bluuebunny/crossref_metadata_2025 huggingface dataset:
Application of XFEM to Model Stationary Crack and Crack Propagation for Pressure Containing Subsea Equipment
Authors: Lee Sam (Kwok Lun) , Martin David
Published: December 2016

The extended finite element method (XFEM) [1, 2] can approximate the discontinuous displacement field near cracks independently of the finite element mesh through the use of interpolation functions, which can describe the displacement field near cracks in the structure. Therefore, XFEM is appearing as an alternative to conventional contour integral method for crack growth analysis. Most benchmarks for XFEM in published literature were mostly for the through-wall crack or loadings were under tension or bending moment. Subsea equipment provided by FMC Technologies is used under the most extreme conditions of pressure and temperature. Due to the fact that most equipment in subsea is pressure containing, assessment of cracks and crack growth under pressure cycles is an essential part of the integrity assessment. The implementation of fracture mechanic based fatigue to analysis can extend equipment life while improving model accuracy.

In this paper, crack growth modelling in a thick wall cylinder was performed by XFEM embedded in Abaqus [3]. The case study of semi-elliptical axial surface crack at internal surface was selected. The fracture and fatigue results based on LCF analysis were compared with contour integral approach and published experimental data [4]. One of the advantages of XFEM is it can easily accommodate changes in the crack propagation direction for mixed mode loading.
A Review of the Shell Buckling and Stiffener Ring Design for Cylindrical Steel Storage Tanks
Authors: Azzuni Eyas , Guzey Sukru
Published: December 2016

Cylindrical steel storage tanks are shells designed to store different types of products such as liquids or grain. The thickness of the shell is calculated to withstand the circumferential stress resulting from the hydrostatic pressure due to the stored product.

A unique situation when there is no stored product leads to the vulnerability of the shell to buckle when there is wind load due to external pressure.

There are two major types of buckling modes: local and general. The local buckling mode is studied analytically in various studies and is easy to mitigate. The general buckling mode can be more damaging to the tank and more costly to mitigate.

The prevention of general buckling due to wind load pressure is achieved through the addition of stiffener rings. However, the stiffener rings design procedure used by various design standards has little known background. This paper reviews the current design approach’s origin and explains a semi-analytical justification for it. The unfolding of the design expressions can lead to more freedom in design variables selection leading to more economical designs.
Assessing the Condition and Estimating the Remaining Lives of Pressure Components in a Methanol Plant Reformer: Part 2 — Engineering Evaluation
Authors: Jaske Carl E. , Shannon Brian E. , Miranda Gustavo , Prewitt Thomas J.
Published: December 2016

Statoil Tjelbergodden operates a 2,400 ton/day methanol plant in Norway. Part 1 of this paper described the advanced non-destructive examination (NDE) technologies that were applied to obtain data for engineering evaluation of radiant catalyst tubes, outlet pigtails, and outlet collection headers. The inspection results were compiled along with data on materials properties and plant operating conditions for use in a series of life prediction studies. This paper describes the assessment methodologies that were applied in evaluating the remaining life of the in-service components. The special purpose WinTUBE™ finite element software was applied to predict remaining catalyst tube creep life based on the computed creep stress-strain response and creep damage accumulation under simulated future operating conditions. Outlet headers and pigtails were modeled using general purpose finite element software to compute stresses and strains during operation. Following the methodology of API 579-1/ASME FFS-1 the computed stresses and strains were used to predict remaining creep life. Using the remaining life estimates to decrease the potential of in-service failures and increase the reliability of future reformer operations is discussed.
Creep Buckling and Post Buckling Behaviors of Stainless Steel Tube Columns Under External Pressure at Extremely High Temperatures
Authors: Jo Byeongnam , Okamoto Koji , Kasahara Naoto
Published: December 2016

Creep buckling failure of a stainless steel tube column was investigated at three temperature conditions (800, 900, and 1000 °C). 304 grade stainless steel was used as a test material in this study. In creep tests, external pressure was increased to a target value, temperature of the tube column was quickly increased to a target temperature, and failure time was measured maintaining the pressure and the temperature. Based on the experimental results of the creep buckling failure time, an empirical correlation was developed by the Larson-Miller parameter. Moreover, post buckling experiments were performed to examine buckling-induced boundary failure at extremely high temperature more than 1300 °C. Additional heating was applied to the specimen which already buckled by external pressure. In the additional heating tests, temperature was increased until boundary failure was formed on the surface of the tube columns. The results showed that the creep buckling failure time was shorter than those in other tensile stress-induced creep tests. The empirical correlation obtained by Larson-Miller parameter predicts well the creep buckling failure time. Finally, boundary failure was obtained in the post buckling under the additional heating.
RCC-MRx Code: Context, Overview, On-Going Developments
Authors: Pétesch C. , Lebarbé T. , Pascal C. , Blat-Yrieix M.
Published: December 2016

RCC-MRx Code (ref 1) is the result of the merger of the RCC-MX 2008 developed in the context of the research reactor Jules Horowitz Reactor project, in the RCC-MR 2007 which set up rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER.

