FUKUSHIMA DAIICHI BWR REACTOR SPRAY AND FEED WATER SYSTEMS EVALUATION FOR EARLY FAILURE Dean Wilkie
|
|
- Matthew Pearson
- 6 years ago
- Views:
Transcription
1 FUKUSHIMA DAIICHI BWR REACTOR SPRAY AND FEED WATER SYSTEMS EVALUATION FOR EARLY FAILURE Dean Wilkie The BWR reactor vessel spray(core spray) and feed water spray systems are designed to inject water into the spray header loops which encircle the inside of the reactor vessel at two different levels. The feed water injects water into the feed water spray loop at an elevation just above the shroud head. The vessel spray water injects water into the spray loop just below the top of the reactor shroud. Each system is independent of the other and operates at different pressures to cover high and low pressure conditions in the reactor vessel as well as varying vessel water level. Each spray system injects water into spray spargers which then direct the flow to the loops within the vessel that have multiple spray heads attached to direct the flow to the reactor core. (see figures later in paper) CORE SPRAY SPARGER Two independent core spray loops (core spray and feed water spray) are installed in the vessel above the upper core grid (top guide) and within the core shroud. The loops are connected to the Core Spray System which is used for core cooling under loss of coolant accident conditions. THE FEED WATER SPRAY SPARGER The feedwater spargers are mounted to the reactor vessel wall in the upper part of the downcomer or annulus region. The spargers, each supplied by one of the two feedwater nozzles, complete a half, circle of the vessel interior and discharges water radially inward. A number of 1 inch holes in each sparger permits the cooler feed water to mix with downcomer recirculation flow before coming in contact with the vessel. THE CORE SPRAY SYSTEM The Core Spray System: Is a low pressure system which supplies cooling water after reactor pressure is reduced to 285 psig (1.96 MPa) Will prevent the reactor from overheating following intermediate or, large break Accommodates some intermediate to small pipe breaks when feedwater is not available, the automatic depressurization system (ADS) will depressurize the reactor thus permitting the Core Spray System to provide core cooling The Core Spray System consists of two identical loops. Each loop contains: Two main pumps, a spray sparger, and associated piping, instrumentation and controls
2 Pumps rated at 3400 gpm full flow capacity. Two booster pumps Two sets of parallel isolation valves, one set inside and the other outside the drywell Water is supplied to the system from the suppression pool. The Fire Protection System is connected to each of the core spray loops to provide a backup supply of water. Each loop has a test recirculation line to the suppression pool for full flow testing without discharging into the reactor vessel. The piping up to the test valve is carbon steel, designed for 400 psig (2.75 Mpa) and 350 F (177 C). From the injection isolation valves to the reactor vessel, the piping is stainless steel designed for 1250 psig (2.75 Mpa and 575 F (302 C). A core spray filling system maintains the Core Spray System at full capacity to preclude any danger of water hammer when the system goes in operation. The discharge from each of the main pumps flows through a check valve to a common header that supplies water to the booster pumps and a bypass line around the booster pumps. The booster pumps discharge piping contains motor operated isolation valves outside the drywell and air operated testable check valves inside the drywell. Flow from each loop is directed from the pumps through two parallel normally closed motor operated valves, a single line at the containment penetration, two parallel check valves, one locked open manually operated valve and then into the sparger. Both Core Spray systems and their diesel generators will automatically start upon the detection of one high drywell pressure, or one low reactor vessel water level condition. These conditions generally indicate a pipe break. The system can also be manually initiated by the control room operators.
