bluuebunny/crossref_metadata_2025_split huggingface dataset:
Dissecting the Dinosaur; Problems With B16.5 and B16.47 Flange Standards
Authors: Brown Warren 
Published: January 2014

ASME B16.5 and B16.47 flanges have been in existence for a very long time with very few changes made since their creation. That might lead the casual observer to think that the designs contained in those documents have been optimized and represent best in technology. While that may have been the case when they were first introduced, it is anything but the case at the moment. In fact, in many cases the design of these standard flanges is far from optimal. However, even worse than that, there are presently several significant issues with the flange standards that could be easily resolved, without adding cost to the flanges or requiring dimensional modifications.
These changes would eliminate some existing or potential joint integrity failures. This paper outlines a list of improvements, complete with justification for the need, which the author would like to see made to the ASME standard flange standards. Some of these suggestions have already gained support in industry and been relayed on to the B16.5 committee, however they have not yet been incorporated. These simple improvements can, for the most part, be implemented within existing company standards in order to immediately circumvent potential or existing joint leakage issues. The paper details suggestions for modification of the standard B16.5, B16.47 and B16.20 wording, both for consideration by the B16 committees and incorporation into corporate standards in the interim.
Applicability of Miniature C(T) Specimen to Evaluation of Fracture Toughness for Reactor Pressure Vessel Steel
Authors: Yoshimoto Kentaro , Hirota Takatoshi , Sakamoto Hiroyuki , Sugihara Takuji , Sakaguchi Shohei , Oumaya Toru 
Published: January 2014

Irradiation embrittlement of Japanese reactor pressure vessels (RPV) is usually monitored by conducting tests on irradiated RPV material according to surveillance test program. Although fracture toughness specimens are contained in Japanese PWR surveillance capsule, the number of specimens is limited due to capacity of capsule. In order to evaluate lower bound of fracture toughness considering its scatter with higher reliability, it is expected to obtain additional fracture toughness data using remaining broken specimens of irradiated materials.
One of solutions to this problem is specimen reconstitution technique. However, it is difficult to make numbers of specimens by reconstitution because of need for specific equipments and time-consuming machining operations. As an alternative method, fracture toughness test using miniature C(T) specimens with dimension of 4×10×10mm, which can be taken from broken halves of Charpy specimen, is proposed and the studies to verify the reliability and robustness of evaluation method have been conducted in the Japanese round robin program since 2010.
In this study, fracture toughness tests were performed on Japanese SA 533 Gr.B Cl.1 steel using miniature C(T) specimens and the effect of specimen size on reference temperature T0 was studied by the Master Curve approach. In addition, the issues related to application to irradiated materials were discussed.
Study on Piping Response Under Multiple Excitation: Part 2 — Validation for Multiple Excitation Analysis of Piping
Authors: Kai Satoru , Watakabe Tomoyoshi , Kaneko Naoaki , Tochiki Kunihiro , Moriizumi Makoto , Tsukimori Kazuyuki 
Published: January 2014

The piping in a nuclear power plant is laid across multiple floors of a single building or two buildings, which are supported at many points. As the piping is excited by multiple-inputs from the supporting points during an earthquake, seismic response analysis by multiple excitations is needed to obtain the exact seismic response of the piping. However, few experiments involving such multiple excitation have been performed to verify the validity of multiple excitation analysis. Therefore, analysis of the seismic design of piping in Japan is performed by the enveloped Floor Response Spectrum (FRS), which covers all floor response spectra at all supporting points. The piping response estimated by enveloped FRS is conservative in most cases compared with the actual seismic response by multiple excitations. To perform rational seismic design and evaluation, it is important to investigate the seismic response by multiple excitations and to verify the validity of the analytical method by multiple excitation test.
This paper reports the validation results of the multiple-excitation analysis of piping compared with the results of the multiple excitations shaking test using triple uni-axial shaking table and a 3-dimensional piping model (89.1mm diameter and 5.5mm thickness). Three directional moments from the analysis and the shaking test were compared on the validation. As the result, it is confirmed that the analysis by multiple time history excitation corresponds with the test result.
Technical Basis for Proposed Code Case of Alternative Rules for Cladding Repair by Underwater Electrochemical Deposition in Class 1 and 2 Applications
Authors: Kim Myong-Jin , Kim Dong Jin , Kim Joung Soo , Kim Hong Pyo , Hwang Seong Sik 
Published: January 2014

