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Accelerator driven subcritical reactors / H. Nifenecker, O. Meplan and S. David

Nifenecker, H.

Bristol ; Philadelphia : IOP Publishing, 2003

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Lo trovi qui: Univ. della Basilicata

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Éléments de sûreté nucléaire - Les réacteurs à eau sous pression / / Jean Couturier

Couturier Jean

[Place of publication not identified] : , : EDP SCIENCES, , 2021

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Lo trovi qui: Univ. Federico II

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Proceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors : [Newport Beach, California, August 1-5, 1999] / / sponsored by the Minerals, Metals and Materials Society, American Nuclear Society, National Association of Corrosion Engineers International ; edited by Steve Bruemmer, Peter Ford, Gary Was

Warrendale, Pennsylvania : , : Minerals, Metals & Materials Society, , [1999]

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The Risks of Nuclear Energy Technology : Safety Concepts of Light Water Reactors / / by Günter Kessler, Anke Veser, Franz-Hermann Schlüter, Wolfgang Raskob, Claudia Landman, Jürgen Päsler-Sauer

Kessler Günter

Berlin, Heidelberg : , : Springer Berlin Heidelberg : , : Imprint : Springer, , 2014

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Safety of sodium-cooled fast reactors : particle-bed-related phenomena in severe accidents / / Songbai Cheng, Ruicong Xu

Cheng Songbai

Gateway East, Singapore : , : Springer, , [2021]

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Safety of sodium-cooled fast reactors : particle-bed-related phenomena in severe accidents / / Songbai Cheng, Ruicong Xu

Cheng Songbai

Gateway East, Singapore : , : Springer, , [2021]

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Sodium fast reactors with closed fuel cycle / / Baldev Raj, P. Chellapandi, P.R. Vasudeva Rao

Raj Baldev <1947, >

Boca Raton, Florida : , : CRC Press, , [2015]

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Sodium fast reactors with closed fuel cycle / / Baldev Raj, P. Chellapandi, P.R. Vasudeva Rao

Raj Baldev <1947, >

Boca Raton, Florida : , : CRC Press, , [2015]

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Sodium fast reactors with closed fuel cycle / / Baldev Raj, P. Chellapandi, P.R. Vasudeva Rao

Raj Baldev <1947, >

Boca Raton, Florida : , : CRC Press, , [2015]

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Structural materials for heavy liquid metal cooled fast reactors : proceedings of a technical meeting / / International Atomic Energy Agency

Vienna, Austria : , : International Atomic Energy Agency, , [2021]

