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Since 60Co gamma peaks in the gamma energy spectrum were not revealed in the electronic noise by the Multi-Pixel Photon Counter (MPPC), an experimental setup was proposed to reduce the noise, and to unveil the peaks and the gamma spectrum. For this purpose, constant fraction timing discrimination method was utilized. The timing signals were coincidentally gated in the multichannel analyzer with the signals directly from the MPPC. Obtained result spectrum showed that the asserted setup was quite successful in reducing the noise components, and in obtaining the neat gamma peaks of the radioisotope through the MPPC.

In this study, energy resolution dependent on K-Parameters (Trapezoidal filter system optimizing the count rate) of EXOGAM2 detector with newly developed NUMEXO2 digital electronics using 152Eu radioactive source measured and analysed. This experimental study was carried out at the GANIL (Grand Accelerature National d'Ions Lourd) nuclear research center in France. The measurement and analysis results for different K parameters using a 152Eu radioactive source with low and high gamma energy levels were performed for two EXOGAM2 detectors. Without applying any K parameter, the resolution values expressed by FWHM for the 1.EXOGAM2 detector are measured as 2.360% FWHM for (344.2785 keV), 2.820% FWHM for (1408.006 keV), and 2.451% FWHM for (344.2785 keV), 2.975%FWHM for (1408.006 keV) for the 2.EXOGAM2 detector, respectively. For different K parameters (K=2µs, K=5µs ve K=10µs) detail results are shown below., Bu çalışmada, yeni geliştirilen dijital elektronik NUMEXO2 dijital elektronik ile EXOGAM2(EXOtic GAMma array spectrometer) dedektörünün K-Parametrelerine (sayım oranını optimize eden Trapezoidal filtre sistemi) bağlı enerji çözünürlüğü için 152Eu radyoaktif kaynak kullanılarak ölçüm ve analiz yapılmıştır. Yapılan bu deneysel çalışma Fransa’da bulunan GANIL (Grand Accelerature National d’Ions Lourd) nükleer araştırma merkezinde gerçekleştirilmiştir. Düşük ve yüksek gama enerji seviyesine sahip 152Eu radyoaktif kaynak kullanılarak farklı K-parametreleri için gerçekleştirilen ölçüm ve analiz sonuçları iki EXOGAM2 dedektörü için gerçekleştirilmiştir. Herhangi bir K-parametresi uygulamadan 1.EXOGAM2 dedektörü için FWHM ile ifade edilen çözünürlük değerleri sırası ile (344. 2785 keV) için 2.360% FWHM, (1408.006 keV) için 2.820% FWHM olarak ölçülürken 2.EXOGAM2 dedektörü için ise sırasıyla 2.451% FWHM ve 2.975% FWHM olarak ölçülmüştür. Farklı K-parametreleri (K=2µs, K=5µs ve K=10µs) için ise sonuçlar detayları ile aşağıda gösterilmiştir.

A code is developed to correct for the coincidence summing effect in detecting a voluminous gamma source, and this code is applied to a152Eu standard source as a test case. The source is 1000 mL of liquid in a cylindrical shape. To calculate the coincidence summing effect, the cylindrical source is considered as 10(radial) × 8(height) sectional sources. For each sectional source, the peak efficiency and total efficiency are obtained by Monte Carlo simulation at each energy for 10 energies between 50 keV and 2000 keV. The efficiencies of each sector are then expressed as polynomials of gamma energy. To calculate the correction coefficients for the coincidence summing effect, the KORSUM code is used after modification. The magnitudes of correction are 4%–17% for the standard 152Eu source measured in this study. The relative deviation of 4.7% before the coincidence correction is reduced to 0.8% after the correction is applied to the efficiency based on the measured gamma line. Hence, this study has shown that a new method has been developed that is applicable for correcting the coincidence effect in a voluminous source, and the method is applied to the measured data of a standard 152Eu cylinder source.

