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This study looks for innovative methods to improve the overall performance of the U-20% Zr metallic fuel. The first solution is to swap out the helium gap for a ternary liquid metal bonded gap while the second involves minimizing the helium gap’s thickness to 0.04 mm in order to minimize its thermal resistance. The proposed solutions have been subjected to neutronic, thermal-hydraulic, and solid structure investigations, and their performance has been contrasted with that of a typical U-20% Zr metallic alloy with a 0.08 mm He-gap. According to neutronic analysis, the investigated fuel materials have almost identical neutronic performance. After using the LM bonded gap, both thermal-hydraulic and solid structure performance improved significantly. The performance of the U-20% Zr with 0.04 mm He-gap was moderate and unattractive to be used since it was deduced that its drawbacks outweighed its benefits.

The main objective of the presented work is to investigate the effectiveness of homogenization of the GFR-2400 core in simulating its neutronic characteristics. To explore and evaluate the neutronic behavior of the large scale Gas cooled Fast Reactor GFR-2400 core, two computer models (homogeneous and heterogeneous) were designed using the MCNPX code. The designed heterogeneous model has been validated by comparing its results with a previously published paper. The results of both models were compared with each other to study the effect of fuel homogeneity on the radial flux and power distribution through the core. As part of a safety analysis of the reactor core for both designs, the reactivity worth of control rods, Effective delayed neutron fraction (ß eff), Doppler constant and Depressurization effect have been analyzed and compared. The variations of the effective multiplication factor (k eff), the minor actinides concentration and the most important fission products as a function of operation time have been investigated for the presented designs.

This study discusses the effect of using 232Th instead of 238U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO2 with (Th, 235U)O2 and (Th, 233U) O2. Firstly, the infinity multiplication factor (Kinf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd2O3 and Er2O3 on the Kinf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of 232Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of 135Xe, 149Sm, and 151Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρCR), the effective delayed neutron fraction βeff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

A transmutation process of three long-lived fission products (79Se, 99Tc and 107Pd) in the experimental fast reactor JOYO is postulated. The possibility of increasing the transmutation rate utilizing the high neutron flux present in the JOYO reactor by loading neutron-moderating subassemblies in the reflector zone has been investigated. A cluster of reflector subassemblies was replaced with beryllium or zirconium hydride (ZrH1.65) moderated subassemblies. These moderated subassemblies surrounded one central test subassembly that would contain the three long-lived fission products (LLFP) simultaneous and without isotopic separation. ChainSolver 2.34 code is used to calculate the transmutation rates. In this study, the new characteristics of LLFP transmutation in a fast reactor using moderator materials were shown for future applications.

This work presents a comparison between the homogeneous and heterogeneous [seed-blanket (SB)] fuel assembly used in the VVER-1200 core. The MCNPX 2.7 code with the ENDF/B-VII.0 data library was use...

The broad goal of this work is to use the accumulated plutonium around the world in peaceful uses of energy production rather than for military purposes. This work investigates the feasibility of burning the Pu vector that is extracted from the spent fuel of light water reactor or from the nuclear warheads stockpile. MCNPX code version 2.7 has been used to design two assembly models, conventional assembly and Blanket-seed (BS) assembly. The conventional assembly is fueled with UO2 while the BS assembly is fueled with ThO2 in the blanket region and Pu vector either in the form of reactor grade plutonium (rgPu) or weapon grade plutonium (wgPu) in the seed region. Using of Th-232 as a fertile material in the blanket region prevents from more Pu breeding and enables a longer fuel burnup rate. Five different fuel models are studied to investigate the optimum concentration of Pu vector that is used in the seed region. MCNPX code is used to simulate the neutronic characteristics of the presented fuels. The multiplication factor Keff values of the suggested fuel in the BS assembly have been analyzed and compared with that obtained in the conventional assembly. Some parameters related to safety operation such as moderator temperature coefficient (MTC) and effective delayed neutron fraction (βeff) have been calculated. The Pu vector concentrations have been investigated with burnup to distinguish the effectively of the suggested models to burn how much of the used Pu in the seed region. Using of Th-232 as a fertile material in the blanket region enhance the ability of BS assembly to bred more fissile material (U-233). Therefore, the suggested fuel assemblies achieve a higher conversion ratio (CR). The radioactivity of the actinides and non-actinides elements has been investigated with burnup. Due to the poising effect of fission products on the reactivity of the reaction, it is important to investigate the concentrations of the most important fission product at the presented fuel.

