Acta Polytechnica https://doi.org/10.14311/AP.2025.65.0237 Acta Polytechnica 65(2):237–242, 2025 © 2025 The Author(s). Licensed under a CC-BY 4.0 licence Published by the Czech Technical University in Prague URANIUM FOIL NEUTRON ACTIVATION ANALYSIS Ján Kozica,b,∗, Milan Štefánika,b a Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors, V Holešovičkách 2, 180 00 Prague, Czech Republic b Czech Academy of Sciences, Nuclear Physics Institute, Řež 130, 250 68 Řež, Czech Republic ∗ corresponding author: kozicjan@cvut.cz Abstract. At the Department of Nuclear Reactors of the Faculty of Nuclear Sciences and Physical Engineering of the Czech Technical University (FNSPE CTU) in Prague, there is a need to implement a neutron activation analysis as a tool for the separation of uranium foils depending on various 235U and 238U contents. Experiments were carried out on the training reactor VR-1 “Sparrow”, which is operated at the institute mentioned above. The uranium foils were analysed using the neutron activation analysis and gamma-ray spectrometry, and were separated into two groups, the first group contained foils of depleted uranium and the other contained natural uranium foils. The uranium foils can be further used for educational purposes or experimental applications, such as the measurement of neutron spectral indexes, neutron activation analysis of uranium ore, or delayed neutron analysis. The neutron activation analysis has been extended to determine whether foils contain natural or depleted uranium. Keywords: Production rate, uranium foils, neutron activation analysis, gamma-ray spectrometry, Training Reactor VR-1, HPGe detector. 1. Introduction Neutron activation analysis can be used as a powerful non-destructive tool for the composition analysis of investigated samples. It can be used in various nuclear facilities, such as nuclear reactors, nuclear generators, or accelerator-driven neutron sources. The low-power reactor VR-1 “Sparrow” [1] has been operated at the Department of Nuclear Reactors of the Faculty of Nuclear Sciences and Physical Engi- neering of the Czech Technical University (FNSPE CTU) in Prague [2] since 1990 [1]. Its main purpose is to educate students, however, it can also be used to train qualified personnel or conduct research [1]. Various experimental equipment is available at the VR-1 reactor which allows its utilization for experi- mental tasks such as neutron activation analysis [3–7] or neutron radiography [8]. There is a long-standing neutron activation analysis program at the institute mentioned above [3–7]. This paper deals with the neutron activation analysis of uranium foils and is focused on an extension of the neutron activation analysis to determine whether foils contain natural or depleted uranium. The goal of this research is both the extension of the neutron activa- tion analysis mentioned above and the separation of uranium foils depending on various 235U and 238U contents. There are multiple studies considering the determi- nation of uranium concentration based on the neutron activation analysis measurements, e.g. [9–13]. All the research mentioned measures uranium concentration based on the neutron activation of 238U. Even though the studies analyse the uranium concentration, none consider uranium’s isotopic composition. The need of the Department of Nuclear Reactors is to separate the uranium foils depending on various 235U and 238U contents, therefore, the NAA of uranium contents needs to be further extended to be able to determine whether the uranium foils are made from depleted or natural uranium. 2. Materials and methods 2.1. Neutron activation analysis Neutron activation analysis is an experimental method which is used for qualitative and quantitative exami- nation of the composition of an unknown sample [14]. The sample is irradiated in a well-described neutron field and then a gamma-ray spectrometry method is employed for further analysis. The qualitative exami- nation means that the composition of the sample is determined from the measured gamma-ray spectrum via identification of the characteristic gamma radia- tion of the irradiated sample and further traction of reaction channels based on the energy distribution of the neutron field in which was the sample irradiated. Light water or pool-type training or research reac- tors have well-moderated neutron energy spectrum in specific