Search

The annual production of dried citrus pulp within the Florida Citrus Industry has risen during the past twenty five years from a modest beginning to a total of over 330,000 tons per year. To this must be added another 52,000 tons of citrus meal and citrus molasses produced last year so that the combined total by-products production for cattle feed purposes during the 1958–59 season was approximately 380,000 tons. This was estimated to be worth over 14 million dollars. Paper published with permission.

Most of you have probably noticed the new addition to the Kraft Citrus Products Division Plant here in Lakeland. This is their new sectionizing plant for the preparation of fruit salad, which has only been in operation for a few weeks. Paper published with permission.

Almost four million tons of citrus fruit were used by the Florida processing plants last year. Juice obtained from citrus represents only approximately half of the weight of the fruit. Two million tons of peel, pulp and seeds, accumulated yearly at processing plants, presents a challenge and an opportunity. This is a large amount of raw material, readily available, concentrated in a relatively small number of locations, and produced over a period of time which makes possible economical handling and processing. Paper published with permission.

There has been a significant amount of discussion of the appropriateness of the design margins from ASME Boiler and Pressure Vessel Code Section VIII Division 3 being applied to the design of HPHT equipment. A technical report has been issued with a first revision which references the use of ASME Section VIII Division 2 and Section VIII Division 3 for design of HPHT equipment above 15 ksi. Additionally, there has been recent pressure testing which attempted to validate the margins used in ASME Section VIII Division 3 and API 17TR8. There has also been a significant amount of discussion that the equipment in HPHT service may be very different from the pressure equipment which utilizes ASME Section VIII Division 3 and the design methods discussed in the standards. This paper will discuss the background of ASME Section VIII Division 3, including an overview of how the standard began, and the types of equipment which were being considered when the document was first drafted. It will also discuss the background of the design margin used in ASME Section VIII Division 3 over the years and how the current margin came into play. It will also discuss the margins used in addition to the margins on the structural capacity of a piece of equipment and how the margins interrelate within the Code. This paper will demonstrate the robust nature of ASME Section VIII Division 3, and its applicability to the design of HPHT equipment.

This paper deals with the numerical analysis of an air spring that consists of two tanks connected by a long pipe. Two resonance points may appear in the frequency response of a vibratory system supported by this type of air spring despite the fact that the system has an apparent single mass. This phenomenon is caused by the presence of a secondary mass as reported in our previous paper. It was found that the secondary mass is the mass of air contained in the pipe. The magnitude of this mass is extremely small, but the acceleration of the air in the pipe — and therefore the inertia force generated from it — becomes very large. The generated force is further amplified by the Pascal’s principle and is transmitted to the supported mass. There are obvious nonlinear characteristics in this type of air spring; whereas the previous studies were based on linear assumptions. In this study, the governing equations for the air stream expressed by a nonlinear partial differential equation were solved by using the finite difference method. In particular, the pressure loss is evaluated due to air vortex being generated behind the orifice installed in the pipe. As a result of this study, it was found that the orifice is effective in suppressing the height of the secondary resonance point. Of course, it has become possible to accurately estimate the amplitude dependency of the dynamic characteristics of the air spring supported system by this non-linear analysis.

The objectives of this paper are to determine the structural and functional bending capacity, for an 18-3/4 inch 10 ksi Dry Completion Unit (DCU) external tieback connector (ETBC), utilizing more advanced analysis design verification procedures of API 17TR8 [1]. This requires the evaluation using a 3D nonlinear elastic-plastic finite element analysis (FEA). Although the design verification procedures of API 17TR8 [1] are written for high pressure high temperature (HPHT) equipment, this paper demonstrates the feasibility of applying it to lower pressure equipment which have more critical requirements for bending capacity vs. pressure capacity. The capacity results are compared between the existing linear elastic criteria of API 17G [2] and the requirements of API 17TR8 [1]. The application of this evaluation demonstrates that the analysis methodology for HPHT equipment can be applied to subsea tieback connectors to obtain maximum bending capacity of the connection.