The VVER-1200/AES-2006 is one of the most promising nuclear reactors for power generation. This work provides an in-depth analysis of the void fraction effect on the operation of VVER-1200 using MCNPX code. The void fraction in the VVER-1200 may be result from the loss of coolant accident (LOCA) or boiling the moderator/coolant materials. The fission multiplicities (υ), the estimated of the recoverable energy per fission (Q) and the effective delayed neutrons fraction (βeff) have been investigated at different fuel enrichment and a different void fraction. It is beneficial to calculate the βeff at the different void fraction and different enrichment, due to its impact on the reactor power change rate. The excess reactivity in the reaction can be controlled using the gadolinium dioxide or boron carbide. Increasing the void fraction decrease the burnup rate of gadolinium dioxide and boron carbide in the VVER-1200 assembly. The Doppler coefficient has a large effect on the reactivity of the reaction. Therefore, it is important to calculate the Doppler coefficient at different fuel temperature. The existence of the void fraction increases the concentration of the most hazardous radioactive nuclear waste products.

In this paper thorium fuel feasibility in large scale Gas Cooled Fast Reactor (GCFR) is investigated. The neutronics benchmark used in this study, GFR2400, corresponds to a 2400 MWth GFR concept proposed by the French CEA. MCNPX computational code is used to design a 3D heterogeneous model of the GFR2400 core. A detailed feasibility analysis of the performance of thorium fuel cycle is performed by using thorium as an alternative fertile fuel for the natural uranium vector of the reference core design. The most essential neutronic parameters characterizing the core are determined both for beginning of life (BOL) conditions as well as during burnup. Also, a three-dimensional core cycle-by-cycle simulations is performed to allow explicit characterization of the core behavior and safety-related parameters during both open and equilibrium cycles. The thorium-based core shows favorable neutronic characteristics with an acceptable control and safety parameters for both BOL and open cycle states. The depressurization reactivity effect and core expansion coefficients (axial and radial) show improvement compared to the uranium-based core. However, this improvement is compensated by the deterioration in the effective delayed neutron fraction (β-eff) and the Doppler reactivity Effect. The results of isotopic transmutation and fuel burnup confirm the capability of the core to work in both open and closed cycles and to self-recycle its own MA vector and plutonium of LWRs. However-as is the case in other fast reactors including the uranium-based GFR2400 core, the fuel cycle closure causes safety related parameters to degrade.

This work investigates the possibility of increasing the incineration rate of reactor-grade plutonium (rgPu) in the pressurized water reactor (PWR). In an attempt to reduce the production of plutonium isotopes from the conventional fuel and increase the rate of burning the accumulated plutonium, thorium-232 is suggested to be used as a fertile material in the standard fuel assembly and a suggested Blanket-seed assembly. The neutronic properties of Th-rgPu in both the standard fuel assembly and the BS assembly were analyzed and compared with the standard fuel to distinguish which of them is more effective in burning the rgPu. The performance of Th-rgPu in the standard fuel assembly and the BS assembly has been assessed by tracking the infinity multiplication factor (Kinf), the amount of the rgPu incinerated, reactor safety parameters, the most important fission products, the radioactivity of the spent fuel and the radial power distribution through the assembly. The neutronic analysis confirmed the viability of BS assembly in burning a large quantity of rgPu, maintaining the neutronic safety parameters at acceptable values.

