CANDU Safety #3 - Nuclear Safety Characteristics Dr. V.G. Snell Director Safety & Licensing
|
|
|
- Arline Dalton
- 5 years ago
- Views:
Transcription
1 CANDU Safety #3 - Nuclear Safety Characteristics Dr. V.G. Snell Director Safety & Licensing 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 1
2 What Makes A Safe Nuclear Design? inherent nuclear safety is ideal all transients terminated safely by inherent negative neutronic feedback without any moving parts or fluids SLOWPOKE research reactor is in this category (20kW) passive nuclear safety is nice all transients terminated safely by inherent neutronic feedback and/or movement of fluids, without any external motive power or valves PIUS mostly in this category none of the major power reactors are of these types because of economics & power density requirements 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 2
3 Neutronic Time-scales prompt neutron lifetime 10-4 seconds (LWRs( LWRs) 10-3 seconds (CANDU) cannot control on prompt neutrons delayed neutrons (0.6% of all neutrons) lifetimes range from tenths to tens of seconds basis of safe control 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 3
4 Other Time-scales fuel temperature feedback - milliseconds coolant temperature feedback - ~7 seconds moderator temperature feedback: ~7 seconds in LWRs minutes in CANDU practical control system response - few seconds practical shutdown system response - 2 seconds 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 4
5 Sign of Reactivity Coefficients Coefficient CANDU LWR Fuel Temperature ( ) Coolant temperature + Coolant density + Moderator temperature + + if highly poisoned 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 5
6 Power Coefficient power coefficient reactivity change following a change in reactor power combination of fuel, temperature and moderator reactivity near zero for CANDU, negative for LWRs negative power coefficient is convenient since it simplifies the control system negative power coefficient is not required for safety in CANDU a large positive power coefficient would be unsafe since it would require a very fast control system for stability a large negative power coefficient requires deep shutdown systems to hold the reactor shut down 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 6
7 Response to Accidents Accident CANDU LWR Withdrawal of control rod - Small, negative Negative prompt feedback Withdrawal of control rod - long-term feedback Slow power increase terminated by shutdown Power increase stabilizes at higher power level; may Loss of coolant (rapid decrease in coolant density) Cold H 2 O injection (decrease in coolant temperature & purity) Steam Line Break (decrease in coolant temperature & density) or control systems Fast rise in power requiring shutdown systems Drop in power, H 2 O is a neutron absorber Drop in power, shutdown required in longer term require shutdown system Drop in power, shutdown required in longer term Increase in power, cold water must be borated Increase in power requiring shutdown Control rod ejection Not physically possible Rapid rise in power above prompt critical, stopped by fuel temperature feedback and eventually shutdown system 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 7
8 Other Neutronic Safety Characteristics avoid large amounts of reactivity compensation CANDU achieves this via on-power refueling LWRs compensate by using boron in coolant, recent designs use burnable poisons in fuel avoid large and fast positive reactivity insertions: CANDU - large LOCA LWR - rod ejection, steam line break ensure shutdown systems are effective for any core state possible during operation (major lesson from Chernobyl) 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 8
9 CANDU Fuel Temperature Reactivity 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 9
10 CANDU Moderator Temperature Reactivity 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 10
11 CANDU Coolant Temperature Reactivity 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 11
12 Coolant Void Reactivity 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 12
13 Reactor Period and Reactivity for Varying l* 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 13
14 Reactivity Control primary short term control: light-water zone compartments spatial and bulk power control flux shaping & xenon override: adjuster rods (normally inserted) rapid control power reduction on upsets: absorber rods (normally out of core) long term control: on-power refueling fresh fuel reactivity compensation: boron / gadolinium in moderator emergency shutdown: rods and poison injection 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 14
15 Reactivity Control zone controllers, absorber and adjuster rods come in from the top of the core into the moderator shutoff rods also come in from the top liquid poison injection (shutdown system 2) comes into the moderator from the side 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 15
16 Xenon CANDU is a large reactor (6m. long cylinder by 7.6m diameter) potential for spatial instability due to xenon decay (28mk) First Azimuthal A mk First Azimuthal B mk First Axial mk Subcriticality 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 16
17 Spatial Control xenon instabilities (first azimuthals) ) can occur; they are slow (hours) but require spatial detection and control spatial control also needed for local flux variations due to refueling (~8%) and global flux tilts due to asymmetric refueling 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 17
Reactivity Coefficients
Reactivity Coefficients B. Rouben McMaster University Course EP 4D03/6D03 Nuclear Reactor Analysis (Reactor Physics) 2015 Sept.-Dec. 2015 September 1 Reactivity Changes In studying kinetics, we have seen
(1) The time t required for N generations to elapse is merely:
19 Changes In Reactor Power With Time The two preceding modules discussed how reactivity changes increase or decrease neutron flux and hence, change the thermal power output from the fuel. We saw how the
20.1 Xenon Production Xe-135 is produced directly in only 0.3% of all U-235 fissions. The following example is typical:
20 Xenon: A Fission Product Poison Many fission products absorb neutrons. Most absorption cross-sections are small and are not important in short-term operation. Xenon- has a cross-section of approximately
Lesson 14: Reactivity Variations and Control
Lesson 14: Reactivity Variations and Control Reactivity Variations External, Internal Short-term Variations Reactivity Feedbacks Reactivity Coefficients and Safety Medium-term Variations Xe 135 Poisoning
Shutdown Margin. Xenon-Free Xenon removes neutrons from the life-cycle. So, xenonfree is the most reactive condition.