Studies performed decades ago in the canine and nonhuman primate established the dose response relationships for the hematopoietic acute radiation syndrome in response to mixed neutron/gamma, x-radiation, and Co gamma radiation. There were no published studies that determined the dose response relationships for the gastrointestinal acute radiation syndrome in response to either noted radiation quality. This analysis of a retrospective, unpublished study provided the dose response relationships in a canine model for the acute gastrointestinal syndrome relative to the acute hematopoietic syndrome due to mixed neutron/gamma radiation. Canines were exposed to total-body, steady state, bilateral, 0.40 Gy min, mixed neutron/gamma (5.4:1) radiation from a TRIGA reactor. The average neutron/gamma energy (MeV) was 0.85/0.9, and exposure was reported as midline tissue dose. Medical management was not administered. The mixed neutron/gamma exposure resulted in an estimated LD50/6 of 2.83 Gy [2.76, 2.94] and LD50/30 of 2.16 Gy [2.01, 2.24] for the GI- and H-ARS respectively. The mean survival times for decedents after mixed neutron/gamma exposure approximate to the LD50/6 were 8.5 d, 10.5 d, and 4 d for 2.75 Gy, 2.80 Gy, 3.00, and 3.12 Gy exposures, respectively. The mean survival times for decedents for mixed neutron/gamma exposure approximate to the LD50/30 were 21.3 d and 15.6 d for 2.00 Gy and 2.25 Gy, respectively. Furthermore, the dose response relationships for the acute hematopoietic syndrome due to mixed neutron/gamma exposure (0.85/0.9 MeV; 5.4:1) resulted in an estimated relative biological effectiveness of 1.2 as compared with reference Co gamma radiation.

Using a portable microcomputer multi-channel gamma spectrometer to directly measure the radioactive content in the soil, the on-site gamma energy spectrum measurement was carried out on the soil of 7 parks in a certain city. The survey results show that the specific activity of 40k in the urban park soil ranges from 470.88 bq/kg to 640.2 bq/kg. The specific activity of 238u ranges from 26.77 bq/kg to 33.31 bq/kg. The specific activity of 232th ranges from 42.14 bq/kg to 54.72 bq/kg. The content of the three radionuclides in the soil in the park is lower than the national average, and far lower than the radioactivity of ordinary building materials.

Transparent heavy metal oxide glasses are becoming increasingly important because of their effectiveness in reducing unwanted occupational and public radiation exposure in the medical settings; and for the clear view it offers when properly used during the construction of diagnostic radiology facilities, mobile shields and eye wears. WinXCom computer program was used to determine the mass attenuation coefficients of the studied glass systems; while other shielding parameters were determined by using appropriate equations. All these shielding parameters were determined at 20 keV, 30 keV, 40 keV and 60 keV; these are the average energies of photons commonly used in diagnostic radiology. The outcome of this work shows that increment in Bi2O3 led to better MAC and HVL at 20 and 30 keV while increment in TeO2 led to better MAC and HVL at 40 and 60 keV B1, B3 and B4 showed different improved shielding properties at 20, 40 and 60 keV zBi2O3–30B2O3-(70-z) TeO2 glass system showed better shielding qualities when compared with commonly used materials in diagnostic radiology.

The self-attenuation correction coefficients (Csa) for 137Cs counting in twelve soil samples were determined by MCNP simulation of a high-purity germanium detector. The correction coefficients were defined as the simulation and experimental efficiency ratio for gamma radiation with 661.66 keV. The experimental efficiency of the detector for 661.66 keV of gamma energy in soil samples was obtained using the IAEA reference materials. The exact gamma radiation efficiency for soil samples was calculated on the correct height and soil density using the MCNPX simulation code. The result of the measurements, including the self-attenuation correction coefficient, was compared with the given value and indicates an increase in accuracy.

IIn this study, pure silicone rubber and iron ore concentrate added silicone rubber materials were performed against Cs-137 gamma source. Cs-137 is accepted medium energy level gamma source in nuclear technology. Up to 67 wt. % iron ore concentrate added into silicone rubber materials were used in the experiments. Linear attenuation graphs were carried out for the studied samples against Cs-137 gamma source which has unique peak at 0.662 MeV energy. Increasing iron ore ratio in the materials caused the higher radiation shielding performance for Cs-137 gamma source. In addition, increasing iron ore ratio decrease the total volume air bubbles in the samples. Decreasing air bubbles in the samples has contributed to gamma ray shielding. 0.5 mmPbE and 1 mmPbE Lead equivalent (PbE) values were determined for the samples. 0.5mmPbE standard has been provided by using 2.83 mm thick of 67 wt. % iron ore concentrate imbedded silicone rubber at 0.662 MeV gamma energy. In conclusion, iron ore concentrate imbedded silicone rubber materials are among the promising economic radiation shielding materials which could be alternative to lead.