This paper summarizes recommendations made as part of the work considering modernization of ASME Section I (Power Boilers). It is not intended to provide a comprehensive review of issues regarding design guidelines for components. However, to introduce the context of the approaches that are recommended, background is provided on a number of topics related to Design-By-Analysis, Design-By-Formula, Design-For-Safety and Design-For-Lifetime. Benefits, shortcomings or inconsistencies associated with particular approaches are discussed. It is apparent that no one current Code or Standard has a method that can be universally adopted. This paper outlines the features of a Design-By-Analysis approach that is both relevant, technically consistent and which considers the key modes of structural behavior and material response.

NORSOK U-001 4th edition was released October 1st, 2015. It replaces edition 3 from 2002. The 4th edition is based on industry experience and use of the previous version. With focus on Norwegian Continental Shelf (NCS) operations and experience, NORSOK U-001 provides normative reference and amendments to selected parts of ISO 13628 (equivalent to API 17 series). The focus areas of this paper are the informative annexes to NORSOK U-001: The revised Annex A (informative), and the new Annex B (informative). These annexes provide NCS related amendments to Part 4 (API 17D), Part 7 (API 17G), and Part 15 (API 17P) of ISO 13628. Annex A (informative) provides loads for design and selection of subsea trees, wellheads and templates and has been completely revised with new standard design loads. Annex B (informative) gives recommendations for structural design, manufacture and testing of subsea wellhead and tree equipment (ISO 13628-4/API 17D) and completion/workover riser systems (ISO 13628-7/API 17G). This paper provides a brief introduction and background to Annex A and Annex B. Some differences between NORSOK U-001 Annex B, ISO 13628-7 (API 17G), ISO 10423 (API 6A), API 6X, ISO 13628-4 (API 17D), ASME VIII Div. 3 and ASME VIII Div.2 are also highlighted.

The design of a brand new reciprocating compressor cylinder is a difficult and time expensive task, involving CAD, FEA and CFD tools, which can take several months. Following standard approach, usually several trials are needed in order to achieve the most suitable solution. Every time it is necessary to change some design/model features, all the following operations have to be repeated: modifying the CAD model; meshing the model; calculating new loads and/or boundary conditions; configuring the results analyses; post process the results. Following standard methods each described operation has to be performed manually by a user. This paper describes a method to achieve the automation of most of these manual operations, transferring the workload from human to hardware resources, with a consequent saving in time needed to achieve results and related quality. Paper will describe the process (including details about FEA and CFD analyses) to design a cylinder with this method, compared with a traditional one. The comparison will show differences in delivery time, weight and total cost (engineering + production) between the cylinders.

The UK Small Modular Reactor (UKSMR) programme has been established to develop an SMR for the UK energy market. Developing an SMR is a multi-disciplinary technical challenge, involving nuclear physics, electrical, mechanical, design, management, safety, testing to name but a few. In 2016 Upadhyay & Jain performed a literature review on modularity in Nuclear Power. They concluded that although modularisation has been utilised in nuclear to reduce costs, more work needs to be done to “create effective modules”. Hohmann et al also concluded the same for defining modules in the chemical process plant industry. The aim of this paper is to further define modules with a particular focus on an SMR for the UK market, the UKSMR. The methods highlighted may be relevant and applied to other international SMR designs or other types of plant. An overview and examination of modularisation work in nuclear to date is provided. The different configurations are defined for the Nuclear Steam Supply System (NSSS) in primary circuits and then for Balance of Plant (BOP) modules. A top level design process has been defined to aid in the understanding of design choices for current reactors and to further assist designing balance of plant modules. The paper then highlights areas for additional research that may further support module design and definition.

