Experiment and analysis of induced radioactivity in large SS-316 stainless steel shielding material bombarded with 14 MeV neutrons

ELSEVIER

Fusion Engineering and Design 28 (1995) 449-456

Fusion Engineering and Design

Experiment and analysis of induced radioactivity in large SS-316 stainless steel shielding material bombarded with 14 MeV neutrons Y. Ikeda, C. Konno, F. Maekawa, Y. Uno, Y. Oyama, K. Kosako, J. Maekawa Department of Reactor Engineering, Japan Atomic Energy Research Institute, Tokai, Ibaraki 319-11, Japan

Abstract

Radioactivities induced in SS-316 samples in a thick SS-316 shield material bombarded with 14 MeV neutrons were measured after cooling times ranging from several hours to 8 months. Experiments focusing in particular on deep locations where the (n, 7) reactions are dominant were analyzed by activation inventory codes ACT4 and REAC*3, and the activation cross-section libraries the JENDL activation file, the data library associated with the REAC*3 code system and FENDL/A1.1. The ratio of calculated to experimental results gave a practical uncertainty range for the calculations with the presently available code systems. It was shown experimentally that not only activation cross-sections but also neutron spectrum calculations are sources of large uncertainties in the calculation of radioactivities.

1. Introduction

The importance of induced radioactivity in the D - T fusion environment is recognized as the most critical issue from the safety point of view [1-3]. Extensive efforts have been carried out in terms of development of codes for calculating inventories and associated development of data libraries [4-9]. To make development of the code system effective, validation by experimental data is essential to assure objectively the reliability of the calculations. The importance of integral experiments was demonstrated by the extensive experimental program on induced radioactivity within the framework of J A E R I - U S D O E collaboration [ 10-12]. Experimental analysis provided invaluable information on the present status of radioactivity cross-sections. The raElsevier Science S.A. SSDI 0920-3796(94)00326-2

dioactivities of concern, however, were produced by mostly threshold reactions with D - T neutrons, because the neutron spectra tested were rather hard. Still there is a need for experimental verification of calculations of radioactivities induced by low energy neutrons. This paper describes a new integral experiment on the radioactivity induced in a large SS-316 shield assembly [13]. The primary objective of this experiment was to provide integral experimental data of induced radioactivity to validate the data and methods associated with SS-316 which is the primary candidate for blanket shield structural materials in the on-going ITER project. The parameters investigated were cooling time, neutron spectrum, and activation cross-section libraries, which should be factored into the experimental analysis. Irradiation positions were selected deep inside the

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Y. Ikeda et al. / Fusion Engineering and Design 28 (1995) 449-456

i

813mm(32")

=[_-

1118mm(44")

d +-

2(

:-~--406.4 mm - - ~

i

• Experimental hole

609.6 mm

Fig. 1. Cross-sectional view of the SS-316 assembly and sample position.

assembly, resulting in softer neutron spectra with less important D - T primary neutron flux. As neutron transport codes, DOT3.5 [14] and MCNP [15] were employed coupled with JSSTDL [16] and FSXLIB-J3 [17] cross-section libraries based on the JENDL-3 nuclear data file [18]. For calculations of the activation, we used ACT4 of the T H I D A code system [4] with the JENDL activation library [7] and the F E N D L activation library [9], and the REAC*3 code [8] with the associated cross-section library REAC175.

2. Description of the experimental system

Fig. 1 shows the SS-316 assembly used for fusion shielding experiments [13]. Dimensions of the assembly are 1200 mm diameter and 1118 mm thickness for the test region, to which a source reflector cavity 203 mm thick is attached. The experimental holes for sample insertion were filled with the same SS-316 material. The D - T neutron source was located in the source reflector cavity at 300 mm from the front surface of the test region. The whole assembly was suspended by a boxtype structure to align the symmetry axis to the incident d ÷ bean direction. The height from ground level to the axis was 1800 mm.

Table 1 Chemical compositions of the SS-316 sample Element

Concentration (wt.%)

Fe Ni Cr Mn Mo Si C P S W

68.49 10.09 17.03 1.35 2.08 0.28 0.05 0.32 0.26 0.05

3. Measurement of irradiation and activity

The D - T neutrons were generated via the 3T(d, n)4He reaction using the FNS facility [19]. The incident dueteron beam current and energy were about 2 mA and 350 keV respectively. Nominal neutron yield at the source was about 2 x 1011 s -1. The samples of SS-316 placed at 406.4 and 609.6 mm from the surface of the test region along the central axis of the assembly

Y. Ikeda et al. /Fusion Engineering and Design 28 (1995) 449-456

Table 2 Cooling time and collection time Run

Cooling time (s)

