bluuebunny/crossref_metadata_2025_split huggingface dataset:
Activities of the GIF Safety and Operation Project of Sodium-Cooled Fast Reactor Systems
Authors: Yamano Hidemasa , Chenaud Marie-Sophie , Kang Seok-Hun , Sumner Tyler , Tsige-Tamirat Haileyesus , Wang Jin , Rozhikhin Evgeny 
Published: October 2021

Abstract
The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishment of the safety approaches and validate the performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 “Methods, Models and Codes” is devoted to the development of tools for the evaluation of safety. WP-SO-2 “Experimental Programs and Operational Experience” includes the operation, maintenance and testing experiences in experimental facilities and SFRs (e.g., Monju, Phenix, BN-600, EBR-II and CEFR), and WP-SO-3 “Studies of Innovative Design and Safety Systems” relates to safety technologies for GEN-IV reactors such as active and passive safety systems and other specific design features. This paper reports recent activities within the SO project.
Solving the Challenges of Early Storage of Spent Fuel: the SENTRY™ Spent Fuel Management System
Authors: Lloyd Timothy M. 
Published: October 2021

Abstract
Placing spent fuel into dry cask storage as soon as possible allows decommissioning plants to lower site risk, reduce costs associated with management and maintenance of the spent fuel pool, and accelerate decommissioning. The SENTRY™ Spent Fuel Management System1 is a dual-purpose system satisfying requirements for both storage and eventual transportation of the spent fuel. This paper describes the features and characteristics of the SENTRY system and explains the key areas where the product family represents a step change from existing industry offerings.
Fuel being loaded into casks within times as low as 1.5 to 2 years presents substantial new challenges to electric utilities and cask vendors. Decay heat falls off as a sum of exponentials and is much higher at early times. By the 1.5 to 2 year timeframe, heat loads are decreasing on the order of 10% per month. Radiological source terms are not only falling in strength, but are also changing in energy intensity and type. Harder neutron and photon spectra at these earlier times must also be addressed.
Thermal and radiological source term data for the SENTRY system have been established through the use of an NRC-approved Westinghouse data set characterizing fuel over a wide range of burnup, enrichment, and decay time values. This data is used to construct cooling tables which provide loading rules in the form of the minimum cooling time required to load the assembly in a particular zone of a SENTRY canister. No further plant-specific calculations or demonstrations are required by the utility.
Unique to Westinghouse is the use of adjoint methods to develop shielding results, which are de-coupled from the source term associated with specific assemblies. As the industry’s leading performer of Reactor Vessel Surveillance Programs, Westinghouse makes routine use of a combined forward/adjoint approach to radiation transport. The adjoint shielding methodology allows for the explicit calculation of dose rates over a very large range of source term values with essentially no additional computation. As a result, Westinghouse can make use of its ADSORB computer code to develop shielding results for the full set of enrichment, burnup, and decay times described above. ADSORB also uses an equivalent process to determine the earliest time that ensures the fuel’s decay heat will fall below the analyzed values used to demonstrate acceptable thermal performance. The use of this methodology allows Westinghouse to establish the shortest overall time for which fuel can safely be placed into defined regions of a spent fuel canister.
Additive Manufacturing at Westinghouse
Authors: Cleary William , Armstrong Clinton , Huegel David , Pomorski Thomas 
Published: October 2021

Abstract
Additive manufacturing (AM) is an enabling technology for novel designs and complex shapes that cannot be produced using traditional manufacturing methods. For many nuclear applications, AM could help streamline manufacturing and the supply chain, and could potentially reduce production costs while achieving higher performance through improved heat transfer, thermal hydraulic (T/H) performance, material life and accident tolerance. These benefits would improve fuel reliability and operating margins. Additionally, there are a significant number of potential applications for light water reactors (LWRs) and next generation reactors.
AM is also opening the potential to produce obsolete and legacy components which could enable plants to continue operations expediently as well as economically. The use of reverse engineering to digitize components lends itself to AM as this is the first step in producing a component with AM.
The NRC (Nuclear Regulatory Commission) is actively engaged in the evaluation of AM as well as other Advanced Manufacturing Techniques to better regulate their usage as needed. Engagement with the NRC is important to ensure regulations are grounded in understanding these technologies.
Several examples of additive manufacturing use, to improve performance and capabilities, are presented.
Sealing Evaluation of Tubesheet-Flanged Joint Under Internal Pressure and Non-Axisymmetry Thermal Loading
Authors: Tang Hui , Shi Qianyu , Li Qi , Wang Zhijian , Ma Ming 
Published: October 2021