Kuosheng BWR-6 plant, owned by Taiwan Power Company, is currently operating at core thermal power of 2 971 MWt. The purpose of this study is to develop a method used to calculate the radiation dose at the reactor auxiliary building and at various rooms accommodating safety-related equipment during the design basis Loss of Coolant Accident (LOCA). The method integrates various aspects, including core inventory calculation using ORIGEN2 code, radioactive material transport and removal calculation using RADTRAD code, activity release calculation based on Technical Specification leakage limit and RADTRAD output, atmospheric dispersion calculation using ARCON96 code, and dose calculation using gamma energy and spectrum based finite cloud model. The assumptions used are based on the Regulatory Guide 1.89 revision 1. The calculation shows that the original Environmental Qualification(EQ) radiation parameters defined in the FASR, for the safety-related electrical and instrumental control equipment in the reactor auxiliary building, can be relaxed significantly and the radiation environment during accident could be reclassified, from “harsh environment” to “mild environment”.

Radioisotope indium-113m ( 113m In) with half-life, T 1/2 = 1.7 hours and gamma energy, Eg = 391 keV is suitable and meets the criteria as radiotracer in industry. 113m In radioisotope was obtained from tin-113 decay ( 113 Sn, T 1 / 2 = 115 days) of 112 Sn (n,g) 113 Sn neutron activation in nuclear reactor. The process of separation of radioisotope 113m In using column chromatography method based on silica gel matrix using 0.05 M HCI solution. Radionuclide and radiochemical purity tests were performed using the gamma-spectrometry method and paper chromatography. The final product specification in the form of 113m InCl 3 is clear solution, pH 2, obtained yield of 81.83%, radionuclide purity of 90.22%, radiochemical purity of 91.61 ± 0,29% and stable for 3 days at room temperature.

To suppress spectral drift in lengthy gamma spectrum measurements, a convenient and practical spectrum stabilization method with software-only adjustment is studied. The gamma energy spectra are recorded in consecutive small time intervals. The drifts in different time intervals relative to the first reference spectrum are estimated with a full spectrum nonlinear optimization technique, and the drifted spectra are corrected and accumulated to form a drift-free spectrum. This method needs neither hardware adjustment such as high voltage or electronics gain, nor peak location. Presented experimental results show that the stabilization accuracy can be significantly improved using this approach.

Silicon photomultipliers (SiPMs) are gaining popularity over conventional photomultiplier tubes (PMTs) in various radiation detection applications. For this application, it is necessary to optimize the scintillator material and its geometry. In order to develop a compact, low power gamma spectrometry system based on a scintillator coupled SiPM detector, energy deposition in Gd3Ga3Al2O12(Ce) (abbreviated as GGAG(Ce)), NaI(Tl) and CsI(Tl) scintillators of different dimensionswas calculated using Monte Carlo simulation code FLUKA for different gamma energies from 60 keV to 2614 keV. It is observed that energy deposition is highest in GGAG(Ce) followed byCsI(Tl) and NaI(Tl) respectively. By increasing thepath length of the scintillator, energy deposition is seen to increase for energies 662 keV, 1250 keV and 2614 keV. However, for 60 keV, the increase in longitudinal dimension above 3 mm does not affect the energy deposition. Based on the simulation data,GGAG(Ce) scintillator with 3×3×24 mm3 geometry was chosen for making a scintillator coupledSiPM detector.

The National Physical Laboratory Vinten 671 chamber was selected as a proving ground for a new radionuclide source model in the EGSnrc software. The computational Vinten model is validated against measurements of radionuclide artifacts whose activities were determined by absolute methods. The response of the Vinten chamber is first calculated as a function of gamma energy, but more strikingly, an explicit simulation of radionuclide decay was implemented and now permits the direct determination of a calibration factor, including additional effects due to all decay paths of the radionuclide. The Monte Carlo and experimental calibration factors are found to agree at the percent level, in absolute terms.

