Innovation Pathway Study: U.S. Commercial Nuclear Power
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WASH-1097.Pdf
WASH 1097 UC-80 THE USE OF THORIUM IN NUCLEAR POWER REACTORS JUNE 1969 PREPARED BY Brookhaven National Laboratory AND THE Division of Reactor Development and Technology WITH THE ASSISTANCE OF ARGONNE NATIONAL LABORATORY BABCOCK & WILCOX GULF GENERAL ATOMIC OAK RIDGE NATIONAL LABORATORY PACIFIC NORTHWEST LABORATORY For sale by the Superintendent of Documents, U.S. Government Printing Office Washington, D.C. 20402 - Price $1.25 FOREWORD This report on "The Use of Thorium in Nuclear Power Reactors" was prepared under the direction of the Division of Reactor Development and Technology, U.S.A.E.C., as part of an overall assessment of the Civilian Nuclear Power Program initiated in response to a request in 1966 by the Joint Committee on Atomic Energy. It represents the results of the inquiry by the Thorium Systems Task Force whose membership included representatives of Babcock & Wilcox Company, Gulf General Atomic Company, the Argonne National Laboratory, the Brookhaven National Laboratory, the Oak Ridge National Laboratory, the Pacific Northwest Laboratory, and the U.S. Atomic Energy Commission. Publication of this report, which provides information basic to the AEC reactor development program, completes one phase of the evaluation effort outlined in the 1967 Supplement to the 1962 Report to the President on Civilian Nuclear Power, issued in February 1967. The 1967 Supplement outlined changes since 1962 in the technical, economic and resource picture and provided background for further study. Specifically, this report represents the consensus of the task force on the potential use of the thorium cycle and the specific thorium fueled reactor designs which have been proposed. -
An Introduction to Nuclear Power – Science, Technology and UK
sustainable development commission The role of nuclear power in a low carbon economy Paper 1: An introduction to nuclear power – science, technology and UK policy context An evidence-based report by the Sustainable Development Commission March 2006 Table of contents 1 INTRODUCTION ................................................................................................................................. 3 2 ELECTRICITY GENERATION ................................................................................................................. 4 2.1 Nuclear electricity generation ................................................................................................. 4 2.2 Fission – how does it work?..................................................................................................... 4 2.3 Moderator ................................................................................................................................. 5 2.4 Coolant...................................................................................................................................... 5 2.5 Radioactivity ............................................................................................................................. 6 3 THE FUEL CYCLE: FRONT END ............................................................................................................ 7 3.1 Mining and milling ................................................................................................................... 7 3.2 Conversion and -
Nuclear Fuels and Reprocessing Technologies: a US Perspective
INL/EXT-20-59106 Review – Nuclear Fuels and Reprocessing Technologies: A U.S. Perspective March 2021 Guy Fredrickson Tae-Sic Yoo 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, trademark, 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-20-59106 Review – Nuclear Fuels and Reprocessing Technologies: A U.S. Perspective Guy Fredrickson Tae-Sic Yoo March 2021 Idaho National Laboratory Pyrochemistry & Molten Salt Systems Department Idaho Falls, Idaho 83415 http://www.inl.gov Prepared for the U.S. Department of Energy Office of Nuclear Energy Under DOE Idaho Operations Office Contract DE-AC07-05ID14517 Page intentionally left blank ABSTRACT Reprocessing and/or waste management issues are of concern to the “back end” of the nuclear fuel cycle. Of course, there are a great many “nuclear fuel cycle” scenarios to consider; if not in practice, then at least in theory. The simplest conceptually is the “once through” fuel cycle in which the spent fuel is discarded. -
NRC Collection of Abbreviations
