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)
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
Flux density [1/cm2]
1E+7
1E+5
1E+3
1E+1
1E-1
1E-3
1E-9
1E-7
1E-5
1E-3
1E-1
1E+1
Energy [MeV]
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
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
>1MeV
Near baffle
Near gap
Inner
Corner
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
0.06163
0.0253eV
0.06214
0.06220
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
0.75%
0.50%
ratio [-]
0.25%
0.00%
-0.25%
-0.50%
-0.75%
0
2
4
6
8
10
Energy [MeV]
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 VVER1000 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)
C/E
•
Comparison of symmetrical pins used
for verification of experiment
1.35
1.30
1.25
1.20
1.15
1.10
1.05
1.00
0.95
0.90
0.85
0
5
10
15
MOBY DICK
–
•
25 30 35
pin position
MCNPX
40
45
50
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
0.25
0.20
0.15
0.10
0.05
0.00
-0.05
-0.10
-0.15
-0.20
0
5
10 15 20 25
55
1s uncertainty
Near baffle, better agreement with
MCNP than with MOBY DICK
P/P(inv)-1
•
20
30 35 40 45
50 55
Pin position
Experiment
MCNPX
1s experimental interval
1s calculational interval
Axial profile of power density C/E
• Discrepancies in distant grids locations
5500
5000
Power [a.u.]
4500
4000
3500
3000
2500
2000
1500
1000
0
25
50
75
100
125
Axial position [cm]
MCNPX & ENDF/B VI.2.
Benchmark data
Measurement
Neutron fluxes in reflector
Neutron flux density [a.u.]
1E+1
1E+0
1E-1
1E-2
1E-3
1E-4
0.1
1
Energy [MeV]
Pt-2
pt2 -calc
Pt-21
pt-21 calc
Pt-22
pt-22 calc
10
Pt-23
pt-23 calc
Neutron fluxes in RPV
Neutron flux density [a.u.]
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
0.1
1
Energy [MeV]
10
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
1.006
1.004
– keff
– Slight variations if used
– ENDF/B VII & S(α, β)
results closer to experiment
Keff
1.002
1
0.998
0.996
0.994
2.75
ENDF VI
3.25
ENDF VII
3.75
H3BO3 [g/kg]
4.25
ENDF VI free gas
ENDF VII free gas
• Fe
0.3
0.25
Attenuation ratio
– Photon flux density (18cm Fe)
– Notable variations if used
– ENDF/B VII & S(α, β)
results closer to experiment
0.2
0.15
0.1
0.05
0
>1MeV
>3MeV
>5MeV
Energy group
Free gas
TSL
Experiment
>7MeV
Nuclear data library effect - fuel
H [cm]
ρ [g/kg]
1.006
1.004
1.002
Keff
• 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
1
0.998
0.996
0.994
2.75
3.25
3.75
H3BO3 [g/kg]
4.25
4.75
ENDF VI.2
ENDF VII
JEFF 3.1.
JENDL 3.3.
JENDL 4
RF
CENDL
1S
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
• Neutrons (thick layers)
– Most notable discrepancies
(4–7 MeV) for JENDL 4
(C-E)/E
Nuclear data library effect – Fe (18 cm slab)
and TENDL 2009
30%
25%
20%
15%
10%
5%
0%
-5%
-10%
-15%
-20%
1
2
3
4
5
6
7
8
9
10
• Photons
– Most notable discrepancies
(>7MeV) for JEFF 3.1
and TENDL 2009
Photon flux density [cm-2.s-1]
Energy [MeV]
ENDF VI.2
ENDF VII
JEFF 3.1.
JENDL 3.3.
JENDL 4
ROSFOND 2009
CENDL 3
TENDL 2009
1s uncertainty
1E+5
1E+4
1E+3
1E+2
1E+1
1E+0
0
1
2
3
4
5
6
7
8
9
10
Energy [MeV]
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 theVVER1000 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
• The variation are smaller
than related uncertainties
M.D./MCNP -1
Influence of power distribution on results
=> Diffusion approximation
power density may be
used in following
transport calculations
0.8%
0.6%
0.4%
0.2%
0.0%
-0.2%
-0.4%
-0.6%
-0.8%
-1.0%
0
1
2
3
4
5
6
7
8
9
Energy [MeV]
Pt-2
Pt-3
Pt-7
2.0%
M.D./MCNP-1
1.5%
1.0%
0.5%
0.0%
-0.5%
-1.0%
-1.5%
1
3
5
7
Energy [MeV]
P-3
P-7
9
10
3He reaction rate attenuation
• In RPV simulator of VVER-1000
<0.55eV
ENDF VII
>0.55eV
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 with PE liner
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
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
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%
• 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
6.34%
2.90%
0.32%
-0.08%
-0.16%
12.03%
5.49%
0.61%
-0.15%
-0.30%
near baffle
corner