Gamma-ray shielding effectiveness of different types of natural marble tiles commonly used for cladding the exterior walls of residential and non-residential buildings in Riyadh, Saudi Arabia has been investigated in the energy range 59.5–1332.5 keV. To this end, linear attenuation coefficients (μℓ) have been obtained by applying narrow-beam technique. Mass attenuation coefficient (μm), transmission factor (TF) and half-value layer (HVL) have been employed to study the shielding performance of marbles and also compared with those of calcite. The experimental findings revealed that HVL parameter correctly determined the shielding effectiveness of the dissimilar types of marble. Compared with other marbles, Carrara marble (Italy) showed preferable shielding effectiveness. Moreover, a formula, μm = aE−b, was proposed to evaluate and compare the shielding properties of the marble tiles over a broad energy range. In this formula, a and b were determined empirically. In conclusion, natural marble proves to be more reliable in reducing gamma radiation when used for external building cladding than ordinary concrete, and it could be a good alternative to lead shield against high gamma energy because it is 24% heavier than lead. Keywords: Gamma-ray shielding effectiveness, Attenuation coefficients, HVL, Transmission factor, Natural marble

Gamma imager capable of monitoring, detection and localization of radioactive sources is of critical importance for nuclear safety and homeland security with the increasing use of radioactive sources in applications of nuclear technology. A 4π view gamma imager based on gamma ray occlusion technique is under development in our lab. The key component of the gamma imager is a 3D position-sensitive detector, which enables 4π field of view of the imager. In this work, we propose a 3D positionsensitive detector design for the 4π view gamma imager based on dual-ended readout (DER) technique. The detector was composed of a GAGG block coupled to an SiPM array at both crystal ends. DOI calibration was conducted with a collimated fan beam 18F source and DOI resolution was evaluated. Energy calibration was performed for each 3D position region of the detector by incorporating 3D position information to correct for the energy response inconsistency at different 3D position regions inside the detector. Various radioisotopes including 99mTc, 131I, 137Cs and 22Na were tested. An exponential factor was applied to compensate for energy response non-linearity due to SiPM saturation effect at higher gamma energy. The measured average DOI resolution was 1.96 ± 0.21 mm for all the four DER detector modules, indicating excellent 3D positioning accuracy of the detector design. With the proposed energy calibration approach, characteristic photopeaks of each radioisotope can be clearly identified at the correct energy position on the measured energy spectra. The achieved energy resolution was 9.76% at 662 keV, demonstrating good energy performance of the detector design, which could benefit radioisotope identification capability of the gamma imager. To conclude, the proposed 3D position-sensitive detector design based on DER technique features good 3D positioning and energy performance and is feasible for the 4π view gamma imager.

Accurate determination of the proppant placement and propped fracture height are important for evaluating and optimizing stimulation strategies. A technology using non-radioactive proppant and a pulsed neutron gamma energy spectra logging tool to determine the placement and height of propped fractures is proposed. Gd2O3 was incorporated into ceramic proppant and a Monte Carlo method was utilized to build the logging tools and formation models. Characteristic responses of the recorded information of different logging tools to fracture widths, proppant concentrations and influencing factors were studied. The results show that Gd capture gamma rays can be used to evaluate propped fractures and it has higher sensitivity to the change of fracture width and traceable proppant content compared with the exiting non-radioactive proppant evaluation techniques and only an after-fracture measurement is needed for the new method; The changes in gas saturation and borehole size have a great impact on determining propped fractures when compensated neutron and pulsed neutron capture tool are used. A field example is presented to validate the application of the new technique.

