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In the 1999 Kargil War, India defended its territory from a Pakistani incursion but--in a departure from its historical behavior and standing war plans--chose not to expand the war with counter-attacks into Pakistan. Many observers attribute this restraint to nuclear deterrence, since India and Pakistan had become declared nuclear powers just a year earlier. In fact, India's restraint was rooted not in deterrence, but specific strategic conditions. Those conditions no longer apply--and in a future conflict India may be encouraged to take especially risky and escalatory wartime action, which would pose an unprecedented test for nuclear deterrence. [ABSTRACT FROM AUTHOR]

Accurate assessment of uncertainties in cross-section data is crucial for reliable nuclear reactor simulations and safety analyses. In this study, we focus on the interpolation procedure of the partial derivatives (PD) cross-section model used to evaluate nodal parameters from pregenerated multigroup libraries. Our primary objective is to develop a systematic methodology that enables prediction of the incurred errors in the cross-section model, leading to the development of optimal case matrices, more efficient cross-section models, and informed case matrix construction for reactor types lacking substantial data and experience. We make progress toward this objective through a rigorous analytic error analysis enabled by the derivation of error expressions and bounds for the PD model based on the discovery that the method is a form of Lagrange interpolation. Our investigations reveal distinct outcomes depending on the chosen cross-section functionalizations, achieved by identifying the sources of error. These error sources are found to include interpolation error, which is always present, and model form error, which is a property of the supplied case matrix. We show that simply increasing grid refinement through the addition of branches may not always lead to decreased cross-section errors, particularly in cases where the model form error predominantly contributes to the total error. We present numerical results and verification in a companion paper, showing these different error characteristics for various cross-section functionalizations. Although applied to current light water reactor environments, our methodology offers a means for advanced reactor analysts to develop case matrices with quantified error levels, advancing the goal of a general methodology for robust two-step reactor analysis. Future work includes exploring different lattice types and functionalizations, extending reactivity bounds to multilattice problems, and investigating historical effects within the macroscopic depletion model. [ABSTRACT FROM AUTHOR]

The technology of nuclear reactors is evolving rapidly, driven by the pursuit of more powerful and efficient systems. The Small Lead-based Reactor (SLR) represents an advanced nuclear reactor design that holds great promise for delivering enhanced power and efficiency. In the context of the SLR's fuel rod, the high-speed coolant flow within the reactor can induce vibration, potentially causing fretting wear and damage to the cladding. This study utilized the Burgreen correlation to establish an equivalence relationship between water and Lead-Bismuth Eutectic (LBE). An experiment was then conducted to simulate axial flow-induced vibration (FIV) of a simply supported fuel rod, employing an equivalent water loop. Flow-induced vibration characteristics in both the time and frequency domains were investigated at different flow speeds. The experimental results revealed a positive correlation between flow velocity and amplitude. The fuel rod exhibited low-frequency vibrations with a random pattern, registering frequencies around 14 Hz. These experimental findings can be leveraged to enhance the accuracy of numerical simulations for axial flow-induced vibration (FIV) of the fuel rod. Subsequently, the revised numerical model was applied to simulate FIV in the LBE environment using computational fluid dynamics (CFD), and the numerical results were found to be in agreement with the experimental data. [ABSTRACT FROM AUTHOR]

The electrochemical corrosion potential (ECP) data of nickel-base alloy 82 weld metal (Alloy 82) in both O2 and H2O2 environments were measured. Although Alloy 82 is used at various welds in boiling water reactors (BWRs), the ECP data of Alloy 82 under BWR conditions are limited. The ECP data measured in this study revealed that Alloy 82 had higher ECPs than 304 and 316 L stainless steels and nickel-base alloy 182 weld metal (Alloy182) at low oxidant concentrations in both O2 and H2O2 environments. The ECPs of Alloy 82 calculated with an analytical ECP model that was developed in this study agreed with measured values in the cases that it was acceptable for the model to have an error range of ± 0.1 V in the tested range of O2 and H2O2 concentrations. The possible reason for the ECPs of Alloy 82 to be higher than those other materials at low oxidant concentrations is given as follows: the anodic polarization curve of Alloy 82 exhibits a lower current density at low potentials because of its higher Ni and Cr contents. The developed ECP model validated by the ECP data will contribute to evaluations of the effectiveness of mitigation technologies for stress corrosion cracking environments. [ABSTRACT FROM AUTHOR]

AbstractThe aging process or flow-induced vibration of reactor cores may lead to increased mechanical vibrations, affecting the reliability of in-core sensors and necessitating a robust solution for robust field reconstruction. This work tackles the challenges of reconstructing multiphysics fields from sparse and movable measurements by introducing an advanced framework that integrates various machine learning models with Voronoi tessellation. Our approach, building upon the Voronoi tessellation-assisted Convolutional Neural Network (VCNN), expands the capabilities to include a wider array of neural network architectures such as Convolutional Neural Networks (CNNs), Fourier Neural Operator (FNO), Dilated ResNet Encode-Process-Decode (DilResNet), Dilated Convolution Neural Operator (DCNO), Galerkin Transformer (GT), U-shaped Neural Operator (UNO), and Multiwavelet-based Operator (MWT). The effectiveness of these models is evaluated and validated through numerical tests based on the International Atomic Energy Agency benchmark, particularly noting average relative errors below 5% and 10% in the $${L_2}$$L2 norm and $${L^\infty }$$L∞ norm, respectively, within a 5-cm amplitude around sensor nominal locations. The developed software toolkit encapsulates these architectures, providing a versatile option for nuclear engineers to reconstruct different types of physical fields efficiently. [ABSTRACT FROM AUTHOR]

This investigation has significant applications in the fields of mechanical, industrial, and biomedical engineering, biosciences, and technology, particularly in areas such as blood pumping, drug delivery, glass-fiber production, paper production, and nuclear reactors. This study explores the numerical analysis of a mathematical model of entropy generation in convective magnetohydrodynamic (MHD) flow of Williamson nanofluid model over a stretching sheet with Cattaneo–Christov heat flux and suction/injection. Furthermore, heat generation, viscous dissipation, Joule heating, radiation, and chemical reaction are considered. The appropriate transformations are used to transform the nonlinear partial differential equations (PDEs) into nonlinear ordinary differential equations (ODEs). The transformed equations are solved numerically using bvp5c MATLAB package. The effects of the involved physical parameters on the flow quantities and the entropy generation are presented and discussed in detail with figures and tables. It is observed that the thermal field is enhanced by increasing the Eckert number, the Joule heating, and the thermal relaxation parameters. Also, the concentration field is observed to be a decreasing function of the augmented chemical reaction parameter. Further, the increasing magnetic field and the Williamson parameter led to an increase in the skin friction coefficient. Moreover, the entropy generation increases due to an increase in the diffusion parameter and the Brinkman number. [ABSTRACT FROM AUTHOR]

