DOCSLIB.ORG

Light Water Reactor Hydrogen Manual

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Light Water Reactor Hydrogen Manual NUREG/CR-2726 SAND82-1137 R3 Printed August 1983 Light Water Reactor Hydrogen Manual Allen L. Camp, John C. Cummings, Martin P. Sherman, Chester F. Kupiec, Robert J. Healy, Jeffrey S. Caplan, John R. Sandhop, James H. Saunders Prepared by Sandia National Laboratories. W' , 0 Albuquerque, New Mexico 87185 and Livermore, California 94550 ' ,'j, , for the United States Department of Energy under Contract DE-AC04-76DP00789 i , and I, j, General Physics Corporation, 1000 Century Plaza , , Columbia, Maryland 21044 .f. tjI !i I:'• Vtj - I,; -,' Prepared for1 U. S. NUCLEAR REGULATORY COMMISSION NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employ- ees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. Available from GPO Sales Program Division of Technical Information and Document Control U.S. Nuclear Regulatory Commission Washington, D.C. 20555 and National Technical Information Service Springfield, Virginia 22161 NUREG/CR-2726 SAND82-1137 R3 LIGHT WATER REACTOR HYDROGEN MANUAL SANDIA NATIONAL LABORATORIES GENERAL PHYSICS CORPORATION Allen L. Camp Chester F. Kupiec John C. Cummings Robert J. Healy Martin P. Sherman Jeffrey S. Caplan John R. Sandhop James H. Saunders Manuscript Completed: January, 1983 Date Published: June, 1983 Sandia National Laboratories Albuquerque, NM 87185 Operated By Sandia Corporation for the U. S. Department of Energy Prepared for Division of Accident Evaluation Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Under Memorandum of Understanding DOE 40-550-75 NRC FIN No. A 1246 Author's Note While there are no current plans for updating this manual, persons who wish to point out errors or suggest revisions should send comments to the address shown below. If time and funding permits, we will attempt to pass on these comments to other recipients of the manual in an appropriate manner. Reactor Safety Studies Division 9441 Sandia National Laboratories P. 0. Box 5800 Albuquerque, NM 87185 ii ABSTRACT A manual concerning the behavior of hydrogen in light water reactors has been prepared. Both normal operations and accident situations are addressed. Topics considered include hydrogen generation, transport and mixing, detection, and combustion, and mitigation. Basic physical and chemical phenomena are described, and plant-specific examples are provided where appropriate. A wide variety of readers, including operators, designers, and NRC staff, will find parts of this manual useful. Different sections are written at different levels, according to the most likely audience. The manual is not intended to provide specific plant procedures, but rather, to provide general guidance that may assist in the development of such procuedures. iii TABLE OF CONTENTS Pag e Chapter 1 - Introduction and Objectives 1-1 1.1 Statement of the Problem 1-1 1.2 Hydrogen Problems in LWRs 1-1 1.3 Hydrogen Research Efforts 1-2 1.4 Description of Manual 1-4 Chapter 2 - Hydrogen in Light Water Reactors 2-1 2.1 Production of Hydrogen 2-1 2.1.1 Summary 2-1 2.1.2 Steam:Zirconium Reaction 2-1 2.1.3 Steam:Steel Reaction 2-3 2.1.4 Radiolysis of Water 2-7 2.1.5 Core:Concrete Interactions 2-10 2.1.6 Other Sources 2-13 2.1.7 Time Sequence of Hydrogen Production 2-16 2.2 Hydrogen Solubility, Transport, and Mixing 2-19 2.2.1 Introduction 2-19 2.2.2 Solubility of Hydrogen in Water 2-19 2.2.3 Transport and Mixing Within Containment 2-24 2.2.4 Air:Steam:Hydrogen Concentration Within Containment 2-26 2.3 Combustion of Hydrogen 2-28 2.3.1 Introduction 2-28 2.3.2 Example Calculation 