International Journal of Science and Advanced Technology (ISSN ) Volume 3 No 12 December 2014

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1 Monte Carlo Simulation Approach for Generating NaI Detector Response Functions (DRFs) in Consideration of Delayed Gamma-rays due to Detector Activation Hamed Panjeh, Reza Izadi Najafabadi Department of Physics Faculty of Science, Ferdowsi University of Mashhad Mashhad, Iran John Kettler Germany Institute of Nuclear Fuel Cycle RWTH Aachen University Aachen, Germany Motahhareh Abbasi Department of Physics Faculty of Science, Guilan University Rasht, Iran Abstract During neutron irradiation in Prompt Gamma Neutron Activation Analysis (PGNAA), absorption of thermal neutrons by the detector is inevitable. Therefore, the final gamma spectrum will be a mixture of natural background, prompt and delayed gamma-rays originated from different setup parts including moderator, sample and detector itself. A Delayed-Gamma-Monte-Carlo Code (DGMC) was written to determine delayed gamma neutron activation spectrum arising from an activated detector. Delayed gamma-rays spectral response of 3" 3" Na(Tl) crystal to thermal neutron beam was also simulated while the neutron beam ''on". Keywords-Prompt Gamma Neutron Activation Analysis, PGNAA, Delayed Gamma Neutron Activation Analysis, DGNAA, Delayed-Gamma-Monte-Carlo Code, DGMC, Detector Response to Neutron, Neutron Activation, Delayed Gammarays, Detector, NaI, FORTRAN, MCNP, Monte Carlo I. INTRODUCTION Neutron activation is a non-destructive method for analyzing composite materials. Depending on the methods, it is divided into two categories: Delayed Gamma Neutron Activation Analysis (DGNAA) and Prompt Gamma Neutron Activation Analysis (PGNAA) [1, 2]. The PGNAA technique uses both Thermal Neutron Analysis (TNA) and Fast Neutron Analysis (FNA) [3]. In PGNAA the gamma spectrum is taken during the irradiation [4, 5], while in DGNAA the gamma spectrum is gotten when the neutron beam is off [6, 7]. Usually in DGNAA the sample is irradiated in a source room or in a nuclear reactor core and the spectrum will be acquired in the spectroscopy room. In all PGNAA applications, both sample and detector simultaneously receive thermal neutrons during neutron irradiation [8, 9]. In PGNAA, gamma spectrum acquisition of the prompt gamma-rays has to be in simultaneous with neutron irradiation. In these conditions, depending on the 9 setup geometry, the distance between the source and detector, and also the neutron source spectrum, detector may receive threshold damage-dose of thermal neutrons. Crystal defects are also the consequence results, which cause undesirable effects on the spectrum and detector resolution. Inadequate shielding of the gamma-ray detector in a high intensive field of neutron can cause deterious effects on the detector responses and the acquired spectra [10]. Long time irradiation leads usually to disturbance in calibration and the detectors have to be replaced with the new ones. In addition, the main problem in online analysis is the interfering between the prompt gamma-rays emitted from the active sample and those prompt and delayed gammarays originated from the detector cell and other setup parts. Hence, the final gamma spectrum will be a mixture of natural background, prompt and delayed gamma-rays originated from different setup parts including moderator, sample and detector itself. In the high flux of thermal neutrons like accelerator fields or high intensive radioisotope neutron source, disturbing effects on the spectrum causing from these interfering phenomena will be more. These disturbing effects become more important when the energy of gamma-rays emitted from the activated detector and the other setup parts is approximately near the interested energy range of prompt gamma-rays emitted from the sample under study. Accumulation of these gamma-rays in the Region of Interest (ROI) is unavoidable and discrimination between them is an important challenge in PGNAA [11]. In the most cases, the peaks are not resolved and only one broad peak appears. Hence PGNAA technique, contrary to DGNAA, has some disadvantages as mentioned, but in some conditions, PGNAA is the only inevitable choice such as cases; 1) where there is no sufficient time to activate material and then get the spectrum, 2) when the signature of the material is only prompt gamma-rays, 3) when the intensity of the delayed

