COMPARISON OF RADIOACTIVE DOSES BEYOND THE FINAL SHIELDING
INSIDE REACTOR BUILDINGS OF RUSSIAN VVER-1000 AND GERMAN
PWR-1300
N. M. Sharifloo
Imam Hossein University, Tehran -Iran
M. Gholampour*
Moscow Power Engineering Institute (Technical University)
Various shielding layers are used in pressurized water reactors (PWR and VVER) to
reduce the radioactive doses released from reactor core. Through each shielding layer, a
portion of the neuron and Gamma doses is shielded. To prevent and decrease risks of
LOCA event and considering the vibrations of the pressure vessel, an air gap between the
pressure vessel and concrete shield is envisaged in the design. Such gap causes the axial
streaming of neutron and gamma radiations. Since intensive neutron and gamma doses in
the core couses damage of systems and restrict of access to the zones, estimation of the
maximum permissible radiation dose is too important.
MCNP code was used to calculate radiactive doses in both VVER-1000 and PWR1300. The results of running MCNP code have been shown in tables and diagrams. Results
of comparing radioactive doses beyond the biological shielding layer inside reactor
building in Russian VVER-1000 (Bushehr Nuclear Power Plant) and German PWR-1300
(German ex-design of BNPP) are presented.
To ensure the accuracy of computations, obtained results have been compared with
certified and approved data given in the PSAR and FSAR (Preliminary and final Safety
Analysis Report) of Bushehr Nuclear Power Plant. Comparison depicts the high accuracy
of computations using MCNP and approves validity of modeling process. The obtained
results indicate that in PWR-1300 reactors (such as German ex-design of BNPP) personnel
can sustain an annual duration of 109.97 hours inside the reactor pit, while this value for
VVER-1000 reactors rates up to 113.29 houres per year. This shows that the post-shielding
radiation in VVER-1000 reactors has lower rate in compare to PWR-1300’s.
The total flux in PWR- 1300 is a bit higher than VVER- 1000. The shielding layers
in Busheher NPP is disgned and constructed for PWR-1300, but after reconstruction of this
unit, PWR – 1300 replaced by VVER – 1000. Considering total flux of these reactor types,
it is concluded that Busheher reactor is shielded better than typical Russain NPPs with
VVER -1000 and the received dose beyond shielding layers of Busheher reactor is
proportionally less than the PWR-1300.
1. Shielding in the pressurized reactors
High amounts of neutron and gamma rays in the upper parts of the core while reactor
is in operation, damages the existing systems or actuates them and restricts the access to
those parts. Since personnel shall attend these parts for testing purposes, accurate
estimating of radiation dose and shielding effectiveness is of particular importance [1, 2].
In this research, streaming of the neutron and gamma radiations within different shielding
layers along the reactor axis during operation for VVER-1000 and PWR-1300 reactors has
been computed by MCNP code. The flux and dose rate values in different shielding layers
of these two reactors have been calculated with MCNP codes, too.
2. Calculation of flux and dose in various shielding layers of reactor
The neutron and gamma rays radiation through the core and shields of VVER-1000
and PWR-1300 reactors during reactor operation was modeled with MCNP code and the
flux and dose values in working conditions were calculated. Table 1 showes specifications
of the protection layers which constitute shielding in the VVER-1000 and PWR-1300
reactors of BNPP [3,4,5,6]. To find the mass percentage of the elements in the core (and
shielding layers), knowing the density and volume of each element, we measure the
materials mass and having determined the mass of each element, we calculate the total
1
mass for all elements of the core and shielding layers. Then we can find the mass
percentage (proportion) for each of those elements.
Table 1
Specification of protection layers which constitute shielding in the VVER-1000 (BNPP1)
and PWR-1300 reactors (exdesign of BNPP).
