Nuclear Engineering

In order to increase the accuracy of measurement of radioactive waste drum assay by Segmented Gamma Scanning technique a new measurement technique was proposed. Results of the study with noise data of measuring factors are presented. The results confirm that the technique can be successfully developed for assay of radioactive waste drums.

I have adapted the nuclear materials storage location to align with mudstone bedrock for either or both above or below sea level tunnels.

Please find a revised scheme to locate a nuclear material storage scheme in mudstone bedrock with the option for both above and below sea level tunnels.

A trial proving tunnel is included to enable characterization to be determined to address risks and experiences tunnelling in the area.

It has significant capacity and can be constructed in stages.
Cumbria Nuclear Material Storage Scheme Constructed in Mudstone Bedrock Proposal
Considering the UK Government has a preference for installing the Nuclear Material Storage tunnels in Mudstone Bedrock the following location is an alternative to locating the work site further from Millom.  The selected area enables the tunnels to be either above sea level or below sea level.  The proposed scheme provides both above and below sea level tunnels to be constructed for the different performances to be assessed.
Each level within the mudstone material area can accommodate up to 11 tunnels each 60 metres apart.  This will enable up to 100 Km of tunnel per level.
The location is suitable for locating a jetty for import of TBMs and materials and export of removed crushed bedrock.
The proposed scheme includes a proving tunnel to enable the characteristics of the sites to be evaluated.  This tunnel will be used to test the logistics, storage methods and trial the equipment and machines required for operating the storage facilities.
The providing tunnel will enable the ground and bedrock and below sea level water movement characteristics and management to be evaluated.
The following images are:
Image 1 Image 1 Proposed Scheme location enabling the construction of on land and offshore tunnels in Mudstone
Image 2 Location of Portals and Storage Footprint
Image 3 Proposed 250m Contour Area
Image 4 Black Combe and Fell Grey Mudstone Area

This paper proposes a theoretical design for an electro-nuclear reactor based on the Al-Hamed Electro-Nuclear Equation, which integrates electrical energy input with nuclear mass-energy conversion while accounting for the mass of secondary particles. The reactor is designed to optimize the balance between electrical stimulation and nuclear efficiency. Core calculations incorporate both the classical and revised models for comparative evaluation. This model aims to guide the future development of hybrid energy systems that merge electrical and nuclear processes into a unified, efficient, and safe reactor concept.

Results from fie current physics, materials and blanket R&D programs are combined with physics and engineering design constraints to characterize candidate tokamak demonstration plant (DEMO) designs. Blanket designs based on the principal structural materials, breeding materials and coolants being developed for the DEMO were adapted from the literature. Neutron flux and activation calculations were performed, and several radioactive waste disposal indices were evaluated, for each design. Of the primary low-acfivation structural materials under development in the US, it appears that vanadium and ferritic steel alloys, and possibly silicon carbide, could lead to DEMO designs which could satisfy realistic low-level waste (L,LW) criteria, provided that impurities can be controlled within plausible limits. Allowable LLW concentrations are established for the limiting alloying and impurity elements. All breeding materials and neutron multipliers considered meet the LLW criterion. This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employets, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or sefvice by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, mommendation. or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. 1. DSCu -Dispersion Strengthened Copper * Structural Impurity

negative ion sources at relevant parameters in ITER-like conditions: source and accelerator located in the same vacuum where the beam propagates, immunity to electromagnetic interferences of multiple radio-frequency (RF) antennas, avoidance of RF-induced discharges on the outside of the source. Three years of experiments on SPIDER have addressed to the necessary design modifications to enable full performances. The source is presently under a long shut-down phase to incorporate learnings from the experimental campaign, in particular events/issues occurred during operation, which led to the identification of improvement opportunities/necessities (e.g. RF discharges, local burns, water leaks, other damages, configuration/design upgrades to maximize chances/margin to quest target parameters). Parallelly, developments on MITICA, the full-scale prototype of the ITER NBI featuring a 1 MV accelerator and ion neutralization, are underway including manufacturing of the beam source, accelerator and the beam line components, while power supplies and auxiliary plants, already installed, are under final testing and commissioning. Integration, commissioning and tests of the 1MV power supplies are essential for this first-of-kind system, unparalleled both in research and industry field. 1.2 MV dc insulating tests of high voltage components were successfully completed. The integrated test to confirm 1MV output by combining invertor systems, DC generators and transmission lines extracted errors/accidents in some components. To realize a concrete system for ITER, said events have been addressed and solutions for the repair and the improvement of the system were developed. Hence, NBTF is emerging as a necessary facility, due to the large gap with existing injectors, effectively dedicated to identify issues and find solutions to enable successful ITER NBI operations in a time bound fashion. The lessons learned during the implementation on NBTF and future perspectives are here discussed.

