The goal of this study was to experimentally examine the spatial and temporal variations of air and helium concentrations and temperature fields within simulated reactor cavities of a High Temperature Gas-cooled Reactor (HTGR) following helium discharge into an initially air-filled reactor cavity system. The research scenario involved a hypothetical small pipe break in the Reactor Pressure Vessel (RPV), resulting in the release of pressurized high-temperature helium into the surrounding cavity. A scaled five-compartment experimental facility, modeled after the General Atomics Modular High Temperature Gas-cooled Reactor (GA-MHTGR) design, was constructed for helium and air mixing experiments. Oxygen sensors and thermocouple probes were installed in all five cavities to measure the oxygen concentrations and temperature distributions of the gas mixture. The helium concentration could then be determined from the measured oxygen concentration in the helium-air mixture. Custom helium injection nozzles were fabricated enabling the investigation of circular and rectangular break geometries as well as horizontal and upward or downward helium injection directions. Detailed transient temperature maps were generated using a combination of a fiber optics temperature sensor and multiple thermocouple probes. The experimental findings highlighted the significant impact of the injected helium jet velocity and direction on the gas mixing process. Lower gas injection velocities resulted in a higher buoyancy effect on the gas jet leading to helium mixing mostly with air in the upper regions of the cavity. Conversely, higher injection velocities exhibited a lower buoyancy effect and uniform mixing of helium and air throughout the cavity. The findings also demonstrated how the direction of the injected helium jet influences the air-helium temperature profiles within the cavities.

A high frequency stabilization mount to compensate for small attitude fluctuations is developed for enhanced imaging and pointing systems on Unmanned Aircraft Systems (UAS). This system consists of a custom camera mount, piezoelectric actuators, and a digital controller to actively control flight vibrations. Payload designs that acquire views of the Earth surface and stationary or moving targets require stable cameras for precision viewing. Placing cameras onboard these small aircraft are vital to the Intelligence, Surveillance, and Reconnaissance (ISR) mission of many payload designs. It is necessary to have real time precision viewing imaging systems while in flight, but it is increasingly difficult as the small planes reach higher altitudes. A slight change in the camera angle at high altitudes results in a large shift from the designated target. This project focuses on high frequency analysis for small oscillations rather than large attitude changes that are accomplished with a gimbal. The two systems work together to handle both the high frequency oscillations due to engine vibration and turbulence as well as large low frequency attitude changes. Results will include laboratory testing and simulation data to further prove the effectiveness of this specialized stabilization system.

Fundamental streamwise pressure distributions were measured at Idaho State University to address a primary objective of obtaining insight into the apparent evolution of air flow regimes at moderate Reynolds numbers in thin, wide rectangular ducts with converging inlets.Ducts of 3.030 and 4.077 mm spacings s (called three and four mm), having non-dimensional lengths of (L/s) ≈ 500.8 and 372.2 and widths (W/s) ≈ 114.6 and 83.2, respectively, were compared.Flow rates ranged from Reynolds numbers based on hydraulic diameter (Re Dh ) of about 980 to 10,000 and corresponding non-dimensional downstream pressure gradient parameters (-K p) covered the range 0.038 to 0.0061, respectively.At low Re Dh , downstream results for apparent friction factors f ds from both test sections agreed with the laminar fullydeveloped predictions and for the high Re Dh they showed approximate agreement with the correlation of Beavers, Sparrow and Lloyd [J.Basic.Eng 1971] for fully-developed turbulent flow in comparable ducts.For three mm, apparently transition started within the test section length for Re Dh greater than about 2700 and for the four mm test section at Re Dh ≈ 3000.The downstream pressure gradient parameter data indicated that the flow would remain laminar if (-K p,ds ) is 0.022 or greater for the three mm test section or 0.021 or greater for the four mm one.The hydraulic entry behavior of the local apparent friction factor f app {x} of the constant crosssection duct downstream of the end of the inlet bevel showed apparently laminar flow initially for all experiments.These entry data followed (above) the numerical predictions of Schade andMcEligot [IJHMT 1971] for laminar flow in the entry of an infinitely-wide parallel plate duct with a uniform entry velocity.Examining the data with the streamwise Reynolds number as the ordinate demonstrated that, for Re Dh ≈ 4000 or greater, transition onset (chosen at the minimum f app ) is near constant around Re x,to ≈ 1.7 x10 5 .As a consequence, in laminar coordinates the location x to *{Re Dh } = (x to -x bevel ) /(sRe Dh ) systematically decreases as Re Dh increases.A secondary objective ---to assess design-style computational fluid dynamics (CFD) predictions from a popular "standard k-ε two-layer turbulence model" ---was also pursued.The turbulent kε CFD predictions over-predict the entry f app {x}, by as much as a factor of two, because it is treated as a turbulent flow rather than the laminar entry flow implied by the experiments.The kε calculations predict a turbulent entry length of about forty spacings but the experiments indicated it can be much longer.It appears that ---for Re Dh > ~ 4000 ---transition would have occurred in the growing near-wall boundary layer.We can obtain further insight into the developing flow phenomena in the hydraulic entry region at these higher Re Dh by considering the transition as "bypass transition," i.e., transition primarily induced by freestream turbulence, and by employing available direct numerical simulations (DNS) of representative bypass transition boundary layers such as those of Zaki and colleagues [JFM 2013[JFM , 2016]].For the most part, the present data for flow in the hydraulic entry of a wide rectangular duct show the same flow friction features as the DNS of bypass transition, possibly indicating the presence of the same phenomena / structures.