The lifetime extension of the nuclear power plants is considered as an energy challenge worldwide. That is why, the risk analysis and the study of various effects of different factors that could potentially represent a hazard to a safe long term operation are necessary. These structures, often of great dimensions, are subjected during their life to complex loading combining varying mechanical loads, multiaxial, with non-zero mean values associated with temperature fluctuations and also PWR environment. Based on more recent fatigue data (including tests at 300°C in air and PWR environment, etc...), some international codes (RCC-M [2], ASME and others [3][4][5]) have introduced a modification of the austenitic stainless steels fatigue curve combined with a calculation of an environmental penalty factor, namely Fen, which has to be multiplied by the usual fatigue usage factor [6]. Unfortunately, experimental data on this issue are rare. In order to obtain fatigue strength data under structural loading, biaxial test means with and without PWR environment were developed at LISN in collaboration with EDF and AREVA [6]. Two kinds of fatigue device have been developed. Within the same specimen geometry, structural loads can be applied in varying only the PWR environment. The first device (FABIME2) is devoted to study the effect of biaxiality and mean strain/stress on the fatigue life [9]. A second and new device called FABIME2e is for the study of the environmental effect. With these new experimental results, the PWR environment effect on the fatigue life of stainless austenitic steels will be quantified accurately on semi-structure specimen. This device combines the structural effect like equi-biaxiality and mean strain and the environmental penalty effect with the use of PWR environment during the fatigue tests. The aim of this paper is to present the numerical interpretation of the results obtained with these two devices “FABIME2” and “FABIME2e”. Two important aspects will be addressed. The first concerns the mechanical behavior of austenitic stainless steel and the capabilities of the numerical model to reproduce the hardening of the material. And the second concerns the study of the crack growth during the equibiaxial fatigue test.

High pressure testing and operation of industrial tools and equipment can be hazardous to personnel and property in the event of an accidental mechanical failure or release of contained pressure. Hazard barricades are commonly installed around equipment to protect nearby personnel or property from projectile impacts or overpressure. The ASME Standard PCC-2 “Repair of Pressure Equipment and Piping” allows the use of hazard barricades for this purpose when a safe standoff distance cannot be satisfied, but it currently provides minimal guidance for engineered design. Other references provide guidance for preventing projectile perforation of a barricade. While perforation prevention is a key component of shield design, properly anchoring a shield and inter-connecting the shield components will make the difference between an effective barricade application and a barrier that could potentially compound the consequences of an accidental failure. This paper investigates the importance of engineered structural connections and consideration of global structural response in the design of protective barricades. The structural models focus on impact loading of steel plates and bolted connections, and the results are directly compared to test results in terms of effective barrier response.

Considering “Pipeline Industry’s Need of the Hour - Self Reliance and Emerging Technology Trends” as theme of pipeline conference, my submission of paper is on several use cases on technology and innovation for pipeline’s different life cycle phases. There are continual developments in pipeline industry and several of technologies were successfully implemented or under consideration phase for implementation for pipeline industry. In all phases of pipeline lifecycle of material management, project implementation as well in Operation & Maintenance (O&M) of pipeline. Several technological changes as well several of innovations in existing systems are of extensive use for enhancing pipeline safety, reliability, monitoring system as well ease of operations and maintenance. Many of systems been used for other purposes and are of great use for pipelines. This paper discusses on few of such technologies and their implementation for pipelines.

In the current paper the reliability of spring interconnections for the use in automotive applications was investigated. A commercially available connector was used for contacting a piezo-jet printed silver pad on polycarbonate. The structures were tested using vibration tests at three different temperatures (−25°C, 25°C and 85°C) and measured using four-point probe measurement system and white light spectrometry. The obtained results showed the reliability of the printed structure and the contact, despite of the contact resistance change. The reasons of this change and experimental results are discussed in the paper. The investigations described in the paper provide an initial approach to the development of spring interconnections for printed electronics.

This paper discusses the design of a removable hardened test pit lid for conducting high pressure pneumatic testing under extreme temperature variations. The high pressure extreme temperature (HPET) pit configuration requires a robust lid to contain the energy release that could occur through testing a variety of oil services items placed within a frangible thermally insulated enclosure. The insulated and hardened enclosures include an inlet and outlet for airflow. These energetic release scenarios represent postulated test item failures resulting in significant fragment and shock loads within the insulated enclosure. A 618in3 gate valve charged with 20,000psi of compressed nitrogen under a temperature range of −100°F to 600°F was used to produce the worst-case energy release in the event of a catastrophic failure within the test pit. In order to assess the structural response of the hardened pit lid, two sets of computational fluid dynamics (CFD) analyses were performed. A 2-dimensional axisymmetric CFD model of the gate valve was used to estimate the fragment velocity of its blind flange. A 3-dimensional CFD model of the test pit, insulated enclosure, and blast lid was used to estimate the applied blast loads. The test pit enclosure was assessed using non-linear dynamic finite element analysis using single-integration hexahedral continuum elements. The braces were modeled with layered shell elements and the bolts were explicitly modeled with beam and spring elements. All of the materials are modeled using a plasticity-based piecewise linear hardening constitutive model, with rate dependency and softening, fit to each material yield, tensile, and elongation limit.

