Risks of Operating Candu 6 Nuclear Power Plants: Gentilly Unit 2 Refurbishment and Its Global Implications
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CHAPTER 13 Reactor Safety Design and Safety Analysis Prepared by Dr
1 CHAPTER 13 Reactor Safety Design and Safety Analysis prepared by Dr. Victor G. Snell Summary: The chapter covers safety design and safety analysis of nuclear reactors. Topics include concepts of risk, probability tools and techniques, safety criteria, design basis accidents, risk assessment, safety analysis, safety-system design, general safety policy and principles, and future trends. It makes heavy use of case studies of actual accidents both in the text and in the exercises. Table of Contents 1 Introduction ............................................................................................................................ 6 1.1 Overview ............................................................................................................................. 6 1.2 Learning Outcomes............................................................................................................. 8 1.3 Risk ...................................................................................................................................... 8 1.4 Hazards from a Nuclear Power Plant ................................................................................ 10 1.5 Types of Radiation in a Nuclear Power Plant.................................................................... 12 1.6 Effects of Radiation ........................................................................................................... 12 1.7 Sources of Radiation ........................................................................................................ -
SAFETY RE-ASSESSMENT of AECL TEST and RESEARCH REACTORS D. J. WINFIELD Chalk River Nuclear Laboratories ATOMIC ENERGY of CANADA
309 IAEA-SM-310/ 94 SAFETY RE-ASSESSMENT OF AECL TEST AND RESEARCH REACTORS D. J. WINFIELD Chalk River Nuclear Laboratories ATOMIC ENERGY OF CANADA LIMITED 310 IAEA-SM-310/94 SAFETY RE-ASSESSMENT OF AECL TEST AND RESEARCH REACTORS ABSTRACT Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor MAPLE-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration/test reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX and NRU). ZED-2 and PTR are operated by the Reactor Physics Branch, all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault tree analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis and indicates the plan for future analysis. -
Atomic Energy of Canada Limited BUCKLINGS of HEAVY WATER
B Atomic Energy of Canada Limited BUCKLINGS OF HEAVY WATER MODERATED LATTICES OF ZEEP RODS by r.N. MCDONNELL and A. OKAZAKI Chofk River, Ontario September 197! AECL-3998 BUCKLINGS OF HEAVY WATER MODERATED LATTICES OF ZEEP RODS By *F.N. McDonnell and A. Okazaki Reactor Physics Branch * NRC Post-Doctorate Fellow attached to Reactor Physics Branch. ABSTRACT The bucklings of heavy water moderated lattices of ZEEP rods (3.25 cm diameter natural U metal contained in aluminum tubes) were measured by the flux mapping method in the ZED-2 reactor. The isotopic composition of trie moderator was 99.46 atom percent D2O. The bucklings, corrected for the 0.34 atom percent difference in isotopic composition, are in excellent ag.eement with earlier measurements made in the ZEEP reactor. Chalk River Nuclear Laboratories Chalk River, Ontario September, 1971 AECL-3998 Laplaciens de réseaux de barres ZEEP modérés par eau lourde par F.N. McDonnell* et A. Okazaki** •Boursier Post-Doctorat du Conseil National de Recherches détaché à la Sous-Division de Physique des réacteurs **Sous-Division de Physique des réacteurs Résumé Les iaplaciens des réseaux de barres ZEEP modérés par eau lourde (barreaux d'uranium naturel métallique ayant 3,25 cm de diamètre et gainés dans des tubes d'aluminium) ont été mesurés par la méthode du cadrage du flux dans le réacteur ZED-2. La composition isotopique du modérateur était de 99j46% atomes de D-0. Les laplaciens, corrigés pour la différence de 0,34% atomes dans la composition isotopique, sont en excellent accord avec les mesures précédentes effectuées dans le réacteur ZEEP. -