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Pubbl/distr/stampa	Warrendale, Pennsylvania : , : Minerals, Metals & Materials Society, , [1999]
Descrizione fisica	1 online resource (1252 p.)
Disciplina	621.48 621.4834
Altri autori (Persone)	BruemmerS. M FordF. P (F. Peter) WasGary S <1953-> (Gary Steven)
Soggetto topico	Nuclear power plants - Corrosion Water cooled reactors - Corrosion Nuclear power plants - Materials - Effect of radiation on
ISBN	1-118-78777-3 1-118-78761-7 1-118-78795-1
Formato	Materiale a stampa
Livello bibliografico	Monografia
Lingua di pubblicazione	eng
Nota di contenuto	Cover; Title Page; Copyright Page; FOREWORD; TABLE OF CONTENTS; PWR Primary-1: Mechanisms; An Overview of Internal Oxidation as a Possible Explanation of Intergranular Stress Corrosion Cracking of Alloy 600 in PWRs; Methodology to Understand the Mechanisms of PWSCC; Hydrogen Effects on PWR SCC Mechanisms in Monocrystalline and Polycrystalline Alloy 600; Insights into Environmental Degradation Mechanisms from Analytical Transmission Electron Microscopy of SCC Cracks; Measurement of the Fundamental Parameters for the Film-Rupture/Oxidation Mechanism-The Effect of Chromium Comparison of Hydrogen Effects on Alloy 600 and 690Comments on a Proposed Mechanism of Internal Oxidation for Alloy 600 as Applied to Low Potential SCC; Internal Oxidation and Embrittlement of Alloy 600; PWR Primary-2: Chemistry and Failure Analysis; The Effect of Primary Coolant Zinc Additions on the SCC Behaviour of Alloy 600 and 690; PWSCC of Alloy 600: A Parametric Study of Surface Film Effects; Modelling of Stress Corrosion Crack Initiation on Alloy 600 in Primary Water of PWRs; Effect of Water Chemistry on Environmentally Assisted Cracking in Alloy in Simulated PWR Environments Unique Primary Side Initiated Degradation in the Vicinity of the Upper Roll Transition in Once Through Steam Generators from Oconee Unit 1PWR Primary-3: Hydrogen Effects & Microstructure; On the Possibility of Forming Ordered Ni2Cr in Alloy 690; Hydrogen Embrittlement of PH 13-08 Mo Stainless Steel in PWR Environment Effect of Microstructure; The Effect of Special Grain Boundaries on IGSCC of Ni-16Cr-9Fe-xC; Fracture Behavior of Nickel-Based Alloys in Water; Hydrogen-Assisted Failure of Alloys X-750 and 625 under Slow Strain-Rate Conditions An Experimental Study of the Hydrogen Embrittlement of Alloy 718 in PWR Primary WaterA Study of Corrosion Mechanisms and Kinetics of Alloy 718 in PWR Primary Water; Stress Corrosion Crack Propagation Rate of Alloy 600 in the Primary Water of PWR: Influence of a Cold Worked Layer; PWR Primary-4: Crack Growth & Creep; Stress Corrosion Crack Growth Rate Measurements in Alloys 600 and 182 in Primary Water Loops Under Constant Load; Initial Results on the Stress Corrosion Cracking Monitoring of Alloy 600 in High Temperature Water Using Acoustic Emission Stress Corrosion Crack Propagation Rates in Reactor Vessel Head Penetrations in Alloy 600Stress Corrosion Life Assessment of Alloy 600 PWR Components; Influence of Chromium Content and Microstructure on Creep and PWSCC Resistance of Nickel Base Alloys; A Simplified Model for SCC Initiation Susceptibility in Alloy 600, with the Influence of Cold Work Layer and Strength Characteristics; Creep of Nickel Base Alloys in High Temperature Water; An Investigation of Alloy 182 Stress Corrosion Cracking in Simulated PWR Environment; BWR-1: Cracking Response Characteristics of Crack Propagation Through SCC under BWR Conditions in Stainless Steels Stabilized with Titanium or Niobium
Altri titoli varianti	Environmental degradation of materials in nuclear power systems--water reactors
Record Nr.	UNISA-996213874203316

Autore	Kessler Günter
Edizione	[1st ed. 2014.]
Pubbl/distr/stampa	Berlin, Heidelberg : , : Springer Berlin Heidelberg : , : Imprint : Springer, , 2014
Descrizione fisica	1 online resource (365 p.)
Disciplina	621.4834
Collana	Science Policy Reports
Soggetto topico	Nuclear energy Radiation protection Radiation—Safety measures Energy systems System safety Nuclear Energy Effects of Radiation/Radiation Protection Energy Systems Security Science and Technology
ISBN	3-642-55116-5
Formato	Materiale a stampa
Livello bibliografico	Monografia
Lingua di pubblicazione	eng
Nota di contenuto	From the Contents: Inherent safety characteristics of Pressurized Water- and Boiling Water Reactors (PWRs and BWR s) -- Safety design concepts of present and future PWRs and BWR s -- Radiation protection and emission of radioactivity during normal operation of PWRs and BWRs -- Accident and risk analysis as well as additional severe accident measures to be initiated after core cooling accidents -- Safety design concepts against external hazards, e. g. earthquakes, chemical explosions, flooding.
Record Nr.	UNINA-9910299619903321

