Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor
|
|
- Ralf Hutchinson
- 6 years ago
- Views:
Transcription
1 Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor Yuliang SUN Institute of Nuclear and New Energy Technology, Tsinghua University Beijing , PR China 1 st Workshop on PBMR Coupled Neutronics/Thermal Hydrolics Transient Benchmark The PBMR-400 Core Design, June 2005, NEA/OECD, Paris
2 Contents Introduction of HTR-10 Experiments carried-out during commissioning and operation Safety demonstration experiments Experiments in progress and planned for the future
3 The HTR-10 Test Reactor I I
4 The HTR-10 Test Reactor 700/ /3.43 ST G 2.5MW SG 250/0.5 HTR /3.04 Heating grid 104/6.1 Condenser LP Heater Cooling Tower He Blower Deoxidiser HTR-10 Process Flow
5 HTR-10 Key Design Parameters Reactor thermal power MW 10 Reactor core diameter cm 180 Average core height cm 197 Graphite reflector thickness cm 100 Primary helium pressure MPa 3.0 Helium mass flow rate at full power kg/s 4.3 Average helium temperature at reactor outlet C 700 Average helium temperature at reactor inlet C 250 Main steam flow rate t/h 12.5 Main steam pressure at turbine inlet MPa 3.43 Main steam temperature at turbine inlet C 435 Feed water temperature C 104 Number of control rods in side reflector 10 Number of absorber ball units in side reflector 7 Nuclear fuel UO2 Heavy metal loading per fuel element g 5 Enrichment of fresh fuel element % 17 Number of fuel elements in equilibrium core 27,000 Fuel loading mode Multipass 5
6 HTR-10 Key Milestones Mar. 1992: HTR-10 project formally approved by government Jun. 1995: First concrete of reactor building Apr. 2000: Reactor internals and RPV upper head installed May 2000: Turbine-generator systems installed Dec. 2000: Initial criticality reached Jan. 2003: Full power operation Since Jan. 2003: operation, experiments, safety demonstration
7 HTR-10 In-core and Vessel Thermocouples
8 Experiments carried-out during commissioning Commissioning phases Phase A: Pre-operational tests 36 items Phase B: Fuel loading and low power physics test 44 items Phase C: Power operation test 20 items
9 Commissioning Phases Phase A A1: Subsystem or component function test A2: process system function test A3: preparation before fuel loading Phase B B1: fuel loading and first criticality B2: zero power physics experiment B3: low power experiment Phase C C1: 0~30% rated power experiments C2: 30~100% rated power experiments
10 1.5 Test Items Planed on Phase C Number Title of test item Phase 1 Verifying the function of cavity ventilation C1 2 Verifying the function of secondary circuit isolation valve C1 3 Impurity and radiation measurement C1 4 Measurement of neutron and gamma fields 5 Steam safety valve performance test C1 6 Hot functional test of secondary circuit C1 7 Startup/shutdown circuit and steam generator circuit switch test C1 8 Verifying the capacity of primary circuit cavity cooling C1, C2 9 Measurement of power coefficient C1, C2 10 Verifying the main design parameters C1, C2 11 Verifying the direction of natural circuit after shutdown C1, C2 12 Transient behavior of main helium circulator trip C1 13 Transient behavior of loss of off-site power C1 14 Grid synchronization C1 15 Thermal power calibration C1, C2 16 Power regulation test C2 17 Behavior test of loss of generator load C1, C2 18 Rated power level test C2 19 Verify HTR-10 shutdown means C2 20 Shut down margin measurement C2
11 Commissioning Milestones DATE April, 2000 November 20 th, 2000 December 26 th, 2000 July, 2002 August 15 th, 2002 December 31 st, 2002 January 1 st, 2003 January 26 th, 2003 Pre-operation test First fuel loading EVENT Reach first criticality in air atmosphere Criticality in helium atmosphere and zero power physics tests Low power and hot function tests Power escalation to 3MW Connecting to electric grid Reaching rated power of 10MW
12 Initial Criticality: Calculation Approaches Diffusion Approach with VSOP code system GAM: calculation of fast and epithermal spectrums THERMOS: calculation of thermal spectrum CITATION: finite mesh diffusion code solving the eigenvalue problem in 4 energy groups ZUT-DGL: generation of cross-sections of the resolved and unresolved resonances Nuclear data: ENDF/B-V and JEF-I Monte Carlo Approach with MCNP4A Modeling: three dimensional reactor model, hexahedron lattice of particles Nuclear data: ENDF/B-V Number of cycles: 140 number of source neutrons per cycle: 10,000 number of cycles skipped for collecting K eff : 5
13 Initial Criticality To be calculated and measured: Amount of core loading for the first criticality K eff =1.0 in terms of core loading height or number of loaded balls. Calculation Results: VSOP: cm MCNP: cm
14 Initial Criticality: Comparison between calculation and Experiment The first criticality experiment of HTR-10 was made in December 2000 with the extrapolation approach. In the experiment, first criticality is reached when a total number of 16,890 balls are loaded into the reactor core, of which 9,627 are fuel balls and 7,263 are dummy graphite balls, namely to the ratio of 57:43. This loading corresponds to a loading height of cm. (The core atmosphere is 15ºC air when first criticality is achieved.) Calculation with VSOP and MCNP predicts first criticality at a loading of mixed balls of 16,821 and 16,864 at 27ºC, respectively. The real air temperature is 15ºC while the initial criticality is achieved. After temperature coefficient correction, VSOP predicts a critical loading of 16,759 mixed balls which corresponds to a loading height of cm. The experimental critical loading is 16,890 mixed balls or cm in terms of loading height. The calculation error is very small and satisfying.