The development of the systems analysis codes in use today was a very challenging task, stemming from the interplay of multiple physical phenomena, special components and control systems, and particularly the wide thermodynamic state envelope for a typical design basis accident scenario that includes single and two-phase behavior of the water working fluid within both the reactor vessel and the steam generator for the indirect cycle pressurized water reactor systems and also for the direct cycle boiling water reactor systems. The major developmental work leading to the current systems analysis codes was performed between the 1970s through the 1990s—and today these analysis tools are used throughout the world by organizations that design, submit their designs for licensing reviews, build, and operate light water reactor nuclear power plants. Differences in form of the discretized equations and closure relationships used within the systems analysis codes versus those in higher-fidelity computational fluid dynamics (CFD) codes lead to correspondingly different techniques to verify and validate (V&V) the equations in these two classes of codes. Systems analysis codes use a fundamental approach which has been developed over the years and which has been approved by the regulatory agencies whereas the CFD codes use high-fidelity V&V techniques as described in the ASME V&V standards for computational fluid mechanics and heat transfer codes. Because of the wide usage of high-fidelity CFD codes together with systems analysis codes, it is advisable to normalize the techniques for verifying, validating, and performing code adequacy assessments of these tools within the methodology that is presently available in the U.S. Nuclear Regulatory Commission’s Regulatory Guide 1.203. A strategy to begin closing the gap between the fundamental approach used to V&V systems analysis codes and the high-fidelity techniques used for modern CFD codes is outlined. It is postulated that the gap can be closed to the extent that some of the “high-fidelity” techniques may be used for systems analysis codes and thus enhance the quality of the code adequacy determination process for systems analysis codes.

