Overviewof Gen IV Reactor Systems Development

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1 Overviewof Gen IV Reactor Systems Development

2 Status by project 2

3 Generation IV GFR - Summary Helium coolant Fast neutron spectrum High outlet temperature Longer term alternative to SFR + Transparent coolant + High temperature/efficiency + Strong Doppler effect + Weak void effect + Chemically and neutronically inert coolant + Zero activation cooant - Decay heat removal (LOCA) - High power density - Low thermal inertia - High coolant pumping power Thermal power 2400 MWth Coolant in/out 400 C/850 C System pressure 70 bar 3

4 Status of GFR System Cooperation GFR System Arrangementsigned by Euratom, France, Switzerland and Japan Project Arrangement on Conceptual Design & Safety signed by Euratom, France and Switzerland Project Plan was intended updated for but this has proved to be difficult because no funding available for the foreseeable future in Euratom and Switzerland and only small funds available in France to support ALLEGRO development. 4

5 SCWR System Agreement (year of sign.) and Representatives Canada (2006) L. Leung, D. Brady Euratom(2006) T. Schulenberg, J. Starflinger Japan (2006) H. Matsui Russia (2011) A. Sedov, A. Churkin China (2014) Y.P. Huang, L. Zhang Projects: Thermal-Hydraulics and Safety, TH&S, signed (EU, CA, JP), RU and CN expressed interest to join Materials and Chemistry, M&C, signed (EU, CA, JP), CN expressed interest to join Fuel Qualification Test, FQT, provisional (EU, CA, CN) System Integration and Assessment, SI&A, provisional (EU, CA, JP) 5

6 Project Thermal-Hydraulics and Safety Progress in 2013: Canada: 8 deliverables on heat transfer, choking flow, safety systems, thermal insulation Euratom: 8 deliverables on turbulence modelling and heat transfer Similar in 2014 Joint benchmark exercise completed 2014 Flow and heat transfer of supercritical water in a 7 rod bundle Tests by JAEA, Japan Blind predictions by 10 organizations from EU and Canada Organized by M. Rohde, TU Delft 6

7 Thermal-Hydraulics and Safety, Updated Project Plan Planned future contributions 2015 to 2019, e.g. Heat transfer to supercritical water in tubes, annuli, sub-channels and rod bundles (CA, CN, RF) Heat transfer to supercritical CO 2 and Freon in tubes, annuli and rod bundles; analysis of fluid-to-fluid scaling laws (CA, CN, EU, RF) Pressure loss of supercritical water flow in rod bundles (CN, RF) Test of rod cladding ballooning (RF) Blow-down experiments with supercritical water (CA, CN, RF) Flow instabilities (CA, CN, EU, RF) SCWR safety requirements and evaluation (CA, EU, CN, RF) System code development (CA, CN) CFD and turbulence modelling (CA, CN, EU, RF) 7

8 Project Materials and Chemistry Canada: total of 18 deliverables Progress in 2014: Euratom: total of 14 deliverables; commissioning tests of in-pile supercritical water loop in Rez completed CA, EU, JP: Round-robin corrosion tests and characterization of identical alloys; development of Materials Databases in SCW CA, EU: Development of coatings and surface modification Updated project plan 2015 to 2019 under negotiation, e.g. Tests of un-irradiated material: corrosion, SCC, creep, effect of coatings and surface modification, ODS-materials (CA, CN, EU) Radiolysis and water chemistry: corrosion tests with an in-pile supercritical water loop (EU), supported by modelling (CA), and outof-pile test (CN) 8

9 EXHAUST TO VENT STACK 7m (23') IHX X-SECTION (FLATTENED FOR CLARITY) CONTROL RODS (7) PLAN VIEW OF IHX AND PUMPS IHX (2) 2 1.7m EACH PUMPS (2) ON Ø 142.5" B.C. DRACS (2) 2 0.4m EACH SECONDARY CONTROL RODS Na-CO 2 HEAT EXCHANGER SODI UM DUMP TANK Ø 2.5 m x 3.8 m LONG (Ø 7.5' x 12.6'LONG) PLAN VIEW OF THE CORE PRIMARY CONTROL RODS CORE BARREL Ø 266 / 268 cm (104.7" / 105.5") METERS 10 TURBINE/GENERATOR BUIL DING 3.25m (10'-8") 7m [23FT] (29.5") 0.75m THERMAL SHIELD 1m TRAVEL DISTANCE OF THE CONTROL RODS 4.57m Primary Ves sel I.D. [15FT] IHX Guard Ves sel I.D. 5.08m [16.7FT] m 3,186 gal. 1.89m [6.2FT] (Ø 25.5') Ø 7.7m SECTION A - A Na-Air HEAT EXCHANGER (2) ELEVATOR 3.5m 1m (11'-8") ( 39.4") GUARD VESSEL (1" THICK) PRIMARY VESSEL (2" THICK) CONTROL BUILDING 12.72m [41.7FT] 14.76m [48.4FT].61m [2FT] Ho t P o ol No r ma l s o di um le v el P um pof f Sodium Le v el Co l dpo o l Nor ma l s o di um le v e l 2.29m 1.93m [ 7.5FT] [6.3F T] Sodi um fa ult ed l ev e l Gen IV SFR System Options and Design Tracks Loop Pool Small Modular JSFR ESFR PGSFR SMFR AHX Chimney PDRC piping IHTS piping Steam Generator IHX DHX PHTS pump Reactor core IHTS pump In-vessel core catcher BN-1200 will be presented by Russia as new Gen-IV SFR design track for the next SIA meeting 9