positions of the reactor core, therefore, the ra- diative capture dominates as the reaction channel and the determination of the composition of the sample becomes quite certain [15]. Often there is a need to analyse quantitatively a par- ticular element, therefore, a production rate is intro- 237 https://doi.org/10.14311/AP.2025.65.0237 https://creativecommons.org/licenses/by/4.0/ https://www.cvut.cz/en Ján Kozic, Milan Štefánik Acta Polytechnica duced by equations [15]: PR = S(Eγ )λ treal tlive Iγ (Eγ )εγ (Eγ )(1−e−λtirr )e−λtcool (1−e−λtreal ) , (1) PR =N0 ∫ +∞ 0 ϕ(E)σ(E)dE, (2) where S(Eγ) is an area under the peak of full absorption in the measured gamma spectra, λ is a decay constant of the produced nuclide, tlive is the duration of the gamma spectrum measure- ment, treal is the duration of the gamma spectrum measure- ment with addition of a dead time, Iγ(Eγ) is an intensity of the particular gamma radi- ation, εγ(Eγ) is a detection efficiency of an apparatus for the particular gamma radiation, tirr is the duration of the irradiation, tcool is the cooling time, N0 is a number of stable nuclei in the sample, ϕ(E) is a spectral neutron flux density, σ(E) is microscopic cross section of the particular reaction. From a combination of the Equations (1) and (2) an expression for the number of neutral nuclei in the sample N0 could be derived: N0 = S(Eγ )λ treal tlive Iγ εγ(1−e−λtirr)e−λtcool(1−e−λtreal)∫ +∞ 0 ϕ(E)σ(E)dE , (3) N0 = mNA MM , (4) where m is a weight of the analysed nuclide, NA is an Avogadro’s number, MM is a molar weight of the analysed nuclide. A combination of the Equations (3) and (4) gives an expression for the determination of the weight of the analysed nuclide m. Usually the information about the neutron field is not available, therefore, a comparative neutron ac- tivation analysis is introduced. In this method, an unknown sample and a standard is irradiated using the same experimental arrangement. Then a fraction of production rate of the unknown sample PR, sample and the standard PR, standard is analysed. For example, the weight of the observed nuclide in the unknown sample could be determined from the fraction men- tioned above and the weight of the observed nuclide in the standard by an expression [15]: msample = PR, sample PR, standard · mstandrad. (5) Figure 1. Microscopic cross sections for the most important interactions of the gamma radiation with germanium [16]. Figure 2. The Training Reactor VR-1 hall [22]. 2.2. Gamma-ray spectrometry The gamma-ray spectrometry is a method which is used for a qualitative and quantitative analysis of the gamma radiation. The qualitative analysis means the identification of the gamma radiation and the quanti- tative analysis stands for a counting of the identified detected gamma photons by a detection apparatus. The most important interactions of the gamma radia- tion with a matter are photoelectric absorption, Comp- ton’s scattering and electron-positron pair production. Microscopic cross sections of the gamma interactions mentioned above with germanium are illustrated in Figure 1, since germanium crystal is manufactured in HPGe detectors utilized in this research and they have an excellent energy resolution, therefore, HPGe detectors are widely used in the world. The data were obtained from online database [16–21]. 2.3. Training reactor VR-1 The Training Reactor VR-1 “Sparrow” (see Figure 2) is a light water pool-type reactor with a nominal thermal power of 100 W which is from the point of view of a relatively high volume of a coolant a zero power reactor. The VR-1 reactor is used for the ed- ucation of students, the training of qualified staff of the nuclear industry and experimental applications such as neutron activation analysis, neutron radio- graphy, the development of new shielding materials or the development of an advanced nuclear fuel. For the experimental applications, the VR-1 reactor is 238 vol. 65 no. 2/2025 Uranium foil neutron activation analysis Figure 3. The reactor core of the VR-1 reactor [22]. equipped with various dry vertical channels, a radial and a tangential channel. The samples for the neu- tron activation analysis are placed into the chosen dry vertical channel. A reactor core of the VR-1 reactor is assembled on a core support plate with a square lattice of 8 × 8 cells (see Figure 3). IRT-4M nuclear fuel with 235U enrichment of 19.7 % is utilized in the core. The number of fuel cells in the reactor core varies in the range from 16 to 24 depending on the actual core configuration. Absorption rods UR-70 con- taining a cadmium absorber are employed to control the reactor. The number of absorption rods varies in the range from 5 to 7 depending on the actual core configuration. Maximal neutron flux density reaches a value of the order 109 cm−2 s−1 [1]. 