Pulsed Neutron Imaging for Non-destructive Testing ... - Science Direct

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ScienceDirect Physics Procedia 88 (2017) 89 – 94

8th International Topical Meeting on Neutron Radiography, Beijing, China, 4-8 September 2016

Pulsed neutron imaging for non-destructive testing using simulated nuclear fuel samples Daisuke Itoa*, Tadafumi Sanoa, Jun-ichi Horia, Yoshiyuki Takahashia, Hiroyuki Hasemib, Takashi Kamiyamab, Ken Nakajimaa a

Research Reactor Institute, Kyoto University, 2-1010 Asashironishi, Kumatori, Sennan, Osaka 590-0494, Japan b Faculty of Engineering, Hokkaido University, Sapporo 060-8628, Japan

Abstract An integrated assessment method for a nuclear fuel with high decay heat and high radioactivity is required to establish fast reactor system with Trans-Uranium (TRU) fuel containing minor actinides. In addition, a Pu quantitation method with rapidity and accuracy is also necessary in a viewpoint of nuclear security. For these demands, a quantitative evaluation technique for nuclei concentration, thermal property and physical information of such fuel has to be developed. The present study focuses on the non-destructive imaging using pulsed neutrons. Experiments are carried out at Hokkaido University Neutron Source (HUNS) and a gas electron multiplier (GEM) is applied to obtain 2-D information of time-of-flight (TOF). To simulate a nuclear fuel pellet, a sample with equivalent thermal neutron cross-section to the enriched uranium fuel is prepared and the transmitted images of the simulated sample are acquired. Furthermore, a small piece of In, which simulates the Pu spot in the actual fuel, is inserted into the sample and the detectability of the small spot is discussed. © 2017 The Authors. Published by Elsevier B.V. This is an open access article under the CC BY-NC-ND license © 2017 The Authors. Published by Elsevier B.V. (http://creativecommons.org/licenses/by-nc-nd/4.0/). Peer-review under responsibility of organizing the organizing committee of ITMNR-8. Peer-review under responsibility of the committee of ITMNR-8 Keywords: nuclear fuel; non-destructive testing; simulated fuel pellet; GEM

* Corresponding author. Tel.: +81-72-451-2373; fax: +81-72-451-2431. E-mail address: [email protected]

1875-3892 © 2017 The Authors. Published by Elsevier B.V. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/). Peer-review under responsibility of the organizing committee of ITMNR-8 doi:10.1016/j.phpro.2017.06.011

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1. Introduction Nuclear power generation is one of the options for long-term stable energy supply. However, for reduction of the environmental load due to the high-level radioactive waste from the nuclear power plants, a fast reactor system using TRU fuel with minor actinides (MA) should be developed (Salvatores et al., 2009). An integrated assessment method for a nuclear fuel with high decay heat and high radioactivity is required to establish the fast reactor system. In addition, a Pu quantitation method with rapidity and accuracy is also necessary in a viewpoint of nuclear security. Especially, Pu spots attributed to the insufficient mixing in fuel production process is very important for safety operation of the reactor and its size and local distribution has to be quantified non-destructively. For these demands, a quantitative evaluation technique for nuclei concentration, thermal property and physical information of such fuel has to be developed. Non-destructive testing method using neutrons is very effective tools to study materials including heavy metals such as nuclear fuel. Previously, some studies applied conventional neutron radiography to visualized nuclear fuels (Yasuda et al., 2005; Grosse, et al., 2011; Craft et al. 2015). The shape and the structure inside the fuel could be understood from the neutron radiographs. On the other hand, a pulsed neutron resonance absorption spectroscopy with computed tomography has been developed, and it has been applied to quantify the temperature distribution (Kamiyama et al., 2005) and a nuclide density distribution in the sample (Kamiyama et al., 2009). This technique would be applicable to non-destructive testing of the nuclear fuel and also to detecting the Pu spot in the fuel pellet. In the present study, the pulsed neutron resonance absorption spectroscopy with a position sensitive detector was applied to study the feasibility of the Pu spot detection in the nuclear fuel pellet. For this purpose, a simulated fuel sample with similar neutron transmission characteristics to the actual UO2 fuel was prepared and the small Indium pieces, which have relatively similar resonance absorption properties to 240Pu, were inserted into the sample. Then, the possibility of the small piece detection was investigated from the obtained radiographs and calculated distributions. 2. Experimental details To prepare a simulated nuclear fuel sample with a thermal neutron absorption cross-section equivalent to that of 3.4% enriched 235UO2 fuel pellet, bismuth oxide (Bi2O3) and neodymium oxide (Nd2O3) were mixed. Monte-Carlo simulation code (MVP2.0 + JENDL-4.0) was applied to calculate the neutron capture cross-section. The concentration of the mixing ratio between Bi2O3 and Nd2O3 was adjusted and the thermal neutron capture cross1.E+01 UO2 (3.4wt%) UO2 (3.4wt%)

Macroscopic cross section (/cm)

Simulated fuel ໝ٘೫ྋ

1.E+00

1.E-01

1.E-02

1.E-03

1.E-04 1.0E-03

1.0E-01

1.0E+01 1.0E+03 Energy (eV)

1.0E+05

1.0E+07

Fig. 1. Calculated macroscopic capture cross-section of UO2 and simulated fuel.


