ASTRID. Presented by A. VASILE (CEA) ADVANCED SODIUM TECHNOLOGICAL REACTOR FOR INDUSTRIAL DEMONSTRATION

Transcription

1 ASTRID ADVANCED SODIUM TECHNOLOGICAL REACTOR FOR INDUSTRIAL DEMONSTRATION Presented by A. VASILE (CEA) IAEA Seminar on Fast Reactors, Bariloche, Argentina, 1 5 October 2012 Alfredo Vasile OCTOBER 2012 CEA JUNE 2012 PAGE 1

2 OUTLINE ASTRID objectives and organization of the project General features of the reactor Safety: Examples of major improvements Conclusions CEA OCTOBER 2012 PAGE 2

3 OBJECTIVES AN ORGANIZATION CEA JUNE 2012 PAGE 3

4 ASTRID OBJECTIVES Industrial prototype (could be a step before a First Of A Kind) Including French and international SFRs feedback A GEN IV system Safety Level at least equivalent to GEN III systems Progress on specific Na reactors issues Including FUKUSHIMA accident feedback Operability Load factor of 80% or more after first learning years Significant progress concerning In Service Inspection & Repair (ISIR) Waste transmutation Demonstration of minor actinides transmutation according to June 28, 2006 French Act on Waste Management A mastered investment cost Irradiation services and options test CEA OCTOBER 2012 PAGE 4

5 PROJECT ORGANIZATION CEA/Nuclear Energy Division is the leader of the ASTRID project Industrial partnerships to cover main design engineering batches AREVA NP: nuclear island (core and fuel stays at CEA) Support to the owner by EDF ALSTOM: turbine island BOUYGUES:civil engineering COMEX NUCLEAIRE: batches in robotics and mechanics TOSHIBA: development of large electromagnetic pumps On going discussions for other partnerships Manpower: 300 CEA, 200 industry For technical developments and experimental facilities, CEA is willing to develop international partnerships CEA OCTOBER 2012 PAGE 5

6 SCHEDULE FOR ASTRID AND ASSOCIATED FACILITIES Preliminary choice of options Decision to continue Decision to build Fuel loading ASTRID Pre-conceptual design Conceptual design Basic design Detailed design & Constructi on Facilities Feasibility Report on minor actinides partitioning Position Report on minor actinides partitioning and transmutation Core manufacturing workshop (AFC) MA bearing fuels fabrication facility Commercial deployment: from CEA OCTOBER 2012 PAGE 6

7 GENERAL SCHEDULE CEA OCTOBER 2012 PAGE 7

8 MAIN MILESTONES Mid of 2010: preliminary selection of ASTRID characteristics for launching the preconceptual design Preconceptual design: 2010 to end of 2012: The preconceptual design is considering some open options. Innovation and technological breakthroughs are favored, while maintaining risk at an acceptable level During the preconceptual design phase, start of the interactions with the Safety authorities on safety objectives and orientation First estimation of ASTRID investment cost, including the different open options Schedule of next steps and their associated costs Safety orientation report: report delivering and first advices of the French Safety Authorities End 2014: Conceptual design Safety Options File CEA OCTOBER 2012 PAGE 8

9 GENERAL FEATURES OF ASTRID CEA JUNE 2012 PAGE 9

10 PRELIMINARY DESIGN CHOISES / OPEN OPTIONS Main features 1500 thmw - ~600 emw pool type reactor With an intermediate sodium circuit High level expectations in terms of safety demonstration Preliminary strategy for severe accidents (core catcher ) Diversified decay heat removal systems Oxide fuel UO2-PuO2 for starting cores Transmutation capability Fuel handling in sodium. Open options Core design Energy conversion system Reference : water/steam Alternative : N2 Number of loops Devices to eliminate severe accidents (i.e. 3rd shutdown level) Core catcher technology SGs materials and technology Innovative technologies for Na fires detection and mastering I&C Innovative options to be tested Many of these options will be decided during the pre-conceptual design up to end 2012, Carbide fuel SiC-SiC materials CEA OCTOBER 2012 PAGE 10

11 DESIGN OF THE NUCLEAR ISLAND CEA JUNE 2012

12 MATERIAL SCOPE 316 LN 800H 9 Cr? 316 LN 9 Cr? SGs, Energy conversion System C ageing, weldings, compatibility Pipes and circuits C creep, fatigue, creepfatigue, thermal fatigue,ageing weldings 316 LN Ni alloy Fixed structures Life time years Hot structures 550 C + low irradiation creep, seal coef., weldings Core structures C nominal C accidental High irradiation No swelling First core EM10->T91/T92? AIM1->AIM2? Long term R & D F/M ODS->Base V? Ceramics->SiC/SiC? Specific field of investigation : long term process, nuclear facilities Cold structures Heat exchanger, Pumps 400 C + low irradiation no deformation 316 LN 316 LN Vessel 400 C no deformation negligeable creep