High pressure heater (HPH) 2 merupakan salah satu peralatan yang menunjang kinerja feed water sistem pada siklus air dan uap di PLTU. HPH 2 berfungsi untuk memanaskan feed water dengan menggunakan ekstraksi dari turbin sebelum air tersebut dipanaskan dalam boiler sehingga kerja boiler menjadi lebih ringan dan dapat meningkatkan efisiensi pembangkit listrik. Namun HPH 2 pada PLTU UJP Jabar 2 Pelabuhan Ratu telah mengalami kebocoran sejumlah 532 tube dari 1429. Hal ini merupakan akibat off service HPH 3 dan over extraction flow masuk HPH 2. Kebocoran pada HPH 2 membuat keluaran temperature feed water pada HPH 1 yang masuk boiler menurun. Hal tersebut menyebabkan konsumsi bahan bakar pada boiler meningkat, sehingga efisiensi siklus pembangkit menurun. Oleh karena itu, perlu dilakukan proses redesign dan recalculation pada HPH 2 secara Thermodynamics-Heat Transfer agar didapatkan desain HPH 2 dengan performa yang terbaik. Proses redesign dan recalculation dilakukan dengan menggunakan an...

The primary aim for the in-core fuel management in a nuclear reactor is to achieve higher fuel utilization without compromising the safety during operation. Loading pattern optimization during each refueling stage is the main challenge in light water reactors like pressurized water reactors (PWRs) and boiling water reactors (BWRs). Whereas, in case of heavy water reactors like pressurized heavy water reactors (PHWRs), the development of refueling scheme is relatively simple mainly due to on-line refueling, use of small length bundle and natural uranium as fuel. The use of small length bundle and flexibility of multi-bundle shift scheme helps in controlling the ripples in power due to refueling. Even though the PHWRs possess best neutron economy, the use of natural uranium fuel in PHWRs limits the achievable burn-up to 7500 MWd/Te and as a result large amount of waste management is a serious concern. In Indian advance heavy water reactor (AHWR), both the features of PWRs (high discharge burn-up fuel) and PHWRs (on-power refueling) are being included along with advanced safety features. The AHWR shall possess good neutron economy and owing to higher discharge burn-up, lower waste is expected. However, during on-power refueling with full length channel and high enrichment fuel, the challenge is to control the refueling ripples and maintaining the operational parameters under their design limits. This necessitates the implementation of a special refueling strategy to control the power peaking due to on-power refueling. This paper describes about the special refueling strategy for this kind of problems. A computer code based on the heuristic approach has been developed for facilitation of on-power refueling in automatic manner for high discharge burn-up fuel in AHWR core. Further, parallel processing on shared memory architecture (personal computer) has been employed to enhance the efficiency of this code.

Alamos National Laboratory (LANL) in the late 1980s has been done using information available in the open literature. The Monte Carlo results that were obtained for critical test configurations have been quite good with k eff values within 0.5% of the critical value of unity. The material worths calculated in this study were found fairly close to the measurement values. The lack of information on design data uncertainties and the inconsistency in the design data from different sources make it difficult to quantify the uncertainty in the core integral parameters arising from data uncertainties. The evaluated configurations could however be useful as numerical benchmarks that are based on the currently available information.