Heu Repatriation Project
HEU REPATRIATION PROJECT RATIONALE In April 2010, the governments of Canada and the United States (U.S.) committed to work cooperatively to repatriate spent highly- enriched uranium (HEU) fuel currently stored at the Chalk River Laboratories in Ontario to the U.S. as part of the Global Threat Reduction Initiative, a broad international effort to consolidate HEU inventories in fewer locations around the world. This initiative PROJECT BACKGROUND promotes non-proliferation This HEU is the result of two decades of nuclear fuel use at the by removing existing weapons Chalk River Laboratories for Canadian Nuclear Laboratories (CNL) grade material from Canada research reactors, the National Research Experimental (NRX) and and transferring it to the National Research Universal (NRU), and for the production of U.S., which has the capability medical isotopes in the NRU, which has benefitted generations of to reprocess it for peaceful Canadians. Returning this material to the U.S. in its existing solid purposes. In March 2012, and liquid forms ensures that this material is stored safely in a Prime Minister Harper secure highly guarded location, or is reprocessed into other forms announced that Canada and that can be used for peaceful purposes. the U.S. were expanding their efforts to return additional Alternative approaches have been carefully considered and inventories of HEU materials, repatriation provides the safest, most secure, and fastest solution including those in liquid form. for the permanent disposition of these materials, thereby eliminating a liability for future generations of Canadians. For more information on this project contact: Email: [email protected] Canadian Nuclear Laboratories 1-866-886-2325 or visit: www.cnl.ca persons who have a legitimate need to PROJECT GOAL know, such as police or emergency response To repatriate highly-enriched uranium forces. -
Reactor Physics Calculations in the Nordic Countries
ESPOO 2003 VTT SYMPOSIUM 230 The eleventh biennial meeting on reactor physics calculations in the Nordic VTT SYMPOSIUM 230 countries was arranged by VTT Processes in Otaniemi, Espoo and on board Tallink´s m/s Romantika on April 9–10, 2003. General reactor physics, calculational methods, a code system adapted for RBMK reactor analyses, and transmutation of nuclear waste were presented by representatives of universities and programme developers. Computer programmes are the most important tools of reactor physics. At the meeting there were presentations of VTT Processes’ new deterministic 3- dimensional radiation transport code MultiTrans and BWR simulator ARES based upon the AFEN model, and also of new features in internationally wellknown codes like CASMO-4E and POLCA (POLCA-T) together with Reactor physics calculations in the Nordic countries results obtained by these programmes. A code for PWR loading pattern search, called LP-fun, is being developed by Westinghouse and others. On the subject of code validation, measurements on SVEA-96+ fuel bundles in the PROTEUS facility had been analyzed with the PHOENIX4 code, reactor scram experiments in the Loviisa and Mochovce VVER reactors using CASMO-4, MCNP4B and HEXTRAN, results of gamma scanning by the PHOENIX4/POLCA7 combination. Some difficulties in predicting the power distribution in the reactor core with sufficiently good accuracy using any of the available code systems were reported by OKG. Heating of non-fuel regions by gamma radiation and neutrons had been investigated using the HELIOS lattice code. Calculational results for heat deposition from gamma radiation in the moderator tank of the Forsmark-1 reactor were reported by Risø. -
Canada 176 Canada Canada
