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The present paper summarises the results of the simulation of heat transfer to supercritical water in rod-bundle geometries by a CFD code, using wall function models. Two different sets of experimental data were considered, concerning both relatively high and low mass flux conditions and inlet temperature spanning from low to near-critical values. In past analyses, the unsuitability of low-Reynolds number turbulence models was observed in predicting heat transfer in rod bundles, when both high mass and heat flux values were imposed; in fact, large overestimations of wall temperatures were reported in such conditions. This motivated to try simpler models, such as the wall function approach, in order to investigate if, though expectedly not very accurate, they could at least reproduce the experimental data at an acceptable level. As reported in the present paper, the selected models, which adopt a “high y + ” wall treatment (indicating wall functions in the STAR-CCM+ code), seem reasonably able at reproducing the general observed experimental trends. The present understanding of the phenomena and the available modelling techniques unfortunately do not allow obtaining better results when the wall or fluid temperature approach the pseudocritical value. However, the comparison with results obtained by low-Reynolds models shows that, at least when facing operating conditions similar to the ones considered in the present work, a downgrading of the adopted modelling techniques may be beneficial and allows obtaining reasonable results. Analyses were also performed considering conditions far from the pseudo-critical temperature, in which low-Reynolds models had provided good performance, reporting that the wall function approach seems effective also in these cases for obtaining first guess results. [ABSTRACT FROM AUTHOR]

Abstract: Gamma ray tracking is an efficient detection technique in studying exotic nuclei which lies far from beta stability line. To achieve very powerful and extraordinary resolution ability, new detectors based on gamma ray tracking are currently being developed. To reach this achievement, the neutron–gamma discrimination in these detectors is also an important task. In this paper, by suitable layered feedforward neural networks (LFNNs), we have constructed novel and consistent empirical physical formulas (EPFs) for some highly nonlinear detector counts measured in neutron–gamma discrimination. The detector counts data used in the discrimination was actually borrowed from our previous paper. The counts used here had been originally measured versus the following parameters: energy deposited in the first interaction points, difference in the incoming direction of initial gamma rays, and finally figure of merit values of the clusters determined by tracking. The LFNN–EPFs are of explicit mathematical functional form. Therefore, by various suitable operations of mathematical analysis, these LFNN–EPFs can be used to derivate further physical functions which might be potentially relevant to neutron–gamma discrimination performance of gamma ray tracking. [Copyright &y& Elsevier]

Highlights • The effect of time-dependent delayed energy is investigated for slow and fast transient problems. • For operational depletion, the effect of the time-dependent delayed heat is small. • For a PWR rod ejection, the time-dependent model increases the overall energy release by 7%. • The accuracy of scaling the spatial energy deposition of delayed gamma rays is affirmed. Abstract In reactor core analysis, equilibrium delayed fission energy is often assumed except for estimating the decay heat during reactor shutdown or post-accident scenarios. In this paper, we investigated a few slow and fast transient operational conditions in which the assumption of equilibrium delayed energy could be questionable. A model of lumping the delayed heat source into groups has been adopted in this study to calculate the time-dependent delayed power in the CASL neutronics code MPACT. For the normal operational depletion, the time-integrated effect of the delayed heat shows a small impact in calculating the flux normalization factor, and thus eigenvalue and depletion isotopics. The effect of delayed energy for the reactivity-initiated accident (RIA) problems is more interesting. For a PWR 3-D full core simulation of rod ejection, the peak core power is reduced by about 3% and the decay heat is significantly higher after SCRAM, with an overall energy release increased by 7%. Therefore, the time-dependent delayed model is more beneficial for fast transient analysis. In addition to the temporal effect, the spatial distribution of energy deposited from delayed gamma rays are compared with that from prompt gamma rays for a few typical LWR mini-lattices, and the results affirm that using the scaling approach for spatial energy deposition of delayed gamma rays has good accuracy. [ABSTRACT FROM AUTHOR]