This code has been issued in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires) in 2012, and a new edition was published at the end of 2015.

A significant work has been performed for this edition to improve the code in order to facilitate its use and understandability, and also to have a better fit with the feedbacks of the users.

In parallel, in compliance with the EC’s objectives and its own policy of openness, AFCEN proposes to make its codes evolve, taking into account the needs and expectations of European stakeholders (operators, designers, constructors, suppliers...) through a workshop called CWA phase 2.

This paper gives an overview of the realized work and also will identify the work to be done for an opening of a standard such as RCC-MRx code.
Tests on SENT and CT Specimens to Study Geometry Effects in the Brittle to Ductile Transition
Authors: Le Delliou Patrick , Geniaut Samuel
Published: December 2016

The accurate prediction of ductile fracture behaviour plays an important role in structural integrity assessments of critical engineering structures under fully plastic regime, including nuclear reactors and piping systems. Many structural steels and aluminium alloys generally exhibit significant increases in fracture toughness, characterized by the J-integral, over the first few mm of stable crack extension (Δa), often accompanied by large increases in background plastic deformation. Conventional testing programs to measure crack growth resistance (J–Δa) curves employ three-point bend, SEN(B), or compact, CT. However, laboratory testing of fracture specimens to measure resistance curves (J–Δa) consistently reveals a marked effect of absolute specimen size, geometry, relative crack size (a/W ratio) and loading mode (tension vs. bending) on R-curves. These effects observed in R-curves have enormous practical implications in defect assessments and repair decisions of in-service structures under low constraint conditions. Structural components falling into this category include pressurized piping systems with surface flaws that form during fabrication or during in-service operation.

A research program was launched by EDF R&D to study geometry effects (e.g. triaxiality effects) in the brittle to ductile transition of carbon-manganese steels using Single-edge notch tension (SENT) specimens, by comparing the results obtained on these specimens with the results obtained on CT specimens.

This paper presents the results of the tests conducted between −40°C and −100°C on a large number of specimens of both types. The toughness values of the SENT specimens appear to be included in the scatter of the CT12.5 ones, so the geometry effect between CT and SENT specimens in the brittle to ductile region is not significant. Moreover, the results of the CT12.5 cut in the L-S direction are not very different of those of the specimens cut in the T-S direction. The Master Curve methodology fits rather well the CT12.5 results, whereas the SENT results are not well covered by this methodology.

The energetic approach called GP has been applied to the analysis of some tests. This approach shows that the geometry effect between both types of specimens is limited, in agreement with the experimental observations.
Two Case Studies in 3-D Residual Stress Prediction for Nuclear Power Applications
Authors: Benson Michael L. , Raynaud Patrick A. C. , Brust Frederick W.
Published: December 2016

Residual stress prediction contributes to nuclear safety by enabling engineering estimates of component service lifetimes. Subcritical crack growth mechanisms, in particular, require residual stress assumptions in order to accurately model the degradation phenomena. In many cases encountered in nuclear power plant operations, the component geometry permits two-dimensional (i.e., axisymmetric) modeling. Two recent examples, however, required three-dimensional modeling for a complete understanding of the weld residual stress distribution in the component. This paper describes three-dimensional weld residual stress modeling for two cases: (1) branch connection welds off reactor coolant loop piping and (2) a mockup to demonstrate the effectiveness of the excavate and weld repair process.
Fatigue Management During LTO of Nuclear Power Plant Borssele
Authors: Hannink M. H. C. , de Bont C. G. M. , Blom F. J. , Quist P. W. B. , de Jong A. E. , Besuijen W.
Published: December 2016

Fatigue is an important ageing effect to manage for long term operation (LTO) of nuclear power plants (NPPs). One of the steps in the process of LTO assessment is the revalidation of TLAAs (time limited ageing analyses). The safety margins of NPP Borssele with respect to fatigue were demonstrated by projecting the fatigue analyses to the end of the intended period of LTO. Besides this, it has to be ensured that the analyses remain valid during the entire period of LTO. Periodic verification of the load assumptions that are made in the analyses is therefore an important aspect of adequate fatigue management. In this paper, an integrated fatigue management approach is presented, coupling load monitoring, transient counting and fatigue assessment.

