SFR CORE DESIGN PERFORMANCE AND SAFETY

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1 SFR CORE DESIGN PERFORMANCE AND SAFETY A. VASILE European Nuclear Education Network Association Gen IV - INSTN Alfredo Vasile 19 SEPTEMBER SEPTEMBRE 2012 CEA 10 AVRIL 2012 PAGE 1

2 OUTLINE GEN IV specifications Overview of the Design Methodology SFR core design: example of results, key issues Application to ASTRID Conclusions 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 2

3 GEN IV CRITERIA 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 3

4 GOALS FOR GEN IV FUEL CYCLES Optimization of the use of natural resources Waste minimization Proliferation resistance Recycling options: All paths should be kept available, they could be used in a sequence. DepU DepU DepU R T FP MA R T FP R T FP MA U Pu U Pu U Pu MA U & Pu recycling Heterogeneous MAs recycling Homogeneous MAs recycling (GenIV) 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 4

5 SFR CORE LAYOUT 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 5

SFR CORE SUBASSEMBLIES Central sub-assemblies Fuel sub-assemblies (containing the fissile material), Blanket sub-assemblies & reflector sub-assemblies, Sub-assemblies supported by the diagrid,

6 SFR CORE SUBASSEMBLIES Central sub-assemblies Fuel sub-assemblies (containing the fissile material), Blanket sub-assemblies & reflector sub-assemblies, Sub-assemblies supported by the diagrid, Peripheral sub-assemblies Control rods Lateral neutron shielding, Sub-assemblies on a dummy diagrid, Function: optimize the neutron balance, Material: enriched boron with absorbing 10 B, Spread out in the fissile zone. 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 6

7 SFR CORE DESIGN REQUIREMENTS Main core functions: Producing & controlling the power level, Ensuring the fissile material filling density, Guaranteeing & controlling the temperature level, Stabilising and maintaining overall core consistency. Design issues related to these functions: Neutronics Power density, safety parameters Thermal-hydraulics Cooling and monitoring, Thermomechanics (& materials) Stability and consistency, Fuel physicochemistry Temperature stability, Instrumentation Monitoring. 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 7

8 SFR CORE DESIGN PROCESS Pre-design S/A Design Design Core definition 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 8

SFR CORE DESIGN PROCESS Simplified modeling Detailed modeling Preliminary physical studies Pre design studies Design studies Neutronics Physical behavior Way of interest Neutronics

9 SFR CORE DESIGN PROCESS Simplified modeling Detailed modeling Preliminary physical studies Pre design studies Design studies Neutronics Physical behavior Way of interest Neutronics Pre-design Feasibility domain Neutronics Thermo Hydraulic Mechanic Fuel Behavior Materials Severe accidents assessments S/A Pre design Core detailed image 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 9

10 SFR CORE DESIGN PROCESS 1st phase: parametric study Neutronic assessment Study of the physical phenomena Most important parameters to improve Safety (Doppler, Void, ) Performances (BU,..) Elementary physical studies Neutronic 2nd phase: images Compilation of most promising options from the 1 st phase Core images focusing on: safety, looking for cores with limited void effect self sustainability (BG, Pu Inventory) Economic performances (Cycle length, BU,..) Pre design studies Neutronic Pre-design 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 10

11 SUBASSEMBLY DESIGN PROCESS Neutronics Spatial Power Distribution within the S/A and neutronic performances Fuel Element Fuel and Clad Temperatures Pressurization Elementary Design S/A Geometrical Design New Volume Fractions Thermal hydraulic behavior The S/A design need a multi-discipline process : Neutronic: irradiated fuel characteristics, neutronic performances (depletion, power distribution ) Mechanics: pressurization Thermal hydraulics: fuel and pin temperatures Iterative design process involving the multidiscipline criteria New methodologies are being developed using multicriteria optimization approach NO criteria YES S/A candidate 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 11

