bluuebunny/crossref_metadata_2025 huggingface dataset:
Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China
Authors: Chen Haiying , Wang Shaowei , Tian Xinlu , Liu Fudong
Published: October 2021

The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.
Validation of Evaluation Method of Feedback Reactivity for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors
Authors: Yoshimura Kazuo , Doda Norihiro , Tanaka Masaaki , Yamano Hidemasa , Igawa Kennichi
Published: October 2021

Negative reactivity caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In this paper, the numerical results of the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests in the pool-type experimental SFR in the United States, EBR-II (BOP-302R and BOP-301) are discussed in order to validate the evaluation method of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD. During the transient of the ULOHS tests as one of the representative issues, the reactor power decreased to the decay heat level due to the negative reactivity caused by the radial expansion of the core. By comparing the numerical results and the experimental data, the profiles of the increase of the core inlet temperature and the decrease of the reactor power calculated by Super-COPD were comparable with those of the experimental data. Applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS event. The refinement of the plenum model of the cold pool by taking account of the thermal stratification was indicated for the future work in order to improve the profile of the core inlet temperature.
Corrosion Behavior of Iron-Chrome Alloys in Liquid Bismuth
Authors: Takai Toshihide , Furukawa Tomohiro , Watanabe Shigeki , Ishioka Noriko S.
Published: October 2021

For the mass production of astatine-211, a promising radiopharmaceutical for cancer treatment, the National Institute for Quantum and Radiological Science and Technology has proposed the innovative “Liquid Bismuth Target System.” The target window in this system must be made from a material that resists the highly corrosive liquid bismuth environment. To meet this requirement, a promising target window material was selected in corrosion experiments performed in stagnant liquid bismuth. Based on knowledge of corrosion in liquid lead–bismuth eutectic gained during the development of fast reactors and accelerator-driven subcritical systems, FeCrMo–alloy, FeCrAl–alloy, and austenitic stainless steel (as a reference) were selected as the specimen materials. Experiments were carried out under saturated dissolved oxygen and low oxygen conditions, and the corrosion behaviors of the specimens were evaluated, mainly by scanning electron microscopy. The FeCrAl–alloy exhibited the most excellent corrosion resistance, followed by FeCrMo–alloy. Both materials are suitable candidates for the target window. Although austenitic stainless steel was less corrosion resistant than the former two materials, it is a likely applicable for the target window under appropriately limited operation conditions (such as irradiation current and exposure time) of the liquid bismuth target system.
Study on the Small Pressurized Water Reactor Based on Fully Ceramic Microencapsulated Fuel
Authors: Huang Jin Feng
Published: October 2021

After Fukushima nuclear power plant disaster, the efforts to overcome these defects of PWRs were carried out, such as replacing the cladding and fuel materials. One of these feasible efforts is using Fully Ceramic Microencapsulated (FCM) fuel replacement traditional UO2 pellets fuel into PWR.

The FCM fuels are composed of Tri-structural-isotropic (TRISO) particles embedded in silicon carbide matrix. The TRISO fuel can hold its containment integrity and without fission production releases under the design temperature limit of 1600 °C. Furthermore, the silicon carbide matrix will benefit the thermal conductivity, radiation damage resistance, environmental stability and proliferation resistance. Consequently, the safety of the reactor could be significantly improved with FCM fuel instead of the conventional UO2 pellet fuel in PWR.

We also analyzed the temperature distribution for the FCM fuel compared the traditional UO2 pellets, the calculation indicated that the centerline temperature is lower than UO2 pellets due to FCM higher thermal conductivity.

The calculation demonstrated that, utilizing FCM replacement of conventional UO2 fuel rod is feasible and more security in a small pressurized water reactor.

In this paper, a small pressurized water reactor utilizing FCM fuel is considered. A 17 × 17 fuel assemblies with Zircalloy cladding was applied in conceptual design through a preliminary neutronics and thermal hydraulics analysis. The reactor physics is accomplished, including the refuel cycle length, the effective multiplication factor, power distribution analysis being discussed. The Soluble Boron Free (SBF) concepts are introduced in small PWR, as a result, it makes the nuclear power plant more simpler and economical. FCM fuel loading has a very high excess reactivity at the beginning of reactor core life, and it is important to flat reactivity curve during operation, or to minimize excess reactivity during the core life. Consequently, conventional burnable poison configurations were introduced to suppress excess reactivity control at beginning of cycle.
Research on Nuclear Turbine Control and Protection System Based on DCS Integrated Technical Solution
Authors: Shi Guilian , Wang Jikun , Gao Jingbin
Published: October 2021