CANADA 176 CANADA CANADA 1. GENERAL INFORMATION 1.1. General Overview 2 Canada, occupying about 10 million km and having a population of over 30 million (Table 1), is one of the least densely populated countries in the world. Canada’s birth rate, at present, is 12 per 1,000 whereas death rate is seven per 1,000 with the result that the rate of natural population increase now stands at five per 1,000 persons. Canada has strong seasonal changes and large regional variations in temperature. The rigorous climate, the energy intensive nature of the country's industries, and the large distances between population centres produce a high per capita energy use. TABLE 1. POPULATION INFORMATION Growth rate (%) 1980 1960 1970 1980 1990 2000 2001 To 2001 Population (millions) 17.9 21.7 24.5 27.7 30.8 31.0 1.1 Population density (inhabitants/km²) 1.8 2.2 2.5 2.8 3.1 3.1 Predicted population growth rate (%) 2001 to 2010 7.2 Area (1000 km²) 9976.1 Urban population in 2001 as percent of total Source: IAEA Energy and Economic Database. 1.2. Economic Indicators Table 2 gives the statistical Gross Domestic Product (GDP) data and the GDP by sector. TABLE 2. GROSS DOMESTIC PRODUCT (GDP) Growth rate (%) 1980 1970 1980 1990 2000 2001 To 2001 GDP (millions of current US$) 266,002 572,676 687,752 711,912 4.8 GDP (millions of constant 1990 US$) 286,186 434,401 572,676 748,108 779,872 3 GDP per capita (current US$/capita) 10,850 20,674 22,361 22,954 3.6 Source: IAEA Energy and Economic Database. -
Vver and Rbmk Cross Section Libraries for Origen-Arp
VVER AND RBMK CROSS SECTION LIBRARIES FOR ORIGEN-ARP Germina Ilas, Brian D. Murphy, and Ian C. Gauld, Oak Ridge National Laboratory, USA Introduction An accurate treatment of neutron transport and depletion in modern fuel assemblies characterized by heterogeneous, complex designs, such as the VVER or RBMK assembly configurations, requires the use of advanced computational tools capable of simulating multi-dimensional geometries. The depletion module TRITON [1], which is part of the SCALE code system [2] that was developed and is maintained at the Oak Ridge National Laboratory (ORNL), allows the depletion simulation of two- or three-dimensional assembly configurations and the generation of burnup-dependent cross section libraries. These libraries can be saved for subsequent use with the ORIGEN-ARP module in SCALE. This later module is a faster alternative to TRITON for fuel depletion, decay, and source term analyses at an accuracy level comparable to that of a direct TRITON simulation. This paper summarizes the methodology used to generate cross section libraries for VVER and RBMK assembly configurations that can be employed in ORIGEN-ARP depletion and decay simulations. It briefly describes the computational tools and provides details of the steps involved. Results of validation studies for some of the libraries, which were performed using isotopic assay measurement data for spent fuel, are provided and discussed. Cross section libraries for ORIGEN-ARP Methodology The TRITON capability to perform depletion simulations for two-dimensional (2-D) configurations was implemented by coupling of the 2-D transport code NEWT with the point depletion and decay code ORIGEN-S. NEWT solves the transport equation on a 2-D arbitrary geometry grid by using an SN approach, with a treatment of the spatial variable that is based on an extended step characteristic method [3]. -
Nuclear Fuel Optimization and Problem of Xa9953256 Increasing Burnup and Cost Efficiency of Light Water Reactors in Russia