2 gamma-rays is so weak to be detected among the high intensive prompt gamma-rays and the prompt pulses are prominent. In all kinds of PGNAA applications, the architecture of the PGNAA setup needs to be aware about the conditions and effects of each interfering factor in the final results to optimize the geometry, using each kind of neutron source and the detector shields. II. GAMMA-RAY SOURCES IN ONLINE ANALYSIS As mentioned already, in PGNAA both detector and setup structure materials (including source housing, sample, shield and moderator) will be activated during neutron irradiation. Principally, indication of the role of detector, sample, moderator and other parts of setup in the final spectrum needs separated investigations, including measurement of the detector materials prompt and delayed gamma-rays and also sample and other parts prompt and delayed gamma-rays, one by one. One of the most important challenges is simulating and measuring the delayed gamma-rays produced in the detector itself, during irradiation. In the present work, only the detector and its activation are part of interest. A new approach for the simulation of delayed gamma-rays due to detector neutron activation has been proposed. Recently some new features beside delayed gamma calculation feature are added to MCNPX 2.6 and upper version, which are hoped to predict the delayed gamma-rays in the final spectrum. The time needed for mass calculations is also an important factor [12, 13]. In our recent project (under review in ARI Journal) entitled "Direct Simulation Monte Carlo Code for Time Dependent Foil Activation Measurements", a straightforward method is introduced which was called here to be used in DGNAA. The present work proposes a direct simulation method to calculate delayed gamma spectrum based on Monte Carlo technique. A Delayed-Gamma- Monte-Carlo Code (DGMC) was developed for determination of delayed gamma-rays produced in the sample and detector materials. Considering the sample activation, interference of delayed gamma-rays with the prompt gamma-rays arising from the sample material, and activation of the other parts are subjected in another work. The main goal of the study was the identification of the detector material activation contribution to the total delayed gamma-rays and total obtained spectrum. In order to simulate the detector activation and measure its contribution to the acquired spectrum, period of irradiation and spectrometry interval have to be known exactly. Knowledge about setup configuration and the setup running conditions are also needed. For example in a land mine detection system, activation will run for a short period of time, while the condition governed on the PGNAA system used in cement factory is completely different. The size of whole system and the source type are also different. The online analyzer system in cement factory (installed after stone crusher) is usually turned on continuously. The system and the detectors used in this setup receive continuous flux of neutrons for a long period of time. Obviously in this case, activation of the detector and its effects on the spectrum will be significant in comparison with a land mine detection system. 10 III. NAI DETECTOR ACTIVATION Nowadays, the usage of NaI detectors in online analyzers is so common. Among the gamma detectors, NaI has some benefits such as lower costs, high efficiency, easy installation and is also commonly used in industry and laboratory setups [14]. NaI detector has the advantages of being efficient for high-energy gamma-rays, rugged, and also can be used without cooling [11, 15, 16]. Therefore, NaI detector was chosen and its activation was investigated here. NaI detectors are used frequently in industrial PGNAA applications such as total body nitrogen estimations like BCCA, bulk material analysis, oil well logging [17-23], explosive detection, landmine detection system and for raw material analysis in coal and cement factory. Spectrum analyzing is so crucial in the success of PGNAA method and reliability of the results. So that getting knowledge about the entire region of energy and the origin of each peak are of interest. Activation of the NaI can affect the output results. Hence, knowledge about those ROI that are affected due to the detector activation is so important. Although some studies are done on the neutron effects on the NaI Detector [24, 25], there was only one experimental attempt to extract the delayed gamma-rays contributions to the total spectrum. Originally, Gardner [11] used the experimental method accompanying the least square technique for obtaining detector activation spectral data. In order to study the detector activation and extract the delayed gamma-rays of the detector, the pulses resulting from neutrons interacting