Copper benchmark experiment at the Frascati Neutron Generator for nuclear data validation

Fusion Engineering and Design 109–111 (2016) 843–847

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Copper benchmark experiment at the Frascati Neutron Generator for nuclear data validation M. Angelone ∗ , D. Flammini, S. Loreti, F. Moro, M. Pillon, R. Villari ENEA Dipartimento Fusione e Tecnologie per la Sicurezza Nucleare, C.R. Frascati, via E. Fermi 45, 00044 Frascati, Italy

h i g h l i g h t s • • • • •

A benchmark experiment was performed using pure copper with 14 MeV neutrons. The experiment was performed at the Frascati Neutron Generator (FNG). Activation foils, thermoluminescent dosimeters and scintillators were used to measure reactions rates (RR), nuclear heating and neutron spectra. The paper presents the RR measurements and the post analysis using MCNP5 and JEFF-3.1.1, JEFF-3.2 and FENDL-3.1 libraries. C/Es are presented showing the need for deep revision of Cu cross sections.

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Article history: Received 27 July 2015 Received in revised form 16 December 2015 Accepted 26 January 2016 Available online 6 February 2016 Keywords: Benchmark experiment 14 MeV neutrons Copper cross sections Frascati Neutron Generator Activation technique MCNP Monte Carlo code

a b s t r a c t A neutronics benchmark experiment on a pure Copper block (dimensions 60 × 70 × 60 cm3 ), aimed at testing and validating the recent nuclear data libraries for fusion applications, was performed at the 14MeV Frascati Neutron Generator (FNG) as part of a F4E specific grant (F4E FPA-395 01) assigned to the European Consortium on Nuclear Data and Experimental Techniques. The relevant neutronics quantities (e.g., reaction rates, neutron flux spectra, doses, etc.) were measured using different experimental techniques and the results were compared to the calculated quantities using fusion relevant nuclear data libraries. This paper focuses on the analyses carried-out by ENEA through the activation foils techniques. 197 Au(n,␥)198 Au, 186 W(n,␥)187 W, 115 In(n,n )115 In, 58 Ni(n,p)58 Co, 27 Al(n,␣)24 Na, 93 Nb(n,2n)92 Nbm activation reactions were used. The foils were placed at eight different positions along the Cu block and irradiated with 14 MeV neutrons. Activation measurements were performed by means of High Purity Germanium (HPGe) detector. Detailed simulation of the experiment was carried-out using MCNP5 Monte Carlo code and the European JEFF-3.1.1 and 3.2 nuclear cross-sections data files for neutron transport and IRDFF v1.05 library for the reaction rates in activation foils. The calculated reaction rates (C) were compared to the experimental quantities (E) and the C/E ratio with relative uncertainties was assessed. © 2016 Elsevier B.V. All rights reserved.

1. Introduction Copper is largely used in tokamaks in heat sink components (e.g., divertor, FW, etc.), magnets, diagnostics, microwave waveguides and mirrors. However, few experiments were performed so far with Cu in the neutron energy range relevant to fusion: measurement of leakage spectra [1] and, integral experiment [2,3], the latter performed at the FNS facility and aiming at validating the JENDL-3.1 [4] nuclear data libraries for copper. In the FNS work, the authors concluded that the calculations reproduced the mea-

∗ Corresponding author. E-mail address: [email protected] (M. Angelone). http://dx.doi.org/10.1016/j.fusengdes.2016.01.065 0920-3796/© 2016 Elsevier B.V. All rights reserved.

