The Potential for Graphite Oxidation in Dry Wall ICF Chambers

Size: px
Start display at page:

Download "The Potential for Graphite Oxidation in Dry Wall ICF Chambers"

Transcription

1 The Potential for Graphite Oxidation in Dry Wall ICF Chambers El S. Mogahed, I. N Sviatoslavsky, J. G. Murphy, H. Y. Khater, M. H. Anderson, and G. L. Kulcinski Fusion Technology Institute University of Wisconsin May 31, 2001

2 Outline of Presentation Why are we concerned? Fundamental assumptions and model of worst accident Time dependent temperature of graphite first wall after accident (without oxidation) Time dependent availability of oxygen in the chamber Potential first wall carbon oxidation rates Time dependent temperature of graphite first wall after accident (with oxidation) Calculated erosion rates (with oxidation) Methods to mitigate carbon oxidation Conclusions and recommendations

3 Why are we concerned about possible oxidation of dry wall graphite structures? Woven graphite structures run at 800 to 1500 C in a 0.5 torr Xe atmosphere in SOMBRERO. The introduction of air as a result of an accident could release radioactive tritium and result in structural degradation if oxidation were to take place. The fundamental question is Will the graphite first wall temperature be high enough, when the oxygen reaches it, to cause significant oxidation? 3

4 Worst Accident Scenario A 1 m 2 hole is introduced in the 1.3 m thick outer building wall while the reactor is operating. The air rushes through the 1 m 2 opening and through the laser ports in the outer chamber, and finally through the beam ports in the inner chamber. The reactor and all mechanical equipment shuts down and the Li 2 O coolant drains from the chamber and upper manifold by gravity. The first wall radiates heat to the rest of the chamber and it also loses heat out the back of the blanket and shield. The ultimate heat sink is the wall of the outer building 4

5 1 m 2 Hole OUTER BUILDING SOMBRERO BUILDING INNER BUILDING VOLUME = 2.6x10 4 m 3 OUTER BUILDING Atmospheric Pressure 50m 60 x m 2 15m 10 m CHAMBER 60 x m 2 VOLUME = 9x10 5 m 3 VOLUME = 1.73x10 3 m 3 Atmospheric Pressure 1 m 2 Hole OUTER BUILDING 9.54 m 2 Hole INNER BUILDING 2.28 m 2 Hole CHAMBER Model of SOMBRERO Building as it is used in the MELCOR code 5

6 Air Pressure History in SOMBRERO Chamber After the Accident -The pressure in the chamber increases with time and reaches one atmosphere in about 4000 s (1.1 Hr) (MELCOR Calculations). Pressure (10 4 Pa) - The oxygen partial pressure follows same build-up pattern if there is no oxidation Pressure In the Chamber Time History Atmospheric Pressure Air Pressure Maximum O Partial Pressure 2 (No Oxidation ) Time (10 3 s) 6

7 SOMBRERO Finite Element Thermal Model Assumptions: Axi-symmetric model. Transient solution,that allows conduction, convection, and radiation heat transfer that varies with time. Back of chamber radiates to the inner building wall that is initially at 350 C. The back of the inner building wall radiates to the outer building wall at 20 C. 7

8 Transient Thermal Analysis of SOMBRERO First Wall With No Oxidation Li 2 O, T= 550 C Assumptions: Cooling due to Li 2 O drainage during the first 130 s after the accident Only conduction and radiation heat transfer are in force after 130 s. Min. Midplane Max. SOMBRERO CHAMBER 8

9 Transient Thermal Analysis of SOMBRERO First Wall With No Oxidation Assumptions: Temperature History of SOMBRERO FW Maximum Initial Temp. = 1450 C Convection Cooling Conduction Radiation Max. ( C) Midplane (C ) Min. (C ) Conduction Radiation Cooling due to Li 2 O drainage during the first 130 s after the accident Only conduction and radiation heat transfer are in force after 130 s. SOMBRERO FW Temperature ( C) Min. Midplane Max. SOMBRERO CHAMBER Temperature Equalization Time (s) 9

10 Transient Thermal Analysis of SOMBRERO FW/Chamber With No Oxidation Cooling due to Li 2 O Drainage During the First 130 s After the Accident. Only Conduction and Radiation Heat Transfer is in effect after 130 s. - The Back of the Chamber Remains Hot ( 780 C) During the First 130 s. Maximum First Wall Temperature ( C) Max. Mid. Min. Cooling Due to Li 2 O Drainage 130 s Min. Midplane Max. SOMBRERO CHAMBER Time (s) 1 Day 10 1 Month

11 Maximum First Wall Temperature (No Oxidation) ( C) The Oxygen Partial Pressure is Low While the First Wall Temperature is High FW Temperature Oxygen Partial Pressure Time (s) Air Pressure Pressure (Pa) 11

12 Different Graphites Have Very Different Oxidation Rates Experimental data is available for temperatures over 800 C Month FW Max. Temperature Union Carbide (CIT) Union Carbide graphite oxidation rate is two times higher than GraphNOL N3M. Time (s) - Pfizer pyrolytic graphite oxidation rate is much less than both Union Carbide Graphite and GraphNOL N3M One Week GraphNOL N3M Extrapolation Pfizer One Day Oxidation Rate (g/cm Pyrolytic Experimental Data s) First Wall Maximum Temperature ( C) 12

