APPLICATION OF SERPENT 2 FOR SODIUM FAST REACTOR NEUTRONICS AND SAFETY ANALYSIS

Transcription

1 WIR SCHAFFEN WISSEN HEUTE FÜR MORGEN Alexander Ponomarev Paul Scherrer Institute, Switzerland APPLICATION OF SERPENT 2 FOR SODIUM FAST REACTOR NEUTRONICS AND SAFETY ANALYSIS 8 th International Serpent User Group Meeting, Espoo, Finland, May 29 June 1, 2018

2 Outline Benchmark of Superphénix reactor (SPX) SPX core specification for static neutronics Selected results on core performance Application for XS generation for SFR transient analysis with spatial kinetics 1

3 SPX Benchmark: Introduction A new neutronics and thermal-hydraulics Benchmark exercise for the large-power SFR core will be performed based on the available data for the Creys-Malville Nuclear Power Plant Superphénix (SPX1) start-up core Phase 1 Static neutronics to validate static neutronics codes evaluating the core performance including comparison with experimental data Phase 2 Transient to validate system codes simulating the start-up transients actuated during the commissioning phase Contribution to Horizon-2020 ESFR-SMART project European Sodium Fast Reactor Safety Measures Assessment and Research Tools in-kind activity 2

4 Creys-Malville Nuclear Power Plant: Superphénix reactor Largest Sodium-cooled Fast Reactor in the world ( ) Only ~4.5 years of operating (7.9 billion kwh produced) Two INES level 2 incidents: Leak of the storage drum (1987) Pollution of primary sodium (1990) NERSA consortium logo Thousands of experiments were conducted at start-up and operation 3

5 Creys-Malville Nuclear Power Plant: Superphénix reactor Parameter Unit Value Thermal / electric power MW 2990 / 1242 Average fissile / fertile fuel temperature C 1227 / 627 Primary sodium inlet / outlet temperature C 395 / 545 Fissile fuel - MOX Fissile fuel enrichment in the inner / outer subcore % 16.0 / 19.7 Total mass of plutonium in the fissile core kg 5780 Volume of the fissile core m Equivalent diameter of the fissile core m 3.70 Height of the fissile pellet stack m 1.00 Height of the lower/upper breeder blanket m 0.30 / 0.30 Height of the radial blanket fertile pellet stack m 1.60 Figure: View of the SPX reactor core: - dummy fuel sub-assemblies - fertile S/A and neutron shielding Number of subassemblies in the IC/OC - 193* / 171* Number of subassemblies in the RB - 234* Number of control rods (CSD/DSD) - 21 / 3 Subassembly pitch in the diagrid mm (*) Differs from considered start-up core configuration 4

6 SPX Benchmark: Core specification for neutronics Reproduce core characteristics at SPX1 start-up configuration Based on open literature sources Evaluation of core performance: Criticality at different states Feedback effects and coefficients The spatial flux and power distributions in-core (including comparison with data as measured with the fission chambers) CR worth curve Kinetics parameters 5

7 SPX Benchmark: Core specification for neutronics Core specification constructed: as fabricated pin, fuel SA, CSD control rods design available homogeneous/heterogeneous approaches for non-fuel zones temperature expansion laws defined fuel specification developed from scratch based on assumptions initial core criticality of about 3700 pcm at 180C reproduced (with JEFF311) criticality level at HZP and HFP reproduced CSD rods worth reproduced Distributed as Serpent 2 input deck for different core states Sources: Nuclear Science and Engineering, Vol.106 (1990) Fast Reactor Database, IAEA, ISBN , Vienna, 2006 Superphenix Benchmark used for Comparison of PNC and CEA Calculation Methods, and of JENDL-3.2 and CARNAVAL IV Nuclear Data, O-Arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation,

