bluuebunny/crossref_metadata_2025 huggingface dataset:
Preliminary Verification Calculation and Sensitivity Analysis on PISAA Code Compared to MELCOR
Authors: Song Mingqiang , Zhang Ningna , Yang Xiaoming , Ma Rubing , Yang Zhiyi , Ding Chao
Published: October 2021

Since the Three Mile Island accident happened in the United States in 1979, the nuclear industry has accelerated the development of calculation codes for simulating severe accident. Especially after the Fukushima nuclear accident, the nuclear industry and nuclear safety regulatory authorities have paid more attention to the study of severe accidents in nuclear power plants. Based on advanced pressurized water reactors in Generation II, typical severe accident scenario, like small break loss of coolant accident (SBLOCA) is selected, and verification calculations on the Program Integrated for Severe Accident Analysis (PISAA), the severe accident integrated code developed by China with MELCOR, is conducted in this paper. The results of flow rates in break, pressure, water level and hydrogen production are obtained, showing that PISAA has the same trend compared with MELCOR reasonably, and PISAA has a completed calculation function which can simulate severe accident phenomena. Uncertainty analysis is also performed, 7 representative parameters related to hydrogen production are selected, and Latin hypercube calculation is performed. Hydrogen production distribution is around 300kg∼520kg, indicating that there is no cliff effect in PISAA. In the future, the researchers plan to do some works to improve the accuracy and functions of PISAA.
Numerical Simulation of Thermo-Hydraulic Characteristics of 7-Pin SFR Test Fuel Bundle With Variable-Pitch Helical Wire
Authors: Li Siyuan , Zhang Aimin , Ji Songtao , Li Yanlin
Published: October 2021

This paper is focused on the applicability of variable-pitch helical wire in a 7-pin SFR test fuel bundle by using CFD method. The differences between the variable-pitch helical wire scheme and the even-pitch helical wire scheme in the pressure loss, velocity, temperature field of the test fuel bundle are compared and analyzed. The simulation results show that it is reliable to carry out the thermo-hydraulic analysis of SFR test fuel bundle by using CFD numerical simulation method. The helical wire mixing ability to the coolant and the pressure drop is influenced by the helical wire pitch. Compared with the even-pitch helical wire scheme on fuel rod surface, the variable-pitch helical wire doesn’t drastically change the flow field structure while the pitch changes, but gradually increases its mixing ability to the coolant along the axial direction, which is more conducive to accelerating balance the thermal inhomogeneity and inhomogeneity ratio of the coolant, solve the problem of temperature rise caused by the coolant congestion on the contact structure between the helical wire and fuel rod surface. The stronger the mixing ability of the helical wire to the coolant, the more effective it is to lower the maximum temperature of the coolant. The application of the variable-pitch helical wire can well match the demand of the coolant for thermal and hydraulic needs, and the safety and economy of the assembly can be correspondingly improved.
Flow Induced Vibration Analysis and Remediation Using a Cartesian Grid Flow Solver
Authors: Boschitsch Alexander , Danilov Pavel , Kaufman Andrew , Bilanin Alan
Published: October 2021

Flow through pipe junctions produces unsteady vortex shedding processes that can generate acoustic pressures whose modal amplification contributes to structural fatigue. Such flow-induced vibration (FIV) is an enduring concern in nuclear power plants (NPPs) and has led to delays in extended power uprate licensing, produced strong fluctuating loads in piping, and contributed to cracks in low pressure turbine blades. Modern computational fluid dynamics (CFD) analysis can analyze FIV phenomena, but the practical challenges of setting up and adequately meshing multiple candidate designs continue to deter routine and reliable CFD-based FIV assessment. Recently, CFD solvers implemented on Cartesian grids and offering autonomous mesh generation about complex geometries have made significant strides in addressing these challenges. Here, the successful application of a CFD solver (CGE) used in combination with subscale testing to rectify FIV in NPP piping, is described. The article reviews the role of CFD in FIV analysis and design, summarizes the principal analysis components of CGE, outlines the 1/8th subscale testing program used in conjunction with CFD to assess FIV, and concludes with postinstallation assessments of the modification. The concepts contained here have helped address FIV in several NPPs resulting in significant reductions in operation & maintenance (O&M) costs.
Production of Cyclotron-Based Gallium-68 With Low Energy Protons: Preliminary Target Design and Cyclotron Shielding Considerations
Authors: Salas-Tapia Luis Fernando , Zhang Tian
Published: October 2021