with the crystal are categorized into two classes: 1) the prompt pulses, which appear instantaneously with the neutron interaction and 2) the delayed pulses, which may appear long after the neutron interaction. The delayed pulses create an additional problem, because their intensity depends on the pervious exposure history of the crystal. These delayed pulses are, of course, due to the buildup of radioactive isotopes in the NaI crystal under neutron irradiation. Table 1 shows the most probable interactions occurring in the NaI crystal. In principal, the detector activation analysis is a time dependent problem. Various materials with different halflives confirm this fact. The DSMC is written based on 5 separated phases. DSMC simulates all the processes exactly from activation up to spectroscopy. It also includes the delay between stopping irradiation and staring spectroscopy. The code itself is a combination of two coupled MCNP [26] codes joined to complementary FORTRAN programs. Monte Carlo FORTRAN codes are necessary for the parts which are not included in the MCNP, such as time calculations. IV. SIMULATION METHOD AND ASSUMPTIONS The condition assumed here is that the detector is irradiated in a period of time with a thermal neutron beam. Delayed gamma-rays spectral response of 3" 3" Na(Tl) crystal to thermal neutron beam is investigated while the neutron beam ''on". The condition of simulation is that irradiation and spectrum acquisition are in simultaneous. Detector materials will be activated and in consequence, the

3 TABLE I. MOST PROBABLE INTERACTIONS OCCURRING IN THE NAI CRYSTAL (NUCLEAR CHART SOFTWARE) Reaction Threshold (MeV) 23 Na(n,p) 23 Ne Na(n,α) 20 F Na(n,γ) 24 Na 0 23 Na(n,2n) 22 Na I (n,γ) 128 I I (n,2n) 126 I I(n,p) 127 Te I(n,α) 124 Sb 0 Product Nuclide Halflife Decay products Beta (MeV) Intensity (%) Gamma (MeV) Intensity (%) 23 Ne , , , 32.9, s F 11.1 s Na , 100, h Na , , y I m 1.665, I d 0.371, 0.862, a 15.5 a, 76.0 a 3.6, 32.1, a, a a a 0.388, 0.491, 0.511, 0.666, 0.753, , 1.78, 0.31, , 2.85, 2.32, 33.1, 4.20, Te 9.35 h Sb 60.2 d 0.210, 0.610, 8.86, , 22.3 For beta decay end point energy 127 I (n,γ) 128 I data has been reported from [27] , 0.645, 0.722, 1.690, , 7.46, 10.81, 47.79, 5.51 Residual Nuclide 23 Na 20 Ne 24 Mg 22 Ne 93.1% ( 128 Xe) 6.9% ( 128 Te) 43% ( 126 Xe) 57% ( 126 Te) 127 I 124 Te produced radio-nuclides start to decay simultaneously with irradiation. The DGMC is developed based on 5 separated steps. It simulates all the processes from activation up to spectroscopy. The software directly simulates the physics involved with particle interactions. The written code is a combination of two coupled MCNP codes joined to the complementary FORTRAN programs. FORTRAN codes are indispensable for time dependent parts (steps 2 and 3) which are not included in the MCNP. The neutron interactions with different energies cause different kinds of induced active nuclei. In the high energy range of neutrons, some new channels which need the threshold energy will be turned on. Therefore, for simplicity, only thermal neutron energy is defined in the simulation to keep all the other threshold interactions off. Tables 2 and 3 show all the open channels up to 11 MeV for neutrons entered the NaI detector cell containing 23 Na and 127 I. Cross sections of 23 Na and 127 I isotopes are shown in Figs. 1 and 2. Considering cross sections and Tables 2 and 3, all the interactions in the thermal energy range lead to only (n,γ) and (n,n) for both 23 Na and 127 I isotopes. According to the cross sections, (n,γ) is the most probable interaction for both isotopes. Decay schemes of 24 Na and 128 I isotopes, and the subsequent gamma-rays and beta emission spectra are shown in Figs. 3 to 6. For example, 24 Na disintegrates only by the emission of beta-particles (100%). The main transition (99.944%) has a maximum energy (end point) of 1393 kev (Table 1) and populates the 4123 kev level of 24 Mg (Fig. 3). This process occurs in femto second and is followed by two gamma-rays in a cascade (1369 kev in each transition and 2754 kev in % of the transitions) which leads through the 1369 kev level to the ground state of 24 Mg. 11 TABLE II. Reaction Products OPEN CHANNELS FOR REACTION: N+ 23 NA UP TO 11 MEV Q-Value (MeV) Threshold (MeV) 24 Na + γ Na + n Ne + p F + α Ne + d Ne + n + p F + n + α Figure 1. Neutron cross-section for 23 Na (n,γ), 23 Na (n,n), and 23 Na (n,total) for low energy region. European JEFF Libraries [28]. Figure 2. Neutron cross-section for 127 I (n,γ), 127 I (n,n), and 127 I (n,total) for low energy region. European JEFF Libraries [28].

