Activities for Safety Assessment of Fast Spectrum Systems A. Seubert Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh Forschungszentrum, D-85748 Garching, Germany 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors IAEA HQ, Wien, 23-24 June 2015 Work supported by German Federal Ministry of Economic Affairs and Energy due to a resolution of the German Bundestag.
Content R&D for the safety assessment of liquid metal cooled fast spectrum systems Thermal hydraulics of liquid metals (ATHLET) Simulation of time-dependent distributed neutron sources (PARCS) Modeling of radial and axial core thermal expansion (FEM/PARCS) Simulation of spallation neutron sources (MCNPX) MAXSIMA (MYRRHA) ESNII+ (ASTRID) neutronic and thermal-hydraulic simulation 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 2
Generic ASTRID Core Design Radial core layout: Axial core layout: 301cm 248cm 210cm 166cm 146cm 121cm 91cm Flat-to-flat assembly pitch at nominal operation: 17,611 cm 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 3
ASTRID Neutronic Modeling with HELIOS HELIOS 1.12 190 energy group library, unadjusted 112 energy group library 8 energy group structure: Energy group index Lower limit (ev) Energy group index Lower limit (ev) 1 2.2313E+6 5 4.0868E+4 2 8.2085E+5 6 1.5034E+4 3 3.0197E+5 7 7.4852E+2 4 1.1109E+5 8 1.0000E-4 For selected comparative test calculations Serpent v2.1.21 JEFF-3.1 nuclear data 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 4
HELIOS Models (examples) Fissile/fertile assemblies: CSD/DSD assemblies: 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 5
p''' in W/cm^3 Evaluation of Core Safety Parameters (HELIOS/PARCS/FEM) 2000 1500 Reactivities (pcm) for 9 voiding scenarios 350 Radially integrated axial power density 1000 300 500 250 0-500 -1000-1500 -2000-2500 1 2 3 4 5 6 7 8 9 Institute A, Code 1 Institute A, Code 2 Institute B GRS HELIOS, FEM Institute B Institute C, Code 1 Institute C, Code 2 200 150 100 PARCS v32m16co 50 FEM-Diff-3D 0 90 110 130 150 170 190 210 230 Axial elevation (cm) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 6
Evaluation of Core Safety Parameters Radial power distributions 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 7
ASTRID Thermal-Hydraulic Modeling with ATHLET 3.0b Three single SAs Inner hot SA CORE-IH Inner average SA CORE-I1 Outer SA CORE-O1 Mass flow controlled by common fill FILL-IN entering DIAGRID Mass flow distribution in SAs according to individual flow loss coefficients Pressure controlled by time dependent volume BOUND SAs Geometric data, mass flows, diagrid porosity etc. according to ESNII+ specifications One representative heated rod group for each SA pipe One unheated rod group for each SA pipe in lower section (-IH, -I1, -O1) (only structure of cladding) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 8
ASTRID Steady State TH Simulation with ATHLET 3.0b 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 9
Diagrid Thermal Expansion Modeling In an SFR, increasing coolant inlet temperature causes thermal expansion of the diagrid plate, i.e. the assembly core support structure. V non-deformed state state with expanded diagrid Enlarged spacing between adjacent subassembly wrappers spacing is filled by coolant. Associated reactivity changes may present a large (negative) contribution to the total reactivity feedback. Assembly pitch changes affect the radial spatial meshing of the core simulator. Aim: Treat pitch changes with the core simulator s fixed radial meshing. 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 10
Diagrid Thermal Expansion Modeling Mapping the meshing of the radially deformed core (x, y, z) to the meshing of the non-deformed core (,, z): (K. Azekura, T. Hayase, J. Nucl. Sci. Techn. 26 (1989) 374) Diffusion equation of the non-deformed core: Approximate diffusion equation of the deformed core: D g r 2 x 2 g 2 D g r 2 y 2 g 2 + σ g r r g r = q g r q g r = r g r s σ gg D g r g r + σ g r r g r = q g r Change of cross sections to account for enlarged inter-assembly gap g g + g k g g Different multiplication factor σ f g r g r 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 11
