Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 1

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1 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 1

2 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 2

3 Established in 1937, Bachelor, master and PhD study, About students, 7 faculties: o Faculty of Civil Engineering, o Faculty of Mechanical Engineering, o Faculty of Chemical and Food Technology, o Faculty of Architecture, o Faculty of Materials Science and Technology in Trnava, o Faculty of Informatics and Information Technologies, o. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 3

Electrical Engineering, Institute of Electronics and Photonics, Institute of Robotics and Cybernetics,

4 Institute of Communication and Applied Linguistics, Institute of Power and Applied Electrical Engineering, Institute of Computer Science and Mathematics, Institute of Automotive Mechatronics, Institute of Electrical Engineering, Institute of Electronics and Photonics, Institute of Robotics and Cybernetics, Institute of Telecommunication,, Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 4

5 Established in 2011 by association of two departments: o Department of Physics, o Department of Nuclear Physics and Technology, More than 50 years tradition, Fields of activities: o education, o Research, o Development, 60 members: o 10 professors and 10 associate professors, o 7 educationalists, o 16 PhD. students, o 6 other staff. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 5

6 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 6

Vladimír Slugeň Professor, Director of the Institute of Nuclear and Physical Engineering.

Gabriel Farkas Associate professor, coordinator of the reactor-physics working group.

student, member of the reactor-physics working group. Branislav Vrban PhD.

7 Vladimír Slugeň Professor, Director of the Institute of Nuclear and Physical Engineering. Ján Haščík Associate professor, Team leader of the reactorphysics working group. Gabriel Farkas Associate professor, coordinator of the reactor-physics working group. Štefan Čerba PhD. student, member of the reactor-physics working group. Jakub Lüley PhD. student, member of the reactor-physics working group. Branislav Vrban PhD. student, member of the reactor-physics working group. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 7

8 DANTSYS : o 3D deterministic discrete ordinates transport theory code, o ONEDANT, TWODANT, TWOHEXD, TWODANT/HQ, THREEDANT, DIF3D: o Multi-group diffusion theory solver, o 1,2 and 3 dimensional orthogonal, triangular and hexagonal geometries, o Variational nodal method for 2 and 3 dimensional hex, and cartesian geometries, o Experience: SFR, GFR, R-Z, Hex-Z, higher harmonics, ERANOS code system: Extensive sets of data libraries, Cell and lattice code, Diffusion, SN transport and variational nodal transport flux solvers, Experience: 2,3 D diffusion calculations. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 8

9 SCALE code system: o Verified and validated computational tool, o Continuous energy and multi-group data libraries, o KENO-3D, NEWT, ORIGEN, MAVRIC, TSUNAMI-3D, o Experience: criticality safety and burnup (VVER, GFR), uncertainty analysis (GFR), MCNP(X) Stochastic steady state Monte Carlo computational code, Criticality safety, radiation shielding and protection, dissymmetry, medical physics, detector design applications, Typically point-wise cross section data is used but MG is also avalilable, Experience: Criticality safety, medical applications and activation analysis. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 9

MCAM : o Monte Carlo Modeling interface program, Fortran and C++ utilities: o Local multiplication factors, mesh fluxes, ACE library

10 NJOY, TRANSX, CRSRD : o Nuclear data processing tools, o CE and MG MCNP ACE, MATXS and ISOTXS libraries, NESTLE: Few-Group Neutron Diffusion for Steady-State and Transient Problems, MICROSHIELD, VISIPLAN, GOLDSYM: o Radiation shielding and dose assessment codes, MCAM : o Monte Carlo Modeling interface program, Fortran and C++ utilities: o Local multiplication factors, mesh fluxes, ACE library handling, Bash scripts for data pre and post processing. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 10

11 KAERI (1 year) Development of utilities aimed on optimization of neutron cross section data sets for design studies, reactor core and radiation shield optimization of the Korean space reactor, CEA (3 month) Investigation of the new capabilities of proposed SFR core designs, NPP Mochovce (3 years) Implementation of new fuel type to the operation of WWER reactors, GoFastR (1year) ALLEGRO and GFR2400 core analyses, Scientific papers - FR13, Nuclear Data Sheets, Core Neutronics Calculation of the GFR2400 Gas Cooled Fast Reactor Progress in Nuclear Energy. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 11

12 APSTRACT Analyzer of Perturbation and Sensitivity with TRAnsport Calculation application of standard perturbation Theory to determine sensitivity of k eff on cross section data. Identification of important processes and evaluation of the influence of variation in input parameter based on the estimation of a change of the system response due to change in this parameter. Capable to determine sensitivities to reactivity response and to decompose the integral reactivity in terms of space or material. ATCROSS Adjustment Tool for CROSs Section data. Conventional cross section adjustment method was applied to evaluate cross section data as much as possible within their error limits and taking into account the correlations, in such way, that a better agreement between calculated results and measured integral data is obtained. In addition, the determination of cross section uncertainty propagation on evaluated results was implemented to the code. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 12

13 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 13

14 Main objectives: Optimization of safety feedback coefficients by introducing a moderator into the SFR reactor core : o An effort was made to limit the excess reactivity and the void effect of the SFR reactor core, since these effects are associated with transient initiating general core melting. Minor Actinides Transmutation in a depleted UO 2 filled blanket zone of the SFR: o For 3 basic minor actinides compositions the utilization of BeO, MgO, Li2O, SiC and ZrH 2 moderators was investigated. Transmutation rates, decay heat production, He production, source terms and many other parameters were analysed and compared to the reference composition,without moderator. The French deterministic ERANOS2.2 code was used for these simulations. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 14