A reactor vessel consists of carbon/low alloy steel (SA508) as the structural material, and a cladding layer (SS309) as a protective layer. Cladding layer experiences damage by accidents such as thermal sleeve detachments. Carbon/low alloy steel is weak to corrosion when exposed to the primary water especially during refueling time, which is a potential threat against the integrity of a reactor vessel. Therefore, the damaged cladding part should be repaired. Repairing the inside of the reactor vessel is difficult because of high radiation levels, poor accessibility, and the underwater condition. One promising repair technology is the electrochemical deposition (ECD) of nickel.
Nickel plating through ECD has a lot advantages such as excellent corrosion resistance, no heavy thermal effect (near room temperature process), and proper mechanical properties for long-term nuclear operation. Moreover, the plating layer can be easily removed and is applicable to underwater conditions. The ECD technology used to repair the inside of the steam generator tube has been approved in the ASME code case N-569, titled “Alternative Rules for Repair by Electrochemical Deposition of Class 1 and Class 2 Steam Generator Tubing.” It is expected that an ECD layer can protect the corrosion/erosion of the exposed surface of damaged cladding/ carbon steel.
The ECD process consists of many processes, such as surface cleaning, strike layer formation for the adhesion between ECD layer and cladding/carbon steel, activation, and a main electrodeposition process. “Alternative Rules for Cladding Repair by Underwater Electrochemical Deposition in Class 1 and 2 Applications” is proposed based on the Code Case N-569. This proposed code case includes material requirements, qualification, essential variables, and examination requirements.
Applicability of Optical Fiber Bragg Grating Sensors in the Structural Health Monitoring of Composite Overwrapped Pressure Vessels
Authors: Pereira Gilmar , Figueiredo Joana , Faria Hugo , Marques A. Torres 
Published: January 2014

Composite overwrapped pressure vessels (COPV) have been increasingly pointed to as the most effective solution for high pressure storage of liquid and gaseous fluids. Reasonably high stiffness-to-weight ratios make them suitable for both static and mobile applications. However, higher operating pressures are sought continuously, to get higher energy densities in such storage systems, and safety aspects become critical. Thus, reliable design and test procedures are required to reduce the risks of undesired and unpredicted failures. An in-service health monitoring system may contribute to a better product development, design and optimization, as well as to minimize the risks and improve the public acceptance.
Within the scope of developing different COPV models for a wide range of operating pressures and applications, optical fiber Bragg grating (FBG) sensors were embedded in the liner-composite and composite-composite interfaces during their manufacture in order to allow the online strain monitoring during preliminary testing and service-life. The ability of these measuring systems to effectively assess the strain fields was to be investigated. Simultaneously, a finite element analysis (FEA) was made using the ABAQUS® platform. In this numerical analysis, accurate and realistic simulation of the different materials, geometry and loading conditions was approached. Particularly, the anisotropic nature of the wound laminate and the varying orientation of the fibers were attained. However, the cohesive zones were not attributed independent properties.
Comparison between experimental and numerical data was addressed. In general, although the experimental-numerical data agreement was not as good as desired, a preliminary insight to both the structural health monitoring (SHM) system and the numerical modeling approaches was actually achieved. Full characterization and validation shall be further addressed in the continuation of the present work.
The first set of results and difficulties on the development and implementation of this SHM system to COPV are presented and discussed in this paper.
A Parametric CFD Analysis of the Valve Pocket Losses in Reciprocating Compressors
Authors: Balduzzi Francesco , Ferrara Giovanni , Maleci Riccardo , Babbini Alberto , Pratelli Guido 
Published: January 2014