2-28 2.3.3 Conditions Necessary for Combustion 2-36 2.3.4 Deflagrations 2-42 2.3.5 Combustion of a Steam:Hydrogen Jet 2-47 2.3.6 Detonation of Hydrogen 2-49 2.3.6.1 Detonation Limits 2-49 2.3.6.2 Transition to Detonation 2-54 2.3.6.3 Detonation Pressures and Temperatures 2-54 2.3.6.4 Local Detonations 2-55 2.3.6.5 Missile Generation 2-59 2.4 Mitigation Schemes 2-60 2.4.1 Deliberate Ignition 2-60 2.4.2 Water Sprays, Fogs and Foams 2-61 2.4.3 Pre-Accident Inerting 2-61 2.4.4 Post-Accident Inerting 2-65 2.4.5 Partial Pre-Inerting 2-65 2.4.6 High-Point-Vent Flaring 2-66 2.4.7 Other Methods 2-66 iv Page Chapter 3 - Hydrogen Behavior and Control--Normal Operation 3-1 3.1 Introduction 3-1 3.2 Hydrogen Considerations Common to PWRs and BWRs 3-1 3.2.1 Radiolysis 3-1 3.2.2 Radiation Synthesis 3-2 3.2.3 Control of Hydrogen from Radiolysis and Radiation Synthesis 3-3 3.2.4 Corrosion 3-4 3.2.5 Electrical Generators 3-8 3.2.6 Hydrogen Bulk Storage Facility 3-11 3.2.7 Gas Chromatography 3-12 3.2.8 Battery Room 3-12 3.2.9 Resin Beds 3-13 3.3 Hydrogen Considerations Specific to PWRs 3-13 3.3.1 Reactor Coolant System 3-14 3.3.1.1 Pressurizer 3-14 3.3.1.2 Control Rod Drive Mechanisms 3-16 3.3.1.3 RCS Leakage 3-18 3.3.2 Chemical and Volume Control System 3-18 3.3.3 Containment Entry 3-19 3.3.4 Gaseous Waste Disposal 3-19 3.4 Hydrogen Considerations'Specific to BWRs 3-22 3.4.1 Reactor Coolant System 3-22 3.4.2 Off-Gas System 3-22 3.4.3 Chemistry 3-27 3.4.4 Containment Entry 3-27 Chapter 4 - Hydrogen Behavior and Controi--Accident Conditions 4-1 4.1 Introduction 4-1 4.2 Containment Structure and Integrity 4-1 4.2.1 LWR Containment Comparison 4-1 4.2.2 PWR Containments 4-4 4.2.3 BWR Continments 4-7 4.3 Indications of a Loss-of-Coolant Accident 4-10 4.3.1 Introduction 4-10 4.3.2 PWR Indications Internal to the Reactor Coolant System 4-13 4.3.2.1 Temperature 4-13 4.3.2.2 Flow 4-13 4.3.2.3 Pressure 4-15 4.3.2.4 Pressurizer Level 4-15 4.3.2.5 Reactor Vessel Water Level 4-17 4.3.2.6 Nuclear Instrumentation 4-17 4.3.3 PWR Indications External to the Reactor Coolant System 4-22 V Page 4.3.3.1 Containment Pressure 4-23 4.3.3.2 Containment Temperature 4-23 4.3.3.3 Containment Humidity 4-23 4.3.3.4 Containment Activity 4-23 4.3.3.5 Containment Sump Level 4-23 4.3.4 BWR Indications Internal to the Reactor Coolant System 4-23 4.3.4.1 Temperature 4-23 4.3.4.2 Flow 4-23 4.3.4.3 Pressure 4-24 4.3.4.4 Reactor Vessel Level 4-24 4.3.4.5 Nuclear Instrumentation 4-28 4.3.5 BWR Indications External to the Reactor Coolant System 4-28 4.3.5.1 Drywell/Containment Pressure 4-28 4.3.5.2 Drywell/Containment Temperature 4-29 4.3.5.3 Drywell/Containment Activity 4-29 4.3.5.4 Suppression Pool Water Level 4-29 4.4 Indications of Hydrogen Production 4-29 4.4.1 Indications of Hydrogen Production Internal to the Reactor Coolant System of a PWR 4-30 4.4.1.1 Temperature 4-30 4.4.1.2 Flow 4-30 4.4.1.3 Pressure 4-32 4.4.1.4 Pressurizer Level 4-34 4.4.1.5 Reactor Vessel Water Level 4-34 4.4.1.6 Reactor Coolant Sampling 4-34 4.4.1.7 Reactor Coolant Activity 4-34 4.4.2 Indications of Hydrogen Production External to the Reactor Coolant System of a PWR 4-38 4.4.2.1 Containment Pressure 4-38 4.4.2.2 Containment Temperature 4-38 4.4.2.3 Containment Humidity 4-38 4.4.2.4 Containment Sampling 4-39 4.4.2.5 Steam Line Radiation 4-40 4.4.2.6 Gross Radiation Level Detection 4-40 4.4.3 Indications of Hydrogen Production Internal to the Reactor Coolant System of a BWR 4-41 4.4.3.1 Temperature 4-41 4.4.3.2 Flow 4-41 4.4.3.3 Pressure 4-42 4.4.3.4 Reactor Vessel Water Level 4-42 4.4.3.5 Reactor Coolant Activity 4-44 4.4.4 Indications of Hydrogen Production External to the