Collection time (s)

Position A

1 2 3 4

6.0654 × 2.2992 x 4.7702 x 2.1070 x

l0 s 105 107

1381 21927 38654 86478

5.8258 x 104 2.5255 x 105 2.1227 x 107

2239 51483 164130

104

Position B

1 2 3

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Fig. 2. Measured radioactivities at different cooling times and decay curves calculated using ACT4/JENDL at (a) position A and (b) position B.

were irradiated with D - T neutrons for 10 h. Hereafter, the positions at 406.4 mm and 609.6 mm are denoted positions A and B respectively. The size of the SS-316

451

sample was 9 x 10 x 40 mm 3. The weight of the sample was around 90 g. The chemical composition of the sample is given in Table 1. After irradiation, induced radioactivities were measured by 7-ray spectroscopy with Ge detectors after cooling times ranging from several hours to about 8 months. The effective self-absorption coefficient for the particular sample-detector configuration was derived experimentally using SS-316 absorbers with different thicknesses. Even though such a massive sample was used, the radioactivities produced were considerably weakened owing to large attenuation of neutrons in the shield materials in front of the samples. The natural background y-rays were carefully subtracted from the measured spectra. Experimental data were derived as the radioactivity intensity per unit weight of SS-316 (Bq cm 3). The experimental values were normalized for the source neutron intensity of 1.0 x 1011 s -1. The radioactivities were identified by y-ray energies and their intensity relationship. The decay rate was derived from y-ray counts, detector efficiency, 7-ray emission probability and other corrections needed, e.g. y-ray collection time, half-lives, etc. The decay property data were taken from Table of Radioactive Isotopes [20]. When multiple yrays were associated with a particular radionuclide, the most intense y-ray line was used for data processing. In Table 2, the cooling times and collection times for y-ray measurements are summarized.

4. Experimental results

The observed y-ray lines were mostly associated with the decay of radioactivities produced via (n, y) reactions. This is simply because the high energy neutron flux was attenuated by at least 400 mm thick SS-316 in front of the sample location, resulting in relatively small reaction rates of the threshold type reactions. However, the fractional contribution of the low energy neutron flux increases with depth in SS-316. The major radioactivities were SlCr (T~/2 = 27.7 d), S4Mn (312.5 d), 56Mn (2.579h), 59Fe (44.6d), 57Ni (36.0h), 57Co (271 d), 58Co (70.8d), 6°Co (5.27y), 99Mo (66.0h), 99mTc (6.02 h), and ~87W (23.9 h). Except for 54Mn, 57Ni, 57Co and 58Co, all the radioactivities were produced via (n, y) reactions. In addition to those radionuclides, low level activities of 92rnNb (10.15d), 9SNb (35.0d), 9SZr (64.0 d) and 76As (26.3 h) were identified. These minor activities along with 99mTC which is the decay product of 99M0 f l - decay were, however, discarded from the experimental analysis.

452

Y. Ikeda et al. / Fusion Engineering and Design 28 (1995) 449-456

Table 3 Ratios of (n, 7) reaction contributions to (n, threshold) reaction contributions

10 ~ 10 -5 10 -6 "~

I 0 "7

~-

10 -8

Depth in SS-316 (mm) 200 400 600

I 0 "s

I 0 -~° 10 -11 1 0 -7

1 0 -s

10-3

1 0 -1

Ratio (n, 7) to (n, threshold) 5~Cr

56Mn

99Mo

0.076 0.718 6.260

0.213 2.063 17.65

0.390 3.222 24.84

101

Neutron Energy (MeV) Fig. 3. Neutron flux spectra at positions A and B, calculated with DOT3.5 and MCNP.

In Figs. 2(a) and 2(b), experimental data at positions A and B respectively are plotted with corresponding decay curves calculated using ACT4 with the JENDL activation library. At a cooling time shorter than 1 day after irradiation, 56Mn was the most dominant activity at both positions, followed by 99M0 and its daughter nuclides 99mTc and 187W. At a cooling time of several days, 99M0 and 99mTc maintained the largest contribution, followed by 5ICr. The S6Mn disappeared quickly because of its short half-life. After 8 months cooling time, only four activities with appreciably long halflives, S4Mn, 58Co, 59Fe and 6°Co, were observed. The experimental error comprises mainly 7-ray counting statistics, the detector efficiency error, and the self-absorption correction. Fortunately, there is no need to consider the neutron self-shielding effect in the SS316 samples because the whole system could be treated as homogenized SS-316 material. The self-shielding effect was treated in the experimental analysis. The experimental error ranged from _+5% to 20% in most cases. Some cases with poor counting statistics gave more than _+50% uncertainties.

