Michal Košťál PhD thesis Transport of Neutrons and Photons in Construction Parts of VVER‑1000 Reactor Michal Košťál PhD thesis Department of experimental reactor physics at LR-0, Research Center Řež Czech Technical University in Prague Faculty of Nuclear Sciences and Physical Engineering Department of Nuclear Reactors

The objects of PhD thesis and supporting references Compilation of the calculation model for neutron and photon transport in VVER-1000 transport benchmark (with prospect of calculations in biological shielding) Determination of neutron emission density, across the reactor core and assessment of link between neutron emission density and fission density Determination of neutron emission spectra of various fuel pins Estimation of related uncertainties Estimation of sensitivity to the selection of specific nuclear data library Estimation of sensitivity to the selection of specific transport model (in case of Fe and H2O) The main aim of the thesis and references was the compilation and validation of calculation model of VVER-1000 transport benchmark. The topic was divided info few sub.-goals ….

VVER-1000 benchmark The experiments were performed in VVER-1000 benchmark. It is a radial full-scale physical model of approximately 30° azimuthal symmetry sector of the core periphery. In axial range is shorten with regard to real reactor to 1.25m due to construction parameters. The displacer, a steel sector tank containing an appropriate air gap, simulates the down-commer water density reduction, which takes place under the VVER-1000 power reactor operating conditions. The RPV simulator is modular one, consist of 4 layers. The It allows to determine the neutron / photon parameters in its various parts. It is placed in LR-0 reactor. Radial full scale VVER-1000 transport benchmark (RPV, baffle, barrel) Baffle is not is not unruffled - milled cooling holes in vertical and in horizontal plane as well For simulation of the water density reduction displacer is used RPV consist of four 5cm steel blocks, the first one consist of 1cm of stainless (RPV cladding simulator) and 4cm low alloy steel

LR-0 Reactor Light water moderated zero-power reactor Maximal nominal power 1 kW, thermal neutron flux density ~ 1013 n.m-2 s-1 Core in Al tank, inner diameter 3500 mm, thickness 16mm, height 6600 mm Power control realized by means of moderator level change or control-cluster position Demineralized water with or without diluted boric acid is used as moderator Dismountable fuel elements VVER type fuel, length of pins is shortened (125cm) with regard to LR-0 construction

Upper view on VVER-1000 core inside LR-0 For illustration, this slide shows the upper view on the mock-up core without moderator.

4 points in reflector 5 points in positions In front or water The mock-up construction allows to determine the fluxes in its various parts. Measuring points 4 points in reflector In front or water Behind 5cm, Behind 10cm Behind 15cm 5 points in positions In front of RPV In ¼ of RPV In ½ of RPV In ¾ of RPV Behind RPV As stated, the benchmark construction allows to determine the fluxes in various parts of mock-up. This figure illustrate the points, where the fluxes were determined.

Pin power distribution Radial profile Fission density ~ ( generally not proportional to emission density) Model verified on keff results, being 0.99462 (ENDF/B VI.2.) The fission density was determined using defined source model, the keff value was used for it verification. The result ensures, the uncertainty in fission source estimation less than 0.6%. The fission density is generally not proportional to emission density and power density, due to variation of ν (number of neutron per fission) and energy per fission. The Χ (fission spectra ) varies with energy for both 235 and 238U as well.

Various position incident neutron spectra The variations are caused by different properties of boundary materials, and are reflected in different neutron spectra in various mock-up parts Different properties of steel causes considerably harder spectra near baffle than in other regions The neutron spectra vary across the core

Variations in fission products and energy generation Near baffle Near gap Inner Corner <1eV 81.7% 86.0% 85.2% 79.2% 1eV - 1keV 9.2% 6.5% 6.9% 10.9% 1keV-0.1MeV 0.94% 0.7% 0.70% 1.12% 0.1-1MeV 0.46% 0.33% 0.36% 0.54% >1MeV 0.43% 0.35% 0.37% 238U 7.23% 6.15% 6.47% 7.72%   140Ba 140La near baffle 0.06163 0.06167 near gap 0.06171 0.06176 inner 0.06168 0.06173 corner 0.06159 0.0253eV 0.06214 0.06220 This effect changes the fission distribution in various pins – more fission caused by higher energy neutrons may be observed in peripheral regions. But this effect has only slight effect on both fission product or energy generation from fission. Thus the fission density may be dealt as power density.   near baffle near gap inner corner Neutrons/fission [-] 2.42055 2.42050 2.42062 2.42058 Energy/fission [MeV] 203.184 203.043 203.086 203.250

Various position neutron emission spectra Only small variations between corner pin emission spectra and inner pin emission spectra both are similar with Watt emission spectra for 235U and thermal neutron

Comparison with diffusion approach There are considerable discrepancies between both Possible reasons of such discrepancies Incorrect boundary conditions (i.e. approximation of full core, but benchmark is just 1/6 of VVER-1000 core Peripheral regions (near baffle) seems to be reflection of innacuracies from diffusion approach This lead to comparison between previously calculated power density, using diffusion approach – Moby Dick code – and recent calculation using Monte Carlo approach. There are considerable discrepancies between both.