The characteristic parameters of a material relation to photon interactions such as: mass attenuation coefficient, effective atomic number, effective electron density are required to provide essential data in diverse works such as nuclear diagnostic and cancer radiotherapy, industrial irradiation, radiation dosimetry, radiation protection and shielding, analyzing of the concentration of elements and radioactive isotopes. In this paper, the theoretical models such as non-relativistic form factor (NRFF), relativistic form factor (RFF), and modified form factor (MFF) were used to calculate the ratio Rayleigh-Compton for elements with at gamma energy 59.5 keV. The results showed that there was a discrepancy between the theoretical modes at a high atomic number. The mean value of the Rayleigh-Compton ratio depends on the atomic number, which shows the quadratic function of the correlation coefficient R2 = 0.996 as well. Besides, the experimental system was set-up and measured for some targets such as aluminum, copper, and lead at a scattering angle 150o using 241Am source by Si(Li) detector to confirm the theoretical values. The preliminary result showed that there was a good agreement with experimental and theoretical results is lower than 20%. Further investigation will be measured by the samples for more detailed evaluation.

Gamma energy spectrum is a way to confirm the type and content of radioactive nuclides in substances. In order to obtain effective data of gamma spectrum, it is necessary to analyze the data of gamma spectrum obtained by the detector. Nevertheless, due to the influence of statistical fluctuation and gamma generated by the environment itself during the measurement of gamma spectrum, the obtained gamma spectrum is not accurate enough as a result of needing spectral analysis of gamma spectrum, and noise reduction can make the spectral data smoother. Therefore, this paper studies the noise reduction algorithm for gamma spectrum analysis, analyzes the design and application of different algorithms in the process of noise reduction, and summarizes the advantages and disadvantages of this algorithm and others.

The aim of this study is to develop the Effective Dose Calculation Program (EDCP) for the usage of Naturally Occurring Radioactive Material (NORM) added consumer products. The EDCP was developed based on a database of effective dose conversion coefficient and the Matrix Laboratory (MATLAB) program to incorporate a Graphic User Interface (GUI) for ease of use. To validate EDCP, the effective dose calculated with EDCP by manually determining the source region by using the GUI and that by using the reference mathematical algorithm were compared for pillow, waist supporter, eye-patch and sleeping mattress. The results show that the annual effective dose calculated with EDCP was almost identical to that calculated using the reference mathematical algorithm in most of the assessment cases. With the assumption of the gamma energy of 1 MeV and activity of 1 MBq, the annual effective doses of pillow, waist supporter, sleeping mattress, and eye-patch determined using the reference algorithm were 3.444 mSv year−1, 2.770 mSv year−1, 4.629 mSv year−1, and 3.567 mSv year−1, respectively, while those calculated using EDCP were 3.561 mSv year−1, 2.630 mSv year−1, 4.740 mSv year−1, and 3.780 mSv year−1, respectively. The differences in the annual effective doses were less than 5%, despite the different calculation methods employed. The EDCP can therefore be effectively used for radiation protection management in the context of the usage of NORM-added consumer products. Additionally, EDCP can be used by members of the public through the GUI for various studies in the field of radiation protection, thus facilitating easy access to the program.

The capabilities of indirect radiobioassay by urine and fecal sample analysis were compared with the direct radiobioassay methods of whole body counting and lung counting for the most common radionuclides and inhalation exposure scenarios encountered by Hanford workers. Radionuclides addressed by in vivo measurement included 137Cs, 60Co, 154Eu, and 241Am as an indicator for plutonium mixtures. The same radionuclides were addressed using gamma energy analysis of urine samples, augmented by radiochemistry and alpha spectrometry methods for plutonium in urine and fecal samples. It was concluded that in vivo whole body counting and lung counting capability should be maintained at the Hanford Site for the foreseeable future, however, urine and fecal sample analysis could provide adequate, though degraded, monitoring capability for workers as a short-term alternative, should in vivo capability be lost due to planned or unplanned circumstances.

In 2011, a network of five thallium-doped sodium iodide (Nal(Tl)) detectors was installed on Aragats Space Environmental Center (ASEC) and was included into ASEC detectors system. Along with monitoring of different species of secondary cosmic rays, ASEC detectors register several thunderstorm ground enhancements (TGEs). NaI(Tl) detector integration in the ASEC detector system is of great importance for the study of thunderstorm phenomena for the reason that NaI(Tl) detectors have a higher efficiency of gamma rays detection compared with plastic ones. In this article, the design and characteristics of NaI(Tl) detectors are described. Simulations of detector response are performed. Comparison of simulation results with experimental data showed good agreement between simulations and experimentally observed distributions for analog-to-digital converter (ADC) channels (codes) of NaI(Tl) detectors at two depths of the atmosphere, thus, indicating the correctness of the detector׳s response determination. A procedure for reconstruction of gamma energy spectrum was developed and approximation of the energy spectrum of recorded TGE event was carried out by a power function under the assumption that the recorded fluxes consist mainly of gamma quanta.