10 System Integration & Assessment Project Objectives Integration of the results of R&D Projects Performance of design and safety studies Assessment of the SFR System against the goals and criteria set out in the Gen IV Technology Roadmap Integration Role Specific tasks have been developed and refined Identify Generation-IV SFR Options» General system options» Specific design tracks» Contributed trade studies Maintain comprehensive list of R&D needs Review Generation-IV SFR Technical Projects Unlike the technical Projects, based on synthesis of results produced by other Projects 10

11 Safety and Operation Project Partners CHINA INSTITUTE OF ATOMIC ENERGY COMMISSARIAT À L'ÉNERGIE ATOMIQUE ET AUX ÉNERGIES ALTERNATIVES EUROPEAN ATOMIC ENERGY COMMUNITY DEPARTMENT OF ENERGY OF THE UNITED STATES OF AMERICA JAPAN ATOMIC ENERGY AGENCY KOREA ATOMIC ENERGY RESEARCH INSTITUTE STATE ATOMIC ENERGY CORPORATION ROSATOM Activities in three areas 1. Methods, models and codes, 2. Experimental programs and operational experiences 3. Studies of innovative design and safety systems. 11

12 CD & BOP Project Subjects for (1) In-Service Inspection & Instrumentation (ISI) technology Ultrasonic inspection in sodium using different approaches and technologies, codes and standards (CEA, Euratom, JAEA, KAERI) (2) Repair experience Phénix, Monju, (CEA, JAEA) (3) Leak Before Break (LBB) Assessment technology Creep, fatigue, and creep-fatigue crack initiation & growth evaluation for Mod. 9Cr-1Mo (Grade 91) steel, Na leak detection by laser spectroscopy (JAEA, KAERI) (4) Supercritical CO 2 BraytonCycle Energy Conversion S-CO2 compressor tests, S-CO 2 cycle demonstration tests, Compact heat exchanger tests, Material oxidation tests in S-CO 2, Sodium-CO 2 reaction tests, S-CO 2 SFR plant dynamic analyses and control strategy development, Computer code analysis, S-CO 2 SFR design study, Validation of S-CO 2 plant dynamic analyses with S-CO 2 loop data, Sodium plugging tests (CEA, DOE, Euratom, JAEA, KAERI) (5) Steam Generator design and associated safety & instrumentation (since 2011) Na/water reaction, thermal-hydraulics, thermal performance, DWT structural evaluation and heat exchange performance, DWT-SG fabrication (CEA, JAEA, KAERI) 12

13 Overview of GACID Conceptual Scheme Objective: to demonstrate, using Joyo and Monju, that FR s can transmute MA s (Np/Am/Cm) and thereby reduce the concerns of HL radioactive wastes and proliferation risks. A phased approach in three steps. Material properties and irradiation behavior are also studied and investigated. Step-1 Np/Ampin irrad. test Step-2 Np/Am/Cm pin irrad. test Step-3 Np/Am/Cm bundle irrad. test GACID overall schedule Joyo Joyo Planning Monju Test fuel fabrication Monju MA-bearing MOX fuel pellets Fuel pin fabrication Monju (Final Goal) Monju Irradiation test MA raw material preparation The Project is being conducted by CEA, USDOE and JAEA as a GIF/SFR Project, covering the initial 5 years since Sep. 27,

