Reactor physics (Physics)

Civilian marine reactors face a unique set of design challenges in addition to the usual irradiation and thermal-hydraulic limits affecting all reactors. These include requirements for a small core size, long core lifetime, a 20% cap on fissile loading, and limitations on the use of soluble boron. One way to achieve higher burnup/longer core life is to alter the neutron spectrum by changing the hydrogen-to-heavy-metal ratio, thus increasing the conversion of fertile isotopes in the fuel. In this reactor physics study, we optimize the two-dimensional lattice geometry of a 333 MWth soluble-boron-free marine PWR for 18% 235 U enriched micro-heterogeneous ThO 2 -UO 2 duplex fuel and 15% 235 U enriched homogeneously mixed all-UO 2 fuel. We consider two types of coolant: H 2 O and mixed 80% D 2 O + 20% H 2 O. We aim to observe in which spectrum discharge burnup is maximized in order to improve uranium utilization, while satisfying the constraint on moderator temperature coefficient. It is observed that higher discharge burnup for the candidate fuels is achievable by using either a wetter lattice or a much drier lattice than normal, while epithermal lattices are distinctly inferior performers. The thorium-rich duplex fuel exhibits higher discharge burnup potential than the all-UO 2 fuel for all moderation regimes for both coolants. The candidate fuels exhibit higher initial reactivity and discharge burnup with the mixed D 2 O-H 2 O coolant than with the H 2 O coolant in the under-moderated regime, whereas these values are lower for the D 2 O-H 2 O coolant in the over-moderated regime.

We present a comprehensive framework for a vector-tensor extension of General Relativity termed Vortical Geometrodynamics Theory (VGT). By introducing an effective coupling between a dynamical background vector field and stellar/plasma vorticity , we demonstrate the emergence of non-reciprocal metric terms analogous to the shift vectors in Alcubierre-type metrics. We further explore the transition from sub-luminal reactive propulsion to a super-luminal "warp" regime through a non-linear vacuum potential. This transition allows for the formation of a metastable vacuum bubble characterized by reduced effective mass and localized negative effective pressure. Finally, we propose a falsifiable experimental setup utilizing a tri-toroidal MHD configuration to detect these anisotropic metric signatures.

Nuclear reactors are commonly described as stable systems operating under controlled critical conditions. This work challenges that view by showing that reactor stability is not an inherent property of nuclear fission, but the result of constraints imposed on fundamentally unstable neutron dynamics. By analyzing reactor kinetics, delayed neutron effects, and feedback mechanisms within a unified framework, it is demonstrated that controllability arises from timescale separation rather than intrinsic stability. The system remains operable only within a narrow regime where physical processes evolve slowly enough for control to act. This perspective reframes nuclear reactors as dynamically constrained systems operating at the edge of instability, with important implications for reactor control and safety analysis.

The nuclear fusion threshold, Coulomb barrier energy, critical density, and chain reaction conditions are derived entirely from the universal closure condition C = g², without nuclear equation-of-state models or empirical fitting. The barrier energy Ec = (α · αs/2) mpc² is expressed purely in terms of the electromagnetic and strong coupling constants with no free parameters — a result with no equivalent in standard fusion physics. Nuclear density, the D-T viability ratio of 20.6, and a minimum ignition radius of 3–4mm follow from the same geometry.

The question of whether the cosmos has a finite beginning or an infinite past remains a central issue in both philosophy and contemporary cosmology. This study investigates the metaphysical and physical plausibility of a past-eternal universe by integrating philosophical analysis with conceptual examination of modern cosmological models. Employing qualitative philosophical methodology and critical analysis of theoretical physics, the paper evaluates temporal ontology (Atheory and B-theory), the implications of special relativity, and paradoxes associated with actual infinities, including Thomson’s Lamp and a formalized infinite traversal analogy. The analysis argues that an actual infinite temporal regress generates logical and metaphysical difficulties, particularly under a dynamic conception of temporal becoming. Furthermore, the paper examines the Friedmann–Lemaître–Robertson–Walker model, the Hawking–Penrose singularity theorems, and the Borde–Guth–Vilenkin theorem, assessing their implications for past completeness. While certain speculative models attempt to avoid a temporal boundary, they remain mathematically constrained and philosophically contentious. The study concludes that both metaphysical reasoning and contemporary cosmology provide convergent support for the thesis that the cosmos is temporally finite and had a definite beginning.