13 MATERIAL SCOPE ASTRID FOCUS To extend R&D programs on accidental conditions (from the Fukushima accident): high temperature material behaviour: creep, creep-fatigue, Stellites replacements: aims to gain benefits for decommissioning, Welded seals behaviour, Material behaviour justification up to lifetimes of 60 years CEA OCTOBER 2012 PAGE 13

14 LAY-OUT EXAMPLES 4 secondary loops CEA OCTOBER 2012 PAGE 14

15 TRANSMUTATION CAPABILITY A progressive approach flexible to meet the decisions according to June 28, 2006 French Act on Wastes Management 1 st milestone end levels of demonstration: experimental stage Pellets A few pins A subassembly reserved positions for several subassemblies 2 ways considered : homogeneous and heterogeneous Transmutation of Am, eventually Np and Cm Specification for ASTRID in 2012 CEA OCTOBER 2012 PAGE 15

16 SAFETY: EXEMPLES OF MAJOR IMPROVEMENTS CEA JUNE 2012 PAGE 16

17 ASTRID SAFETY APPROACH CEA OCTOBER 2012 PAGE 17

18 A CORE WITH ENHANCED SAFETY CFV core Absorbing protection SFRv2 core Sodium plenum zone Upper inner fissile zone Inner fertile zone Outer fissile zone Feedback experience from SPX and EFR : To reduce the fuel reactivity loss per cycle To reduce the sodium void worth 2008 : concept with larger pin and smaller-diameter spacing wire Increase of the fuel fraction Low reactivity loss during cycle Decrease of the Na fraction lower Na voiding effect Progress on control rod withdrawal Lower inner fissile zone Fertile blanket Neutronic protection 2009 : heterogeneities added to SFRv2 core Sodium void worth strongly reduced Quantification of the potential benefit from safety point of view is on going Choice of the core: september 2012 CEA OCTOBER 2012 PAGE 18

19 EANOS/PARIS CALCULATIONS CORE DESIGN OPTIONS: SFRV2 AND CFV LAYOUTS SFRv2 AIM MWth CFV V1 - AIM MWth Spatial distribution of the sodium void effect Spatial distribution of Doppler CEA OCTOBER 2012 PAGE 19

20 CORE DESIGN OPTIONS: SFRV2 AND CFV CHARACTERISTICS Core 1500 MWth SFRV2B CFV-V1 Number of fuel pin / SA Fuel pin diameter (mm) 9,43 8,45 Pu enrichment E1/ E2 (%) 13,9 / 17,6 23,5 / 20 Height H1 / H2 (cm) / 90 Number of SA C1 / C2 144 / / 114 Number batch / Fuel cycle lenght 4 x 390 JEPP 4 x 360 JEPP Void effect ($) - RZ +5,1-0,5 Breeding gain -0,05-0,02 TCT average fissile C1/C2 (GWj/t) 76 / / 69 Dr / cycle (pcm/jepp) -2,2-4,3 Number of CR (*) Core diameter (cm) Plin max BOL (W/cm) Amplification based on individual effects CEA OCTOBER 2012 PAGE 20

21 COMPARISON OF PRELIMINARY RESULTS ON UNPROTECTED LOSS OF FLOW SITUATIONS Best estimate calculations trends (uncertainty analysis is on-going) CFV V1 V2B ULOSSP Margin to Na boiling of 55 C Na boiling in 100s ULOF Na boiling in 3500s Na boiling in 100s ULOHS Temperature of neutronic shutdown : 700 C Temperature of neutronic shutdown : 800 C LIPOSO 680 C / 45%Pn 736 C / 43%Qn CFV core: a promising core for an improved intrinsic behaviour in case of unprotected situations and control rod withdrawal Analysis of severe accidents conditions are on-going On-going definition of version 2 of CFV core for improving inherent behaviour (with the objectives to increase the sodium boiling margin and robust demonstration of no fuel melting in case of CRW) CEA OCTOBER 2012 PAGE 21

22 CORE DESIGN OBJECTIVES Natural behavior favorable for transients of unprotected loss of flow and loss of heat sink Target criteria : no sodium boiling for a ULOSSP transient Sodium void effect minimized Target criteria : Na void effect < 0 Natural behavior favorable for a complete control rod withdrawal (with no detection) Target criteria : no fuel melt Improved performances Target criteria : Cycle length 480 efpd, High fuel burnup, and breeding gain 0 Core design extrapolable to higher power CEA OCTOBER 2012 PAGE 22

