Yousry Gohar

Fusion solution to dispose of spent nuclear fuel, transuranic elements and highly enriched uranium.

Abstract The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D–T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

Monte Carlo and deterministic calculation of the Bell and Glasstone spatial correction factor

Abstract The Bell and Glasstone correction factor is used in subcritical assembly pulsed neutron source experiments to correct the spatial dependency of the measured reactivity on the detector position. The correction factor is defined as the ratio between the reactivity obtained by computer codes in criticality mode and that obtained by computer codes in source mode. In the area method, the reactivity (in dollar units) of a subcritical assembly is given by the ratio between the prompt and the delayed areas; these areas are obtained by integrating the detector reaction rate over the pulse period. This work illustrates different methods to calculate the Bell and Glasstone spatial correction factor using both Monte Carlo (MCNPX) and deterministic (PARTISN) computer codes. The different calculation methods include: (1) the one-simulation dynamic method (which has been applied by MCNPX computer simulations); (2) the two-simulation static method (which has been applied by both MCNPX and PARTISN computer simulations); (3) the one-simulation static method (which has been applied by MCNPX computer simulations). In the one-simulation dynamic method: (1) the external neutron source is time dependent; (2) the detector reaction rate is obtained from a single pulse and it is superimposed until the delayed neutron contribution reaches the asymptotic value; (3) the prompt area is obtained as the difference between the total and delayed areas. In the two-simulation static method: (1) the external neutron source is time independent; (2) the total and prompt areas are obtained by two separate computer simulations (one with and the other without delayed neutrons); (3) the delayed area is obtained as the difference between the total and prompt areas. In the one-simulation static method, first introduced in this study, the prompt and delayed areas are tallied in the same MCNPX simulation, which halves the computing time and reduces the statistical error relative to the two-simulation static method.

Characterization of lead-bismuth eutectic target material for accelerator driven transmuters

Abstract Lead–bismuth eutectic (LBE) is under consideration as a target material with high-energy protons for generating neutrons to drive actinide and fission product transmuters. A characterization has been performed to study the performance of this target material as a function of the main variables and the design selections. The characterization includes the neutron yield, the spatial energy deposition, the neutron spectrum, the beam window performance, and the target buffer requirements. The characterization has also considered high-energy deuteron particles to study the impact on the target neutronic performance. The obtained results quantify the LBE target material performance with proton or deuteron particles as a function of the target variables and selections.

Monte Carlo and deterministic computational methods for the calculation of the effective delayed neutron fraction

Abstract The effective delayed neutron fraction β eff plays an important role in kinetics and static analysis of the reactor physics experiments. It is used as reactivity unit referred to as “dollar”. Usually, it is obtained by computer simulation due to the difficulty in measuring it experimentally. In 1965, Keepin proposed a method, widely used in the literature, for the calculation of the effective delayed neutron fraction β eff . This method requires calculation of the adjoint neutron flux as a weighting function of the phase space inner products and is easy to implement by deterministic codes. With Monte Carlo codes, the solution of the adjoint neutron transport equation is much more difficult because of the continuous-energy treatment of nuclear data. Consequently, alternative methods, which do not require the explicit calculation of the adjoint neutron flux, have been proposed. In 1997, Bretscher introduced the k -ratio method for calculating the effective delayed neutron fraction; this method is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Using Monte Carlo calculation Bretscher evaluated the β eff as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as the k -ratio method). In the present work, the k -ratio method is applied by Monte Carlo (MCNPX) and deterministic (PARTISN) codes. In the latter case, the ENDF/B nuclear data library of the fuel isotopes ( 235 U and 238 U) has been processed by the NJOY code with and without the delayed neutron data to prepare multi-group WIMSD neutron libraries for the lattice physics code DRAGON, which was used to generate the PARTISN macroscopic cross sections. In recent years Meulekamp and van der Marck in 2006 and Nauchi and Kameyama in 2005 proposed new methods for the effective delayed neutron fraction calculation with only one Monte Carlo computer simulation, compared with the k -ratio method which require two criticality calculations. In this paper, the Meulekamp/Marck and Nauchi/Kameyama methods are applied for the first time by the MCNPX computer code and the results obtained by all different methods are compared.

Monte Carlo and Deterministic Neutronics Analyses of YALINA Thermal Facility and Comparison with Experimental Results

Abstract This study compares Monte Carlo and deterministic neutronics analyses of the zero-power YALINA Thermal subcritical assembly, which is located in Minsk, Belarus. The YALINA Thermal facility consists of a subcritical core that can be driven by either a californium neutron source or a deuterium-deuterium (D-D) neutron source. The californium neutron source is generated by the natural decay of 252Cf; the D-D neutron source is generated by a deuteron accelerator. The MCNPX, MONK, NJOY, DRAGON, PARTISN, and TORT computer programs have been used for calculating the neutron spectrum, the neutron flux, and the 3He(n,p) reaction rate set by californium and D-D neutron sources. These parameters have been computed in different experimental channels of the assembly for different fuel loading configurations. The MCNPX and MONK computer programs modeled the facility without any major approximation; the PARTISN and TORT computer simulations used 69 energy groups, S16 angular quadrature set, linear anisotropic scattering, and approximately 60 homogenized material zones. The results calculated by different computer programs are in good agreement; in addition, they match the 3He(n,p) reaction rate from experimental measurements obtained by californium and D-D neutron sources.