sured data for neutrons above 10 MeV and reaction rates of high threshold reactions to within 10%. Instead, for the neutron below 10 MeV, in particular for 115 In(n,n ) and 197 Au(n,␥) reaction rates, the discrepancy between measurements and calculations was very large (C/Es ∼0.2÷2.5). For these reasons it is important to verify the status of the nuclear data library for copper, not only for the latest release, such as JEFF 3.2 [5], but also for earlier data set such as JEFF 3.1.1 [6] and FENDL 3 [7]. The new experiment has been conducted at the 14 MeV Frascati Neutron Generator (FNG) as part of a F4E specific grant (F4E FPA395 01) assigned to the European Consortium on Nuclear Data and Experimental Techniques composed by ENEA, KIT, CCFE, JSI, AGH and NPI and led by ENEA. The copper benchmark experiment comprised three steps: (a) the pre-analysis, devoted to study the best

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M. Angelone et al. / Fusion Engineering and Design 109–111 (2016) 843–847

Table 1 Used activation reactions and main physical parameters. Reaction

Effective threshold (MeV)

T1/2

E␥ (keV)

P␥ (%)

197

12.3 10.8 8.5 2.9 0.1 Thermal Thermal

6.183 d 10.2 d 14.96 h 70.86 d 4.49 h 23.9 h 2.695 d

355.7 934.5 1368.6 810.8 336.2 685.7 411.8

86.9 99.0 100.0 99.45 45.9 26.4 95.54

Au(n,2n) 93 Nb(n,2n) 27 Al(n,␣) 58 Ni(n,p) 115 In(n,n ) 186 W(n,␥) 197 Au(n,␥)

experimental conditions and block size; (b) the experimental campaign, in which the block was first assembled and then irradiated by 14 MeV neutrons. The relevant neutronics quantities (e.g., reaction rates, neutron flux spectra, doses, etc.) were measured using different experimental techniques; (c) the post-analysis, in which the nuclear quantities calculated (C) using fusion relevant nuclear data libraries, were compared to the experimental results (E) in order to validate the used copper nuclear data libraries. The experiment was complemented by sensitivity/uncertainty analysis performed both using deterministic and Monte Carlo SUSD3D and MCSEN codes, respectively [8,9]. The results of these analyses are presented in companion paper presented in this symposium [10]. The pre-analyses of the experiment provided also the activation reactions (Table 1) to be used, which cover the whole neutron energy range relevant to fusion neutronics. Furthermore, the pre-analyses pointed out that for the proposed block dimension (60 × 70 × 60 cm3 ) the effect of the neutron room background due to the neutrons backscattered on the FNG bunker walls is negligible up to about 55 cm depth in the Cu block. Thus an additional polyethylene or Li2 O3 shield around the block was not needed [11]. MCNP-5 [12] and JEFF-3.1.1 library for transport and IRDF2002 [13] for activation were used. The DT neutron source subroutine developed by ENEA for FNG [14] was employed in MCNP5 (this applies also to post-analysis in Section 4). This paper reports on the activation foils measurement and related analysis performed by ENEA using MCNP5. The C/E values are presented and discussed. 2. Experimental activity 2.1. The Copper block The Copper block was composed by seven 60 × 60 cm2 and 10 cm thick oxygen free high conductivity copper (OFHC) plates. The total weight was about 2.2 t. OFHC was used because of its high purity in Cu (>99.95%). The plates were placed on an aluminum support 2 cm thick (100 × 80 cm2 surface) and packed by 4 aluminum pillars placed at the four corners of the block and connected along two of the side of the block by 4 bolted aluminum bars (Fig. 1). All the plates, except of the last one, have one or two vertical cylindrical

Fig. 1. Picture of the Cu block, the Al pillars are shown.