Nuclear reactors, cyclotrons and generators are being used to produce nuclear medicine radionuclides. Radiopharmaceuticals based on Gallium-68 are widely used both experimentally and clinically in Positron Emission Tomography (PET). Currently Ga-68 is mainly produced via 68Ge/68Ga generators, which are expensive. An alternative method is by cyclotron using high enriched Zinc-68 via the 68Zn(p,n)68Ga reaction. Investigations have been conducted for producing Ga-68 using a solid target impacted with a proton beam in energies between (13–14) MeV or higher. The main interest of this work is to produce Ga-68 using energies within a lower range, therefore Ga-68 can be accessed to locations with cyclotrons of lower energy. In this article, an approach to a novel target design to get Ga-68 with low energy protons is conducted. Adequate shielding to avoid extra doses outer the cyclotron is also considered. Monte Carlo code FLUKA is used to simulate the processes of interaction of radiation with matter and the radioactive isotope production. For a beam energy of 11 MeV and current on target of 40 μA the simulated produced activity is about 69.1 GBq. The results show the feasibility of producing Ga-68 with low energy protons and self-shielding system guarantees the regulation-based dose exposition outer the cyclotron.
Experimental Study of Characteristics of Flow Field in Rod Bundle Channel Under Blocking Conditions
Authors: Yu Xiaoyong , Zhang Yonghao , Qi Peiyao , Liu Yusheng , Qiao Shouxu , Tan Sichao
Published: October 2021

The rod bundle fuel is characterized by compact structure and narrow flow passage. The fragments and corrosion products, flowing with the coolant, can cause local blockage accident, threaten the integrity of the fuel cladding. Therefore, it is necessary to use the Particle Image Velocimetry (PIV) to visualize and measure the flow fields downstream of the blockages. The results show that partial blockages will cause flow reversal. In the backflow zone, vortices are generated downstream of the blockage, causing increase in the resistance. The length of backflow zone increases with the increase of the Reynolds number. The wake area formed downstream of the blockage presented periodic changes with the time and the period is about 0.8s. For the blockage of the interior subchannels, in the backflow zone, two rows of asymmetrically distributed vortices, and the vortices interfere with each other and cause squeeze deformation. For the blockage of the side and corner sub-channels, the formed vortices have irregular shape and nonuniformed distribution, and the flow field is more complex and changeable. This is believed to be caused by the high intensity turbulence and the influence of the wall.
Experimental and Numerical Investigation on Debris Bed Quenching With Additional Injection of Non-Condensable Gas
Authors: Petroff Markus , Kulenovic Rudi , Starflinger Jörg
Published: October 2021

Severe accidents of light water reactors with core degradation can lead to the formation of a so-called debris bed inside the reactor cavity. In the scenario of depleted residual water, the bed can partially melt and interact with the concrete underneath generating non-condensable gases (NCG) at the bottom of the particle bed, which will flow through the debris bed. The impact of additional gas on the quenching process can be considered in thermal-hydraulic system codes like ATHLET. However, there is still a need for experimental validation of respective models or verification of corresponding simulation results. In the frame of this work, specific quenching experiments on a monodispersed particle bed with various NCG injection rates and initial bed temperatures are carried out employing the FLOAT test facility to extend the existing experimental database. Furthermore, numerical simulations are carried out with COCOMO-3D for the prediction of the thermo-hydraulic conditions inside the debris bed, the quench front propagation as well as the total quenching times.
Design of a Novel Test Section for the Lead Fast Reactors Development: The CIRCE-THETIS Facility
Authors: Lorusso Pierdomenico , Di Piazza Ivan , Martelli Daniele , Musolesi Andrea , Tarantino Mariano
Published: October 2021

Within the R&D program of Generation IV lead fast reactors and Accelerator-Driven Systems, a new project named PATRICIA (EC – H2020) has been launched by the European Commission to support innovative solutions for the development of MYRRHA ADS concept and lead fast reactor technologies in general. In this framework, the ENEA Brasimone Research Centre (Italy) supports such research activities through experimental campaigns, involving CIRCE, a large scale pool-type facility using lead-bismuth eutectic as the primary coolant and pressurized water as the secondary fluid.