Richard R. Schultz Chair, NEDThe ASME Nuclear Engineering Division (NED), the sponsoring organization for both the creation of the Journal of Nuclear Engineering and Radiation Science and key supporter of Professor Igor Pioro as the Editor-in-Chief of the Journal, is delighted to note that the journal has been successfully publishing high impact technical articles for the benefit of the nuclear community since 2015. As is true from the very beginning, the ASME Nuclear Engineering Division once again wishes the Journal and in particular Editor-in-Chief I. Pioro every success in the publication of the excellent Journal of Nuclear Engineering and Radiation Science.In keeping with the NED support of the Journal, and of course NED policy to implement and complement the position and practices of the American Society of Mechanical Engineers, the NED is acting to tailor its organization to better serve the needs of both. During the most recent meeting of the NED Executive Committee (EC), the EC voted to create a flexible and dynamic organizational structure capable of providing timely and comprehensive support for nuclear engineering-related conferences and summits (see Fig. 1). A major ingredient in the NED organization is NED-lead technical committees focused on current developments and industry-related topics that support the nuclear community, for example, centered on advanced reactors, thermal-hydraulics (including computational fluid dynamics as well as verification and validation), and codes and standards, etc. The focus areas for the technical committees are reflected in the track content of the NED-sponsored International Conference on Nuclear Engineering (ICONE) each year. The administrative committees will focus on the organization of conferences/meetings as well as the administrative needs of NED. Presently the NED cosponsors the Small Modular and Micro-Reactor Summit (SMMR), and Advanced Clean Energy Summit (ACES) in addition to sponsoring ICONE.The NED is positioned to have a place for all the membership in the technical committees and meetings/summits. As usual, the NED organization and focus areas are designed for maximum member involvement. The latest example is ICONE28, held virtually between August 4 and 6, 2021. Keynotes and Plenaries were aimed directly at major topics of interest in the nuclear community: “operating plant issues and experience,” “climate change,” and “advanced reactors.” Between 95 and 136 people attended these events virtually (of course even more conference attendees viewed the videos of these events at later times: not surprising since one of the quirky characteristics of international virtual conferences is the time ingredient, for example, 8 AM EDT, when ICONE28 was begun each day, was 2 PM the same day in Rome, 8 PM in Beijing, China, where the Chinese Nuclear Society (CNS) is based, and 9 PM in Tokyo, Japan, where the Japan Society of Mechanical Engineers (JSME) is based. CNS and JSME partner with ASME to host ICONE. ICONE28 featured nine panels, three workshops, and 14 tracks as listed below:ICONE28 Panel SessionsICONE28 Workshops:ICONE28 Tracks:Attendees were from 20 countries with China, Japan, the European Community, and the United States especially well represented.ICONE29 is scheduled to be held in Shenzhen, China between Aug. 8 and 12, 2022. The theme of ICONE29 is Nuclear Energy Innovation Power a Carbon Neutral Future. Text-only abstracts are due by Nov. 30, 2021, on the ICONE29 website.1 ICONE29 is slated to be a hybrid (virtual and face-to-face) conference.

Dr. Romney Duffey is an internationally recognized multi-disciplinary scientist, consultant, manager, speaker, author, and poet. Born on June 26, 1942, and educated in England, Dr. Duffey has over 50 years of unique experience in the UK, USA, and Canada on nuclear technology development, risk assessment, industrial safety, nuclear-system design, and accident analysis. As an applied physicist, his career has included a wide span of senior power-industry and government positions as researcher, executive advisor, senior manager, published author, lecturer, and consultant. Dr. Duffey is globally known as a developer of new concepts and designs with innovation advantages and market potential, for contributions to risk management and reliability applications, and to the enhancement of our understanding of the physical world. In addition to working in the USA, Canada, and the UK his international industrial, laboratory, and technical connections are worldwide.

Because of the important role that fracture plays in the behavior of ceramic UO2 fuel in a nuclear reac-tor environment, fracture models are a major component of fuel performance codes. As with any aspect of fuel performance, it is crucial to validate these fracture models against experimental data; however, obtaining well-controlled data for conditions representative of a reactor environment is difficult. Quench-ing is proposed here as a relatively simple approach for using in a laboratory environment to achieve conditions that approximate those of a reactor environment. In this paper, an experimental apparatus containing a single instrumented fuel pellet in a sealed section of copper tubing is developed and ap-plied to a series of seven experiments in which the apparatus is first heated to a high temperature (580- 680 degrees C) by immersion in a molten salt bath, then quenched in a cold bath (-10-4 degrees C). Development of these experiments was guided by numerical simulations, and post-test simulations were performed to aid in understanding the experimental behavior and assessing the accuracy of the predictions of fracture initiation and propagation. In addition, similar experiments were performed on solid copper rods to pro-vide temperature-dependent heat transfer coefficients for use in simulations of these experiments. This study demonstrates that quenching is a viable approach for generating thermal gradients representative of those in prototypical light-water reactor conditions at powers of about 5-10 kW/m. Fracture is ex-pected to begin at these power levels, and moderate amounts of radial and axial cracking was observed in these quenching tests.(c) 2022 Elsevier B.V. All rights reserved.