Reactor Coolant System of a BWR 4-44 4.4.4.1 Drywell/Containment Pressure 4-44 4.4.4.2 Drywell/Containment Temperature 4-46 4.4.4.3 Drywell/Containment Humidity 4-46 4.4.4.4 Drywell/Containment Sampling 4-46 4.4.4.'5 Steam Line Radiation Monitoring 4-47 4.4.4.6 Gross Radiation Level 4-47 4.4.4.7 Sump Leak Detection 4-47 4.4.4.8 Suppression Pool Temperature 4-48 vi Page 4.5 Hydrogen Detection 4-48 4.5.1 Combustion Detector 4-49 4.5.2 Solid-State Detector 4-50 4.5.3 Electromechanical Detector 4-50 4.5.4 Thermal Conductivity Detector 4-50 4.5.5 Absorption Detector 4-50 4.5.6 Detector Availability 4-51 4.6 Transport and Mixing of Hydrogen 4-51 4.6.1 Introduction 4-51 4.6.2 Hydrogen Transport Pathways in a PWR 4-53 4.6.2.1 Hydrogen Transport Through the Reactor Vessel Head Vent 4-53 4.6.2.2 Hydrogen Transport Through the Hot Leg and Cold Leg Piping 4-57 4.6.2.3 Hydrogen Transport Into the Pressurizer and Pressurizer Relief Tank 4-58 4.6.2.4 Hydrogen Transport From Containment to Other Buildings 4-60 4.6.2.5 Considerations to Minimize Hydrogen Transport to Other Buildings 4-62 4.6.3 General Hydrogen Transport Pathways in a BWR 4-63 4.6.3.1 Hydrogen Transport Through a Reactor Coolant System Break 4-63 4.6.3.2 Hydrogen Transport Through a Stuck Open Safety/Relief Valve to the Suppression Pool 4-66 4.6.3.3 Primary Containment Leakage to the Secondary Containment (Mark I and II BWR Designs) 4-66 4.6.3.4 Hydrogen Transport via the Drywell Purge System and Vacuum Breaker 4-68 4.6.3.5 Hydrogen Transport Through the Bypass Leakage Paths From the Drywell 4-70 4.6.3.6 Considerations to Minimize Hydrogen Release to Other Buildings 4-72 4.6.4 Air Handling Fans 4-74 4.7 Hydrogen Combustion 4-74 4.7.1 introduction 4-74 4.7.2 Potential Sources of Accidental Hydrogen Ignition 4-75 4.7.3 Evidence of Hydrogen Combustion in PWRs 4-76 4.7.3.1 Containment Pressure Effects 4-77 4.7.3.2 Pressurizer Pressure Effects 4-77 4.7.3.3 Steam Generator Pressure Effects 4-77 4.7.3.4 Containment
Load more

Recommended publications
  • Dickins' Sony Deal Opens New Chapter by Ajax Scott Rob Dickins Bas Teame Former Rival Sony Music Three Months After Leaving Warn
    Dickins' Sony deal opens new chapter by Ajax Scott later building a team of up to 20. Rob Dickins bas teame Instant Karma will sign around five rA, ^ threeformer months rival Sony after Music leaving warner ri likelyartists to a be year, from most the UKof orwhom Ireland. are hisMusic music to launchindustry the career. next stage of Dickins says he will not specifically S Dickins' new record label Instant goclose chasing to expiring. after those Tm notacts. going If they to with the hasfirst got single off torelease a fiying under start the handling partnership Interscope - My releasesName Is inby the US UK they'reare interested going to in call," being he part says. of that therapper single, Eminem released - enjoying next Monday a huge initial(March shipout. 29), have Around been 220.000 ordered unitsby of months of negotiations between runningDickins, Warner's who spent UK ninepublishing years theretaiiefsVwhile Top 40 of the the Airplay track waschart. yesterday "Ifs appealing (Sunday) right on acrosscourse theto move board," into says Dickins, his lawyer Tony Russell arm before moving to head its Polydor head of radio promotionl" TU.Ruth ..U.._ Parrish. A V.UK promo visit will inelude wasand afinaliy list of whittled would-be down backers to Sony that New partners: Dickins (second isrecord in no company hurry toin 1983,issue addsrecords. he im_ rrShady LPes - thiswhich week debuted on The at Bignumber Breakfast twoThe in and the CD:UK.States earlierThe this (l-r)left) Sony in New Music York Worldwide last week with "Ifs really not important when the onth - follows on April 12.