This article focuses on apartheid South Africa’s nuclear sector and the regime’s attempts to cooperate with overseas energy companies to provide financial investments and crucial technologies. It highlights the connectedness of the South Africans in the global nuclear marketplace and their ability to secure technical support from Western states during the Cold War between 1953 and 1976. The article analyses the parallel negotiations with French and German firms to engage with the regime in sharing of sensitive knowledge and bargaining lucrative contracts. Using newly discovered archival records it is shown that French-German competition was at a boiling point at least twice in the 1970s. Moreover, the regime in Pretoria managed to garner enough support in the nuclear field to further the growth of its domestic industry, ultimately being capable of enriching uranium and obtaining a turnkey nuclear power reactor. The white minority government in Pretoria however failed to position itself as an important uranium supplier on a global scale, because of sanctions targeting domestic racial apartheid policies and a more robust international non-proliferation regime towards the 1980s. [ABSTRACT FROM AUTHOR]

In a part of the historical reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the condition of such fracture pattern change. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of ℇz/ℇθ = 0.5–1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio ℇz/ℇθ >0.75 and a concurrent hoop strain < 1.7% at the failure instant. [ABSTRACT FROM AUTHOR]

Plate-type fuel elements, widely used in high flux reactor (HFR), may experience collapse or vibration under high coolant flow rates. The study examined the behaviors of curved fuel plates under varying coolant velocities. The deformations of curved plates with varying curvature radii and flat-plate assemblies were computed to facilitate comparison. The fluid domain was simulated using Ansys Fluent, while the solid domain was analyzed using Ansys Mechanical. The study employed a partitioned, two-way, and explicit coupling strategy. The simulation results have shown that the critical velocity for assemblies of curved plates is 20 m/s, whereas for flat plates it is 6 m/s. As the curvature radius or width increases, the plates become increasingly susceptible to collapse. The critical velocity is associated with the nonlinearity between the maximum plate deflections and the square of flow velocities. It also involves plate deflections that extend from the leading edges to the entire length of the fuel at the critical velocity. Under axial flow conditions, the deflection directions of multiple curved plates exhibit regular patterns. The leading edges of the plates adjacent to the walls deflect towards the walls, while the adjacent plates deflect in opposite directions. [ABSTRACT FROM AUTHOR]

The survivability of the domestic nuclear power industry depends on the cost-competitiveness of safe and secure nuclear power generation. Advanced reactor design concepts aim to have increased safety margins over traditional large light water reactors (LWRs). With increased safety margins comes the potential for a corresponding decrease in off-site risk to the general public from a hypothetical release of radioactivity due to sabotage or theft. Without sacrificing safety or security, advanced reactor designers may be able to achieve operational cost improvements over current LWRs in part by designing less burdensome physical protection systems (PPSs) and by replacing on-site response forces with off-site response forces. To accommodate these developments, the U.S. Nuclear Regulatory Commission is drafting new rulemakings for physical security when licensing through the current frameworks in 10 CFR 50 or 10 CFR 52 along with drafting an entirely new licensing framework: 10 CFR 53. A novel technology-inclusive consequence-informed methodology for the selection of the optimal licensing path for the design of PPSs at advanced fixed-site commercial nuclear power facilities is presented herein. This methodology proposes integrating security considerations at the beginning of a reactor facility design effort to streamline the licensing process. Off-site total effective dose equivalents at the exclusion area and low population zone boundaries were identified as the key metrics when determining a design's most appropriate licensing path that in turn affects the design requirements placed upon the PPS. Given these metrics, source-term generation of potential adversary-induced physics-based sabotage actions utilizing severe accident modeling software and off-site plume dispersal modeling were identified as appropriate for determining siting constraints, potential target sets for hypothetical sabotage events, and their subsequent off-site dose consequences. The methodology proposes using the consequence results from the sabotage modeling, in combination with desired cost-saving PPS characteristics, to help inform the licensing path selection. Once a licensing path is chosen, the methodology utilizes the Design and Evaluation Process Outline to evaluate an effective PPS following the licensing requirements placed on the facility. This paper also presents examples of hypothetical commercial nuclear power facilities with varying consequence levels and demonstrations of how to select the optimal licensing pathways for each. [ABSTRACT FROM AUTHOR]

Recently declassified material and other information that has never before appeared in the public domain allow the authors to explain some of the workings of the Nuclear Emergency Support Team (NEST)—often one of the first units to respond whenever there is a nuclear incident, whether it involves a nuclear reactor or a nuclear weapon. Long the subject of mystique, NEST is often depicted on screen as a secretive government unit with highly specialized capabilities and harrowing missions. The reality is at once more mundane and more remarkable. Formed in the 1970s in response to a spate of nuclear blackmail attempts, NEST has been at the center of every major nuclear event from the accident at Three Mile Island to the disaster at Fukushima. Other operations, unknown to the public, are described here for perhaps the first time. Historical accounts provide a glimpse into the breadth of the organization's missions, from neutralizing terrorist nuclear devices to responding to nuclear reactor accidents. The diversity of NEST's missions and the uniqueness of its scientific capabilities set the unit apart as a national asset. [ABSTRACT FROM AUTHOR]

The inverse reactor period α is a fundamental mode eigenvalue of the α-mode nonlinear Boltzmann eigenvalue equation that considers delayed neutron contributions. Thus far, several Monte Carlo methods, including the α-k, weight balancing, and transition rate matrix methods, have been developed to calculate α. This study presents a new Monte Carlo method for predicting α by using the derivatives of the k-eigenvalue with respect to α. Formulae are derived to calculate the first and second derivatives using the differential operator sampling method. The key feature of the new proposed method is its ability to estimate the uncertainty of the predicted α by considering the uncertainty of the k-eigenvalue and its derivatives with respect to α. [ABSTRACT FROM AUTHOR]