22.05 Reactor Physics - Part Thirty-One Shutdown Margin 1. Shutdown Margin: Shutdown margin (abbreviated here as SDM) is defined as the amount of reactivity by which a reactor is subcritical from a given
Chem 481 Lecture Material 4/22/09
Chem 481 Lecture Material 4/22/09 Nuclear Reactors Poisons The neutron population in an operating reactor is controlled by the use of poisons in the form of control rods. A poison is any substance that
Lectures on Applied Reactor Technology and Nuclear Power Safety. Lecture No 5. Title: Reactor Kinetics and Reactor Operation
Lectures on Nuclear Power Safety Lecture No 5 Title: Reactor Kinetics and Reactor Operation Department of Energy Technology KTH Spring 2005 Slide No 1 Outline of the Lecture (1) Reactor Kinetics Reactor
Introduction to Reactivity and Reactor Control
Introduction to Reactivity and Reactor Control Larry Foulke Adjunct Professor Director of Nuclear Education Outreach University of Pittsburgh IAEA Workshop on Desktop Simulation October 2011 Learning Objectives
Operational Reactor Safety
Operational Reactor Safety 22.091/22.903 Professor Andrew C. Kadak Professor of the Practice Lecture 3 Reactor Kinetics and Control Page 1 Topics to Be Covered Time Dependent Diffusion Equation Prompt
Hybrid Low-Power Research Reactor with Separable Core Concept
Hybrid Low-Power Research Reactor with Separable Core Concept S.T. Hong *, I.C.Lim, S.Y.Oh, S.B.Yum, D.H.Kim Korea Atomic Energy Research Institute (KAERI) 111, Daedeok-daero 989 beon-gil, Yuseong-gu,
Reactivity Coefficients
Revision 1 December 2014 Reactivity Coefficients Student Guide GENERAL DISTRIBUTION GENERAL DISTRIBUTION: Copyright 2014 by the National Academy for Nuclear Training. Not for sale or for commercial use.
Power Changes in a Critical Reactor. The Critical Reactor
Chapter 8 Power Changes in a Critical Reactor n For very small reactivity increases n For small reactivity increases n For large reactivity increases/decreases The Critical Reactor < k = hfpel f L t =
Lecture 27 Reactor Kinetics-III
Objectives In this lecture you will learn the following In this lecture we will understand some general concepts on control. We will learn about reactivity coefficients and their general nature. Finally,
but mostly as the result of the beta decay of its precursor 135 I (which has a half-life of hours).
8. Effects of 135Xe The xenon isotope 135 Xe plays an important role in any power reactor. It has a very large absorption cross section for thermal neutrons and represents therefore a considerable load
Reactor Kinetics and Operation
Reactor Kinetics and Operation Course No: N03-002 Credit: 3 PDH Gilbert Gedeon, P.E. Continuing Education and Development, Inc. 9 Greyridge Farm Court Stony Point, NY 0980 P: (877) 322-5800 F: (877) 322-4774
Lectures on Applied Reactor Technology and Nuclear Power Safety. Lecture No 4. Title: Control Rods and Sub-critical Systems
Lectures on Nuclear Power Safety Lecture No 4 Title: Control Rods and Sub-critical Systems Department of Energy Technology KTH Spring 2005 Slide No 1 Outline of the Lecture Control Rods Selection of Control
The moderator temperature coefficient MTC is defined as the change in reactivity per degree change in moderator temperature.