    [Show full text]
  • Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment Volume 2, Detailed Analysis
    Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within A Factory Environment Volume 2, Detailed Analysis Xuan Chen, Arnold Kotlyarevsky, Andrew Kumiega, Jeff Terry, and Benxin Wu Illinois Institute of Technology Stephen Goldberg and Edward A. Hoffman Argonne National Laboratory August 2013 This study continued the work, supported by the Department of Energy’s Office of Nuclear Energy, regarding the economic analysis of small modular reactors (SMRs). The study team analyzed, in detail, the costs for the production of factory-built components for an SMR economy for a pressurized-water reactor (PWR) design. The modeling focused on the components that are contained in the Integrated Reactor Vessel (IRV). Due to the maturity of the nuclear industry and significant transfer of knowledge from the gigawatt (GW)-scale reactor production to the small modular reactor economy, the first complete SMR facsimile design would have incorporated a significant amount of learning (averaging about 80% as compared with a prototype unit). In addition, the order book for the SMR factory and the lot size (i.e., the total number of orders divided by the number of complete production runs) remain a key aspect of judging the economic viability of SMRs. Assuming a minimum lot size of 5 or about 500 MWe, the average production cost of the first-of-the kind IRV units are projected to average about 60% of the a first prototype IRV unit (the Lead unit) that would not have incorporated any learning. This cost efficiency could be a key factor in the competitiveness of SMRs for both U.S.
    [Show full text]
  • Scaling Analysis of the Osu High Temperature Test Facility During a Pressurized Conduction Cooldown Event Using Relap5- 3D
    AN ABSTRACT OF THE THESIS OF Juan A. Castañeda for the degree of Master of Science in Nuclear Engineering presented on May 21, 2014. Tittle: SCALING ANALYSIS OF THE OSU HIGH TEMPERATURE TEST FACILITY DURING A PRESSURIZED CONDUCTION COOLDOWN EVENT USING RELAP5- 3D Abstract approved: Brian G. Woods In early 2000, the Generation IV International Forum (GIF) was created to perform research and development for the next generation nuclear systems. Among the selected nuclear systems was the Very High Temperature Gas-Cooled Reactor (VHTR). Then in 2008, the U.S. Department of Energy (DOE) decided that the Next Generation Nuclear Plant (NGNP) would be the VHTR. The VHTR was chosen because it has the capability to produce electricity, hydrogen and may be used for other high-temperature process heat applications. In support of licensing and validation of the VHTR, Oregon State University was tasked to develop a high temperature apparatus that will be able to capture the thermal fluids phenomena of the VHTR and perform integral effects tests for validation of existing safety codes. The design has been called the High Temperature Test Facility (HTTF), which is a scaled design of the Modular High Temperature Gas-Cooled Reactor (MHTGR). The objective of this study was to investigate the ability of the HTTF to simulate the Pressurized Conduction Cooldown (PCC) Event during the natural circulation phase in the MHTGR. This was achieved with the aid of a thermal hydraulic systems code, RELAP5-3D. ©Copyright by Juan A. Castañeda May 21, 2014 All Rights Reserved SCALING ANALYSIS OF THE OSU HIGH TEMPERATURE TEST FACILITY DURING A PRESSURIZED CONDUCTION COOLDOWN EVENT USING RELAP5-3D by Juan A.