I Nuclear Regulatory Commission c ElLc LI El LIL El, EEELIILE El ClV. El El, El1 ....... I -4 PI AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources: 1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2. The Superintendent of Documents, U.S. Government Printing Office, P. 0. Box 37082, Washington, DC 20402-9328 3. The National Technical Information Service, Springfield, VA 22161-0002 Although the listing that follows represents the majority of documents cited in NRC publica- tions, it is not intended to be exhaustive. Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu- ments and correspondence. The following documents in the NUREG series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro- ceedings, international agreement reports, grantee reports, and NRC booklets and bro- chures. Also available are regulatory guides, NRC regulations in the Code of Federal Regula- tions, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Information Service Include NUREG-series reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. Documents available from public and special technical libraries include all open literature items, such as books, journal articles, and transactions. -
Consolidated Safety Analysis Report for IF-300 Shipping Cask
REVISION CONTROL SHEET TITLE: Consolidated Safety DOCUMENT NO.: NEDO-10084 Analysis Report for IF-300 Shipping Cask AFFECTED | DOC. REMARKS PAGE(S) I REV. I __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .I.I by General Electric Company. 1-i 3 Last revision prepared Incorporates all C of C 9001, Revision 29 references 1-1 & 1-2 2-i & 2-ii From 2/8/84 through 5/10/85. A vertical line on the right hand margin indicates a 2-1 - 2-15 Revision. "N" denotes new information while "E" 3-i & 3-ii Denotes an editorial change. 3-1 - 3-16 4-i - 4-ii 4-1 - 4-21 5-i - 5-vi 5-1 - 5-311 6-i - 6-iv U 6-1 - 6-82 7-i - 7-ii 7-1 - 7-22 8-i & 8-ii 8-1 - 8-22 n 9-i & 9-ii 9-1 - 9-6 10-i & 10-ii 10-1 - 10-15 A-i & A-ii V1-i - V1-iv V1-1 - V1-52 V1-A-i/ii V1-A-1 - V1 -A- 3 Vl-B-i/ii V1-B-1 & V1-B-2 Vi-C-i/ii V1-C-1 - 1 of 4 V1-C-8 PAGE J. I 9906220178 990608 PDR ADOCK 07109001 iii-1 B PDR REVISION CONTROL SHEET TITLE: Consolidated Safety DOCUMENT NO.: NEDO-10084 Analysis Report for IF-300 Shipping Cask AFFECTED DOC. REMARKS PAGE(S) REV. Vl-D-i - 3 Vl-D-vi Vl-D-1 - Vl-D-132 Vl-E-i & Vl-E-ii V1-E-l - V1-E-34 U V2-i - V2-iv V2-1 - V2-64 V3-i - V3-iv V3-1 - V3-32 VI-i & VI-ii VI-1 - VI-6 i - viii First revision prepared by VECTRA Technologies, Inc. -
Identifying Preliminary Criteria for Safeguarding Advanced Nuclear Reactors
A. NILSSON et al. IDENTIFYING PRELIMINARY CRITERIA FOR SAFEGUARDING ADVANCED NUCLEAR REACTORS A. NILSSON AN & Associates Uppsala, Sweden [email protected] C. JORANT SDRI Consulting Paris, France [email protected] A. FINAN Nuclear Innovation Alliance Cambridge, MA, USA [email protected] E. REDMOND Nuclear Energy Institute Washington, DC, USA [email protected] K. LUONGO Partnership for Global Security Washington, DC, USA [email protected] Abstract Climate change and the urgent need to reduce carbon emission globally forecast a significant rise in the interest of advanced nuclear reactors. The contributed paper presents the results of a dedicated study of proliferation resistance of the main technologies for advanced nuclear reactors. Much of the current safeguards system has been developed and implemented for the water-cooled reactors that dominate the nuclear energy landscape. Advanced nuclear reactors are distinctly different compared with light-water reactors, with the use of different coolants and a broader spectrum of fuel compositions. Some advanced reactor designs employ a fast neutron spectrum, often associated with an ability to create plutonium. The variety of advanced reactor designs and their coolants present challenges to the existing safeguards regime. Some of the designs allow for on-line fuel processing and unique refuelling schemes. These features may require updated or new tools for nuclear accountancy and control as well as built-in technological features to prevent malicious use of the facility. The paper communicates the results of an overview of the most prominent advanced nuclear reactor designs, in which their proliferation resistance is assessed in comparison with the implementation of IAEA safeguards at a light water-cooled reactor (LWR) fuelled with low enriched Uranium, as established in States with comprehensive safeguards agreements and additional protocols. -
Fundamentals of Boiling Water Reactor (Bwr)