Fig. 2. Pulsed neutron imaging samples with (a) three In pieces ( I2 mm x 2mm), (b) three In pieces (I1 mm x 1mm), (c) three different In pieces.

Table 1. Measurement times. TOF range

0-600 μs (Δt =150 ns)

0-20 ms (Δt =5 μs)

with sample

11.4 h

2.0 h

without sample

10.1 h

2.8 h

section of the mixture was fitted to be that of UO2 fuel. The calculated cross-sections are shown in Fig.1. As a result, the ratio of the simulated fuel was 42 wt% for Bi2O3 and 58 wt% for Nd2O3. The mixture of Bi2O3 and Nd2O3 was encapsulated to aluminum holder and small Indium pieces were inserted into the mixture, as shown in Fig. 2. The size of the samples was 15 mm in diameter and 5 mm in depth. Three Indium pieces with 2 mm and 1 mm size were inserted into the mixture in Fig. 2(a) and (b), respectively, and three different size pieces were inserted in Fig. 2(c). The experiments were performed at HUNS (Hokkaido University Neutron Source). As the operation condition of the linac, the energy is 35 MeV and the current is 35 μA. Pulse width and repetition frequency are 3 μs and 50 Hz, respectively. The neutron flux in the energy range from 0.01 to 0.1 eV is 4.5×104 n/cm2/s at the sample position. As the position sensitive detector, a GEM detector (THIN-GEM; Bee Beans Technologies Co., Ltd.) was used to obtain two-dimensional information of time-of-flight (TOF). The spatial resolution is 0.8 mm and the total detecting area is 100 x 100 mm2. The measurements were carried out for two different neutron energy regions. One focuses on the high-energy resonance absorption region and another is thermal neutron region. Measurement times for each region were summarized in Table 1. In the imaging of the sample, the direct beam image was subtracted from the imaging data. 3. Results and discussion The transmission spectrum of 0.5 mm Indium plate covering the detecting area is shown in Fig. 2. The resonance absorption peaks at 1.46, 3.82 and 9.07 eV were observed from the figure. In particular, 1.46 eV resonance absorption is seen around TOF = 400 μs, and this attenuation is large and has broad shape. Thus, this region was used to make energy-resolved radiograph for Indium pieces. Figure 4 shows the energy-resolved transmission neutron radiographs of the test sample illustrated in Fig. 2. In the thermal neutron region, the transmitted neutron flux decreases due to the absorption of the sample which is the mixture of Bi2O3 and Nd2O3, and it is found that 1 mm and 2 mm pieces could be detected in the radiograph. In the resonance absorption region, the attenuation of the mixture was less and the pieces were observed more clearly, as shown in Fig. 4(b). The existence of the small Indium piece with 0.5 mm size which is smaller than the pixel size of the detector could be also detected in the resonance absorption radiograph. From these results, it is found that not

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Daisuke Ito et al. / Physics Procedia 88 (2017) 89 – 94 㻌 4.0x10-3 3.5x10-3

In 0.5mm thick Direct beam

-3

2.5x10-3 2.0x10-3

㻌

Counts [-]

3.0x10

1.5x10-3 1.0x10-3 5.0x10-4 0.0

0

100

200

300

400

500

600

TOF [Ps] Fig. 3. Neutron transmission spectrum of In.

a

b 0.5mm In

Fig. 4. Neutron radiographs in (a) thermal neutron region (TOF = 0 ~ 20 ms) and (b) resonance absorption region of In (TOF = 360 ~ 510 μs).

