SCWR Research in Korea. Yoon Y. Bae KAERI
|
|
|
- Ami Beatrix Clarke
- 5 years ago
- Views:
Transcription
1 SCWR Research in Korea Yoon Y. ae KAERI
2 Organization President Dr. In-Soon Chnag Advanced Reactor Development Dr. Jong-Kyun Park Nuclear Engineering & Research Dr. M. H. Chang Mechanical Engineering & Research Dr. K.. Park Fluid Engineering & Research Dr. Y. Y. ae Thermal HYdraulic & Safety Research Dr. H. D. Kim PSA Dr. J. J. Ha Advanced Fuel Development Dr. D. S. Son PWR Fuel Development Dr. Y. H. Chung GEN IV KALIMER, Dr. D. H. Han SMART. Dr. S. K. Zee Nuclear Data Evaluation lab. 2002/11/19 2
3 Activities Feasibility study sponsored by KISTEP Provide information for decision making of MOST State of art Identify research items Lay foundation of SCWR research Draw attention of research community into SCWR 2002/11/19 3
4 Activities Feasibility study of SCWR concept Design Limits Core Primary and Safety Systems System performance and efficiency (KAERI Internal project) 2002/11/19 4
5 Activities I-NERI Radiolysis and Heat Transfer Studies for Supercritical-Water-Cooled Reactors (SCWRs) Mid to Long Term Nuclear R&D? 2002/11/19 5
6 Preliminary calculation 2002/11/19 6
7 S/G Sizing - Assumptions Use Griem s correlation Nu = Re Pr ρw ρ Assume U-tube length 10m ω Assume 12% of tube length is economizer Overall HTC for economizer and evaporater are 3,355 and 7,565 W/m 2 K /11/19 7
8 S/G Sizing Procedures h i = Nu d 1 = U k tube 1 h i + t k tube + 1 h o ( ) Q = UA T T sat Repeat until Q reaches 4000 MWt ( A = 2πd x) T w = T Q h A i T = T, i+ 1, i + Q m C SG P 2002/11/19 8
9 S/G Sizing - Results 2002/11/19 9
10 S/G Sizing - Results 2002/11/19 10
11 Core Sizing - Assumptions Use ishop s correlation Pitch/Diameter = 1.15 Nu Pr = Re C µ P =, k C P 2/3 Hydraulic diameter 3.26 mm Fuel diameter = 9.16 mm 0.9 Pr h = T W W ρ W ρ h T Tight lattice hexagonal fuel assembly 160 ass y 397 rods(324 fuel)/ass y D L 2002/11/19 11
12 Core Sizing - Procedures T b, i T = b, i 1 + q A m C c p Repeat until T reaches T hot = 565C T w = T b, i + q h i [ h i = f ( Tw,...)] 2002/11/19 12
13 Core Sizing - Results 2002/11/19 13
14 Inputs and Outputs Technical Data System pressure = MPa Steam Generator T COLD = ºC Core power = MWt Heat transfer per S/G = MWt Mass flow rate per S/G = kg/s Total flow area = m 2 Number of tubes = Tube length = m Total heat transfer area = m 2 Mass flux = kg/m 2.s Average heat flux = MW/m /11/19 14
15 Inputs and Outputs Core Fuel diameter = mm Pitch of fuel rod = mm Pitch to fuel diameter = mm Gap between fuel rods = mm Pitch of fuel assembly = mm Hydraulic diameter = mm Number of rod per fuel ass y = 397 Number of fuel rod per fuel ass y = 324 Number of fuel ass y = 160 Cross sect. area of all fuel ass y = m 2 Cross sect. area of all fuel rod = m 2 Net fuel channel area = m 2 Core mass flow rate = kg/s Core mass flux = kg/m 2 s Core heat flux = MW/m 2 Core length = m Average specific power = kw/m RCP power = kw T HOT = ºC 2002/11/19 15
16 Thermal Efficiency From SG MS Reheater Reheater HP TN A C LP TN CON C A To SG FWP COP 2002/11/19 16
17 Thermal Efficiency A: Primary loop, SC condition + secondary of APR1400 (6.895 MPa saturated steam) : Option A + secondary of 20 MPa saturated steam C: Option A + secondary of 20 MPa superheated steam ( T = 100 ) D: Option A + secondary of SC condition, 24 MPa, 500 ) E: High pressure turbine at primary + secondary of APR /11/19 17
18 Thermal Efficiency A C D E Core Power (MWt) Flow rate(kg/s) Primary SG Inlet Pres. (MPa) SG Outlet Pres. (MPa) SG Inlet Temp. (C) SG Outlet Temp. (C) Steam Pres. (MPa) Secondary Steam Temp. (C) Steam Flow rate (kg/s)) Feedwater Temp. (C) Feedwater Pres. (MPa) Results Elec. Power (MWe) Pump Power (MW) ( ) 66 (39+27) Efficiency (%) /11/19 18
The Research of Heat Transfer Area for 55/19 Steam Generator
Journal of Power and Energy Engineering, 205, 3, 47-422 Published Online April 205 in SciRes. http://www.scirp.org/journal/jpee http://dx.doi.org/0.4236/jpee.205.34056 The Research of Heat Transfer Area
NPP Simulators for Education Workshop - Passive PWR Models
NPP Simulators for Education Workshop - Passive PWR Models Wilson Lam (wilson@cti-simulation.com) CTI Simulation International Corp. www.cti-simulation.com Sponsored by IAEA Learning Objectives Understand
The exergy of asystemis the maximum useful work possible during a process that brings the system into equilibrium with aheat reservoir. (4.
