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Activity on Fast Reactor in Slovakia
Ján Haščík, [email protected]
Institute of Nuclear and Physical Engineering STU in Bratislava
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 2
Content
Experience of neutronic group
Computational schemes and tools
Strategies for further research of fast reactor systems
Further activity
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 3
Experience in the field of design studies
NPP Mochovce (3 years) – Implementation of new fuel type to the operation of WWER reactors
CEA (3 month) – Investigation of the new capabilities of proposed SFR core designs,
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
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 4
Experience in the field of design studies 2
GoFastR (1year) – ALLEGRO and GFR2400 core analyses
Scientific papers - FR13, Nuclear Data Files, in publication
Near Future – studies related to reactivity control systems, calculation bias, uncertainties, reactivity feedback effects and kinetics.
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 5
CEA – Reactivity Coefficients and Transmutation
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.
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 6
CEA – Reactivity Coefficients and Transmutation 2
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.
For the simulations the French deterministic ERANOS2.2 code was used
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 7
NPP Mochovce- Criticality Safety analyses
Criticality safety analysis of spent fuel storage of NPP MOCHOVCE using MCNP5,
Verification of neutron power density distribution of the 2nd unit of the Mochovce NPP for the 12. fuel loading campaign, 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.)
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 8
International GoFastR project
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 on the neutronic benchmark
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 9
International GoFastR project 2
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
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 10
KAERI - Sensitivity & uncertainty, cross section adjustment
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.
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 11
KAERI - Sensitivity & uncertainty, cross section adjustment 2
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.
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 12
International Conference on Fast Reactors and Related Fuel Cycles, Safe Technologies and Sustainable Scenarios – FR13
Investigation of the Coupled effects of Movable Reflector and Safety Control Rods in the GFR 2400
Calculation of Void and Doppler Coefficients Identification and analyses of decoupled neutronic
zones in the GFR 2400 core Performed by the MCNP5 and KENO6 codes with
ENDF/B VII.0 libraries
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 13
Sensitivity and Uncertainty Analysis of the ALLEGRO MOX Fuel Subassembly
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.
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 14
Available computational tools
DANTSYS – 3D deterministic transport theory code DIF3D – 3D deterministic diffusion theory based
computational tool ERANOS code system SCALE code system MCNP(X) – stochastic steady state Monte Carlo computational
code NJOY, TRANSX, CRSRD – nuclear data processing tools NESTLE/PARCS – few-group neutron and hydro-mechanic
codes MICROSHIELD, VISIPLAN, GOLDSYM, radiation shielding
and dose assessment codes MCAM – Monte Carlo Modeling interface program
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 15
SAFETY Control Systems
Studies of the effectiveness of available control rod designs for GFR2400
Reactivity worth Integral characteristics amplification and interferences
Proposal of heterogeneous CR design Optimal pin radius and position Material composition Utilization of neutron absorber as a part of CR
design
Application of deterministic approaches Covariance matrices of CR amplification factors
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 16
SAFETY Control Systems 2 Reactivity management by introducing a movable
reflector Localization of local hot spots areas, local multiplication
factors › 1 To flatten the neutron flux distribution by reflector
removal Investigation of various reflector S/As and ring
configuration
Possibilities of third level control system
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 17
Coupled neutron and Thermal-hydraulics - Kinetic calculations
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
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 18
Coupled neutron and Thermal-hydraulic - Kinetic calculations 2
VR-1 Training/academic reactor of CTU in Prague Pool type, square lattice and zero power Thermal neutron spectrum
LR-0 Research reactor operated by Research Centre
Řež Pool type, triangular lattice and zero power Thermal neutron 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.
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 19
ALLEGRO
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, ALLEGRO design & safety VUJE (Slovakia) Technology related experiments UJV (Czech
Republic) Closed fuel cycle and fuel issues MTA EK
(Hungary) Industrial application of high temperature gases
NCBJ (Poland)
FEI STU
47th IAEA Technical Working Group on Fast Reactors Vienna
2014 20
Further activities
Neutronic analyses of Gas Cooled Fast Reactor in the frame of APVV 0123-12 project (Slovak Agency for Science a Research) till November 2017
Analyses oriented mostly to GRF 2400 and ALLEGRO
As associated member in GoFastR -cooperation with VÚJE and group of BUTE
Cooperation with CTU (VR1) and CV Řež (LR 0) benchmark testing simulation tools