    [Show full text]
  • Section 3.9.5, "Reactor Pressure Vessel Internals," Revision 4
    NUREG-0800 U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN 3.9.5 REACTOR PRESSURE VESSEL INTERNALS REVIEW RESPONSIBILITIES Primary - Organization responsible for mechanical engineering reviews Secondary - None I. AREAS OF REVIEW Reactor pressure vessel (RPV) internals consist of all structural and mechanical elements inside the reactor vessel in nuclear power plants. The U.S. Nuclear Regulatory Commission (NRC) regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, “Domestic Licensing of Production and Utilization Facilities,” Appendix A, “General Design Criteria (GDC) for Nuclear Plants,” GDC 1, “Quality Standards and Records,” GDC 2, “Design Bases for Protection against Natural Phenomena,” GDC 4, "Environmental and Dynamic Effects Design Bases," and GDC 10, "Reactor Design,"; 10 CFR 50.55a, “General Provisions”; and 10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants”; require that structures and components important to safety be constructed and tested to quality standards commensurate with the importance of the safety functions performed and designed with appropriate margins to withstand effects of anticipated operational occurrences and normal operation, natural phenomena such as earthquakes, postulated accidents including loss-of-coolant accidents (LOCAs), and events and conditions outside the nuclear power unit. Revision 4 – March 2017 USNRC STANDARD REVIEW PLAN This Standard Review Plan (SRP), NUREG-0800, has been prepared to establish criteria that the U.S. Nuclear Regulatory Commission (NRC) staff responsible for the review of applications to construct and operate nuclear power plants intends to use in evaluating whether an applicant/licensee meets the NRC's regulations. The SRP is not a substitute for the NRC's regulations, and compliance with it is not required.
    [Show full text]
  • Integral Pressurized Water Reactor Simulator Manual
    65 @ Integral Pressurized Water Reactor Simulator Manual Simulator Reactor Pressurized Water Integral Integral Pressurized Water Reactor Simulator Manual TRAINING COURSE SERIES TRAINING ISSN 1018-5518 VIENNA, 2017 TRAINING COURSE SERIES 65 TCS-65_cov_+Nr.indd 1,3 2017-05-24 12:15:04 INTEGRAL PRESSURIZED WATER REACTOR SIMULATOR MANUAL The following States are Members of the International Atomic Energy Agency: AFGHANISTAN GEORGIA OMAN ALBANIA GERMANY PAKISTAN ALGERIA GHANA PALAU ANGOLA GREECE PANAMA ANTIGUA AND BARBUDA GUATEMALA PAPUA NEW GUINEA ARGENTINA GUYANA PARAGUAY ARMENIA HAITI PERU AUSTRALIA HOLY SEE PHILIPPINES AUSTRIA HONDURAS POLAND AZERBAIJAN HUNGARY PORTUGAL BAHAMAS ICELAND QATAR BAHRAIN INDIA REPUBLIC OF MOLDOVA BANGLADESH INDONESIA ROMANIA BARBADOS IRAN, ISLAMIC REPUBLIC OF RUSSIAN FEDERATION BELARUS IRAQ RWANDA BELGIUM IRELAND SAN MARINO BELIZE ISRAEL SAUDI ARABIA BENIN ITALY SENEGAL BOLIVIA, PLURINATIONAL JAMAICA SERBIA STATE OF JAPAN SEYCHELLES BOSNIA AND HERZEGOVINA JORDAN SIERRA LEONE BOTSWANA KAZAKHSTAN SINGAPORE BRAZIL KENYA SLOVAKIA BRUNEI DARUSSALAM KOREA, REPUBLIC OF SLOVENIA BULGARIA KUWAIT SOUTH AFRICA BURKINA FASO KYRGYZSTAN SPAIN BURUNDI LAO PEOPLE’S DEMOCRATIC SRI LANKA CAMBODIA REPUBLIC SUDAN CAMEROON LATVIA SWAZILAND CANADA LEBANON SWEDEN CENTRAL AFRICAN LESOTHO SWITZERLAND REPUBLIC LIBERIA SYRIAN ARAB REPUBLIC CHAD LIBYA TAJIKISTAN CHILE LIECHTENSTEIN THAILAND CHINA LITHUANIA THE FORMER YUGOSLAV COLOMBIA LUXEMBOURG REPUBLIC OF MACEDONIA CONGO MADAGASCAR TOGO COSTA RICA MALAWI TRINIDAD AND TOBAGO CÔTE D’IVOIRE
