Enhanced Accident Tolerant Fuel at AREVA NP. Dr. Elmar Schweitzer, Dr. Jeremy Bischoff COP23, Bonn, 11/08/2017

Similar documents
AREVA NP S ENHANCED ACCIDENT TOLERANT FUEL DEVELOPMENTS: FOCUS ON CR- COATED M5 CLADDING

Involvement of EDF in the Halden Reactor Project: a long-term cooperation in R&D

ON-GOING STUDIES AT CEA ON CHROMIUM COATED ZIRCONIUM BASED NUCLEAR FUEL CLADDINGS FOR ENHANCED ACCIDENT TOLERANT LWRS FUEL

Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Westinghouse ACCIDENT TOLERANT FUEL PROGRAM

Westinghouse Advanced Doped Pellet Characteristics and Irradiation behaviour

In-pile testing of CrN, TiAlN and AlCrN coatings on Zircaloy cladding in the Halden Reactor

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Advanced Zirconium Alloy for PWR Application

CEA ACTIVITIES SUPPORTING THE OPERATING FLEET OF NPPS

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Design bases and general design criteria for nuclear fuel. 1 General 3. 2 General design criteria 3

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Dry storage systems and aging management

3D Printing of Components and Coating Applications at Westinghouse

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Status of NEA Nuclear Science activities related to accident tolerant fuels

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

A RIA Failure Criterion based on Cladding Strain

Task 3: Licensing Plan for Accident Tolerant Fuel

OUT-OF-PILE R&D ON COATED NUCLEAR FUEL ZIRCONIUM BASED CLADDINGS FOR ENHANCED ACCIDENT TOLERANCE IN LWRS

Advanced LWR Fuels Research in the United States Shannon Bragg-Sitton Idaho National Laboratory

Topic 1: Fuel Fabrication. Daniel Mathers and Richard Stainsby

Appendix 1: Development of LWR Fuels with Enhanced Accident Tolerance; Task 1 Technical Concept Description

DRAFT PROJECT PLAN TO PREPARE THE U.S. NUCLEAR REGULATORY COMMISSION TO LICENSE AND REGULATE ACCIDENT TOLERANT FUEL

Sandrine BOUTIN Stéphanie GRAFF Aude TAISNE Olivier DUBOIS REVIEW OF FUEL SAFETY CRITERIA IN FRANCE

Oxide Surface Peeling of Advanced

Fuel Reliability (QA)

PRELIMINARY ASSESMENT OF THE PERFORMANCE OF SIC BASED ACCIDENT TOLERANT FUEL IN COMMERCIAL LWR SYSTEMS

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Simulating Pellet and Clad Mechanical Interactions of Nuclear Fuel Rod for Pressure Water Reactors

Assessment of the V&V Challenges of Accident Tolerant Fuels

Behavior of high burnup fuel during LOCA - Key observations and test plan at JAEA -

Damage Build-up in Zirconium Alloys Mechanical Processing and Impacts on Quality of the Cold Pilgering Product

NUCLEAR FUEL AND REACTOR

The Norwegian Thorium Initiative

Investigations of Structural Integrity at AREVA

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Application of Coating Technology on the Zirconium-Based Alloy to Decrease High-Temperature Oxidation

Developing Fuels with Enhanced Accident Tolerance. Fiona Rayment and Dave Goddard

CONTRIBUTION OF RESEARCH REACTORS TO THE PROGRAMMES FOR RESEARCH AND TECHNOLOGICAL DEVELOPMENT ON SAFETY

Regulatory Challenges. and Fuel Performance

SIMULATION OF FUEL BEHAVIOURS UNDER LOCA AND RIA USING FRAPTRAN AND UNCERTAINTY ANALYSIS WITH DAKOTA

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

SAFETY ENHANCEMENT TECHNOLOGY DEVELOPMENT WITH COLLABORATIVE INTERNATIONAL ACTIVITY

"DESIGN ISSUES CONCERNING COMPOSITE MATERIAL FUEL-ELEMENT JACKETS BASED ON SILICIUM CARBIDE WITHIH A MATTER OF SAFETY CONCEPT OF WATER- COOLED

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Material characterization Capabilities at IFE Kjeller (NMAT)

RECENT ACTIVITIES AND PLAN WITH FRAPCON/FRAPTRAN

How innovative approaches & technologies throughout the Fuel Cycle are supporting NPP Operations while anticipating future back-end Challenges