The approach for periodic verification of the load assumptions in the fatigue analyses consists of two parts. First, it is verified whether the occurred numbers of cycles of the different load cases remain smaller than the numbers of cycles assumed in the fatigue analyses. Secondly, it is verified whether the thermal transients defined for the load cases in the fatigue analyses conservatively represent the occurred thermal transients. For the verification, the application LEAF (Load Evaluation Application for Fatigue) was developed. At NPP Borssele, thermal transients occurring during different load cases (e.g. start-up, shutdown, reactor trip) are registered by a temperature monitoring system. The load evaluation application processes the measurement data and verifies the conservatism of the load conditions in the fatigue analysis of the fatigue relevant components in the system. This paper explains the steps that are followed for the load evaluation and gives a demonstration of the results. The presented procedure is an essential part of adequate fatigue management during LTO.
Thermomechanical Fatigue Response and Constitutive Modeling for Haynes 230
Authors: Morrison Machel , Ahmed Raasheduddin , Hassan Tasnim
Published: December 2016

Design by analysis is usually performed by commercially available finite element analysis (FEA) software. Constitutive models available in the FEA software are developed and validated using limited experimental data. Hence, a broad set of thermomechanical fatigue experiments with strain dwell at compressive peaks are performed to understand local fatigue failure responses of high temperature components. This study developed a unified viscoplastic model based on nonlinear kinematic hardening of Chaboche type with added features of strain range dependence, rate dependence, temperature dependence, static recovery, and mean stress evolution. The robustness of the constitutive model is demonstrated by comparing its simulations against the experimental responses.
Effect of Carbide Content on Temper Embrittlement of 2.25Cr1Mo Steel
Authors: Wang Yian , Xie Guoshan , Zhang Zheng , Qian Xiaolong , Zhou Yufeng , Li Jing
Published: December 2016

Temper embrittlement is a common damage mechanism of pressure vessels in the chemical and petrochemical industry serviced in high temperature, which results in the reduction of roughness due to metallurgical change in some low alloy steels. Pressure vessels that are temper embrittled may be susceptible to brittle fracture under certain operating conditions which cause high stress by thermal gradients, e.g., during start-up and shutdown. 2.25Cr1-Mo steel is widely used to make hydrogenation reactor due to its superior combination of high mechanical strength, good weldability, excellent high temperature hydrogen attack (HTHA) and oxidation-resistance. However, 2.25Cr-1Mo steel is particularly susceptible to temper embrittlement.

In this paper, the effect of carbide on temper embrittlement of 2.25Cr-1Mo steel was investigated. Mechanical properties and the ductile-brittle transition temperature (DBTT) of 2.25Cr-1Mo steel were measured by tensile test and impact test. The tests were performed at two positions (base metal and weld metal) and three states (original, step cooling treated and in-service for a hundred thousand hours).

The content and distribution of carbides were analyzed by scanning electron microscope (SEM). The content of Cr and Mo elements in carbide was measured by energy dispersive X-ray analysis (EDS). The results showed that the embrittlement could increase the strength and reduce the plasticity. Higher carbide contents appear to be responsible for the higher DBTT. The in-service 2.25Cr-1Mo steel showed the highest DBTT and carbide content, followed by step cooling treated 2.25Cr-1Mo steel, while the as-received 2.25Cr-1Mo steel has the minimum DBTT and carbide content. At the same time, the Cr and Mo contents in carbide increased with the increasing of DBTT. It is well known that the specimen analyzed by SEM is very small in size, sampling SEM specimen is convenient and nondestructive to pressure vessel. Therefore, the relationship between DBTT and the content of carbide offers a feasible nondestructive method for quantitative measuring the temper embrittlement of 2.25Cr-1Mo steel pressure vessel.
Comparison of Leak-Before-Break Assessment of Main Loop Piping Lines Fabricated of Different Materials
Authors: Ma Linwei , Zheng Xiaotao , Wang Yan , He Jiasheng , Shu Anqing
Published: December 2016

Leak-Before-Break (LBB) assessment is used for the design of nuclear reactor coolant system main loop piping to lower the cost of construction and operation in China. In these applications, the materials of main loop piping lines are cast austenitic stainless steel (CASS) or wrought stainless steel (WSS) due to the different type of reactor design. According to US.NRC SRP3.6.3, LBB assessment includes two major calculations, such as critical crack size calculation and leakage flaw size calculation. The elastic-plastic instability analysis or plastic instability analysis is chosen to perform critical size calculation depending on material properties, especially fracture toughness. In this paper, LBB assessment in the guidance of SRP 3.6.3 was performed to evaluate main loop piping lines of CASS and WSS. The JR curve tests and the adjustment due to thermal aging are performed to achieve reasonable material properties. J integral/tearing modulus approach is used to determine critical crack size of CASS pipe and net section collapse (NSC) approach is used to determine critical crack size of WSS pipe. Leakage flaw size under 1gpm leakage detection capability is determined based on Henry’s homogeneous nonequilibrium critical flow model. In order to demonstrate that fatigue crack growth is not a potential source of pipe rupture for the evaluated piping lines, the fatigue crack growth of a postulated circumferential part-through-wall crack under nuclear power plant full life time operating transients and the fatigue crack growth of a circumferential through-wall crack under one time safe shutdown seismic are analyzed. And the LBB assessment procedure and results of CASS pipe and WSS pipe are compared.
R5 Procedure Based Damage Estimation in a Steam Turbine Valve Under In-Service Conditions
Authors: Cai Zhenwei , Wang Weizhe , Hong Hui , Liu Yingzheng
Published: December 2016