NUCLEAR FUEL OPTIMIZATION AND PROBLEM OF XA9953256 INCREASING BURNUP AND COST EFFICIENCY OF LIGHT WATER REACTORS IN RUSSIA M.I. SOLONIN, Yu. K. BIBLASHCVLI, F.G. RESHETNIKOV, F.F. SOKOLOV, A.A. BOCHVAR All-Russia Research Institute of Inorganic Materials, Moscow, Russia Abstract Brief analysis is given of the results of WWER-1000 and WWER-440 fuel assembly (FA) and fuel rod operation to the average burn-up of 44 and 34 MW-day/kg U, respectively. The reliable operation of the fuel is noted which made it possible to outline the paths of further improvements in their fuel cycles. Based on a series of improvements reactors are being switched on a 4-year cycle; a 5-year cycle is contemplated for the WWER- 440. The fuel cycle of the boiling RBMK reactors is subjected to essential alterations. Specifically, the use of erbium oxide as an integral burnable absorber allows a higher enrichment of uranium and bum-up. To extend the burn-up a new zirconium alloy E-635 is assumed to be used as fuel rod claddings and other FA components; the alloy has higher corrosion and irradiation resistance. INTRODUCTION The current technical level of the PWR and BWR fuel production and the perfected system of the reactor management have allowed to provide the successful operation of PWR NPP to the average burn-up of ~45 MW* day/kg U in all countries of developed nuclear power. The differences in the reliability and FA serviceability as well as in the number of leaky fuel rods are rather small and the attention is usually not focused on this issue. -
Research Branch
CA9600028 Background Paper BP-365E THE CANADIAN NUCLEAR POWER INDUSTRY Alan Nixon Science and Technology Division December 1993 Library of Parliament Research Bibliothèque du Parlement Branch The Research Branch of the Library of Parliament works exclusively for Parliament conducting research and providing information for Committees and Members of the Senate and the House of Commons. This service is extended without partisan bias in such forms as Reports, Background Papers and Issue Reviews. Research Officers in the Branch are also available for personal consultation in their respective fields of expertise. ©Minister of Supply and Services Canada 1994 Available in Canada through your local bookseller or by mail from Canada Communication Group -- Publishing Ottawa, Canada K1A 0S9 Catalogue No. YM32-2/365E ISBN 0-660-15639-3 CE DOCUMENT EST AUSSI PUBUÉ EN FRANÇAIS LIBRARY OF PARLIAMENT BIBLIOTHÈQUE OU PARLEMENT TABLE OF CONTENTS Page EARLY CANADIAN NUCLEAR DEVEOPMENT 2 THE CANDU REACTOR 4 NUCLEAR POWER GENERATION IN CANADA 5 A. Background 5 B. Performance 6 C. Pickering Nuclear Generating Station 8 D. Bruce Nuclear Generating Station 9 1. Retubing 9 2. Pressure Tube Frets 10 3. Shut Down System Design Flaw 12 4. Steam Generators 12 E. Darlington 13 1. Start-up Problems 13 2. Costs 14 AECL 15 A. Introduction 15 B. CANDU-Design and Marketing 16 1. Design 16 2. Marketing 17 a. Export Markets 17 b. Domestic Market ! 18 C. AECL Research 19 D. Recent Developments 20 OUTLOOK 21 * CANADA LIBRARY OF PARLIAMENT BIBLIOTHÈQUE DU PARLEMENT THE NUCLEAR POWER INDUSTRY IN CANADA Nuclear power, the production of electricity from uranium through nuclear fission, is by far the most prominent segment of the nuclear industry. -
Th, PU)02 BENCHMARK EXPERIMENTS in ZED-2 SENSITIVITY and UNCERTAINTY ANALYSIS OF
TSUNAMI ANALYSIS OF (Th, PU)02 BENCHMARK EXPERIMENTS IN ZED-2 SENSITIVITY AND UNCERTAINTY ANALYSIS OF (Th, PU)02 BENCHMARK EXPERIMENTS IN ZED-2 USING TSUNAMI BY TING ZHU, B.A.Sc. A THESIS SUBMITTED TO THE DEPARTMENT OF ENGINEERING PHYSICS AND THE SCHOOL OF GRADUATE STUDIES OF MCMASTER UNIVERSITY IN PARTIAL FULFILMENT OF THE REQUIREMENTS FOR THE DEGREE OF MASTER OF ApPLIED SCIENCE © Copyright by Ting Zhu, March 2011 Master of Applied Science (2011) McMaster University (Engineering Physics) Hamilton, Ontario, Canada TITLE: Sensitivity and Uncertainty Analysis of (Th, Pu)Oz Benchmark Experiments in ZED-2 Using TSUNAMI AUTHOR: Ting Zhu Bachelor of Applied Science in Engineering Sci ence (Physics) University of Toronto, Toronto, Canada SUPERVISOR: Dr. Adriaan Buijs NUMBER OF PAGES: x,88 ii Abstract In 1984, the ZED-2 research reactor was used to study five (Th, Pu)Oz fuel bundles with the goal to provide both benchmark tests for future reactor code validation and experimental measurements for a possible thorium fuel cycle in CANDU. In this work, the neutronic models of these critical exper iments were investigated by TSUNAMI, a sensitivity and uncertainty (SjU) analysis tool, part of the SCALE6 reactor physics package from the Oak Ridge National Laboratory. TSUNAMI consists of different modules that are capable of calculating the values of keff and the uncertainties in keff due to uncertainties in the nu clear data. It generates energy-dependent sensitivity coefficients from which the percentage change in keff due to perturbations in nuclear data values can be determined. The calculated keff has a bias which is the difference between calculation and measurement. -