3 HIGH PRESSURE CORE SPRAY (HPCS) This system provides high pressure emergency core cooling for small, intermediate, and large line breaks. It is a single loop system and consists of: o A suction shutoff valve o One motor drive pump o Discharge check valve o Motor operated injection valve o Minimum flow valve o Full flow test valve to the suppression pool o Two high pressure flow test valves to the condensate storage tank o Discharge sparger and o Associated piping and instrumentation. o HPCS takes suction from the condensate storage tank or suppression pool The HPCS behaviors: Pumps the water into a sparger located on the upper core shroud Spray nozzles mounted on the sparger are directed at the top of the fuel assemblies to remove decay heat following a loss of coolant accident (LOCA) The suppression pool is the alternate source of Water for the HPCS system Normal power for the HPCS is provided by Standby Power System division 3 Diesel Generator The HPCS initiates automatically on high pressure in the drywell or low water level in the reactor vessel (level 2) FIGURES
4 Core spray loops and position relative to the shroud for the vessel spray line Feedwater sparger attachment and distribution header
5 Reactor vessel Cutaway with component labeling
6 Reactor vessel cutaway with component labeling
7 CORE SPRAY LINE ISSUES The feedwater inlet nozzle on some boiling water reactor (BWR) vessels contains a stainless steel cladding over its carbon steel base material, the combined total being approximately 6" (15 cm) thick. Significant thermal stresses on the inner radius of such nozzles have in some cases caused in service thermal fatigue cracking of the cladding, having the potential to propagate into the base material. This problem was first defined in 1975 by the US Nuclear Regulatory Commission (NRC) and associated BWR suppliers. Two remedies were proposed, one being the removal of the cladding and the other being a modification to the feedwater nozzle sparger or both. At this time, cracks were identified at a given BWR plant using penetrant and ultrasonic testing (PT & UT) and removed. Grind outs to remove the cracks resulted in divots in the cladding. Mock up of the BWR reactor vessel penetration nozzle for the feed lines to the spray systems. OBSERVATION As newer systems are developed for testing for these stress related cracks more refined data can reveal not only existing, but also potential crack formation. These nozzles are assumed to undergo the normal stresses for reactor operation and not the extreme temperature and stress conditions that existed at the Fukushima reactors during the course of the accidents. Potential piping degrading is likely to happen not only on the feed water but also the internal spray lines for each core spray system.
8 These spray lines were counted on by TEPCO at the Fukushima reactors to deliver water critical to ensuring the reactor pressure vessel (RPV) and fuel remained cooled. TEPCO made connections to the feed water lines which ultimately would reach the feedwater and core spray spargers and onto the loop nozzles. There has been no confirmation that these feedwater lines were still intact outside or inside the reactor vessel or in containment. Accident analyses for severe accidents often treat the normal components in the BWR reactor such as the steam separator and dryer as general core material. portal.org/smash/get/diva2:224035/fulltext01.pdfand The analyses for severe accidents including meltdown and fuel dry out conditions assume the fuel melt temperature levels increase to a point where these excessive temperatures heat the core material to extreme levels which can act to deform or melt. The conditions at the Fukushima reactor units 1 3 resulted in this very melt condition and as a result have likely deformed or melted the internal piping and spray loops which have been borne out by leakage further down to the torus room as well as water running out of the MSIV doorway on unit 3. TEPCO has kept water flowing through both the core spray and feedwater lines since they initially hooked up these temporary systems. They have claimed that the reactor vessel and systems have been kept cool however, there is no known instrumentation other than that of the reactor vessel walls which can be reliably counted on to demonstrate sufficient cooling. TEPCO has used selected temperature instrument readings to claim that units 1 3 are in a cold shutdown mode. Stating this condition is illogical because cold shutdown mode consists of many other factors other than temperature to declare the reactor has entered into the cold shutdown official mode. DISCUSSION np/plant data/f1_2_chart3.pdf Paper tape recordings were checked on Units 1 3 and they don't show any flow for either core sprayer. The units are in 50 liters/second. Mind you that what Tepco injects would only be about 1 liter/s and below the resolution of the tape recordings. So if the gauge worked and any water was injected through the core sprays at all, the rate must have been much less than anticipated when the recordings were dimensioned. (quote from simplyinfo.org member)
9 Since the structure temperatures close above the core were high (T=l100 K K) prior to lower head dryout content/attachment/cla udiotorregrosa_mss_thesis.pdf (Note: the report is talking about the injection systems that were not available at normal capacity at the Fukushima reactors 1 3) In the event that any of these injection systems is available and core cooling is not regained, the water level in the vessel is expected to be below the core lower plate but still filling the reactor lower plenum. The uncovered fuel rods and reactor core materials (including reactor internal structures, control rods, instrumentation tubes, etc.) will overheat, melt and drain by gravity to the lower plenum does a melted nuclear core look like/ This link contains video footage of the inside of the TMI 2 reactor for initial inspection as well as preparations and fuel debris removal. The footage shows melting of structural components just above the reactor core as well as other peripheral structural melting during the accident. TMI was able to, in time, restore water to the core thus preventing more significant damage as that incurred at the Fukushima reactor unit 1 3.