13 The Choice of the Wrong Graphite for an IFE First Wall Could Cause Problems if The Building is Breached FW graphite temperature rises as a result of exothermic reaction with oxygen. Maximum First Wall Temperature ( C) Oxidation of GraphNOL graphite is very rapid, it will oxidize the 1 cm FW in about 3.5 Hr. Oxidation of pyrolytic graphite is very low and raises FW temperature about 10 C at most Temp. of Pfizer Pyrolytic with Oxidation Temp. of GraphNOL N3M with Oxidation Temp. of FW without Oxidation Oxidation thickness (Pfizer Pyrolytic) Oxidation thickness GraphNOL N3M Oxidation thickness( cm) Time (s)

14 Methods of Mitigating Air Ingress Accident Effect Shutters on chamber beam-ports triggered by pressure imbalance Reduce inner building wall temperature SiC coating everywhere except first wall Thin ( 30 µ) SiC undercoating on the first wall Flood chamber with CO 2 Plug hole in building! 14

15 Conclusions The survival of the graphite first wall, in the event of an unlikely ingress of air, is critically dependent on the type of graphite used. If the wrong type of graphite is chosen, the first wall could be completely oxidized in a few hours If pyrolytic, or H-451 graphite is chosen, there should be little oxidation (even after a few months) An experiment is needed to measure oxidation rates at reduced oxygen levels and at 200<T< 800 C. 15

LOSS OF COOLANT ACCIDENT AND LOSS OF FLOW ACCIDENT ANALYSIS OF THE ARIES-AT DESIGN

LOSS OF COOLANT ACCIDENT AND LOSS OF FLOW ACCIDENT ANALYSIS OF THE ARIES-AT DESIGN LOSS OF COOLANT ACCIDENT AND LOSS OF FLOW ACCIDENT ANALYSIS OF THE ARIES-AT DESIGN E. A. Mogahed, L. El-Guebaly, A. Abdou, P. Wilson, D. Henderson and the ARIES Team Fusion Technology Institute University

More information

FUSION TECHNOLOGY INSTITUTE

FUSION TECHNOLOGY INSTITUTE FUSION TECHNOLOGY INSTITUTE SOMBRERO A Solid Breeder Moving Bed KrF Laser Driven IFE Power Reactor W I S C O N S I N I.N. Sviatoslavsky, M.E. Sawan, R.R. Peterson, G.L. Kulcinski, J.J. MacFarlane, L.J.

More information

GA A23414 GENERAL ATOMICS/SAN DIEGO STATE UNIVERSITY TEAM INVESTIGATES IFE TARGET HEATING DURING INJECTION

GA A23414 GENERAL ATOMICS/SAN DIEGO STATE UNIVERSITY TEAM INVESTIGATES IFE TARGET HEATING DURING INJECTION GA A23414 GENERAL ATOMICS/SAN DIEGO STATE UNIVERSITY TEAM INVESTIGATES IFE TARGET HEATING DURING INJECTION by PROJECT STAFF JULY 2000 DISCLAIMER This report was prepared as an account of work sponsored

More information

S. Reyes. Lawrence Livermore National Laboratory. ARIES Meeting April 22-23, 2002 U. Wisconsin, Madison

S. Reyes. Lawrence Livermore National Laboratory. ARIES Meeting April 22-23, 2002 U. Wisconsin, Madison S. Reyes Lawrence Livermore National Laboratory ARIES Meeting April 22-23, 2002 U. Wisconsin, Madison Work performed under the auspices of the U. S. Department of Energy by University of California Lawrence

More information

ARIES-AT BLANKET AND DIVERTOR

ARIES-AT BLANKET AND DIVERTOR ARIES-AT BLANKET AND DIVERTOR A. R. Raffray, M.S.Tillack, X. Wang L. El-Guebaly, I. Sviatoslavsky S. Malang University of California, San Diego University of Wisconsin Forschungszentrum Karlsruhe EBU-II,

More information

High performance blanket for ARIES-AT power plant

High performance blanket for ARIES-AT power plant Fusion Engineering and Design 58 59 (2001) 549 553 www.elsevier.com/locate/fusengdes High performance blanket for ARIES-AT power plant A.R. Raffray a, *, L. El-Guebaly b, S. Gordeev c, S. Malang c, E.

More information

FUSION TECHNOLOGY INSTITUTE

FUSION TECHNOLOGY INSTITUTE FUSION TECHNOLOGY INSTITUTE Apollo-L2, An Advanced Fuel Tokamak Reactor Utilizing Direct Conversion W I S C O N S I N G.A. Emmert, G.L. Kulcinski, J.P. Blanchard, L.A. El-Guebaly, H.Y. Khater, J.F. Santarius,

More information

Reflections on Fusion Chamber Technology and SiC/SiC Applications Mohamed Abdou UCLA

Reflections on Fusion Chamber Technology and SiC/SiC Applications Mohamed Abdou UCLA Reflections on Fusion Chamber Technology and SiC/SiC Applications Mohamed Abdou UCLA Presented at CREST Conference, Kyoto, Japan, May 21, 2002 The Region Immediately Surrounding the Plasma Divertor / First

More information

Naturally Safe HTGR in the response to the Fukushima Daiichi NPP accident

Naturally Safe HTGR in the response to the Fukushima Daiichi NPP accident IAEA Technical Meeting on on Re evaluation of Maximum Operating Temperatures and Accident Conditions for High Temperature Reactor Fuel and Structural Materials, 10 12 July 2012, Vienna, Austria Naturally

More information

Laser-Driven IFE Power Plants Initial Results from ARIES-IFE Study

Laser-Driven IFE Power Plants Initial Results from ARIES-IFE Study Laser-Driven IFE Power Plants Initial Results from ARIES-IFE Study Farrokh Najmabadi For the ARIES Team Third US-Japan Workshop on Laser-Driven Inertial Fusion Energy Jan.25-27, 2001 Lawrence Livermore