Abstract
Tubesheet-flanged joint is usually used in some situations for its detachability, such as some small scale heat exchangers. This type of structure is convenient for maintenance. But leakage will become a major failure mode for flanged joint, especially for heat exchangers with multiple passes. Temperature gradient of different passes may cause non-axisymmetric thermal loading, stress and deformation distribution. In this paper, a detailed three dimensional finite element model of a tubesheet-flanged joint of an oil-water heat exchanger is established with a flexible graphite metallic spiral wound gasket. This kind of gasket has a good compression resilience performance. The compression-unloading curve of gasket under different temperatures are used. The stress and temperature distribution of the joint and gaskets are simulated under preload condition, pressure condition and operation condition. The seating performance are evaluated using gasket compression stress and flange stiffness criteria. The results show that the gasket compression stress distributions under preload and pressure condition are similar. Under operation condition, the gasket compression stress in local area decreases significantly and the gasket compression stress distribution along the circumferential direction becomes non-uniform because of different pass temperature. The Taylor-water method, typically used for flange design, provides a design bolt load for sizing the bolt and flange, does not provide an explicit calculation for the assembly bolt load. At some locations the compression stress is much smaller than minimum seating stress using Taylor-water method to determine bolt preload. However, under preload of half bolt yield strength recommended by ASME PCC-1, all assessment parameters are within the allowable limits.
A Proposed Methodology for ECV Fitness-for-Service Evaluation
Authors: Duffey Thomas A. , Gibson Joshem C. 
Published: October 2021

Abstract
A methodology of fitness-for-service evaluation (FFSE) for explosive containment vessels (ECVs) is introduced that utilizes change-in-thickness measurements pre- and post-test to determine the propensity of the structure to ratchet or to shake down. The method focuses on ductile failure and complements previously developed brittle failure methodologies associated with fatigue-fracture of flaws introduced during manufacture or subsequent service. The methodology is illustrated using measured thickness changes on a spherical vessel and is intended to eliminate or diminish the need for detailed, challenging finite element calculations of ratcheting and shakedown. An example is presented, based upon measured thickness changes in an explosively loaded containment vessel. Current limitations of the procedure are discussed. Applicable consensus code requirements and issues with the numerical modeling of ratcheting are briefly presented.
Fracture Mechanics Assessment of Hydrogen Storage Vessel
Authors: Jia Xiaoliang , Chen Zhiwei , Ji Fang 
Published: October 2021

Abstract
High strength steel is usually used in fabrication of hydrogen storage vessel. The fracture toughness of high strength steel will be decreased and the crack sensitivity of the structures will be increased when high strength steels are applied in hydrogen environment with high pressure. Hence, the small cracks on the surface of pressure vessel may grow rapidly then lead to rupture. Therefore, this paper makes a series of research on how to evaluate the 4130X steel hydrogen storage vessel with fracture mechanics. This study is based on the assumption that there is a semi-elliptic crack on internal surface of hydrogen storage vessel. First of all, based on linear elastic fracture mechanics, the stress intensity factors and crack tolerance of 4130X steel hydrogen storage vessel have been calculated by means of finite element method based on interaction integral theory and polynomial-approximated approach from GB/T 34019 Ultra-high pressure vessels. Then, a comparative study has been made from the results of above methods to find out the difference between them. At last, the fatigue life of a 4130X steel hydrogen storage vessel has been predicted based on linear elastic fracture mechanics and Paris formula. The calculation methods and analysis conclusion can be used to direct the design and manufacture of hydrogen storage vessel.
Evaluation of the Japanese Fatigue Test Data in Gr.91 for Elevated Temperature Design
Authors: Ando Masanori , Toyota Kodai , Hashidate Ryuta , Onizawa Takashi 
Published: October 2021

Abstract
The ASME Boiler and Pressure Vessel Code (ASME BPVC) Section III, Division 5, Subsection HB, Subpart B provided only one design fatigue curve for Grade 91 steel (Gr.91) at 540 °C (or 1000 °F) in 2019 and earlier versions. To overcome this disadvantage, The ASME Section III Working Group on Creep-Fatigue and Negligible Creep (WG-CFNC) had taken an action to incorporate the temperature-dependent design fatigue curves for Gr. 91 developed by Japan Society of Mechanical Engineers (JSME) into ASME BPVC Section III Division 5. As a result, the temperature dependent design fatigue curves are provided in the 2021 edition of the ASME BPVC. To clear the features of the best-fit fatigue curve equation developed by the JSME, 305 data stored in the database were analyzed. Details of the database and relationship between the best-fit fatigue curve equation and the data including the statistic values and the values of 95% and 99% lower confidence bound calculated by failure probability assessment were clarified through analysis. In addition to the best-fit fatigue curve equation, an equation for dynamic stress-strain response showing the behavior of Gr.91 steel under cyclic loading of is also provided based on the same database. Moreover, some additional available data of fatigue and creep-fatigue tests obtained in Japan are also provided for considering the creep-fatigue damage evaluation under elevated temperature condition.
Application of Ultrasonic Guided Wave Testing for Overhead Pipelines in Service
Authors: Liu Shuhong , Ding Ju , Wang Shenghui 
Published: October 2021