Prototypic neutronic parameters of thermal reactors were emulated in materials testing assemblies located in a 600 MWth metallic-fueled sodium fast reactor. Part of the outer reflector is composed of a moderating material such as graphite; all of the test assemblies were located in this designated thermalized region. Possible options for flux shaping assemblies and materials irradiation test assemblies are described. Both active and passive test assemblies are described. This paper focuses on PWR and VHTR in-core conditions. It is possible to modify the fast spectrum streaming from the driver to softer spectrum regions. In all configurations considered, the suggested options lead to significant neutron fluxes supporting accelerated neutron-induced damage studies. [ABSTRACT FROM AUTHOR]

The use of simulation within the nuclear industry has been limited by the perceived cost and benefits that such tools offer. In this paper, we present the development of a simulator based on the use of master-slave-manipulators to dismantle a small research reactor core assembly. We discuss the design of the simulator hardware and the different software elements that make up the system. We show how this type of tool can be used to explore various options during the development of a nuclear decommissioning process. Finally, we discuss the benefits that have been observed from using the system to support the decommissioning of the CONSORT research reactor. [ABSTRACT FROM AUTHOR]

Abstract: This paper develops the work of Cattle et al. [Cattle, B.A., Fellerman, A.S., West, R.M., 2004. On the detection of solid deposits using gamma ray emission tomography with limited data. Measurement Science and Technology 15, 1429–1439] by considering a generalization of the model employed therein. The focus of the work is the gamma ray tomographic analysis of high-level waste processing. The work in this paper considers a two-dimensional model for the measurement of gamma ray photon flux, as opposed to the previous one-dimensional analysis via the integrated Beer–Lambert law. The mathematical inverse problem that arises in determining physical quantities from the photon count measurements is tackled using Bayesian statistical methods that are implemented computationally using a Markov chain Monte Carlo (MCMC) approach. In a further new development, the effect of the degree of collimation of the detector on the reliability of the solutions is also considered. [Copyright &y& Elsevier]

This paper presents the capabilities of a measurement setup built to measure the axial distribution of neutron emitters in spent fuel pins from nuclear power plants. The measurement system uses a custom gamma-blind detector based on a novel combination of silver-activated zinc sulfide scintillators and wavelength-shifting fibers for direct fast neutron detection while having a low sensitivity for gamma rays even in the strong gamma fields present during the inspection of spent fuel. A collimator with an opening of 2 cm is used to create axial resolved information. The system has been tested using UOX and MOX spent fuel samples with burn-ups between 40 and 90 GWd/t and with long cooling times (20–25 y). Measurements of the total neutron emission of the samples relative to each other achieved a 1 σ uncertainty of 2%. Absolute values were measured by comparison with a calibrated 252Cf source, achieving a 1 σ uncertainty of 5%. The neutron emission of the samples was previously measured using a thermal neutron detector. The results from that previous measurement agree within 2. 5 σ with the new results. The reconstruction of the axial neutron emission profile was performed for one of the samples. Based on the edges of the sample, the resolution of the reconstruction was estimated to be 2.5 cm. These promising results are the basis for future development of the measurement setup which should not only improve the accuracy of the results but also enable the measurement of spent fuel with much shorter cooling times of ∼ 1 y. • Neutron emission of spent fuel rods was measured with a novel detector. • Relative neutron emission was measured with a standard deviation of 2%. • Relative and absolute mission values agree with previous measurements. • Absolute neutron emission was measured with a standard deviation of 5%. • Axial resolution was estimated at 2.5 cm. [ABSTRACT FROM AUTHOR]