Moderator Temperature Coefficient MTC 1 Moderator Temperature Coefficient The moderator temperature coefficient MTC is defined as the change in reactivity per degree change in moderator temperature. α
APPLICATION OF THE COUPLED THREE DIMENSIONAL THERMAL- HYDRAULICS AND NEUTRON KINETICS MODELS TO PWR STEAM LINE BREAK ANALYSIS
APPLICATION OF THE COUPLED THREE DIMENSIONAL THERMAL- HYDRAULICS AND NEUTRON KINETICS MODELS TO PWR STEAM LINE BREAK ANALYSIS Michel GONNET and Michel CANAC FRAMATOME Tour Framatome. Cedex 16, Paris-La
Figure 22.1 Unflattened Flux Distribution
22 Neutron Flux Control If nothing were done to flatten the flux in our reactors, it would look something like Figure 22.1. The flux would be a maximum in the efuel of the reactor (where neutrons are moving
Chemical Engineering 412
Chemical Engineering 412 Introductory Nuclear Engineering Lecture 18 Nuclear Reactor Theory IV Reactivity Insertions 1 Spiritual Thought 2 Mosiah 2:33 33 For behold, there is a wo pronounced upon him who
3. State each of the four types of inelastic collisions, giving an example of each (zaa type example is acceptable)
Nuclear Theory - Course 227 OBJECTIVES to: At the conclusion of this course the trainee will be able 227.00-1 Nuclear Structure 1. Explain and use the ZXA notation. 2. Explain the concept of binding energy.
Fundamentals of Nuclear Reactor Physics
Fundamentals of Nuclear Reactor Physics E. E. Lewis Professor of Mechanical Engineering McCormick School of Engineering and Applied Science Northwestern University AMSTERDAM BOSTON HEIDELBERG LONDON NEW
Reactor Operation Without Feedback Effects
22.05 Reactor Physics - Part Twenty-Six Reactor Operation Without Feedback Effects 1. Reference Material: See pp. 363-368 of the article, Light Water Reactor Control Systems, in Wiley Encyclopedia of Electrical
Lecture 28 Reactor Kinetics-IV
Objectives In this lecture you will learn the following In this lecture we will understand the transient build up of Xenon. This can lead to dead time in reactors. Xenon also induces power oscillations
XV. Fission Product Poisoning
XV. Fission Product Poisoning XV.1. Xe 135 Buil-Up As we already know, temperature changes bring short-term effects. That is to say, once a power change is produced it is rapidly manifested as a change
Development of 3D Space Time Kinetics Model for Coupled Neutron Kinetics and Thermal hydraulics
Development of 3D Space Time Kinetics Model for Coupled Neutron Kinetics and Thermal hydraulics WORKSHOP ON ADVANCED CODE SUITE FOR DESIGN, SAFETY ANALYSIS AND OPERATION OF HEAVY WATER REACTORS October
SUB-CHAPTER D.1. SUMMARY DESCRIPTION
PAGE : 1 / 12 CHAPTER D. REACTOR AND CORE SUB-CHAPTER D.1. SUMMARY DESCRIPTION Chapter D describes the nuclear, hydraulic and thermal characteristics of the reactor, the proposals made at the present stage
N U C L : R E A C T O R O P E R A T I O N A N D R E G U L A T O R Y P O L I C Y, I
N U C L 6 0 6 0 : R E A C T O R O P E R A T I O N A N D R E G U L A T O R Y P O L I C Y, I FALL 2013 INSTRUCTORS: Gregory Moffitt & Ryan Schow LECTURES: MONDAY & WEDNESDAY 11:50 AM 1:10 PM MEB 1206 OFFICE
Coupled Neutronics Thermalhydraulics LC)CA Analysis
Coupled Neutronics Thermalhydraulics LC)CA Analysis B.Rouben, Manager Reactor Core Physics Branch,AE,CL Presented at Chulalongkorn University Bangkok, Thailand 9 1997 December RFSP -Reactor Fuelling Simulation
Reactivity Power and Temperature Coefficients Determination of the TRR
Reactivity and Temperature Coefficients Determination of the TRR ABSTRACT Ahmad Lashkari Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran Tehran 14399-51113,
Reactor Operation with Feedback Effects
22.05 Reactor Physics - Part Twenty-Nine Reactor Operation with Feedback Effects 1. Reference Material: See pp. 368 372 in Light Water Reactor Control Systems, in Wiley Encyclopedia of Electrical and Electronics
"Control Rod Calibration"
TECHNICAL UNIVERSITY DRESDEN Institute of Power Engineering Training Reactor Reactor Training Course Experiment "Control Rod Calibration" Instruction for Experiment Control Rod Calibration Content: 1...