Abstract
In the detection of overhead pipelines in petrochemical enterprises, ultrasonic guided wave detection technology has become an important technical means and development direction for pressure pipeline in-service detection and evaluation. In the ultrasonic guided wave inspection of overhead pipelines, it is very important to identify defect signals and suppress noise signals. In this paper, wavelet transform (WT) and empirical mode decomposition (EMD) are used for signal processing, time-frequency transition, signal decomposition, superposition, etc., to reduce noise and extract important signals successfully. Through experiments, it is found that the decomposed signal in WT can better preserve defect information and reduce the interference of noise signals. But the signal processed by the EMD is better than WT. Finally, in the ultrasonic guided wave signal processing, the defect information in the raw signal is retained, so as to reduce the interference of the noise signal and assist inspector in the identification of the defect in the raw signal.
Numerical Analysis of the Crack Driving Force of Mismatched Girth Welded Pipes Subject to Large Plastic Deformations
Authors: Wu Kai , Zhang Hong , Yang Yue , Liu Xiaoben 
Published: October 2021

Abstract
Strength mismatched pipes with part-through cracks can suffer large plastic deformation from permanent ground deformations caused by geohazards. Thus, the crack driving force involved in engineering critical assessments plays an important role in guaranteeing pipeline integrity when pipes are subjected to complex loads induced by a hostile environment. In this paper, Python scripts are developed to generate up to 200 finite element models of strength mismatched pipes with various crack sizes under large plastic deformations based on the commercial software ABAQUS. The effects of crack length, crack depth, and strength mismatch factors on the evolution of crack tip opening displacement (CTOD) and global strain were investigated. An approximately linear relationship was observed in all cases tested with global strain values varying from 0.5% to 3%. Meanwhile, the value of the CTOD increased with the increase of crack length and crack depth, and decreased with increasing mismatch factor from the undermatch to the overmatch conditions. The effect of the crack depth on the CTOD is comparatively larger than the crack length, which presented an obvious change of the CTOD for deep cracks coupled with undermatched conditions. Overmatched welding only affected the value of CTOD slightly, while a drastic increase of CTOD value was observed for the undermatched welding conditions, especially for deep and long cracks.
A Study on Fluid-elastic Instability of Tube Array with Different Stiffness Asymmetry in Cross Flow
Authors: Guo Kai , Tan Wei , Cao Liqin , Zhang Hongsheng 
Published: October 2021

Abstract
The fluid-elastic instability is probably to occur in both transverse and streamwise directions. The studies of streamwise instability are mainly performed with the tubes are fixed in the transverse direction, which may not happen in the practical tube bundles. In this study, the effect of asymmetry of the tube stiffness on the streamwise instability was investigated through water tunnel tests and numerical simulations. The vibration was controlled by changing the stuffiness of tubes in the two directions. The results show that the streamwise vibration can affect instability threshold in the transverse direction in single tube tests, and the streamwise vibration was suppressed when the tubes fall into instability in transverse direction. The asymmetry of the tube stiffness can have an effect on streamwise and transverse instability, but the effect is not pronounced in the multi-flexible tube array.
Developing and Implementing Ch. VII O&M Programs at CCGT Plants: Case Studies
Authors: Jackson Peter , Rosario Robert , Fabricius Andreas , Johny Anita , Wholey Alexandria 
Published: October 2021

Abstract
We will present the results from several projects from the USA and other jurisdictions where ASME B31.1 Ch. VII O&M Covered Piping System (CPS) Programs have been implemented at several types of natural gas-fired CCGT plants. Common elements of programs for different plants will be summarized as well as plant-specific considerations for high energy piping condition assessment for newer plants. Pros and cons between a common program for a thermal fleet and plant-specific programs will be discussed including advantages and disadvantages of each approach. Effective implementation of parts of the Nonmandatory Appendix V guidance within the CPS Program will be described and recommendations for best practices.
A brief overview of degradation-specific mechanisms for high energy piping and approaches for planninng/scheduling NDE inspections will be described. This overview will include: creep, fatigue, corrosion (erosion-corrosion - E/C and flow accelerated corrosion - FAC) as well as mechanisms that are commonly responsible for high energy piping leaks, failures and repairs including thermal quench cracking of HRSG interstage, terminal desuperheaters and turbine bypass attemperators. A brief summary of Gr. 91 inspection planning in Ch. VII O&M Programs will also be included as well as corrosion under insulation (CUI) and common inspection scopes for high temperature steam drains.
Resolution of constant force and variable spring pipe supports on high pressure/high temperature piping that are not accommodating thermal expansion as per their engineering design can be evaluated using pre-outage pipe stress models and data obtained from field walkdowns to support rapid decisions for repair/replacement in the field.
Finally, experiences with long term scheduling the need for adaptive management of the CPS Programs will be summarized with typical management oversight actions described for effective implementation.
Assessment of Flaws in Non-Stress Relieved Carbon Steel Welds Caused by Hydrogen Attack
Authors: Bagnoli K. E. , Cater-Cyker Z. A. , Hay C. A. , Holloman R. L. , Hioe Y. , Wilkowski G. , Rollins B. C. , Nikbin K. M. 
Published: October 2021