10 image capture from TMI video
11 image capture from TMI video
12 image capture TMI video COMMENTS and CONCLUSIONS TEPCO, in the haste and crisis reactions knew they needed to get some sort of water fed into the reactor vessels of Units 1 3. In their desperation they set up portable and temporary systems as they chose locations to tie into the reactor core spray and feedwater systems. This decision, with no knowledge of the statues of the systems within the reactor building including the reactor vessel was the best they thought could be done at the time. Since these temporary hookups were made there have been varying claims, although not proven, that the water is somehow keeping the core and melted fuel cooled. Final Thoughts: The spray and feedwater lines at units 1 3 may be feeding water toward the reactor nozzles but very likely have s suffered damage from the earthquake and explosions to the point of developing leakage outside the reactor vessels. The core internal materials associated with the separators and core spray lines have
13 degraded, slumped or melted and are not delivering water sufficient to cool as designed. The upper portion of the steam separator and reactor feedwater spray loops have also degraded, slumped partially and have reached levels of cracking or failure although not as severe as the lower areas including the core spray lines. Water delivery through the spray nozzles in the feed water system has been minimized to less that designed Neither of the systems are providing design flow and are operating in a very degraded state at low pressure.
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 informationAP1000 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 informationIsolation Condenser; water evaporation in the tank and steam into the air. Atmosphere (in Severe Accident Management, both P/S and M/S)
Loss of Ultimate Heat Sink ANS AESJ AESJ Fukushima Symposium, March h4, 2012 Hisashi Ninokata, Tokyo Institute of Technology Available ultimate heat sinks at 1F1~3 1F1 (Fukushima Dai ichi Unit 1) Sea water
More informationWestinghouse 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 informationStudy on Severe Accident Progression and Source Terms in Fukushima Dai-ichi NPPs
Study on Severe Accident Progression and Source Terms in Fukushima Dai-ichi NPPs October 27, 2014 H. Hoshi, R. Kojo, A. Hotta, M. Hirano Regulatory Standard and Research Department, Secretariat of Nuclear
More informationNSSS Design (Ex: PWR) Reactor Coolant System (RCS)
NSSS Design (Ex: PWR) Reactor Coolant System (RCS) Purpose: Remove energy from core Transport energy to S/G to convert to steam of desired pressure (and temperature if superheated) and moisture content
More informationAP1000 European 15. Accident Analysis Design Control Document
15.2 Decrease in Heat Removal by the Secondary System A number of transients and accidents that could result in a reduction of the capacity of the secondary system to remove heat generated in the reactor
More informationMajor Influential Issues on the Accident Progressions of Fukushima Daiichi NPP
1/12 Major Influential Issues on the Accident Progressions of Fukushima Daiichi NPP Summary M. Naitoh, M. Pellegrini. H. Suzuki, H. Mizouchi, and H. Okada The Institute of Applied Energy, Japan This paper
More informationNPP 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 informationEnsuring Spent Fuel Pool Safety
Ensuring Spent Fuel Pool Safety Michael Weber Deputy Executive Director for Operations U.S. Nuclear Regulatory Commission American Nuclear Society Meeting June 28, 2011 1 Insights from Fukushima Nuclear
More informationRadiation Control in the Core Shroud Replacement Project of Fukushima-Daiichi NPS Unit #2
General Radiation Control in the Core Shroud Replacement Project of Fukushima-Daiichi NPS Unit #2 Yasunori Kokubun, Kazuyuki Haraguchi, Yuji Yoshizawa, Yasuo Yamada Tokyo Electric Power Co, 22 Kitahara
More informationVerification 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 informationThe 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 informationNuclear 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 informationPost-Fukushima Assessment of the AP1000 Plant