Applicability of the first-order perturbation (FOP) theory method to reactivity estimation for sourcedriven systems is examined in this paper. First, the formally exact point kinetics equations have been derived from the space-dependent kinetics equations and the kinetics parameters including the dynamic reactivity have been defined. For the dynamic reactivity, exact and first-order perturbation theory expressions for the reactivity change have been formulated for source-driven systems. It has been also shown that the external source perturbation itself does not change the reactivity if the initial λ -mode adjoint flux is used as the weight function. Using two source-driven benchmark problems, the reactivity change has been estimated with the FOP theory method for various perturbations. By comparing the resulting reactivity changes with the exact dynamic reactivity changes determined from the space-dependent kinetics solutions, it has been shown that the accuracy of the FOP theory method for the accelerator-driven system (ADS) is reasonably good and comparable to that for the critical reactors. The adiabatic assumption has also been shown to be a good approximation for the ADS kinetics analyses.

The isotopic inventories of a 2000 MWt Advanced Burner Reactor (ABR) core have been estimated to support the ABR accident analysis to be reported in the Appendix D of the Programmatic Environmental Impact Statement (PEIS). Based on the Super-PRISM design, a preliminary core design of 2000 MWt ABR was developed to achieve a one-year cycle length with 3-batch fuel management scheme. For a bounding estimation of transuranics (TRU) inventory, a low TRU conversion ratio (~0.3) was targeted to increase the TRU enrichment. By changing the fuel compositions, isotopic inventories of mass and radioactivity were evaluated for four different core configurations: recycled metal fuel core, recycled oxide fuel core, startup metal fuel core, and startup oxide fuel core. For recycled cores, the TRU recovered from ABR spent fuel was used as the primary TRU feed, and the TRU recovered from 10-year cooled light water reactor spent fuel was used as the makeup TRU feed. For startup cores, weapons-grade plutonium was used as TRU feed without recycling ABR spent fuel. It was also assumed that a whole batch of discharged fuel assemblies is stored in the in-vessel storage for an entire irradiation cycle. For both metal and oxide fuel cores, the estimated TRU mass at beginning of equilibrium cycle (BOEC), including spent fuel TRU stored in the in-vessel storage, was about 8.5 -8.7 MT for the recycled cores and 5.2 MT for the startup cores. Since a similar power was generated, the fission product mass are comparable for all four cores: 1.4 MT at BOEC and about 2.0 MT at end of equilibrium cycle (EOEC). Total radioactivity at BOEC is about 8.2×10 8 curies in recycled cores and about 6.9×10 8 curies in startup cores, and increases to about 1.1×10 10 curies at EOEC for all four cases. Fission products are the dominant contributor (more than 80%) to the total radioactivity at EOEC for all four cases, but the fission product radioactivity decreases by 79% after one-year cooling. The heavy metal radioactivity in recycled cores is higher than that of the startup core because of the higher TRU inventory at BOEC. The leading contributors at EOEC are U-239, Np-239, Am-242 and Pu-243 at EOEC, but these short-lived nuclides decay out quickly during the refueling time. As a result, the leading contributors at BOEC are

This paper presents an assessment of applicability of the multigroup cross sections generated with Monte Carlo tools to the fast reactor analysis based on transport calculations. 33-group cross section sets were generated for simple one-(1-D) and two-dimensional (2-D) sodium-cooled fast reactor problems using the SERPENT code and applied to deterministic steady-state and depletion calculations. Relative to the reference continuous-energy SERPENT results, with the transport corrected P 0 scattering cross section, the k-eff value was overestimated by 506 and 588 pcm for 1-D and 2-D problems, respectively, since anisotropic scattering is important in fast reactors. When the scattering order was increased to P 5 , the 1-D and 2-D problem errors were increased to 577 and 643 pcm, respectively. A sensitivity and uncertainty analysis with the PERSENT code indicated that these large k-eff errors cannot be attributed to the statistical uncertainties of cross sections and they are likely due to the approximate anisotropic scattering matrices determined by scalar flux weighting. The anisotropic scattering cross sections were alternatively generated using the MC 2 -3 code and merged with the SERPENT cross sections. The mixed cross section set consistently reduced the errors in k-eff, assembly powers, and nuclide densities. For example, in the 2-D calculation with P 3 scattering order, the k-eff error was reduced from 634 pcm to À223 pcm. The maximum error in assembly power was reduced from 2.8% to 0.8% and the RMS error was reduced from 1.4% to 0.4%. The maximum error in the nuclide densities at the end of 12-month depletion that occurred in 237 Np was reduced from 3.4% to 1.5%. The errors of the other nuclides are also reduced consistently, for example, from 1.1% to 0.1% for 235 U, from 2.2% to 0.7% for 238 Pu, and from 1.6% to 0.2% for 241 Pu. These results indicate that the scalar flux weighted anisotropic scattering cross sections of SER-PENT may not be adequate for application to fast reactors where anisotropic scattering is important.