The use of a small-bore highly reflective metallic tube has been demonstrated at a pool-type nuclear reactor for transmitting Cherenkov radiation emitted from a quartz cylinder placed near the reactor core. The study revealed the promising prospect of using a metallic tube to remotely make localized measurements in order to independently monitor reactor power using Cherenkov light generated in the water pool or inside an attached radiator. This study confirmed qualitatively and quantitatively that commercially available stainless steel tubes with highly reflective inner surfaces can adequately transport Cherenkov light, produced in the proximity of a SLOWPOKE-2 reactor core operated at 20 kW, with acceptable transmission losses over significant distances. It was demonstrated that this Cherenkov-based instrument has a fast response, high fidelity, and the Cherenkov-induced output current varies linearly with the reactor power. Scans in the orthogonal directions showed the possibility of using the Cherenkov-based instrument for 3-D mapping of the general gamma field around the reactor core. Analytical and numerical models were developed and validated by experiments to predict the transmission efficiency of light in the highly polished small-bore light guides. This paper provides details of calculations and experimental validation under laboratory conditions and at a pool-type SLOWPOKE-2 reactor. [ABSTRACT FROM AUTHOR]

In transport theory, adjoint-based partial derivatives with respect to mass density are constant-volume derivatives. Likewise, adjoint-based partial derivatives with respect to surface locations (i.e., internal interface locations and the outer system boundary) are constant-density derivatives. This paper derives the constant-mass partial derivative of a response with respect to an internal interface location or the outer system boundary and the constant-mass partial derivative of a response with respect to the mass density of a region. Numerical results are given for a multiregion two-dimensional ( r - z ) cylinder for three very different responses: the uncollided gamma-ray flux at an external detector point, k eff of the system, and the total neutron leakage. Results from the derived formulas compare extremely well with direct perturbation calculations. [ABSTRACT FROM AUTHOR]

The laser-Compton scattering (LCS) phenomenon is one known method for producing energetic and high-intensity gamma rays with variable photon spectrums. Availability of high-intensity gamma rays has made probing the (γ,n)-based photonuclear transmutation possible and worth consideration. Parametric studies of LCS have shown that the gamma ray’s spectrum is tunable and dependent upon various facility-related parameters. In this paper, a systematic optimization study of the LCS spectrum has been performed in view of the incident laser energy, electron beam energy and collimation angle. This optimization has been done with the aim to maximize the transmutation reaction rates of long-living fission products (LLFPs) such as 129 I, 135 Cs, and 137 Cs that have an isotopic composition of typical LWR spent fuel. In addition, the (γ,n) and (γ,2n)-reaction rates have been simultaneously calculated in order to evaluate the feasibility of photonuclear transmutation using the (γ,n)-reaction. It has been shown that careful optimization of the LCS spectrum noticeably enhances the (γ,n)-reaction rate without requiring any isotopic separation of the iodine and cesium targets. Furthermore, in order to enhance the energy efficiency of the LCS-based photo-transmutation system, a novel multiple laser-Compton scattering extraction (MULEX) concept has been introduced in this work, and it has been shown that the total production of the LCS photons can easily be increased by a factor of 10 or 20 in a single accelerator system. [ABSTRACT FROM AUTHOR]

The variation in dosimetric parameters such as mass attenuation coefficient, half value layer factor, exposure buildup factor, and the photon mean free path for different oxide glasses for the incident gamma energy range 0.015–15 MeV has been studied using MicroShield code. It has been inferred that the addition of PbO and Bi 2 O 3 improves the gamma ray shielding properties. Thus, the effect of chemical composition on these parameters is investigated in the form of six different glass compositions, which are compared with specialty concrete for nuclear radiation shielding. The composition termed ‘Glass 6’ in this paper has the highest mass attenuation and the smallest half value layer and may have potential applications in radiation shielding. An example dry storage cask utilizing an additional layer of Glass 6 as an intermediate shielding layer, simulated in MicroShield, is capable of reducing the exposure rate at the cask surface by over 20 orders of magnitude compared to the case without a glass layer. Based on this study, Glass 6 shows promise as a gamma-ray shielding material, particularly for dry cask storage. [ABSTRACT FROM AUTHOR]