This paper discusses the design of a hardened enclosure for high pressure extreme temperature (HPET) pneumatic testing of seals. The seal test fixture is located within a blast-fragment barrier (BFB) in case of a failure of the test fixture. The BFB is located within a hardened test room since the BFB has a 1-in annulus air gap to allow for cabling access to the test fixture, through which shock and gas pressure can propagate out into the room. A 30.5in3 (0.51) volume charged with 15,000psi (1034bar) of compressed nitrogen under a temperature range of −50°F to 400°F was used to produce the worst-case energy release in the event of a catastrophic failure of the seal fixture. In order to assess the structural capacities of the BFB and test room, an equivalent axisymmetric computational fluid dynamics (CFD) model of the test fixture, BFB, and test room was used to estimate the fragment velocity and applied blast load time histories to the BFB and test room. The BFB was assessed using non-linear dynamic finite element analysis. All of the materials are modeled using a plasticity-based piecewise linear hardening constitutive model, with rate dependency and softening, fit to each material yield, tensile, and elongation limit. The seal fixture generated fragment interactions with the interior surfaces of the BFB model using slideline and pinball contact algorithms.

The National Bureau of Standards (NBS) had undertaken a comprehensive study of underground soil corrosion of iron pipes and plates. The maximum pit depth data for different types of wrought iron and carbon steel pipes have been widely analyzed and utilized in the corrosion literature. There is another important but relatively obscure data set about the testing of pipes with bituminous coating that NBS carried out in collaboration with the American Petroleum Institute (API). This program tested dozens of coatings on operating line pipes as well as short sections of pipes at 15 soil sites over a 10 year period (1930–1940). This paper presents an overview of this data and presents statistical analysis of protection offered by coatings.

Detailed analysis of dissimilar metal welds (DMWs) is difficult and the accuracy of the results is often limited by the availability of materials data and knowledge of operating conditions. There are few design or assessment procedures that explicitly take account of the presence of DMWs under high temperature creep and fatigue conditions. One that does is the UK R5 procedure, which provides a simplified assessment approach based on laboratory testing and stress analysis supported by service experience. This paper briefly summarises the R5 approach. Then, more recent work on performance of DMWs operating at high temperature and issues for development of analysis methods is discussed.Copyright © 2016 by ASME

In geohazard areas, buried pipelines are subjected to permanent ground deformations, which constitute major threats for their structural safety. Geohazards include seismic fault movement, liquefaction-induced lateral spreading and slope instability, and the corresponding deformations induce severe strains in the pipeline. Calculation of these strains is necessary for assessing pipeline integrity. In the present paper, an analytical methodology is presented and compared with existing analytical and numerical methodologies for stress analysis of buried pipelines. The proposed methodology is also compared with full-scale available experimental results and offers significant intuition on the behavior of buried pipelines subjected to permanent ground deformations. Using the proposed analytical methodology, one may predict the strains developed in the pipeline wall quite efficiently and with good accuracy.

Electrical power is a resource humans heavily rely on, and it has become a basic human need. Today, the major sources of electricity generation are fossil fuels, renewable energy, and nuclear power. This paper concentrates on electricity generated through nuclear power and compares it to the other electricity generation technologies. The objective behind this paper is to discover the impact that nuclear power has on the total electricity generated in Canada, and in addition on a global scale. The paper presents the current role that nuclear power plays in the global electricity generation, and also the expansions that need to be made in the nuclear power industry to fulfill the future electrical power demands. A number of projections have been made based on the current rate of nuclear reactors being put into operation, which is approximately 4 reactors per year, and current term of reactor operation, which is 45 years. These projections were made for the nuclear power in the world. A major outcome of this analysis projects that between 2030 and 2035, the number of operating nuclear reactors in the world can drop by 50%. If this dangerous trend is not addressed, we can lose a viable, and reliable source of energy. The datasets that were analyzed during the process were taken from multiple open literature sources such as journals, reports, and online databases. The paper presents a comparison between nuclear power and other energy sources, and the positive impact nuclear power can have on the world if needed advancements were made in building new nuclear power plants.Copyright © 2016 by ASME