8 SPX Benchmark: Core specification for neutronics Inner core subassemblies (190 SAs) Outer core subassemblies (168 SAs) Radial breeder blanket subassemblies (225 SAs) Radial steel shielding subassemblies (294 SAs) Diluent steel subassemblies (18 SAs) Control rods (21 CSDs) Shutdown rods (3 DSDs) Superphenix start-up core subassemblies arrangement 7

9 SPX Benchmark: Core specification for neutronics Cross section of geometry: Fuel SA Radial Breeder Blanket SA Subassemblies axial structure in Superphénix core model Control rod absorber (CSD) 8

10 SPX Benchmark: Selected results Core criticality for different states: Reference Serpent 2 / JEFF311 solution Case ID CSD insertion* [cm] Temperature for XS / geometry [K] Fissile Fertile Other K-eff [-] Calculated Reactivity [pcm] Measured reactivity [pcm] C E [pcm] 180 C / / / C / / / HFP / / / Ref.HZP / / / (*) From the top of fissile height Isothermal Temperature Coefficient and its components Parameter Benchmark* Calculation Experiment 3.34 Isothermal Temperature Coefficient Kiso (3.18**) ( C) [pcm/ C] 2.63 ± ± 0.14 (3.10***) Expansion component k [pcm/ C] ± ± 0.15 Doppler component KD [pcm] ± ± 118 (1334***) (*) Calculations do not consider inserted CRs and differential effect of CR movement due temperature expansion of the vessel and its inner structures (**) Calculated value as sum of two individual components (***) Critical core configuration with CRs inserted by 40cm 9

11 SPX Benchmark: Selected results z [mm] Reaction rates for Ref. HZP configuration compared to measured data Blanket Fissile Blanket Control rod Experiment Calculation Normalized U-235 fission rate [-] (a) Axial profile of U-235 fission rate Normalized U-238 fission rate [-] Experiment Calculation Subassembly row number [-] (b) Radial profile of U-238 fission rate in fissile region Normalized Pu-239 fission rate [-] Experiment Calculation Subassembly row number [-] (c) Radial profile of Pu-239 fission rate in upper breeder blanket 10

12 Application for XS generation: SFR transients with spatial neutron kinetics Serpent 2, continuous energy Monte Carlo neutron transport code PARCS, deterministic neutron diffusion code TRACE, thermal-hydraulics system code 11

13 Application for XS generation: SFR transients with spatial neutron kinetics ULOF scenario: Loss of the primary pumps, no SCRAM Scenario studied with TRACE and built-in Point Kinetics model Ultimate goal: Analyze sensitivity to 3D feedback effects by coupled TRACE/PARCS calculation for ULOF analysis of SFR core Time-dependent XS using derivatives XS generation with Serpent 2 Derivatives parametrise the XS for perturbed reactor states 12

14 Application for XS generation: Serpent 2 model of SFR Define universes such that they correspond to PARCS nodes Neutron spectrum in each sub-assembly is exact 10 8 histories : 100 inactive cycles, 1000 active cycles with 10 5 neutrons 35 to 45 hours on PSI cluster, using 12 parallel nodes JEFF3.1.1 nuclear data library 13

15 Application for XS generation: New method for XS improvement (sampling) Motivation: use PARCS as surrogate model with 1-group XS Task: Generate 1-group XS with Serpent Find set of 1-group transport XS and albedos with which PARCS can reproduce as close as possible Serpent solution, in terms of k-eff 1-group scalar flux distribution New sampling method was developed for generating one-group XS s with Monte-Carlo code and adjusting transport XS s and albedos for reactor kinetics calculations with nodal diffusion code 14

16 Application for XS generation: New method for XS improvement (sampling) 15

17 Application for XS generation: New method for XS improvement (sampling) Results of PARCS calculation: k-eff, power, flux, best-fit transport cross-sections and albedo coefficient Axial and radial flux shapes are close Maximum deviation of SA power (PARCS vs Serpent) is 4.5% Perturbed states: match of Doppler constants, match of the change in reactivity for different voiding scenarios between PARCS and Serpent Derivatives calculated zone by zone using the reference state XS and perturbed state XS from Serpent, use the corrected transport XS from PARCS Linear parametrisation of axial top albedo as function of voided fraction of the whole plenum 16

