Substantiation Safety Approaches & Safety Design Goals of JSFR*

Transcription

1 Substantiation Safety Approaches & Safety Design Goals of JSFR* Advanced Nuclear System Research & Development Directorate Japan Atomic Energy Agency (JAEA) Shoji KOTAKE IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June 2010 * Japan Sodium-cooled Fast Reactor

2 Contents Japan Sodium-cooled Fast Reactor Progress in Safety Features of SFR in Japan Development Targets Design Requirements on Safety Event Categorization & Design Principle JSFR Safety Features Design Approach (RSS, HTS) Design Measures against CDA Design Measures against Chemical potential of Na R&D Statuses Concluding Remarks IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

3 Progress in Safety Features of SFR in Japan Monju Prototype DFBR JSFR Demonstration / Commercialization Design for Demonstration DBE, BDBE and Site evaluation events - Reinforced passive safety features and In Vessel Retention with re-criticality free core Operation Start by 2025 Power :1785MWt / 750MWe(not yet decided), Temperature: 550ºC Innovative Technologies for Safety Enhancement DBE, BDBE and Site evaluation events - Passive shutdown system in BDBE and accommodation with CDA energetics Design Work during 90s Power :1,600MWt / 660MWe, Temperature: 550ºC DBE, BDBE and Site evaluation events - Mechanistic approach and design accommodation with CDA energetics Reactor Initial Criticality in 1994, & Restarted in 2010 Power :714MWt / 280MWe, Temperature: 529ºC Joyo Experi mental DBE and Site evaluation events - CDA evaluation by the Bethe-Tait model Reactor Initial Criticality in 1977 Power :50MWt 100MWt 140MWt (Mk-III Core) Temperature : 435ºC 500ºC 500ºC IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

4 JSFR Innovative Reactor Technologies For Economic Competitiveness Reduction of plant materials and reactor building volume 1) Shortened piping with high chromium steel 2) Two-loop cooling system 3) Integrated pump-ihx component 4) Compact reactor vessel 5) Simplified fuel handling system 6) CV with steel plate reinforced concrete building Long operation with high burnup to reduce the fuel/ operational cost 7) Advanced fuel material Secondary Pump SG Integrated Pump-IHX Reactor Vessel 1,500 MWe Large Scale FBR For Improved Reliability Sodium Technology 1) Sodium leak tightness with double wall piping 2) Higher reliable SG with double wall tube 3) Design accommodation with In-service inspection requirements with innovative technologies For Enhanced Safety Core Safety 1) Passive shutdown and decay heat removal 2) In Vessel Retention with Re-criticality free core Seismic Reliability 3) Integrity of fuel subassemblies during Earthquake IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

5 Development Targets of JSFR Safety and Reliability SR-1 Ensuring safety equal to future LWR cycle SR-2 Ensuring reliability equal to future LWR cycle Sustainability Environment Protection EP-1 Radioactive influence through normal operation no more than future LWR cycle EP-2 Emission control of environment transfer substances Waste Management WM-1 Reduction of an amount of radioactive waste compared with future LWR cycle WM-2 Improvement of waste manageability equal to or more than future LWR cycle WM-3 Reduction of radio-toxicity compared with future LWR cycle Efficient Utilization of Nuclear Fuel Resources UR-1 Breeding ratio to enable transition from LWR cycle to FBR cycle & its flexibility Economical Competitiveness EC-1 Electric generation cost which can match other power plants Nuclear Non-Proliferation NP-1 Adoption of institutional measures and application of technical features NP-2 System design of physical protection and its development IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

6 JSFR Design Requirements on Safety Ensuring a safety level equivalent to future LWRs & related cycle facilities Ensuring design margin for Design Basis Events (DBEs) Adopting passive safety features and consideration of Accident Management Procedure against Beyond DBEs not entering into CDA Development of Re-criticality free core* against Core Disruptive Accidents (CDA) Assurance of In-Vessel Retention (IVR) against thermal consequences Ensuring reliability level equivalent to future LWRs & related cycle facilities Easy accessibility to maintain and repair components and structures under sodium * To avoid severe energetics during CDA IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