12 NEUTRONIC PRE DESIGN STUDIES Fast implementation and reduced time calculations Assessment of global parameters Homogeneous cells calculations without self shielding Pu content, BG, mass of HN Cycle length, BU calculations 2D RZ core geometry, homogeneous compositions DPA, Max and average power Core safety parameters : Void effect, Doppler, 33 group for flux calculation in diffusion approximation Delayed Neutrons fraction. Calculations are performed without uncertainties at this stage 13 SEPTEMBRE 2012 PAGE 12 CEA 19 SEPTEMBER 2012

CONCEPTUAL DESIGN Detailed core study (from a pre-design core definition) Detailed core description Implementation of control rods and backup systems Overall optimization Transfer data

13 CONCEPTUAL DESIGN Detailed core study (from a pre-design core definition) Detailed core description Implementation of control rods and backup systems Overall optimization Transfer data to other codes Thermal hydraulics Fuel behavior mechanics Neutronics System Transients & Severe Accidents Neutronics Reference Code (MC) 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 13

SFR CORE DESIGN STUDIES Most accurate calculations and time ratio Specific calculation scheme Heterogeneous cells calculations Pin and S/A description with fine self shielding treatment Detailed core

14 SFR CORE DESIGN STUDIES Most accurate calculations and time ratio Specific calculation scheme Heterogeneous cells calculations Pin and S/A description with fine self shielding treatment Detailed core description (3D Hex Z) Loading Batch Management Detailed characterization Global performances Reaction rates distribution per S/A and meshes Feedback coefficients per S/A and meshes Individual control rod worth and detailed monitoring of safety criteria (10$,cold and hot shutdown, handling error,,..) 33 group for flux calculation in transport theory Uncertainties depends on the qualification domain 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 14

15 DESIGN METHODOLOGY APPLIED TO SFR SYSTEMS ICAPP'12, Chicago, USA 24,28 june 2012 PAGE 15

16 EXEMPLE OF PARAMETRIC STUDIES Ways to improve the breeding gain Ways to improve the Void/Doppler ratio Geometry modifications Increase of the fuel volume fraction Decrease of the sodium volume fraction Increase of the height / diameter ratio (0<H/D<1) Decrease of the core volume Increase of the core volume Decrease of the height / diameter ratio (0<H/D<1) Increase of the sodium volume fraction Increase of the fuel volume fraction Increase of the core radius Modification of the core radius «options» addition Minor actinides addition Sodium plenum addition Sodium plenum addition Addition of CaH 2 moderator Annular core Addition of B 4 C moderator Addition of B 4 C moderator Addition of an internal fertile axial blanket Addition of an internal fertile axial blanket Annular core Addition of CaH 2 moderator Minor actinides addition Fuel type Nitride Carbide Carbide Oxide Metallic Nitride Oxide Metallic Large deterioration Large improvement No Effect 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 16

17 SFR CORE DESIGN PROCESS Key parameters Loss of reactivity per cycle Coolant void effect Doppler Core pressure drop Pu enrichment Detection Safety oriented optimization Decrease Decrease Optimization Decrease or increase Decrease Decrease + early detection system and high performance core monitoring Major expected benefits Decrease Control Rods efficiency so as to reduce consequences of a sudden withdrawal of rod (s) accident (UTOP) To minimize energy release potentially leading to severe accident Favorable feedback in ULOF transient Safety margin in case of gas ingress in the core Favorable feedback in ULOF Reactivity insertion accident (sudden rod withdrawal) (reactivity insertion) To favor natural convection Favorable natural behavior in case of ULOF (LIPOSO) To limit consequences of core meltdown accident? Fusion limited to few fuel assemblies (7 being the criteria in the past) Fast delayed n detection in case of instantaneous total blockage (IBT) Acoustic detection of local boiling (IBT) Enhanced detection system for Sudden rod (s) withdrawal accident Enhanced detection of fuel handling mistakes 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 17

SFR / PWR Reactor PWR SFR SFR S/A Thermal power (MW) 4250 3600 Electrical power (MW) 1450 1450 Fuel UOX (~4%), MOX (~8%) MOX (~20%) Coolant water