Reactor physics methods at ANL were traditionally developed primarily for fast-spectrum liquid metal reactor (LMR) design and analysis. The success of systematically derived nodal methods for light water reactor (LWR) analysis, to which A. F. Henry made essential contributions, motivated the development at ANL of analogous methods applicable to fast reactors and other systems. Henry's research and technical guidance 1-5 strongly influenced these methods. In this paper, we review the small part of Henry's legacy embodied in the ANL nodal methods for static, depletion, and transient reactor physics analyses.

Modern physics already admits that mass and energy are interchangeable, E=mc^2⇔m=E/c^2, and that geometry, not force, ultimately governs free-fall motion through the geodesic equation d^2 x^μ/dτ^2+Γ_αβ^μ u^α u^β=0. My program simply takes this to its logical engineering conclusion: use compact reactors and mass–energy conversion (m=E/c^2) to generate extreme energy densities, then channel that energy into engineered field configurations—superconducting coils, chiral ferromagnetic–piezo stacks, and vacuum-structured metamaterials—to realize a synthetic stress–energy tensor T_μν^(("engine" )). Through Einstein’s equation G_μν=8πGT_μν/c^4, this engineered T_μνbecomes a tool for sculpting Γ_αβ^μitself, so that the vehicle interior remains in permanent free-fall (no g-force, mass-independent motion), while its trajectory relative to Earth reflects an effective geometric “takeoff” and reentry rather than a Newtonian, thrust-limited flight. The mass–energy inversion is not a slogan; it is the fuel source for a geometry-first propulsion architecture. The missing piece between “reactor energy” and “specific spacetime geometry” is a 4D holographic navigation system that treats curvature, not thrust, as the control variable: it measures the local metric and CST phase, solves for the target Γ_αβ^μthat carries the ship from A to B at the desired effective velocity with bounded tidal forces, and then drives the geometry core—superconducting coils, chiral torsion stacks, and metamaterials—so that a^μ=-Γ_αβ^μ u^α u^βreplaces F = ma as the law of motion inside the ship. If a craft possessed a geometry-based propulsion system with a 4-dimensional holographic navigation layer that controls its local spacetime metric Γ_αβ^μ, then instantaneous directional changes, high-acceleration maneuvers without g-force, drag-free atmospheric traversal, and effective FTL motion would be natural outcomes of geodesic control, rather than thrust or aerodynamic dynamics. Under this model, reported UAP flight characteristics are consistent with engineered control of curvature rather than conventional propulsion in theory.

The use of optical manipulations that changes the mechanical properties of light has been crucial across multiple fields. These techniques are being used in various optics and phonics applications. By following the same physical theory we can manipulate an atomic sized particle in the quantum domain. We have proposed an asymmetric configuration model to reverse quantum pulling force on a Xenon atom at a certain angle resulting from the interaction energy of the matter wave (Helium) near a copper substrate. Several full-wave quantum simulations and on this quantum mechanical configuration with two counter-propagating plane matter waves at different angles were performed. By utilizing a solution of time independent Schrödinger equation and stress tensor formula we calculated the force on the quantum particle. It implies physical adsorption scatterer phenomena that can be used for the development of modern quantum applications such as particle trapping and inertial confinement fusion.

Konsep bahan bakar wallpaper dikembangkan untuk menurunkan temperatur puncak bahan bakar dengan memberikan modifikasi yang baru pada bahan bakar pebble dimana material grafit dengan radius tertentu ditempatkan di pusat bahan bakar sehingga zona bahan bakar hanya tersebar di bagian permukaan dalam pebble. Penelitian ini difokuskan pada perhitungan dan analisis parameter neutronika reaktor yang disebut koefisien reaktivitas temperatur teras. Reaktor pebble bed HTR-10 dipilih sebagai objek reaktor dalam penelitian. Seluruh perhitungan dilakukan dengan menggunakan program transport Monte Carlo MCNPX 2.6.0 dan pustaka data nuklir energi kontinu ENDF/B-VI. Dalam studi ini, koefisien reaktivitas temperatur dihitung dalam tiga tahap perhitungan, yakni perhitungan koefisien reaktivitas temperatur bahan bakar, koefisien reaktivitas temperatur moderator dan kombinasi koefisien reaktivitas temperatur bahan bakar dan moderator. Hasil analisis memperlihatkan bahwa kenaikan temperatur bahan bakar ...

