY APSRA A.K. Nayak, PhD Reactor Engineering Division Bhabha Atomic Research Centre Trombay, Mumbai 400085, India INPRO Consultancy Meeting

More information

DEVELOPMENT AND APPLICATION OF PROBABILISTIC SAFETY ASSESSMENT PSA IN DAYA BAY NUCLEAR POWER STATION

DEVELOPMENT AND APPLICATION OF PROBABILISTIC SAFETY ASSESSMENT PSA IN DAYA BAY NUCLEAR POWER STATION 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18) Beijing, China, August 7-12, 2005 SMiRT18-A01-2 DEVELOPMENT AND APPLICATION OF PROBABILISTIC SAFETY ASSESSMENT PSA

More information

Status report Advanced Power Reactor (APR1000)

Status report Advanced Power Reactor (APR1000) Status report 103 - Advanced Power Reactor (APR1000) Overview Full name Acronym Reactor type Coolant Moderator Neutron spectrum Thermal capacity Gross Electrical capacity Design status Designers Advanced

More information

Safety enhancement of NPPs in China after Fukushima Accident

Safety enhancement of NPPs in China after Fukushima Accident Safety enhancement of NPPs in China after Fukushima Accident CHAI Guohan 29 June 2015, Brussels National Nuclear Safety Administration, P. R. China Current Development of Nuclear Power Mid of year 2015

More information

OPG Proprietary Report

OPG Proprietary Report N/A R001 2 of 121 Table of Contents Page List of Tables and Figures... 5 Revision Summary... 6 Executive Summary... 7 1.0 INTRODUCTION... 9 1.1 Objectives... 10 1.2 Scope... 10 1.3 Organization of Summary...

More information

Steam Power Station (Thermal Station)

Steam Power Station (Thermal Station) Steam Power Station (Thermal Station) A generating station which converts heat energy into electrical energy through turning water into heated steam is known as a steam power station. A steam power station

More information

AP1000: The PWR Revisited

AP1000: The PWR Revisited AP1000: The PWR Revisited IAEA International Conference on Opportunities and Challenges for Water Cooled Reactors in the 21st Century aolo Gaio estinghouse Electric Company tober 27, 2009 1 Background

More information

SELECTED VALIDATION CASES RELATED TO NUCLEAR SAFETY ANALYSES

SELECTED VALIDATION CASES RELATED TO NUCLEAR SAFETY ANALYSES VTT Jukka Ylijoki, Ismo Karppinen, Eija-Karita Puska, Ari Silde May, 2015 FORTUM Kari Porkholm, Harri Kontio APROS VALIDATION SELECTED VALIDATION CASES RELATED TO NUCLEAR SAFETY ANALYSES and TRAINING SIMULATORS

More information

HPR1000: ADVANCED PWR WITH ACTIVE AND PASSIVE SAFETY FEATURES

HPR1000: ADVANCED PWR WITH ACTIVE AND PASSIVE SAFETY FEATURES HPR1000: ADVANCED PWR WITH ACTIVE AND PASSIVE SAFETY FEATURES D. SONG China Nuclear Power Engineering Co., Ltd. Beijing, China Email: songdy@cnpe.cc J. XING China Nuclear Power Engineering Co., Ltd. Beijing,

More information

RESULTS OF THE GRADUAL UPGRADING AT BOHUNICE WWER - 440/230 NPP

RESULTS OF THE GRADUAL UPGRADING AT BOHUNICE WWER - 440/230 NPP RESULTS OF THE GRADUAL UPGRADING AT BOHUNICE WWER - 440/230 NPP P. Krupa Ingeneer, e-mail: Krupa_Peter@ebo.seas.sk Bohunice NPPs Introduction The centre of upgrading activities in VVER NPP is clearly in

More information

Balance of Plant Requirements and Concepts for Tokamak Reactors

Balance of Plant Requirements and Concepts for Tokamak Reactors Balance of Plant Requirements and Concepts for Tokamak Reactors Edgar Bogusch EFET / Framatome ANP GmbH 9 th Course on Technology of Fusion Tokamak Reactors Erice, 26 July to 1 August 2004 1 Contents Introduction

More information

Chapter 2.6: FBC Boilers

Chapter 2.6: FBC Boilers Part-I: Objective type questions and answers Chapter 2.6: FBC Boilers 1. In FBC boilers fluidization depends largely on --------- a) Particle size b) Air velocity c) Both (a) and (b) d) Neither (a) nor

More information

ENERGY RECOVERY IMPROVEMENT USING ORGANIC RANKINE CYCLE AT COVANTA S HAVERHILL FACILITY

ENERGY RECOVERY IMPROVEMENT USING ORGANIC RANKINE CYCLE AT COVANTA S HAVERHILL FACILITY Proceedings of the 18th Annual North American Waste-to-Energy Conference NAWTEC18 May 11-13, 2010, Orlando, Florida, USA Paper Number: NAWTEC18-3563 ENERGY RECOVERY IMPROVEMENT USING ORGANIC RANKINE CYCLE