holes of 28 mm or 52 mm diameter inside which a OFHC rod containing the activation foils is inserted. The rods of 52 mm diameter were realized to allow the inclusion of a NE-213 scintillator for neutron and gamma-ray fluxes measurements. The foils lie along the central axis of the block. The activation foils were hosted inside copper holders able to host up to 2 mm thick foils of 18 mm diameter (Fig. 2). 2.2. Foils activation and experimental results The activation foils were of 18 mm diameter but gold foils were 15 mm diameter. Au foils were 50 ␮m thick while W foils were 25 ␮m thick. As far as threshold detectors are concerned, the foils thickness was 1 mm for the first six positions while in the last two positions the thickness was of 2 mm. The goal was to have experimental activities with error <±10% at 1 ␴ level in the deepest position. For the other positions the experimental error on the measured foils activity ranged from <±1% up to ±4% at 1 ␴ level. After irradiation, each foil was measured by mean of an HPGe detector 60% relative efficiency, absolutely calibrated in efficiency respect to secondary national standard gamma-ray sources. The quoted error on the efficiency curve is ±3.0% at 1 ␴ level. The measured foils activity was corrected for: the decay from the end of irradiation, the decay during the counting period and the decay during the irradiation time (temporal factor, calculated by using the actual time dependent neutron emission recorded by the FNG neutron monitors). The source-to-detector geometry (point-to-disc correction) and the gamma self-absorption in the foil (which depends upon foil’s material, thickness and gamma-ray energy) were also accounted for. For each measured reaction rate a total error ranging

Fig. 2. (Left) Cu holder for foils; (right) copper rods.

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1.3 1.2

Ratio

1.1 1.0 0.9 0.8 0.7 1

2

3

4

5

6

Irradiation # Fig. 3. Ratio between neutron yield measured by Nb foils and the associated alpha particle monitor.

from ±4.5% to ±10% (1 ␴) was quoted. This total error was calculated by using the quadratic propagation law. The overall source of uncertainty contributing to the total error are: the FNG absolute neutron yield calibration (±3.0%), the HPGe detector calibration (±3.0%), the foils activity (<±1.0 up to ±10.0%), relative efficiency calibration, <±2.0% (for foils measured at different positions of the HPGE detector respect to the absolute calibrated one). It ought to be stressed that Nb foils located at the center of the first Cu plate, in front of the FNG target were routinely used for cross checking the neutron yield measured by the absolute associated alpha particle monitor. Fig. 3 shows the comparison between FNG neutron yields measured by Nb foils and alpha particles. Fig. 4 summarizes the results for all the measured reaction rates. The reaction rates (RR) for 197 Au(n,␥) and 186 W(n,␥) reactions decrease of about one order of magnitude from the first (3.5 cm) to the last (57 cm) experimental position. Instead, the RR for high threshold reactions (27 Al(n,␣), 58 Ni(n,p), 93 Nb (n,2n), and 197 Au(n,2n)) decrease of about 5 orders of magnitude, but 115 In (n,n ), which has a lower threshold (Ethr = 0.1 MeV), decreases of about 4 order of magnitude.

Fig. 4. Experimental RR versus penetration depth in the Cu block.

3. MCNP calculation The simulation of the experiment was carried-out using MCNP5 Monte Carlo code and JEFF-3.1.1, JEFF 3.2 and FENDL-3 nuclear data for neutron transport and IRDFF v1.05 library for the reaction rates of the activation foils. A detailed 3D MCNP model of the Copper block and experimental assembly was prepared (Fig. 5). MCNP5 was run in parallel mode using MPI on CRESCO cluster. The spatial and energy distribution of the neutron source is generated using an external routine already tested and used in previous benchmark experiments [14]. The weight windows (WW) variance reduction technique has been adopted to obtain statistical uncertainties below ±1.5% for RR of the threshold reactions and below ±3% for RR of non-threshold reactions. The weight windows have been generated with the MCNP built-in WW generator. In order to obtain low statistical uncertainties in all the experimental positions the WW were optimized to reduce the variance in the last detector position for threshold detectors. Instead, three different sets of WW were used for thermal detectors (Au and W), each of which was generated to optimize the variance in a different range of penetration depth along the block.

Fig. 5. Geometry of the MCNP model used for the calculations. (Left) side view; (right) top view.