The pursuit for an optimum burnable absorbers (BAs) takes a great interest due to their importance for regulating and control the reactivity of the reactor. MCNPX code is used to simulate the effect of Europium and Pyrex on the neutronic characteristics of PWR and compare it with the conventional burnable absorber. The thermal neutron flux and the power distribution through PWR assembly are analyzed to investigate the effect of gadolinium and europium on it. The PWR assembly is filled in a different fuel burnup process with 2% Gd 2 O 3 and 2% Eu 2 O 3 as integral burnable absorbers (IBAs). Despite the fact that the absorption cross section of gadolinium is larger than europium, the effective multiplication factor in the case of (UO 2 + Gd 2 O 3 ) IBAs rods increases more than that in the case of (UO 2 + Gd 2 O 3 ) IBAs rods. The Pyrex alloy is investigated in the Burnable poison rods (BPRs) to compare its effect on the neutronic characteristics of PWR with the conventional material such as B 4 C and (Ag-In-Cd). The effect of these materials on the Keff values has been analyzed. Due to the high absorption cross section of 135 Xe and 149 Sm resulting from the fission process, it was important to study the effect of different BAs materials on their concentration. The 239 Pu and 241 Pu concentrations resulting from the burning process have been calculated for the suggested BAs materials.

Fast reactor concept like the 2400 MW GCFR is considered a promising candidate for future sustainable and economic nuclear power systems. It is a highly innovative system with advanced geometry and fuel materials (Ceramic fuel pellets of mixed uranium-plutonium carbide within fuel pins). The active fuel region of this reactor is divided radial to two core zones namely, inner core fuel assemblies (IC), and outer core fuel assemblies (OC). In the present study Monte Carlo transport code, MCNPX is used to design heterogeneous and homogeneous models for inner and outer core fuel assembly. The deigned models have been used to simulate the neutronic behavior and fuel transmutation during open cycle operation and to study the effect of Homogeneity inside the fuel assembly. The results of heterogeneous and homogeneous models were compared with each other and good agreement was found.

This study aims to find an economic method to decrease the excess reactivity and increase the fuel cycle length. The neutronic feasibility to replace the 238U with 232Th and the traditional assembly with Blanket-Seed assembly was analyzed. MCNPX code version 2.7 was used to design a three-dimension model of PWR to simulate the neutronic characteristics of the investigated fuel types and designs. The radial thermal neutron flux and the radial thermal power distribution through the assembly have been studied. The radial thermal neutron flux and thermal power distribution provide an insight view about the proposed fuel types and designs. The effect of the proposed fuel types and design on the infinity multiplication factor (Kinf), the breeding fissile concentration, the reactor-grade plutonium (rgPu), the minor actinides (MAs), conversion ratio (CR) and the concentration of the most radioactive fission products have been analyzed and compared with that of the standard fuel. From the neutronic point of view, the results showed the economic feasibility of using 232Th in regulating the initial excess reactivity.

This paper presents an integrated study about some accident-tolerant fuels compared to the standard fuel in trying to find an alternative fuel that enhances the safety, competitiveness, and economics of commercial nuclear power. The neutronic, thermal–hydraulic and solid mechanics analysis of Uranium Nitride (UN) and Uranium Carbide (UC) were performed in a single fuel assembly of the VVER-1000 nuclear power reactor and compared with that of Uranium dioxide (UO2). MCNPX 2.7 computer code and COMSOL-Multiphysics computer software were used to make a deep analysis for the investigated fuel types. In the neutronic analysis, the fuel burn-up, the safety parameters and the power peaking factor (PPF) for the hot channel were simulated. The thermal–hydraulic and solid mechanics analyses were performed on the hot channel to make sure that the maximum temperature of the clad and the coolant would not exceed the safety limits.

This work focuses on the viability of using liquid metal bonded gap instead of the traditional helium (He)-gap through investigating its effects on the neutronic, thermal–hydraulic, and solid mechanics analyses. The shortcoming of using a He-gap is the high thermal resistance of helium, which causes overheating and much heat accumulation inside the fuel, and hence causes many problems with the fuel‘s solid mechanics behavior. To reduce both the heat accumulation and overheating inside the fuel material, LM boned gap was applied. The obtained results showed a slight difference in the various neutronic parameters, while in the thermal–hydraulic analysis, an excellent enhancement in the maximum fuel temperature happened, which reduced the heat accumulation in the fuel material. Also, in solid mechanics analysis, the enhancement that came from decreasing the fuel temperature decreased both the thermal expansion of the fuel and the von-Mises stress that acts on it.