Threshold reactions induced by reactor fast neutrons are well recognized. The geological samples consisting of five mafic (gabro) and four ultramafic (serpentinite) rocks were collected from El-Sid gold district, Central Eastern Desert, Egypt. The concentration of Fe and Ni were determined in these samples by activation analysis with fast neutrons, using the threshold reactions 54 Fe(n,p) 54 Mn and 58 Ni(n,p) 58 Co respectively. The fast neutron flux was determined using the 92 Mo(n,2n) 92m Nb reaction. The determined concentrations of Fe and Ni were checked by determining them in the GSJ JB-1 reference material using the same (n,p) reactions. Good agreements were obtained between the measured and recommended values. The concentrations of Fe were also determined by (n,γ) reactions in the samples using, the k 0 -NAA method. There are good agreements between the determined concentrations from (n,p) and (n,γ) reactions.

This paper explores the binding energy curve, provides computational examples for specific nuclei (e.g., helium-4, carbon-12, iron-56, anduranium-238), and details the semi-empirical mass formula (SEMF) as an approximation tool for nuclearmasses.

The binding energy per nucleon is a critical metric in nuclear physics that elucidates the stability of atomic nuclei and underpins phenomena such as nuclear fusion and fission.

Using established parameters and computational methods, this analysis demonstrates the peak stability near iron-56 and the implications for energy-releasing reactions.

Data are drawn from theoretical models and experimental validations, with citations to peer-reviewed sources. This framework might support further research in nuclear astrophysics and energy applications.—

VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK. Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choosed because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction and Power Peaking Factor (PPF), and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly ...

This study presents calculation of gamma attenuation parameters for in-house developed Polyboron and Borated polyethylene shielding materials and locally available shielding materials like Pure polyethylene, water and ordinary concrete which can be used as potential radiation shielding materials. The study was carried out using GEANT4 which is a software arrangement, a combination of tools which can be used to simulate the particle transportation accurately through matter. The results were compared and in good match have been observed with the database of XCOM taking photon energy range 40 keV-20 MeV. For each energy 10000 particles were transported through the slab. Relaxation-lengths have been also calculated from obtaining linear attenuation coefficients. The obtained result shows Mass attenuation coefficient, Linear attenuation coefficient and relaxation length density of the shielding materials on the photon energy. The total Mass attenuation coefficient , Linear attenuation coefficient , Half Value Layer (HVL) and Tenth Value Layer (TVL) of each material for some mostly used gamma radiation sources has been calculated. The study's findings can be used to understand how effective shielding is in reactor shielding, containment buildings, nuclear spent fuel casks and other shielding scenarios. These materials are engineered solutions intended to maximize safety, utility, and cost-effectiveness in radiation situations; they are more than merely barriers. As radiation technologies advance, it is crucial that they be continuously studied, optimized, and combined.

The use of general descriptive names, registered names, trademarks, etc. in this publication does not imply, even in the absence of a specific statement, that such names are exempt from the relevant protective laws and regulations and therefore free for general use.

The concept of Integral Small Modular Reactor (SMR) isn't new but it seems that the proper time for using this idea has been coming. According to the International Atomic Energy Agency (IAEA), the reactors with electrical power lower than 300 MW have been defined as small reactors, although SMRs are categorized by this fact that more advantages and design features are attained when intentionally make reactors small. In fact, these reactors use their size as advantage to attain more design purposes. The scalability, modularity, improved safety characteristics and more important than other, lower up-front cost of the SMRs, offer great advantages over large common nuclear power plants. According to the IAEA reports there are many interests all over the world to move toward using of these kinds of reactors. There are many different type of SMRs under various stages of design, licensing and construction. Nowadays, there are many initiatives to use thorium in nuclear reactors and fuel cycles. Thorium is three times more abundance than Uranium, however, despite of several initiatives and researches on Th-232 utilization in many types of reactors, this fuel hasn't been commercialized yet. Most of The SMRs have been designed to have long cycle, so they must use a lot of poisoning material in the beginning of the cycle. Taking in the account that Thorium can be used as a absorber in the beginning of the cycle and also be used as a fertile material during the cycle, it seems to be a good option to use mixed (U-Th)O 2 as SMR's fuel. This paper briefly is going to review the research about Thorium utilization as a nuclear fuel and the possibilities of using mixed (U-Th)O 2 fuel as an alternative option for SMRs fuel. The Korean System Integrated Modular Advanced Reactor (SMART) categorized as SMR that has received its standard design approval, was chosen as reference core for our calculations. The calculations have been performed by MCNPX code as a well-known Monte Carlo code. Geometry and all materials were kept the same as the SMART core, and the only variable was the fuel pin material, in which we use several mass proportion of uranium and thorium but keeping the enrichment in U-235, lower than 5 wt%. The results confirm that it's possible to use mixed (U-Th)O 2 with lower burnable absorber at the beginning of the cycle and have a longer burnup cycle.