[Display omitted] • Provision to handle thick shields up to 100 mean free path. • Incorporation of modified Kalos' formula for double layer gamma ray exposure buildup factor. • Validation of the code in handling bulk and stratified shielding configurations. Salient modifications and validations performed on a Point Kernel method based code QADCGPIC, which are important for optimizing shield design is presented in the paper. The geometric progression method based build–up factors (BUFs) for penetration depths of medium up to 100 mean free path is incorporated into the modified QADCGPIC–M code. Further improvement to the code is made by incorporating the modified Kalos' formula of double layer gamma ray exposure build–up factor (DBUF) for the shielding materials used in the design of nuclear and radiological facilities. The modified code is validated by estimating the BUFs, DBUFs, and gamma dose rates at the exit of the shield with Monte Carlo simulation for various shielding configurations of lead, tungsten, iron, and concrete and published values. The validation exercise indicates a good agreement between the computed values by different methods. The incorporation of geometric progression fitting formula and modified Kalos' formula into the QADCGPIC–M code increases the ability of the code to handle bulk shielding and optimization of thicknesses of double layered shielding configurations. [ABSTRACT FROM AUTHOR]

In this paper we are proposing a re-evaluation of the thermal-neutron induced capture cross section of 147 Nd. A unique measurement exists from which this cross section was calculated in 1974. This original calculation is based on an assumed value for a specific gamma-ray fraction (called F 2 ), taken from the neighboring nucleus 145 Nd. With the availability of reaction codes such as TALYS, such fraction can nowadays be calculated using specific reaction models and parameters. The new value of F 2 indicates a decrease of the thermal cross section by 45%, leading to 243 barns, instead of the 440 barns previously reported. This new cross section impacts the calculation of the number density for the well-known burn-up indicator 148 Nd, but as shown, the change is close to the usual experimental uncertainties for the 148 Nd number densities, thus having a limited impact on burn-up calculation. [ABSTRACT FROM AUTHOR]

• We have proposed a UNET model based on deep neural network (DNN), and built the source term activity reconstruction system using hardware. • Through the measurement of the radiation field, the source term activity can be solved quickly and flexibly. • Visualize the reconstruction results of the source term activity to determine the location of radiation hot spots. • Portable source term activity reconstruction systems can optimize decommissioning activity dose uptake and workforce in engineering applications. The radiation damage weakens the efficiency of decommissioning nuclear facilities, and the source term is hard to be located and measured because it often distributes in or inside objectives. In practice, the hot spot and source intensity are measured by the gamma camera then the radiation field is estimated by particle transport algorithm as a reference in scheme planning. This technology is expensive in measuring and simulating. Hence, this paper presents a source term activity reconstruction (hereinafter referred to as STAR) method based on deep learning to solve this problem. Firstly, a UNET liked framework is constructed to establish the correlation between the radiation field and source activity. Secondly, a source activity reconstruction problem is used to validate the framework. Then we deploy it to the Raspberry Pi with a γ-ray detector to verify it can be applied in practical works. The verification results show that Raspberry Pi can complete source term activity reconstruction in a few seconds without consuming too much computing resources. In framework validating, 5000 samples consist of randomly generated activity distribution and its grid dose value which is calculated by the Monte-Carlo program. The results show that the average reconstruction error is less than 15% and the trained framework is performing well in Raspberry Pi. This method reduces the requirements for instruments and the dose detection is parallelable. Therefore, it can be widely used in nuclear facility decommissioning to improve the efficiency of source reconstruction. [ABSTRACT FROM AUTHOR]

