National Research Centre "Kurchatov Institute" PETERSBURG NUCLEAR PHYSICS INSTITUTE Neutronic calculations of reactor PIK using SERPENT code M.S. Onegin 8 th International Serpent UGM 29 May 1 June 2018 1
PETERSBURG NUCLEAR PHYSICS INSTITUTE Content Reactor PIK - introduction Serpent model of reactor PIK Criticality calculations Burnup of PIK fuel Doppler effect Temperature dependence of reactor reactivity Steady state calculations 2
Reactor PIK PETERSBURG NUCLEAR PHYSICS INSTITUTE 2011 Criticality reached 2013 Construction finished 2014 2017 Licensing and neutron stations and installations construction 2018 Power operation scheduled The maximum heat output Heat transfer agent Reflector Number of horizontal experimental channels Number inclined experimental channels Number of vertical experimental channels 100 MW water heavy water 10 6 6 3
Vertical cut of reactor PIK PETERSBURG NUCLEAR PHYSICS INSTITUTE 1. Fuel handling machine 2. Control rod disconnect driveline 3. Hydraulic lock 4. Central experimental channel 5. Transfer device cylinder 6. Cold neutron source 7. Biological shielding 8. Control rod 9. Reactor core 10. Driving gear of channel gate 4
PIK core PETERSBURG NUCLEAR PHYSICS INSTITUTE 1. Hafnium central control rod 2. Water displacers 3. Fuel assemblies Zr case 4. Profiled Fuel elements 5. Fuel elements 6. Fuel assemblies with irradiation volumes 7. Volumes for samples irradiation Heat-transfer agent light water Presser 50 bar Inlet temperature 50 C Outlet temperature 90 C Inlet water velocity 10 m/s Power heat load 2 MW/l (mean) density 6.6 MW/l (max) 5
А - А 1 8 0 59 5. 2 3 1 5 6 4.53 65.2 65.9 5 9 8 116 2 1 4 1085 1 0 5 0 А 5 0 0 А 3 FA of reactor PIK PETERSBURG NUCLEAR PHYSICS INSTITUTE S cell = 23.688 mm 2 S pin = 10.15 mm 2 S meat = 7.23 mm 2 3 2 3 6 Hexagonal lattice with 5.23 mm pitch 6
SERPENT model of reactor PIK 12 hexagonal FA 6 square FA First 2 layers of hexagonal FA are profiled (1/3 fuel load) 3858 fuel pins 10 Horizontal neutron channels 6 Vertical neutron channels Weight of channels ~0.3 0.4 % The calculation time increased ~4 times! 7
Lattice calculations Shape #1 Shape #2 Parameter Shape #1 NEWT SERPENT ENDF/B-7.0 SERPENT ENDF/B-6.8 SERPENT-2 ENDF/B-6.8 k 1.596567 1.59787(8) 1.59543(8) 1.59543(6) ε 1.865195 1.85832 1.85947 p 0.464947 0.46812 0.46741 f 0.955383 0.95474 0.954782 η 1.927000 1.92491 1.92585 Parameter Shape #2 NEWT SERPENT-2 SERPENT-2 ENDF/B-6.8 ENDF/B-7.0 k 1.603214 1.59573(6) 1.59840(6) ε 1.851426 p 0.470570 f 0.954866 η 1.927165 k pf NEWT deterministic 2d code from SCALE 6.2 package 8