18 To conclude Acknowledgments: The work has been prepared within EU Project ESFR-SMART which has received funding from the EURATOM Research and Training Programme under the Grant Agreement No Serpent 2 calculations have been performed on Cray XC40 supercomputer supported by a grant from the Swiss National Supercomputing Centre (CSCS) 17

19 Backup slides

20 New H2020 project: ESFR-SMART European Sodium Fast Reactor Safety Measures Assessment and Research Tools Budget ~10 MEUR (5 MEUR from EU) Use legacy experiments -- SFR operational data (SPX1/CEA) -- sodium boiling (KNS-37/KIT) -- molten fuel ejection (CABRI/IRSN) -- molten pool behaviour (SCARABEE/IRSN) -- aerosols in containment (FAUST/KIT) -- aerosols in containment (NALA/KIT) -- aerosols from sodium fire (FANAL/CEA) Develop new instrumentations -- eddy current flowmeter (HZDR) Coordinator-- Dr. K. Mikityuk (PSI) PSI /CH CEA /FR CIEMAT /SP Chalmers /SW EDF /FR ENEA /IT Framatome/FR GRS /DE HZDR /DE IPUL /LV IRSN/FR JRC /EU KIT /DE LEMTA /FR LGI /BE NNL /UK UCAM /UK UPM /SP WOOD/UK Calibrate and validate codes Assess new safety measures for ESFR -- low void effect core design -- corium discharge tubes -- passive decay heat removal -- large-inertia and passive pumps -- improved natural circulation, etc Conduct new experiments -- MOX fuel properties (CEA ITU) -- forced-to-natural convection (KASOLA/KIT) -- sodium boiling (KARIFA/KIT) -- chugging boiling (CHUG/PSI) -- corium jet/catcher (JOLO/LEMTA) -- corium/catcher (LIVE/KIT) -- corium jet/concrete (MOCKA/KIT) Establish new networks -- students mobility grants to work at EU Na facilities -- workshops and summer school Demonstrate new reactor concept features -- iso-breeder (produces fissile fuel for own needs) -- safer than LWRs (no core meltdown in Fukushima-like accident) -- safer than SFRs (low void effect)

21 European Sodium Fast Reactor CAD design Horizon-2020 ESFR-SMART project coordinated by PSI started on September Janos Bodi completed MS thesis on CAD models for ESFR and started his PhD on November 1, CAD drawings are design specifications used to develop code inputs.

22 Code-to-code comparison and validation on basis of experimentally obtained data Serpent 2 (PSI), MCNP (CIEMAT), ERANOS (IRSN), WIMS (WOODS) Reactivity coefficients set to be used in PK reproduce Isothermal Temperature Coefficient (from 180C to 400C) KK iiiiii = kk + KK DD /(TT ) optionally: feedback coefficients obtained experimentally for different power levels kk(tt ii ), gg ΔΔΔΔ, hh PP optionally: data for spatial kinetics solutions Source: Outcome of Phase 1 M. Vanier, Superphenix Reactivity and Feedback Coefficients, Nuclear Science and Engineering, Vol.106 (1990) List of PK reactivity coefficients (define branch calculations from Ref. HZP): Doppler constant (zone-wise), pcm Sodium expansion (zone-wise), pcm/k Axial fuel expansion (zone-wise), pcm/k Clad expansion (zone-wise), pcm/k Wrapper expansion, pcm/k Diagrid expansion, pcm/k CRDL expansion, pcm/k Strongback expansion, pcm/k Vessel wall expansion, pcm/k ddρρ CCCC = kk ddtt ii + gg dd TT + h PP