• Capable of reconstructing the 3-D radiation field of different numbers of radiation sources. • Works with both randomly and regularly sampled data. • Verified with multiple Monte Carlo simulations and validated against a 1-D experiment. • Takes no more than 3.125% of simulation data for 3-D field reconstruction. • Mean relative errors can reach 2.18% for simulation and 7.87% for experiment. The reconstruction of 3-D gamma radiation field is important for radiation protection in the presence of unknown radiation sources and limited measurements, such as the decontamination and decommissioning of nuclear facilities. While there may be multiple radioactive sources in these applications, the current 3-D reconstruction methods do not take the multiple radiation sources into consideration. This paper develops a method that has the potential to reconstruct the 3-D gamma radiation field of multiple radioactive sources with limited measurements of the field. This method solves a 3-D fourth-order TV-H-1 equation to promote the continuity of the reconstructed 3-D radiation field, which simultaneously fills the missing dose rates and recovers local details near the radioactive sources. The proposed method was verified against three Monte Carlo simulations with different numbers of sources and validated against a 1-D experiment. The method was tested with different numbers of combinations of the input data, including both randomly and regularly distributed data sets. The results reveal that the method accurately reconstructs the 3-D radiation field and achieves a relative error level as low as 2.18%, using only 3.125% of the data for the simulation cases. The 1-D experiment shows a relative error range between 5.57% and 65.28%. These results preliminary demonstrate the potential of the proposed method for 3-D radiation field reconstruction in practice. [ABSTRACT FROM AUTHOR]

• The photon-induced multiple neutron generation reaction rate ratio (PMNRR) method was proposed to detect 237Np. • A combination of photon energies ranging over 12–15 MeV and 6–11 MeV was selected. • The feasibility of the detections of the separated 237Np and the 237Np above a certain composition ratio in UO 2 was verified. • The PMNRR method is robust to the energy distribution of incident photons. This paper presents a novel method for the detection of 237Np in its separated form and advanced fuel by applying the principle of the photofission reaction ratio (PFRR). In this method, 237Np was identified using the photon-induced multiple neutron generation reaction rate ratio (PMNRR) with multiple high-energy gamma rays. The feasibility of the discrimination between the separated 237Np and uranium (235U, 238U) and the detection of the 237Np above a certain composition ratio in UO 2 was examined by numerical analysis simulation. The proposed method has a wide range of incident photon energy selectivity because it utilizes the (*, 2n) reaction, which is considered an inevitable measurement obstacle in the PFRR method. The results show the detectability of 237Np even with a 10% uncertainty in the nuclear data of the (γ, fission/2n) reaction, and the robustness of the PMNRR method to the energy distribution of incident photons. In the future, the proposed method is expected to be applied for detection and verification in the field of nuclear non-proliferation and security. [ABSTRACT FROM AUTHOR]

Abstract: A new computer code named as SHIELD DESIGN has been developed in Visual Basic Version 6.0 for precise estimation of shielding thickness in gamma ray shielding designs. In the conventional forward calculations using point kernel method, the sources and geometry are specified along with the shield thickness to obtain dose rates at desired locations. But in this code, a new computational technique based on inverse calculation and iterative procedures have been employed to estimate the shield thickness using the dose rate criteria as input information. This new technique enables quicker and easier way of shield design optimization by computing shield thickness required in a single run whereas in the forward shield design optimization method, dose rate computations need to be carried out for various shield thicknesses. The shielding thickness arrived with this code provides a first hand information to the designer to carry out forward calculations for accurate optimization. Also, the code is very useful to regulators for quicker verification of shield calculations submitted by the designers for approval. Further, various options available in this code extend its flexibility and user friendliness. This paper briefly presents the development and validation of the code, SHIELD DESIGN. The validation has been carried out by performing analyses of the gamma shielding design for some standard problems. The results were compared with those of renowned forward calculations, experimental measurements, NCRP-51 benchmarks and other gamma radiation shielding utilities like RAD-PRO and satisfactory agreement was observed. These validation tests demonstrate the precise estimation of shield thickness. [Copyright &y& Elsevier]