23 CORE DESIGN APPROACH ASTRID core design is mainly guided by safety objectives: 1. Prevention of the core meltdown accident by a natural behavior of the core and the reactor (as 3rd line of defense in case of no actuation of the two shutdown systems) Natural behavior favorable for transients of unprotected loss of flow and loss of heat sink Target criteria : no sodium boiling for a ULOSSP transient for CFV type core (CFV = Low Na void worth core) Sodium void effect minimized Target criteria : Na void effect < 0 for CFV type core Natural behavior favorable for a complete control rod withdrawal (with no detection) Target criteria : no fuel fusion by adding passive complementary systems if natural behavior is not sufficient for some transient cases 2. Mitigation of the core meltdown Absorbing protection Upper inner fissile zone Inner fertile zone Lower inner fissile zone Fertile blanket Sodium plenum zone Outer fissile zone Neutronic protection To garantee that core meltdown accidents don t lead to significant mechanical energy release, whatever initiator event by a favorable natural core behavior (negative sodium void worth for CFV type core) by adding specific mitigation dispositions in case of natural behavior is not sufficient CEA OCTOBER 2012 PAGE 23

24 ISSUES TO BE INVESTIGATED BY EXPERIMENTS Measures for In-Vessel-Retention (IVR) have priority but 3 options are open Internal core-catcher Core-catcher between 2 vessels External core-catcher Design studies for ASTRID aim at preventing the risk of core melting. However according to the WENRA 2010 «Safety Objectives for New Nuclear Power Plants», 4th level of in depth prevention request that this accident will be taken into account in the design process. CEA OCTOBER 2012 PAGE 24

25 DECAY HEAT REMOVAL ARCHITECTURE Sodium/air loop, using secondary circuit (connected circuit on secondary pipes, or dedicated heat exchanger integrated in Intermediate Heat Exchanger) Circulation of air along steam generators vessels Dedicated circuit set outside of reactor vessel, using vessel wall as heat exchange surface Dedicated sodium/air loop, with its own heat exchanger located in the primary circuit 4 Na/air loop, dedicated loop in the main vessel 1 Na/air loop, using secondary loop 2 air circulation along steam generator 3 cooling through the vessel CEA OCTOBER 2012 PAGE 25

26 SODIUM WATER REACTIONS Sodium water reaction Violent and exothermic reaction Main reaction : Na + H 2 O NaOH + ½ H kj/ mole of water (at 500 C) Effects of a sodium-water reaction in a Steam Generator Chemical effects Global corrosion in polluted sodium environment Local erosion / corrosion («wastage») self-evolution of the tube leak orifice damage of the nearest tubes Mechanical effects For large leaks(>100 g/s) Fast over-pressure associated to pressure wave propagation Slow over-pressure associated to massive introduction of water in secondary sodium circuit Thermal effects For large leaks(>100 g/s) and due to exothermal reaction effects on tubes : heating, creeping, swelling, burst 2 ways to reduce the SWR risk: Improve SG design of the steam PCS (Rankine cycle) in order to : reduce the risk of SWR occurrence limit the consequences of an hypothetical violent reaction PCS (Brayton cycle with pure nitrogen at 180 bar) in place of steam cycle to eliminate de facto the SWR risk CEA OCTOBER 2012 Feasibility to be demonstrated.

27 STEAM POWER CONVERSION SYSTEM (1/2) Innovation : modular SGs Protection of secondary piping and intermediate heat exchanger integrity in case of simultaneous failure of all the tubes of a SG module (accidental envelope case) Imply a maximal SG power of about 150 MWth Improvement of detection Hydrogen detection by means of permeation through very sensitive nickel membrane but : Complicated fabrication and operation Response time to be optimized An electro-chemical flow meter has been tested recently in PHENIX Simpler Possible optimization of response time Study of diversified detection method based on acoustic principle in progress Protection against large sodium-water reactions ensured by : Passive fast draining of the sodium loop by means of rupture disk Fast insulation and depressurization of the water-steam loop CEA OCTOBER 2012 PAGE 27

28 STEAM POWER CONVERSION SYSTEM (2/2) Reverse SG (sodium inside the tubes) Innovative approach with SG no sensitive to wastage phenomena occurring in classical SGs Reduce the probability of crack to leak evolution (tubes with external pressure) Speeds up the leak detection Slow down drastically the propagation of a possible SWR Feed-back : 2 designs of reverse SG operated in BOR-60 reactor Preliminary design of 125 MWth modules achieved Issues : Dimensioning of external pressurized shell Modeling of the sodium-water reaction General design of the reverse SG In service inspection CEA OCTOBER 2012 PAGE 28