Abstract
Cracking due to high temperature hydrogen attack (HTHA) has been observed in non-PWHT’d carbon steel process equipment at conditions of temperature and hydrogen partial pressure below the original design limits recommended in API RP 941, necessitating changes to that standard. Consequently, flaw assessment procedures are needed to manage defects detected during inspection, or to establish appropriate inspection frequency. The latter typically involving estimation of the time for a detected or postulated crack to reach a critical size. This type of evaluation has been difficult to perform owing to the scarcity of data of fracture toughness as well as crack growth rate for steels in high temperature hydrogen. To address this gap, an experimental program was undertaken to help describe the ductile tearing characteristics of steel removed from service with various levels of HTHA damage. Near full thickness single edge notched tension SEN(T) specimens were machined from field samples and tested using existing “natural” cracks as the starter-crack. This provided insight into the behavior of real flaws subject to constraint conditions closely matching circumferential flaws in piping. Tests of undamaged steel were also performed in hydrogen at conditions designed to produce HTHA and compared with tests run in nitrogen. Crack growth tests obtained from the literature have been used to develop an empirical crack growth law for use in fitness for service assessments. The C* integral was also explored as a parameter for describing crack growth rate due to the strong similarity of HTHA damage to creep. The key results show substantial reduction in tearing resistance resulting from HTHA damage. A crack growth law similar to the Nikbin-Smith-Webster (NSW) model using the C* integral was found to show promise in describing the combined effects of creep and HTHA on crack propagation, although additional testing is needed to validate the correlation.
Thermal Stress Analysis of a Spent Nuclear Fuel Canister
Authors: Jensen Ben J. , Campbell Charlton F. , Klymyshyn Nicholas A. 
Published: October 2021

Abstract
The potential for chloride induced stress corrosion cracking (CISCC) in spent nuclear fuel dry storage canisters is a current topic of research by the US Department of Energy Spent Fuel and Waste Science and Technology program. One of the important prerequisites for CISCC is a tensile residual stress state. This study utilizes computational models to provide an initial analysis of thermal stresses in a generic spent nuclear fuel canister. A STAR-CCM+ thermal fluid model provides a temperature profile analysis of the canister based on four different ambient temperature conditions. An ANSYS APDL finite element model incorporates the temperature profiles to analyze the thermal stresses in the canister. The calculated thermal stress magnitudes are in the range of 10 MPa to 80 MPa, which could be significant for crack propagation through the canister wall. The maximum thermal stress is not a concern for the structural failure of the canister, but it is high enough that it is a significant contribution to the total stress state of the canister wall, which also includes residual stress from fabrication, internal pressure from helium cover gas, and potential transient mechanical loads, such as earthquakes. This paper describes the initial finite element thermal stress analysis that was completed in 2020, reports the initial findings, and identifies areas where model refinement is still needed to complete this work in the future.
Structural Dynamics Modeling of Spent Nuclear Fuel During Hypothetical Package Drop Events
Authors: Kadooka Kevin , Fitzpatrick James , Klymyshyn Nicholas 
Published: October 2021

Abstract
The response of spent nuclear fuel (SNF) to hypothetical package drop events is of particular interest in the scope of spent fuel storage and transportation because of the mechanical shock encountered in such scenarios. Previous testing and modeling by the U.S. Department of Energy have demonstrated that the shock and vibration environment of normal shipping and handling conditions (excluding package drop events) is relatively benign and does not challenge the integrity of spent nuclear fuel. Cask drop events are worth considering because SNF packages are required to withstand free drops onto unyielding surfaces as part of their licensing basis. The acceleration experienced during drop events can be orders of magnitude higher, and thus more advanced models are needed to encompass potential nonlinear behavior of the fuel, such as spacer grid buckling and rod-to-rod impact. This work describes a number of finite element models developed to calculate the response of spent nuclear fuel to various hypothetical drop events that have been validated by package and fuel assembly drop tests conducted in the last decade. Sensitivity of the model response to factors such as package drop orientation, secondary impacts, and irradiated material properties, as well as their potential impacts to fuel cladding integrity, was also investigated. Cask drops are not expected as a regular occurrence during SNF transportation, but this work helps raise the understanding of SNF mechanical loads to the point of consistency with the package design requirements.