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.
A Sodium-Cooled Fast Reactor Simulation System and its Application In Teaching Research Based on VPOWER Platform
Authors: Ji Chengzhi , Xie Biheng , Guo Xiaoyu , Han Wenbin , Hao Yisheng , Chen Junyi , Huang Shanfang , Wang Kan , Wei Hongbin , Liang Yanming
Published: October 2021

To ensure the safe operation of nuclear power plant, employees must complete nuclear power plant-related safety training. College students who are majored in nuclear engineering in China cannot deeply participate in nuclear power plant work during the college period. So, they do not have a deep understanding of the theoretical concepts of nuclear power systems in class. For research and teaching work on Sodium-cooled fast reactor (SFR), the Department of Engineering Physics at Tsinghua University has developed the sodium-cooled fast reactor nuclear power plant simulation software (ECFR) with Neoswise Science & Technology Co. Ltd. ECFR is the dynamic process model software of the SFR nuclear power plant which is based on VPOWER platform. This paper presents the application of ECFR in teaching. Experimental courses have been conducted for Tsinghua students. Students can observe the main system structure of the power plant and perform accident analysis through ECFR. Those experiments can help students grasp the latest nuclear power technology better and motivate them to study effectively. Furthermore, the future improvement direction of ECFR is also discussed, which includes the improvement measures of human-computer interaction and system design.
Application of High Accuracy Numerical Methods for the Natural Circulation Problem
Authors: Chao Fei , Li Longze , Yang Wen , Qiu Jinrong , Tai Yun , Shan Jianqiang
Published: October 2021

Natural circulation refers to the phenomenon of fluid circulation driven by the driving force formed by the density difference and height difference. Natural circulation is very common in nuclear power system, which has an important impact on reactor safety. Therefore, accurate simulation of natural circulation phenomenon is one of the most important research contents in current reactor thermal hydraulic analysis. The low order difference method is widely employed to discretize basic conservation equations for current main thermal-hydraulics analysis code such as RELAP5, CATHARE, TRACE, etc. However, the low order difference scheme has the characteristics of large numerical truncation error and low calculation accuracy when simulating natural circulation problem. In this paper, the high-order accuracy numerical algorithm for two-fluid two-pressure model is used to simulate Welander natural circulation problem. Numerical results imply that the higher order difference scheme can accurately predict the unstable boundary and chaotic behavior with fewer grids, while the lower order difference scheme may get wrong results when the mesh is not fine enough, that is, the unstable mode of natural circulation can be predicted as stable mode. Therefore, high-order accuracy numerical algorithm could prevent excessive numerical diffusion effectively and improve the prediction accuracy, which demonstrates the advantage of using high-order schemes.
Towards Malicious Action Detection for Nuclear Security via Integrated Deep Learning Based Image Recognition and Natural Language Processing
Authors: Demachi Kazuyuki , Sudo Masaki , Chen Shi
Published: October 2021

In recent years, deep learning has been very successful in various fields. In particular, since image data contains a lot of information, it is expected that deep learning will be applied to nuclear security technology for detecting the malicious action. However, most human action recognition using deep learning only sets OK / NO in advance for the combination of object identification result, so it is difficult to apply them to complicated nuclear security situation. On the other hand, in most nuclear facility, the nuclear security rules are determined as a detailed text. So, it is desirable to detect malicious actions in captured images according to these rule text. If it is possible to flexibly determine whether the captured image scene violates or matches the nuclear security rules, it is necessary to greatly improve its practicality. However, this requires different deep learning interfaces for image identification and natural language processing.

Most of the deep learning research currently being conducted is closed only within each field, and there is no mutual access, that is, an interface. The human brain, on the other hand, usually links and processes all kinds of cognitive information, including language, sight, speech, and touch. Like the brain, deep learning abilities should develop anew when different types of deep learning are interconnected. One of the reasons this is difficult is that there is no common data format between different types of deep learning.

In this study, it was determined that the logical expression is the most suitable as a common data format. The logical expression is a method of expressing the correlation between an object and a person. In this research, a deep learning interface method between image recognition and natural language processing was developed via the logical expression. This allows flexible judgment of wrong actions according to the rule text.
Hybrid Dynamic Response Test Focusing on the Support Structure of Piping Systems
Authors: Okuda Yukihiko , Nishida Akemi , Sakai Michiya , Shiogama Yuzo , Li Yinsheng
Published: October 2021

To develop a more realistic seismic response evaluation method of nuclear power plants, it is necessary to evaluate the seismic behavior considering the joints of each component that are treated as independent models during design evaluation, such as buildings, equipment, and piping systems. Particularly, the piping support structure, which is the joint between the building and piping, is important in the seismic evaluation of the piping system. While the current seismic evaluation of piping support structures is performed within the elastic range, it is important to consider the realistic elastic-plastic response of piping support structures for fragility assessment in seismic probabilistic risk assessment. However, the seismic evaluation method that considers the elastic-plastic response of piping support structures has not yet been established, and there is a need to improve the seismic response evaluation method.