Studsvik Report. SCIP IV Technical Description. Public. Compiled by Hans-Urs Zwicky DRAFT AS A BASIS FOR DISCUSSION

steam oxidation and post-quench mechanical

Thorium-Plutonium LWR Fuel

Impact of the irradiation damage recovery during transportation on the subsequent room temperature tensile behavior of irradiated zirconium alloys

Overview of ATF research and ongoing experiments at the Halden reactor project

The DENOPI project: a research program on SFP under loss-of-cooling and loss-of-coolant accident conditions

Fuel Reliability Guidelines. EPRI Fuel Reliability Program

Fuel data needs for Posiva s postclosure. B. Pastina (Posiva) IGD-TP 5th Exchange Forum Kalmar

Effect of Hydrogen on ZIRLO and Zr-1.0Nb Irradiation Creep and Irradiation Growth

Post Quench Ductility of Zirconium Alloy Cladding Materials

Sustaining Material Testing Capacity in France: From OSIRIS to JHR

Joint Research Centre

PROPOSAL OF A GUIDE TO PERFORMANCE ASSESSMENT OF FUEL RODS FOR NUCLEAR POWER PLANTS

EP-450 Steel as Cladding Material for Fast Neutron Reactor Fuel Rods

European LEad-Cooled TRAining reactor: structural materials and design issues

Electric Power Research Institute. Fuel Reliability. Program Overview

R&D activities related to nuclear fuel performance and technology at the DG JRC. Paul VAN UFFELEN

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

Physical Properties. Can increase the strength by cold working but the recrystallization temperature is 400 to 500 C

Sampling of Reactor Pressure Vessel and Core Internals Ralf Oberhäuser

Safety design approach for JSFR toward the realization of GEN-IV SFR

Cladding embrittlement, swelling and creep

The international program Phebus FP (fission

INFLUENCE OF STEAM PRESSURE ON THE HIGH POST-COOLING MECHANICAL PROPERTIES OF ZIRCALOY-4 AND M5 CLADDING (LOCA CONDITIONS)

Nuclear Fission Renaissance: Opportunities for Research

Materials Issues Related to Reactor Design, Operation & Safety

Material Selection According to ALARA during Design Stages of EPR. P. Jolivet, A. Tamba, F. Chahma AREVA

Mixed-oxide (MOX) fuel performance benchmarks

Irradiation Assisted Stress Corrosion Cracking. By Topan Setiadipura [09M51695] (Obara Lab., Nuclear Engineering Dept., Tokyo Tech.

THE FUKUSHIMA ACCIDENT: IMPLICATIONS FOR NUCLEAR SAFETY. Edwin Lyman Union of Concerned Scientists May 26, 2011

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

Design of the Reactor Core for Nuclear Power Plants

OUTLINE OF THE ROKKASHO MOX FUEL FABRICATION PLANT

Structural materials for Fusion and Generation IV Fission Reactors

Fuel Rod Mechanical Behaviour Under Dynamic Load Condition on High Burnup Spent Fuel of BWR and PWR

ACTIVITIES in NUCLEAR FUEL BEHAVIOUR

A New Method Taking into Account Physical Phenomena Related to Fuel Behaviour During LOCA

Effects of Pre-Irradiation on Irradiation Growth & Creep of Re-Crystallized Zircaloy-4

A Brief Summary of Analysis of FK-1 and FK-2 by RANNS

Understanding the effects of reflooding in a reactor core beyond LOCA conditions

Preliminary Neutronic Assessment for ATF (Accident Tolerant Fuel) based on Iron Alloy

nuclear science and technology

available online at MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION Martin Sevecek a,b

AREVA: Experience in Dismantling and Packing of Pressure Vessel and Core Internals

Acceptance Criteria in DBA

MYRRHA FUEL TRANSIENT TESTS PROJECT AT THE TRIGA-ACPR REACTOR

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

ZRO 2 AND UO 2 DISSOLUTION BY MOLTEN ZIRCALLOY

SFR-Pyroprocessing Development Plan

Transcription:

Enhanced Accident Tolerant Fuel at Dr. Elmar Schweitzer, Dr. Jeremy Bischoff COP23, Bonn, 11/08/2017