    [Show full text]
  • Selling Or Selling Out?: an Exploration of Popular Music in Advertising
    Selling or Selling Out?: An Exploration of Popular Music in Advertising Kimberly Kim Submitted to the Department of Music of Amherst College in partial fulfillment of the requirements for the degree of Bachelor of Arts with honors. Faculty Advisor: Professor Jason Robinson Faculty Readers: Professor Jenny Kallick Professor Jeffers Engelhardt Professor Klara Moricz 05 May 2011 Table of Contents Acknowledgments............................................................................................................... ii Chapter 1 – Towards an Understanding of Popular Music and Advertising .......................1 Chapter 2 – “I’d Like to Buy the World a Coke”: The Integration of Popular Music and Advertising.........................................................................................................................14 Chapter 3 – Maybe Not So Genuine Draft: Licensing as Authentication..........................33 Chapter 4 – Selling Out: Repercussions of Product Endorsements...................................46 Chapter 5 – “Hold It Against Me”: The Evolution of the Music Videos ..........................56 Chapter 6 – Cultivating a New Cultural Product: Thoughts on the Future of Popular Music and Advertising.......................................................................................................66 Works Cited .......................................................................................................................70 i Acknowledgments There are numerous people that have provided me with invaluable
    [Show full text]
  • Design and Development of Gas Cooled Reactors with Closed Cycle Turbinesgas
    IAEA-TECDOC-899 Design and development of gas cooled reactors with closed cycle turbinesgas Proceedings of a Technical Committee meeting held in Bei/ing, China, 30 October - 2 November 1995 INTERNATIONAL ATOMIC ENERGY AGENCA / Y The IAEA does not normally maintain stocks of reports in this series. However, microfiche copies of these reports can be obtained from IN IS Clearinghouse International Atomic Energy Agency Wagramerstrasse 5 P.O. Box 100 A-1400 Vienna, Austria Orders shoul accompaniee db prepaymeny db f Austriao t n Schillings 100,- in the form of a cheque or in the form of IAEA microfiche service coupons which may be ordered separately from the IN IS Clearinghouse. originatine Th g Sectio f thino s publicatio IAEe th An i was- Nuclear Power Technology Development Section International Atomic Energy Agency Wagramerstrasse 5 0 10 x P.OBo . A-1400 Vienna, Austria DESIG DEVELOPMEND NAN COOLES GA F TDO REACTORS WITH CLOSED CYCL TURBINES EGA S IAEA, VIENNA, 1996 IAEA-TECDOC-899 ISSN 1011-4289 IAEA© , 1996 Printed by the IAEA in Austria August 1996 FOREWORD lons Iha tg been recognized that substantial generatioe gainth n si f electricitno y from nuclear e obtainefissiob n ca nd throug e directh h t s turbincouplin ga a hig a o t ehf o g temperature helium cooled reactor. This advanced nuclear power plant is unique in its use of the Brayton cycle to achieve a net electrical efficiency approaching 50% combined with the attendant features of low initial capital costs due to plant simplification, public acceptance from the safety attributes of the high temperature gas cooled reactor (HTGR), and reduced radioactive wastes.