In this study, a hybrid dynamic response test for simulating the seismic behavior of the piping support structure, including the elastic-plastic response, has been conducted. Specifically, static cyclic loading tests and hybrid dynamic response tests were conducted using four types of piping support structures to understand the basic mechanical behavior. This report presents the details of the tests and test results.
Research on Quality Assurance of Raw Materials for Nuclear Power Equipment
Authors: Pu Chenghao , Jin Gang , Yi Zilong , Huang Jiaqi , Han Dongao , Li Maolin , Wu Qi
Published: October 2021

In terms of quality management of raw materials for nuclear power equipment, equipment manufacturers implement quality management for raw materials according to different standards, to ensure the quality and reliability of nuclear power equipment. However, through some typical cases, it is found that many manufacturers do not effectively operate the quality assurance system in the equipment raw material quality management, and there are not enough anti-counterfeiting measures, resulting in unqualified raw material quality affecting the quality of nuclear power equipment. For example, when a manufacturer purchased ASME code grade bars for manufacturing nuclear valves, the raw material supplier forged part of the test reports from the third-party laboratory to make the results meet the requirements of the standard. However, the manufacturer did not find the problem in the procurement process. By comparing the quality assurance requirements of nuclear power units under different standards, this paper analyzes the problems in the process of raw material purchasing and quality management, combined with some measures and requirements to prevention CFSI, puts forward some suggestions for optimizing the quality assurance of raw materials for nuclear power equipment.
Characteristics of Debris From Simulated Molten Fuel Coolant Interaction Experiments
Authors: Hemanth Rao E. , Shukla Prabhat Kumar , Varghese Paulson , Polaki S. R. , Vetrivendan E. , Das Sanjay Kumar , Ponraju D. , Athmalingam S. , Venkatraman B.
Published: October 2021

Studies on Molten Fuel Coolant Interaction (MFCI) resulting from a hypothetical core melt accident in Sodium cooled Fast Reactor is important as the MFCI phenomena is associated with generation of dynamic pressure which has a potential for mechanical damage to the reactor vessel and structures. During MFCI, molten core at high temperature interacts with relatively cooler liquid sodium and undergoes fragmentation. The size, shape and physical state of debris depend on the associated fragmentation mechanism, super heat of the melt and sodium temperature. The fragmented particles settle on core catcher and form as debris bed, which continue to generate decay heat. Characteristics of debris govern the debris bed coolability and needs thorough investigation. Experiments have been conducted by releasing a high temperature melt at 2400 °C into liquid sodium at 400 °C to study the fragmentation phenomena. Molten mixture of alumina and Iron generated by aluminothermy process was used for simulating the corium. After the experiment, sodium was drained and debris was retrieved by vacuum evaporation method. Size distribution and mass median diameters of the resultant debris was determined by sieve analysis. The debris was segregated into metallic/ nonmetallic particles using magnetic separation method and subjected to microscopic investigation for obtaining morphological characteristics. This paper outlines about the characteristics of debris generated by interaction of simulated corium with sodium towards assessment of fragmentation mechanism and associated debris coolability.
Development of Effective Momentum Model for Steam Injection Through Multi-Hole Spargers: Unit Cell Model
Authors: Wang Xicheng , Grishchenko Dmitry , Kudinov Pavel
Published: October 2021

The Steam injection through multi-hole spargers into the pressure suppression pool (PSP) is used in light water reactors to prevent containment over-pressure. The development of thermal stratification in the PSP can reduce its cooling capacity and results in higher containment pressures compared to completely mixed pool conditions. Explicit modelling of direct contact condensation (DCC) of steam at the steam-water interface is a challenge for contemporary codes. Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed to enable the prediction of thermal stratification and mixing transients induced by steam condensation in a large pool. The general idea of the EHS/EMS is to resolve the effect of the DCC phenomena on a large pool, instead of explicit modelling of the small-scale phenomena at steam-water interface.

The EHS/EMS models can be implemented using (i) respective boundary conditions at the boundary of the Steam Condensation Region (SCR) or (ii) using source terms in the heat and momentum transport equations. In previous work, EHS/EMS models were implemented using the second approach and validated against data from PPOOLEX and PANDA tests. It was found that results are sensitive to the spatial distribution of the source terms. Since the current data are not sufficient to provide a reasonable distribution, a preliminary study of the first method was done in this paper.