15 Full Power Operation Performance Parameter Design Test value value Thermal power (MW) Pressure of primary circuit (MPa) Outlet/inlet helium temperature of SG ( C) 250/ /700 Mass flow rate of water through SG (kg/s) Pressure of steam (MPa) Temperature of main steam ( C) Temperature of feed water ( C) Temperature of PRV ( C) <= Temperature in the helium blower cavity ( C) <= Temperature in the control rod drives cavity ( C) <= Temperature of RPV supports ( C) <=70 50
16 Measurements of Doses At power of 1MW,3MW,5MW,7.5MW and 10MW,measurements of the radiation levels within the plant are performed to verify the adequacy of the shielding design Measured are neutron, gamma dose rates in process rooms and working places, the gamma dose rates around the plant site and beta concentration of discharged gas from stack Results: all dose rates are far below the operation limits
17 Lose of Off-site Power Test The purpose of this test is to confirm The protection system can respond correctly, Required safety functions can be performed, The emergency power can provide electricity within one hour, The reactor and other main equipment such as helium blower are safe and to examine the adequacy of accident management procedures Initial test condition Power of 3MW the three entrance switches of central power supply were turned off
18 Lose of Off-site Power Test Result The protect system was triggered. Safety functions performed The outlet temperature of helium decreased rapidly. The temperature of Hein, PRV and SGPV decreased slowly. The temperatures of shielding, surface cooler outlet and air cooler outlet basically remained unchanged within the experiment time frame. The temperature of PV support increased by about 1 C. power (kw) temperature( C) :52 15:21 15:50 16:19 16:48 17:16 tim e 14:52 15:21 15:50 16:19 16:48 17:16 time Heout Hein RPV SGPV PVSuport Sh ield in g Surface out Air cooler out
19 ATWS Safety Demonstration Tests The purpose of conducting these tests is to demonstrate the inherent safety features as well as to obtain the core and primary cooling system transient data for validation of transient analysis codes. Test items: Two ATWS scenarios were simulated: Loss of primary cooling (helium blower trip) without scram Reactivity insertion by withdrawal of one control rod without scram Test condition: 30% rated power
20 Loss of primary cooling without scram Power(kw) power(kw) speed(rpm) Speed (rpm) Time(s) -200 Transient of power after helium circulator trip
21 Loss of primary cooling without sram Temperature ( ) Time (s) Transients of helium inlet and outlet temperatures after circulator trip
22 Reactivity insertion without scram Power(kw) power(1mk) power(5mk) position(5mk) position(1mk) Time(s) Rod poaition (mm Short term power transient during reactivity insertion without scram
23 Reactivity insertion without scram Power (kw) position(5mk) position(1mk) power(5mk) power(1mk) Tme (s) Position (mm Long term power transient during reactivity insertion without scram
24 Reactivity insertion without scram outlet mixing room Temperature( ) side reflector underside bottom reflector side reflector upside Top bottom reflector reflector metal support Time(s) Core temperature transients during reactivity insertion ATWS simulation
25 Experiments in progress and under planning Safety demonstration tests under rated power of 10MW Correlation between fuel fission product retention modeling and radioactivity intensity in primary helium Installation of He turbine-generator unit to the HTR-10 reactor
26 HTR-10 GT Objective To gain experience of HTRs with gas turbine cycle Steps Joint design with OKBM Installation of gas turbine cycle system Operation of gas turbine cycle
27 HTR-10 GT t=752.1 C Q=4.55 kg/s Reactor N=10 MW t=461.4 C Gas cooler N=0.17 MW Water for cooling t=20 C Q=2.213 kg/s t=501.5 C P= MPa t=750 C G=4.66 kg/s P= MPa Recuperator N =8.63 MW t= 97.8 C P=1.58 MPa Frequency converter Ne=2.02 MW Generator Ne=2.1 MW Turbine n=250 r/s N =5.727 MW Bypass valve HPC N=1.765 MW P= MPa t= 330 C Steam to turbine Steam generator N=3.12 MW t= C P= MPa Precooler N=2.98 MW t=26 C P=1.0 MPa Intercooler N =1.68 MW Water for cooling t=20 C Q=21.1 kg/s Feed water Water for cooling G=35.0 kg/s t=20 C t=23.9 C G=4.76 kg/s P= MPa t= 94.3 C P= MPa LPC N=1.743 MW Flow diagram of HTR-10 GT
28 HTR-10 GT Parameters for HTR-10 with the gas turbine cycle Value POWER CONVERSION UNIT Thermal power transferred to PCU, MW Thermal power transferred to the gas-turbine cycle, MW Thermodynamic efficiency, % Gross efficiency (el.) of the RP gas-turbine part, % Total relative pressure loss, % Total relative helium leaks, % PCU mass, t PCU height, mm Water temperature at the PCU inlet, C REACTOR Temperature at the core inlet/outlet, C Pressure at the inlet/outlet, MPa Helium flow rate, kg/s 330/ /
29 HTR-10 GT HTR-10 with He gas turbine cycle
30 Ideas and Comments about Possible Experiments on HTR-10 are welcome and appreciated!
HTR Research and Development Program in China
HTR Research and Development Program in China Yuanhui XU Institute of Nuclear and New Energy Technology Tsinghua University, Beijing, China 2004 Pacific Basin Nuclear Conference And Technology Exhibit
More informationChapter 4 THE HIGH TEMPERATURE GAS COOLED REACTOR TEST MODULE CORE PHYSICS BENCHMARKS
Chapter 4 THE HIGH TEMPERATURE GAS COOLED REACTOR TEST MODULE CORE PHYSICS BENCHMARKS 4.1 HTR-10 GENERAL INFORMATION China has a substantial programme for the development of advanced reactors that have
More informationThermal Fluid Characteristics for Pebble Bed HTGRs.