Small Modular Reactors (SMRs) with their excellent safety and economic features will be in high demand in the near future. Most SMR designs have longer burn-up cycle length with more fuel enrichment and smaller core size in comparison to the large conventional nuclear reactors. The small size of these reactors causes more neutron leakage (less core radius results in a higher area to volume ratio and more relative leakage). This feature of SMRs causes high values of maximum Power Peaking Factors (PPFs) through the core, so optimizing the safety parameters is of high necessity. Also, long burn-up cycle length needs a high initial excess reactivity, which results into use of some materials and methods to control this high excess reactivity. One of these methods is using a high number of Integral Fuel Burnable Absorber (IFBA) rods.In the present designs of IFBA rods, usually some amounts of fuel with lower enrichment are used at the top and bottom parts of the IFBA rods (known as cutback fuel) to flatten the axial PPFs. The small size of the SMRs (using a lower number of FAs) helps to have much less possible radial loading patterns (in comparison to the large reactors) and provides the possibility to optimize the axial variations in amounts of cutback fuel in IFBA rods simultaneously. Accordingly, the best axial and radial loading pattern according to the objective functions could be achieved. At the present work, the main goal is to optimize radial core loading pattern and axial variations of cutback fuel lengths at the IFBA rods of an SMR simultaneously using a multi-objective neutronic and thermal-hydraulic fitness function. The multi-objective fitness function includes burn-up cycle length, Minimum Departure from Nucleate Boiling (MDNBR), maximum and average radial and axial PPFs during the entire cycle lengths. The Cuckoo Optimization Algorithm (COA) as a new robust metaheuristic algorithm with high convergence speed and global optima achievement has been used. For the thermo-neutronic calculation, DRPACO package consists of the coupling system of DRAGON/ PARCS/COBRA codes have been used. Finally, the results of SMR core axial and radial loading pattern optimization using COA presents a core configuration with improvement in the core safety and economic parameters in comparison to the reference SMR core.

This paper analyzes the behavior of the entropy of a nuclear fuel rod under deep burnup conditions, beyond standard operational ranges, reaching up to 60 years. The evolution of the neutron source distribution in a pressurized water reactor (PWR) fuel pin was analyzed using the Monte Carlo method and Shannon information entropy. To maintain proper statistics, a novel scaling method was developed, adjusting the neutron population based on the fission rate. By integrating reactor physics with information theory, this work aimed at the deeper understanding of nuclear fuel behavior under extreme burnup conditions. The results show a “U-shaped” entropy evolution: an initial decrease due to self-organization, followed by stabilization and eventual increase due to degradation. A minimum entropy state is reached after approximately 45 years of pin operation, showing a steady-state condition with no entropy change. This point may indicate a physical limit for fuel utilization. Beyond this point, entropy rises, reflecting system degradation and lower energy efficiency. The results show that entropy analysis can provide valuable insights into fuel behavior and operational limits. The proposed scaling method may also serve to control a Monte Carlo simulation, especially for the analysis of long-life reactors.

Simulation of criticality accident transients offers the ability to confirm understanding of critical configurations, bound accident scenarios and aid comprehensive emergency planning. Computational ability to recreate excursion power histories in fissile solution is sought due to the predominance of fissile solutions in process criticality accidents [1]. Applicable solution transient physics methodologies are developed for neutron kinetics, heat transfer, and radiolytic gas transport. This work is the first step in validating the use of COMSOL [2] as a flexible and powerful computational tool to simulate criticality excursions. Previously, characterization of excursion power history profiles has been limited to hand-calculations and simulation tools with 1D/0-D physics treatments for specific geometries [3, 4, 5, 6]. This work demonstrates COMSOL’s capability for an accurate, generalized approach to excursion simulations for nuclear safety applications.

In this paper, a study has been carried out to design a fuel pebble for utilizing thorium in a molten salt cooled and graphite moderated and reflected Innovative High Temperature Reactor (IHTR20MWth). An attempt is made to maximize the conversion of 232 Th to 233 U under the constrain of maintaining negative reactivity coefficient like Fuel Temperature Coefficient (FTC), Coolant Void Reactivity (CVRC) coefficient. This is done by first studying the effect of heavy metal loading and 233 U weight fraction on the infinite multiplication factor by varying the radius of fuel kernel and packing density. The variation of conversion ratio and reactivity coefficients with fuel depletion has also been studied.