Chapter 7 & 8 Control Rods Fission Product Poisons. Ryan Schow
Chapter 7 & 8 Control Rods Fission Product Poisons Ryan Schow Ch. 7 OBJECTIVES 1. Define rod shadow and describe its causes and effects. 2. Sketch typical differential and integral rod worth curves and
Xenon Effects. B. Rouben McMaster University Course EP 4D03/6D03 Nuclear Reactor Analysis (Reactor Physics) 2015 Sept.-Dec.
enon Effects B. Rouben McMaster University Course EP 4D03/6D03 Nuclear Reactor Analysis (Reactor Physics) 2015 Sept.-Dec. 2015 September 1 Contents We study the importance of e-135 in the operation of
Optimization of CARA Fuel Element with negative coolant void coefficient
Optimization of CARA Fuel Element with negative coolant void coefficient H. Lestani +, P. Florido *, J. González +, A. Marino 0 + Instituto Balseiro, CNEA - CONICET * Instituto Balseiro - Florestan technology
Neutron reproduction. factor ε. k eff = Neutron Life Cycle. x η
Neutron reproduction factor k eff = 1.000 What is: Migration length? Critical size? How does the geometry affect the reproduction factor? x 0.9 Thermal utilization factor f x 0.9 Resonance escape probability
AN ABSTRACT OF THE THESIS OF. Justin R. Mart for the degree of Master of Science in Nuclear Engineering presented on June 14, 2013.
AN ABSTRACT OF THE THESIS OF Justin R. Mart for the degree of Master of Science in Nuclear Engineering presented on June 14, 2013. Title: Feasibility Study on a Soluble Boron-Free Small Modular Reactor
Nuclear Theory - Course 227 REACTIVITY EFFECTS DUE TO TEMPERATURE CHANGES
Nuclear Theory - Course 227 REACTIVITY EFFECTS DUE TO TEMPERATURE CHANGES In the lesson on reactor kinetics we ignored any variations ln reactivity due to changes in power. As we saw in the previous lesson
PWR CONTROL ROD EJECTION ANALYSIS WITH THE MOC CODE DECART
PWR CONTROL ROD EJECTION ANALYSIS WITH THE MOC CODE DECART Mathieu Hursin and Thomas Downar University of California Berkeley, USA mhursin@nuc.berkeley.edu,downar@nuc.berkeley.edu ABSTRACT During the past
Solving Bateman Equation for Xenon Transient Analysis Using Numerical Methods
Solving Bateman Equation for Xenon Transient Analysis Using Numerical Methods Zechuan Ding Illume Research, 405 Xintianshiji Business Center, 5 Shixia Road, Shenzhen, China Abstract. After a nuclear reactor
«CALCULATION OF ISOTOPE BURN-UP AND CHANGE IN EFFICIENCY OF ABSORBING ELEMENTS OF WWER-1000 CONTROL AND PROTECTION SYSTEM DURING BURN-UP».
«CALCULATION OF ISOTOPE BURN-UP AND CHANGE IN EFFICIENCY OF ABSORBING ELEMENTS OF WWER-1000 CONTROL AND PROTECTION SYSTEM DURING BURN-UP». O.A. Timofeeva, K.U. Kurakin FSUE EDO «GIDROPRESS», Podolsk, Russia
Chain Reactions. Table of Contents. List of Figures
Chain Reactions 1 Chain Reactions prepared by Wm. J. Garland, Professor, Department of Engineering Physics, McMaster University, Hamilton, Ontario, Canada More about this document Summary: In the chapter
R.A. Chaplin Department of Chemical Engineering, University of New Brunswick, Canada
REACTVTY CHANGES R.A. Chaplin Department of Chemical Engineering, University of New Brunswick, Canada Keywords: Reactivity Coefficient, Fuel Burnup, Xenon, Samarium, Temperature Contents 1. ntroduction
Evaluating the Safety of Digital Instrumentation and Control Systems in Nuclear Power Plants
Evaluating the Safety of Digital Instrumentation and Control Systems in Nuclear Power Plants John Thomas With many thanks to Francisco Lemos for the nuclear expertise provided! System Studied: Generic
The Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit
The Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit ABSTRACT Peter Hermansky, Marian Krajčovič VUJE, Inc. Okružná
SLOWPOKE-2 31 YEARS AND STILL GLOWING
Slide 1 SLOWPOKE-2 31 YEARS AND STILL GLOWING Canadian Nuclear Society Conference 2012 Slide 2 SRC s SLOWPOKE-2 Reactor 31 Years and Still Glowing Slide 3 Safe LOW POWer Kritical Experiment Slowpoke is
Reactivity Balance & Reactor Control System