More information

Status Report and Documentation of DCLL Design

Status Report and Documentation of DCLL Design Status Report and Documentation of DCLL Design He primary and secondary loops footprint at TCWS DCLL design evolution DCLL, DEMO inboard routing assessment Documentation of DCLL design Clement Wong, Dick

More information

Summary of Major Features of ARIES- ST and ARIES-AT Blanket Designs

Summary of Major Features of ARIES- ST and ARIES-AT Blanket Designs Summary of Major Features of ARIES- ST and ARIES-AT Blanket Designs Presented by A. René Raffray University of California, San Diego with the Contribution of the ARIES Team and S. Malang APEX Meeting,

More information

Progress and critical issues for IFE blanket and chamber research

Progress and critical issues for IFE blanket and chamber research Fusion Engineering and Design 51 52 (2000) 1095 1101 www.elsevier.com/locate/fusengdes Progress and critical issues for IFE blanket and chamber research B. Grant Logan a, *, Wayne R. Meier a, Ralph W.

More information

The design features of the HTR-10

The 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 information

CERAMIC BREEDER BLANKET FOR ARIES-CS

CERAMIC BREEDER BLANKET FOR ARIES-CS CERAMIC BREEDER BLANKET FOR ARIES-CS A.R. Raffray 1, S. Malang 2, L. El-Guebaly 3, X. Wang 4, and the ARIES Team 1 Mechanical and Aerospace Engineering Department and Center for Energy Research, 460 EBU-II,

More information

Maintenance Concept for Modular Blankets in Compact Stellarator Power Plants

Maintenance Concept for Modular Blankets in Compact Stellarator Power Plants Maintenance Concept for Modular Blankets in Compact Stellarator Power Plants Siegfried Malang With contributions of Laila A. EL-Guebaly Xueren Wang ARIES Meeting UCSD, San Diego, January 8-10, 2003 Overview

More information

Thermal-Hydraulic Study of ARIES-CS Ceramic Breeder Blanket Coupled with a Brayton Cycle

Thermal-Hydraulic Study of ARIES-CS Ceramic Breeder Blanket Coupled with a Brayton Cycle Thermal-Hydraulic Study of ARIES-CS Ceramic Breeder Blanket Coupled with a Brayton Cycle Presented by A. R. Raffray With contributions from L. El-Guebaly, S. Malang, X. Wang and the ARIES team ARIES Meeting

More information

Overview of ARIES ACT-1 Study

Overview of ARIES ACT-1 Study Overview of ARIES ACT-1 Study Farrokh Najmabadi Professor of Electrical & Computer Engineering Director, Center for Energy Research UC San Diego and the ARIES Team Japan-US Workshop on Fusion Power Plants

More information

IFE TARGET FABRICATION AND INJECTION

IFE TARGET FABRICATION AND INJECTION IFE TARGET FABRICATION AND INJECTION 1951 µm 1690 µm 1500 µm 0.8 µm of CH + 5% Au CH (DT) 64 Fuel DT Vapor Fast Gas Valve Gas Reservoir Target Loader Gun Barrel Transport from Target Factory Rotating Shield

More information

Maintenance method for blanket and final optics of dry wall fast ignition ICF power plant

Maintenance method for blanket and final optics of dry wall fast ignition ICF power plant 1 Maintenance method for blanket and final optics of dry wall fast ignition ICF power plant Japan-US Workshop on Fusion Power Plants and Related Advanced Technologies with participation of EU February

More information

FUSION TECHNOLOGY INSTITUTE

FUSION TECHNOLOGY INSTITUTE FUSION TECHNOLOGY INSTITUTE An Improved First Stability Advanced Fuel Tokamak, Apollo-L3 W I S C O N S I N G.A. Emmert, G.L. Kulcinski, J.P. Blanchard, L.A. El-Guebaly, H.Y. Khater, C.W. Maynard, E.A.

More information

Task III presentations

Task III presentations Task III presentations Nelson (ORNL) Nygren (SNL) Rognlien (LLNL) Youssef (UCLA) Smolentsev (UCLA) ALL Sn-CLiFF Configuration and Status (15 min.) Divertor Integration Status (20 min) FW and Diveror Temperature

More information

Operation of DIII-D National Fusion Facility and Related Research Cooperative Agreement DE-FC02-04ER54698 (GA Project 30200)

Operation of DIII-D National Fusion Facility and Related Research Cooperative Agreement DE-FC02-04ER54698 (GA Project 30200) May 10, 2010 Dr. Mark Foster U. S. Department of Energy Office of Science General Atomics Site/Bldg. 7 Rm. 119 3550 General Atomics Ct. San Diego, CA 92121 Reference: Operation of DIII-D National Fusion

More information

Corrosion of Structural Materials in Molten Fluoride and Chloride Salts

Corrosion of Structural Materials in Molten Fluoride and Chloride Salts Corrosion of Structural Materials in Molten Fluoride and Chloride Salts Stephen Raiman, James Keiser Oak Ridge National Laboratory USA 1st IAEA workshop on Challenges for Coolants in the Fast Spectrum:

More information

Safety Design of HTGR by JAEA in the light of the Fukushima Daiichi accident

Safety Design of HTGR by JAEA in the light of the Fukushima Daiichi accident Technical Meeting on the Safety of High Temperature Gas Cooled Reactors in the Light of the Fukushima Daiichi Accident, 8-11 April 2014, IAEA Head quarters, Vienna, Austria Safety Design of HTGR by JAEA