Experimental and simulated energy spectra of 137Cs, 60Co, and 241Am standard gamma sources were acquired. A spectrometer consisting of a NaI(Tl) scintillation detector was used to obtain experimental spectra. The theoretical energy spectra were acquired by means of FLUKA Monte Carlo (MC) simulations, and the comparison between the experimental and simulated spectra was performed with the ROOT program. Spectral results such as energy resolution, counting efficiency, peak to total, peak to Compton and peak to valley values were compared. Good agreement between the experimental and simulated spectra was performed with the ROOT program. Good agreement between the experimental and simulated results was obtained.

In the present work, several commonly used aluminum alloys were investigated for their protective properties against gamma radiation. The gamma transmission technique was used to study the gamma attenuation behavior of the alloys. Cs-137 (0.662 MeV) and Co-60 (1.25 MeV) gamma radioisotope sources, which have relatively medium and high gamma energy levels, were used as gamma sources. The linear and mass attenuation coecients of the aluminum alloys were measured. The mass attenuation coecients of the samples were compared with the theoretical values which were calculated using XCOM computer code. The dierence between the experimental and theoretical values was below 5%. In addition half-value layer (HVL) values for the studied aluminum alloys were calculated using the linear attenuation coecients. The attenuation coecients of the dierent aluminum alloys were compared. The biggest HVL was observed for 1050 alloy, for both gamma isotope sources, which means the smallest gamma attenuation capability among the studied alloys. It is concluded that the alloys were applicable for the gamma radiation shielding applications.

The fibrous self-compacting concrete is a high performance concrete with uniformly distributed iron fibers, developed and characterized by Maragon (2006 and 2009). Transmission measurements, with 137Cs and 60Co sources were performed for the attenuation coefficients determination for both ordinary and fibrous self-compacting concretes. The results were compared to each other and to the values found in the literature for ordinary concrete. The mass attenuation coefficient for the fibrous self-compacting concrete showed to be higher than those for ordinary concrete of about 5%, depending on the gamma energy. However, it should be noted that the density of fibrous self-compacting concrete is higher than ordinary concrete, 2.4 g/cm3 and 1.9 g/cm3 respectively, increasing still further the difference in mass attenuation coefficient. In addition to that, by using Monte Carlo simulations, with MCNP5 Monte Carlo computer code, the data was extended to the 50-3000 keV gamma energy range.

Compton-suppression spectrometer is well suited to the analysis of low levels of radioactive nuclides. Monte Carlo simulations can be a powerful tool in calibrating these types of detector systems, provided enough physical information on the system is known. A simplified Compton-suppression spectrometer model using the Geant–4 simulation toolkit was discussed. The spectrometer model was tested to evaluate photo peak efficiency of detecting point and disk sources. The efficiency calibration was calculated for incident gamma energy from 200 to 3000 keV in both the suppressed and unsuppressed mode of operation. The applicability of the efficiency transfer method in various measurement geometries was tested successfully. It can save time and avoid tedious experimental calibration for different samples geometries. Key words: Compton-suppression spectrometer, photopeak efficiency, Geant–4 Monte Carlo simulation Compton-suppression spectrometer, photopeak efficiency, Geant–4 Monte Carlo simulation.

Level densities are required at excitation energies where discrete level information is not available or incomplete to get a reliable theoretical analysis of cross sections, spectra, and angular distributions. The total reaction cross sections of some lanthanides (141Pr, 142Nd, 144Sm, 153Eu, 160Gd, 159Tb, 165Ho, 175Lu) were calculated using TALYS 1.2 code for gamma-induced reactions through the five level density models in the incident gamma energy range from 5 to 30 MeV. All calculations from the present study were compared with each other and with the experimental data obtained from EXFOR library. The total photo-neutron reaction cross-section values obtained from the model calculations and experimental measurements taken from EXFOR confirm the presence of systematical disagreements reported in the literature, except 165Ho case where one can see a satisfactory agreement.

