Radiation shielding analysis of a large helical device

ELSEVIER

Fusion Engineering and Design 28 (1995) 515-524

Fusion Engineering and Destgn

Radiation shielding analysis of a large helical device Hiroyuki Handa a, Katsumi Hayashi a,,, Hirokuni Yamanishi b, Yoichi Sakuma b, Hiroshi Kaneko b, Haruo Obayashi b, Osamu Motojima b, Yuichi Ogawa c, Koubun Yamada d, Teruo Abe d a Hitachi Engineering Co. Ltd., 3-2-1 Saiwai-cho, Hitachi, Ibaraki 317, Japan b National Institute for Fusion Science, Furo-cho, Chikusa-ku, Nagoya 464-01, Japan ° Department of Quantum Engineering and Systems Science, University of Tokyo, Hongo, Bunkyo-ku, Tokyo 113, Japan d Business Automation Co. Ltd., 1-24-10 Toranomon, Minato-ku, Tokyo 105, Japan

Abstract Radiation shielding analysis of a large helical device (LHD) has been performed to provide information on a radiation environment needed for equipment design and operation planning. First, the applicability of the JENDL-3 library which was applied to the shielding calculations of the LHD was tested by means of a benchmark experiment on the shielding efficiency of ordinary concrete carried out at the fusion neutronics source facility at JAERI. Benchmark analysis was carried out with the MCNP continuous energy Monte Carlo code to test the JENDL-3 library itself. The same benchmark calculation was carried out with the DOT3.5 code to examine the groupwise cross-section library of FUSION-40 processed from JENDL-3. Secondly, the evaluation of the following items with the DOT3.5 code was revised, using FUSION-40 instead of the formerly used G~CX-40 library: (1) dose distribution inside and outside the L H D building; (2) dose distribution inside the cellar of the LHD building resulting from steamed radiations; (3) Environmental dose distribution resulting from sky-shine effect; (4) activity and dose rate levels of vacuum vessel and superconducting magnet.

1. Introduction The large helical device (LHD) [1] project is aimed at exploring the feasibility of helical plasmas for fusion applications; in particular, to demonstrate the steadystate currentless plasmas confined in the fields generated by superconducting helical and poloidal coils. At present, the machine design is in its final stage, and the building is under construction.

The radiation protection concepts of the LHD in the building design phase [2] are revised and developed in this study. The actual structures of the building and device are taken into account. The newly released nuclear data library J E N D L - 3 [3] was applied.

2. Analysis of benchmark problem 2.1. Objectives

* Corresponding author.

A benchmark calculation is necessary to test the nuclear data of concrete, i.e. the major shield of the

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H. Handa et al. / Fusion Engineering and Design 28 (1995) 515-524

516

~:;Qi~!,<~i:i:~ii!!:!ii!!i!iiiiiii~ii~!iU!i~i~ I ,i ¸ ~ ii ~ ~ i! ~ i~i ~i~i:!il ii ~iiili ~ii!!iiil)

Void

:

:

Table 1 Composition of the concrete assembly for FNS concrete benchmark experiment

-<

~ !<

~6~er~te

iIi 6oo ..............

0.0

20.025.0 22.5 30.0

40.0

60.0

75"080 0

}~...

(cm)

Fig. 1. Geometry of FNS concrete benchmark experiment, LHD. The benchmark experiment [4] was performed at the fusion neutronics source (FNS) facility of the Japan Atomic Energy Research Institute (JAERI). The JENDL3 library and the derived group constant set FUSION-40 [5] were tested by analyzing the experimental data. The old GICX-40 library [6] and the FUSION-J3 [5] detailed group constant set produced from JENDL-3 also were tested for comparison.