High temperature gas cooled reactor (HTGR) which is one of generation IV reactor has been widely given attention in many countries since the sixties of the last century because of its inherent safety and high efficiency. Currently, the HTGR commonly uses regenerative Brayton cycle. However, as reactor outlet temperature (ROT) rising, regenerative Brayton cycle has a higher reactor inlet temperature (RIT) than 500°C and is limited by reactor materials. Combined cycle of HTGR not only can solve the problem of high RIT, but also can get a higher cycle efficiency than 50%. In this paper an accurate model of combined cycle consisting of topping Brayton cycle, bottoming Rankine cycle and heat recovery steam generator (HRSG) was established. In terms of new model of combined cycle, this paper analyzed the main properties of simple combined cycle. And put forward two optimization schemes improving the cycle efficiency of combined cycle.Copyright © 2016 by ASME

Modern industrial instrumentation and control systems (I&Cs) used in nuclear power plants (NPP) are facing more with cybersecurity threats and vulnerabilities, which were neglected before. Cybersecurity incidents are a subject to grow into more complex attacks with worse consequences than before. The use of field programmable gate arrays (FPGA) in such critical systems causes specific risks for ensuring of safety, as the master-property of such kind of systems, and security as a subordinate property primarily to the NPP reactor trip systems (RTS). Cybersecurity assessment results of industrial I&Cs are mainly based on subjective assessment of the expert judgment and they do not take into account all features of propagating FPGA technology. Nowadays there is a big gap in understanding how to assess and assure the security of FPGA-based NPP I&Cs (FNI&Cs). Conformance of FNI&Cs to security requirements, their verification to high-level standards often is subjective and depends on particular expert. Regulatory and certification bodies, developers and end-users of FNI&Cs are missing the understandable methodology for security assurance of such kind of systems taking into account specific context of the operating environment which allows decreasing time-to-market and thus providing benefits for all interested parties. The paper describes cybersecurity assurance technique of multi-version FNI&Cs. Requirements profile is formulated using the best practices from the following international regulations. The goal of the paper is presentation of the case-based methodology and tool of FNI&Cs cybersecurity assurance based on international regulations. Proposed methodology provides comparable and repeatable process of assurance.

In the paper, a new test bench for characterization and durability assessment of multi disc wet clutches for motorbike applications is presented. The design of the new test bench is inspired by the current SAE standards J2490 and J286 that refer to the testing of friction material for automatic transmissions wet clutches. Differently from the test bench described in the two standards, which is mainly designed to test the friction material, the test bench presented in this paper has been designed for testing the whole clutch, comprising all discs and the actuation mechanism. The lubrication system of the clutch utilized on the test bench is very similar to the one present on the actual motorbike. The design and instrumentation of the new test rig is analysed and discussed in the paper. The test bench has been used for the characterization of different motorbike clutches. The bench has proved to be able to measure the clutch characteristic for different temperatures and clutch wear levels. A very good repeatability of the tests has been achieved allowing for a comparison of different clutches design, friction material properties and operating conditions (temperature, ..). Some experimental results are presented and discussed in the paper.

Incorporating user trust in the development of intelligent systems is one of the new challenges in engineering design field. Trust in human-intelligent system interaction determines how much user relies on the system and directly influences the benefits that an intelligent system provides to human decision-making. This paper reviews the existing literature on trust in human-AI interaction to highlight key areas for engineering system design research to address the overarching issue of user trust in system development. We present how trust influences the use of an intelligent system, describe multiple contexts that users interact with intelligent systems and categorize ways in which trust is formed based on the literature. We classify the key factors that are critical in the formation of user trust in three categories of human user attributes, design of the intelligent system, and task characteristics. We also present the analytical models that exist in the literature used to evaluate and predict trust. This paper is not intended to be a thorough literature review but rather a position paper that provides a structure to the existing literature as a reference for engineering design research. We propose future directions for engineering system design community based on the gaps and open questions identified in the literature.