Shielding layer
1
2
3
4
5
6
7
8
9
10
11
Core
Core baffle
First Coolant
Core barrel
Second
Coolant
Reactor
pressure vessel
First air gap
Insulation
inside
biological
shielding
Second air gap
Outside
biological
shielding
Material composition
PWR1300 & VVER1000
Fuel + Zr
Steel
Coolant
Steel
Density
VVER
PWR
4.1
4.1
7.58
78.5
0.71
0.71
7.58
7.85
Thickness
VVER
PWR
158
174.5
2.5
2.5
35.5
35.5
8.0
8
Coolant
0.71
0.71
31.5
31.5
Steel
7.58
7.85
25.6
25.6
Air
Polyethylene
0.0
0.1
0
0.9
84
2.4
84
2.4
Concrete
2.14
2.14
55
55
Air
0.0
0
60
60
Concrete
2.2
2.2
140
140
2.1. Calculation of the group spectrum for the reactor core gamma rays
These rays fall into two (2) parts [7, 8]:
2.1.1.
Measuring the group spectrum of the fission prompt gamma rays
Volume density of the fission prompt gamma sources in the energy state “G” is
measured by the following relations:
Sγfg=3.13x1010gfρ

cm 3 .s
(1)
where the volume density of power is defined per W/cm3 and gf as follows:
gf=gf(E)dE
(2)
From the equation No.2, total number of the prompt gamma rays generated in the state “g”
is measured for each fission. In this equation, F(E) in various energy ranges is measured
by the following experimental equations:
F ( E)  60.94 0.1  E  0.6 Mev

Fission  Mev
F ( E)  20.2 exp( 1.78E) 0.6  E  1.5Mev
F ( E)  7.2 exp( 1.09E) 1.5  E  10.5Mev
(3)

Fission  Mev
(4)
Fission  Mev
(5)

2.1.2. Calculations of the group spectrum of the gamma rays generated by the decay
of fission fractions
Volume density of the gamma sources generated by the fission fractions decay in
the energy state “g” is measured with the following equations:
2
SFP
g
 3.13 x1010 FP 

g
(6)
cm 3 .s
Where ρ is the volume density of power per W/cm3, and FPg is defined as:
FP g 

g FP ( E )dE
(7)
FP ( E )  7.4 exp(1.1E )

(8)
Fission  Mev
Values of Fg and FPg for 13 energy groups of gamma rays have been calculated and
represented in table 2.
Table 2
g
g
Values of Prompt Fission Gamma (F ) and Decay Gamma (FP ) for 13 energy
groups of gamma rays.
Group
No.
1
2
3
4
5
6
7
8
9
10
11
12
13
14
Energy
Band
(Mev)
8-14
6-8
4-6
3-4
2.5-3
2-2.5
1.5-2
1-1.5
0.7-1
0.45-0.7
0.3-0.45
0.15-0.3
0.1-0.15
0-0.1
Total
Prompt Fission
Gamma (No. Per
Fission)Fg
1.008 E-3
8.465E-3
7.489E-2
1.677E-1
1.821E-1
3.140E-1
5.415E-1
1.129E0
1.352E0
1.679E0
1.042E0
1.042E0
3.473E-1
Neglected
7.880E0
Decay Gamma
(No. Per
Fission) FPg
9.494 E-4
8.140E-3
7.346E-2
1.656E-1
1.821E-1
3.156E-1
5.470E-1
9.481E-1
8.762E-1
9.870E-1
7.363E-1
8.648E-1
3.229E-1
Neglected
6.028E0
Total (No. Per
Fission)
y
1.975 E-3
1.661E-2
1.484E-1
3.323E-1
3.642E-1
6.296E-1
1.089E0
2.077E0
2.228E0
2.666E0
1.778E0
1.907E0
6.702E-1
Neglected
1.3908E1
Fraction
1.407E-4
1.194E-3
1.067E-2
2.389E-2
2.619E-2
4.527E-2
7.826E-2
1.493E+1
1.602E-1
1.917E-1
1.279E-1
1.371E-1
4.819E-2
Neglected
1.000
Also, quantitative values and fluctuation curves of Fg FPg corresponding to energy state
are given in Fig.1.
Values given in table 2 which indicate the energy state, have been used as the input data
for MCNP code in order to calculate the core gamma rays flux and dose within various
shielding layers while operating German PWR-1300 and Russian VVER-1000 reactors.
3
Fig. 1 - Energy spectrum of fission, decay and total gamma in core.
2.2. Measurement of the neutron and photon source intensity during reactor operation
[9,10]
Considering that the majority of neutrons generated in the core are prompt
neutrons, therefore only prompt neutrons have been covered by the calculations. The Watt
Spectrum has been used for calculating the flux, neutron dose, and gamma capture rays.