In the current study results from an experimental investigation on an oxygen/methane multi-injector combustion chamber are presented. They provide detailed information about the thermal loads at the hot inner walls of the combustion chamber at representative rocket engine conditions and pressure ranges up to 40 bar. The present study aims to contribute to the understanding of the thermal transfer processes and of the interaction between the injectors and the injector-wall. Furthermore, the test results are used as a test case for the validation of the in-house engineering tool Thermtest. Due to the complex flow phenomena linked to the use of cryogenic propellants, like extreme variation of flow properties and steep temperature gradients, in combination with intensive chemical reactions, the problem has been partially simplified by injecting the propellants in gaseous form.

This study investigates the application of the Quantum Harmonic Resonance Framework (QHRF) in enhancing fusion plasma heating, ignition, and cumulative energy output within a controlled Tokamak plasma simulation environment. Two plasma scenarios were compared: a Standard Plasma utilizing traditional heating mechanisms, and a QHRF-Enhanced Plasma utilizing dynamic parametric resonance fields. Comparative results demonstrate that QHRF fields: • Accelerate ignition time, • Increase maximum plasma temperature, • Substantially boost total fusion energy output, • Enable longer and more stable self-heating phases. Controlled variables, theoretical justification, and reproducible outcomes provide robust evidence that QHRF fields constitute a breakthrough technology for practical fusion energy systems.

This paper presents a numerical investigation of thermal-fluid dynamics processes through the gaps of the spherical fuel elements (fuel pebbles) in the core of a Pebble Bed Reactor (PBR), using Computational Fluid Dynamics (CFD). The PBR is a Very High Temperature Reactor (VHTR) design to the Generation IV International Forum (GIF). The PBR is one of the most promising projects of the seven(originally six) classes of GIF reactors. The set of pebbles that constitute the core was modeled by representations of crystalline structure with different packing factors. The results shown the importance of simulation of heat conduction inside the pebble fuel as well as the need to better assess the influence of the arrangement formed by pebbles fuel in PBR reactors thermal-fluid dynamics behavior.

The Comprehensive Nuclear-Test-Ban Treaty (CTBT) bans the testing of nuclear explosive devices underground, in the atmosphere and underwater. Two main technologies, radionuclide and seismo-acoustic monitoring, are deployed in the International Monitoring System used for the verification of the CTBT. Medical isotope production from fission-based processes is the dominant contributor to a worldwide background of radioxenon. This background can make the discrimination of nuclear tests from legitimate nuclear activities very challenging. Even if emissions from medical isotope producers experienced a large reduction, there remain other important sources of radioxenon that contribute to the global background such as research reactors and nuclear power plants. Until recently, the largest producer of medical isotopes was located in Canada, at the Canadian Nuclear Laboratories (CNL) facility. The characterization of CNL emissions and its research reactor can provide valuable information for effective verification of the CTBT.

Nuclear energy is considered a clean, reliable, and an inexhaustible energy source for power generation. Nuclear power is harnessed from nuclear fission reactions in a dedicated power plant. The by-products (produced in the nuclear power plant) are radioactive and pose a threat to the environment. The safe disposal of nuclear waste is vital to ensure the sustainable use of the nuclear energy. The immobilization of radioactive waste before final disposal is essential for the interim storage and transportation. This review summarizes the recent work on glass, ceramics, and glass-ceramics matrices to immobilize high-level waste. The synthesis methods, leaching behavior, and radiation resistance of matrices are discussed briefly.

This paper presents and explains test results showing the effect of ion species on the single event burnout (SEB) failure voltage using a SEB sensitive engineering power double diffused metal oxide silicon field effect transistor (DMOSFET). The analyses show the determining factor of tested SEB failure voltage is the ion species itself rather than test or beam conditions such as initial beam energy, surface linear energy transfer (LET), ion range, or ionized charge. Also, results from five test facilities and five test setups are compared to determine if there will be differences in test results when different test setups or different heavy ion accelerator facilities were used.