The increasing demand for high efficiency in the field of energy production has also had an impact on reciprocating compressors. In this case, the need to reduce losses is one of the most important challenges. In the past, most studies were focused only on the analysis of valve pressure losses. More recently, interest has been extended to the prediction of the losses through pocket valves and cylinder ducts, these losses being a crucial aspect of a more accurate estimate of the absorbed power.
The use of CFD simulation has shown great potential for the study of the entire reciprocating compressor, but is still limited by high computational costs. Recently the authors have presented a simplified CFD approach: the actual valve geometry is replaced with an equivalent porous region, which has significantly increased the speed of calculation while ensuring accuracy as well.
Based on this approach, a new methodology for the evaluation of pocket valve losses is presented. Since the pocket losses have been proven to be independent of the actual valve installed, a set of CFD simulations using a parameterized geometry of the pocket valve was performed and the relationship between the losses of the pocket and its geometrical features was obtained. An analytical response surface was defined using the values of the geometrical dimensions as inputs and the pocket flow coefficient as output. Finally, the response surface was validated through a set of test cases performed on different geometries with the actual valve and the results have shown good predictability of the tool.
Proposed Modifications to the EN 1591-1 Flange Calculation Procedure
Authors: Schaaf Manfred 
Published: January 2014

At present, the European Standards EN 1591-1 which is used for the calculation of bolted flanged joints and EN 13555 in which the determination of the required gasket characteristics are defined are under revision.
Several studies on gasket testing and on flange calculation have been performed in order to verify the suggested modifications of the standardization working groups. As a result of these studies several optimizations in respect of the testing procedures, the definition of the gasket characteristics and the calculation procedure can be derived.
In this paper, an advanced calculation procedure will be presented which is based on the use of additional or modified gasket characteristics such as the stiffness-independent gasket deflection caused by creep of the gasket material or formulas for the description of the gasket compression curves are presented.
Improved Gasket Testing
Authors: Schaaf Manfred 
Published: January 2014

At present, the European Standards EN 1591-1 which is used for the calculation of bolted flanged joints and EN 13555 in which the determination of the required gasket characteristics are defined are under revision.
Several studies on gasket testing and on flange calculation have been performed in order to verify the suggested modifications of the standardization working groups. As a result of these studies several optimizations in respect of the testing procedures, the definition of the gasket characteristics and the calculation procedure can be derived.
In this paper, additional or modified gasket characteristics such as the stiffness-independent gasket deflection, caused by creep of the gasket material or formulas for the description of the gasket compression curves, are presented. Several tests with diverse gasket materials were performed in order to generate generally admitted procedures to determine these required characteristics which are the basis for the advanced calculation procedure which will be presented in a second paper.
Method for Estimation of Pressure Vessel Internals Lifetime on IASCC Criterion
Authors: Fedorova Valentina , Margolin Boris 
Published: January 2014

Austenitic stainless steels are used extensively as structural materials in the internal components of reactor pressure vessels. However, high neutron doses lead to a significant reduction in the fracture resistance of these steels in water environment. Irradiation assisted stress corrosion cracking (IASCC) of internals has been observed in pressurized water reactors (PWRs).
In the present work the IASCC model of the irradiated austenitic steels in PWR water has been developed.
On the basis of analysis of available experimental data IASCC mechanism is proposed. Based on this mechanism, the dependence of fracture stress under IASCC on neutron dose is derived. For its construction the following assumptions were made.
1. Creep rate due to grain boundary sliding does not depend on neutron dose.
2. Fracture strain due to grain boundary sliding decreases when neutron dose increases.
3. There is an apparent stress threshold below which IASCC initiation does not occur in PWR environment.
Life prediction analysis for IASCC is performed on the basis of linear rule of damage accumulation.
Brittle Fracture Analysis of a Dissimilar Metal Weld
Authors: Blouin A. , Chapuliot S. , Marie S. , Bergheau J. M. , Niclaeys C. 
Published: January 2014