ABSTRACT Post-Fukushima Assessment of the AP1000 Plant Ernesto Boronat de Ferrater Westinghouse Electric Company, LLC Padilla 17-3 Planta 28006, Madrid, Spain boronae@westinghouse.com Bryan N. Friedman,
More informationThe 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 informationSMR/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 informationJoint 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 informationThe Westinghouse AP1000 : Passive, Proven Technology to Meet European Energy Demands
BgNS TRANSACTIONS volume 20 number 2 (2015) pp. 83 87 The Westinghouse AP1000 : Passive, Proven Technology to Meet European Energy Demands N. Haspel Westinghouse Electric Germany GmbH, Dudenstraße 6, 68167
More informationCANDU 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 informationRELAP 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 informationNUCLEAR 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 informationSummary. 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 informationModule 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 informationHPR1000: 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 informationRELAP5/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 informationConcepts and Features of ATMEA1 TM as the latest 1100 MWe-class 3-Loop PWR Plant
8 Concepts and Features of ATMEA1 TM as the latest 1100 MWe-class 3-Loop PWR Plant KOZO TABUCHI *1 MASAYUKI TAKEDA *2 KAZUO TANAKA *2 JUNICHI IMAIZUMI *2 TAKASHI KANAGAWA *3 ATMEA1 TM is a 3-loop 1100
More informationThe 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 informationShutdown and Cooldown of SEE-THRU Nuclear Power Plant for Student Performance. MP-SEE-THRU-02 Rev. 004
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 TABLE OF CONTENTS 1. PURPOSE...3
More informationApplication 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 informationNuclear 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 informationStatus report Chinese Supercritical Water-Cooled Reactor (CSR1000)
Status report Chinese Supercritical Water-Cooled Reactor (CSR1000) Overview Full name Chinese Supercritical Water-Cooled Reactor Acronym CSR1000 Reactor type Pressurized Water Reactor (PWR) Coolant Supercritical
More informationSafety 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 informationRisks 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 informationNuclear 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 informationExperiences from Application of MELCOR for Plant Analyses. Th. Steinrötter, M. Sonnenkalb, GRS Cologne March 2nd, 2010
Experiences from Application of MELCOR 1.8.6 for Plant Analyses Th. Steinrötter, M. Sonnenkalb, GRS Cologne March 2nd, 2010 Content Introduction MELCOR 1.8.6 Analyses for the Atucha II Power Plant Modeling
More informationACR-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 informationSpent Fuel and Spent Fuel Storage Facilities at Fukushima Daiichi
Spent Fuel and Spent Fuel Storage Facilities at Fukushima Daiichi The National Academies Lessons Learned from the Fukushima Nuclear Accident for Improving Safety and Security of U.S. Nuclear Plants: Phase
More informationModule 06 Boiling Water Reactors (BWR)
Module 06 Boiling Water Reactors (BWR) 1.10.2015 Prof.Dr. Böck Vienna University oftechnology Atominstitute Stadionallee 2 A-1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Contents BWR Basics
More informationFrequently Asked Questions: Japanese Nuclear Energy Situation
NUCLEAR ENERGY INSTITUTE Frequently Asked Questions: Japanese Nuclear Energy Situation 1. What will be the impact of the Fukushima Daiichi accident on the U.S. nuclear program? It is premature to draw
More informationCRFA System B BASES
CRFA System B 3.7.2 APPLICABLE SAFETY ANALYSES LCO and 15 ( Refs. 3 and 4, respectively). The isolation mode of the CRFA System is assumed to operate following a DBA. The radiological doses to CRE occupants