More information

Strategy of WCCB-TBM Testing in ITER

Strategy of WCCB-TBM Testing in ITER Third IAEA DEMO Programme Workshop 13 May 2015 Hefei, China Strategy of WCCB-TBM Testing in ITER JAEA Blanket technology group Hisashi Tanigawa, T. Hirose, Y. Kawamura and M. Enoeda Outline DEMO blankets

More information

ACTIVATION, DECAY HEAT, AND WASTE DISPOSAL ANALYSES FOR THE ARIES-AT POWER PLANT

ACTIVATION, DECAY HEAT, AND WASTE DISPOSAL ANALYSES FOR THE ARIES-AT POWER PLANT ACTIVATION, DECAY HEAT, AND WASTE DISPOSAL ANALYSES FOR THE ARIES-AT POWER PLANT D. Henderson, L. El-Guebaly, P. Wilson, A. Abdou, and the ARIES Team University of Wisconsin-Madison, Fusion Technology

More information

Concept and preliminary experiment on ILE, Osaka. protection of final optics in wet-wall laser fusion reactor

Concept and preliminary experiment on ILE, Osaka. protection of final optics in wet-wall laser fusion reactor Concept and preliminary experiment on protection of final optics in wet-wall laser fusion reactor T. Norimatsu, K. Nagai, T. Yamanaka and Y. Izawa Presented at Japan-US Workshop on Power Plant Studies

More information

ARIES-ACT-DCLL NWL Distribution and Revised Radial Build

ARIES-ACT-DCLL NWL Distribution and Revised Radial Build ARIES-ACT-DCLL NWL Distribution and Revised Radial Build L. El-Guebaly and A. Jaber Fusion Technology Institute University of Wisconsin-Madison http://fti.neep.wisc.edu/uwneutronicscenterofexcellence Contributors:

More information

Design Characteristics of the Cold Neutron Source in HANARO Process System and Architecture Design

Design Characteristics of the Cold Neutron Source in HANARO Process System and Architecture Design Design Characteristics of the Cold Neutron Source in HANARO Process System and Architecture Design Sang Ik Wu, Young Ki Kim, Kye Hong Lee, Hark Rho Kim Korea Atomic Energy Research Institute (KAERI), Daejeon,

More information

CHAPTER 7 THERMAL ANALYSIS

CHAPTER 7 THERMAL ANALYSIS 102 CHAPTER 7 THERMAL ANALYSIS 7.1 INTRODUCTION While applying brake, the major part of the heat generated is dissipated through the brake drum while the rest of it goes into the brake shoe. The thermal

More information

Coolant Cleanup and Process Requirements

Coolant Cleanup and Process Requirements Coolant Cleanup and Process Requirements I. Sviatoslavsky and L. El-Guebaly Fusion Technology Institute University of Wisconsin - Madison With input from R. Moir (LLNL) ARIES Project Meeting Princeton

More information

HAPL Blanket Strategy

HAPL Blanket Strategy HAPL Blanket Strategy A. René Raffray UCSD With contributions from M. Sawan and I. Sviatoslavsky UW HAPL Meeting Georgia Institute of Technology Atlanta, GA HAPL meeting, G.Tech. 1 Outline Background Strategy

More information

ELFR The European Lead Fast Reactor DESIGN, SAFETY APPROACH AND SAFETY CHARACTERISTICS. Alessandro Alemberti

ELFR The European Lead Fast Reactor DESIGN, SAFETY APPROACH AND SAFETY CHARACTERISTICS. Alessandro Alemberti ELFR The European Lead Fast Reactor DESIGN, SAFETY APPROACH AND SAFETY CHARACTERISTICS Alessandro Alemberti Alessandro.Alemberti@ann.ansaldo.it TECHNICAL MEETING ON IMPACT OF FUKUSHIMA EVENT ON CURRENT

More information

FUSION TECHNOLOGY INSTITUTE

FUSION TECHNOLOGY INSTITUTE FUSION TECHNOLOGY INSTITUTE LIBRA-LiTE: A 1000 MWe Reactor W I S C O N S I N G.L. Kulcinski, R.L. Engelstad, E.G. Lovell, J.J. MacFarlane, E.A. Mogahed, G.A. Moses, R.R. Peterson, S. Rutledge, M.E. Sawan,

More information

TOPIC: KNOWLEDGE: K1.01 [2.5/2.5]

TOPIC: 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 information

ARIES: Fusion Power Core and Power Cycle Engineering

ARIES: Fusion Power Core and Power Cycle Engineering ARIES: Fusion Power Core and Power Cycle Engineering The ARIES Team Presented by A. René Raffray ARIES Peer Review Meeting University of California, San Diego ARIES: Fusion Power Core and Power Cycle Engineering/ARR

More information

Sheet) Graphite Sheet

Sheet) Graphite Sheet PGS(Pyrolytic Graphite Sheet) Graphite Sheet Panasonic Electronic Device Co.,Ltd Panasonic Electronic Device Hokkaido Co.,Ltd PGS Graphite Sheet PGS (Pyrolytic Highly Oriented Graphite Sheet) is made of

More information

RESIDUAL STRESS AND DISTORTION ANALYSIS IN LASER BEAM WELDING PROCESSES

RESIDUAL STRESS AND DISTORTION ANALYSIS IN LASER BEAM WELDING PROCESSES Ind. J. Sci. Res. and Tech. 015 3():0-5/Kanthi ISSN:-31-96 (Online) RESIDUAL STRESS AND DISTORTION ANALYSIS IN LASER BEAM WELDING PROCESSES Rajesh Goud Kanthi School of Mechanical Engineering,CEST, Fiji