The goal of this work is to develop a ‘Unit Cell’ model by using respective boundary conditions for steam injection through multi-hole sparger. The condensed turbulent jet is resolved by introducing the liquid jet with the same effective momentum and heat as the injected steam. A uniform velocity profile solved by EMS model and the temperature boundary solved by EHS model is provided on each injection hole of the sparger wall. Validation is conducted against sparger test in PANDA facilities.
Study on Influence of Rolling and Heaving Motions on Differential Pressure and Flow Rate Measurements
Authors: Zhang Biao , Liu Jingyu , Li Xin , Qiao Shouxu , Li Dongyang , Tan Sichao
Published: October 2021

Ocean conditions will cause ships to be affected by rolling, heaving and other motions, which introduce additional inertial forces to the marine nuclear power system, introduce additional pressure drop to the fluids in each loop, bring errors to the measurement of thermal hydraulic parameters such as pressure difference and flow rate. Therefore, the measurement of pressure difference and flow rate under ocean conditions is studied and corrected in this paper. We built a rolling platform and a heaving platform and measured the differential pressure and flow rate under various working conditions. Through analysis, the relationship between differential pressure, flow fluctuation and swaying, undulating and other movements under moving conditions was obtained. Besides, the calculation method of attached pressure drop was analyzed, and the following conclusions were obtained: The influence of motion on differential pressure and flow rate is mainly composed of two parts: The additional pressure drop caused by the swinging motion of working medium in the pipeline and the pressure lead-in pipe, and the periodic fluctuation of friction pressure drop caused by the periodic fluctuation of flow velocity caused by swinging motion. The attached pressure drop caused by motion can be calculated through theoretical analysis. Through the analysis of experimental data and the comparison of theoretical calculation methods, it is found that the pressure drop measured by the experiment is in good agreement with that fitted by the theoretical calculation, and the modification scheme of pressure difference under ocean conditions is feasible to some extent.
Review of Determining Transformed Temperature Over 325°C Based on Monte Carlo Simulation for Environmental Fatigue Evaluation
Authors: Lee Bonghee , Nam Ilkwun
Published: October 2021

The NUREG/CR-6909, which had provided the technical background for the Environmental Fatigue Evaluation (EFE), was revised in May 2018 to incorporate the Japan Nuclear Energy Safety organization (JNES)’s extensive fatigue ε-N data and to address concerns from interested stakeholders over the original NUREG/CR-6909. But the revised NUREG/CR-6909 sets limit of the temperature ranges to 325°C for the application of the environmental fatigue correction factor (Fen). Therefore, the revised NUREG/CR-6909-based EFE is not applicable to the materials whose temperatures are higher than 325°C.

In this paper, Monte Carlo simulations have been performed to determine the transformed temperatures over 325°C for the materials which are not covered in the revised NUREG/CR-6909. To determine the distributions of fatigue lives in LWR environments, Monte Carlo simulations were performed using the fatigue life models in the LWR environments and transformed parameters provided in the revised NUREG/CR-6909 and JNES-SS-1005. The transformed parameters which affect the fatigue lives in the LWR environments were selected randomly using the uniform distribution. And a total of 100,000 simulations were performed for each material. The distributions of fatigue lives in LWR environments obtained from Monte Carlo simulations were fit to lognormal curves to estimate the cumulative distributions.

The estimated cumulative distributions were individually determined in the temperature ranges of 100 ∼ 325°C and 325 ∼ 350°C to investigate the temperature effects (FT,effect) over 325°C. The temperature effect is defined as ratio of fatigue lives in the LWR environments in the temperature range of 100 ∼ 325°C to that in the temperature range of 325 ∼ 350°C in this paper.

As a result of reviewing the temperature effects between the revised NUREG/CR-6909 and JNES-SS-1005, the transformed temperatures over 325°C for each material have been suggested for the application of the revised NUREG/CR-6909.
Statistical and Lifetime Characterization of PTFE Materials for Extreme Environments
Authors: Shinde Sannmit , Devito Dominic , Kenny Meghan , Gordon Ali P. , Poust Zachary , Drago Jim , Yang Ming-Hang , Nichols Paul
Published: October 2021

Most industrial manufacturing or process plants use bolted connections to transfer media from one location to another. Gaskets are often used to seal these systems as they offer high levels of leak mitigation; however, despite their nearly universal usage, testing done on gaskets has been limited. Historically, while industrial applications use gaskets at various temperatures and pressures, load retention tests have been conducted at a set temperature, time duration and compressive stress and report the result as load lost or retained. They do not account for longer time periods, the characteristics of the load loss over time or internal pressure the gasket is sealing. This current study seeks to change that. Traditional flat gaskets at thicknesses of 1/16” and 1/8” are leveraged against novel textured PTFE gaskets to compare load retention behavior across a range of styles. Different filler materials are tested as well, from aluminosilicate microspheres to barium sulfate and silica. Some design of experiments (DOE) methods are applied to characterize the factors that are most responsible for the load relaxation response of any particular type of gasket. A parameter from a previous study, efficiency, is used as an overall metric to quantify gasket performance. Furthermore, overall gasket deformation is examined to validate the statistical findings. Lastly, an empirical equation is presented that may be able to characterize gasket behavior across multiple loading and environmental conditions, as well as estimate the residual load carrying capacity of a gasket.
Background of Addition of Fatigue Crack Growth Rate Factors for Intermediate Strength Ferritic Steels in KD-4 of ASME Section VIII Division 3
Authors: Terada Susumu , Yoshida Toshio
Published: October 2021