4 TABLE III. OPEN CHANNELS FOR THE REACTION: N+ 127 I UP TO MEV Reaction Products Q-Value (MeV) Threshold (MeV) 128 I + γ Sb + α Te + p I + n In + 2α Sb + n + α Sn + p + α Te + d Ag + 3α In + n + 2α Te + n + p Sn + d + α Te + t Sb + 3 He Sn + n + p + α Cd + p + 2α Sn + t + α Sb + 2p I + 2n Rh + 4α Figure 5. Decay scheme of 128 I. Figure 6. β spectrum of 128 I E(ave)=746.6 kev, E(max)= kev [29]. V. SIMULATION STEPS Figure 3. Decay scheme of 24 Na. Figure 4. Beta spectrum of the 24 Na. E(ave)=500 kev, E(max)=1393 kev. A. Step I: MCNP code using FM card; In this step, a program was written to calculate the total neutron capture reactions in the whole volume of 3" 3" NaI detector cell. Tally output is subdivided into 77 cylindrical thin segments with the thickness of 0.1 mm (Fig. 7). A thermal neutron beam impinging upon the NaI detector is defined in source (sdef) card. NaI crystal geometry and its construction materials are also introduced in the geometry and material cards. As shown in Fig. 7, each segment is placed in a special distance from the detector surface and consequently takes different flux of neutrons. Figure 8 shows the simulation results of the thermal neutron flux intensity for each segment. Clearly the front segments take more flux of neutrons than the segments positioned in depth. Since fast neutrons turn on the threshold interactions (Tables 2 and 3), the detector is only irradiated with thermal neutrons. Although in general cases, this code is inclusive to deal with various interactions and different decay schemes of interaction products, for simplicity, the other 12

5 channels of the interactions have been forced to keep turn off only via thermal neutrons irradiation. In this step, total neutron capture reactions per volume unit (production/cm 3 ) are obtained using the F4 tally together with the FM4 card, for a bunch of neutrons (NPS card) emitted simultaneously from the neutron source. Each segment output tally is then multiplied by its volume value to calculate the total neutron capture reactions in that segment. For each segment, output tally (M i ) is normalized per neutron source. Note that tally calculations have to be done for both 23 Na and 127 I isotopes, separately. In the simulation, the cross sections are needed in the thermal energy range, using proper databases with suitable information about neutron cross sections like the ENDF Library. Figs. 9 and 10 show the production amount of the 24 Na and 128 I isotopes per volume unit, respectively in the segments. Obviously the production rate will decrease for the segments positioned in depth. It means segmentation is necessary and the results satisfy us to avoid dealing with the NaI as a large block in our calculation. Due to the difference between the neutron cross-sections of 23Na (n,γ) and 127 I (n,γ) in thermal energy range (Figs. 1, 2), the production rate of the 128I in each segment is more than the 24 Na. Figure Na production in NaI in each segment. Data are calculated for one second and normalized to the source neutrons. Figure 7. Tally segmentation sketch for the NaI detector. The 3'' 3'' NaI detector is subdivided into 77 segments longitudinally (0.1 thickness). Figure 8. MCNP simulation of thermal neutron flux in each detector segment. B. Step II: The first complementary FORTRAN program; The MCNP program described above can only calculate the total neutron capture reaction for a single moment of each cylindrical segment (M i per neutron source, i is the segment number). In real conditions in the PGNAA setup, a large number of neutrons (depending on the source activity) 13 Figure I production in NaI in each segment. Data are calculated for one second and normalized to the source neutrons. are emitted from the neutron source simultaneously. So, the detector is in continuous exposure of the neutrons. This condition longs for a period of time. The question arises, what is the number of active nuclei of both kinds of isotopes ( 24 Na and 128 I) at the subsequent times of irradiation? And then, how many disintegrations take place in each moment? The first question is answered in the current step and the second one will be answered in the following step. The tools needed to answer these questions are not included in the features of present MCNP code, so complementary FORTRAN programs are mandatory. Clearly, the whole production of active nuclei in one second in each segment is M i A, where M i is the total production number of radio-nuclides per source neutrons for the ith segment and A is the source activity. These number of active nuclei produced in a moment do not remain constant in the progress of time and there will be a rate of decay besides the rate of production. After thermal neutron activation of the target nuclei in the NaI detector, they will be active and undergo decay. Based on the physics of active nuclei, they may decay in a short or long period of time. Therefore, there will be a number of active nuclei disintegrations beside the active nuclei productions in each segment volume, during the irradiation time. The residual amount of active nuclei of each isotope at the subsequent times of irradiation is calculated by a Monte Carlo code written in FORTRAN. Fig. 11 shows these processes during 7 seconds (typically) irradiation time of the first segment of the detector. It also illustrates the Monte Carlo algorithm performed in this step.

