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 PragueFaculty of Nuclear Sciences and Physical EngineeringDepartment 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)

VVER-1000 benchmark

• 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

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

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.)

Various position incident neutron spectra

1E-3

1E-1

1E+1

1E+3

1E+5

1E+7

1E-9 1E-7 1E-5 1E-3 1E-1 1E+1

Energy [MeV]

Flu

x de

nsity

[1/

cm2]

near baffle pin near gap pin inner pin

• 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

Neutrons/fission [-] 2.42055 2.42050 2.42062 2.42058

Energy/fission [MeV] 203.184 203.043 203.086 203.250

  140Ba 140La

near baffle 0.06163 0.06167

near gap 0.06171 0.06176

inner 0.06168 0.06173

corner 0.06159 0.06163

0.0253eV 0.06214 0.06220

235U 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% 0.46%

238U Near baffle Near gap Inner Corner

>1MeV 7.23% 6.15% 6.47% 7.72%

Various position neutron emission spectra

-0.75%

-0.50%

-0.25%

0.00%

0.25%

0.50%

0.75%

0 2 4 6 8 10

Energy [MeV]

ratio

[-]

N(out)/N(0.0253eV)-1 N(in)/N(0.0253eV)-1 N(corner)/N(0.0253eV)-1 N(in)/N(out)-1

• 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

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

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

0.850.900.951.001.051.101.151.201.251.301.35

0 5 10 15 20 25 30 35 40 45 50 55

pin position

C/E

MOBY DICK MCNPX 1s uncertainty

-0.20

-0.15

-0.10-0.05

0.00

0.05

0.100.15

0.20

0.25

0 5 10 15 20 25 30 35 40 45 50 55

Pin position

P/P

(inv)

-1

Experiment MCNPX

1s experimental interval 1s calculational interval

Axial profile of power density C/E• Discrepancies in distant grids locations

1000

1500

2000

2500

3000

3500

4000

4500

5000

5500

0 25 50 75 100 125

Axial position [cm]

Pow

er [a

.u.]

MCNPX & ENDF/B VI.2. Benchmark data Measurement

Neutron fluxes in reflector

1E-4

1E-3

1E-2

1E-1

1E+0

1E+1

0.1 1 10Energy [MeV]

Neu

tron

flux

den

sity

[a.u

.]

Pt-2 Pt-21 Pt-22 Pt-23pt2 -calc pt-21 calc pt-22 calc pt-23 calc

Neutron fluxes in RPV

1E-6

1E-5

1E-4

1E-3

1E-2

1E-1

0.1 1 10

Energy [MeV]

Ne

utr

on

flu

x d

en

sity

[a

.u.]

Pt-3 Pt-4 Pt-5 Pt-6 Pt-7

Pt-3 Calc. Pt-4 Calc Pt-5 Calc Pt-6 Calc. Pt-7 Calc.

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– ENDF/B VII & S(α, β)

results closer to experiment

0.994

0.996

0.998

1

1.002

1.004

1.006

2.75 3.25 3.75 4.25H3BO3 [g/kg]

Ke

ff

ENDF VI ENDF VII ENDF VI free gas ENDF VII free gas

0

0.05

0.1

0.15

0.2

0.25

0.3

>1MeV >3MeV >5MeV >7MeVEnergy group

Atte

nuat

ion

ratio

Free gas TSL Experiment

Nuclear data library effect - fuel

H [cm] ρ [g/kg] ENDF/B VI ENDF-VII JEFF 3.1. JENDL 3.3. JENDL 4ROSFOND

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

• 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

0.994

0.996

0.998

1

1.002

1.004

1.006

2.75 3.25 3.75 4.25 4.75H3BO3 [g/kg]

Kef

f

ENDF VI.2 ENDF VII JEFF 3.1. JENDL 3.3.

JENDL 4 RF CENDL 1S

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

-20%-15%-10%

-5%0%5%

10%15%20%25%30%

1 2 3 4 5 6 7 8 9 10

Energy [MeV]

(C-E

)/E

ENDF VI.2 ENDF VII JEFF 3.1.

JENDL 3.3. JENDL 4 ROSFOND 2009

CENDL 3 TENDL 2009 1s uncertainty

1E+0

1E+1

1E+2

1E+3

1E+4

1E+5

0 1 2 3 4 5 6 7 8 9 10

Energy [MeV]

Ph

oto

n fl

ux

de

nsi

ty [c

m-2

.s-1

]

ENDF/B VI.2. ENDF/B VII JEFF 3.1.

JENDL 3.3 JENDL 4 ROSFOND 2009

CENDL 3.1 TENDL 2009 Experiment

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

-1.0%

-0.8%

-0.6%

-0.4%

-0.2%

0.0%

0.2%

0.4%

0.6%

0.8%

0 1 2 3 4 5 6 7 8 9 10

Energy [MeV]

M.D

./MC

NP

-1

Pt-2 Pt-3 Pt-7

-1.5%

-1.0%

-0.5%

0.0%

0.5%

1.0%

1.5%

2.0%

1 3 5 7 9

Energy [MeV]

M.D

./MC

NP

-1

P-3 P-7

3He reaction rate attenuation<0.55eV >0.55eV  

  ENDF VII ENDF VII+TSL CENDL 3.1 experiment 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 1.55 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

  ENDF VII ENDF VII+TSL experiment ENDF VII ENDF VII+TSL experiment

3 / 4 2.232 2.195 2.218 1.541 1.541 1.559

4 / 5 1.467 1.551 1.455 1.386 1.455 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

• In RPV simulator of VVER-1000 with PE liner

Pin power measurement

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 6.19E-2

near baffle 0.06% 2.05% 0.19% 0.01% -0.04% -0.04%

corner 0.11% 3.89% 0.36% 0.01% -0.07% -0.07%

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%

• La-140 – 1596keV (fraction 0.954)– Long irradiation time => long decay time => many measured pins

• Sr-92 – 1383keV ( fraction 0.9)– Short irradiation time => short decay time => few measured pins