In Table KD-430 and KD-430M of ASME Section VIII Division 3 (hereinafter called ASME Div. 3), there were no fatigue crack growth rate factors and threshold value of stress intensity factor range for carbon and low alloy steels with yield strength less than or equal to 620 MPa. These fatigue crack growth rate factors and threshold value of stress intensity factor range for ferritic steels with intermediate strength were also necessary for designing ASME Div. 3 vessels. We investigated the fatigue crack growth rates given in various standards. Especially Bloom’s paper related to ASME Sec. XI was investigated in detail. The test results on fatigue crack growth rate under various stress intensity range ratio in Bloom’s paper were compared with test results in other references. An equation for fatigue crack growth corrected by the stress intensity factor ratio was developed based on our investigation. The equation developed for fatigue crack growth was confirmed to agree with the test data in Bloom’s paper for negative and positive R ratios. Hence this equation, which was appropriate for a wide range of positive and negative R ratios, was proposed for ASME Div. 3. The addition of the threshold value of the stress intensity factor range for intermediate strength ferritic steels was also proposed. The fatigue crack growth rate factors at room temperature were provided in Table KD-430 and KD-430M of ASME Div. 3. As the operating temperature is higher than room temperature, the temperature correction is necessary for calculating fatigue crack growth. The temperature correction method in KD-4 of ASME Div. 3 was also proposed. These proposed changes except minimum threshold value were approved by Board in 2018 and they were reflected in 2019 Edition. The minimum threshold value was approved by the Board in 2021. It will be reflected in 2021 Edition. The background of these proposed changes is shown in this paper.
Thick-Wall Cylinder Reference Stress Crack Solutions for API 579-1/ASME FFS-1 Annex 9C
Authors: Thorwald Greg , Parietti Lucie
Published: October 2021

A new set of reference stress solutions for cracks in thick-wall cylinders were computed for addition in the next edition of the API 579-1/ASME FFS-1 standard, and are described in this paper. The geometry cases used ratios for the cylinder radius, wall thickness, crack depth, and crack length. The crack locations included axial, circumferential, internal, and external cracks. 3-D crack meshes were generated for each case to compute J-integral versus pressure result trends, which were used to determine the reference stress.

The Failure Assessment Diagram (FAD) method uses reference stress solutions to compute the Lr ratio on the FAD x–axis to evaluate cracks for plastic collapse; the FAD y-axis Kr ratio evaluates fracture failure. The elastic-plastic J-integral reference stress method will be briefly reviewed using results from this project.

A stress-strain curve was selected to represent typical material used for high-pressure components. The computed reference stress was shown to depend on the yield strength to tensile strength ratio, and a ratio of 90% was selected for use in this project.

Some shallow internal cracks in the thicker cylinder cases showed unexpected behavior in the J-integral versus pressure results, which prevented the reference stress from being computed. An alternative method was developed to use the maximum converged pressure as the nominal load to obtain reference stress solutions for those cases.
Proposed Method to Map the Elastic Follow-Up 3D Distribution Time History Using Finite Element Methods
Authors: Breach Michael
Published: October 2021

Elastic follow-up (EFU) is a complex and influencing phenomenon in pressure vessels and piping systems. It affects the performance of structural components at elevated temperatures. Quantification of elastic follow-up is challenging since it’s still not clearly defined in the ASME Boiler & Pressure Vessel Code. Pressure vessels and piping under operating load over time can exhibit mixed elastic follow-up trajectories due to inelasticity. Typically, secondary, and primary load are present at the onset which can redistribute over time due to strain accumulation and stress relaxation during elastic-follow-up. ASME Boiler & Pressure Vessel Code Div. 5, limits strain accumulation via criteria in regions outside of the elastic core. However, a method that directly addresses Elastic follow-up, would be advantageous in establishing actual margins against elastic follow-up during the design phase. The phenomenon has been well studied with various assemblages of uniaxial models, showing load and displacement controlled and mixed responses.