The objective of this paper is to give an overview of the capabilities of Eulerian bifluid approach to meet the needs of studies for nuclear safety regarding hydrogen risk, boiling crisis, and pipes and valves maintenance. The Eulerian bifluid approach has been implemented in a CFD code named NEPTUNE_CFD. NEPTUNE_CFD is a three-dimensional multifluid code developed especially for nuclear reactor applications by EDF, CEA, AREVA, and IRSN. The first set of models is dedicated to wall vapor condensation and spray modelling. Moreover, boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. The paper aims at presenting the generalization of the previous DNB model and its validation against 1500 validation cases. The modelling and the numerical simulation of cavitation phenomena are of relevant interest in many industrial applications, especially regarding pipes and valves maintenance wher...

This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an ...

A Monte Carlo procedure has been applied in this work in order to address the impact of activation cross sections (XS) uncertainties on contact dose rate and decay heat calculations for the outboard first wall (FW) of a magnetic fusion energy (MFE) demonstration (DEMO) reactor. The XSs inducing the major uncertainty in the prediction of activation related quantities have been identified. Results have shown that for times corresponding to maintenance activities the uncertainties effect is insignificant since the dominant XSs involved in these calculations are based on accurate experimental data evaluations. However, for times corresponding to waste management/recycling activities, the errors induced by the XSs uncertainties, which in this case are evaluated using systematic models, must be considered. It has been found that two particular isotopes, 60 Co and 94 Nb, are key contributors to the global DEMO FW activation uncertainty results. In these cases, the benefit from further improvements in the accuracy of the critical reaction XSs is discussed.

In this study, the diffraction model is employed to study the clus- ter stripping nuclear reaction on the basis of the Regge-pole method. A for- mula for the angular distribution is derived. The angular distribution spectr a for the cluster stripping nuclear reactions 40Ca(3He,p)42Sc, 40Ca(t, p)42Ca and 54 Fe(t, p)56Fe at various energies have been well reproduced with projectile energies varying between 10 MeV and 15 MeV. The predicted theoretical re- sults for the cluster stripping reactions 40Ca(3He,p)42Sc with 10 MeV, 12 MeV and 15 MeV incident 3He; 40Ca(t, p)42Ca with 10.1 MeV incident triton; and 54 Fe(t, p)56Fe with 12 MeV incident triton were compared with the experimen- tal data and good agreement was observed. Also a comparison between the diffraction model prediction of the angular distribution values for the cluste r stripping reaction 40Ca(3He,p)42Sc at energy 12 MeV of 3He and that of the distorted-wave Born approximation (DWBA) is implemented and a satisfactory agreem...

Considerable progress has been made on the design of the ITER electron cyclotron emission (ECE) diagnostic over the past two years. Radial and oblique views are still included in the design in order to measure distortions in the electron momentum distribution, but the oblique view has been redirected to reduce stray millimeter radiation from the electron cyclotron heating system. A major challenge has been designing the 1000 K calibration sources and remotely activated mirrors located in the ECE diagnostic shield module (DSM) in the equatorial port plug #09. These critical systems are being modeled and prototypes are being developed. Providing adequate neutron shielding in the DSM while allowing sufficient space for optical components is also a significant challenge. Four 45-meter long low-loss transmission lines transport the 70-1000 GHz ECE from the DSM to the ECE instrumentation room. Prototype transmission lines are being tested, as are the polarization splitter modules that separate O-mode and X-mode polarized ECE. A highly integrated prototype 200-300 GHz radiometer is being tested on the DIII-D tokamak in the USA. Design activities also include integration of ECE signals into the ITER plasma control system and determining the hardware and software architecture needed to control and calibrate the ECE instruments.

The liner structure in nuclear power plants provides containment for the operation and therefore the study of its durability and integrity during its service life is an important issue. There are several causes for the deterioration of the liner, which in general involve corrosion due to its metallic nature. The present paper is aimed at describing the assessment of corrosion problems of two liners from two different nuclear power plants, which were evaluated using non-destructive electrochemical techniques. In spite of the testing difficulties arisen, from the results extracted it can be concluded that the electrochemical techniques applied are adequate for the corrosion evaluation. They provide important information about the integrity of the structure and allow for its evolution with time to be assessed.