29 GAS POWER CONVERSION SYSTEM (1/2) Very innovative concept with feasibility to be demonstrated Nitrogen selected as gas, at a pressure of 180 bars Encouraging first results Net efficiency of the reactor plant about 38% possible Turbomachinery Turbine Two possible designs: single flow and split flow with high isentropic efficiency (94%) Turbine design challenging but not unfeasible (no showstopper identified) Compressor Two technologies (axial and radial) with equivalent isentropic efficiency (90%) Radial technology should be put forward : simpler, cheaper, no performance test required HP and LP compressors with same technology CEA OCTOBER 2012 PAGE 29

30 GAS POWER CONVERSION SYSTEM (2/2) Key points on sodium/gas heat exchangers Compact heat exchangers PCHE heat exchanger technology choosen Technological issue of the gas PCS: codification, fabrication control and In Service Inspection Tubes and shell heat exchangers More robust back-up option, but very heavy components and large floor space requirements, feasibility and integration to be detailed. General architecture Investigation in progress to optimize piping layout, performance (pressure losses), accessibility, maintenance and operation. CEA OCTOBER 2012 PAGE 30

31 SURVEILLANCE AND ISI&R: A FOUR LEVEL STRATEGY Looking for Considered options * Operating parameters variation improve the prevention level L1: Continuous monitoring Statutory ISI * Abnormal deformation of structures * Vibrations * Leakage * Safety Robust and redundant detection systems Innovative instrumentation Up to date technologies * Excessive deformations ISI&R oriented design ASTRID ISI&R L2: Periodic examination Statutory ISI L3: Exceptional interventions Doubts / warnings * Fatigue (or creep-fatigue) cracks next to the welded junction * Corrosion / loss of thickness * Erosion on rotating parts * * The same type as for Level 2 Localization : everywhere! Under Na telemetry Under Na volumetric NDT Under sodium robotic carriers The same type as for Level 2 Repair * Noxious cracks Removable components L4: Repair Investment protection * Loss of parts * Stuck mechanisms * Out of use primary components Specific repair tools * CEA OCTOBER 2012 PAGE 31

32 SURVEILLANCE AND PROTECTION OF THE CORE Inlet and outlet core temperature DT, CRW (TIB?) TC, optical fiber, flowmeter Surveillance ciel de pile Gaseous FP Clad failure Localisation in gaz Ionisation chamber, Gaseous FP Spectro g, (CRDS) Core CFV Acoustic Detection Abnormal noise (TIB?) Innovative Line of Defense A ( ) R&D Traducteurs US haute température Inlet and outlet core flowrate TIB, loss of flow Active Acoustic Det. Gas rate SONAR, TUSHT Neutronic monitoring Fonctionnement, CRW High Temperature Fission Chambers with large dynamics Clad Failure Detection in Na Open CF, BTI HTFC / IHX US telemetry under Na Core movment, (Temperature) Clad failure Localisation in Na FP neutronic detection FC with B CEA OCTOBER 2012 PAGE 32

33 CONCLUSIONS R&D results [CEA-AREVA-EDF] obtained from 2007 to 2009 have contributed to ASTRID mid 2010 choice of options ASTRID has the objective to demonstrate at the industrial scale progress in the identified domains of SFR weakness (safety, operability, economy) and to perform transmutation demonstrations A lot of improvements are related to safety The first very important milestone is 2012 (June 2006 French Act on wastes management) : ASTRID pre-conceptual design studies : First investment cost evaluation First safety Authorities advice on the orientations for ASTRID safety With the ASTRID program funded by the French government, France has the opportunity to develop a GEN IV Sodium Fast Reactor CEA OCTOBER 2012 PAGE 33

34 Commercial reactor ASTRID RAPSODIE

35 FAVOURABLE CHARACTERISTICS OF SFR Phénix ( ) Super-Phénix ( ) Rapsodie ( ) Easy to operate: no pressurization of the primary coolant, high thermal inertia, control by single rod position, no xenon effect, no need of soluble neutron poison Radiation protection : higher level of protection than LWR Few effluents and little radioactive waste High thermal efficiency Large sodium boiling margin Natural convection Diversification of heat sink by using air CEA OCTOBER 2012 PAGE 35

36 PAGE 36 CEA JUNE 2012 Commissariat à l énergie atomique et aux énergies alternatives Centre de Cadarache Saint Paul Lez Durance T. +33 (0) Nuclear Energy Directorate Reactor Studies Department Etablissement public à caractère industriel et commercial RCS Paris B