Nuclear in Canada NUCLEAR ENERGY a KEY PART of CANADA’S CLEAN and LOW-CARBON ENERGY MIX Uranium Mining & Milling
Nuclear in Canada NUCLEAR ENERGY A KEY PART OF CANADA’S CLEAN AND LOW-CARBON ENERGY MIX Uranium Mining & Milling . Nuclear electricity in Canada displaces over 50 million tonnes of GHG emissions annually. Electricity from Canadian uranium offsets more than 300 million tonnes of GHG emissions worldwide. Uranium Processing – Re ning, Conversion, and Fuel Fabrication Yellowcake is re ned at Blind River, Ontario, PELLETS to produce uranium trioxide. At Port Hope, Ontario, Nuclear Power Generation and Nuclear Science & uranium trioxide is At plants in southern Technology TUBES converted. URANIUM DIOXIDE Ontario, fuel pellets are UO2 is used to fuel CANDU loaded into tubes and U O UO URANIUM Waste Management & Long-term Management 3 8 3 nuclear reactors. assembled into fuel YUKON TRIOXIDE UO2 Port Radium YELLOWCAKE REFINING URANIUM bundles for FUEL BUNDLE Shutdown or Decommissioned Sites TRIOXIDE UF is exported for 6 CANDU reactors. UO enrichment and use Rayrock NUNAVUT 3 CONVERSION UF Inactive or Decommissioned Uranium Mines and 6 in foreign light water NORTHWEST TERRITORIES Tailings Sites URANIUM HEXAFLUORIDE reactors. 25 cents 400 kg of COAL Beaverlodge, 2.6 barrels of OIL Gunnar, Lorado NEWFOUNDLAND AND LABRADOR McClean Lake = 3 Cluff Lake FUEL PELLET Rabbit Lake of the world’s 350 m of GAS BRITISH COLUMBIA Cigar Lake 20% McArthur River production of uranium is NVERSION Key Lake QUEBEC CO mined and milled in northern FU EL ALBERTA SASKATCHEWAN MANITOBA F Saskatchewan. AB G R University of IN IC ONTARIO P.E.I. IN A Saskatchewan The uranium mining F T E IO 19 CANDU reactors at Saskatchewan industry is the largest R N TRIUMF NEW BRUNSWICK Research Council NOVA SCOTIA private employer of Gentilly-1 & -2 Whiteshell Point Lepreau 4 nuclear power generating stations Rophton NPD Laboratories Indigenous people in CANDU REACTOR Chalk River Laboratories Saskatchewan. -
CANDU Fundamentals
CANDU Fundamentals CANDU Fundamentals CANDU Fundamentals Table of Contents 1 OBJECTIVES ............................................................................. 1 1.1 COURSE OVERVIEW ............................................................... 1 1.2 ATOMIC STRUCTURE.............................................................. 1 1.3 RADIOACTIVITY – SPONTANEOUS NUCLEAR PROCESSES ....... 1 1.4 NUCLEAR STABILITY AND INSTABILITY................................. 2 1.5 ACTIVITY ............................................................................... 2 1.6 NEUTRONS AND NEUTRON INTERACTIONS............................. 2 1.7 FISSION .................................................................................. 2 1.8 FUEL, MODERATOR, AND REACTOR ARRANGEMENT............. 2 1.9 NUCLEAR SAFETY.................................................................. 3 1.10 NUCLEAR POWER REACTORS................................................. 3 1.11 CANDU REACTOR CONSTRUCTION ...................................... 4 1.12 MODERATOR AND MODERATOR SYSTEM............................... 4 1.13 MODERATOR COVER GAS SYSTEM & MODERATOR AUXILIARY SYSTEMS......................................................................... 5 1.14 HEAT TRANSPORT SYSTEM .................................................... 6 1.15 HEAT TRANSPORT AUXILIARY SYSTEMS ............................... 6 1.16 REACTOR FUEL ...................................................................... 7 1.17 NEUTRON LIFE CYCLE ...........................................................