2.2. Outline o f benchmark experiment Fig. 1 shows the geometry of the F N S concrete benchmark experiment. The cylindrical concrete assembly was 600 m m in diameter and 600 m m thick, which was a suitable configuration for checking the two-dimensional calculation. The assembly was made of mortar equivalent in composition to ordinary concrete, in order to avoid the heterogeneity inherent in aggregates. The composition of the assembly is listed in Table 1, as determined by chemical analysis. D e u t e r i u m - t r i t i u m neutrons were produced by bombarding a water-cooled titanium target containing 7 Ci of tritium with a deuteron beam of energy 330 keV. The average neutron (n) strength at the target was 2 x 1011 n s -1 with a deuteron beam current of about 2 mA. Seven kinds of foil, the sensitive energy regions of which varied from thermal to 14 MeV, were selected for measurement of the reaction rates. The set of foils was placed along the central axis of the assembly, as shown in Fig. 1. The positions were 0, 25, 50, 100, 200, 400 and 550 m m from the front surface.

2.3. Method o f analysis The continuous energy Monte Carlo code MCNP4.2 [7] was used to test JENDL-3, by comparing calculated results with those obtained in experiments. Because there are no cross-section data for barium and zinc in

Component

Atomic number density

Estimated error (%)

Silicon Aluminum Iron Calcium Magnesium Sulfur Sodium Potassium Titanium Phosphorus Manganese Barium Vanadium Cobalt Zinc Copper Nickel Carbon Hydrogen Oxygen

1.1370 x 1022 2.4650 x 1021 7.0330 x 102o 5.2209 x 1021 5.3073 x 102° 1.1689 x 1020 6.5574 x 1020 4.3714 x 1020 7.0278 x 1019 2.4489 x 1019 3.0154 × 1019 1.0463 × 1019 5.249 x 1018 2.27 x 1017 1.43 × 1018 6.52 x 10t7 2.73 × 1017 2.1264 x 1020 9.3507 x 1021 3.9484 x 1022

(--+2) (-+2) (-+2) (-+2) (-+2) (_+2) (-+2) (-+2) (-+2) (-+2) (-+2) (-+2) ( -+ 10) ( -+ 10) ( -+ 10) ( -+ 10) ( _+10) (-+2) (-+2) (_+2)

JENDL-3 substitution of copper data was made. Since they were very minor as components, only a little change in the results would be expected. The two-dimensional tin code DOT3.5 [8] was used to test FUSION-40 (neutron 42 group, photon 21 group, P5 Legendre expansion), i.e. the group constant set produced from JENDL-3. The GRTUNCL [8] code was applied to the calculation of the first collision source for the DOT3.5 calculation, to avoid the ray effect resulting from a point source. The calculation was performed using the GRTUNCL and DOT3.5 codes as in the case of using FUSION-40 to test GICX-40 (neutron 42 group, photon 2l group, P5 Legendre expansion), i.e. the group constant set mainly produced from ENDF/B-III. Because cross-section data for sulphur, phosphorus, manganese and cobalt are not included in the GICX-40 library, the sulphur and phosphorus were replaced by silicon and the manganese and cobalt by iron. The FUSION-J3 (neutron 125 group, photon 40 group, P5 Legendre expansion) library, produced from JENDL-3 was also tested in the same way as FUSION-40.

2.4. Results and discussion The C/E (calculated/experimental) values for the reaction rates of the activation foils are summarized in

H. Handa et al. / Fusion Engineering and Design 28 (1995) 515-524 Table 2. T h e JENDL-3 library gives good agreement between the m e a s u r e d d a t a a n d calculated results, except for Ni-58(n,2n) a n d Au-197(n,7). W e m a y d o u b t the activation cross-section d a t a for Ni-58(n,2n), because the C/E value remains at 0.9 or less everywhere. The Au-197(n,~,) is mainly the result o f t h e r m a l n e u t r o n reaction. A m a x i m u m s t a n d a r d deviation for Au197(n,7) is a b o u t 0.2, which is n o t very good. The variance reduction technique applied in this analysis should be i m p r o v e d w h e n the analysis from 14 M e V to t h e r m a l n e u t r o n s is p e r f o r m e d by the M o n t e Carlo method. It was confirmed t h a t JENDL-3 itself is applicable to t r a n s p o r t analyses.