Abstract: KAERI has developed a lattice transport calculation code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and its library generation system. Recently, the library generation system has been extended to include a gamma cross section generation capability and a gamma transport/diffusion calculation module has been implemented in KARMA 1.2. The method of characteristics for the neutron transport calculation to estimate eigenvalue has been utilized to predict gamma flux distribution and energy deposition. In addition, the coarse mesh finite difference method with diffusion approximation has also been utilized to estimate gamma flux distribution and energy depositions for each coarse mesh with homogenized pins as a computationally efficient alternative. This paper describes the procedure to generate neutron induced gamma production and gamma cross section data, and the methods to predict gamma flux distribution, gamma energy deposition and gamma smeared pin power distribution. The computational results for benchmark problems show that the gamma library and gamma simulation in KARMA are reasonable. And it is noted that gamma smeared power distributions predicted by coarse mesh diffusion calculation are very accurate compared to the results of transport calculation. [Copyright &y& Elsevier]

Abstract: This paper represents the evaluation of the gamma-ray spectrum associated with the 252Cf spontaneous fission source, based on Monte Carlo simulation. In this study, with the help of the fission yields and gamma-ray spectrum emitted from each of the products, the probability of gammas in 252Cf fission has been obtained. Furthermore, the probabilities of emitted gammas following the alpha decay from 252Cf have been computed, given that in this isotope 96.91% of decays are in the form of alpha. Finally, about 450 gammas with different energies and probabilities have been extracted and after regulating their normalization, they are given to the MCNP computer code as input. Some addition gammas due to (n, γ) interaction in the detector have also been considered. The obtained results are in relatively good agreement with experiment data. [Copyright &y& Elsevier]

Abstract: We present an empirical model that describes the yield of gamma rays emitted by fission in the time interval from 20 to 958ns following a fission event. The analysis is based on experimental data from neutron-induced fission of 235U and 239Pu. The model is devised by first using regression analysis to identify likely patterns in the data and to choose plausible fitting functions. We provide statistical and physical arguments in support of time and energy independence. The intensity of the emitted gamma rays can be described as a bivariate distribution that is the product of independent variates for energy and time. We test several plausible distribution families for the energy and time variates and use maximum likelihood and minimum χ 2 to estimate distribution parameters. Because of the uncertainty in the experimental data, multiple combinations of variate pairs give rise to a surface that plausibly well fits the observations well. The best-fit variate turns out to be lognormal in energy and F in time. The findings illustrated in this paper can be used to simulate gamma ray de-excitation from fission in Monte Carlo codes. [Copyright &y& Elsevier]

Abstract: This paper describes a new radiation shielding material, amethyst ore. We have determined the elemental composition of amethyst using WDXRF spectroscopy technique. To see the shielding capability of amethyst for several photon energies, these results have been used in simulation process by FLUKA Monte Carlo radiation transport code. Linear attenuation coefficients have been calculated according to the simulation results. Then, these values have been compared to a fine shielding concrete material. The results show that amethyst shields more gamma beams than concrete. This investigation is the first study about the radiation shielding properties of amethyst ore. [ABSTRACT FROM AUTHOR]

Abstract: The Schwinger method for solving inverse transport problems is applied to the problems of interface location identification, shield material identification, source isotope weight fraction identification, and material mass density identification (separately) in multilayered two-dimensional cylindrical gamma radiation source/shield systems. The method is iterative and estimates unknown interface locations, source isotope weight fractions, and material densities directly, while the unknown shield material is identified by estimating its total macroscopic gamma-ray cross sections. The energies of discrete gamma-ray lines emitted by the source are assumed to be known, while the unscattered flux of the lines is assumed to be measured at points external to the system. In numerical test cases, the Schwinger method correctly identifies the unknowns when the same deterministic ray-tracing code is used for both the parameter estimation process and simulation of the measured data. With realistic simulation of the measured data using a Monte Carlo code, the method produces more ambiguous results for interface location, shield material identification, and material density identification. The method works well for source weight fraction identification with measured data simulated by Monte Carlo. In addition to the application to more realistic (two-dimensional) problems, this paper extends previous work on the Schwinger inverse method by using surface formulas for unknown interface locations, automatic correction attempts for violated constraints, and ray-tracing instead of discrete-ordinates for transport calculations. [Copyright &y& Elsevier]