7 Event Categorization & Design Principles (1) Prevention of abnormal operation & failures (2) Control of abnormal operation & detection of failures (3) Control of accidents within design basis (4) Control of severe plant conditions, including prevention of accident progression & mitigation of severe accident consequences Rational design margin Quality assurance Preventive maintenance (Inspection, On-line monitoring, and so on) Unreliabilit y Reactivity Control RSS [Reactor Shutdown System] Heat Removal DHRS [Decay-Heat Removal System] Containment For DBE 10-2 /d 10-4 /d 10-6 /d Primary RSS Backup RSS Prevention For BDBE 10-1 ~10-2 /d Passive RSS SASS [Self-Actuated Shutdown System] > Coolant retention By Guard Vessel & Guard Pipes > Redundant & diverse passive operation AM [Accident Management] Mitigation For BDBE 10-1 ~10-2 /d Re-criticality free + In-vessel debris cooling IVR In-Vessel Retention At typical CDAs (e.g. ULOF, UTOP) Pressure-resistant & leak-tight Containment Vessel Against chemical potential of sodium Sodium leak leak-tight guard vessel & pipes (Double Boundary) SG tube leak double-wall tube, early detection & rapid depression of steam-water side IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

8 Design Approaches related to Safety Reactor Shutdown System [RSS] Two independent RSS with Flexible Joint & Restraint Core Heat Transport System [HTS] Short-piped Two-Loop System for Primary Main Pipings Design measures against Pipe Break Concepts & Countermeasures against CDA Passive Safety Features [PSF] for CDA Prevention: Self-Actuated Shutdown System Natural Circulation Decay Heat Removal System In-Vessel Retention [IVR] for CDA Mitigation: Core/Fuel Characteristics & FAIDUS for Re-criticality free Design for In-vessel debris cooling Others Design measures against Sodium Leakage in CV Design measures against Sodium-Water Reaction in SG IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

9 Reactor Shutdown System [RSS] Two independent rapid RSS [Primary & Backup] Each capable of settling subcritical state in DBEs Diversities Latch/de-latch mechanisms & Insertion force Primary RSS: insertion by gas pressure Backup RSS : electromagnets for de-latch, and inserted by gravity Detection signal 1 st signal assigned to Primary RSS, 2 nd signal assigned to Backup RSS for abnormal transients & accidents 3 rd signal assigned for Primary RSS for abnormal transients Assurance of absorber insertion Flexible joint at connecting part, for enhanced insertion capability Robust restraint against seismic requirements Regular testing on logic circuits and rod insertion, etc IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

10 RSS: Flexible Joint & Restraint Core Flexible joint De-latching under deformation condition Restraint core Supported by core barrel under deformation condition Enhanced insertion capability under core & UIS displaced conditions in earthquake Normal condition Flexible joint Control rod is de-latched easily. Deformation condition Restraint core Reactor core is surrounded by core barrel. Deformation condition Normal condition Fuel subassembly SASS Flexible joint Active core Support Pads Core barrel Control rod Upper Core Core support structure Reactor vessel IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

11 HTS: Short-piped Two-Loop System Technical features on two-loop system Large pipe diameter Small # of loops Increase of coolant flow rate per loop Increase of impact on assuring core coolability DHRS with high reliability Enlargement of pump / heat exchangers; IHX and SG Higher coolant velocity in pipe Flowinduced vibration in pipe Experimental study (underway) Impact on in-vessel fluid dynamics Optimization of fluid dynamics in upper plenum Experimental study (underway) Pump seizure accident Decrease of core pressure loss, higher resistance against inverse flow, etc. Confirmation by safety evaluation Pipe break Provision of guard pipe Confirmation by safety evaluation Heat removal by NC with design margin 2PRACS+ DRACS 2 by 2 damper, non-safety class blowers Confirmation by PSA Compact design by use of high Cr steel; Fabricability, and Function Confirmation Confirmation by; Trial Manufacturing for key structures, and Function testing, e.g. vibration test and T/H test (underway) IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

12 HTS: Design Accommodations Reduction in core flow rate by primary pump failure Large influence on short-term core cooling Design accommodations to cope with this accident Pump trip delay time Coast down time Response time of scram signal Higher resistance to prevent reverse flow Restricted pressure loss at core IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

13 HTS: Measure against Pipe Break Double-walled vessel and pipes (i.e. guard vessel and guard pipe) No penetration in RV and GV Coolant positioned above Na surface in RV Pressure: Secondary > Primary systems R&D progresses on LBB technology for High-Cr steel Improved Sodium leak detector Improved Inspection and Repair technology Guard Piping Reactor Vessel Sodium Leak Detectors Bellows and Partition Structure Intermediate Heat Exchanger Guard Vessel Nitrogen Atmosphere Double wall IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

14 DBE: Loss-of-Offsite Power accident Abnormal transient category 100% DRACS and 200% PRACSs Accident category 150% PRACSs (1 damper failure assumed) Capability of Natural circulation DHRS effectiveness confirmed IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