The objective of this work is to determine which of the main three fuel types (UO2, (Th,U)O2 and MOX fuel) can burn minor actinides (MAs) effectively and at the same time sustain the main operating parameters at an acceptable level. The effect of adding 1 wt % of MAs to the main three fuel types on the infinity multiplication factor (Kinf), thermal neutron flux and the actinides’ concentration is investigated using MCNPX code version 2.7. The MAs can be used to regulate the excess reactivity of the reactor. The variation of the MAs concentration with burnup is studied to distinguish the fuel type that has the ability to burn a large amount of MAs. The transmutations of MAs in the pressurized water reactor (PWR) affect the level of radioactivity of the spent fuel. The main operating parameters are calculated to assess the possibility of burning MAs in the PWR. A comprehensive mathematical study of the results is performed to assess the performance of three fuel types after adding the MAs. Hence, determine the best fuel type. The mathematical study utilizes one of the known multi-criteria decision-making methods, the technique of order preference by similarity to an ideal solution (TOPSIS).

This research discusses the neutronic characteristics of VVER-1200 assembly fueled with five different fuel types based on thorium. These types of fuel based on mixing thorium as a fertile material with different fissile materials. The neutronic characteristics of these fuels are investigated by comparing their neutronic characteristics with the conventional uranium dioxide fuel using the MCNPX code. The objective of this study is to reduce the production of long-lived actinides, get rid of plutonium component and to improve the fuel cycle economy while maintaining acceptable values of the neutronic safety parameters such as moderator temperature coefficient, Doppler coefficient and effective delayed neutrons (β). The thorium based fuel has a more negative Doppler coefficient than uranium dioxide fuel. The moderator temperature coefficient (MTC) has been calculated for the different proposed fuels. Also, the fissile inventory ratio has been calculated at different burnup step. The use of Th-232 as a fertile material instead of U-238 in a nuclear fuel is the most promising fuel in VVER-1200 as it is the ideal solution to avoid the production of more plutonium components and long-lived minor actinides. The reactor grade plutonium accumulated in light water reactor with burnup can be recycled by mixing it with Th-232 to fuel the VVER-1200 assembly. The concentrations of Xe-135 and Sm-151 have been investigated, due to their high thermal neutron absorption cross section.

This paper aims to reach the most optimizing design of the VVER-1200 assembly to achieve a significant consumption of weapons-grade plutonium (wgPu) vector. Different assembly pitches based on the Blanket-Seed (BS) design were suggested to consume the wgPu vector and their neutronic results were compared with the results of the traditional pitch. The power mismatch problem between the Blanket rods and the Seed rods has been solved by using new design (concentric rods BS assembly). The infinity multiplication factor (Kinf) has been analyzed for the five suggested pitches. Using of Th-wgPu as a fuel in the BS assembly increases the nuclear fuel life cycle. The main related safety parameters were calculated for the investigated pitches to identify the most appropriate safety pitch. The neutronic results proved that the concentric rod BS assembly can be used to consume a large amount of wgPu vector safely.

There is an urgent need to find compact burnable BA materials as an alternative to the traditional materials (B4C and Ag–In–Cd). These materials must possess higher density, good thermophysical properties, radiation resistance, and swelling resistance. In this article, the most interesting rare earth nuclides for neutron absorption (gadolinium, erbium, dysprosium, hafanium and dysprosium) are studied and compared from the point of view of their possibilities for the control of core reactivity in PWRs to increase the cycle length. MCNPX code is used to design a three dimensional model for PWR assembly. The PWR assembly is loaded in different operation process with 2% Gd2O3 and 2% Er2O3 as integral burnable absorbers (IBAs). The effects of gadolinia and erbia on the thermal neutron flux and normalized power are investigated. The study presents that erbium is more effective IBA than gadolinium for a reactor with a very long cycle. The dysprosium titanate, dysprosium hafnate and hafnium diboride are investigated as material for a control rod in PWR assembly. The effects of these control rod materials on the multiplication factor and the concentrations of fission products are calculated. From the viewpoint of the achievable fuel discharge burnup, it was found that the suggested control rod materials achieve a satisfactory control rod worth (CRW), i.e., the CRW of the presented materials is comparable with the CRW of the standard AIC control rods.