517

FUSION-J3 gives generally good agreement between the m e a s u r e d data a n d calculated results, because FUSION-J3 has so m a n y groups. However, it is observed t h a t F U S I O N - J 3 tends to underestimate in t h i n n e r concrete regions, a n d better agreement with the m e a s u r e d data is attained in thicker concrete regions. T h e FUSION-40 a n d the GICX-40 codes give almost the same C / E values, except for Au-197(n,7). However, because the g r o u p structure o f the activation cross-section is n o t as fine, the calculated results do n o t show good agreement with the m e a s u r e d data, especially for high energy threshold reactions such as Zr-90(n,2n), Ni-58(n,2n) a n d Nb-93(n,2n). This is assumed to be

Table 2 C/E values of activation foils for FNS benchmark experiment analysis Position (ram)

Experimental

C/E MCNP(JENDL-3)

DOT3.5(FUSION-J3)

DOT3.5(FUSION-40)

DOT3.5(GICX-40)

0.99 0.97 1.00 0.96 0.99 1.04 1.12

0.92 0.92 0.93 0.94 0.99 1.04 1.03

0.73 0.71 0.71 0.70 0.71 0.71 0.68

0.73 0.71 0.71 0.70 0.72 0.74 0.73

10 29 10 29 10 30 10 TM 10 31

1.03 1.01 1.02 0.97 0.98 1.02 0.84

0.90 0.90 0.88 0.89 0.93 0.95 0.97

0.90 0.88 0.86 0.84 0.85 0.84 0.84

0.90 0.88 0.86 0.85 0.87 0.88 0.89

Fe56(n,p) Mn56 0 2.258 X 10 - 2 9 25 1.701 x 10 29 50 1.219 X 10 - 29 100 6.747 X 10 - 3 0 200 2.127 × 10 30 400 2.452 × 10 -31 550 5.158 × 10 -32

1.00 0.98 1.03 1.01 0.98 1.04 0.94

0.84 0.84 0.88 0.88 0.91 0.96 0.99

0.92 0.90 0.92 0.90 0.89 0.89 0.89

0.92 0.90 0.92 0.89 0.89 0.91 0.92

Inl 15(n,n')In115m 0 2.709 25 2.707 50 2.233 100 1.624 200 7.102 400 1.164 550 2.678

0.87 0.94 1.04 1.01 1.04 0.99 1.06

0.77 0.81 0.90 0.91 0.96 1.04 1.10

0.80 0.85 0.93 0.93 0.96 1.02 1.05

0.82 0.86 0.93 0.92 0.93 0.94 0.95

Zr90(n,2n) Zr89 0 1.752 25 1.241 50 8.761 100 4.463 200 1.227 400 1.200 550 2.384 Nb93(n,2n)Nb92m 0 9.816 25 7.196 50 5.354 100 2.862 200 8.518 400 9.343 550 1.904

x 10 -28 x 10 -28 × 10 - 2 9

× 10 29 x 10 - 2 9 × 10 - 3 °

x 10 -3I X 10 - 2 9 X 10 - 2 9

x × × x x

X 10 -29 × 10 -29

;:< l0 29 x 10 -29 x 10 3o × 10 -30 × 10

3]

(1.00) a (0.97) (0.97) (0.96) (0.97) (0.97) (0.93)

(1.00) a (0.97) (0.97) (0.96) (0.99) (1.01) (1.00)

518

H. Handa et al. / Fusion Engineering and Design 28 (1995) 515-524

Table 2 (continued) Position (mm)

Experimental

Ni58(n,2n) Ni57 0 9.372 × 25 6.620 × 50 4.614 x 100 2.327 × 200 6.266 x 400 6.228 × 550 1.104 × A127(n,c~)Na24 0 2.537 × 25 1.879 × 50 1.363 × 100 7.564 × 200 2.377 x 400 2.716 × 550 5.698 x Ni58(n,p) Co58 0 7.227 × 25 5.994 × 50 4.806 x 100 3.010 × 200 1.154 × 400 1.732 × 550 3.820 × Au197(n,y) Au198 0 1.328 × 25 2.919 × 50 4.238 x 100 5.971 × 200 4.932 × 400 1.196 X 550 2.356 ×