An experimental model of a photovoltaic (PV) module-proton exchange membrane (PEM) electrolyser system has been built. A model has been developed for each device separately based on the experimental results. Output current-voltage (I-V) characteristics of the PV module are modelled in respect to different irradiance and temperature conditions by experimental tests. Similarly, input I-V characteristic and hydrogen formation characteristic of the PEM electrolyser are measured and modelled. After these studies, combined PV module-PEM electrolyser system model is defined. There is a good agreement between model predictions and measurements. At 18-100% irradiance interval, operating points of PEM electrolyser on the PV module are predicted with relative errors of 0.1-0.8%. Furthermore, the study shows that these simple model system devices can easily be defined in MATLAB/Simulink and used to model similar systems of different size.

This paper describes a radiation source that can be used to actively interrogate containers, trucks, trains, cars, etc to determine the presence and location of chemical explosives and special nuclear materials such as uranium and plutonium. Active interrogation methods using high energy photon or neutron sources to induce fission are the only feasible option for detection of highly enriched uranium (HEU) because passive detection methods are easily compromised by even moderate amounts of shielding. For detection of chemical explosives, the same active interrogation device can be used to produce resonant photons that can detect nitrogen that is used in most chemical explosives. The accelerator based system described here produces a penetrating beam of high energy photons or neutrons that can "see" inside a sealed container. If chemical explosives or special nuclear materials are present, they will emit a characteristic signal that is detected and interpreted by electronic sensors. Shielded "dirty bombs" can be detected by the attenuation of high energy photons caused by the density of the shield material. The interrogating source of radiation is based upon a new high current negative ion source and high current tandem accelerator. The accelerator accelerates ions and projects them onto an appropriately designed target. The target converts the energy of the ion beam into a high energy highly penetrating photon or neutron beam. The beam is made to pass through the container. If explosives, special nuclear materials or shielded dirty bombs are present, the beam together with a suitable detection system uniquely identifies the location, amount and density of material.

This work proposes a paradigm shift in nuclear safety. Fission and fusion phenomena are treated as inertial processes. This simple statement has profound implications when applied to the thermodynamic understanding of a nuclear engine given that its MeV release is: independent of terrestrial reference; not affected by temperature, kinetic, physical potential or gravitational influences; a single phenomenon describes the macro; and has no enthalpic meaning. This approach leads, for example, to fission engine analytics resulting in matrix solution of: average neutron flux, useful power output and system heat rejection, these coupled to reactor coolant mass flow. The only appropriate descriptive vehicle for nuclear phenomena is Second Law exergy analysis, treating MeV release as a thermodynamic potential. This approach demands reinterpretation of Einstein's 𝛥𝐸 = 𝑐^2 𝛥𝑚 by describing his ΔE, not as a ΔEnergy, but an exergetic potential, an ultimate "Free Exergy." Free Exergy consists of both recoverable and irreversible portions of the MeV release. In transference to a coolant, the recoverable release produces an exergetic increase (ṁΔg); the fluid exergy's 𝑇 𝑅𝑒𝑓 being explicitly computed from an Inertial Conversation Factor (Ξ). Importantly, Ξ also transforms recoverable nuclear release to an explicit, and consistent, thermal power (ṁΔh). Asserted is that nuclear phenomena traditionally have had no direct computational nexus with thermodynamics. Since N Reactor days, there has been no direct nexus between nuclear power and its resultant coolant energy flow. N Reactor's neutron flux was not well measured but was the motive force behind delivering 4000 MWt to the Columbia River. It is well known that neutron flux in a 1270 MWe PWR is approximately 1.0x10 13 1 n0cm-2 sec-1. This flux a product of Neutron Transport Theory (NTT). However, a back-calculated flux based on the PWR's rated 3640 MWt produces twice the NTT flux. From the time of the Manhattan Project, nuclear engineers had treated flux as a relative parameter: hard to measure, a relative value to be normalized, is not uniform in the axial or radial, etc.