Thermal Fluid Characteristics for Pebble Bed HTGRs. Frederik Reitsma IAEA Course on High temperature Gas Cooled Reactor Technology Beijing, China Oct 22-26, 2012 Overview Background Key T/F parameters
More informationHTR-PM Project Status and Test Program
IAEA TWG-GCR-22 HTR-PM Project Status and Test Program SUN Yuliang Deputy Director, INET/ Tsinghua University March 28 April 1, 2011 1 Project organization Government INET R&D, general design, design of
More informationHTGR PROJECTS IN CHINA
HTGR PROJECTS IN CHINA ZONGXIN WU and SUYUAN YU * Institute of Nuclear and New Energy Technology, Tsinghua University Beijing, 100084, China * Corresponding author. E-mail : suyuan@tsinghua.edu.cn Received
More informationREACTOR TECHNOLOGY DEVELOPMENT UNDER THE HTTR PROJECT
REACTOR TECHNOLOGY DEVELOPMENT UNDER THE HTTR PROJECT Takakazu TAKIZUKA Japan Atomic Energy Research Institute The 1st COE-INES International Symposium, INES-1 October 31 November 4, 2004 Keio Plaza Hotel,
More informationCOMPARISON OF FUEL LOADING PATTERN IN HTR-PM
2nd International Topical Meeting on HIGH TEMPERATURE REACTOR TECHNOLOGY Beijing, CHINA, September 22-24, 2004 #Paper C23 COMPARISON OF FUEL LOADING PATTERN IN HTR-PM Fu Li, Xingqing Jing Institute of
More informationHTTR test program towards coupling with the IS process
HTTR test program towards coupling with the IS process T. Iyoku, N.Sakaba, S. Nakagawa, Y.Tachibana, S. Kasahara, and K.Kawasaki Japan Atomic Energy Agency (JAEA) 1 Contents 1. Outline of the HTTR (High
More informationTechnologies of HTR-PM Plant and its economic potential
IAEA Technical Meeting on the Economic Analysis of HTGRs and SMRs 25-28 August 2015, Vienna, Austria Technologies of HTR-PM Plant and its economic potential Prof. Dr. Yujie Dong INET/Tsinghua University
More informationScenarios of Heavy Beyond-Design-Basis Accidents in HTGRs N.G. Kodochigov, Yu.P. Sukharev
Scenarios of Heavy Beyond-Design-Basis Accidents in HTGRs N.G. Kodochigov, Yu.P. Sukharev IAEA Technical Meeting on the Safety of High Temperature Gas Cooled Reactors in the Light of the Fukushima Daiichi
More informationThe design features of the HTR-10
Nuclear Engineering and Design 218 (2002) 25 32 www.elsevier.com/locate/nucengdes The design features of the HTR-10 Zongxin Wu *, Dengcai Lin, Daxin Zhong Institute of Nuclear Energy and Technology, Tsinghua
More informationHTGR Safety Design Fundamental Safety Functions Safety Analysis Decay heat removal Criticality
HTGR Safety Design Fundamental Safety Functions Safety Analysis Decay heat removal Criticality Frederik Reitsma IAEA Course on High temperature Gas Cooled Reactor Technology Oct 22-26, 2012 Content / Overview
More informationJoint ICTP-IAEA Workshop on Nuclear Reaction Data for Advanced Reactor Technologies May 2008
1944-1 Joint ICTP-IAEA Workshop on Nuclear Reaction Data for Advanced Reactor Technologies 19-30 May 2008 Gas-Cooled Reactors International Reactor Physics Experimental Benchmark Analysis. J.M. Kendall
More informationNUCLEAR HEATING REACTOR AND ITS APPLICATION
NUCLEAR HEATING REACTOR AND ITS APPLICATION Zhang Yajun* and Zheng Wenxiang INET, Tsinghua University, Beijing China *yajun@dns.inet.tsinghua.edu.cn Abstract The development of nuclear heating reactor
More informationThermal Response of a High Temperature Reactor during Passive Cooldown under Pressurized and Depressurized Conditions
2nd International Topical Meeting on HIGH TEMPERATURE REACTOR TECHNOLOGY Beijing, CHINA, September 22-24, 2004 #Paper F02 Thermal Response of a High Temperature Reactor during Passive Cooldown under Pressurized
More informationDESIGN OF A PHYSICAL MODEL OF THE PBMR WITH THE AID OF FLOWNET ABSTRACT
NUCLEAR ENGINEERING AND DESIGN VOL.222, PP 203-213 2003 DESIGN OF A PHYSICAL MODEL OF THE PBMR WITH THE AID OF FLOWNET G.P. GREYVENSTEIN and P.G. ROUSSEAU Faculty of Engineering Potchefstroom University
More informationAdvanced High Temperature Reactor Project PBMR relaunch
Advanced High Temperature Reactor Project PBMR relaunch D.R. Nicholls Chief Nuclear Officer, Eskom Africa Utility Week, CTICC May 2017 Potential for Pebble Bed Modular Reactor - PBMR PBMR was based on
More informationSteady State Temperature Distribution Investigation of HTR Core
Journal of Physics: Conference Series PAPER OPEN ACCESS Steady State Temperature Distribution Investigation of HTR Core To cite this article: Sudarmono et al 2018 J. Phys.: Conf. Ser. 962 012040 View the