EXIFON 2.0 code has been used to calculate the (α,n) and (α,2n) reactions of 122Sb for the cyclotron production of 123,124I which have application in positron emission tomography imaging. For both reactions, the threshold for the reaction obtained from the evaluation compared well with measured values. Results obtained shows good agreement with measurements from EXFOR. It was observed that there is a need for optical model contribution to the pre-equilibrium mechanism of EXIFON at 16MeV for (α,2n) and 27MeV for (α,n) reactions. However, the results show good agreement with available measured data at lower energy level.

Today, small modular reactors have received considerable attention in various countries. The ABV reactor is a PWR small modular reactor that has various applications. This reactor has been used silumin metal fuel with a 16.5% enrichment. In the present work, the efficiency of the conventional UO2 fuel with enrichment of less than 10% to be used as the main fuel of ABV reactor has been investigated, and four different patterns for the reactor core have been proposed. To perform the calculations, the ABV reactor is modeled using the PARCS neutronic code and the RELAP5 thermohydraulic code. Finally, using computational codes for the proposed patterns of the reactor core, various quantities including reactor cycle length, reactivity, burnup, power distribution, fuel, coolant temperature distribution, and feedback coefficients have been calculated.

Different types of nuclear fuels are being used in nuclear reactors. For selecting a suitable nuclear fuel, it is required to conduct neutronics analysis of fuels in a specific reactor condition. This study investigates the pellet models of three different material combinations of nuclear fuel where light water with two different densities has been used as neutron moderators. The fuel elements used are UO 2 , UN, and UC 2. Analysis is performed using the open-source nuclear code OpenMC-0.13.3 which is based upon the Monte-Carlo method, incorporating both the ENDF/B-VIII.0 and JEFF-3.3 nuclear data libraries. 10,000 particles in 100 batches were taken for the simulation. Values of Effective Multiplication Factor (keff) for each fuel element were calculated using both JEFF-3.3 and ENDF/B-VIII libraries. The dependency on the enrichment of k eff value for the pin cell of UO 2 fuel was studied. Finally, for a total of three models in our analysis, which includes the UO 2 fuel pellet moderated by light water, the neutron flux spectrum was produced. Time dependency of fission reaction rate and k eff were studied. Analysis of depletion was done for U-235 for each model and the behavior of two most significant neutron poisons Xe-135 and Sm-149 were observed.

https://www.osti.gov/biblio/4716426
A cross section library has been set-up to be used for burn-up calculations
in the presence of Gd. The available cross section data are briefly discussed, a new
reduced energy scheme is suggested and its validity is tested; the values resulted for
the materials of interest ate also given. A few assumptions are discussed which allow
to simplify the model to be used in bum-up calculations

Thorium does not have intrinsic fissile content unlike uranium. 232 Th has nearly three times thermal absorption cross section compared to 238 U and hence requires much larger externally fed fissile content compared to uranium based fuel. These factors give a permanent economic competitive edge to uranium. Thus thorium is not inducted in any significant measure in present day power reactors, despite the fact that thorium is three times more abundant in the earth's crust than uranium. Uranium reserves vary from country to country and there is also difficulty in having equitable distribution of uranium. Thus when 235 U would get exhausted, perhaps much sooner in countries having limited uranium reserve, there will be a need to switch over from the today's open fuel cycle programme based on 235 U feed to closed fuel cycle based on Pu feed. At that stage thorium and (depleted) uranium would become equal candidates to form the fertile base. All economic considerations would have to be readdressed. The size and growth of the nuclear power programme based on closed fuel cycle would be dependent on maximizing the fissile conversion rate in those reactors. In this paper we reemphasize the principles and the details of the thermal reactor concept 'A Thorium Breeder Reactor' (ATBR), in which the use of PuO 2 seeded thoria fuel is found to give excellent core characteristics like two years cycle length with nearly zero control maneuvers, fairly high seed output to input ratio and intrinsically safe reactivity coefficients [Jagannathan V, Ganesan S, Karthikeyan R. Sensitivity studies for a thorium breeder reactor design with the nuclear data libraries of WIMS library update project. In:

There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., 235 U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either depleted uranium or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia to come. This has to be done by innovating means of large scale fertile to fissile conversion and then using the man-made fissile material for sustenance as well as growth of fission nuclear power. This paper attempts to give a broad picture of the available options and the challenges in realizing the theoretical possibilities.