M. Angelone et al. / Fusion Engineering and Design 109–111 (2016) 843–847

0.01 1E-3

58

1.1

1E-4

0.9 0.8

1E-5 0

10

20

30

40

50

0.7

60

0

10

1.1

1.05

20

93

1.0

1.00

30

40

50

60

Penetration Depth (cm)

Penetration Depth (cm)

92

Nb(n,2n) Nb

m

0.9

0.95

C/E

JEFF3.1.1/JEF3.2

58

Ni(n,p) Co

1.0

C/E

-2

1.2

JEFF 3.1.1 JEFF 3.2

-1

-1

Total Flux (cm *s *n )

846

0.90

0.8 0.7 0.6

0.85 0

10

20

30

40

50

0

60

3.1. Results and discussion Neutron spectra have been calculated in all the eight experimental positions. JEFF-3.1.1 and FENDL-3 produce similar results as these files are very similar. Differences in neutron spectra have been noticed between JEFF-3.1.1 and JEFF-3.2. It clearly arises that for energy below 1 keV, the neutron flux spectrum calculated with JEFF 3.1.1 has a higher intensity than the one calculated using JEFF-3.2 and this behavior becomes more pronounced for larger penetration depths. Moreover, in the intermediate energy range (1 keV < E < 1 MeV) the JEFF 3.2 library produces, in turn, a more intense neutron spectrum. In the high energy region (E > 1 MeV), again the neutron fluxes calculated using JEFF 3.1.1 result higher than those obtained using JEFF-3.2. However, as the intermediate energy region contributes to the total flux for more than 95%, the total neutron flux at deep penetration results higher when calculated with JEFF-3.2, as shown in Fig. 6. Fig. 7 shows the C/Es calculated using the different nuclear data libraries. The C/E at zero penetration depth represents the Nb foil placed in front of the FNG target used for independent neutron yield measurement (see Fig. 3). For this foil the C/Es are within the quoted uncertainty. Instead, the remaining C/Es for Nb are underestimated by ≈15% for calculation performed with JEFF-3.1.1. The underestimation of the calculation is more severe with JEFF 3.2, and the discrepancies in this case increase with the penetration depth. For 27 Al(n,␣) the discrepancies are smaller with respect to the Nb case (≈−10% dif-

27

C/E

30

40

50

24

Al(n, ) Na 10

20

30

40

50

60

Penetration Depth (cm)

1.1 1.0 0.9 0.8 0.7 0.6 0.5 0.4

115

115

In(n,n')

0

10

20

30

40

In

50

m

60

Penetration Depth (cm)

1.1 1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3

186

0 1.1 1.0 0.9 0.8 0.7 0.6 0.5 0.4

20

Penetration Depth (cm)

1.1 1.0 0.9 0.8 0.7 0.6 0.5 0

C/E

All the MCNP models have also the already mentioned Nb foil, 13 mm diameter and 100 ␮m thick, located at the center of the first surface of the Copper block right in front of the FNG target. The calculated RR for this Nb foil is also compared to the same experimental data and used to get an independent absolute neutron yield, as already discussed in Section 2. The calculated (C) RR in activation foils and neutron spectra were compared to the experimental quantities (E) and the C/E ratio with relative uncertainties was assessed. The discrepancies between the calculated and the measured RR have been investigated also by studying some of the parameters of the MCNP model: uncertainties on the experimental positions, Cu density, modeling of the stuff around the Cu block and the position of the block respect to the FNG target. However, the variation of the investigated parameters could not justify the disagreement as they introduce negligible perturbation (<±2% max).

C/E

Fig. 6. Comparison between total flux calculated using JEFF-3.1.1 and JEFF-3.2. The lower figure shows the JEFF-3.1.1/JEFF-3.2 ratio which is decreasing with the penetration.