This special section of the ASME Journal of Nuclear Engineering and Radiation Science contains selected papers from the 27th International Conference On Nuclear Engineering (ICONE-27)1 held in Tsukuba, Ibaraki, Japan, from May 19 to 24, 2019. Since the first meeting in 1991, the Nuclear Engineering Division (NED) of the American Society of Mechanical Engineers (ASME) and Japan Society of Mechanical Engineers (JSME) sponsored the ICONE meetings between USA, Japan, and Europe. In 2005, Chinese Nuclear Society (CNS) formally joined ASME and JSME and hosted ICONE meetings. As a result, the ICONE is now the premier global conference on nuclear-reactor technology. ICONE is for nuclear professionals, who want to stay technologically current and on top of industry trends and developments.The success of ICONE is due to the contribution of numerous professionals from industry, government, and academia from around the globe. Through the ICONE student program, the conference also fosters the development of future nuclear professionals. A major factor is ensuring high technical quality, with all papers being reviewed before acceptance, and the attraction of major plenary and panel speakers on key topics.The Japan Society of Mechanical Engineers with technical sponsorship from numerous national and international societies and organizations hosted the ICONE-27 conference. Atomic Energy Society of Japan (AESJ), Canadian Nuclear Society (CNS), Korean Nuclear Society (KNS), Atomic Energy Association of Vietnam (AEAV), Korean Society of Mechanical Engineers (KSME), and UK Institution of Mechanical Engineers (IME) provided supporting sponsorship for the meeting. The theme of the 27th ICONE was “Nuclear saves the world.” Coincidently, just one month ahead of the conference on Apr. 6, 2019, the New York Times published an opinion editorial “Nuclear Power can save the World.” The success of ICONE-27 is demonstrated by the contribution of 758 presentations and papers that were shared with an attending audience of more than 950 engineers, developers, power-plant maintenance professionals, researchers, regulators, and students representing 35 countries.The papers submitted covered the breath of topics within the multidisciplinary nature of nuclear engineering. To provide a manageable selection of thin quality work captured in the conference proceedings, the editors for this special edition decided to focus on papers that were initially recommended by the conference reviewers and session chairs as having archival value and then further screened by members from the international technical-program committee. Approximately, 50 authors were invited to update their papers and to submit for another round of double peer review process.Because of the in-depth review process, which has ensured high-quality papers, and the important topics that are discussed, we hope that this Special Section of the ASME Journal of Nuclear Engineering and Radiation Science will be a valuable addition to the book of knowledge. This is an appropriate time and place to mention and to thank a very large number of people, who give their time and expertise to make continued growth of the ICONEs. The success of the ICONE series of conferences would be impossible without them. First, thanks to all authors, their research and development work, ingenuity, and dedication to the nuclear industry. Reviewers volunteer their talent, time, and objectivity as they fulfill their responsibility. Reviewers may also be assigned to multiple papers within relatively short time. Of course, ICONE would be impossible without the track and session organizers as they were often responsible for inviting prospective authors, planning and organizing multiple sessions within the topical tracks. We wish to express our deepest gratitude to all of you for outstanding and continued support of ICONE series of conferences and for making possible the publication of this special topical section of the Journal of Nuclear Engineering and Radiation Science.Guest Editorial team:

The objective of the calculation performed and discussed in this report was to examine the behavior of the helium flow through a single cooling channel in the modular high-temperature gas-cooled reactor (MHTGR) core during a pressurized conduction cooling scenario (PCC). The region of interest for the single cooling channel flow calculation is in the upper reflector blocks. Flow from the upper reflector blocks, during the PCC scenario, moves upward and into the upper plenum.