High temperature electrolysis of water using solid oxide electrochemical cells (SOEC) is a promising technology for hydrogen production with high energy efficiency and may promote decarbonization when coupled with renewable energy sources and excess heat from nuclear reactors. Over the past several decades there have been extensive scientific and engineering studies on cell materials and degradation behaviors that have greatly improved current density, decreased total resistance, and lowered degradation rates. Although the technology is now at a near-commercial level, maintaining consistency in cell testing and minimizing variance in practical testing environments is an often overlooked but crucial topic. To promote high quality data collection, testing procedures and balance of plant component details are extremely important to consider. This work discusses some key factors affecting the reproducibility of practical SOEC testing on the button cell level, namely, current collection ...

The ARIES team is currently proposing two tungsten divertor concepts for their tokamak designs and performed extensive analysis to optimize their thermal and structural performance . Due to the high divertor operating temperatures and the low ductility of tungsten, thermal creep and fracture will be important failure mechanisms to consider. This paper presents a series of finite element analyses addressing the viable operating ranges of these tungsten plate divertor concepts with respect to creep and fracture. For fracture, the J-Integral, a path-independent contour integral that estimates the strain energy release rate for a crack of assumed geometry, is used to address crack propagation. Elliptical surface cracks are introduced both inside and outside the coolant channel and steady state calculations are carried out for both full power and cold shutdown conditions. It is determined that the critical crack is on the inside of the coolant channel with the highest stress intensities at full power operation. Also, transient creep simulations are performed to predict the high temperature thermal deformations and creep strains at various surface flux levels. Finally, transient thermal calculations are carried out to simulate edge localized modes (ELMs) in the plasma, and conclusions are drawn with respect to the severity and frequency of these events with respect to surface melting for the two concepts.

•ARIES-IFE study concluded in 2003 -focused on evolving parameter design space -laser and heavy ion drivers -direct and indirect-drive targets -dry wall, wetted wall and thick liquid wall chambers •Termination of effort on heavy ion, indirect-drive target, thick liquid wall chamber studies (HYLIFE) •Currently IFE technology is primarily funded through the following two programs: -HAPL study (multi-year, multi-institution effort led by NRL) -Z-pinch study (started in 1999 as LDRD, now much broader)

The first wall of a laser fusion chamber will experience high heat loads pulsed at 5-10 Hz with pulse widths on the order of a few microseconds. This poses a challenging problem for dry wall designs, as the wall will be susceptible to a variety of failure modes. The primary design concept of the High Average Power Laser (HAPL) project is a ferritic steel first wall coated with tungsten armor. Due to the extreme heat loads, the armor will experience high temperatures, extensive yielding, and surface cracking. In order to evaluate the ability of this design to provide a suitable lifetime, a series of experiments to simulate chamber conditions using ions, xrays, infrared heating, and lasers is under way. These experimental efforts have been coupled with numerical modeling to help determine likely failure modes and establish design criteria for chambers. This paper compares models for the thermomechanical effects seen in the tests to those expected in a full power chamber, in order to assess the ability of the tests to mimic the actual chamber performance. The tests are found to have some limitations, but they still offer excellent approximations of the true behavior.

TITAN is a study to investigate the potential of the reversed-field pinch concept as a compact, high-power density energy system. Two reactor concepts were developed, a self-cooled lithium design with vanadium structure and an aqueous solution loop-in-pool design, both operating at 18 MW/m 2. The key safety features of the TITAN-I lithium-vanadium blanket design are in material selection, fusion power core configuration selection, lithium piping connections, and passive lithium drain tank system. Based on these safety features and results from accident evaluation, TITAN-I can at least be rated at a level 3 of safety assurance. For the TITAN-II aqueous loop-in-pool design, the key passive feature is the complete submersion of the fusion power core and the corresponding primary coolant loop system into a pool of low temperature water. Based on this key safety design feature, the TITAN-II design can be rated at a level 2 of safety assurance.

TITAN-II is a compact, high-power-density reversed-field pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO 3 and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the reversed-field pinch reactor to operate at a neutron wall loading of 18 MW/m 2. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a level 2 of passive safety assurance design.

This paper covers the results of a scoping study to assess the possible application of magnetic diversion to a laser IFE reactor. Its impact on the engineering design and performance of the reactor is discussed, key issues are identified, and the findings from this assessment are summarized.