23 Core and primary circuit specification for thermal hydraulics Channel approach for the core (multi-1d, up to 1 channel/sa) Unique boundary conditions for every transient inlet temperature, primary sodium flow, outlet pressure, vessel wall temperature available for considered transients Open sensitive issues: SA flow rates, fuel-clad gap, assumptions on thermal expansion of diagrid, strongback, vessel wall TRACE (PSI), SIM-SFR (KIT), CATHARE (ENEA), ASTEC-Na (IRSN), home-made neutronics/thermal-hydraulics tool (CIEMAT) Approach has been applied in TRACE and SIM-SFR Source: K. Mikityuk and M. Schikorr, New Transient Analysis of the Superphénix start-up Tests, Proceedings of International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), Paris, France, 4-7 March 2013 Fissile fuel region CR drive diagrid P=f(t) Pipe Upper plenum Fissile fuel rod Lower plenum Pipe vessel strongback T=f(t) & m& =f(t) Fertile fuel region Fertile fuel rod TRACE model of SPX1 primary system

24 Selected transients Validation of modelling of operational transients with focus on thermal expansion effects 6 operational transients considered -74 pcm stepwise reactivity insertion at 1542 MWth -50 pcm reactivity insertion at 692 MWth +30 pcm reactivity insertion at hot zero power (HZP) -10% primary mass flowrate reduction at 1415 MWth -10% primary mass flowrate reduction at 666 MWth +10% secondary mass flowrate increase at 692 MWth optionally: reproduce (in transient?) feedback coefficients kk(tt ii ), gg ΔΔΔΔ, hh PP obtained experimentally in range of operation powers 0-80% Source: Nuclear Science and Engineering, Vol.106 (1990)

25 Selected results List of PK reactivity coefficients which defines branch calculations from HZP core state: Fuel Doppler constant (zone-wise), pcm Sodium expansion (zone-wise), pcm/c Axial fuel expansion (zone-wise), pcm/c Clad expansion (zone-wise), pcm/c Hexcan expansion(zone-wise), pcm/c Diagrid expansion, pcm/c CRs position related, pcm/cm Strongback (and diagrid) axial expansion, pcm/c Vessel wall expansion, pcm/c

26 Selected results Doppler constant [pcm] Effect as result of fuel isotopes temperature rise from 600K to 1500K Zone-wise: IF Fiss, OF Fiss, LAB, UAB, RB + Total

27 Selected results Sodium density coefficient [pcm/c]: As result of pin bundle sodium heat-up by 400C (up to 800C) Zone-wise: 3 radial zones (total height from bottom of fuel column): IF, OF, RB Fissile core: IC+OC-Fiss Optionally: IF-Fiss, OF-Fiss, ILAB, OLAB, IUAB+Above, OUAB+Above Optionally: SVE [pcm] for all zones + global (not to be used in transients)

28 Selected results Axial fuel expansion coefficient [pcm/c]: As result of fuel pellet stack elongation by 1% Zone-wise: IF, OC Optionally: Contributions with CRs inserted, IF-Fiss, IF-OC

29 Selected results Cladding expansion coefficient (axial and radial) [pcm/c]: As result of cladding temperature increase by 490C (+1% in dimensions) Increase of pin outer diameter, accounts for ejection of sodium (~3.3% ) out of the pin bundle Zone-wise: IF, OF Assumption: perturbation only within fuel height, thus NO other dimensions change (height and axial positions of LGP and UGP, total pin length, ) Steel Doppler: strong effect!

30 Selected results Hexcan expansion coefficient (axial and radial) [pcm/c]: As result of hexcan temperature increase by 490C (+1% in dimensions) Zone-wise: IF, OF Accounts for ejection of sodium (~0.5% ) due increase of hexcan volume in the core Assumption: perturbation only within fuel height, thus NO other dimensions change (height and axial positions of fuel column, height of sodium plenum..) Steel Doppler: strong effect!

31 Selected results Diagrid radial expansion coefficient [pcm/c]: As result of diagrid temperature increase by 489C (+1% in subassembly pitch in diagrid) Accounts increase of sodium volume in the core 18