Autore	Raj Baldev <1947, >
Edizione	[1st ed.]
Pubbl/distr/stampa	Boca Raton, Florida : , : CRC Press, , [2015]
Descrizione fisica	1 online resource (886 p.)
Disciplina	621.4834
Soggetto topico	Sodium cooled reactors Fast reactors
ISBN	0-429-16779-2
Formato	Materiale a stampa
Livello bibliografico	Monografia
Lingua di pubblicazione	eng
Nota di contenuto	Front Cover; Contents; Foreword; Preface; Authors; Chapter 1: Nuclear Fission and Breeding; Chapter 2: Fast Spectrum Reactor vis-à-vis Pressurized Water Reactors; Chapter 3: Description of a Fast Spectrum Reactor; Chapter 4: Unique Worthiness of SFR; Chapter 5: Design Objectives for the Efficient Use of Natural Uranium and Plutonium; Chapter 6: Prospect of Various Types of FSRs; Chapter 7: Choice of Materials and Their Performance; Chapter 8: System and Components; Chapter 9: Design Basis; Chapter 10: Design Validations; Chapter 11: Design Analysis and Methods Chapter 12: Safety Principles and PhilosophyChapter 13: Safety Criteria and Basis; Chapter 14: Event Analysis; Chapter 15: Severe Accident Analysis; Chapter 16: Sodium Safety; Chapter 17: Computer Codes and Validation; Chapter 18: Test Facilities and Programs; Chapter 19: Safety Experiments in Reactors; Chapter 20: Severe Accident Management; Chapter 21: Safety Analysis of PFBR : A Case Study; Chapter 22: Specific Aspects of Civil Structures and Construction; Chapter 23: Manufacturing and Erection of Mechanical Components; Chapter 24: Illustrations from International SFRs Chapter 25: Commissioning Issues : Various Phases and ExperiencesChapter 26: SFR Program in Countries; Chapter 27: Feedback from Operating Experiences; Chapter 28: Innovative Reactor Concepts for Future SFRs; Chapter 29: Fuel Cycle for SFRs; Chapter 30: Decommissioning Aspects; Chapter 31: Material Science and Metallurgy; Chapter 32: Chemical Sensors for Sodium Coolant Circuits; Chapter 33: Robotics, Automation, and Sensors; Chapter 34: Operator Training Simulators for Fast Breeder Reactors; Chapter 35: Economics of SFRs with a Closed Fuel Cycle; Back Cover
Record Nr.	UNINA-9910817048403321