A submersible annular linear induction pump (ALIP) design with an outer diameter of 66.8 mm with appropriate materials is developed for circulating molten lead and alkali liquid metals of sodium and sodium-potassium-78 (NaK-78) alloy in test loops at temperatures up to 500°C. These loops investigate the compatibility of these liquid coolants with nuclear fuel and structure materials to support the development of advanced, Generation IV nuclear reactors. The present ALIP, which employs high-temperature ceramic-insulated coil wires and Hiperco-50 center core and stators, fits in Type 316 stainless steel, 2.5-in. standard schedule 5 pipe. This pipe, considered for the riser tube of the Versatile Test Reactor (VTR) in-pile test cartridge loop, has an inner diameter of 68.8 mm permitting 1.0-mm radial clearance for the present ALIP. An improved equivalent circuit model (ECM) is developed to analyze the performance of the present ALIP design. The accuracy of the model predictions is successfully validated using reported experimental measurements by other investigators for a low liquid sodium flow ALIP at 200°C and 330°C. The improved ECM calculates the performance characteristics of the present ALIP design and investigates the effects of varying the terminal voltage, current frequency, winding wire diameter, center core length, width of the liquid flow annulus, and working fluid properties and temperature on the pump operation. For circulating molten lead, the calculated peak efficiency of the present ALIP design of 6.7% occurs at a flow rate of 9.5 kg/s and pumping pressure of 263 kPa. The calculated peak efficiency for circulating liquid sodium is much higher, 26.3%, and occurs at a lower flow rate of 2.2 kg/s but a higher pumping pressure of 364 kPa. The calculated peak efficiency for circulating NaK-78 (23%) is lower than for sodium and occurs at a lower flow rate and pumping pressure of 1.9 kg/s and 310 kPa, respectively. [ABSTRACT FROM AUTHOR]

Extreme benchmarks of 10 or more places for the point kinetics equations for time-dependent nuclear reactor power transients are rare. Therefore, to establish an extreme benchmark, we employ a Taylor series (TS) with continuous analytical continuation to solve the ordinary differential equations of point kinetics including feedback. Nonlinear Wynn-epsilon convergence acceleration confirms the highly precise solutions for neutron and precursor densities. Through adaptive partitioning of time intervals, the proposed Converged Accelerated Taylor Series, or CATS algorithm in double precision, automatically performs successive mesh refinement to obtain high-precision initial conditions for each subinterval, with the intent to reduce propagation error. Confirmation of 10 to 12 places comes from comparison to the BEFD (Backward Euler Finite Difference) algorithm in quadruple precision also developed by the author. We report benchmark results for common cases found in the literature including step, ramp, zigzag, and sinusoidal prescribed reactivity insertions and insertions with nonlinear adiabatic Doppler feedback. We also establish a suite of new prescribed reactivity insertions and insertions with feedback, based on reactivities with Taylor series representations as suggested by the CATS algorithm. [ABSTRACT FROM AUTHOR]

Several preliminary conceptual designs of nuclear thermal rocket reactor cores are presented that use tin-bonded monolithic ceramic [mononitride (UN), monocarbide (UC), and uranium dioxide (UO2)] fuel plates or pins with molybdenum-tungsten alloy clad. Neutron moderation is provided by a block of Be metal or composite materials using metal hydrides such as ZrH1.6 or YH1.6 with different matrices (MgO or Be). Mainly high-assay low-enriched uranium is considered, but highly enriched uranium is also assessed for a few configurations. Nominal core thermal power is 300 MW corresponding to about 66 kN (15 klbf) of thrust, and with minimal modifications, 500 MW may be possible (25 klbf of thrust). Depending on the configurations, the amount of 235U needed for criticality is 30 to 90 kg, and reactor weight is 2.5 to 3.8 tonnes. [ABSTRACT FROM AUTHOR]

The total fission energy released in a criticality accident involving fissile solution boiling tends to be high because the relatively high fission power continues during boiling. Simulating fission power change correctly during boiling seems essential to estimate the total fission energy. Fission power during boiling changes depending on fissile concentration and volume as the solution evaporates. In this study, we investigated the effect of concentration and volume change on estimated total fission energy for a long time of boiling. We introduced a model calculating the evaporation of fissile solution into the modified quasi-steady-state method to simulate power change during boiling. Three CRAC experiments and the Idaho Chemical Processing Plant (ICPP) criticality accident in 1959 were analyzed. As a result, the calculated energy considering concentration and volume change during boiling reproduced the measured energy well. [ABSTRACT FROM AUTHOR]

To validate the nuclear data in JENDL-4.0 and obtain information to improve their accuracy, the analysis of the measured nuclide inventories of the high-burnup PWR UO2 and BWR MOX fuels in the REBUS program was performed by using MVP-BURN with the nuclear library and burnup chain based on JENDL-4.0, and the ratios of the calculated and measured inventories (C/Es) were compared with the previous results obtained with JENDL-3.2 or JENDL-3.3. As a result, an improvement in the accuracy of the neutron cross-sections and cumulative fission yields (CFYs) in JENDL-4.0 was confirmed for the inventories of 236U, 238Pu, 239Pu, 241Pu, 242Pu, 241Am, 243Cm, 244Cm, 133Cs, 145Nd, 148Sm, 149Sm, 151Sm, and 153Eu of the PWR UO2 fuel and 239Pu, 241Pu, 242Pu, 241Am, 243Am, 243Cm, 244Cm, 246Cm, 134Cs, 144Ce, 145Nd, 147Sm, and 154Eu of the BWR MOX fuel. Based on the C/Es obtained with JENDL-4.0, the directions to improve the neutron cross-sections and CFYs were tentatively proposed. They included the neutron capture cross-sections of 234U, 236U, 237Np, 238Pu, 241Am, 242Cm, 245Cm, 143Nd, 147Pm, 154Eu, and 155Eu, the fission cross-sections of 243Cm and 245Cm, and the CFYs of 144Nd, 147Pm, and 155Eu. [ABSTRACT FROM AUTHOR]

For all the nuclear reactor systems, quantitative assessment of the accident management (AM) effects against long-term external hazards became one of the essential issues after the lesson learned from the Fukushima Daiichi Nuclear Power Plant accident. However, the influence from the safety systems' stochastic and dynamic shifting between multiple working states, which is related to the interaction with the adjacent components/systems in general, has not been accounted for yet. Therefore, this research aims to develop a dynamic probability risk assessment tool considering repairable multi-component interdependency for investigating the AM influences on the multi-state safety systems under long-term external hazards. Based on the newly proposed methodology in this research via integrating the Petri net (PN) model with the continuous Markov chain Monte Carlo (CMMC) method, a framework applying PN-CMMC methodology to a severe accident analysis code, SPECTRA, had been originally constructed. Different AM influences on the multi-state decay heat removal systems against long-term volcanic ashfall were also quantitatively confirmed, indicating that halving the repairing time is more influential in suppressing the core damage frequency than doubling the number of adjacent electricity support systems. Therefore, the PN-CMMC-SPECTRA framework can further assess the uncharted dynamic multi-state concerns, leading to a safer AM strategy. [ABSTRACT FROM AUTHOR]