6 Suppose a uniform activation, where each dot representing an activated nucleus, is uniformly distributed in the first segment of the detector. The total number of active nuclei produced in the first moment is M 1 A. In the progress of time, the 1 st bunch of active nuclei will decay and the number of existed active nuclei produced in the 1st "generation" will decrease according to the speed of disintegration. At 2s, the same number of active nuclei is produced. And also the 2 nd generation starts to disintegrate, but obviously after 7 s from the beginning of the irradiation, the remained active nuclei of the 2 nd generation will be more than the 1 st generation. Fig. 11 is only a didactic diagram to explain the interplay between irradiation and decay. The full concept of the algorithm and the stochastic nature of the decay are not shown exactly in Fig. 11, but are taken into account in simulations correctly. Because of the stochastic nature of the decay, decays of each generation are not exactly the same. So the 1 st and 2 nd generations have to be different in shape. For this reason, simulations for each generation have been done separately. The total remained active nuclei in the first segment of the detector after 7 s neutron irradiation are the integration of the presented nuclei of 7 generations from the 1st to the 7 th generation. As shown in Fig. 11, the total number of remained active nuclei after 7 s is the sum of the contents of the last column. These calculations have been done for all the segments and both kinds of isotopes ( 24 Na and 128 I). C. Step III: The second complementary FORTRAN program; What is needed to account for delayed gamma-rays in the detector has not obtained yet. Total disintegrations of the active nuclei which are the source intensity of the delayed gamma-rays in the period of irradiation are of interest. As shown in Fig. 11, in each second of irradiation time, the segment is activated and undergoes decay. What were calculated in step 2 are the total remained active nuclei after a specific time (typically 7 s) for a particular segment for each kind of isotopes. Now a program is written to get the total decays during the irradiation. Each column in Fig. 11 shows the remained active nuclei at its corresponding time in the considered segment. For a better explanation of the algorithm, each column and row have been labeled, A1 for the first column and B1 for the first row (1 st generation) and so on (Fig. 12). If the irradiation time is 30 minutes, then the full illustration of the 1800 generations for a segment would be an upper triangular matrix. This procedure is calculated twice for each segment due to the existence of two kinds of isotopes ( 23 Na and 127 I) in the structure of the crystal. The subtraction of the (T+1) th column (A t +1) from the T th column (A t ) gives the total disintegrations in the T th second of delayed gamma emission time in a segment. The lower part of Fig. 12 shows the subtracted results in a column format. Assume that the delayed gamma emission is started 1 s after the beginning of the irradiation. The first column (calculated by (A 1 B 1 - A 2 B 1 )) in the lower part of Fig. 12 is the total delayed gamma-rays emitted at this moment. Evidently, if the irradiation time is 7 s, the period of delayed gamma-rays emission is 6 s. The 1 s as the time offset shown in the lower part of Fig. 12 is chosen for this reason. Finally the total number of disintegrations of each isotope in the considered segment during the 7 s of detector irradiation is the integration of the 6 columns in the lower part of Fig. 12. Definitely, depending on the segment position, the total delayed gamma-rays emission during the period of irradiation will be different for each segment. Based on the first assumption, the detector spectrum acquisition is in simultaneous with irradiation. The detector Figure 11. Illustration of the first segment activation for 7 s and active nuclei production in each generation. The bunch of dots in a uniform shape in the left below corner shows the activations occur in that segment, which is in each second formed all over the segment volume. Total number of dots in a bunch is M 1 A. Total integration is the sum of dots put in the right column of the diagram. It is the total active nuclei remained after 7 s of radiation. 14 Figure 12. Illustration of the algorithm performed to obtain the total number of disintegrations taking place in a segment, during 7 s irradiation.