10 .30

10 -3° 10 -30 10 -30 10 -31 10 .32 10 .32 10 - 2 9

10 -29

10 .29 10 3o 10 .3o 10 -31

10 .32 10 .29 10 .29 10 .29 10 .29 10 -29 10 -30 10 -31

10 -27

10 27 10 .27 10 .27 10 .27 10 .27

10 -28

C/E MCNP(JENDL-3)

DOT3.5(FUSION-J3)

DOT3.5(FUSION-40)

DOT3.5(GICX-40)

0.90 0.88 0.90 0.86 0.88 0.89 0.74

0.87 0.86 0.87 0.88 0.92 0.92 1.00

0.57 0.55 0.55 0.55 0.56 0.53 0.57

0.57 0.55 0.56 0.55 0.57 0.56 0.61

0.95 0.94 0.98 0.94 0.92 0.99 0.89

0.81 0.82 0.84 0.84 0.87 0.91 0.93

0.88 0.87 0.88 0.85 0.85 0.85 0.85

0.88 0.87 0.87 0.85 0.85 0.87 0.59

0.92 0.97 1.01 1.02 0.96 0.96 1.10

0.78 0.82 0.84 0.86 0.90 0.93 1.00

0.98 0.98 0.98 0.97 0.96 0.94 0.99

0.98 0.97 0.96 0.94 0.92 0.88 0.91

0.87 0.81 0.98 1.13 0.84 0.96 0.78

1.09 1.09 1.03 0.95 0.98 1.18 1.35

1.12 1.11 1.05 0.96 0.98 1.17 1.31

0.97 0.92 0.85 0.75 0.70 0.70 0.75

(1.00) (0.96) (0.96) (0.96) (0.98) (0.93) (1.00)

(1.00) (0.96) (0.98) (0.96) (1.00) (0.98) (1.07)

a The values in parentheses are normalized as the values of position 0 to be unity.

caused by a few groups in the high energy region, because these discrepancies begin f r o m the front surface and continue. I f the activation cross-sections o f the FUSION-J3 group structure are collapsed to the FUSION4O g r o u p constant, using the source spectrum as the weighting function, the reaction rates f r o m the original FUSION-40 result in a b o u t 65% for Ni-58(n,2n) and 78% for Zr-90(n,2n) c o m p a r e d with the reaction rates f r o m the collapsed constants respectively. Because it has turned out that the activation cross-section f r o m J E N D L - 3 w a s 9 0 % o f the data value measured by M C N P

analysis, the cross-section takes the value 0.9 x 0.65 = 0.59 and can explain C / E = 0.57. F o r Zr-90, the activation cross-section takes the value 0.87 a n d can almost explain C / E = 0.7-0.74. The difference between FUSION-40 and GICX-40 can be observed only in Au-197(n,7) data within this benchm a r k analysis. GICX-40 gives systematic underestimation, as the n e u t r o n penetrates deeply t h r o u g h the concrete. As a result o f using activation foils, the data concerning intermediate neutrons c a n n o t be obtained directly f r o m this b e n c h m a r k analysis. However, it is

H. Handa et al. / Fusion Engineering and Design 28 (1995) 515-524

._o

1.1

........

I . . . . . . . . J . . . . . . . . . . . . . . . . . J ......... i ......... t .........

z

^

! -

1.0

!

!

i

519

197(n,7) is affected by the higher energy neutrons. The calculational accuracy of GICX-40 can be estimated approximately from the C/E values of the Au-197(n,?) data in relatively thin concrete regions. If C/E = 0.7 in concrete 40 cm thick is extrapolated to the calculational accuracy for concrete 2 m ( = 200 cm) thick, 200/40 = 5, which gives (0.7) 5 = 1/6. In contrast, FUSION-40 gives fairly good agreement, even in the Au-197(n,7) data, and the validity of the FUSION-40for the LHD shielding evaluation was confirmed.