For neutron sources in the energy state “g”, we have
n
S g  3 / 13 E10 X g v 3
(9)
cm . sec
Where “ν” is equal to 2.45 neutrons per fission and ρ is the power volume density
(Watt/cm3) Value of Xg is obtained by the following equation:
X g   g x( E )dE
X (E) 
2

(10)
Sinh 2 E exp(  E )
(11)
Where E is measured with Mev.
Calculations were made assuming the PWR-1300 reactor running in 3300 MW thermal
power, while it was assumed 3000 MW for VVER-1000. Due to the MCNP code
computing the flux and dose per one (1) neutron/ photon, the neutron source intensity (S)
for both reactor types is obtained this way:
Neutron source in PWR-1300:
S  3300 MV 
2.45 n 10 6 Watt 3.13 EFission


 2.53 E 20 n / Sec
Fission
1MW
Watt  Sec
(12)
Neutron source in VVER-1000:
S  3000 MW 
2.45 n
10 6 Watt 3.13 E10 Fission


 2.35 E 20 n
Sec
Fission
1MW
Watt  Sec
(13)
Having Watt Spectrum and the neutron source intensity, the MCNP code has been used for
computing the neutron flux and dose and capturing gamma rays in various shielding layers
of the said two types of reactor.
4
For calculating the photonic source, we have:
S  3.131010  T P
(14)
Where P is the reactor power measured by W and T is the total sum of gamma rays
generated per fission, which is in turn obtained by the following:
 

g

g
(15)
Where:
 g   f  FP
g
g
(16)
Fg and
FPg
Now, substituting the values of
from Table 2 would give:
γ=Fg+FPg= 7.880EO+6.028EO=13.908
Using equation No.14, we calculate the neutron source (S) value for VVER-1000 and
PWR-1300 reactors.
Photon source in PWR-1300:
S= 3.13E10x13.908x3300E6=1.436E21 γ/sec
(17)
Photon source in VVER-1000:
S= 3.13E10X13.908X3000E6=1.306E21 γ/sec
(18)
With the group spectrum of gamma rays from Table 2 and photon source intensity
values, it is possible to run MCNP code for obtaining the in-core gamma rays flux and
dose in various shielding layers of both types of reactors in working conditions.
2.3. The flux and dose rates calculated for various shielding layers
Having calculated the neutron flux and gamma rays, we can measure the
equivalent dose rate for each ray (neutron or gamma) by the following
N
relation: D 
 F
i
(19)
i
i 1
Where N represents the number of energy group with i as neutron (photon) flux in each
of energy group. Fi is the flux to dose conversion factor for each energy group. Values of
the neutron flux, gamma flux, and the total gamma and neutron flux for PWR-1300 and
VVER-1000 reactor are given in Table3 and 4 for comparison.
Table 3
The neutron and gamma fluxes in different shielding layers for PWR-1300 (German)
Neutron & Gamma
Neutron flux
Total Gamma flux
Distance from center
flux
1
1
of core in horizontal
1
axial (cm)
cm 2 . sec
cm 2 . sec
2
cm . sec
177
220.5
252
277.6
364
419
619
E13 7.9973
1.4107E11
8.8659E8
7.9172E8
6.1627E8
3.3907E7
1.2607E2
9.8957E12
1.1366E11
6.2136E9
1.7963E8
5.5280E8
2.6458E7
1.4931E2
5
8.9869E13
2.5473E11
7.1002E9
9.7135E8
11.6907E8
6.0365E7
2.7538E2
Table 4
Compare of Neutron and Gamma fluxes in different shielding layers of VVER-1000
(Russian)
Neutron & Gamma
Neutron flux
Total Gamma flux
Distance from center
flux
1
1
of core in horizontal
1
axial (cm)
cm 2 . sec
cm 2 . sec
2
cm . sec
160.5
204
235.5
260.6
375
402
602
6.9947E13
1.5063 E11
2.4971 E9
8.9153 E8
8.7250 E8
3.4152 E7
1.8265 E2
1.6127 E13
1.1771 E11
5.0631 E9
2.8186 E8
2.7125 E8
2.4717 E7
1.1496 E2
8.6074 E13
2.6834 E11
7.5602 E9
11.7339 E8
11.4375 E8
5.8869 E7
2.9761 E2
To compare the total flux and effects of the shielding layers in PWR-1300 and
VVER-1000 reactors, the neutron, gamma and total flux and neutron dose, gamma dose
and total dose rate have been calculated with MCNP code for various shielding layers.