More information

Scenarios 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 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 information

Research Article Normal and Accidental Scenarios Analyses with MELCOR and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept

Research Article Normal and Accidental Scenarios Analyses with MELCOR and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept Science and Technology of Nuclear Installations Volume 2015, Article ID 865829, 9 pages http://dx.doi.org/10.1155/2015/865829 Research Article Normal and Accidental Scenarios Analyses with MELCOR 1.8.2

More information

AP1000 European 19. Probabilistic Risk Assessment Design Control Document

AP1000 European 19. Probabilistic Risk Assessment Design Control Document 19.39 In-Vessel Retention of Molten Core Debris 19.39.1 Introduction In-vessel retention of molten core debris through water cooling of the external surface of the reactor vessel is a severe accident management

More information

Safety Methodology Implementation in the Conceptual Design Phase of a Fusion Reactor

Safety Methodology Implementation in the Conceptual Design Phase of a Fusion Reactor Safety Methodology Implementation in the Conceptual Design Phase of a Fusion Reactor Lina Rodríguez-Rodrigo Joëlle Uzan-Elbez 1 Safety Methodology Implementation in the Conceptual Design Phase of a Fusion

More information

Development of Low Activation Structural Materials

Development of Low Activation Structural Materials Materials Challenge for Clean Nuclear Fusion Energy Development of Low Activation Structural Materials T. Muroga National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan Symposium on Materials

More information

Lecture (3) on. Nuclear Reactors. By Dr. Emad M. Saad. Mechanical Engineering Dept. Faculty of Engineering. Fayoum University

Lecture (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 information

Severe Accident Progression Without Operator Action

Severe Accident Progression Without Operator Action DAA Technical Assessment Review of the Moderator Subcooling Requirements Model Severe Accident Progression Without Operator Action Facility: Darlington Classification: October 2015 Executive summary After

More information

IFE Reactor Chamber Design and Blanket Issues Some Comments

IFE Reactor Chamber Design and Blanket Issues Some Comments IFE Reactor Chamber Design and Blanket Issues Some Comments Presented by A. René Raffray UCSD TITAN Workshop on System Integration Modeling on MFE and IFE University of California, Los Angeles Los Angeles,

More information

Thin coatings using combined magnetron sputtering and ion implantation technique

Thin coatings using combined magnetron sputtering and ion implantation technique Thin coatings using combined magnetron sputtering and ion implantation technique E. Grigore, C. Ruset National Institute for Lasers, Plasma and Radiation Physics PO Box MG-36, Magurele, Romania Outline

More information

Technology Challenges in the Development of Inertial Fusion Energy

Technology Challenges in the Development of Inertial Fusion Energy Technology Challenges in the Development of Inertial Fusion Energy Seventh IAEA Technical Meeting on Physics and Technology of Inertial Fusion Energy Chamber and Targets March 19, 2015 Thomas Anklam LLNL-PRES-668442

More information

Computational Fluid Dynamics Analysis for loads definition during accidental scenarios

Computational Fluid Dynamics Analysis for loads definition during accidental scenarios IDM UID TF35GS VERSION CREATED ON / VERSION / STATUS 15 Sep 2016 / 1.4 / Approved EXTERNAL REFERENCE / VERSION Technical Specifications (In-Cash Procurement) Computational Fluid Dynamics Analysis for loads

More information

The Effects of Temperature Cycling on Cheddar Cheese

The Effects of Temperature Cycling on Cheddar Cheese The Effects of Temperature Cycling on Cheddar Cheese By D. Reinemann J. Mitchell A. Hade J. Leggett 7 May, 1998 This work was supported by a grant from the Energy Center of Wisconsin. 1 Introduction The

More information

Laser Fusion Chamber Design

Laser Fusion Chamber Design Laser Fusion Chamber Design J. Blanchard 1, A.R. Raffray 2 and the Team 1 University of Wisconsin, Madison, WI 53706 2 University of California, San Diego, La Jolla, CA 90093 TOFE Albuquerque, NM 1 IFE

More information

Kinetics of Silicon Oxidation in a Rapid Thermal Processor

Kinetics of Silicon Oxidation in a Rapid Thermal Processor Kinetics of Silicon Oxidation in a Rapid Thermal Processor Asad M. Haider, Ph.D. Texas Instruments Dallas, Texas USA Presentation at the National Center of Physics International Spring Week 2010 Islamabad

More information

PRESENT STATUS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR (HTTR)

PRESENT 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 information

Preliminary Lessons Learned from the Fukushima Daiichi Accident for Advanced Nuclear Power Plant Technology Development

Preliminary Lessons Learned from the Fukushima Daiichi Accident for Advanced Nuclear Power Plant Technology Development Preliminary Lessons Learned from the Fukushima Daiichi Accident for Advanced Nuclear Power Plant Technology Development A. Introduction The IAEA Report on Reactor and Spent Fuel Safety in the Light of

More information

Application for Permission to Extend the Operating Period and Application for Approval of Construction Plans of Unit 3 at Mihama Nuclear Power Station

Application for Permission to Extend the Operating Period and Application for Approval of Construction Plans of Unit 3 at Mihama Nuclear Power Station November 26, 2015 The Kansai Electric Power Co., Inc. Application for Permission to Extend the Operating Period and Application for Approval of Construction Plans of Unit 3 at Mihama Nuclear Power Station