More informationFuel Management Effects on Inherent Safety of Modular High Temperature Reactor
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 26[7], pp. 647~654 (July 1989). 647 Fuel Management Effects on Inherent Safety of Modular High Temperature Reactor Yukinori HIROSEt, Peng Hong LIEM, Eiichi SUETOMI,
More informationPresent Status and Future Plan of HTTR Project
Present Status and Future Plan of HTTR Project Tatsuo Iyoku *, Toshio Nakazawa, Kozou Kawasaki, Hideyuki Hayashi, and Seigo Fujikawa Department of HTTR Project, Japan Atomic Energy Research Institute Narita-cho,
More informationModule 11 High Temperature Gas Cooled Reactors (HTR)
Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Module 11 High Temperature Gas Cooled Reactors (HTR) 1.3.2017 Development
More informationStatus report 96 - High Temperature Gas Cooled Reactor - Pebble-Bed Module (HTR-PM)
Status report 96 - High Temperature Gas Cooled Reactor - Pebble-Bed Module (HTR-PM) Overview Full name Acronym Reactor type Coolant Moderator Neutron spectrum Thermal capacity Gross Electrical capacity
More informationModule 11 High Temperature Gas Cooled Reactors (HTR)
Prof.Dr. H. Böck Atominstitute of the Austrian Universities Stadionallee 2, 1020 Vienna, Austria boeck@ati.ac.at Module 11 High Temperature Gas Cooled Reactors (HTR) 1.10.2013 Development of Helium Reactor
More informationTOPIC: KNOWLEDGE: K1.01 [2.5/2.5]
KNOWLEDGE: K1.01 [2.5/2.5] P283 The transfer of heat from the reactor fuel pellets to the fuel cladding during normal plant operation is primarily accomplished via heat transfer. A. conduction B. convection
More informationModule 06 Boiling Water Reactors (BWR)
Module 06 Boiling Water Reactors (BWR) 1.3.2017 Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Contents BWR Basics Technical
More informationModule 09 High Temperature Gas Cooled Reactors (HTR)
c Module 09 High Temperature Gas Cooled Reactors (HTR) Prof.Dr. H. Böck Vienna University of Technology /Austria Atominstitute Stadionallee 2, 1020 Vienna, Austria boeck@ati.ac.at Development of Helium
More informationAdvanced Methods for BWR Transient and Stability Analysis. F.Wehle,S.Opel,R.Velten Framatome ANP GmbH P.O. BOX Erlangen Germany
Advanced Methods for BWR Transient and Stability Analysis F.Wehle,S.Opel,R.Velten Framatome ANP GmbH P.O. BOX 3220 91050 Erlangen Germany Advanced Methods for BWR Transient and Stability Analysis > Background
More informationDESIGN, SAFETY FEATURES & PROGRESS OF HTR-PM. Yujie DONG INET, Tsinghua University, China January 24, 2018
DESIGN, SAFETY FEATURES & PROGRESS OF HTR-PM Yujie DONG INET, Tsinghua University, China January 24, 2018 Meet the Presenter Dr. Dong is a Professor in Nuclear Engineering at the Tsinghua University, Beijing,
More informationCore Management and Fuel Handling for Research Reactors
Core Management and Fuel Handling for Research Reactors A. M. Shokr Research Reactor Safety Section Division of Nuclear Installation Safety International Atomic Energy Agency Outline Introduction Safety
More informationModule 06 Boiling Water Reactors (BWR)
Module 06 Boiling Water Reactors (BWR) 1.10.2015 Prof.Dr. Böck Vienna University oftechnology Atominstitute Stadionallee 2 A-1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Contents BWR Basics
More informationIAEA Course on HTR Technology Beijing, October 2012
IAEA Course on HTR Technology Beijing, 22-26.October 2012 Safety and Licensing HTR Module Siemens Design of the 80ies Dr. Gerd Brinkmann AREVA NP GMBH Henry-Dunant-Strasse 50 91058 Erlangen phone +49 9131
More informationGT-MHR OVERVIEW. Presented to IEEE Subcommittee on Qualification
GT-MHR OVERVIEW Presented to IEEE Subcommittee on Qualification Arkal Shenoy, Ph.D Director, Modular Helium Reactors General Atomics, San Diego April 2005 Shenoy@gat.com GT-MHR/LWR COMPARISON Item GT-MHR
More informationEconomic potential of modular reactor nuclear power plants based on the Chinese HTR-PM project
Available online at www.sciencedirect.com Nuclear Engineering and Design 237 (2007) 2265 2274 Economic potential of modular reactor nuclear power plants based on the Chinese HTR-PM project Zuoyi Zhang,
More informationHTR-PM of 2014: toward success of the world first Modular High Temperature Gas-cooled Reactor demonstration plant