After calculating the neutron and gamma flux while the flux conversion factor to
the equivalent dose rate for neutrons and gamma rays are available, values of these doses
were found for both PWR-1300 and VVER-1000 reactors. Using the flux to dose rate
conversion factor given in Table 5 [11] and, we calculated the equivalent dose for neutron
and gamma rays (Table 6 and 7).
Table 5 A.
Neutron Flux -to - Dose Rate conversion factors and Quality Factors
Neutron
E (Mev)
2.5E-08
1.0E-07
1.0E-06
1.0E-05
1.0E-04
1.0E-03
1.0E-02
1.0E-01
5.0E-01
1.0
2.0
2.5
5.0
7.0
10.0
14.0
20.0
NCRP-38, ANSI/ANS-6.1.1-1977*
DF (E)
Quality
2
(rem/hr)(n/cm .s)
Factor
3.67E-06
2.0
3.67E-06
2.0
4.46E-06
2.0
4.54E-06
2.0
4.18E-06
2.0
3.76E-06
2.0
3.56E-06
2.5
2.17E-05
7.5
9.26E-05
11.0
1.32E-04
11.0
ICRP-2
DF (E)
Quality Factor
(rem/hr)(n/cm2.s)
3.85E-06
2.3
4.17E-06
2.0
4.55E-06
2.0
4.35E-06
2.0
4.17E-06
2.0
3.70E-06
2.0
3.57E-06
2.0
2.08E-05
7.4
7.14E-05
11.0
1.18E-04
10.6
1.43E-04
9.3
1.25E-04
9.0
1.56E-04
8.0
1.47E-04
7.8
1.47E-04
7.0
1.47E-04
6.5
1.47E-04
6.8
2.08E-04
7.5
2.27E-04
8.0
1.54E-04
6.0
*Extracted from American National Standard ANSI/ANS-6.1.1-1977
6
Table 5. B
Photon Flux -to - Dose Rate conversion factors
ANSI/ANS-06.1.1-1977
E (Mev)
DF (E)
(rem/hr) (photons/cm2.s)
0.01
3.96E-06
0.03
5.82E-07
0.05
2.90E-07
0.07
2.58E-07
0.1
2.80E-07
0.15
0.79E-07
0.2
5.01E-07
0.25
6.31E-07
0.3
7.59E-07
0.35
8.78E-07
0.4
9.85E-07
0.45
1.08E-06
0.5
1.17E-06
0.55
1.27E-06
0.6
1.36E-06
0.65
1.44E-06
0.7
1.52E-06
0.8
1.68E-06
1.0
1.98E-06
1.4
2.51E-06
1.8
2.99E-06
2.2
3.42E-06
2.6
3.82E-06
2.8
4.01E-06
3.15
4.41E-06
3.75
4.83E-06
4.25
5.23E-06
4.75
5.60E-06
5.0
5.80E-06
5.25
6.01E-06
5.75
6.37E-06
6.25
6.74E-06
6.75
7.11E-06
7.5
7.66E-06
9.0
8.77E-06
11.0
1.03E-05
13.0
1.18E-05
15.0
1.33E-05
E (Mev)
0.01
0.015
0.02
0.03
0.04
0.05
0.06
0.08
0.1
0.15
0.2
0.3
0.4
0.5
0.8
1
1.5
2
3
4
5
6
8
10
7
ICRP-2
DF (E)
(rem/hr) (photons/cm.2.s)
2.78E-06
2.78E-06
1.11E-06
5.88E-07
2.56E-07
1.56E-07
1.20E-07
1.11E-07
1.20E-07
1.47E-07
2.38E-07
3.45E-07
7.69E-07
9.09E-07
1.14E-06
1.47E-06
1.79E-06
2.44E-06
3.03E-06
4.00E-06
5.56E-06
6.25E-06
7.69E-06
9.09E-05
Table 6
Compare of Neutron and Gamma dose rates in different shielding layers of PWR-1300
German Design.