More information

DEMO Concept Development and Assessment of Relevant Technologies

DEMO Concept Development and Assessment of Relevant Technologies 1 FIP/3-4Rb DEMO Concept Development and Assessment of Relevant Technologies Y. Sakamoto, K. Tobita, H. Utoh, N. Asakura, Y. Someya, K. Hoshino, M. Nakamura, S. Tokunaga and the DEMO Design Team Japan

More information

COMPARISON OF STEADY-STATE AND PULSED-PLASMA TOKAMAK POWER PLANTS

COMPARISON OF STEADY-STATE AND PULSED-PLASMA TOKAMAK POWER PLANTS COMPARISON OF STEADY-STATE AND PULSED-PLASMA TOKAMAK POWER PLANTS F. Najmabadi, University of California, San Diego and The ARIES Team IEA Workshop on Technological Aspects of Steady State Devices Max-Planck-Institut

More information

Nuclear Research Institute. Nuclear Research Institute CONTENTS

Nuclear Research Institute. Nuclear Research Institute CONTENTS CONTENTS Nuclear Organizational Structure in VN Brief Introduction to the DNRR: - History of the reactor - Operation and Utilization of the reactor Decommissioning Planning 1 THE ORGANIZATION CHART OF

More information

Space Environmental Hazards For Space and Launch Systems

Space Environmental Hazards For Space and Launch Systems Space Environmental Hazards For Space and Launch Systems Dr. Joseph E. Mazur The Aerospace Corporation Presentation to the 2012 FAA Commercial Space Transportation Conference Physical Sciences Laboratory

More information

Safety 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 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 information

Status of the FRM-II Project at Garching. Hans-Jürgen Didier, Gunter Wierheim Siemens AG, Power Generation (KWU), D Erlangen

Status of the FRM-II Project at Garching. Hans-Jürgen Didier, Gunter Wierheim Siemens AG, Power Generation (KWU), D Erlangen IGORR7 7 th Meeting of the International Group on Research Reactors October 26-29, 1999 Bariloche, Argentina Status of the FRM-II Project at Garching Hans-Jürgen Didier, Gunter Wierheim Siemens AG, Power

More information

Decay Heat Removal studies in Gas Cooled Fast Reactor during accidental condition - demonstrator ALLEGRO

Decay Heat Removal studies in Gas Cooled Fast Reactor during accidental condition - demonstrator ALLEGRO Decay Heat Removal studies in Gas Cooled Fast Reactor during accidental condition - demonstrator ALLEGRO S. Bebjak 1, B. Kvizda 1, G. Mayer 2, P. Vacha 3 1 VUJE, Trnava, Slovak Republic 2 MTA-EK, Budapest,

More information

Demonstration Test Program for Long term Dry Storage of PWR Spent Fuel

Demonstration Test Program for Long term Dry Storage of PWR Spent Fuel Demonstration Test Program for Long term Dry Storage of PWR Spent Fuel 16 November 2010 K.Shigemune, The Kansai Electric Power Co., Inc. The Japan Atomic Power Company Kyushu Electric Power Co., Inc. Mitsubishi

More information

Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor

Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor Yuliang SUN Institute of Nuclear and New Energy Technology, Tsinghua University Beijing 100084, PR China 1 st Workshop on PBMR Coupled

More information

COLD NEUTRON SOURCE AT CMRR

COLD NEUTRON SOURCE AT CMRR COLD NEUTRON SOURCE AT CMRR Hu Chunming Shen Wende, Dai Junlong, Liu Xiankun ( 1 ) Vadim Kouzminov, Victor Mityukhlyaev / 2 /, Anatoli Serebrov, Arcady Zakharov ( 3 ) ABSTRACT As an effective means to

More information

Fast and High Temperature Reactors for Improved Thermal Efficiency and Radioactive Waste Management

Fast 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 information

Operation of the helium cooled DEMO fusion power plant and related safety aspects

Operation of the helium cooled DEMO fusion power plant and related safety aspects Operation of the helium cooled DEMO fusion power plant and related safety aspects 1 st IAEA workshop on Challenges for coolants in fast spectrum system: Chemistry and materials, Vienna, July, 5-7 2017

More information

Thermal Analysis of a Mobile Hot Cell Cask

Thermal Analysis of a Mobile Hot Cell Cask Thermal Analysis of a Mobile Hot Cell Cask Jay Kalinani 1, Arvind Kumar R 2 P.G. Student, Department of Physics, MS Ramaiah University of Applied Sciences, Bangalore, Karnataka, India 1 U.G. Student, Department

More information

Development of a Large-scale Monoblock Divertor Mock-up for Fusion Experimental Reactors in JAERI

Development of a Large-scale Monoblock Divertor Mock-up for Fusion Experimental Reactors in JAERI Development of a Large-scale Monoblock Divertor Mock-up for Fusion Experimental Reactors in JAERI S. Suzuki, K. Sato, K. Ezato, M. Taniguchi, M. Akiba Japan Atomic Energy Research Institute, Ibaraki-ken,

More information

THE ARIES-I TOKAMAK REACTOR STUDY

THE ARIES-I TOKAMAK REACTOR STUDY THE ARIES-I TOKAMAK REACTOR STUDY Farrokh Najmabadi, Robert W. Conn, and The ARIES Team 16th SOFT London, September 3-7, 1990 ARIES Is a Community-Wide Study U. W. UCLA ANL U. IL. FEDC ORNL RPI ARIES GA

More information

This is a repository copy of A Two-Dimensional, Finite-Difference Model of the Oxidation of a Uranium Carbide Fuel Pellet.