Void

........ i ......... i,, ~ .................................. 0.0

10.6

21.2

31,9

RXIRL

42.5

53.1

63.7

DISTRNCE

74.4

85.0

3. Shielding analysis of the LHD

CM

Fig. 2. Distribution of the ratio of the neutron flux over 10 MeV obtained by GtCX-40to that obtained by vusroN-40. 1.0

o.6 g

Shileding evaluation for the LHD design was performed in the following manner. First, the neutron and gamma-ray distributions in the experimental room were obtained. Then, the duct streaming contributions were calculated to estimate the doses in the cellar. Finally, the neutron and gamma-ray sky-shine were calculated. The activation of the device and the biological dose rate also were estimated for accessibility considerations after shutdown.

0.7

3. I. Dose distribution

z

~ 0.6 0.5

i .................. 4...............

i i i i .......... i .................. i ....

•

'

i

; ~ 21.2

! . 42.5

: : O.O

Void I 10.6

31.9

RXIRL F i g . 3. D i s t r i b u t i o n

Concrete

.

,

i . . 53.1

i

i

.

63.7

OISTRNCE

of the ratio of total neutron

b y GICX-40 t o t h e o b t a i n e d

74.4

: : : 85,0

[CM)

flux obtained

by FUSION-40.

possible that GICX-40 underestimates the intermediate neutrons as well. Therefore, we tried to compare the calculated neutron fluxes. Figs. 2 and 3 show the distribution of the ratio of the neutron flux obtained by GlCX-40 to that attained by FUSION-40. The figures show the ratio of the neutron flux over 10 MeV and the total neutron flux, respectively. For the neutron flux over 10 MeV, all the high energy threshold reactions except for Ni-58(n,p) and In-115(n,n') occur mainly in this energy region. As shown in Fig. 2, the difference between the GICX-40 and the FUSION-40 results is small, and the reasonable small discrepancies in the high energy region can be explained by this result. However, as shown in Fig. 3, the total neutron flux ratio of GICX-40 to that of FUSION-40 gives values lower than unity, as the neutrons penetrate deeply through the concrete. GICX-40 underestimates the total neutron flux compared with FtJS[ON-40. The underestimation of Au-

The dose distribution of the LHD building was obtained using the DOT3.5 code, changing the calculational conditions from the previous calculation, i.e. from G I C X - 4 0 t o F U S I O N - 4 0 f o r the cross-section library; from the asymmetric $146 to the symmetric S160 for the number of angles; from P3 to P5 for the order of scattering; from 3416 to 12 084 for the number of mesh intervals, as shown in Table 3. The conditions for the source neutrons generated from the plasma are described in a previous paper [2]. The calculation model of the experimental room is shown in Fig. 4. The dose rate at the inner surface of the building wall is 3 x 103 mSv per shot, and that outside the building wall is 8.7 x 10 4 mSv per shot. However, in the previous work [2], values of 10 4 and 2.4 x 10 -4 mSv per shot were obtained for the dose rates at the inner surface and outside of the building wall respectively. For the dose rate at the inner surface, the discrepancy by a factor 3 between the previous and present work resulted from the difference in modelling of the main device. For the outside dose rate, the relative attenuation rate in concrete 2 m thick was changed from 2.7 x 10 -8 to 2.9 x 10 -7, and this results from the effect of changing the cross section library from GICX-40 to F U S I O N - 4 0 ( a s discussed in Section 2) and the difference in the calculation conditions as shown in Table 3.

520

H. Handa et al. / Fusion Engineering and Design 28 (1995) 515 524

Table 3 Comparison of the conditions of transport calculation around the LHD building between the previous and present work No.