Radius (cm)
Neutron dose rate
rem/h
177
220.5
252
277.6
364
419
619
4.5302E9
1.4238E7
2.2940 E5
4.8306 E4
4.0158 E4
4.5479 E2
3.1060 E-3
Total Gamma
dose rate
rem/h
2.5927E8
5.1018E6
1.3749E6
2.9714E3
2.3087E3
1.7335E2
1.9627E-2
Neutron and Gamma
dose rate
rem/h
4.7894E9
1.9339 E7
1.6043 E6
5.1277 E4
4.2466 E4
6.2814 E2
2.2733 E-2
Table 7
Compare of Neutron and Gamma dose rates in different shielding layers of VVER-1000
(Russian) during operation.
Neutron dose
Total Gamma dose Neutron and Gamma
Radius (cm)
rate (rem/h)
rate (rem/h)
dose rate (rem/h)
160.5
4.5013 E9
2.4886 E8
4.7501 E9
204
1.4981 E7
5.0571 E6
2.0038 E7
235.5
2.0479 E5
2.0791 E6
2.2838 E6
260.6
4.8913 E4
2.8965 E3
5.1809 E4
375
4.1653 E4
1.7635 E3
4.3416 E4
402
4.5987 E2
1.5790 E2
6.1777 E2
602
3.1671 E-3
1.8900 E-2
2.2067 E-2
Table 8 represents total (neutron and gamma) flux, as well for various shielding
layers of PWR-1300 and VVER-1000 reactor. Neutron and gamma doses have also been
calculated for various shielding layers of PWR-1300 and VVER-1000 reactors and the
relevant values have been given in the Tables 9 &10.
Table 8
Compare of total flux (Neutron and Gamma) in different shielding layers of German
PWR-1300 and Russian VVER-1000, during their operation.
Calculated points
Beyond Core shroud (baffle)
Beyond Core Barrel
Before RPV
Beyond RPV
Before inner Concrete layer
Beyond inner Concrete layer
Beyond biological shielding
Total Flux
(PWR-1300) 1/cm2.sec
8.9869E13
2.5473E11
7.1002E9
9.7135E8
11.6907E8
6.0365E7
2.7538E2
8
Total Flux
(VVER-1000) 1/cm2.sec
8.6074E13
2.6834E11
7.5602E9
11.7339E8
11.4375E8
5.8869E7
2.9761E2
Table 9
Compare of neutron dose rate in different shielding layers for German PWR-1300 and
Russian VVER-1000.
Calculated points
Beyond Core shroud (baffle)
Beyond Core Barrel
Before RPV
Beyond RPV
Before inner Concrete layer
Beyond inner Concrete layer
Beyond biological shielding
Neutron dose rate
(PWR-1300) rem/h
4.5302E9
1.4238E7
2.2940E5
4.8306E4
4.0158E4
4.5479E2
3.1060E-3
Neutron dose rate (VVER1000) rem/h
4.5013E9
1.4981E7
2.0479E5
4.8913E4
4.1653E4
4.5987E2
3.1671E-3
For comparison values of the calculated doses for both reactors are given in the Table 11.
Table 11
Compare of total dose rate in different shielding layers for German PWR-1300 and
Russian VVER-1000
Total dose rate
Total dose rate
Calculated points
(PWR-1300) rem/h
(VVER-1000) rem/h
Beyond Core shroud (baffle)
4.7894E9
4.7501E9
Beyond Core Barrel
1.9329E7
2.0038E7
Before RPV
1.6043E5
2.2838E6
Beyond RPV
5.1277E4
5.1809E4
Before inner Concrete layer
4.2466E4
4.3416E4
Beyond inner Concrete layer
6.2814E2
6.1777E2
Beyond biological shielding
2.2733E-2
2.2067E-2
After finding values of the neutron and gamma fluxes and corresponding doses of these
rays for PWR-1300 and VVER-1000 reactors, the values were compared with those given
in the BNPP PSAR as benchmark of the calculations. Figs.2-6 show the results of
comparing total flux and dose values for both reactors with the ones presented in the
Benchmark. As seen in the figures, the values found through the calculations are very
close to those given in the PSAR.