, cm -2 s -1 Criticality Neutron data library Keff Shape #1 Shape #2 ENDF-B/VI-8 0.99852(8) 0.99796(7) -0,06(1) ENDF-B/VII-0 0.99961(8) 0.99918(8) -0,04(1) +0,11(1) +0,12(1) 6x10 7 5x10 7 4x10 7 3x10 7 2x10 7 SERPENT Experiment There was no difference in the calculation time of the variant when we used more complicated fuel pin crosssection form. 1x10 7 0-30 -20-10 0 10 20 30 40 Z, см 9
Weight of control rods CR CR weight, CR weight, CR exp /CR calc experiment β eff calculation, β eff АЗ1 0.62 0.58(2) 1.07 АЗ2 0.55(5) 0,56(1) 0.98 КС1 0,52 0,48(2) 1.08 КС2 0,53(5) 0,49(1) 1.08 КС3 0,52(5) 0,55(1) 0.95 КС4 0,47(5) 0,45(1) 1.04 КС5 0,49(5) 0,49(1) 1.0 КС6 0,58(5) 0,47(5) 1.23 Mean 0.54(5) 0.51(5) 1.06 10
Heat peaking factor а б в г д е ж з и к л м н о 2,00 2,21 2,42 2,58 2,65 2,60 2,70 2,51 2,32 2,32 2,43 2,39 1,89 1,38 1,52 1,561,57 1,68 1,64 1,69 1,69 1,64 1,68 1,57 1,56 1,52 1,38 2,14 2,25 2,39 2,48 2,46 2,53 2,53 2,72 2,49 2,47 2,53 2,38 2,48 2,23 2,12 1,71 1,70 1,72 1,85 1,78 1,88 1,82 1,92 1,92 1,82 1,88 1,78 1,85 1,72 1,70 1,71 1,57 1,45 1,55 1,54 1,53 1,54 1,58 1,57 1,55 1,57 1,58 1,54 1,53 1,54 1,55 1,45 1,57 1,44 1,36 1,38 1,33 1,39 1,41 1,41 1,35 1,45 1,45 1,35 1,41 1,39 1,33 1,38 1,38 1,36 1,44 1,43 1,37 1,40 1,36 1,42 1,32 1,32 1,35 1,39 1,391,39 1,35 1,32 1,32 1,42 1,36 1,40 1,37 1,43 1,44 1,33 1,33 1,29 1,33 1,30 1,29 1,28 1,30 1,30 1,30 1,30 1,28 1,29 1,30 1,33 1,29 1,33 1,33 1,44 1,59 1,37 1,34 1,31 1,26 1,27 1,29 1,26 1,30 1,25 1,28 1,25 1,30 1,26 1,29 1,27 1,26 1,311,34 1,37 1,59 1,77 1,46 1,30 1,25 1,27 1,21 1,25 1,26 1,25 1,22 1,30 1,30 1,22 1,25 1,26 1,25 1,21 1,27 1,25 1,30 1,46 1,77 1,76 1,531,40 1,36 1,32 1,27 1,25 1,25 1,26 1,30 1,28 1,30 1,26 1,25 1,25 1,27 1,32 1,36 1,40 1,53 1,76 1,62 1,44 1,25 1,37 1,25 1,21 1,22 1,22 1,21 1,21 1,22 1,22 1,21 1,25 1,37 1,25 1,44 1,62 1,62 1,44 1,32 1,27 1,24 1,27 1,27 1,19 1,27 1,27 1,24 1,27 1,32 1,44 1,62 1,47 1,35 1,25 1,26 1,26 1,31 1,27 1,26 1,26 1,25 1,33 1,47 K V K V K Z K Z (expe (calc) (experi (calc) riment) ment) 1 layer 2,41(1) 2,53(7) 1,68(1) 1,63(5) 2 layer 1,50(1) 1,65(3) 1,60(1) 3 layer 2,49(1) 2,51(3) 1,52(1) 1,57(4) 4 layer 1,78(1) 1,85(2) 1,44(1) 1,44(3) 5 layer 1,50(1) 1,55(1) 1,40(1) 6 layer 1,37(1) 1,38(2) 1,37(1) 7 layer 1,31(1) 1,36(2) 1,36(1) 1,33(3) - 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 8 layer 1,27(1) 1,30(1) 1,34(1) 1,33(3) 9 layer 1,25(1) 1,28(1) 1,33(1) 10 layer 1,24(1) 1,25(2) 1,32(1) 1,31(3) 11 layer 1,24(1) 1,29(2) 1,33(1) 12 layer 1,23(1) 1,25(2) 1,32(1) 13 layer 1,24(1) 1,25(2) 1,32(1) 14 layer 1,33(2) 1,27(2) 1,32(1) 11