Dominant phenomena are identified and characterized for the loss-of-forced-convection (LOFC) scenario, the depressurized loss-of-forced-convection (DLOFC) scenario, and the moisture ingress scenario with relief valves available scenario. These scenarios are considered for the modular high temperature gas-cooled reactor (MHTGR) baseline design—with both a prismatic core and a pebble-bed core. The objectives of this effort, described in the above paragraph, were achieved by following a methodology identified by the U.S. Nuclear Regulatory Commission (NRC) and recommended in their Regulatory Guide 1.203. The methodology employed is described. For the above scenarios, baseline phenomena identification and ranking tables (PIRTs) generated by a NRC-sponsored effort in 2008 for the LOFC and DLOFC scenarios and a baseline PIRT generated by a Department of Energy-sponsored effort in 2011 by their Next Generation Nuclear Plant (NGNP) Moisture Ingress Assessment Committee were used to identify the dominant phenomena and existing levels-of-knowledge. Ten dominant phenomena (with only medium or low levels-of-knowledge) were identified for the LOFC and DLOFC scenarios including: (1) Inlet plenum stratification and plumes, (2) core coolant flow distribution, (3) radiant heat transfer from the top of the core to the upper vessel head, (4) reactor cavity air circulation and heat transfer, (5) reactor cavity cooling system behavior, (6) decay heat distributions, (7) core effective thermal conductivity, (8) molecular diffusion in air ingress DLOFC, (9) duct exchange flow, (10) phenomena that affect cavity gas composition for air-ingress. Five dominant phenomena, with high levels of knowledge, were identified for the moisture-ingress scenario with relief valves available including: (1) flow through break, (2) moisture level in the primary, (3) transport to moisture monitor, (4) moisture monitor instrument response, and (5) heat removal by shutdown cooling system. These phenomena are discussed and characterized. A listing of experiments—both completed and ongoing—is given with a view toward including data useful for further more detailed examination of the dominant phenomena listed in the previous paragraph. Note that only data for the LOFC and DLOFC scenarios are listed—since the moisture ingress scenario with relief valves available –have been identified as having adequate levels-of-knowledge by the PIRT committee. Finally, a preliminary computational fluid dynamics (CFD) calculation of inlet plenum stratification and plumes is described. Techniques employed for this calculation are typical for preliminary scoping calculations performed to validate numeric tools.

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

A reference design for the Next Generation Nuclear Plant (NGNP) is to use General Atomics Modular High Temperature Gas-cooled Reactor (MHTGR). For such a configuration in normal operation, the helium coolant flow proceeds from the upper plenum to the lower plenum principally through the core coolant channels and the interstitial gaps (bypass flow) that separate the prismatic blocks from one another. Only the core prismatic blocks have coolant channels. The interstitial gaps are present throughout the core, the inner reflector region, and the out reflector region. The bypass flows in a prismatic gas-cooled reactor (GCR) are of potential concern because they reduce the desired flow rates in the coolant channels and, thereby, can increase outlet gas temperatures and maximum fuel temperatures. Consequently, it is appropriate to account for bypass flows in reactor thermal gas dynamic analyses. The objectives of this project include the following: fundamentally understand bypass flow and heat transfer at scaled, undistorted conditions and with geometry distortions; develop improved estimates of associated loss coefficients, surface friction and heat transfer for systems and network codes; and obtain related data for validation of CFD (computational fluid dynamic) or system (e.g., RELAP5) codes which can be employed in predictions for a GCR for normal power, reduced power, and residual heat removal operations.

The results of this project are best described by the papers and dissertations that resulted from the work. They are included in their entirety in this document. They are: (1) Jeff Harris PhD dissertation (focused mainly on forced convection); (2) Blake Lance PhD dissertation (focused mainly on mixed and transient convection). This dissertation is in multi-paper format and includes the article currently submitted and one to be submitted shortly; and, (3) JFE paper on CFD Validation Benchmark for Forced Convection.

In VHTR, helium from the reactor vessel is conveyed to a power conversion unit through a hot duct. In a hypothesized Depressurized Conduction Cooldown event where a rupture of the hot duct occurs, pressure waves will be initiated and reverberate in the hot duct. A numerical model is developed to quantify the transients and the helium mass flux through the rupture for such events. The flow path of the helium forms a closed loop but only the hot duct is modeled in this study. The lower plum of the reactor vessel and the steam generator are treated as specified pressure and/or temperature boundary to the hot duct. The model is based on the conservation principles of mass, momentum and energy, and on the equations of state for helium. The numerical solution is based on the method of characteristics with specified time intervals with a predictor and corrector algorithm. The rupture sub-model gives reasonable results. Transients induced by ruptures with break area equaling 20%, 10%, and 5% of the duct cross-sectional area are described.