only the spot detection but also the resonance nuclide identification is possible by using this method. However, the noise level is very high in Fig. 4(b) and there are some small dark pixels even though no absorption materials exist. Thus, more neutron flux or time is required for the detection of the pieces smaller than the pixel size. The spatial distribution of the transmitted neutron flux was calculated by PHITS2 + JENDLE-4.0 in order to compare with the imaging results. The calculated macroscopic capture cross-section in Fig. 1 was used and three In pieces with different size were put into the simulated sample. Tally face was 20 mm downstream of the sample. In this calculation, the generated particle number was 3.6×1010 (equivalent to about 1000 hours’ measurement at HUNS). The mesh size was 0.8 mm which is the same with the experiments using GEM detector. The calculated results are shown in Fig. 5. Two energy regions were selected to compare the effect of the resonance absorption of In. Large attenuation of 2 mm piece was observed in the resonance absorption region of In, as shown in Fig. 5(a). In addition, not only 1 mm but also 0.5 mm size could be detected. This is consistent with the experimental result in Fig. 4. However, in higher neutron energy region, it is difficult to find the smaller piece in Fig. 5(b). As a consequence, the calculated results represent the different images in different energy domain. Therefore, appropriate energy domain should be determined when the resonance imaging is performed.

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Neutron lethargy//source source Neutronflux flux // letharegy

Fig. 5. Calculated spatial distributions of transmission neutron flux in the range of (a) E=1.0~10 eV (TOF = 158 ~ 492 μs) and (b) E=10~100 eV (TOF = 53 ~ 158 μs).

2.E-04 2.E-04 2mmφ 2mmφ

1mmφ 1mmφ

0.5mmφ 0.5mmφ

none ໃ͢

2.E-04 2.E-04

1.E-04 1.E-04

5.E-05 5.E-05

0.E+00 0.E+00 0.1 0.1

1.0 1.0

Energy [eV] Energy (eV)

10.0 10.0

100.0 100.0

Fig. 6. Line profiles of calculated neutron transmission properties of In pieces.

In the calculated results, neutron transmission spectra were estimated at each pixel which has In piece. The spectra are shown in Fig. 6. The size of In pieces is 0.5, 1.0 and 2.0 mm. The resonance absorptions of the In were observed at 1.46, 3.82 and 9.07 eV as with the experimental result in Fig. 3. It is found that the transmission neutron flux is varied by the size or amount of the In in the measurement area. Therefore, the spot size or the concentration of the In might be quantified by using the shape of the neutron spectrum. 4. Conclusions The pulsed neutron resonance absorption spectroscopy with the GEM detector was applied to study the feasibility of the Pu spot detection in the nuclear fuel pellet. The simulated fuel sample with similar neutron transmission characteristics to the actual UO2 fuel was prepared and the small In pieces were inserted into the sample. The possibility of the small piece detection was investigated from the measured radiographs and calculated distributions, and the following conclusions were obtained. x The In spot with larger size than the spatial resolution of the detector could be recognized in the thermal neutron radiograph. x The 0.5 mm spot was detected in the energy-resolved radiograph. So, the pulsed neutron imaging focusing

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on the resonance absorption is required for smaller size detection. The calculated results by PHITS2 + JENDLE-4.0 showed similar tendency with the imaging results. This study showed the possibility for not only the spot detection but also the resonance nuclide identification. As a next step, a sample with multiple nuclides will be prepared and measured by this method. Also a detector with higher spatial resolution and the resistance to radiations should be developed for irradiated nuclear fuel imaging.

x

Acknowledgements This study includes the result of “Development of Non-Destructive Methods Adapted for Integrity test of Next generation nuclear fuels” entrusted to the Kyoto University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT). References Craft, A.E., Wachs, D.M., Okuniewski, M.A., Chichester, D.L., Williams, W.J., Papaioannou, G.C., Smolinski, A.T., 2015, Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory, Physics Procedia, 69, 483-490. Kamiyama, T., Ito, J., Noda, H., Iwasa, H., Kiyanagi, Y., Ikeda, S., 2005, Computer tomography thermometry-an application of neutron resonance absorption spectroscopy, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 542, 258-263. Kamiyama, T., Miyamoto, N., Tomioka, S., Kozaki, T., 2009, Epithermal neutron tomography using compact electron linear accelerator, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 605, 91-94. Grosse, M., Steinbrueck, M., Kaestner, A., 2011, Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 651, 315-319. Salvatores, M., Chabert, C., Fazio, C., Hill, R, Peneliau, Y., Slessarev, I., Yang, W.S., 2009, Fuel cycle analysis of TRU or MA burner fast reactors with variable conversion ratio using a new algorithm at equilibrium, Nuclear Engineering & Design, 239, 10, 2160–2168. Yasuda, R., Matsubayashi, M., Nakata, M., Harada, K., Amano, K., Sasajima, F., Nishi, M., Horiguchi, Y., 2005, Application of neutron imaging plate and neutron, CT methods on nuclear fuels and materials, IEEE Transactions on Nuclear Science, 52, 1, 313-316.