Energy Equation Entropy equation in Chapter 4: control mass approach The second law of thermodynamics Availability (exergy) The exergy of asystemis the maximum useful work possible during a process that
HEAT TRANSFER AT SUPERCRITICAL PRESSURES (SURVEY) 1
HEAT TRANSFER AT SUPERCRITICAL PRESSURES (SURVEY) 1 Igor Pioro*, Hussam Khartail and Romney Duffey Chalk River Laoratories, AECL, Chalk River, ON, Canada K0J 1J0 Keywords: Supercritical pressure, forced
ENGINEERING OF NUCLEAR REACTORS
22.312 ENGINEERING OF NUCLEAR REACTORS Monday, December 17 th, 2007, 9:00am-12:00 pm FINAL EXAM SOLUTIONS Problem 1 (45%) Analysis of Decay Heat Removal during a Severe Accident i) The energy balance for
22.06 ENGINEERING OF NUCLEAR SYSTEMS OPEN BOOK FINAL EXAM 3 HOURS
22.6 ENGINEERING OF NUCLEAR SYSTEMS OPEN BOOK FINAL EXAM 3 HOURS Short Questions (1% each) a) The specific power in a UO 2 pellet of a certain LWR is q"'=2 W/cm 3. The fuel 235 U enrichment is 4 % by weight.
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
DEVELOPMENT OF COMPUTATIONAL MULTIFLUID DYNAMICS MODELS FOR NUCLEAR REACTOR APPLICATIONS
DEVELOPMENT OF COMPUTATIONAL MULTIFLUID DYNAMICS MODELS FOR NUCLEAR REACTOR APPLICATIONS Henry Anglart Royal Institute of Technology, Department of Physics Division of Nuclear Reactor Technology Stocholm,
Lectures on Applied Reactor Technology and Nuclear Power Safety. Lecture No 6
Lectures on Nuclear Power Safety Lecture No 6 Title: Introduction to Thermal-Hydraulic Analysis of Nuclear Reactor Cores Department of Energy Technology KTH Spring 2005 Slide No 1 Outline of the Lecture
Department of Engineering and System Science, National Tsing Hua University,
3rd International Conference on Materials Engineering, Manufacturing Technology and Control (ICMEMTC 2016) The Establishment and Application of TRACE/CFD Model for Maanshan PWR Nuclear Power Plant Yu-Ting
Improvement of Critical Heat Flux Performance by Wire Spacer
Journal of Energy and Power Engineering 9 (215) 844-851 doi: 1.17265/1934-8975/215.1.2 D DAVID PUBLISHING Improvement of Critical Heat Flux Performance by Wire Spacer Dan Tri Le 1 and Minoru Takahashi
The average velocity of water in the tube and the Reynolds number are Hot R-134a
hater 0:, 8, 4, 47, 50, 5, 55, 7, 75, 77, 8 and 85. 0- Refrigerant-4a is cooled by water a double-ie heat exchanger. he overall heat transfer coefficient is to be determed. Assumtions he thermal resistance
ENGINEERING OF NUCLEAR REACTORS. Fall December 17, 2002 OPEN BOOK FINAL EXAM 3 HOURS
22.312 ENGINEERING OF NUCLEAR REACTORS Fall 2002 December 17, 2002 OPEN BOOK FINAL EXAM 3 HOURS PROBLEM #1 (30 %) Consider a BWR fuel assembly square coolant subchannel with geometry and operating characteristics
EasyChair Preprint. Numerical Simulation of Fluid Flow and Heat Transfer of the Supercritical Water in Different Fuel Rod Channels
EasyChair Preprint 298 Numerical Simulation of Fluid Flow and Heat Transfer of the Supercritical Water in Different Fuel Rod Channels Huirui Han and Chao Zhang EasyChair preprints are intended for rapid
ATLAS Facility Description Report
KAERI/TR-3754/2009 기술보고서 ATLAS Facility Description Report ATLAS 실험장치기술보고서 한국원자력연구원 제출문 한국원자력연구원장귀하 본보고서를 2009 연도 APR1400/OPR1000 핵심사고열수력종합 효과실험 과제의기술보고서로제출합니다. 2009. 4. 주저자 : 강경호공저자 : 문상기박현식조석최기용 ATLAS
Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 41, No. 7, p (July 2004)
Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 41, No. 7, p. 765 770 (July 2004) TECHNICAL REPORT Experimental and Operational Verification of the HTR-10 Once-Through Steam Generator (SG) Heat-transfer
ECE309 INTRODUCTION TO THERMODYNAMICS & HEAT TRANSFER. 13 June 2007
ECE309 INTRODUCTION TO THERMODYNAMICS & HEAT TRANSFER 13 June 2007 Midterm Examination R. Culham This is a 2 hour, open-book examination. You are permitted to use: course text book calculator There are