5. Experimental analysis Characterization of the impact of the neutron flux spectrum on prediction of the induced radioactivity is essential for ensuring overall accuracy of the prediction. In the present analysis, two neutron transport calculation codes, DOT3.5 and MCNP, were employed in order to assess the adequacy of each approach as a calculation system. Both the cross-section libraries for those transport calculations were based on JENDL-3.

For the DOT3.5 calculation, a library with self-shielding effect, namely JSSTDL, was used. As noted, the self-shielding effect in the sample materials was included in the flux spectrum calculations. The experimental assembly was modeled as an R - Z two-dimensional system. For the DOT3.5 calculation, Ps, S~6 approximations were adopted. The neutron spectrum of the D - T neutron target, calculated by the MCNP, was used as the source neutron spectrum. The calculated neutron spectra at two sample locations are given in Fig. 3. On the whole, the calculations gave similar spectra. However, for the thermal neutron energy group, there are considerable discrepancies. ACT4 and REAC*3 were used as the inventory codes. The activation cross-section library based on the JENDL activation file with 125 neutron energy groups and the 175 energy group library installed in the REAC*3 code system were used for the induced radioactivity calculations. In addition to these libraries, the FENDL activation file version 1.1 was tested for comparison, utilizing ACT4 as the inventory code. The multi-group FENDL library with 125 energy group structure was created from the pointwise FENDL file for application to the ACT4 calculation and to be compared directly with calculations of the JENDL activation cross-section. Induced radioactivities corresponding to the experimental conditions were calculated by these codes to be compared with the measured values. The D - T neutron source intensity was assumed to be 1.0 x 10~ s ~.

6. Discussion In this section, the calculations are compared with experimental values. Based on the C/E (calculation/experiment) values, the adequacy of the calculation, impact of the neutron flux calculation, uncertainty of the activation cross-section and other associated topics are

453

Y. Ikeda et al. / Fusion Engineering and Design 28 (1995) 449-456 2.0

2.0

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(d)

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60

80

Fig. 4. (a) C/E for 51Cr production. (b) C/E for 54Mn, S6Mn and 59Fe productions. (c) C/E for 57Ni, 57C0 and 58C0 productions. (d) C/E for 99M0 and 187W productions.

discussed for each radionuclide. It should be noted that there are several radioactivities which are produced by both threshold reactions and the (n, 7) reaction. In Table 3, the ratio of reaction rates of the threshold type reaction to those of (n, 7) reactions of concern are given for the isotopes of importance. The reaction rate calculation was based on a neutron flux calculated using DOT3.5 and the JENDL activation data library. slCr. The activity is produced by the reactions 52Cr(n, 2n) and S°Cr(n, 7). Although 52Cr is 20 times more abundant than 5°Cr, the (n, 7) reaction contributes about 42% of the total 51Cr production at position A as shown in Table 3. The 51Cr production at a depth of 610 mm was dominated by the 5°Cr(n, 7) reaction with a contribution of more than 85% to the total. The C/E values for all cases are plotted in Fig. 4(a). For the DOT3.5 flux, calculations with both JENDL and REAC175 gave reasonable C/E values. For the MCNP flux, however, systematically higher C/E values were found, in particular for position B. The

values calculated using F E N D L are systematically lower than those calculated using JENDL by 10%. Nevertheless, all C/E values range from 0.8 to 1.3. It can be concluded that all calculations are adequate if an uncertainty of a factor of two is allowed in the estimation. S4Mn. S4Fe(n, p) and SSMn(n, 2n) are two major contributing reactions for 54Mn production in SS-316. The two contributions are comparable according to both cross-section values and abundances of 54Fe and 55Mn. The C/E values shown in Fig. 4(b) are in the range 0.83-1.28 for all cases. The ACT4 calculation using the MCNP flux gives better C/E than that using the DOT3.5 flux when JENDL is used. The REAC*3/ REAC175 calculation gives a systematically lower value by 20% than ACT4/JENDL for both neutron fluxes. The A C T 4 / F E N D L calculation was almost identical with the ACT4/JENDL calculation. From the range of C/E and rather large experimental error of more than _+10%, the present results confirm that all calculations satisfy the criteria of a factor of two uncertainty.