One important part of the integrity demonstration of large ferritic components is based on the demonstration that they could never undergo brittle fracture. Connections between a ferritic component and an austenitic piping (Dissimilar Metal Weld — DMW) have to respect these rules, in particular the Heat Affected Zone (HAZ) created by the welding process and which encounters a brittle-to-ductile transition.
Within that frame, the case considered in this article is a Ni base alloy narrow gap weld joint between a ferritic pipe (A533 steel) and an austenitic pipe (316L stainless steel). The aim of the present study is to show that in the same loading conditions, the weld joint is less sensitive to the brittle fracture than the surrounding ferritic part of the component. That is to say that the demonstration should be focused on the ferritic base metal which is the weakest material.
The bases of this study rely on a stress-based criterion developed by Chapuliot et al., using a threshold stress (σth) below which the cleavage cannot occur. This threshold stress can be used to define the brittle crack occurrence probability, which means it is possible to determine the highest loading conditions without any brittle fracture risk.
Long Term Operation of Nuclear Power Plants: Status in the U.S.
Authors: Young Garry G. 
Published: January 2014

As of January 2013, the U.S. Nuclear Regulatory Commission (NRC) has renewed the operating licenses of 73 nuclear units out of a total of 104 licensed units, allowing for up to 60 years of safe operation. In addition, the NRC has license renewal applications under review for 15 units and more than 13 additional units have announced plans to submit applications over the next few years [1]. This brings the total of renewed licenses and plans for renewal to over 97% of the 104 operating nuclear units in the U.S.
This paper presents the status of the U.S. license renewal process and issues being raised for possible applications for subsequent renewals for up to 80 years of operation. By the end of 2013 there will be 26 nuclear plants in the U.S. (or 25% of the 104 units) that will be eligible to seek a second license renewal and by the end of 2016 this number will increase to about 50% of the 104 licensed units. Although some nuclear plant owners have announced plans to shutdown before reaching 60 years, the majority are keeping the option open for long term operation beyond 60 years. The factors that impact decisions for both the first license renewals and subsequent renewals for 80 years of safe operation are presented and discussed in this paper.
Prediction of the Crack Growth Rate Under Creep and Neutron Irradiation for Austenitic Stainless Steels in Initial, Aged and Irradiated Conditions
Authors: Buchatsky Andrey , Margolin Boris , Gulenko Alexander , Kashtanov Alexander 
Published: January 2014

This work is aimed at investigating the crack growth rate in austenitic steels under creep and neutron irradiation for a material in initial and aged conditions.
The crack growth rate under creep is studied for 18Cr-9Ni steel and its welds. This steel and its welds are the base structural materials for manufacturing BN-600 fast reactor components. Experimental investigations of crack growth rate have been performed for the base metal, weld metal and heat-affected zone which were subjected to thermal ageing during operation. It is shown that the fastest crack growth rate is observed for specimens from the base metal.
The crack growth rates are compared for the base metal in initial and aged conditions. The initial condition of metal is modeled by annealing thermally aged metal taken from the components after its operation. The investigations show that the crack growth rate for the initial condition is faster than for the aged condition.
A method for predicting the crack growth rate under creep and irradiation has been developed on the basis of a physical-and-mechanical intergranular fracture model proposed before [1]. The crack growth rate for the material in initial condition is used as input data to calculate this rate under creep and irradiation according to the developed method.
International Harmonisation: A Key Challenge for Mechanical Components Codes and Standards
Authors: Faidy Claude 
Published: January 2014