2.4. Data acquisition Two experiments were carried out on the VR-1 reactor. The first and the second experiments were conducted on 2nd of June 2023 and 9th of June 2023 respectively. Totally nine uranium foils (listed in Table 1) were irradiated in a dry vertical channel on the position A4 (see Figure 4). The reactor core configuration C20 [23] was utilized and the parameters of the reac- tor VR-1 were the same for both of the experiments, particularly the power of the reactor reached a value of 107 imp s−1 (ca. 10 W) and the irradiation lasted ca. 20 minutes. The foils U-dep2 and U-dep3 were utilized as standards from depleted uranium. After the irradi- ation, the foils were moved to a gamma-spectrometric laboratory where their gamma spectra were measured two or three times. Three semiconductor HPGe de- tectors Canberra [24] were employed simultaneously, therefore, even fission products with a half-life of the order of tens of minutes were measured. 3. Results Since the fact that the analysed uranium foils have different molecular structures (U, UO2, U3O8) the standard comparative neutron activation analysis for the 239U to analyse 238U can not be employed. How- ever, the fraction of production rates of the unknown uranium foil and the standard could be compared to the fraction of the sum of the fission yields of 235U, Foil Weight [mg] Type U-dep2 422.4 Standard U-10 778.2 Sample U-11 707.3 Sample S 442.7 Sample J 393.9 Sample U-dep3 320.8 Standard k0 390.5 Sample U-1 250.4 Sample U-X 147.3 Sample Table 1. Analysed uranium foils. Figure 4. A scheme of the reactor core configuration C20 [23]. 238U and the fission yield of 238U since the standards were made of depleted uranium. The fission products were chosen based on the following conditions: • the yield of the fission product is higher than 5 % for 235U, • the yield of the fission product is higher for 235U than for 238U, • the fission product emits gamma radiation and has a half-life such that it is possible to measure its gamma spectrum at least two times even after the necessary cooling time between the end of irradia- tion and the start of gamma spectra measurement. All of the chosen fission products are summarized in Table 2 together with fission product yield for 235U y235 and 238U y238, fraction y238+y235 y238 , charac- teristic gamma-ray energy Eγ , intensity Iγ and half- life T1/2. The fission product yields were obtained from ENDF/B-VIII.0 nuclear database [25] and in- formation about the gamma radiation was obtained from an online database [26]. The production rates for the chosen fission prod- ucts in all of the foils were determined using Equa- tion (1) from quantities obtained from the measured 239 Ján Kozic, Milan Štefánik Acta Polytechnica Product y238 y235 y238+y235 y238 Eγ [keV] Iγ [%] T1/2 Unit 142Ba 0.09 0.15 2.74 255.30 20.50 10.60 min 138Xe 0.10 0.15 2.51 258.41 31.50 14.08 min 101Tc 0.12 0.14 2.15 306.86 89.00 14.22 min 131Sb 0.07 0.08 2.15 943.40 47.00 23.03 min 146Pr 0.06 0.08 2.38 453.88 48.00 24.15 min 138Cs 0.11 0.18 2.67 1 435.80 76.30 33.41 min 134Te 0.12 0.16 2.36 767.20 29.50 41.80 min 91Y 0.08 0.16 3.07 555.57 95.00 49.71 min 134I 0.14 0.21 2.48 847.03 95.40 52.50 min 97Nb 0.11 0.18 2.62 658.08 98.00 72.10 min 87Kr 0.03 0.08 3.28 402.59 49.60 76.30 min 139Ba 0.11 0.18 2.64 165.86 23.70 83.06 min 142La 0.09 0.16 2.80 641.29 47.00 91.10 min 92Sr 0.08 0.17 3.06 1 383.93 90.00 2.71 h 90Y 0.06 0.16 3.46 202.51 97.30 3.19 h 135I 0.12 0.17 2.35 1 260.41 28.90 6.57 h 135Xe 0.13 0.19 2.47 249.77 90.00 9.14 h 97Zr 0.11 0.17 2.58 743.36 93.00 16.91 h 133I 0.13 0.19 2.47 529.87 87.00 20.80 h 133Xe 0.13 0.19 2.48 81.00 38.00 5.24 d Table 2. Fission products which were chosen for analysis [25, 26]. Foil J Foil S Foil U-10 Foil U-11 Det. PR PR,standard Result PR PR,standard Result PR PR,standard Result PR PR,standard Result y238+y235 y238 DET 1 0.97 DU 1.22 DU 3.16 NU 12.28 NU 2.66 DET 2 0.92 DU 1.11 DU 11.79 NU 1.85 DU 2.61 DET 3 0.90 DU 1.07 DU 