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RELAP5-3D Activities at the ENEA - SIMING Lab C. Parisi, E. Negrenti, M. Sepielli [UTFISST–SIMING] A. Del Nevo [UTIS-TCI] RELAP5-3D Users Seminar Marriot.

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Presentation on theme: "RELAP5-3D Activities at the ENEA - SIMING Lab C. Parisi, E. Negrenti, M. Sepielli [UTFISST–SIMING] A. Del Nevo [UTIS-TCI] RELAP5-3D Users Seminar Marriot."— Presentation transcript:

1 RELAP5-3D Activities at the ENEA - SIMING Lab C. Parisi, E. Negrenti, M. Sepielli [UTFISST–SIMING] A. Del Nevo [UTIS-TCI] RELAP5-3D Users Seminar Marriot Downtown Hotel, Salt Lake City (UT), USA 25-28 July 2011

2 Contents ENEA activities in the field of nuclear fission research The SIMING Lab Recent Activities AER DYN-002 WWER-440 3D NK Benchmark OECD/NEA KALININ-3 Benchmark Fukushima Daiichi Unit-1 Simulation Conclusions

3 ENEA & the SIMING Lab  ENEA is the Italian National Agency for New Technologies, Energy & Sustainable Economic Development  A relevant branch of ENEA actively involved in researches for fission nuclear power technology  Research Reactors (e.g., TRIGA & Tapiro)  Experimental Engineering (e.g., AP600 components testing)  Nuclear Material Characterization  Models & Simulation  Integrated Services (e.g., waste management)  Metrology & Radioprotection  Advanced Nuclear Systems (e.g., Gen IV, LFR)  The ENEA-SIMING Lab was established in 2010 for performing R&D activities for a establishing a LWR engineering simulator in the Casaccia Research Center (Rome) TRIGA Reactor (1 MW) TAPIRO Fast Neutron Source SPES-2 Vapore TH facility

4 AER DYN-002 WWER-440 3D NK Benchmark  AER (Atomic Energy Research) is an association of institutions performing researches & safety analyses on WWER technology  AER DYN-002 benchmark: asymmetric CR ejection transient with Doppler feedback in VVER-440 geometry  CR worth ~ 2β  Prompt critical transient  Fuel @ HZP BOL  Power = 1.375 KW  Homogeneus Temperature = 260 °C (553 K)  No heat exchange between FA and coolant has to be assumed  adiabatic transient  Power surge quenched by the negative Doppler feedback in the fuel  Two types of calculations possible:  Absorber defined by:  2-group diagonal albedo matrix  Equivalent XSec  calculations by R5-3D/NESTLE & PARCS  OK for PARCS, use of AER-DYN-001 equivalent XSec for R5-3D/NESTLE  Reflector BC defined by 2-group diagonal albedo matrices  OK for PARCS, use of AER-DYN-001 equivalent XSec for R5-3D/NESTLE

5 AER DYN-002 WWER-440 3D NK Benchmark  Half core simulation (180° geometry)  184 NK nodes  only FA specified  BC for Reflector specified by Albedo Matrix  3 Types of Fuel  “1”: 1.6%  “2”: 2.4%  “3”: 3.6%  Xsecs of composition #4 are for the CR absorbers  Delayed neutrons groups specified: β = 0.005 (maximize reactivity insertion)

6 AER DYN-002 WWER-440 3D NK Benchmark  Core 250 cm height  CR partially inserted (2m)  Eccentric CR ejected in 0.16 sec., speed = 12.5 m/s  Adiabatic feedback described by the following dependence: CR partially inserted To be ejected  T f,0 = 260 °C  γ = -7.228e-4 (°C) 1/2  2.0 seconds simulation to be run

7 AER DYN-002 WWER-440 3D NK Benchmark  Full core, coupled model  22 Axial layers:  Active zone of 20x12.5 cm  Bottom and Top reflector of 25 cm  Reflector modeled using AER-DYN-001 reflector XSecs  XSecs for modeling the CR absorber  TH feedback by simplified R5-3D nodalization (2 channels model) NESTLE original model Channel & HS for Ejected CR Channel & HS for the other FAs

8 AER DYN-002 WWER-440 3D NK Benchmark  Core Total Power trend  Larger power surge predicted by R5-3D/NESTLE  simplified feedback model (?)

9 AER DYN-002 WWER-440 3D NK Benchmark  TH nodalization was improved  11, 51 and full core (349+1) channels were implemented 11 channels 51 channels 349+1 channels Nodalization sensitivities Power Comparison v

10 AER DYN-002 WWER-440 3D NK Benchmark  Calculation of eigenvalue  Ejected CR in  Ejected CR out  Good prediction of Keff by PARCS code, slightly overestimation by R5-3D/NESTLE  Compared to DYN3D, they overestimate or underestimate the static reactivity of the ejected CR

11 AER DYN-002 WWER-440 3D NK Benchmark Power comparisons with independent solutions (DYN-3D code solutions, HEXNEM1 & 2) Power Trend