Why Develop eatf Solutions? Zr alloy eatf solution p.2

eatf Program u Evolutionary Concept (Near-term solution) Cr-coated Zr cladding + Cr 2 O 3 -doped UO 2 fuel u Revolutionary Concept (Long-term solution) SiC f /SiC cladding Multi-layer composite cladding with a metallic liner to ensure the leak-tightness Cr-coating provides significant reduction in HT steam oxidation Cr 2 O 3 -doped fuel shows high fission gas retention in transient conditions - Very low oxidation kinetics in HT steam - High strength at high temperature - High melting temperature p.3

Description of s Near-Term eatf Fuel Rod Concept Cr-Coated Zr Cladding Cr 2 O 3 -doped UO 2 Fuel Large Grain Microstructure Thin coating (5-20µm) Dense coating 100µm Enhanced viscoplasticity No modification of the underlying Zr substrate microstructure p.4

Normal Operation: Benefits of Cr-Coated Cladding with Cr 2 O 3 -doped UO 2 Fuel Cr-Coated Cladding FGR (Rod Pressure) Improved PCI Behavior Cr 2 O 3 -doped UO 2 Fuel Corrosion + H Pickup Increased Fuel Density Fuel Failures Increased Fuel Utilization Increased Margins p.5 Cladding Wear Provides: - RELIABILITY - FLEXIBLE OPERATION - ECONOMIC BENEFITS Chipping

Accidental Conditions: Benefits of Cr-Coated Cladding with Cr 2 O 3 -doped UO 2 Fuel Cr-Coated Cladding HT Steam Oxidation HT Creep & Ballooning FGR (Rod Pressure) Cr 2 O 3 -doped UO 2 Fuel Increased Post-Quench Ductility Heat and H 2 Production Improved Coolable Geometry Release of Radionuclides Fuel Fragmentation Increased: - COPING TIME - MARGINS - SAFETY p.6

Licensing and Performance Confirmation Performance Data Licensing Confirmation Reactor Sample Type PIE Timeline OSIRIS - CEA IMAGO - Commercial Reactor (Europe) Cr-coated tubes and flat specimens Cr-coated tubes, flat specimens, strained samples Coating performance, adherence Corrosion kinetics, coating performance, mechanical properties, microstructural eval 2015-2018 2016-2019 2019-2022 Commercial Reactor (US) Lead Fuel Rods Cr 2 O 3 -doped pellets Pool side examination 2017 HALDEN - Norway Cr-coated rodlets with UO 2 pellets Coating performance, pellet-clad interaction 2017-2021 ATR - INL Cr-coated rodlets with Cr 2 O 3 -doped pellets Microstructural evaluation, oxidation, hydrogen pickup, coating performance 2017-2021 p.7

The road towards eatf 2016 2017 2018 2019 2020 2021 2022 2023 2024 2025 2026 2027 2028 2029 IMAGO 2016 Irradiation of Material Test Samples + PIE LTR programs Irradiation of lead test rods + PIE LTA programs (planned) commercial eatf FA Base M5 Cladding Chromia-doped UO 2 pellets Cr-coating p.8

Excellent Behavior of Cr-coated Cladding after 1 Cycle of Irradiation IMAGO irradiation project (European PWR) u First irradiation of ATF materials in a commercial PWR Slight golden color = very thin oxide layer (<1µm) Uncoated segment Cr-coated segment Smooth transition coated - uncoated No delamination = very adherent coating Excellent behavior of Cr-coated cladding Important input data to support lead fuel rod justification p.9

Fabrication of Full-Length PVD Prototype Machine to Manufacture LFR Cladding Lab Scale Feasibility Full-Length Feasibility Industrial Scale Production Goals: p.10 u Demonstrate feasibility of full-length Cr deposition by Physical vapor deposition u Produce Cr-coated cladding for lead fuel rods Full-length Cr-coated cladding fabrication on track for lead fuel rod insertion in 2019 Fabrication of 6m long prototype completed Start-up and first tube production on-going

is fast-tracking the implementation of its near-term eatf solution: u Cr-coated Zr cladding + Cr 2 O 3 -doped UO 2 fuel Take Away Message First commercial irradiation feedback OK Full-length coating prototype completed LFR insertion 2019 p.11

Any reproduction, alteration, transmission to any third party or publication in whole or in part of this document and/or its content is prohibited unless AREVA has provided its prior and written consent. This document and any information it contains shall not be used for any other purpose than the one for which they were provided. Legal action may be taken against any infringer and/or any person breaching the aforementioned obligations. p.12