Edizione	[1st ed.]
Pubbl/distr/stampa	Vienna, Austria : , : International Atomic Energy Agency, , [2021]
Descrizione fisica	1 online resource (228 pages)
Disciplina	621.4834
Soggetto topico	Liquid metal cooled reactors Nuclear reactors - Materials
ISBN	92-0-128721-6 1-5231-4991-4
Formato	Materiale a stampa
Livello bibliografico	Monografia
Lingua di pubblicazione	eng
Nota di contenuto	Intro -- 1. Introduction -- 1.1. Background -- 1.2. Objective -- 1.3. Scope -- 1.4. Structure -- 2. Summary of Meeting Sessions -- 2.1. Session I: HLM Compatibility with Structural Materials: Phenomena, Modelling and Operational Experience -- 2.2. Session II: Corrosion Mitigation Measures: Coating, New Structural Materials, Environmental Conditioning -- 2.3. Session III: Qualification Programmes of Structural Materials for HLM Fast Reactors -- 3. Summary of Group Discussions -- 3.1. Group Discussion i: Outstanding Research Challenges -- 3.2. Group Discusion II: New Materials and Coating Techniques -- 3.3. Group discussion III: Industrialization -- 4. Conclusions -- ABBREVIATIONS -- PAPERS PRESENTED AT THE MEETING -- SESSION I: HLM COMPATIBILITY WITH STRUCTURAL MATERIALS: PHENOMENA, MODELLING AND OPERATIONAL EXPERIENCE -- KINETICS AND MECHANISM OF CRACK INITIATION OF LIQUID METAL EMBRITTLEMENT -- 1. INTRODUCTION -- 2. EXPERIMENTAL -- 2.1. Materials -- 2.2. Specimens -- 2.3. Test technique -- 2.4. Environment -- 2.5. Post-test evaluation -- 3. RESULTS I - SENSITIVITY TO LME/EAC CRACKING IN HLM -- 3.1. T91 & -- liquid LBE -- 3.2. T91 & -- liquid Pb -- 4. RESULTS II - CONDITIONS FOR LME/EAC CRACKING IN HLM: INITIATION -- 4.1. T91 & -- liquid LBE -- 4.2. T91 & -- liquid Pb -- 4.3. Initiation of LME/EAC of T91 in HLM - summary -- 4.4. 15-15Ti in liquid LBE -- 4.5. 15-15Ti in liquid Pb -- 5. RESULTS III - CONDITIONS FOR LME/EAC CRACKING IN HLM: KINETICS -- 5.1. Fracture resistance of T91 in liquid LBE -- 5.2. Crack Growth Rate of T91 in HLM -- 6. Discussion -- 7. Conclusions -- RATEN ICN STATUS ON MECHANICAL PROPERTIES INVESTIGATION OF 316L GENERATION iv CANDIDATE MATERIAL -- 1. INTRODUCTION -- 2. EXPERIMENTAL METHOD -- 3. RESULTS AND DISCUSSION -- 4. PLANNED WORKS -- 5. CONCLUSIONS. SIMULATIONS OF SOME STRUCTURAL MATERIALS BEHAVIOR UNDER NEUTRON IRRADIATION -- 1. INTRODUCTION -- 2. REACTOR CONFIGURATION AND COMPUTATIONAL TOOLS -- 3. RESULTS AND DISCUSSION -- 3.1. Molybdenum Alloys -- 3.2. Vanadium Alloys -- 4. CONCLUSIONS -- VACANCY TYPE DEFECTS BEHAVIOR IN MATERIALS FORESEEN FOR LIQUID METAL COOLED FAST REACTORS -- RESEARCH OF CORROSION BEHAVIOR OF STEAM GENERATOR TUBES FOR LEAD-COOLED POWER UNIT -- 1. INTRODUCTION -- 2. CORROSION RESISTANCE IN LIQUID LEAD -- 3. CORROSION RESISTANCE IN WATER AND SUPERHEATED STEAM -- 4. INTERGRANULAR CORROSION RESISTANCE -- 5. THE MAIN RESULTS AND CONCLUSIONS -- TENSILE TESTING OF SUB-SIZED T91 STEEL SPECIMENS IN LIQUID LEAD -- 1. INTRODUCTION -- 2. THE LILLA FACILITY AND TEST SECTIONS -- 2.1. The description of the LILLA facility -- 2.2. The description of the test sections -- 3. SSRT TESTS OF T91 IN ARGON AND LIQUID LEAD -- 3.1. Material and test conditions -- 3.2. Tests in argon and liquid lead -- 4. CONCLUSIONS -- SESSION II: CORROSION MITIGATION MEASURES: COATING, NEW STRUCTURAL MATERIALS, ENVIRONMENTAL CONDITIONING -- CORROSION AND MECHANICAL TESTING OF A LOW ALLOYED ALUMINA FORMING AUSTENITE FOR LIQUID LEAD APPLICATIONS -- 1. INTRODUCTION -- 2. MATERIALS AND EXPERIMENTS -- 2.1. Materials -- 2.2. Liquid lead corrosion exposures -- 2.3. Small punch testing -- 3. RESULTS -- 3.1. Liquid lead corrosion exposures -- 3.1.1. 550 C exposures -- 3.1.2. 600 C exposures -- 3.2. Small punch testing and aging -- 4. DISCUSSION -- 5. CONCLUSIONS -- ALUMINA NANOCERAMIC COATINGS: AN ENABLING TECHNOLOGY FOR HEAVY LIQUID METAL-COOLED FAST REACTORS -- 1. INTRODUCTION -- 1.1. Lead Fast Reactor (LFR) technology: materials perspective -- 1.2. Mitigation strategy for LFR development: state of the art -- 2. ALUMINA COATINGS BY PULSED LASER DEPOSITION AND BASIC PROPERTIES. 