15 DBE: Primary Pump Seizure accident (a) Activation of primary RSS by the signal of low primary pump speed, i.e., 80% normal speed (b) Activation of secondary RSS by the signal of low primary flow rate, i.e., 50% rated flow Applicable design envelop on RSS activation signals for sufficient safety margins IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

16 Design Measures against CDA CDA Prevention CDA Mitigation Event Initiation (0) Just After Initiation CDA phases Key points for Prevention Passive-featured RSS Passive-featured DHD Key points for Mitigation Design Measures in JSFR SASS PRACS & DRACS Design Measures in JSFR In-Vessel Retention Concept (1) Initiating Phase (2) Early-Discharge Phase (3) Material Relocation & Decay Heat Removal Phases Restriction of Sodium Void Worth Active core length Enhancement of Early Fuel Discharge Core [Debris] catcher Space for quenching and stable cooling Core & Fuel Configurations Re-criticality free core F/A with Inner DUct Structure In-vessel debris cooling In-vessel core catcher Space for molten fuel relocation and cooling IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

17 Concepts & Design Measures against CDA Concept: Prevention: Passive Safety Features [PSF] Both Shutdown & Cooling Mitigation: Re-Criticality Free Core to ensure In-Vessel Retention Prevention of excursion due to sodium voiding and/or molten pool compaction Assurance of long term cooling System & Component level: SASS [Self-Actuated Shutdown System] 3 rd shutdown system + Robust core support structure Natural Circulation Decay Heat Removal Systems Loop system ensures sufficient core flow to prevent fuel failures Accommodation of Core Design i.e., Sodium void worth, Active Core Height, and so on FAIDUS [Fuel Assembly with Inner Duct Structure] Enhancement of early fuel discharge not to leading to whole core pool In-vessel Core Catcher Material relocation of molten fuel, quenching the molten fuel in the sodium pool, and long term stable cooling of debris at the multi layer core catcher IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

18 PSS: SASS (Self-Actuated Shutdown System) Failure of Control Rod insertion Scram failure Failure of Reactor Protection System Requirements to Passive RSS Effective for all types of ATWS Minimum repercussion on core & plant design Robust Robust restraint restraint core core supporting supporting configuration configuration against against seismic seismic Flexible Flexible joints joints for for enhanced enhanced insertion insertion capability capability Passive Passive de-latch de-latch and and gravitational gravitational insertion insertion of of Control Control Rods Rods Large negative reactivity insertion with quick response Less uncertainty by experimental demonstration Ease of maintenance [e.g. function testing during regular inspection] Curie point electromagnet SASS [Self-Actuated Shutdown System] IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

19 PSS: SASS Mechanisms The 3 rd RSS with Negative reactivity insertion around 1$ Core outlet coolant Temp. increase near CR Sensing alloy temperature reaching the Curie point, ca 680deg.C At the same time, the maximum clad temperature in the core reaches higher temperature Passive de-latch due to decreasing magnetic force Passive insertion of absorber rods by gravity IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

20 BDBE Transient Analysis 2.0 Coolant limit Detachment temp.: 670oC ULOHS UTOP Coolant limit Detachment temp.: 680oC Time (second) Detachment temp.: 680oC IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June Cladding and coolant temperature( ) Reactivity ($) Coolant limit 0.0 Cladding and coolant temperature ( ) 2.0 Reactivity ($) SASS detachment condition The detachment temperature : 670oC-680oC 0 Reactivity Cladding temperature Coolant temperature Cladding and coolant temperature ( ) Safety criteria Areal fuel melt fraction < 30% Core coolant temperature < 1020oC (Boiling point of the coolant) ULOF 4.0 Reactivity ($) Analyzed ATWS using Super-COPD code Unprotected Loss-Of-Flow : ULOF Unprotected Transient Overpower : UTOP Unprotected Loss-Of-Heat-Sink : ULOHS

21 PSS: DHRS [Decay Heat Removal System] Passive System by Natural Circulation with sufficient elevation difference between heat source & heat sink, without pumps, blower or motor Simplified System Configuration Air Cooler Redundant Dampers No Pump No Blower * PRACS: 2 units No Pump Air Cooler No Pony Motor No Blower DRACS: 1 unit DHX PHX Primary Cooling System Secondary Cooling System IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

22 IVR: Core & Fuel Characteristics Parameters related to CDA sequence Sodium void reactivity Less than around 6$ including uncertainty Core Height Less than around 1m Specific Power High enough for milder power sequence in transient Fuel smear density High failure threshold with annular pellet Cut view of reactor core Core region Legend Inner core Outer core Radial blanket SUS shield Zr-H shield Control rod IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