English text only NUCLEAR ENERGY AGENCY COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS
Unclassified NEA/CSNI/R(2008)6/VOL2 NEA/CSNI/R(2008)6/VOL2 Unclassified Organisation de Coopération et de Développement Économiques Organisation for Economic Co-operation and Development 26-Nov-2008 English
CONCEPTUAL DESIGN FOR A RE-ENTRANT TYPE FUEL CHANNEL FOR SUPERCRITICAL WATER-COOLED NUCLEAR REACTORS
CONCEPTUAL DESIGN FOR A RE-ENTRANT TYPE FUEL CHANNEL FOR SUPERCRITICAL WATER-COOLED NUCLEAR REACTORS by Jeffrey Samuel A Thesis Submitted in Partial Fulfillment of the Requirements for the Degree of Master
A DIRECT STEADY-STATE INITIALIZATION METHOD FOR RELAP5
A DIRECT STEADY-STATE INITIALIZATION METHOD FOR RELAP5 M. P. PAULSEN and C. E. PETERSON Computer Simulation & Analysis, Inc. P. O. Box 51596, Idaho Falls, Idaho 83405-1596 for presentation at RELAP5 International
Lectures on Applied Reactor Technology and Nuclear Power Safety. Lecture No 7
ectures on Nuclear Power Safety ecture No 7 itle: hermal-hydraulic nalysis of Single-Phase lows in Heated hannels Department of Energy echnology KH Spring 005 Slide No Outline of the ecture lad-oolant
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
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
Heat Transfer And Fluid Flow Analysis Of Pressure Tube In Candu 6 Nuclear Reactor Using Supercritical Water
Heat Transfer And Fluid Flow Analysis Of Pressure Tube In Candu 6 Nuclear Reactor Using Supercritical Water Lakshmana Kishore.T #1, Dr. Kiran Chaudhari *2, Dr. G. Ranga Janardhana #3 #1 Assistant Professor
STUDY OF HEAT TRANSFER IN A 7-ELEMENT BUNDLE COOLED WITH THE UPWARD FLOW OF SUPERCRITICAL FREON-12
STUDY OF HEAT TRANSFER IN A 7-ELEMENT BUNDLE COOLED WITH THE UPWARD FLOW OF SUPERCRITICAL FREON-12 by Graham Richards A Thesis Submitted in Partial Fulfillment of the Requirements for the Degree of Master
ME 354 THERMODYNAMICS 2 MIDTERM EXAMINATION. Instructor: R. Culham. Name: Student ID Number: Instructions
ME 354 THERMODYNAMICS 2 MIDTERM EXAMINATION February 14, 2011 5:30 pm - 7:30 pm Instructor: R. Culham Name: Student ID Number: Instructions 1. This is a 2 hour, closed-book examination. 2. Answer all questions
Heat Transfer Predictions for Carbon Dioxide in Boiling Through Fundamental Modelling Implementing a Combination of Nusselt Number Correlations
Heat Transfer Predictions for Carbon Dioxide in Boiling Through Fundamental Modelling Implementing a Combination of Nusselt Number Correlations L. Makaum, P.v.Z. Venter and M. van Eldik Abstract Refrigerants
Experiences of TRAC-P code at INS/NUPEC
Exploratory Meeting of Experts on BE Calculations and Uncertainty Analysis in Aix en Provence, May 13-14, 2002 Experiences of TRAC-P code at INS/NUPEC Fumio KASAHARA (E-mail : kasahara@nupec or jp) Institute
PROBLEM h fg ρ v ρ l σ 10 3 T sat (kj/kg) (kg/m 3 ) (N/m) (K)
PROBLEM 10.9 KNOWN: Fluids at 1 atm: mercury, ethanol, R-1. FIND: Critical heat flux; compare with value for water also at 1 atm. ASSUMPTIONS: (1) Steady-state conditions, () Nucleate pool boiling. PROPERTIES:
Determination of heat losses of a concrete silo for sugar and a fan project
Determination of heat losses of a concrete silo for sugar and a fan project VLP - Processing lines of food industry - example, 5/2000. Given data Silo diameter Height of a cylindrical part of silo Concrete
PREDICTION OF MASS FLOW RATE AND PRESSURE DROP IN THE COOLANT CHANNEL OF THE TRIGA 2000 REACTOR CORE
PREDICTION OF MASS FLOW RATE AND PRESSURE DROP IN THE COOLANT CHANNEL OF THE TRIGA 000 REACTOR CORE Efrizon Umar Center for Research and Development of Nuclear Techniques (P3TkN) ABSTRACT PREDICTION OF
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