This is a repository copy of A Two-Dimensional, Finite-Difference Model of the Oxidation of a Uranium Carbide Fuel Pellet. This is a repository copy of A Two-Dimensional, Finite-Difference Model of the Oxidation of a Uranium Carbide Fuel Pellet. White Rose Research Online URL for this paper: http://eprints.whiterose.ac.uk/101208/

More information

Lithium and Liquid Metal Studies at PPPL

Lithium and Liquid Metal Studies at PPPL Lithium and Liquid Metal Studies at PPPL LTX R. Maingi, M. Jaworski, R. Kaita, R. Majeski, J. Menard, M. Ono PPPL High Power Devices NSTX-U Surface Analysis EAST Test stands IAEA TM on Divertor Concepts

More information

Overview of Fission Safety for Laser ICF Fission Energy. Per F. Peterson, Edward Blandford, and Christhian Galvez

Overview of Fission Safety for Laser ICF Fission Energy. Per F. Peterson, Edward Blandford, and Christhian Galvez Overview of Fission Safety for Laser ICF Fission Energy Per F. Peterson, Edward Blandford, and Christhian Galvez University of California at Berkeley, Berkeley, California,peterson@nuc.berkeley.edu This

More information

Fluid Thrust Chamber Design. Kevin Cavender, Den Donahou, Connor McBride, Mario Reillo, Marshall Crenshaw

Fluid Thrust Chamber Design. Kevin Cavender, Den Donahou, Connor McBride, Mario Reillo, Marshall Crenshaw Kevin Cavender, Den Donahou, Connor McBride, Mario Reillo, Marshall Crenshaw 1 1 2 2 Kevin Cavender, Den Donahou, Connor McBride, Mario Reillo, Marshall Crenshaw 3 3 Fuel Selection Fuel Mixture Ratio Cost

More information

Advanced power core system for the ARIES-AT power plant

Advanced power core system for the ARIES-AT power plant Fusion Engineering and Design 82 (2007) 217 236 Advanced power core system for the ARIES-AT power plant A.R. Raffray a,, L. El-Guebaly b, S. Malang c, I. Sviatoslavsky b, M.S. Tillack a,x.wang a, The ARIES

More information

Technical Report ARAEW-TR A THERMAL PERFORMANCE STUDY OF THE 155MM XM297 ACTIVELY COOLED BARREL

Technical Report ARAEW-TR A THERMAL PERFORMANCE STUDY OF THE 155MM XM297 ACTIVELY COOLED BARREL AD Technical Report ARAEW-TR-622 A THERMAL PERFORMANCE STUDY OF THE 155MM XM297 ACTIVELY COOLED BARREL MARK D WITHERELL NOVEMBER 26 ARMAMENT RESEARCH, DEVELOPMENT AND ENGINEERING CENTER Armaments Engineering

More information

Safety Requirements for HTR Process Heat Applications

Safety Requirements for HTR Process Heat Applications HTR 2014 Conference, Oct. 2014, Weihai, China Norbert Kohtz (TÜV Rheinland), Michael A. Fütterer (Joint Research Centre): Safety Requirements for HTR Process Heat Applications Outline 1. Introduction 2.

More information

Clearance Issues and Radwaste Volume for ARIES-AT

Clearance Issues and Radwaste Volume for ARIES-AT Clearance Issues and Radwaste Volume for ARIES-AT L. El-Guebaly, D. Henderson, A. Abdou, P. Wilson, E. Mogahed Fusion Technology Institute - Madison With inputs from M. Sawan (UW), D. Petti (INEEL) Web

More information

Safety Issues in TBM Program

Safety Issues in TBM Program Idaho National Engineering and Environmental Laboratory Safety Issues in TBM Program Brad Merrill, Dave Petti 1 Hans-Werner Bartels 2 1 Fusion Safety Program 2 ITER IT, Safety Group APEX/TBM Meeting, UCLA,

More information

Loss of Coolant Flow Accident Analysis for the Fluoride Salt Cooled High Temperature Reactor

Loss of Coolant Flow Accident Analysis for the Fluoride Salt Cooled High Temperature Reactor Loss of Coolant Flow Accident Analysis for the Fluoride Salt Cooled High Temperature Reactor Yao FU, Yang YANG, Yang ZOU, Qiang SUN and Jie ZHANG Key Laboratory of Nuclear Radiation and Nuclear Energy

More information

THERMAL AND STRESS ANALYSIS OF HYPER TARGET SYSTEM *

THERMAL AND STRESS ANALYSIS OF HYPER TARGET SYSTEM * THERMAL AND STRESS ANALYSIS OF HYPER TARGET SYSTEM * T.Y. Song, N.I. Tak, W.S. Park Korea Atomic Energy Research Institute P.O. Box 105 Yusung, Taejon, 305-600, Republic of Korea J.S. Cho, Y.S. Lee Department

More information

Environmental and Economical Assessment of Various Fusion Reactors by the Calculation of CO 2 Emission Amounts

Environmental and Economical Assessment of Various Fusion Reactors by the Calculation of CO 2 Emission Amounts Environmental and Economical Assessment of Various Fusion Reactors by the Calculation of CO 2 Emission Amounts Satoshi UEMURA, Kozo YAMAZAKI, Hideki ARIMOTO, Tetsutarou OISHI and Tatsuo SHOJI Department

More information

Design of Solid Breeder Test Blanket Modules in JAERI

Design of Solid Breeder Test Blanket Modules in JAERI Design of Solid Breeder Test Blanket Modules in JAERI Presented by: S. Suzuki, Blanket Engineering Lab., Japan Atomic Energy Research Institute, JAERI Contents 1. Outline of blanket development in JAERI