14 VHTR PMBs 4 active VHTR Projects: Hydrogen Production (HP)» Chair/co-chair: Francois LE NAOUR (FR) / Sam SUPPIAH (CA) Fuel and Fuel Cycle (FFC)» Chair/co-chair: David PETTI (US) / LIU Bing (CN) Materials (MAT)» Chair: William R. CORWIN (US)» 3 Working Groups: Metals, Graphite, Ceramics Computational Methods Validation and Benchmarks (CMVB) restarted since Oct 2014» Chair/co-chair: SHI Lei (CN) / Hans GOUGAR (US) System Integration and Assessment Project (SIA)» under discussion (limited resources) 14

15 Development targets Very High Temperature is not a goal per se Two stages for VHTR Near-term: He outlet temperatures C for process steam applications prepare for construction and licensing of a demonstrator/foak Longer-term: New materials and fuels should enable higher temperatures up to above 1000 C; bulk H 2 from thermochemical processes 15

16 R&D objectives Qualification of TRISO fuel»uo 2 and/or UCO fuel Fuel cycle: Disposal of fuel and graphite Metal, Graphite, Composites»Pressure vessel materials, Components (SG / IHX), core internals, valves H 2 Production»HTSE, S-I, Cu-Cl Computer tool validation 16

17 MSR system Memorandum of Understanding effective 6 Oct 2010 JRC ( Euratom ), CEA ( France ) Russia (Nov 2013) Japan, China, Korea, US observers LFR system Memorandum of Understanding effective 22 Nov 2010 JRC ( Euratom ), Tokyo Institute of Technology ( Japan ) Russia (Rosatom) joined 18 Jul 2011 US, Korea, China observers 17

18 Studied MSR Concepts Two reactor concepts using molten salt are discussed in GIF MSR meetings MSFR MOSART Molten salt reactors, in which the salt is at the same time the fuel and the cooling liquid» MSR MOU Signatories France and EU work on MSFR (Molten Salt Fast Reactor)» Russian Federation works on MOSART (Molten Salt Actinide Recycler & Transmuter). Russian Federation joined the Memorandum of Understanding (11/2013) Solid fuelled Reactors cooled by molten salt» USA and China work on FHR(fluoridesalt-cooled high-temperature reactor) concepts and are Observers to the PSSC FHR 18

19 GIF MSR Project A Provisional Project Management Board has been set up Two meetings per year where members and observers report on their activities and recent progresses The project is devoted to Molten Salt Reactors Information is also exchanged on solid fuelled reactors cooled by molten salt The various molten salt reactor projects like FHR, MOSART, MSFR, and TMSR have common themes in basic R&D areas, of which the most prominent are: o liquid salt technology, o materials behavior, o the fuel and fuel cycle chemistry and modeling, o the numerical simulation and safety design aspects of the reactor 19

20 GIF LFR REFERENCE SYSTEMS Three reference systems of GIF LFR are: ELFR (600 MWe), BREST (300 MWe), and SSTAR (small size) CLOSURE HEAD CO2 OUTLET NOZZLE (1 OF 8) CO 2 INLET NOZZLE (1 OF 4) Pb-TO-CO 2 HEAT EXCHANGER (1 OF 4) CONTROL ROD DRIVES CONTROL ROD GUIDE TUBES AND DRIVELINES THERMAL BAFFLE SG Primary Pump Reactor Vessel Safety Vessel FAs DHR dip cooler 3 BREST system of intermediate size 2 4 FLOW SHROUD RADIAL REFLECTOR ACTIVE CORE AND FISSION GAS PLENUM FLOW DISTRIBUTOR HEAD GUARD VESSEL REACTOR VESSEL SSTAR system of small size with long core life ELFR system for central station power generation Core 2 - Steam Generator 3 - Pump 4 - Refueling Machine 5 - Reactor Vault 20

21 Status of the main activities: SRP, White Paper, SDC, ToR SYSTEM RESEARCH PLAN (SRP): Substantial revision of SRP was started by mid-2012 and is now completed. Final draft of SRP has been issued by pssc and the report is currently being reviewed by EG LFR White Paper on safety: Review of White Paper on safety (based on ALFRED as an example of an LFR to apply ISAM to) was completed by EG. The final version of the paper has already been published on the GIF web-site by RSWG LFR Safety Design Criteria: Safety Design Criteria (SDC) for LFR will be developed on the basis of SDC for SFRs Work is still ongoing, first draft is expected to be availableby spring 2015 GIF LFR abstract was sent to GIF Symposium held in conjunction with the ICONE23 conference in Japan (May 2015) Preparation of draft Terms of reference for GIF system safety assessment is currently ongoing and the draft is expected to be available shortly (by March 2015) 2014 Annual report was sent to the GIF Secretariat in the first week of December Revision of LFR information on the GIF web-site will be available by spring

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