Reactivity Balance & Reactor Control System K.S. Rajan Professor, School of Chemical & Biotechnology SASTRA University Joint Initiative of IITs and IISc Funded by MHRD Page 1 of 6 Table of Contents 1 MULTIPLICATION
LECTURE 4: CORE DESIGN ANALYSIS
Dr. Giovanni (John) Brenciagfia Lecture 4: Core Design Analysis LECTURE 4: CORE DESIGN ANALYSIS MODULE OBJECTIVES: At the end of this module, you will be able to specify the combination of numerical methods
17 Neutron Life Cycle
17 Neutron Life Cycle A typical neutron, from birth as a prompt fission neutron to absorption in the fuel, survives for about 0.001 s (the neutron lifetime) in a CANDU. During this short lifetime, it travels
PHYSICS AND KINETICS OF TRIGA REACTOR. H. Böck and M. Villa AIAU 27307
PHYSICS AND KINETICS OF TRIGA REACTOR H. Böck and M. Villa AIAU 27307 *prepared for NTEC Overview This training module is written as an introduction to reactor physics for reactor operators. It assumes
A Hybrid Deterministic / Stochastic Calculation Model for Transient Analysis
A Hybrid Deterministic / Stochastic Calculation Model for Transient Analysis A. Aures 1,2, A. Pautz 2, K. Velkov 1, W. Zwermann 1 1 Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh Boltzmannstraße
Fission Reactors. Alternatives Inappropriate. Fission Reactors
Page 1 of 5 Fission Reactors The Polywell Reactor Nuclear Reactions Alternatives Inappropriate Hidden Costs of Carbon Web Site Home Page Fission Reactors There are about 438 Neutron Fission Power Reactors
Power Installations based on Activated Nuclear Reactions of Fission and Synthesis
Yu.V. Grigoriev 1,2, A.V. Novikov-Borodin 1 1 Institute for Nuclear Research RAS, Moscow, Russia 2 Joint Institute for Nuclear Research, Dubna, Russia Power Installations based on Activated Nuclear Reactions
ULOF Accident Analysis for 300 MWt Pb-Bi Coolled MOX Fuelled SPINNOR Reactor
ULOF Accident Analysis for 300 MWt Pb-Bi Coolled MOX Fuelled SPINNOR Reactor Ade afar Abdullah Electrical Engineering Department, Faculty of Technology and Vocational Education Indonesia University of
Lamarsh, "Introduction to Nuclear Reactor Theory", Addison Wesley (1972), Ch. 12 & 13
NEEP 428 REACTOR PULSING Page 1 April 2003 References: Lamarsh, "Introduction to Nuclear Reactor Theory", Addison Wesley (1972), Ch. 12 & 13 Notation: Lamarsh (2), "Introduction to Nuclear Engineering",(1975),
Nuclear Theory - Course 127 EFFECTS OF FUEL BURNUP
Nuclear Theory - Course 127 EFFECTS OF FUEL BURNUP The effect of fuel burnup wa~ considered, to some extent, in a previous lesson. During fuel burnup, U-235 is used up and plutonium is produced and later
Sensitivity Analyses of the Peach Bottom Turbine Trip 2 Experiment
International Conference Nuclear Energy for New Europe 2003 Portorož, Slovenia, September 8-11, 2003 http://www.drustvo-js.si/port2003 Sensitivity Analyses of the Peach Bottom Turbine Trip 2 Experiment
Safety Analysis of Loss of Flow Transients in a Typical Research Reactor by RELAP5/MOD3.3
International Conference Nuclear Energy for New Europe 23 Portorož, Slovenia, September 8-11, 23 http://www.drustvo-js.si/port23 Safety Analysis of Loss of Flow Transients in a Typical Research Reactor
REACTOR PHYSICS FOR NON-NUCLEAR ENGINEERS
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2011) Rio de Janeiro, RJ, Brazil, May 8-12, 2011, on CD-ROM, Latin American Section (LAS)
Course Reactor & Auxiliaries - Module 1 - Moderator Heavy Watec. Module 1 MODERATOR ..." Y "... I
Approval Issue Module 1 MODERATOR I-ICA\/V \AIATeD..." Y. 1111 "... I Course 233 - Reactor & Auxiliaries - Module 1 - Moderator Heavy Watec ~t'\tcc 11_ CCl:I:CC~I~CC...,-... ~"...-...- OBJECTIVES: Mter
12 Moderator And Moderator System
12 Moderator And Moderator System 12.1 Introduction Nuclear fuel produces heat by fission. In the fission process, fissile atoms split after absorbing slow neutrons. This releases fast neutrons and generates
Nuclear Theory - Course 227
Lesson 227.00-2 NEUTRON BALANCE Nuclear Theory - Course 227 DURNG STEADY REACTOR OPERATON We have seen, in the previous lesson, what type of neutrons are produced and how they are produced in a reactor.