Boron carbide (B4C)–stainless steel (SS) eutectic reaction behavior is one of the most important issues in the core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). In this study, the immersion experiments using B4C pellets with molten SS were conducted to evaluate the CDA sequences such as contact event of solid B4C with degraded core materials including SS at very high temperature. The immersion experiment aims at understanding the kinetic behavior of solid B4C–liquid SS reaction based on the reduced thickness of B4C pellet after the experiment in the temperature ranges from 1763 to 1943 K, which is higher than the temperature of solid B4C–solid SS reaction. Based on the kinetic consideration of the reaction rate constants for solid B4C–liquid SS reaction, it was found that similar temperature dependency was identified between solid B4C–liquid SS and solid B4C–solid SS. Besides, the reaction rate constants of solid B4C–liquid SS were smaller than those of solid B4C–solid SS extrapolated in higher temperature region by two or more orders of magnitude due to two different evaluation method for B4C side/SS side. It was confirmed that this difference was reasonable through the consideration of previous reaction tests in solid–solid contact for B4C side/SS side. [ABSTRACT FROM AUTHOR]

Following the severe accident at the Fukushima Daiichi Nuclear Power Plant in 2011, the revised nuclear safety regulation in Japan requires continuous safety improvement and states that PRA methods that reflect the latest knowledge should be used in activities related to continuous safety improvement. In this context, the construction of PRA models for the digital RPS (DRPS) has been addressed as an important issue within the Working Group on Risk Assessment (WGRISK) of the OECD/NEA, and several studies have been conducted. And there are challenges in aligning them with the conventional probabilistic risk assessment methodology. In a previous study, the authors developed a simultaneous differential equation describing the relationship between state transitions and state probabilities based on Markov state transition diagrams to calculate them numerically. However, the analytical method for uncertainty analysis commonly used in conventional PRA evaluations is not explicitly presented. The purpose of this study is to provide a methodology for more accurate evaluation of core damage frequency in nuclear power plants equipped with digital RPS, taking into account the uncertainties, and to contribute to the continuous improvement of safety. [ABSTRACT FROM AUTHOR]

In the decommissioning of the Fukushima Daiichi Nuclear Power Station, understanding the distribution of radioactive substances and dose-equivalent rates is crucial to develop detailed decontamination plans and minimize worker exposure. In this study, we remotely visualized radioactive hotspots and dose-equivalent rate distribution in Unit 1 reactor building of the station using a Mecanum wheel robot equipped with a Compton camera, simultaneous localization and mapping device, and survey meter. We successfully visualized high-concentration radioactive hotspots on the U-shaped piping of the drywell humidity control system and the atmospheric control piping in the ceiling in front of the transverse in-core probe room. Furthermore, the hotspot location was identified in three dimensions using the Compton camera used to analyze the atmospheric control piping. By simultaneously analyzing the dose-equivalent rate data acquired by the survey meter and the hotspot locations visualized by the Compton camera, it was confirmed that the hotspots caused elevated dose-equivalent rates in the surrounding area. In the future, if this robotic system is used in unexplored areas, such as the upper floors of reactor buildings, it can provide information about the locations of radioactive hotspots and the distribution of dose-equivalent rates. [ABSTRACT FROM AUTHOR]

The highly loaded design method of helium compressors can effectively solve the difficulty in compressing helium in High Temperature Gas-cooled Reactors (HTGR). But it also causes obviously different attack angle characteristics of blade surface loads in a highly loaded helium compressor compared to air compressors. This difference inevitably affects separation characteristics and flow loss within the compressor. In the current study, the effects of highly loaded design methods and changes in attack angle on the separation characteristics of the compressor cascade are analyzed by applying a numerical simulation method first. Then, the influence of Mach number on the loss characteristics of the cascade for a highly loaded helium compressor is systematically analyzed. Finally, the effect of differences in the material properties of working fluid on the separation characteristics is discussed. The results indicate that the proportion of secondary flow loss to the total loss in highly loaded compressor cascades is 2.46 times larger than that in conventionally loaded ones. While the properties of working fluid have an effect on the performance of the compressor cascade, their effects on the weight factor of vortex loss are highly limited. [ABSTRACT FROM AUTHOR]

For its applications in nuclear reactors, food processing, chemical engineering, water emulsions and thermal power generating systems, the significance of irregular heat source and Arrhenius energy on electro-magnetohydrodynamic hybrid nanofluid flow over a rotating stretchable disk with nonlinear radiation have been investigated. The flow problem has been modeled utilizing the modified Buongiorno model and the thermophysical characteristics of water-based Cu − F e 3 O 4 hybrid nanoliquid. Effects like passive control of nanoparticles, hydrodynamic slip and convective boundary conditions are also heeded to boost the realistic nature of this work. Further, engineering quantities like moment coefficient and pumping efficiency of the disk are also elucidated which boosts the novelty of this research work. The modeled governing equations are transmuted into a system of first-order ODEs, with the help of apposite similarity transformations, which are then numerically resolved using the finite-difference based bvp5c algorithm. It is noticed that per unit increase in the electric parameter decreases the skin friction coefficient by 41.75% and increases the heat transfer rate by 15.31%. It is also observed that the entrainment velocity is directly proportional to the changes in electric field parameter and is inversely proportional to the changes in volume fraction of copper and magnetite nanoparticles. [ABSTRACT FROM AUTHOR]

Film boiling occurs when an overheated sample is rapidly plunged into the liquid. The study of film boiling is paramount in many engineering applications, like, cooling of nuclear reactor fuel rods, regenerative cooling of rockets, and wet treatment of blast slag furnaces. The objective of the present study is to explore the influence of boric acid at various concentrations (0.5%, 1%, 3%, and 5% vol.) and subcooling temperature on film boiling heat transfer using stainless steel rod (304 L). We performed vertical immersion quenching experiments in subcooled (50 °C) and saturation temperature at atmospheric conditions. Under saturated conditions, the sample quenched in deionized water took 73 s to cool down to 100 °C, while in the subcooled condition, the sample only took 13 s. At both saturated and subcooled conditions, 1% vol. boric acid solution accelerated the quenching rate and facilitated lowest film boiling time than other solutions considered. Conversely, film boiling time is higher when the sample is quenched in deionized water with 5% vol. of boric acid. The coolant temperature and additive concentration influence the film boiling time and Leidenfrost temperature, with a local build-up of boric acid concentration at the liquid–vapor interface being the governing mechanism. [ABSTRACT FROM AUTHOR]