The properties of the ground, beta and negative parity state bands of 146Nd–156Nd isotopes have been studied. The ratio of gamma energy over spin has been calculated and drawn as a function of spin (I) to show the ground-state evolution for each isotope. The ratio between the excitation energies of successive (I+2) and (I) states as a function of spin (I) has been calculated and drawn to determine the property of the ground-state bands of 146Nd–156Nd isotopes. The odd-even staggering has been drawn to determine the energy difference of the states of the ground-state bands and the negative-parity bands. The energy levels for the ground-state band of 146Nd–156Nd have been calculated using the Bohr-Mottelson Model, the Interacting Boson Model, the Interacting Vector Boson Model and the Doma–El-Gendy Relation. The energy levels of the beta band for the measured experimentally levels of 148Nd, 150Nd and 152Nd have been calculated using the Interacting Boson Model. The energy levels of the negative parity band of the 146Nd, 148Nd, 150Nd, 152Nd and 156Nd isotopes have also been calculated using the Bohr-Mottelson Model and the Interacting Vector Boson Model. The calculated energy levels compared with the experimental data indicate the quality of the fit presented in this paper.

The particular physics of Laser Produced Plasmas (LPP) needs some diagnostic requirements. Specifically, the X-ray monitoring of the plasma is known to be difficult since typically X-ray emissions are concentrated in bursts from a few tens of ps to few ns, based on the power and pulse time width of the laser. Therefore, the energy measurement of the radiation coming from a single experimental run is basically unfeasible using conventional techniques. Additional particles can be produced from LPP experiments, especially high energy gamma photons and electrons. As a case study in recent experiments, carried out on VEGA-2 laser facility (CLPU, Salamanca, Spain), the aim was to produce neutrons through photonuclear reactions on different types of solid targets. We have used the Timepix3 chip, in a "side-on" configuration, in order to produce a quick estimate of the gamma photons energy involved in the reactions. This detector, based on silicon, is realized with a single chip of 256 × 256 pixels bump-bonded with a 14 mm × 14 mm × 300 μm silicon layer. Interaction of gammas with the detector in this configuration produces some characteristic clusters of pixels and, for each cluster, a variety of physical and morphological parameters can be defined. Based on some of these parameters, we have characterized the detector response using some known laboratory gamma sources and the related Geant4 simulations. This allows quick energy discrimination for the gamma photons coming from different experimental runs.

Data on nuclear level densities extracted from transmission data or gamma energy spectrum store the basic statistical information about nuclei at various temperatures. Generally this extracted data goes through model fitting using computer codes like CASCADE. However, recently established semiclassical methods involving no adjustable parameters to determine the level density parameter for magic and semi-magic nuclei give a good agreement with the experimental values. One of the popular ways to paramaterize the level density parameter which includes the shell effects and its damping was given by Ignatyuk. This damping factor is usually fitted from the experimental data on nuclear level density and it comes around 0.05 $MeV^{-1}$. In this work we calculate the Ignatyuk damping factor for various nuclei using semiclassical methods., Comment: 9 Figures, 1 Table

124I has a complex decay scheme with high gamma energy and low positron abundance. In this study, comparative performance measurements of 124I were performed in terms of spatial resolution, sensitivity, and image quality. All measurements were performed using both 2D and 3D PET in both brain mode and whole body mode. The transverse and axial spatial resolutions at 1 cm were 5.56 and 6.07 mm for 124I, and were 4.58 and 4.77 mm and for 18F, respectively. Sensitivities were 0.5 kcps/MBq (2D) and 3.4 kcps/MBq (3D) for 124I, and 1.8 kcps/MBq (2D) and 9.8 kcps/MBq (3D) for 18F. The %contrast of 3D was higher than that of 2D in 124I. For 124I PET imaging, 3D acquisition with brain mode was highest achievable imaging acquisition mode with finer spatial resolution and higher contrast. This result will be useful for 124I PET imaging.