12 AER DYN-002 WWER-440 3D NK Benchmark  Integral Power vs. time – comparison with independent solutions  Solution obtained in the codes solutions uncertainty band Power Trend

13 OECD/NEA KALININ-3 Benchmark OECD/NEA KALININ-3 Benchmark: switching-off of one of the four operating main circulation pumps at nominal reactor power Test performed during KALININ-3 NPP (Russia) commissioning rapid rearrangement of the coolant flow through the reactor pressure vessel & coolant temperature change (spatially dependent ) Lot of data recorded & available through OECD/NEA

14 4 Excercises, of which: 3 rd ) Best-estimate coupled code plant transient modeling 4 th ) Uncertainty analysis for the purpose of PHASE-III (System Phase) of OECD Benchmark for Uncertainty Analysis in Best –Estimate Modelling (UAM) Data available: Reactor Power, Cold Leg & Hot Legs Temperatures, Pressures, etc.. + Coolant temperature at the exit of selected 96 FA + Detectors for axial profile power distribution (SPND) OECD/NEA KALININ-3 Benchmark

15  3D TH nodalization mainly developed during the V1000CT-2 phase [e.g., see Frisani, Parisi, D’Auria, “3D NK TH VVER1000 MSLB analysis by RELAP5-3D© code”, ICONE15 conf.]  Large model, very detailed: whole primary side is modeled  RPV  4 SG  PRZ  Particular care devoted to the RPV modelling  3D modelling of Downcomer (20 azimuthal sectors = 18 deg, 10 axial levels)  3D Modelling of LP, UP, Lower and Upper Core Plate (always 20 azimuthal sectors)  1D modelling for the UH  Secondary Side modeled up to the turbine  Nodalization Statistics:  Heat Structures = 6500  Mesh Points = 68340  Volumes = 8503  Junctions = 9300  NK nodes = 6752

16 OECD/NEA KALININ-3 Benchmark Primary side and SGs 4 SGs 3D TH RPV PRZ

17 OECD/NEA KALININ-3 Benchmark  Special modelling for 14 selected FA  Water for TC measurement take into account  Water flowing into the CRGT  Water entering from three small holes  Heat exchange through dedicated HS To the UP FA Head FA TH CHANNEL CRGT CRGT HS Fuel HS Flow Path at the FA exit v

18 OECD/NEA KALININ-3 Benchmark  FA versus azimuthal sectors (1-20)  TH model coupled with the 3D NK

19 OECD/NEA KALININ-3 Benchmark  Calculation expected to be performed by PSU by HELIOS Lattice Physics Code  New database was composed by 28 composition divided by 10 layer  28 x 10 = 280 Fuel Compositions  CR modeled by 110 Rodded Compositions (2 Lib. for B4C and Dy-Ti parts)  3 Compositions (# 281, 282, 283) used for Bottom, Radial and Top Reflector  ADF directly included in the Cross Section values 1/6th core symmetry 280 composition libraries

20 OECD/NEA KALININ-3 Benchmark NEMTAB & NEMTABR text files for Rodded & UnRodded Fuel Compositions OBTAINING XSEC IN NESTLE & PARCS FORMAT Cross Section reference values and variation coefficients automatically calculated by FORTRAN program NESTLECONV 3D Linear Interpolation for reference value Minimum Least Square method for variation coefficients User can select : Libraries Dimension & Reference conditions for the interpolation  Available RELAP5-3D© version does not allow to perform ONLINE XSEC interpolation  need to be performed OFFLINE + VARIATION COEFFICIENTS calculations  XSEC trend re-constructed by NESTLE using first/second degree polynoms

21 OECD/NEA KALININ-3 Benchmark  6752 NK nodes  Radial, Upper and Bottom Reflectors modeled  32 axial layers  2 for top/bottom reflector  30 for the core  Core Axial mesh of 18 cm ca. FA Type location NESTLE Composition Map

22 OECD/NEA KALININ-3 Benchmark  Different structure of the CR taken into account thanks to the NESTLE capability to model driver-part of a CR NESTLE CR Map NESTLE CR Modelling K3 CR

23 OECD/NEA KALININ-3 Benchmark  Excercise No. 3 is considered (full 3D NK TH simulations)  Null-transient run for achieving SS  checking main parameters for verifying consistency  Transient: MCP-1 off, CR #10 & #9 moving according to the Benchmark specifications  Preliminary results  qualitative check for assesing the 3D NK TH model implementation

24 OECD/NEA KALININ-3 Benchmark  3D NK Parameters  CR X in (82% ca. from the bottom)  Keff: 0.99425  OK, e.g. other solution: 0.99983 (KORSAR/GP code) 2D Power Axial Power

25 OECD/NEA KALININ-3 Benchmark  Some results for TH paramenters NPP RKT Power R5-3D CL Temperatures R5-3D RKT power NPP CL mean temperatures

26 OECD/NEA KALININ-3 Benchmark  3D NK Parameters  Radial Power perturbation (no sensible change after 60 secs.) @ 15.0 s @ 55.0 s @ 155.0 s Thermo-couple response calculation  Testing capability in reproducing thermocouple measurement at FA exit