2 PWR S/A Bundle geometry square hexagonal Pins per S/A 264 331 Fissile hight (cm) 420 100 Number of S/A in the core 241 387 Core volume (m 3 ) 46 12

18 SFR / PWR Reactor PWR SFR SFR S/A Thermal power (MW) Electrical power (MW) Fuel UOX (~4%), MOX (~8%) MOX (~20%) Coolant water sodium Primary pressure (b) Cladding diameter (mm) PWR S/A Bundle geometry square hexagonal Pins per S/A Fissile hight (cm) Number of S/A in the core Core volume (m 3 ) Uranium, Plutonium mass (t) Core inlet temperature ( C) Core T ( C) Power density (W/cm 3 ) Reactivity control Control rods (~80), Soluble Boron Control rods (~30) 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 18

19 SFR SUBASSEMBLY Axial blanket bundle Wrapper tube Fuel bundle Sodium inlet 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 19

20 PHENIX, SUPER PHENIX AND EFR Reactor PHENIX SUPER PHENIX EFR Thermal power (MW) Pellet diameter (mm) (central hole 2 mm) 6.94 (central hole 2 mm) Cladding diameter (mm) Pins per S/A Fissile hight (cm) Blanket zone hight (radial/upper/lower) (cm) 52/22/30 60/30/30 40/15/25 S/A width across flats (cm) S/A pitch (cm) Equivalent core fissile radius (cm) Core fissile volume (m 3 ) Number of fuel S/As Zone Enrichment 1 Zone Enrichment 2 Zone Enrichment 3 Number of control rods Number of safety control rods Number of blanket S/As (2 rings) 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE (3 rings) (1 rings)

21 REACTIVITY EFFECTS - INTRINSIC dρ DOP = K DOP dt T comb. α comb. 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 21

22 REACTIVITY EFFECTS - LOCAL 2) LOCAL EXPANSION Sodium density Cladding and wrapper tube radial Cladding and wrapper tube axial Fuel axial (free or linked to the cladding) Remarks: d ρ = k. dt i i i No radial fuel expansion effect because fuel radial expansion is limited by the cladding When the fuel is linked to the cladding the axial expansion is driven by the cladding axial expansion. 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 22

23 REACTIVITY EFFECTS - GLOBAL 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 23

24 REACTIVITY EFFECTS Type of effect Effect Related to Formulation Value (SPX) pcm / C Intrinsec Doppler T fuel ( ) ρ Sodium density T sodium ) ρ Cladding radial T cladding ) ρ Cladding axial T cladding ) ρ Local Wrapper tube radial T wrapper tube ( ρ Wrapper tube axial T wrapper tube ( ρ Fuel axial (linked fuel) T cladding ) ρ Fuel axial (free fuel) T fuel ( ) ρ Global Diagrid radial T diagrid ( ) Relative expansion core/vessel/control rods T core, T control rods, T vessel ρ. ρ mm/ C 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 24

25 LOSS OF SECONDARY FLOW TRANSIENT Primary flow Secondary flow Core power Sodium and fuel temperatures 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 25

26 LOSS OF SECONDARY FLOW TRANSIENT Reactivity Doppler Sodium Core/Vessel/CR Diagrid Core inlet temp., ρ Diagrid < 0, P, ρ DOPPLER > 0, ρ Sodium > 0, ρ Fuel < 0, ρ Core/Vessel/Control rods> 0 Core power could be stabilised but isothermal sodium temperature >1000 C) 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 26

27 ADVANCED SODIUM TECHNOLOGICAL REACTOR FOR INDUSTRIAL DEMONSTRATION Industrial prototype (step before a First Of A Kind) Integrating French and international SFRs feedback A Generation IV reactor Safety : Level at least equivalent to GENIII systems With significant improvements on sodium specific issues Operability : Validate on the long term an ambitious load factor Significant improvements concerning ISI&R Ultimate waste transmutation : Continue experimentation of minor actinides transmutation, up to large scales if so decided An investment cost under control Irradiation services and testing of longer term options 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 27