The Space and Endo-Atmospheric Nudet (Nuclear Detonation) Surveillance Experimentation and Risk-Reduction (SENSER) project, scheduled to launch to geosynchronous orbit in 2019, contains several experimental and demonstration systems developed by Los Alamos National Laboratory and Sandia National Laboratories to assess next-generation technologies for space-based national security missions. The CLYC experiment in SENSER will examine Cs 2 LiYCl 6 (CLYC) scintillators in the space environment. CLYC enables neutron and gamma-ray detection in a single instrument and better gamma energy resolution than traditional scintillators. The CLYC experiment will also investigate silicon photomultiplier (SiPM) photodetectors for additional SWaP reduction. In this paper, we present the CLYC experiment objectives, design, and results of initial laboratory tests of the partially-built system.

The aim of this study is to develop the assessment technique of the effective dose by calculating the organ equivalent dose with a Monte Carlo (MC) simulation and a computational human phantom for the naturally occurring radioactive material (NORM) added consumer products. In this study, we suggests the method determining the MC source term based on the skin-point source enabling the convenient and conservative modeling of the various type of the products. To validate the skin-point source method, the organ equivalent doses were compared with that by the product modeling source of the realistic shape for the pillow, waist supporter, sleeping mattress etc. Our results show that according to the source location, the organ equivalent doses were observed as the similar tendency for both source determining methods, however, it was observed that the annual effective dose with the skin-point source was conservative than that with the modeling source with the maximum 3.3 times higher dose. With the assumption of the gamma energy of 1 MeV and product activity of 1 Bq g −1 , the annual effective doses of the pillow, waist supporter and sleeping mattress with skin-point source was 3.09E-16 Sv Bq −1 year −1 , 1.45E-15 Sv Bq −1 year −1 , and 2,82E-16 Sv Bq −1 year −1 , respectively, while the product modeling source showed 9.22E-17 Sv Bq −1 year −1 , 9.29E-16 Sv Bq −1 year −1 , and 8.83E-17 Sv Bq −1 year −1 , respectively. In conclusion, it was demonstrated in this study that the skin-point source method could be employed to efficiently evaluate the annual effective dose due to the usage of the NORM added consumer products.

International audience; One of the features of the spectra obtained by airborne gamma spectrometry is the low counting statisticsdue to a short acquisition time (1 s) and a large source-detector distance (40 m) which leads tolarge statistical fluctuations. These fluctuations bring large uncertainty in radionuclide identification anddetermination of their respective activities from the window method recommended by the IAEA,especially for low-level radioactivity. Different types of filter could be used on spectra in order to removethese statistical fluctuations. The present work compares the results obtained with filters in terms oferrors over the whole gamma energy range of the filtered spectra with the window method. These resultsare used to determine which filtering technique is the most suitable in combination with somemethod for total stripping of the spectrum.

The recently developed Cs2LiYCl6:Ce (CLYC) crystals are interesting scintillation detectors not only for their gamma energy resolution (

Gamma attenuation coefficients of host rocks belonging to Moghra Formation of Eocene era in one of two candidate sites located at the North Western Coast of Egypt in Al-Dabaa locality were estimated using Monte-Carlo Code for Nuclear Power (MCNP). Carbonates rock samples were taken to represents several depth intervals. A virtual mono-energetic gamma source with increasing energies was used. The gamma attenuation properties were found to be totally independent on the lithology of rock type, provided that roch sample density remains constant. Ultimately it is found that, gamma attenuation coefficient showed an inverse exponential dependence on the gamma energy regardless rock type and their related depth.

Concentration of 212Bi, 212Pb and 228Ac were tracked at different time intervals for 33 days in IAEA-RGTh-1 reference material having 2.89% Th and soil collected from ambient environment. In case of reference material, mean concentration of 232Th evaluated using gamma energy line of 228Ac (912 keV) and 212Bi (728 keV) shows a deviation of only 0.05%, and 5.66% respectively from the certified mean value of 232Th, whereas a large negative deviation of 29.9% was observed in the case of 212Pb (239 keV). Th-232 concentration in soil sample as evaluated using above three different gamma lines shows a variation of only 10-13%.