FUNDAMENTALS OF BOILING In this light a general description of BWR is presented, with emphasis on the reactor physics aspects (first lecture), and a survey of methods applied in fuel WATER REACTOR (BWR) and core design and operation is presented in order to indicate the main features of the calculation^ tools (second lecture). The third and fourth lectures are devoted to a review of BWR design bases, S. BOZZOLA reactivity requirements, reactivity and power control, fuel loading patternc. Moreover, operating 'innts are reviewed, as the actual limits during power AMN Ansaldo Impianti, operation and constraints for reactor physics analyses (design and operation). Genoa, The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality Italy and low power operation are illustrated in the fifth lecture. The banked position withdrawal sequence mode of operation is indeed an example of safety and design/operation combined approach. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact Abstract between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the These lectures on fundamentals of BWR reactor physics are a synthesis startup of a commercial BWR. A control .od calibration is also illustrated. of known and established concepts. These lecture are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. -
Tracg Analysis of Boiling Water Reactor Control Rod Drop Accident to Optimize Analysis Methodology
TRACG ANALYSIS OF BOILING WATER REACTOR CONTROL ROD DROP ACCIDENT TO OPTIMIZE ANALYSIS METHODOLOGY D. Miranda*, C. McElroy, D. G. Vreeland, J. Yang, J. Banfield, and J. Vedovi GE Hitachi Nuclear Energy 3901 Castle Hayne Road, Wilmington, NC 28402, USA [email protected]; [email protected]; [email protected]; [email protected]; [email protected]; [email protected] ABSTRACT During transient events, it is important to confirm that the plant maintains compliance to regulatory standards that provide safety criteria to prevent the release of fission products. The design basis Reactivity Insertion Accident (RIA) considered for Boiling Water Reactors (BWRs) is the Control Rod Drop Accident (CRDA). During this postulated scenario, a control blade becomes decoupled from the control rod drive mechanism as the drive is withdrawn and remains lodged in place. At a later time, the blade falls out to the position of the drive causing a reactivity insertion into the core. This leads to a power excursion and terminates due to the Doppler Effect or upon completion of the SCRAM. Many BWR plants adopt a Banked Position Withdrawal Sequence (BPWS) methodology that was developed to minimize the control rod worth and mitigate the consequences of the CRDA event. The BPWS methodology places restrictions on control rod movement to ensure that no CRDA could exceed the applicable event limits by reducing the incremental control rod reactivity worth to acceptable values. In this paper, the best estimate model code TRACG, the GE Hitachi (GEH) Nuclear Energy proprietary version of the Transient Reactor Analysis Code, TRAC, is used to model CRDA event to ensure that the plant is compliant with the new more restrictive regulatory standards. -
Occupational Exposure at Nuclear Power Plants in Tarragona Spain in April 2000
INFORMATION SYSTEM ON OCCUPATIONAL EXPOSURE Ninth Annual Report OCCUPATIONAL EXPOSURES AT NUCLEAR POWER PLANTS 1999 ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT Pursuant to Article 1 of the Convention signed in Paris on 14th December 1960, and which came into force on 30th September 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies designed: - to achieve the highest sustainable economic growth and employment and a rising standard of living in Member countries, while maintaining financial stability, and thus to contribute to the development of the world economy; - to contribute to sound economic expansion in Member as well as non-member countries in the process of economic development; and - to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance with international obligations. The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, Germany, Greece, Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The following countries became Members subsequently through accession at the dates indicated hereafter: Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971), New Zealand (29th May 1973), Mexico (18th May 1994), the Czech Republic (21st December 1995), Hungary (7th May 1996), Poland (22nd November 1996) and the Republic of Korea (12th December 1996). The Commission of the European Communities takes part in the work of the OECD (Article 13 of the OECD Convention). NUCLEAR ENERGY AGENCY The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEEC European Nuclear Energy Agency. -
BWRX-300 (GE Hitachi and Hitachi GE Nuclear Energy) USA DATE (2019/9/30)