23 IVR: FAIDUS [Fuel Assembly with Inner Duct Structure] Re-criticality free core features characteristic to avoid severe energetics due to excursion in CDA sequences Core melting Early fuel discharge FAIDUS For Enhancing Molten-Fuel Discharge Wrapper tube Molten fuel No fuel discharge Grid spacer UAB Core LAB Avoid large scale fuel compaction Compacting motion Inner duct Support for inner duct No large power excursion Possibility of large power excursion by re-criticality FAIDUS* Cross section of sub-assembly Modified FAIDUS IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

24 IVR: In-Vessel Debris Cooling System (Example) Protection against direct melt jet attack Reactor core Multi-layered Debris Tray for debris retention within limit bed height of cooling and sub-critical state core support structure Enlargement of Coolant Inventory Guide Tubes for fuel debris settling to lower plate for molten fuel quenching and fragmentation into small particles Chimney for effective coolant circulation IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

25 Measures Against Sodium Leakage Containment System = Double Boundary Configurations Guard Vessels & Guard Pipes for Primary & Secondary heat transport systems in Containment Vessel, including in-vessel cold leg piping Primary piping Guard pipe Cover Gas line Containment Boundary Steam Generator Steam Secondary piping Air conditioning line Water IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

26 Measures Against Na-Water Reaction Steam Generator Mechanically-bonded Double-wall Tube Crack propagation halt, Better Heat transfer performance, Easier Inspection Straight Tubes Easier Inspection, High reliability (without tube-to-tube welding) Steam Inner tube Outer tube Interface Steam outlet Na Na inlet Mechanically-bonded double-wall tube Na outlet Water inlet IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

27 Example of JSFR Safety R&D For Safety Evaluation DBE/BDBE Transient & Accident Analysis Code SASS Magnetic Irradiation Test in Joyo (Experimental Reactor) Fully Passive Natural Circulation DHRS For CDA CDA Analysis Code EAGLE tests at Impulse Graphite Reactor (IGR), Kazakhstan For PSA Level-1 PSA, including Seismic PSA Level-2 PSA on Source term & Sodium-Fire For Safety and Technical Standardization R&D for Safety & Technical Standards Establishments IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

28 SASS Magnetic Irradiation Test in Joyo Magnetic characteristic by online measurement Test condition Test temperature 600 C Targets of neutron fluence Temperature sensitive alloy : 5x10 17 n/cm 2 Iron core : 6x10 18 n/cm 2 Test specimen B-H ring temperature sensing alloy iron core Φ 45 mm 3 mm Key parameters Maximum magnetic flux density : Bm Residual magnetic flux density : Br Coercivity : Hc B(T) Beginning of irradiation End of irradiation Irradiation Test in Experimental FR Joyo Result : 30Ni-32Co-Fe in 600 C Br Hc Bm H(A/m) Magnetic characteristic in 600 C is not affected by irradiation IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

29 Fully Passive Natural Circulation DHRS Water & Na Experiments 3D-calculation of Loss of offsite power 1D model 1/10 scaled water test 1/5 scaled partial sodium test 9.00E E E+02 温度( ) コールドレグ配管 ホットレグ配管 6.00E+02 統合コード 5.00E+02 DHX PHX FAST PHX 3D model 4.00E+02 ２次冷却材 出口 ２次冷却材 出口 3.00E E E E E E E E+03 時間(秒) 有効伝熱部 有効伝熱部 IHX IHX 炉容器 ２次冷却材 入口 ２次冷却材 入口 炉心ヒータ 下部プレナム 1/10 scale water test facility Temperature IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June 2010 Flow velocity 29

30 EAGLE tests on FAIDUS & Scenario Sequence In-pile downward fuel discharge tests IGR in Kazakhstan EAGLE out-of-pile facility tests Initial state Fuel melting Duct failure Fuel discharge Experiments successfully simulated molten fuel discharge through steel duct filled with sodium IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June

31 Concluding Remarks JSFR features: Innovative technologies for reactor & core safety, with achieving economical competitiveness Safety Concepts/Systems/Components are applied in conformity to event categorization and design principles In safety design work, especial concerns on: Reactor Shutdown System, Heat Transport System, Countermeasures against CDA Safety-related R&D are in progress for: Realizing safety features to meet the design principle, e.g. SASS, Natural Circulation, CDA scenario, Analysis Codes, and PSA IAEA-GIF Joint Workshop on Safety Aspects of SFR, Vienna, Autriche, June