EXPERIMENTAL DETERMINATION OF NEUTRONIC PARAMETERS IN THE IPR-R1 TRIGA REACTOR CORE
2011 International Nuclear Atlantic Conference - INAC 2011 Belo Horizonte,MG, Brazil, October 24-28, 2011 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-04-5 EXPERIMENTAL DETERMINATION
9.1 MODULE OVERVIEW MODULE OBJECTIVES NEUTRON SOURCES IN THE CANDU REACTOR SOURCE MULTIPLICATION IN A SUBCRITICAL REACTOR...
- Module 9 Source Neutron Effects 9.1 MODULE OVERVEW 2 9.2 MODULE OBJECTVES 2 9.3 NEUTRON SOURCES N THE CANDU REACTOR 3 9.4 SOURCE MULTPLCATON N A SUBCRTCAL REACTOR...5 9.5 STABLZATON TME AFTER REACTVTY
3. Detector Systems. This chapter describes the CANDU detector systems.
3. Detector Systems This chapter describes the CANDU detector systems. 3.1 Zone-Control Detectors To vary the amount of water in the zone-control compartments, the Reactor Regulating System utilizes the
Fast Feedback Effects ofreactivity
Fast Feedback Effects ofreactivity Fuel Temperature Coolant Temperature Moderator Temperature Core Voiding Feedback - The Reactivity Chan~es as Power Changes When Power changes, the temperatures of reactor
Neutronic Calculations of Ghana Research Reactor-1 LEU Core
Neutronic Calculations of Ghana Research Reactor-1 LEU Core Manowogbor VC*, Odoi HC and Abrefah RG Department of Nuclear Engineering, School of Nuclear Allied Sciences, University of Ghana Commentary Received
Delayed neutrons in nuclear fission chain reaction
Delayed neutrons in nuclear fission chain reaction 1 Critical state Temporal flow Loss by leakage Loss by Absorption If the number of neutrons (the number of fission reactions) is practically constant
Subcritical Multiplication and Reactor Startup
22.05 Reactor Physics - Part Twenty-Five Subcritical Multiplication and Reactor Startup 1. Reference Material See pp. 357-363 of the article, Light Water Reactor Control Systems, in Wiley Encyclopedia
OECD/NEA Transient Benchmark Analysis with PARCS - THERMIX
OECD/NEA Transient Benchmark Analysis with PARCS - THERMIX Volkan Seker Thomas J. Downar OECD/NEA PBMR Workshop Paris, France June 16, 2005 Introduction Motivation of the benchmark Code-to-code comparisons.
Mechanical Engineering Introduction to Nuclear Engineering /12
Mechanical Engineering Objectives In this lecture you will learn the following In this lecture the population and energy scenario in India are reviewed. The imminent rapid growth of nuclear power is brought
PHYS-E0562 Ydinenergiatekniikan jatkokurssi Lecture 5 Burnup calculation
PHYS-E0562 Ydinenergiatekniikan jatkokurssi Lecture 5 Burnup calculation Jaakko Leppänen (Lecturer), Ville Valtavirta (Assistant) Department of Applied Physics Aalto University, School of Science Jaakko.Leppanen@aalto.fi
Research Article Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code
Hindawi Science and Technology of Nuclear Installations Volume 2017, Article ID 5151890, 8 pages https://doi.org/10.1155/2017/5151890 Research Article Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal
Metropolitan Community College COURSE OUTLINE FORM LAB: 3.0
Metropolitan Community College COURSE OUTLINE FORM Course Title: Nuclear Plant Operation II Course Prefix & No.: LEC: PROT - 2420 3.0 COURSE DESCRIPTION: LAB: 0 Credit Hours: 3.0 This course introduces