This study evaluated the gamma and neutron-shielding parameters, electron and charge particles' interaction properties of three different stainless steel-based alloys (2507SS, 304SS and AISI1018) with densities of 7.82, 7.81 and 7.82, respectively for deployment in nuclear reactor construction. Gamma shielding parameters were evaluated with Phy-X/PSD, Py-MLBUF and GRASP computer programs within the energy range of 0.2 MeV to 15 MeV. The relative deviations (RD %) between two of these computational platforms were calculated. Fast Neutrons Effective Removal Cross-Sections (FNRCS) and Macroscopic Removal cross-section (MRCS) values for fast neutrons were calculated with Phy-X/PSD and MRCS software, while thermal/fast neutron attenuation factors were calculated with NGCals software. The continuous-slowing-down approximation (CSDA) range and total stopping power (TSP) values of H+ and He++ ions were estimated with SRIM Monte Carlo code in a wide energy range of 1–20 MeV. The projectile range of electrons was computed with ESTAR NIST software within the kinetic energy of 0.01–1000 MeV. Gamma-ray, thermal neutron and fast neutron transmission factors (TFs) values were calculated for all samples for a range of well-known energies. From the obtained results, all evaluated parameters were dependent on the composition of shielding materials, the type of radiation and the photon energy of the radiation. The maximum MAC values obtained from the three alloy samples are 0.1505 cm2/g for 2507SS, 0.1453 cm2/g for 304SS and 0.1459 cm2/g for AISI1018 at 0.02 MeV. These results projected 2507SS as a better shielding alloy when compared with other investigated alloy samples. Considering the closeness of densities of investigated alloy samples the shielding and interaction properties of all alloy samples gave excellent and similar results implying that all investigated alloys will effectively serve as a radiation-shielding material in a nuclear reactor. [ABSTRACT FROM AUTHOR]

This work describes the research of high-fidelity multiphysics models for the MegaPower nuclear reactor, a megawatt-level heat pipe reactor. Combining the Monte Carlo neutronics model, the heat pipe analysis model, the fuel analysis model, and the thermoelasticity model produces the Multi-Physics Coupling code for Heat pipe nuclear reactors (MPCH) code platform. Using the heat pipe analysis model, a database of heat pipes is generated to save computing costs. Comparison is made among four calculating modes with differing degrees of coupling. It was discovered that the thermal expansion effect reduces core reactivity by 537 ± 11 pcm and the temperature feedback coefficient by 61%. With the incorporation of the heat pipe module, a temperature difference arises between the wall of heat pipes, which can reach a maximum value of 80 K at steady state. Simultaneously, the global fuel rod temperature difference increases from 34 K (under the assumption of uniform heat pipe wall temperature) to 93 K, and the monolith temperature variance increases from 34 to 108 K. At the periphery of the monolith, the increased temperature variation causes a monolithic stress of 188.6 MPa. To further investigate the safety of the reactor, three-heat-pipe-failure scenarios are evaluated. The heat pipe analysis model reveals that a single heat pipe failure results in a monolith peak temperature of 1046 K, giving a maximum monolith stress of 237 MPa. The maximum monolith stresses and temperatures for the two-heat-pipe-failure scenario and the three-heat pipe-failure scenario are 330 MPa/1128 K and 471 MPa/1233 K, respectively. In steady-state operation, the stresses exceed the yield tensile strength (131MPa) whereas those generated by the failure of three heat pipes exceed the ultimate tensile strength (345 MPa) in high temperature. These results illustrate the necessity of including coupled multiphysics models into the design and safety evaluation of innovative nuclear reactors. [ABSTRACT FROM AUTHOR]

Composition of deuterium (heavy) water in light water was drastically reduced to one-half by one stage of a newly invented device that can create a nonequilibrium vapor state driven by a dynamic system. Vapor density of heavy and light water is arbitrarily controlled. This system works at 20–60°C, even under atmospheric pressure. The mechanism should equally work even for tritium water, as tritium-contaminated water is a problem in the treatment of nuclear reactors as well as fusion reactors. [ABSTRACT FROM AUTHOR]

The typical initiating events of a core disruptive accident (CDA) in small sodium-cooled fast reactors (SFR) are evaluated with the computational code, SAS4A. CDA is one of the hypothetical events in the safety assessment of the SFR, as the SFR has the potential to experience power excursion due to the sodium voiding and the core degradation. To improve the safety of future SFRs, the development of SFRs with low void reactivity has been promoted. Small SFRs can have a negative void coefficient of reactivity. The analysis of the CDA event sequence in small SFRs is valuable for the investigation of the reactor characteristics for the future R&D of SFRs. The event progression during ULOF and UTOP, which are the typical initiating events of CDA, is investigated through the numerical analysis. The event progression of these accidents in the low void reactivity reactor is found to be slow due to the effective insertion of the negative reactivity feedback and the absence of significant positive reactivity insertion. No power excursion occurs in the initiating phase. In ULOF, the cladding melt and motion behavior becomes more important for the evaluation of the event progression due to its positive reactivity. [ABSTRACT FROM AUTHOR]

In this review, a highly pertinent and debatable issue in the maritime industry has been addressed. Although fracture mechanics theories have expanded substantially in recent decades, researchers of the present era in the marine industry still pay little attention in incorporating fracture characteristics into account while designing the structural components. Besides the fundamental strength characterization methodologies, which include establishing the ultimate strength, hardness and impact characteristics, the authors firmly advocate that the fracture parameters of the marine-grade materials must be carefully considered during structural design phase, decidedly for weldments of similar and distinct materials. Nonetheless, these types of paradigms are an inevitable part in prominent sectors, viz., nuclear reactors, pipeline industry and space missions. Owing to this, fracture mechanical investigations performed in marine fields incorporating the standard fracture parameters like stress intensity factor (K), J-integral (J), etc. are thoroughly discussed in this review. The application of different specimen types, viz., compact tension (CT) specimens, single edge notched bend (SENB) and single edge notched tension (SENT) specimens, is also succinctly summarized. The shortcomings of the experimental studies are suggested, and a thorough discussion is also done regarding the potential application of fracture mechanical characterization on marine grade materials and their weldments. [ABSTRACT FROM AUTHOR]

A large amount of cesium hydroxide (CsOH) is generated during a light water reactor severe accident (SA) and transported through leaky parts to the environment. During that process, some CsOH may interact with oxidized structural materials and change their physicochemical properties. Accurate examination of this interaction is required by source term analysis to derive consistent and appropriate source term transport models, i.e., for SA, decommissioning, and dismantling work of a nuclear reactor. To obtain detailed interaction characteristics, in this study CsOH was exposed to Fe3O4/Fe2O3 and Cr2O3 under a simulated SA environment over a wide temperature range, from 300°C up to 1050°C. As a result, Cs2FeO4, CsFeO2, and Cs2CrO4 were observed at respective temperatures. Cs2FeO4 is stable only at low temperatures and decomposes to form CsFeO2 at about 591°C. However, both Cs2FeO4 and CsFeO2 could react with Cr2O3 to form more stable Cs2CrO4, which melts at 957°C and then completely evaporates at higher temperatures. [ABSTRACT FROM AUTHOR]