Status Report – BWRX-300 (GE Hitachi and Hitachi GE Nuclear Energy) USA DATE (2019/9/30) The BWR-300 is the 10th generation Boiling Water Reactor (BWR) crated by GE Hitachi Nuclear Energy (GEH). It is a SMR evolution of the ESBWR which is licensed by the US NRC and utilizes many of the components for the operational ABWR. The first BWRX-300s are expected to start construction in in 2024 and 2025 and enter commercial operation in 2027 and 2028. INTRODUCTION Indicate which booklet(s): [ ] Large WCR [√] SMR [ ] FR GE Hitachi Nuclear Energy’s (GEH’s) BWRX-300 is a designed-to-cost, 300 MWe water- cooled, natural circulation Small Modular Reactor (SMR) utilizing simple, natural phenomena driven safety systems. It is the tenth generation of the Boiling Water Reactor (BWR) and represents the simplest, yet most innovative BWR design since GE, GEH’s predecessor in the nuclear business, began developing nuclear reactors in 1955. The BWRX-300 is an evolution of the U.S. NRC-licensed, 1,520 MWe ESBWR. It is designed to provide clean, flexible energy generation that is cost competitive with natural gas fired plants. Target applications include base load electricity generation, load following electrical generation within a range of 50 to 100% power and district heating. GEH, a world-leading provider of advanced reactor technology and nuclear services, is a global alliance created by the General Electric Company (GE) and Hitachi, Ltd. to serve the global nuclear industry. Development Milestones 2014 ESBWR DCD Issued 2017 BWRX-300 Evolution from ESBWR -
TRANSMUTATION of AMERICIUM and CURIUM in a LANTHANIDE MATRIX COMPANY WIDE CW-123700-CONF-018 Revision 0
Conference Paper TRANSMUTATION OF AMERICIUM AND CURIUM IN A LANTHANIDE MATRIX COMPANY WIDE CW-123700-CONF-018 Revision 0 Prepared by Rédigé par Hyland Bronwyn Reviewed by Vérifié par Edwards Geoffrey W R Approved by Approuvé par Hyland Bronwyn 2012/07/04 2012/07/04 UNRESTRICTED ILLIMITÉ Atomic Energy of Énergie Atomique du Canada Limited Canada Limitée Chalk River, Ontario Chalk River (Ontario) Canada K0J 1J0 Canada K0J 1J0 Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018 Transmutation of Americium and Curium in a Lanthanide Matrix B. HYLAND1, E.D COLLINS2, R. J. ELLIS2, G. DEL CUL2 and M. MAGILL1 1Chalk River Laboratories, Atomic Energy of Canada Limited, Canada, K0J 1J0 Tel: +1(613)584-9243 ex. 44707 , Fax: +1(613)584-8198 , Email: [email protected] 2Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee, U.S.A. Abstract – With world stockpiles of used nuclear fuel increasing, the long-term remediation of used fuel is a growing concern. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. Partitioning and transmutation of the spent fuel is one possibility to deal with the long-lived transuranic actinides. It is desirable to develop a partitioning and transmutation scheme that is as straightforward as possible for both the reprocessing step and the transmutation in reactor. This study looks at a reprocessing scheme that is simplified to keep the americium, curium and lanthanides together in one recycle stream. These actinides can be transmuted to fissionable isotopes and then to shorter-lived fission products in a CANDU reactor. -
Limerick Generating Station, Units 1 & 2, Revision 19 to Updated Final
LGS UFSAR CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS 1.1 INTRODUCTION 1.2 GENERAL PLANT DESCRIPTION 1.2.1 Site Characteristics 1.2.1.1 Location 1.2.1.2 Site Environs and Access 1.2.1.3 Geology 1.2.1.4 Seismology 1.2.1.5 Hydrology 1.2.1.6 Meteorology 1.2.2 Principal Design Criteria 1.2.2.1 General Design Criteria 1.2.2.2 System Design Criteria 1.2.2.2.1 Nuclear System Criteria 1.2.2.2.2 Safety-Related Systems Criteria 1.2.2.2.3 Process Control Systems Criteria 1.2.2.2.4 Power Conversion Systems Criteria 1.2.2.2.5 Electrical Power Systems Criteria 1.2.2.2.6 Auxiliary Systems Criteria 1.2.2.2.7 Radioactive Waste Management Systems Criteria 1.2.2.2.8 Shielding and Access Control Criteria 1.2.2.2.9 Fuel Handling and Storage Facilities 1.2.3 General Arrangement of Structures and Equipment 1.2.4 System Description 1.2.4.1 Nuclear System 1.2.4.1.1 Reactor Core and Control Rods 1.2.4.1.2 Reactor Vessel and Internals 1.2.4.1.3 Reactor Recirculation System 1.2.4.1.4 Residual Heat Removal System 1.2.4.1.5 Reactor Water Cleanup System 1.2.4.1.6 Nuclear Leak Detection System 1.2.4.2 Safety-Related Systems 1.2.4.2.1 Reactor Protection System 1.2.4.2.2 Neutron Monitoring System 1.2.4.2.3 Control Rod Drive System 1.2.4.2.4 Control Rod Velocity Limiter 1.2.4.2.5 Control Rod Drive Housing Supports 1.2.4.2.6 Nuclear System Pressure Relief System 1.2.4.2.7 Reactor Core Isolation Cooling System 1.2.4.2.8 Primary Containment CHAPTER 01 1-i REV.