7 records those delayed gamma-rays originated from all of its segments which have interaction with the origin segment or the others. However, some of the produced delayed gammarays escape from the detector cell without any interaction. A set of Monte Carlo programs (described in the following steps) were written to simulate all the photon interactions and obtain the segments-pulse height tally. According to what is shown in Fig. 12, the total production of delayed gamma-rays in a segment of the detector increases by time. Depending on the irradiation time and the isotope half-life, saturation may take place or may not. In the saturation time, the production rate is equal to the disintegration rate. So, during the period of irradiation after saturation, a constant number (M i A) in each second is added to the total number of produced delayed gamma-rays in the i th segment. D. Step IV: Segments weight and segments-pulse height tally calculations; So far, the whole volume of detector was geometrically subdivided and activation tallies of each isotope were obtained by simulation in step 1. The total disintegrations of each isotope (the source intensity of the delayed gammarays) were also obtained for each segment in the period of irradiation. For a better explanation, two symbols (α and β) are reserved for the total number of disintegrations of each kind of 24 Na and 128 I isotopes, respectively. Since there are 77 segments, 77 α i are chosen, where i varies from 1 to 77 corresponding to the segment number. α i is the total decays of 24 Na taking place during the irradiation time in the ith segment. In other words, α i is the 24 Na-segment weight of the ith segment, and β i is the 128 I-segment weight of the ith segment, either. Each segment has its own weight in the formation of the total spectrum. In the current step, a set of MCNP programs (154; 77 for each isotope) were written to obtain the output pulse height tally distributions of the induced radio-nuclides in the 77 segments. The source geometry definitions were according to the segment geometry definitions in step 1. Also the particle type (PAR) and energy of the source definitions corresponded to the decay schemes of 24 Na and 128 I isotopes (Figs. 3, 5). The resulting spectra were obtained with F8 tally card, accompanying a proper FT card [4, 26, 30, 31]. Using proper GEB parameters (obtained in our previous work [4]) gives a broadened photoelectric peak corresponding to the real condition. The duration of each run was selected 500 min, so the tally relative error of each energy bin maintained below 1%, in accordance with an emission of 10 7 photons (NPS). The output tally intensity is normalized to the source photons. Therefore, as will be discussed in the following step, the spectra have to be multiplied by the weights (α i and β i ) obtained earlier. E. Step V: spectra accumulation; In step 4, 154 detector spectra were simulated for 77 segments with 2 kinds of sources (two decay schemes). The total delayed gamma spectrum is obtained by the integration of the whole spectra considering their weights (α i and β i ). Fig. 13 explains the procedure performed for obtaining the integrated spectrum resulting only from 24 Na isotope disintegrations in all segments. A similar procedure with the weights of β i was also done to get the spectrum of 128 I isotope disintegrations. The final spectrum is the summation of the whole segments spectra. VI. RESULTS AND DISCUSSION The simulations were performed for a 3" 3" NaI detector with 77 cylindrical segments. The flux intensity of the thermal neutron beam was also chosen 10 3 n/cm 3. Another assumption is that NaI detector was irradiated 30 min continuously. In these conditions, NaI detector starts to record the delayed gamma-rays spectrum of its active nuclei. The spectrum acquisition period is also equal to the irradiation period (30 min) and starts from the beginning of irradiation. Figs. 14 and 15 show the delayed gamma spectra due to the disintegrations of 24 Na and 128 I