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Gohar, Y.

Publications and source records attributed to Gohar, Y..

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Electron Target Cooling Analyses of the KIPT ADS Using MCNP and Ansys Fluent

This study presents multiphysics analyses of the electron target cooling system of the accelerator-driven system (ADS) of the Kharkiv Institute of Physics and Technology (KIPT) using MCNP and Fluent computer programs. MCNP has been used to transport electrons, gammas, and neutrons, and to calculate the energy deposition in the target materials. The MCNP mesh-tally data have been imported into Fluent by a C subroutine that has been compiled and linked to Fluent as a user-defined function.The KIPT ADS is located in Ukraine and was in operation until February 2022. The Fluent model is based on the computer-aided design files from the manufacturing process of the target assembly. Here, the Fluent results for the reference case match very well the literature results obtained by STAR-CCM+ during the design phase. Other cases that differ from the reference one have been analyzed; in these cases, it is assumed a malfunction of the electron accelerator or of the water cooling system. The target cooling system operates normally for all the analyzed cases except when the inlet water mass flow rate is decreased. The transient analysis showed that the target cooling system can operate for 180 s with full power when the inlet water mass flow rate is decreased down by 75%.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Stereolithographic geometry model of the IBR-2M experimental facility

The IBR-2M is a fast research reactor that operates in supercritical condition for ∼ 800 μs every 200 ms. Two reflector parts in nickel rotate in opposite directions generating 1.8 GWth peak power when they align with the fuel zone changing the reactor status from deep subcritical to supercritical. The reactor core uses high-enriched plutonium fuel and is cooled by sodium. This reactor has been modeled by MCNP and SERPENT computer programs. The MCNP model uses combinatorial geometry, whereas the SERPENT model employs the Stereolithographic (STL) geometry representation that can be used by 3D printers. The STL geometry was constructed using the CUBIT computer program. The CUBIT program was also used for a three-dimensional visualization of the Monte Carlo models. SERPENT and MCNP models use the same geometry, material specifications, and nuclear data. The latter are based on the ENDF/B-7.0 library. SERPENT and MCNP using same geometry and same material specifications produce similar k{sub eff} values within 120 pcm.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Kinetics calculation of fast periodic pulsed reactors using MCNP6

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for similar to 400 mu s, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

Zhong, Z.↗
DOE OSTI2017

CUDA Computation of the Feynman Distribution

In 2006, NVIDIA Corporation introduced the Computer Unified Device Architecture (CUDA) for C, C++, and Fortran programming languages. The CUDA parallel computing platform uses the Graphics Processing Unit (GPU) of NVIDIA graphics card, and it is designed for parallel computations. In these computations, the same algorithms are executed on the data in parallel. The GPU contains many Streaming Multiprocessors (SM) that can independently execute blocks of threads. A thread is a sequence of serial programmed instructions. The SM executes threads in groups of 32 and a group of 32 threads is called warp. The threads within a block can be synchronized and have access to a shared memory that is visible only to them. Conversely, blocks of threads cannot be synchronized and have access only to the global memory of the graphics card. This memory is called device memory and is independent of the host memory controlled by the Central Processing Unit (CPU). A single thread has access to a local memory (register) that is not visible to any other thread. The part of the program that is executed on the graphics card is called kernel. Each thread of the kernel has a unique identification index that is accessible by the in-built variable threadIdx; the latter is a 3-component vector. A threads block can contain up to 1024 threads. Blocks of threads have a unique identification index that is accessible by the in-built variable blockIdx; the latter is a 3-component vector. The in-built variable blockDim contains the number of threads per block; this variable is a 3-component vector. Threads and blocks represent different levels of parallelism; a kernel is executed as a grid of blocks of threads. Figure 1 illustrates a grid of thread blocks for a GPU with 2 SM. A SM can execute multiple blocks of threads in parallel; a block of threads is executed on a single SM. A GPU can execute multiple kernels concurrently. This work has used the Quadro M5000 graphics card with the following technical specifications: - 2048 cores with 861 MHz frequency, - 16 streaming multiprocessors, - 5.2 trillion transistors, - 8.2 Gb device memory, - 64 kb register memory (per thread), - 48 kb shared memory (per block) - 211.6 GB/s memory bandwidth, and - 3.5 GFLOPS floating point performance.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