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO2) and uranium zirconium hydride (UZrH1.6) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO2 contains 8 × 8 fuel rods while that fueled with UZrH1.6 contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH1.6 fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

The main target of this research was to estimate the possibility for improving the performance of Boiling Water Reactor (BWR) cores by using uranium thorium dioxide fuels instead of the commonly used oxide fuel. MCNPX (Monte Carlo N-Particle eXtend) computer code based on Mont Carlo method, is used to design two models for a BWR fuel bundle in a typical operating temperature and pressure conditions, the first is the oxide fuel bundle model and the second is the blanket–seed (BS) bundle model. A test case was compared with a previously published data calculated by HELIOS code and good agreement was found. The two models are used to determine the multiplication factor, pin-by-pin power distribution and thermal neutron flux distribution for both UO 2 and (UZr/ThO 2 ) fuels. BS bundle model improves the performance of BWR in different ways: (a) by increasing the energy extracted per fuel assembly, (b) by reducing the uranium ore, (c) reducing the plutonium accumulated in BWR through burn-up and continues to grow throughout the world, and (d) by increasing fuel burn-up and reactor lifetime. The spent fuel in the reactor can be recycled, and the plutonium and its isotopes can be extracted. The extracted plutonium can be used as a Mixed Oxide (MOX) fuel.

This paper presents a design of boiling water reactor BWR model using MCNPX to develop benchmarks for checking the fuel management computer code packages. MCNPX code based on Monte Carlo method, is used to design a three dimensional model for BWR fuel assembly in typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. This design is used to study the thermal neutron flux and the pin by pin power distribution through the BWR core assemblies. The fuel used in BWR core is UO 2 with three different types of enrichment (0.711%, 1.76% and 2.19%). This enrichment is distributed in such a way as to flatten the power. The effect of different enrichment values on the radial normalized power distribution is analyzed. The spent fuel in the reactor can be recycled, and plutonium and its isotopes can be extracted.

This work presents the neutronic characteristics of advanced boiling water reactor (BWR) which fueled with hydride fuel. MCNPX Computer code based on Mont Carlo Method, is used to design a three dimensional model for BWR fuel assembly in a typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. This model is used to determine attainable discharge burnup, pin-by-pin power distribution, axial power distribution and reactivity effect of voiding. It is found that hydride fuel bundle design can be simplified by reducing the volume of water in-between the fuel bundles. The increase in the number of fuel rods per given core volume enables an increase in the BWR power density relative to oxide fueled core design.

This article discusses the effect of different Integral Burnable Absorber numbers (IBAs) on the neutronic characteristics of a Pressurised Water Reactor (PWR) to provide a suitable safety level. MCNPX code version 2.6 was used to design a three dimensional model for PWR assembly. The designed model has been validated by comparing the output values of the infinity multiplication factor (Kinf) with a previously published value. The designed MCNPX model was used to analyse the radial distribution of thermal neutrons and the power through PWR assembly with and without IBA. Due to the high absorption cross-section of gadolinium, it has been used as a burnable absorber material in the IBA rods. The gadolinium isotopes suppressed the power in the regions where they were distributed. The existence of IBA rods has a large effect on the Kinf. This effect decreases gradually with burnup due to the degradation of gadolinium. The gadolinium isotopes degradation was analysed with burnup. Different numbers of IBA rods were investigated to optimise the suitable number that can be used in the PWR assembly. The reactivity has been investigated at different numbers of IBAs. The gadolinium effect on the concentration of 135Xe and 149Sm resulting from the fission process was analysed.