Fuel burnup, % 7 6 5 MCNP + MONTEBURNS SERPENT-2 MCU-REA 4 3 2 1 0 0 10 20 30 40 50 60 70 80 90 100 110 120 130 t, days 8 burnable fuel material zones with division along height (10 layers) 10 MW reactor power As a result we get 80 burn-up material zones. To use burn-up FA again we need to include the axial division of the core explicitly in the reactor model and describe all 80 burn-up materials. So the profit from automatic division of fuel disappeared after fist operation cycle. But after that the user have a direct control over the fuel burning. 12
Doppler and temperature effects 0,0-0,5, % -1,0-1,5-2,0 water in central trap is hot water in central trap is cold -2,5 20 40 60 80 100 120 140 160 T water, o C Red curve NEWT code Green curve SERPENT-2 cell Circles SERPENT-2 full core calculations T = 323-343 K: T = 300-343 K: Experiment T=297 345 K: / T 1.50 cent / / T 1.14 cent / K K 13
Steady state calculations Material zones: Hexagonal inner hip 1 st profiled layer hip2 2 profiled layer hi3l 3 layer hi 4 14 layers Hexagonal outer ho 1-13 layers holl 14 layer Square outer box - 1-13 layers bll - 14 layer Axial 10 layers Total - 80 PC 16 cores; 32 Gb memory SUSE Linux; OpenMP 20 threads Cold state var. Memory 18.3 Gb 5x107 histories 1.42 hour 8-th International UGM; Espoo, May 29 14
Heat load calculations We couldn t use standard code interface for the fuel pins heat release data because of non standard form of our fuel pins. So we used detectors to calculate heat release Q i in every fuel division bin: det 2 du 2 dm fuelprof dr -8 void dz -25. 25. 10 15
T, K Water heating t i t Q i i1, Gcpi 400 380 hi hip hip2 hi3l ho holl box bll 360 340 cold hot ~ 0.6% (0.79 ) mat waterhi -0.99016 tft 320 420 moder lwtr 1001 therm lwtr 0 lwj3.00t lwj3.01t lwj3.03t lwj3.05t eff 320 0 2 4 6 8 10 layer 2 waterhi 0 5 0 0 40-25 25 1 1 10-0.9889 325.7-0.9874 329.4-0.9849 334.4-0.9817 340.4-0.9777 347-0.9735 353.7-0.9696 359.8-0.9663 364.6-0.9643 368.3-0.9622 371 16
2 fuelz4 0 5 0 0 40. -25.0 25.0 1 1 10-8.753 0. -8.753 0. -8.753 0. -8.753 613. -8.753 0. -8.753 0. -8.753 0. -8.753 0. -8.753 0. -8.753 0. Fuel heating t t clo t w q S Ko conv t t K q cli clo cl S fuel Nu d кон g 0, 11 0, 8 0, 4 в 3, 6 0, 023Re Pr 1 0, 357 w Fr Nu t K q cli fuel V Pr Pr cl cl 2 dt 16 fuel 17
Cold state var. Optimization level 3 Memory 18.3 Gb 5x10 7 histories 1.42 hour Hot state var. Fuel mat: tft 300 900 Optimization level 1 Memory 0.6 Gb 5x10 7 histories 1.44 hour Summary Advantages: SERPENT -2 correctly predict all important neutronic parameters of the reactor It can be used for steady state calculations of the reactor Code correctly reproduce parameters significant for reactor safety Good FOM Need to improve: - neutron tracing in heavy reflector with neutron channels - steady state interface for thermohydraulic calculations 18
Thank you for the attention! 19