MAIN FEATURES Pool type reactor : 1500 MWth Sodium cooled 3 Primary pumps, 4 IHX IHX DHX REDAN PP P P CORE S/A : UPuO 2 fuel (annular pellets) Fuel pins with spacer Hexagonal wrapper tube General

28 MAIN FEATURES Pool type reactor : 1500 MWth Sodium cooled 3 Primary pumps, 4 IHX IHX DHX REDAN PP P P CORE S/A : UPuO 2 fuel (annular pellets) Fuel pins with spacer Hexagonal wrapper tube General performances objectives for the core Average burn-up > 100 GWd/t Break-even core : breeding gain 0 (without fertile radial blanket Fuel cycle length : efpd Transmutation capabilities for MA transmutation Safety improvement 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 28

29 SAFETY PERFORMANCE TARGET Favorable natural behavior during loss of Flow transients Target criteria : no sodium boiling for a ULOSSP transient (Unprotected Loss Of Station Supply Power) Sodium void effect minimized Target criteria : Na void effect < 0 Natural behavior favorable for a complete control rod withdrawal (without detection) Target criteria : no fuel fusion Improvement of behavior in case of sub-assembly Total Instantaneous Blockage Ensure sub -criticality of 7 melted adjacent sub-assemblies Elimination of Transient of Power Compaction or gas flow are " practically eliminated" by design. 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 29

30 CFV CONCEPT ICAPP'12, Chicago, USA 24,28 june 2012 PAGE 30

31 CFV CORE LAYOUT 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 31

32 CFV CORE LAYOUT Absorber zone Plenum zone Theses Features give a negative Sodium void worth Fissile zone Blanket zone 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 32

CORE DESIGN OPTIONS: SFRV2 AND CFV LAYOUTS SFRv2 AIM1 1500 MWth

calculations were performed with CEA s reference system codes

33 CORE DESIGN OPTIONS: SFRV2 AND CFV LAYOUTS SFRv2 AIM MWth CFV V1 - AIM MWth Neutronic and thermal-hydraulic core calculations were performed with CEA s reference system codes ERANOS and CATHARE. 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 33

34 CORE PERFORMANCES ICAPP'12, Chicago, USA 24,28 june 2012 PAGE 34

35 CFV AND V2B PERFORMANCES 1/2 Core design CFV V2B 1500 Thermal Power 1500 MW Elect. Power 600 MW Fuel Residence Time 1440 EFPD 1560 EFPD High Cycle length, positive to competitiveness Fuel Cycle Length 360 EFPD 390 EFPD ρ (cycle) pcm pcm ρ (efpd) - 4,3 pcm - 2,2 pcm Batch 4 Weak reactivity loss, favorable to limit CR withdrawal Fissile zone diameter 340 cm 312 cm S/A Pitch 17,5 cm 16,8 cm Nb fuel elements C1/C2 177 / / 144 Pin diameter 9,7 mm 10,73 mm Pins/Assembly CSD/DSD Nb 12 / 6 18 / 6 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 35

36 CFV AND V2B PERFORMANCES 2/2 Core design CFV V2B 1500 Pvol (Fiss.+ int. fertile) 226 W/cm3 194 W/cm3 Pvol (Fissile)) 258 W/cm3 194 W/cm3 Fuel burnup C1/C2 105/ 69 GWd/t 76 / 67 GWd/t Pu enrichment C1/C2 23,5 / 20 % 13,9 / 17,6 % DPA max Void effect EOC -0,5 $ +5,1 $ βeff 364 pcm 373 pcm Breeding Gain -0,02-0,05 Pu inventory (HN) 4,9 t 5,3 t Max linear power rate 483 W/cm 407 W/cm Core pressure drop 2,6 b >3b Total core flow rate 7990 ks/s Inlet core temperature 400 C Outlet core temperature 550 C Criteria reached for CFV, good for safety and public acceptance Break Even Core, durability Favorable for natural convection 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 36