This study will present a method of assessing elastic follow up in finite element models. The method is based on the R5 elastic follow-up factor (EFF) expression that has been readily derived in the uniaxial case and employed by authors. Essentially the Von Mises equivalent strains are obtained from both a linear elastic and inelastic 3D finite element models subject to thermal (secondary) and pressure (primary) loading. The linear elastic model establishes the hypothetical initial elastic stress and the inelastic model will map the inelastic response through time. The Von Mises equivalent strains are then substituted into R5 expression, which are obtained from the onset linear-elastic model and the time varying inelastic model to map the elastic follow up and through time. The results are benchmarked against uniaxial results having combined primary and secondary loading.
Accurate Evaluation of Bolt and Clamped Members Stiffnesses Of Bolted Joints
Authors: Rousseau Rashique Iftekhar , Bouzid Abdel-Hakim , Zhao Zijian
Published: October 2021

The axial stiffnesses of the bolt and clamped members of bolted joints are of great importance when considering their integrity and capacity to withstand external loads and resist relaxation due to creep. There are many techniques to calculate the stiffnesses of the joint elements using finite element (FE) modeling, but most of them are based on the displacement of nodes that are selected arbitrarily; therefore, leading to inaccurate values of joint stiffness. This work suggests a new method to estimate the stiffnesses of the bolt and clamped members using FE analysis and compares the results with the FE methods developed earlier and also with the existing analytical models. A new methodology including an axisymmetric finite element model of the bolted joint is proposed in which the bolts of different sizes ranging from M6 to M36 are considered for the analysis to generalize the proposed approach. The equivalent bolt length that includes the contribution of the thickness of the bolt head and the bolt nominal diameter to the bolt stiffness is carefully investigated. An equivalent bolt length that accounts for the flexibility of the bolt head is proposed in the calculation of the bolt stiffness and a new technique to accurately determine the stiffness of clamped members are detailed.
Application of Load Release Factor Method in Thermal Expansion Load Imposed on Vessel Nozzle
Authors: Zhang Xiangbing , Duan Chenghong , Lu Mingwan , Luo Xiangpeng
Published: October 2021

The pressure vessels are connected by pipelines to form a system. Thermal expansion of the pipeline imposes an additional load on the nozzle of the connected vessel. There are two methods to deal with the thermal expansion load of pipeline in the design of pressure vessel: the partition method and the integrated method. A new Load Release Factor Method (LRFM) is proposed in this paper based on theoretical derivation. A spherical head with central nozzle is taken as an FEM (Finite Element Method) analysis model. The results show that a conservative design will be obtained by the traditional partition method, and the integrated method is the closest to the actual situation in spite of the large amount of calculation. However, compared with the traditional two methods, the LRFM can not only ensure the design margin but also reduce the calculation. This paper could be a reference for the analysis of pipeline thermal expansion load in the vessel design.
Materials Compatibility Concerns for Hydrogen Blended Into Natural Gas
Authors: Ronevich Joseph A. , San Marchi Chris
Published: October 2021

Hydrogen additions to natural gas are being considered around the globe as a means to utilize existing infrastructure to distribute hydrogen. Hydrogen is known to enhance fatigue crack growth and reduce fracture resistance of structural steels used for pressure vessels, piping and pipelines. Most research has focused on high-pressure hydrogen environments for applications of storage (>100 MPa) and delivery (10–20 MPa) in the context of hydrogen fuel cell vehicles, which typically store hydrogen onboard at pressure of 70 MPa. In applications of blending hydrogen into natural gas, a wide range of hydrogen contents are being considered, typically in the range of 2–20%. In natural gas infrastructure, the pressure differs depending on location in the system (i.e., transmission systems are relatively high pressure compared to low-pressure distribution systems), thus the anticipated partial pressure of hydrogen can be less than an atmosphere or more than 10 MPa. In this report, it is shown that low partial pressure hydrogen has a very strong effect on fatigue and fracture behavior of infrastructure steels. While it is acknowledged that materials compatibility with hydrogen will be important for systems operating with high stresses, the effects of hydrogen do not seem to be a significant threat for systems operating at low pressure as in distribution infrastructure. In any case, system operators considering the addition of hydrogen to their network must carefully consider the structural performance of their system and the significant effects of hydrogen on structural integrity, as fatigue and fracture properties of all steels in the natural gas infrastructure will be degraded by hydrogen, even for partial pressure of hydrogen less than 0.1 MPa.
Reducing the Cost of Fatigue Crack Growth Testing for Storage Vessel Steels in Hydrogen Gas
Authors: Nibur Kevin , Somerday Brian
Published: October 2021