When supercritical fluids absorb heat energy through channels, their thermophysical properties rapidly change, resulting in an enhanced, deteriorated, or normal heat transfer phenomenon. Understanding the phenomena of heat transfer is essential for applications involving supercritical fluids, particularly nuclear power generation. As a result, the Nusselt number correlation is developed and used to characterize the heat transfer performance under a variety of operating conditions, geometries, and flow directions. Unfortunately, for supercritical fluid heat transfer, there are now over 50 Nusselt number correlations, which create difficulties to comprehend all of the Nusselt number correlations due to their complex structures and distinctive formulations of the modifying factors. Therefore, this review article is devoted to providing a comprehensive yet succinct overview of the key components of the majority of supercritical Nusselt number correlations. The supercritical properties of water, carbon dioxide, and helium are briefly introduced, taking supercritical carbon dioxide as an example. The potential use of supercritical fluid in engineering applications, such as Generation IV nuclear reactors, waste heat recovery, and concentrated solar power, is presented. The origin and properties of the variants of the Reynolds number, the Prandtl number, and the reference temperature modified for the Nusselt number correlation are categorized and examined. [ABSTRACT FROM AUTHOR]

The growing threat of climate change and global tensions resulting in energy shortages has led many countries to reconsider the merit of nuclear energy to meet national emissions reduction targets and provide domestic energy security. Australia's unique nuclear experience has led to a prohibition on building nuclear reactors, distinctive from other G20 countries. With Australia's energy transition underway, an exploration of the discourse surrounding nuclear energy in the national debate is timely to understand the barriers and opportunities for nuclear power in our energy future. This article investigates how nuclear energy is framed in media and by debate stakeholders, how this potentially constraints the energy source, and the influence of an increasingly carbon-conscious world on the debate. A critical discourse analysis of nuclear energy in Australia and semistructured interviews are used to understand the discourse landscape. The study finds that the discursive practises of politically conservative social actors engaged in the debate significantly shape how nuclear energy is perceived and received by the public, contributing to a polarised and politicised debate. It highlights the significance of climate change and energy security discourses in sustaining media salience whilst exposing the exclusionary nature of select nuclear energy rhetoric which limit the debate. Further research into public perceptions and discourses around nuclear energy in Australia is required given the need to decarbonise. [ABSTRACT FROM AUTHOR]

The Idaho National Laboratory (INL) is a government research facility that has been conducting scientific research on nuclear energy and military applications for over 70 years. INL has made significant contributions to the nuclear energy industry, including generating the first usable electricity from nuclear power and developing nuclear propulsion systems for Navy submarines and aircraft carriers. However, there is also a dark side to the INL, as it has been responsible for worker health issues, pollution, and radioactive waste. The article highlights concerns about worker safety, hazardous conditions, and retaliation against those who report safety concerns. [Extracted from the article]

The investigation of the microbial community change in the biofilm, growing on the walls of a containment tank of TRIGA nuclear reactor revealed a thriving community in an oligotrophic and heavy-metal-laden environment, periodically exposed to high pulses of ionizing radiation (IR). We observed a vertical IR resistance/tolerance stratification of microbial genera, with higher resistance and less diversity closer to the reactor core. One of the isolated Bacillus strains survived 15 kGy of combined gamma and proton radiation, which was surprising. It appears that there is a succession of genera that colonizes or re-colonizes new or IR-sterilized surfaces, led by Bacilli and/or Actinobacteria, upon which a photoautotrophic and diazotrophic community is established within a fortnight. The temporal progression of the biofilm community was evaluated also as a proxy for microbial response to radiological contamination events. This indicated there is a need for better dose-response models that could describe microbial response to contamination events. Overall, TRIGA nuclear reactor offers a unique insight into IR microbiology and provides useful means to study relevant microbial dose-thresholds during and after radiological contamination. [ABSTRACT FROM AUTHOR]

Molten salt nuclear reactors (MSRs) constitute a promising technology to produce safe, reliable, abundant low-carbon energy. To design MSR systems and perform safety analyses on them, numerical simulation is a powerful tool. Here, we implemented a coupling between several solvers of the deterministic neutronics code APOLLO3® (the MINARET SN transport and the MINOS diffusion and SPn-simplified transport solvers) and the computational fluid dynamics (CFD) code TRUST/TrioCFD, both developed at the French Alternative Energies and Atomic Energy Commission (CEA). The code coupling is orchestrated using the dedicated C3PO library of the open-source SALOME platform. A new code-coupling strategy is employed whereby the delayed neutron precursor concentrations are computed by the CFD code, which eases the use of traditional deterministic neutronics codes. We verified the correctness of our implementation by performing a numerical benchmark dedicated to fast spectrum MSRs originally devised by the French National Center for Scientific Research. The numerical results we obtained are in excellent agreement with those obtained by recent MSR-dedicated multiphysics simulation tools. This study provides a new convenient neutronic–thermal-hydraulic coupling strategy for MSR core simulation. [ABSTRACT FROM AUTHOR]

A coupling framework named Multi-Physics CORE (MPCORE) is developed to analyze the multiphysics phenomenon in a nuclear reactor. MPCORE performs two-way coupling between two physics modules. A rod ejection accident (REA) is an important design-basis accident that results in an instantaneous power surge in the case of prompt criticality. Hence, this technical note studies the passive response of a nuclear reactor core in the case of a similar rapid reactivity insertion. Stand-alone calculations by neutronics, thermal-hydraulic (TH), or fuel performance (FP) modules use conservative options for other physics modules. Thus, multiphysics analysis provides a more realistic assessment of actual prospective damage. MPCORE employs an adaptive time-step feature to reduce execution time. Moreover, it performs in-memory transfer of data between different modules. This technical note evaluates the performance of the TH module with cross flow (subchannel) and without cross flow (one-dimensional). For the FP module, the effect of dynamic and static gap heat transfer coefficient models is also quantified. Hence, four combinations with these two TH and FP options are simulated. The following are the safety parameters compared for different models: departure from nucleate boiling ratio, linear power, fuel enthalpy, fuel centerline temperature, cladding outer surface temperature, and coolant temperature. [ABSTRACT FROM AUTHOR]

In recent years, there has been much discussion of small modular reactors. Companies developing such designs have received large amounts of government funding. Lower power outputs of these reactors will likely result in higher costs in comparison to large nuclear reactors, and even if they achieve parity, will fail economically, since large reactors are themselves struggling to compete with renewable sources of electricity. Mass manufacture is unlikely to reduce costs adequately and might itself become a source of problems, including the possibility of recalls. The history of problems with non-traditional nuclear reactor designs indicates that they will likely take longer to commercialize than light-water small modular reactor designs. The problems related to radioactive waste and nuclear weapon proliferation will persist, though in different technical configurations depending on reactor design. Small modular reactors fail the tests of time and cost, which are of the essence in meeting the challenge of climate change. Even the official schedules indicate that their contributions will be negligible by 2030 and remain small by 2035, when the grid needs to be nearly completely decarbonized. [ABSTRACT FROM AUTHOR]