The researchers analyzed the essential and beneficial heavy elements in soil that improve plant growth, development and other functions in this study using k 0­ – INAA technique. The use of k 0 – Instrumental Neutron Activation Analysis technique has become a veritable tool for multi-element analysis. This method has been applied in this present study to assess essential mineral elements and beneficial heavy elements that supports plants growth of fifteen samples collected from Safana, Dan Musa and Kankia, LGAs of Katsina State. The samples were properly prepared together with the standard reference material (NIST Coal Fly Ash 1633b) and were irradiated at a thermal power level of 31.0 kW corresponding to a neutron flux setting of 2.5 x 10 11 cm -2 s -1 and 5.0 x 10 11 cm -2 s -1 for the outer and inner irradiation channels respectively using the Nigeria Research Reactor – 1 (NIRR – 1) at the Centre for Energy Research and Training, Zaria. After irradiation, samples were counted usi...

OpenNode is a sophisticated reactor core simulation tool, meticulously designed for educational purposes. Its main function is to solve the complex partial differential equation governing multi-group neutron diffusion in a Cartesian geometry. OpenNode relies on the second- and fourth-order Nodal Expansion Method (NEM) to provide accurate solutions to three-dimensional steady-state problems, encompassing both the direct and adjoint modes. At the heart of this method, the neutron balance equation serves as the fundamental framework for calculating the nodal neutron flux and the effective multiplication factor, enabling the complex evaluation of neutron flux distributions within individual nodes. The NEM methodology, carefully programmed in Fortran-90 in OpenNode, meticulously decomposes the reactor core into nodes of reasonable size. This approach balances computational efficiency and accuracy, achieving outstanding precision even when using a reduced number of computational meshes. To further increase the number of equations while maintaining the balance between unknowns, OpenNode judiciously incorporates two essential approximations: the Weighted Residual Method (WRM) and the Quadratic Transverse Leakage (QTL) approximation. OpenNode transcends simple computational prowess by adopting an intuitive Graphical User Interface (GUI), meticulously developed by combining Python 3, Qt Designer, and Blender software. This Python-powered graphical interface offers users a seamless, user-friendly experience, facilitating tasks such as data entry, execution of complex simulations, calculation of results, data visualization, 3D design of nuclear core geometry, and in-depth analysis of results. One of OpenNode’s key features is its innate flexibility, enabling it to solve problems in both 2D and 3D Cartesian geometries. This adaptability makes it a versatile tool suitable for a wide range of research and power reactor applications, offering users the ability to model and design reactors with unrivaled ease. To substantiate OpenNode’s reliability, comprehensive criticality calculations are carried out in various scenarios, including single-cell verification, BIBLIS, KOEBERG cores in two dimensions (2D), and IAEA cores in three dimensions (3D). The results clearly demonstrate OpenNode’s robustness and reliability in multidimensional criticality analysis. In particular, OpenNode adheres to an open-source philosophy, making it readily available for download. This commitment to open access fosters collaboration and paves the way for wider adoption within the education and research communities, nurturing a thriving ecosystem of shared knowledge and exploration.

The researchers analyzed the essential and beneficial heavy elements in soil that improve plant growth, development and other functions in this study using k 0­ – INAA technique. The use of k 0 – Instrumental Neutron Activation Analysis technique has become a veritable tool for multi-element analysis. This method has been applied in this present study to assess essential mineral elements and beneficial heavy elements that supports plants growth of fifteen samples collected from Safana, Dan Musa and Kankia, LGAs of Katsina State. The samples were properly prepared together with the standard reference material (NIST Coal Fly Ash 1633b) and were irradiated at a thermal power level of 31.0 kW corresponding to a neutron flux setting of 2.5 x 10 11 cm -2 s -1 and 5.0 x 10 11 cm -2 s -1 for the outer and inner irradiation channels respectively using the Nigeria Research Reactor – 1 (NIRR – 1) at the Centre for Energy Research and Training, Zaria. After irradiation, samples were counted usi...

In order to ascertain the level of flux variation in one of the inner irradiation channels of the Nigeria Research Reactor-1 (NIRR-1), the irradiation container used for routine activation analysis was employed with copper wires as flux monitors. Measurements were carried out with these wires arranged in axial direction to determine the thermal neutron flux at selected positions using absolute foil activation method. Our results show that there exists a slight flux variation from one position to another ranging from (4.57 ± 0.24) x 10 11 to (5.20 ± 0.20) x 10 11 cm-2 s-1. Individual foil shows slight flux variation from one position to another in the same irradiation container but they all pointed toward a level of stability in spite of the recent installation of the cadmium lined irradiation channel. The values obtained in this work are in good agreement with the previously measured value of (5.14 ± 0.24) x 10 11 cm-2 s-1 after commissioning of NIRR-1. This shows that the cadmium lined installation does not affect the flux stability. In order to improve the accuracy of neutron activation analysis (NAA) using NIRR-1 facility, there is need for flux corrections to be made by miniature neutron source reactor (MNSR) users during NAA particularly long irradiation, where more than six samples are irradiated simultaneously in the same container.