Many Code Comparisons have been done by different organizations. The differences are better and better known by Code Development Organizations like ASME, AFCEN, JSME, KEPIC, CSA and NIKIET.
This paper uses the last MDEP (Multinational Design Evaluation Program) comparison for class 1 components to classify differences in different aspects:
- National regulatory requirements
- Technical differences
- Scope differences.
Few examples are selected to confirm needs of harmonization in different areas like: documentation, design rules, materials, product specifications, welding and non destructive examination personal qualifications.
The paper concludes on selected topics proposed for harmonization by AFCEN.
The objectives of this paper is to discuss technical harmonization of Nuclear Codes and Standards, based on French long experience in Codes and Standards used for design-fabrication and operation of nuclear components (mainly pressure retaining components).
After a long period of use of ASME Section III code, during the Westinghouse licensing process, AFCEN (AREVA, EDF and the major manufacturers) decided to develop their own AFCEN French Codes. The 1st version has been issued in 1980 and the last one in 2007, completed by annual addendum. During more than 20 years the 2 Codes, RCC-M and ASME Section III, have leave separately, with different constraints like industrial history, localisation of fabrication, more new plants in France than in USA, different R&D programs to support Code improvement…
Recently a detailed review of differences for class 1 vessel has showed under a “general global quality equivalence”, a lot of differences in the Code development process, in the Code organization, in the scopes, in the State of the Art fulfillment, in ageing consideration at the design stage, in relation with national or international regulations, in term of standards used or complementary specification needs…
The harmonization of Codes and Standards is possible under an important effort to move toward new ideas, more international rules and with a strong support of national safety authorities.
A Study of the Effect of Long-Term Thermal Aging on the Creep-Rupture Properties of 18Cr-9Ni Steel
Authors: Gulenko A. G. , Margolin B. Z. , Buchatsky A. A. , Kashtanov A. D. 
Published: January 2014

Experimental investigations of the creep-rupture properties of metal of intermediate heat exchangers (IHX) from fast breeder reactor BN-600 after its operation during ∼ 170 thousands hours over temperature range of 500–550°C were carried out. IHX was produced from 18Cr-9Ni steel. To study the role of thermal aging part of the IHX metal was annealed at 1000–1050°C for 0.5 hours. Comparison of experimental results of aging metal and annealed metal was carried out and the microstructure and fractography of both materials examined. The effect of aging on the creep-rupture properties of 18Cr-9Ni steel was revealed and described by the physical and mechanical model of intergranular fracture.
On the basis of experimental and calculative results an explanation of the effect of long-term aging on the mechanical short-term and creep-rupture properties of 18Cr-9Ni steel has been provided.
Acoustic Damping in Smooth and Corrugated Pipes With and Without Liquid Injection
Authors: Golliard J. , Belfroid S. P. C. , Vijlbrief O. 
Published: January 2014

From previous experiments, it was shown that small quantities of liquids reduce or eliminate the singing phenomenon in corrugated tubes. The singing mitigation was attributed to filling up of the corrugations, interference with the boundary layer or due to the increased acoustic damping.
To evaluate the influence of the presence of liquids on the acoustic damping, this damping was measured in a smooth and a corrugated tube (internal diameter 49 mm). The damping was measured using an active source (loudspeaker) and measuring amplitude of sound waves on both sides of the pipe. This was done at different gas velocities (U = 0, 15, 30 m/s) and different liquid injection rates (0, 150, 250 350 ml/min) for a frequency range between 0–2200 Hz.
The smooth pipe results compared well with literature models such as the Kirchhoff model. In the presence of liquid, the acoustic damping increased linearly with the liquid rate. This increase was higher than expected based on the assumptions of mist flow or overall pressure drop. In case of the corrugated tube, the measured damping is dominated by a large fluctuation due to the source behavior (negative damping) at flowing conditions. At low velocities a decrease in source strength was measured with increasing liquid. At higher gas velocities, it was observed that liquid has almost no effect on the damping.