More informationPreparedness at PFBR Kalpakkam to meet the challenges due to Natural events Prabhat Kumar, Project Director, PFBR, Director construction BHAVINI
BHARATIYA NABHIKIYA VIDYUT NIGAM LIMITED (A Government of India Enterprise) Preparedness at PFBR Kalpakkam to meet the challenges due to Natural events Prabhat Kumar, Project Director, PFBR, Director construction
More informationNuclear 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 informationResearch 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 informationAssessing and Managing Severe Accidents in Nuclear Power Plant
Assessing and Managing Severe Accidents in Nuclear Power Plant Harri Tuomisto Fortum, Finland IAEA Technical Meeting on Managing the Unexpected - From the Perspective of the Interaction between Individuals,
More informationAdvanced LWRs Jacopo Buongiorno Associate Professor of Nuclear Science and Engineering
Advanced LWRs Jacopo Buongiorno Associate Professor of Nuclear Science and Engineering 22.06: Engineering of Nuclear Systems Outline Performance goals for near-term advanced LWRs Technical features of
More informationSmall 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 informationOVER VIEW OF ACCIDENT OF FUKUSHIMA DAI-ICHI NPSs AND FUTURE PLANNING TOWARD D&D
OVER VIEW OF ACCIDENT OF FUKUSHIMA DAI-ICHI NPSs AND FUTURE PLANNING TOWARD D&D 16 NOVEMBER, 2011 Hiroshi RINDO JAPAN ATOMIC ENERGY AGENCY Table of Contents 1. What s happened at Fukushima Dai-Ichi NPSs
More informationSide Event of IAEA General Conference. Severe Accident Analyses of Fukushima-Daiich Units 1 to 3. Harutaka Hoshi and Masashi Hirano
Side Event of IAEA General Conference Severe Accident Analyses of Fukushima-Daiich Units 1 to 3 Harutaka Hoshi and Masashi Hirano Japan Nuclear Energy Safety Organization (JNES) September 17, 2012 Side
More informationMain 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 informationSafety Design Requirements and design concepts for SFR
Safety Design Requirements and design concepts for SFR Reflection of lessons learned from the Fukushima Dai-ichi accident Advanced Nuclear System Research & Development Directorate Japan Atomic Energy
More informationOPERATING 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 informationLessons Learned from Fukushima Daiichi Nuclear Power Station Accident and Consequent Safety Improvements
Hitachi Review Vol. 62 (2013), No. 1 75 Lessons Learned from Fukushima Daiichi Nuclear Power Station Accident and Consequent Safety Improvements Masayoshi Matsuura Kohei Hisamochi Shinichiro Sato Kumiaki
More informationPassive 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 informationRegulatory Actions and Follow up Measures against Fukushima Accident in Korea
Int Conference on Effective Nuclear Regulatory Systems, April 9, 2013, Canada Regulatory Actions and Follow up Measures against Fukushima Accident in Korea Seon Ho SONG* Korea Institute of Nuclear Safety
More informationASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like
ASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like Lucas Albright a,b, Dr. Polina Wilhelm b, Dr. Tatjana Jevremovic a,c a Nuclear Engineering Program b Helmholtz-ZentrumDresden-Rossendorf
More informationwww.flexonics.com 1 800 854 2553 Senior Flexonics Pathway Division 2400 Longhorn Industrial Dr. New Braunfels, Texas 78130 Tel Int: 1 830 629 8080 Fax Int: 1 830 629 6899 E-mail: sales@pathway.flexonics.com
More informationCONTAINMENT STRUCTURES
CONTAINMENT STRUCTURES M. Ragheb 11/16/2017 INTRODUCTION A misconception about nuclear power plants containment structures is that their massive concrete construction is a protection against the release
More informationThe Materials Initiative NEI Overview
The Materials Initiative NEI 03-08 Overview Robin Dyle EPRI Annual Materials/NRC Technical Exchange June 2, 2015 Overview Primary system materials integrity is vital to plant performance and reliability
More informationAnalysis 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 informationIV. Occurrence and Development of the Accident at the Fukushima Nuclear Power Stations