radio-nuclides, respectively. The final spectrum which is the summation of these two spectra is also shown in Fig. 16. Fig. 16 shows the contribution of delayed gamma-rays of detector materials activation to the total obtained spectrum. The delayed gamma spectrum energy range, due to the detector activation by thermal neutrons, is extended to only 3 MeV. Hence, the contribution of the activated detector s delayed gamma-rays to the total spectrum is defined by this method of simulation. The calculated spectrum is usually considered to be part of backgrounds, which is not discriminated experimentally. In the present study, a new approach for simulating and determining the effect of one of the background sources in the PGNAA method was introduced. The problem with NaI detector in the PGNAA setups is that some fast or thermal neutrons may enter the detector and lead to the activation of NaI crystal through the measurement. Shielding the detector is not usually 100% successful and the concern about detector activation and disruption in the obtained spectrum exists forever. Beside the attempts to prevent entering the neutrons into the detector, knowledge about the full physical processes occurring during the irradiation and the consequent results are indeed valuable and help the designer of the setup to choose the shield and moderator materials better. Since for thermal neutrons, there are not threshold interactions, thermal neutron irradiation was studied first. In near future, we are going to determine the effects of fast neutrons and calculate other kinds of background. Also we will study the activation of other types of detectors and compare the neutron activation setups. Figure 13. The procedure is performed to obtain the integrated spectrum of 24 Na disintegrations in all segments. Each segment spectrum is multiplied by its weight (weights were calculated in step 3). 15

8 Figure 14. Simulated delayed gamma spectrum of 24 Na decay in a 3" 3" NaI detector cell. Irradiation longs for 30 min. Figure 15. Simulated delayed gamma spectrum of 128 I decay in a 3" 3" NaI detector cell. Irradiation longs for 30 min. Figure 16. Simulation of the total delayed gamma spectrum of 3" 3" NaI detector activation. The thermal neutron beam is "on" for 30 min. In simulation process, it is considered the spectroscopy and detector irradiation are in simultaneous ("online spectroscopy"). REFERENCES [1] Z. Révay, R. M. Lindstrom, E. A. Mackey, and T. Belgya, Neutron- Induced Prompt Gamma Activation Analysis (PGAA), in Handbook of Nuclear Chemistry, A. Vértes, et al., Editors. Springer US., 2011, p [2] F. A. Dilmanian, D. A. Weber, S. Yasumura, Y. Kamen, L. Lidofsky, S. B. Heymsfield, R. N. Pierson, Jr., J. Wang, J. J. Kehayias, and K. J. Ellis, Performance of the Delayed- and Prompt- Gamma Neutron Activation Systems at Brookhaven National 16 Laboratory, in In Vivo Body Composition Studies, S. Yasumura, et al., Editors. Springer US, 1990, p [3] Y. J. Park, B. C. Song, H. J. Im, and J. Y. Kim, Performance characteristics of a prompt gamma-ray activation analysis (PGAA) system equipped with a new compact D D neutron generator, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, vol. 606, pp , May [4] H. Miri Hakimabad, H. Panjeh, and A. Vejdani-Noghreiyan, Evaluation the nonlinear response function of a 3" 3" NaI scintillation detector for PGNAA applications, Applied Radiation and Isotopes, vol. 65, pp , March [5] S. H. Cohn, In vivo neutron activation analysis; a new technique in nutritional research, The Journal of Nutritional Biochemistry, vol. 3, pp , Aug [6] J. J. Toth, R. Wittman, R. E. Schenter, and J. A. Cooper, Candidate reactions for mercury detection induced by neutron and alpha particles, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, vol. 572, pp , Jan [7] J. H. Moon, S. H. Kim, Y. S. Chung, J. M. Lim, G. H. Ahn, and M. S. Koh, U determination in environmental samples by delayed neutron activation analysis in Korea, Journal of Radioanalytical and Nuclear Chemistry, vol. 282, pp , July [8] J. Kettler, E. Mauerhofer, and