The commercial nuclear power plant industry initiated the licensing modernization project (LMP) to enhance the risk-informed and performance-based (RIPB) regulatory basis for advanced nuclear power reactors. The LMP framework relies heavily on RIPB concepts and approaches that together integrate the defense-in-depth philosophy. One example approach for seismic design is to align the LMP concepts with the performance targets described in the American Society of Civil Engineers (ASCE) standard, ASCE 43-19. The underlying strategy of this approach is to consider the performance of individual structures, systems, and components (SSCs) in seismic design, as well as the role they play in an accident event sequence. This approach contrasts with current regulations, in which every individual safety-related SSC is designed to the same seismic criteria irrespective of the role the SSC plays in the overall system performance. This new philosophy envisions more flexible seismic design options for each SSC, such that the overall seismic design can meet system-level acceptability criteria as well as plant-level acceptability criteria. The objective of this paper is to illustrate the flexibility and benefits of this proposed approach to the seismic design of SSCs in terms of reduced SSC demands (by reducing the design ground motions for SSCs) and improved SSC capacities (by allowing for alternative damage state limits). A simple shear wall was designed using ASCE 43-19 for a hard rock site in the Central Eastern United States considering alternate seismic design category and limit state combinations to examine the physical designs and functional fragilities of these combinations and their impact on seismic performance. The flexibility of this proposed approach is illustrated by an example that shows reduced SSC demands, while the SSC capacities and margins remain consistent with the required safety performance without any loss in overall plant safety. [ABSTRACT FROM AUTHOR]

This work summarizes the radiation transport–based design for a new D2O-moderated ex-core irradiation facility in the Washington State University (WSU) TRIGA reactor for optimization of 135Xe sources used for calibration and quality control testing of Xe gas detection equipment in support of the Comprehensive Test Ban Treaty (CTBT). Three-dimensional (3-D) particle transport analysis characterizing the WSU reactor core using MCNP6.2 (3-D Monte Carlo) and PENTRAN (3-D deterministic parallel SN) form the basis for the computational optimization. Excellent agreement between MCNP6.2 and PENTRAN predictions is observed. A fundamental fuel bundle depletion analysis is applied to enable a more accurate prediction of neutron flux and neutron spectrum distribution, which drives production rates of 135Xe and 133Xe. The results of various model simulations were used to inform recommendations for the final irradiation chamber design, which has been optimized for safe placement in the reactor tank prior to startup and will allow for insertion and rotation of xenon "bean" samples using existing WSU irradiation equipment, while remaining within operational parameters. The irradiation chamber is expected to produce samples that will remain viable for use in CTBT standards applications for durations 70% to 80% longer than samples produced using current procedures. Thus, this design is expected to improve CTBT-related calibrations and performance testing and to support the continued stability of the CTBT monitoring network. [ABSTRACT FROM AUTHOR]

Two of the experiment vehicles being developed for the Versatile Test Reactor (VTR) are presented here. The first is a rabbit system that will enable rapid insertion of small test capsules into the high fast flux of the VTR core for relatively short durations. The rabbit concept development includes the construction/demonstration of a near-full-scale system in a deep-water pool to demonstrate functionality, development of a concept of operations and initial procedures, and validation of thermal-hydraulic modeling. In addition, modeling efforts are underway to simulate the thermal and neutronic environment of a rabbit capsule. The second type of experiment vehicle presented here is a driver fuel test assembly for inserting fuel and materials tests into the core by replacing a driver fuel assembly. A novel design for dismountable test assemblies is proposed for the VTR. [ABSTRACT FROM AUTHOR]

In this paper, it is considered the existence of a solution for a nonlocal equation involving the p-Laplacian operator which is related to several applications such as vibration problems, flow of fluids through a homogeneous isotropic rigid porous medium, nuclear reactor's dynamics and in population dynamics. The approach is based on sub-supersolution arguments and Schauder's fixed point theorem. An important fact is that our result is valid for nonlocal terms which may exhibit a growth higher than p. [ABSTRACT FROM AUTHOR]

As part of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Rod Bundle Heat Transfer (RBHT) project, an experimental study was performed to investigate the entrained droplet sizes and velocities in a rod bundle under reflood conditions. Experimental results were obtained from the U.S. Nuclear Regulatory Commission/The Pennsylvania State University RBHT test facility using advanced dual laser measurement systems that allow for the simultaneous measurement of droplet behaviors at two axial locations during reflood transients. The RBHT facility is highly instrumented and contains a 7×7 electrically heated bundle with dimensions matching those in commercial pressurized water reactors. The combination of the measurement capabilities of the RBHT facility and the choice of appropriate experimental conditions allows for the measurement of unique droplet size and velocity distributions under different transient reflood conditions. [ABSTRACT FROM AUTHOR]

Zirconium is an important structural and cladding material in nuclear industry. Zr and Hf coexist in nature, their separation is important for use in nuclear reactors. An in-house synthesized Alkyl Phosphine Oxide (APO) ligand has been identified which provides better separation factor and extraction compared to Tributyl Phosphate (TBP) at lower feed acidity. There is high amount of silica in zirconium nitrate solution which results in formation of emulsion in mixer-settler. A novel differential contactor, pulsed stirred column (PSC), is explored to prevent emulsion formation during selective extraction of Zr from nitrate medium. Experimental studies on extraction of Zr using APO based organic phase are carried out in the PSC having 20 mm inner diameter (ID) and 1 m active section height. Aqueous phase is zirconium nitrate solution containing 75 g/L Zr in nitrate medium. The effects of stirring speed and residence time on percentage extraction are investigated. 99.99% pure Zr is obtained from zirconium nitrate solution without any emulsion formation. At optimized operating conditions (aqueous phase flowrate 10 ml/min, organic phase flowrate 45 ml/min, pulsing velocity 1.25 cm/s, rotational speed 1000 rpm), extraction of Zr is found to be 54%. 100% stripping of loaded organic using 0.5 M oxalic acid as strippant is also achieved in the PSC. [ABSTRACT FROM AUTHOR]