IV. Occurrence and Development of the Accident at the Fukushima Nuclear Power Stations 1. Outline of Fukushima Nuclear Power Stations (1) Fukushima Daiichi Nuclear Power Station Fukushima Daiichi Nuclear
More informationNuScale SMR Technology
NuScale SMR Technology UK IN SMR; SMR IN UK Conference - Manchester, UK Tom Mundy, EVP Program Development September 25, 2014 Acknowledgement & Disclaimer This material is based upon work supported by
More informationModule 06 Boiling Water Reactors (BWR) Vienna University of Technology /Austria Atominstitute Stadionallee 2, 1020 Vienna, Austria
Module 06 Boiling Water Reactors (BWR) Prof.Dr. H. Böck Vienna University of Technology /Austria Atominstitute Stadionallee 2, 1020 Vienna, Austria Contents BWR Basics Technical Data Safety Features Reactivity
More informationImplementation 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 informationP Boiling Water Reactor Vessel and Internals Program (BWRVIP)
2018 Research Portfolio P41.01.03 - Boiling Water Reactor Vessel and Internals Program (BWRVIP) Program Description As boiling water reactors (BWRs) age, various types of materials degradation mechanisms
More informationEXPERIMENTS 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 informationTHREE 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 informationUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C August25, 2016
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August25, 2016 Mr. Jon A. Franke Site Vice President Susquehanna Nuclear, LLC 769 Salem Boulevard NUCSB3 Berwick, PA 18603-0467 SUBJECT:
More informationThe DENOPI project: a research program on SFP under loss-of-cooling and loss-of-coolant accident conditions
The DENOPI project: a research program on SFP under loss-of-cooling and loss-of-coolant accident conditions NAS meeting March 2015 N. Trégourès, H. Mutelle, C. Duriez, S. Tillard IRSN / Nuclear Safety
More informationVALIDATION OF RELAP5/MOD3.3 AGAINST THE PACTEL SBL-50 BENCHMARK TRANSIENT ABSTRACT
VALIDATION OF RELAP5/MOD3.3 AGAINST THE PACTEL SBL-50 BENCHMARK TRANSIENT J. Bánáti 1 *, V. Riikonen 2 ; V. Kouhia 2, H. Purhonen 2 1 Chalmers University of Technology, SE-41296 Gothenburg, Sweden 2 Lappeenranta
More informationAGEING MANAGEMENT PROGRAM TO REACTOR PRESSURE VESSEL INTERNALS COMPONENTS IN A BWR NUCLEAR POWER PLANT
AGEING MANAGEMENT PROGRAM TO REACTOR PRESSURE VESSEL INTERNALS COMPONENTS IN A BWR NUCLEAR POWER PLANT C. R. Arganis J. a, J. A. Aguilar T. a, M. A. Sanchez M. b a Instituto Nacional de Investigaciones
More informationContents of summary. 1. Introduction
Contents of summary 1. Introduction 2. Situation regarding Nuclear Safety Regulations and Other Regulatory Frameworks in Japan before the Accident 3. Disaster Damage in Japan from the Tohoku District -
More informationDesign 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 informationThe design features of the HTR-10
Nuclear Engineering and Design 218 (2002) 25 32 www.elsevier.com/locate/nucengdes The design features of the HTR-10 Zongxin Wu *, Dengcai Lin, Daxin Zhong Institute of Nuclear Energy and Technology, Tsinghua
More informationSafety Implication for Gen-IV SFR based on the Lesson Learned from the Fukushima Dai-ichi NPPs Accident. Ryodai NAKAI Japan Atomic Energy Agency
Safety Implication for Gen-IV SFR based on the Lesson Learned from the Fukushima Dai-ichi NPPs Accident Ryodai NAKAI Japan Atomic Energy Agency Contents Introduction Japanese Government Report to the IAEA
More informationRisks of Nuclear Ageing
Risks of Nuclear Ageing Technical characteristics of ageing processes and their possible impacts on nuclear safety in Spain S. Mohr, S. Kurth Greenpeace, Valencia, November 2014 Age of European NPPs (grid
More informationImprovements Needed in Nuclear Power Plant Probabilistic Risk Assessments: Lessons Learned from Fukushima
Improvements Needed in Nuclear Power Plant Probabilistic Risk Assessments: Lessons Learned from Fukushima Mohammad Modarres Professor of Nuclear Engineering Department of Mechanical Engineering University