Fig. 2- Comparison between calculated flux and benchmark
in German PWR-1300.
9
Fig.3- Comparison between calculated flux and benchmark
in Russian VVER-1000.
Fig.4- Comparison between total calculated dose rate and benchmark in PWR-1300.
Fig.5- Comparison between total calculated dose rate and benchmark
in VVER-1000
10
Fig. 6- Comparison between neutron and gamma fluxes in
German PWR-1300 during operation.
A comparsion between neutron and gamma fluxes and dose rates in German PWR-1300
and VVER-1000 carraied out. The results have showned in Figs 7-10.
Fig.7- Comparison between neutron and gamma doses in
German PWR-1300 during operation.
Fig. 8- Comparison between neutron and gamma fluxes in Russian VVER-1000 during
operation.
11
Fig. 9- Comparison between neutron and gamma doses in Russian VVER-1000 during
operation.
Fig.10- Comparison between total flux in different shielding layers of
Russian VVER-1000 and German PWR-1300 during operation.
12
3. Comparison of the calculation results
Before proceeding with comparison of the results obtained by utilizing MCNP code, it is
necessary to review the radiation working regulations.
It is known that the personnel’s occupational exposure shall be controlled in such a way
not to exceed maximum rate of 50 msv per year [12] and exceeding this amount is not
permitted in working conditions. Therefore, using the output dose of the PWR-1300 and
VVER-1000 reactors obtained by MCNP code, we determine the permissible working
hours for the radiation workers of both reactors and compare the results with each other,
having the current standard as benchmark.
As shown in Table 11, the output dose rate beyond the final concrete layer of the PWR1300 and VVER-1000 reactors is 2.2733 E-3 rem/h and 2.2067E-2 rem/h respectively.
Table 11
Compare of gamma dose rate in different shielding layers for German PWR-1300 and
Russian VVER-1000
Calculated points
Beyond Core shroud (baffle)
Beyond Core Barrel
Before RPV
Beyond RPV
Before inner Concrete layer
Beyond inner Concrete layer
Beyond biological shielding
Gamma dose rate
(PWR-1300) rem/h
2.5927E8
5.1018E6
1.3749E6
2.9714E3
2.3087E3
1.7335E2
1.9627E-2
Gamma dose rate
(VVER-1000) rem/h
2.4886E8
5.0571E6
2.0791E6
2.8965E3
1.7635E3
1.5790E2
1.8900E-2
The following calculations have been performed to find the permissible working hours
beside each of the reactors.
3.1. PWR-1300
Total Dose Rate of Final Concrete= 2.2733 E-2 Rem/h=2.2733 E-4 sv/h =
E-1 msv/h=1991.41 msv/year
Allowable working hours= Standard Permissible Dose Received per Year
Output Dose of the Reactor Final Concrete
= 50 msv/year
=219.94 h/year × 1/2 (Safety Factor) = 109.97 h/year
2.2733 E-1 msv/h
3.2. VVER-1000
Total Dose Rate of Final Concrete= 2.22067 E-2 Rem/h=2.22067 E-4 sv/h =
2.22067 E-1 msv/h=1933.07 msv/year
Allowable working hours= Standard Permissible Dose Received per Year
Output Dose of the Reactor Final Concrete
= 50 msv/year
=226.58 h/year × 1/2 (Safety Factor) = 113.29 h/year
2.22067 E-1 msv/h
Considering the calculation results and the output dose upon the final concrete
shielding layer of VVER-1000 reactor being lower than that of the PWR-1000, maximum
permissible working hours beside the Russian VVER-1000 reactor seems to be more
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convenient than the German PWR-1300 reactor (113.29 hours per year versus 109.97
hours).
The “1/2” factor in the above-mentioned relations is the safety factor for the radiation
workers (those who are under radiation exposure). The purpose of using this factor is to
reduce the permissible working hours for those who work adjacent to the reactor core by
half. Since these personnel may be exposed to radiation outside their working site, i.e.
when they are subject to medical X-Ray (MRI and etc.), within borders of nuclear power
plant, or by natural background radiation, their annual exposure rate exceeds the
permissible 50 msv and therefore their permissible working hours are reduced to half, so
that their received dose shall not exceed the permissible rate in any case.