Recently, one of the outer irradiation channels of the Nigeria Research Reactor-1 (NIRR-1), which is Miniature Neutron Source Reactor (MNSR) was cadmium-lined. In order to ascertain level of flux disturbance on the inner and the remaining outer channels and also to establish flux profiles in sample container as experienced by samples for Neutron Activation Analysis (NAA), a foil activation technique was used to determine the radial and axial neutron flux distribution of the reactor. Measurements were carried out with copper wire flux monitors arranged in radial and axial directions in a typical irradiation container used for routine NAA in NIRR-1. Our results show that for a preset neutron flux of 5.0 Â 10 11 cm À2 s À1 the axial and radial neutron flux of NIRR-1 ranges from 4.47 Â 10 11 to 5.16 Â 10 11 cm À2 s À1 and 4.80 Â 10 11 to 5.55 Â 10 11 cm À2 s À1 respectively, with mean value of 4.79 Â 10 11 cm À2 s À1 for axial and 4.99 Â 10 11 cm À2 s À1 for radial. The mean value for each position was compared with our previously measured values after commissioning of NIRR-1 (Jonah et al., 2005) and the one done recently at the Ghana MNSR (Abrefah et al., 2010). The values all pointed towards a level of consistency in variation despite the recent installation of cadmium-lined irradiation channel in NIRR-1. This is an indication that the installed cadmium line did not affect NIRR-1 flux stability. However, our individual foils shows slight flux variation from one foil position to another in the same irradiation container as against the claims of MNSR manufacturers and some users for perfect flux stability across the container. There is therefore the need for flux corrections to be made by MNSR users during NAA particularly for samples in the axial position. This can be achieved by sandwiching flux monitors between samples during irradiations.

ABSTRAK ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM. Telah dilakukan analisis simulasi pengaturan posisi batang-batang kendali untuk melanjutkan operasi reaktor daya eksperimental (RDE) paska scram setelah beroperasi pada periode waktu tertentu. Pengendalian reaktivitas pada reaktor RDE yang akan dibangun di Indonesia dengan rujukan high temperature gas reactor (HTR) 10 MWt, dilakukan dengan 10 pasang batang-batang kendali atau control rod (CR). Apabila terrjadi kondisi abnormal maka CR secara otomatis akan jatuh tersisip ke dalam reflektor reaktor sehingga reaktor scram dan berada pada kondisi subkritis. Untuk melanjutkan operasi reaktor pasca scram diperlukan analisis terkait pengaruh reaktivitas negatif dari Xenon dan suhu. Pada makalah ini disajikan hasil simulasi yang dilakukan untuk penentuan posisi CR paling optimum untuk melanjutkan operasi reaktor, menggunakan simulator PCTRAN-HTR. Simulasi dilakukan pada vari...

For muon-catalyzed fusion to be of practical interest, a very efficient means of producing muons must be found. We describe here some schemes for producing muons that may be more energy efficient than any heretofore proposed. There are, in particular, some potential advantages of creating muons from collisions of high-energy tritons confined in a magnetic mirror configuration. If one could catalyze 200 fusions per muon and employ a uranium blanket that would multiply the neutron energy by a factor of ten, one might produce electricity with an overall plant efficiency (ratio of electric energy produced to nucleazc energy released) approaching 30%.

This is a preprint of a paper intended for publication in a Journal or proceedings. Since changes may be made before publication, this preprint is made available with the cnderstandtng that it will not be cited or reproduced without the permission of the author.

For muon-catalyzed fusion to be of practical interest, a very efficient means of producing muons must be found. We describe here some schemes for producing muons that may be more energy efficient than any heretofore proposed. There are, in particular, some potential advantages of creating muons from collisions of high-energy tritons confined in a magnetic mirror configuration. If one could catalyze 200 fusions per muon and employ a uranium blanket that would multiply the neutron energy by a factor of ten, one might produce electricity with an overall plant efficiency (ratio of electric energy produced to nucleazc energy released) approaching 30%.