    [Show full text]
  • Reactor Coolant System
    NUCLEAR SYSTEMS ENGINEERING Cho, Hyoung Kyu Department of Nuclear Engineering Seoul National University 4월6일보 강가? 능 Contents of Lecture Contents of lecture Chapter 1Principal Characteristics of Power Reactors Nuclear system . Will be replaced by the lecture note . Introduction to Nuclear Systems . CANDU . BWR and Fukushima accident . PWR (OPR1000 and APR1400) Chapter 4 Transport Equations for Single‐Phase Flow (up to energy equation) Chapter 6Thermodynamics of Nuclear Energy Conversion Systems: Nonflow and Steady Flow : First‐ and Second‐Law Applications Chapter 7 Thermodynamics of Nuclear Energy Conversion Systems : Nonsteady Flow First Law Analysis Thermodynamics Chapter 3 Reactor Energy Distribution Heat transport Chapter 8 Thermal Analysis of Fuel Elements Conduction heat transfer 3. PRESSURIZED WATER REACTOR References Lecture note 한수원 계통교재 Contents History of PWR Plant Overall Reactor Coolant System Steam and Power Conversion System Auxiliary System Plant Protection System Other systems History of PWR 1939: Nuclear fission was discovered. 1942: The world’s first chain reaction Achieved by the Manhattan Project (1942.12.02) 1951: Electricity was first generated from nuclear power EBR (Experimental Breeder Reactor‐1), Idaho, USA Power: 1.2 MWt, 200 kWe History of PWR 1950s: R&D USA: light water reactor . PWR: Westinghouse . BWR: General Electric (GE) USSR: graphite moderated light water reactor and WWER (similar to PWRs) UK and France: natural uranium fueled, graphite moderated, gas cooled reactor Canada: natural uranium fueled, heavy water reactor. 1954: initial success USSR: 5 MWe graphite moderated, light water cooled reactor was connected to grid. USA: the first nuclear submarine, the Nautilus (PWR) History of PWR 1956~1959 UK: Calder Hall‐1, 50 MWe GCR (1956) USA: Shippingport, 60 MWe PWR (1957) France: G‐2, 38 MWe GCR (1959) USSR: the ice breaker Lenin (1959) .
    [Show full text]
  • Neutron Kinetics Model Development for VVER-1000 Reactor
    Development of coupled thermohydraulics-neutron kinetics models with RELAP5-3D code for VVER reactors A. Shkarupa, N. Trofimova, I. Kadenko, A. Kharitonov, R. Yermolenko, T. Galatyuk Taras Shevchenko National University of Kyiv Sixth International Information Exchange Forum “Safety Analysis for NPPs of VVER and RBMK Types” 08-12 of April, 2002, Kyiv, Ukraine Introduction RELAP5-3D code models for Rivne NPP unit 1 (VVER-440), Zaporizhzhya NPP unit 5 (VVER- 1000) and Kozloduy NPP unit 6 (VVER-1000) have been developed within the framework of the Ukraine VVER Special Transient Analysis project [1]. For the development of RELAP5-3D VVER reactor models the corresponding RELAP5/Mod3.2 code models, developed and validated under the US DOE International Nuclear Safety Program, were used. One-dimensional models of the reactor pressure vessel and its internals have been replaced with their three-dimensional models for original complete plant models. More detailed description of RELAP5-3D VVER models is presented in this paper. RELAP5-3D code [2] has multi-dimensional thermal-hydraulic and neutron kinetics modeling capabilities. This removes any restrictions with respect to applicability of this code to full scope operational transients. Besides RELAP5-3D users may activate a new set of CHF correlations based on data available in the Czech Republic data bank [3]. These correlations replace the “CHF Table Look-up” method. Differences in the output of the PG and the table lookup method were estimated. One of the main purposes of the project is the application of the multi-dimensional coupled thermohydraulics-neutron kinetics models to analyzing the specific VVER transients in support of the In- Depth Safety Analysis (ISA) projects in Ukraine.