The Pennsylvania State University. The Graduate School. Department Of Mechanical & Nuclear Engineering
The Pennsylvania State University The Graduate School Department Of Mechanical & Nuclear Engineering FEASIBILITY STUDY FOR THE CONCEPTUAL DESIGN OF A DUAL-CORE BOILING SUPERHEAT REACTOR A Thesis in Nuclear
Safety analysis on beam dump Luca de Ruvo LNL - INFN Safety group SSTAC meeting, July 23, 2015
Safety analysis on beam dump Luca de Ruvo LNL - INFN Safety group SSTAC meeting, July 23, 2015 Topics: 1. Overview on Beam Dump 2. Description of cooling circuit 3. Safety philosofy 4. BD safety system:
Thermal Hydraulic Considerations in Steady State Design
Thermal Hydraulic Considerations in Steady State Design 1. PWR Design 2. BWR Design Course 22.39, Lecture 18 11/10/05 1 PWR Design Unless specified otherwise, all figures in this presentation are from:
ME 201 Thermodynamics
ME 0 Thermodynamics Solutions First Law Practice Problems. Consider a balloon that has been blown up inside a building and has been allowed to come to equilibrium with the inside temperature of 5 C and
Chapter 11: Heat Exchangers. Dr Ali Jawarneh Department of Mechanical Engineering Hashemite University
Chapter 11: Heat Exchangers Dr Ali Jawarneh Department of Mechanical Engineering Hashemite University Objectives When you finish studying this chapter, you should be able to: Recognize numerous types of
Title: Development of a multi-physics, multi-scale coupled simulation system for LWR safety analysis
Title: Development of a multi-physics, multi-scale coupled simulation system for LWR safety analysis Author: Yann Périn Organisation: GRS Introduction In a nuclear reactor core, different fields of physics
Comparison of Silicon Carbide and Zircaloy4 Cladding during LBLOCA
Comparison of Silicon Carbide and Zircaloy4 Cladding during LBLOCA Prepared By: Kwangwon Ahn Prepared For: 22.314 Prepared On: December 7 th, 2006 Comparison of Silicon Carbide and Zircaloy4 Cladding during
Chapter 5. Mass and Energy Analysis of Control Volumes
Chapter 5 Mass and Energy Analysis of Control Volumes Conservation Principles for Control volumes The conservation of mass and the conservation of energy principles for open systems (or control volumes)
MAE 11. Homework 8: Solutions 11/30/2018
MAE 11 Homework 8: Solutions 11/30/2018 MAE 11 Fall 2018 HW #8 Due: Friday, November 30 (beginning of class at 12:00p) Requirements:: Include T s diagram for all cycles. Also include p v diagrams for Ch
Validation of MARS-LMR Code for Heat Transfer Models in the DHRS of the PGSFR
Validation of MARS-LMR Code for Heat Transfer Models in the DHRS of the PGSFR Chiwoong CHOI, Taekeong Jeong, JongGan Hong, Sujin Yeom, Jong-Man Kim, Ji-Yeong Jeong, YongBum Lee and Kwiseok Ha Korea Atomic
OF A HELIUM-COOLED MODULE
GA-A21783 DESIGN, FABRICATION, AND TESTING OF A HELIUM-COOLED MODULE FOR THE ITER DIVERTOR RECEIVED by C.B. BAXI, J.P. SMITH, and D. YOUCHISON ' MAR 2 9 1996 0 s-q-i AUGUST 1994 GENEHL ATOMfCS DISCLAIMER
Memorial University of Newfoundland Faculty of Engineering and Applied Science
Memorial University of Newfoundl Faculty of Engineering Applied Science ENGI-7903, Mechanical Equipment, Spring 20 Assignment 2 Vad Talimi Attempt all questions. The assignment may be done individually
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
ESO201A: Thermodynamics
ESO201A: Thermodynamics First Semester 2015-2016 Mid-Semester Examination Instructor: Sameer Khandekar Time: 120 mins Marks: 250 Solve sub-parts of a question serially. Question #1 (60 marks): One kmol
PROBLEM ρ v (kg/m 3 ) ANALYSIS: The critical heat flux can be estimated by Eq with C = 0.
PROBLEM 10.10 KNOWN: Fluids at 1 atm: mercury, ethanol, R-14a. FIND: Critical heat flux; compare with value for water also at 1 atm. ASSUMPTIONS: (1) Steady-state conditions, () Nucleate pool boiling.