Diagrid Thermal Expansion Modeling p(t) Thermal expansion correlation for diagrid Subassembly pitch thermal expansion: p T = p 20 C 1 + ε SS316 T Thermal expansion of SS316 ε SS316 T = a T 20 C + b T 20 C 2 +c T 20 C 3 with coefficients a, b, c given within ESNII+ project WP6 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 12
Multiplication factor Results for Diagrid Expansion of the ASTRID Generic Design Implementation: Fortran module Made available first to an FEM few-group diffusion code Application to ASTRID generic design within ESNII+ project Ref.: A. SEUBERT ET AL., ANNUAL MEETING ON NUCLEAR TECHNOLOGY, BERLIN, GERMANY, 2015 Could be also implemented in PARCS in future 1,005 1,004 1,003 1,002 1,001 1,000 Exact 0,999 Approximate 0,998 300 400 500 600 700 800 900 1000 1100 Diagrid temperature (K) Diagid temperature 400 K (nom.) Flat-to-flat pitch (cm) exact k eff Reactivity change ρ (pcm) approx k eff Deviation from exact ρ (pcm) 17.611 1.00466 600 K 17.674 (+0.36%) 1.00280 185 1.00277 3 800 K 17.741 (+0.74%) 1.00089 375 1.00076 13 1000 K 17.809 (+1.1%) 0.99893 570 0.99873 20 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 13
p'''approx./p'''exact - 1 p''' in W/cm^3 Results for Diagrid Expansion of the ASTRID Generic Design Axial power density profiles in individual subassemblies 350 300 250 200 150 100 50 Axial power density profiles Outer zone close to reflector Inner zone close to center 0 90 110 130 150 170 190 210 Axial elevation (cm) 0,15% Axial power density deviations 0,10% 0,05% 0,00% -0,05% -0,10% -0,15% Outer zone close to reflector Inner zone close to center -0,20% 90 110 130 150 170 190 210 Axial elevation (cm) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 14
p'''approx./p'''exact - 1 p''' in W/cm^3 Results for Diagrid Expansion of the ASTRID Generic Design Axial power density profiles in individual subassemblies compared with nominal diagrid temperature 350 300 250 200 150 100 50 Axial power density profiles Outer zone close to reflector Inner zone close to center Outer zone at Tdiagrid nominal Inner zone at Tdiagrid nominal 0 90 110 130 150 170 190 210 Axial elevation (cm) 14,0% Axial power density deviations 12,0% 10,0% 8,0% 6,0% 4,0% 2,0% Outer zone close to reflector Inner zone close to center Outer zone at Tdiagrid nominal Inner zone at Tdiagrid nominal 0,0% -2,0% 90 110 130 150 170 190 210 Axial elevation (cm) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 15
Axial fuel/cladding expansion modeling Problem: Keep axial nodalization of the core simulator unchanged Assembly pitch unchanged, simultaneous radial pin expansion treated by cross section library parameter Not expanded (nominal state) z 4 axially expanded Expanded state mapped on nodalization of nominal state z 4 z 4 z 4 σ 4 σ 4 z 3 z 3 z 3 z 3 σ 3 σ 3 σ 3 = f σ 2, σ 3 σ 2 z 2 z 2 σ 2 z 2 z 2 Interpolation using either: Linear, quadratic, cubic neutron flux FMFD diffusion neutron flux solution z 1 z' 1 z 1 z' 1 σ 1 σ 1 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 16
Axial fuel/cladding expansion modeling Test calculations of axial expansion modeling for ASTRID Nuclear material properties fixed consistent changes in nuclide density and radial pin dimensions will be treated by a few-group nuclear cross section library parameter Linear thermal expansion according to ESNII+ specifications 1,0075 1,007 1,0065 Multiplication factor exact T/ C k r r/t 75 1,00496 275 1,00559 62 0,312 475 1,00624 127 0,321 675 1,00687 189 0,311 1,006 1,0055 1,005 1,0045 exact Linear interpolation Quadr. Interpol. (F'li as calc.) Quadr. Interpol. (F'li = F're = 0) FMFD diffusion solution (model 900 C) 0 100 200 300 400 500 600 700 800 T ( C) FMFD diffusion solution (model 900 C) delta_z = 1,0 cm, nz_min = 2, nz_max = 4 k r w.r.t. exact 1,00499 3 1,00573 14 1,00648 24 1,00718 31 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 17