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Y. Ikeda et al. / Fusion Engineering and Design 28 (1995) 449-456

S6Mn. As shown in Table 3, two reactions SSMn(n, 7) and S6Fe(n, p) contribute to S6Mn production in SS316. Although S6Fe was more than 50 times more abundant than 55Mn in SS-316, the 5SMn(n, 7) contributes 70% of the total even at position A. The contribution at position B increases to more than 94%. The C/Es are plotted in Fig. 4(b) along with results for 54Mn. It is obvious that all calculations overestimate the experimental values by more than a factor of two. As it is commonly accepted that the 55Mn(n, 7)S6Mn cross-sectoin is known with reasonable accuracy, these large C/E values indicate a serious problem not only in the neutron flux calculation but also in the process of multi-group cross-section formation. In particular, the cross-section at resonacne and thermal energy regions should be carefully examined. Uncertainty of the 5SMn abundance in the SS-316 sample is another possible reason for the uncertainty in 56Mn production. Therefore, experiments focusing on this particular reaction product are required.

-ff e~ O

°4

r~

L~

] 0 "~

1 0 -1

101

10 3

10 s

10 z

Neutron Energy (eV)

(a)

o

,4

SgFe. S9Fe is produced mainly by the reaction SSFe(n, 7). Although there is a minor contributing reaction 62Ni(n, ~) in SS-316, it is expected to be very small compared with the SSFe(n, 7) contribution. If Co impurity exists, the 59C0(n, p) is considered to be another possible contributor, though the contribution should be small because of the relatively low cross-section for the high energy neutron flux fraction. Thus, it is reasonable that 5SFe(n, 7) dominates ~gFe production. The C/Es are also plotted in Fig. 4(b). The ACT4/ JENDL calculations with DOT3.5 flux gave good agreements with experiments. The ACT4/JENDL with MCNP gave 10%-30% overestimations. The REAC*3/ REAC175 with DOT3.5 and MCNP fluxes showed systematic overestimations of 50%-70% and a factor of 2-2.3 respectively. ACT4/FENDL gave the same trend as the REAC*3/REAC175. It was clearly shown that the different fluxes resulted in more than 30% discrepancy even though the same cross-section was used. 57Ni. The 58Ni(n, 2n)57Ni reaction is the unique source

of 57Ni in SS-316. The C/E values are shown in Fig. 4(c). All calculations of ACT4/JENDL, REAC*3/ REAC175 and ACT4/FENDL with DOT3.5 flux gave reasonable C/E values, whereas calculations with MCNP flux show systematic underestimations of 200/025%. These results suggest that not only a low energy neutron flux, but also a high energy neutron flux is the source of uncertainty in the overall radioactivity calculation. Nevertheless, the fact that all calculations re-

10 "~

(b)

10-'

10 ~

10 '~

10 s

10"

Neutron Energy (eV)

Fig. 5. (a) Cross-section for 98Mo(n, 7)99M0. (b) Cross-section for 186W(n,7) 187W-

sulted in a C/E in the range 0.72-1.04 demonstrates the adequacy of prediction of 57Ni production.

57C0. 5 7 C 0 is produced both via the 58Ni(n, np)S7Co reaction and by the decay of 57Ni. The cross-section of 58Ni(n, np)57Co in the 14 MeV region is higher by a factor of 10 than 58Ni(n, 2n)SYNi, so it is assumed that the 58Ni(n, np)SYCo reaction is the major contributor. The C/Es are also shown in Fig. 4(c). The best C/E is observed for the REAC*3/REAC175 with DOT3.5 calculation. As shown for 57Ni, calculations with the MCNP flux tended to give systematically lower values by 10%-15% than those with the DOT3.5 flux. The ACT4/FENDL with DOT3.5 gave the lowest C/E value of 0.65. It is suggested that the F E N D L cross-section should be re-examined. 58Co is identified as the product of 58Ni(n, p)SSm+gCo reaction. Fig. 4(c) also shows the 5SCo.