A new test section named THETIS is currently under design and it will be installed in the main vessel of the CIRCE facility. The test section will include a vertical mechanical pump for the primary coolant circulation and a new prototypical helical coil steam generator (HCSG). This type of steam generator turns out to be very interesting for nuclear power plants, since the helical geometry is very compact and it assures a high power removed, taking up the minimum amount of space.

The experimental tests will aim at: (i) investigating the thermal-hydraulic behavior of the system in steady state operation (forced circulation regime), during operational and accidental transients (postulated scenarios) and in natural circulation regime, considering as heat sink the HCSG (acting as Decay Heat Removal system, DHR) and the reactor vessel auxiliary cooling system, in stand-alone or coupled operation, (ii) characterizing the performances of the helical coil steam generator.

The present paper aims at presenting the preliminary layout of the THETIS test section, describing in details the main components, the instrumentation installed and the main operating conditions of the facility.
Study on Water Cooler Performance for High Temperature Helium Experimental System
Authors: Hu Qingxiang , Peng Wei , Zhao Gang , Wang Jie
Published: October 2021

Very-high-temperature gas-cooled reactor (VHTR) has good inherent safety and is one of the six reactor technologies of Generation IV nuclear energy systems. The ultra-high core outlet temperature makes VHTR suitable for process heat applications such as hydrogen production. The intermediate heat exchanger (IHX) is the key equipment for energy transfer between VHTR and the process heat application system. To verify the high temperature performance of IHX, it is generally necessary to set up a high temperature helium test system to conduct a series of performance tests. In this experimental system, a well-behaved water cooler needs to be designed to cool down the high-temperature helium in the experimental loop to ensure the safety of the experimental system. In the present study, the effect of the cooling water flow rate G, the cooling water injection position L, and the lower heat exchange tubes material were analyzed. The results show that increasing the flow rate of the cooling water, the maximum temperature of the tube sheet Tmax decreases and the temperature gradient in the flow direction ∇Ty increases. Tmax increases first and then decreases with the increasing L. There exists an optimal cooling water injection position between L = 175mm and 225mm. The variation of ∇Ty is the opposite. The use of ceramic instead of steel reduces Tmax by about 295K and reduces ∇Ty by half. Subsequent research and optimization on structure parameters of the water cooler are expected to further improve the cooling performance.
Study on Flow Characteristics of Double Loop Natural Circulation System Under Asymmetric Conditions
Authors: Wang Shuang , Li Xin , Liu Yongchao , Qiao Shouxu , Tan Sichao
Published: October 2021

The load distribution and coupling feedback between parallel loops is one of the issues in the multi-loop nuclear reactor system. In the present study, a dual-loop reactor natural circulation experimental system with typical structural features was designed and built based on the characteristics of a floating nuclear power plant. The system is used to study the system response characteristics under the conditions of heat load disturbance, resistance disturbance and system response time. The results show that the flow of the primary loop will be redistributed when the load disturbance was introduced into the natural circulation loop. The primary flow rate changes quickly when the heat load disturbance reaches the critical point. When the heat load disturbance is lost on one side, the total natural circulation flow rate of the loop decreases to 60% of the value in the steady-state flow. The response of the system after introduction of resistance disturbance is similar to the introduction of the load disturbance. The response of the system can be divided into two stages, i.e., the rapid response period and the slow adjustment period after the introduction of the disturbance. With the increase of disturbance, the negative slope of stage 1 increases, and the slope of stage 2 gradually changes from positive to negative. With the experimental, a model is developed to predict the response of the system. A characteristic number S is defined to characterize the symmetry of the system. Based on the experimental data, it is found that when S is between 0.5–2, the external disturbance will not cause asymmetric response of the system.
Study on Collapsed Cross Section for Radial Reflector in LWR
Authors: Shibano Ryosuke , Kawano Tatsuya , Takeda Satoshi , Kitada Takanori , Masaoka Yoshitada , Nagano Hiroaki , Hyoudo Hideaki , Kodama Yasuhiro
Published: October 2021