The Center for Nuclear Fuel Technology (PTBBN) in BATAN needs to strengthen its strategy in light of the enhanced tasks of BATAN as the Implementing Organization under the National Research Agency (BRIN) to carry out research, development, assessment and application as well as inventions and innovations in an integrated manner. Following the experience and lessons learned in the past 3 decades and change of workforce generation, a strategic assessment is performed to evaluate and strengthen the fuel program for light water reactor (LWR) to support the development of a Prototype Commercial Scale Nuclear Power Plant, as outlined in the National Research Priority 2020-2024. Collaboration with the national and international stakeholders is essential to further enhance the knowledge and know-how of fuel technology and to improve the quality of the fuel program in areas with gaps and proprietary nature (e.g. fuel specifications, qualification requirements, benchmarking). The study reviews the current status and challenges of the fuel program in Indonesia and revisits the business process of PTBBN. Mapping of existing and planned capabilities as well as potential collaboration with relevant stakeholders is performed. Consultation is being conducted with the national nuclear experts. Fuel strategies, experience and standard practice in other countries are examined. The results show that mastery of nuclear fuel technology requires strong government support in terms of manpower, materials and funding, and collaboration with international partners. Preliminary considerations for a strategy framework for the development of LWR fuel prototype are presented. Considerations for domestic fuel fabrication should take into account the economics and strategic value, and further studies must be performed on feasibility, risks, cost and benefit.

In this work, the reaction cross-section for neutron-induced reactions on 127 I isotope was calculated using EXIFON code in the energy range of incident particle from 0 MeV to 30 MeV. The code is based on an analytical model for statistical multistep direct and multistep compound reactions (SMD/SMC model). In order to see the effect of nuclear structure on cross sections, the calculation was done using nuclear shell structure effect and without considering shell structure effect. Obtained results and statistical analysis showed that shell structure effect does not give significant changes to the cross-section at considered energy ranges. This shows that EXIFON code is a good tool for investigation of nuclear reaction cross section and is useful in the production of the radioisotopes of Iodine, Antimony and tellurium of high purity and in an efficient manner using cyclotron or nuclear reactors, these isotopes have potential application for field of medical science especially for diagnostics and therapeutic purposes.

In this work, the reaction cross-section for neutron-induced reactions on 127 I isotope was calculated using EXIFON code in the energy range of incident particle from 0 MeV to 30 MeV. The code is based on an analytical model for statistical multistep direct and multistep compound reactions (SMD/SMC model). In order to see the effect of nuclear structure on cross sections, the calculation was done using nuclear shell structure effect and without considering shell structure effect. Obtained results and statistical analysis showed that shell structure effect does not give significant changes to the cross-section at considered energy ranges. This shows that EXIFON code is a good tool for investigation of nuclear reaction cross section and is useful in the production of the radioisotopes of Iodine, Antimony and tellurium of high purity and in an efficient manner using cyclotron or nuclear reactors, these isotopes have potential application for field of medical science especially for diagnostics and therapeutic purposes.

For the utilization of the NIRR-1 to include the k 0-standardization method, the following neutron spectrum parameters in one inner and one outer irradiation channels were determined using the "Cd-ratio for multi-monitor method": The thermal to epithermal flux ratio, f, and the epithermal flux shape factor, α. These parameters determined in the inner irradiation channel A 1 ,are: α =-0.046 ± 0.005 and f = 18.4 ± 0.34 and for the outer irradiation channel A 2 α = + 0.024 ± 0.002 and f = 49.5 ± 0.96. The results compare well with reported values obtained from other reactor facilities with similar core configuration such as the Slowpoke and Miniature Neutron Source Reactor facilities available in literature.

The Center for Nuclear Fuel Technology (PTBBN) in BATAN needs to strengthen its strategy in light of the enhanced tasks of BATAN as the Implementing Organization under the National Research Agency (BRIN) to carry out research, development, assessment and application as well as inventions and innovations in an integrated manner. Following the experience and lessons learned in the past 3 decades and change of workforce generation, a strategic assessment is performed to evaluate and strengthen the fuel program for light water reactor (LWR) to support the development of a Prototype Commercial Scale Nuclear Power Plant, as outlined in the National Research Priority 2020-2024. Collaboration with the national and international stakeholders is essential to further enhance the knowledge and know-how of fuel technology and to improve the quality of the fuel program in areas with gaps and proprietary nature (e.g. fuel specifications, qualification requirements, benchmarking). The study reviews the current status and challenges of the fuel program in Indonesia and revisits the business process of PTBBN. Mapping of existing and planned capabilities as well as potential collaboration with relevant stakeholders is performed. Consultation is being conducted with the national nuclear experts. Fuel strategies, experience and standard practice in other countries are examined. The results show that mastery of nuclear fuel technology requires strong government support in terms of manpower, materials and funding, and collaboration with international partners. Preliminary considerations for a strategy framework for the development of LWR fuel prototype are presented. Considerations for domestic fuel fabrication should take into account the economics and strategic value, and further studies must be performed on feasibility, risks, cost and benefit.