MCNP6 and SERPENT2 Performances with Unstructured Mesh Geometry

Traditionally, Monte Carlo codes model a nuclear reactor by combinatorial geometry. In this model, volume primitives (e,g, spheres, cylinders, parallelepipeds...) are combined together by Boolean operations (e.g. union, intersection, and complement) and replicated by lattice definitions to build the core geometry. Recently, high performances computing platforms allowed using unstructured mesh geometry in Monte Carlo codes. When this option is chosen, any type of core geometry can be modeled since the restriction of using volume primitives is removed. In an unstructured mesh geometry, Boolean operations and lattice definitions are not used since each volume is defined by triangular surfaces (facets). This study focuses on two Monte Carlo codes: MCNP (version 6.1.1 beta) developed at Los Alamos National Laboratory since 1957 and written in FORTRAN90, C, and C++ (267,000 total source code lines), and SERPENT (version 2.1.27) developed at VTT Technical Research Centre since 2004 and written in C (195,000 total source code lines). These two codes follow different approaches for the unstructured mesh geometry modeling: MCNP uses ABAQUS format, whereas SERPENT uses the STL format. In the STL (STereoLithographic) geometry model only the surfaces of the volume are meshed; the interior part of the volumes is not meshed. In the STL geometry model, the whole geometry is defined by the sequence of SOLID and ENDSOLID sections. Each volume of the geometry must have a SOLID and ENDSOLID section. The STL file contains no information specifying the materials of the volumes. MCNP and SERPENT can model any kind of complicated geometry using unstructured meshes. MCNP and SERPENT use the ABAQUS and STL file formats, respectively. The STL file format does not have tetrahedrons and has only triangular facets on the volumes boundary. This allows SERPENT simulations to run 10 times faster than MCNP. However, if Monte Carlo results have to be coupled to thermal-hydraulics codes, the interior of the geometry volumes must be meshed, which may reduce SERPENT performances.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Advances in the computation of the Sjöstrand, Rossi, and Feynman distributions

This study illustrates recent computational advances in the application of the Sjöstrand (area), Rossi, and Feynman methods to estimate the effective multiplication factor of a subcritical system driven by an external neutron source. The methodologies introduced in this study have been validated with the experimental results from the KUKA facility of Japan by Monte Carlo (MCNP6 and MCNPX) and deterministic (ERANOS, VARIANT, and PARTISN) codes. When the assembly is driven by a pulsed neutron source generated by a particle accelerator and delayed neutrons are at equilibrium, the Sjöstrand method becomes extremely fast if the integral of the reaction rate from a single pulse is split into two parts. These two integrals distinguish between the neutron counts during and after the pulse period. To conclude, when the facility is driven by a spontaneous fission neutron source, the timestamps of the detector neutron counts can be obtained up to the nanosecond precision using MCNP6, which allows obtaining the Rossi and Feynman distributions.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

MCNP6 Modification for the Calculation of Rossi and Feynman Distributions

The Rossi distribution is given by the number of captured neutron pairs as a function of the time interval between the two capture events. This distribution contains a random part, which is given by neutron captures caused by neutrons of different fission chains, and a correlated part, which is given by neutron captures caused by neutrons from the same fission chain. The correlated part exponentially vanishes with the {Delta}t time interval following the decay of prompt neutrons. The Rossi distribution peaks when the time interval between two capture events is small since this condition is satisfied for neutrons generated from the same fission chain. In the Poisson probability distribution, the variance is equal to the mean and the next event probability is independent of the elapsed time since the previous event. Neutron captures events do not follow the Poisson probability distribution because they are correlated events via fission chains. Consequently, the Feynman distribution Y is defined as the variance to mean ratio minus one, of the neutron counts array as a function of a fixed time interval. The MCNP6 source code has been modified to allow the calculation of the Rossi and Feynman distributions. The source code modification takes into account the time when a neutron is emitted by the external neutron source and gives a precision in the timestamp lower than a fraction of nanosecond. The modified MCNP6 version has been used to model KUKA facility driven by a californium neutron source. The prompt neutron decay constants obtained by the Rossi and Feynman methods are in agreement.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Mesh Adaptive Search Optimization in SERPENT2 Stereolithography Geometry

SERPENT2 Monte Carlo program has the capability to model reactor cores by stereolithography (STL) geometry. In this geometry, volumes are defined by the union of triangulated surfaces. Using the STL geometry offers unique benefits, including: 1) the model can be printed on 3D printers; 2) the possibility to use directly geometrical models from Computer-Aided Design (CAD); 3) the capability to perform multi-physics simulations coupling neutronics, thermal-hydraulics, and material science computations. In this work the STL geometry has been created by the CUBIT software, which is designed for finite elements mesh generation. The materials in the STL geometry have been assigned by a simple C program using the colors of the volumes in the CUBIT model. The goal of this work is to minimize the SERPENT2 computing time using mesh adaptive search algorithms. The basic idea is to split the geometry of the core into two zones: one zone containing the fuel elements (including the moderator and the coolant) and the other containing the reflector (and the target zone of the core, if it is driven by an external neutron source). By performing this splitting, two separate meshes are used, which results in a considerable computing time saving. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Dead-Time Correction of Pulsed Neutron Source Experiments with the Backward Extrapolation Method