Item

Previous work

Present work

1 2

Code Library

3 4 5

Number of angles Order of scattering Source neutrons

6 7

Geometry Number of mesh intervals

8 9

Flux convergence Flux calculation model

DOT3.5 GICX-40 (n, 42; 7, 21) Asymmetric S146 P3 D - D 2.4 x 1017 n per shot D 1 2.4 x 1015 n per shot (Constant source) R-Z 3416 (61 x 56) 1% Weighted difference

DOT3.5 FUSION-40 (n, 42; 7, 21) Symmetric S160 P5 D D 2.4 x 1017 n per shot D - T 2.4 x 1015n per shot (Constant source) R Z 12 084 (106 x 114) 1% Weighted difference

(1)

Concrete

3O 25

oE 8 20 "~: <

15.

/

Plasma Axis

[~~ lljar

10-

Poloidal Coil Shell Support

L~Vacuum

Vessel

Plasma 0 -I------r r ~r ~ R 0 5 10 15 20 Radial distance from plasma axis (m)

Fig. 4. Two-dimensional calculational model of the LHD experimental room.

3.2. Duct streaming analysis The duct streaming effect resulting from many holes on the ceiling should be calculated to estimate the dose rate in the cellar. We chose the following four types of duct for typical cases:

vertical p o r t - - p o s i t i o n r = 3.9 m; area 0.5 m in diameter; 10 ducts; (2) coil feeder d u c t - - p o s i t i o n r = 5.6 m; area 0.95 m in diameter; nine ducts; (3) outer belljar d u c t - - p o s i t i o n r = 7 . 5 m ; area, lmx2m, lmxlm;nineducts; (4) N B I d u c t - - p o s i t i o n r = 17.3 m; area, 2 m x 3 m; eight ducts. Here, r is t h e distance between the plasma axis and the duct. The dose distribution from each duct was estimated using the DOT3.5 code, These ducts are modelled to be annular in shape, with a center axis that coincides with the plasma axis. First, calculations for a single annular duct were modelled and performed. These results of the dose distribution for four types of duct were reduced according to the ratio of their real duct area to the annular area. Then, the whole dose at a space point was obtained by summing up the contributions from all ducts, with the reduced results for the four types of duct. In this manner, the total neutron and gamma-ray dose distribution in the cellar was obtained, as shown in Fig. 5. However, the unreduced results were directly applied to the shielding design near the ducts, because this correction method may lead to underestimation of the peak value of the streamed dose. The dose rate in the cellar is nearly constant, i.e. about 100 mSv per shot. This result is almost the same as the previous estimation [2]. This is because the streamed neutrons from the ducts are dominant in the cellar, and the neutron penetration through the concrete is not significant in this case.

H. Handa et al. / Fusion Engineering and Design 28 (1995) 515-524

13 12 11 10 __j-Vert,cal Port I OuterBeIIj~,rDuct / 9 OoitFeederDuct ~ , 1 0 ; 8I

t

521

"~

NBIDuct

E q~

7i 6~

Plasma Axis

5 <

Cellar

4

/

3i 2 1 0

0 1 2 3 4 5 6 7 8 £ 10 11 1"_ 13 ',4 15 16 17 18 19 2C 21 Radial distancefrom plasma axis (m) Fig. 5. Total neutron and gamma-ray dose distributions in the cellar, resulting from the duct streaming effects.

3.3. Sky-shine

To estimate the dose rate at the site boundary, the direct radiation dose from the building wall and the sky-shine radiation dose from the building roof should be considered. The GRTUNCL code was applied to calculate the first collision source, to avoid the ray effect. The equivalent point sources, by angular direction and energy group, for GRTUNCLcalculation of the sky-shine evaluation were obtained by editing the angular fluxes at the upper surface of the roof from the two-dimensional transport calculation around the LHD building. The point sources were evaluated by multiplying the angular fluxes by the mesh area and the angle weight in each mesh point, and summing these to obtain the total leakage from the upper surface of the roof. The correction for the difference between the roof area of the real LHD building and that in the two-dimensional calculational model was considered in this evaluation. The point sources were located at the upper roof surface of the building center. However, the point sources for the direct contribution calculation were estimated from the angular fluxes at the outer surface of the concrete wall, in the same manner as for the sky-shine calculation. The point source was located 20 m above the ground level of the building center. The transport calculation in air was performed using the DOT3.5 code, using the first collision source. Fig. 6 shows the direct and sky-shine dose distributions vs. the radial distance from the side wall in the case of 100