The fatigue behavior in steam turbine valve is generally considered to be associated with thermal transients during start and stop phase. However, the recent analysis has shown that steam parameter fluctuations under in-service steady state operation could induce significant increase in the damage. In this paper, the effect of in-service steam parameter fluctuations on stress-strain behavior in the valve was analyzed based on FE analysis under in-service data. The rain-flow cycle counting method was applied to get effective stress-strain cycle numbers and cycle amplitudes to classify the types of fatigue cycle based on in-service data. The creep-fatigue damage during steady state operation of the valve was estimated by using R5 (Volume 2/3) high temperature assessment procedure together with the FE results. Frequent steam pressure fluctuations at steady state operation were identified as the most influential factor for the fatigue life of the steam turbine valve.
Comparison of Fatigue Property Obtained Under High Pressure Hydrogen Gas Service and Cathodic Hydrogen Charging Service
Authors: Takagi Shusaku , Okano Hiroshi , Nagao Akihide
Published: December 2016

Evaluations of mechanical properties under high pressure hydrogen gas environments are difficult because special testing equipment which can obtain data under high pressure hydrogen gas is necessary and the cost of installing those machines is very high. Therefore, a method for predicting mechanical properties under high pressure hydrogen gas without special equipment was investigated in this study. A JIS SCM435 steel, of which the typical chemical composition is Fe-0.35%C-0.2%Si-0.75%Mn-1.1%Cr-0.2%Mo, was used. Material with the size of 22mm in square and 200mm in length was heat-treated by quenching in oil and tempering, followed by water cooling. The material microstructure was tempered martensite, and the tensile strength of the steel was about 1000MPa. The fatigue property was evaluated in cathodic hydrogen charging service and high pressure hydrogen gas service. The fatigue property obtained by the cathodic charging method was almost the same as the property under the high pressure hydrogen gas condition. The reason is conjectured to be that the hydrogen fugacity on the specimen surface and the surface state of the specimen under those test services was substantially the same.
Steel-Concrete Composite Vessel for Stationary High-Pressure Hydrogen Storage
Authors: Wang Yanli , Feng Zhili , Ren Fei , Lim Yong Chae , Chen Jian , Jawad Maan
Published: December 2016

A novel Steel Concrete Composite Vessel (SCCV) was designed and engineered for stationary high-pressure gaseous hydrogen storage applications. SCCV comprises four major innovations: (1) flexible modular design for storage stations for scalability to meet different storage pressure and capacity needs, flexibility for cost optimization, and system reliability and safety, (2) composite storage vessel design and construction with an inner steel vessel encased in a pre-stressed and reinforced outer concrete shellshell, (3) layered steel vessel wall and vent holes to address the hydrogen embrittlement (HE) problem by design, and (4) integrated sensor system to monitor the structural integrity and operation status of the storage system. Together, these innovations form an integrated approach to make the SCCV cost competitive and inherently safe for stationary high-pressure hydrogen storage services.

A demonstration SCCV has been designed and fabricated to demonstrate its technical feasibility. Capable of storing approximately 89 kg of gaseous hydrogen at 6250 psi (430 bar), the demonstration vessel was designed to include all major features of SCCV design and fabricated with today’s manufacturing technologies and code/standard requirements. Two crucial tests have been performed on this demonstration vessel. A hydro-test was successfully carried out to 8950 psi per ASME VIII-2 requirements. The cyclic hydrogen pressure test between 2000 psi and 6000 psi is currently being performed to validate its use for high-pressure hydrogen storage. Multiple sensors, such as pressure sensors and strain gages, were incorporated in the demonstration SCCV to collect information to validate the design and operation of SCCV. Key design parameters and test data on its performance are summarized in this paper.
Evaluation of Ductile Crack Propagation Behavior for Subsurface Crack in Cladded Plates Using XFEM
Authors: Nagai Masaki , Murai Kiminori , Nagashima Toshio , Miura Naoki
Published: December 2016

In petrochemical and nuclear power plants, inner surface of components, such as pressure vessels and piping, which are made of carbon steels or low-alloy steels, is often cladded by austenitic stainless steels to improve the corrosion resistance of those components. In the evaluation of a crack postulated near the inner surface of cladded components, the following two different kinds of cracks are often assumed: One is a surface crack penetrated through cladding and the other is a subsurface crack under cladding. In order to evaluate the structural integrity of those components in a rational manner, it is important to investigate the crack propagation behavior. In particular, it should be evaluated whether a subsurface crack penetrates through cladding or is arrested at the interface between cladding and base metal. In this study, ductile crack propagation analysis for a subsurface elliptical crack in cladded plates was performed using elastic-plastic XFEM, which can model the crack independent of finite elements.
Mechanism of Hydrogen-Assisted Surface Crack Growth of Austenitic Stainless Steels in Slow Strain Rate Tensile Test
Authors: Matsuoka Saburo , Yamabe Junichiro , Matsunaga Hisao
Published: December 2016