Application of Computational Fluid Dynamics to the Flow Mixing and Heat Transfer in Rod Bundle
Tenth International Conference on Computational Fluid Dynamics (ICCFD10), Barcelona, Spain, July 9-13, 2018 ICCFD10-045 Application of Computational Fluid Dynamics to the Flow Mixing and Heat Transfer
Research Article Subchannel Analysis of Wire Wrapped SCWR Assembly
Science and Technology of Nuclear Installations, Article ID 301052, 8 pages http://dx.doi.org/10.1155/2014/301052 Research Article Subchannel Analysis of Wire Wrapped SCWR Assembly Jianqiang Shan, Henan
Numerical Simulation of the MYRRHA reactor: development of the appropriate flow solver Dr. Lilla Koloszár, Philippe Planquart
Numerical Simulation of the MYRRHA reactor: development of the appropriate flow solver Dr. Lilla Koloszár, Philippe Planquart Von Karman Institute, Ch. de Waterloo 72. B-1640, Rhode-St-Genese, Belgium,
THERMAL HYDRAULIC ANALYSIS IN REACTOR VESSEL INTERNALS CONSIDERING IRRADIATION HEAT
THERMAL HYDRAULIC ANALYSIS IN REACTOR VESSEL INTERNALS CONSIDERING IRRADIATION HEAT Sungje Hong, Kunwoo Yi, Jin Huh, Inyoung Im and Eunkee Kim KEPCO Engineering and Construction Company. INC. NSSS Division.
ANALYSIS OF THE OECD MAIN STEAM LINE BREAK BENCHMARK PROBLEM USING THE REFINED CORE THERMAL-HYDRAULIC NODALIZATION FEATURE OF THE MARS0MASTER CODE
ANALYSIS OF THE OECD MAIN STEAM LINE BREAK BENCHMARK PROBLEM USING THE REFINED CORE THERMAL-HYDRAULIC NODALIZATION FEATURE OF THE MARS0MASTER CODE THERMAL HYDRAULICS KEYWORDS: MARS/MASTER code, coupled
Tutorial 1. Where Nu=(hl/k); Reynolds number Re=(Vlρ/µ) and Prandtl number Pr=(µCp/k)
Tutorial 1 1. Explain in detail the mechanism of forced convection. Show by dimensional analysis (Rayleigh method) that data for forced convection may be correlated by an equation of the form Nu = φ (Re,
CONDENSATION HEAT TRANSFER COEFFICIENT CORRELATION BASED ON SLIP RATIO MODEL IN A HORIZONTAL HEAT EXCHANGER
CONDENSATION HEAT TRANSFER COEFFICIENT CORRELATION BASED ON SLIP RATIO MODEL IN A HORIZONTAL HEAT EXCHANGER Seok Kim, Sung Uk Ryu, Seung Tae Lee, Dong-Jin Euh, and Chul-Hwa Song Korea Atomic Energy Research
Presented at The 5th International Symposium Supercritical CO2 Power Cycles March 28-31, 2016, San Antonio, Texas
1 Use of 1-D Finite Enthalpy Method for a High-Temperature Recuperator Made of Polymer Derived Ceramic Composite for a Supercritical Carbon Dioxide Power System Mahmood Mohagheghi, Husam Zawati, Jihua
Experimental Study of Heat Transfer Analysis in Vertical Rod Bundle of Sub Channel with a Hexagonal on Small Modular Reactor
International OPEN ACCESS Journal Of Modern Engineering Research (IJMER) Experimental Study of Heat Transfer Analysis in Vertical Rod Bundle of Sub Channel with a Hexagonal on Small Modular Reactor Syawaluddin
Preventing Thermal Runaways of LENR Reactors. Jacques Ruer sfsnmc
Preventing Thermal Runaways of LENR Reactors Jacques Ruer sfsnmc 1 Temperature activated reactions Several authors report that the LENR power increases with the temperature. 2 Temperature activated reactions
Study on the improved recuperator design used in the direct helium-turbine power conversion cycle of HTR-10
Study on the improved recuperator design used in the direct helium-turbine power conversion cycle of HTR-10 Wu Xinxin 1), Xu Zhao ) 1) Professor, INET, Tsinghua University, Beijing, P.R.China (xinxin@mail.tsinghua.edu.cn)
CFD STUDIES IN THE PREDICTION OF THERMAL STRIPING IN AN LMFBR
CFD STUDIES IN THE PREDICTION OF THERMAL STRIPING IN AN LMFBR K. Velusamy, K. Natesan, P. Selvaraj, P. Chellapandi, S. C. Chetal, T. Sundararajan* and S. Suyambazhahan* Nuclear Engineering Group Indira
Investigation of CTF void fraction prediction by ENTEK BM experiment data
Investigation of CTF void fraction prediction by ENTEK BM experiment data Abstract Hoang Minh Giang 1, Hoang Tan Hung 1, Nguyen Phu Khanh 2 1 Nuclear Safety Center, Institute for Nuclear Science and Technology
ANALYSIS OF METASTABLE REGIMES IN A PARALLEL CHANNEL SINGLE PHASE NATURAL CIRCULATION SYSTEM WITH RELAP5/ MOD 3.2
13 th International Conference on Nuclear Engineering Beijing, China, May 16-20, 2005 ICONE13-50213 ANALYSIS OF METASTABLE REGIMES IN A PARALLEL CHANNEL SINGLE PHASE NATURAL CIRCULATION SYSTEM WITH RELAP5/
EFFECT OF DISTRIBUTION OF VOLUMETRIC HEAT GENERATION ON MODERATOR TEMPERATURE DISTRIBUTION
EFFECT OF DISTRIBUTION OF VOLUMETRIC HEAT GENERATION ON MODERATOR TEMPERATURE DISTRIBUTION A. K. Kansal, P. Suryanarayana, N. K. Maheshwari Reactor Engineering Division, Bhabha Atomic Research Centre,
If there is convective heat transfer from outer surface to fluid maintained at T W.