Simulations were used in lecture-based courses, Thermodynamics I and Thermodynamics II, for demonstration purposes and as digital assignments. Digital assignments were separate from the customary pencil & paper homework given once a week. The primary objective was to model the flow and heat transfer in a system, reinforce theoretical concepts, and allow study of more complex two- and three-dimensional problems. Students encounter fluid flow and heat transfer in the sophomore course, Thermodynamics I, for the first time. Velocity, pressure and temperature fields are quite complicated even for simple geometries, hard to visualize, and difficult to understand. Digital assignments go beyond basic theoretical concepts and relatively simple pencil & paper problems. Initially, there were 6 digital assignments in each course. After the first offerings and assessments, the number of assignments is expected to grow to 8–10. Scaffolding of digital assignments ensured that simulations gain in complexity as students become more familiar with both the theoretical underpinnings and the software interface. Our Mechanical Engineering program, has a freshmen graphic communication course incorporating AutoCAD and a junior CAD course with SolidWorks and Ansys. This provided both a challenge and an opportunity for lecture-based thermo-fluids courses positioned ‘in between.’ Modern computational skills were obtained outside the classroom, and that was accomplished with effective use of an on-line environment. Ansys Fluent was chosen to present and reinforce thermo-fluids fundamentals. Each digital assignment had a mesh, detailed grading criteria, and supplementary documentation. Students worked on the: (1) problem set up in the preprocessor-stage, (2) solver, and (3) postprocessor. This approach helped students (1) understand the flow and heat transfer inside the system as well as the application of conservation of mass and energy, (2) understand and interpret results by comparing them with theoretical and experimental data, (3) develop modern technical skills relevant to the demands of Industry 4.0, and (4) develop research capabilities. Simulations in the classroom and as digital assignments are a representation of the real world, and provide an egalitarian and unrestricted way for students to interact with it, acquiring skills and an appreciation for subject matter as well as the engineering profession. Semester long discussions on various aspects of digital assignments help develop a mentor-mentee bond. This approach to teaching, research and mentoring does not depend on access to laboratory facilities and funding that can reach only a limited number of students. It encourages students’ spirit of inquiry, and ultimately leads to a professional development opportunities beyond the classroom setting.

The accelerator-driven subcritical system (ADS) can transmute nuclear waste by using high-energy neutrons, which is the most effective nuclear waste disposal scheme at present. The zero-power lead-bismuth alloy device accurately constructed the interaction mode between nuclear fuel and lead-bismuth alloy coolant material, and more accurately simulated the core physical characteristics of lead-bismuth reactor. Moreover, the inherent safety of the reactor is enhanced by installing several passive safety shutdown systems. However, due to the particularity of the potential harm caused by radioactive material leakage, nuclear-grade equipment should maintain high safety performance under many working conditions, such as earthquake working conditions. To ensure the safe operation of the ADS zero-power device and the safe shutdown under seismic conditions, the motion of the control rod of the ADS zero-power device must be analyzed under seismic conditions. At present, most studies on the control rod drop analysis focus on water reactor, but this paper will analyze and calculate the control rod drop of the lead reactor. In this paper, the physical mechanism of rod falling of the control rod component is analyzed, and the influence of contact collision between the control rod component and the guide tube of the control rod on the whole rod falling time is considered. In this paper, the finite element method is used for transient analysis of ADS lead-bismuth alloy zero power reactor, and the time history diagram of the control bar position is obtained under the seismic conditions, thus verifying that ADS lead-bismuth alloy zero power reactor can be safely shut down under the safe shutdown earthquake (SSE).