The service life of high-pressure hydrogen storage vessels at fueling stations is dictated by fatigue crack growth. Standards such as the ASME Boiler and Pressure Vessel Code (BPVC) provide a methodology for calculating the fatigue-limited design life of high-pressure hydrogen storage vessels, in which one essential input is the fatigue crack growth rate (da/dN) vs. stress-intensity factor range (ΔK) relationship measured for the material of construction in the service environment, i.e. hydrogen gas. These measurements must also be conducted at sufficiently slow cyclic loading frequency since decreasing the frequency usually results in faster crack growth rates. Generation of complete fatigue crack growth data sets according to standard test methods becomes very time consuming and expensive when these environment and frequency criteria are met. Two modifications to standard test procedures may reduce the time and associated costs related to this testing. One approach is to accelerate the rate at which ΔK changes with respect to crack extension. This can be accomplished by controlling the normalized K-gradient, C, where C = 1/K·dK/da. In addition, by using negative values of C (i.e. decreasing ΔK), multiple da/dN vs ΔK segments can be generated from a single test specimen. This paper summarizes the status of a project designed to identify the limits to which these two strategies may be employed to measure fatigue crack growth relationships for pressure vessel steels in gaseous hydrogen environments. These limits are defined as the bounds where loading history effects begin to alter the measured da/dN vs ΔK relationships.
Influence of Resistor Wire on the Maximum Stress In Electrofusion Joint
Authors: Hu Yurui , Xiang Guangte , Shi Jianfeng , Zheng Jinyang
Published: October 2021

Polyethylene (PE) pipes are widely used in oil and gas transportation, municipal construction. Electrofusion (EF) welding is one of the most commonly used connection method for PE pipes. The welding quality of EF joints is crucial to the safety of PE pipeline. Resistor wire is an important part of pipe fittings, and has great influence in final welding quality and strength of EF joints. In this paper, we discussed the influence of resistor wire for EF fittings in the aspect of selection of resistor wire materials and determination of structural dimensions. Resistivity and temperature coefficient of resistance (TCR) are two factors for material selection. In addition, the influence of resistor wire structure, i.e., resistor wire spacing, resistor wire diameter, on the stress distribution of EF joint was discussed through finite element analysis. The results revealed that the maximum stress in EF joint increase with the diameter of resistor wire, and is barely affected by resistor wire spacing.
Modeling the Fracture Behavior of SEN(T) Specimens With Crack In Heat-Affected-Zone
Authors: Kalyanam S. , Hill L. , Wilkowski G. M. , Hioe Y.
Published: October 2021

Single-Edge-Notch Tension, SEN(T) specimens have been used for fracture toughness determinations of base, weld, and heat-affected-zone (HAZ) materials. While the experimental results and computational finite element analysis (FEA) of SENT specimens with cracks in base and weld metals are relatively straight-forward, the analysis of SENT specimens with cracks in the HAZ poses several complexities that are highlighted through computational FEA in this work. SENT specimens have been found to provide good similitude for surface cracks in pipes, where a surface-cracked structure has lower constraint condition than bend-bars, SEN(B) and compact tension, C(T), specimens. The lower constraint condition gives higher upper-shelf toughness values, and also a lower brittle-to-ductile transition temperature. Also, the SEN(T) specimen eliminates concern of material anisotropy since the crack growth direction in the SEN(T) specimen is the same as in a surface-cracked pipe. While traditional methods of evaluation allow for a direct interpretation and application of elastic-plastic fracture mechanics approaches such as those that have evolved over the last five decades, for SEN(T) specimens with a flaw/crack in base materials and to some extent in weld materials, the presence of cracking in the HAZ complicates such routine approaches.

This work initially highlights findings from an earlier published SEN(T) experiment that was investigated and reported in support of the development of recommended practices and worldwide standards. From that experiment the presence of mode-mixity in HAZ fracture experiments was identified, and their characterization and determination of how to account for them in the recommended practices are under development. This work supplements the earlier experimental findings through computational FEA that shows the effects of mode-mixity and also highlights the differences when the initial crack is located in different regions of the HAZ and is of different lengths so as to keep the initial crack and the crack growth within the HAZ region. Of particular interest is the extent of crack growth that can be used for J-R and CTOD-R curves depicting fracture resistance curves for HAZ material cracks. Towards such an end, this work starts with an experimental discussion, followed by computational FEA to provide initial guidelines and a framework for future discussions. The goal being to create a better procedural analysis and treatment of the elastic-plastic crack growth in HAZ materials which will benefit the industry. The effects of mode-mixity in HAZ testing is critical to the development of crack growth resistance, CTOD-R and J-R curves employed in Engineering Critical Assessment (ECA) of pipelines.
Full-Scale Hydrostatic Test of Internal CFRP Repair on a Degraded Pipe With a Postulated Flaw for Short-Term Loading in Safety Related Nuclear Applications
Authors: Uddin M. , Orth F. , Hioe Y. , Krishnaswamy P. , Smith L. , Basavaraju C.
Published: October 2021

The use of Carbon Fiber Reinforced Polymer (CFRP) in nuclear industry is very limited and has not been used for nuclear safety related application until recently. In 2019, a new ASME Boiler and Pressure Vessel (BPV) Code Case N-871 was approved for internal repair of buried Class 2 and 3 nuclear safety related piping using CFRP for Service Levels A, B, C and D for a service period of 50 years. However, US Nuclear Regulatory Commission (NRC) has not yet accepted the Code Case. This Code Case provides a very detail guidance on materials, fabrication, installation, and design of CFRP repair for Class 2 and 3 nuclear safety related piping.