Heat Transfer 1. What are the different modes of heat transfer? Explain with examples. 2. State Fourier s Law of heat conduction? Write some of their applications. 3. State the effect of variation of temperature
R13 SET - 1 '' ''' '' ' '''' Code No RT21033
SET - 1 II B. Tech I Semester Supplementary Examinations, June - 2015 THERMODYNAMICS (Com. to ME, AE, AME) Time: 3 hours Max. Marks: 70 Note: 1. Question Paper consists of two parts (Part-A and Part-B)
Onset of Flow Instability in a Rectangular Channel Under Transversely Uniform and Non-uniform Heating
Onset of Flow Instability in a Rectangular Channel Under Transversely Uniform and Non-uniform Heating Omar S. Al-Yahia, Taewoo Kim, Daeseong Jo School of Mechanical Engineering, Kyungpook National University
CFD and Thermal Stress Analysis of Helium-Cooled Divertor Concepts
CFD and Thermal Stress Analysis of Helium-Cooled Divertor Concepts Presented by: X.R. Wang Contributors: R. Raffray and S. Malang University of California, San Diego ARIES-TNS Meeting Georgia Institute
NUMERICAL HEAT TRANSFER ENHANCEMENT IN SQUARE DUCT WITH INTERNAL RIB
NUMERICAL HEAT TRANSFER ENHANCEMENT IN SQUARE DUCT WITH INTERNAL RIB University of Technology Department Mechanical engineering Baghdad, Iraq ABSTRACT - This paper presents numerical investigation of heat
Computational and Experimental Studies of Fluid flow and Heat Transfer in a Calandria Based Reactor
Computational and Experimental Studies of Fluid flow and Heat Transfer in a Calandria Based Reactor SD Ravi 1, NKS Rajan 2 and PS Kulkarni 3 1 Dept. of Aerospace Engg., IISc, Bangalore, India. ravi@cgpl.iisc.ernet.in
CFD SIMULATION OF SWIRL FLOW IN HEXAGONAL ROD BUNDLE GEOMETRY BY SPLIT MIXING VANE GRID SPACERS. Mohammad NAZIFIFARD
CFD SIMULATION OF SWIRL FLOW IN HEXAGONAL ROD BUNDLE GEOMETRY BY SPLIT MIXING VANE GRID SPACERS Mohammad NAZIFIFARD Department of Energy Systems Engineering, Energy Research Institute, University of Kashan,
SENSITIVITY ANALYSIS FOR ULOF OF PGSFR
Proceedings of the Asian Conference on Thermal Sciences 2017, 1st ACTS March 26-30, 2017, Jeju Island, Korea SENSITIVITY ANALYSIS FOR ULOF OF PGSFR Sarah Kang 1, Jaeseok Heo 1, ChiWoong Choi 1, Kwi-Seok
DESIGN AND EXPERIMENTAL ANALYSIS OF SHELL AND TUBE HEAT EXCHANGER (U-TUBE)
DESIGN AND EXPERIMENTAL ANALYSIS OF SHELL AND TUBE HEAT EXCHANGER (U-TUBE) Divyesh B. Patel 1, Jayesh R. Parekh 2 Assistant professor, Mechanical Department, SNPIT&RC, Umrakh, Gujarat, India 1 Assistant
SYSTEM ANALYSIS AND ISOTHERMAL SEPARATE EFFECT EXPERIMENTS OF THE ACCIDENT BEHAVIOR IN PWR SPENT FUEL STORAGE POOLS
SYSTEM ANALYSIS AND ISOTHERMAL SEPARATE EFFECT EXPERIMENTS OF THE ACCIDENT BEHAVIOR IN PWR SPENT FUEL STORAGE POOLS H. Chahi 1, W. Kästner 1 and S. Alt 1 1 : University of Applied Sciences Zittau/GörlitzInstitute
Coolant. Circuits Chip
1) A square isothermal chip is of width w=5 mm on a side and is mounted in a subtrate such that its side and back surfaces are well insulated, while the front surface is exposed to the flow of a coolant
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,
Enhancement of Critical Heat Flux in Tight Rod Bundle with Wire Spacer
Journal of Energy an Power Engineering 10 (2016) 109-115 oi: 10.17265/1934-8975/2016.02.005 D DAVID PUBLISHING Enhancement of Critical Heat Flux in Tight Ro Bunle with Wire Spacer Dan Tri Le 1 an Minoru