C/E

Penetration Depth (cm)

10

10

W(n, )

20

30

187

W

40

50

60

Penetration Depth (cm) FENDL-3 JEFF-3.1.1 JEFF-3.2

197

0

198

Au(n, ) Au 10

20

30

40

50

60

Penetration Depth (cm) Fig. 7. C/E for the copper experiment. Legenda: Square (black) is FENDL-3; circle (red) is JEFF-3.1.1; triangle (blue) is JEFF-3.2. (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)

ference from unity), however the same qualitative considerations hold. As far as the C/Es for 58 Ni(n,p)58 Co is concerned, its value is around unity within the uncertainties, this is found for all the libraries used. However, the calculation with JEFF-3.2 still produces worse agreement. For 115 In(n,n ) the C/Es are not only lower than unity, but a decreasing trend versus the penetration depth is

M. Angelone et al. / Fusion Engineering and Design 109–111 (2016) 843–847

observed. For JEFF 3.1.1 results C/E is 0.85 at 3.5 cm penetration depth and drops to 0.7 at 57 cm. JEFF-3.2 has still lower C/E with respect to the other libraries and, again, the discrepancy increases with the penetration depth. The main difference of 115 In(n,n ) with respect to the reactions mentioned so far is in the lower energy threshold of the used reaction, about 0.1 MeV, while the other reactions have thresholds ≥1 MeV (Table 1). Indeed, for the non threshold reactions 197 Au(n,␣) and 186 W(n,␣) the same steep decrease of the C/E is observed with penetration depth. All the used nuclear data libraries are in good agreement among them, but the C/Es are very far from 1. In the case of 197 Au(n,␥) C/E = 0.85 in the first position while C/E = 0.6 in the last one. Not substantial differences are between JEFF-3.1.1 and JEFF-3.2. For 186 W(n,␥) the results are even worse. In the first point C/E = 0.65 and the C/E drops to 0.4 in the last position using JEFF-3.1.1 and a slight down using JEFF-3.2. Sensitivity/uncertainty analysis was performed to complement the MCNP results. Discrete ordinate SUSD3D and Monte Carlo MCSEN codes were used. Very good agreement was observed between the sensitivities calculated using both codes. Very high sensitivity of the threshold reaction (from 1% up to more than 10% of change in reaction rate per 1%) and capture reactions (from 1% up to 6%) to Cu cross sections, was found. The threshold detectors (Al, Ni, and In) were found to be sensitive to the inelastic scattering, 63,65 Cu(n,2n) and 63 Cu(n,n p) cross sections. The thermal reactions 197 Au(n,␥) and 186 W(n,␥) are most sensitive to 63,65 Cu elastic scattering and (n,␥) capture reaction at lower energies. Other contributions are less significant. The JEFF-3.2 covariance evaluation of 63 Cu and 65 Cu are problematic at low energies of the (n,␥) reaction, and need to be re-evaluated 4. Conclusions A benchmark experiment on a block of pure copper has been performed at FNG using 14 MeV neutrons. 197 Au(n,␥)198 Au, 186 W(n,␥)187 W, 115 In(n,n )115 In, 58 Ni(n,p)58 Co, 27 Al(n,␣)24 Na, 93 Nb(n,2n)92 Nbm activation reactions were used. The calculations have been performed with MCNP5, using JEFF 3.1.1 and JEFF 3.2 nuclear data library for transport and IRDFF v1.05 dosimetry file for the reaction rates calculation. The analysis of the RR has demonstrated that underestimation up to 15% is found for high threshold reactions when using JEFF-3.1.1 and FENDL-3 (which indeed are very similar files) and up to 20% using JEFF-3.2. The underestimation of the low threshold reaction 115 In(n,n ) is still more severe. Furthermore, a steep decreasing trend versus penetration depth is observed. C/Es for 115 In(n,n ) range from 0.87 to 0.69 (first and the last experimental position, respectively) with JEFF 3.1.1. Also for this reaction the JEFF 3.2 library provides the largest underestimation. The observed behavior is consistent with that reported by the FNS team with JENDL-3.1. The non-threshold reactions 197 Au (n,␥) and 186 W (n,␥) show even more larger underestimation, with the C/Es decreasing with