There are two aspects in designing CFRP repair for nuclear safety related piping application in terms of service duration — short-term use condition and end-use (end of service life) condition. The final design must satisfy both conditions in terms of available safety margin for the entire period of service. In contrast to end-use condition, some of CFRP materials’ degradation phenomena such as time effect factor due to sustained loading, material adjustment factors due to environmental exposure do not affect design calculation for short-term use. However, while end-use condition design criteria may control the final CFRP repair design, it is equally important to understand the available safety margin for short-term use.

In this paper, the available safety margin in ASME BPV Code Case N-871 was evaluated for internal CFRP repair of nuclear safety related piping for short-term use by conducting a full-scale hydrostatic test. The full-scale hydrostatic test was conducted on 40” outside diameter steel pipe. The pipe contains a postulated flaw (cut-out) with an internal CFRP repair designed according to ASME BPV Code Case N-871. The internal CFRP repair consisted of three unidirectional CFRP layers applied over the flaw 360° around the circumference at the inside surface of the pipe. Two glass fiber reinforced polymer (GFRP) layers were also applied as dielectric barrier and watertightness layer. The design pressure and temperature for the hydrostatic test was 84 psi and 72F. The pipe with internal CFRP repair was pressurized until failure (leak) occurred. The details of CFRP repair installation and hydrostatic test are presented in this paper. Finally, the available safety margin for short-term use was evaluated. Detailed analysis of strain gage data, bulging measurements and the video revealed that potential failure mechanism in the full-scale hydrostatic test of pipe with an internal CFRP repair and a postulated flaw (cut out).
Calculation of Outer Crack Stress Intensity Factors for Nozzle Junctions in Cylindrical Pressure Vessels Using FCPAS
Authors: Bozkurt Murat , Nash David , Uzzaman Asraf
Published: October 2021

Pressure vessels can be subjected to various external local forces and moments acting in combination with main internal pressure. As a result of the stress system set up, and in the presence of attachment welds, surface cracks can occur on the interior and exterior walls. If these cracks cannot be detected at an early stage, there is a real potential for the vessel to rupture with obvious dangerous consequences. The behavior of fractured or geometric discontinuity structures can be investigated with linear elastic fracture mechanics (LEFM) parameters. The stress intensity factor (SIF) is the leading one, and with correct calculations, it can produce the stress intensity in the crack tip region. In cylinder-cylinder intersections subject to local loads, the maximum stress distribution occurs in and around these opening areas and failure in the system usually occurs in this region.

Using this approach, the present study develops three-dimensional mixed mode stress intensity factor solutions on for external cracks on nozzle joints in cylindrical pressure vessels nozzle junctions for a variety of geometrical configurations. This was undertaken using a finite element approach and employing a bespoke software tool and solver, FCPAS - Fracture and Crack Propagation Analysis System — to create the finite element mesh and propagation characteristics. From this, a parameter study examining the influence of the crack shape, size and position was carried out with a fixed pressure vessel nozzle cylinder intersection geometry configuration and the appropriate stress intensity factors identified and reported. The FCPAS tool is shown to be an effective approach to modelling and characterizing cracks in pressure vessel nozzles.
Control Strategy for the Open Loop Control of Piezo Actuators in High Frequency Machine Hammer Peening Processes
Authors: Hiekel Alexander , Fiedler Maik , Kolouch Martin
Published: October 2021

Machine hammer peening is a production technology for the surface finishing of metal parts. A hammer peening tool is utilized to smooth and harden the work piece surface by an incremental micro forming process. Due to the production requirements of a high process force and a working frequency, piezo actuators represent a feasible solution for the process. Achieving high surface quality also requires the implementation of control structures to avoid the resonance frequencies. This paper presents the system prototype of a piezo-based tool for machine hammer peening (“PiezoPeening”). By introducing the system components and identifying relevant parameters, the elements of the control structure are determined and the integration of smart material into the tool is illustrated. Due to the relation between the working frequency, system resonance frequency and surface quality, an analysis by using different working parameters is performed for properly setting the working frequency and actuator pre-load. Based on the results, a control sequence is proposed to deal with different positioning systems.