slovak university of technology in bratislava · pdf filetechnical meeting on priorities in...

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

SLOVAK UNIVERSITY OF TECHNOLOGY IN BRATISLAVA Faculty of Electrical Engineering and Information Technology

Institute of Nuclear and Physical Engineering




Established in 1937,

Bachelor, master and PhD study,

About 18 000 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 Faculty of Electrical Engineering and Information Technology.




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,

Institute of Nuclear and Physical Engineering,




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.




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

Ján Haščík – Associate professor, Team leader of the reactor-physics 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.




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.




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.




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.




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.




APSTRACT Analyzer of Perturbation and Sensitivity with TRAnsport Calculation application of standard perturbation Theory to determine sensitivity of keff 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.




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 UO2 filled blanket zone of the SFR:

o For 3 basic minor actinides compositions the utilization of BeO, MgO, Li2O, SiC and ZrH2 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.




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 3rd and 4th unit of the Mochovce NPP using MCNP5,

Verification of spatial neutron power distribution densities of the 2nd 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.)




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.




Δρ= 2581 pcm

Δρ=1345 pcm




0

11

C

ABSr

CC 1

2

pinN

CC 2

3




Δρ [pcm] *Af [-]

Homogenous 12364 10.1

Heterogeneous 11273 8.4




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.




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 keff and specific ENDF reactions,

Investigation of the spatial distribution of integral sensitivities.




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.




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.




Demonstrator of the GFR,

75 MWth 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),

oClosed fuel cycle and fuel issues MTA EK (Hungary) ,

o Industrial application of high temperature gases NCBJ

(Poland).

slovak university of technology in bratislava · pdf filetechnical meeting on priorities in...

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