Abstract: This paper demonstrates a parameter identification problem encountered in the assay of intermediate-level nuclear waste vaults. A lumped parameter model of the vault contents is employed in which a likely geometry is assumed so as to reduce the number of unknown quantities in the resulting inverse problem. The inverse problem is tackled using Bayesian statistics, and the statistical distributions are sampled using a Markov chain Monte Carlo approach. The Bayesian approach affords flexibility in the specification of prior information: i.e. information that is known in advance of recording any measurements. Two parameterizations of the problem are considered: one involving five parameters and the other involving three parameters. It is shown that in order to obtain a satisfactory solution to the five-parameter problem a very specific type of prior information has to be specified, whereas for the three-parameter problem much less specific prior information is required. The inverse problem of identifying the model parameters in both the three-parameter and five-parameter formulations is solved with limited, but quantifiable precision. [Copyright &y& Elsevier]

Abstract: An important issue in nuclear safeguards is to verify operator-declared data of spent nuclear fuel. Various techniques have therefore been assigned for this purpose. A nondestructive approach is to measure the gamma radiation from spent nuclear fuel assemblies. Using this technique, parameters such as burnup and cooling time can be calculated or verified. In this paper, we propose the utilization of gamma rays from 137Cs, 134Cs and 154Eu to determine the consistency of operator-declared information. Specifically, we have investigated to what extent irradiation histories can be verified. Computer simulations were used in order to determine limits for detecting small deviations from declared data. In addition, the technique has been experimentally demonstrated on 12 PWR fuel assemblies. A technique for determining burnup and cooling time for fuel assemblies where no operator-declared information is available is also presented. In such a case, the burnup could be determined with 1.6% relative standard deviation and the cooling time with 1.5%. [Copyright &y& Elsevier]

Abstract: Artificial neural network (ANN) has recently been used for the analysis of gamma-ray spectrum. The ANN can provide a computational model which has a cost in terms of the time comparable to that of more conventional mathematical models. In this paper, the gamma-ray spectra measured for 7 different mediums were available in the training data set to ANN which was developed 11-input layer, 1-output layer model with three hidden layer. The input parameters were atomic percent of elements constituted the mediums, Compton cross-section, photoelectric cross-section and channel number. The output parameter was counts per channel. The network has been trained using Kohonen and back propagation algorithm with the hyperbolic tangent transfer function in hidden layers and sigmoid transfer function in output layer. After the network was trained, mean squared error was found to be 0.00008. When the network was tested by untrained data, the linear correlation coefficient was found to be 99%. [Copyright &y& Elsevier]

This paper presents the study of the gamma exposure to the spent fuel pool walls from spent fuel assemblies. The Monte Carlo code, MCNP4C, was used to model and analyze the spent fuel pool walls. In addition, the gamma source from the spent fuel assembly was modeled by the isotope generation and depletion code, ORIGEN2.2. Both PWR (TMI-I) and BWR (PB-2) spent fuel pool models were investigated. The primary objective of the study was to determine the arrangement of the spent fuel assemblies to minimize the gamma radiation to the spent fuel pool walls. Different cases of spent fuel assembly configurations in the spent fuel pool were studied. The study results show the amount of gamma radiation deposited on each section of the spent fuel pool walls. Finally, the spent fuel assembly with axial peaking factor was also studied. [Copyright &y& Elsevier]

In this paper the calculation of direct heat generation and energy savings due to the penetration of 1.37 and 2.75 MeV energy photons, emitted from a Na-24 radiation facility, through double layer shielding slabs of aluminium, steel and lead is described. A comparison is being made among six different shielding material combinations in order to assess the optimum shield related to the maximum energy captured due to γ-rays penetration through the combined shielding materials. [Copyright &y& Elsevier]