More information

SAFETY AND PERFORMANCE ACHIEVEMENT OF INDIAN NUCLEAR POWER PLANT. Randhir kumar NPCIL Shift charge engineer, TAPS 3&4, NPCIL, INDIA

SAFETY AND PERFORMANCE ACHIEVEMENT OF INDIAN NUCLEAR POWER PLANT. Randhir kumar NPCIL Shift charge engineer, TAPS 3&4, NPCIL, INDIA POSTER PRESENTATION IAEA-CN-164 International Conference on opportunities and challenges for water cooled reactors in the 21 st century- Vienna, Austria. 27-3 October 29 SAFETY AND PERFORMANCE ACHIEVEMENT

More information

Application of the Defense-in-Depth Concept in the Projects of New-Generation NPPs Equipped with VVER Reactors. JSC ATOMENERGOPROEKT Moscow

Application of the Defense-in-Depth Concept in the Projects of New-Generation NPPs Equipped with VVER Reactors. JSC ATOMENERGOPROEKT Moscow Application of the Defense-in-Depth Concept in the Projects of New-Generation NPPs Equipped with VVER Reactors Yu. Shvyryaev V. Morozov A. Kuchumov JSC ATOMENERGOPROEKT Moscow Content General information

More information

Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor

Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor Yuliang SUN Institute of Nuclear and New Energy Technology, Tsinghua University Beijing 100084, PR China 1 st Workshop on PBMR Coupled

More information

Reclaimed Water Treatment, Distribution, Pumping and Storage Engineering Report Form

Reclaimed Water Treatment, Distribution, Pumping and Storage Engineering Report Form Oklahoma Department of Environmental Quality Water Quality Division Phone: 405-702-8100 Construction Permitting Section 707 N. Robinson, OKC, OK 73102-6010 P.O. Box 1677, OKC, OK 73101-1677 Reclaimed Water

More information

System Analysis of Pb-Bi Cooled Fast Reactor PEACER

System Analysis of Pb-Bi Cooled Fast Reactor PEACER OE-INES-1 International Symposium on Innovative Nuclear Energy Systems for Sustainable Development of the World Tokyo, Japan, October 31 - November 4, 2004 System Analysis of Pb-Bi ooled Fast Reactor PEAER

More information

National Maritime Center

National Maritime Center National Maritime Center Providing Credentials to Mariners U.S.C.G. Merchant Marine Exam (Sample Examination) Page 1 of 24 Choose the best answer to the following Multiple Choice Questions. 1. On a main

More information

Safety design approach for JSFR toward the realization of GEN-IV SFR

Safety design approach for JSFR toward the realization of GEN-IV SFR Safety design approach for JSFR toward the realization of GEN-IV SFR Advanced Fast Reactor Cycle System R&D Center Japan Atomic Energy Agency (JAEA) Shigenobu KUBO Contents 1. Introduction 2. Safety design

More information

The Commission Errors Search and Assessment (CESA) Method

The Commission Errors Search and Assessment (CESA) Method P A U L S C H E R R E R I N S T I T U T PSI Bericht Nr. 07-03 May 2007 ISSN 1019-0643 Laboratory for Energy Systems Analysis (LEA) The Commission Errors Search and Assessment (CESA) Method Bernhard Reer,

More information

LBLOCA AND DVI LINE BREAK TESTS WITH THE ATLAS INTEGRAL FACILITY

LBLOCA AND DVI LINE BREAK TESTS WITH THE ATLAS INTEGRAL FACILITY LBLOCA AND DVI LINE BREAK TESTS WITH THE ATLAS INTEGRAL FACILITY WON-PIL BAEK *, YEON-SIK KIM and KI-YONG CHOI Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute 1045 Daedeokdaero,

More information

Status report 97 - Advanced Boiling Water Reactor (ABWR)

Status report 97 - Advanced Boiling Water Reactor (ABWR) Status report 97 - Advanced Boiling Water Reactor (ABWR) Overview Full name Acronym Reactor type Coolant Moderator Neutron spectrum Thermal capacity Gross Electrical capacity Design status Designers Advanced

More information

Steam Cooling Systems and Hybrid Cooling. Andrew G. Howell Xcel Energy

Steam Cooling Systems and Hybrid Cooling. Andrew G. Howell Xcel Energy Steam Cooling Systems and Hybrid Cooling Andrew G. Howell Xcel Energy Steam Cooling Systems Once-through Recirculating Cooling Tower Direct Dry Cooling (air-cooled condenser) Indirect Dry Cooling (Heller)

More information