Fuel pins selection for C/E comparison Selection reflects the pins with expected discrepancies The experimental uncertainties prevail in C/E uncertainty Peripheral pins uncertainty unanswerable problem in this selection – power density in center (As-27) ~20x higher than in periphery (As-4) and reasonable doses must be ensured Power determined by means of La-140 fission product activity measurement Experiment realized 16 days after irradiation – enough time for setting of La-Ba equilibrium Construction of LR-0 fuel allows to dismantle the fuel element and handled with each pin independently. This allows the fission density determination. The La-140 activity was used as indicator of fission density in each pin. The figure presents the selected pins used for comparison. The selection is focused on pins with expected discrepancies when diffusion approach is used for calculations. Namely it is pins near baffle, in corners and near lateral reflector. For data validation, the values were determined in symmetrical positions.

Pin power density C/E Selection of pins in positions with expected discrepancies near the core and baffle (1 – 31) assemblies corners (32 – 46) near lateral reflector (47 – 52) Comparison of symmetrical pins used for verification of experiment Near baffle, better agreement with MCNP than with MOBY DICK diffusion approximation insufficiency appears in the boundary regions (high neutron flux gradient, different material boundary Near water gap (corner pins, near lateral reflector pins), both MCNP and MOBY DICK results in similar agreement with experimental values The comparison between both calculated and experimental values is shown in first figure. Especially near baffle the considerably better agreement was met in case of MCNPX than Moby Dick calculation. The comparison between power densities in symmetrically located pins shows second figure.

Axial profile of power density C/E Discrepancies in distant grids locations The comparison in axial profile was performed for non peripheral pin – due to sufficient activity of measured nuclide activity.

Neutron fluxes in reflector In case of neutron fluxes in lateral reflector the discrepancy in region 2.35MeV can be observed. It reflect the negative resonance on 16O.

Neutron fluxes in RPV In case of neutron fluxes in RPV simulator, the good agreement between both calculated and measured values reflects the absence of smooth structure, which cannot be registered by stilbene detector.

Transport model effect H2O keff Slight variations if used ENDF/B VII & S(α, β) results closer to experiment Fe Photon flux density (18cm Fe) Notable variations if used The effect of used nuclear data on the results was studied as well. Its worth noting, more notable variations may be observed in case of iron than in case of water. In case of water the effect of transport model was studied on keff results. In case of ENDF 7 better agreement is met when TSL law is employed. In case of iron, the effect of transport model was studied on photon flux density attenuation coeffcient results. Better agreement is met, when the TSL is not employed.

Nuclear data library effect - fuel Only slight variations Except ENDF/B VI.2 discrepancies less than related uncertainties Best C/E agreement CENDL 3.1 Only ENDF 6 calculations differ from experiments more than related uncertainty The effect of library was studied for both – fuel description as well as construction parts of mock-up. In case of fuel, the effect was studied for keff results. The resulted variations are small – only calculation which differs from experimental one more than related uncertainty, are those obtained with ENDF 6. Nevertheless the discrepancy is not higher than 160% of sigma. H [cm] ρ [g/kg] ENDF/B VI ENDF-VII JEFF 3.1. JENDL 3.3. JENDL 4 ROSFOND 2009 CENDL 3.1 51.34 2.85 0.99559 1.00154 1.00093 0.99926 1.00164 1.00153 0.99946 65.91 3.63 0.99562 1.00256 1.00079 0.99938 1.00253 1.00205 0.99921 79.11 4.06 0.99596 1.00291 1.00151 0.99942 1.0028 1.00222 0.99979 96.71 4.44 0.99616 1.00314 1.00129 0.99968 1.00392 1.00245 0.99965 103.37 4.53 0.99607 1.00265 1.00075 0.99967 1.00226 1.00186 0.99941 150 4.68 0.99462 1.00137 0.99936 0.99842 1.00133 1.0009 0.99863

Nuclear data library effect – Fe (18 cm slab) Neutrons (thick layers) Most notable discrepancies (4–7 MeV) for JENDL 4 and TENDL 2009 Photons Most notable discrepancies (>7MeV) for JEFF 3.1 and TENDL 2009 More notable discrepancies may be found in case of iron. The results obtained with JENDL 4 and TENDL 2009 considerably differ from experimental values. In case of photons notable discrepancies are in upper energy region and for JEFF and TENDL 2009.