3DGShield, A new 3D gamma ray shielding code, based on the method of point kernel technique, has been developed for arbitrary regular & irregular shaped source and shield geometries. The 3D geometry modeling and its mesh generation are performed using Gmsh, a free open source finite element mesh generation software. The output of the Gmsh software is in Gmsh MSH (.msh) file format which will be used as an input of geometry description of our code. The salient features of 3DGShield are (i) ability to handle any type of complex source geometries including regular source shapes like point, line, disc, box, cylinder, sphere, hexagonal cylinder, cone etc. and irregular source shapes like pieces of broken glass, piece of brick etc., (ii) handling discrete as well as distributed photon energy sources, (iii) modeling any kind of regular and irregular complex multilayered shield geometries, (iv) having provision to export the output results into ParaView, an open source data analysis and visualization software, for quick visualization and analyzing of the results. Buildup factor is the correction factor which takes account the contribution of scattered photon. Various appropriate multilayer build-up factor formulas have been incorporated in our code. Out of these, the most popular methods are Blizard's method, Goldstein method of an effective atomic number and Broders's method. Verification and validation of the code have been done by a number of benchmark problems available in literature. The results are presented in this paper which demonstrate the accuracy and efficiency of the code in complex geometries. This code is written in Python-2.7 language and compatible both in Windows and Linux operating system. [ABSTRACT FROM AUTHOR]

Abstract: The electron paramagnetic resonance (EPR) and X-ray diffraction (XRD) investigations of non-irradiated and gamma-irradiated Trommel Sieve Waste (TSW) have been carried out at room temperature. TSW is obtained during the process of manufacturing borax from tincal. In this paper, we aimed to study the magnetic properties of TSW before and after irradiation. The EPR spectrum of non-irradiated TSW showed a naturally present free radical EPR singlet, g =2.0040±0.0005, and six hyperfine lines of Mn2+. After gamma-irradiation, it is observed that only the signal intensity has changed. Also, XRD analysis of non-irradiated and gamma-irradiated TSW has been performed to determine whether TSW is a crystalline material. [Copyright &y& Elsevier]

• The dry cask was reproduced in SCALE 6.2.3 and validated with measurement data. • The cask surface dose rate was calculated for a 300-year storage time. • The variations of radiation source dose rate contributions to the total dose were obtained. • Neutron and gamma ray dose rate percentages according to energy group were analyzed. The dry storage cask became a semi-permanent solution for spent nuclear fuel since no permanent solution has been finalized over the past three decades. Considering the difficulty of deep geological repository projects in many aspects, prolonging the service life of dry casks longer than a few decades has become an option for spent nuclear fuel storage. It must be ensured that dry cask safety objectives, including radiation shielding, criticality safety, structure health, decay heat removal, and containment be fulfilled during the prolonged service life. This paper presents numerical modeling results of the radiation dose rate distribution on a TN-32 cask surface aged up to 300 years. The numerical modeling method was validated with measurement data, and the difference between the calculated results and measurement data in the total dose rate is 7%. The detailed radiation distribution variations over cask surfaces along the axial and radial directions for 300 years of storage were obtained, and their azimuthal angular dependence was established. The maximum total dose rate over the cask surfaces is less than 10 mrem/hour (9.258 ± 0.212 mrem/hour) after 40 years, and this outcome fulfills all dose rate requirements from 10 CFR 71.47(b). After 300 years of storage, the average total dose rates on the side surface and top surface are 0.150 ± 0.007mremmrem/hour and 0.057 ± 0.003mremmrem/hour, respectively, while the maxima are 0.601 ± 0.014 mrem/hour and 0.112 ± 0.006 mrem/hour, respectively. The variations in the fuel gamma ray, structure gamma ray, neutron and secondary gamma-ray contributions to the total dose rate are presented and analyzed in this work. The neutron and gamma-ray dose rate contributions by energy group are presented and analyzed at the cask top, mid-plane, and bottom at the selected storage times. [ABSTRACT FROM AUTHOR]