ANALYSIS OF THE OECD PEACH BOTTOM TURBINE TRIP 2 TRANSIENT BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE TRAC-M/PARCS
ANALYSIS OF THE OECD PEACH BOTTOM TURBINE TRIP 2 TRANSIENT BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE TRAC-M/PARCS Deokjung Lee and Thomas J. Downar School of Nuclear Engineering
Core-Physics Aspects of Safety Analysis. B. Roubeo Manager, Reactor Core Physics Branch AECL
Core-Physics Aspects of Safety Analysis by B. Roubeo Manager, Reactor Core Physics Branch AECL Presented at Chulalongkorn University Bangkok, Thailand, 1997 December Abstract To illustrate the physics
MEASUREMENT AND COMPARISON OF CONTROL ROD WORTH OF BTRR USING INHOUR EQUATION AND PERIOD REACTIVITY CONVERSION TABLE
J. Bangladesh Acad. Sci., Vol. 41, No. 1, 95-103, 2017 MEASUREMENT AND COMPARISON OF CONTROL ROD WORTH OF BTRR USING INHOUR EQUATION AND PERIOD REACTIVITY CONVERSION TABLE MD. IQBAL HOSAN, M. A. M. SONER
HIGH TEMPERATURE THERMAL HYDRAULICS MODELING
HIGH TEMPERATURE THERMAL HYDRAULICS MODELING OF A MOLTEN SALT: APPLICATION TO A MOLTEN SALT FAST REACTOR (MSFR) P. R. Rubiolo, V. Ghetta, J. Giraud, M. Tano Retamales CNRS/IN2P3/LPSC - Grenoble Workshop
DEVELOPMENT OF A COUPLED CODE SYSTEM BASED ON SPACE SAFETY ANALYSIS CODE AND RAST-K THREE-DIMENSIONAL NEUTRONICS CODE
DEVELOPMENT OF A COUPLED CODE SYSTEM BASED ON SPACE SAFETY ANALYSIS CODE AND RAST-K THREE-DIMENSIONAL NEUTRONICS CODE Seyun Kim, Eunki Lee, Yo-Han Kim and Dong-Hyuk Lee Central Research Institute, Korea
Challenges in Prismatic HTR Reactor Physics
Challenges in Prismatic HTR Reactor Physics Javier Ortensi R&D Scientist - Idaho National Laboratory www.inl.gov Advanced Reactor Concepts Workshop, PHYSOR 2012 April 15, 2012 Outline HTR reactor physics
NUCLEAR SCIENCE ACAD BASIC CURRICULUM CHAPTER 5 NEUTRON LIFE CYCLE STUDENT TEXT REV 2. L th. L f U-235 FUEL MODERATOR START CYCLE HERE THERMAL NEUTRON
ACAD BASIC CURRICULUM NUCLEAR SCIENCE CHAPTER 5 NEUTRON LIFE CYCLE 346 RESONANCE LOSSES p 038 THERMAL NEUTRON 2 THERMAL NEUTRON LEAKAGE 52 THERMAL ABSORBED BY NON-FUEL ATOMS L th 07 THERMAL f 965 THERMAL
Year 11 Physics booklet Topic 1 Atomic structure and radioactivity Name:
Year 11 Physics booklet Topic 1 Atomic structure and radioactivity Name: Atomic structure and radioactivity Give a definition for each of these key words: Atom Isotope Proton Neutron Electron Atomic nucleus
Ph.D. Dissertation. Khaled Sayed Mahmoud. Department of Nuclear Techniques Institute of Nuclear Techniques Budapest University of Technology
NUMERICAL MODELING OF REACTIVITY EXCURSION ACCIDENTS IN SMALL LIGHT WATER REACTORS Ph.D. Dissertation Khaled Sayed Mahmoud Department of Nuclear Techniques Institute of Nuclear Techniques Budapest University
ANALYSIS OF THE OECD MSLB BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE RELAP5/PARCS
ANALYSIS OF THE OECD MSLB BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE RELAP5/PARCS T. Kozlowski, R. M. Miller, T. Downar School of Nuclear Engineering Purdue University United States
NEW FUEL IN MARIA RESEARCH REACTOR, PROVIDING BETTER CONDITIONS FOR IRRADIATION IN THE FAST NEUTRON SPECTRUM.