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
ME 331 Homework Assignment #6
ME 33 Homework Assignment #6 Problem Statement: ater at 30 o C flows through a long.85 cm diameter tube at a mass flow rate of 0.020 kg/s. Find: The mean velocity (u m ), maximum velocity (u MAX ), and
Design constraints Maximum clad temperature, linear power rating
Design constraints Maximum clad temperature, linear power rating K.S. Rajan Professor, School of Chemical & Biotechnology SASTRA University Joint Initiative of IITs and IISc Funded by MHRD Page 1 of 7
UPDATES ON DESIGN AND ANALYSES OF THE PLATE-TYPE DIVERTOR
UPDATES ON DESIGN AND ANALYSES OF THE PLATE-TYPE DIVERTOR X.R. Wang 1, S. Malang 2, M. S. Tillack 1 1 University of California, San Diego, CA 2 Fusion Nuclear Technology Consulting, Germany ARIES-Pathways
THERMAL HYDRAULIC REACTOR CORE CALCULATIONS BASED ON COUPLING THE CFD CODE ANSYS CFX WITH THE 3D NEUTRON KINETIC CORE MODEL DYN3D
THERMAL HYDRAULIC REACTOR CORE CALCULATIONS BASED ON COUPLING THE CFD CODE ANSYS CFX WITH THE 3D NEUTRON KINETIC CORE MODEL DYN3D A. Grahn, S. Kliem, U. Rohde Forschungszentrum Dresden-Rossendorf, Institute
WTS Table of contents. Layout
Table of contents Thermal and hydraulic design of shell and tube heat exchangers... 2 Tube sheet data... 4 Properties of Water and Steam... 6 Properties of Water and Steam... 7 Heat transfer in pipe flow...
ME Thermodynamics I
Homework - Week 01 HW-01 (25 points) Given: 5 Schematic of the solar cell/solar panel Find: 5 Identify the system and the heat/work interactions associated with it. Show the direction of the interactions.
A Study on Hydraulic Resistance of Porous Media Approach for CANDU-6 Moderator Analysis
Proceedings of the Korean Nuclear Society Spring Meeting Kwangju, Korea, May 2002 A Study on Hydraulic Resistance of Porous Media Approach for CANDU-6 Moderator Analysis Churl Yoon, Bo Wook Rhee, and Byung-Joo
ECE309 THERMODYNAMICS & HEAT TRANSFER MIDTERM EXAMINATION. Instructor: R. Culham. Name: Student ID Number:
ECE309 THERMODYNAMICS & HEAT TRANSFER MIDTERM EXAMINATION June 19, 2015 2:30 pm - 4:30 pm Instructor: R. Culham Name: Student ID Number: Instructions 1. This is a 2 hour, closed-book examination. 2. Permitted
I. (20%) Answer the following True (T) or False (F). If false, explain why for full credit.
I. (20%) Answer the following True (T) or False (F). If false, explain why for full credit. Both the Kelvin and Fahrenheit scales are absolute temperature scales. Specific volume, v, is an intensive property,
FUEL PIN BUNDLE EXPERIMENTAL CHARACTERIZATION IN HLM LARGE POOL SYSTEM
FUEL PIN BUNDLE EPERIMENTAL CHARACTERIZATION IN HLM LARGE POOL SYSTEM Martelli D., Barone G., Forgione N., Angelucci M. Department of Civil and Industrial Engineering, University of Pisa Largo Lucio Lazzarino
Neutronic analysis of nanofluids as a coolant in the Bushehr VVER-1000 reactor
NUKLEONIKA 2012;57(3):375 381 ORIGINAL PAPER Neutronic analysis of nanofluids as a coolant in the Bushehr VVER-1000 reactor Ehsan Zarifi, Gholamreza Jahanfarnia, Farzad Veisy Abstract. The main goal of
THERMODYNAMICS, FLUID AND PLANT PROCESSES. The tutorials are drawn from other subjects so the solutions are identified by the appropriate tutorial.