Several criteria based on reduction in area (RA) or relative RA (RRA) are proposed for determining the hydrogen compatibility of austenitic stainless steels; however, the mechanism of hydrogen-induced degradation in RA and RRA is not necessarily clear. The degradation in the RA and RRA of the austenitic stainless steels is attributed to hydrogen-assisted surface crack growth (HASCG) accompanied by quasi-cleavages; therefore, a mechanism of the HASCG should be elucidated to establish novel criteria for authorizing various austenitic stainless steels for use in high-pressure gaseous hydrogen. To elucidate the HASCG mechanism, this study performed slow strain rate tensile (SSRT), elasto-plastic fracture toughness (JIC), fatigue crack growth (FCG) and fatigue life tests on Types 304, 316 and 316L in high-pressure hydrogen gas. Experimental results of Type 304 were provided in this paper as a representative of Types 304, 316 and 316L. The results demonstrated that the SSRT surface crack grew via the same mechanism as for the JIC and fatigue cracks, i.e., these crack growths could be uniformly explained on the basis of the hydrogen-induced successive crack growth (HISCG) model, which considers that cracks successively grow with a sharp shape under the loading process, due to local slip deformations near the crack tip by hydrogen. Accordingly, the HIS crack is ductile, not brittle.
Development of Mechanism for Preventing Derailing of Material Handling Equipment
Authors: Iwamoto Kosuke , Sato Yuji , Otoyo Teruyoshi , Horiuchi Munenori , Sakai Hidetoshi , Arai Haruki
Published: December 2016

All of the traveling cranes at the port of Kobe were damaged by the strong ground motion of the Southern Hyogo Prefecture Earthquake in 1995. Seismic isolation systems for traveling cranes were developed to increase earthquake resistance against strong ground motions. In general, the isolation systems for traveling cranes for strong ground motions are very large. Quays which are not reinforced have a risk of being damaged by strong ground motions, and require traveling cranes to prevent derailing during middle strong motions, because derailing leads to collapse of the traveling cranes. A mechanism for preventing derailing of material handling equipment was developed for middle strong ground motions.

A middle rocker beam of the new mechanism is divided into two beams, and uplift motion of a leg of a crane can be absorbed by rotating motions of the two beams. This allows the mechanism for preventing derailing to be easily installed into the established equipment. Seismic analyses using an finite element model of the traveling crane and vibration tests were conducted to verify the validity of the design and derailing prevention performance. The results of analyses and vibration tests showed that the mechanism for preventing derailing can prevent wheels from derailing during middle strong earthquakes. The configuration of the mechanism and the results of the analyses and vibration tests are presented in this paper.
A Study of the Stress Intensity Factor and Crack Opening Displacement Relationship Between Uniform Thickness Pipe Bends and Non-Uniform Thickness Pipe Bends
Authors: Bae Kyung-Dong , Kim Chul-Goo , Kim Seung-Jae , Lee Hyun-Jae , Kim Yun-Jae
Published: December 2016

This paper proposes the relationship of stress intensity factor and crack opening displacement between pipe bends with uniform thickness and those with non-uniform thickness. In actual case, pipe bends have thickness variations. Unlike typical pipe bends, heat induction bend pipes have significant thickness variations (non-uniform thickness) because of manufacturing process. When the ratio of radius of curvature and pipe radius is 3 for heat induction bend pipes, the thickness at intrados and extrados can be calculated by 1.75 times and 0.875 times of nominal thickness which is original thickness before manufacturing process, respectively. In this situation, it is difficult to apply existing elastic stress intensity factor and crack opening displacement results [1, 2] and it is essential to modify existing solution or to create new solution. In this paper, to find effect of pipe bends thickness variation, 90° through-wall cracked pipe bends with not only uniform thickness but also non-uniform thickness are considered. The ratios of radius and thickness are 5, 10 and ratios of pipe radius of curvature and radius are 3, 4 and 5. Loading condition is in-plane opening bending for intrados crack and closing bending for extrados crack. The through-wall crack sizes are 12.5%, 25% and 37.5% of circumferential cross section. Material of pipe bends is assumed to follow elastic behavior. The proposal is made by extensive finite elements analyses using ABAQUS [3], predicted elastic stress intensity factors for pipe bends with non-uniform thickness are compared with finite element results. The results show a good agreement. It may be useful to calculate elastic stress intensity factor for bends with non-uniform thickness without complex modeling and finite analyses.
Heat Removal Performance in Accordance With Mesh Size of Screen Installed at Opening of the Inlets and Outlets of Concrete Storage Cask
Authors: Bang Kyoung-Sik , Yu Seung-Hwan , Lee Ju-Chan , Seo Ki-Seog , Choi Woo-Seok
Published: December 2016