shots per year. The direct gamma-ray dose and the sky-shine neutron dose are the main contributions at a short distance from the building wall (up to about 200 m). The direct gamma-ray dose and the sky-shine neutron and gamma-ray doses give almost the same level up to about 700 m; the gamma-ray doses of both sky-shine and direct radiation will be the main contributions to the environmental dose beyond that distance. In the case that 100 full shots are given each year, the dose is 0.024mSv per year at about 125 m from the outer surface of the building wall (supposedly the nearest distance to the site boundary), which is under the 0.05 mSv per year design target for the site boundary and is one-twentieth of the limit provided in Japanese law. Furthermore, the two-dimensional transport calculation around the LHD experimental room provides conservative estimation from the viewpoint of the modelling. The opening of the vertical and horizontal ports, which are installed in palces in the real design, are modelled as being opened in all azimuthal directions as shown in Fig. 4. The other structures (except the LHD machine itself) are neglected in the calculational model of the LHD experimental room. Therefore, it is expected that the direct and sky-shine doses can be decreased to about one-quarter of their values, by considering the more detailed streaming effect of the vertical and horizontal ports, and the shielding effect of the girder of considerable thickness actually installed in the lower surface of the roof.

522

H. Handa et al. / Fusion Engineering and Design 28 (1995) 515-524 100 10-1 ~ ( : " ~

I

10-2 ~

g

10-3 10_4 10-5 10-6

",

TOTAL J . . . . . . . . Direct(photon) ........... Skyshine(Neutr0n) ' Skyshlne(photon) I Direct(Neutron , _ _ . ~.~...j

" ~'-~z.<~ . ~ , ~ .

x

.

~¢'~a~ ~_.~.~ ~ ~'""'-,'Zf: #.'r~..~ ~

""""

"'-,

10-7

"'-..--

10_8

~'-'-~':~ ""~--~.

- ....... ~......

10-1010-9

"'-......~.

10-11 0

200

400

600

800

1000

1200

1400

1600

1800

2000

DistancefromSideWall(m) Fig. 6. Environmental dose distributions around the LHD building. 3.4. Activation o f the L H D

It is important to estimate the activation of the device from the radiation exposure point of view in maintenance and repair. The activation levels of a vacuum vessel (SS316) and superconducting magnet (SCM) were estimated using the THIDA-2 [9] code system. The neutron flux at the vacuum vessel, obtained from two-dimensional calculation (see Section 3.1), was applied to the activation calculation. The compositions of the SS316 and SCM are given in Table 4.

Some 100 shots per year for 6 years of operation were assumed as the irradiation pattern for the activation calculation. Figs. 7 and 8 show the radioactivity of 1 g of SS316 and SCM (including the can). The main radiation sources 1 day after shutdown are Co-58, Fe-55, Mn-54 and Co-60 for SS316, and Cu-64, Co-58, Co-57, Sn-125 and Sb-125 for the SCM. Fig. 9 shows the dose rates at the surfaces of the vacuum vessel and SCM. The dose rates at the surfaces of the vacuum vessel and SCM are 0.18 mSv h -1 and 0.074 mSv h -1 1 day after shutdown respectively.

Table 4 Composition of SS316 and SCM for activation calculation 4. Conclusions No.