M. Steinbusch, Detection of unexploded ordnance by PGNAA based borehole-logging, Journal of Radioanalytical and Nuclear Chemistry, vol. 295, pp , Sep [9] H. Miri-Hakimabad, H. Panjeh, and A. Vejdani-Noghreiyan, Experimental optimization of a landmine detection facility using PGNAA method, Nuclear Science and Techniques, vol. 19, pp , March [10] M. H. Hadizadeh Yazdi, A. A. Mowlavi, M. N. Thompson, and H. M. Hakimabad, Proper shielding for NaI(Tl) detectors in combined neutron-γ fields using MCNP, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, vol. 522, pp , Jan [11] R. P. Gardner, E. Sayyed, Y. Zheng, S. Hayden, and C. W. Mayo, NaI detector neutron activation spectra for PGNAA applications, Applied Radiation and Isotopes, vol. 53, pp , Nov [12] J. W. Durkee Jr, M. R. James, G. W. McKinney, H. R. Trellue, L. S. Waters, and W. B. Wilson, Delayed-gamma signature calculation for neutron-induced fission and activation using MCNPX, Part I: Theory, Progress in Nuclear Energy, vol. 51, pp , Nov [13] J. W. Durkee Jr, M. R. James, G. W. McKinney, H. R. Trellue, L. S. Waters, and W. B. Wilson, Delayed-gamma signature calculation for neutron-induced fission and activation using MCNPX. Part II: Simulations, Progress in Nuclear Energy, vol. 51, pp , Nov [14] R. Proctor, S. Yusuf, J. Miller, and C. Scott, Detectors for on-line prompt gamma neutron activation analysis, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, vol. 422, pp , Feb [15] K. D. Ianakiev, B. S. Alexandrov, P. B. Littlewood, and M. C. Browne, Temperature behavior of NaI(Tl) scintillation detectors, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, vol. 607, pp , Feb [16] R. P. Gardner, and C. W. Mayo, NaI detector nonlinearity for PGNAA applications, Applied Radiation and Isotopes, vol. 51, pp , Jan [17] V. E. Darwin, and M. S. Julian, Well Logging for Earth Scientists. Elsevier, [18] J. H. Marshall, and J. F. Zumberge, On-line measurments of bulk coal using prompt gamma neutron activation analysis, Nuclear Geophysics, vol. 3, pp. 445, March [19] H. Miri-Hakimabad, R. Izadi-Najafabadi, A. Vejdani-Noghreiyan, and H. Panjeh, Improving the safety of a body composition analyser based on the PGNAA method. Journal of Radiological Protection, vol. 27, pp. 457, Nov [20] R. Khelifi, A. Amokrane, and P. Bode, Detection limits of pollutants in water for PGNAA using Am Be source, Nuclear

9 Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, vol. 262, pp , June [21] R. Khelifi, Z. Idiri, L. Omari, and M. Seghir, Prompt gamma neutron activation analysis of bulk concrete samples with an Am Be neutron source, Applied Radiation and Isotopes, vol. 51, pp. 9-13, Jan [22] J. L. Pinault, Use of new spectral analysis methods in gamma spectra deconvolution, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, vol. 305, pp , Jan [23] J. L. Pinault, and J. Solis, The optimization of gamma spectra processing in prompt gamma neutron activation analysis (PGNAA), Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, vol. 267, pp , Feb [24] D. Sudac, and V. Valkovic, Irradiation of 4 4 NaI(Tl) detector by the 14.0 MeV neutrons, Applied Radiation and Isotopes, vol. 68, pp , April [25] L. C. Thompson, Neutron effects in a 2 2 NaI(Tl) scintillation spectrometer, Nuclear Instruments and Methods, vol. 25, pp , Feb [26] J. F. E. Briesmeister, MCNP-A General Monte Carlo N-Particle Transport Code. Version 4C, LA-13709M, [27] N. Benczer, B. Farrelly, L. Koerts, and C. S. Wu, Investigations of I^{128}, Physical Review, vol. 101, pp , Feb [28] JEFF, European JEFF Libraries [29] R. B. Firestone, and L. P. Ekström. 2013, Available from: [30] D. Schulze, and C. Segebade, ACTIVATION ANALYSIS Photon Activation, in Encyclopedia of Analytical Science (Second Edition), W. Editors-in-Chief: Paul, T. Alan, and P. Colin, Editors. Elsevier: Oxford, 2005, pp [31] C. M. Salgado, L. E. B. Brandão, R. Schirru, C. M. N. A. Pereira, and C. C. Conti, Validation of a NaI(Tl) detector's model developed with MCNP-X code, Progress in Nuclear Energy, vol. 59, pp , May