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A concrete storage cask developed for this purpose should be able to adequately emit the decay heat from the spent nuclear fuel. Moreover, the concrete storage cask must ensure that the temperatures of the spent nuclear fuel assemblies are maintained within the allowable values under normal and offnormal conditions and during an accident. However, the thermal conductivity of concrete is low and the allowable temperature of concrete is lower than that of steel. Therefore, the concrete storage cask must be designed to have heat removal capabilities with appropriate reliability. Accordingly, a passive heat removal system was designed to maintain the temperatures of the fuel assembly cladding material and concrete storage cask components within the allowable limits. The passive heat removal system consists of four air inlets and four air outlets with openings that are covered by mesh screens to prevent debris or wildlife from entering the ventilation ducts. The effect of each screen on the heat removal of the concrete storage cask depends on its mesh size. Therefore, thermal tests were performed to evaluate the heat removal performance in accordance with the mesh size of the screen. The screen mesh size was estimated to have an insignificant effect on the temperature rise of the canister surface and the over-pack surface, but it had a considerable effect on the temperature rise of the components of the over-pack body. As the screen mesh size decreased, the heat removal by the natural convection cooling through the passive heat-removal system was reduced, and the temperature of the concrete storage cask rose.
C(t) Estimation of Cracked Pipe Under Radial Thermal Gradient or Axial Displacement Control Load
Authors: Je Jin-Ho , Lee Kuk-Hee , Kim Yun-Jae , Oh Young-Ryun
Published: December 2016

The present study analyzes the elastic follow-up factor under various secondary loads in the piping with cracks through finite element analysis. The secondary loads being considered included axial tensile load caused by the displacement applied in form of uniform stress through thickness and the load due to the thermal gradient in the radial direction applied as bending stress. According to this study, under displacement control loads, elastic follow-up factor can increase greatly.
Development of a Core Seismic Analysis Method for a Fast Reactor
Authors: Iwasaki Akihisa , Hirota Kazuo , Monde Masatsugu , Matsubara Shinichiro , Ikarimoto Iwao
Published: December 2016

A fast reactor core consists of several hundreds of core assemblies, which are hexagonal flexible beams embedded at the lower support plate in a hexagonal arrangement, separated by small gaps, and immersed in a fluid. Core assemblies have no support for vertical fixing in order to avoid the influence of thermal expansion and swelling. These days, in Japan, it has become necessary to postulate huge earthquakes in seismic evaluations. If a great earthquake occurs, the large displacement and impact force in each core assembly may cause problems with control rod insertability and core assembly strength. So, it is necessary to grasp the vibration behavior of the core elements during an earthquake in order to appropriately design the core support structures and core elements of a fast reactor. Thus, considering horizontal and vertical forces (impact forces and fluid forces) acting on the core elements during an earthquake, a core seismic analysis method has been developed to evaluate 3D core vibration behavior considering fluid structure interaction and vertical displacements (rising). This paper summarizes the details of the core element vibration analysis code in 3D (REVIAN-3D) that has been developed.
Fracture-Mechanics Analysis of the Reactor Pressure Vessel Beznau 1 Based on Thermal-Hydraulics Input Data From KWU-MIX and CFD Analyses
Authors: Nicak Tomas , Trewin Richard , Keim Elisabeth , Cremer Ingo , Blasset Sebastien , Hägeli Heinz
Published: December 2016

The integrity of a reactor pressure vessel (RPV) has to be ensured throughout its entire life in accordance with the applicable regulations. Typically an assessment of the RPV against brittle failure needs to be conducted by taking into account all possible loading cases. One of the most severe loading cases, which can potentially occur during the operating time, is the loss-of-coolant accident, where cold water is injected into the RPV nearly at operating conditions. High pressure in combination with a thermal shock of the ferritic pressure vessel wall caused by the injection of cold water leads to a considerable load at the belt-line area known as Pressurized Thermal Shock (PTS).

Usually the assessment against brittle failure is based on a deterministic fracture-mechanics analysis, in which common parameters like J-integral or stress intensity factor are employed to calculate the load path for an assumed (postulated) flaw during the PTS event.

The most important input data for the fracture-mechanics analysis is the transient thermal-hydraulics (TH) load of the RPV during the emergency cooling. Such data can be calculated by analytical fluid-mixing codes verified on experiments, such as KWU-MIX, or by numerical Computational Fluid Dynamics (CFD) tools after suitable validation. In KWU-MIX, which is the standard used for TH calculations within PTS analyses, rather conservative analytical models for the quantification of mixing and, depending on the water level, condensation processes in the downcomer (including simplified stripe and plume formations) are utilized. On the contrary, the numerical CFD tools can provide best-estimate results due to the possibility to consider more realistically the stripe and plume formations as well as the geometry of the RPV in detail.

In order to quantify the safety margin inherent to the standard approach, two fracture-mechanics analyses of the RPV Beznau 1 based on thermal-hydraulic input data from KWU-MIX and CFD analyses were performed. Subsequently the resulting loading paths were compared between each other and with material properties obtained from the irradiation surveillance program of the RPV to demonstrate the exclusion of brittle-fracture initiation.
Optimization of High Pressure Components for LDPE Manufacturing
Authors: Blome Matthias , Rueckriem Stefan
Published: December 2016

Coned and threaded type connection valves, fittings and tubing up to sizes of 9/16” are typical high pressure components used on the initiator injection system of a Low Density Polyethylene LDPE facility. The process pressures of up to 360 MPa (52,200 psi) on the reactor side require even higher pressures on the initiator injection lines. Based on the differential pressures and the challenging environment 414 MPa (60,000 psi) has been defined as an appropriate pressure rating for such components.