THERMODYNAMICS, FLUID AND PLANT PROCESSES The tutorials are drawn from other subjects so the solutions are identified by the appropriate tutorial. THERMODYNAMICS TUTORIAL 2 THERMODYNAMIC PRINCIPLES SAE
ARIES-AT Blanket and Divertor Design (The Final Stretch)
ARIES-AT Blanket and Divertor Design (The Final Stretch) The ARIES Team Presented by A. René Raffray and Xueren Wang ARIES Project Meeting University of Wisconsin, Madison June 19-21, 2000 Presentation
COMPARISON OF MEASURED AND ANALYTICAL PERFORMANCE OF SHELL-AND-TUBE HEAT EXCHANGERS COOLING AND HEATING SUPERCRITICAL CARBON DIOXIDE
The 4th International Symposium - Supercritical CO Power Cycles September 9-10, 014, Pittsburgh, Pennsylvania COMPARISON OF MEASURED AND ANALYTICAL PERFORMANCE OF SHELL-AND-TUBE HEAT EXCHANGERS COOLING
Chapter 5. Mass and Energy Analysis of Control Volumes. by Asst. Prof. Dr.Woranee Paengjuntuek and Asst. Prof. Dr.Worarattana Pattaraprakorn
Chapter 5 Mass and Energy Analysis of Control Volumes by Asst. Prof. Dr.Woranee Paengjuntuek and Asst. Prof. Dr.Worarattana Pattaraprakorn Reference: Cengel, Yunus A. and Michael A. Boles, Thermodynamics:
Managing Thermal Gradients on a Supercritical Carbon Dioxide Radial Inflow Turbine. David W. Stevens
Managing Thermal Gradients on a Supercritical Carbon Dioxide Radial Inflow Turbine David W. Stevens dstevens@, Turbine Cooling Historically, a consistent engineering challenge Machine durability Traditional
ECE309 INTRODUCTION TO THERMODYNAMICS & HEAT TRANSFER. 20 June 2005
ECE309 INTRODUCTION TO THERMODYNAMICS & HEAT TRANSFER 20 June 2005 Midterm Examination R. Culham & M. Bahrami This is a 90 minute, closed-book examination. You are permitted to use one 8.5 in. 11 in. crib
CHAPTER 5 MASS AND ENERGY ANALYSIS OF CONTROL VOLUMES
Thermodynamics: An Engineering Approach 8th Edition in SI Units Yunus A. Çengel, Michael A. Boles McGraw-Hill, 2015 CHAPTER 5 MASS AND ENERGY ANALYSIS OF CONTROL VOLUMES Lecture slides by Dr. Fawzi Elfghi
Steady-State and Transient Neutronic and Thermal-hydraulic Analysis of ETDR using the FAST code system
Steady-State and Transient Neutronic and Thermal-hydraulic Analysis of ETDR using the FAST code system Sandro Pelloni, Evaldas Bubelis and Paul Coddington Laboratory for Reactor Physics and Systems Behaviour,
Development and Validation of the Wall Boiling Model in ANSYS CFD
Development and Validation of the Wall Boiling Model in ANSYS CFD Th. Frank, C. Lifante, A.D. Burns PBU, Funded CFD Development ANSYS Germany Thomas.Frank@ansys.com E. Krepper, R. Rzehak FZ Dresden-Rossendorf
SINGLE-PHASE AND TWO-PHASE NATURAL CONVECTION IN THE McMASTER NUCLEAR REACTOR
SINGLE-PHASE AND TWO-PHASE NATURAL CONVECTION IN THE McMASTER NUCLEAR REACTOR A. S. Schneider and J. C. Luxat Department of Engineering Physics, McMaster University, 1280 Main St. West, Hamilton, ON, L8S
CHAPTER 7 ENTROPY. Copyright Hany A. Al-Ansary and S. I. Abdel-Khalik (2014) 1
CHAPTER 7 ENTROPY S. I. Abdel-Khalik (2014) 1 ENTROPY The Clausius Inequality The Clausius inequality states that for for all cycles, reversible or irreversible, engines or refrigerators: For internally-reversible
Idaho National Laboratory Reactor Analysis Applications of the Serpent Lattice Physics Code
Idaho National Laboratory Reactor Analysis Applications of the Serpent Lattice Physics Code By Frederick N. Gleicher II, Javier Ortensi, Benjamin Baker, and Mark DeHart Outline Intra-Pin Power and Flux
ENGINEERING OF NUCLEAR REACTORS. Tuesday, October 9 th, 2014, 1:00 2:30 p.m.
.31 ENGINEERING OF NUCLEAR REACTORS Tuesday, October 9 th, 014, 1:00 :30 p.m. OEN BOOK QUIZ 1 (solutions) roblem 1 (50%) Loss o condensate pump transient in a LWR condenser i) Consider the seaater in the