More informationDEVELOPMENT 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 informationUnit 2 Primary Containment Vessel Internal Investigation
Unit 2 Primary Containment Vessel Internal November 30, 2017 Tokyo Electric Power Company Holdings, Inc. 1. Conditions inside the Unit 2 PCV According to accident development analysis, it is assumed that
More informationReview 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 informationPost-Fukushima Actions in Korea
Post-Fukushima Actions in Korea IAEA TWG-LWR Vienna, June 18-20, 2013 Presented by Jong-Tae Seo 1. Impacts on National Energy Policy and NPP Plan 2. Actions Taken after Fukushima Accident 3. Findings and
More informationPLANT VOGTLE UNITS 3 AND 4
PLANT VOGTLE UNITS 3 AND 4 ZERO GREENHOUSE GASES Nuclear energy facilities release zero greenhouse gases while producing electricity. A single uranium pellet the size of a pencil eraser produces as much
More informationSEVERE ACCIDENT FEATURES OF THE ALTERNATIVE PLANT DESIGNS FOR NEW NUCLEAR POWER PLANTS IN FINLAND
SEVERE ACCIDENT FEATURES OF THE ALTERNATIVE PLANT DESIGNS FOR NEW NUCLEAR POWER PLANTS IN FINLAND Risto Sairanen Radiation and Nuclear Safety Authority (STUK) Nuclear Reactor Regulation P.O.Box 14, FI-00881
More informationChapter 3 Accident of Fukushima Daiichi Nuclear Power Plant: Sequences, Fission Products Released, Lessons Learned
Chapter 3 Accident of Fukushima Daiichi Nuclear Power Plant: Sequences, Fission Products Released, Lessons Learned Jun Sugimoto Abstract The nuclear accident that occurred at the Fukushima Daiichi Nuclear
More informationLBLOCA 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 informationPressurized Thermal Shock Potential at Palisades
Pressurized Thermal Shock Potential at Palisades Prepared by Michael J. Keegan Coalition for a Nuclear Free Great Lakes (July 8, 1993, Rekeyed August 3, 2005) History of Embrittlement of Reactor Pressure
More informationCalculate the Costs of Piping System Elements
Calculate the Costs of Piping System Elements by Ray Hardee, Engineered Software, Inc. Last month s column described the process of creating an energy cost balance sheet for a piping system (see Figure
More informationCAREM: 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 informationThe 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 informationWhat 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 informationAppendix B. Aging Management Programs and Activities
Appendix B Aging Management Programs and Activities This page is intentionally blank. TABLE OF CONTENTS Table of Contents...i Appendix B: Aging Management Programs and Activities... B.1-1 B.1 INTRODUCTION...
More informationINVESTIGATION 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 informationGENERAL CONTENTS SECTION I - NUCLEAR ISLAND COMPONENTS
- June 2013 Addendum GENERAL CONTENTS SECTION I - NUCLEAR ISLAND COMPONENTS SUBSECTION "A" : GENERAL RULES SUBSECTION "B" : CLASS 1 COMPONENTS SUBSECTION "C" : CLASS 2 COMPONENTS SUBSECTION "D" : CLASS
More informationC. ASSE 1013 Performance Requirements for Reduced Pressure Principle Backflow Preventers.
PART 1: GENERAL 1.01 Purpose: A. This standard is intended to provide useful information to the Professional Service Provider (PSP) to establish a basis of design. The responsibility of the engineer is
More informationDownsizing 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 informationOPG 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 informationBWR Safety Improvement as a Lesson Learned from Fukushima Accident
BWR Safety Improvement as a Lesson Learned from Fukushima Accident M. M. Zaky and S. A. Kotb ETRR-2, Atomic Energy Authority, Cairo, Egypt Received: 20/5/2016 Accepted: 15/7/2016 ABSTRACT The serious accident
More informationPDHonline Course E438 (3 PDH) Nuclear Accidents. Charles A. Patterson, P.E. PDH Online PDH Center
PDHonline Course E438 (3 PDH) Nuclear Accidents Charles A. Patterson, P.E. 2014 PDH Online PDH Center 5272 Meadow Estates Drive Fairfax, VA 22030-6658 Phone & Fax: 703-988-0088 www.pdhonline.org www.pdhcenter.com
More informationEPR: 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