The design and safety assessment of nuclear reactors rely on a combination of calculations performed by several simulation packages, each dedicated to modeling a specific ensemble of phenomena. To treat the complexity of the physical problem, appropriate software architectures and methodologies to trace and implement user needs are of paramount importance to fulfill the needs of all the possible stakeholders. This work proposes a systematic approach to break the complexity of constructing a lattice neutronics platform that is one of the simulation packages needed in nuclear reactor analysis. After reviewing the state of the art of current methods applied in reactor physics engineering, the work concentrates on identifying the applicable software architecture strategies and on discussing advantages and drawbacks. While the specific target is the neutronics code APOLLO3®, the subsequent categorization and analysis of user needs written in the form of formal requirements allow for defining a unified approach to design an effective, industrial-grade, and future-proof calculation platform. Subsequent presentation of typical use cases involved in developing deterministic lattice calculation schemes allows linking the formal definition of use cases and software architecture with the actual application to a specific calculation setting. This work aims, therefore, at proposing an innovative viewpoint to tackle large software developments applicable in the nuclear industry. The research presented in this paper has been developed at Framatome in the context of the lattice neutronics work package of the H2020 CAMIVVER project. [ABSTRACT FROM AUTHOR]

The article discusses the importance of technological innovations in nuclear civil engineering, particularly in the context of climate change and the global energy crisis. It highlights the need for better modeling and construction techniques in nuclear projects, considering the growing threat of extreme weather events and the need for protection against malevolent acts. The article also mentions the success of the Technological Innovations in Nuclear Civil Engineering conference and the wide range of topics covered, including structural behavior, natural and technological hazards, monitoring and performance over time, and new construction technologies. The article concludes by emphasizing the ongoing research needs in both existing and new nuclear power plants. [Extracted from the article]

The adoption of multiplication eigenvalue is a well-established approach for the design of nuclear reactors. However, despite its popularity and nice physico-mathematical properties, this eigenvalue formulation is not able to provide quantitative information about what parameters the designer has to modify. In this paper, a novel generalized eigenvalue formulation is introduced to disclose the full potential of the neutron transport equation for core design applications. To illustrate the advantages of this new design-oriented approach with respect to traditional methods, some relevant problems arising in the physics of reactors are solved, such as the determination of the absorber density in the control rods and of the fissile concentration in the molten salt fast reactor. [ABSTRACT FROM AUTHOR]

The current study describes a simulation-based analysis of the atmospheric dispersion of radionuclide fission product particles in the near-field and far-field of a generic, conceptual microreactor, which is a small nuclear reactor with a power output typically ranging from 1 to 20 MW(thermal) and generally lower than 50 MW(electric). The near-field is a distance of up to 100 m from the microreactor while the far-field is a distance of 300 m or beyond from the microreactor. The generic microreactor operates at a pressure close to the ambient pressure. Therefore, in the event of a postulated accident that causes the leakage of radionuclide particles from the microreactor containment into the environment, the radionuclide particles are unlikely to travel too far from the reactor, as opposed to conventional nuclear reactors. The current paper provides estimates of average and 95th-percentile values of the normalized effluent concentration of the atmospheric radionuclide particle dispersion with respect to the source strength in the near-field and far-field of the conceptual microreactor. The computer code Atmospheric Relative CONcentrations in Building Wakes (ARCON96) was used to perform all simulations for the current study. It was observed that the 95th-percentile values of the normalized effluent concentration decrease by an order of magnitude as the receptor distance increases, i.e., from the near-field to the far-field. The dispersed aerosol concentration also decreases with time. A parametric study was performed to understand which input parameters affect the normalized effluent concentration values the most, and a definitive screening design was employed for this purpose. The atmospheric stability class and the distance between the reactor and the receptor were the parameters found to affect the aerosol dispersion characteristics by the greatest extent. The study recommends that the computer code RADTRAD (Radionuclide Transport and Removal And Dose Estimation) be used to estimate the actual dosage over distance using the outputs from ARCON96 as inputs, along with reactor-specific core inventories. [ABSTRACT FROM AUTHOR]

Kinetic neutron parameters are of fundamental importance in the field of nuclear reactor dynamics and control. Moreover, the precursor yield fraction and the neutron generation time for a given nuclear reactor are dependent on the properties of the reactor. Thus, in-pile experiments, such as oscillation experiments and noise experiments, are commonly conducted to measure those values. In this work, a method for determining the kinetic parameters of a reactor along with their covariance data from in-pile experiments is presented. It is performed by combining values of the reactor's response function obtained from both oscillation and noise experiments over a wide range of frequencies. The method is carried out for the MINERVE zero power reactor (ZPR) using a reanalysis of both oscillation and noise experiments that were conducted in the MINERVE reactor in 2013 and 2014. Moreover, various advantages and disadvantages of performing multiple in-pile experiments and combining their results in order to obtain a single set of kinetic parameters along with their covariance data are considered. Some suggestions for the design of such in-pile experiments are also discussed. [ABSTRACT FROM AUTHOR]

The French National Center for Scientific Research (CNRS) is carrying out design studies on a nuclear electric propulsion (NEP) engine based on a molten salt reactor (MSR). A NEP engine based on liquid nuclear fuel could allow developing a core design with relatively high power densities and temperatures while using simple reactivity control systems and keeping low pressure and temperature gradients in the fuel. Nevertheless, the design work of such an engine poses significant technical challenges and requires the use of advanced numerical simulation tools. Different MSRs for space are currently being studied. In this work, a MSR concept using a fast neutron spectrum is investigated using a multiphysics tool based on a numerical coupling between the OpenFOAM (computational fluid dynamics) and SERPENT 2 (Monte Carlo neutronics) codes. The analysis of this paper is focused on the reactor core coupled neutronic and thermal-hydraulic phenomena. Steady state full-power conditions are calculated for two different fast MSR designs using low-enriched uranium (LEU) and highly enriched uranium. The results show that the proposed core layout and materials allow obtaining a satisfactory temperature distribution in the core (maximal values and gradients) without significant penalization of the reactor operating conditions. A reactivity control strategy excluding the use of control rods is studied for the LEU concept. Transient and safety studies are also performed and show acceptable performance. [ABSTRACT FROM AUTHOR]

Metal hydrides are being seriously considered for advanced nuclear reactor or microreactor applications due to their solid physical state and high hydrogen density. Using hydrides for autonomous applications poses several research and development challenges, one of which relates to neutron upscattering in the thermal energy regime. These hydrides, including zirconium hydride and yttrium hydride, result in a positive temperature coefficient of reactivity for several advanced reactor designs. In this study, we consider one such design that exhibits positive feedback from metal hydrides and thoroughly investigate the neutronic aspects of the core. Temperature reactivity coefficients for four fuels and two hydride moderator configurations are studied, and the total temperature coefficients are found to be positive for all designs, showing that this issue cannot be resolved simply by material variations. Accordingly, five epi-thermal absorbers were evaluated to demonstrate the feasibility of the excess positive feedback suppression in the core instigating from neutron energy spectrum shift. Following which, two promising burnable poison candidates are selected to investigate further throughout the core discharge. Promising results are shown for this core design, which can be extended to other hydride-moderated remote special-purpose reactor designs. [ABSTRACT FROM AUTHOR]