Numerical Simulation of a Completely Passive Spent Fuel Pool: Lessons Learned

As part of the design and safety analyses of the KIPT accelerator driven subcritical assembly system of Ukraine, a passive cooled spent fuel pool has been conceived, designed and analyzed numerically. The total decay power of the pool is low and The maximum heat load is 0.5 kW. Air cooling of the spent fuel pool tank through a natural convection thermo-siphon is deemed sufficient to provide sufficiently low temperatures. Natural convection of the water within the tank removes the decay heat from the fuel elements to the tank surface. The present work discusses the numerical simulations of such facility by the means of CFD. While the system has low power and it is relatively simple, it poses significant challenges for the CFD simulations. In fact the presence of two natural convection patterns is a source of numerical instability at such low power. These issues and the obtained solutions are discussed in this paper. Since the problem (the simulation of two coupled natural convection systems) is general and likely to be of significant relevance to the design of future power plants, this paper is targeted to a broader audience. Rather than the specific design the focus will be on the theoretical and the practical problems involved with this kind of simulations. The problem is analyzed theoretically and numerically. For CFD simulations, the range of meshes used ranges from 1 million points to 40 million points. Several turbulence models and wall modeling approaches have been tried and tested. Several set of simulations have been performed: sets of simplified simulations considering only the external air thermo-siphon assuming a constant heat flux at the tank wall as well as a set of simulations of the coupled system using a porous medium approach in the fuel tank. All simulations provided consistent predictions and helped confirm that the temperature within the pool is below boiling point.Copyright

Design analysis and optimization of self-cooled lithium blankets and shields

Abstract A study of self-cooled lithium blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main design parameters considered during the course of the study were the tritium breeding ratio, the blanket energy multiplication factor, the energy fraction lost to the shield, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Another study was carried out to determine materials, compositions, arrangements, and thickness of the shield zone for the reference blanket. Helium and water-cooled shields were optimized for the inboard and outboard sections of the reactor. Based on the above two studies, the reference blanket and shield configurations were developed for the ANL Tokamak Power Systems Study. The helium-cooled shield was selected for use with liquid metal blankets to reduce safety concerns related to lithium—water reactivity. This helium-cooled shield provides shielding characteristics similar to a conventional water-cooled shield. The analyses and results from these studies are the subject of this paper.

Nuclear data needs for fusion reactors

Abstract The nuclear design of fusion reactor components (e.g., first wall, blanket, shield, magnet, limiter, divertosr, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems.

BIOLOGICAL SHIELD DESIGN AND ANALYSIS OF KIPT ACCELERATOR-DRIVEN SUBCRITICAL FACILITY

Abstract Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology of Ukraine have been collaborating on the conceptual design development of an electron accelerator-driven subcritical facility. The facility will be utilized for performing basic and applied nuclear research, producing medical isotopes, and training young nuclear specialists. This paper presents the design and analyses of the biological shield performed for the top section of the facility. The neutron source driving the subcritical assembly is generated from the interaction of a 100-kW electron beam with a natural uranium target. The electron energy is in the range of 100 to 200 MeV, and it has a uniform spatial distribution. The shield design and the associated analyses are presented including different parametric studies. In the analyses, a significant effort was dedicated to the accurate prediction of the radiation dose outside the shield boundary as a function of the shield thickness without geometrical approximations or material homogenization. The MCNPX Monte Carlo code was utilized for the transport calculation of electrons, photons, and neutrons. Weight window variance-reduction techniques were introduced, and the dose equivalent outside the shield can be calculated with reasonably good statistics.

Multiplier, moderator, and reflector materials for advanced lithium–vanadium fusion blankets

Abstract The self-cooled lithium–vanadium fusion blanket concept has several attractive operational and environmental features. In this concept, liquid lithium works as the tritium breeder and coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V–4Cr–4Ti) is used as the structural material because of its superior performance relative to other alloys for this application. However, this concept has poor attenuation characteristics and energy multiplication for the DT neutrons. An advanced self-cooled lithium–vanadium fusion blanket concept has been developed to eliminate these drawbacks while maintaining all the attractive features of the conventional concept. An electrical insulator coating for the coolant channels, spectral shifter (multiplier, and moderator) and reflector were utilized in the blanket design to enhance the blanket performance. In addition, the blanket was designed to have the capability to operate at average loading conditions of 2 MW/m2 surface heat flux and 10 MW/m2 neutron wall loading. This paper assesses the spectral shifter and the reflector materials and it defines the technological requirements of this advanced blanket concept.