Thank you for your attention

Published results Thermal scatter treatment of iron in transport of photons and neutrons, M. Košťál, František Cvachovec, Bohumil Ošmera, Wolfgang Hansen, Vlastimil Juříček, Annals of Nuclear Energy, Volume 37, Issue 10, October 2010, pp 1290–1304 The Pin Power Distribution in the VVER-1000 Mock-Up on the LR-0 Research Reactor, M. Košťál, V. Rypar, M. Svadlenkova, Nuclear Engineering and Design, Volume 242, January 2012, pp 201– 214 Determination of AKR-2 leakage beam and verification at iron and water arrangements, M. Košťál, F. Cvachovec, J. Cvachovec, B. Ošmera, W. Hansen Annals of Nuclear Energy, Volume 38, Issue 1, January 2011, pp 157-165 Calculation and measurement of neutron flux in the VVER-1000 mock-up on the LR-0 research reactor, M. Košťál, F. Cvachovec, V. Rypar, V. Juříček: Annals of Nuclear Energy, 40 (2012), pp 25–34, The Power Distribution and Neutron Fluence Measurements and Calculations in theVVER-1000 Mock-Up on the LR-0 Research Reactor, Košťál, M., Juříček, V., Novák, E., Rypar, V., Švadlenková, M., Cvachovec, in press, ISRD-2011, Bretton woods, USA Transport of neutrons and photons through iron and water layers, Košťál, M., Cvachovec, F., Ošmera, B., Noack, K., Hansen, W.,. Proceedings of the 13th International Symposium on Reactor Dosimetry, Ackersloot, Netherlands. pp. 269 – 279 Results send for review: Neutron and photon transport in Fe with the employment of TENDL 2009, CENDL 3.1., JENDL 4 and JENDL 4 evolution from JENDL 3.3 in case of Fe, M. Košťál, F. Cvachovec, J.Cvachovec, B. Ošmera, W. Hansen, Nuclear Engineering and Design Thermal neutron transport in the VVER-1000 mock-up on the LR-0 research reactor, Nuclear Engineering and Design, M. Košťál, V. Juříček, J. Milčák, A. Kolros The criticality of VVER-1000 mock-up with different H3BO3 concentration, M. Košťál, V. Rypar, V. Juříček, Progress in Nuclear Energy

Influence of power distribution on results The variation are smaller than related uncertainties => Diffusion approximation power density may be used in following transport calculations

3He reaction rate attenuation In RPV simulator of VVER-1000 <0.55eV >0.55eV   ENDF VII ENDF VII+TSL CENDL 3.1 experiment 3 / 4 21.89 7.80 18.41 18.68 2.69 2.665 2.55 2.54 4 / 5 9.55 5.70 9.32 3.99 2.01 2.058 1.90 1.85 5 / 6 1.55 2.15 1.82 1.23 1.521 1.53 1.42 6 / 7 0.10 0.26 0.12 0.28 1.01 0.991 1.00 1.02 3 / 7 32.72 24.45 38.38 25.86 8.48 8.27 7.40 6.77 In RPV simulator of VVER-1000 with PE liner   ENDF VII ENDF VII+TSL experiment 3 / 4 2.232 2.195 2.218 1.541 1.559 4 / 5 1.467 1.551 1.455 1.386 1.389 5 / 6 1.347 1.350 1.316 1.337 1.306 1.311 6 / 7 1.328 1.274 1.223 1.346 1.322 1.267 3 / 7 5.859 5.857 5.196 3.843 3.873 3.597

Pin power measurement La-140 – 1596keV (fraction 0.954) Long irradiation time => long decay time => many measured pins Te-140 I-140 Xe-140 Cs-140 Ba-140 La-140 T 1/2 0.304 s 0.86 s 13.6 s 63.7 s 12.75 d 1.678 d yield 1.70E-4 2.04E-3 3.74E-2 5.73E-2 6.19E-2 near baffle 0.06% 2.05% 0.19% 0.01% -0.04% corner 0.11% 3.89% 0.36% -0.07% Sr-92 – 1383keV ( fraction 0.9) Short irradiation time => short decay time => few measured pins Se-92 Br-92 Kr-92 Rb-92 Sr-92 T 1/2 0.093 s 0.343 s 1.84 s 4.492 s 2.71 h yield 1.74E-6 4.11E-4 1.74E-2 4.77E-2 5.83E-2 near baffle 6.34% 2.90% 0.32% -0.08% -0.16% corner 12.03% 5.49% 0.61% -0.15% -0.30%