Y. Ikeda et al. / Fusion Engineering and Design 28 (1995) 449-456

C/E for SSCo production. The discrepancies for calculations with different fluxes were about 15%, the trends being the same as for 57Ni and S7Co. The calculation with MCNP tended to give lower C/E values than that with DOT3.5. The best C/E was given by ACT4/ JENDL with MCNP for position A. the REAC*3/ REAC175 calculations with DOT3.5 and MCNP fluxes also overestimated the measurements by 20% and 10% respectively. For position B, all calculations underestimated the results by up to 35%. However, as the error in experimental data at position B was large owing to poor 7-ray counting statistics, further experimental measurement is needed for testing the data for this reaction product at the deep location of the shield material. The overall prediction accuracy for all calculations seems adequate. 6°Co. 6°Ni(n, p)6°Co is the principal reaction for 6°Co production in SS-316. The reaction is threshold type and cross-sections at the 14 MeV neutron energy region are around 140 rob. From the simple calculation, considering the abundance of 6°Ni and D - T neutron flux at the sample positions, the 6°Ni(n, p)6°Co reaction could not be the major sourace of 6°Co. The large underestimations of all calculations by three to four orders of magnitude, implied that there was a large amount of cobalt impurity in the SS-316 sample. In order to estimate the fraction of cobalt in the sample, a calculation with 0.2% w/o Co was carried out. The result showed good agreement with the measured value within _+10%. This strongly suggested that 0.2%-0.3% cobalt always exists as an impurity in SS-316. After 8 months cooling, the 6°Co became the dominant activity in SS-316, contributing 90% of the total. The present result clearly demonstrates that control of cobalt impurity in SS-316 is a critical problem for reduction of the radioactivity. 99Mo. 99Mo is produced via both the 98Mo(n, 7) and

~°°Mo(n, 2n) reactions; the ratios of the contributions are shown in Table 3. Although the abundance of Mo in SS-316 is about 2%, 99Mo is one of the largest radioactivities in SS-316 for rather short cooling times of up to several days after irradiation. This is due to the l a r g e cross-sections for both contributing reactions. From Table 3, it is understood that the 9SMo(n, 7) reaction dominates 99Mo production at both positions A and B. All C/E values are plotted in Fig. 4(d). The ACT4/ JENDL with DOT3.5 and MCNP overestimated experiments by 30%-40% and 75% 85% respectively. The REAC*3/REAC175 with DOT3.5 and MCNP overesti-

455

mated the results by 67% 90% and a factor of 2.1-2.3 respectively. It is clearly shown that the difference in neutron spectrum caused a 40% difference in the radioactivity calculations. It should also be pointed out that the different cross-section libraries gave large discrepancies in the calculations. Fig. 5(a) compares the 98Mo(n,j reaction cross-section in JENDL and REAC175. The data are quite different. As a whole, REAC175 shows a higher cross-section at the resonance region of 500 eV than JENDL. In particular, the cross-section at 10 eV resonance in REAC175 is two orders of magnitude larger than that in JENDL. The difference in calculation is strongly reflected by this differeence is cross-section. These cross-section data should be examined further. The ACT4/JENDL with DOT3.5 showed the best C/E values of around 1.2-1.3. 187W. Although tungsten is not a primary element, it is commonly accepted that there is a very small amount of tungsten impurity in SS-316, of the order of several hundred ppm. However, it makes a proportionately large contribution to the induced radioactivity in SS316 in the soft neutron spectrum field deep in the shield material. The C/Es are shown in Fig. 4(d) along with results for 99Mo. There is a systematic large underestimation in ACT4/JENDL with both DOT3.5 and MCNP by a factor of 4-5. In contrast, REAC*3/ REAC175 overestimated experiments by 40%-70%. The ACT4/JENDL with DOT3.5 gave the worst C/E of 1.8-2.3. The results indicate that the flux spectrum dependence was dominated by the cross-section differences, Fig. 5(b) demonstrates the difference in cross-sections of the libraries. In the giant resonance region at 20 eV, the cross-sections differ by more than one order of magnitude. The extreme underestimation with JENDL can be explained by this fact. Therefore, it is recommended that the cross-section in JENDL should be re-evaluated.

7. Conclusions

Induced radioactivities in SS-316 at soft neutron spectral fields were measured for cooling times ranging from several hours to 8 months after irradiation. The experiments were analyzed by ACT4 with JENDL and FENDL, and REAC*3 with REAC175. The analysis gave the following results. (1) The uncertainty due to the neutron spectrum affected the accuracy of prediction for the (n, 7) reaction products. From the present analysis with DOT3.5

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Y. Ikeda et al. / Fusion Engineering and Design 28 (1995) 449-456

and M C N P , the difference in calculated results ranged from 10% to 30%. (2) The uncertainty due to ambiguity in the number densities of minor constituents should be seriously taken into account in the overall prediction uncertainty so far as SS-316 is concerned. Even for elements at ppm level, the induced radioactivity due to (n, 7) reactions dominates the activities. (3) There were large discrepancies in the cross-sections of 9SMo(n, 7)99Mo and 186W(n, ~)lSTW between J E N D L and REAC175. The present experiments and analysis demonstrate the importance of integral experiments using a realistic neutron environment expected for the SS-316 I T E R shield blanket to determine the adequacy of predictions of induced radioactivity.

Acknowledgments The authors wish to acknowledge J. Kusano, C. Kutukake, S. T a n a k a and Y. Abe for operation of the F N S accelerator.

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