27 Fukushima  After the Fukushima NPP accident, UTFISST SIMING Lab decided to develop a computational model of the Fukushima-Daiichi Unit 1 to perform events simulation  Unit 1 suffered the major damages (more than 14 hrs w/o cooling) & large part of core melt down to the bottom of the RPV  Work has been carried out with a collaboration between UTFISST SIMING Lab (Rome) & UTIS TCI Lab (Brasimone)  RELAP5-3D model set-up  capability to simulate accident up to the onset of clad degradation  Possibility to perform a successive Severe Accident calculation by the use of RELAP/SCDAP code

28 Fukushima Daiichi Unit 1  BWR-3 + Mark I containment  Reactor Thermal Power: 1380 MWth  In commercial operation since 1970

29 Fukushima Daiichi Unit 1  BWR-3 with 2 recirculation circuits and 20 Jet Pumps (JP)  Core Features:  400 Fuel Assemblies (FA)  97 CR blades RPV Section Core Section

30 Fukushima Daiichi Unit 1  Engineered Safety Features (ESF) in case of reactor isolation  At high pressure, Decay Energy removal by :  Safety/Relief Valve (S/RV)  steam discharge in Wet-Well (WW) & Dry Well (DW)  Isolation Condenser (IC)  High Pressure Core Injection (HPCI) system  Containment Cooling System (CCS)  In order to have the actuation of the above systems, AC/DC power, compress air, etc., is needed Unit-I ESF

31 Fukushima Daiichi Unit 1  2 IC per unit  Decay energy removal by natural circulation  Capacity of 100 m 3 of water each (secondary side at atmospheric pressure)  1-1.5 hr for a complete SS evaporation  ! DC power needed in order to keep open the valves on the RPV-IC lines

32 Sequence of Events Reconstruction

33 RELAP5-3D Modelling 1/3  RELAP5-3D TH Modelling  RPV  Steam Lines up to the turbine & SV/SRV  Recirculation circuits  Isolation Condensers  Turbine Bypass  Coupled to the containment nodalization  Nodalization statistics:  1076 Volumes  1013 Junctions  536 HS  Real time simulation


35 RELAP5-3D Modelling 3/3  400 FA modelled using 5 equivalent TH channels  5 radial zones with different radial peaking factors  Axial power distribution (imposed)

36 Model validation: Steady State  No operating parameters available for Fukushima Daiichi Unit 1  Model Validation performed using public data from Santa Maria de Garona NPP (Spain)  Similar unit: BWR-3, Mark I, 1380 MWth  Steady state validation results

37 On Transient Validation  On transient validation performed using Turbine Trip event occurred during 1992  Results are in good qualitative agreement with NPP data

38 Contents  Base case simulated (reference analysis for TEPCO & NISA)  t= 0.0 sec  EQ  Reactor Scram, Turbine Stop Valve closure  t= +30.0 sec  Bypass valve opening  t=+60.0 sec  Bypass valve closure, MSIV closure  Reactor Isolation  t=+360 sec  IC opening  t=+1060 sec  IC closure  t=+1860 to 2880 sec  IC openings/closures (3 times)  t=+3060 sec  TSUNAMI flooding  loss of AC/DC power, compress air & instrumentation (only Safety Valves available)

39 Contents Phase 1: Scram, Bypass valve opening/closure (pressure control), Reactor Isolation (MSIV closure)  from t= 0.0 secs to 60 secs Phase 2: Energy removal by Isolation Condensers  from t= +360 secs. to t= ~3060 sec (~ 51 min) Phase 3: End of cooling (stop of energy removal), loss of RPV inventory (water level decrease up to TAF)  from t= ~ 3060 secs to t=+6800 secs (~2hr) Phase 4: Core uncovery & degradation, H2 formation, core voiding….  from t=+2hr to +3.4/4 hrs

40 Base Case Results Reactor Power RPV Dome Pressure IC A & B Mass Flow Containment pressure (DW & WW)

41 Base Case Results SRV/SV mass flow DC & In-shroud Levels above TAF Hot Channel Clad Temperatures  Comparison with Japanese data & analyses showed some agreement/similarities EVENTRESULT [ENEA]NISATEPCO CORE EXPOSURE 11/3 h 16.46 2 hrs 3 hrs CORE DAMAGE 11/3 h 17.46 3 hrs 4 hrs

42 Sensitivity case: seal pump leakage Comparison In-shroud Levels above TAF [Ref. vs Sens.] Containment Pressures [Ref. vs Sens.] Peak Clad Temperature [Ref. vs Sens.]  Sensitivity case run including seal-pump leakages (~20/25 gpm per pump)  Clad damage anticipated by ~1/2 hr ca.  Greater containment pressurization

43 Conclusions  ENEA [Siming Lab] being involved in NPP simulation technology by using BE state of the art code RELAP5- 3D  Participation to international benchmarking activities + models development for internal analyses  Kalinin-3 & AER WG D activities presented, results submitted/being submitted  Fukushima Daiichi Unit 1 model  available for being further integrated/used in future ENEA activities

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