There are two important points worthy of note regarding the said working hours:
A) Assuming that each of these personnel has an average two months leave and works
5 days a week (each working day includes 6 working hours) during a whole year
(365 days), each person generally attends the work place for duration of 201 days
or total 1206 hours in the power plant. Also considering the working nature of this
part, depending on the working conditions (repairs, inspections), it is not possible
to predict any preset time for attending beside the reactor core and depending on
the daily tasks, the duration of attending beside the reactor core can vary from a
few minutes to tens of minutes a week. Working hours in the direct exposure
conditions is anyway quite limited, so that even if in the worst conditions the
working hours beside the core reaches to two hours per week, it will sum up to 80
hours of working beside the core during the whole year. It shall be noted, of course,
that due to the system’s high safety level and the least possible need to continuous
repairs and beside that, assigning numerous personnel to simultaneous work beside
the core and reducing the exposure rate by reducing the repairs time this way, such
amount of working hours would never realize and this amount is usually half the
value mentioned above.
Considering the above, it can be concluded that both reactors are looking desirable
in view of permissible working hours versus shielding, and the calculations depict
proper assignment of the tasks to the personnel working in the core vicinity.
Applying the safety factor of “1/2” in calculating the working hours can even
enhance the personnel’s protection ratio.
B) Comparison of the said two reactors in view of the permissible working hours
beside the core leads the following conclusion:
While the initial dose in both reactors is almost the same (table 4-39), it can be seen that
the final output dose of the BNPP reactor is proportionally less than the PWR-1300
(2.2067×10-2 Rem/h versus 2.2733×10-2 Rem/h).
4. Conclusion:
- Results of running MCNP code have been shown with view to the tables and
diagrams. Analysis of the above-mentioned calculations have been performed
bearing in mind the importance of knowing the permissible working hours of the
personnel near the core (beyond the final concrete), which makes the ultimate goal
of this research as well.
- In-core radiation neutron and gamma flux values of the PWR-1300 and VVER1000 reactors, and gamma dose, neutron dose and the total dose for each of the
shielding layers were calculated by MCNP code.
- Considering the importance of knowing the permissible working hours of the
personnel near the core (beyond the final concrete), this duration has been
determined for both reactor types.
- The effective permissible exposure value for the radiation workers is 50 msv per
year. Based on this, the annual permissible working hours has been determined to
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be 109.97 hours for PWR-1300 reactor and 113.29 hours for VVER-1000 ( in case
of BNPP).
The total flux in PWR- 1300 is a bit higher than VVER- 1000. The shielding layers
in Busheher NPP is disgned and constructed for PWR-1300 , but after
reconstruction of this unit, PWR – 1300 replaced by VVER – 1000 . Considering
the total flux of these reactor types and taking in acount that the German NPPs with
PWR -1300 is better shielded than Russian NPPs with VVER-1000, it is
concluded that The Busheher-I NPP is shielded better than typical Russain NPPs
with VVER -1000 and the received dose beyond shielding layers of Busheher
reactor is proportionally less than the PWR-1300.
References
1- C. Francis, etal. (1979): "Radiation streaming and Reactor cavity shield Design at
TVA PWR Plants; "ORNL/RSIC-43 PP 149-162.
2- Gilles champion etal. (1985); "New Trends in shielding Designs for PWR in
France"; Tran sections of the American Nuclear society USA; USA, Vol. 50, PP
480-481.
3- "Technical Assignment for Reconstruction and completion project of the Busherhr
NPP unit 1"; Radiation safety main buildings and structures of power unit.
Moscow, 1999.
4- Atomic Energy organization of Iran (AEOL); "Preliminary Report on safety
Assurance"; PSAR; 2000.
5- Preliminary safety Analysis Report"; Chap.12, Radiation protection. Moscow,
1999.
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Nuclear Technology"; physics, Trieste, Italy (12 Feb-16 Mar 1990). World
scientific publishing Co., pte ltd. Singapour.
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Report ORNL/RSIC-52, PP.5-20.
10- C.Devillers and J.P. Payen (1977); "Shielding against Neutron steaming in
pressurized-water Reactors, in proceedings of National Transportation safety Board
meeting"; Washington D.C., USA PP.1-7.
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Department of Iran, National Radiation Protection Division.
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