MYRRHA Critical Core Modeling with SERPENT and PARCS Reflector Assembly Withdrawn Safety Rod Inserted Safety Rod Fuel Assembly (MOX) 26 different materials Hexagonal geometry Complex geometry for safety and control rod Dummy Assembly (Lead) Withdrawn Control Rod Inserted Control Rod 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 18
Reference calculation: Monte-Carlo calculation Preparation of XS for PARCS calculations JEFF31 library (continuous energy) Exact 3D geometry (No approximation) SERPENT: 3D modelling Macro XS k-eff, Power distribution Number of source neutrons per cycle: 200 000 Number of active cycles run: 2000 Number of inactive cycles run: 200 Calculated with the 3D neutron flux Collapsed into 8 energy groups for every type of material 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 19
2 methods to generate few-group cross sections for PARCS Serpent 2.1.23 with JEFF31 library (continuous energy) Method 1: 3D whole core modelling Method 2: 2D single assembly modelling Realistic 3D geometry 3D neutron flux Macro XS collapsed into 8 groups Macro XS collapsed into 8 groups infinite lattice periodic boundary condition 8 energy groups structure Lower energy limits (ev) 2.23E+06 8.21E+05 3.02E+05 1.11E+05 4.09E+04 1.50E+04 7.49E+02 1.00E-04 PARCS k-eff, Power distribution PARCS k-eff, Power distribution 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 20
K-eff Multiplication factor results (preliminary) 1,05 1,013381,01347 1,00726 1 0,97869 0,97562 0,96893 Serpent 3D PARCS Method 1 PARCS Method 2 0,95 0,95155 0,94832 0,94213 0,92811 0,92335 0,91709 0,9 0,85 0,8 Control Rods position Withdrawn Inserted Withdrawn Inserted Safety Rods position Withdrawn Withdrawn Inserted Inserted 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 21
Average power per ring (in % of the total power) Radial power distribution comparison (preliminary) 3 Radial power distribution 2,5 2 PARCS 1,5 SERPENT 1 0,5 Relative difference between PARCS (Method 1) and Serpent results for each FA (in %) 0 0 1 2 3 4 5 6 Distance from center [ring number] 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 22
Extension of PARCS for External Neutron Source Simulation 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 23 Objective: Simulation of ADS Experience from implementation in TORT-TD and validation by YALINA-Thermal Transport equation: Example for a single localized source pulse: l l l d gl g g fg g g g gg g g tot g g t r c t r r t r r d t r q t r r t v,, ) (1,,,,,,, ˆ 1 ' ' ' ' 4 ' 0 1 2 3 4 5 6 7 8 0 2 4 6 8 10 12 14 16 18 20 normalized to steady state power Time in sec Power Level
Neutron source strength (a.u.) Neutron source strength (a.u.) Neutron source strength (a.u.) Simulation of Spallation Neutron Sources MCNPX model of the MYRRHA spallation target 0,08 0,07 0,06 0,05 0,04 0,03 0,02 0,01 0,00 0,001 0,01 0,1 1 10 100 Energy (ev) 9,0E-02 8,0E-02 7,0E-02 6,0E-02 5,0E-02 4,0E-02 3,0E-02 6,0E-05 5,0E-05 4,0E-05 3,0E-05 2,0E-05 2,0E-02 1,0E-05 1,0E-02 Energy group 8 Energy group 7 0,0E+00 0,0E+00 100 110 120 130 140 Axial position (cm) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 24
Summary Activities at GRS for Safety Assessment of Fast Spectrum Systems Systems: Sodium cooled fast reactors (e.g. ASTRID) Liquid metal cooled fast spectrum systems, including source driven sub-critical systems (e.g. MYRRHA) Codes: ATHLET, PARCS, HELIOS, SERPENT, MCNPX Modeling extensions: Thermal hydraulics of liquid metals (ATHLET) Simulation of time-dependent distributed neutron sources (PARCS) Modeling of radial and axial core thermal expansion Simulation of spallation neutron sources (MCNPX) Few-group nuclear cross-section generation (HELIOS, SERPENT) Whole core neutronics modeling (PARCS, SERPENT) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, 23-24 June 2015 25