    [Show full text]
  • Scaling Study of the Depressurized Conduction Cooldown Event in the High Temperature Test Facility Using RELAP5-3D/ATHENA
    AN ABSTRACT OF THE THESIS OF Robert J. Aldridge for the degree of Master of Science in Nuclear Engineering presented on May 29, 2013. Title: Scaling Study of the Depressurized Conduction Cooldown Event in the High Temperature Test Facility Using RELAP5-3D/ATHENA Abstract approved: ________________________________________________________________________ Brian G. Woods In 2008, the Department of Energy (DOE) and Nuclear Regulatory Commission (NCR) decided that the Next Generation Nuclear Plant (NGNP) would be the Very High Temperature Reactor (VHTR). In support of the licensing and validation effort of the VHTR, Oregon State University was tasked with designing, building, and operating a thermal hydraulic test facility, the High Temperature Test Facility (HTTF). It is expected that the HTTF will provide data on the systemic response of the Modular High Temperature Gas Reactor (MHTGR) during the Depressurized Conduction Cooldown (DCC) event to validate safety analysis codes and the safety concept of the VHTR. The objective of this thesis was to determine the ability of the HTTF to model the DCC event compared to the MHTGR. To do this, a thermal hydraulic systems code, RELAP5-3D/ATHENA, was used to compare the systemic response of the MHTGR and HTTF during the DCC event. Two phases of the DCC event were modeled, the molecular diffusion phase and natural circulation phase. Each phase was modeled as a Separate Effects Test (SET) and as an Integral Effects Test (IET). For each test, the radial temperature profile, peak fuel temperature as a function of time, time to peak fuel temperature, air front velocity, time to onset of natural circulation, and natural circulation flow rates were compared to the MHTGR response.
    [Show full text]
  • Next Generation Nuclear Plant Reactor Pressure Vessel Acquisition Strategy
    INL/EXT-08-13951 Revision 0 Next Generation Nuclear Plant Reactor Pressure Vessel Acquisition Strategy R. E. Mizia April 2008 DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness, of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trade mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof. INL/EXT-08-13951 Revision 0 Next Generation Nuclear Plant Reactor Pressure Vessel Acquisition Strategy R. E. Mizia April 2008 Idaho National Laboratory Idaho Falls, Idaho 83415 Prepared for the U.S. Department of Energy Assistant Secretary for the Office of Nuclear Energy Under DOE Idaho Operations Office Contract DE-AC07-05ID14517 EXECUTIVE SUMMARY The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 850 to 950°C and a plant design service life of 60 years.
    [Show full text]
  • Preliminary Results of the Seventh Three-Dimensional Aer Dynamic Benchmark Problem Calculation
    PRELIMINARY RESULTS OF THE SEVENTH THREE-DIMENSIONAL AER DYNAMIC BENCHMARK PROBLEM CALCULATION. SOLUTION WITH DYN3D AND RELAP5-3D CODES. Marek Beník, Jan Hádek Nuclear Research Institute ež plc, 250 68 ež, Czech Republic E-mail: [email protected] ABSTRACT The paper gives a brief survey of the 7th three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at NRI ež. This benchmark was defined at the 20th AER Symposium in Hanassari (Finland). It is focused on investigation of transient behaviour in a VVER-440 nuclear power plant. Its initiating event is opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. The VVER-440 plant is at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations were performed with the code DYN3D. Transient calculation was made with the system code RELAP5-3D . The two-group homogenized cross sections library HELGD05 created by HELIOS code was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP th Dukovany was adopted for the 7 AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector channels connected with the three-dimensional model of the reactor vessel. The detailed nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5-3D system code and a short description of NPP input deck and reactor core model.
    [Show full text]