More information

Reactor Technology: Materials, Fuel and Safety. Dr. Tony Williams

Reactor Technology: Materials, Fuel and Safety. Dr. Tony Williams Reactor Technology: Materials, Fuel and Safety Dr. Tony Williams Course Structure Unit 1: Reactor materials Unit 2. Reactor types Unit 3: Health physics, Dosimetry Unit 4: Reactor safety Unit 5: Nuclear

More information

ITER CONVERTIBLE BLANKET EVALUATION

ITER CONVERTIBLE BLANKET EVALUATION GA-C21941 ITER CONVERTIBLE BLANKET EVALUATION by C.P.C. WONG and E. CHENG* \ *TSI Research Inc. Prepared Under Subcontract No. ITER-GA-4002 Raytheon Engineers and Constructors Ebasco Division DISTRIBUTION

More information

In Cheol BANG, Ji Hyun Kim School of Energy Engineering Ulsan National Institute of Science and Technology Republic of Korea

In Cheol BANG, Ji Hyun Kim School of Energy Engineering Ulsan National Institute of Science and Technology Republic of Korea Rod-Type Quench Performance of Nanofluids Towards Developments of Advanced PWR Nanofluids-Engineered Safety Features In Cheol BANG, Ji Hyun Kim School of Energy Engineering Ulsan National Institute of

More information

Tritium building. Port cell. Hot cell. Ancillary systems TBM

Tritium building. Port cell. Hot cell. Ancillary systems TBM Hot cell Port cell Tritium building US DCLL TBM Ancillary systems TBM Reported by: Arnie Lumsdaine M. Abdou, M. Dagher, A. Ying, N.B. Morley, S. Smolentsev, S. Sharafat, A. Aoyama, M. Youssef UCLA C. Wong,

More information

ACR Safety Systems Safety Support Systems Safety Assessment

ACR 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 information

Molten Salts: Common Nuclear and Concentrated-Solar- Thermal-Power Technologies

Molten Salts: Common Nuclear and Concentrated-Solar- Thermal-Power Technologies Molten Salts: Common Nuclear and Concentrated-Solar- Thermal-Power Technologies Charles Forsberg Department of Nuclear Science and Engineering (NSE) Massachusetts Institute of Technology 77 Massachusetts

More information

ARIES Issues and R&D Needs Technology Readiness for Fusion Energy

ARIES Issues and R&D Needs Technology Readiness for Fusion Energy page 1 of 16 ARIES Issues and R&D Needs Technology Readiness for Fusion Energy M. S. Tillack ARIES Project Meeting 28 May 2008 page 2 of 16 The goal of the TWG s is to translate advisory committee recommendations

More information

Full Submersion Criticality Accident Mitigation in the Carbide LEU-NTR

Full Submersion Criticality Accident Mitigation in the Carbide LEU-NTR Full Submersion Criticality Accident Mitigation in the Carbide LEU-NTR Paolo Venneri, Yonghee Kim Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon, Korea, 35-71 paolovenneri@kaist.ac.kr,

More information

Update on APEX Safety Issues

Update on APEX Safety Issues Update on APEX Safety Issues Presented by: K. A. McCarthy With contributions by: L. C. Cadwallader, K. A. McCarthy, B. J. Merrill, and R. L. Moore Fusion Safety Program APEX Meeting May 12-14, 1999 PPPL

More information

Design Windows and Roadmaps for Laser Fusion Reactors

Design Windows and Roadmaps for Laser Fusion Reactors US/Japan Workshop on Power Plant Studies April 6-7, 2002, San Diego Design Windows and Roadmaps for Laser Fusion Reactors Yasuji Kozaki Institute of Laser Engineering, Osaka University Outline 1. Design

More information

Improved Design of a Helium- Cooled Divertor Target Plate

Improved Design of a Helium- Cooled Divertor Target Plate Improved Design of a Helium- Cooled Divertor Target Plate By X.R. Wang, S. Malang, R. Raffray and the ARIES Team ARIES-Pathway Meeting University of Wisconsin, Madison April 23-24, 2009 Critical Design

More information

Controlled management of a severe accident

Controlled management of a severe accident July 2015 Considerations concerning the strategy of corium retention in the reactor vessel Foreword Third-generation nuclear reactors are characterised by consideration during design of core meltdown accidents.

More information

Current Status of Research and Development on System Integration Technology for Connection between HTGR and Hydrogen Production System at JAEA

Current Status of Research and Development on System Integration Technology for Connection between HTGR and Hydrogen Production System at JAEA Current Status of Research and Development on System Integration Technology for Connection between HTGR and Hydrogen Production System at JAEA Hirofumi Ohashi, Yoshitomo Inaba, Tetsuo Nishihara, Tetsuaki

More information

Ensuring Spent Fuel Pool Safety

Ensuring Spent Fuel Pool Safety Ensuring Spent Fuel Pool Safety Michael Weber Deputy Executive Director for Operations U.S. Nuclear Regulatory Commission American Nuclear Society Meeting June 28, 2011 1 Insights from Fukushima Nuclear

More information

Influence of Alloy Microstructure on Oxide Growth in HCM12A in Supercritical Water

Influence of Alloy Microstructure on Oxide Growth in HCM12A in Supercritical Water Mater. Res. Soc. Symp. Proc. Vol. 1125 2009 Materials Research Society 1125-R06-05 Influence of Alloy Microstructure on Oxide Growth in HCM12A in Supercritical Water Jeremy Bischoff 1, Arthur T. Motta

More information