Few-group cross sections of fuel and reflector region are generally used for the core design of LWR. The power distribution and reactivity depend on the few-group cross sections in the core design. The few-group cross sections depend on the fuel adjacent to the reflector region. Therefore, it is necessary to calculate the few-group cross sections. However, the effect of changing fuels adjacent to the radial reflector has not been clarified. Therefore, the purpose of this study is the evaluation of the effect of changing fuels adjacent to the radial reflector on the power distribution and the reactivity in order to clarify the effect of changing fuels adjacent to the radial reflector on the radial reflector cross sections. First, the effect of changing fuels adjacent to the radial reflector on the power distribution is evaluated and is compared to the effect of changing temperature or boron concentration which are generally taken into account in the conventional model of the radial reflector cross sections. In this evaluation, the effect of changing fuels adjacent to the reflector on the radial reflector constant is evaluated by using UO2 and MOX fuels. Furthermore, the effect on the radial reflector constant is evaluated based on the perturbation theory. The results show the effect on power distribution is about 0.3 % at maximum, and the effect on the reactivity is quite low (< 1 × 10−7Δk/k).
Research on Application of Additive Manufacturing Technology In Nuclear Fuel Assembly Field
Authors: Li Hua , Yue Ti , Zhu Fawen , Peng Yuan , Li Yun , Huang Chunlan , Zhang Youjia
Published: October 2021

As one of the key technologies of the fourth industrial revolution, Additive Manufacturing (AM, also called 3D printing), which is the basis of innovation design with its high degree of manufacturing freedom, has been widely used in the construction field, aerospace field, automobile manufacturing field, biomedical field, etc. With the increasing maturity of additive manufacturing technology, it has been gradually used in the field of nuclear energy in order to realize design reform. Public reports have shows that additive manufacturing for the pressure vessel fuel, some kinds of fuel pellets, parts of fuel assembly and related core components has been realized. In this paper, the advantages and challenges of additive manufacturing technology used in nuclear fuel assembly design and manufacturing are analyzed, the application status of additive manufacturing technology in the field of nuclear fuel assembly design and manufacturing is summarized, and the application development trend of the additive manufacturing technology for the fuel assembly is also clarified. In general, although the application of the additive manufacturing technology in the field of nuclear fuel assembly design and engineering is still at the stage of research, the application scope of this innovative technology will be greatly expanded with the continuous progress and improvement of the additive manufacturing technology in the near future.
Applicability Assessment of Accident Analysis Codes and Determination of Testing Facility for Validation of the CAP1400
Authors: Cai Xiaoyu , Shi Guobao , Yan Jinquan , Fan Pu , Zhao Dongjian
Published: October 2021

The approach in uprating the AP1000 to the CAP1400 was to increase the power capability of the plant while retaing the creadibility of proven components and substantial safety margins. Therefore, the CAP1400 retains the AP1000 licensing basis. A comparison of the CAP1400 major design with the AP1000 is provided in Tab. 1.
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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Spent Nuclear Fuel Mechanical Loads in the General Package Drop Scenario
Authors: Klymyshyn Nicholas , Spitz Casey , Ivanusa Pavlo , Kadooka Kevin , Fitzpatrick James
Published: October 2021

The US Department of Energy Spent Fuel and Waste Science and Technology program is performing research to determine the mechanical loading conditions applied to spent nuclear fuel (SNF) during normal conditions of transportation to inform mechanical tests of SNF and close an important knowledge gap related to the practical disposition of SNF in the US. A recent multi-national collaborative test campaign measured SNF assembly impact response to the 30 cm horizontal package drop scenario, which is a common regulatory basis test of SNF package design. Researchers at Pacific Northwest National Laboratory (PNNL) are using the test data to validate explicit finite element models to calculate the mechanical loads and structural response of spent nuclear fuel assemblies in the as-tested 30 cm horizontal package drop scenario. Once the as-tested package drop model is validated, the next step is to apply the model to the general 30 cm drop scenario, which includes all impact angles, all fuel assembly types, and all burnup conditions. PNNL is developing a damage model that will incorporate the results of finite element parametric studies to establish trends in SNF mechanical loading to various input parameters, like impact orientation and burnup. The damage model will have the capability to estimate mechanical loads for any single set of input parameters, but it will first be used to describe the upper bounds of potential SNF loading in the 30 cm drop scenario. This paper describes PNNL’s progress toward developing the general solution to the problem of spent nuclear fuel mechanical loads in the 30 cm package drop scenario, and it describes the next steps in closing the knowledge gap.
Prediction of Ultimate Load of Pipeline With Defects Based On Material Hardening
Authors: Fang Zhuoyu , Dong Shaohua
Published: October 2021