Turbine Control System (DEH) and Turbine Protection System (ETS) are important auxiliary systems for turbines. Normally DEH and ETS are supplied by turbine manufacturers. This paper investigates the operation experience in Nuclear Power Plants (NPPs) in China, there are many problems including information security, maintenance inconvenience, long supply cycle guarantee, high cost and so on in DEH and ETS. This paper analyzes the problems and tries to get the reasons why DEH&ETS have these problems. The Distribute Control System (DCS) technology has been widely used in nuclear power plant in now days, so this paper puts forward a technical solution based on the safety DCS and non-safety DCS platform to realize the DEH and ETS, and according to the feasibility analysis of products, and the test results based on engineering prototype, the solution can solve the problems of DEH and ETS effectively, some performances of the DEH&ETS have been improved, and the solution has a good prospect in further.

Nuclear technology, as a high quality, clean and reliable energy supply, is attracting broad interest from countries across the world. F321 austenitic stainless steel (F321SS) is widely utilized in key components of nuclear power plant due to its excellent corrosion resistance and high temperature mechanical properties. Irradiation can easily lead to the degradation behaviors of materials, such as irradiation hardening, irradiation embrittlement and high-temperature He embrittlement, etc. Understanding such degradation is important for predicting the evolution of material behavior under irradiation and extending the lifespan of existing nuclear reactors. Ion irradiation is most commonly used to model neutron-induced damage since the irradiation conditions (temperature, flux, spectrum, etc.) can be regulated more accurately and flexibly. In this paper, the Fe-ion irradiation experiments of F321SS at different temperatures and doses were carried out, and the nanoindentation experiments under different conditions were further conducted. Irradiation hardening is observed in all specimens and strongly depending on irradiation temperature and damage dose. The hardness after irradiating increases with doses and saturates for at least 1dpa under low temperature regimes (< 300°C). However, at higher temperature (450°C and 560°C), nano-hardness reaches the peak at ∼0.5dpa and then declines. Moreover, the hardness of all specimens has a similar trend with temperature, that is, it first increases, reaches the peak, and then decreases.

With the increasing demands of versatile and long-lasting requirements of deep space missions, space nuclear reactor (SNR) power system is becoming the most potential energy source compared to conventional solar-battery and chemical energy in the future large-scale and long-life space missions. Since the very first successful launch of SNR system (SNAP-10A) in 1965, the United States, the Soviet Union / Russia, the European Union, Japan, Brazil and China have proposed various technical routes and schemes for SNR technologies over the past few decades. This paper presents a historical review of several key technologies (i.e., space nuclear reactor system and thermoelectric conversion system), especially focus on technical progress for recent decade and discusses on-going development activities. The paper also presents a forecast of potential future space applications of these key technologies.

The rules for the design of high temperature reactor components in Section III, Division 5, Subsection HB, Subpart B (HBB) of the ASME Boiler and Pressure Vessel Code contain two options for evaluating the deformation-controlled design limits on strain accumulation and creep-fatigue: design by elastic analysis and design by inelastic analysis. Of these options design by inelastic analysis tends to be less overconservative and produce more efficient designs. However, the HBB currently does not provide approved material models for use with the inelastic analysis rules, limiting their widespread use. A nonmandatory appendix has been developed to provide general guidance on appropriate material models and provide reference material models suitable for use with the design by inelastic analysis approach. This paper describes a viscoplastic model for Alloy 617 suitable for use with the HBB rules proposed for incorporation into the new appendix. The model represents the high temperature creep, creep-fatigue, and tensile response of Alloy 617 and accurately accounts for rate sensitivity across a wide range of temperatures. The focus in developing the model was on capturing key features of material deformation required for accurately executing the HBB rules and on developing a relatively simple model form that can be implemented in commercial finite element analysis software. The paper validates the model against an extensive experimental database collected as part of the Alloy 617 Code qualification effort as well as against specialized experimental tests examining the effect of elastic follow up on stress relaxation and creep deformation in the material.