Criticality safety analysis of spent fuel storage of NPP MOCHOVCE using MCNP5, Calculation of the thermal reactivity feedback coefficients fort the first fuel loading of the 3 rd and 4 th unit of the

15 Criticality safety analysis of spent fuel storage of NPP MOCHOVCE using MCNP5, Calculation of the thermal reactivity feedback coefficients fort the first fuel loading of the 3 rd and 4 th unit of the Mochovce NPP using MCNP5, Verification of spatial neutron power distribution densities of the 2 nd unit of the Mochovce NPP for the 12. fuel loading using the MCNP5 and SCALE codes. Figure : MCNP model of compact grid of the spent fuel storage pool horizontal cross section. (Variant C2 (D2) loading with 4.87 % enriched FAs and four rows of 45 MWd/kg (50 MWd/kg) burned FAs.) Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 15

Associate member of the 7-th framework GoFastR project,

in the neutronic benchmark, Calculations performed on the

effect, o Depressurization reactivity, o Control rod

16 Associate member of the 7-th framework GoFastR project, Individual and cooperative works with BME Budapest and VUJE, a.s. Trnava, Calculations performed on GFR2400: Participation in the neutronic benchmark, Calculations performed on the MOX and ceramic pin type ALLEGRO reactor cores: o Doppler effect, o Depressurization reactivity, o Control rod efficiency, o Burn-up calculations. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 16

17 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 17

18 Δρ= 2581 pcm Δρ=1345 pcm Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 18

19 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 19

C 1 1 0 C C 1 2 r ABS C 2 C3 N pin Technical Meeting on

20 C C C 1 2 r ABS C 2 C3 N pin Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 20

Δρ [pcm] *A f [-] Homogenous 12364 10.1 Heterogeneous 11273 8.

21 Δρ [pcm] *A f [-] Homogenous Heterogeneous Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 21

22 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 22

Optimization of the heterogeneous CR design: o Optimal pin pitch and position, o Material composition, o Wrapper position and thickness, o Heat generation studies, o Utilization of neutron absorber

23 Optimization of the heterogeneous CR design: o Optimal pin pitch and position, o Material composition, o Wrapper position and thickness, o Heat generation studies, o Utilization of neutron absorber as a part of CR design, Application of deterministic approaches: o Reevaluation of the CR design, o Covariance matrices of CR amplification factors. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 23

Utilization of the TSUNAMI-3D module of the SCALE6 system, 238-group ENDF/B

Identification of appropriate criticality-safety benchmark experiments,

Investigation of the spatial distribution of integral sensitivities.

24 Utilization of the TSUNAMI-3D module of the SCALE6 system, 238-group ENDF/B VII.0 incident-neutron data libraries with 44-group SCALE covariance data, Identification of appropriate criticality-safety benchmark experiments, Calculation of sensitivity profiles for k eff and specific ENDF reactions, Investigation of the spatial distribution of integral sensitivities. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 24

25 To achieve the goals of this task it is necessary to face the following challenges: Processing of nuclear data to few group macro sets, Evaluation of reactivity effects of steady state calculations, Conjunction uncertainties (sources of uncertainties) of investigated parameters to transient calculation, Verification of the achieved results by the means of experimental data, Be able to truly predict the kinetic behavior of the investigated system, Based on experience, proceed from kinetics to dynamics. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 25

these institutes parallel effort will be made with focus to validate our methods and improve our skills in determining the calculation bias,

26 VR-1 o Training/academic reactor located at the Charles University in Prague, o Pool type, square lattice and zero power, o Thermal spectrum, LR-0 o Research reactor operated by Research Centre Rez, o Pool type, triangular lattice and zero power, o Thermal spectrum, In the frame of cooperation with these institutes parallel effort will be made with focus to validate our methods and improve our skills in determining the calculation bias, uncertainties, reactivity feedback effects and other important parameters. Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 26

27 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 27

28 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 28

Demonstrator of the GFR, 75 MW th thermal power, 2010 memorandum of understanding, ALLIANCE - EURATOM 2012 FP7, ALLegro Implementing Advanced Nuclear fuel cycle in Central Europe,: o ALLEGRO design &

29 Demonstrator of the GFR, 75 MW th thermal power, 2010 memorandum of understanding, ALLIANCE - EURATOM 2012 FP7, ALLegro Implementing Advanced Nuclear fuel cycle in Central Europe,: o ALLEGRO design & safety VUJE (Slovakia), o Technology related experiments UJV (Czech Republic), o Closed fuel cycle and fuel issues MTA EK (Hungary), o Industrial application of high temperature gases NCBJ (Poland). Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 29

30 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 30

SENSITIVITY AND PERTURBATION THEORY IN FAST REACTOR CORE DESIGN

Journal of ELECTRICAL ENGINEERING, VOL. 65, NO. 7s, 214, 25 29 SENSITIVITY AND PERTURBATION THEORY IN FAST REACTOR CORE DESIGN Jakub Lüley Branislav Vrban Štefan Čerba Ján Haščík Vladimír Nečas Sang-Ji