37 CFV AND V2B FEEDBACK COEFFICIENTS CFV SFRV2 V0 Axial clad expansion 0,061 0,064 pcm/ C Radial clad expansion 0,089 0,147 pcm/ C Sodium expansion 0,09 0,492 pcm/ C Fuel expansion -0,23-0,254 pcm/ C Favorable Sodium worth feedback coefficient for TH transient Doppler Constant Fissile zone (KD) pcm Doppler constant Fertile slab (KD) pcm Plate expansion -0,88-0,797 pcm/ C Favorable Doppler constant, upgrade TOP behavior 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 37

38 TRANSIENT BEHAVIOR Thermal hydraulic transients Control rod withdrawal ICAPP'12, Chicago, USA 24,28 june 2012 PAGE 38

39 THERMAL HYDRAULIC TRANSIENTS To investigate the core capabilities, 3 mains unprotected transients were assessed, in comparison to classical SFR core design. ULOSSP (Unprotected Loss Of Station Supply Power) : blackout transient without scram and without starting up of ultimate emergency systems or decay heat removal ULOF (Unprotected Loss of Flow): loss of primary pumps without scram, secondary pumps remaining at nominal flow. ULOHS (Unprotected Lost Of Heat Sink): Secondary pumps tripped in 100s without scram, the primary pumps still remaining in normal operation Hypothesis for comparison: Halving time for the Primary Pumps is 20s No optimization of flow rate between S/A All results presented are given for the hot S/A 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 39

40 CFV ULOSSP Boiling Temp Outlet Temp Inlet Temp Na density Doppler Power 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 40

41 CFV ULOSSP VS USUAL SFR DESIGN Boiling Temp Outlet Temp Inlet Temp Na density 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 41

42 CFV VS CLASSICAL SFR DESIGN (ULOHS) 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 42

43 SUMMARY ON THERMAL HYDRAULIC TRANSIENTS Core design CFV V2B 1500 ULOSSP 55 C of marging Na boiling ~100s ULOF Na boiling ~3500s Na boiling ~100s ULOHS Temp. of neutronic shutdown 700 C Temp. of neutronic shutdown 800 C These results are given without uncertainties and margin were evaluated for the hot S/A but no for the hot sub channel of the hot S/A 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 43

44 CONTROL ROD WITHDRAWAL P ( t ) = P [ 1+ k '. ρ ( t )][. 1+ b 0. ( t )] lin lin o i ρ Plin0, initial linear power of the considered fuel sub-assembly, k i, relative variation of linear power density per unit of inserted reactivity, ρ total reactivity worth of the control rod between its initial position in the core and the parking position at the end of the withdrawal. b0 is the relative variation of the total power of the core per unit of reactivity inserted All the CRW that can occur on the CFV core can be detected by two devoted independent systems of core detection which stop the reactor by scram. The first system prompted is the core temperature monitoring; the second is the neutron detection In case of a total control rod withdrawal (corresponding to the failure of two strong lines of defense) the outer rods do not comply with the criterion of no melting 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 44

45 CONCLUSION Sodium cooled Fast Reactors core design is an iterative and integrated process. It must accommodate different requirements mainly related to safety, fissile materials balance and economic aspects. It includes neutronics, thermal-hydraulics, mecanics, fuel behaviour for the core itself but also whole plant behavior under normal and accidental conditions. Pre-conceptual design studies of ASTRID prototype core shows significant improvements related to previous SFRs designs (SPX, EFR). 13 SEPTEMBRE 2012 CEA 19 SEPTEMBER 2012 PAGE 45

Thank you for your attention PAGE 46 CEA 10 AVRIL 2012 Commissariat à l énergie atomique et aux énergies alternatives Centre de Cadarache 13108 Saint Paul Lez Durance T.

46 Thank you for your attention PAGE 46 CEA 10 AVRIL 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 13 SEPTEMBRE 2012 Etablissement public à caractère industriel et commercial RCS Paris B

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