Element

Weight fraction (wt.%) SS316 (p = 7.98 g cm-3)

1 2 3 4 5 6 7 ,8 9 10 11 12 13 14 15 16

Iron 65.15 nickel 12.0 Chromium 17.0 Molybdenum 2.5 Manganese 2.0 Cobalt 0.2 Carbon 0.08 Silicon 1.0 . Phosphorus 0.045 Sulphur 0.3 Copper Titanium Niobium Aluminum Lead Tin

SCM (p = 4.52 gcm -3) 23.54 5.94 6.02 0.89 0.71 3.51 × 10-3 0,028 0.354 0.016 0,011 26,01 3.96 3.44 23.3 2.28 3.42

Detailed and updated shielding analysis of the LHD facility was performed, using the new library data and reflecting the design of the device. The following results were obtained through shielding analysis. (1) The 6Icx-40 library used in the previous work underestimates the neutron flux penetrating the concrete wall and should be changed to the other new library data. (2) The validity of FUSION-40, which was newly applied to the shielding calculation of the LHD, was confirmed The following results also were obtained through shielding analysis of the LHD. (3) The dose distribution in the experimental room was estimated by two-dimensional calculation. (4) The duct streaming effect was considered in estimating the dose distribution of the cellar. (5) The sky-shine and direct doses at the site boundary were estimated.

H. Handa et al. / Fusion Engineering and Design 28 (1995) 515-524 IM

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524

H. Handa et al. / Fusion Engineering and Design 28 (1995) 515-524

(6) The dose rate resulting from the activation of the device was estimated. We are planning three-dimensional estimation using the TORT [ 10] code to obtain more details on the radiation environment for equipment design as a next step.

References [1] O. Motojima, T. Akaishi, M. Asao, K. Fujii, J. Fujita et al. and LHD design group, Engineering design study of superconducting large helical device, Proc. 13th Int. Conf. on Plasma Physics and Controlled Nuclear Fusion Research, Washington, DC, 1990, IAEA, Vienna 1991. [2] H. Handa, K. Hayashi, Y. Ogawa, Y. Sakuma, H. Obayashi et al., Radiation protection concepts of large helical device (LHD), Fusion Eng. Des. 17 (1991) 335-342. [3] K. Shibata et al. and JENDL compilation group, Japanese Evaluated Nuclear Data Library, Version-3, JENDL-3, Rep. JAERI-1319, Japan Atomic Energy Research Institute, June 1990. [4] K. Oishi, Y. Ikeda, H. Maekawa and T. Nakamura, Experiment and analysis of neutron spectra in a concrete assembly bombarded by 14-MeV neutrons, Nucl. Sci. Eng 103 (1989) 46 58.

[5] K. Maki, K. Kosako, Y. Seki and H. Kawasaki, Nuclear Group Constant Set FUSION-J3 for Fusion Reactor Nuclear Calculations based on JENDL-3, Rep. JAERI-M 91-072, Japan Atomic Energy Research Institute, May 1991. [6] Y. Seki and H. Iida, Coupled 42-group neutron and 21-group gamma ray cross section sets for fusion reactor calculations, Rep. JAERI-M 8818, Japan Atomic Energy Research Institute, 1980. [7] J.F. Briesmeister (ed.), M C N P - - A General Monte Carlo Code for Neutron and Photon Transport, Version 3A, LA-7396-M, Rev. 2 (September 1986); also available from ORNL/RSIC as CCC-200/MCNP4. [8] W.A. Rhoades and F.R. Mynatt, The DOT-III two dimensional discrete ordinates transport code, Tech. Memo. ORNL/TM-4280, Oak Ridge National Laboratory, Oak Ridge, TN, 1973. [9] Y. Seki, H. Iida, H. Kawasaki and K. Yamada, THIDA2: an advanced code system for calculation of transmutation, activation, decay heat and dose rate, Rep. JAERI 1301, Japan Atomic Energy Research Institute, 1986. [10] W.A. Rhoades and R.L. Childs, The TORT three-dimensional discrete ordinates neutron/photon transport code, Rep. ORNL-6268, Oak Ridge National Laboratory, Oak Ridge, TN, November 1987; also available for ORNL/ RSIC as CCC-543/TORT-DORT.