3. MATERIAL QUALIFICATION FOR lfr APPLICATIONS -- 3.1. Heavy ion irradiation tests -- 3.2. Gas permeation tests for tritium confinement -- 3.3. Corrosion tests -- 4. SUMMARY AND CONCLUSIONS -- EVALUATION OF THE HIGH-ENTROPY CR-FE-MN-NI ALLOYS COMPATIBILITY WITH A LIQUID LEAD COOLANT -- 1. INTRODUCTION -- 2. MATERIALS AND METHODS -- 3. RESULTS AND DISCUSSION -- 4. CONCLUSIONS -- DESIGN AND MATERIAL SELECTION FOR LEAK-BEFORE BREAK NATURE OF DOUBLE WALLED ONCE THROUGH STEAM GENERATORS IN LEAD-BISMUTH COOLED FAST REACTORS -- 1. INTRODUCTION -- 2. STEAM GENERATOR DESIGN AND MATERIALS SELECTION -- 2.1. Double-walled Once-through Steam Generator Design -- 2.2. Lead-bismuth side Materials Selection for Double Wall Tubes -- 2.3. Advanced Corrosion Control -- 2.3.1. Operation Temperature Reduction -- 2.3.2. Oxygen Control -- 2.4. Alumina Forming Austenitic Steel Development -- 3. LEAK BEFORE BREAK APPROACH -- 4. SUMMARY AND FUTURE WORK -- COMPATIBILITY EVALUATION ON STRUCTURAL MATERIALS FOR CLEAR IN OXYGEN CONTROLLED LEAD-BISMUTH EUTECTIC AT 500 C AND 550 C -- 1. INTRODUCTION -- 2. EXPERIMENTAL -- 2.1. Specimen preparation -- 2.2. Post-test analysis -- 3. EFFECT OF OXYGEN CONCENTRATIONS IN STATIC LBE ON CORROSION BEHAVIOR -- 4. LONG-TERM CORROSION BEHAVIOR -- 5. RESEARCH AND DEVELOPMENT OF SI-CONTAINED STEEL -- 6. CONCLUSIONS -- DEVELOPMENT OF ALUMINA FORMING MATERIALS FOR CORROSION MITIGATION IN HEAVY LIQUID METAL COOLED NUCLEAR REACTORS -- 1. INTRODUCTION -- 2. CORROSION OF METALLIC MATERIALS IN LEAD ALLOYS AT ELEVATED TEMPERTAURES -- 3. ADVANCED MITIGATION STRATEGIES -- 3.1. Surface alloying -- 3.2. AFA - alumina forming austenitic steels -- 3.3. HEA - Alumina forming High Entropy Alloys -- 3.4. Corrosion tests - Set-up and conditions -- 4. RESULTS AND DISCUSSION -- 4.1. GESA FeCrAl-surface alloys -- 4.2. AFA alloys after corrosion test. 4.3. HEA alloys after corrosion test -- 5. SUMMARY AND OUTLOOK -- CORROSION OF STEEL CLADDINGS OF FAST REACTORS FUEL ELEMENTS IN THE INTERACTION WITH URANIUM-PLUTONIUM NITRIDE FUEL -- 1. INTRODUCTION -- 2. THE EFFECT OF THE CARBON AND OXYGEN IMPURITIES -- 3. CONCLUSION -- SESSION III: QUALIFICATION PROGRAMMES OF STRUCTURAL MATERIALS FOR HLM FAST REACTORS -- QUALIFICATION PROGRAMME OF CANDIDATE MATERIALS FOR ALFRED -- 1. INTRODUCTION -- 2. MATERIAL CHOICE AND COOLANT CHEMISTRY STRATEGY -- 3. MATERIALS QUALIFICATION AND RESULTS -- 3.1. Al2O3 by PLD -- 3.2. FeCrAl Aluminizing -- 3.3. AlTiN coating by Physical Vapor Deposition (PVD) -- 3.4. Al2O3 by Detonation Gun Spray -- 3.5. AFA steels -- 4. SUMMARY AND R& -- D NEEDS -- SOME NEW R& -- D FOCUS IN STRUCTURE MATERIALS LICENSING FOR THE SVBR-100 REACTOR FACILITIES -- 1. INTRODUCTION -- 2. MAIN RESULTS OF MATERIALS CORROSION RESISTANCE JUSTIFICATION FOR SVBR-100 REACTOR -- 3. MAIN RESULTS OF MATERIALS REACTOR TESTING -- 4. RESEARCH PROGRAM ON COMPLETION OF JUSTIFICATIONS FOR THE USE OF MATERIALS -- STATUS OF HLMC TECHNOLOGY AND RELATED MATERIALS RESEARCH IN CHINA INSTITUTE OF ATOMIC ENERGY -- 1. INTRODUCTION -- 2. LBE DyNAMIC CORROSION TEST LOOP -- 2.1. Characteristic of loop design -- 2.2. Main parameters -- 2.3. Main components and auxiliary system -- 2.3.1. Main circuit -- 2.3.2. Storage tank and LBE supply system -- 2.3.3. Covering gas and vacuum system -- 2.3.4. Sampling system -- 2.3.5. Instrumentation control system -- 2.4. Current state of the loop -- 3. LBE ThermoNVECTION LOOP -- 3.1. Characteristics of loop -- 3.2. Main parameters -- 3.3. Current studies performing in the loop -- 4. LBE STATIC CORROSION TEST APPARATUS -- 4.1. Characteristic of the apparatus -- 4.2. Main parameters -- 4.3. Studies performed in the apparatus -- 5. CORROSION TESTS IN LBE -- 5.1. Parameters for corrosion test. 5.2. Test results -- 5.3. Summary -- 6. LME EFFECTS ON FM STEEL T91 -- 6.1. Parameters for SSRT test -- 6.2. Test results -- 6.3. Summary -- 7. CONCLUSIONS -- Appendix -- MEETING PROGRAMME -- A.1. Meeting Organization -- A.2. Meting Sessions -- A.3. Group Discussions -- LIST OF PARTICIPANTS -- CONTRIBUTORS TO DRAFTING AND REVIEW.
Record Nr.	UNINA-9910795871203321