HTR-PM of 2014: toward success of the world first Modular High Temperature Gas-cooled Reactor demonstration plant ZHANG/Zuoyi Chief Scientist, HTR-PM project Director, INET of Tsinghua University Vice
More informationPRESENT STATUS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR (HTTR)
PRESENT STATUS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR (HTTR) Shusaku Shiozawa * Department of HTTR Project Japan Atomic Energy Research Institute (JAERI) Japan Abstract It is essentially important
More informationnuclear science and technology
EUROPEAN COMMISSION nuclear science and technology Co-ordination and Synthesis of the European Project of Development of HTR Technology (HTR-C) Contract No: FIKI-CT-2000-20269 (Duration: November 2000
More informationThe HTR/VHTR Project in Framatome ANP
The HTR/VHTR Project in Framatome ANP Framatome ANP Dominique HITTNER HTR-VHTR Project R&D manager Framatome ANP Framatome ANP The reference concept of ANTARES programme: a flexible heat source for heat
More informationNeutronics, Thermal Hydraulics and Safety Parameter Studies of the 3 MW TRIGA Research Reactor at AERE, Savar
Neutronics, Thermal Hydraulics and Safety Parameter Studies of the 3 MW TRIGA Research Reactor at AERE, Savar Md. Quamrul HUDA Energy Institute Atomic Energy Research Establishment Bangladesh Atomic Energy
More informationVVER-440/213 - The reactor core
VVER-440/213 - The reactor core The fuel of the reactor is uranium dioxide (UO2), which is compacted to cylindrical pellets of about 9 height and 7.6 mm diameter. In the centreline of the pellets there
More informationLEU Conversion of the University of Wisconsin Nuclear Reactor
LEU Conversion of the University of Wisconsin Nuclear Reactor Paul Wilson U. Wisconsin-Madison Russian-American Symposium on the Conversion of Research Reactors to Low Enriched Uranium Fuel 8-10 June 2011
More informationThe Next Generation Nuclear Plant (NGNP)
The Next Generation Nuclear Plant (NGNP) Dr. David Petti Laboratory Fellow Director VHTR Technology Development Office High Temperature, Gas-Cooled Reactor Experience HTGR PROTOTYPE PLANTS DEMONSTRATION
More informationA. the temperature of the steam at the turbine exhaust increases. B. additional moisture is removed from the steam entering the turbine.
P77 Overall nuclear power plant thermal efficiency will decrease if... A. the temperature of the steam at the turbine exhaust increases. B. additional moisture is removed from the steam entering the turbine.
More informationUNIT-5 NUCLEAR POWER PLANT. Joining of light nuclei Is not a chain reaction. Cannot be controlled
UNIT-5 NUCLEAR POWER PLANT Introduction Nuclear Energy: Nuclear energy is the energy trapped inside each atom. Heavy atoms are unstable and undergo nuclear reactions. Nuclear reactions are of two types
More informationSafety Issues for High Temperature Gas Reactors. Andrew C. Kadak Professor of the Practice
Safety Issues for High Temperature Gas Reactors Andrew C. Kadak Professor of the Practice Major Questions That Need Good Technical Answers Fuel Performance Normal operational performance Transient performance
More informationDesign, Analysis and Optimization of the Power Conversion System for the Modular Pebble Bed Reactor System. Chunyun Wang
Design, Analysis and Optimization of the Power Conversion System for the Modular Pebble Bed Reactor System By Chunyun Wang B.S.M.E. Tsinghua University, 1991 M.S.N.E. Tsinghua University, 1994 Submitted
More informationREACTOR TECHNOLOGY DEVELOPMENT UNDER THE HTTR PROJECT TAKAKAZU TAKIZUKA
ELSEVIER www.elsevier.com/locate/pnucene Progress in Nuclear Energy; Vol. 47, No. 1-4, pp. 283-291,2005 Available online at www.sciencedirect.com 2005 Elsevier Ltd. All rights reserved s =, E N e E ~)
More informationOverview and Progress of High Temperature Reactor Pebble-bed Module Demonstration Project (HTR-PM)
Overview and Progress of High Temperature Reactor Pebble-bed Module Demonstration Project (HTR-PM) FU Jian JIANG Yingxue CHENG Hongyong CHENG Wei (Huaneng Shandong Shidao Bay Nuclear Power Co., Ltd.) Abstract
More informationPhysics Design of 600 MWth HTR & 5 MWth Nuclear Power Pack. Brahmananda Chakraborty Bhabha Atomic Research Centre, India
Physics Design of 600 MWth HTR & 5 MWth Nuclear Power Pack Brahmananda Chakraborty Bhabha Atomic Research Centre, India Indian High Temperature Reactors Programme Compact High Temperature Reactor (CHTR)