This paper introduces a novel procedure to correct pulsed neutron source experimental data for dead-time effects. The procedure is based on the application of the backward extrapolation method to the reaction rate value in each time interval (bin) of the neutron counts distribution as a function of time. The theoretical framework has been successfully applied to the experimental data (timestamps) from the YALINA Thermal facility driven by a pulsed D-D neutron source. Dead-time effects may have more than 500 pcm (1 pcm=1 E-5) impact on the effective multiplication factor calculated by the area method. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

{sup 99}Mo Production at KIPT Neutron Source Facility

Argonne National Laboratory (ANL) and Kharkov Institute of Physics and Technology (KIPT) have cooperated on the construction of the Neutron Source Facility (NSF) at KIPT, Kharkov (Ukraine). The NSF facility consists of an electron accelerator driven system (ADS) generating 300 kW of thermal power for multiple applications, including medical isotope production. ANL performed extensive analyses including medical isotopes production, reactivity monitoring, operations and control, refueling, cold neutron source design, target design,and radiation shielding. The present study focuses on the use of the KIPT NSF for medical isotope production. Medical isotopes are radioactive isotopes with short half-lives used in diagnostic and therapeutic medical procedures. For example, brain cancer is difficult to treat by ordinary surgery and only brachytherapy with {sup 131}Cs, which has 9.7 days half-life, has shown effective results on the patients. {sup 99m}Tc is the most common medical isotope used in medical diagnoses because the wavelength of its gamma radiation (8.8 pm) is about the same as the conventional X-ray diagnostic devices. {sup 99m}Tc has 6 hours half-life, therefore it cannot be stored because it decays into {sup 99}Tc. The parent of {sup 99m}Tc is {sup 99}Mo, which has 66 hours half-life. Consequently, the delivery of {sup 99m}Tc occurs through shipping {sup 99}Mo samples to the medical facilities. The Neutron Source Facility of KIPT is the first ADS in the world operating at non-zero power and allows producing several isotopes for medical applications. The total power of the facility is about 300 kW, including ∼200 kW from fission reactions. The construction of the NSF has been already completed and operations will start during year 2016. After 20 irradiation days, {sup 82}Br, {sup 64}Cu, {sup 165}Dy, {sup 166}Ho, {sup 192}Ir, {sup 194}Ir, {sup 186}Re, {sup 188}Re and {sup 153}Sm have specific activities, larger than 100 MBq/mg without taking into account the self-shielding effect and therefore they are good candidates for production. The self-shielding effect plays an important role for all nuclide parents that have a large microscopic capture cross section. This effect reduces the reaction rate and therefore the specific activity. For instance, the analyses of molybdenum trioxide (MoO{sub 3}) irradiation samples have shown that the self-shielding effect may reduce the specific activity by a factor 1.4. The metallic form of natural molybdenum produces a higher specific activity, relative to the trioxide form. Analyses on the 6-day Ci {sup 99}Mo activity have shown that is possible to perform between 40 to 310 {sup 99m}Tc medical procedures per MoO{sub 3} irradiation week. The number of medical procedures depends on the diagnosed organ and the weight of the patient. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

COUPLING SJÖSTRAND AND FEYNMAN METHODS IN PROMPT NEUTRON DECAY CONSTANT ANALYSES

The Sjöstrand and Feynman methods have been widely applied to subcritical assemblies analyses. Within the point kinetics framework, the Sjöstrand method determines the system reactivity by the ratio between the prompt and delayed areas from pulsed neutron source experiments. In addition, the slope of the prompt area gives the prompt neutron decay constant. The latter parameter can be also obtained from fitting the Feynman curve. Consequently, the goal of this work is to combine the Sjöstrand and Feynman methods in order to find the best fitting domain for calculating the prompt neutron decay constant. The experiments and MCNP simulations performed to accomplish this goal have analyzed the KUKA subcritical assembly facility of Japan driven by 100 MeV protons.

Feynman↗
DOE OSTI2015

MONTE CARLO ANALYSES OF THE YALINA THERMAL FACILITY WITH SERPENT STEREOLITHOGRAPHY GEOMETRY MODEL

This paper analyzes the YALINA Thermal subcritical assembly of Belarus using two different Monte Carlo transport programs, SERPENT and MCNP. The MCNP model is based on combinatorial geometry and universes hierarchy, while the SERPENT model is based on Stereolithography geometry. The latter consists of unstructured triangulated surfaces defined by the normal and vertices. This geometry format is used by 3D printers and it has been created by: the CUBIT software, MATLAB scripts, and C coding. All the Monte Carlo simulations have been performed using the ENDF/B-VII.0 nuclear data library. Both MCNP and SERPENT share the same geometry specifications, which describe the facility details without using any material homogenization. Three different configurations have been studied with different number of fuel rods. The three fuel configurations use 216, 245, or 280 fuel rods, respectively. The numerical simulations show that the agreement between SERPENT and MCNP results is within few tens of pcms.

CUBIT↗