Valves, fittings and tubing with this pressure rating are more or less standardized and available from multiple sources on the market. However, the technical challenges in designing and manufacturing components for this duty are sometimes under-estimated and design improvements can lead to a higher process safety and a reduced downtime. This paper covers the design requirements and has a specific focus on a filter with screwed element and a check valve with ceramic ball.

The majority of high pressure in-line filters on the market have a sintered element that is press-fitted into the filter body. This design has some disadvantages in case of fast depressurizations as they occur when a safety valve or a rupture disk blows off. The sudden differential pressure may cause the element to move inside the body, and this can be accompanied by cracking of the element. Parts of the broken sintered element are a risk to the process since they would move downstream followed by a contamination of the reactor.

A new generation of such filters with screw-in elements was designed and tested. The R&D work resulted in an extended element that fits into a modified filter body with standard outside dimensions. Before performing a field test at a LDPE plant three prototype samples went through intensive testing at the manufacturer’s facilities.

Check valves are typical components used in order to maintain the flow in one direction. For liquids the metal to metal sealing ball type is the most common. Since the process pressure exposes the mating parts — ball and orifice — to high compression forces, imprints and marks on the ball surface are sometimes experienced. The material of these types of balls is typically bearing steel which has high mechanical strength but limited corrosion resistance. In order to improve the cycle life of high pressure ball check valves alternative materials were looked at. During the further product improvement a check valve with ceramic ball was designed and lab tested at the manufacturer’s facilities.
Fitness-for-Service of a Column Equipment Containing Dent Defects Based on API 579 Level 3 Analysis and Considering the Dynamic Loads
Authors: Han Zhiyuan , Xie Guoshan , Sun Liang , Zhao Minzhen , Qian Haiyan , Cao Luowei
Published: December 2016

Fitness-for-service assessment of a pressurized component containing dents or other mechanical damage is important to ensure the operational safety and structural integrity of the damaged equipment. In the present study, the API 579 level 3 fitness-for-service analysis were performed for a column equipment containing dent defects, and the main difficulty is to determine the applied dynamic load, especially when considering wind load and earthquake load at high order mode shapes. To solve this problem, a simplified method combining the analytic calculation and finite element (FE) method were proposed in this study. At the first step, the wind load and earthquake load on different segments of the column without defect were calculated by analytic methods according to Chinese code GB4710. Then the critical load combination determined in step one were applied on a whole column FE model containing 6 dent defects with different dimensions. Based on the FE simulation, the stress linearization were performed for strength check, and buckling analysis were performed buckling collapse failure respectively. The results of strength check and critical buckling load showed that the column containing dents under static load and dynamic load conditions was acceptable for continued operation based on API 579 level 3 analysis. Thus, the methodology developed in this study provides an available fitness-for-service assessment for dents in column equipment and enables the consideration of dynamic loads.
On the Applicability of Neuber’s Rule for Low-Cycle Fatigue
Authors: Spring Daniel W. , Gassama Edrissa , Stenta Aaron , Cochran Jeffrey , Panzarella Charles
Published: December 2016

Neuber’s rule is commonly applied in fatigue analysis to estimate the plasticity of purely elastic FEA results. In certain cases, this is more efficient than running elastic-plastic models. However, the applicability of Neuber’s rule is not well understood for complex models and may not always be appropriate. In this paper, the applicability of Neuber’s rule is investigated. The background of Neuber’s rule is discussed, theoretical limitations are derived, and algorithmic outlines of the procedures are presented. Neuber’s plasticity correction procedure is applied to both the Ramberg-Osgood elastic-plastic constitutive relation and the advanced Chaboche isotropic/kinematic nonlinear hardening relation. Throughout the manuscript, the aspects of each model are discussed from an educational perspective, highlighting each step of the implementation in sufficient detail for independent reproduction and verification. This level of detail is often absent from similar publications and, it is hoped, may lead to the wider dissemination of Neuber’s rule for plasticity correction. The final component of the paper presents a multiaxial correction of the Chaboche hardening model. To the best of the authors’ knowledge, this is the first published application of Neuber’s rule to the multiaxial plasticity correction of the Chaboche combined isotropic/kinematic hardening model. Examples are used to illustrate the behavior of the method and to present some of the commonly overlooked components when assessing the applicability of Neuber’s method.
Assessment of RPV Surveillance Materials Program Data of a BWR
Authors: Hernández Callejas Rogelio , Medina-Almazán A. Liliana , Merino Caballero Fco. Javier , Vázquez Belmont Salvador
Published: December 2016