NEW FUEL IN MARIA RESEARCH REACTOR, PROVIDING BETTER CONDITIONS FOR IRRADIATION IN THE FAST NEUTRON SPECTRUM. M. LIPKA National Centre for Nuclear Research Andrzeja Sołtana 7, 05-400 Otwock-Świerk, Poland
NUCLEAR SAFETY AND RELIABILITY WEEK 8
Nuclear Safety and Reliability Dan Meneley Page 1 of 1 NUCLEAR SAFETY AND RELIABILITY WEEK 8 TABLE OF CONTENTS - WEEK 8 Loss of Primary Coolant Analysis...1 (1) Plant analysis...1 Potential Leaks or Breaks...2
Conceptual Design of Inherently Safe Fast Reactor (ISFR)
GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1015 Conceptual Design of Inherently Safe Fast Reactor (ISFR) Yoshiro ASAHI* Japan Atomic Energy Research Institute, Tokai, Ibaraki, Japan 319-1195
CHAPTER NEUTRON DETECTORS
73 CHAPTER 6 CONTROL AND OPERATONS The basic and unique dynamic characteristics associated with the reactor core have been discussed in the preceding chapter; here we will focus upon certain reactor core
Some Results in studying NPP accidents using PCTRAN-2 LOOP and WWER-1000 simulators
Technical Meeting on Effective Utilization of Nuclear Power Plant Simulators as Introductory Educational Tools Some Results in studying NPP accidents using PCTRAN-2 LOOP and WWER-1 simulators Vo Hong Hai,
Correlation between neutrons detected outside the reactor building and fuel melting
Attachment 2-7 Correlation between neutrons detected outside the reactor building and fuel melting 1. Introduction The Fukushima Daiichi Nuclear Power Station (hereinafter referred to as Fukushima Daiichi
A METHOD TO PREVENT SEVERE POWER AND FLOW OSCILLATIONS IN BOILING WATER REACTORS
A METHOD TO PREVENT SEVERE POWER AND FLOW OSCILLATIONS IN BOILING WATER REACTORS Yousef M. Farawila Farawila et al., Inc. Nuclear@Farawila.com ABSTRACT This paper introduces a new method for preventing
XA04C1727. Abstract. 1 Introduction
XA04C1727 Validation of Computer Codes Used in the Safety Analysis of Canadian Research Reactors by W. E. Bishop and A.G. Lee AECL 2251 Speakman Drive Mississauga, Ontario K IB2 Phone: 905)823-9040 FAX:
Physics Codes and Methods for CANDU Reactor
Physics Codes and Methods for CANDU Reactor by Zhu XingGuan*, Cai Jianping* and Chow H.C.** *Shanghai Nuclear Engineering Research & Design Institute ** Atomic Energy of Canada Limited Abstract This paper
Control Rod Calibration and Worth Calculation for Optimized Power Reactor 1000 (OPR-1000) Using Core Simulator OPR1000
World Journal of Nuclear Science and Technology, 017, 7, 15-3 http://www.scirp.org/journal/wjnst ISSN Online: 161-6809 ISSN Print: 161-6795 Control Rod Calibration and Worth Calculation for Optimized Power
Nuclear Fission. 1/v Fast neutrons. U thermal cross sections σ fission 584 b. σ scattering 9 b. σ radiative capture 97 b.
Nuclear Fission 1/v Fast neutrons should be moderated. 235 U thermal cross sections σ fission 584 b. σ scattering 9 b. σ radiative capture 97 b. Fission Barriers 1 Nuclear Fission Q for 235 U + n 236 U
FUEL PERFORMANCE EVALUATION THROUGH IODINE ACTIVITY MONITORING K.ANANTHARAMAN, RAJESH CHANDRA
6A-46 FUEL PERFORMANCE EVALUATION THROUGH IODINE ACTIVITY MONITORING K.ANANTHARAMAN, RAJESH CHANDRA CA9800600 Refuelling Technology Division Bhabha Atomic Research Centre Bo m ba,-4000 85,INDIA ABSTRACT
Estimation of Control Rods Worth for WWR-S Research Reactor Using WIMS-D4 and CITATION Codes
Estimation of Control Rods Worth for WWR-S Research Reactor Using WIMS-D4 and CITATION Codes M. S. El-Nagdy 1, M. S. El-Koliel 2, D. H. Daher 1,2 )1( Department of Physics, Faculty of Science, Halwan University,
NATURAL CONVECTION HEAT TRANSFER CHARACTERISTICS OF KUR FUEL ASSEMBLY DURING LOSS OF COOLANT ACCIDENT
NATURAL CONVECTION HEAT TRANSFER CHARACTERISTICS OF KUR FUEL ASSEMBLY DURING LOSS OF COOLANT ACCIDENT Ito D*, and Saito Y Research Reactor Institute Kyoto University 2-1010 Asashiro-nishi, Kumatori, Sennan,
NEUTRONIC CALCULATION SYSTEM FOR CANDU CORE BASED ON TRANSPORT METHODS
Romanian Reports in Physics, Vol. 63, No. 4, P. 948 960, 2011 NEUTRONIC CALCULATION SYSTEM FOR CANDU CORE BASED ON TRANSPORT METHODS V. BALACEANU 1, M. PAVELESCU 2 1 Institute for Nuclear Research, PO
INTRODUCTION TO NUCLEAR REACTORS AND NUCLEAR POWER GENERATION. Atsushi TAKEDA & Hisao EDA
INTRODUCTION TO NUCLEAR REACTORS AND NUCLEAR POWER GENERATION Atsushi TAKEDA & Hisao EDA 1 CONTENTS The first step toward nuclear power Physics of nuclear fission Sustained chain reaction in nuclear reactor