More informationConversion of MNSR (PARR-2) from HEU to LEU Fuel
Conversion of MNSR (PARR-2) from HEU to LEU Fuel Malik Tayyab Mahmood Nuclear Engineering Division Pakistan Institute of Nuclear Science & Technology, Islamabad PAKISTAN Pakistan Institute of Nuclear Science
More informationCore Management and Fuel handling for Research Reactors
Core Management and Fuel handling for Research Reactors W. Kennedy, Research Reactor Safety Section Division of Nuclear Installation Safety Yogyakarta, Indonesia 23/09/2013 Outline Introduction Safety
More informationLecture (3) on. Nuclear Reactors. By Dr. Emad M. Saad. Mechanical Engineering Dept. Faculty of Engineering. Fayoum University
1 Lecture (3) on Nuclear Reactors By Dr. Emad M. Saad Mechanical Engineering Dept. Faculty of Engineering Fayoum University Faculty of Engineering Mechanical Engineering Dept. 2015-2016 2 Nuclear Fission
More informationCANDU Fundamentals. Table of Contents
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
More informationNUCLEAR ENERGY MATERIALS AND REACTORS - Vol. II - Advanced Gas Cooled Reactors - Tim McKeen
ADVANCED GAS COOLED REACTORS Tim McKeen ADI Limited, Fredericton, Canada Keywords: Advanced Gas Cooled Reactors, Reactor Core, Fuel Elements, Control Rods Contents 1. Introduction 1.1. Magnox Reactors
More informationSafety Evaluation of VHTR Cogeneration System
Safety Evaluation of VHTR Cogeneration System Hiroyuki Sato, Tetsuo Nishihara, Xinglong Yan, Kazuhiko Kunitomi Japan Atomic Energy Agency IAEA International Conference on Non-electric Applications of Nuclear
More informationConcept Design and Thermal-hydraulic Analysis for Helium-cooled ADS
Concept Design and Thermal-hydraulic Analysis for Helium-cooled ADS Tianji Peng 1, Zhiwei Zhou 1, Xuanyu Sheng 1 and Long Gu 2 1: Institute of Nuclear and New Energy Technology (INET), Collaborative Innovation
More informationUNIT- III NUCLEAR POWER PLANTS Basics of Nuclear Engineering, Layout and subsystems of Nuclear Power Plants, Working of Nuclear Reactors: Boiling Water Reactor (BWR), Pressurized Water Reactor (PWR), CANada
More informationThe Simulator Development for RDE Reactor
Journal of Physics: Conference Series PAPER OPEN ACCESS The Simulator Development for RDE Reactor To cite this article: Muhammad Subekti et al 2018 J. Phys.: Conf. Ser. 962 012054 View the article online
More information6.1 Introduction. Control 6-1
Control 6-1 Chapter 6 Control 6.1 Introduction Although the bulk of the process design of the Heat Transport System is done prior to the control design, it is helpful to have the key features of the control
More informationBN-1200 Reactor Power Unit Design Development
International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13) BN-1200 Reactor Power Unit Design Development B.A. Vasilyev a, S.F. Shepelev a, M.R.
More informationHeat exchanger equipment of TPPs & NPPs
Heat exchanger equipment of TPPs & NPPs Lecturer: Professor Alexander Korotkikh Department of Atomic and Thermal Power Plants TPPs Thermal power plants NPPs Nuclear power plants Content Steam Generator
More informationA. Kaliatka, S. Rimkevicius, E. Uspuras Lithuanian Energy Institute (LEI) Safety Assessment of Shutdown Reactors at the Ignalina NPP
A. Kaliatka, S. Rimkevicius, E. Uspuras Lithuanian Energy Institute (LEI) Safety Assessment of Shutdown Reactors at the Ignalina NPP Outline Introduction Specific of heat removal from shutdown RBMK-type
More informationAREVA HTR Concept for Near-Term Deployment
AREVA HTR Concept for Near-Term Deployment L. J. Lommers, F. Shahrokhi 1, J. A. Mayer III 2, F. H. Southworth 1 AREVA Inc. 2101 Horn Rapids Road; Richland, WA 99354 USA phone: +1-509-375-8130, lewis.lommers@areva.com
More informationMCNP5 CALCULATIONS COMPARED TO EXPERIMENTAL MEASUREMENTS IN CEA-MINERVE REACTOR
U.P.B. Sci. Bull., Series D, Vol. 74, Iss. 1, 2012 ISSN 1454-2358 MCNP5 CALCULATIONS COMPARED TO EXPERIMENTAL MEASUREMENTS IN CEA-MINERVE REACTOR Mirea MLADIN 1, Daniela MLADIN 21 The paper describes the
More informationABSTRACT. 1. Introduction
Improvements in the Determination of Reactivity Coefficients of PARR-1 Reactor R. Khan 1*, Muhammad Rizwan Ali 1, F. Qayyum 1, T. Stummer 2 1. DNE, Pakistan Institute of Engineering and Applied Sciences
More informationJournal of American Science 2014;10(2) Burn-up credit in criticality safety of PWR spent fuel.