The effects of radiation on the structural performance of fusion reactor structures is recognized as a major issue for the development of fusion reactor technology. Neutron irradiation changes the mechanical properties of structural components resulting in a general degradation of these properties. In addition to the mechanical loads (pressure and weight) and the thermal strains, non-uniform inelastic strain fields are induced by radiation swelling and creep in fusion structures. In this paper, we describe a new computer code, STAIRE, for ST_Tress Analysis Including Radiation Effects. This code is based on standard beam theory for pipe-bends. The theory is modified in two areas: (1) consideration of the pipes' cross-section deformation as the radius of curvature changes; and (2) inclusion of inelastic radiation and thermal strains (swelling and creep). An efficient analytical/numerical approach is developed for the solution of indeterminate beam problems. As an application of the method, the stress distribution and deflections of toroidal blanket pipes in the Mirror Advanced Reactor Study (MARS) are evaluated. Swelling strains are identified as a major source of stress and deformation in the proposed blanket design, and possible soliJtions to the problem are outlined.

Aim of this work is the preliminary thermal-fluid-dynamic assessment of a subcritical system to perform integral measurements on transmutation processes, designed in the frame of EU CHANDA project: in particular, a low power Accelerator Driven System (ADS) may represent an attractive intermediate step to fill the gap between existing and future facilities like MYRRHA (or possibly EFIT). The activity finds its place within the roadmap for the evaluation of transmutation processes in ADSs, where currently measurements are performed on the zero-power ADS Guinevère at SCK•CEN (Mol, Belgium), while in the future it is expected that MYRRHA will be the major high-power facility for performing this kind of studies and assessing the performance of a fast ADS. The neutronic characterization and burn-up simulations based on MCNP6 code of the reactor core allowed the definition of the geometry and the composition of the fuel assembly and, therefore, the thermal power to be removed. On this basis a thermo-fluid-dynamic assessment, through simple analytical accounts and detailed 3D CFD calculations by OpenFOAM-v1612+ and ANSYS FLUENT v17.0 was made. The present paper is an extended version including the main conclusions and observations emerged in the 2nd AIGE-IIETA International Conference in Genoa.

Consideration is given to the optimization peculiarities of the flow parts (FP) of the high-pressure cylinder (HPC) and the medium-pressure cylinder (MPC) of the K-540-23.5 turbine. The optimal options of the FPs of the HPC and MPC were imple-mented into the thermal scheme taking into account the effect of the underheating tem-perature of the feed water and steam parameters on the heaters during the steam bleed-offs.

A lithium-ion battery employing a novel phase change material (PCM) thermal management system was designed for an electric scooter. Passive thermal management systems using PCM can control the temperature excursions and maintain temperature uniformity in Li-ion batteries without the use of active cooling components such as a fan, a blower or a pump found in air/liquid-cooling systems. Hence, the advantages of a compact, lightweight, and energy efficient system can be achieved with this novel form of thermal management system. Simulation results are shown for a Li-ion battery sub-module consisting of nine 18650 Li-ion cells surrounded by PCM with a melting point between 41 and 44 • C. The use of aluminum foam within the PCM and fins attached to the battery module were studied to overcome the low thermal conductivity of the PCM and the low natural convection heat transfer coefficient. The comparative results of the PCM performance in the presence of Al-foam and Al-fins are shown. The battery module is also simulated for summer and winter conditions. The effect of air-cooling on the Li-ion battery was also studied. These simulation results demonstrate the successful use of the PCM as a potential candidate for thermal management solution in electric scooter applications and therefore for other electric vehicle applications.

A passive thermal management system is evaluated for high-power Li-ion packs under stressful or abusive conditions, and compared with a purely air-cooling mode under normal and abuse conditions. A compact and properly designed passive thermal management system utilizing phase change material (PCM) provides faster heat dissipation than active cooling during high pulse power discharges while preserving sufficiently uniform cell temperature to ensure the desirable cycle life for the pack. This study investigates how passive cooling with PCM contributes to preventing the propagation of thermal runaway in a single cell or adjacent cells due to a cell catastrophic failure. Its effectiveness is compared with that of active cooling by forced air flow or natural convection using the same compact module and pack configuration corresponding to the PCM matrix technology. The effects of nickel tabs and spacing between the cells were also studied.