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the penetration depth. The C/Es calculated with JEFF 3.1.1 range from 0.87 to 0.59 for 197 Au (n,␥), and from 0.68 to 0.41 for 186 W (n,␥) (first and last position, respectively). However, in this case all the libraries produce the same results. The present results call for a deep revision/re-evaluation of the copper cross sections. The new release JEFF-3.2 for Cu provided the highest disagreement in the C/E analysis and must be revised. To this end the results of the companion sensitivity/uncertainty postanalysis will help in identifying the main causes of uncertainty in the Cu cross sections. It worth to note that the largest discrepancy among the C/E values was observed for the thermal (capture) reactions suggesting problems and uncertainties in the 63,65 Cu capture and elastic cross sections at lower energy rather than at high energy. Acknowledgment The work leading to this publication has been funded partially by Fusion for Energy (F4E) under the Specific Grant Agreement F4E395-01. This publication reflects the views only of the author, and F4E cannot be held responsible for any use which may be made of the information contained therein. References [1] C. Ichihara, et al., Measurement and analysis of leakage neutron spectra from spherical assemblies of chromium, manganese and copper with 14 MeV neutrons, J. Nucl. Sci. Technol. 37 (4) (2000) 358–367. [2] I. Murata, et al., Neutron–nuclear data benchmark for copper and tungsten by slab assembly transmission experiments with DT neutrons, Fusion Eng. Des. 58–59 (2001) 617–621. [3] C. Konno, et al., Benchmark experiment on copper with DT neutrons for verification of neutron transport and related nuclear data of JENDL-3.1, Fusion Eng. Des. 28 (1995) 745–752. [4] K. Shibata, et al., JENDL-3: Japanese evaluated nuclear data library, version-3, JAERI-1319 (1990). [5] JEFF-3.2 Evaluated Data Library—Neutron Data, OECD-NEA (2014), https:// www.oecdnea.org/dbforms/data/eva/evatapes/jeff 32/. [6] The JEFF-3.1 Nuclear Data Library, JEFF Report 21, in: A. Koning, R. Forrest, M. Kellett, et al. (Eds.), OECD/NEA, Paris, 2006. [7] Nuclear Data Libraries for Advanced Systems—Fusion Devices (FENDL 3.0), Summary Report of the Third Research Coordination Meeting IAEA, Vienna, Austria, 6–9 December, (2011). [8] R.L. Perel, Upgrading of the MCSEN sensitivity software to comply with the current standard of the MCNP-5 Monte Carlo code, Final Report on Task 3.1 of the F4E Grant F4E-FPA-168.01, February 2014. [9] I. Kodeli, The SUSD3D code for cross-section sensitivity and uncertainty analysis—recent development, Transactions of the American Nuclear Society, vol. 104, Hollywood, Florida, June 26–30, 2011. [10] I. Kodeli et al., Cross-section sensitivity and uncertainty analyses of the FNG copper benchmark experiment, Fusion Eng. Des. this symposium. [11] D. Flammini, et al., Pre-analysis of the copper neutronics benchmark experiment for nuclear data validation, 28th SOFT 2014, San Sebastián, Spain, September 29–October 3, Fus. Eng. Des. 98–99 (2015) 1964. [12] X5 Monte Carlo Team, MCNP—A General Monte Carlo N-particle Transport Code: Version 5 User’s Guide, LANL Report LA-CP-03-0245, October 2005. [13] L.R. Greenwood, R. Paviotti-Corcuera, Summary Report of the Technical Meeting on International Reactor Dosimetry File: IRDF-2002, IAEA Report INDC (NDS)-435, August 2002. [14] A. Milocco, The D-T Source Routine Developed at ENEA and available in SINBAD Data Base, IJS Internal Report IJS-DP-9988 (2008).