Burn-up credit in criticality safety of PWR spent fuel Rowayda F. Mahmoud 1, Mohamed K.Shaat 2, M. E. Nagy 3, S. A. Agamy 3 and Adel A. Abdelrahman 1 1 Metallurgy Department, Nuclear Research Center, Atomic
More informationCompact, Deployable Reactors for Power and Fuel in Remote Regions
Compact, Deployable Reactors for Power and Fuel in Remote Regions James R. Powell and J. Paul Farrell Radix Corporation, Long Island, New York Presented by Jerry M. Cuttler Dunedin Energy Systems, LLC
More informationPROBABILISTIC SAFETY ASSESSMENT OF JAPANESE SODIUM- COOLED FAST REACTOR IN CONCEPTUAL DESIGN STAGE
PROBABILISTIC SAFETY ASSESSMENT OF JAPANESE SODIUM- COOLED FAST REACT IN CONCEPTUAL DESIGN STAGE Kurisaka K. 1 1 Japan Atomic Energy Agency, Ibaraki, Japan Abstract Probabilistic safety assessment was
More informationPartial Load Characteristics of the Supercritical CO2 Gas Turbine System for the Solar Thermal Power System with the Na-Al- CO2 Heat Exchanger
The 6th International Symposium - Supercritical CO2 Power Cycles March 27 29, 2018, Pittsburgh, Pennsylvania Partial Load Characteristics of the Supercritical CO2 Gas Turbine System for the Solar Thermal
More informationFast and High Temperature Reactors for Improved Thermal Efficiency and Radioactive Waste Management
What s New in Power Reactor Technologies, Cogeneration and the Fuel Cycle Back End? A Side Event in the 58th General Conference, 24 Sept 2014 Fast and High Temperature Reactors for Improved Thermal Efficiency
More informationOperation of the High-Temperature Engineering Test Reactor
Operation of the High-Temperature Engineering Test Reactor Nozomu Fujimoto, Naoki Nojiri, Yukio Tachibana and Toshihiko Mizushima Department of HTTR Japan Atomic Energy Agency (JAEA) -1- Contents 1. 1.
More informationA Research Reactor Simulator for Operators Training and Teaching. Abstract
Organized and hosted by the Canadian Nuclear Society. Vancouver, BC, Canada. 2006 September 10-14 A Research Reactor Simulator for Operators Training and Teaching Ricardo Pinto de Carvalho and José Rubens
More informationTechnology and Prospect of Process Heat Application of HTR(High temperature gas cooled reactor) Applications in Oil Refining Industry
Technology and Prospect of Process Heat Application of HTR(High temperature gas cooled reactor) Applications in Oil Refining Industry Dr. Min Qi, Associate Professor Institute of Nuclear and New Energy
More informationNatural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactor
Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactor Ade Gafar Abdullah 1,2,*, Zaki Su ud 2, Rizal Kurniadi 2, Neny Kurniasih 2, Yanti Yulianti 2,3 1 Electrical
More informationRELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING
Science and Technology Journal of BgNS, Vol. 8, 1, September 2003, ISSN 1310-8727 RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING Pavlin P. Groudev, Rositsa V. Gencheva,
More informationOECD Transient Benchmarks: Preliminary Tinte Results TINTE Preliminary Results
OECD Transient Benchmarks: Preliminary Tinte Results Presentation Overview The use of Tinte at PBMR Tinte code capabilities and overview Preliminary Tinte benchmark results (cases1-6) The use of Tinte
More informationSafety design approach for JSFR toward the realization of GEN-IV SFR
Safety design approach for JSFR toward the realization of GEN-IV SFR Advanced Fast Reactor Cycle System R&D Center Japan Atomic Energy Agency (JAEA) Shigenobu KUBO Contents 1. Introduction 2. Safety design
More informationAbility of New Concept Passive-Safety Reactor "KAMADO" - Safety, Economy and Hydrogen Production -
GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1092 Ability of New Concept Passive-Safety Reactor "KAMADO" - Safety, Economy and Hydrogen Production - Tetsuo MATSUMURA*, Takanori KAMEYAMA, Yasushi
More informationR.A. Chaplin Department of Chemical Engineering, University of New Brunswick, Canada
NUCLEAR REACTOR STEAM GENERATION R.A. Chaplin Department of Chemical Engineering, University of New Brunswick, Canada Keywords: Steam Systems, Steam Generators, Heat Transfer, Water Circulation, Swelling
More informationModule 06 Boiling Water Reactors (BWR) Vienna University of Technology /Austria Atominstitute Stadionallee 2, 1020 Vienna, Austria
Module 06 Boiling Water Reactors (BWR) Prof.Dr. H. Böck Vienna University of Technology /Austria Atominstitute Stadionallee 2, 1020 Vienna, Austria Contents BWR Basics Technical Data Safety Features Reactivity
More informationClosing Remarks and Action Plan PBMR COUPLED NEUTRONICS/THERMAL HYDRAULICS TRANSIENT BENCHMARK THE PBMR-400 CORE DESIGN
Closing Remarks and Action Plan PBMR COUPLED NEUTRONICS/THERMAL HYDRAULICS TRANSIENT BENCHMARK THE PBMR-400 CORE DESIGN OECD Interlaken, PBMR400 Switzerland T5.0 September 14 September 14, 2008 2008 1
More informationDevelopment Status of the Fission Surface Power Technology Demonstration Unit
Development Status of the Fission Surface Power Technology Demonstration Unit Maxwell Briggs, Marc Gibson, and Steven Geng NASA Glenn Research Center 1 Fission Power Systems Presentation Outline The Fission
More informationChannel Type Reactors with Supercritical Water Coolant. Russian Experience.
International Conference Nuclear Energy for New Europe 2003 Portorož, Slovenia, September 8-11, 2003 http://www.drustvo-js.si/port2003 Channel Type Reactors with Supercritical Water Coolant. Russian Experience.