Pipelines are important facilities for the development of oil and gas, and it is of great significance to ensure their safe and efficient operation. The existing pipeline ultimate load evaluations and researches lack the consideration of pipe characteristics and failure mechanism, there are certain errors, and the types of defects involved are relatively single. In order to give full play to the potential of pipes, consider the strain hardening effect of pipeline steel during the true failure process of pipelines, and study the influence of different types of defects on the ultimate bearing capacity of the pipeline. This paper uses the twin shear stress yield criterion and based on the Ramberg-Osgood material strain hardening model to derive an equation which is used to predict the ultimate load of the defect-free X60 pipeline. Using finite element method to study different kinds of defects on the ultimate load of X60 pipeline. A model for predicting the ultimate load of the X60 pipeline with a flat-bottomed rectangular single-point external defect is fitted, and the accuracy of the model is verified by comparison with real data. The failure modes of pipelines with different defects are also studied. On the basis of the above prediction model, the ultimate load prediction models of X60 pipeline with axially adjacent double-point external defects were fitted. Besides, the critical distance of interaction between the double-point defects has been discussed. The above analysis provides a basis for the safety assessment of X60 pipeline.
Application and Limitation of Safety Valves Inspection Codes in China
Authors: Xing Jian , Han Zhiyuan , Ai Shuping
Published: October 2021

Safety valves are important devices to protect system by relieving pressure in oil refining and chemical equipment. According to API 576-2017 and API 510, the inspection interval for safety valves should not be more than 10 years. Also, TSG ZF001–2006 in China specifies the inspection interval for safety valves is only 1 year. However, the maintenance interval for petrochemical units normally is 3∼5 years, which thus results in the inconsistency between its inspection interval and the maintenance interval. This study firstly conducts contrastive analysis about the difference of major regulations and standards for safety valves inspection between China and API standards, and then performs thorough investigation of the operational management and regulation about the extension for safety valves inspection interval in the typical Chinese petrochemical enterprises. Besides, the failure modes and risk level of safety valves are investigated by quantitative analysis of 44,692 safety valves engineering application data in oil refining and chemical plants based on Risk-Based Inspection (RBI) technology. The result indicates that the overall risk level of safety valves is low, as the percentage of safety valves with a high-risk level is less than 1%, and the accident rate caused by safety valves failure is less than 0.005%. In addition, this study analyzes the limitations of specification such as TSG ZF001–2006 in terms of safety valves inspection and the application of regulation extension. Based on above analysis, the specifications and standards about safety valves should be revised and introduce the RBI technology to extend the inspection interval.
Study the Integrity of Flange to Gasket Connection Using ASME PCC-1 Under Bolt Tightening and Operating Condition
Authors: Kalra Balram , Kumar Arun
Published: October 2021

Scope of this paper is to study the flange to gasket joint under bolt tightening and operating conditions as per the guidelines specified in ASME PCC-1 Appendix O using FEA approach with ANSYS as the FEA tool. Contact pressure for various transient loading conditions is obtained and compared against the required minimum contact pressure for the leak proof joint. ASME-PCC 1 gives the guidelines for bolted flange assembly for pressure vessel and piping joints in a standardized manner. With the help of FEA tools simulation can be done and broad picture of the assembly can be looked upon so that a designer engineer can have better insights for the stress and deformation during seating and operating conditions. Thorough investigation of the flange joint has been carried out by finite element analysis in this paper and acceptance criteria as recommended in ASME PCC-1 Appendix-O have been used to establish the design in a safer and optimized way.