1.85 DU 1.66 DU 2.63 Result DU DU NU DU Table 3. Results from the first experiment. Foil k0 Foil U-1 Foil U-X Det. PR PR,standard Result PR PR,standard Result PR PR,standard Result y238+y235 y238 DET 1 2.53 NU 1.23 DU 0.86 DU 2.44 DET 2 1.95 DU 1.29 DU 0.47 DU 2.76 DET 3 1.97 DU 0.78 DU 0.50 DU 2.65 Result DU DU DU Table 4. Results from the second experiment. gamma spectra. Then fractions of production rates of the analysed foils and the standard PR PR,standard were calculated. The fractions of production rates were determined for measurements on all three HPGe de- tectors used. The fractions of production rates were compared to the fraction of the fission product yields y238+y235 y238 . The relative difference of 20 % was allowed to determine whether the foils contain natural or de- pleted uranium. The results from both of the exper- iments are listed in Tables 3 and 4 where depleted and natural uranium are denoted as DU and NU respectively. As a verification of the results a fraction of the production rates for the chosen fission products and production rate for 239U PR PR,239U within the analysed uranium foils were determined. Averaged values for all of the uranium foils are listed in Tables 5 and 6. The fractions for uranium foils with an unknown content of 238U and 235U were compared to the same fraction for the standards containing only depleted uranium. As can be seen from Tables 3, 4, 5 and 6 there is agreement in between the newly imposed methodology and the validation method. The processing of data and its analysis is an exten- sive task. The preparation of data for chosen fission product measured on chosen detector for the first and the second experiment are summarized in Tables 7 and 8 respectively. The fraction y238+y235 y238 which serves 240 vol. 65 no. 2/2025 Uranium foil neutron activation analysis Foil U-dep2 Foil J Foil S Foil U-10 Foil U-11 Det. PR PR,239U PR PR,239U Result PR PR,239U Result PR PR,239U Result PR PR,239U Result DET 1 0.08 0.07 DU 0.08 DU 0.22 NU 0.83 NU DET 2 0.04 0.02 DU 0.04 DU 0.28 NU 0.03 DU DET 3 0.08 0.08 DU 0.04 DU 0.06 DU 0.05 DU Result DU DU NU DU Table 5. Verification of the results from the first experiment. Foil U-dep3 Foil k0 Foil U-1 Foil U-X Det. PR PR,239U PR PR,239U Result PR PR,239U Result PR PR,239U Result DET 1 0.07 0.10 DU 0.07 DU 0.21 NU DET 2 0.03 0.02 DU 0.05 DU 0.04 DU DET 3 0.05 0.06 DU 0.05 DU 0.02 DU Result DU DU DU Table 6. Verification of the results from the second experiment. Foil U-dep2 Foil J Foil S Foil U-10 Foil U-11 Prod. PR PR,239U PR PR,standard PR PR,239U PR PR,standard PR PR,239U PR PR,standard PR PR,239U PR PR,standard PR PR,239U 134Te 0.12 0.97 0.13 1.22 0.14 3.16 0.22 12.28 0.83 Table 7. Data preparation from the measurement on the DET 1 for the first experiment. Foil U-dep3 Foil k0 Foil U-1 Foil U-X Prod. PR PR,239U PR PR,standard PR PR,239U PR PR,standard PR PR,239U PR PR,standard PR PR,239U 87Kr 0.02 2.02 0.02 0.78 0.02 0.47 0.02 Table 8. Data preparation from the measurement on DET 3 for the second experiment. as comparative condition for value of PR PR,standard could be found in Table 2. 4. Conclusion In the present work, the neutron activation analysis have been extended to separate uranium foils depend- ing on various 235U and 238U contents. Furthermore, the uranium foils have been separated into two groups of foils. The first group contains natural uranium foils and the other contains depleted uranium foils. The results are summarized in Tables 3 and 4. The results were verified with another method and the results of the verification method are listed in Tables 5 and 6. The verification method is in excellent agreement with the newly imposed methodology. The analysed foils can be used further in the education of students or in experimental applications such as measurements of neutron spectral indexes, neutron activation analysis of uranium ore or delayed neutron analysis. Finally, two new experimental methods for the analysis of the 235U and 238U contents were imposed. The experiments are a part of the program devoted to the neutron activation analysis measurements at the Department of Nuclear Reactors of FNSPE CTU in Prague. Acknowledgements Research activities and irradiation experiments at the Training Reactor VR-1 were supported by the project Large Research Infrastructures (project no. 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