More information1. INTRODUCTION. Corresponding author. Received December 18, 2008 Accepted for Publication April 9, 2009
DEVELOPMENT OF A SIMPLIFIED MODEL FOR ANALYZING THE PERFORMANCE OF KALIMER-600 COUPLED WITH A SUPERCRITICAL CARBON DIOXIDE BRAYTON ENERGY CONVERSION CYCLE SEUNG-HWAN SEONG *, TAE-HO LEE and SEONG-O KIM
More informationSafety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor
FR13 - TECHNICAL SESSION 3.5: Fast reactor safety: post-fukushima lessons and goals for next-generation reactors Paper n. IAEA-CN-199/260 Safety Analysis Results of Representative DEC Accidental Transients
More informationInnovations and Safety Ensuring in WWERs on the Base of Collaboration on the National and International Levels
IAEA INPRO DF9, Vienna 21 November 2014 Innovations and Safety Ensuring in WWERs on the Base of Collaboration on the National and International Levels Grigory Ponomarenko OKB GIDROPRESS Podolsk, Russian
More informationnuclear science and technology
EUROPEAN COMMISSION nuclear science and technology Fast-Acting Boron Injection System (FABIS) Contract No: FIKS-CT-2001-00195 Final report (short version) Work performed as part of the European Atomic
More informationResearch Article A Small-Sized HTGR System Design for Multiple Heat Applications for Developing Countries
International Nuclear Energy Volume 21, Article ID 918567, 18 pages http://dx.doi.org/1.1155/21/918567 Research Article A Small-Sized HTGR System Design for Multiple Heat Applications for Developing Countries
More informationThermal Hydraulic Simulations of the Angra 2 PWR
Thermal Hydraulic Simulations of the Angra 2 PWR Javier González-Mantecón, Antonella Lombardi Costa, Maria Auxiliadora Fortini Veloso, Claubia Pereira, Patrícia Amélia de Lima Reis, Adolfo Romero Hamers,
More informationSmall Modular Nuclear Reactor (SMR) Research and Development (R&D) and Deployment in China
Small Modular Nuclear Reactor (SMR) Research and Development (R&D) and Deployment in China Danrong Song, Biao Quan Nuclear Power Institute of China, Chengdu, China songdr@gmail.com Abstract Developing
More informationDevelopment of a DesignStage PRA for the Xe-100
Development of a DesignStage PRA for the Xe-100 PSA 2017 Pittsburgh, PA, September 24 28, 2017 Alex Huning* Karl Fleming Session: Non-LWR Safety September 27th, 1:30 3:10pm 2017 X Energy, LLC, all rights
More informationSuper Critical CO 2 Gas Turbine Cycle FBRs
The First COE-INES International Symposium at Keio Plaza Hotel, November 3, 2004 Super Critical CO 2 Gas Turbine Cycle FBRs Yasuyoshi Kato Research Laboratory for Nuclear Reactors Tokyo Institute of Technology
More informationTemelin 1000 MW Units Active Testing Stage - Commissioning Experience
International Conference Nuclear Energy for New Europe 2003 Portorož, Slovenia, September 8-11, 2003 http://www.drustvo-js.si/port2003 1 ABSTRACT Temelin 1000 MW Units Active Testing Stage - Commissioning
More informationModule 12 Generation IV Nuclear Power Plants. Atominstitute of the Austrian Universities Stadionallee 2, 1020 Vienna, Austria
Module 12 Generation IV Nuclear Power Plants Prof.Dr. H. Böck Atominstitute of the Austrian Universities Stadionallee 2, 1020 Vienna, Austria boeck@ati.ac.at Generation IV Participants Evolution of Nuclear
More informationHTR reactors within Polish strategy of nuclear energy development Cooperation with Japan
HTR reactors within Polish strategy of nuclear energy development Cooperation with Japan Taiju SHIBATA Senior Principal Researcher Group Leader, International Joint Research Group HTGR Hydrogen and Heat
More informationCONTENTS CO-GENERATING WATER-DESALINATING FACILITY POWERED BY SVBR-75/100 NUCLEAR REACTOR DESIGN ORGANIZATIONS INVOLVED IN THE PROJECT
CONTENTS CO-generating water-desalinating facility powered by SVBR-75/100 NUCLEAR 1 REACTOR Design organizations involved in the project 1 Layout of co-generating nuclear-powered water desalinating facility
More informationACR Safety Systems Safety Support Systems Safety Assessment
ACR Safety Systems Safety Support Systems Safety Assessment By Massimo Bonechi, Safety & Licensing Manager ACR Development Project Presented to US Nuclear Regulatory Commission Office of Nuclear Reactor
More informationModule 05 WWER/ VVER (Russian designed Pressurized Water Reactors)
Module 05 WWER/ VVER (Russian designed Pressurized Water Reactors) 1.3.2016 Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at
More informationAnalysis of HTR-10 First Criticality with Monte Carlo Code Tripoli-4.3
2nd International Topical Meeting on HIGH TEMPERATURE REACTOR TECHNOLOGY Beijing, CHINA, September 22-24, 2004 #Paper C11 Analysis of HTR-10 First Criticality with Monte Carlo Code Tripoli-4.3 Hong CHANG
More informationSafety Design Requirements and design concepts for SFR
Safety Design Requirements and design concepts for SFR Reflection of lessons learned from the Fukushima Dai-ichi accident Advanced Nuclear System Research & Development Directorate Japan Atomic Energy
More informationDesign and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Core Configurations
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 32[9], pp. 834-845 (September 1995). Design and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Core Configurations
More information