The ASME Code, Section III, provides rules for the evaluation of cyclic loading that include evaluations of fatigue. These rules were first developed in the 1960s along with the original development of Section III. Since then, technology and analytical advancements have improved our understanding of the fatigue phenomenon. As an example, the developments in environmentally assisted fatigue (EAF) have become a major consideration in the low-cycle fatigue for light water reactors and plant life extension. Fatigue usage has an outsized impact on the overall cost and complexity of nuclear plants, which goes beyond just the qualification of ASME components. For example, pipe break locations, pipe whip restraints, and inservice inspections are, in part, based on fatigue analysis results. Therefore, it is important that fatigue analyses provide the most realistic assessment of fatigue with appropriate conservatism. The Section III Standards Committee has been working to keep the Code up to date with the latest research, analysis tools and fatigue data. In 2017, a Fatigue Steering Committee was formed to provide a venue for stakeholders to identify and coordinate development of updated fatigue design criteria, analysis methods, and their associated physical and thermo-mechanical properties for application to Section III of the ASME BPV Code. The stakeholders include many international organizations including the Japan Society of Mechanical Engineers. Many PVP papers have been used as the basis for the Fatigue Action Plan items. The Steering Committee completed its work in 2019 with the adoption of a Fatigue Action Plan, which provided a vision for advanced fatigue approaches. The Fatigue Action Plan recommended specific Code actions and activities to guide and coordinate the Code actions in the area of fatigue. This paper outlines the overall Fatigue Action Plan and provides an update on the implementation progress. Section III has already accomplished several actions on the Fatigue Action Plan by publishing numerous Code Cases and Code revisions that have improved the analysis tools available to designers to address fatigue. This paper provides an overview and introduction to these Code Cases and other tools for designers. The Fatigue Action Plan was largely based on addressing the need for the light water reactor fleet to address low cycle thermal fatigue. The Committee also recognizes that going forward the Fatigue Action Plan must address the needs of the growing advanced reactor community as well as other users of the Code. The paper discusses these needs and Section III plans to meet these future needs.

For many nuclear power plants worldwide the operation period will be extended to 60 or 80 years in the coming years. As the operation period increases, the importance of knowledge of ageing mechanisms increases. In the framework of LTO there is limited knowledge about ageing and structural integrity of concrete structures. Recent developments have shown that ageing of civil structures receive more attention internationally. In order to increase the knowledge in the field of civil structures, this paper focusses on investigation of ageing of civil structures and determining an ageing management strategy. Knowledge of the ageing mechanisms of civil structures and especially concrete, will lead to improvement of ageing management and assessment methods of concrete. As a first step international information was gathered on civil structures ageing issues and management thereof (see PVP2019-93029). In addition a highlevel assessment methodology was proposed. In the next step the initially proposed assessment methodology has been tested by application to a nuclear reactor. The resulting list of relevant AMPs has been verified with the outcome for another PWR with a steel containment. With this experience the assessment methodology is tested, compared and improved (see PVP2020-21838). The results indicated that the method can be used to obtain a list of plant specific AMPs. What was added to the assessment method is the link to the TLAAs for civil structures. In this follow up step the transition is made from a high level of IGALL AMPs to a practical AMPs that will deal with the right mechanism at the right location. The detailing to a level of practical work instructions for the maintenance of the plant has to be made in order to make real life implementation possible. In this step studying of relevant degradation mechanisms, relevant AMPs (like IAEA AMP305 and AMP306 ) and applicable literature in combination with the practical knowledge from operation of a reactor, has taken place. The international developments on ageing management of concrete will be included. The goal of the project is to obtain more knowledge on ageing management of civil structures and especially concrete. It will lead to an assessment method for civil ageing management and ageing management programs dealing with the relevant mechanisms at the various locations in a practical manner. The results of this ongoing work are presented in this report. For the research reactor all SSCs in scope of the Continued Safe Operation could be linked to the relevant AMP(s) and a resulting set of plant specific AMPs for civil ageing management was obtained. Including the international developments, literature and guidelines, a more general applicable list was created (Table 5 through Table 13). The conclusion is that Figure 2 represents a practical method for obtaining a set of plant specific civil AMPs ready for implementation. For representation in this paper the final outcome is given in as a generic list of actions for a generic reactor (Table 5 through Table 13). In these tables the relevant SSCs, ageing mechanisms and actions are listed. The tables represent an generic list of actions for civil ageing management that might others help develop their ageing management program. Future steps are shifting the focus from the general but practical assessment methodology to finite element modelling techniques for concrete. The assessment criteria for concrete (e.g. in ASME III, ASME XIII or Eurocode) will be investigated and investigation on the modelling of the concrete for ageing are planned.