Irradiation embrittlement is a limiting condition for the long-term safety operation of a nuclear Reactor Pressure Vessel (RPV). When a Boiling Water Reactor (BWR) is approaching its initial licensing, in order to operate the reactor for another 20 years and more, it should be demonstrated that the irradiation embrittlement of the reactor vessel materials will be adequately managed by ensuring that the fracture toughness properties are above a certain level of the required safety margin. In this work the Charpy specimens recovered from two surveillance capsules of two BWRs (fluence 3.58×1017 – 9.03×1017 n/cm2) were impact tested at temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials. The measured transition temperature shifts (ΔRTNDT) and the Upper Shelf Energy (USE) for the plate and weld materials were compared to the predictions calculated according to Regulatory Guide 1.99 Rev.2. The credibility of surveillance data were analyzed according with the five criteria established in the Regulatory Guide 1.99, Revision 2. The Master Curve (MC) approach and the instrumented impact tests using pre-cracked Charpy specimens were implemented in order to fully validate this techniques that can be used for embrittlement monitoring during life extension periods.
Influence of Residual Stress due to Prior Overloading on the Fatigue Life of a Notched Plate
Authors: Karpanan Kumarswamy
Published: December 2016

During autofrettage, pressure vessels are subjected to high internal pressure, causing the internal wall to yield plastically. When the internal pressure is released, the inner wall of the vessel develops compressive residual stress. Similarly, when a subsea component is hydrotested, some of the highly stressed regions yield during hydrotesting and, when the pressure is released, these regions develop compressive residual stress. Fatigue life is greatly influenced by local stress on the component surface. Fatigue crack initiation primarily depends on the cyclic stress or strain and the residual stress state. Tensile residual stress decreases fatigue life and the compressive residual stress significantly increases fatigue life. This is true for both fatigue crack initiation and propagation.

In this paper, effects of residual stress on a notched plate are studied by subjecting it to an initial overload cycle and subsequent low loading cycles. Tensile and compressive overloads on the notched plate induce compressive and tensile residual stresses, respectively. An elastic-plastic finite element analysis (FEA) was performed to simulate the overload and low loading cycles on the notched plate. The stress and strain from the FEA is used to perform strain-based fatigue analysis. ASME VIII-3, Brown-Miller (B-M), Maximum shear strain, Socie-Bannantine, and Fatemi-Socie methods are used for calculating the fatigue life of the notched plate. Fatigue life predicted by both stress and strain methods matches well with the test fatigue data.
The Method for Structural Reliability Estimation of the Heat Exchanger Tubes of Steam Generator of WWER NPPs
Authors: Orynyak Igor , Zarazovskii Maksym , Borodii Mykhaylo
Published: December 2016

After the replacement due to the well known problem of weld seam #111 the steam generators (SG) of the Ukrainian WWER-1000 type nuclear power plants (NPP) have been in operation for almost 25 years. So now, the main practical problem related to the SG is the leakage/breaking of heat exchanger tubes (HET). If the leakage criterion is not satisfied the NPP Unit has to be stopped until the corresponding HET is plugged, which leads to unexpected huge financial losses.

Every year at least 12.5% of the SG HETs are controlled by eddy current inspection, after which some of the HETs are plugged according to the appropriate plugging criteria (in most cases – 65% loss of tube wall thickness). Thus, every SG has its own statistics of plugging.

From the practical point of view the Utility is interested in the following questions: whether the failure of the HET happens during the normal operation mode (NOM) up to the next NPP stop or not; what is the probability of this event; what is the defect growth rate? Answers on these questions could give us the opportunity to increase the effectiveness of the HET in-service inspection and to improve the plugging criterion.

A simple statistically-based method for HET integrity assessment is proposed. The method based on an exponent distribution law and HET plugging statistics of specific SG, taking into account defect growth. Based on the history of the tubes plugging (year of operation versus quantity of pluggings) the three statistical parameters inherent to this specific SG have to be found: initial number of defects, parameter of the exponential distribution (initial defect size) and the defect growth rate.

The developed method was used for the prediction of HET leak/break number for all Ukrainian WWER-1000 SG. It is shown that for those SG which has less than 2% HET pluggings the accumulated pluggings as of the end of the next year of NOM will not exceed this value.

Based on the Ukrainian history of WWER-1000 SG operation the database of HET plugging is presented.

This method is also used for justification of pressure reduction of periodic hydrostatic test (HT) for primary circuit of WWER-1000 and WWER-440 NPPs from 24.5 MPa to 20.3 MPa and from 19.1 MPa to 15.7 MPa, correspondingly. To justify a reduction of the HT pressure, a quantitative, risk-informed assessment of HT effectiveness has been performed with taking into account the HT pressures change. The HET failure probability is calculated as the proportion of defects which exceed critical sizes for NOM and HT. Limit load models are used for the determination of the critical defects size. The variation of reliability is calculated as the difference between fracture probability during NOM after HT at routine and reduced pressures. It is shown that HT pressure reduction does not practically increase the fracture probability during operation, and satisfies the criteria of risk change.